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Proceedings of the International Symposium % URANiUM TECHNQLOqY
Organised by A ENGINEERING SCIENCES COMMITTEE / BOARD OF RESEARCH IN NUCLEAR SCIENCES DEPARTMENT OF ATOMIC ENERGY GOVERNMENT OF INDIA Proceedings of the International Symposium on URANIUM TECHNOLOGY
VOLUME II
BHABHA ATOMIC RESEARCH CENTRE, TROMBAY, BOMBAY, 400 085
DECEMBER 13-15, 1988
Organised by ENGINEERING SCIENCES COMMITTEE BOARD OF RESEARCH IN NUCLEAR SCIENCES DEPARTMENT OF ATOMIC ENERGY GOVERNMENT OF INDIA SYMPOSIUM ORGANISING COMMITTEE
Shri 3. Sen, 6ARC - Chairman Shri R.K. Garg - IRE Ltd. Bombay Shri M.K. Batra - UCIL Jaduguda Shri J.L. Bhasin - UCIL Jaduguda Shri K. Balaramamoorthy, - NFC Hyderabad Shri P.R. Roy - BARC Shri A..N. Prasad - BARC Dr.R.M. Iyer - BARC Shri T.K.S. Murthy - IRE Ltd. Prof. S.L. Narayanamurthy - IIT, Bombay Shri T.A. Menon - FACT, Cochin Comdo. K.C. Chatterjee- - OCL Bombay Shri CM. Das - NPC, Bombay Dr.C.K. Gupta - BARC Shri K.S. Kop'piker - BARC Dr. Ashok Mohan - BARC Dr.V. Venkat Raj - BARC Shri G.R. Balasubramanian - IGCAR, Kalpakkam Dr.S. Viswanathan - AMD,Hyderabad Shri M.R. Balakrishnan ' - BARC Shri U.R. Marwah - Member-.Secretary
1 TECHNICAL COMMITTEE
1. Shri K.S. Koppiker, BARC - Chairman 2. Shri V.S. Keni, BARC 3. Shri S.K. Chandra, IRE Ltd, Bombay 4. Dr.T.K. Mukherjee, BARC 5. Dr. A. Rarnanu jam, BARC to. Shri S.N. Bagchi, BARC 7. Shri U.R. Marwah, BARC - Member Secretary The National Symposium on Uranium Technology was held at BARC, Bombay during Dec. 13-15,1989 under the auspices of Engi neering Science Committee of Board of Research in Nuclear Science, Department, of Atomic. Energy.
In the context of expanding nuclear power programme in India, the need for production of large quantity of uranium fuel from indigeneous'resources hardly needs any elaboration. After three decades of experience in :this area, it is thus appropriate to pool together the expertise and experience gained in the country during this period in various aspects of uranium technol ogy like exploration, mining, ore processing, refining and con version to. oxide and metal. This symposium has provided an opportunity to the uranium technologists to interact and exchange their experience and plan future strategies.
Being the first symposium on this toppic the response has been excellent as evidenced by the extensive participation and contribution of papers covering almost all aspects of uranium technology. Altogether 69 papers, including invited lectures have been presented and the proceedings, have been brought out in two volumes.
It is our hope that this technical coverage of the pro ceedings and panel discussion would serve as a valuable reference material for uranium technologists in the coming years; We wish to record our sincere thanks to the members of the organising and various other committees, authors of invited' lectures, and contributed papers, and panel members for making it possible to bring out this proceedings. The cooperation extended by Head, Library and Information Division, BARC is gratefuly acknowledged.
TECHNICAL COMMITEE CONTENTS
INAUGURAL SESSION
Introductory Remarks by Shri S. Sen, Charlman, Organizing Committee
Welcome Address by Dr. P.K. Iyengar, Director, BARC
Presidential Address by Dr. M.R. Srinivasan, Chairman, AEC
Inaugural Address by Dr. H.N. Sethna, Chairman, TOMCO & Tata Electric Compani
Vote of thanks by U.R. Marwah, Member Secretary, Organizing Committee
Keynote Address by .Shri R.K. Garg, CMD, IRE Ltd.
TECHNICAL SESSION I Plenary Lectures
Uranium mining in India - Past, present and the future M.K. Batra, Adviser, UCIL, Jaduguda.
TECHNICAL SESSION II II A Uranium Prospecting
Contributed Papers Structure as a guide for uranium exploration in the Turamdih-Mohuldih area, Singbhum Dt.Bihar. R. Mohanty, M.B. Verma Prospecting for uranium in carbonate rocks of 19 Vempalli formation, Cuddappah Basin, Andhra Pradesh M. Vasudeva Rao, J.C. Nagabhushana, A.V. Jeygopal and M. Thimmiah
Evaluation of favourable structural features for 36 uranium from airborne geophysical surveys over parts of Madhya Pradesh K.L. Tiku, S.V. Krishna Rao and Bipan Behari
Integrated geophysical investigation for uranium 49 A case study from Jamini, West Kamerg Dt. R. Srinivas, J.K. Dash, S. Sethuram, K.L. Tiku, Bipan Behair K.L. Tiku, Bipan Behari i
Natural thermoluminescence of whole- 74 rock as a potential tool in the exploration for sandstone type uranium deposits: Application to the lower Mahadek sandstone of Meghalaya R. Dhana Raju, R.C. Bhargava, A.Paneerselvam and S.N. Virnave
Hydrogeochemical exploration for uranium: 90 A case study from the Cuddappah Basin, Andhra Pradesh R.P. Singh, P.K. Jain, B.R.M. Kumar, S.S. Rao, A.V. Patwardhan and S.G. Vasudeva - An Alpha-gamma counting integrating device 109 for uranium exploration G. Jha, M. Gaghavayya, M.N. Srinivasan', S. Sastry.
Geostatistical study of Bhaten ore deposit 126 C.V.L. Bajpai and P.P. Sharma II B Analytical Techniques in Uranium Technology
Uranium analysis using an on-lone background 147 correction programme with carrier-distillation technique by a computer controlled spectrometer R.K. Dhumwad, A.B. Patwardhan, V.T. Kulkarni, K. Radhakrishnan
Determination of trace metals in uranium oxide 132 by ICP-MS S. Vijayalakshmi, R. Krishna Prabhu, T.R. Mahalingam and C.K.. Mathews
Development of flow-injection analysis technique for 157 uranium estimation A.H. Paranjape, S.S. Pandit, S.S. Shinde, A. Ramanujam and R.K. Dhumwad
Standardisation of DC Arc carrlr-distillation 166 procedure on a direct reading spectrometer for the determination of B, Cd etc. in nuclear grade uranium S.S. Biswas, P.S. Murty, S.M. Marathe, A. Sethuinadhavan V.S. Dixit R. Kaimal and A.V. Sankaran
Spectrographs determination of B,Cd and NI in 182 magnesium fluoride A. Sethumadhavan, V.S. Dixit and P.S. Murty.
.j Estimation of uranium in leach liquors of low 189 iron content: Modification of a spectro photometry method using 4-(2-pyridil a 20) resorcinol G. Suryaprabhavati, Leela Gopal, G.S. Chawdary and Radha R. Das.
Discussions 200 TECHNICAL SESSION III III A Mining and Ore Benefeciation
Contributed Papers 204
Development of mining at Jaduguda J.L. Bhasin
Role of support services at Jaduguda uranium mine 232 S.D. Khanwalkar, V.N. Radhakrishnana, M.N. Srinivasan, Pinaki Roy, S.N. Bannerji
Recovery of uranium concentrate from copper tailings 254 S. Chakraborty, U.K. Tiwari and K.K. Beri
Significance of petrology in the ore processing 284 • technology with special reference to the uranium processing from the copper tailings of Singbhum Thrust Belt N.P. Subrahmanyam, T.S. Sunilkumar, D. Narasimhan and N.K. Rao
Improved gravity flow sheet for the recovery of 300 uranium values from the copper tailings R. Natarajan, R.S. Jha, U. Sridhar, N.K. Rao
Magnetic separation for preconcentration of uranium 318 values from copper plant tailings R.S. Jha, T. Srinivasan, R. Natarajan, U. Sridhar and N.K. Rao
Preliminary beneficiation studies on liranium ore 332 from Tummalapalli, Andhra Pradesh N.P.H. Padmanabhan, U.Sridhar,.N.K. Rao
, Discussions 3« TECHNICAL SESSION HI III 11 Analyt.ie.-il Techniques in Uranium Technology-11
Contributed Papers 349
Rapid determination of uranium in uranyl nitrate solution by Gamma Spectrometry T.K. Shankaranarayanan and D.S. Gupta
Modification of fluorimetric method of uranium 356 analysis for Jaduguda Plant Samples A.B. Chakraborty and V.M. Pandey
Determination of uranium in sea war t* by 369 adsorptive differential pulse voltametry R.N. Khandekar and Radha Raghunath
Difficulties in preparing a standard sample 376~ of uranium metal having traces of nitrogen R.S.D. Toteja, B.L. Jangida, M. Sundaresar.
Estimation of manganese In tailings plant 382 effluents by ICP-AES Joydeb Roy and V.M. Pandey
Voltamnetric studies of uranium (VI) 389 reduction i l
Discussions 402 TECHNICAL SESSION IV Uranium Ore Process Technology
Invited Lecture Technologies for processing low-grade uranium 403 ores and their relevance to Indian Situation T.K.S. Murthy
Contributed Papers ,
Jaduguda Uranium Mill-Rich experience 431 for future challenges K.K. Beri
Grinding and leaching characteristics 463 of the Indian uranium ores. V.M. Pandey and R.U. Choudhary
Recovery of uranium by direct low-acid 477 leaching from copper concentrator tailings V.M. Pandey, R.U. Choudhary, A.K. Sarkar, A.P. Bannerjee, A.B. Chakraborty, N. Malty
Selection of ion exchange resin for uranium 485 adsorption from Jaduguda leach liquors D.P. Saha and V.M. Pandey
Apprlication of advanced technologies for uranium 498 mining and processing at Narwa Pahar and Turamdih Projects R.C. Purl and R.P. Verma
Impounding of tailings at Jaduguda - 528 Planning, design and management of tailings dam S.N. Prasad and K.K. Berl TECHNICAL SESSION V Uranium Ore Process Technology-contd and Byproduct Uranium
Contributed Papers
Nuclear pure uranium from ores using weak 555 base ion-exchange resins S.V. Parab, S.S. Gharat, G. Cherian and K.S. Koppiker
Development of an integrated process for recovery 570 of uranium from ore and its refining at the •• location of new uranium mill at Turamdih R.A. Nagle, S.V. Parab, S.S. Gharat, A.P,. Giriyalkar and K.S. Koppiker
Preparation of nuclear grade uranium oxide 582 from Jaduguda leach liquor V.M. Pandey, A.B. Charkraborty and N. Malty
Uranium recovery from phosphoric acid 592 G. Sivaprakash
On-site tests for recovery of uranium from wet 621 process phosphoric acid at FACT H. Singh, R.A. Nagle, A.B. Giriyalkar, M.F. Fonseca and K.S. Koppiker
Recovery of uranium from nitro-phos acid °28 R.A. Nagle, A.B. Giriyalkar and K.S. Koppiker
Recovery of uranium from monazite - a fresh 635 look at the current practice S.L. Mishra and K.S. Koppiker Recovery of uranium from sea water- 643 A, laboratory study D.V. Jaywant, N.S. Iyer and K.S. Koppiker
TECHNICAL SESSION VI Uranium Refining
Contributed Papers
Operating experience in the refining of uranium by 653 solvent extraction using mixer settlers SMT. S.B. Roy, H. Singh, K. Kumar, A.M. Meghal, V.N. Krishnan, K.S. Koppiker
CALMIX-Innovative mixer-settler system 659 C.K.R. Kaimal, B.V. 'Shah, I.A.. Siddiqui, S.V. Kumar
Precipitation of ammonium diuranate-a study 666 T.S. Krishnamoorthy, N. Mahadevan, M. Sankar Das
Continuous reactor system for precipitation of 688 uranium from uranyl solution \ I.A. Siddiqui, B.V. Shah, S.H. Tadphale, S.V. Kumar
Preparation of metal grade uranium trioxide 695 through ammonium diuranate precipitation route S.R. Ramachandran, P.D. Shringarpure and A.M. Meghal
Studies on preparation and characterisation of 701 ammonium uranyl carbonate (AUC) V.N. Krishnan, M.S. Visweswariah,. P.D. Shringarpure and K.S. Koppiker
Batch precipitation technique-a process for 708 UO2 powder procution. A.K. Sridharan, G.V.S.R.K. Somayaji, N. Swaminathan and K. Balaramamoorthy Development of AUC route for production of U0_ Powder 712
U.C. Gupta, Smt. Meena R. and N. Swaminathan
Analytical technique in uranium dioxide 728 fuel production stream. T.S. Krishnan, S. Syamsundar, B. Gopalan, R. Narayanaswamy and C.K. Ramamurthy
TECHNICAL SESSION VII Uranium Metal Production
Contributed Papers
Improvements in process technology for 750 uranium metal production at UHP A.M. Meghal, H. Singh, A.V. Vedak, K.S. Koppiker
Improvements in equipment design for 756 hydrofluorination of U0_to UF, A.V. Vedak, R.N. Kerkar, and A.M. Meghal
Magnesio-thermic reduction of Ufy to 762 uranium metal - plant operating experience S.V. Mayekar, H. Singh, A.M. Meghal, K.S. Koppiker
Recovery of uranium from magnesium fluoride 770 slag at UMP P.K. Bandopadhyay, B.M. Shadakshari, H. Singh and A.M. Meghal
Future trends in the processing of 777 uranium slag generated suring production of uranium metal. Keshav Chandra, Mahesh Singh, II. Singh, A.M. Meghal, K.S. Koppiker and S. Sen Quality assurance during uranium metal production at 790 UMP V.N. Krishnan, R.D. Shukla, M.S. Visweswariah
Novel surface chemical treatment to improve the 796 quality of scintered U0_ pellet B. Venkataramani and R.M. Iyer
P.C. based uranium enrichment analyser 805 V.K. Madan, K.R. Gopalakrishnan and B.R. Bairi
Discussions. 810
TECHNICAL SESSION VIII
Environmental aspects, Health & Safety
Contributed Papers
Treatment of uranium tailings vis-a-vis radium 811 containment P.M. Markose, K.P. Eappen, M. Raghavayya, K.C. Pillai
Radon problems in uranium industry 833 A.H. Khan and M. Raghavayya
Effective dose evaluation of uranium mill workers at 848 Jaduguda G. Jha and M. Raghavayya
Radiological and environmental safety aspects of 857 uranium fuel fabrication plants at Nuclear Fuel Complex at Hyderabad S. Viswanathan, B. Surya Rao, A.R. Laxman and T. Krishna Rao Limits of plutonium contamination in reprocessed 865 uranium for handling in natural uranium plants V.K. Sundaram and M.R. Iyer
Biosorption of uranium by yeast 874 A.K. Mathur, N. Muralikrishna, V. Krishnamurthy and R. Sankaran
Discussions 885
TECHNICAL SESSION IX Health and Safety Aspects-contd General Chemistry of uranium technology
Contributed Papers
Operational health physics experience at uranium 888 metal plant, Trombay P.P.V.J. Nambiar, Pushparaja, J.V. Abraham
Radio activity levels in the process streams of 897 uranium metal plant (UMP) at Trombay Pushparaja, S.G. Sahasrabhude, J.V. Abraham and M.R. Iyer
Radiological and conventional safety aspects of 902 machining operations of uranium ingots V.B. Joshi, I.K. Oomen, S. Sengupta, T.S. Iyengar
Radiation risks, medical surveillance programme and 910 radiation protection in the mining and milling of uranium ores Dr. A.K. Rakshit
Separation of uranium VI, Chromium and zirconium by 924 solvent extraction with crown ethers N.V. Deorkar and S.M. Khopkar Uranyl ion transport across tri-n-butyl phosphate-n 939 dodecane liquid membranes J.P. Shlkla and S.K. Mishra
TECHNICAL SESSION X Project Management
Invited Lecture
Consultancy, project engineering service for the 947 uranium industry A.K. Bhattacharya, Vice Chairman DCL
Contributed Papers
Project Management-proglems in execution • 958 D.G. Nair, FACT
Conditions required for opening of a commercial 985 mineral deposit S. Sastry, UCIL Management of uranium mining and process wastes at 998 Turamdih Project R.C. Purl and R.P. Vernia, UCIL
TECHNICAL SESSION XI
Panel discussion on A021
"Present status and future strategies on uranium technology" SESSION IV
URANIUM ORE PROCESS TECHNOLOGY
Chairman i Shri D.V. BHATNAGAR (Retired) UCI.L Be po rt e'ur * S hr i FARVT NDERPAL SI N5H B A R C - 403 -
TECHNOLOGIES FOB PROCESSING LOW-GRADE URANIUM ORES AND THEIR RELEVANCE TO THE INDIAN SITUATION
T.K.S. MURTHY
1. IDEA OF A MINERAL RESOURCE
There are many ways of defining a ' mineral resource* but none- of them are unambiguous. One of them which is easy to comprehend is that given by Cohen a mineral resource must contain useful oomponents (uranium in the context of this, talk) thai can be extracted by means of 'available technology' and at a 'reasonable cost'. The first condition is selfevident. In many cases the decision as .to what is a'reasonable cost' is not an easy one and can be quite involved. In the case of many metals there exists some prevailing world-market price and that can indicate the 'reasonable cost'.However,there are many cases where a nation -may for socio-economic arid political reasons, produce .•a ' metal for domestic consumption or even for export even if its cost of production, is above the 'reasonable cost'. In India about 50V of the copper and zinc demand is -met by indigenous production but at a unit cost almost double -. 404 -
of that prevailing in the world market. In the case, of uranium this criterion is all the more difficult to adhere to. The reasons are many. The number, of producers and consumers are too few to evolve any thing like a free-market price. Uranium finds its main application as fuel in nuclear reactors. The possibility of its use in nuclear explosives makes uranium a material of great strategic importance. The absence of bilateral or multilateral safeguard agreements, or other governmental approvals, therefore, comes in the way of certain producer countries from supplying uranium to some consumer .countries. Hence, the question of 'reasonable cost' has different connotations in different cpuntries.
2. URANIUM RESOURCES OF THE WORLD
' The OECD Nuclear Energy Agency (NEA) and IAEA periodically publish information on the uranium, resources of the world, outside the centrally planned economies. These resources are divided into 'Reasonably Assured' and .'Estimated Additional* categories. Under each category, : again, there is a subdivision according to the estimated cost of..production of concentrates by application of -available technologies. These subdivisions are: (a) less than $ 80 per • . ' per kg U; (b) between.$80 and 130/kg U and (c) above $130'per kg. U. Broadly speaking, category (a) is good for exploiting under the present, demand conditions; category (b) are - 405
standby reserves which can be exploited if tho mounting demand for uranium pushes up its price and category (c) is essentially a speculative type which is unlikely to be exploited in the foreseeable future. Keeping in mind the relatively high cost of uranium that is likely to be produced from presently known ores in this country the NEA-IAEA data are given in an abridged form in Table-I, mentioning reserves upto a cost of $ 130 per kg U.
The countries of the world can be grouped into five broad categories depending on their uranium resources and consumption:
(i) those' with large/rich ore reserves and relatively large consumption, e.g..U.S.A.They may be exporting as well as importing concentrates;
(ii) those with relatively large resources but • with moderate or no' consumption, e. g., Canada, South Africa, Niger and Namibia. They are essentially exporters of uranium;
(iii) those with, relatively low or practically no uranium resources but with major nuclear energy generation plants, e.g., Japan and West Germany. They are major' importers of uranium. - 406 -
(iv) those countries with moderate uranium resources and moderate consumption, e.g., India and Brazil.
Over 85% of the uranium resources are concentrated in a relatively small number (five to six) of countries. A point • which has not been adequately brought out in Tabla-I is that the uranium grades and resources - of the present major producers enable them to obtain uranium at much lower cost than the $130 per kg U limit which is chosen for this table. In fact by the end of 1988 the NUEXCO spot prices were quoted as low as $30 per kg U. Another point of relevance is, though .the.total quantity of Indian resources is' about, the same (about 60,000 t U) as given in NEA-IAEA report, the present indications are that.most of it falls in the price range much higher than $130 per kg U. It may be closer to $200 or Rs.3000-5000: A third point of significance is 'that,' though on a 'global basis the . presently known resources of low to medium cost uranium -are adequate to: meet the projected demand for this important nuclear raw material for the foreseeable future-it is perhaps of not. much' i • . solace to countries .like India which are not signatories to the 'Non-proliferation treaty* and cannot freely draw on the world surpluses. :They have to depend on indigeneous production only'. - 407 -
3. TECHNOLOGY OF URANIUM ORE PROCESSING
The hydrometallurgical technology for uranium ore processing and production of the yellow-cake was well established by mid 1950's to early 1960's. A near selective leaching by dilute sulphuric acid digestion of the uranium bearing minerals in a variety of ores all over the world is at the root of the .hydroraetallurgy of uranium. The sulphuric acid leaching is an entirely post world-war II development. Two revoluntionary techniques have played a vital role in establishing sulphuric acid leaching as an economically successful and universally applicable ' operation for processing low-grade uranium ores. They are the resin ion.-" exchange (IX) process introduced in early. 1950's and the liquid ion-exchange or solvent extraction, process (SX) in 1960's. A simplified flow-sheet which is broadly followed in most of.the operating plants in the world is shown in Fig.l. Though there are many variations of the techniques and types of equipment used for carrying out each of the unit operations like size-reduction, leaching, concentration'arit) final 'precipitation and recovery of yellow cake, the general flow-sheet has not undergone any profound changes over the
years.: During the past 30-40 years, of its existance .the
i uranium industry was rocked by wild variations of concentrate price.. It was as high, as $110 per. kg U in the lata 70's, - 408 -
following the global oil crisis, but fall to $50 In early 80's and Is, at present, as low as $30. The major uranium producers coped with the falling prices not so much by adopting revolutionary technologies, except in a few cases, but by more cqraraonsense measures such as increasing cut off grades in the mines, reducing capital costs etc.
4. SIGNIFICANCE OF TECHNOLOGICAL IMPROVEMENTS ON REDUCTION OF COST OF CONCENTRATES
The economic viability of a deposit depends on tonnage of ore in the deposit, its grade, ore body characteristics which define the mining capacity, and cost of mining, roineralogical characteristics which decide the process parameters and the physical location which play a role in overhead and transportation costs etc. It is not possible to discuss in detail, in a brief talk like this,-the quantum of effect that each one of these factors has on the final cost of concentrate production. As a rough indication it can be.said that in modern- industrial countries the cut off grade for uranium ores is 0.IX U if the mill is nearby and 0.2% if it is more than 200 km from the mine. When foreign ventures are started in industrially underdeveloped, countries ' like Africa the value is .set at even 0.3% due to poor infrastructural and technological support available. - 409 -
4.1. Subeconoroic ores, Particularly' those in India
The subeconomic ores may fall in any one of the following categories : (i) Grade normal but; deposit is small, (ii) Grade not; "too poor, deposit not too small, but distance from the mill is too large, {iiiT" Grade quite poor but deposit is very big, (iv) Poor grade. deposit not big but exploitation is obligatory.
To bring each of these categories into an economically exploitable range different strategies have to be adopted. It is beyond the scope of this talk to cover all these alternatives. The. presently known deposits in this country fall in category (iv) . .Generally,the grades are in the range 0.04-0.07% 03O8. Average uranium available from a deposit is 5,000-10,000t U308 -*nd the total proved and anticipated reserves, are about 8Q,GO0t U308. As,a consequence of all these factors it is over optimistic to expect any amount of R&D effort to. bring those ores to a category'.that would yield uranium ctonaontrates at rates anywhere near the prevailing world market rates or even within the! range of $C0 per kg U:
But some improvements in the processes at present a being. practiced in this country and adoption of better equipment in
i some of the unit operations holds hope of significant though not dramatic decrease in production cost. - 410 -
4.2. Contribution of Different operations to uranium Production cost
To understand the impact of technological modifi cations on the cost of processing low-grade ores it Mill be Helpful to have a rough idea of the contribution of each of the unit operations involved in the process towards the overall cost. Table-II gives the distribution of capital and operating costs for individual operations. Columns 1,2,3)4 provide the percentage contribution of each item. However, the contribution from mining is not included in the original set of figures. The economics of milling can never be .considered in isolation as mining costs alone can place' a project in a cost category where no amount of innovation in process technology can render the project attractive or profitable. To bring out this aspect, in colum(5) of Table-II rough estimate;of. the distribution of costs including mining are given. They; are to be taken as indicative and are given' only to. help, further discussion of the impact of alttornative strategies. it is important to . note that where, deep underground reining of the ore is involved about 40% of the concentrate cost is accounted by the mining operation itself.
4.3. Making Subeconoraic ores workable
Thara ara several ways of taoklirig the problem of
subeaonoraia uranium ores arid make them workable. They are J - 411 -
(i) increasing the cut off grade; (ii) reducing the feed mass by selective rejection of gangue mass, that is. by adoption of physical benefi- ciation techniques; (ill) Estimation of one or more steps in the general flow sheet for ore-processing'(Fig.1); (iv) Modification of process techniques or equipment; (v) Recovering valuable byproducts from uranium ores; (vi) Obtaining uranium as byproduct.
It is now proposed to consider if there are any cases, in the world, of uranium ores which are comparable to the type found in India (in terms of grade, . tonnage and mining .characteristics) yet processed economically. Then the possibility of taking advantage of any of the alternatives (i)-(yi) to bring down the cost of production of uranium in India.can be looked into.
5. CASES IN THE WORLD THAT ARE SIMILAR TO THAT OF INDIA
There is \ hardly any case in the world where an ora similar, to that existing in India (with respect to grade and.tonnage of uranium) are processed and at the same time uranium is produced at an attractive price. The cut off grade in most of the operating mines is above 0.1% U308. s - 412 -
For example, Jabiluka I and II ore bodies in Australia contain total uranium reserves of 207,400 t grading 0.39% U308. Ranger is another big open pit mine with grade ranging 0.15-0.4% U308 with a capacity to produce about 3,000 t U308 annually. There are even cases like the recently opened Cigar Lake deposit in Canada with reserves, of 180,000t U308 at an average grade as high as 14.6% U30B. These cases are all in direct contrast to the low-grade and small deposits of this country.
However, a case, somewhat akin to that of India, is the uranium deposit in the black shales in Sweden. Even here the similarity does not extend beyond the grade. The grade in Sweden is 0.034% U308 but the resources are very large and. the plan was to start a mill to produce about 500t U per year. However, due to environmental constraints large saale mining could not be undertaken and the deposit remains unexploited.
The other case of somewhat similar nature is that of Rossing deposit in Namibia. The average grade is 0.03-0.05% U3O8, but the uranium reserves are vast—151,000t U308. It is an open pit mining at a rate of 35,000t per day and concentrate production capacity of 3,500t U308 per year. The process followed is a standard sulphuric acid leaching, at relatively coarse size, Poter continuous IX unit for uranium aonaontration, followed by ELUEX proaess. Namibia is one of the major exporters o£ uranium at-competitive prices. - 413 -
From these two cases, particularly that of Rossing, it appears that open pit mining and large throughputs are essential to make low-grade deposits economic. This is in contrast to the deep underground mining and low throughputs, 1,500-3, OOOt ore per day, which the mines in India can sustain.
6. OTHER MEASURES FOR REDUCING COST OF PRODUCTION Let us now consider the feasibility of measures (1)- (iv) suggested in para 4.3. (i) Increase the cut-off grade : In view oflow average grade of the ores and relatively small deposits it will not be possible to raise the cut-off grade in any mine without seriously sacrificing the available resources.
(ii) Adoption of Physcial Beneficiation Techniques : In principle it is possible to obtain a high grade concentrate of any metal from a low-grade ore by using one or more of the physical beneficiation techniques. If successful, such a process reduces the tonnage of material to be put through tho ahomioal recovery proaess and hence makes the overall metal production highly economic. In its early years of growth uranium industry carefully investigated this option. However, except in isolated instances, it was not possible to achieve a high recovery and a high concentration ratio at the same time. Even those few successful cases are those in which uranium ia obtained as a byproduct and not as - 4H -
a primary product^ So far, the investigations on ore dressing 'of ores mined or likely to be mined primarily for uranium in this country have not yielded any better results. In view o£ our limited uranium resources any major sacrifice of recovery (above 20%) to obtain a high concentration ratio (mora than 4) may be unacceptable. If, on the other hand, a small concentration factor (of the order of 2) is obtained with high recovery (near 90%) the economic benefit may be doubtful. This is clear from the cost distribution shown in Table-II. The mining, comminution, precipitation (of concentrate) and drying, tailings disposal and lab. and services which account for about 75% of the cost of the final product remain the same whether the mined ore is directly subjected to chemical leaching or. it goes through a beneficiation step and the concentrate is leached.. In case only a low concentration factor (not beyond
2-3 fold)>is achieved, out of the remaining 25%: of the cost '0% only a marginal saving not more than / can be . expected. Therefore, if the loss of uranium in the precpncentration step exceeds 10% not only does it affect the precarious resource situation it will also make the uranium costlier than that obtained by direct leaching.
(iii) Elimination of some Steps in the Flow-sheet '•
Under favourable conditions some.of the steps in the : general flow-rahest (Fig.l.) oan be eliminated or modified, ljeading to reduction in aapital and operating - 415 - •
.costs. Considering from the product end, if the IX system can ifunction with unclarified leach liquors there, will, be some "saving of cost. This has been demonstrated to be feasible if a fluidized. bed continuous ion-exchange system (CIX) is adopted as in the Rossing mine, Namibia. However, continuous IX systems are usually recommended where the throughput of the solution (and uranium) is large. If the IX or SX can be carried put on leach pulps, without solid-liquid separation the cost advantage is significant. However, inspite of. a number o^f investigations carried out in Canada and elsewhere no industrially successful system.has yet been evolved. The. ' ore characteristics and economics-] have made .agitation leaching of.finely ground ore the most widely used extraction technique for uranium. . With a view to reduce the processing costs modified leaching operations are carried out in some parts of the world! Among them are—'•- vat leaching where a coarsely ground ore (1-3 mm size) contained,in .vats is leached by spraying"of acid, heap leaching where ore crushed to 50-100 mm size is leached by .intermittant or continuous spraying of acid, in-place leaching : Where the "shattered' ore is leached.in the minerstopes by. spraying' or .flooding with acid and finally solution mining where the leaching . reagents are injected into the mineral '.deposit through specially prepared injection wells and 'the percolated •• • ' solution,, is pumped out.through collection wells. In -this - 416 -
case the ore is not physically removeed from where it is naturally placed. Vat leaching, heap leaching and in-place leaching, if feasible, confer increasing order of economic benefit. However, their success- depends on sufficient degree of porosity and permeability to reagents" existing in the ore. As a general rule the time required for leaching uranium increases (from days to years) and its recovery decreases in the order—vat leaching, heap leaching, in-place leaching.
In 1977 a large in-place leaching operation was started in the Agnew Lake mine, Canada. The ore was broken in the underground stopes by blasting and the leach solutions were applied by spray pipes at the top of the stopes. The liquors uere collected by a sump system below the' stopes and pumped to the surface where the uranium was recovered by IX. The test work indicated that about 70X of the uranium could be recovered in one year from the ore which averages 0.05% U30B. This recovery,, however, was not achieved in the large scale run consisting of llO.OOOt ore due to the larger than expected particle size of the fractured rock. The operation was finally discontinued after about 18 months. In the case of most of the ores from Slngbhum area (Bihar) it is observed that they consist of hard rock with low porosity and hence none of these methods can be used
i without severly sacrificing the recovery. r Solution mining (KOM times called in-situ leaching ISL)
is limited to only one type of uranium deposits, a roll front - 417 -
(deposited from moving ground water) deposit in permeable sand stones that roust be on an aquifer (sand stone filled with water). It is difficult to predict the performance of an ISL project. There have been a great many more unsuccessful than successful ISL projects. The unconsolidated sand stone deposits in U.S.A. have been found to be amenable to leaching by this technique. The first commercial ISL operation was carried out by Utah Construction and Mining Co., in U.S.A. With fall in uranium prices more and more operations of this type are launched in U.S.A. supplying, at the present time, about lOOOt. of uranium which is about 20% of total U.S. output.
Though solution mining has the potential to produce uranium at lower costs than other mining methods, the major deposits in this country (in Bihar) are not of the type that can be treated by this technique. . However', it' is to be seen if the recently found sand stone type deposits in ' Meghalaya and elsewhere can ever be brought under this mode of extraction. (iv) Modification of Process Techniques or Equipment:
In some of the operating uranium mills autogeneous or semiautogeneous grinding of the ore is practiced in place of ball milling. However, this method has succeeded only in the case of sand stone ores. A few examples where high speed thickeners and belt fiters have been used with advantage for solid-liquid separation, following acid leaching of the ground ore ar* described in literature. Loss of soluble - 418 -
uranium 'from leached pulps is reported in many cases to be less than 2%. However, in the Jadugoda uranium mill this loss is 6-8%. There is a strong case to study this aspect in detail and reduce the soluble loss to 2% in the present and future' mills.. Assuming a mill capacity of 2,000t ore per day, 300 working days, average 0.05% U308 and cost of production Rs.4,000 per kg U308, the impact of reducing soluble loss of uranium by 5% can be approximated as a saving of Rs. 6 crores per annum. Scope of Producing a high grade Product : ——————————————^———————————————— ————— The yellow-cake produced normally in the Uranium mill needs refining by solvent extraction (TBP) to make uranium However, it is pessibl* fo p*oeU*Ct. -nuclear g-ra-Ot. -ma/cytai i-b >ntli &t*' oxide of nuclear purity;/ For example,, the well known BOFFLKX process followed in some of the South African plants consists of the following main steps, starting with the ore leach liquor: IX concentration of uranium, elution of loaded resin with 10X sulphuric.acid, extraction of the uranium from the eluate by a tertiary amine like Alaroine-336, stripping of the' loaded amine with a strong solution of ammonium sulphate at pH 4.5 and final: precipitation of uranium as ammonium diuranate .(ADU). The ADU, thu3 obtained, is described a3 'of near nuclear purity'. ; While this material may be eminently suited for conversion.to UF6, for isotopic enrichment, it may not meet the purity specification of natural oxide or. metallic fuel. The Uranium Extraction Division of B.A.R.C. carried out;.laboratory investigations on a process which includes a TBP refining of uranium from the nitrate strip - 419 -
solution of the loaded amine. The final ADU produced by this route is expected to be of nuclear purity. By adopting such an integrated recovery cum refining process it is anticipated that a reasonable saving in the cost of uranium oxide can be achieved. (v) Recovery of. valuable byproducts from uranium ores:
There are very few instances in the world in which the uranium ores yield byproducts in such quantity and of such value that,the economy of uranium production is materially improved. In this country also none of the known ores have this advantage. Though some copper, nickel and molybdenum are presently recovered as byproducts from the Jadugoda ore the operation is not profitable enough to benefit uranium production. (yi) Obtaining .uranium as byproduct: Producing'uranium as byproduct is; a potentially attractive proposition. There are many striking cases in the world in which this' strategy is successfully implemented.
Olympic' Dam "• One of - the recent and vary, successful byproduct uranium recovery is carries out in the Olympic Dais • Project (Australia). The deposit carried 2,000 million tons of ore bearing 1.68% Cu, 0.06% U308, 0.6g per ton gold and some silver.: The mining rate is 20,000t ore per'day and the annual production is 3,000 t U308,. 150,000t Cu, '3,400 kg Au and 23,000 kg Ag. . - 420 -
Phalaborwa : Phalaborwa deposit in Northern Transval . (South africa) is another example of a successful byproduct uranium production. The main products are chalcopyite and baddeleyite - (Zr02). The abridged flow-sheet shown in Fig.2 also, brings'out the success of the beneficiation technology which upgrades uranium present at. a level of 0.035X U308 in the copper tailings to a uranothorianite concentrate bearing 3.OX U308. At a throughput of 330.000t ore per day Phalaborwa is one of the major operations of this type. South Africa : In South Africa•practically the whole of the 4,000 t uranium produced annually is derived as byproduct from gold ores. .The Witwatersrand Group of mines which account for about 60% of the world's gold production carries 0.02-0.03 X 0308. The first mill.to recover uranium from gold •tailings was commissioned at West - Rand Consolidated mines in 1952. By 1957 a total of 17 plants had been erected leading to a production of about 3,800 t U308 per annum. The flow sheet followed is more or less the sane as the one in Fig-1. The main steps are : Milling-.. >Cyanidation——>Acid leaching—> Flo tat ion (Fine) For Gold For Uranium For Pyrite
Later a reverse leach process in which uranium is first leached was introduced in some mills :
Milling >Acid Leaching—>Cyanidation ^Flotation (Fine) For Uranium For Gold For Pyrite The second process led to better gold recoveries as the acid laaoh exposed the gold particles for a better cyanide attack. - 421 -
Byproduct uranium, though in considerable amount, contributes only a small part of the profits of the combined gold-uranium operations.
A considerable amount of gold tailings material is still available from the preuranium extraction era. These slimes for which the cost of mining and comminution have already been borne, have become an important feed stock. In the Ergo Project about 380 million tons of tailings material, with an average grade of 60 ppm (0.006%) U308 is being processed at the rate of more than 1.5 million tons per month (a.very high throughput). 50,000 tons of slimes per- day is pumped 9 km to the plant. 3% of the material is removed as flotation concentrate, which carries a major portion of the uranium, before the balance is pumped a further 11 km to a new tailings disposal site. Apart, from the demonstration ' of successful byproduct uranium recovery the operations involve a massive transport of the- material to and from the mill and a highly efficient flotation process which concentrates uranium by a factor of 25.
Phosphate rock ' A vary important source of byproduct uranium in many parts of tha world is tha .wet-process phosphoric acid which derives 100-150 ppm of uranium from the raw material, rock phosphate. U.S.A. has been a pioneering country in development of technology for uranium recovery from this source. The motivation is easy to understand. The phosphate deposits of that country alone contain 4 million - 422 -
tons of uranium. By 1980) about. 27' million tons, of rock was converted into wet-process acid, annually,, setting, the potentially recoverable uranium at 3,000t. per /ear.. After 2- 3 decades of research a' very effective', so 11 vent, extraction process has emerged. In. this, process a mixture of di 2-athyl hexyl Phosphoric acid (DEHPA) and trioctyl phosphine oxide (TOPO) in kerosene is used as a. selective extractant for uranium. After a reductive- stripping, from, the loaded solvent - uranium is concentrated and! it, is. subjected! to> a. second cycle of extraction and final stripping; with' ammonium' carbonate to> obtain ammonium uranyl carbonate. (AUG)) whichi oni calcination gives an, oxide with about. 90%. UL By 1982'. eight byproduct uranium recovery plants were' in< operation) im U'.S'.. with, ant installed capacity of about. l.OOOt 07 per- year.. Other countries-like France, Belgium, Spain-, Canada, and! Yugoslavia, are also reported to have set. up> plants: for- Che- same- purpose..
Scope of Byproduct Uranium in India: There La only a limited scope for byproduct uranium. The raonazite that, is. being processed in the. Bare Earths, plant, of IRE! carries, about. 005% U308. At present a small part of it. ('less; than 25%)) is recovered. By modifying the process flow sheet- for- roonazite there is .a possibility of obtaining about. IQt of! (1308! '. annually as byproduct. But it needs considerable* amount of. development work' to mobilise) this quantity.. . Another source is the' copper' flotation! plant! tailings £m Sinabhum area. The three.plants at MbaabahiL,, Surdw and Rakha •! generate about 5,000t j?er- day: o£ tailings; wi.thi .*» avaxag* - 423 -
U308 content of 0.009%. The case is similar to that of the gold tailings in South Africa., though the scale, of operation is much smaller. Over the past 2-3 decades several investigations have been carried out in D.A.R.C. and U.C.I.L. to upgrade the uranium content of the copper tailings by physical beneficiation using gravity based methods. It has been found possible to obtain concentrates bearing close to v 0.1% 0308 but the recovery is about 40%. These concentrates are at present being processed in the Jadugoda uranium, mill for uranium recovery. Investigations also revealed that by direct sulphuric acid leaching over 70% of the uranium can be recovered from the tailings. this alternative is being considered by the Department of Atomic Energy. - An attractive proposition Mill be pumping the tailings from the three plants to a centrally located, mill, acid leach the uranium and recover the uranium from unclarifled liquors by a CIX system coupled.to an ELUEX process, - There i's a limited scops for_ recovering uranium . from wet-proce33 phosphoric .acid in India. The Indian Phosphate rock mined in Udaipur area is low in uranium (< 50 ppm U). However, about 2 million 'ton's of imported rock distributed in a number of acid manufacturing plants appears to be.a batter source. The output capacity of most ;of the plants is less than 400t acid (30% P205) per. day. There are only 2-3 plants (FACT plant in Cochin being one) with somewhat higher capacities. The feasibility of recovering uranium in theia plants by. solvent extraction is;being assessed. Prellminnv/'
/ - 424 -
estimates have shown that uranium produced from this, source on this scale costs around Rs.2000 per kg. This being lower than the cost at which the product will be available from the low-grade ores it is worth-considering.
7. SUMMARY
The technology for uranium ore processing is well established. Various estimates have shown that on a global basis uranium resources are adequate to meet the foreseeable demand. However, there is an uneven distribution of these resources countrywise. The Indian resources are estimated-to be about 60,000t U. Not. only is the grade of ores low, the individual deposits are also relatively small. A combination of these two factors make the indigeneously produced uranium very costly. As India is not.a signatory of the Nuclear Non- proliferation Treaty it is not possible for this country to /take advantage of the prevailing low price of uranium concentrates and import the requirement. The nature of the deposits, precarious, resource, position and relatively sroaH capacity of the mines do not permit" the country to take advantage of large throughputs in the mills to achieve substantial cost reduction. However, by resorting to as high a scale of milling as the mines would permit, by reducing the loss of solubilised uranium after leaching and by undertaking production of nuclear grade final product at the mill site ai{jnifiaanfc, though not a major, economic benefit can. be - 425 -
derived. For this country it i3 worth considering the possibility of deriving uranium as byproduct from copper flotation plant tailings and from wet-process phosphoric acid. Though the cost of uranium from these sources will not be as low as it is in other countries where byproduct uranium is obtained due to relatively small scale of operation, it uiXl be lower than the one at which the poor grade ores will yield in this country.
A broad Justification for indigeneous production of high cost uranium for nuclear power generation is that the U308 -contributes not more than 8% of the unit power cost. This is clear from Table-Ill which is based on the figures provided by the Japanese Ministry of International trade and industry.
/ - 426 - '!
Fig.l : SIMPLIFIED FLOW SHEET FOR URANIUM ORE PROCESSING
MINE -^ ORE I CRUSHING
GRINDING
ll?S04/Na2C03
LEACHING Oxidant -
SOL1D-LJQUIL Sol id SEPARATION Reiect Leach Liquor J±L ION\EXCHANGE SOlV.EJCrRACrKW ION. EXCHANGE
Eluate '
EJuate Strip SOLVENr . Soln. EXTRACTION
t "> . Strip remprwriDN . *
• > *'
FI LIBATION j: CAKE YELLOW CAKE - 427 -
O^t3RE (330,00(3: 0 tpd) Ground (50% - 74 micron)
-^» Chal.copyri te FLOTATION
Tailings (0.035Z U30g) —n ^ Slimes DESLIMING T Coarse MAGNETIC -^Magnetite SEPARATION
Non magnetics
.GRAVITY > Tailings CONCENTRATION Concentrate.(0.0851-U,0 j
—1^. SHraes CABLING
Table Concentrate
URANOTHORIANITE (3Z U30g) + DADDELEYITE (77X Zr02)
Fig.2 SIMPLIFIED FLOW SHEET FOR URANIUM CONCENTRATION AT PJIALADORWA - 428 -
Table-I
URANIUM RESOURCES OF MAJOR PRODUCING COUNTRIES
Data* as on 1.1.1981, cost range- US$130/kg U . Country Reasonably Assured Estimated Additional '000 t U • -'000 t U
Australia 317 285
Brazil 119 81
Canada 258 760
France 74.9 46.5
India 32 25
Namibia 135 53-
Nigeria 160 53
South Africa 356 175
U.S.A. 605 1,095
Others 237 147
Total 2,294 2,720
* only for countries outside the centrally planned economies
Source:. Abridged from the Joint report by the OECD Nuclear Energy and the IAEA, 1983. - 429 -
Table-Il
DISTRIBUTION OF CAPITAL & OPERATING COSTS IN URANIUM ORE PROCESSING
Capital Costs Overall Operating Costs OPERATION Max.% Min.% Av. Max.% Min.%
Comminution 20 40 20 15 25
Leaching 5 25 10 10 25
L/S Separation 15 35 8 10 20
Concn.& purification 8 20 8 .7 .25
Pptn. Calcination & • 4 10 5 8' 15 packing
Tailings disposal 3 10 6 ,3 40
Lab. & Services 4 10 8 6 10
Mining 35
I) Joint report by OECD Nuclear Energy Agency & IAEA, 1983. II) Guesstimates by Author. - 430 -
' JIable-IlI
CONTRIBUTION OF FUEL AND NONFUEL
COMPONENTS TO NUCLEAR POWER COST
S t a. g e %
U308; . .7.8 Conversion' & 5.2
Enrichment 25.0
Fabrication 4.5
Reprocessing 7.5
Nonfuel 75.0 - 431 -
3AOUGU0A URANIUM MILL RICH EXPERIENCES FOR FUTURE CHALLENGES
K.K. B E R I*
India's only uranium ore processing plant at Daduguda was commissioned In 1967-68 utilising low grade uranium ore. The flowsheet of the mill was based on studies done at Bhabha Atomic Research Centre* The mill In general worked upto expectations, keeping in view that this was the first aill. Subsequently based on experience and further..techno logical advances, changes in different areas ware done from •;ime to time, for better results end better operating practices. In recent past mill was expanded and commissionsd in October 1987 to augment the capacity by 40% to take care of. uranium mineral concentrates recovered from copper tail ings of copper plants and ors from new Bhatin mines.. The mill expansion which was designed, erected and commissioned by. our internal expertise, has given rated capacity and is working satisfactorily.
PROCESS AND FLOUSHEET
-300 ram ore received from mines is crushed in two stages. The ore; is fed to 115 mm opening scalper screen and the oversize goes to 1*25 x 0.4 M jaw crusher for primary crushing. The jaw crusher dischargs along with scalper undersize goes to screening section having 1.8 x 4.8 II long tripple deck screen of opening 115 mm,. 63.5 mm and 25 mm.
* Chief Mill Super in ten dent Uranium Corpn. of India Ltd P.O. 3aduguda Mines Singhbhum, Bihar, PINx832 102 - 432 -
-25 mm is collected as fine ore, —115 + 63.5 mm is collected as pabblsa for secondary grinding and rest of the sizes are returned back for secondary crushing by 1.2 PI short head cone crusher. Cone crusher uorks In close circuit with screen.
The fine ore is ground to +48 mesh -1% maximum and -200 mesh -60% by tuo stage uat grinding. The fine ore is ground in 2.7 1 x 4.8 PI long primary rod mill. The slurry is classi fied in Akins 2.1 M double pitch single spiral classifier. Tha oversize fraction is recycled to 2.7 Pi x 4.8 PI long secondary mill where ore pebbles are used as grinding media. Secondary mill uorks in closed circuit with the classifier.
Tha ground pulp from mill contains 40 - 41% solid. This is thickened to 60% in 25 PI dia thickener and filtered by 10' dia 7 disc size, 2 nos disc filters (one acting aa standby) to give a 82% solid cake. The extra uater ia recycled back to mill. The cake from disc filter ia re- pulped to 60% solid by secondary filtrate rrom down stream process. .
This slurry goes tor leaching* Tor leaching, sulphuric acid and manganese di-oxlds as oxidising agent are used. 10 nos, 3 n dia x 12.3 PI ht. air agitated PISRL pachucas with central air lift column have been employed as reaction vessels. Around 94 - 95% uranium present gets leached at following parameters: 1. Leeching time : 12 hrs. 2. Temperature : 38°C 3. pH j 1.7 - 1.8 4. Redox potential t 440 iv.
The leached slurry is filtered by primary vacuum string die- - 433 -
charge drum filter to gat uranium solution knoun as pregnant solution. The cake is repulpad again uith acidified water and filtered again by secondary filter ( 3 nos each vacuum string discharge drum filters of size 3.7 M x 4.8 tl face are provided in each stags) for batter recovery* The se condary filtrate so obtained is sent back to disc filter section for rBpulping neutral oake for solution balance, flocculating agents are used for better filtration_chara- cteristics.
The unolarified pregnant liquor is clarified by 2 nos pre. coat filters of size 2.4 II dia x 2.4 M to gat clarified liquor of less than 10 ppm. slimes. The uranium concentra tion of liquor at this stage is around 0.5 - 0.6 gm/l, along uith sulphate, ferrous and ferric irons.
For purification and. concentration two column fixed bed ion exchange system is being used. The section consists of 2 parallel sets each consisting of 3 nos 2.5 II dia x 4.3 M ht. FISRL column - ona set acting, aa standby. Acidi fied salt solution is used for slution.
The uranium as magnesium di-uranate is recovered from strong eluate by double precipitation method. The pH is first raised to 3.8 by adding lime slurry to precipitate the sulphate and ferric iron, followed by precipitation of magnesium di-uranate by increasing pH to 7.0 by addition of magneaia. The magnesium di-uranate precipitate known as yellow cake is washed, filtered, dried and packed in drums.
The barren cake obtained from filter section and barren liquor obtained from ion exchange section is neutralised by lima to pH 10. This slurry is classified by hydrooyclone - 434 -
to.separate out sand uhich is sent to mines for back filling and slimes are pumped to slime dam uhere solids settles and clear water is decanted off and disposed.
DILL EXPANSION
The mill uas expanded to increase the processing, capacity by 40jS to take care of ore from neu mino-at Bhatin and . uranium mineral concentrates recovered from.copper tailings from Rakha, Surda and Plosaboni copper plants. The follow ing additions were done. A separate circuit uas provided for neu feed upto neutral thicken in (j end after that both the slurries were combined for further processing. i. The mineral concentrate is in pouder form, but needed further grinding hence a separate ground hopper uas made, so that this can be carried direct to bins for - 'feed to mill bypassing crushing plant. ii. Existing crushing plant capacity uas :• planned -to increase by 30^ by addition of one more stage of: crushing. This is being discussed later in detail. iii. Aball mill! of size 2.134 PI dia x 3.66 PI long with a hydrocyclpne uas provided for grinding of concen trate and Bhatin ore. iv. A neu thickener of size 15 II uas provided to take care of extra slurry from ball mill* v. For ion exchange 1 no. column uas provided'in each set uith the idea of running both'sets in parraliel and extra column to act as standby to take care of regeneration requirement* vi. For other steps additional units, of similar sizes uere provided such as - - 435 -
a. 1 no. disc filter b. 3 nos leaching pachucas c. 1 no. drum filter each in primacy and secondary circuit. d. 1 no. tank each for iron and product precipitation. e. 1 no. neutralising pachuca f. 1 no. hydrocyclone. vii. Host of the existing pump3 needed replacement due to wearing off. Such pumps were changed uith higher capacity. .For others, additional pumps were provided.
The commissioning uas absolutely smooth and no difficulty uas experienced to get increased installed capacity.
PLANT PERFORMANCE AND EXPERIENCES
For selection and design of a process, certain base datas are required to be generated at lab scale test and the success of plant operation depends lot on these datas. For uranium processing plant following informations are re quired in general: - i. Type of ore ii. Best suited leaching route . iii. .Parameters for.optimum leaching efficiency i.e. grind size, reagents and its consumption, rate of reaction, temp. etc. iv. Grindability of ore and uork index. v. Filtration characteristics, rate: and losses in the filtration operation, vi. Route for concentration and purification step and • parameters . for same. vii'. Grade of product* viii. Total overall recovery, ix. The expected release of radioactivity to environment. - 436 -
For 3aduguda the above informations wore generated front studies done at BARC. The plant as a uhole gave expected performance except at feu places uhere corrective actions ware taken. In following lines our experiences with the plant, corrective action taken and points to be taken into consideration for future plants have been dealt uith.
A. ORE RECEIVING AND CRUSHING
a. Spillages;- The Jaduguda ora contains 8 - 115* moisture and 15# fines. Heavy spillages were experienced throughout the crushing section, resulting in lower life of conveyor and idlers, poor plant availability, heavy manpower require ment for cleaning of spillage and operating problem due to frequent jamming of chutes and equipment. The following steps ware taken to overcome the problem: .
i. Changing of troughing idler angle from 20° to 30° as flatten belt gives mors spillages. li. Providing mora number of self! aligning idlers*. iii.- The smaller width bait with higher speed ara more susceptible of running out. The belt speed upto 1 fl/sec. was found most ideal. Widening of belt which needs widening of galleries also is a major job. This has been taken cere for worst affected belts* iv. Modification of major effected chutes to have steep angle in place of low angle to avoid accumulation of fines and sudden release of same resulting in jamming of baits, running out of belts and hsavy spillages* v* Providing of ultra high density polythene liners In chutes for easy, flouability and. for avoiding - 437 -
Tinas accumulation uhoro particlao less than 1" is being handled.
The material handling system uhich involves . handling of large amount of spillage and clean ing of same can poae major problem. The design should take cars of this aspect. b. Crushing plant flowsheet:- (Unbalanced orushing load on primary and secondary crushers). For collection of pebbles the setting of jsu crusher is kept around 89 mm to get good yield of -115 +63.5 mm pebbles. But thiei leaves higher work load on secondary cone crusher uhich has to perform a reduction ratio of about 4.5 to 1, compared to 1.76 to 1 of jsu crusher apart from receiving a feed containing too much material of. higher size than designede This results in uneven manner of liner wear and hence less life of same. This extra load on machine gave more vibrations resulting in frequent break-down of machine.
The crushing plant in normal course uas de signed to run for 10 hours • day. With Mill expansion either us had to have a new crushing plant for 25JC extra capacity or the existing plant had to run 25jC extra, leaving less tine for maintenance.
On close scrutiny it was found that if one more stage of crushing is introduced, then this uill not only Increase the c:\osoity of plant but uill also take care of heavy load on existing oone - 438 -
crusher, resulting in batter officiency of same. It uas decided to install ona 1.2 H standard head cons crusher to crush material of size +115 mm and -115 +63.5 ram (when these pebbles are not collected) and are recycled for secondary crushing leaving only -63.5, mm +25 mm size to be crushed by existing short head cone crusher. The advantages of three stage crushing are: i. In place of -25 mm size, -20 mm size fine ore can be collected, making grinding circuit more efficient, ii. Overall capacity even uith -20 mm'fine ore size uill be 30% more, iii. Balance the uork load on all stages ensuring better life of liner and overall better life of machine, iv. Most economical system for enhancing the crushing plant capacity.
The expected performance appears in drawing No.II.
A standard head cons crusher has been ordered and . related uork has been taken up. It is much better' to take care of such items in the dasign stags itself as any modification afteruard3 becomes difficult due to matching problem and space constraints.
GRINDING
Au^topeneous grinding to rodmlll grinding for primary grinding:- The, primary grinding uas originally planned as autogensous grinding using +115 mm size ore as grinding media. But on commissioning the mill it - 439 -
uas Tound gatting overloaded frequently and running of the plant uas near impossible. This uas dua to bulk ora pabblas uara not as hard as original samples. The primary mill uas converted into rod mill. 4" dia high carbon steel rounds (C-0.85) were employed as grinding media* The mill speed/uas reduced from 22 rpm to 12 rpm and linars Ufa re changed from wave type to lifter bar type.
In general autoganeous grinding should not be em ployed unless extensive tests are done at pilot plant scale on truly representative ore.
Uith the change of primary grinding from pabbla to rod mill, the mill size provided became too big of capacity required and hence mill speed and rod charge level uers reduced to get required capacity and matching uith secondary mill. After 10 years of use the shell uas found uorn out and needed replace^ ment. At this stags it uas decided to replace the same uith smaller proper aiza mill of siza 2.134 H dia x 3.66 II long. This mill had I"ln. Steel uava type liner and key type liner holding system which needed less number of holes in the shell. Though capacity and grind siza uas as per requirement uhen liners uera nau, but uava type.liner design had life of: only 32 days. Uith the uearlng out of uave tha capacity used to fall end more fines uara being generated. Changing the material of construction from Pin. steal to. high chrome nickel liner increased the life to only 42 days.
Ultimately design of liners uas developed by us based on lifter bar type (drauing No. Ill), the - uo -
lift could not ba given much as mora lirt uould have required bolting of each liner, nasding drill ing of mora holes* The mill spsad had to be reduced from 20 RPPI to 18 RPM. The capacity was slightly higher for complete liner life cycle and grind size was satisfactory and consistent. The life of liner went up by four tines.
The lift of liner could haue been increased note to get still batter life, but then Bill spaed had to be reduced still further resulting in capacity reduction of Mill. If will disaster uould have been higher, us could have taken trials for most optimum design of liners. It is always advisable to haue adequate aill dia for proper liner design.
In secondary mill initially wave type liner of high pitch type fin, steel was being used which was giving average 3£ months life. The design was changed to less pitch liners with an idea to reduce slippage and aroaion in between the waves. The life was found 40% higher without effecting adversely the capacity of grind size. "A still'lower pitch can give still better life but it nay effect capacity. For trial, liner with still reduced pitch have been procured for half the mill and'trials are going on. The initial indications are that life of liner may bs mors but full shall liner ney effect the capacity. It may ba worthwhile to have these liners for feed end side where user is mors. The final conclusion will'be drawn only after trials ars over.
Any increase'in liner lire definitely reduces shut-down for liner ohange and hence is preferred. - 441 -
Trials wars taken uith still harder materials i.e. high chroma, high nickel alloy which gave almost double the life than of tin. steal of same design at virtually sans operating cost i.e. Ra./fIT ore ground. This material is definitely preferable as this will reduce shut-down by 3 days/annum. The trials for this material is also going on for fixing up the data.
The crushing and grinding section needs maximum shut-douo for maintenance and these units accounts for minimum ZQ% of total cost of processing. Hence liner life of grinding mill needs extra care in selection as these will influence a lot on stoppa ges because of liner.change and cost involvement.
0. RUBBER LINER
At the time of mill expansion the small primary rod mill (2.134 fl dia x 3.56 n) uas converted tu ball mill for grinding Bhatin ore and mineral concentrate and a nsu mill uas procured of size 2.7 fl dia x 4.8 H uith rubber liners. Rubber liners .reduces capacity by 10% than steel liners. This uas taken care off in selecting mill size. The rubber liners wore found overall better than steel liners sxcept once uhen these liners failed miserably due to poor bonding between base metal and rubber moulding, a manufacturing fault.
The advantages and disadvantages of rubber liners are as follows: i. Free from any leakage near mill shell liner bolt which used to result in frequent stoppage - 442 -
of mill for bolt tightening, lass chance of wearing of shall near bolt holes, no chance of slurry going to drive units and hence bettor life of drive pinnion and girth gear. ii. Lass noise. iii. Grinding rate uas 8 - 10$ logs than Pin. steel liners, but did not effect to us as this factor uas taken into account while selacting the sizs of the mill. iv. The liners are marginally costlier in terms of cost/MT of ore ground; v. The feed end side liners uear uas much mors needing inbetuesn replacement and if repla cement is not done at proper time, uneven height of lifter bars can create relatad problem because of zigzaging of rods in mill.
The rubber liner suppliers' principal has coma out uith mBtal claded liner .particularly suitable for rod mill. Feed end side shell liners uhich it claims uill give increased life of liners. The same seams quite promising, the trial of which is still to be taken.
As regards secondary mill liners, rubber liners is not being considered because of two major.reasons; i. The secondary mill is .running at full capacity. Any reduction in capacity because of use of " ••- rubber liners cannot be tolerated. ii. The rubber liners are of lifter bar typo and lifter bar typo liners uill induce higher lift resulting in breakage of pebbles which will overload the mills. - 443 -
LEACH ING
The leaching efficiency was found equivalent to the designed one at designed parameters. However, it uas seen that the following changes of parameters did not make any difference in leaching efficiency! As per dBsiqn Modified pH 1.5-1.6 1.7-1.8 emf -460 480 -430 4A0 Reactive time 12 hours 10.5 hours
The above changes made the process more economical. As regards grind size ue could have got same leaching at coarser grind eize, keeping controlling parameters same as designed. But any grinding coarser than being main tained gave problem in filters by uay of jamming of agitator.- Hence coarser grin'dsize could not be adopted. Though coarser grirvd size gives some problem at thickener and pachuca also, but'they could have been controlled. A horizontal belt filter could have solved this problem, Operationuise pachuca operation is simple but lou density, lou air pressure and coarser grinding uere main causes of occasional sanding in pachucas. Stricter control and cutting off of pachucas.after one month in.operation virtually eliminated such problems.
Najor mechanical problem uas collapse of internal central air lift pipe supporting structures. Uhile draining of the jammed pachuca,the supporting structure used to collapoe due to dounuard flou of virtually solid columns. The support made of 53 316 inside the pachuca uas not very strong. MS structure, uae made on top of pachuca and central pipe uas supported by tie rods. This system eliminated collapsing of costly 35 316 supporting structure. Nou. tie-rod gives uay; sometime but the same is easy to repair. - 444 -
Control oT parameters*- From the vary beginning the auto matic pH control system in leaching pachuca had not worked properly due to following reasonsx- i. Suitable electrodes uero not available. The life of electrodes was 7 - 10 days* Tha response used to be slow end erratic due to sulphate deposition* The physical impact of vigorously agitated slurry also effected the electrodes, ii. The long co-axial cable used between electrodes and pH amplifiers was • source of major problem due to moistures related cable fault and other electrical interference which are very common in high' impedance pH measurement* Slowly within a short time, manual control was found more suitable and automatic control-was bypassed* The search for good system is going .on and hops to get reliable system shortly. The following steps are being taken in this regard: i. As regards electrode rugged gel filled electrodes with much more mechanical strength has come in' the market and has been ordered for trial. Another A-ntimony ring electrode with rotating carborandum continuous cleaning system, though costly, is being considered for trials* ii. pH amplifier using field preamplifier,there by eliminating the use of long coaxial cable, has been ordered* These emplifiere are narrow range ampli fiers hence pH can be controlled more accurately. us havo to solve this problem and make the system automatic. Manually controlling pH ia difficult. In Deduguda pH variation even upto 1.6 - 1.9 does not make any difference in leaching except acid consumption will be higher. But where leaohlng efficiencies are low and any - change in pH grossly afreets the efficiency, manual control can be of much concern* - 445 -
SOLID LIQUID SEPARATION »•• " • *
This section gave maximum problems, all uere of complex nature and interrelated. The main point of concern uere} i. Poor filter performance due to poor pick up* ii. High uranium soluble losses iii. Fro quant break-do un of filters. iv. Heavy spillage of slurry resulting in related problems.
The following points uere attended and corrective action uere taken to bring tha plant to manageable level:
i. Slurry*- The slurry uas found having more of fines than expected resulting in poor filtration rate with poor washing efficiency. The problem used.to get aggrieved in caaa of any problem in grinding. The problem of fines uere maximum uhen small primary mill with wave type liner ware in operation. Such slurry gave denser cake and U3s mainly responsible for higher soluble losses of 5 - Q% in place of designed 2.5J&. A stricter control on grinding only has avoided cases of fines generation. Though characteristic of slurry uill depend mainly on ore. but controlled grinding can control grind size to some extent. For this better instrument control- is to ba provided at grinding. Studies for the same is still going on.
ii. • Clothl- . .The material, porosity and weave are main factors effecting suitability of cloth on filtration. Earlier nylon filter cloth was being used, nylon gets effected by dilute acid at 1.6 - 1.7 pH. Polyatar and polypropelena would be more suitable. First lots of trial with polyester cloth was taken, but abandoned as polyster cloth uas found mechani cally weaker and holes used to develop. Polyester - 446 -
spun uoave cloth Pound getting blind uithin hours.
Uith the availability of suitable polypropelene yarn, ue could develop cloth similar to nylon and this gave better results'than even nylon. Uith sudden increase in nylon yarn cost, the polypropelene cloth ha3 become still more economical.
Earlier tuill weave cloth uas being used for primary filters and chain weave for secondary filters. But performance of tuill used to get severly effected uith any change in characteristic of slurry. The problem became acute uith processing of uranium mineral concentrate along uith ore. Ultimately chain weave of polypropylene yarn has been stand ardised. The cloth selection is very important for any filtration process and has to be tailormade for particular duty. iii.: Floculant:- Floculation plays major role in filtra tion. Separan NP-10 non ionic synthetic polyacry-' lamide uas initially used and standardised.'at 0.31 Kg/MT of ore.. This uas imported and costly. This una later replaced by indigeneously available gummy product. "FLOCAL" uith consumption rate of 0.062 K^/MT of ore. This floculant uas not very effective in caao of variation of grind size to fineness. However, after commissioning of byproducts plant the chemical mainly sodium silicate used there hod adverse effect on filtration and need.uas felt for further studies of Ploculating agents.
Secondly the gummy product' flocal gave less porous cake resulting in poor washing efficiency. Trials uith different floculating agent indicated that a - 447 -
combination of gummy as well as synthetic poly- acrylamide bass product gave best results. Lots of floculating agents have come in the market. Ue have recently tried RAN 4607 and this has given u& excellent results. Floculating agents has to be specially selected for individual slurry.
The following step3 uere taken to take care of other problems': a. In case of malfunctioning of filtrate pump, liquor used to go to drain via vacuum pump resulting in uranium solution going to drain apart from corroding vacuum pump. The moisture trap height was raised to 10 0 ht. and a baro metric leg uas jprov/ided uhich ensured liquor getting drained in contained area through barometric leg uhich can-be pumped back to . system. b. 3 nos filtrate pumps uere provided - 2 nos for filtrate and one for uash for each filter i.B. in total 18 nos pumps. In' place of these 3 nOs pump uere provided for all primary filters (2 in uorking, 1 standby) and four no. pumps ,uere provided in secondary filters ( 2 for filtrate, 1 for uash'and- 1 .standby) i.e. total 7 pumps for expanded capacity. This resulted in better control and less equipment for maintenance. c. In place of original 'system of draining of •filter vat slurry on floor and subsequently cleahing and pumping of same to secondary filters, 1 no. drain tank uith pump uas pro vided each for primary and secondary filters. ' .-.This resulted in cleaner operation, avoided - 448 -
short circuting of leached slurry to secondary filter and dilution by water consumed for cleaning the floor* d. Agitator and drum drive gear boxes ware found under capacity and hence higher size sturdier gear boxes were provided. Cast iron pedestal blocks wars replaced by IIS fabricated block which virtually eliminated lot of break-douns.
The solid liquid separation if not properly designed and planned can realy become bottle neck for pro duction. As-such string discharge drum niters are not very suitable for uranium ore slurry. For better recovery and for handling wide size distri bution of particle, horizontal belt filter is more su itab le .
CLARIFICATION "
No problem uaa faced except lou clarification rate which mede this section bottleneck for the plant. Oicalite 478 imported product had low clarification rate. Different grades of this product ware triad and grade 4158 which has coarse particles than 478 gives higher rate without compromising much with clarity. Apart from higher through put the consumption per (IT also came down by 8 - 10£,
Now two products ere available locally being menufactured with foreign collaboration. Oicamol and'Indica both products are suitable and are being used, stopping import. - 449 -
ION EXCHANGE
Ion exchange system operation, mechanical and instrumen ts tionuise behaved exactly as designed'and planned. But for continuous and sustained operation needs expertise as day to day problems are interrelated with many factors connected uith composition of liquid.
The Main problem faced and action taken can be summarised as belou: i. The loading characteristics depend a lot on liquor quality. Lou acidity, lou radox potential liquor gave best results as competing ions we're less. ii. The resin should be suitable for this process. It should not be sluggish and silica build up should be less. For this, resin should be tested minimum for 20 cycles. During finding alternative resin, ue came across resins for which silica accumulation waa too high within 8 cycles. iii. The decision for regeneration of resin in the operating plant should be taken on optimum cost basis. Uith fouling of resin the resin becomes sluggish resulting in minute increase in loss and Increased eluant consumption. Hence even if plant throughput can be maintained, resin regeneration should be done if eluant consumption goes high and makes process uneconomical. iv. The resin and silex bed should be cleaned periodi cally to avoid channeling. v. On operation side only problem was leakage through hydraulic valves. The Imported automatic hydraulic Saunders 0N-0FF valves gave trouble free service. The valve diaphragms were developed indigeneously, -450
but proper quality could not bo davalopod. Thin resulted in malfunctioning of values. The local tuo type or values (i) Gramvik & (ii) BOK valves were procured but ware not found satisfactory. Ue era again trying to import feu original Saunders valves. The effort for alternative indigeneous valve is still going on. vi. Ue are using Indian ARU 101 resin supplied by Ion Exchange India in place of imported Oiacidite FF 530 of Permutit. The Indian resin is sluggish and haa 13# lass capacity. Ue havs procured fresh resin Duolite A 101 D/U from M/s Henkel and order for improved quality of Indian ARU 101 is placed. Ouollte A 101 D/U resin though had initial loading similar to sample tested in lab but loading found deteriorating very fast. The column u-as cut off for regeneration. It was seen that resin has become very fine and ]Z% resin has already gone out with backwash. The resin sample will be tested for silica build up before regeneration and after regeneration. Duolite resin is uorlduide accepted as one of the best resins. It seems Indian manu facturers are not able to reproduce same type. Final conclusion uill be done after doing further tests. This is one area uhere still ue uill have to depend on import. vii. The automatic control system ia obsolete and imported apare parts are very costly as this is specially manufactured for us. These days better electronic micro processor based controls have come in the market. Ue have already got the studies done and have received budgetary offer. The cost involved is higher but is still worth replacing. The decision on same uill be taken shortly. - 451 -
PRECIPITATION AND RECOVERY i. Grade:- The grade of product after maintaining 3.B pH at iron precipitation stags and 7.0 pH at P1DU preci u pitation 3taga uaa coming around 71 - 72% 30g» Magnesia was beinn procured of specification of 90% reactive MgO of size 90% passing through -200 mesh. The specification uas changed to finer size
i0e. 90% passing through 325 mesh. The change reduced the MgO consumption and grade of product uet to uet as high as 76 - 78%. Though now 90% reactive MgO is not available in the market instead 85% MgO material is available which has reduced the grade marginally lower. The comparative analysis is as follows: 90% MgO 90% PlgO 85% FtgO 90% .-200 90% -325 90% -325 mesh mash mesh
U308 7T<,5% 77% 73% FeT 0.55 0.55 0.50
Si02 4.0 3.0 6.0 Ca+Mg as ff^gO 7.50 4.75 6.0
He are trying to have, alternative source, of good quality MgO for improvement in grade.
ii. Fa & Product Precipitation nitration;- Initially press filters uere employed for above precipitate needing manual handling of radioactive material. These filters uere replaced by vacuum drum filters of size 2' x 2' for product precipitation and 3' x 3' for iron precipitation. This has made process cleaner nee;ding less man power.
iii. Product Drying & Packlnoi- Initially wet cake from - 452 -
press niters of 40% moisture used to be packed in orums uith polythene bags and despatched. In this, oftar lying for sometime a part of water used to gat separated and many times used to corrode the drum needing repacking apart from high cost of transporting uat product and messy operation of packing.
To take care of above problem it was decided to put up spy ing unit for P1DII. An electrically heated jhain type belt drier uas installed. The chain belt carries the fIDU loaded trays through electrically heated drying chamber followed by air cooling space. The dried MDU is manually discharged to the double roll crusher hopper and crushed PIOU diractly discharged to drum. A small uet scrubber in com bination uith exhaust fan is installed for dust control.
Though drying system is working satisfactorily but is not entirely satisfactory as still some job needs raanul hand ling. ' Ue are still in search of soma automatic fully enclosed system. At present moisture content is kept between 7 - 9% to avoid extra generation of dust.
TAILINGS DISPOSAL AND ENVIRONMENT CONTROL
The main elements which needs special attention is fin. and radium. In the system for pollution control only neutrali sation of tailings uith lime is envisaged. After neutrali sation to pH 10 though Pin. values come to within limit of 0.1 ppm but Ra values ware found above limit of 500 Bq/tl3.
In the pond due to self oxidation of sulphide radicals pH used to come down to the range of 7.5 - 8.00. This used to redissolve ths alssady precipitated manganese and pH - 453 -
values were often more than 0*1 ppm. The effluent discharge values were kept under control by dilution by other streams of mill as vacuum pump, compressor cooling water and sewage treatment plant drain.
The main elements which requires special attention is fin. and radium* Neutralisation upto 10 pH was fixed to com pletely precipitate Mn. but there was no control on Ra. Even after proper neutralisation of tailings slurry, due to self oxidation of sulphide radicals pH in pond used to go low resulting in redissolution of Mn and getting leaked with slime dam overflow* A-part-from slime dam overflow, mine water overflow also contained high fin and Ra* Fin end Ra content of overall effluent going out of UCIL jurisdiction generally used to be within limits due to dilution by aewage treatment discharge* mill general discharge of vacuum cool ing water etc*
After mill expansion to meat extra demand of water the only option alternative was to reclaim water and recycle. Any recycling, of water would, have resulted in build up of ions and lass water available for dilution resulting in fin and Ra values going higher than stipulated. Hence it was decided to have effluent ietreatment plant to keep the effluent within stipulated limit.
For this a combined scheme for reclamation of watar and effluent ratreatment plant was made. Erection work on same is going on. The plant is due to commission in Danuary 1990.
EFFLUENT RETREATFIENT PLANT
The water requirement - potable and industrial aa per - 454 -
dasign before expansion and after expansion is as follows (cu.m./day):
Designed Before ex pn • After axpn. • 1,500 1,600 dines 2,160 mill 4,120 5,120 8,430 Bhatin *»^ 200 400 Colony .2,750 4,300 5,500-
9,030 11,120 15,930
Extra increase from before expansion S.ovel of 4,810 cu.ra./day was to be planned from reclamation.
Uatar could have been reclaimed from following streams:
1. Vacuum pump cooling water 1,800 cu.ro. 2* Compressor cooling water 600 cu.m. 3. nine water 1,00.0 cu.m. 4* Uatar used in magnetite plant 1,320 cu.m. 5. Tailings pond water 600 cu.tn.* 5,320 cu.m.
* As. the mine water, magnetite water and tailings pond water contains contaminated radicals chloride, hence could be U3ed only for secondary cake rspulping and down streams of ion exchange. Hence only-600 cu.m. of tailings pond uatar was planned Air recycling depending upon requirement.
Reclamation Schema
1. The vacuum pump and compressor cooling uatar will be pumped to. industrial water pond from whore it. will, ba used as industrial, uater, .
2. Magnetite plant, effluent containing around 5%.slime - 455 -
. and turbid mines water will ba purapad to a 15 M dia thickener, tha undarflou uill be sent to tailingo pond along uith plant tailings. Tha clear overflou uill be collected in a tank and uill be pumped to meat the requirement of secondary filter cake repul- ping, tailings plant and magnetite plant. Any shortfall uill be met by tailings pond affluent. Effluent re-treatment plant
The main purpose of this plant is to reprecipitate Ra and Pin by treating uith BaCl^ and then increasing pH to 10.
A separate drain uill ba constructed to collect tailings pond overflow to be brought to re-trettment plant. This water uill be pumped to clarifier. Tha ovarflou uill ba .reservoir collected in clarified uator/, ana a part of clarified uatsr uill be pumped to mill for re-use.
Tha remaining uater uill ba pumped to a flash' mixer, where barium chloride uill bo added and uill flow to barium radium sulphate precipitation tank. This slurry yill be neutralised by line in neutralising tanks and uill be clarified in elarifier., The overflow pH of this will be brought to 8 by addition of sulphuric acid and discharged to environment. The underflow of this clarifier along uith rau effluent clarifier uill be pumped back to slime dam.
Operationuisa following two points are worth mentioning!
1. Lime neutralisation:- The limB is mostly procured from area near Katni (MP). The transit time is around a month and due to erratic availability, supply is also erratic. % CaO in lime from 60 - 65% at starting time used to deteriorate . to 40 - 42% - 456 -
apart from lumps getting converted to pouder. In rainy season the deterioration uas too much apart from other handling problem. The lima is ussd to be procured in closed uagona. The procurement of lime in polythene lined baga improved the. condition marginally.
Uith all these problems proper quality and quantity • „ of lime uas not available for neutralisation and proper pH uas virtually impossible to sustain.
The oxygen acetylene-plants ussd clecium carbida for acetylene production generating Ca{OH)_ as uasta. This material was triad and found very - suitable. The same material is being used now. This is oust free and consumption is less and vary economical too.
Pumping of tailing* J- The first stage tailings dam uas 1 KM away uith initial elevation difference, of around 5 metres. Pump uith 30 H head uaa provided and uas suitable. The distance and elevation diff erence gradually increased and presently 2nd stage dam is 2,700 metres away uith head difference of 35 II. To cope up uith the situation till now tin more stages of pumps have been provided of 30 n head each.
The three stage pumping created lot of correlation problem apart from making system costly. Finally trial uas taken with Warman high pressure pumps. Two nos. nSRL pumps with high chrome, high nickel impeller, working/series could develop 90 H head needed for disposal, the pumps arm uorklng since - 457 -
last tuo years and is found overall economical. The Impeller life is around Pour months and rubber lined around 1 year. The major saving is pouer and raanpouer.
One more set is being-procured to replace the entire system. Nou tuo sets-comprising of tuo pumps each of 100 HP will be employed - one set acting as standby. This has replaced three stags of pumping each stage consisting of three pumps of 50 HP each; tuo in operation - one as standby.
CONCLUSION
Daduguda ore processing plant has worked well. The plant production has been sustained uith increasing capacity utilisation. The .Increase in processing cost is also within limit and less than overall inflation. This has been achieved by putting continuous effort for improvements uheraever possible.
Though', there are certain areas where better systems have come up but the same could not be adopted due to cost constraints and/or'as the same cannot be adjusted in the present system. The full automation of grinding circuit and horizontal belt filter in place of vacuum filter comes in this category. The areas where we are still looking for better alternatives are ion exchange control system including on stream analyser and MDU drying plant*
I would like to comment on expansion. While designing new plant certain precautions should be taken so that expansion if required can be taken up* It 4s easy to plan and execute new plant or separate stream rather than expand the plant by incorporating equipment that - 458
too uithout hampering the currant production. In Daduguda uhere ua expanded the plant capacity by 40%, not a single day shut-doun was taken only for hooking the neu system uith existing one.' Such uork war a planned in scheduled shut-do un of liner change etc. only.
The main hurdle uas space and civil foundations.- The layout should be made taking into consideration distance required to be laft betuaen neu and old foundation.
Finally the layout uhether it is a neu plant or plant to be expanded, layout is equally important for successful running of plant. This ia important not only for economic point of vieu but is also important for operation point of vieu. The bad layout can make the system un~ manageable apart from poor uorking condition* this is one area which will be difficult, 'to improve after the installation of plant.
, i-.ooOoo «,. , - 459 -
PROCESSING OF URANIUM ORE - "OM MINES TO YELLOW CAKE
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KilHAVE TYPE — A tll'l LIFTER TY/>£- 8 SlCPtBBLE MILL l-7*
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{Vil 2 fcMvf — E - 462 - - 463 -
GRIMDING AMD LEACHING CHARACTERISTICS OF SOME OF TH3 INDIAN PRAMIUM ORES
V.M. Pandey, R.U.Cboudbary, A.K.Sarkar, A.P.Banerjee A.B.Chataaborty and N. Malty
CRANIUM OOBIOiyXION OF INDIA LIMITED JADTOUDA MINES SINGHBHCM BIHAR
Ihe Uranium requirement in. India has to be met by exploiting low grade do-
posits, assaying 0.01 - 0,0856 050Q. Ihe grinding and leaching character istics of »»e of such low grade ores were studied. The ores were ground •• under triMIs* conditions. It was observed that the ores of Bodal, Jajwal and Narvapahar are harder to grind than the o*es of Jaduguda, Bagjata and Turandih. The Jaduguda deposit, which is tinder operation at present, and Bagjata deposit are the softest amongst all.
The leaching studies for Bodalf tfarwapahar, Turaadih and Jajwel ores wore . mado under acid condition and the effect of various parameters such as grind siae, time, temperature and pH on leaching efficiency were studied, Onder the optimum conditions the leaching efficiency of the above ores were compar ed with the leaching.efficiency of Jaduguda and Bagjata ores. The Jaduguda and Bagjata ores were found to be the best from grinding as well as dissoln- tlon points of view. In oase of Jeduguda and Bagjata about 95 - 96* of D^Og can be solnbilised in 12 hours of leaching time at the following leaching conditions* pH 1.7 - 1.8,,emf -475 to -500 mv, Temp. 55 - 40°G. In case of Turamdih, leaching efficiency was 88 - .89* where as in Narwa, Bodal and Jajwal ores the leaching efficiency varied from 81 - 84*. The reagents consumption was found to be the lowest in case of Jeduguda and Bagjata ores and the hlghost in case of Bodal and Jajwol ores,
AH ore contained refractory minerals union wore not leached under conven tional leaching conditions limiting uranium extraction to 82 jto 95*. • - 464 -
mrscaccxioft
The demand for uranium In India will Increase substantially as the nuclear reactors envisaged la the D.A.E. nuclear Energy flan come Into working. The exploitation of -the Indian uranium Ores will help to meet this in creasing demand.
The most Important uranism ore body known la the country Is In the Singhbmsn Thrust Belt* The Dranlun from this ore body Is being recovered since 1967. In its recent exploration JHD have discovered many new areas of Cranium Ore body. The ores from these new areas present some grinding and leaching characteristics that are different from those of the Jaduguda ore. The selection of suitable "Hwh^ng conditions requires a balance bet ween uranium recovery and processing costs vM.cn are determined by the con sumption of reagents, energy requirements and capital cost of the equipment, leaching characteristics depend on the nature of host reck, the uranium mineralisation and the mode of Its occurence. Because of the many factors Involved, leaching characteristics must bo determined for each ore sepa rately.
This paper describes the studies carried out to establish the optimum parameters for the uranium extraction from Harwapebar, Turamdih, Bcdal and Bagjata uranium ores. 4cn> ramus uranlnite, the principal ore mineral of uranium, has the ideal chemical composition U0o> but material of this composition does not exist in nature as it is always-partly oxidised with conversion of U to 0 • The degree of oxidation ranges from 17 to 60 percent . Braninite Is readily dissolved in dilute sulphurio sold at low temperatures provided that oxidising con ditions ate maintained to convert all uranium to the soluble uranyl form. In order to aaxlaise uranium extraction, a suitable oxidant must be added to or be present in the leach slurry dosing acid leaching of a uraniun ore* It is also well known that tetravalent uranium can only he dissolved la the presence of iron ions in the solution, though one. ess use oxidants - 465 - such as manganese dioxide,' eoditm chlorate, etc., and that It ie necessary to maintain a minimal oxidation potential of about -590 mv to turn tetra> valent uranium to hexavalent form. Equations (l) and (2) show the ororall oxidation process of uranium. 4+5+ 6f S+ 0 +2Fe = 0 +2Fe (l) 2+ 5+ p 2 « +iOg = 2Fe +0g (2)
Generally, the oxidant ia added to the leach circuit In stages* Its use and effectiveness is controlled by afi? neasureaents. The oxidation takes place through a surface electrochemical mechanism In which the concentration of 5+ « *e ions adsorbed on the surface determineo the rate of reaction • She ferric ion acts more effectively In the form of ferric sulphate complexes, FeSO. in particular, vhloh exchange electrons more rapidJy than the uncomplexei ion** In practice, fera-lo sulphate is produced from dissolution of iron minerals In the ore or iron introduced during grinding, addition of sufficient oxidant, usually pyrolusito (&0o) or sodium chlorate, ensures that iron is 5+ present predominantly asFe • leaching rates are increased by increasing the temperature* Increase in tem peratures also increase a the dissolution of acid soluble gangue constituents resulting in a greater consumption of reagents and fcighor concentration of Impurities In solution* Therefore the final choice depends on whether impro vement in extraction can justify the additional costs arising from reagent usage and heating requirements.
EggtlHBffiL
.PTUNIPM ORBi- The ores used in this investigation were provided by A.M.D.
Description of the aqmplei- In case of Narwapabar and Turandlh composited drill core samples of each weighing approximately 30 kgs, was prepared talcing samples from the different bore hole oores in proportion to ttoir contribution in the total ore reserve* In oaee of Bodal, Jajwal and Bagjata a oproposito sample of 500 kg crushed to -1" also was taken* The bulk composite - 466 -
\ wa\3 coned and quartered to produce a 30 kg 'tost work sample. It was then jaw crushed to -6 mm size and further roll crushed to lOOjJ -10 mash (2mm)» After thorough mixing they were refilled down to approximately 1 kg fraction a which were Individually bagged to be kept as feed samples for testing.
GRINDING TESTS
In order to optimise the grind required for leaching, a series of batch grinding tests were carried out in the rod mill of the dimensions J
length of the mill - 400 mm Diameter of the mill - 160 mm r,p,m. of the mill - 100
as followBJ-
1) 1 kg feed sample was ground in the rod mill at 6o£ solids with a rod charge consisting of 2 nos of 25 mm and .3 nos. each of 20 mm and 12 mm din, weighing in total 6.7 kg*
2) The grinding was carried out for 5f 10, 20, 25, 30 and 35 minutes*, The ground product of different times were analysed - for their arize distribution.
LEACHING TESTS To ascertain the leaching characteristics, leaching tests ware conducted on the samples of each deposit varying the grind time, temperature and time of leaching. In each case one Kg dry sample was taken in a 2 L glass boaker with calculated amount of water to make a slurry of 605C solids. The contents of the boaker was stirred with 4 bloded glass propellers. Required quantities of HgSO^ and pyrolusite wore added and pH and Emf Were monitored continuously to maintain the pH and redox potential at the desired levels. In general, oxidant was not added until leaching hod boon in progress for an hour to allow for completion of initial.aold reactions with metallic iron and sulphides which would consume oxidant needlessly. - 467 -
Tha progress of the leaching reaction was chocked by withdrawing slurry samples at appropriate intervals. The scjnples were filtered and the residues washed repeatedly and finally dried at U0°C. leaching effici encies were determined by assaying the dried residues for uranium,
BESUIffS AND DISCUSSIOH
Mlneralogical and Chemical Analysis:
Mineralogical and Chemical Analysis of typical ore samples of Jaduguda, Harwapahar, Turaudih and Bodal deposits are given in Table I and Table-II respectively, TABLE- I
Minerals J_ Percent Content f Jaduguda Harwapahar \ Tursodih I Bodal OBS j OBS J ORE
Quartz 60.0 51,00 65.0 50.3 Chlorite & 20.0 38.00 27.0 Sericite Apatite 3.0 4.8 3.50 Magnetite 11,0 4.3 2.6 0.3 Sulphides iyrite 1.0 0.7 0.2 lyrolusite - 0.2 Tourmaline . 3.5 Traces Oranlnite 0.1 0.055 Amphiboles Mica - 42.8 Total opaques - 1.6 5.6 Refractory uranium minerals 0.3 0.6 others 1.1 0.945 0.5 0.2 - 468 -
TiBLB- n
Elements I P e r c e n t C o n t e n t | Jaduguda Ore I Harvapahar .Or© 1 Turandih Ore £ Bodal Ore
D3°8 0.07 0.047 0.045 0.0986
SiDa 67.20 51.24 52.50 51.41 *£* mm 34.68 28.9 5.50 13.82 32.8 12.88 12.94 10.4 ?e0 6.37 . - 11.57 F^°5 7.87 - • 5.38 CaO 5.40 1.88 4.69 6.20 HgO 5.20 1.07 2.55 3.60 MnO 0.15 0.01 . - 0.10
Ii0g 0.66 0.71 0.385 IjCfi P2°5 1.04 0.81 0.91 0*51 1*26 2.82 101 - 0.71 Cu 0.09 0.019 0.055 Ho 0.025 0.008 0.0033 Hi 0.11 0.025 0.022 S 0.79 0.05 0.1B Undertarmii lei Test rest - ... 333S3333S33333333333333S333B33333333 Uranium contribution from uranlnito is maxlnaw in Jaduguda Ore i.e. 81.7JC. In Narvapaha? the uranium contribution from uranlnite ia 75,22. psm iNjmcsis The ores passing through 2ma sice obtained after jaw crushing and roll crushing were analysed for its else distribution and for %0Q content in each siza fraction. The results are recorded la Xable.ni. - 469 -
TABLE- HI
Meah | Narwapahar | Turaadlh I Bodal \ Jajval slae i*3 *Vs J*** **?*[ vtS MSpE3—JTB^J + 10 0.46 0.056 0.48 0.050 - - 8.50 0.051 + 14 21.80 0.056 22.64 0.046 9.49 0.10 19.50 0.051 + 20 25.50 0.0555 26.93 0.047 25,49 0.1020 21.00 0.046 + 50 16.12 0.0555 16.91 0.047 22.89 6.097 7.00 0.044 + 48 8.55 0.0582 8.75 0.048 13.71 0.105 8.50 0.047 + 65 3.25 0.0580 3.40 0.050 4.3G 0.115 7.50 0,343 + 100 1.94 0.0580 2.04 0.054 5..51 0.118 7.00 Q&B& + 150 2.96 0.0570 2.10 0.0475 3.U O.liO 4.50 0.047 . + 200 2.17 0.0550 1.76 0.0518 2.25 O.IiC 5,0 0.054 • 325 3.82 0.0490 3,28 Q.052 15.21 0.1H 7.0 OoGOO (-200 nei&) - 525 13.63 0,059 11.75 0.0529 - ' - d.So' 0.089
ssscssassss&ssss.SBEiassssssssassx'.BitsaSBss Iron the above data it ie ales?, that Uranium is mora or loss evenly dis- tribute! in + 200 aeah fraction*. However, Uraaitoa content of th© finest fraction is nore than average in all earn*.
QTOHDIHG CHABACEgiiatlOa
The above mentioned feed ground for different tine periods under identical conditions was analyse! for its else distribution. The remits are given in Table-IV. - 470 -
TiSlK- IV
OHSS Mesh Hos. (Tyler Sarieo) jing !• + 65 j1 + 100' !3 + 150 . J + 200 1I - 200 SMinutes! i Norwa 5 51.25 6.40 4.77 4.09 33.49 10 25.50 11.20 8.70 6.80 48.00 IS 8.20 9.30 10.50 9.00 63.00 20 6.20 8.40 9.00 9.40 67.00 25 2.10 5.72 5.92 8.85 77.41 30 1.25 2.53 4.96 6.25 85.01
Turamdih 5 44.50 8.70 5.70 4.90 35.20 10 17.1 10.60 7.90 6.20 58.20 15 6.7 5.50 6.80 7.20 71.80 20 1.0 2.70 4.60 6.10 85.60 25 0.50 2.10 3.20 4.60 89.60 30 0.40 1.40 2.10 3.50 92.60
Bodal 5 _ 71.88 4.14 2.59 21.59 10 - 64.24 5.09 3.27 27.40 15 . - 55.55 6.31 4.51 33.83 20 * - 45.68 8.22 S.16 40.94 25 '- 34.10 U.16 6.67 .48.07 30 - 20.15 14.94 8.95 55.91 35 - 7.29 13.08 11.57 68.06 40 - 2.13 6.49 9.87 81.51
JAJWAL 5 60.0 1D.00 8.00 5.00 17.00 10 47.5 14.50 9.50 5.00 23.50 15 35.5 17.00 U.00 6.50 30.00 20 14.5 22.00 16.00 9.50 38.00 25 2.00 19.00 12.00 21.5o 45.50 30 2.00 7.00 13.50 20.00 57.60 35 0.50 4.50 5,00 27.50 62.50 40 0.20 3.30 10.00 19.00 67.50 - 471 -
From the above table It is clear that the Jajval and Bodal ores are harder to grind than the Naruapahax and Turamdih Ores. The work index for Narvapahar and Turamdih ores are more or less same as for Jaduguda Ore (ll.O K»h/T) bub in case of Bodal and Jajval ores it is more than Jaduguda ore (12.90 KWh/T).
LEACHING STUDIES,
Leaching tests were conducted on Narvapahar, Tura&dib, Bodal and Jajval ores to study the effect of variation of the following parameter at (a) Grind size lb) Leaching time (c) Leaching temperature
(a) Effect of variation of .grind aire Leaching tests were conducted, on the ore samples, ground for different time periods, under the.following leaching conditions!
pH - 1.6 to 1.7 Snf _ -500 + SO mv
temperature - 45°C - 50°C
Time - 8 HrB.
The Head assay of the samples verer
Narwa - 0.056* U 0g Turamiih-0.048 " Bodal - 0.0986 » Jajwal - 0.052 •
The leaching efficiency obtained at particular grinding is shown in Table-V. - 472 -
TJBLK-V
Grinding Time j Percent OJOQ leeched Harwa T Turandlh i Bodal 1 Jojual
0 55.50 66.84 55.55 56.75 5 67.60 83.95 55.57 59.62 10 77.1* 86.45 56.56 67.31 15 78.92 87.04 57.40 71.15 20 80.71 88.75 76.47 75.08 25 81.60 89.16 78.60 75.00 30 82.14 89.79 79.41 75.96
3S333S33S S S S 3 S: 8 :S 3 3 3 3 3 3 s s 3 S S S 3 s s 3 S 8 8 3 9 The Harwa and Turondih ores air* Tory fragile and even on grinding Tor 15 ' ats the -200 aeah fraction le aare than 6o£ which is tba optima grind also required for Jaduguda ore. Bub the difference In lwMtg between 15 and 50 s£o« grinding la about 2$ in caee of Horwa end about 556 in ease of Xurandlh ore.
The differonc* in leaching In ease of Varna can be narrowed to Iff by grind*, leg it for 20v Bta. on which the ore does not beeono Tory fin*. Tba orea of Bodal and Jajwal are Tory bard to grind and even on grinding for 50 ate the -200 fraction la leva then the -200 fraction of Harm, and Turaaila orea ground only for 15 nts. Therefore 50 at a. grinding was taken aa optlam grind else for leaching of Bodal and Jajwal ©rea.
(b) Iff ect of line Tariattlon on ^twMlpgt -, ill tbe erea ground to 60 to 66* -200 neah ware leached under feast*, lag eondltlonat
08 - 1.6 to 1.7 *aT - -500 • 60 BT Pulp dcaalty - 60} aollda '•aparaturo -- 45* . 50*0 - 473 -
The variation in leaching efficiency with time is shown in Table-VI. TiPLB - VI
Leaeblng j Percent Leaching tins | Narva I Turasdih I Bodal I Jajval I Begjata iJaduguda •"• I f 1 I • I I z 75.67 80.29 71.60 70.50 92.24 90.66 4 78.% 82.91 77.68 80.76 93.51 . 95.55 6 80.00 84.79 80.52 82.11 94.00 94.51 8 80.71 87.04 82.75 83.26 . 94.70 .95.13 12 81.60 88.21 84.57 ~ 84.61 95,20 95.50 24 82.32 88.54 86,80 85.13 95.80 * 98.0
3 S a S 9 s a s s a 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 It is dear from the above tabla that the dissolution of uraniua is ndnlinun in ease of Narvapahar and maximum in Jaduguda and Bagjata. In aU the eases more than 703* Cranium is leached out in first two hours. la ease of Jaduguda and Bagjata oore than 9G6f vtraoium is leached out in first two hours then dissolution booo&es very slow. It nay also be con cluded that 12 hours Is optimal leaching tine for all the ores because there is very little increase even after leaching for another 12 hours.
(o) Effect of Tanperaturet All the ores ground in the range of 60 to 65$ -200 mean were ^eaohei under identical conditions at different temperatures. The results are shown In Table-VH,
P.O. *.-, 60? solid pH « Jo6 - 1.7 faf w -aop + esar Laanhlng tJao a (lit} teours. 474 -
Temperature i P e x cent L e aehln g °C Naxvapahar Turendih i Bodal J&jwal J 1 i i 50 74.82 84.57 79.41 71.5 40 78.57 85.00 81.74 82.00 SO 80.71 87.04 .?2.75 83.65 60 81.45 88.0 80.26 84.70
In all the caeeB dissolution of uranium increases with temperature but the reagents consumption and gangue dissolution also increases. On leaching above 50°C the difference in leaching efficiency is very email bub the reagent consumption and gangue dissolution go very high. Therefore the optimum temperature for leaching has been recommended as 45°C - 50°C.
CONCLUSION
The grinding and leaching characteristics and the reagent consumption in leaching of different ores have been compared and the results are summarised in Table-VIII. For conparision of the grinding characteristics of different ores the size analyses were done on all the ores ground for 50 mts. All the ores were ground in the size range of 60 to 6556 -200 me ah and were loacbed under following conditions*
Leaching time = 8 hours pH = 1.6 t 0.1 1st hour 1.8 + o.i for r«ot 7 hours &nf = -500 + 50 mv
Temperature = 50»C for conpariuion of leaching characteristics. - 475 -
TABLE - VIII
Recovery of Uranium from Jaduguda, Narvapahar, Turamdih, Bodal, Jajwal and Bagjata Uranium Ores. .'
Size06) ! Jaduguda J Jajwal j Narwapahar j Turandih j Bodal | Bagjata
+ 48 NIL NIL 0.53 NIL NIL - + 65 O.20 2.00 0.72 0.40 NIL -
+ 100 4.20 7.00 2.53 1.40 13.98 -
+ ISO 4.80 13.50 4VS3 2.10 14.28 - + 200 14.70 20.00 6.25 3.50 9.86 -
- 200 76.0 57.5 85.01 92.60 61.88 -
Head Assay » W 0.06 0.052 0.056 0.048 0.0986 0.053 Leaching 95.0 83.26 82.14 89.79 83.16 . 96.0 Efficiency (*) Reagent Con sumption (KgA) V°4 16.0. 54.20 27,70 16.66 ; 36.05 17.0 lyroluaite (MnOg-58.67*) 5.0 15.25 11.55 2.54 14.00 3.7
It is setfi from the table that Narwapahar and Turamiih ores are soft to grind whereas ores of Bodal and Jajwal are hard to grind. The dissolution of uranium is more than 95% in case of Jaduguda and Bagjata. In case of Narwapahar, Bodal and Jajwal the dissolution is only of the order 82 - QZ% whereas, in Turamiih it is more than Narvapahar, Bodal and Jajwal but less than Jaduguda and Bagajata. This difference in uranium dissolution probably, depends on the association of uranium with refractory minerals. The leachant - 476 -
and oxidant consumptions are less in the case of ores where leaching efficiency is more. ".' -.'"; ;,
ACKNDUUDGBlHfl:
The authors wish to thank Chairman and Managing Director, Uranium Corporation of India Limited for his teen interest in the research activities of C.R.&D Department.
REgffiBTOES
1. Fronde 1, C. " Systematic Mineralogy of Uranium and Thorium ", . U.S. Geol. Survey Bull. 1064(1958).
2. Clegg W.J. and Foley D.D., • Uranium Ore Processing ", Jddi'son - Wesley, Beading, Mass, 1958. •
3. Eligwe C.A. and Torma A.E., n Leaching of Uranium Ores with the
n^02 -.Ha_S04 - H^SO^ system. « J. gydrometallorgy 9 , 83-95'
(1982). ' t : ; .
4. Bobbert J. Ring^J? Leaching Characteristics of/.Australia n Uranium Ores. ", Proc. Jbistralias. Inst. MinJletallKoi272, December, 1979. - 477-
wsansu OP ORAHirM BI pmaa LOW ACID LSACHINS EBOH QOPSSt. OOSCBgBjgoa TAILItCS
. V.H. Panday, B.O. Choudhary, A.K, Barker, A.P. Bsnerjee AJB. Chakraborty ar.i H. Matty
PEUHIUM ODBPOHJCIOM OF IKPIA UMETH) JAOTQUDA MINES
BIHiS
It is already established tbst eoppar ores of Singhbhum Thrust Bolt coo-
•tain on an average 0.01* ^GQ. Uranium ia being recovered from tbe tail ings of Bakha, Surd* and Mosaboni Copper Concentrators of HCL by tabling method but tbe recovery is low. Since the proven resources in the country are United therefore to meet the uranium requirement it has to be obta ined from all the available sources of Uranium and has to be recovered to the marl rami possible extent. The recovery of Uranium by direct leaching method from the Copper Concentrator tailings in the Slnghbhun thrust belt will be a, step In'this direction.
This paper deals with the leaching characteristics of above mentioned copper: concentrator tailings. The leaching efficiencies were determined both under low-acid, condition i.e. at pH 2 to 2,;"i and also under standard conditions fixed for leaching of Jaduguda ore on laboratory scale. The leaching efficiency was found to be acre or less the sane but under Jadugoda ore leaching conditions the reagent consumptions vera high. There, fore it was deolded to conduct large scale tests under low sold conditions and the tests voire conducted at 200 kg real., Suitable leaching coalition were temp, 40 - 46*0, pB S to 2.2, enf -500 «v and leaching tine 8 Hrs. under these conditions ttjOg ^aching efficiencies were 80-82* from Bakha Oopper tailings, 88-90* from Surda copper tailings and 78 - 90* froia Kosabonl Oopper tsiHnas, mtaoauccioM The •asentlallty of the development of uuoltar power generation has in- - 478 -
creased because of maldistribution of the fossil energy resources, their ultimate shortage and harmful effect of their burnings upon the environ ment. The requirement of uranium to meet the demand of power generation in India has also increased. As far as J>3 known at present, this has to be obtained from the low grade deposits, assaying 0.01 - 0.085t UJOQ. Most of these deposits are located in South Bihar in the Singhbhum Thrust belt. The Jaduguda deposit is one of the best out of these.
At present the uranium is recovered from Jaduguda and Bhotin Ores. A nmall quantity is also recovered as a by-product from the tailings of Copper Concentrator Plants, in Singhbhum belt, by tabling method. The recovery of Uranium by tabling method is poor. Since the decision has been taken to treat the low grade ores of B hat in, Narwapahar and Turamdih, some incen tive exists to examine the possibility of improving Uranium recovery from Copper Concentrator tailings by direct aoid leaching, because in this case the crushing and grinding cost is eliminated. This paper attempts to describe the present status of study on extraction of uranium from copper concentrator tailings by direct low acid leaching method.
SXPggMBffAL
1. Monthly composites of the copper ores of Singhbhum Thrust belt mined and processed by HCL are being analysed for O^g content in our lab oratory since the past several years. Monthly composite samples of copper concentrator plant tailings are also being analysed for its
size and UgOg OOQtoaba
2. Leaching experiments were conducted first on SCO gms scale in Laboratory Laboratory data were confirmed by doing the leaching tests on 200 kg scale in rubber lined leaching tanks.
3. Ion exchange studies m also oorxlucted for determining the loading
to W 0hAiOa ZTlcopper tail£ s Tleach° liquor^ . *«*••*•"•• 0^ th^resin with the - 479 -
BESPITS AND DISCUSSIONS
1. Chemical Assay and size .Analyst BI-
The monthly oompo sites of Copper Concentrator tailings from Surd a, Rakha and Mosaboni have been analysed for several years and thoy
have boon found to vary from 0.0073$ to 0.013$ t^0Q in caso 0f Surda,
0.0078$ to 0.015$ 0308 in case of Hakha and 0.00656 to 0.009$ Us°8 in caso of Moeaboni. The size analysis of a typical sample from Stkrda, Rakha and Mosaboni Copper Concentrator tailings io givan in TablB-I. TABLS ~ I SIZE ANALYSIS Qg IHB OOPPBt CONCaffBATOB. TAIUHGS
Mesh Size i Percent Weight (Tyler Standard / Surda Rakfaa Mosaboni Sieve) ! 1 + 48 9.2 2.0 2.1 48+65 4.0 4.5 5.1 65+100 8.7 9.5 10.2 100+150 11.8 16.5 14.1 150+200 16.2 16.0 17.5 200+325 36.5 40.5- 38.3
3ssaa = = = — s —
It la clear from the table that 50$ -200 mo ah (Tyler standard Sieve) is the optimum grind size for the recovery of copper by floatation method..
The tailings of above size has to be leached for uranium recovery. In Singhbhum belt 60$ -200 mesh is the optimum grinding, for optimum diab solution of uranium in the ore. Further grinding has very little effect on leaching of uranium from Uranium ores.
2. leaohinqi.. leiohing testa were conducted on 500 gm scale under varying 480 - coalitions keeping the palp density constant "*t 6036 solids. Leaching tests vote conducted for about a month on fresh samples collected over/ day. She average leaching obtained In caso of Surda, Sakha, sod Mosaboni Is sonarieed In Xablo-.II.
It is clear from tablo-II that the leaching efficiency-is more or lass the sans at pH 1.6 to 1.8 and pH 2.1 to 2.2 but the difference in re agent consumption is substantial. It can also be concluded that about 82 - 83$ uranium can be leeched out from Sakha and Surde Copper tailings and 7835 from Mosaboni tailings under low sold leaching condition.
From the data given in above Table-Ill it is clear that about 78 - 805* of Uranium can be dissolved from Rakha Copper tailings, about 88 - 9G0C from Surda and about 77 -. 7031 from Mosaboni Copper Tellings. On Pilot plant scale leaching teste the uranium leaching efficiency in ease of Surda is appreciably. higher than the results obtained on laboratory scale leaching tests. While analysing separate bore hols samples we have seen that leaching efficiency differ* from one bore bole to another, Probably while conducting pilot plant seal* teat the ore was being mined from different some and in this tone the leeching efficiency varied from 88 - 90*. . .
Comparison of the Beoovcry of Uranium by Two Alternative Bootes, i.e. by' Tabling and by Direct lev idd leaching*
The recovery of uranium by direct low add leaching has been compared with the recovery by tabling method. The results are summarised in Table-UT. It Is clear from the table that by adopting direct low acid if«M..g route more than double of the uranlm recovered by tabling method can be recover, ed from Bakha and Ssrda end more than three and half times can be recovered from Hosaboni. The total recovery from three concentrators «*•»! be about 520 kg per day i.e. about 8 H.T. par month, about half of the Jaiuguda -"I production. 481 -
MWffS OP UB.SCUS LBiCHDIG TESTS OH SUBPA. HJLKHA UP HOSIBONI OOPPHl OOHOBgRilOB. TAHJSG6
Copper tails Bipari-IEaal Assay Leaching Conditions Beagents oonroaptlon £ Leaching moatMOt I Trap, pa flsT "iisr Acid fyrolualta/ SgA Hrs. 1 Rakha Coppar 0.0102 BOOB teap. 1st far.. 2.5 450-500 7. 5.6 1.5 77.9 Tails (23«C) 6 hrs.2.1^2.2 0.012 8 1st far.-.1,8 450-500 7 7.5 1.9 77.4 6 fare. 1.6 -1.7 0.0102 45-50 1st hr, 1*8 450-500 7 8.5 2.8 83.0 6 fare, 1,6- 1.7 . 0.0102 45-50 1st far. 2.3 450-500 7 5.6 2.4 82.6 6 fara. 2.0 - 2.2 Snrda Copper Tails 1 0.Q09 55«C let far. .2,2-2.5 450-500 7 4.5 1.80 82.20 6 bra. 2.0-2.2 0.0090 45-50 let far. 2.2 - 2.5 400-420 7 5.80 1.40 80.00 6 brs. 2.0 -' 2.2 3 0.0090 " • 440-470 7 4.0 1.85 82.20 4 0,0090 • 1st far. 1,8; 6 fare.1.6 480-550 7 6.0 5.0 85.5 Kosaboni Copper 1 0,0074 45«C l6t far.l.8;6 hra.1.6-1.7 450-500 7 15.52 4,18 82.43 Tails 2 0.0074 • 1st far.2.3j6 hre.20-2.2 450-500 7 8X0 4.46 78.57 3 0.0082 32»C lit far.2.4:6 fars.2.5 475 7 7.0 4.0 67.07 - 482 -
LARGE SCALE. IDW ACID LBACMKG TESTS OH COPPER CONCENTRATOR TAILINGS FBOM RAKHA, SUK3A AMD MOSA80NI
After confirming "the leaching parameters in Laboratory, the experir-ontB were conducted on 200 kg. scale in pilot plant. The data are summarised in Table-IU.
TABLE - III Leaching Tests on 200 kg 3cale on Copper Concentrator Tailings frpn Rqkha, Surda and Mosaboni.
Common Leaching Conditions
Pulp density = 60S solid a pH 1st hr. = 2.3 6 hra. = 2.1 - 2.2 en£ - -450 -500 mv
I \ \. Copper Concert-JExpt.j Head As say IT eop. of uf H S o i Pyrolusito 1* Leaching trator Tails | No. \ % OjOg jleachiag| c£nsuni-| con sumption J I ption | KgA KgA J
Bakha 1 0.0102 R.T. 4.2 2.0 78.4 (28°C) 78.5 2 0.0102 45-50 4.7 3.0 3 0.0103 45 - 50 5.6 2.8 80.0 4 0.0105 45 - 50 6.6 3.0 80.6
Surda 1 0.0114 R.T. (30«C) 9.0 4.0 88.4 2 0.0114 45 - 50 9.5 5.5 91.1
Mosabpiii 1 0.0075 R.T. (25«C) 8.3 4.5 74.6 2 0.0075 45-50 8.75 6.0 77.3 3 0.0092 45 - 50 11.11 6.5 77.2 - 483 -
T43LS- IV
COMP/RISION Qg IHB RECOVERS OF URANIUM BY AUiKNATIVB BQUTES failings of } Rakha I ftirda I Mosabonl Cu Conoenfcratesj^y J By low |~By {By low acid I By fay lo. j tao-^gl acid j tabllngj • leaching I tabling leaching | | leachingj j <
l.Cu Ore treated oooo 1000 1000 1000 2500 2500 S/dey
2. Cu tails 940 940 940 940 2350 2350 available T/day
3. % U308 in tails 0.011 0.011 0.011 0.011 0.009 0.009
4, % Recovery in 38 81 4,2 91 23 78 I abllng/leaching
5. % Recovery at MDU stage ^35'1 W-1 36-5 85.5 20 73,3
6* Ua°8 P'OBent in 940/2350 T of lo3«4 103.4 103.4 103.4 211.5 211.5 tailings, Kg
7. UgOg recovered per day; % 34.2 78.7 37.7 88.4 42.3 l£5.0
8. T D508 in MDU,
produceprodu d per io.2 23.6 u.S 26#5 u#? 46.5 annum(30ariP'TO 0 working daye) - 484 -
IOH EXCHANGE STUDIES
In general about 200 litre of leach liquor and washings were obtained per 200 kg of tailings leached. The adsorption and elation studies vers conducted in a 50 ml column using De-Acedite FF-530 resin. Xha contact time of 3 minutes were used during adsorption. The elution waB done with 1 M NaCl + 0.1 N sulphuric acid with a contact time of 10 minutes.
Saturated loading for Mosaboni, Surda and Rakha was 47.6 g UjOg/l, 60.7 g UJOQ/1, 65.6 g u^Og/1 respectively.
M.D.U. PREOIECTAtlOHt- M.D.U. precipitated from eluted solution assayed
64% to 68% OJOQ on dry basis in all the three cases. Alternatively a crude cake of uranium can be precipitated directly from leach liquor and may be sent to Jaduguda for dissolution and processing alongwith Jaduguda leach liquor,
C0NCL3SI0N
Uranium from tailings of Copper Concentrators of Singhbhum can be recovered by direct low acid leaching method. Jbout 78 - 80% Uranium can be dissolved from Rakha Copper tailings; 88 - 90% from- Surda Copper tailings and about 77 - 78% from Mosaboni tailings. Jkb M.K.U. stage the urahlun recovery will be 76%, 85*5% and 73.3% from Rakha, Surda and Hosabonl respectively.
ACKMOMLEDGBiBfl
The authors wish to thank Chairman & Managing.Director, Uranium Corporation of India United for hie keen Interest in the Research activities of C.R.&D Department. - 485 -
SELECTION OF ION EXCHANGB RESIN FOR URANIUM ADSORPTION FROM JADUGUDA LSACH LiqUOR
D.P. Saba and V.M. Pandey
URANIUM CORPORATION OF INDIA IIHirSD ' JJDOGUDA MINES SINGHCSHM BIHAR
The ion exchange reains used to purify and' concentrate uranium in treat ment plants are strong base resins of the quarter nary amroniura type.
They fix the uranium in the form of anions UO (OO3)- or D0g (S04)3 . In general, they are used to recover uranium from leach solutions conta ining between 5 and JO00 ppm uranium. The resin UBed for the adsorption of uranium should, have the following properties: (l) Higher Osmotic stability (2) Higher reaction rates (3) Higher resistance to organic fouling (4) Smaller swelling difference in polar and non polar solvents.
These superior properties along with the performance of resin culminate in the realization of maximum operating life and consistency in the performance over extended period of operation.
Since the'quality of resin plays, a very important role in the' has been found a batter substitute for Indion ARU-103 resin. INTRODUCTION Ion exchange has found wide use in the treatnent of water, effluent a and in the field of hydrometallurgy. In hydroraetallurgy, recovery and • purification of Uranium using synthetic organic ion exchange resins la. by far the largest single application. The use of anion exchange resinB for the recovery of uranium from the sulphate laach liquor has been one of tho most interesting contribution of chemistry to the mining and metallurgical industries. In a period of only 8 years it ha3 grown from the laboratory scale to an established process in largo scale use in mines all over the world. The first small scale column to recover uranium was set up at the Western Reefs mine, Transvaal. The results of this work* 'showed that a practical recovery process for uranium hod been found, Tho uranium complex was selectively adsorbed giving a high resin loading inspite of competition from high concentration of other ions in solution. On elation with ammonium nitrate solution a relatively rich and pure uranium solution was produced. It has since been shown that many other salt solutions may be used as eluting agents. Following the • pilot plant test of two south African mines, tho first full seals plant was brought into action in 1952; and tho bull/ling of new units continued without interruption. In UCIL uranium recovery plant, the concentration and. purification steps in tho recovery of uranium from Jeduguda uranium ore is being done in fixed beds of strong base ion exchange resin. Tho method of application has genor i.y taken the form of pressure vessel having a liquid collecting system filted with strainers across tho bottom of tho vessel upon which a bed of ion exchange reain iB imposed. Tho liquid flows down through tho bed whore masB trannfer occurs and the treated liquid is collected at tho bottom. Tho selection of proper resin is vary important for-tho ion exchange proceso. In tho beginning an imported rosin I>e-Acidito FF-530 WUB used for purification and concen tration of uranium from Jaduguda loach liquor. All along effort was mode in this laboratory to find an alternative lndigonous resin to re- - 487 - place the imported resin. To achieve tbo sane the imported resin samples were supplied to Indian manufacturers and they were asked to produce the similar resin. Ultimately M/B Ion exchange India Ltd., came out with a resin namely Indion JUIU-103 which was found at par with Oe—Acidite FP-530 and since then this resin is being used in the plant. To find the alternative sources other manufacturers were approached toiproduce resins suitable for uraniun adsorption, ill the resins indigenously produced for uranium adsorption were tested in our laboratory and some of them have been found suitable for use in the plant. This paper deal3 with the adsorption and elution characteristics and other details of the resins tested in our laboratory. Indion ARU-103 and Duolite JU101 D/U are two such resins that have been found suitable froa the study of eight indigenous resins. THEORETICAL Ion exchange technology is usually attractive for uranium hydroraetallnrgy since there are a host of ion exchange materials available which will selectively remove uranium from both .acid (HgS04) and carbonate (HagCOj) leach liquors containing any impurities and which can readily be stripped with relatively cheap reagents. Highly pure concentrate can be precipd^ tated out from the stripped sjlution after the ranoval of iron from it. The adsorption of uranium in a colunn is controlled by such operating conditions as tho flow rate of uranium leach liquor, amount of adsorbent and tho adsorption ability of tho adsorbent. Uranium forms' anionic sulphate complexes in tho VI Oxidation state in acidio aquas sulphate solutions. In tho absence of other anion forming to) impurities, tho complexes of uranium sulphate can bo written as follows 2+ SO. » it/, _. ™z + so4 -_ ^ w> so4 (1) JZ 4 - 488 - 2~ IJCsSO,. + SO. N. uo2(so4)2 ...... (2) "w2 4 4 X w>2(so4)2 + so4 N + - 2 V H4.SO4 N Reaction(4) occurs in dilute acid solution, i.e. below about 0.5 M and this applies to uranium recovery since the pH value is normally adjusted to the range 0*5 - 2.0* The process involved adsorption of uranyl sulphate complex on a strongly basic anion exchange resin. The following are the possible ion exchange reactions: 2- 2- BjS^ f UDj, (S04)2 _ >, R»g (S04)2 + S04 ...(5) ^ . 8R 80 + 30 a (S0 • fe 4 ^z^ ^ ^ 4% 4)5 + 2So/~ *••& Hare R denotes the resin matrix structure and the over bars denote the species with the resin phase. The uranyl sulphate can be stripped from the resin.'/i with nitrate, chloride or sulphate solutions* In Theory, it is possible for the neutral uranyl sulphate complex to diffuse into the resin matrix, hut equstion(7) is unlikely since-the extent of hydration of the ion is high and it is therefore unlikely to make much actual progress into the resin. Therefore equation(5) and (6) are the most likely and it is possible that they occur simultane ously to a varying degree depending on solution eruditions. Irrespective of which of these reactions occur, the mechanian will involve counter .diffusion of ions through a nugtber of possible resistances. . - 189 - In laboratory absorption and elation tests were performed using 50 ml vet settled samples of different resin In a glass oolunn of 35 aa long and 2*5 aa diameter. Anion exchange resins were supplied In sulphate form or chloride form. Tie resins in chloride form were converted to sulphate form by dilate sulphuric add solution(pH 1,5). Before use the resins were kept under acidified water for 24 boors for swelling and So ml of this wet settled resin was packed in glass columns of dimensions mentioned above. The resin was fashed with five bed volumes of 1.5 pH water which was passed through the column at the rate of 5 ml/minute, ifter washing the resin clarified lsach liquor was passed through the column at the rate of 5 ml/minute keeping the contact time 4 minutes. Samples of effluent liquor (barren) was collected at regular Intervals for estimation of UgOQ leakage. The number of bed volumes of Isach liquor passed at which the UJOQ concentration in the effluent was l£ of 0*0g concentration in the pregnant solution was noted as break - through point. The passing of the pregnant through the column was con tinued till the U3°a concentration of the effluent was found to be equal to that of the influent. This was the point at which the resin was satu rated to its full extent with respect to uranium. The flow of -pregnant liquor was stopped and the residual liquor from the coluan was removed by passing 5 bed volumes of acidified water (pH 1.5). The saturated column was eluted with 1M brine solution acidified with sulphuric acid (acidity 0*12 N) by passing at the rate of 2 ml/mlnute keeping the contact time 10 minutes, till the resin was practically free from uranium. Each bed volume of the eluted solution was collected and estimated for uraniwn to find out the loading capacity. The residual eluting solution was removed by passing 5 bed volumes of acidified water (pM 1,5). The column was ready for the next cyole. On several cycles of operation the resin was ologged by tiny suspended particles present in the leach liquor and was pressed to some extent, - 490 - causing obstruction to the freo flow of the liquor through the resin. The column was then back washod by passing acidified water (pH 1,5). Size ana3ysie of the resin was done by wot serving method using the Tyler standard screen. To estimate the silica content of the resin it was ignited and silica was estimated from the residuo(ash). RESULTS AND DISCUSSIONS 1. Analysis of Leach Liquor;- The analysis of leach liquor which was used for adsorption and elution studies is given in Tablfi-1. Z. Size Analysis of Resin:- The results of wet sieve analysis of 10 resins are given inTable-II, 3. Adsorption Characteristics of Resins; Adsorption characteristics of 10 different rosins wore studied. After the first cycle of adsorption it was found that some of the resins were not suitable for our use. The resins which gave encourag~ ing results in the first cycle of operation, their adsorption charat>. teristics wore studied for several more cycles. Figure 1 gives the adsorption behaviour of resins in first cycle of operation. The adsor ption behaviour of the resin after continuing for many cyclBs is given in Figuro-Z. From Figuro-1 it is clear that the adsorption characteristics of De- Acidite FF-530, Indion AttU-103, Duolite A-101 D/U, Tulsion A 36 MP and Agrion A-600 aro more or loss similar. These rosinB have lata break through and quicker adsorption. In case of other rosins the break through is early and the adsorption 1B slow which can be saon from the flat nature of the curve. Therefore it was decided to carry on more cycles of operation in the caoo of above mentionod 5 rooins. The adsorption curves after several cycles of operation are given in Figure-2. From the figure it is observed that in 15 cycles of operation in the case of Tuloion A 36 HP rosin the adsorption becomes alowor - 491 - though it8 loading remained the some. The change in adsorption chara cteristics may be attributed to high silica odsorption by the rosin during the length of operation. The adsorption characteristics of other resins did not deteriorate much. 3. SLution Characteristics of Realnst The elution curves for first cycle of operation is plotted in Figure- in. Elution curves for 5 promising resins after different number of cycles of operation are given in Figure-IA'. Resins like Duolite A-101 D/U, Tuleion «i 36 MP have high peak values and their curves have sharp fall indicating quick elution. The resinB like Indion ARU-103, De-Jkcidite EF-530, 4grion A 600 have moderate peak values. In Indion J&U-103 and De-Aoidito FF-530 elation was quits fast while that in the casa/Agrion JU600 it was slow requiring long elution time resulting in lower concen tration of U308 in the eluted solution. The peak value for Dosion 4SB 108 was very low and the elution curve was sluggish. The results of the Ion exchange studies are summarised in Table-Ill. CONCLUSION Ion exchange studies wore done on eight indigenous resins for finding out their suitability in our plant. For comparision of characteristics of Indian resins two imported resins were also studied along with them. Anong the eight Indian resins studied six were not found suitable. Thus we have found an indigenous alternative to Indion ARU-103 resin. ACKN0MLED3B-1KNT The authors wish to thank Chairman and Managing Director, Uranium Corporation of India Limited for his keen interest in the research activities of C.R.&D Department. REFglENOBS 1. Ardon T.V., Brunner J.J., Macdonold R.D. and Olson R.S. GML progrooa report (leaching No,42, 1950. 2. Streat M and Tokel C.M.J., J. Inorg, Nucl. Chem.43, 807 - 830(1981). -492 - TJBLE- I JWALYSIS OP THB LEAGUED LIQUOR pH . 1.7 to 1.8 Bnf = -450 to -520 wr %°8 = 0.43 to 0.61 g/1 P2°5 = 0.41 to 0.52 g/1 CI = 0.31 to 0.4 g/1 so4 = 12.0 to 15.0 g/1 Fa* = 1.0 to 1.5 g/1 SiO = 0.86 to 1.02 g/1 z TJBLS - II SIZE ANALYSIS OP THB MSI US Size In Name of Resin Mesh 6 8 9 10 + 14 (£) NIL NIL NIL NIL NIL NIL NIL NIL NIL HIL * 20(55) 42.0 23.99 55.0 98.69 2.78 1.70 15.37 26.01 49.49 42.67 + 35($) 55.1 68.11 41.0 1.31 94.15 93.87 82.91 72.60 50.18 56.49 - 35(35) 2.9 7.90 4.0 Trace 3.09 4.43 1.72 1.39 0.33 0.84 50 ml of each MSR was wot solved through standard sieves. - 493 - TCLS- m APSOBgPH AMD &LPTIPH CHARlCTggSriC8 0? THE BK3IH3 t H«e of SMISI 1 2 5 4 5 6 7 8 9 10 Cycle Ho. 1 1 1 21 1 16 1 1 15 1 1 1 15 1 20 1 Break through 50 70 66 55 50 75 75 55 95 55 85 80 70 65 50 (0.005 gA OjOg) BV loot saturation 200 ISO ZOO 160 190 180 170 240 200 280 190 200 ISO 200 220 B7 loading capacity UjO, g/1 KSa 61.58 81*0 75.5 75.7 71.0 85.74 73.5 73.0 62.78 82.24 74.19 75.0 70.2 69.9 74.68 Silica content - 0.4 0.92 Trace 0.05 - Trace 10.04 Trace Trace Trace 0.05 Trace 0.80 Trace Button peak Yalne g>85 14>6 M<5 ^ ^ ^^ ^j ^^ ^^ w#26 1?<88 14^ 14 • Hame of the Basin i Huaber l to 10 stand as for toe following resins 1. Dofdon 1S8 108 5. Tulaion X 56 (KP) Cl~ strong inlon Exchanger Type-II 2. igrion 4-600 6. Tulsion 4-27 (GEL) Cl~ (UGS) Strong inlon Exchanger Type-I. 5. Be_4cidlte ff-530 7# TulBlon A-27 wok anionic exchanger 8. Duolite 4-101 D/U chloride form 4. JBi-430 s> Tnrf-;«n JBII_103 10. Contex B 200 strong base anionic exchanger Type-I. - 494 - APS OR P.fi QN CH A RACT ERISTICS ----- ~£Jk$T^C%-CL£ 0- S&B'fc&'i A *t£ U>3 7ft--'0€AC+O+TE- FT 530 JTJ AGftlOH A 600 V JOUOLUE A lOt D/U ;iAJ;-|rUL^J0N-.Av2-7- j W- JDOStOH ASB 108 >T IRA: 4io. L —t- & ;^^ iV:.^0 : . 100 »20 MO . 18 0^ 2^0. 260 .".. t"E JiVCiHT | U Q(JO fc PASSED .t •• -FJG-*!- •araiaMin '-ill. 495 - H- ADSORPTION CHARACTERISTICS OF PEStNS AFTER SEVERAL CYCLES 100- 2 _; < Q IN DION ARU 103 ^ 20 TH. CYCLE 60 ; < A DEACIDLTE FF 530— 16 TH. CYCLE 4Jft- • Q "AGRION '-. A $0© : T—r. 2» ST.1 CY.C.U E 2 • ;' ~ • i r U y: JDUOLffe-A »0I 0/U-I3TH. CYCyE Q: • "rUL'siON A 36 MP —15 TH.. CYCLE •ah i& • *%» W8 • "tefl 't4^ 'ito m •>«» '*zo u24 0 *260 2*Q LEACH iLI^UOf? PASSED (BVJ > FIG.-- 2. '2. > " i96> ELUTON "CHARACTERISTICS OF RESINS FIRST CYCLE t .. . 9 INOION ARU 103 £ DEACIDiTE FF 930 ...... - AG RIO N A 600 r:-:i*' EJ :.. ^ DUOLITE A 101 D/U I*. • •:- JULSION A 36 M P ...„• : J|4- f TULSION A 27 : po$iON AS.B: ioa ' P •" 1- ns iff. T- r.ft.A: 490 :^K »••*• •••• »'•• ' I -.11 I'.:'. lO : i''-i • a... '/ 4 -. I* tf«. W. 1 ELUA^IT PASSED feV-<-^ . !. : ••;.. 1: •• •• j • •• .(.:..• .:.,„ * .» » , , - <&l - ELUTfON CHARAGBRrSTICS OF RESINS AFTER ! SEVERAL CYCLES •4 O IN010N ARU 103 20TH. CYCLE A DE ACIDISE rr 530 I«TH.CYCLE ( * t fTJ AGRION, fl 600 21 ST. CYCLE , \z - V DUOLJTE A"IOr 0/U IS TH. CYCLE 11 £ TULSiOM A 36 MP ISTH CYCLE | ":-4 - • »• v ^T£! * .. '•: • ' * •7 '/• .•iv.-; Y< • '.-.'• ' • : 1 .1 a .1: * 1" if • 1 ' » » ., • > •• - 2 -44— t • 10 12 14 It IK ELUANT PASSED (BV)- -'.FIG .-4 - 49» - APPLICATION OF ADVANCED TECHNOLOGIES FOR URANIUM MINING AND PROCESSING AT NARWAPAf-'AR AND TURAMDIH PROJECTS R.C.PURI-CHIEF SUPT(PROJ)-UCIL P...? .VERMA-ADDL .SUPT(MILL)-UCIL UCIL has started construction work on two mines, one.each at Narwapahar and Turamdih Projects and a combined proc essing plant at Turaradih as a part of the country's ambi tious Atomic Energy Programme. The adoption of latest co ncept of declines as mine entries will enable completion of project in 4 years only and will also allow large mech anisation underground. Use of latest world technology of L P D trucks, L H D vehicles, Drill Jumbos, Scissor lift, Passenger Carrying and Service Vehicles will result in rapid development progress rates and large production from concentrated work places. Mine layout providing access ways in waste to orebodies and use of high capacity high pressure fans for ventilation will enable adequate control on radon in mine workings. Process Plant has been desi gned based on experience of Jaduguda Operations and inf ormation/Data of several most modern operation of over seas countries such as Canada, USA, South Africa,France and Australia. Use of Horizontal Belt Filters for filt ration, Draught Tube Circulators for leaching and Himsley Continuous Counter Current Fluidized Bed Ion Exchange Sy stem provide high efficiency and flexibility for extrac tion of uranium together with low capital as well as ope ration and maintenance costs. The paper details the va rious methods, processes and equipments giving the bene fits derived from each. - 499 - INTRODUCTION Depletion of fossil fuels is a burning problem world wi de. India is no exception. Hence she has prudently op ted for development of alternative sources of energy. At omic energy is one of the prime nonconventional targets and requires concomitant nuclear fuel industries based on uranium cycle. The premier step in the cycle consists of raining of uranium ore and extraction of uranium present in the ore by physico-chemical process. Uranium Corpora tion of India Limited (UCIL), has been engaged in such operations since the sixties. Its established its first mine and processing plant for uranium ore at Jaduguda and started production in the year 1967. During the early ei ghties, the company took up new projects to increase pro duction of uranium to meet country's requirement of fuel for Atomic Power Plants as a part of programme of genera tion of 10,000 MW of power by 2000 AD. Bhatin Mine loca ted 3 KM west of Jaduguda was constructed and made opera tional for production from October,1986. Narwapahar Pro ject located about 10 KM west of Jaduguda and Turamdih Pr oject located about 9 KM further west of Narwapahar and about 5 KM south of Jarashedpur are the two new sites being developed. Figure I is the location map showing these uranium projects. Each of Narwapahar Mine and Turamdih Mine is rated to produce 1500 tonnes of ore per day. Ore from Narwapahar Mine will be transported to Turamdih site by monocable aerial ropeway and a combined process plant will treat the ore to produce magnesium diuranate yellow cake. Uranium is a mixture of U238, U235 and U234 as they occur in nature. Uranium is radioactive and its ores contain a suite of radioactive elements. During disintegration of - 500. uranium* radium is formed. The disintegration of radi um results in liberation of radon gas. This radon .gas along with its shorlived daughter products which are po tential alpha emitters are of concern in uranium mining and processing. Water from uranium mines and process pl ants thus contain radionuclides of uranium and radium out of which radium is of prime concern. MINING In mining industry the world over, there have been great technological advances during the last quarter of century. Uranium mining is similar to other non-coal mining. So far it is restricted to underground mines only because of the occurence of deposits in depths. Indian metal mining industry has also seen.advances in mechanisation. You have already heard about advances made at Jaduguda Mine and Pr ocess Plant. Modern advanced technology has been selected for Nazwapahar and Turamdih Mines. Though some of the eq- ipment would be used for the first time in the country,ca re has been taken to see that its use elsewhere in the wo rld is well proven. The application of advanced technolo gies is detailed herein broadly. TRACKLESS MINING USING DECLINE ENTRIES The most important advanced technology to be used is the introduction of trackless mining methods using the dec line system for mine construction. Declines have been co nstructed so far-in the country at ..two mines only. These provide easy means for rubber tyred trackless equipment to enter the mine directly from the surface. The figures 2&3 show the mine entry system at Narwapahar Mine and Turamdih Mine respectively. The size of the decline is primarily - 501 - dependent upon the size of the largest vehicle which has to enter the mine through It. The figure 4 shows the cross-section of the decline having dimensions selected according to the size of low profile dumper to be used at the mines. For lowering such large size equipments th rough, shafts, these would have to be dismantled and lowered part by part and then assembled again underground. It would be very difficult to bring these equipments to surface for maintenance, repairs, running etc. time and again. So declines provide means of large scale mechani sation. Further in case of declines, you need not wait for its complete construction and equipping before under taking development work off it. Access cross-cuts to the ore-body can be driven as soon as the requisite depths are reached and development or production work can be started at higher levels simultaneously with construction of de cline lower down. So the mines can be commissioned for production early. These advantages of declines can be understood easily from figure 5 which shows various modes of mine entries* USE OF RAMPS INSIDE STQPE WORKINGS At Narwapahar, for ore-body thicknesses over 2.5m, step mining methods and for thickness over 7.5m post pillar mining methods are proposed. The modern feature of the se mining methods is the use of ramps inside the stope workings. The LHD vehicles, and drill jumbos can enter through these ramps speedily and efficiently without di smantling. Through these ramps the equipment can be ta ken from one stope to the other and from one level to the other quickly. The production from such stope is envis aged at 375 tonnes per day.with O.M.S. of over 10. Such concentrated production with so high stope O.M.S. has so far not been achieved in the country. The figures 6 & 7 - 502 - show the step mining layout and post pillar mining lay out using such ramps. MODERN MINING MACHINERY Diesel driven drill jumbos, load haul and dump vehicles (LHD), low profile dumper 1. Drill Jumbo The jumbo will be two boomed for simplicity and will drill parallel Holes. It will be possible to angle the drill feed fr>r drilling vertical holes for rock bolting or service anchors. The arrangement of the booms will allow for the collaring of a cross-cut at right angles. Such a Jumbo is to be used for the first time in the country. This provides advantages of faster drilling, . precison parallel hole drilling, use of 3200mm long holes and consequently greater pull per blast and av oids hard arduous physical work which worker in gene ral als*o do not want to do. The use of Jumbo also gi ves maximum flexibility of operations. 2. LHD Two sizes of LHD's are to be used. 1.7m size is to 3 be used at Narwapahar. 3.8m size would be used at - 503 - Turamdih. Though 1.7m LHDs are being used at Kolihan Mine of M/S H.C.Ltd., the use of 3.3m size would be for the first time in the country. Its use will give large outputs in Room & Pillar mining where rooms of 5m x 5m shall be made and sub-level stoping methods wh ere ring holes will be blasted. The figure 8 shows the Room & Pillar Stoping Method (Turamdih Mine) making use • of the modern mining machineries. 3. LPD The dump trucks.are to be utilised in the development of declines, in trackless level development^ or ini tial ore-production from development and stoping in trackless areas.. A standard 22.7 tonnes rear dump truck has been selected for use. 4. Underground Service Vehicles The various vehicles to be used are as follows :- i) Supply Truck ii) Lubrication/Service Truck iii) Explosive Truck iv) Water Bowser v) Motor Grader vi) Personnel Carrier vii) Scissor Lift Truck. The use of these equipments are considered essential for the type of mining to be adopted and would be in troduced for the first time in the country.,?. These shall initially be imported but later on it is expec ted that local manufacturers would be able to develop such vehicles indigenously. With the use of personnel carriers, the work force can be transported direct to the working places, even to the stoping areas through - 504 - ramps from the surface. 5. I.T.H. Drills These will be used for drilling 115mm dia holes for stopes and pillars. Slot raise will also be mined using these drills by drop raising. These drills ha ve so far been used by only one company in the cou ntry. 6. High Pressure Main and Auxiliary Ventilation Fans. Because of the large scale, use of big diesel equip ment, ventilation requirements for the mine are high. Ventilation systems have been designed with the help qf computers. At each of the two mines, three fans with capacity of 50m /Sec. at 1.7 KPa each with 120 (BHP) KW electric motors will be used for main vent ilation. The total circulating quantity of air of 150m /sec is equivalent to 0.091m /sec per tonne mi ned for a total daily tonnage of 1500 tonnes of ore plus an allowance of 1096 adciitJonal waste. This much quantity of air required to control heat, dust and diesel emissions is .sufficient to remove radon and its daughters. For auxiliary ventilation, each trackless heading is to be ventilated with a minimum of 20.0m /sec of air. In order to achieve this standard, it is proposed to use twin axial flow fans each capable of producing 10m /sec of air at a pressure of 1.9 KPa. These would be installed in ventilation ducting of 760 mm diameter. The mine layout provide accessways in waste and entr ies to ore-bodies are made with x-cuts at suitable in tervals so that fresh air intake is provided to each - 505 - ore block to be mined. Figures 9 & 10 show such mine layouts with details of ventilation arrangements. PROCESSING The ore processing methodology being adopted is essen tially the same which is being used at Jaduguda process plant, the details of which are given in the paper on the subject. The details relevent to Narwapahar and Turamdih Projects are discussed in this paper. MINERALOGICAL AND CHEMICAL COMPOSITIONS Mineralogical and Chemical composition of Narwapahar and Turamdih ore is given in.table No.I & II with Jaduguda deposit for comparative information. WORK INDEXES Work Index of Narwapahar ore varies considerably from zone to zone but in Turamdih deposits variation is not much as shown in Table No.III. The Jaduguda ore work index is also mentioned for comparison. CRUSHING The run of mine ore is expected to contain about 5ft mo isture and fines generated during mining and haulage wi ll give rise to wet sticky ore for crushing. Hence Pri mary crushing will be carried out by a jaw crusher prior to transportation to the stockpiles to be created for Turamdih and Narwapahar ore separately. This will provide surge capacity in the plant equivalent to 3 days produc tion requirement. The stockpiled ore to be released for treating in crushing - 506 - plant by a reclaimer and hopper, conveyor belt system and shall be crushed by a standard head cone crusher followed by tertiarry crushing by two short head cone crushers operating in closed circuit to a vibrating screen producing fine ore of nominal size 20 mm. Adoption of standard Head Cone Crusher is for better power utilisation, higher reduction ratio and uniform product size. MILLING Not much changes have occured in the Milling practice of uranium ore, but in recent years many big size plants have adopted Semi-Autogenous Grinding system. The run of mine ore is charged directly to grinding mill alongwith steel balls comprising 5% of the charge. The use of such mills result in significant savings both in operating and capital cost by way of elimination of crushing circuits and near elimination of grinding media which results in reduced oxidants and acid consumptions. Other . advantages are capacity to handle wet sticky ore. But SAG system have certain disadvantages over conve ntional grinding circuits which arB; 1) Fairly accurate and extensive pre-design testing. 2) Determination of structural competency of ore to form suitable grinding media. 3) Substantial demonstrative Pilot Plant test work. 4) Fairly large tonnage of ore to be treated. These mills are not suitable for smaller tonnage like 3000T/day envisaged in Turamdih Plant. 5) Diameter to length ratio being high (D:L = 3:1) for such mills the size of mill would be very large nece ssitating import of the equipment. - 507 - The cost benefit analysis heavily weighed in favour of a single processing plant with flexibility to mix-up ores of both deposits. This did not allow consideration for Autogeneous or Semi-Autogenaous system. Moreover, import of mill drive units and control equipments would have been necessary for SAG system. Hence proven grinding circuit- Rod-Ball mill combination operating in closed circuit with hydrocyclones has been choosen for Turamdih Plant. High degree of Automotion and Instrumentation has been propo sed for the circuit by including variable speed drive for hydrocyclone pumps, particle size monitor, weightometer, mass and density meters etc. These will provide flexibi lity in optimum working conditions, achieve high through put of ore, control on quality of ground products as well as reduction in cost by way of reduced manpower requireme nt and loss of energy in over-grinding of the product. DEWATERING The ground pulp obtained from the milling circuit as Hy drocyclone overflow product will contain about 30-35% solids, where as leaching circuit's requirement would be for 60-62% solids. To reduce excess water, Thickener and Horizontal Belt filters have been selected in place of Thickner and Disc filter system in Jaduguda Plant, The choice of Horizontal Belt Filter over the Disc Fil ter was based on following reasons : o 1) Only two Belt filters of 72M area would be required against more Disc filters requirement for same tonn age handling. 2) The smaller number of belt filters result in superior operational supervisions, reduced maintenance and simpler layout. - 508 - 3) The dewatering and the leached pulp filters will be identical thus helping in standardisation and reduced . spare and maintenance costs. 4) Belt filters do not have feed pan, hence emergency power is not required, 5) Control on pulp density would be better entailing maximum consumption of acidic wash and filtrate from leached pulp filters, LEACHING Following three types of leaching equipment were con sidered ; 1) Air Agitated pachucas 2) Mechanically Agitated tanks 3) Drought tube circulators Based on cost benefit analysis and similarity in leach- ing characteristics of both Narwapahar and Turamdih ore, mixing of both ore at leaching'stage would-be attractive. The drought tube circulators are adopted to carryout this operation and major advantages considered are the fo llowing ; 1) Drought tube circulators are able to resuspend sett led, solids after power failure, without need for emer gency power or compressed air necessary for other two system. 2) Fume and Acid mist emission from draught tube circu lators is neglegible which results in improved work ing conditions and reduced corrosion on the leaching tanks and structures. - 509 - 3) Power requirement for circulators is the lowest among all the three systems hence lower operating costs. 4) Mixing in draught tube circulators is superior to the other systems, consiquently better leaching, 5) Lower over all cost. After extensive laboratory studies on leaching characteris tics of Narwapahar and Turamdih ores individually and in mixed form following leaching conditions have been fixed for the circuit :- PH = 1.6 - 1.8 Redox Potential = 500 + 50 rav Residence time = 12 Hrs. Average leached pulp temp = 30° C Temperature after heating • u * ' = 60° C by steam Increase in leaching Efficiency; a) Turamdih Ore by = 5% b) Narwapaharr" ore by = 2.5% . Acid consumption a) Narwapahar Ore = 27-28 Kgs/tone b) Turamdih Ore =14 - 16 Kgs/tone Pyrdlusite consumption (100% Mno2) a) Narwapahar Ore =5-7 Kgs/tone b) Turamdih Ore = 0.4 - 0.9Kgs/tone LEACHED PULP FILTRATION Traditionally liquid solid separation after leaching of ground ore in uranium extraction plants has been acco mplished by vaccum Drum Filters and Counter Current De- cantation. In recent past* uranium plants have been turning to high - 510 - capacity thickeners and Horizontal Belt Filters in an effort to improve the efficiency of Liquid-Sdlid separa tion and reduce operational and capital costs. However, Belt Filters have the following advantages ; 1) Relatively big filter duty and low cake moisture. 2) Counter current washing facility for the cake. 3) Flood wash at all times lead to better recovery of disolved uranium values. 4) Throughput can be varied by changing belt speed. 5) Requirement of small floor area. 6) Lower capital cost for given duty condition. 7) Variation and flexibility with regard to filter speed and wash pattern. 8) Ability to filter coarser pulp. 9) No emergency power requirement. CLARIFICATION The Uranium bearing solution produced after filtration will have high content of suspended solids about 1500- 20O0 ppm. One of the principal considerations in the process selection has been to eliminate the costly li quor clarification step, ahead of" Ion Exchange Circuit. Options available are ; 1) Precoat Filters - as followed in Jaduguda plant 2) Sand Bed Filters 3) Reactor clarifiers 4) Hopper clarifier The hopper clarifier has been considered due to their - 511 - low operational cost and on recommendation of SELTRUST U.K, the consultant who have developed and operated su ch equipments in few Uranium plants. The clarification section will have two clarifiers ope rating in parallel. The unclarified liquor from Belt Filters will be pumped at constant rate to the clarifiers and the clear pregnent solution with suspended solids around 50-200 PPM will overflow to storage tanks* Flo- culants will be injected in.the feed line to the clari- fier for promoting floe formation, a blanket of which will form in the equipment thereby trapping suspended solids contained in the rising feed solution. Solids settled at the clarifier bottom shall be removed by a timer actuated automatic value. Operating cost of this system compared to pre-coat filters would be considera bly low, while any operational malfunctioning will not affect the subsequent process due to ability of Himsley Ion Exchange columns to accept high degree of turbidity in the feed liquor. URANIUM RECOVERY AND PURIFICATION. Solvent extraction and Ion exchange are the two systems employed in Uranium Industry for this operation. The solvent extraction system, because of its obvious impact on environment and expected dillute pregnant Uranium solution (about 0.4 - 0.5 gms/lit) due to low grade ore was not considered. The final choice rested on adoption of Ion Exchange technology for the purification and reco very of uranium from the clarified pregnent liquor. The more recent inovations in Ion Exchange technology are in the field of continuous Ion Exchange System. The abili ty of this system to treat low. grade Uranium bearing sol utions without prior clarification has made it an attrac- - 512 - tive route for treatment of turbid feed stocks from the leaching of low grade ore. For Turamdih process plant, Himsley continuous counter current Ion Exchange System has been selected. This system operating on fluidised bed counter current absorption principle, controlled resin transfer in bat ches by hydraulic means and multistage counter current packed bed downflow elution procedure will produce pure and concentrated solution resulting in high grade'Yellow cake' after precipitation. Advantages of Contlneous Ion Exchange system are ; 1. Ability to tolerate high turbidity liquor. 2. Efficient use of the resin's exchange capacity resul ting in substantially lower resin inventory than in a fixed bed system of comparable U^OQ output capacity. 3. Elution procedure assures complete elution and pro vides more concentrated eluate (25 gms/lits). 4. Barren liquor^yalues can be controlled as low as lppm of UOOQ! value. 5. Arrangement for external regeneration of resin. High degree of automation shall be incorporated to con trol all operations sequentially,eliminating operational error and deployment of more operating staff. IRON PRECIPITATION As the conventional LAMIX process would be followed,pre cipitation of iron from strong cluate will be carried out in three tanks equipped with mechanical agitators, PK probe to monitor PH and addition of regents. Since the eluant for CCIX himsley columns will be 10-12$ Hg S04, - 513 - pollutants like nitrates, chlorides etc. will not get . released to environment alongwith excess barren cluate likely to be generated during precipitation process. The precipitation process will be improved further by air injection into the precipitation tanks which will oxidise residual ferrous iron into feric, thus impro ving quality of the final product* Parameters to be mentained in iron precipitation circuit are the follo wing :- 1. Precipitation tanks : 3 Nos. 2. Volume of each tank : 3.3M3 3. Final PH of precipitation ; 3.5 4. Lime slurry addition : In first two tanks 5. Magnesia or caustic soda addi tion.. . : In third tanks 6. Residence time : 5 hours YELLOW CAKE PRECIPITATION • The circuit will comprise of three Agitator tanks ^in series. The clarified liquor from the iron precipita tion section will be pumped to the first tank where Magnesia slurry addition in controlled way will raise the PH of- liquor from 3.5 to 7.3. The Uranium present in liquor will precipitate out as Magnesium Diuranate (MDU) and will be washed in filter feed surge tanks by syphoning with water to remove excess sulphate ions, parameters to be maintained in MDU precipitation circuit shall be as follow :- 1. Precipitation tanks : 3 nos. 2. Volume of each tank : 15M 3. Total residence time : 15 hrs. 4. Magnesia addition to raise PH to : 7.5 5. MDU thickener Dia : 10 meter - 5H - 6. Filter feed surge tanks : 3 nos. DRYING & PACKING Various options are availa ble for this operation,such as Automisation, Band/Chain Drier.Steam driers etc. How ever, present layout is based on electrically heated oven. The section will comprise of a rotary vacuum drum filter, a preforrner and drier system. The thickened MDU slurry after wash in the surge tanks will be pumped to the dr um filter. The filtered cake at about 50# moisture will fall through a vertical chute on to the preformer moun ted on the feed end of the driers The extruder type preformer will produce cylindrical pa llets to increase the surface area for better drying to 2% moisture content by weight in the electrically heated contineous conveyor band drier. The dried pallets after storage in bins will be packed into 200 litrs M.S.,drums for onward shipment. This system of drying and packing will generate substantial amount of fine MDU dust affec ting the working environment in the section, hence few other systems better than this are under consideration and one of them is drying by automisation. This system have distinct advantage over other drying systems beca use ; 1. Very regular product grain size within particle range of 20 to 100 micros is obtained by agglomeration du ring drying with a small percentage of particles sm aller than 10 micros. Healthier environment for op erating staff due to reduction in quantity of fines and to the use of totally enclosed drying circula tion system. - 515 - UCIL is considering to import if necessary a totally dust free automatic plant for this purpose. TAILINGS NEUTRALISATION Maximum use has been made of recycling liquid solution, like barren liquor from Ion Exchange columns which needs to be neutralised as being acidic in nature before dis posal. The Horizontal Belt filters' cake being the solid waste will be repulped with Mines Water or excess Barren liquor and the resultant pulp to be neutralised to 10.5PH in Draught tube circulators*three numbers in series pro vided for this purpose. Provision would also be made to inject small quantity of compressed air in circulators to oxidise ferrous to ferric ions to ensure complete iron precipitation and to assist in manganese removal. Neut ralised tailings product will be classified to sand and slime fractions, sand being used as backfill in mines, will be stored while fines will be impounded in a speci ally built tailings pond. To reduce water input in the pond, slimes prior to disposal shall be thickened to 50$ solid content. ACKNOWLEDGEMENT The authors are thankful to the Chairman & Managing Dire ctor, Uranium Corporation of India Limited for his guidance and permission to present/publish this paper. REFERENCES Detailed Project Report, Uranium Ore Mining and Processing -—Project at Narwapahar and Turamdih. - 516 - TABLE - I Mineralaaical Composition Items Narwapahar Turamdih Jaduguda Ore % Ore % Ore % Quartz 52.5 65.0 65.0 Chlorite 36.0 27.0 22.0 Magnetite 3.8 2.6 3.0 Other Opaques 1.3 1.6 - Apatite 5.3 3.3 3.0 Tourmaline 0.8 - 3.5 Other transp. minerals 0.3 - 3.5 Sulphides - - 2.0 Others 0.5 1.5 Ilmanite 1.5 TABLE-III » Samoie Work" Index Kwh/T(ann e .Narwa Main Band I 8.79 Narwa Main Band II 9.71 Narwa Main Band III 13.08 Narwa Khundungri Block 12.35 Turamdih Sample I 12.3 Turamdih Sample II 12.7 Jaduguda ore 10.8 - 517 - TABLE - II Chemical Analysis Items Narwapahar Tvramdih Jaduguda Ore % Ota % Ore ft Cu 0.019 - 0.062 0.08 Mo 0.006 0.002 0.02 Ni 0.023 Trace 0.15 Fe 12.94 9.410 - S 0.05 0.100 0.-79 P2°5 1.87 0.80 1.04 Si02 51.24 57.41 67.2 Ti02 0.71 - 0.66 CaO 1.88 1.75 5.4 MgO 1.07 •» 2.2 A12°3 13.82 - 5.5 FeO a# - 6.37 Fe203 ^ *• 7.87 Undetermined ^ - 2.77 R «# 2°3 19.30 - 51ft LOCATION MAP SHOWING URANIUM PROJECTS NPEX 1. URANIUM PROJECT 1 -.*. ' 1 2 SUBARNAREKHA RIVER 3 GARA NALA 4 OUDH NAOI I IM| 5 NATIONAL HIGH-WAY BSBS9 6 P.W.D. ROAD 7 DISTRICT BOARD ROAC 8 RAILWAY LINE L»€g| { - 519 - ENTRY SYSTEM (NARWAPAHAR MINE) LONGITUDINAL SECTION AIN VERTICAL VENTILATION SHAFT SHAFT VENTILATION NW SHAFT \ SE 103M - l«0M MJM • amP* 5S0M ORE-BODY LIMIT 0RE-BO0Y. LIMIT UNDERGROUND UNDERGROUND - INCLINED SHAFT INCLINEO SHAFT FIG.-2 520 - ENTRY SYSTEM (TURAMDIH MINE) LONGITUDINAL SECTION NW SW j/C VENTILATION SERVICE At/C VENTILATION "' SHAFT SHAFT-i P SHAFT - S2I - CROSS SECTION OF DECLINE Ft 6-4 •}22 VARIOUS MODES OF MINE ENTRIES FIS.-5 e.?^ VENT AIR W , HAULAGE BELLOW STOPjE j VENT AR OUT THROUGH STEP MINING PILLAR VENT RAISES WHEN , HOLED (BOTH ENDS AT STOPE) LAYOUT ** NOTE:- 1. THE PLAN SHOWS THE COMPLETED RAMP SYSTEM WITH STOPING BLOCK IN ACTUAL ILLAR UNES OPERATION, RAMPING HEADING DRIVING AND SLIPING TO FULL WIDTH OF "ROOM" WOULD PROCEED SEOUENTLY. 2. RAMPS HAVE A NOMINAL U'/<(8') GRADE 3. FINAL PILLAR DIMENSIONS NOMINALLY 5M X 5M PLAN • ENTRY FROM ACCESS *|C KCESS-xlC PILLARS, ! ORE PA.SS 1 ' ' ^"==^JAGLAGE j^-OBE PASS EXTRACTION- X)C ENTRCWTDYV FROFonMu ACCESsTAi-rcecl I T *-EXTRACTION-xl^EXTRACTIONx|Cc ' HAULAGE c ydte PASS J- ^ X/ l^OROREE PPASA S TO EXTRACTION TO EXTRACTIONS " BELOW x/c BELLOW ELEVAffoN SECTION F16,-6 ELEVATION - 525 ROOM ANDP1LLAR. STOP1NG METHOD (TURAMD1H MINE) VENTILATION DRIFT P • • • • TEMPORARY PILLAR FROM STOM 3 OUT SECTION VENTILATION CROSS OVER TAKE AIR + ++++*+ + +++-H-VH-++H. + 4. + -H-*** 4+•»+*•»-• • ^^-^ I INTAKE AIR • ••••I50MM ORE TO BE STOPED IOOMM • MAIN VALVES FIG.-8 - - •??$ - DIAGRAM SHOWING PART OF MAIN VENTILATION SYSTEM VIEW AT INTERSECTION OF VENT RAISE AND CROSSCUT VENTILATION DOOR BETWEEN SECTIONS OF VENT RAISE X/CUT TO STOPE VENT RAISE AIR PASSING UP STOPE VENT RAISE MAY BE DIVERTED IN-TO EITHER VENTILATION CROSS CUT OR CONTJNI TION OF THE VENT RAISE F/w HAULAGE RETURN AIRWAY VENT SHAFT (STOPING HAS CEASED IN THE BLOCK IMMEDIAT- BELOW THIS LEVEL) ...V HAULAGE CMR INTAKE To STOPING BLOCK BELOW) RETURN AIR PASSING FROM STOPE HEADING TO STOPE VENT RAISE jffiTj '7 -CV— FM HAULAGEfAIR INTAKE TO STOPING BLOCK BELOW) F I 6.- 9 ( 527 SHOWING MINE DEVELOPMENT AND STOPING BLOCKS 2 4 t I-ACCESS X/C FROM F/W HAULAGE TO ST0PIN6 BLOCKS C-POST PILLAR MINING I-Xyfc TO STOPE VENTILATION RAISES ^-INTERSECTION WITH INDICATED RESERVES 3- X/C TO STOPE ORE PASSES A-WCLINEO ROOM AND PILAR Fl 6-10 B- STEP MINING . - 528 - - IMPOUNDING OF TAILINGS AT JADUGUDA PLANNING, DESIGN AND MANAGEMENT S.N.PRASAD* & K.K.BERI** The tailings are the waste product of the ore processed by the mineral industry which consists of ground up rock that remains after the mineral values have been recovered from the ore* Tailings dams are constructed to impound this tailings for pollution hazards, the degree of which depends upon the waste, being most severe for the radio active waste associated with the uranium ore processing* Tailings dam including disposal is one of She most important step due to following three points :- i. Cost - The cost of tailings disposal as a whole is substantial compared to .total processing .cost and on top. this expenditure does .not add to profitability in any-way* '.Hence the cost in this head should be as low as possible* 2. .Stability - The tailings dam should be stable. This is an important structure in view of the * Asstt.Supdt.(Civil) ** Chief Mill Superintendent Uranium Corporation of India Ltd. P.O. Jaduguda Mines Singhbhum, Bihar, PIN J 832 102. - 529 - structural stability, as its failure will release a very large amount of semi fluid tailings and pollutants which can not only cause the severe threat to life and property of the nearby area but also can cause extensive down stream pollution* The stability is important not only for the life of the plant, but also for the time to come* 3« Environment - The area should be least inhabited requiring minimum uprooting of population and effluent discharge should have least effect on surrounding as well as down environment* The planning, design and management of the dam is done to achieve above objectives. PLANNING AND SITE SECTION For planning and site selection following points need consideration s- 1, Capacity Requirement »- The capacity of the dam " is tobe decided based on ore reserves. If the ore reserves axe expected to increase in future, provision of the same has tobe kept either in the initial design or in way of future expansion. 2, The type and quantity of effluent discharge J- This information is required to assess the.effect of the discharge on down stream life and environ ment. Thin information is also necessary to take clearance from pollution board for permission to discharge the effluent. If need be effluent has tobe retreated to comply with pollution board requirement. 3, Proximity to plant »- It is desirable that pond should be nearest to the plant to reduce pumping cost. - 530 - 4. Minimum Embankment requirement :- The site should be preferably surrounded by hills so that the requirement of dam tobe constructed becomes minimus The bigger the length of dam will not only increase the cost of dam, but will also make the system less stable, as natural hills are definitely more stable, 5. The land and inhabitation :- The selected site should have minimum inhabitation and agriculture needing minimum uprooting of inhabitation* The cost of land should also be taken into considera tion. 6. The earthquake :- The area should not be severaly prone to earthquake. The higher the magnitude of earthquake more is the danger of liquefication and dam failure* 7. The type of earth j- The rate of percolation of earth should be minimum. The effect of seepage on surrounding environment should also be consi dered. In case of seepage above desirable limit the base has tobe lined or land treated to take care of the same. DESIGN In site selection, cost and environment has played major role. In design, the stability of dam is the main consi deration followed by economical consideration. For the stability the care should be taken for major causes of dam failure which are as follows. - 5 51 - Slope failure :- Slope instability can result from earthquakes, loss of shear strength due to an increase of pore water pressure in the soil or excess internal stresses produced by increasing the height of a dam without flatt ering its slope. Liquefaction :- Liquefaction of a.soil is a temporarV' fi'W.te in which the structure of the soil is disturbed, causing the particles to lose contact, and to transfer the weight of the soil and any super-imposed loads on to the water in the voids. This causes the pore pressure to rise and the soil behaves like a dense fluid. It is unable to support loads or to resist significant shear forces, result' ing in collapse of the dam. Liquefaction can be caused by changes in static stress - conditions, particularly in very loose soils. However, the more general cause of structural disturbance leading to liquefaction is dynamic loading such as can occur during an earthquake, or .vibration due to any reason even by a moving train. Overtopping failures :- One of the most common causes of failure is overtopping by flood waters. Where adequate provision is not made to tackle major floods over-topping, breaching and total loss of semiliquid slimes can be expected. Collaose of dewaterlng conduit or decant conduit failure :.- Many tailings deposits have experienced problems with the dewatering conduits, ranging from minor leaks to total - 532 - evacuation of the slimes from a deposit. Proper conduit specification and support is essential to take care of above problem. &• Erosion failure :- In areas of heavy rainfall some form of protection against erosion is usually required. As a single storm by itself* rarely causes major damage, the problem is generally considered to be one of maintenance, when erosion is severe, repairs are tobe made immediately to the slope with some preventive maintenance at the areas of worst damage. TYPE & CONSTRUCTION OF DAM For construction of tailings dam, three methods are in common use :- (Drg.No.1) 1• Upstream method 2. Downstream method 3. Centreline method i. Upstream method :-. With upstream method, an ini tial starter dam is constructed at down stream toe. The centre line of the top of the embank- raent is shifted towards the pond area as the height of dam increases. - The downstream toe of each of the subsequent dyke rests over the pervious dyke while the upstream portion is supported over tailings. 2. Downstream method s- In down stream method of construction, the centre line of the top of the - 533 - dam shifts towards downstream side as the height of the dam increases* In this case also a starter dam is constructed. In this case, each subsequent stage of dam construction is supported on the top of the downstream slope of the previous section* Prior to each downstream construction a pervious under drain layer, at least 3* thick or alternative drainage system must be placed to minimise the chance for build up of pore water reduces the'shear strength* To-minimise seepage through the dam, it is always recomm ended that the upstream face be sealed with some silt or clay* 3* Centreline method t- The centreline method of tailings dam construction is actually a varia tion of downstream method* The only difference is that instead of the crest of the dam moving downstream as the dam is built, the crist is raised vertically* Off all. the system, upstream method is most eco nomical but least stable* It has disadvantages of being built on top of previously deposited unconsolidated tailings* Under static loading conditions, there is a limiting height to which such a dam can be built without danger of a sheer failure occuring in downstream direction* The height will depend on (i) the downstream slope of tailings dam (ii) the location of the phreatic line within the dam and the strength of the tailings within the zone of shearing. A rise in phreatic line due to heavy rainfall can cause failure* - 534 - In contrast down stream method has following adva ntages - 1. None of the embankment is built on previously deposited, loose, fine tailings. 2. Under drainage system can be provided as per requirement as the dam is built* The under drainage permits control of the line of satura tion through the dam and hence, increases stabi lity. 3. The dam can be designed and subsequently const ructed to whatever degree of competency that may be required, including resistance to earth quake forces. 4. The dam can be raised above its original ultimate design height with a minimum of problems and design modification. BUILDING MATERIAL The building material should be such which is impervious and should result in most dense deposition. The choice available with the designer is tailings, local soil or borrowed soil in order of economy. A suitable optimum mix has tobe decided on lao icale test which has maximum relative density and gives most compact layer. This gives the fundamental basis and considerations about dam and its design. Now, we will discuss Jaduguda tailings impounding. - 535 - TAILINGS DAM AT JADUGUDA The first stage tailings dam is situated at about 1 km. from the sand-slime seperation unit. The tailings dam has been constructed across two hills to convert the valley into a reservoir of waste tailings. The drg.no. 2 gives the contour of the lands The initial toe dam has been made at axis A-A, the lowest level of the valley. Upstream method of construction of dam has been followed. Present final level of dam is 129.4 metre. The slime deposition area is at present around 82 acres. To augment capacity of tailings dam Ilnd stage dam was planned and constructed at axis B-B in the upstream of 1st stage dam. The present height of dam is 144 MRL with submergence area of 35 acres. The final planned height of this dam is 150 meter level and final submergence area will go upto 36 acres. The decantation wells of Ilnd stage dam discharges effluent in 1st stage dam to be subsequently discharged by the first stage decantation wells. The care has been taken that level of bottom of hume pipe is higher than final slime level of 1st stage. The Ilnd stage dam present capacity is for next two and half years only. For future requirement Illrd stage dam has been planned south of 1st stage dam, at axis C-C. The dam's initial level will be 125 MRL and final height 145 MRL and total catchment area will be 73 acres. In future a bund has tobe made at axis D-D when slimes comes to that level. The dam is planned to follow downstream construction method for better stability and reliability. The effluent water from this dam will meet the present drainage system and will follow same route. - 536 - 1ST STAGE DAM The drawing no*. 3. gives the stage construction of 1st stage dam* The length of this dam at present is 590 meter. The toe or starter dam was made of locally avai lable earth, and subsequent increase was done with sand, locally available clayey earth and slime in the proportion of 1:3x6 respectively* The dam has been raised gradually as per requirement of storing tailings in the pond* The dam was initially designed for 15 meter height* But. decision was taken to raise the dam to final height of around 20 metres as this was found international practice to go to that height ffor such dams and raising of daa would have increased the capacity equivalent to 7 - 8 years for running of the plant* The problem was of decantation wells* (Drawing No*4 ) . The position of these wells were fixed as per 15m* height design* It was seen that.one of the wells was coming within the embankment* So it was decided to close this well and construct another well at firm ground to compensate for the well tobe closed* for further safety, the diameter of pipes of new well was kept 900mm inplace of 750mm dia pipes in the old well* CONSTRUCTION PROCEDURE For raising of each subsequent dyke, sand : earth and slimes have been used as building materials in 1:3s6 proportion* Along upstream and downstream faces, adopted 8lopes are 1s3 and 1*2.as shown in the flg.no..3' earth, slimes and sand are mixed together in the above, propor tion and spread in layers not exceeding 300mm in " - 537 - thickness* Mixing is done manually. While mixing, care is taken that no lumps should left unbroken otherwise the mixture may not be homogeneous. Then it is rolled with 10 - 12 MT roller to make the layer well compacted. The thickness of each layer is limited to 300mm to ensure the proper mixing and compaction.. During mixing, water is sprinkled before spreading and compacting. The quan tity of water for a given valume of the material is predetermined by laboratory test to get maximum density. Finally, after raising the dyke to desired height, the following treatment is done along the slopes t- 1. Upstream face - Earth is spreaded along the entire length in the slope of the dam is well compacted by tapping with the rammers. This is also a manual method hence, care is taken for smooth spreading and compaction. This makes an impervious layer and minimises the seepage through the body of the dam. 2. Downstream face - Waste rock boulder soling is done along this face. If this treatment is not done, during rainfall it will lead to raincut and thus resulting in guilley formation and finally erosion of the face. Thus, this treat ment is necessary to take care of this problem. FUNCTION AND DETAIL OF DECANTATION WELL - (Drawing no.5) The slurry is being discharged into the pond by pumping it from sand-slime soperation unit to the dam through pipe line. The slurry gets settled into the pond behind the embankment and the clear water is decanted through hume pipe situated at the bottom of the wells. The walls of the decantation wells are constructed with the rubble masonry. The forth side has 1200mm long opening. As the slime level of dam rises the 1200X1SOX100mm sleepers - -,yj - are provided in the opening so that only clear effluent gets discharged from wells* The height of well is incre ased alongwith raising of dam as per requirement. 750mm dia RCC hume pipe of specification NP2 has been provided which leads decanted water out of dam. Collars have been used at joints, which are properly supported. This is essential to avoid collapse of pipe under slime loado THE MANAGEMENT OF TAILING DAM Only a stable design of dam can not ensure the stability of dam unless regular maintenance is under taken and precautions are taken in depositing the slime. As explained earlier, the tailings are pumped in the pond and sleeper height in the well is maintained such that only clear water is decanted off. Such slime deposition contains around 20% moisture and are virtually semi-fluid. Such condition can endanger the dam. In actual practice, the deposition was done area wise so that periodically every area was allowed tobe dried to make the same more compact. This has given us nice dividends as will by be explained later on. Though, this could not be done for areas very near to dam as a pool of water had to be maintained for allowing the fine particles to settle. The dam is periodically inspected. The sleeper height are inspected and extra sleepers are provided whereever required. The slope of dam is inspected for any erosion due to rain water and such channels are repaired. Boul der soling provided at downstream slope for arresting erosion is inspected and repaired as per requirement. The dam is inspected for any seepage etc. These pre cautions and inspections are virtually done on every day basis in rainy season. - 539 - The raising of dara also has tobe planned properly. Raising in rainy season is not practicable and hence should be avoided. Minimum free board of 1 meter should be available at the time of next raising. This gives stability for dyke to be raised by acting for that part of the raising as key. The higher the free board the more will be the raising cost. Hence, regular monito ring of level of slime in dam is a must for proper plan ning. As such minimum one metre free board is essential to take care of water accumulation and to safeguard against overtopping in case of unusual heavy rains. Erosion of the tailings dust from the pond is much during summer by the wind and creating pollution problem. vThe effective check is possible by free plantation in the pond. But, the problem is that all the plants can not grow in the vicinity of the slimes. We have tried with the variety of trees. One kind of tree named, 'Akasia* has grown up. This was done on trial basis. As the plant is coming satisfactorily, we have planned to go for plantation of this tree during the coming monsoon, Amari plant also grows satisfactorily in the tailings pondo As the 'Akasia' tree takes time to grow up the 'Amari plant* have been planted inside the pond for early effect. The pond has given greenery scene as this plant has almost covered it. PROBLEM FACED AND EXPERIENCED GAINED Our Jaduguda tailings dam was found quite stable as it had never given any trouble nor had suffered any damage, except incident on morning of 15/9/85 of collapsing of decantation well. T 540 - It had been raining heavily previous night and was still raining* At about 8.0 AM the operator at pump house noticed slurry coming out through one of the decantation well hume pipe* On inspection it was found that a cavity had formed in pond near, particular well, indi cating that some escape route had developed in the well below the normal level of slime. Heavy flow of rain water was flowing through that route and in process. eroding and carrying along already deposited slime. It was decided to remove the sleepers, of other adjoi ning wells, so that.water can pass through them and effected well can be cordoned off to stop further erosion and escape of slimes in the water flow. It took ah hour time to arrange for work force. In the mean time flow from hume pipe reduced. It was seen that the well has collapsed presumably from near the escape route and its debries partially blocked the passage of slurry. With less outlet flow, which ultimately stopped completely* The cavity got filled up with water. But side walls which were virtually vertical at that time were collap sing due to gradual decrease in angle of internal fric tion* . As the well was very near to the embankment, the upstream of embankment also started collapsing throating to make way in dam, which would have resulted in overtopping of dam and could have released the slime deposition completely. Our first task was now to save the embankment by stopping further collapse* It was decided, to put sand . filled gunny bags. By the time bags filled with slime and sand were arranged, in emergency 200 bags of cement were dumped, followed by bags filled with sand and slime. After some time collapsing stopped, and immediate danger was averted. - 541 - In next step the embankment was strengthened by dumping waste rock in upstream and downstream of effected embank ment. It took virtually three days of contineous work to come to this stage* In next step water from the cavity was pumped out by sub- mercible pump with ah idea to revive the well. But all collapsed' side walls and waste rock dumped filled up the cavity to*such an extent, that to dig out of same to level upto which well got damaged was found not advisable as the same was risky. Secondly, we had no idea about the depth upto...,which the well is still intact. Hence decision was • taken to close this well completely. To compensate the draining capacity it was decided to plan alternative arran gement for drainage system. The closing,of well was done from hume pipe. Bags filled with sand were pushed in the hume pipe as far as possible then boulders were filled for 3 meter and then the opening was blocked with concrete. In other words from outer point of hume pipe 2.5 meter of concrete, 3 meter of boulders and 4 meter of slimes were put. A 50mm 0 G.I.Pipe was provided in between for release of seepage water in the humei pipe.* The cavity-was slowly filled with slime upto the original slime level. The cause of the abovar problem, could not be assessed as. the collapse portion of well could not be inspected. It was feared that the same thing can happen to other wells also and created doubt about safety of comp lete dam. It was decided to approach M/3.Central Water and Power Research Station, to conduct the studies about safety of dam and suggest remedial action to be taken if any. - 542 - But another similar incident gave us the reason of above problem. The adjoining area of above well also remained dry because of diversion of water to other wells. In process nearest decantation well also remained dry for this period. Because of this drying deep cleavages were formed in slimes as deep as may be 1 - \Y2 meter. After filling up of above cavity, on 30th November water was released to the adjoining well also. On 1st December night slurry was found coming out from this adjoining well and a cavity of around 1 meter depth was found near the well. As this was dry season water flow was very less. The well was immediately cordoned off. On inspection, well was found absolutely intact and slurry was found having escape route around six meter below the slime level due to Duckling of one of the slee per. At that place in place of 4" thick usual sleeper two 1" planks were found which had buckled under pressure and gave way for slime. On enquiry it was gathered that around seven years before these planks were used in place of usual sleepers due to shortage of same. The expected depth of damage of first well also was at the same depth. On inspection all wells at that particular depth were found having same type of planks. As these sleeper could not be replaced the same were straigthened for every well so as to avoid such problems in future. The bulged sleeper was first straigthened by applying props from the side wall and stiffening was done. Vertical stiffening was also done, to join these planks with other sleepers* Other wells were also strengthened accordingly. - 543 - DRAINAGE SYSTEM All the water of catchment area is supposed to be drained by these decantatlon wells. The load on them becomes very heavy at the time of heavy rains. The water level in dam at that time can go as high as 600mm which can endanger the dam. Due to particular contour it was found possible to drain out all the water from alternative route, so that even in worst circumstances, these wells have not to take any extra load. In case of any eventuality or possible threat, the well area can be completely isolated and made free of water even in heaviest rains. So a drainage system was constructed thro J the hills. The slime level near the embankment was raised slightly, so that the drainage system has the lowest level. The system has been found very satisfactory and had taken complete water flow even in heavy rains. STUDIES BY M/S CENTRAL WATER AND POWER RESEARCH STATION. PUNE, M/s Central Water and Power Research Station, was appointed to study the dam in general and to assess its stability in particular including if the dam height can be raised further. They were asked to suggest remedial action if any tobe taken for the safety of the dam. They were also asked to study the suitability of proportion of mix of building material being used now and suggest any change in same if required. - 544 - They took field sample in first week of May*87, The samples were collected by drilling bore holes from dam and from pond. The report was submitted to us in last week of February'89• THE MAIN POINTS OF CONCLUSION ARE s- 1 • The existing dam is marginally safe with safety factor around 1• Liquefaction analysis has shown dam tobe safe for earthquake magnitude upto 6 and ground acceleration of- 0.06g, 2. The stability is low because of excessive moisture content of slime dam soil ( maximum of the oxder of 4556 ) • The stability of the dam can be increased if this moisture content can be reduced to induce better compaction* 3. The dam height can be raised provided moisture * content can be reduced and safety factor can. be increased. For reducing the moisture content they suggested pumping of extra water from dam by making bore holes. 4. The proportion of mix of building material as 1:3:6 of sand, earth and slime being currently used by us was found optimum and suitable. ACTION TAKEN/BEING TAKEN FOR STABILITY OF DAM - The extensive tests have been done by M/s Central Water - 545 - and Power Research Station and report has been presented. As water pool has always been maintained near the wells for proper settling of slimes, there are chances that moisture content is high in dam soilo But moisture con tent of the order of 45$ seems too high. Secondly we have our own reservation if water can be pumped, by drilling bore hole. Instead we formulated different strategy. As our drai nage system became operative, all the water was diverted and all dam area was allowed to dry. This was wAth the idea that by this moisture content of dam soil will come down due to self percolation and compaction may take place. The slime level of dam area in general got depressed by around 15 - 20mm. A half meter dia by /2 meter depth cavity was observed after some days at the place where well had collapse initially., This area cavity was filled without proper drainage. Hence this may. be having poor compaction. With drying of area mois ture content went down resulting compaction. Further observation could not be done as monsoon,'set in. FUTURE ACTION * . i it''' - . - - - e-j We have again approached to Central Water and Power Research Station with the proposal to check the data once again. We propose to take sample as follows :- 1o Allowing the dam area to dry for one month in dry season and take sample from same places as was taken in first study for moisture content. - 546 - 2. After that, allow the area to get submerged for next one month and then take sample to check variation in moisture content. This will reveal. 1. If moisture content can go down by natural percolation alone* 2. If moisture content of dam soil is a dynamic state and will change as water level in the area changese M/s Central Water and Power Research Station were appro ached for this proposal and they have agreed with the suggestion* The study is scheduled tobe done in first week of April'90. 11 can be said that slime dam at Jaduguda by and large has been well planned, designed and managed. It should be appreciated that in this field there is not much of experience and expertise is available and actions are taken more on the basis of experiences. The importance of many factors which effects a lot on stability of dam were not felt earlier, but has been realised based on better understanding about the dam. UCIL has gained sufficient practical expertise in the field of slime dam. But many factors are still tobe resolved and we hope to have better knowledge of this area after - 547 - studies, of M/s Central Water and Power Research Station concludes. It is worth mentioning that the steps to be taken for closing down the operation of slime dam after the same has been filled to capacity is seperate expertise. The procedure of same has to be formulated. The exercise will be done and if required, outside help will be taken in laying down, the procedure which will be followed for closing down of dam. .. 00O00 .. 548 - rypEsor DAM UP STREAM METHOD DOWN 5TREAMM6JH2S • : J i I . «1« t»l«AIM CENTRE. LINE METHOD SCALt- t?.4QDB PPG NO-- i 550 - l»3g- gu»»l.t ~»T-.i-nr l.{» «!•• IW» TAILINGS - '. - ///I. CdtHII »--'» »tn» .'.^t( DETAILS OF STAGE CONSTRUCION OF FIRST STAGE DAM •5> CMj. - 1 • gOO i I . S -552' »&»» >*V* • .!»« ~»*»=*i.*L_5£2,.V_»_£-: "R7 »»• «IW/M *** 1 SECTION ce *» u * rt or DE«NTMIOW WCLL ft.AUK Mlf r , '••-j •-- srt • -Z2S, x-^n 3—' {• •""••! SECTION flfl I tm«mimitT. PLAN. DpG,*#>.•€ ,-*r.»n;-M - 554 - Session IV DISCUSSIONS Paper No. 4 B.L. JANGIDA : What are the crosslinkage of the ion exchange resins used? D.P. SAIIA : -This information may be available from the technical brochures supplied by the resin manufacturers. T.G^- SRINIVASAN : I would like to make a comment. The manufacturers of macroporous resins in India do not give the cross-linkage value. It is enough if performance of different resins with specific trade names are studied, since it can be safely assumed that different batches of the same resins are unlikely to differ in their chemical properties with respect to capacity, cross linkage etc. SESSION URANIUM ORE PROCESS TECHNOLOGY AND BYPRODUCT URANIUM Chairman : Dr. O.K. GUPTA B A R 0 Reporteur: Dr. T.K. MUJCT3RJEE B A R C \ - 555 - KUCLEAH PURE UKATJIUK PROI.T ORES USING WEAK BA3E ION EXCHANGE RESINS S.V. Parab, S.'J. ^harnt, G. Cherian and K.S. Koppiker Uranium Extraction Division, BARC, Bombay - 400 085. Presently the uranium requirements of our nuclear energy programme are met from uranium nine3 of Uranium Corporation of India Ltd. located in Singbbhum Dt., Bihar. The process followed for treating the ores involves leaching of uranium with dilute sulphuric, acid, followed by separation of uranium by the conventional strong base anion exchange "technique. The final product is yellow cake (magnesium diuranate) which is sent to TTuclear Fuel Complex, Hyderabad for refining to nuclear purity &nd conversion to reactor fuel. There, uranium is refined to nuclear purity by solvent extraction process carried out in a nitric acid medium. Laboratory studies have been carried out on the feasibility of obtaining the nitrate solution directly from the ion exchange resin 30 that the final purification of uranium to nuclear grade can be carried out at the mine-head itself, thus eliminating several process steps. In this paper, the data collected on two resin samples - One Indian and one imported are presented. It has been 3hown that it is possible to obtain nuclear grade uranium product directly from the ore under certain conditions using weak-ba3e union exchangers. A tentative process flow sleet is also presented. - 556 - Introduction Currently, the uranium requirements of our nuclear energy programme are met from the uranium niine3 operated by UCIL, Jaduguda. It is primarily supplied as magnesium diuranate which undergoes chemical refining at Nuclear Fuel Complex (NPC), Hyderabad, before being converted to ceramic grade UOo* During the refining step, appreciable quantity of nitric acid insoluble matter - chiefly silica is generated as waste. Since this waste contains good amount of uranium, it is returned to UCIL for treatment. To meet the increasing needs of nuclear reactors, NPC will be expanding its capacity several folds and with the current practice it would entail disposal of large quantities of active silica waste. One of the options, i3 to locate a nuclear refinery close to the uranium mill. In case, the refining/U02 production facilities are to be co-located with the mining/milling facilities, it would present an opportunity to integrate the process operations of the mill and the refinery. This process integration would help to improve the process economics also. At the very outset it can be deduced that any process integration should essentially eliminate the MDU precipitation step by UOIL and in its place, only uranium solution should be made available to the Fuel Plant, preferably as uranylnitrate /which can then be ^ , „„_ .. . ^processed by solvent extraction by NPC. ouch a scheme of integration would have the following advantages to the Hill and the Fuel Plant. - 557 - A. For the Hill 1. Elimination of the following process steps a) Iron precipitation with lime b) KDU precipitation with magnesia c) Filtration of gypsum and yellow cakei d) Drying and packing of yellow cake in steel drums 2. Energy saving due to elimination of above steps 3. Saving in chemical consumption 4. Elimination of dust problems due to handling of MgO and yellow cake B. Por Fuel Plant 1. Elimination MDU dissolution in nitric acid and associated scrubbing/pollution problems 2. Elimination of filtration step and waste raffinate cake disposal 3. Some saving in nitric acid consumption and reduction of nitric effluent. Such an integration would also result in marginal increase in process recovery both at the Mill and the refinery and reduce investment outlay and operating costs. DAS/UCIL is now in the process of setting up a new uranium mill at Turamdih near Jamshedpur and similar plants are likely to be set up at other locations in the near future. The Turaradih mill is based on the conventional sulphuric acid leaching followed by ion exchange purification/concentration step using strong base anion exchange resin in Himsley continuous contactors. The uranium would be eluted from the - 558 - resin using dilute sulphuric acid to yield a strong eluate containing upto 25g U,0„/l which would be neutralised by an alkali/MgO to yield yellow cake. Any modification of the process in Mill would essentially involve final conversion of uranium from sulphate to nitrate medium. This can be achieved in two ways. 1. The strong sulphuric acid eluate is further processed by solvent extraction using amine (ELUEX Process) and the amine is stripped with a nitrate salt to yield uranyl nitrate solution. 2. The ion exchange process itself is marginally modified by xi.sing weak base anion exchangers to enable nitrate elution of ur?-nium to obtain uranyl nitrate directly from the ion exchange resin. The second alternative has been studied in the laboratory and the data collected is presented in this paper. The first alternative will be described separately. In both the alternatives, uranium is converted from sulphate to nitrate medium and hence uranyl nitrate solution would.contain 3ub3tantial amount of sulphate since uranium is held both by the resin and the amine as a sulphate complex. Presence of this sulphate is a major drawback in the further purification step using Tributyl phosphate. 2.0 V/eak-ba!ie anion exchangers in uranium ore processing Kunin suggested the use of Y/B anion exchan/rer Amberlite X?:=270 in place of strong ba.Je 1RA-4 00, since the fori: or had lower affinity for iron. Also the V.'B renin has 2 fa.Jtrr olution rutp.t with nitrate. This re3in bus bei»n tested earlier tit TIED, for uranium adsorption from sulphate leach lKjuoru. Later, Kohm and !!aaa have introduced Amberlite X5-299 - 559 - as a resin 3pecific for uranium industry . This improved version had higher uranium capacity with low iron retention and had excellent elution properties with neutral salt3. James while working on South African low grade ores proposed alternate process routes based on treatment of unclarified solution in CIX contactors'using WB and SB ion exchangers. CIX Nitrate ADU (V/eak-Base) Elution Precipitation Resin leach Liquor CIX ^ Sulphuric Amine _^NH3(NH4)£04 Acid Extraction Stripping Strong Base Resin Elution ADU Precipitation He had suggested the possibility of obtaining nuclear grade uranium from nitrate eluate after refining with TBP. However this possibility wa3 not followed in practice, and no reliable data wa3 available. Resin3 tested In our laboratory testa two resir. samples have been tested 1. Amberlite XE 299 - supplied by Rohm & Haas, USA 2, Tulsion A-2X (MP) - Supplied by Chemical Division of Thermax Pvt. Ltd., Pune. - 560 - The former is designed for uranium industry, whereas the Indian resin has not been utilised for such'applications. It is quite likely, the manufacturers of this resin may- improvise this resin to meet specific requirements of uranium industry if desired. The description of Tulsion A-2X (IIP) is given in Table T. Ion Exchange Studies The IX studies have been carried out using a synthetic leach liquor having the following chemical composition (in g/l) T 2+ Lyig : 1.10, Fe 2.0, Pe 1*0, P2050.75, So|~ 30,. pH 1.70 Preliminary te3ts were carried out on 10ml war in glass columns. Loading was done at a retention time of 2 mi'n. aid 4 min. and after saturation the column was washed with acidified water (pH 1.5) and the uranium was eluted with ammonium nitrate at a retention time of 10 min. and the loading capacity was determined. Adsorption of uranium from sulphate leach liquors at pH 1.7-2.0 was not expected to pose any problem. But elution of uranium by neutral ammonium nitrate had to be studied in detail lince not much wao known about its effect on resin etc. T'ence elution parameters were studied in detail e.g. effect of HH.NO, concentration, and effect of retention time and pH. Effect of NH,NO, Concentration The Nil.NO, concentration was varied from 1 to 3 M with 4 3 3M solution, elution wa3 fast and the 3trong elunte was confined to 3 BV and the entire elufeion was completed in 9 BV. Since the long range effect of using 3K nitrate \ - 561 - solution was not known, most of the studies were carried out with 1H NH.NO,. Effect of retention time of elution The elution reteniion time was varied from 4 min. to 16 min. and it was found that 10min. would be adequate when using 1M IIH4NO,. Elution Problems During these elution studies, it v/as observed that a yellow precipitate appeared in eluate fractions after 4-6 hours. On testing, it was found to be uranium phosphate. Kumin had earlier made similar observations during his studies.on V/B resins and had suggested the use nitrate salt concentration to above 3M to prevent this phosphate precipitation. Since we had some doubt regarding the stability of the resin at high NH.NO, concentration, this problem was studied in detail. The pH of leach liquor was varied from 1.5 to about 1.9 and no charge in phosphate absorption could be noticed. One of the. wayo available to prevent precipitation of phosphate with uranium is to provide another metal ion whioh can complex the phosphate and keep it in solution and the be3t metal for this purpose is ferric ion and this wa3 studied in depth. - 562 - Effect of ferric iron on phosphate precipitation The ferric iron was added as a nitrate salt and the concentration was varied from 0.025 to 0.1J.T in UI NH.UO,. Tests were carried out on both resin samples. The result3 show that with a split elution pattern (1-3-6), a 3 BV strong eluate with the following composition would be obtained in (g/l) U308 15, Fe 2.0, P205 0.15, S04 15, And it was found that.at 1M NF.NO, and 0.025M Fe, the precipi tation would occur only after 100 hours of storage. Kenoe, for safe operation, the ferric ion concentration was optimised at 0.05M. Increase of MI. NO, concentration from 1M to 3M did not show any improvement 3ince precipitation occurred.in 2-4 hours and hence this approach was not pursued further. Similarly, the contact time during elution was found to have practically little effect on phosphate distribution or precipitation. Retention of nitrate ion on the resin In presence of nitrate ions, adsorption of uranium from sulphate solution is known to be seriously affected. Hence tests were carried out on 10ml resin columns containing Tulsion. After elution, the re3in wa3 thoroughly flushed out and during adsorption cycle, effluent samples of raffinate were collected for every 2 BV upto 20 BV and analysed for both uranium and nitrate contents. It wa3 found that uranium adsorption took - 563 place in the presence of small quantity of nitrate, ^he nitrate content of first 2 BV fraction was around 7-8 g/l and was reduced to 0.5 g/l in 20 BV. However, this aspect has to be studied on a larger column simulating plant practices. ( Stability of resin and Resin-Li fewest The two samples .of resin were subjected to repeated adsorption and elution cycles (35 cycles) and the loading capacities were determined after each cycle. Conditions of tests were - Resin 20ml wsr, Peed 1.0g U,0g/l, Bluant: 1.0M NH^NO, plu3 C.05M Pe(N05),; contact time 4 min. (Ads), 8 min(El). She elution pattern was studied after every 5 cycles, and it was found that loading capacities were steady and no significant change in elution pattern could be noticed„ 2he average loading capacity (g U,0g/l wsr) was - Tulaion, 58 and X5299, 48. Blution was faster with XS 299. Prolonged contact tests with nitrate eluant were also carried out on 20ml column (Tul3ion) by passing 3.0L of 1.0T.T 1IH.N0, at 8.min. retention time. Each such passage corresponding to 15 cycles and the resin was put on adsorption as usual. After 7 3uch repeated cycles, uranium loading capacity was not affected. !Jfce possible accumulation of -iron and phosphate on the resin during repeated loadings was also itudied. It wa3 observed that the residual value3 after 35 cycles were low. "jut Uhii5 i3 another observation to be confirmed on prolonged trials using 'live* leach liquors. - 564 - Di3cusaion of IX tests results The tests have indicated the feasibility of using Y/B anion exchanger e.g. Tul3ion A-2X (MP) with ?JH.NO,- Pe(NO,)i mixture as the eluant. The residual nitrate on resin does not seem to hamper the uranium loading. However, the initial portion of raffirate cannot be recycled for leaching due to its nitrate build up. It is possible to obtain a strong eluate fraction (3 BV) having the composition U508 15 Fe 2.0 S04 : 20 P205 0.15 This uranyi nitrate solution containing 0.2M sulphate would be the starting material for further refining by solvent extraction U3ing TBP. Since the current practice is based on a feed solution assaying 200g U30Q/l and 2M HNO,, laboratory studies on purifying the above solution were also carried out. Befining of uranium to nuclear purity by TBP-3olvent extraction technique In a conventional refining process the I'TDU is dissolved in nitric, acid, ard the insoluble3 are filtered off. The uranyi nitrate (200g U,08/l and 2K HNOj) is extracted with 30# TBP in kerosene in 5-8 stages of contact and the loaded solvent is scrubbed with 2M ENO, to remove all impurities from the extract, and finally uranium i3 stripped with dimineralised water to obtain nuclear pure uranyi .nitrate. Due to the low concentration of uranium in the feed solution obtained via V/B-ion exchange route, the process parameters have to be redifined. - 565 - Effect of solvent concentration Prom:the experimental solution uranium wa3 extracted with solvent containing varying concentration of TBP and it- was found that 30# v/v solution gave the best extraction coefficient. Using 30$ TBP, extraction equilibrium and the McCabe- Thiele plot were drawn and it was found that minimum 7 stages of countercurrent extraction would be required to obtain a raffinate low'in uranium. Batch countercurrent te3t Using separating funnels, a 8-stage countercurrent extraction test was carried out at a phase ratio of o/A=1.1 and it was found that almost all uranium was extracted. f Removal of impurities The countercurrent tests were.repeated after addition of 100 ppm of gadolinium oxide as an added impurity to the synthetic solution. Tbe:loaded extract was scrubbed with 1M 1IH.N0, at a phase ratio 0/A=5«- The uranium after stripping was analysed.for Gd content and was found to be<0.i ppm on U ba3is. Normally Gd content in MDU is 5-10 ppm and is brought down to<0.04 ppm. in our .trial we had deliberately added a ten fold excess as an extreme case. Influence of sulphate ion Extraction tests were carried out with synthetic eluate solution containing varying sulphate concentration. It was found that increasing the NIL MO, content would help to tolerate higher sulphate content during TBP extraction e.g.. in 3M. nitrate - 566 - J with 25g 30./1 the uranium loading is 30g U,Og/l and at 1M it is around 15g U,0Q/l. Hence, an optimum concen- trat.ion of sulphate in eluate would be 20-25 g/1. This can be maintained by a regular sulphate bleed a3 gypsum by addition of calcium nitrate. Bemoval of sulphate from eluate solution Since most of the ammonium nitrate ha3 to be recycled to avoid disposal problem, sulphate build up in recycled solution has to be carefully controlled by gypsum precipita tion. Precipitation tests have shown that even when stoichiometric quantity of calcium nitrate is added sulphate precipitation is not quantitative. This could be due to solubility of calcium sulphate in ammonium nitrate solution. Hence, it is more economical to remove 75-80 per cent of the sulphate leaving about 2.0g Ca/l in solution. The original value of 30g SOVl can be brought to < 10g SO./l. This solution can be reused for next cycle of elution without causing any precipitation problems on the resin bed. However, this aspect will have to be confirmed on large 3cale trials. Based on these tests , a process flow sheet for adsorption of uranium on WB anion exchanger, followed by elution with neutral JIK.1I0, and TBP extraction has been 4 3 finalised and is shown in Pig. 1. This requires final confirmation by large scale trial3 using 'IIVP' leach liquor. The main problem to be faced will be the sulphate build up in the nitrate elution cycle. - 567 - Acknowledgements The authors are grateful, to Shri T.K.3. Murthy, former Head, Uranium Extraction Division under whose overall guidance this study was carried out. References 1. R. Kunin fetal Engineering & Mining Journal, July, 1969, 73. 2. K.D. Kamat et al Divisional *eport BARC-731 (1974). 3. S.V. Parab et al Divisional Report, OSS/1979. 4. H.E. James Proc. Adv. Gr. Meeting, Washington DC IAEA-AG/73-5, Vienna, 1976, page 35. - 568 - Table - I Description of Tulslon A-2X(MP) Type" t Weakly.basic Polymer structure s Macroporous polystyrene Functional Group : Tertiary amine Form supplied t Pree Base Screen size (BSS Mesh) :'-16 + 32 Particle size In mm t 0.3 - 1.2 Total capacity : 80 g/L as CaCO, Kax. Temp. (°C) » 60 pH range : 0-9 Bulk density ': 720-730 g/L - 569 LEACH UQUOR IOgU30e/l WEAK-BASE BARREN UOUOR K)M EXCHANGE- • 20% TO BLEED ADSORPTION REST FOR REUSE ACIOIFIEO WATER UDADED RESIN -> -TO LEACHING TANKS PH IS ~- WASHIN+ G t-OM NH4NO3 LOADED RESIN -^ELUTION A 003 M F«(N03)3 8-10f irin . R.T. I BV 4 BV 5BV STRONG EL o-a«^u^)8 WATErR IOflSQ,/T I9g U3O8/I 4-01504/1 "20gS0;/l _i SATURATED' - GYPPSUS M GYPSUM CAKE TO CofN03>2 SOLN. "PRECIPITATIOATION gon d REMOVAL- I- LEACHING TANS I BV NH*N03 BARREN RAFFttUffE (TO MAKE 3MN0J) _-'SX FOR REUSE •^•EXTRACTION- ^.FORN^NOa so*/, xapj 0 STAGES •*DVCRYSXALU2ATI0A IN KEROSENE 0/A»l»l«nfaw "—t SCRUB RAFFINATE SCRUBBING 3M W SOLft- 5 ->6 STAGES • 0/A*S,lmln. MTRATE RECOVERY EXTRACT I 14-19 g U^l BARREN SOU/ENT STRIPPING FOR RECYCLE 0. M. WATER - POSTAGES- O/x-5, 3mln V URANIUM PPTN. TO IX^PREPNA . FIO.l. SCHEMATIC FLOW SHEET FOR PRODUCTION OF NUCLEAR GRADE UOo FROM ORE VIA WEAK-BASE IX ROUTE. NSK>RN8>0'7-'0& - 570 - DEVELOPI.IINT OP AN INTEGRATED PROCESS POP RECOVERY 0? URANIUM- PROM ORE AND ITS RAINING AT THE LOCATION OP PEW URANIUM MILL AT TURAMDIH, SINC-BHUH DT. R.A. Nagle, S.V. Parab, S.S. Gharat, A.B. Giriyalkar and K.S. Koppiker Uranium Extraction Division BARC, Bombay - 400085. The second uranium ore processing plant of Uranium Corpn. cf India" Ltd. is being set up at Turamdih, near Jamshedpur. This plant would incorporate several new technologies. Instead of fixed-bed ion exchange, a continuous countercurrent ion exchange technique based on Himsley columns is being adopted. The uranium would be eluted from loaded resin with dilute a sulphuric acid, and this acidic eluate assaying around 25g • U,0g/1 would be neutralised with alkali to produce yellow cake. The next fuel fabrication plant is also expected to be located at Turamdih, close to the new uranium mill. To take advantage of this co-location of the two plants, the feasibility of integrating the two proceS3 flow sheets has been inveotigated. The data collected during laboratory trial3 oh development of an integrated flow nteet err presented. The process involves introduction of the conventional . VI.USX process usjrp the tertiary, amine Alamine ^"56 to extract uranium from the oulphurie acid eluate solution followed by nitrntp atrip»iri..- of uranium from the nmir.e phaue to yield a concentrated solution of urunyl nitrate. Jince thbj Jtcp - 571 - involves a change over from a sulphate to a nitrate medium, this solution v/ould contain appreciable amount of sulphate. Since the presence of sulphate has a deleterious effect on the next step of purification, it is eliminated by precipi tation as gypsum by addition of calcium nitrate. Finally, i.iraniuni is purified by solvent extraction with tributyl $ phosphate. The final product of uranyl nitrate solution obtained in laboratory trials has been fovmd to pass the specifications for nuclear purity. A process flow sheet has been finalised and is undergoing large bench-scale trials. INTRODUCTION The conventional process for recovery of uranium from ore3 is generally based on sulphuric acid leaching followed by separation/concentration of uranium by ion exchange/ solvent extraction technique. Finally uranium is converted to the yellow cake or oxide. This is ultimately dissolved in nitric acid at the ursnium refinery and purified to nuclear purity levels by solvent extraction process using tributyl phosphate. During past two decades efforts have been made at several places to obtain the nuclear grade uranium directly at the mine-head it3elf. The earliest efforts were made in 1 2 South Africa where amine extraction route BUFFIEX and FUHLJ3X processes were studied in detail. Thi3 approach was ultimately given up as the final product could not meet the stringent specifications for a few key elements. Jame35 etc pursued these objectives by incorporating continuous IX technique baaed on weak base/strong base resinr? in the process flow sheet. - 572 - The weak base resins had the advantage that neutr 1 nitrate^ solution could be utilised for eluting uranium and the resulting uranylnitrate could be purified further. Later, Garza and Jaoraek while studying El Nopal uranium" ore have proposed amine extraction from sulphate solution followed by nitrate stripping to yield a concentrated uranyl nitrate solution which can be further purified by TEP extraction. UCIIi is setting up a new uranium mill to $reat 3000MT ore/day at Turamdih near Jamshedpur and NFC is setting up its new uranium oxide plant at this location adjacent to the mill. To take advantage of this co-location, possibility of integrating the process flow sheets of these two DAE units has been investigated. The process finalised by UCI1 involves the use of inprovised IX technology based on CIX - Hlmsley contactors. The uranium would be eluted with 10$ w/v H2S0. and finally precipitated. If this eluate is processed by amine extraction (i.e. TTL1JEX process) and the uranium is stripped from the amine with ammonium:nitrate solution, one can obtain'uranyl nitrate which can be further purified by TBP extraction. This option, would eliminate several process steps in UCIL plant as well as in NPC unit and improve the process economics for U02» SOIVWP 73XTRACTTOU WITH- AMINE Use of long chain alkylamine for extraction of uranium frorc sulphate medium is by now a we.ll accepted practice. At Uranium Extraction Division, studies on use of Alamine-336 were made almost two decades back, and adoption of ?LUEX process in the mill at Jaduguda waa strongly recommended. • In this process, the loaded amine is stripped with NH,T(NH«)g30 solution at a pll of 4.0 aixJ the uranium is ultimately precipitated a3 ammonium diurnate. This process has the - 573 - advantage that the dilute sulphuric raffinate is recycled ' for leaching and the ADU obtained would be of high purity ana will not generate any silica-cake wa3te at the N?C plant. In the present proposal under study it i3 proposed to atrip the amine using NH-NO, solution, which can be straightway processed at the NFC plant by solvent extraction with THP. As per the project data, the Himsely - contactors are" supposed to yield a sulphuric acid strong eluate assaying about 25g U,08/l. It is likely that in routine operation, it may be difficult to maintain such high concentration and a value' may be somewhere between 10-25g U,08/l. PlffiLIITIEARY T53TS Saturation loading capacity for 15'^.v/v amine with 10^v/v isodecanol in kerosene was found to be 19g U5Og/l. 'Equilibrium diagram for urnnium extraction using 15$ amine solution was prepared and McCabe-Thiele plot was drawn. It was found that over 99 per cent, of uranium could be extracted from the synthetic 3trong eluate solution in 4 countercurrent stages at a phase ratio of 0/A=1.5. Any uranium left unextracted will not be lost-3ince the barren raffinate would be 'recycled to leaching. Thp. loaded extract assaying about I9g U,0g/l was utilised for carrying out the stripping 3tudie3. Extraction of uranium fron' strong eluate was not expected to p03e any serious problem, wheren? for stripping v.'itli nitrate, onough d*i*a v»a:? not ?v>'Tl-.»hT<>. Hrrea, thiu aspect "/.i-j Studied in rtetailt - 574 - Ammonium nitrate stripping of uramlua from amine In uranium industry, chloride or sulphate have been most commonly used for stripping uranium from loaded amine. Ammonium nitrate has found limited use, partly due to its cost. Though Brown et al in their early studies found this reagent to be effective, they reported unsatisfactory phase separation properties. It was used on large scale operation by Paure, in S. Africa. They used a mixture of ammonium nitrate and nitric acid. The stripping mechanism can be represented by the equation. (R3NH)4U02(SO+)3 + 4 N03 _ > AR^KVSQ^ + U02(S04)|" + So|" i In 3. African trials the authors experienced problems in re-use of the amine after repeated stripping cycles, leading to the loss of extraction capacity and emulsion problems, and ultimately the nitrate stripping was abandoned in favour of ammonia-ammonium sulphate route. However, Garza and Jamrack after extensive studies have advocated the use of ammonium nitrate as stripping agent. In our studies we have used 1.25M NH4N03 (100g NH^NOj/l) containing 0.25-0.5N HNO, as the stripping agent. Stripping Studies The loaded extract assaying 19g U,0g/l was stripped with 1) Neutral 1M NH^NO^ solution 2) 1.25M NH^NO-j + 0,5M HNO, and 3) Neutral 1.5M NH.C1 solution at a phase ratio o/A-5, in multiple contact with fresh agent. In all oases, stripping was complete in 3 contacts, acidified solution being faster than neutral nitrate. The results obtained with acidified - 575 - nitrate are shown in Table I. It was observed that the loaded extract contains about 21g SO."/lgi all of which would report to the atrip solution along with uranium. Thus at a phase ratio of O/A=*5I one can expect the strip product to contain ~~ 90g "UJOQ/l with over 100g S07~/l. Table 1 Distribution of uranium and sulphate during stripping with acidified NH.MO, solution Extract » 19g U,Og/l Strip Agent : 1.25M NH4H03 + 0.5M HN03 Phase Ratio O/A • 5 Contact time J 5 min. contact No. AnalyAnalysii s of strip solution (g/l) U3°8 sof 1 69.8 98.7 2 11.3 . 11.2 3 3.1 1.0 4 1.3 5 0.58 6 . 0.27 The sulphate content of the extract can be partially reduced by scrubbing with acidified (pH 1.0). This .would help to remove any sulphuric acid extracted by the free amine, leaving only the stoichiometric quantity of sulphate bound to uranyl ion. About 10^ of uranium goes in the scrub solution which can be recycled. - 576 - Removal of sulphate The composite solution of uranyl nitrate obtained from nitrate stripping would assay around 70g U50Q/l. Howeverf the high sulphate oontent of this solution is is major obstacle for successful extraction with TBP. Hence, reduction of sulphate value by precipitation as calcium sulphate is unavoidable. Precipitation with stoichiometric quantity of calcium nitrate, added as saturated solution removed about 80 per cent of sulphate, with 2 hour digestion of precipitate* With substoichiometric quantity (0.75) of calcium nitrate, precipitation was effective to 70$, but the residual calcium content was lower, and safer for further processing. In order 2+ to study the effect of Ca ions in recycle strip agent, tests were carried out with 3.0M HH.1TO. + 0.2N HN0- containing calcium ions 2-20g Ca /l. It was observed .that if the Ca content was 2.0 g/l, even upto 45 g/l of sulphate did not cause any precipitate. TJranyl nitrate treated in'this fashion, and after gypsum filtration, would assay (g/l) 2+ U308 . 75 , SoJ" t 16 Ca . 6 The final, cake washings would contain some uranium (»~1.0g U»0Q/1) which haa to be recycled or used for meke up* Ultimately in order to avoid any uranium loss, this gypsum oake will have to be recycled to the leaching stage of the mill* This particular aspect has to be studied in depth on a pilot scale, to define the quantity of nitrate value introduced into the leach liquor* - 577 - Stability of amine In nitrate stripping The amine solution (5f10,20#) was kept in contact with 3.0M and 6.0M NH.NO, solution containing 0.5N HNO,, for varying periods of time upto 60 days with periodic mixing every day. In another experiment 15$ amine solution and 1.5M NH.NO- +-U0» + 0.5N HNO* solution were kept agitated over a magnetic stirrer for 6 hrs./day for 3 weeks. In all these experiments, the amine solution was titrated for the amine content after conversion to free base form and it waB observed that after a slight reduction (<0«01M) in concentration ocourring within 24 hours, thetre was almost no change in the amine concentration. Uranium loading tests with amine also did not indicate any reduction in extraction capacity. However, this aspect will have to be confirmed in a pilot test using live sulphuric acid eluate solution. Amine regeneration The amine in nitrate form would be regenerated by contacting with dilute ammonia. The amine will now.be in free' base form and the nitrate values will be recovered as ammonium nitrate for reuse. The amine in base form will be converted to sulphate form by final contact with dilute sulphurio acid. Purification of uranyl nitrate using TBP Sufficient quantity of synthetic solution having the above composition was prepared for trials using TBP. - 578 - Preliminary tests indicated that the extraction capacity of this solvent could be remarkably improved by increasing the acidity of the.uranium solution to 1.0M HNO,. The saturation loading capacity of 30# TBP in kerosene at this acidity was found to be «-•. 130g U,Og/l» The loaded extract has to be scrubbed to remove the undesirable impurities and is a vital step for obtaining uranium of nuclear purity. Scrubbing the extract with 1.5M lJILNO,c containing 1.0M HNO, was carried out. The scrub raffinate contained around 24-g U,Og/l which can be recycled. Equilibrium isotherm ard McCabe-Thiele plot were drawn and it was observed that the uranium content of aqueous^ phase can be reduced to 0.1g U,Oa/l in 4 countercurrent stages at a phase ratio .A/0-1.25. The loaded extract would assay around 100g U,0Q/l. This is a value which ia normally maintained in the current practice of uranium refining. t Stripping of uranium from TBP phase is achieved using dimineralised water as usual and the strip solution can be obtained to assay 50-00g U.,Og/l to suit the requirements of the further ADU precipitation step. Conclusions These laboratory tests have .shown that it is possible to obtain concentrated solution of urany] nitrate from ths sulphuric acid eluates obtained from CIX operation, using the amine extraction technique. The amine is fairly stable to - 579 - to nitrate stripping and this strip solution can be further processed by TBP extraction to yield a reactor grade uranium product. Based on these trials, a tentative flow sheet has been finalised and is now being tested on a bench-scale using mixer-settlers. The major areas requiring closer look have been identified as 1) Standardisation for control of sulphate build up in the amine-strip cycle 2) Effect of residual calcium in recycle stripping agent 3) long range stability of the amine to nitrate system 4) Long range effect of nitrate recycle, on the TBPT purification step. In general* this flow-sheet involves a few steps like gypsum separation and nitrate concentration make up which could prove to be a hindrance in normal operations of the plant. If these 3teps prove to be unacceptable, one has the option to go for theconventional ELUEX process, using NH,-(NH.)2S0. stripping route to obtain high grade ADU which can then be Shipped to the uranium refinery anywhere in the country since it would not generate any active silica waste for disposal. Acknowledgement The authors are grateful to Shrl T.K.S. Murthy, former Head, Uranium Extraction Division under whose directions this work has been carried out. - 580 References 1. A. Raure etc. J.S. Afr. Inst. Min. Metall J56, March, 1966. 2. B.G. Meytourgh, J.S. Afr. Inst. Min. Metall, Oct., 1970, p 55 3» H.E. James, "Uranium ore Processing", STI/PUB/453, IAEA, Vienna, 1976, p 23. 4» S. Ajuria-Garza, W.D. Jamrack, "Uranium ore Processing" STI/PUB/453/, IAEA, Vienna, 1976, p 107 5« K.D. Kamat, T.K.S. Murthy, Divisional Report Chem. Engg. Divn., BARC, 1973. 6. K.B. Brown, USAEC Report - ORNL-2486, Dee. 1957, p 11. - SRI -> GROUND ORE: SLURRY -CONC. H^04 LEACHING at pH 1-5-2-0 TO REUSE SOLID/LIQUID SEPARATION FOR REUSE (LEACHING) .BARREN C. LX. BLEED RAFFINATE ADSORPTION CYCLE I:5 WASHING WAT|R - IOV. C.I.X. HgSQ, ELUTION CYCLE IOV. AMINE AMINE SX 10 V. HgSQi <> IN KEROSENE (EXTRACTION) AQ. RAFFINATE SOLVENT .SOLVENT AMINE S:t NH4OH REGENERATION (STRIPPING) < AQUEOUS RECOVERED rotjHOg AMINE FOR REUSE NH^N03 STRIP SOLN. HNO-i STRIP FEED CONDITIONING 0-5M HNO3 I NITRATE SOLUTION LNH4^^Jr-;- T.B.R SX (EXTRACTION) (tap REMOVAL) 1 SOLUTION BARREL EXTRACT TO STRIP T.B.R SOLK RECYCLED zap sx DM. WATER- STRIPPING WITH D. M. WATER I NUCLEAR GRADE URANYL NITRATE URANIUM PRECIPITATION A.D.U/A'A.U.C . RLTRAFILTRATIO1 N | _> FILTRATE FOR DISPOSAL TO UO2 PRODUCTIOT N FtQ.l. CONCEPTUAL FLOW SHEET BY STRONG BASE ION EXCHANGE- ELUEX ROUTE. KSK PBBMMCION Cg HUCLBft GRADB qaitHHM OPDK FBQH JlDUGUDA LSACa llQUOR V.M. Pandey, A3. Chakraborty and H.Maity PHAHioM ooaroajgioH OF IHPJA unnm JADUGODA MINES SUmiDHDH BIHAR Studies have been conducted on pilot plant scale for solvent extraction of uranium from sulphuric acid leach liquor of Jaduguda Uranium ore using ALmine-336 in kerosene aa extractant and Iaodecanol aa modifier. A solvent extraction set up of 10 litre/min. capacity, having 4 stages each for extraction and stripping, was used. The Uranium uaa recovered by stripping with 1M sodium chloride solution. The MDU precipitated from strip solution contained 85£ U^O . Since N.F.C. has planned to set up its plant at Turandih and a proposal was also made to integrate the process flow sheet of DCIL and HFC, it was decided to hare a detailed studies on the production of nuclear grade uranium oxide from leach liquor. The work was done in line with the eluex process adopting one mors stage o£ solvent extraction, i.e. after leaching in sulphuric acid and fittratilon, the clarified solution at pH of about 2 was processed through ion exchange system for extraction of uranium. The uranium was eluted with sulphuric acid. ?ron the eXuted solution uranium was again extr^Aed with Alaaino-336 in toros9n*. Ieo- decanol was used as modifier. Ttee stripping was done with ammonium nitrate - nitric acid, which is not a common practice. From the nitrate solution uranium was again extracted with TOP in kerosene and stripped with acidified water. From this strip solution ammonium dinranate was precipitated and ignited to uranium oxide. A parallel work was also conducted by extracting Uranium directly from - 583 - leach liquor by JiLsjnine-336 and processing further as mentioned abovo. In both the cases tbs Cranium oxid© produced contained the seme percen tage (99.8X) of %08. IHtBODUCTIDM The hydrometallurgical processing of Uranium Ore often involves four units of operationsj- (i) Leaching by chemical or Bacterial means, (ii) Separation of leach liquor from the ganguo material, (ill) Concentration and purification of the valuable metal by Ion exchange or Liquid-liquid extraction, and (iv) Precipitation of Uranium Concentrate. The Uranium Corporation of India Ltd. is Involved in the development of all the four operations, although most of the efforts are being directed towards chemical leaching and the concentration &nd purification steps. Since the uranium leach liquor normally contains IOBS than a gram of uranium per litre, separation and concentration by ion exchange or solvent extraction is ideally suited for such a solution. Both the process es have been ueed with ion exchange initially very popular. In the late 1950*8 the ion exchange was gradually replaced by solvent extraction. The recovery of uranium from ores by using solvent extraction began in 1955 with the use of D^EHPA, and since 1957, secondary or tertiary mlnos have been the most popular extract ants, particularly the tertiary aminos in recent years. The amine extraction process 1B known as the Aneoc process . The tertiary amine is selective for uranium in the presence of impurities such as iron, thorium, phosphate and rare earths. The Dapax process using OjKHP* was used originally for tho recovery of uranium from ores., but is not as selective for uranium as the tertiary amino in the presence of impurities such as ferric iron and rare oarths. - -584 - A combination of the two unit processes, *••»• ion exchange and solvent extraction, has also been used In certain circumstance s. The use of solvent extraction employing tertiary amino, following an ion exchange cycle is ia practice in many plants of the world. This has been called the eliieoc process, and is capable of production of uranium compounds of high purity 3'4. In South ifrica, a •tertiary amine was used to extract the uranium in pilot plant operation from the 1J0& sulphuric acid elnate from ion ex change processing . This process known a* Bufflex process was subsaquen- tly installed at the Harmony Gold Mines in South Africa , Also in South Africa, the Purlex process was developed subsequently to the Bufflex process, The pur lax process proved more economical than the Buff lex process?. The pur lax process is similar in many aspects to the Jaex process. The concentration and purification- steps In the recovery of uranian from Jaduguda uranium ore ia being dono in fixed beds of strong bass Ion ex. chang© resin. In most part of the world the fixed bed columns have been replaced either by the pur lax liquid-liquid extraction process or by con tinuous ion exchange process. Zee main advantage of purlax process over the fixed bed strong base resin process are the saving in elnent costs and the saving in capital cost. In addition the process has the advantage of producing purer product. She main advantage of continuous ion exchange system is tbe reduction in resin inventory. Saving in eluent cost and getting more concentrated elated solution are otner advantages. Although eo change was made in Jaduguda plant but in the new project at Turamdih a continuous Ion exchange system (Himslsy Column) has been proposed. Since at Turamdih WC is also going to set up one of its units for fuel preparation, a proposal was made by BJStO to integrate OCIL and NPG flow sheets for the production of nuclear grade uranium oxide. To achieve this objective it was proposed that after elution with left sulphuric aold in fflaeley Column, the elated solution should b* processed by another atep of solvent extraction before processing by TBP, It was deolded in one of the meeting! that elated solution will be processed with Alamine-3o6 - 585 - and the loaded uranium vdll bo stripped with nitric aold-ammonium nitrate eolation, 4b UC1L, pilot plant teats have been done on the purification and concen tration of -uranium from Jaduguda 3flach liquor using Alamine-336. The same work was extended and Alsmine-336 was used to extract uranina from elated eolation and stripping was done, with HH^HOj - HN05 solution. The stripped solution was further processed with IBP and uranium oxide was prepared. The quality of uranium oxide thus prepared was compared with the uranium oxide prepared by extracting uranium directly fros the leach liquor by Alanlne-336, stripping by HH4HD3 - HMD and then process ing with TBP as in the first case. This paper deals with various aspects of the above studies. BPfflSMBNTiL Ion Exchange >- in ion exchange column of 2 litre capacity was set up and 2 litres of wet settled resin was filled In the column. The leach liquor containing 0,47 g U^c/l vas passed through the column till the resin was completely saturated. The adsorbed uranium was eluted with lOSt (V/V) sulphuric acid. The eluted solution was analysed for uranium which contained 18.95g/l UjOg* Extraction!- The extraction of uranium from the elutod solution was studied using 30/C aloaino-356 and 5% ieodeoanol as phase modifier in kerosene in counter ourrent manner keeping the contact time 5 minutes and settling time 6 minutes. The effect of phase ratios on the extraction of uranium was studied with different concentration of U,°Q in the eluted solution. No. of extraction stages were also determined. Stripping^- Stripping of Jooded organio was accomplished in 2, 3 and 4 stages with an aquous ammonium nitrate - nitric aoid strip solution keeping 3 min, contact time and 6 min. settling time. Tho effect of aquous strip solution composition, phase ratios, no. of stages on stripping wore also studied. - 586 - TBP Extraction and. Uranism Oaddo Preparation: Jfaunoniun nitrate-nitric acid stripped solution was further processed with TBP and from the stripped solution of TBP circuit ammonium diuranate was precipitated and ignited at 850°C to OjOg. The uranium oxide thus produced was analysed for UjOg* The uranium oxide was also prepared frora tbi leach solution avoiding ion exchange route using amine extraction from laach liquor, followed by TBP extraction. RE3UI3S AND DISCUSSIOH (1) Extraction with Alamlne~336 (a) Effect of amine concentration on Uraniun loaiingi The extraction of uranium from the elated solution containing different concentration of UgOg with different concentration of amine was studied. The results are shown In Table-I. TALB- I THE ffVSGT OF AMINE CONCENTRATION OH DSANIUM LOADING Phase ratios(AsO) = 2il Contact time = 5 sin. Settling time = 10 min. Amine concentration Loading (U^ g/i) ] Losing (^o g/^ £xm from Elutod solution { eluted solution contaln- containlng 8.40 g/1 UgOgj ^ g/ ^p i 16#9S x 5 2.50 2.80 ^ 4ilC 7.35 6 6.26 9.oo 20 7.50 10.90 25 9.50 13.10 30 10.00 14.88 35 H.40 16#50 - 587 - From the table it is seen that the higher the uraniua concentration in the aquoua solution more is the loading, ilso a solution containing 16 g UgO^lit, requires 30fc amine for extraction. 2. The counter current Extraction at Different Phase Ratio{ The counter current extractions studies using six stages of extraction were done at tvo different aquous to organic phase ratios. The results are shown in tab 3^-11. TABLE- H COUNTER CUBWM? EXTRACTION U_0 in solution = 16.95 g/1 O O Jklattlne-556 = 50$ in kerosene Contact time = 3 min, Settling time = 6 min. Ho. of gtages»6 Phase ratio fhase ratio A$0 Itl AiO 2tl Aquous g/1 U30Q I Organic g/1 UjOj Aquous g/l OgOgjorganie g/l UgOg 16.95 14.64 16,96 14.37/.7 ^14.48 15.50^ ^14,88 12.77 & ;M0.06 15.80*"^ J>14.88 5.48^' ^^7.08 1S.30^[ ]>sL4.44 ^*4 ^ 12.7 Off ^Ul .44 ^*5^ 0.30 0.89 10.64 From the above tablo it is seen that by keeping iquoua to organic ratio 111 the extraction in six stages is more than 993», whereas by keeping aquous to organic ratio 2il only 9055 uranium can be extracted. 3. The Stripping of Uranium iron the Loaded Organic: The stripping of uranium from the loaded organic waB done by using different concentration of ammonium nitrate and nitric acid. The effect of variation of phase ratio was also seen. It was observed " that stripping solution containing 2»0 M M^IO and 1.5 M HJ»3 and organic to aquous ratio 5»1 was suitable for stripping. The counter current stripping was done in 2 adages, 3 stages and 4 stages. The results are given in Table-Ill. TiBLS - III STRIPPING OF URANIUM FROM THE IQADED OROTIC IN DIFFIRBg STAGES. Strip, solution 2.0 M ammonium nitrate ani 1.6 M JfflD3 Phase ratio 0»A = 5il Contact time = 3 min. Settling time = -5 min, — — — — — ••«• — « — — — — — «» — — — ——•>•— — — •-•»*• —_— ,?.#•.— — — — — — — — Uranium distribution g/1 • 8 stage stripping [ 5 stage stripping j 4 stage stripping 7 1 f Organic I Jquous phaeejOrgani iquoua i Organic J Aquous phase I no g/1 .phase n 0 ^/i phase | Phase | ^^ %°8 0^1 I 3 8 n 0 15.87 70.30 15.87 77.30 15.87- 78.00 6.60 30.70 6.93 41.00 8.28 47.50 1.80 - 2.73 12.20 2.32 14.50 0.41 1.26 4.47 0.26 - 589 - 4. Solvent Extraction of Uranium Directly free Jaduguia Leach Liquor. a* Extraction The equilibrium distribution data obtained Tor the extraction of uranium directly from the Jaduguda leach liquor uoing alanine- 336 are given in Table-IV, TJBLE - IV Solvent J- 5/» JU.anine-536 and 4*6 isodecanol as phase modifier in kerosene Phase ratio J- iquous i Organic = 6jl No, of stages = 4 Uranium distribution g/l U^O iquouB Phase J Organic Phase 0.60 3.50 0.414 2.46 0.060 0.674 0.0072 0.32 0.00176 b. Stripping; The uranium was stripped from the loaded solvent with 2.0 M Jtomoniun nitrate and 1.5 M nitric add in counter current manner in 3 stripping stages keeping organic to aquoue ratio 61L, The data obtained are given in Table -V. - 590 - URANIUM DISTRIBUTION, g/J. UgOg. Organic Phase Aquous Phase 3.88 22.0 2.78 13.50 0.35 3.50 0.092 (5) TBP attraction and UjOg Preparation j The stripped solution obtained from alanine circuit from both the extraction routes (in first case uranium was extracted from the' sulphuric acid eluted solution after ion exchange and in the second caBe uranium was extracted directly frco leach liquor) was extracted with IBP and stripped in usual manner. Anoonium diuranate was precipitated, dried and ignited at 850°C. The analysis of U30Q thus produced analysed as follovsj- Peroent UjOg i) lon-oxchange - Anlne extraction - TBP extraction route - 99.86 ii) Aaine extraction . TBP extraction 99.84 route From the above results it can be seen that the purity of uranium oxide, interns of uranium content, obtained from both routes is more or loss the sano. If tv6 steps of solvent extraction has to be adopted, probably it may not be necessary to have the Ion exchange circuit. - 591 - ACKKOWLEDGBlEWr The authors wish to thank Chairman and Managing Director, Uranium Corporation of India Limited for hi8 keen Interest in the research activities of C.R.&D Department. BEggtEHCSS 1, O.J. CroUBe and K.B. Brown, OBNL-2720, Oak Ridge National Laboratory, Oak Ridge, Tennessee, 1359. 2, K.B. Brown and C.F. Coleman solvent extraction in Pre Processing, Progress in Nuclear &*>rgy, Series 111. 2 , PP 3-34, Pub. Pergamon Press (1958). 3. - R. Simard, A. J. GiOmore, V.M. McNamara, H.W. Parsons and H.W. Smith, Can. J. Cbem. Engg., 59, No.6, 229-234 (1961). 4. J.W. Fisher and A.J. Vivyurka., Ion Exchange in the Process Industr ies, Pub. Society of Chemical Industry, London, 1370, PP 163-167. 5. JUFaure, S, Finney, H.P. Hart, C.L. Jordaan, D. Van, Haerden, E.B. ViOJoon, S.E. Robinson and P.J. Lloyd, Proceedings of Panal on Processing Low Grade Uranium Ores, IAEA, Vienna, 1967 PP 119-141. 6. R. Cross, J. South African Inst. KLn and Met all. Nov. 1968 PP 196-201. 7. T.H. Tunby and A. Faure, B The Purlex Process " Presented at Annual Meeting, AIKE, Washington, Feb. 1969. - 592 - URANIUM RECOVERY FROM PHOSPHORIC ACID G. SIVAPRAKASH DESIGN MANAGER FACT ENGINEERING AND DESIGN ORGANISAION Wet process phosphoric acid is a handy and reliable source for uranium. Recovery process is based on solvent extraction technique. DEHPA-TOPO mixture is the preferred extractant. It is possible to design and operate a trouble-free recovery plant if attention is paid to the inherent characteristics of WPA, selection of process steps and equipment. - 593 - 0 INTRODUCTION Phosphoric acid is a basic chemical used for manufacture of fertilizers, detergent intermediates/ pharmaceuticals and animal feed products. Wet process phosphoric acid is a term now used exclusively for the product made by digestion of phosphate rock with sulphuric acid. 1 Presence of uranium in phosphate rock Historically, the existence of uranium in sedimentary phosphate deposits of marine origin was first reported by R.Strutt in 1908. In general uranium is present as a trace constitutent ivs all apatites. Statistical geochemical data shows that sedimentry phosphate deposits of marine origin have a much higher uranium content than those of. an igneous origin. Typical overall ranges are 10-30 ppm U in guano • type, 10-100 in igneous and 50-300 in marine origin sedimentary phosphorites. Besides, the geo-chemical origin, the geographical location of the deposit and its degree of weathering also influence the uranium content of phosphate rock. Table I gives the uranium content of various phosphate rocks. 2 Presence of uranium in phosphoric acid During the acidulation of phosphate rock by sulphuric acid to produce wet process phosphoric acid (WPA) of 25-30% P,0- content, most of the V - 594 - TABLE I AVERAGE URANIUM CONTENT OF PHOSPHATE ROCK Average uranium content ppm U ppm U_0 Algeria 125 147 Christmas Island 80 94 India (Rajasthan) 25 30 Israel 110 129 Jordan 140 165 Morocco 110 129 Senegal 90 106 S. Africa 90 106 Togo 90 106 Tunisia 60 71 USSR 40 47 USA 150(70%) 176 80(30%) 94 Nauru 80 94 - 595 - TABLE II MANUFACTURERS AND LICENSED CAPACITY. OF PHOSPHORIC ACID (1988) Name of factory Capacity of acid as tonnes P^O^per day 1. ' Fertilisers and Chemicals, Travancore Ltd., Cochin Divi sion, Kerala. 360 2.* Fertilisers Corporation of India, Sindri. • 360 3. Coromandel Fertilisers, Vizag 325 4.* Hindustan Copper Ltd., Khetri 210 5. Southern Petrochemical Indu stries Corpn., Tuticorin, Tamil Nadu. 165 6. Gujarat State Fertiliser Co., Baroda. 165 7. Fertilisers & Chemicals Trava core Ltd., Udyogamandal Divi sion, Kerala. 125 8. Hindustan Lever, Haldia, W.B. 140 9. Rashtriya Chemicals and Ferti lisers, Trombay. 100 750 TPD plant of Paradeep Phosphates Ltd. is expected to be commisioned by the middle of 1990. * These units are not operating at present. - 596 - uranium in the rock gets sblubilised. 90-93% of the uranium content of phosphate rock appears in phosphoric acidr and the balance 7-10% goes with the solid residue which is mainly gypsum A typical WPA manufacture flow diagram is shown in fig. 1. A list of major Indian manufacturers and licensed capacity of phosphoric acid is given in table'II. t With the commissioning of the 750 TPD phosphoric acid plant of Paradeep Phosphates Ltd. in 1990, about 0.5 million tonnes of phosphoric acid will be available for uranium recovery. This is after giving due allowance for uneconomically small plants and production of acid in the concentrated form. In addition to indigenous production, India imports more than one million tonnes of phosphoric acid. Table III gives annual import- of phosphoric acid during the past few years. Table III ANNUAL IMPORT OF PHOSPHORIC ACID Year Quantity MT P20 1984-85 6,11,029 1985-86 7,55,812 1986-87 10,74,574 - 597 - 98 V. H2SD4 WATER FOR WASHING FLASH CODLING i_i PHOSPHATE ROCK u REACTION DIGESTION FILTRATION GYPSUM DILUTE ACID RECYCLE WASH WATER w PRDDUCT. TACID 27-30 /. P205 F'S-.l- TYPICAL FLOW SHEET FOR WET PROCESS- PHOSPHOR1C ACID MANUFACTURE -• «593 - Assuming that about 50 percent of imported acid is available for uranium extraction, the combined (indigenous plus imported) quantity of one million tonnes can produce 300 tonnes of U-Og per year. This can satisfy only a fraction of the total needs of the country, but this has the attraction of being a stady and handy source. ECONOMIC VIABILITY OF URANIUM RECOVERY FROM WPA Economic viability of recovery depends on i) Price of U^Og in the market. ii) Government Policy Uranium recovery from phosphoric acid can economicallysustain if the international price of one Kg of uranium is not less than around 70 dollars. International price, in turn, depends on the cost of energy. Thus during the oil crisis in the seventies, recovery from WPA found itself very attractive but afterwards the prospects became bleak. Nevertheless, in national level planning, considerations other than profitability viz. essential nature and dependable source are likely to favour uranium recovery from WPA , even in the face of high costs. Phosphoric acid which remains after extraction of uranium is devoid of most of its sludge and colour and can be easily processed to food grade phosphoric acid. In countries like Belgium, this has been one factor which favoured the extraction - 599 of uranium. PROCESS ROUTES FOR EXTRACTION FROM WPA It is not economical to extract uranium from concentrated, phosphoric acid (more than 35% P20c). For recovery from dilute phosphoric acid a number of routes have been followed in the past. All these routes are based on the principle of solvent extraction. Because of their inherent weaknesses certain routes proved to be non-viable. The OPPA process This is the oldest process. Developed in the fifties, this was used in two plants in Florida. As extractant it uses an alkyl pyrophosphate, commonly octyl pyrophosphoric . acid (OPPA) prepared at the plant site by reaction of octyl alcohol and P2^5* Main steps of this process ares - reduction of uranium in the phosphoric acid to ..+4 by means of Fe - four stage counter-current solvent extraction + 4 of U with a solution of OPPA in Kerosene. +4 - re-extraction of U with H-SO. under simulta neous precipitation of calcium. - Precipitction. of uranium in HjSO. as UF. (green cake)with HF. 600 - - regeneration of the pyrophosphoric acid. This process is currently being used only by Gardinier at their Tampa facility in a modified form. The OPPA process is characterised by low costs and a high distribution coefficient (20-40) of reagent. However, it has a major disadvantage; the extractant is unstable (due to hydrolysis) and can undergo only limited recycle. Because of its consumption, reagent costs are therefore important. The OPAP process This is based on the extraction of tetravalent uranium by a mixture of mono and diesters of octyl phenyl phosphoric acid. This process was originally developed at the uranium Extraction Division of BARC and was later modified by ORNL to take advantage of its chemistry. The "modification involved use of OPAP for the first cycle of SX in the ORNL process. The OPAP process consists of the following steps. - cleaning of.the feed acid. - 4 stage counter-current extraction of the acid at 40-45°C with 0.3 - 0.4 M OPAP in an aliphatic diluent. - Stripping with 10M phosphoric acid containing NaClo, -601 - dilution to 5-6 M. - extraction at 25-45°C with DEHPA/TOPO in an aliphatic diluent. - Scrubbing. - Stripping with ammonium carbonate and formation of ammonium uranyl tricarbonate (AUT). - filtration. - Calcination to U3O3 NaClOo is added to the 10M phosphoric acid used for stripping of the OPAP solution in order to oxidize the uranium to the hexavalent state which is ;less soluble in the organic phase. The stability of OPAP while better than that of OPPA is still less than that of DEHPA-TOPO. Compared with DEHPA-TOPO, OPAP shows the * following, advantages. - OPAP is a more powerful uranium extractant by a factor of 3 or 4, its distribution coefficient being 25-30. ' +4 - it extracts U which is the dominant uranium valence in freshly . produced'' - wet process phosphoric acid. - in the OPAP process a solution containing U is produced and therefore a valence adjustment is not needed for processing in the second cycle. I - 602 - - at process concentrations, OPAP is less expensive by a factor of 2 to 3. This route was tried in two commercial plants including the Earth Sciences Incorporated plant at Calgary in Canada. Due to certain operational problems, these plants have also switched over to two cycle DEHPA + TOPO process. 3.3 DEHPA - TOPO Process This process was originally developed at the Oak Ridge National Laboratories of USA and hence known as ORNL process. It involves two cycles of solvent extra-ction of uranium using a mixture of Di-2-Ethylhexyl phosphoric acid (DEHPA) and TOPO (Tri-n-Octylphosphine oxide). 0 I HO — P OR I OR R DEHPA TOPO R = CH2 CH(CH2)3 CH3 R = C8 H C2 H5 - 603 - These reagents, dissolved in Kerosene, extract hexavalent uranium. Fcr maximum efficiency, all the uranium should be in this valence state. Therefore, the phosphoric acid has to be oxidized prior to the extraction. The process comprises the following steps, (see fig. 2). - cleaning of the phosphoric acid by coagulation and filtration. - 4 stage counter-current solvent extraction of the phosphoric acid with a 0.5M DEHPA/0.125M' TOPO solution at around 40°C. - 3 stage reductive stripping of the organic phase with a small flow of raffinate- reduced by iron metal. - the strip solution, which is about 70 times richer in uranium than the original acid, is oxidized with oxygen of H-O-. - hexavalent uranium is extracted with 0.3M DEHPA/0.07M TOPO. - the organic extract is scrubbed with water to remove extracted phosphoric acid. - from the organic phase uranium is stipped with a 2-3M ammonium carbonate solution. Uranium precipitates as ammonium uranyl tricarbonate (AUT). FIG-2 •FLOW-SHEET OF URANIUM EXTRACTION PHOSPHORIC ACIt> —j, ACID PRE.-TREATMEMT cooling Desupers atu ration Floculalion solid-lictuld Separation >'Gypsum Adsorption •*. OXIDATION CONCENTRATION H2OZ r CYCLE EXTRACTION -*- P-affinale 0 liifi 4 STA6E5 phosphoric acid plant UJ t- LstCYCLE. STRIPPING JZ IRON Ou) 3STAGES >-> V4W OXIDATION HzOz neJ 2 CYCLE EXTRACTION d 4 STAGES -*-2" Cycle rcvfftnale U „3- >- H2O PO4. SCRUBBING N nd (NH4.)2C03 2 CYCUE STRIPPING SOLUTIOS 2 STAGES r H^SOA ^.iNH*)2SO4S0LJTlOM SOUUTIOH SOLVENT REGENERATION TO FERTILIZER PLANT .C02 • Nllj H20 SO^'SCRUBBI NG [- VAPOR CONCENTRATION ABSORPHOH- URAN. SALT PRECIPITATION GAS SOLID.LlQ/. SEPARATION VELLOW CAKE „ I H2SO4. DRYING <& CALCINATIO3N - U308 - AUT is filtered off and calcined to U-,0o. J O DEHPA-TOPO is characterised by a very low distribution coefficient of 5-10. Uranium recovery from the acid is about 95%. The DEHPA-TOPO process is being used by most of the facilities .engaged in uranium recovery from phosphoric acid. Drawback of this process is high cost of reagents. UNIT OPERATIONS IN THE URANIUM EXTRACTION PROCESS In the DEHPA-TOPO process described above, which is presently being employed by majority of the manufacturing firms, following process steps are involved. i) Acid pre-ireatment ii) Oxidation/reduction of acid, iii) Solvent extraction/stripping. iv) Acid post treatment. v) Precipitation jof uranium salt, vi) Filtering and drying of uranium salt. Unit operations involved in these process steps are described in the following paragraphs: Acid pre-treatment WPA contains about 1 percent by weight suspended jsolids and sorn-2 dissolved organic? (humates). Source of humates is the organics present in the - 606 - phosphate rock. Colour of WPA ranges from green to brown to black with increasing amounts of organic matter in the acid. Cejitrai Florida rock which is rich in organics gives a black acid whereas Moroccan rock gives a brown acid and Idaho calcined phosphate gives a green acid. Apart from humates the acid is supersaturated with gypsum and fluosilicates which tend to precipitate during uranium recovery operations. Also, during extraction these humates and solids form emulsions (called 'crud' or 'gunk') at the organic - aqueous interphase of the extractor making phase separation difficult. Poor phase separation shall result in solvent loss. Therefore it is necessary that the acid is cleaned to bring down the suspended solids to around 50 ppm and the dissolved organics to such a level that the acid becomes transparent and light green' in colour» Significantly more pretreatment is required for black acid compared to green acid. A suitable combination of the following operations can be used for this purpose. 4.1.1 Cooling and aging the acid. Cooling of the acid helps in two ways: a) It increases the distribution coefficient of uranium. b) It decreases the solubility of gypsum and silicate solids in acid and therefore more solids crystallize. This reduces the possibility of solid deposition in downstream process equipment. - 607 - Aging is allowing the acid to remain in a vessel with mild agitation. This helps the crystals to grow and eliminates super-saturation of gypsum in" acid. Normally two types of cooling is adopted in industry. a) Flash cooling. This works on the principle of evaporative cooling. Principle of a flash cooler is shown in fig.3. b) Spiral cooler. In this type of cooler there is no direct contact between vapours and cooling water unlike in the case of flash cooling. Flow pattern in a spiral cooler is shown in fig. 4. An important design parameter in these coolers is the AT per pass in the case of the flash cooler and the /S. T between the hot fluid and the cold fluid in the case of the spiral cooler. Higher AT tends to increase scaling in the equipment. 4.1.2 Use of activated clay. It is used as an adsorbent as well as filter aid. It adsorbs humic materials. It also makes up the low solid content of the acid, thus improving the down-stream operation viz. filtration. Activated clay is mixed with the acid in a vessel. 608 - STEAEAMh STEAM EJECTOR COOLING EJECTOR VATER FLASH VESSE CONDENSER HEATED OUT FIG, 3 PRINCIPLE,. _flE.ft RASH COOLER COOLED DUT COLD IN> HOT IN FIG. A FLOV PATTERN TN A SPIRAL (Ml FR - 609 - 4.1.3 Flocculaion Flocculants are chemicals Which can destroy the electrical charge of the suspended particles present in the acid. A suitable type of flocculant (cationic, anionic or nonionic) has to be chosen on the basis of settling studies done in the laboratory. Equipment used for flocculation is an agitated vessel. Minute particles of gypsum floe together and form bigger, easily settleable agglomerates. 4.1.4 Sedimentation or filtration Clarifier is the equipment normally used for sedimentation. It gives the acid enough residence time and imparts a mild rolling action. Solid particles settle at the bottom as a sludge and is removed - through a bottom tapping. Sludge is recycled to the phosphoric acid plant, either to the attack section or to the filtration section. Instead of a clarifier a filter also can be used. A vacuum belt filter or pan filter can be employed. 4.1.5 Gunk removal Another pre-treatment method involves contacting phosphoric acid with Kerosene to form crud deliberately; this crud then being removed in a - 610 - separate step, priorto solvent extraction. After solids removal in a clarifier th-2 cleaned acid passes to a gunk forming mixer settler. Gunk produced is tapped from the interface of the Settler. Gunk has to be filtered or centrifuged for recovering acid/organic adhering to it. The liquid mixture from the filter is phase separated and recycled to clean acid/kerosene streams. By accomplishing humate removal in this mixer-settler/ real extraction mixer settler: are relieved of cruds. In plants where gunk removal is adopted, the normal practice is to use a clarifier instead of a filter and to do away with the activated carbon adsorption step. 4.1.6 Adsorption by activated carbon Granular activated carbon adsorbs the colloidal matter as well as dissolved organics in the acid. Two columns in series plus one stand-by column are sufficient for acids of medium organic content. Freshly back-washed column shall be at the down-stream end. The columns require occasional back-wash to drive out the sediments. Regeneration by dilute caustic is done once in a week. Gradually the carbon gets depleted of its activity irriversibly and needs replacement. (Once in 2-7 years, depending on the impurity level of the acid). For ease of interchanging the - 611 - columns, automatic opening and closing of valves wiit h micro processor control can be employed. 4.2 Oxidation of acid This step is for transforming he valency of uranium from 4 to 6, which gives maximum transfer from the aqueous to the organic medium. Oxidising agents that can be used are: i) Gaseous oxygen Oxygen is stored in the liquid state in bullets. It is expanded through pressure reducing valves. Arrangement for providing the latent heat of evaporation is required. By the use of fine bubble nozzles', tf\e gas is bubbled into the acid. ii) Hydrogen peroxide Advantage of using hydrogen peroxide is that it gets converted to- water and leaves behind no undesirable impurity. But the cost of this chemical is high. Therefore bulk of the oxidation requirement can be met by oxygen and the final correction by hydrogen peroxide. iii) -Sodium chlorate. This is very convenient to handle unlike oxygen, but chlorine liberated can cause corrosion of equipment used in the acid consuming industry. - 612 - v) Nitric acid. This produces acceptable oxidation, but adds nitrates to the phosphoric acid, which is an undesirable additive. The amount of oxidising chemical required can vary substantially from day to day depending on the rock source and the operating conditions of the phosphoric acid attack circuit. The Fe + 2 level is monitored continuously to optimise oxidising chemical usage. 4.3 Solvent Extraction 4.3.1 Mechanism of extraction with DEHPA Di(2-ethyl hexyl) phosphoric acid is described as an acidic extractant which can de-protonate to form an anion. + (C8 H17 0)2 P - OH ^=^ (C8 H17 0)2 P-0~ + H This hydrophobic anion can chelate a cation from an aqueous phase, the chelate being soluble in the organic phase. The extraction is an equilibrium which can shift in either direction dependent on |[H J. As the hydrogen ion concentration decreases, the formation of DEHPA:U complex is favoured. - 613 - Consequently, DEHPA is a pH dependent extractant. Yet, the pH alone is not the only consideration Distribution coefficient for extraction (which is the ratio of uranium concentrations in the org-a- - nic phase an'a in the aqueous phase) is temperature sensitive. Rise in temperature causes fall in distribution coefficient. Temperature has an effect on rate of extraction. The rate of extraction is the rate of transfer of solute from the bulk of the raffinate phase to the interphase and thence into the bulk of the solvent phase. The rate of transfer will be proportional to the interfacial area separating the tv/o phases and the departure of their compositions from the equilibrium situation. This may be represented as Rate = K > a • Ac where the constant of proportionality K is a mass transfer coefficient, a is the interfacial area, and Ac measures the departure of the phase compositions from equilibrium. Rise in temperature increases the rate of extraction, by influencing the mass transfer coefficient. - 614 - Rise in temperature will enhance the rate of coale scence or phase separation. This is quite sensitive and a 10°C rise in temperature can easily halve the thickness of a dispersion band, other factors remaining constant. Lower temperature will help to reduce solvent loss. Normally extraction is carried out at temperatures ranging between 30°C and 40°C. Temperature adopted for a particular extraction plant depends on incoming temperature of acid and temperature of cooling water available. n c Lower P2 c °ntent in the acid favours transfer of uranium to the organic phase. TOPO (Tri-Octyl Phosphine Oxide) is used in combination with DEHPA as a synergist, to improve or accelerate extraction. 4.3.2 Equipment for extraction For solvent extraction rectangular or circular mixer-settlers can be used. Even though the phosphoric acid has been cooled, clarified etc., it still has a tendency to scale and post precipitate to ^ome extent. This must be recognised while designing the vessels, heat exchangers and instruments. Turbine typo mixers used in the mixing compartments provide pumping action. From the settling compartments recycle connection is given to the mixer compartment. - 615 - Outlet port for the mixed phase from mixing compartment should be arranged in such a way as to minimise the width of the dispersion band. In the system, aqueous to organic ratio is maintained at about 2:1, but the last mixer settler shall be operated under organic continuous phase, thus minimising organic entrainment in the acid. Primary stripping Loaded organic from the primary extraction step is stripped with phosphoric acid of 45 to 50% P-Oj. strength- Temperature of the acid is maintained at r »'nd 50°C. The strip acid volume will be about .t. .5 percent of the main acid stream. It is provided with sufficient quantity +2 of Fe ions by reducing with iron. Reduction is done either in a separate cone-shaped 'reduction vessel packed with iron nodules or in the mixer-settler itself by adding iron to the +2 settling compartment. Fe content is monitored by maintaining the redox emf of the stream. Normally 3 stages are sufficient for stripping. Loaded strip acid will contain about 10-12 g/1 of U3°8' . Acid post treatment. The raffinate acid from the primary extractor flows to a -Lamella settler and then to flotation cells where the solvent content in the acid is brought down to 50 ppm. Higher solvent content, if permitted in the acid, may damage the rubber - 616 - lining of WPA concentration plant and result in solvent loss. 6 Secondary extraction and stripping The loaded strip acid is diluted to 30 percent and is oxidised with oxygen and H_0_ and sent through a second DEHPA-TOPO extraction circuit. Rectangular mixer-settlers are used to transfer uranium from the acid to the organic phase. Lean (spent) strip acid is returned to the primary extraction circuit. Loaded organic is sent through a water wash to remove any entrained P_05 before transfer to a secondary stripping circuit. In the secondary strip circuit/ loaded organic is mixed with a dilute ammonium carbonate solution to remove the uranium from the organic/ and convert it into an ammonium salt in the dissolved state. The aqueous solution passes through an after-settler for separating solvent. Barren organic is regenerated with phosphoric acid or sulphuric acid which supplies H ions to the DEHPA, thus re-converting it to the acid form. 7 Uranium precipitation The uranium salt solution is evaporated. During this step some ammonia and carbon dioxide are expelled from the solution and ammonium salt of - 617 - uranium gets precipitated. It is thickened, filtered or centrifuged, dried and packed as 'yellow cake' in drums. 8 Production of U3O0 Yellow cake produced can be transported to a uranium refinery. If the purity of yellow cake is not found -to be satisfactory further purification by re-extraction and stripping is done. Tri-butyl phosphate (TBP) is used as the extractant. The stripped uranium is precipitated, dried and calcined to get U^Og. 0 HOW TO OVERCOME OPERATION PROBLEMS? r Instability in mixer-settler operation Maintaining the aqueous-organic intephase at the correct level is a must for ensuring quality of the separated phases. Steady, controlled flow of organic as well as aqueous stream is a must for this. Good quality metering pumps or installation of constant head gravity flow arrangement can reduce this problem. • For solving the emulsion formation problem, attention is to be paid to optimum speed of mixers, avoidance of large pfiase ratios, satisfactory re-cycle piping and proper positioning of the mixer outlet port. - 618 - 2 Crud formation This is the major problem to be encountered when handling black acid or brown acid. Amount of crud formed is not proportional to organic content; in the sense that after a threshold level crud formation increases drastically. Careful pre-treatment of the acid is the solution for this. 3 Control of iron consumption +2 Iron is used for producing Fe ions. But these +2 Fe ions can get re-oxidised in the mixer-settler because of the agitation and contact with ambient air. If precautions are not taken to minimise re-oxidation/ iron consumption will be high/ resulting in high ppm of Fe in the strip acid (which in turn will result in high iron in the acid returned to the fertiliser plant). Maintaining an inert atmosphere in the mixer-settlers and/or addition of iron at a number of stages rather than as a single shot* can minimise 'iron consumption. 4 Iron precipitation as ferric salt In the secondary extraction • cycle, after oxidation, the acid is likely to contain precipitated ferric oxide. This has to be removed by filtration; otherwise fouling of the interface in secondary extraction settlers can result. The secondary strip solution is also likely to contain precipitated iron. This has to be - 619 - fUtere'd out to present contamination of yellow cake. 5.5 Settler clean-out Occasional flushing and cleaning is required for the settling compartments to remove the settled solids. Frequency of this is found to be roughly once in two weeks.^ 6.0 CONCLUSION , Extraction of uranium from phosphoric acid is a technically viable operation followed in many countries. Setting up of extraction plants depends on economic as well as self-reliance- considerations . REFERENCES 1. Couloheris, A.P. (1979) - "Uranium Recovery from Phosphoric Acid", paper read before The Fertilizer Society ofLonon. 2. McKetta, John J.(Ed.) (1984) - Encyclopedia of Chemical Processing and Design. Vol.21. Marcel Dekker, Inc., New York. 3. Hawyler, S. (1981) - "Uranium Recovery from Phosphates and Phosphoric Acid". Federal Institute for Reactor Research, Wuerenlingen, Switzerland. 4. Koppiker, K.S. - "Design of Solvent Extraction Flow sheet for Metal Recovery - General Laboratory Practice and Industrial Approach".. - 620 - 5. DEHPA Metal Extractant" - Product information bulletin No.5XBl, Albright & Wilson Americas; Inc., Richmond, U.S.A. - 621 - ON-SITE TESTS FOR RECOVERY OF URANIUM FROM WET PROCESS PHOSPHORIC ACID AT F.A.C.T. H.Singh, R.A. Nagle, A.B. Giriyalkar, M.F. Fori sec a K.S.Koppiker Uranium Extraction Division, B.A.R.C. Recovery of uranium -from wet process phosphoric acid has been studied at U.E.D. CBARC) during the last decade. A two cycle DEHPA-TOPO solvent extraction process has been tested on pilot plant scale on -fresh acid produced at FACT, Cochin. Results are summarised in this paper. The acid was subjected to pre-treatmen t involving filteration, activated carbon adsorption, oxidation and heating. The treated acid was ' processed by solvent extraction and raffinate was subjected to post treatment involving settling and coalescence. The organic extract was stripped o-f ura'nium by stray phosphoric acid under reducing conditions. Uranium •from strip solution was recovered by a second cycle o-f extraction a-fter dilution and oxidation, -followed by alkaline strip and caustic precipitation. The e-fect o-f various process parameters has been studied in detail and a ppocesr, -flow-sheet has been suggested for a -full-scale commercial plant. I. INTRODUCTION The wet pocess phosphoric acid (WPA) produced -from imported rocks in. the plants of Fertiliser and Chemical-, Travancore Ltd. (FACT) at Udyogmandal (UDL> and Cochin Division (CD) contains significant amount o-f uranium. Recovery o-f uranium -from phosphoric acid has been studied at UED, BARC -for over a decade(l). The process developed at 0RNL<2) have also been examined. Initial work was done at Bombay using the WPA -from R.'C.F. and A.M.P. Ltd. It was found that 'aging' of acid significantly affects process performance hence a decision was taken to carry out tests at site in F.A.C.T., Alwaye. A small pilot plant for continuous testing was set up at the R&D Division of FACT. Salient results are summarised in this paper. I I . PROCESS FLOWSHEET The recovery of uranium from WPA involves pre-treatment, solvent extraction and post-treatment<3). The process flowsheet for test-work is shown in Figure I. Pre-treatment includes filteration to remove suspended inorganic materials The data from on-site tests was useful in techno-economic evaluation of a full-scale commercial plant. III. TEST RESULTS 3.1 Pre-treatment; The fresh acid contained 22/< P2O5, 1 .26'/. F, 0.1 g/1 U30B, 2.58 g/cc Fe and had a density of 1.25 gm/cm3 with an e.m.f. of 250 mv. It contained about 2'/. suspended solids which separated by sedimentation. The colour was dirty brown having an optical density at 408 nm of 1.? in 1 cm cell. Activated carbon was mixed and acid filtered on gypsum pre-coat filter of batch nutsche type. The clear acid was greenish, with absorbance reading of 0.6. While initial tests were done with oxidation by KC103, subsequent work was done after oxidation with H202. Quantity needed <6 ml/1) was much higher than stoichiometric requirement for conversion of Fe2+ to Fe3+ and U4+ to U6+. The oxidised acid with an emf of 350 mv was suitable for solvent extraction. Analysis confirmed that Fe2+ level was 1ow (0.0 6 gp1). 3.2 Primary Extraction: This was.carried out in a 3 stage mixer settler having provision for recycle of organic phase to maintain organic to aqueous <0/A> phase ratio in mixer of 1.5:1. Agitator was six-blade top -shrouded pneumatically driven turbine. Feed rat.e was varied from 30 1/hr to 66 1/hr and no significant effect on recovery was found. Corresponding residence time for 60 1/hr is 0.6 minutes, which shows rapid rate of mass transfer. The solvent rate was varied and minimum A/0 ratio of 2.4:1 was necessary for high recovery < >?5X) . At A/0 = 4.3 recovery was only 90X. Nature of oxidant (KC103 or H202) had no significant effect. Temperature had a marginal effect on recovery (.2'/.') but \jery significant effect on coalescence flux in settler. At room temperature (29°C) flooding occurred at flow-rate of 54 1/hr with band width of emulsion zone being 52 mm. By comparison at 39°C and 56 1/hr band-width was 20 mm and at 43°C and 66 1/hr it was only 12 mm. NcCabe-Thile diagram was evaluated by stage-to-stage samples. A linear isotherm with si ope of 10.3 was obtained. Crud formation was negligible with stored acid from UDL plant, der i ved f rom Moroccan rock.' However, with fr^'-.o acid from CD plant, extensive crud formation occurred if carbon treatment was not adopted. About 10 litres of crud - 623 - was collected -from 500 1 of acid. Within -few hours the settlers were -flooded and transfer 1 ines choKed. Crud •formed at interphase and grew into organic phase. With carbon treatment no problems were encountered. Phase contiuity was controlled. While raffinate stage operated organic continuous, the -feed stage was operated aqueous continuous. The continuity was readily visible by colour, o-f mixed phase. Phase inversion was observed when overall phase -ratios at start-up were in the range of 1.5:1 to 1:1.5. 3.3 Ra-f-finate Post-Treatmen t * When the extraction was carried out at room temperature without a mixed-phase baffle and inter-phase at low level, the entrainment o-f solvent was high When KC103 was used as oxidant, the chloride level increased -from 70 ppm to 170 ppm. With peroxide o-f-course there was no increase. The acid a-fter treatment was returned to -fertiliser plant and no problems were, reported. 3.4 Primary Stripping, : Stripping was -found to be incomplete with dilute acid even with 75 gm/1 o-f FeS04. Hence strong acid (55X P205) was used. The acid contained suspended and organic impurity. These were removed by passing hot <90°C) acid through a bed of activated carbon. This reduced 3*>sorbance value -from 1.86 to 0.6 and performed satisfactorily in stripping. However, heating was necessary to reduce viscosity and increase clarification rate. With 75 g FeS04/l the emf reading was about 120 mv which remained invariant with storage over weeks. Stripping was carried out in mixer-settler provided with aqueous recycle and anti-vortex baffle plate in mixer. Overall strip phase ratio of 0/A = 16/1 could be used to increase product concentration without loss of efficiency. Residence time needed is high, over 3 minutes. A jelly 1j Ke emulsion formation was observed when operating at room temperature and with aqueous continuous. This is due to high viscosity of aqueous phase. Hence the battery was operated organic continuous at 45-47 C. These results were val idated by batch tests. - 624 - 3.5 Secondary Extraction: The strip product solution -from primary extraction cycle was diluted in 1:1 ratio with water and re-oxidised with 70 ml/1 of H202 to 370 mu. Re-extraction with solvent was carried out at A/0 = 2.2-2.5 and raffinate recycled to -first stage. Recover/ was high over 96'/.. Organic extract was scrubbed with water in 3 stages at A/0 = 1/3 to remove co-extracted phosphate. Uranium loss was negligible. Stripping was carried out with 15X w/v solu t i on of Na2C03 to obtain a strong solution of over 7 g/1 U308. For rapid phase separation 0/A ratio of 1:2 was maintained while mixing. During alkaline strip, a volume increase of organic phase was observed due to neutralisation. Uranium was recovered from strip solution by caustic precipitation and filteration. The crude filter cake was found to contain 46/. U308, 10/. Fe and 3/. P04. Moisture was \9'A. This cake was refined to nuclear purity by re-dissolution in nitric acid and TBP extraction, followed by ammonia precipitation. No problems were encountered in ref i n i ng. 3.6 Solvent Extraction Material Balance: Material balance showing typical flow-rates and uranium concentration at various stages of operation is shown in Figure II. From 60 1/hr WPA feed containing 0.095 g U308 g/1, extraction with solvent at 24 1/hr yields an extract of 0.255 g/1, stripping with strong acid at phase ratio 0/A = 16/1 yields a concentrated solution, of 4.2 gpl. After second cycle extraction and precipitation, over 90'/. recovery is obtained. IK>. CONCLUSIONS Based on the test-work, reagent consumption has also been estimated. Preliminary scale up and economic assessment shows appreciable amount of uranium can be recovered commercially at a cost which is acceptable under Indian conditions. The process feasibility has been demonstrated at site and further action is in progress for detailed engineering. ACKNOWLEDGEMENTS The authors are grateful to Shri T.K.S. Murthy, ex-Director, Chemical Engineering Group,BARC for guidance. Assistance'of FACT staff under the guidance of Dr. A.P. Rao, DGM REFERENCES 1. T.K.S. Murthy, V.N. Pai and R.A. Nagle Recovery of uranium by phenyl phosphoric acids, in Proc. Symp. on Recovery o-f Uranium -from its Ores IAEA, Sao Paulo, Brazil 1970 2. J.Hurst, D.J. Crouse and K.B. Brown Recovery o-f Uranium from Wet Process Phosphoric Acid Ind. Eng. Chem. (Process Des. Develop.) 11(1), 1972, pp 122-128 3. 'Recvery of Uranium -from Wet Process Phosphoric Acid' Parts I & II, Phosphorous and Potassium, III, 1981 pp 31-38. 4. Hand Bok o-f Environmental Data on Organic Chemicals Van Nostrand Reinhold Co. 1983, pp 528 626 Fig. t. PROCESS FLOWSHEET FOR ON-SITE TESTS WET PROCESS ACID MERCHANT GRADE (30 V. P^) (ACID 54 V.) FROM PH0S.AC1D PLANT HEATING TO40'C HEATING 70" C ACTIVATED CARBON ADSORPTION CARBON ACTIVATED CARBON " ADSORPTION 'CARBON GYPSUM . FILTERATION FILTERATION PRE-COAT OXIDATION REDUCTION I . HEATING ... A5*-5rfC HEATING 70" C 1 * DEHPA • • ' TOPO -J ' PRIMARY EXTRACTION PRIMARY LEAN. KEROSENE EXTRACTION STRIP. "SOTJJEW RAFFINATE i DILUTION ~-WATER AFIER-SETTLER 1 OXIDATION —H^ COALESCENCE COLUMN ' ' SECONDARY * EXTRACTION WPA RETURN TO PHOS.AC1D PLANT J JSCRUBBING 1— —1 NaOH| WASTE CRUDE 1 _. WATER URANIUM PRECIPITATION STRIPPING CONCENTR/VT E * LEAN SOLVENI" 1 ACIDIFICATION SOLVENr MS/PNC Drg.No. UP/R/05 - 627 - FIGURE.II SOLVENT EXTRACTION MATERIAL BALANCE WPA MGA 1 \ 60Uhr. 1-5L/hr. 0096g.U3Oa/l. 0-28g.U3 0a/l 100 % DISTRIBUTED SOLVENT. EXTRACT _L LEAN 241/hr. _EXTRACTION-i 1 2AUhr. STRIPPING-1 SOLVENT 0-01g/l O^gLhOg/l 24l./hr. OdgUjOg/l 97'/.DISTRIBUTION RAFFINATE 60 l./hr. OOOag.UsOg/I. 1:1-5 DILUTION H202 STORAGE 60 l./hr. WATER 15*/. NQ2C Soln. 1.6gU3 0s/I. 91./ hr. SOLVENT L EXTRACT LEAN 251/hr. 21 SCRUBBING STRIPPING O-Olg/l EXTRACT10N-2 3^5Tig/g l "SOLVENT 0-01 g/1 RAFFINATE WASTE URANIUM 60/hr. 001g.U3O6/l. SOLUTION O-UgUjOg/l. 7g/l (TO EXTRACTION-1) Drg.No.UP/R/06 - 628 - RECOVERY OP URANIUM FROI.* NITRO-PHOS ACID } R.A. Nagle, A.B. Giriyalkar & K.S. Koppiker Uranium Extraction Division, EARC, Bombay. Uranium recovery from v/et process phosphoric acid obtained via sulphuric acid route is by now a well accepted practice. Several alternate flow sheets essentially based on solvent extraction process are available for this purpose. Use of nitric acid to attack the rock phosphate is gaining popularity since the fertiliser obtained by this route will be a mixture of nitrogen and phosphorus. However, recovery of uranium from the resultant nitro-phosphoric acid ha3 not been attempted anywhere in the world. In India, at present Rashtriya Chemicals & Pertilisers at Bombay, produce bulk of the phosphatic fertiliser by this route. laboratory studies have been carried out on recovery of uranium from nitro-phos acid obtained from RCP Ltd. by solvent extraction technique using various extractant systems and conditions for extraction have b?>en optimised. Based on the results, a conceptual process flow sheet has been suggested. Introduction Fhosphoric acid in produced all over the world mostly by the wet process using 3ulphuric acid. Nitric acid being co3tly is not preferred though it ria:t the advantage that it doea not have the problem of disposing off radioactive gypsum. There is a discernible trond these days of utilising nitric acid au it directly gives a nixed fertiliser. Raahtrlya Chemicals and Pertilisers Ltd. (RCP) in Bombay treats about 180,000 T/Y of rock phosphate by the nitric acid route and 90,000T/Y by the sulphuric acid route. This rock phosphate - 629 - is imported ard contains on an average 120 ppm U,0., which works to about 32T U,0Q/year. Uranium recovery from phosphoric acid produced by sulphuric acid route is by now v.'ell established. Plans are now underway to establish India's first plant based on such a process at FACT, Cochin. But so far technology for recovering uranium from nitro-phos acid is not available. Possibility of extracting uranium from nitrophos acid using trifcutyl phosphate as an extractant has been mentioned in literature. If uranium i3 to be recovered from the rock phosphate processed at RCF, a suitable process to extract uranium from nitro-phos acid has to be established. Laboratory studies have now been carried out to develop a suitable process for this purpose. The data collected is presented and a conceptual process flow sheet has been proposed. Further teats on a continuous basis will confirm its acceptability. The sample of nitro-phos acid obtained from RCF wa3 dark in colour and was viscous. The chemical composition was T P2Cv:26$, KN0,:0.16LT. Fe :1.0g/l. Based on the literature survey, tributyl phosphate diluted with kerosene was selected for preliminary trials. The tests indicated that 40$ TBP solution would be optimum, extracting about 66$ of uranium in a single contact. However, one problem faced was the crysta llisation of calcium nitrate, perhaps due to the extraction of free HNO, by TBP. Prior separation of aclcium nitrate, reduced the extraction of uranium, drasiticRlly. The phase separation after extraction was alno 3low. Of course, once extracted - 630 - there wa3 no problem of stripping uranium with water. In view of the practical problem, tin alternate extraction system was felt to be necessary. Extraction of uranium from nitro-phos acid Some of the well known organo-phosphorus extractants were tested for this purpose. Octylphenyl and nonylphenyl e3ters of phosphoric acid were found to extract uranium better than TBP, but less effectively than DKilPA alone or a mixture of B^ITA + TOPO. DETTA alone gave a KD value of 8.1, (0.5M DSITA, 0/A=l). Further tests were carried out with DHHPA in kerosene. The phase separation was rapid, and 0.5M concentration of IHWPA was found to be sufficient. Rise in temperature from ambient to 50°C was found to reduce the K^ value from 11 to 8. Single contact even at a Org/Aq pha3e ratio of 5 showed further te3t3 about 69 per cent extraction for^a phase ratio of 3 was selected. Equilibrium isotherm and McCabe-Thiele plot were prepared and it wa3 found that it would be possible to extract *•-•> 99 per cent of uranium in 4 counter current 3tages. Batch counter current trials: These trials were carried out using separating funnels. For sake of convenience in analysis, the uranium content of ni tro-pho.3 acid was kept at 0.13g U,0g/l by addition of extra uranium. The raffinate at equilibrium was found to assay 4 ppm U,0fi, in 4 extraction stages. The extract was found to have a dark colour, thi3 colour could be due to the organic matter present in the acid. The - 631 - solvent was tested and found iot to decompose even after exposure to nitrophos acid for prolonged period. Stripping of loaded extract; The uranium present in extract could be stripped with 10M H,PO. or 5M H^PO. containing ferrous iron (—15g Fe +/l). Stripping isotherm and McCabe- Thiele plot indicated that almost quantitative stripping of uranium can be achieved in 4 counter current stages at a phase ratio 0rg/Aq=4. This phase ratio can be maintained even higher during plant operation, by recycling the aqueous phase from the settler to the mixing compartment. On this basis, *„: a single cycle of extraction - stripping, can yield a phosphoric acid assaying upto 4-g u",Og/l. This can be further processed by a second cycle of extraction to yield a strong solution of uranium. Tentative flow sheet Based on these trials, a tentative flow sheet has been outlined as shown in Pig. 1. This flow sheet needs to be tested to fine-tune the process parameters. Two options have' been proposed for uranium recovery from nitro-phos acid produced at ECP, to suit the acid plant flow-sheet. Removal of calcium nitrate crystals at 5°C and its effect of uranium extraction has also been studied, Surprisingly, after separation of calcium nitrate crystals, the k— value of DV.lfPA at 10°C was found to be reduced to 2.6, where as DEMPA-TOPO combination gave a vnlue a3 high as 12. - 632 - Conclusions; Baaed on these preliminary trials, it has been found that it is possible to extract uranium from nitro-phos acid using D^'.PA alone or in combination with TOPO. Ho serious problem in phase separation is anticipated. V/hen trials, on a larger scale using bench or pilot scale equipment are completed, all these process parameters will be optimised. Since the phosphoric acid plant at RCP consists of three separate streams, a plan for recovery of urar.ium from all these streams together, is also presented in Fig. 2. Acknowledgement; The authors wish to thank Shri S. Sen, Associate Director, Chem. Engg. ''roup, BARC for hio keen interest and permission to present this paper. 633 URANIUM RECOVERY K - > •PLANT OPTION , ^ I OPTION NITRIC 1 ACID r' .__4 . *._ *._?._ ~L ACID .NITRO-PHOS. ACIO 350 NITRC^-PHOS SUJRRY PRIMARY I CRYSTALLISER 3'C, Sp.Gr.-l-55, 3 3 TPD" A.N.P PLANT 70 TJ COOLING I 42M /Hr. 5*C 35M /lH r 6V.FREE HN03, FOR 3-6 M. H3PO4 ANP PREPN. -> CALCIUM NITRATE CRYSTALS FOR CONVERSION URANIUM NITRIC RECOVERY ACIO- mi 1 PLANT I I V r GROUND ACD ROCK ^ 200. NITRO-PHOS SLURRY A.N.R _^ FERTILIZER PHOS- TPO N.R PLANT 30-32M3/Hr. PREPARATION PREPARATION -PHATE URANIUM RECOVERY 1 r PLANT 1 SULPHURIC 1 I ACir i I I I 1 1 W P A 3O0. (2SV£S%. P203 ) CONCENTRATOR PHOSPHORIC; * FILTERS •> CLARIFIER TPD I2M3/Hr. (35*/; P205> ACD PLANT 1 50*C L-* GYPSUM TO WASTE FIO.l • CONCEPTUAL FOR FIRST CYCLE URAMUM RECOVERY PLANTS AT R.C.F.-CHEMBUR. - $34 URANIUM STRIP WTRO-PHOS H3P(^ACID PLANT RECOVERY 3 (AN PI T t CYCLE 10M /Hr. 40 U9/Hr. STRIP 3 32M /Hr. H3P04 STRIP URANIUM WTRO-PHOS URANIUM PHOSPHORIC RECOVERY PLANT ^ RECOVERY A n CYCLE ACID D CYCLE IM3/Hr. STRIPPING . (NP) I CYCLE (COMBINED) EXTRACTION 3 0-7SMryHZ T 20 M3/Hr- URANIUM PRODUCT tew'/Hr.2¥7Hr STRIP PHOSPHORIC URANIUM "3P04 .J ACID -U RECOVERY 3 PLANT I CYCLE 0-3 M /Hr. Fia.2. CONCEPTUAL PUN FOR URANIUM RECOVERY FROM PHOSPHORIC ACID AT RCE-CHEMBUR. - 655 - RECOVERY OP URANIUM PROM MONAZITE A FRESH LOOK AT THE CURRENT PRACTICE S.L. Mishra and K.S» Koppiker Uranium Extraction Division, BARC Monazite is being processed at- the Rare-Earth Division of IRE Ltd. at Alwaye for the past three decades. In this process, the bulk of the uranium is locked up in the final thorium hydroxide fraction which is stockpiled, and hence practically not recovered. In view of thp importance of recovering uranium, the current practice followed at IRE has been re-studied. Laboratory studios have shown that by slight modification of the process steps, it is possible to recover up to 85 per cent of uranium as a byproduct of the rare-earths recovery process at Alwaye itself. Recovery of uranium from the monazite hydroxide by nitric acid route has also been attempted and the data collected is presented. Process flow sheets are suggested. INTRODUCTION Monazite found in the beach sands of Kerala has been mentioned in all text books as a source for uranium and thorium the two metals of interest to atomic energy programme. And when the atomic energy commission was established in India, one of its first decisions was to yet up facilities to extract - 636 - uranium from this source. At that time, it was the only known source and hence in spite of the difficult technology, the plant was set up under the aegis of Indian Rare Earth3 Ltd. at Alwaye as early as 1952. The basic technology for processing monazite was obtained from Prance and- over the last three decades this plant has been operated more or less on the same lines, with some minor modifications here and there. A broad outline of thi3 process is described in a flowsheet given in Pig. 1. The steps involved are - 1. Grinding of monazite sand to fine size 2. Attack of thi3 phosphate mineral using caustic lye at about 165°C. This step helps in converting the phosphate into alkali phosphate which is water soluble 3« The fused mass is leached with/to solubilize the-phosphate which is filtered off 4. The hydroxide cake containing uranium, thorium and rare-earth is leached with hydrochloric aeid at controlled pH of 3.0 to preferentially dissolve the rare-earths and the chloride solution is finally evaporated and exported as a solid. Recently new facilities, have been created to separate these rare-earths into various fractions and individual.rare-earth concentrates. 5. The hydroxide, cake called "thorium cake" after rare-earth leaching, is mostly stockpiled since it contains thorium and uranium. URANIUM DISTRIBUTION In this process, uranium gets distributed into various products from where it ha3 to be recovered, by further processing. About half of uranium reports to the thorium cake. This material is further processed at the Thorium Plant located at BARC Trombay - 637 - for recovering thorium as pure thorium nitrate and uranium as crude uranium tetrafluoride. This uranium tetrafluoride is finally processed at the Uranium Ntetal Plant of BARC. The point to be noted is that only a small portion of the thorium hydroxide cake is processed for producing thorium nitrate which has only a limited market. Hence, most of the uranium is locked up in the thorium cake which is stockpiled. Now, proposal to process the stockpile thorium cake for uranium recovery is under study. Under the present context of uranium production technology, the contribution of uranium from monazite is not significant. However, it was felt that the present process needs to be improvised to recover most of the uranium at one particular stage of treatment. With this objective, the process has been re-investigated. ALTBRNATIVE-12 The monazite hydroxide cake is dissolved in nitric acid and filtered to recover the ur.attacked monazite. The nitrate solution containing all the uranium, thorium and rare-earths is processed through a solvent extraction cycle using tributyl phosphate as the extractant to obtain uranium as uranyl nitrate and thorium as thorium nitrate, leaving all the rare-earths a3 nitrate in the final raffinate. The acidity of the solution is kept' at 2.5M nitric acid and 7 per cent TBP solution in kerosene is utilised for extracting uranium quantitatively, in 4 counter-current stages. The extract has to be scrubbed to remove the co-extracted - 638 - thorium and rare-earths contamination. Finally uranium is stripped with water to yield uranyl nitrate. The uranium free thorium hearing solution is acidified to 3.5M nitric acid and using 40 per cent TBP solution in kerosene and a phase ratio Org/Aq. 1:1.2, ahout 99.7 per cent of thorium present could be extracted in 4 counter-current stages. During these test3 problems due to precipitation of titanium phosphate were encountered. This was overcome by addition of a metal like iron or aluminium to complex the phosphate ions which originally are bound by thorium but get released during transfer of thorium to TBP phase. The raffinate solution now containing low concentration of thorium and other metallic impurities and all the rare- earths can be processed for rare-earths recovery. One route would be to remove all the free nitric acid and precipitate the rare-earths as oxalate and process further. ALTERNATIVE •- 2 Another alternative studied is basically similar to the present process. The major cht-.nge being in drying of the monazite hydroxide cake to oxidise the cerium to cerium (IV) state . This helps in obtaining a ceriurc lean rare-earth chloride fraction which is in great demand. At the next step this cerium is once again brought back to Ce(iii) condition and is leached with hydrochloric acid. This cerium rich fraction would contain the bulk of the uranium present in monazite which can be recovered by mere change of pi'. This i3 a major attraction, uince It allows the recovery of 80-85 per cent of urunium at one utage. - 639 - This uranium fraction, containing small amount of thorium can be now further processed by the already established 5 TBP-K nitric acid solvent extraction route to yield high grade uranium and thorium nitrate. The broad outline of the process steps are shown in the schematic flow sheet shown in Fig. 2. '^his is a simple modification of the present process which helps in recovery of bulk of the uranium as a single product. Small amount of uranium which gets leached along with cerium .lean rare-earths chloride fraction is also recover? since product has to undergo a deactivation step as usual. This process has been further tested at IRE, Alwaye and the indications are that the future plant to set up by them would be based on this process flow sheet. REFERENCES . 1. Treatment of monasite sands with special reference to Indian practice. H..N. Sethana and S. Fareeduddin Symposium on Rare-Metals. 1957 page 68 The Indian Institute of Metals. 2. Recovery of uranium, thorium and rare-eartha from mixed hydroxide obtained from monazite by nitric acid route - Jar S.L. Mishra, K.S. Koppiker and T.K.3. Murthy Report No. UED/86/2, BARC. 3« Recovery of uranium, thorium and rare-earths from mixed hydroxide from monazite Part II. S.L. Mishra, K.S. Koppikert and T.K.3. Murthy. Report No. USD/87/1 BARC. - 640 - 4. Production of lanthanon and thorium compounds from monazite (II) B.S. Pilkington and A.W. V/ylie J. appl. Chem., £, 265 (1947). 5. Thorium purification by solvent extraction V.V. Gupta, V.S. Keni and 5.K. Bhosh Proc. Symp. on solvent extraction of metals Feb. 1979, DAE Bombay. 641 - GROUND MOiMAZITE METATHESIS ALKALI - -> at~-l50°C SLURRY ^_ FILTRATION vTRlSODIUM and "^PHOSPHATE WASHING CAKE J&_ LEACHING HCI- -> "at pH3 SLURRY ^k FILTRATION ^RARE-EARTH and "^CHLORIDE WASHING UNLEACHED THORIUM CAKE STOCK MAJOR PORTION "> PILED (-1 V.U308,; \/_ THORIUM and URANIUM RECOVERY URANIUM ^ THORIUM TETRAFLUORIDE*^ NITRATE FIG. I. FLOWSHEET FOR TREATMENT OF MONAZITE. (CURRENT PRACTICE) - 642 - WET MONAZITE HYDROXIDE CAKE V. DRYING I60*C DRIED CAKE HCI LEACHING (I'D pH30,60'C SLURRY iH FILTRATION FILTRATE CERIUM-LESS and . "> RARE-EARTH WASHING CHLORIDE U308<-OIV. UNLEACHED CAKE (U3O8/T.O.) N02SO3- LEACHING pH l-5±0-2 HCI- 90 *C (I'l) SLURRY LOW URANIUM FILTRATION FILTER (0-36% U3Oe/T.O.) and -> THORIUM RICH WASHING CAKE HYDROXIDE CAKE TO STOCK PILE CERIUM RICH FILTRATE SOLUTION (C«,U,Th,RE) URANIUM ALKALI " PRECIPITATION pH 60 SLURRY URANIUM CAKE JL 6-IOV.U308 U and Th FILTRATION RECOVERY by 80-85 V. RECOVERY SX PROCESS CERIUM SOLUTION FOR RECOVERY \if 12 MT UjjOe/YEAR BASED ON 4500 MT MONAZITE/YEAR FIG 2. RECOVERY OF URANIUM FROM MONAZITE BY MODIFIED PROCESS. - 643 - RECOVERY OP URAKIUM PROM BEA WATER - A LABORATORY STUDY D.V. Jayawant, U.S. Iyer & K.S. Koppiker Uranium Extraction Division, BARG Sea water contains traces of uranium, but the volume of aea water being enormous the total quantity of uranium available from this source is very large. Prom time to time claims have been made elsewhere that a breakthrough has been made in developing a technology to recover this uranium at an economic cost. Studies have been carried out at Uranium Extraction Division over a few years to develop a suitable technique to separate the uranium from sea water. Studies were primarily directed towards preparation of suitable inorganic ion exchangers and studying their properties. In this paper preparation of ion exchangers based on hydrous titanium oxide and the data collected in laboratory trials on their application for uranium adsorption from sea water are presented. Introduction The ocean is known to be a vast reserve of uranium readily available in liquid form. The concentration 1 '2 '^3 of uranium is reasonably constant at 3«3 ug/l amounting to a total reserve of 4 x 10 T. The. chemical state of uranium in sea water is reported to be 97.9# as U02(C0j)j , 1.7# as UOgCOH)^ and 0 ^ aB u02(C0»)2~ . Several chemical processes like solvent extraction, ion flotation, coprecipitation to etc have been studied for recovery of uranium from sea water. - 644 - However, all these routes involve transport of vast quantities of sea water and also have adverse environmental effects and hence are not attractive. Only technique which can meet all the requirements is ion echange, using a stable exchanger. Fortunately some of the inorganic exchangers have been found to be ideally suited for this purpose. Throughout the world, work' carried out on uranium recovery from aea water, has been based mostly on the use of hydrated titanium dioxide (HTO). i'his adsorber was first tested in U.K. as early as 1963» However, the major problem is to provide sufficient mechanical strength to HTO, to enable it to withstand the movement of sea water. Hence, when work was initialed in our laboratory the objective was to develop a titanium based adsorber having all the desired properties. This work ii no longer pursued since even a pilot scale - operation for testing, requires large input by way of manpower and investment. Choice of adsorbent As has been mentioned, HTO has been found to be ideally suited for uptake of uranium from sea water*"'. HTO exhibits both cation and anion exchange properties depending on the 8 pH of the solution . At pH 8 it behaves as cation exchanger. Uranium ehich is in the form of carbonate complex at pH 8 in sea water, gets ionised due to acidic surface of HTO forming 2+ •» 2 VO,, and HBO, species , and uranium gets adsorbed-. The main problem with HTO is that on long standing in sea water it crumbles and peptises, and also solubilises in strong solution of ammonium carbonate or dilute sulphuric acid generally used for uranium elution. In the present work, we have tried to prepare four different types of titanium adsorbants for uranium adsorption. These are - 645 - 1) Benefeciated ilmenite 2) titanium impregnated charcoal granules 3) Composite granules consisting of HTO and activated carbon powder 4) titanium silicate granules. Adsorption tests The uranium content in sea water is very low and hence to reduce, the experimental time and for ease of chemical analysis, the sea water was spiked with uranium to the desired level in the laboratory before use. The prepared adsorbents were tested by batch equilibration as well as by columnar operation method. In all these tests, the amount of uranium adsorbed, titanium lost etc. were determined. Benefeciated Ilmenite Ilmenite is found in large quantities on the sea coast of India. It has a composition of 65$ TlOg and 35$ PeO. This is benefeciated by elimination of most of the iron by acid leaohing after converting FeO to metallic iron by carbon reduotion. The leaching of iron leaves small pores in the mineral matrix, thereby converting it into porous, hard titanium compound with increased surface area. This material after treatment with ammonia and subsequent washing was tested for uranium uptake from sea water containing 5mg uV;-/l, by columnar operation. A number of trials under various conditions gave an average loading of 700ug tM)Q per gram adsorbent. The material was quite stable over a number of cycles of adsorption/elution. But it had a major - 646 - drawback of the material being too fine. This effected the flow rate of sea water passing through the column. Efforts to overcome this defect by upflow method led to loss of fines with the raffinate which is not acceptable. Titanium impregnated charcoal granules Activated carbon is known to have very large surface area. The metal hydroxide impregnated in pores of activated Q charcoal has been reported to show better adsorption behaviour compared to active carbon or metal hydroxide alone. Granular activated carbon of mesh size -20 + 32 was soaked in titanium sulphate solution and after draining the solution, the granules were exposed to ammonia fumes to hydrolyse the titanium deposited on the surface, and the granules were finally dried at 120°C for 12 hours. The granules thus prepaxed were tried as uranium adsorbent by column operation, and the adsorbed uranium was eluted with various eluting agents like 0.1 and 0.5M HgSO., 1M (NH^JgCOj, 1M NagCO,, 1.0M NaHCO,, 1I.I(NH+)2C05+1M NaCl. The results obtained are given in Table I. It was found that elution with acid led to severe loss (80$) of titanium during single cycle, whereas with carbonate solution the loss was about 40$. Thus, it was concluded that titanium deposition on carbon granules will not give a stable adsorbent which can withstand chemical action of eluting agent. Composite adsorbent with activated oarbon powder and titanium hydroxide Activated carbon powder which has a large surface area was mixed with HTO powder and granules were prepared with this mixture. Different modes of combining these two constituents - 647 - were tried. Binding agents like ethylpolysilicate as well as combination of sodium silicate and aluminium nitrate were tried. All the samples of adsorbents were studied by passing sea water containing 50ug U,Og/l. The results obtained are shown in Table 2. Out of these samples batch P adsorbent was found to be most Btable. The uranium uptake on the basis of TiOg was also fairly good. Different batches of type P adsorbent varying the ratio of titanium - carbon, silicate- aluminium nitrate were also tested. It was observed that 5g of composite adsorbent (50$ carbon + 50$ HTO) mixed with I5g of sodium silicate and 2ml of saturated aluminium nitrate was optimum ratio for getting suitable hard coarse granules. Number of batches of this composition were prepared to check reproducibility. Adsorption tests have shown that the average loading for this material was 520 ug U,0Q/g adsorbent. Elution with ammonium carbonate and sulphuric acid and dissolution of titanium was found to be minimum. The silica content of these adsorbents was 50$ and titanium was only 5#« Considering the cost of manufacturing this type of adsorbent, it may not be practical to use it for this purpose. Titanium silicate adsorbents Titanium silicate crystal structure is not well defined and it exists as a mixture of TiO, and' Si09 crystals. Baxi and 10 Desai have studied coprecipitated TiOg and SiOg gel for its ion exchange behaviour and claimed that samples with TitSi in the ratio of 1:1 had the highest cation exchange capacity. This material was tested for uranium adsorption from sea water. This sample wa3 supplied by the Central Salt and Marine Chemica Research Institute, Bhavnagar, and had a chemical composition 41$ Ti02 and 40$ SiOp and lose on ignition 17.4$. r 648 - Batch equilibration tests were done on 1g of silicate ( and 1g of HTO for comparison) using sea water assaying 2mg U,0Q/1. Water samples were analysed at periodic intervals of time. In the case of Ti-Silicate equilibrium was reached in 5 min. and for HTO it was 50$ in 5 min and 90$ loading in 30 min. In column operation, with I.Og Ti-Silicate, and 50ug U,Oa/l sea water, 24L of sea water was passed at 2 min contact time. The first 12L of effluent did not show presence of any uranium indicating complete adsorption. The loaded adsorbent was eluted with various eluting agents like 0.01 to 0.1M HgS04 and 0.05 to 1.0M (HH^JgCO-. With acid, elution was fa3t - almost 92$ of uranium gets eluted within first 15 min. whereas it was only 50$ with (NH.)2C0,. Dissolution of titanium in eluting agent was also found to be negligible. The saturation uptake capacity of uranium from sea water containing 3.3ug U^Og/l was found to be 30-40ug U~0g/g adsorbent for the silicate compound as compared to 12ug U^Og/g for HTO. Repeated cycles of adsorption/elution with Ti-silicate did not show any indication of reduction in adsorption capaoity for uranium from sea water. The series of tests have shown that titanium silicate could be more suitable for any large scale experiments on adsorption of uranium from sea water. - 649 - References 1. N. Gogata, N. Inone, Nippon Genshiryoku Gokkaishe Jl, (107), 560, (1971). 2. H. Yamashita etc. Bull. Chem. Soc. Japan, 52.(1), (1980). 3. J.D. Wilson etc. Anal. Chim. Acta 23. , 505 (1960). 4. N. Ogata, H. Kakihana, J. Atomic Energy Soc. Japan H, 82 (1969). 5. N. J. Keen, J. Brist Nuel. Soc. 7 (2), 178 (1968) 6. M. Kanno etc. J. Atomic Energy Soc. Japan, 12, 708, (1970). 7. R.V. Davis and J. Kennedy, Nature, 203_, 1110 (1964). 8. V. Vesely and V. Pekarek, Talanta 19_, 219 (1972). 9. A. Ninomiya etc. Kogyo Kagaku Zasshi Jit 1486 (1971). 10. D.R. Baxi and G.T. Desai Ind. J. Tech. 16, 204 (1978). 11. K. Shoji and T. Keno, Nippon Kagaku Kaishi 9f 1298 (1975). - 650 - Table 1 Adsorption Properties of Titanium Impregnated Granular Charcoal Batch TiOo Concentration TiOg concentration No. of soaked solution . in charcoal TJ,0Q adsorbed 2_o i g/litre mg/g charcoal ug/g charcoal 1 10 4 158 2 15 6.7 226 3 25 6.0 232.8 50 , 6.0 194 - 651 - Table 2 Adsorption of uranium by HTO mixed with activated carbon Ti02 Uptake of Uptake of Loss of content uranium uranium TiOj, (mg/g of t^OgUg/g ug U,08/g during Preparation method sample) aasorbant elution A. HTO dried at 110°C - Mixed with equal weight of carbon, granula 175 100 600 0.66* tion using etbylpolysilieatebinder B. Similar to (1) above but using mixture of sodium silicate and satu rated aluminium nitrate as binder, 40 32 800 12* finally dried at 120°C for 24 hrs. washed, & crushed to size -20 +32 mesh size C. Granules of active carbon prepared sodium silicate ft aluminium nitrate as binder A soaked in Ti solution 12 22 1800 13.2* for 24 hrs. drained, exposed to ammonia fumes, washed and dried -20 +32 mesh siae D. Same as (3) but using ethyl 20 35.4 1700 19* polysilicate ad binder 8. Precipitation of HTO along with activated carbon powder, filtered and air dried and pelletised in 266 1330 4.2# hydraulic press using ethyi polysilicate binder. The pellet dried, and crushed to get granules -20 +32 mesh size f. Same as. (5) but using sodium silicate - aluminium nitrate 105 1121 4.6£ as binder SESSION' VI URANIUM REPINING Chairman : Shrl S.K. CUATTERJ5E KIC Reporteur: Shri D.K. B03E B A R C - 653 - OPERATING EXPERIENCE IN THE REFINING OF URANIUM BY SOLVENT EXTRACTION USING MIXER-SETTLER S.B. Roy, H. Singh, K. Kumar, A.M. Meghal, V.N. Krishnan and K.S. Koppiker Uranium Metal Plant, B.A.R.C., Bombay Uranium purification at Uranium Metal Plant is carried out using solvent extaction process. Equipment used is mixer-settler. The process includes extraction, scrubbing, stripping and solvent processing. Operating experience in these stages is discussed. Effect of silica in feed, fluoride level, solvent quality and extrainment levels are discussed. Experience has shown stripping capacity is limited by low specific settling rate of used solvent with dimineralised water at ambient temperature. Efforts have been made to study these factors in detail. A new stripping and solvent processing unit has been designed. A slurry extractor is being fabricated for over-coming silica problem. Improvements in instrumentation and control are discussed. I. INTRODUCTION Uranium Motal Plant is using solvent extraction process for the purification of crude uranyl nitrate solution as an intermediate step in production of nuclear grade uranium ingots. Magnesium diuranate II. SOLVENT EXTRACTION PROCESS The process flow-sheet is shown in Figure I. The solvent used is tri-n-butyl phosphate (TBP) diluted to 30X v/v in kerosene. The slurry after dissolution is filtered and analysed. Feed solution to extraction contains about 225 g/1 of uranium and has free acidity (FA) of 2 N. The uranium is selectively extracted into solvent phase in 7 stage mixer-settler battery. The organic to aqueous <0/A) ratio is maintained at 2:1 and 80-90VC saturation of solvent is achieved. The loaded solvent containing traces of aqueous entrainment and co-extracted impurities is scrubbed with 0.5 N nitric acid at 0/A rat io of 10/1. The pure ex tract i s - 654 - stripped with dimi neral i sed water (DMW) to tansfer uranium, to the aqueous phase. The phase ratio is controlled at 0/A = 2/3. The lean solvent, -free of uranium, however, contains hydrolysis products o-f TBP and degradation products o-f kerosene. These are removed by washing with soda ash solution. The solvent is re-acidified and recycled. The equipment used -for solvent extraction was initially pulsed columns. During expansion of UMP a decision was taken to switch-over to mixer-settlers for simplicity, design reliability and flexibility in operation. The mixer-settlers used are similar to the units in use at 20P, NFC. III. COMM1SSI ONING EXPERIEMCE During commissioning it was found that the aqueous level in settler of stripping unit used to build-up and flood the unit. As a remedial measure, the aqueous baffle height in stripping unit was reduced from 10 cm.to 2 cm. In addition, inter-phase controllers were redesigned. However, the flooding of aqueous phase still persisted. The design was re-examined and model simulations carried out at different available density data. It was found that it is difficult to maintain adequate inter-phase levels without increasing depth of the unit to 1 metre. The specific settling rate with the used solvent was found to be 2 - 2.5 m3/hr/m2 as against design value of 4 m3/hr/m2. As a compromise for efficient operation, it was decided to derate the whole unit. IU. OPERATING EXPERIENCE 4.1 Silica Accumulation : The feed to solvent extraction process obtained after filteration of MDU slurry generally contains about 0.5 gp1 Si02. This silica slowly accumulates in the extraction battery and it becomes necessary to clean the unit once in a month. This involves draining the entire hold-up and filteration. Spillage and solvent losses, besides demand on man-power, are high. Often, due to mal-functioning in filteration, the silica level goes very high in the feed and immediately the ports get choked and flooding occurs. Several flocculants and filter-aids were tested in lab.scale but without much improvement. The type and quality of filter cloth used was found to have strong effect on the clarity of feed solution. Rotary pre-coat vacuum filteration for silica control was not attempted due to limitations of space and the problem of processing pre-coat cake. A decision is now taken to replace the mixer-settler battery for extraction by a slurry extractor to overcome the silica problem. - 655 - 4.2 . Settling Rate in Stripping ; The settling rate in stripping unit was -found to be the limiting -factor in capacity. At high -flows, coalescence was incomplete, back-mixing occurred across stages, and concentration gradients were poor. Investigations were carried out to improve the rate by increasing temperature and aqueous phase acidity. Batch coalescence times were measured. It was •found at 5Q°C, the coalescence time was less than hal-f its value at 30° C. Similarly use o-f 0.5 M nitric acid instead o-f DM water -for stripping, reduced the coalescence, time without increasing uranium concentration in lean solvent. Based on these results, stripping is routinely carried out now with 0.05 N acid. 4.3 Sol vent Quali ty : Freshly prepared solvent gives high phase separation rate. As the solvent is recycled, the solvent qual i ty degrades and phase separation time increases. This is independent o-f DBP and U concentration. For example a fresh solvent yields complete batch coalescence in 1 min, solvent used -for about a year requires 5 min -for separation and a 'rejected' lot (used over 2 years) does not yield separation even in 15 minutes. Hence a regular check on solvent quality is kept by an emperical batch test. 4.4 Solvent Entrainment Loss ; Solvent is lost by entrainment into raffinate stream, alkali wash stream and pure solution stream -from stripper. The physical solubility loss is comparatively lower. Studies have shown that alkali wash has maximum entrainment <9 - 15 ml /\ > while raffinate loss varies 0.7 to 2 ml/1. The entrainment wM th pure solution is 0.5 to 3 ml/1,, most o-f wh i ch is recovered by providing intermediate hold-up tank -for coalescence. 4.5 'Product Pur i ty : The overall per-formance has been very good, with less than 5'/. o-f total processed batches having high impurity levi-1. The higher level is found to correlate with high impurity in -feed itself resulting from rGcyc\& of raffinate cakes to dissolution. However, even for impure batches, the U03 obtained was pure when EDTA was added during precipitation. No product solution was required to be recycled to starting stage, as was the case with pulse columns. In fact, an evaporator obtained to facilitate this recycle has been rendered redundant. 4.6 Fluoride Level : The fluoride content of the feed in some batches has been found to be high (over 1 gm/1 ). This creates severe corrosion problem in extraction battery where some agitators have shown stress-corrosion cracking at weld joints of impeller to shaft. To avoid this problem, the fluoride bearing solutions are processed separately in polypropylene mixer-settler. 4.7 Extraction from Fluoride Solutions : When magnesium fluoride is processed by acid leaching solution containing 10 - 15 9 U/l, 1 - 3 g F/l and 7 gm tig/1 is obtained. This is highly corrosive. Recovery of uranium from this solution is best achieved by solvent extraction. A 5 stage mixer-settler- battery operating at overall phase ratio of - 656 - 0/A = 1/4 has been in operation -for last 3 years. Internal recycle -from settler to mixer is provided to maintain mixer ratio of 0/A= 1.5/1 -for efficient operation. The unit is made of polypropylene and transfer pumps of magnetic drive type are used. The extract containing 40-60 g U/l is -fed to scrubbing battery o-f main unit. To ensure raffinate level of 0.5 gpl of U, it is'necessary to restrict loading to 65-70 gpl, maintain FA of 2-2.5 N, and ensure that lean solvent has less than 2.5 g U/l. 4.8 Knit-mesh packino ; A dual material woven stainless steel and polypropylene knitted mesh, packing has been installed in stripper. This has reduced emulsion band width by a factor of 2 and reduced entrainment losses by a factor of 5. V. DEVELOPMENT AND DESIGN STUDIES 6.1 New Strippino Unit ; Both batch and continuous single stage testing with used solvent was carried out under various conditions. The data is used for design of high capacity stripping unit. The settler flux of 2.5 m3/hr/m2, top shrouded six-blade turbine mixer, hydraulical 1 y independent stages with internal recirculation are some of the salient features. This unit is under fabrication at Central Workshop of BARC. 5.2 New Solvent Processing Unit ; Matching the capacity of new stripping unit, a high capacity alkali and acid washing mixer-settler battery has been designed and is under fabrication at Central Workshop, BARC. In addition laboratory scale test-work has shown that distillation of solvent in 40°C - 130°C can improve; sol vent quality with" reference to phase separation. The. residue obtained (5/. by volume) is thick viscous black liquid which causes emulsion formation. 5.3 Slurry Extractor : Mixer-settler battery for extraction is being replaced by a 'slurry extractor' developed in Nuclear Fuel Complex (4). The design has been made compact and modular for meeting site-specific constraints. The design of slurry extractor requires close control over phase- ratios, flow-rates and inter-phase level. An automatic control system isunder development jointly with Reactor Control Division of B.A.R.C. - 657 - REFERENCES 1. S. Sen and T.K.S. Murthy 'Ore Processing and Re-fining o-f Uranium in India' in IAEA Symposium on 'Production o-f Yellow Cake and Uranium Fluorides', 1980 2. R.K. Garg, N.Swaminathan and Mohindar Singh "Production o-f nuclear grade zirconia by the solvent extraction process" DAE Symposium.on Solvent Extraction of Metals, Trombay,1979 3. W. Davis, Radiolytic Behaviour of YBP in Science and Technology of TBP, Vol. 1, U.W. Schultz J.D. Navratil and A.F. Ta1bat FIGURE ' = SCHEMATIC PROCESS FLOW SHEET FOR EXTRACTION EXTRACTION EXTRACT^ SCRUBBING PUREEXT.(EP) J STEPPING MAKE l? MIXER SETTLER MIXER-SETTLER MIXER-SETTLER SOLVENT 7 STAGES K i LEAN SOLVENT AQUEOUS . 0-5N 005 N URANYL ,N RAFFINATE FEED HNO3 NITRATE HNO3 TO PURE DM. WATER NEUTRALISATION 0-5 N SOLUTION HN03 UNPS PROCESSING ENTRAPMENT MIXER-SETTLER SEPARATOR U STAGE * ACID f DILUTE UNPS WASH ALKALI ALKALI TO WASH STORAGE NOTE:- FLOW MONITORING BY ROTAMETERS,FLOW CONTROL BY METERING PUMPS. Drg.No.-UP/R/03 MSIPHC - 659 - "CALMIX" - INNOVATIVE MIXER-SETTLER SYSTEM C.K.R.Kaimal, B.V.Shah, I.A.Siddiqui and S.V.Kumar Process Engineering and Systems Division Bhabha Atomic Research Centre, Bombay An improved mixer-settler system using a new device called "CALMIX" (Combined Air Lift and Mixing device) has been developed for meeting the requirements of solvent extraction operations in the processing of Uranium and Thorium. The "CALMIX" device consists of two closely-spaced airlifts which raise the liquid phases into a chamber where intimate mixing of the phases takes place. A six-stage pilot plant scale model of a "CALMIX"-based mixer-settler unit was set up to demonstrate its use in extraction and stripping of uranium. A typical "CALMIX" device and mixer-settler unit are described. The "CALMIX" device has been found to be 3table, reliable and versatile for efficient operation over a wide range of feed and throughput conditions. The system is versatile and possesses inherent compactness, thus offering considerable advantage in various applications. "CALMIX" mixer-settler units for organic-treatment application in Plutonium Plant, Trombay are under fabrication at present. Another series of runs has been made for Uranium-Thorium extraction and stripping in 5% TBP-nitric acid system. Highly encouraging results have been obtained from these tests. The unit operation forms the core of the Purex cycle. Pulse columns have been widely used for solvent extraction in nuclear fuel reprocessing. - 660 - However, 3ome of their disadvantages provide cause for the development of devices offering better efficiency, wider stable operating range and economy of shielded space. Extractive mass transfer in mixer-settler is based upon stage-wise contact and offers the advantages of stable operation with high efficiency requiring little headroom. Mechanicallly agitated mixer-settlers have settling chambers of large size and need maintenance of moving parts (in shielded areas with radiation fields in case of radiochemical plants). Air-pulsed mixer- settler systems are vulnerable to failure of the solenoid valve required for creating vacuum. They also need a hydraulic gradient between, stages and airlifts for inter-stage transfers. The "CALMIX" Device Work was initiated towards development of an improved mass transfer device to meet the requirements of "radiochemical applications. It was decided to use the airlift device in an innovative manner as a mixing device for a mixer-settler system since the airlift is an accepted system for liquid transfer, especially in reprocessing plant.- This work culminated in the development of the mixing device which has been named "CALMIX" (device for Combined Air Lifting and MIXing). Considerable effort has gone in making the device stable, reliable and versatile over a wide range of feed and throughput conditions. Principle of the CALMIX The CALMIX consists of two closely . spaced identical airlifts - one each for organic and aqueous phase. The CALMIX device is schematically illustrated in Fig. 1. - 661 - Both airlifts raise liquid into a common chamber located at the top of airlifts where intimate mixing of the phases takes place in the confined space. The mixing is intensifed by the additive effects of air, change in direction of flow and of discharging the mixed phase through a restricted opening. The CALMIX Mixer-Settler System A typical six-3tage CALMIX mixer-settler unit for a combined organic and aqueous throughput of 360 Iph has an overall size of 950 x 320 x 400 mm (LxWxH). The individual stages are 300 x 150 x 400 mm(LxWxH), the length of -the mixer and -the first and second set-tier in each stage being 100, 80 and 120 mm respectively with 300 mm active height. Off-gas vent, sampling, interface and level measuring instrument probes are provided. Aqueous outlet is controlled by jackleg or by regulating the air supply of airlift using a diaphragm-controlled valve (DCV). The organic outlet is by overflow. The airlifts are specially designed with a very low elevation of the foot-piece without backflow of air. This design ensures uniform distribution of air in -the airlift pipe. During pilot plant testing, the CALMIX'system has been studied in detail with reference to flowrates of air, organic and aqueous, the phase flow ratio, the total throughput and the specific gravity difference between the phases. The system has been continuously re designed, and modified, in order to improve capacity, compactness and uniformity of air flow and improved internals geometry. x Since the internal hydraulics of K C&LMIX mixer-settler differ from those of the puui'-mi;* type equipment, particular effort has been directive to the design of internals for optimising the hydraulics, - 662 - Features have been included -to reduce -the possibility of choking of passages. In general, it has been attempted to design an efficient unit with reliable performance over a wide range of operating conditions. Logs of Diluent Accelerated loss of diluent from the organic phase can be expected due to the airlift. The overall loss of solvent due to evaporation anc. entralnment has been found to be 0.021 ml of solvent pt-.r lit of air passed which is quite acceptable by operating plant considerations. Standardisation The CALMIX device and the CALMIX mixer-settler system have undergone extensive pilot plant testing. Based on thi3 experience, it has been possible to prepare designs of CALMIX devices of standardised dimensions and internals for applications with capacities varying from 60 lph {CALMIX Model CAL20) to 1600 lph for (Model CAL85). Customising to individual process variations can also be done without great difficulty. A typical CALMIX device Model CAL35 is shown in Fig. 2. Applications Plant scale CALMIX mixer-settler units of this type have been designed for inter-cycle alkaline and acid treatment of organic in Plutonium Plant, Trombay (Kef. 1) and purification of tail-end uranium (3D Cycle) in Prefre Plant, Tarapur (Ref. 2). These units have been fabricated and they await commissioning. - 663 - Studies relating to design of CALMIX mixer-settler systems for applications relating to separation of Uranium from irradiated Thorium (Ref. 3) and for Uranium Metal Plant at Trombay are in progress. CALMIX mixer-settler systems are practical and efficient mass transfer equipment of compact size. They are characterised by stability of operation over a wide range of feed conditions and throughput. The mass transfer performance is excellent under varied conditions of operation. Several mixer-settler3 can be connected in series in a continuous process to minimise tankage and pumping requirements. The units are also amenable to in-line instrumentation and automatic control The inherent simplicity of design and fabrication of the CALMIX mixer-settler system, and their good performance under varying conditions make them suitable for a variety of plant applications. Further work is under way to identify other parameters affecting design of the system and prediction of performance. in order to simplify design for specific applications. References 1. "Design of 'CALMIX' Mixer-Settler Unit for Carbonate Treatment of Organic in Plutonium Plant, Trombay - A Summary", - (Document EDF - 88/09 - Internal report) 2. "Design of "CALMIX" Mixer-Settlers for Second Uranium Purification Cycle in PREFRE Plant, Tarapur - A Summary" - (Document EDF - 89/11 - Internal report) 3. "Proposal for Use of "CALMIX" Mixer-Settler System for Thorium Reprocessing" (Document EDF - 88/16 - Internal report) - 664 - 1 IMetAIR S 1 CO Q Q I ( ])4-|-MIXED PHASE DISCHARGE CO OQO MIXING ZONE ORGANIC ORGANIC AIRLIFT AQUEOUS AIRUPT -O u AQUEOUS FlQ.li SCHEMATIC DIAGRAM DF CALMIX - 665 - 'A* NB sen. 40 air line 'C nn dleu anojed holes on 'E' iw trl prtch on -three sides only CALM OCJOQCOC C:-T] Dranaon :r <'<3 50Ca> A 13 B sot m: C 7 D 35 E 1£ F 15 G it H 3+3 I £44 J e K 4 L » N 7« N te» a 157 p •7 0 100 'J'no. of 'K' nn dia. gQuldlj-tnn-t holes Htwd i-nrf ng o,lf ling 23 TH l—JP 11—jA—11—IPS ELEVATION J u. il '. r® ifo LL .„ _.1L. _ .J H i M. — PLAN N Fto. & CALMIX DEVICE HDI3EL CAL35 - 666 - PRECIPITATION OP AMMONIUM DIURANATE - 1 STOUT T.S.Erisbnamoorthy, N.Mahadevan & M.Sankar Das* Analytical Chemistry Division, Bhabha Atonio Hesearoh Centre Trombay, Bombay 400 083 The precipitation of Ammonium diuranate (ADIT) forms the first step in the production of T7CU fuel for our reactors, and hence the quality and consistency of the ADU precipitate is very important in industrial operations. An investigation, on the preoipitation of ADU, was carried out under conditions similar to those In Indus— trial production, to.evaluate the effeot of various variables on the consistency and the quality of AD7. The variables studied were concentration of uranium & ammonia, -pH,temperature and form of ammonia (gas or solution). The properties studied were the settl ing rate of the precipitates, surfaoe area of the ADUs and calcined axide&and compositional characteristics of the ADUs. Multifactorial experiments and ruggedness tests were applied to identify the para meters to vbioh the preoipitation process is most vulnerable, so that such parameters may be controlled effectively. Besides, the effect and the importance of equilibrium conditions.during the preoipitation process, on the quality of the final ADU, was also established* The paper presents the results of these studies and arrives at the best preoipitation conditions. * Since retired. - 667 - ' INTRODUCTION In the production of UOg fuel pellets, the precipitation of ammonium diuranate (ABU) forms the first step* This is followed by drying and calcination of the ABU, and subsequent hydrogen reduc tion to U0_ followed by stabilisation, palletising and sintering. There are several variables in the ADO precipitation, which affect the oonBistency of the preoipitate whioh, in. turn, may possibly be oarried forward, contributing to the final inconsist ency in the finished pellets* Investigations were carried out, on a laboratory scale, to generate information regarding the ABU preoipitation to obtain 'consistent' ADU. The parameters studied were the concentrations of uranium and ammonia, pH, temperature of precipitation, form of ammonia (solution or gas) as well as the effect of washing the preoipitates. The effeot of preoipitating under equilibrium and nonequilibrium conditions were also investigated. The properties measured were the settling rates of ABU precipitates, surface area of the ABUs and the calcined oxides and composition of the ABOs. Precipitation studies were made in a batchwise manner under conditions whioh are close to those^operating plants. These investigations were aimed at identifying the parameter/ parameters to whioh the measured properties are most vulnerable so that they oan be controlled effectively. This paper describes the experimental approach, the results and the inferences that can be drawn. EXPIH DENTAL Reagents and apparatust Stock solution of uranyl nitrate (about 100 g of uranium/l). Ammonia* Serabhai M Chemicals GR (density - 0.91, 25# W/fy) was freshly diluted, as and when required and standardised. All the other reagents are of analytically pure grade. 1. pH matert Elioo digital pH meter, model L1-120 along with Elico glaBB and oalomel eleotrodes. 2. Magnetic stirrers Cenoo make, with adjustable control. 3. Spectrophotometer, Hitaohi - Model 33O. - 668 - > 4* Surface Area Analyser, fabricated in Analytical Chemistry Division, BARC, based on the physioal adsorption of nitrogen gas, at liquid nitrogen temperature, In a dynamic mode from a continu ously flowing mixture of nitrogen and hydrogen; uses a thermal conductivity deteotor for measuring the adsorption and desorption along with a built-in electronic Integrator. Analytical Methods: 1« All the ADU preoipitates were dried In an air oven at 105°C for 4 hours (unless otherwise specified) and preserved in a desiooator* 2. Uranium in the ADUs were estimated by the peroxide oolo- rimetrio method at 420 run. 3. Ammonia was determined by steam distillation, after addition of sodium hydroxide, collecting In hydrochloric acid of known strength and back titrating with standard alkali (Kjeldahl's method). 4* Nitrate in the samples was determined by reduction with Devarda's alloy, followed by steam distillation of ammonia as in 3 above. 5. The surface area of the ADUs as well as the oxides obta ined after caloination were measured In the surface area analyser. RESULTS AND DISCUSSION 1. Miltifactorial experiments and Ruggedness TestBt A ruggedness test, similar to the one suggested by Youden 2 and used by Sarkar and. Sankar Das was adopted in the initial seta of experiments on the study of the precipitation conditions of uranium. The idea behind the ruggedness teat is to make a deli berate, simultaneous change in the speoified conditions and to as certain whether the method (process, here) is sensitive to such ohanges and if sensitive, to identify the one to which the method is most vulnerable. Muoh labour can be saved by using an 'efficient plan1 to single out those factors (variables) affecting the results. In one suoh multifaotaral 'package', seven variables can be investi- 2 gated with just eight experiments • In another, three variables with just four experiments can be u*od, as was done In the following - 669 - sets of experiments. The efficiency oomes from ttao faot that eaoh of the four experiments is used three times* Out of the three identified variables, the effeot of one of the variables at two 'chosen levels' is examined. The do sign of the experiment is such that the two levels of the other two variables are ohanged in suoh a way that their effeot is nullified in the pairs chosen. The use of averages increases the chanoe of deteoting any effeot on the result as a oonsequencfc. of the ohange In the specified condition, 1,1. Table I gives the details of the four experiments Involv ing three conditions (concentration of uranium: 100 or 60 g/L} concentration of ammonia! 10 or SNjiand pH t 3.3 or ^). In eaoh case, to 230 ml of uranium solution, maintained at a constant tem perature of 50°C(+ 1-2°c), ammonia was added at a constant rate of about 0.5 ml/mln. with constant stirring. After the precipitation, the stirring was stopped and the settling rate of the precipitate layer was noted down, 1.1.1. Youden's ruggedness test (Table II) shows that the diff erence in the settling time due to the variation of pH is the highest and the effeot of variation in pn on the settling behaviour of ADUs far outweighs the effeot of variation In the concentration of either uranium or ammonia* These two pB values of 3.3 and 7 were obosen because the first one corresponds to that pH region In whioh the bulk of the preoipitation takes place (plateau region) and the seoond. viz.. pH 7 corresponds to the end of preoipitation* Apparently the role of the concentrations is only minor in comparison to pH. Low pH slurry is very sluggish In settling oompared to the high pH slurry. Low pH slurries will, thus, be decidedly troublesome in their settling behaviour. 1.1.2. On drying at 105°C, all these preoipitates caked. Though six of them were soft cakes, vbioh easily crumbled to powder on pressing in a glass rod, the two preoipitates formed at pH 3*3 and were dried without washing yielded comparatively hard cakes (olinker type). This is also reflected in the surface area values whioh are given in oolumn 6 and 7 of Table I, 4 - 670 - When the preoipltates are unwashed and dried, pH baa the maximum effect en the surface area as brought cut by the ruggedness test (Table II). This is essentially due to the formation of hard calces from the lower pH slurries leading to lower surface areas. When preoipltatos are washed this effect seems to disappear. In industrial operation, formation of suoh clinkers from low pH slu rries, may not be desirable, as in the subsequent stages these clinkers may Introduce thermal gradients during drying and decompo sition. These precipitates were calcined by slow heating to 670*0 and surface area measurements were made on the oxides also. These results are also included in Table I (columns 8 and °)» It is ia~ t ere sting to note that after gradual and slow decomposition, the oxides, In all these oases end up with more or less same surface area* However, it should be borne in mind that these results are based on small samples heated at a slow and constant rate of ICc/min. This will not be the case in plant operation, where heating is quioker and far less uniform, in addition to the cakes being thick. These factors may have a great role in producing powders of differing surface areas in the plant. 1.1.3. Tables III and 17 summarise the data on the composition and the ruggedness test on the compositions of these preoipltates. The unwashed precipitates contain around 6-9$ amnonlum nitrate, whioh ao expeoted, was reduced on washing. In all the unwashed preoipitateB, the NIL, to U mole ratios are all around 0*4 (Table IV). The yellow cake is usually referred to/ammonium din.- ranate (ADU). For a stoichiometric formula of (NH-JoUgO-j, tha NELiU ratio should be 1. However, no ratio reaching even 0.5 was observed in these experiments* Bourns and Watson' had reported, ae early as 1958, that the yellow preoipitate called ADO oan have NH JTJ ratio varying from 0.3 - 0.5, depending on the pH of the supernatant liquid and the amount of washing the preoipitate rece ives. They also reported that it can further lose anmanla when dried above 105°C. Colbarn et al have reported that the molecu- . lax structure of ADU is generally assumed to be represented as - 671 - 0 aC o;f (NHA)2U2Q,7 <3T (^A^A !^ oo^ination these* Septula , in a study of the composition of ammonium uranates, observed that NH,tD ratio varied from about 0.2 - 0.5 as pH increased. No ratio higher than 0.5 could be obtained by him. Our results in Table IV confirm these earlier observations. The yellow cake, though commonly called as diuranate, is basically a non-atoiohiometrio oompound of vari able composition. It is doubtful whether reprodcoible ohemical composition is achievable at all and hence is of no use as a para meter for either academic characterisation or for control purposes. Within these limitations* no composition difference vas noticeable in -the unwashed precipitates due to the variant conditions. When the preoipitatee were washed, there was a decrease In NH,»U ratio, as expected. While there is loss of ammonia in all the cases due to washing, there is a minor indication that the precipitates formed at the lower pH of 3.3 seem to lose ammcula faster than the ones at pE 7* This can he seen from the aotual mole ratios and also from the ruggedness test (Table IV). In view of the above, it may be more appropriate to call these calces as ammonium polyuranates rather than diuranatee. 1.2 * The Initial set of experiments showed that concentration of uranium and ammonia have only minor effeot, whereas pH has decidedly a greater influence on the precipitates. In view of this initial finding, the concentrations of uranium and amnonia were fixed at about 100 g/L and 10N respectively for subsequent experi ments. Attention was paid to a narrower range of pB 6 and 8 (on either side of pH 7), In the next set of multivarlant experiments. Table V gives the experimental conditions. The temperature of preci pitation was chosen as the second variant (50*C and room temperature of 25°c)« The third variant was the form of ammonia used - 10N ammonia solution of' gaseous ammonia. As before, the stirring and the rate of addition of ammonia solution were constant. Where ammo nia gas was used, about 3O-5O bubbles/minute, through an orifice of 3—4 n™ diameter, was maintained. 1.2.1. The settling times, In this set of experiments, are also given In Table V (column 5)* None of the three variables viz. pH - 672 - (in thia range), temperature of precipitation or farm of ammonia have any narked effect on the settling times as shown by the rugged- ness test for settling time (Table VI). Of the-three, the temperature has the least effect. pE as well as ammonia have some-what similar but marginally higher effeot. 1.2.2. The three variants in these experiments do not influence the surfaoe area of the unwashed preoipitates in a significant manner (Table VI). However, it is seen that the influence of tem perature is marginally more than that due to the other two. Preoi- pitation from warm solutions gives a produot of marginally higher surfaoe area. When the precipitateB are washed, pH baa the maximum effeot on the surfaoe area as is dramatically brought out by the ruggednesa test (Table Vl). As pointed out in the earlier set of experiments, ammonia is lost from the precipitate on washing. When the pHs of the water washings were actually measured In this set, the pH increased with each successive washing, going to greater than 9 pH in the oase of the preoipitates obtained at pH 8. Apparently this results in increased surface area of the preoipitates. As In the earlier set, the ADUs were decomposed and the surface area of the oxide residues are given in Table 7 (column 8 and 9). The oxides of all the washed preoipitates tend to reach a 2 —1 surfaoe area of about 5 m g and all the unwashed ones to about 6 m e • In one oase (experiment VIII), the ASH was heated to about 960°C instead of 650°C, and in this case the surface area d-ropped 2 -1 considerably to 1.4 m g • Thermal analysis Indicated that though the weight remained constant after about 650*C (Indicating no change In the composition), the D.T.A. baseline showed a gradual and conti nuous endothermio ohange in this region of 650-960'C (Figure 1). Considering that the surfaoe area at the end has fallen to 1.4 m g" 2 -1 from about 6 m g , it is reasonable to correlate these two and infer that at temperatures above 650*C, a sintering process of the oxide starts. 2 . EQUILIBRIUM CONDITIONS IN THE PRECIPITATION PROCESS The multifactorial experiments showed that of all the para meters studied, pH is the most decisive one. The type of oomposi- - 673 - 'tiona obtained by us, as veil as by the earlier workers, Bhcv that 0tHH_ ratio is ouch higher than would be expected on stoichiometric basis, indicating the presence of polymerised uranium species in the precipitates. This led us to the possibility of nan-attainment of equilibrium ae a faotor to bo looked into. Septula had earlier made a potentiometrio study using about 25 g/l» uranyl nitrate solu tion and 0.2 N ammonia, solution. Using concentrations close to those employed in our plants (100 g/L), we carried out potentio metrio titrations of uranyl solutions against ammonia. To 50 ml of uranyl nitrate solution* 10H ammonia was added and this titration van followed potentiometrically. Fig. 2 elves the potentiometrio titrations* Curve A. represents the case where after every incre mental addition of ammonium hydroxide, the mixture (whioh was const antly and vigorously stirred) was allowed to come to pH equilibrium before the next addition. Generally the time required was of the order of 5-10 minutes for eaoh addition. Hence about 15 minutes was allowed at every point. Curve B represents the titration in which ammonia was added at a oonstant arbitrary rate while the uranium solution was stirred continuously. The titration for curve B took far less time than that for curve A. Curve A and B oan be taken to represent equilibrium and non equilibrium conditions respectively. Titration under equilibrium conditions (Curve A.) shows that there is an initial rise in pH upto a value of about 3, during which the neutralisation of the free acid takes plaoe. Also, possi bly* some polymerisation of the uranyl ion nay be taking place simul taneously without preoipitation (darkening of the colour to brownish yellow from the initial greenish yellow). This is followed by a plateau rBgian at about pH 3*3 - 3»4« during wbioh ADU is preoipi- tated continuously ami gradually. Amnonia is consumed for the pre oipitation of ADU without any rise in pH during this stage* after whioh there is an inflexion point followed by a sudden rise from .3.5 to a pH region of 7-8 Indicating no reaction during this sudden jump. The pH tends to taper off slowly after this Jump, around a pH of about 9-10. It was notioed that the precipitate at this stage became ooaree* granular and tended to eettle faster even during the - 674 - titration. When ammonia was added under nan-equilibrium conditions (curve B - faster constant addition), after the initial increase in pH due to free acid neutralisation, the pH rose gradually hut continuously to a value of about 5 »5» around vhioh it fluctuated in a zig-zag manner, before shooting up parallelly to the other curve, from pH 6 onwards. Under equilibrium conditions, the precipitation takes place at a constant lower pH Whereas under non—equilibrium condi— tions, the plateau region is narrower, zig-zag and is higher by about 2.5 pB units. Curve A (equilibrium) is sucb that it can be expected to be reproduced, whereas the curve B(non-equilibrium) will be more difficult to reproduce. If the rate of addition is decreased and/or the mixing is mare thorough, the curve B will approach the equilibrium curve A and vioe-^versa. Our experimental observations, are very similar, though 5 not exactly identical to that of Deptula • His plateau region was closer to pH 4*2, needed far more time to reaoh equilibrium at the start of preoipitation (30-100 minutes) and the dip In his pH value at the start of precipitation was far more pronounced (about 1 pH unit). All these differences can easily be attributed to the differences in concentrations employed. Based on these curves, the following could be In ferred about the of»«.ra.uwv of plants. 1) 3h the first stage upto pH l/% 3, free acid will get neutralised (2) In the Booond stage (whioh is supposed to operate In the pH range 3-5), the way the ammonium hydroxide is added and mixed, will play a great role in deoiding the type of preoipitation. If non- oquilibrium conditions prevail, the quality of the produot oould be non-reproducible. It is felt that some of the non-roproduoibili- ties - particularly the settling/nonsettllng nature of the preci pitates obtainod can be rationalised through the above laboratory observations. 2.1. Influence* of pH and equilibrium on the Settling Rates of - 675 - the ADUV. Based on the above observations that when preoipitation is made olose to the equilibrium, and that -the preoipltates ao formed at pH about 9 are coarser and granular, the following ex periments were carried out to arrive at the best preoipitation conditione. To 100 ml of the uranium solutions (100 g/L), 10 N ammo nium hydroxide was added from a burette with oonstant vigorous stirring, with the pH being monitored constantly. The slurry at the end of each experiment was transferred to a 100 ml measuring cylinder (ID about 30 mm), and the sedimentation rates were meas ured. Figure 3 reoords the settling curves* In the first instance, the pH was raised to about 6.5 under equilibrium conditions (Incremental addition - about 2-3 hrs) left aside far about 3° minutes and the settling rate measured (1a, Fig.3). After measuring the Bottling rata, this slurry was retraneferred to the original beaker and the pH was raised further to 9, and the settling rate was measured again (ourve 1b)e These two curves dramatically bring out the steep improvement in the settling rate on raising the pH to about 9. The settling rate ia nearly five times faster than -that obtained for the product at pH 6-7. In the next experiment, a oonstant rate of addition of ammonia, of 1 ml/mln. (as against Incremental addition) was main tained and the pH raised to 6*5 in about 20 minutes* It waa noticed that the pE remained oloee to abotit3.3 during preoipitation, i.e. still olose to equilibrium. It was left at pH 6.5 for about an hour and then pH was raised to 9 (settling ourve 2). This ourve is olose to 1b indicating that it is possible to wort close to equili brium using oonstant rate of addition. Curve 3 represents the case where pH was raised direotly to pH 9 (but slowly and working close to equilibrium in about 20 minutes) with no interruption at about pH 6.5» This ourve is also close to the other two ourve a, though Blightly on the higher side. This particular slurry at pH 9 •waa left aside for another hour, was restlrred and the settling rate - 676 - remeasured. There vas no appreciable difference - i.e. allowing extra time at pH 9 does not additionally contribute to any improve ment in the settling rate. When the pH was raised direotly to 9 by addition of ammonia solution in a very fast manner in a few minutes (i.e. under non-equilibrium conditions, the slurry was found not to settle at all (Curve 4). Curve 3 represents a modification of Curve 4J the pH was raised very fast in a few minutes in a non- equilibrium way to pH 6.5 instead of to 9* but allowed to stand for about 100 minutes at this pH, to allow a chance to reaoh equili brium (pH fell to 5.8). The pH was subsequently raised to 9. Thi.3 curve is close to curve 4» shewing the importance of the 'plateau - equilibrium stags'* Two more experiments (curves 6 and 7) were repeated on separata days, similar to curves 1 and 2 and these con firm the above data. They also show the consistently good settling behaviour of these preoipltates. At pH 9» the initial steep portion of -the settling curves works out to be in the region of 4 cm/min. as against IBSB than 1 cm/min at pH 6-7* Besides, at pH 9, there are no fines floating after the settling. Our data on settling curves is also in conformity with those of Septula who used 10 g/1 uranium concentration and about 1N ammonia. The best results, for settling, are obtained when the slurry pH is raised to 9 instead of leaving it at pH 6-7. However, it is essential that the raising of the pH, from the beginning, should be gradual and olose to the equilibrium plateau region. Abrupt addition of ammonia or sudden rise in pH should be avoided. The ammonium nitrate content of the ADU s produced at pH 9 were in the region of about lOjfc. It was found possible to reduoe this to about Jji,by decanting ^he supernatant liquor, repulp- ing with dilute ammonia a couple of timed and discarding the super natant each time. The settling rate continued to be the same after the first two washings (Curve 7)« The surfaoe area values of ADU s produced at pH 9 corres ponding to curves 2,3 6 and 7 of Pig.3 and the corresponding oxide .residues obtained at 650*C axe given in Table VII. The surface - 677 - area values, both for the ADTJs as veil as calcined oxides, are higher than the ones obtained earlier at lower pHa. 5. CONCLUSION 1 • Ammonium polyuranate precipitated at pH 9 under conditions of pH equilibrium will give a product - of high settling rate - which is easily washed by decantation to remove the entrained ammonium nitrate 2 —1 - of high surface area of 9-10m g A3HJ (compared to 2 -1 the surface area of 3-5 n g as obtained at lower pH values). 2. The temperature of decomposition of TO, to V,0o should not exoeed 650°C or else there is the risk of sintering an oxide of high O/d ratio. 3* This uranate on oaloinr'iion below 650°c gives U,0a 2 -1 3 8 of high surface area of 6.5 - 7 m g (oompared to the value of 2 -1 4*5 m g as being obtained at low pH. ACKNOWLEDGEMENTS It is a pleasure to thank Shri C.C.Diaa and^P.V.Ravin- dran for the measurement of surface area of the products involved in this study and the thermal analysis.' ">;- REFERENCES ' ' 1. W.J.Youden, "Statistical Techniques for Collaborative Teats", The Association of Offioial Analytical Chemists, Washington, 1967. 2. B.C.Sarkar and M.Sankar Das, J, Indian Cham. Soo., Vol.LVII, 787-791 (1980) 3. W.T.Bourns and L.C.Watson, Report ABCL-757 (I958)l (CRCE 716 Pt.l). 4. R.P.Colborn, D.L.G., Bayne and M.N.Slabber, "Production of Yellow Cake and Uranium Fluorides", (page 229-260), lAEA,198p. 5. A.Beptula, Nukleonike., VII, 265 (1952) 6. A.Deptula, Nukleonika, VII, 341 (1962) - 678 - TABLE 1 Milt ifac t car ial Experiments Measured property Variables 1 2 3 5 6 7 8 9 Number pH Settling Surface ax«& of MHJ Surface area of oxide °HH.0H 4 tios 2 -1 2 -1 g/L a g H :Teo/3cm m g Jron forasbed from Washed ijlght tfavaehed Washed JDtT I 100 10 5.3 245 0.6+ 7.9 3.9 4.0 II 100 8 7.0 38 4.2 5.4 N.B. H.D. III 60 10 7.0 28 7.7 5.8 4.3 4.1 IV • 60 8 3.3 125 1.6+ 5.0 4.5 4.2 * Boundary vas less sharp 4 Hard calces, values are only indicative N.D. Not determined, oxide was lost. - 679 - TABLE II KUGGEDNESS TESTS Differ en oe in 'D' Values Variable JjJ«J^£ SetiW time Surface area of ASP able UnwaBhed Washed pH 7.0, 3.5 152" 4.9 0.9 Concentration of uranium . g/L 100, 60 65 2.3 1.3 Concentration of ammonia N 10, 8 55 1.2 1.7 38 + 28 245 + 125 - 152 2 2 245 + 38 28 + 125 b. 2 2 -65 245 + 28 38 + 125 2 2 -55 - 680 - TABLE III Coaposition T&washed ALU Washed ABU Expt.No. II III IV I II III IV * Based en estimatiomn «->ofr NOw^, ty reducing with Devarda'e alloys followed "by KB, detn. Corrected far NH.H0- present. - 681 - TABLE IV Effects of variables an composition Composition expressed M aole T*TW». °TT °mx ria r.u ratio of HH, to e/L 4 * U In the ABU E Dnwashad Washed I 100 10 3.3 O.38 0.24 II 100 8 7 O.36 0.31 III 60 10 7 O.39 O.3O IV 60 8 3.3 0.45 0.21 Ruggednesa test on composition Variable Difference in mole ratio due to the variable Unwashed Washed Concn. U 0.05 0.02 NH.OH 0.02 0.01 pH 0.04 0.08 - 682 - TABLE V Miltifaetorial Experiments Variables Measured property I JL Settling Surface area of. Surfa^ area of oxide time djuranatee nrg*"' a g** Ho. PH Temp °C Anmonia Seo/2om ht. Uhvashed Washed from unwashed from washed SSUB JDUB 7 6 50 Gas 57 4.9 6.0 6.0 5.3 8 Solution VI 25 5.2 6.2 6.0 1CK 56 5.3 VII 6 50 Solution 6.2 10H 52 3.7 3.7 5.3 VIII 6 25 GaB 92 3.0 2.6 1.4 4.9 - 683 - TABLE VI Ruggedneea Test Difference in 'D' values Values of the Settling Surfaoe area of Variable variable tim» ESQ Unwashed Washed pH 8, 6 26 0.7 3.0 Temperature 'G 50» 25 10 1.2 0.5 Ammonia Gas,. 10N 30 0.5 0.6 - 684 - TABIE VII 2 •- 1 Surfaoe Area Values m g Sample No. Diuranates Oxide residues 2 9.0 6.7 5 9.5 7.1 6 9.8 6.4 * 7 10.5 7.1 # Washed ADU Sample number corresponds to curve number of Pig.3. - 685 SURFACE AREA VALUES O ARE FOR OXIDES HEATED Z 85.82 " tr TO TEMPERATURES AT UJ POINTS INDICATE BY ARROW i P IG o 84.82 - ^°3 X " \ 6.0 m2g"'! HI UJ cc "DTA "*M ? To.U°C (FOR DTA) f < *-* 83.82 O) X U Z HI E \\i ~\ 3°8 Q. i -j- \ >—<• LU ! I z 82.82 - i UJ 1 1 N \ 1 1 \ WEIGH T :i 81.82 i i \ ?* i i \ i i \ UJ \ \ T 11 \ UJ \ u. 80.82 \Ufn2g'1 u. ii •i \7 79.82 ii i 1 . j 1— 1 1 560 660 760 860 960 TEMPERATURE (°C) FIG. 1 SINTERING EFFECT DUE TO TEMPERATURE - 688 - U-50ml OF 100 g/L A-UNDER pH EQUILIBRIUM CONDITIONS/ SLOW INCREMENTAL ADDITION OF AMMONIA B-UNDER NON-EQUILIBRIUM CONDITIONS/ FASTER ADDITION OF AMMONIA U 6 8 10 12 14 VOLUME OF NH^OH-10N (ml) FIG. 2 POTENTIOMETRIC TITRATION - f*7 - ±f\ A- PH : 9,. NON-EQUILIBRIUM CONDITIONS §15.0 B- pH: 6- 7/ EQUILIBRIUM w CONDITIONS C- pH: 9/ EQUILIBRIUM CONDITIONS O Hi a. 10.0 UL O »- X o ID I 5.0- 10 20 30 TIME (min.) FIG. 3 SETTLING CURVES - 688 - A CONTINUOUS REACTOR SYSTKM FOB PBBCIPITATION OF URANIUM FBOM URANYL SOLUTION I.A.Slddiqul. B.V.Shah, S.H.Tadpbale and S.V.Kumar Process Engineering and Systems Division, Bhabha Atomic Research Centre, Bombay. A continuous reactor system for conversion of Uranium from the Purex tail-end to yellow cake has been developed. This system has been successfully used in the ADU conversion plant of Plutonium Plant, Trombay. Precipitation of Uranium is carried out with gaseous ammonia in a compact continuous reactor with low hold up volumes. The system has been designed for reliable and efficient operation with minimum operator- intervention and i3 very stable to disturbance due to changes in feed quality and feed rate. No elaborate instrumentation is required. The operation is uncomplicated with straight-forward start-up, shut-down and freeze procedures. The reactants" are agitated by air passed into the reactor. The pH of the product overflow is measured in-line and is maintained at the desired value by automatic control of the flowrate of ammonia gas. The reactor is capable of handling Uranyl nitrate feed over a wide range of acidity, uranium content and feed flowrates without difficulty and loss of material. The pilot plant reactor of 22 lit active volume can process 10 to 12 kg uranium per hour. The reactor system and the associated instrumentation are described and design parameters of the system are discussed. - 689 - The Reaction Reprocessing in India is carried out almost entirely by the Purex process. Uranium obtained at the tail end of the Purex process is usually converted into uranium oxide through the ammonium diuranate route by reacting ammonia gas or ammonia liquor with uranyl nitrate solution to precipitate ammonium diuranate. The overall equation of the exothermic reaction is often represented as 2U02(N0s)2 + 6NKs + 3H20 > (NH4)2D207 + 4NH4N03 Precipitation is generally carried "out as a batch process. A continuous set-up would obviously offer several advantages. The Rsanhor KoiH prnnnt Fig. 1 shows the schematic flow diagram of the continuous pilot plant reactor. The pilot plant consists of a 200 mm diam xl000 mm reactor vessel with a conical bottom, feed and product tanks of 1000 lit capacity, a metering pump, pressure regulating valve for control of ammonia gas pressure, diaphragm control valve (DCV) for control of ammonia gas flow and a draft-tube mixer for agitation and mixing of product slurry. A process and instrumentation schematic of the pilot plant is shown in Fig. 2. The reactor is shown in Fig. 3. The product is withdrawn through an overflow line near the top of the reactor. A line is also provided at the bottom for total draining of the reactor contents. Mixing of the contents of the feed tank and the product tank is done with submerged draft tubes. About 50 1pm air flow rate is found to be sufficient for thorough mixing with 100 % submergence. - 690 - The Instrumentation The pH measurement loop consists of a Philips Industrial pH Amplifier Model-PP 9041 U3ed with a combined pH-electrode held in a specially designed electrode chamber. The chamber is located in the product slurry outlet line. Due to continuous refreshing, this closely represents the conditions in the reactor itself. The pH signal is tranduced to a pneumatic signal which is used to drive the DCV controlling the flow of ammonia gas. Fig. 4 shows a schematic diagram of the arrangement of the pH measurement and control loop. An on-line flowrate meter (Ref. 2) In the slurry line gives a direct digital indication of the flowrate. Probes are provided for measurement of level, pressure and temperature in the reactor. The precipitation reaction is exothermic, but this has not been found to be a consideration during design or operation. Rxpfirlmsntal Information Initial runs were conducted with feeds containing 20 to 30 g/1 uranium and approximately 0.5 H nitric acid. Reactor pH was maintained in the range 7.0 to 7.2 on the basi3 of earlier exploratory work. It was possible to control pH within 0.1 pH point. Feed containing 60 to 80 g/1 uranium could be processed without difficulty. Continuous mixing of the reactor contents by sparging of air prevents settling of solids in the reactor. The product quality is continuously maintained even when uranyl nitrate is fed from the top and product withdrawn through an overflow line located at a level below the inlet line, indicating that there is no short circuiting of feed to the product line. This factor has permitted considerable flexibility in the location of the feed nozzles on the reactor. - 691 - Operatinfi the reactor with product withdrawal by overflow has been found advantageous in certain respects as it eliminates the necessity of level control and reduces the possibility of ammonia carry over with the product. The entire outflow of the product is available for continuous measurement of pH. Chances of choking are reduced and there is no need of flow regulation. All these add up to reliable, economical and truly continuous operation, free from the necessity of operator intervention. With the pilot scale reactor of 22 lit active volume it is possible to achieve upto 12 kg/hr Uranium throughput over a range of feed quality and acidity. It was found that uniform throughput was obtained for concentrations of feed ranging from 40 to 240 g/1. Conclusions It can be concluded from prolonged pilot plant testing that a continuous process for the precipitation of uranium is practically advantageous. The continuous reactor required is compactly sized and of simple design, with a small requirement of vessels and tankage. The system does not require elaborate or expensive instrumentation. The reactor system is capable of being operated in a truly automatic mode and the product quality does not depend upon the operator. The process is very flexible with respect to feed quality and maintains the set point within a narrrow band. The operation is uncomplicated with straight forward start-up, shut-down and freeze procedures. The system is stable to disturbance due to changes in the „ flow of feed etc. Application ..,' The continuous system described in this paper can be - 692 - used with advantage in any application requiring precipitation of Uranium from Dranyl nitrate solution. The process was retro-fitted into the ADO Plant at F.E.D., Trorobay and run successfully. More development and design effort is in progress for adapting the system for application at Uranium Metal Plant, Trombay and at Nuclear Fuel Complex, Hyderabad. Refftrennes 1. I.A.Siddiqui, B.V.Shah, S.H.Tadpahle and S.V.Kumar : "A Continuous Process for Precipitation of Ammonium Diuranate from Oranyl Solution. Part-I" - Report BARC- 1422 (1988) 2. B.V.Shah, C.K.R.Kaimal, I.A.Siddiqui and S.V.Kumar : "A Direct Reading On-Line Flowrate Meter for Use in Radiochemical Plant" - Report BARC-1387 (1987) - 693 - pH pHwaaTm U- OKI RUHR t&CQRld SWU. COMERTER RUCTOt w M cotvans PI If Vent rth 0*c4rv4« i* eurnaac avian ill h Ctfllut LAVM Pk DHJNE t 04 POT ROCTO •0-* MMMU rbtUNSGt Cytrrier psnvxr T«« flg.li SCHEMATIC Of CONTINUOUS URANIUM PRECIPITATION flg.Ei CONTINUOUS ADU PRECIPITATION PROCESS PILOT PLANT AND INSTRUMENTATION SCHEMATIC 694 - fU DOJCATOR A** V*At frttMt^t CWTSDL.ES COHVCRItS k- /UfUFEH KCCMCR UttKTL iCTKITt LCVO. tCAURKNT fTO) vr»Do *ouo» o- -w- av h-Un iH mnono Su S^h Win y kit va.e A K Proaurt SN» SUM* ri63. ASU REACTOR Flg.-i' LOOP FOR IH-UNE pH MEASUBEMENT AND CONTROL - 695 - PREPARATION OF METAL GRADE URANIUM TRI OXIDE THROUGH AMMONIUM DIURANATE PRECIPITATION ROUTE S.R. Ramachandran, P.O. Shringarpure, A.M. Meghal Uranium Extraction Division, B.A.R.C. The metal grade uranium trioxide should possess different physical proportions as compared to the ceramic quality product. It should be a granular -free -flowing powder with high tap/pack density and should have high surface area etc. These properties are closely related to the physical properties of ammonium diuranate I. INTRODUCTION The uranium trioxide is an intermediate compound in the production o-f metal grade and ceramic grade uranium di-oxide. The metal grade uranium trioxide should possess di-f-ferent physical properties as compared to the ceramic grade product. It should be a granular free flowing powder with high tap/pack density and should have high surface area with high surface reactivity. These properties are closely related to the properties of ammori i um-di-uranate which in turn depends on the process parameters while precipitating. Uranium trioxide is produced by the denitration of pure uranyl nitrate solution or by the ADU route or AUC route. After- laboratory studies, ADU precipitate route was selected for metal production in Uranium Metal Plant. II. EQUIPMENT i. The precipitation tanks:- These are stainless steel tank's with outer jacket of mild steel for steam heating. The tanks have an agitator and a dip pipe for the dispersion of Amnion i a . ii. Mut;che Filters:- For the f i 1 ter-a t i on , washing and dr «• i ny i>< the cake, uacuum niitsche are used. iii. Calcination Furnace:- Three electrical resistance heating furnace of batch type are used. - 696 - III. PROCEDURE The precipitation, -filtration, washing and dry.ing are carried out batch-wise, 1000 litres of pure uranyl nitrate solution having a uranium concentration o-f ~50 gm U/l is taken. The solution is heated by steam to a temperature o-f 50°C. Ammonia gas diluted with air in the ratio 3:40 M3/hr are passed. The precipitation takes around 4 hours. When the -final pH o-f the slurry reaches 7.0 a -flow o-f ammonia and air is stopped. The slurry is then drained to the nutsche •filter and -filtered. The cake is washed with 3-4 bed volumes o-f demi neral i sed water. The cake is then vacuum dried for 2-3 hours and the cake is manually discharged into stainless steel trays. The cake trays are loaded into the calcination furnace and heated to a temperature o-f 475° C and this temperature is maintained -for 5 hours. The uranium trioxide obtained has a tap density in the range 2.1 to 2.5 gm/cc and the 0/U ratio is 2.78 to 3.0. The uranium trioxide is a -free -flowing powder and is further processed to make uranium metal. In order to improve the tap density and surface area of the oxide, the studies were made by changing the uranium concentration of the solution. The concentration of uranium in the pure uranyl nitrate solution was increased step wise. The tap density of the oxide obtained showed no appreciable change. However, with uranium concentration greater than 60 gm U/L; the tap density of the oxide obtained was around 1.5 gm/cc It was felt that more vigorous mixing and proper distribution of Ammonia - air mixer might help in improving the tap! density. Accordingly the agitator of one process tank was modified. The ammonia air line was extended and ' distributed through a ring at the bottom of the tank. A few batches of precipitation was carried out in this tank by using the pure uranyl nitrate solution. The uranium concentration was in the range of 62-86 gm U/l. The other conditions such as temperature of precipitation, ammonia-air rate were maintained as usual. The time of precipitation was in the range of 5-51/2 hours. The slurry . was filtered, washed and dried in the nutsche filter. Cake obtained was calcined. The tap density of the oxide was in the range 1.4 to 1.8 gm/cc. A few batches of precipitation were carried out at room temperature. The other parameters were maintained as per the normal procedure. The Uranium tri oxide obtained had tap density and 0/U ratio in the range of 1.6-2.4 gm/cc. and 2.9-3.1. The oxide from these batches were further processed through the reduction and hydrofluorination stages. No specific difficulties were encountered. However the tap density of the uranium tetraf1uoride was lower, than normal. - 697 - IV. RESULTS AND DISCUSSIONS The tap density of the oxide obtained> by precipitating pure solution o-f higher concentration was in the range o-f 1 .4 to 1.8 gm/cc and the oxide had brownish lumps as well as fines. By experience one can say this type o-f oxide gives problems in -further processing. The carry over o-f powder to the scrubber etc. in processing through reduction and hydro-f 1 uor i nat i on stages are also higher. The tap density o-f oxide obtained by precipitation of pure solution atroom temperature was also in the range o-f .1.6 to 2.4 gm/cc. These batches did not give any significant problems in the reduction and hydro-f I uor i nat i on stages, but the tap density.of the uranium te tra-f 1 uor i de was lower than normal. V. FUTURE DEVELOPMENT The process carried out in Uranium Metal Plant hitherto meets the quality and purity requirements o-f U03 required -for the production o-f metal grade UF4. However since it- is a batch process without inter stage material handling system, it is more labour intensive. To give an example, about 170 batches o-f ADU are required to be precipitated, -filtered and 700 trays o-f cake have to be handled -for calcination in a month to achieve the full production. It will therefore be more beneficial to develop a continuous integrated process with minimum inter-stage handling. Efforts are. therefore directed to develop continuous and integrated process consisting of precipitation, filtration, calcination and drumming station. They will not only improve the productivity, but reduce the safety problems associated with the handling of radio-active material. The conceptual design is shown in the drawing. In general, ADU route of precipitation of pure uranyl nitrate solution produces large quantity of ammonium nitrate solution, disposal of which becomes a problem because of tightening of environmental safety standard especially for nitrate. To avoid this problem, many uranium refineries in the world are switching over to denitration or AU route. As a long range plan' denitration route for production of U03 would have to be developed. ACKNOWLEDGEMENTS The process outlined in this report has been the result of development act.ivity carried out by a number of our senior colleagues. The final process has been evolved as a result of all their contributions. Ule are thankful to all of them. We also thank a number of staff members of Uranium Metal Plant, who are . regularly carrying out these operat i ons. - 698 - Thanks are due to Shri K.S,. Koppiker, Head, Uranium Extraction Division, B.A.R.C. -for encouragement. REFERENCES 1. Harrington CD. & Ruehle A.E., Uranium Production Technology 2. Shri L.M. Mahajan, P.D. Inamdar and P.D. Sb"ingarpure Internal Report 'UED/UMP/87-11 w. ". - 699 Table- - I Details of batches .precipitated with various Concentrations Uranyl Nitrate Solution Batch No. UNPD Con en. Time of ppn. Tap d 0/U Remarks U gm/L Hrs,.-mts . gm/cc 732 62.27 5 - 10 t .47 2.87 Brown Lumps 741 79.67 5 - 30 1 .39 2.93 Brown Lumps 1 .40 2.95 753* 78.31 4 - 40 1 .72 2.82 Brown Lumps 1 .45 2.94 760 86.62 5 - 15 1 .89 3.07 Brown Lumps 1 .45 2.94 The above batches were carried out in T408 after modifying the agitator of ammonia - air line. * Batch 753 was carried out in T-407 Table - II Details of batches precipitated without heating Uranyl Nitrate Solution UNPD Concn. Tap densi ty 0/U Remarks B. No. U gm/L gm/cc 459 53.27 2.00 2.92 500 41 .6? 2.08 2.97 501 40.48 2.08 3.08 737 54.74 2.22 2.91 738 59.11 2.41 2.90 739 59.955 1.5I .55 5 2.98 Small Brown Lumps 740 58.55 ' 1 .88 3.10 779 49.68 1 .72 2.93 780 46.76 1 .7? 2.90 781 50.12 1 .6? 2.89 782 50.92 2.05 2.90 783 50.222 1.91 .93 3 2.97 Small Brown Lumps 784 51 .73 1 .89 2.82 785 51 .36 1 .97 2.95 - 700 - CONTINUOUS PRECIPITATION, DRYING,CALCINATION. EXHAUST T P.H.C0NTR0LLER STEAM A WASHING I DRYING (K&PRV. X VOLUMETRE MH3 J . •? FLOW-CONTROL GAS KI UNPS HORIZONTAL BELT FILTER VACUUM (ROTARV KILN TYPE) *r-r* FILTERATION ^ B^ TYPE^K^ A.D.U.PPtn BALANCE DRYING 8 CALCINATION trbouu u DRUMMING STATION I 4MMIPNC DRG.NO. UP/R/07 - 701 - STUDIES ON THE PREPARATION AND CHARACTERISATION OF AMMONIUM URANYL CARBONATE V.N. Krishnan, M.S. Visweswaraiah, P.D. Shr i ngarpure. K.S. Koppiker Uranium Extraction Division, B.A.R.C. V.G. Date Atomic Fuels Division, B.A.R.C. Studies have been carried out in the laboratory on the preparation of Ammonium uranyl carbonate (AUO, using concentrated .solution o-f uranyl nitrate. The precipitation o-f AUC has been done by the addition o-f ammonium carbonate solution and secondly by injecting gaseous ammonia and carbon dioxide. The precipitates obtained under varying parameters have been characterised by chemical and XRD analysis and the precipitates obtained under ideal conditions have been -found to have the -formula C INTRODUCTION Uranium dioxide Nukem, in West Germany, has been in the forefront in developing the processes to convert UF^ / uranyj nitrate to UO2 with AUC as an intermediate product <1,2>. Several recent publications also give the improvements made in the AUC process <3-15). AUC is produced by the combination of uranyl nitrate and ammonium carbonate at specific pH and temperature. It is also prepared by passing ammonia and carbon dioxide gases into uranyl nitrate solution.. In an - 702 extractive process,the organic extract of uranium is treated with ammonium carbonate solution (11). In these processes ammonium carbonate solution / ammonia and carbon dioxide gases used were several times the stoichiometric requirement but the powders obtained are crystalline and -free -flowing. The -following are the reactions for the liquid and gaseous precipitations respectively. +3(NH CCI :: + U02(N03>2 6NH3+3C02+3H20= I.S. Chang et al. <13> have reported the preparation of AUC using nuclear grade uranyl nitrate solution together with ammonia and carbon dioxide gases. Similar work has also been reported by N. Swaminathan et al . (14) and A. Boulia and A. Mel 1 ah (15). In the present work, an attempt was made to prepare AUC by following two routes,using ammonia / ammonium carbonate solutons and secondly gaseous additions of ammonia and carbon dioxide to uranyl nitrate solution. The powders obtained were characterised by chemical and XRD analyses and measurement of surface area and particle size. A few samples were also converted to U02 to study their pel 1etisation and sintering properties. ~ EXPERIMENTAL A). Ammonia and ammonium carbonate solutions: To uranyl nitrate solution (50 / 200 g U/l), maintained at 40 C, a mixture of ammonia <1 OX w/v) and ammonium carbonate (25Xw/v) solutions were added with constant agitation. The amount of ammonia and ammonium carbonate were fixed as that required to raise the pH of the solution to 7 and to maintain a C/U ratio of '7.5 respectively. When the addition was complete, the already precipitaed ADU dissolved and slowly the AUC started precipitating out. The slurry was allowed to stand overnight for completion of prec.i p i t at i on . The precipitate was then filtered, washed with ethenol and dried at 80 C. The powder was then analysed and the results are shown in Table I. B). Gaseous precipitation! The experiments, mentioned earlier, were repeated by injecting ammonia and carbon - 703 - dioxide gases into the uranyl nitrate solution under agitation, through a single nozzle, with and without diluting the ammonia gas with air. The temperature of the solution was maintained at 60 C and -final pH at S.5 to 9.0. After reaching this pH, the rate of ammonia gas addition was reduced to 50 7. while the carbon dioxide and air addition rates remained unchanged and the addition o-f gases continued -for a further period of 4-5 hours. The precipitate was then filtered, washed and dried as mentioned in the earlier experiment. The dried powders were then tested and the results are shown in Table II. RESULTS AND DISCUSSION. In the case of experiments where the liquid precipitant was used to precipitate AUC from dilute uranium solutions <50 g/1) , the recovery was only about 55X and it was about 85X from concentrated uranium solutions <200 g/1> . However the recovery was over ??X in the case of gaseous precipitation from concentrated uranium solutions. AUC powders obtained from all the above mentioned experiments showed near stoichiometric chemical. composition. X-ray diffraction (XRD) analysis confirmed the standard pattern for the formula C (NH^) ^02(003) 3] mentioned in the ASTM card 27-1018 in all the cases except where ammonia was not diluted with air. In the later case the crystals obtained were needle like with length to breadth ratio of about 7 to 10 unl ike in the ?arl ier cases where this ratio was only 2 to 3 (see figures 1,2 &3>. Surface area of the powders obtained by gaseous precipitation showed a tenfold increase compared to the powders prepared using solution precipitant. The AUC powders confirming to chemical stoichiometry and adherence to the standard ASTM pattern have, on conversion to UO2, exhibited good compacting and sintering behaviour with no need for grannulation and precompacting. They also gave pellets with green densities 5.6 to 6.3 g/cc and a sintered density of 10.24 to 10.63 g/cc. CONCLUSION The recommended parameters for the preparation of AUC have been found to give powders with the required properties which on conversion to UO2, may reasonably be expected to give oxide powder with good pel 1 etisation and sintering behaviour. However, a few large scale trials will have to be undertaken before the findings of this work is to be put to pract i ce. - 704 - ACKNOWLEDGEMENT The authors would like to extend their thanks to Dr.V.S. Jakkal o-f Water Chemistry Division and Shri. A.R. Biswas o-f Atomic Fuels Division, -for their help in obtaining XRD analysis and SEM micrographs. REFERENCES 1. F. Ploger and H.Vietzke, Chemie. Ing. Techn., 37, 7, <1965> 692. 2. K.G. Hackstein and F. Ploger, Atomwirstsch. Atomtechn. 12, (1967) 524. 3. J. Sondermann, J. Nucl. Mat., 106, <1982> 45. <4. M. Becker, U.S. Patent 3963828 U976). 5; Yi- Ming Pan, Che- Bas Ma and Nien-Nan Hsu, J. Nucl. Mat., 99 (1981) 135. 6. I.S. Chang et al. Annual Report KAERI/ RR 240/20, 1980. 7. J.H. Park et al. " 454/84, 1984. 8. J.H. Park et al. " 511/85, 1986. 9. C.S. Choi et al. J. Nucl. Mat. 153 <1988) 148. 10. V. Baron, Atomic Energy Review, 17 <1979> 891. 11. K>. Baron et al . Iriorg. Chem. Acta. 81 <1984> 83. 12. M.S. Visweswaraiah and P.D. Inamdar and K. Subramanian, Symposium on Sintering and Sintering Products, B.A.R.C. Bombay, Oct. 29-31, 1979. 13. I.S. Chang, S.T. Hwang and J.H. park, * • I.A.E.A., Vienna, 7-10 April, 1986. p 63. 14. N. Swaminathan et a!., Advances in Uranium Re-fin ing and Conversion, Proceedings o-f A Technical Committee Meeting on Advances in Uranium Re-fining and Conversion. I.A.E.A., Vienna, 7- 10 April, 1986. P 21. 15. A. Boulia and A. Mel 1 ah, Hydrometallurgy, 21 <1989> 331. - 705 - TABLE : 1 EXPERIMENTAL PARAMETERS AND RESULTS FOF AUC PRODUCTION BY SOLUTION ADDITION Uriniin concentration: Final pH. Amonia solution » Amoniun carb. solution C/U = 7.5 (200 o/l) (8.5-9.0) (10 '/. vM (25 '/. «/v> AUC CHEMICAL ANALYSIS AUC PHYSICAL ANALYSIS X-RAY D1FFRN. SINTERED DENSITY C00E AS PER OF •/. U02 ZNH4 7. C03 S.AREA P.SIZE DENSITY TAP. DEN. ASTN - 27/1018 PELLETS M/6) (111) (6./CC) •(B./CC) (6./CC.) AUC-13. 48.12 14.8? 31.55 0.24 19.8 2.94 1.42 HATCHES 10.41 - 10.49 AUC-14. 51.54 14.41 32.04 0.18 21.0 2.97 1.43 HATCHES 10.37 - 10.41 AUC-15. 51.84 17.53 29.97 0.19 25.0 3.04 1.18 HATCHES NOT DONE AUC-14. 51.78 14.84 29.44 0.31 23.0 2.84 1.13 HATCHES 10.24 - 10.35 AUC-17. 51.18 14.11 29.29 0.50 20.5 3.04 1.19 HATCHES 10.19 - 10.39 TABLE : 11 EXPERIMENTAL PARAMETERS AND RESULTS FOR AUC PRODUCTION BY INJECTION OF GASES UriniuB concentration s 200 9/1, pH. i 8.5 - 9.0, truptraturf : 40 C. AUC CHEMICAL ANALYSIS AUCPKYSICA L ANALYSIS X-RAY D1FFRN. SINTERED DENSITY AS PER OF 7. U02 7. NH4 X C03 S.AREA P.SI2E DENSITY TAP.DEN. ASTH - 27/1018 PELLETS M/6) AUC-G/20 52.12 13.11 30.0S 3.04 17.5 3.0B 0.97 DOES NOT 10.40 - 10.48 AUC-6/25 53.38 15.41 28.59 3.23 11.9 3.02 1.03 DOES HOT NOT DONE AUC-6/24 53.22 14.54 28.71 3.44 U.4 2.94 0.94 DOES NOT 10.44 - 10.73 AUC-6/27 54.31 13.52 27.07 2.30 13.0 2.89 1.07 DOES NOT NOT DONE AUC-6/28 • 54.40 13.01 30.20 3.32 20.5 2.B7 1.10 HATCHES 10.44 - 10.59 ST0ICH. 51.72 13.79 34.48 i Amoftia was dilutid uith air -lob- 8KU IMMSHIf IMIU f ^ \ Ji. "\£v Fig.J. - SEM Micro^mph of AUC Powdcr.|7 ?ffitj|| 25*MD'40"" S-00000 P-00003 Fig. 2 - SEM Mit-i-ojroph 0-f /UC pOwic\er-Q.20 - 107- Ftq.V SEM Micrograph of AUC poi/Jcler- Q 28 - 708 - BATCH PRECIPITATION TECHNIQUE A PROCESS FOR UO POWDER PRODUCTION J.P.N. Srivastava. A.K. Sridharan. GVSRK Somayajulu. N. Swaminathan Nuclear Fuel Complex, Hyderabad AND K. Balaramamoorthy. Chief Executive. NFC. Hyderabad The problems in the pellet plant with reference to recovery can be summed up as follows: a) Variation in the density of the pellets beyond what could be attributed to heat transfer consideration in sintering. b) Spor idic occurrence of cracks and pits. c) Poor chip resistance for grinding. d) Occasional problems like bulging, blistering, and capping etc. of the pellet. It is believed that the powder quality as received may have a large influence on the first and second aspects. Though for the third the grindina conditions itself may be a cause. the larae variation in the sintered diameter, makes it difficult to adjust the cut within 4 thou, in a single pass. This problem is further compounded due to adjustment of tubes of 6 classes based on ID. All these. point to one thina that homooehity and reproduceability have to be obtained in the powder. The sequence of operations followed at UOP. ' NFC are as follows:. . . . r ' a) Precipitation of ADU in two stages with ION ammonia solution by continuous route. Extent of precipitation is approximately. 80% in first and balance in second vessel. The final pH of slurry is 6.8 to 7.0. • b> Filtration. c) Drying at 250 C continuously for 4 hours. d) Calcination in tunnel furnace at 670 C max. (duration 6 hours). e) Reduction in Rotary furnace at 625/650/650/625 C (1/2 hour). f) Stabilisation - 709 g) Grinding (hammer mill) h) Blending When all the controllable process parameters are maintained constant and" if variations in the quality of powder still occur it can safely be inferred that the non-homogenous nature of the powder can be "design induced". It is also generally believed that the. powder quality is mostly influenced at the precipitation stage. Regular settling test showed large variation in the nature of precipitate. In fact, precipitate quality dictates the design of all the equipments down the line. Thanks to our interaction with BARC. Dr. Sankar Das and his colleagues, studied the precipitation aspects in depth a few years ago and submitted an elegantly worded useful report for our consideration. This formed the base of our efforts to redesign the precipitation set up. They first established that in ADU precipitation, out of all process parameters, pH of precipitatior everts the maximum influence on the quality. They carriec ' : further to discover that precipitation done close to equilibrium conditions yielded a fast settling slurry when the final pH is brought to 8.5 or more and the characteristics are reproduceable and at other pH conditions, they inferred that different polyurantes are found. At the equilibrium condition, they got a NH /U molar ratio close to 0.5 which compares well with the reported values. He further explained that local excursions of pH above the equilibrium due to inadequate mixing inside the vessel, will also cause variations in the quality of the precipitate. The precipitation trials in their lab were done very slowly with vigorous mixing. The suggestions were clear and : only remained to be demonstrated on the large scale and assessed for its performance further upto sintering. Batch precipitation is the obvious choice. Initially 150 Kg batch size was chosen with the available tank, uranium concentration was maintained at 110 gm/lts as obtained in the plant. 10 N ammonia solution was used and added at a low rate and was distributed through many nozzles near the impeller of the agitator. Neutralisation of the FA was done fast by dumping the required quantity of ammonia and precipitation was carried out near completion over a period of 20 hours and the final pK was raised to 8.5 again by dumping excess ammonia. The slurry was filtered as usual and processed further by the normal route. The overall performance was somewhat better and promising. Powder was more free flowing compared to the normal route. - 710 - Encouraged by these results a 500 Ka U biaqer batch was planned as a prototype for expanded capacity of the plant. In this UO (NO ) (5000 1) was taken first and heated to 60 C Then liquor ammonia was added at the rate of 15 1/hr. distributed throuah a rina header located close to the bottom impellor of the agitator. The agitator is fitted^ with 3 paddle type impellers (1.2 m dia). Ammonia addition is done in the maximum turbulence zone. Equilibrium precipitation time was brought down to 16 hours. Precipitation was completed as described above and filtered through a rotary pan filter. with scroll discharge. In this the cake was discharged as fine shavings which is amenable to direct feeding into a fluidised bed or pneumatic drier. A directly heated U tube drier was designed and fabricated. The discharged cakes flowing downwards and then upwards, pass through two well insulated cyclone separators, a trap, a venturi scrubber and demister, and finally pass through one more vaccum receiver and then to vaccum line. The residence time was about 0.4 to 0.5 seconds. A fairly free flowing orange coloured powder amenable for screw feeding into the rotary calcination furnace was obtained. The powder was collected from the two cyclones (95% in the first and balance in the second). This dried powder was then calcined in the Rotary furnace and -reduced in another rotary furnace. Many batches were taken and results shown in the following tables. Results SPECIFIC SURFACE AREA OF ADU PRECIPITATE SI.No. Continuous precipitation Batch precipitation (m /gm) (m /gm) 1 6.97 4.77 2 5.68 4.31 3 11.45 3.82 4 7.78 3.49 5 10.74 4.39 6 11.40 4.81 - 711 - COMPARISON OF PROPERTIES OF BATCH Vs CONTINUOUS MATERIAL Properties Batch Continuous Settled Cake Volume 50-60 60-100 (cc) Settling Time (min) 3-5 15-30 Bulk Density U O 1.3-1.6 0.9-1.6 (gm/cc) SSA U O (m /g) 2.8-4.5 3.0-5.3 Bulk Density • 1.6-2.3 1.4-2.0 UO (gm/cc) SSA UO (m /g) 2.9-4.3 2.2-3.5 REVIEW Even though this powder did conspicuously improve the recovery a fter sintering, it was more consistent in its behaviour, Because of its uniform powder characteristics, choice of s imple equipment, like fluidised bed pneumatic drier and calcination furnaces are possible. This had already been exploited for drying and a similar but somewhat different desian is going to be tried shortly for calcination, Extremely fine precipitate is obtained by continuous precipitation and -this can be handled only after granulation, in fluidised bed equipments. So far the advantages a re mostly for UOP. However, more expenmen ts are required to improve the recovery after sintering Through fluidised bed furnace, control of surface area c an be more effectively established. In a few cases sintered d ensity of 10.65 ave. were obtained under plant conditions, accompanied by high incidence of cracks. But the surfac e area by BET method was only 2.9 M /gm. This is suggestive of the existence of internal pores, Compaction pressure req uired for this powder were also hiaher. It may be no ted that AUC derived powder also behaves similarly. On the whole, it can be said that advantages are yet to be fully realised. - 712 - U02 PRODUCTION VIA AUC ROUTE U.C.Gupta ' Meena.R N.Swaminathan Abstract Currently U02 19 being produced by ADU route on production scale. Recently development work was started in NFC on UOZ production via AUC route. A set-up having a capacity of 6 kgs of AUC per batch has been established. The U02 -powder produced, was subjected to various tests including XRO, SEM and sinterabiiity tests. The XRO results on AUC and UOZ have been found to be matching with the data published by IAEA. The SEM analysis shows that the powder produced has round shape and porosity as reported by other contries. The powder is found to be free-flowing and the green pellets could be made without pre-compacting, granulation and addition of binders. In view of encouraging results a bigger set-up having a capacity of ZO kgs of AUC per batch is under fabrication. The notable feature of the process developed at NFC is that it eliminates effluent generation. In the present work, calcination, reduction and stabilization were done on a batch basis in stationary tubular furnace. Development work involving improved integrated design of the furnace for these operations has been undertaken and fabrication has also been started. - 713 - U02 PRODUCTION VIA AUC ROUTE U.C.Gupta Meena.R N.Swaminathan 1,0 Introduction: Several dry and wet processes are being used for the production of U02 powder on large scale. There are two wet processes for U02 production. These are based on precipitation of Ammonioum-Di-uranate (ADU> or Ammonium Uranyl carbonate (AUC) compounds. The U02 powder produced through AUC route is found to be giving the following advantages: a. High Recovery: It is reported that Nukem, Germany are following AUC route and getting a recovery of greater than 95% where as Canadian companies are following AOU route with a recovery of about 87-88* in their pelletising operations. b. Free Flowability: The powder is free flowing and enables elimination of pre-compaction and granulation steps. c. Easy mechanization: Since the powder produced through AUC route is free flowing it is possible to completely mechanize its transfer through an enclosed system which would effectively avoid airborne activity in the environment during its production. .AOU route is being followed in NFC for production of U02. The recovery achieved on production scale by this route is very low. In view of the advantages with the AUC route, it was decided to develop this process and set up an automated pilot plant for the production of- U02 _powder. Bench-scale, experiments were, therefore, carried o"ut to establish parameters. Based on the experience a small pilot plant with a capacity 10 kgs of U02 per day is proposed to be set up to provide operational data and experience for setting up bigger production plants with mechanized material transfer and automated operations. 2.0 Bench-scale Experiments: The initial experiments were carried out with liquid reagents (solutions) rather than gaseous reagents because of the ease of handling liquids in glass-apparatus in very small scale. About 50 gms of AUC was produced in each batch by reacting UNPS with NH40H and 2.1 UNPS-Feed: To start with UNPS was used as feed solution and reacted with (NH4)2C03. AUC was precipitated out according to the reaction. The filtrate presumably contained mainly NH4N03 and carbonate complex of Uranium.Consequently, when the filtrate was used for recycle and fresh UNPS added for make-up, the- NOJ -ion concentration went on increasing in spite of maintaining a bleed. The increase of NO? -ion concentration after a particular level interfered with the precipitation of AUC and the quality and quantity of the precipitate started deteriorating. 2.2 Lean Solvent as Feed: In the above method only a few recycles were feasible. Experiments were next planned with a low concentration of U and NOJ ion concentration in the feed. The lean solvent from stripping section of solvent-extraction plant(ADU-Process) contains only 5 gms U/ltr. This LS was used as feed and treated with 2.J Extract as Feed: In the next experiments/trials the extract from solvent extraction section having a U concentration of 110 gms/ltr diluted with TBP to give final U concentration of } gms/ltr was. treated with 2.4 ADU as Feed: In view of the above obsesrvations.it was established that in order to enable repeated recycles of filtrate it is necessary that the feed U-solution be free of N03 -ions and impurities. Further experiments were therefore designed with pure AOL) as feed which would contain NO} -ion only in traces.The flow-sheet was modified as given in fig.2. Many trials were conducted and feasibility to recycle 15 times has been established without encountering any difficulty. Many more recycles seem feasible.however trials are yet to be carried out in this direction. A clear solution of ADU was taken as feed for the process. NH4HC0J was added to give a c/u of 10 and AUC was precipitated at a pH of 8-9 and-fiItered.The filtrate was recycled with make up ADU for the subsequent cycles of precipitation. 5.0 Scaled-up experiments -1; The next experiments were carried out at a bigger scale to produce about J00 gms of AUC per batch. A set' up was made for precipitation and filtration, which would also allow for the use of gaseous reagents. About JO trials were conducted with NH4HC03 soln. as reagent and atleast 1? recycles of filtrate were carried out successfully. Even after 1? recycles no difficulty was encountered in further recycling. About Q kgs. of AUC was produced by this method which was further subjected' to thermal treatment steps to produce about J.5 kgs. of U02 powder. Pellets were made and sintered to study the quality of U02. The characterisitics of the powders and pellets produced using this set up is given in Table-I. XRD studies were caried out on both AUC and U02 powders and the patterns are given in Annexure IA and IB. A comparison of the d-spacings of the AUC samples with corresponding d-values in the JCPDS (Joint Committee on Powder Diffraction System) Data Cards No.27-1018 and 27-1018A for AUC reveals a close resemblance. The d-spacings of the U02 sample corresponds to those published in the JCPDS Data Card No.>->50 for UD2. The stoichiometric formulation of the U02 sample was computed to be UO2.04. 4. 0 Scaled-up experiments- ^ Zj_ With the success achieved in the above trials of )00 gms scale, a new set-up was designed and fabricated and the batch size was increased to 6 kgs of AUC.. The powder produced was calcined, reduced and- stabi1ised in conventional furnaces to yield U02. Further, the powder was taken for characterization studies and compactibi1ity and sinterabiIity tests. The results of powder characterization studies (specific surface area,flowabi1ity, o'u. tap density.particle size,shape.SEM analysis) ' are given in Annaxure-III. - 716 - 5.0 Fluidised co1umn furnace for eaIcinat ion: In all the above experiments/trials conventional furnaces were used for the thermal treatment steps. Calcination was carried out in a muffle furnace and reduction in a batch furnace in HZ atmosphere followed by stabilization in the same furnace. The countries following AUC route for U02 production use fluidized bed furnace for thermal treatment of AUC powder. A similar column is envisaged for use at NFC also and development work started in this direction. Various parameters including flow characteristics of powders have been studied in glass columns of 1.0 to 1.5 metre height'. Based on these studies a 3 mtr. long furnace has been designed and fabricated. An automatic feeding device has also been designed and fitted to this furnace.' The salient features of the furnace include high capacity, low power consumption and uniform temperature distribution. This furnace will be used initially far calcination and with the experience gained, a reduction furnace will be designed incorporating safety features to handle H2 gas. 6.0 Results and discussion: A process starting with ADU as U-feed shows promising results. Upto to ADD stage the process of conversion of U-ore will remain same as that being followed at 'present. The powder produced by the AUC process shows very good properties with regard to chemical stoichimetry, particle size and shape, density, flowability, ' compactabi1ity and sinterabi1ity. The precipitation process has been demonstrated in 6 kg batch- size and recovery at this stage is found to be very high. General observations are as follows: 1. In the precipitation step, after 8 recycles NO? -ion build-up (residual) in filtrate was 5.7 gms. per litre.The Nitrate levels after more recycles are being determined. 2. The powder could be easily compacted without granulation, pre-.compaction or addition of binders. J. Green densities of the order of 6.5 gms/cc were achieved without the pellets cracking or chipping. 4. No major defects were visible after sintering and centreless grinding. 5. In one batch sintered density of 10.58 gms/ltr was obtained, however in others a density of 10.42 gms/cc was achieved. Optimization of parameters is going on to achieve higher density. 7.0 Flow-sheet for pi lot plant at NFC: In most of the countries where AUC route is used for U02 powder production, the flow-sheet followed begins with evaporation of UF6 (when starting with UF6) or concentration of UNH "(when starting with UNH) followed by precipitation of AUC using gaseous reagents like CO2 and NHJ. AUC cake is washed with - 717 - methanol befo're being pneumatically transferee! to the. next process. Calcination, reduction and stabilization are carried out in fluidized column furnaces under controlled conditions. After subsequent homogenization of lots,compaction, sintering and centreless grinding operations are carried out. The flow-sheet that will be followed at NFC is given in Annexure 11 -A 4 II-B. It varies from the above mentioned flow sheet, in that, ADU feed will be used instead of IMPS (when starting with natural II). All purification and other sieps prior to ADU production will be the same as that being followed at present in NFC. 7.0 Conclusion AUC route for producing U02 powder has been established successfully at 6 kgs scale. Powder characterization has been carried out and the powder is found to meet the specifications. The experience gained from the work has given confidence for setting up a small Pilot-plant at NFC with a capacity to produce 10 kgs of U02 per. batch. Design work in this direction has already been initiated. - 718 - Table-l Characteristics of powder produced in Laboratory scale set-up AUC Powder c/u 2.95 Reduced powder o/u 2.09 Bulk.density 2.13 gms/cc Tap.density 2.55 gms/cc Sp.surface.area 2.92 m2/gm Pellet Green.density 5.32 gms/cc Sintered.density 10.56 gms/cc - 719 - UNPS DM Water (NH^,)? CO3 I Make-up UNPS— Precipitation Filtration -AUC Crystals Filtrate Recycle -Bleed Filtrate to keep equilibrium level of Nitrate Fig, 1 FLOW SHEET FOR AUC-PROCESS ADU DM Water NH4HCO3 Li I Make up ADU- Precipitator Filtration -AUC Crystals Filtrate Recycle Fig- 2 MODIFIED FLOW SHEET FOR AUC-PROCESS o I l5 I?. rmMTH i'i.i •f.-jrlTtrTMSBCitl^ltlilaJUMKIi"' 1 * r • • ; _ -?r F' 1 ~7T 'ill! I ! i «>I •' Mil iTir.iF.iii irn:j;. nl • I w I i ! ! i! i • I 1111! 11 i 11 * 11 Ill i j! ; i i j | i ! Eii! w JtlllfM^MM! -ir,':ii|| ^ kit! . i 11 ! t! 11 i, i ! i:>_ '.^JJ LI.J1L!' iiiJll!LLiJJj;^iiilijL!j.: JJU.iLifJJIiliii i I* b ? - 722 - Vent Porous melal filter Recycle-tillrote Dried AUC J Gases Packed Spray nriTt'absorbe r absorber Spray absorber Dryiirymmg o> & a> i £ c '0 a "5 c v Hi- E _3 "o Comp. AUC-FLOW SHEET FOR UOp POWDER PRODUCTION - 723 - ANNEXURE-I1 B Stabilized Powder Compaction Sintering Centreless Grinding Inspection CONTINUATION OF FLOW SHEET FROM ANNEXURE-II A - 724 - ANNEXURE-III OBSERVATIONS ON TRIALS IN LAB-SCALE SET-UP 2 1. Precipitation and filtration: a) Very uniform sized AUC particles were obtained with no hard agglomerates. b) Ordinary methanol washing of precipitate did not instantaneously result in dry flowable powder. However, after a few hours of air drying after methanol wash, the powder was very dry and-flowable, that is, ammenable for pneumatic transfer. c) Nitrate ion build-up (residual) in filtrate was found to be 5.7 'gms/ltr after 8 recycles. N03 -ion concentration after 10,1Z,14,16 recycles is being determined 2. Calcination: Colour of calcined powder - black Specific surface area - 5.2 m2/gm Particle size distribution - 10-70 microns J. Reduction and Stabilization: Colour of reduced and stabilized powder - black o/u - 2.09 S.S. area - 5.5 m2/gm Particle size distribution.- 10-70 microns Tap density - 2.6 gms/cc 4. Compaction (without precompaction & granulation): Green density (range studied) - 5.6-6.15 gms/cc 5. Sintering: Sintered density - 10.4-10.>B gms/cc - 725 - "**£*&*£ , CSJZ/^} J>»0TO+fi*r*j ar Juc *»^*»<«* - 726 - flivArcnoKe — ijf (eohiri>\ itrfsj PMoto^.i'f"I or UjOg fO 727 - -•'.v V.-' t cAtv - "J (C»\r •< ( > iioi ./.*-.<' ^liM *n a/ /"' - 728 - ANALYTICAL TECHNIQUES IN URANIUM DIOXIDE FUEL PRODUCTION STREAM AT NFC T.S.Krishnan, S.Syanaunder, B.Gopalan, C.K. Ramamurthy and R. Narayanasv/amy Nuclear Fuel Complex, Hyderabad S00 762. The vital role played by the Chemical Quality Control during the production stage of reactor grade _uranium dioxide need not be overemphasised. A' well structured chemical quality control programme in the Uranium Fuel production industry calls for adoption of infalliable analytical techniques. Starting from ore concentrates and other raw material ingredients, the intermediates and '. final products have to be systematically subjected to chemical analysis. The working methods are to be carefully standardised, periodically reviewed and updated regularly. The present paper discusses such a system dedicated to the production of Uranium dioxide fuel in Nuclear Fuel Complex, Hyderabad. The salient features include sampling, process control analysis, Quality Control analysis and clearance. Laboratory management, etc. Measurement of Uranium concentration, acidity and percentage TBP of the various stream samples constitute the process control analysis. The Quality Control Analysis comprises of determination of trace impurities and physical parameters in the Intermediates and final - 729 - products. Tho experience 'gained over the years of operation and the feedback in restructuring the analytical programme is briefly discussed. New vis.tas in tho field are also indicated. 1.0 Introduction; The Nuclear Fuel Complex, Hyderabad manufactures Uranium dioxide fuel bundles for Indian power reactors starting from the raw material viz., Magnesium diurahate concentrates received-from UCIL, Jaduguda. This is' processed through dissolution, solventextraction, precipitation as ADU, Calcination and reduction to produce nuclear pure uranium dioxide powder. This powder is palletised, sintered and packed into zircaloy fuel tubes (Fig.l and 2). The ends of nineteen of them are welded together, to form', fuel bundles for PHUR. A similar procedure is followed for the manufacture of BUR fuel starting,, from enriched UF 6 of three levels of enrichment viz: 1.6, < 2.1 and 235 2.66% U . . The entire process requires close monitoring, involving various chemical and physical woasurements of samples drawn from appropriate stages. ' Such a monitoring procedure which is essentially the process control analysis provides ample guidance for adjusting the various process parameters. Assessment of impurity contents and physical characteristics of the final product, namely uranium dioxide forms part of Quality - 730 - Control programme. In addition to this, quality assurance methods aro required to be incorporated in all the stages of Quality control. 2.0. Sampling and Quality Control Plan: A well structured sampling programme is essential to give a proper meaning to the analytical results. The sample taken must be true representative of the whole batch/lot. Fig.l indicates the sampling points in the various production stages and Tables 1,11 and III the analytical requirement and the methods adopted. In general, one set of samples is drawn from each batch and in case any mixing or blending of batches are involved, further samples are analysed after such operations. Taking, a representative sample, ensuring its ' hotnogenity and authenticity, sending it to the Control Laboratory and requesting for the parameters to be determined are the responsibilities taken up by the production plants.- Sample receipt and entry, further homogenisation wherever required, distribution to various sections, of the laboratory, arranging for analysis, reporting and releasing the quality clearance of the production batch are the functions of Control laboratory. Depending on the requirement of a particular method, the UO pellets arc powdered and 2 powder samples are oxidised before taking up analysis. 3.0. Separation and proconcentration techniques: Soma of the analytical techniques followed require - 731 - separation and preconcentration steps to achieve lower detection limits and to remove interference or undesirable matrix effects. Thus, the determination of rare earths and certain impurities in Uranium by ICP- OE.S and AAS methods require prior aeparation/preconcentration by solvent extraction by TBP/TOPO. Thorium is determined spectrophotometrically after separation of uranium by Ion-exchange. However, in order to save time, a direct estimation without separation step is desirable wherever possible, as in the case of the determination of F, CI and N in Uranium by using appropriate ion selective electrodes. 4.0. Standardisation: The various analytical procedures had to be standardised initially after taking into account the specification 1imits(Tables IV and V), accuracies expected and available facilities.' The steps involved in standardisation of analytical procedures are: Setting up of instruments, calibration and qualification, running standards, establishment of working curves, trial runs with actual/synthetic samples and calculation of standard deviation and confidence levels, time planning and work schedule for establishing laboratory routine, evolving quality 'control procedures etc. These standard procedures are reviewed and updated periodically in tune with latest developments and also with a view to increase the rate of analytical output, i - 732 - 4.1. Instruments: The choice of instruments depends on tho accuracy required, multielement capability, interference, rapidity, sample size, direct sample analysis capability and operating cost. Mention has to be made about the versatility of emission spectrogrpahic method for analysing a host of Impurities In high purity uranium dioxide samples. Uhile most of the metallic impurities are done by emission spectrogaphy, covering a very limited spectral region, other techniques like Atomic Absorption spectrometry, UV-VIS spectrophotometry, Fluorinetry, Electroanalytical techniques. Inert Gas Fusion etc. have been adopted /or some of the critical impurities. 4.2. Calibration and qualification: Having made a judicious choice of the instruments, these have to bo calibrated by running synthetic and certified standards. In the case of uranium, since the availability of certified standards is rare, analysis of synthetic standards la resorted to. ' Some progress had been made in the preparation of indigenous uranium standards as a part of collaborative work beteen NFC and BARC. The instruments are qualified for routine analysis only after ensuring their proper functioning. 4.3.' Uorklng curves: Establishment of working curves is done by running a range of synthetic powder standards in the case of - 733 - spectrograph!c analysis and with the help of aqueous standards in most of the other cases. For example in the Atomic Absorption and ICP-OES procedures, the uranium matrix is separated out as indicated earlier and the raffinate is ta'ken up for analysis comparine them with aqueous standards. In order to calculate standard deviation. a single standard, often spiked ones, is run on different days repeatedly after subjecting through the routine separation analysis procedure. 4.4. Analytical.Techniques Table VI gives the analytical techniques presently employed in the case of uranium dioxide. A- brief outline of some of these techniques is descr-ibed below: 4.4.1. Emission apectrography: Bulk of the impurities in uranium matrix is analysed by - emission spectrography. Analytical procedures using this method have been standardised on oxide basis for most of the U matrices and hence it is necessary to convert all forms of samples to oxide. Carrier dls_t 11 lat i on method is adopted for the determination of impurities in uranium oxide. 4.4.2. Atomic Absorption spectrometry: This method is employed for the determination of some of the metallic impurities such as Ag, Al, Ca, Fe, Mg, Na, Zn etc. and is resorted to whenever high sensitivities are required for impurities ' which are otherwise difficult to determine by the emission - 734 apectrography because of matrix and spectral interf erencea. 4.4.3. Spectrophotometry: Quantitative semi-micro and nicro chemical analysis invariably involve this technique for the determination of certain selected impurities, especially the non- metallics, after chemically isolating the required specias( ions or complexes)and developing suitable colours. However, these are relatively more time consuming. 4.4.4. Electrochenical methods: Since the advent of select ion electrodes, the procedures for the determination of non-metallic impurities like chlorine, nitrogen have been simplified. Besides these potentiometric methods, other electro chemical methods like polarography and controlled potential coulometry are used for the measurement of uranium concentration. 4.4.5. Isotopic assay by mass spectrometry: This is a unique method applied to the regular 235 determination of U isotope content in BUR fuel. A single focussing mass spectrometer with a thermionic source and Faraday cup/electron multiplier detector units are employed for this purpose. 4.4.6. 0/U ratio determination: For powder samples, where the 0/U ratio la ^.2.OS, the gravimetric procedure is followed which involves - 735 - oxidation of sample to U 0 , weighing the sample before 3 8 and after oxidation and applying correction factor for moisture content. In tho case of pelleta, where the 0/U r-fitio ia near etoichiometry the sample is dissolved in phosphoric acid where U(VI) is determined polarographically and U(IV) is determined volumetrically. 0/U ratio is computed from the ratio of U(VI)/U(IV). 4.4.7. Inert gas fusion method: Hydrogen in UO pellets is determined by the inert 2 gas fusion method. In this method, sample is heated to o about 2500 C in a graphite crucible in Argon atmosphere. The hydrogen evolved is determined with thermal conductivity detector. 4.4.8. Surface area and par tide size determination: The surface area of UO powder is done by the " 2 modified single point BET method employing surface area meter by quantitatively measuring ( using differential manometry) the quantity of nitrogen adsorbed by a known quantity of the sample at liquid nitrogen temperature. The average particle size of UO powder is determined 2 by FiBher Subsieve Analyser by taking 10.9 gms. of UO 2 powder, packing it to the maximum extent and manometrically measuring the drop in pressure caused by tho rutiistance offered by the sample to the flow of air at constant pressure. 4.5. Time planning and work achedule: The process control analysis like uranium acidity in - 736 - feed and raffinate samples, percentage saturation of uranium etc., are carried out immediately on receipt of the samples and the results are communicated within hours. The impurities namely B, Fe,' He and Gd in -iritoi-'Wttdiata aamplaa like, uranyl nitrate pure solution are to be analysed on the same day. The surface area and 0/U ratio in U 0 and UO samples are also 3 8 2 determined on top priority basis. The impurities in final products of plants, like UO powder and pellets, 2 are analysed by various methods in the course of a couple of days. In order to cater to such tight schedules, a careful time planning is carried out and shift routines are drawn up according to priorities. At present, process control analyses have to be carried out in round-the-clock shifts and also on holidays to keep the production plants running. The results are communicatee', over telephone by the supervisor concerned followed by despatch, of confirmatory reports: 4.6. Quality control and clearance; Every UO powder lot is cleared after ensuring the 2 quality of the pevder falling within the specification limits as well as the Boron equivalent limits. In case of lots showing deviations, suitable blendjng procedures are adopted and the blended samples are once again subjected to analysis. On the analysis side various quality assurance procedures are incorporated as for example the frequent - 737 - analysis of "blind" samples other than routine standard samples. In case of deviations in the results obtained for the standard sample analysed in a particular shift, the results of the set of samples analysed are withheld and the causes are investigated. As a part of quality assurance procedures, the raw materials like acids,Oil water and TBP are regularly analysed. 5.0. Improvements: 5.1. Feed-back from Production Plants: Based on the experience gained over the years of operation as well as the review reports of expert Committees on production programme, restructuring of analytical schedules have been carried out from tine to time. These include additional analytical' work like : settling rate and uranium.in clear liquor in the ADU precipitation stage, nitrate in ADU, P, Na in UNPS and UO , Surface area of U 0 etc. 2 3 8. 5.2. New Developments: Lot of effort has already been made to improve the existing techniques and to substitute with the latest developments In the field. These include determination of uranium in solution samples by XRF spectrometry, determination of rare earths and molybdenum by ICP-0E5, determination of nitrogen by select-ion measurements.determination of 0/U by spectrophotometry etc. The Introduction of XRF spectrometry for the analysis of uranium in aqueous and organic process - 738 - stream samples has cut down enormously the tine required for analysis. The results obtained by this direct measurement method compares very well with the conventional methods as shown in Table VII. Similarly rare-earth analysis can now be carried out in a single day. 6.0. New Vistas: The advent of automated laboratory equipments such as micro processor controlled, computer driven and robotics instruments is bringing a revolu .ion in the future of analytical laboratory techniques. Further improved hyphenated techniques such as ICP-HS and ICP- FTS hold lot of promises for the analytical chemist, who henceforth has to be on his toes to catch up with the similar developments in the field. In the Quality Control Laboratory attached to Fuel fabrication industry, the main thrust areas have to be oriented towards quicker,' automated analysis work involving less manpower, time and auxiliary services. Introduction of online analysers such as insitu pH meters, gas sensors and controllers, online spectrophotometers, r - scanners and moisture analysers are* likely to reduce greatly the load on centralised Quality Control Laboratory in addition to optimising the production parameters. Cus toui-bui 11 laboratory information management systems (L1MS) carryout such complex functions like sample 'distribution, allocation - 739 * of priorities, preparation of daily schedule, data collection, report preparation, storage and retrieval of information, maintenance scheduling etc. The ubiquitous sample chancer is giving place to lab assistant robots which carry out normal manual operations like pipetting aliqots, diluting, sample queueing and placing in measuring chamber, rinsing and cleaning of sample cells. Expert systems endowed with artificial intelligence are expected to take decisions such as the best'method of analysis of a particular sample. This will be helpful in conducting trend analysis to aid the production plants. 7.0. Conclusion; Rapid and accurate chemical characterisation involving judicious choice of techniques, instruments, operators, data handling etc., improve the quantity and quality in fuel production centres. The laboratory has to be on the constant lookout for modification of various analytical procedures based on experience and devel opnient. - 740 - NITRIC ACID URANIUM OXIDE PLANT 1. FLOWSHEET UNS UNS*,,, SOLVENT UNP^ MDU AGING EVAPORATION DISSOLUTION EXTRACTION UNBf* SCRAP UNjf J NEUTRALISATION UNPE tfiOj /ADUC CAUSTIC LVE 3 FIlTRATPH 1 IUNRC* NH^DH. ADUC ADUS ADU FILTRATON CALCINATION DRYING •» PRECIPITATION fREJECTS FILTRATE U30fl PELLET / POWDER U02 ^ UOz REDUCTION STABILISATION GRINDING BLENDING rjW2. STORAGE LEGEND MOU ' Magnesium Di Uranatc cake ADUS Ammonium Di Uranatc Slurry UNS Uranyl Nitrate Slurry UNR Uranyl Nitrate Rafflnate UN? Uranyl Nitrate Pure Solvtion UNRF Uranyl Nitrate RaHlnate Filtrate g It •UNPE Uranyl Nitrate Pure solutton-EvaporuWd UNRC Uranyl Nitrate Rallinate Cake * SAMPLIN6, POINTS - 741 - PROCESS FLOW SHEET U&2 ££Li£IS uo, DECOMPACTION FWAL SINTERING UQPPOWPEPI AND GRANULATION COMPACTION AT1700t VISUAL WASHING AND CENTRELESS INSPECTION INSPECTION* ORVING GRINDING (DENSITY & H METALLOGRAPHY) STACKING HOT DM WATER • LOADING U02 TO ASSEMBLY 20STACKS SPRAY WASHING PELLETS W L AT 80*C AND . SECTION IN A TRAY FUEL TUBES DRYING AT 120C R: REJECTS FOR REPROCESSING - 742 - Tabic I SAMPLING AND ANALYTICAL SCHEME UO, PRODUCTION TYPE OF SAMPLE ANALYSIS SAMPLING FREQUENCY 1. UNS, CUN, UNF U & Free acidity Every 2000 lit 2. Extract & EP U & % U saturation Stream sample 3. Scrubbed raffinate U & Free acidity -do- it. UNPS & UNPE U, F.A., Rare Earths and impurities Each shift 5. Lean solvent U <5c TOP % -do- 6. UNR, ADU Filtrate Waste Samples U -do- 7. ADU Cake, UNC, UNRC U & Moisture -do- 8. UoOj. powder O/U ratio, U, TD, Every 500 kg.batch and SA 9. UO_ powder O/U, U, TD, UD,SA Every batch/ APS, MOJSTURE.RE & BLENDED LOT IMPURITIES 10.Sintered UO- Rare Earths, IS, Si, Fe 6 pellets Th & Other impurities from each O/U , & U content 500 kg. lot LEGEND 1. CUN Crude Nitrate '. UNl'S Pure solution 2. UNS Nitrate slurry 5. UNU Raffinate 3. UNF Nitrate feed 6. UNRC Raffinate cake - 743 - Table II PROCESS CONTROL ANALYSIS IMACRO LEVEL ANALYSIS) METHOD CONTROL VOLUHETRY U IN CUN, UNF, UNS; UNPS, EXTRACT, EXTRACT PURE X-RAY FLUORESCENCE U IN CUN, UNF, UNS, UNPS, UNR SPECTROPHOTOMETRY U IN RAFFINATE, UNR, UNRC, WASTE SAMPLES POTENTJOMETRY FREE ACIDITY IN CUN, UNF, UNS, UNPS VOLUMETRY TBP tt & % U SATURATION IN E AND'EP LEGEND Crude Nitrate UNS Nitrate Slurry UNF Nitrate Feed UNPS Pure Solution UNR Ralfinate UNRC Ralfinate cake - 744 - Table III " ' QUALITY CONTROL ANALYSIS FOR UP, (SEMI MICRO AND MICRO LEVEL ANALYSIS) METHOD IMPURITIES EMISSION SPECTROGRAPHY METALLIC IMPURITIES LIKE B, Cd.Co, Crete AND RARE EARTHS LIKE Cc, Dy, Er, Eu, Gd & Sm AFTER SEPARATION ICP RARE EARTHS AFTER SEPARATION ATOMIC ABSORPTION METALLIC IMPURITIES LIKE Ag, AI, Ce, Fe,'Mg, Si, Zn etc. AFTER SEPARATION OF MATRIX MASS SPECTROMETRY ISOTOPE RATIO COMBUSTION - GC CARBON,SULPHUR MICRO KJELDHAL - PHOTOMETRY NITROGEN SPECTROPHOTOMETRY THORIUM SELECT ION ELECTRODE FLUORIDE & CHLORIDE GRAVIMETRY AND POLAROGRAPHY O/U RATIO INERT GAS FUSION HYDROGEN - 745 - Table IV PARAMETERS FOR UO^FOTOER O CHEMICAL COMPOSITION IMPURITIES Vi PPM N U BASIS ELEMENT SPEC MAXIMUM • TYPICAL ANAli AR I < 1 Al 25 . <25 * • D 0.3 0.11 C 200 50 C* 50 35 Cd 0L2 < 0.1 Cr 15 < 10 Cu 10 <6 Dy . 0.15 < 0.1 F • 10 •C 10 Fe 50 « Gd 0.1 < 0.0* MR . 10 9 Mn 5 < 5 Mo 2 <5 Ni 20 2 SI 30 •C 30 Th •— •S EBC 1-1 0.51 O/U RATIO 5. TAP DENSITY SURFACE AREA 6. BULK DENSITY AVERAGE PARTICLE SIZE . 7. SINTERABILITY - 746 - Table V PARAMETERS FOR UOj PELLET CHEMICAL COMPOSITION IMPURITIES IN PPM ONU BASIS ELEMENT MAXIMUM ALLOWED TYPICAL ANALYSIS Ag 1 < 1 Al SO <25 B 03 <0.1 C 100 < 10 Ca X < 36 Cd 0.2 <0.1 Cr 25 <10 Cu 20 < 6 Dy 0.15 <0.1 F 10 -< 10 Fe 100. 60 Cd o.r O/U RATIO METALLOGRAPHY 6. SURFACE INTEGRITY ~ ?47 ~ Tabic VI P ,l.oI 2 ANALYTICAL PROCEDURES FOR NUCLEAR GRADE. URANIUM DIOXIDE • DETECTION METHOD IMPURITIES LIMITStppm) PRECISION 1. Spectrography B,-Cd' 0.1 + 10% Carrier distillation Co, Mn, Ni, Sb 1 AgCl Carrier Pb, Sn • 2 Sample: AgCl-19:l Mo 5 Electrode Charge: 100 Mg V " 10 . Current: 10 amp DC W 50 2. Spectrography after -. -.. 0.04 . +15% separation from 5 gms of .' uranium by Solvent Dy, Er, Sn 0.1 Extraction with TBP-CCM _ and precipitation as 0.2 .oxalates using Lanthanum carrier Direct excitation Sample: Graphite Powder 1:5 Charge: 8mg Current: 10 amp DC 3. ICP Spectrometry • Eu, Gd ~ after separation from 5gm " 0.1 . + 10% of uranium by Solvent Dy, Er, Sm Extraction with TBP- CCW and TOPO. The sample is aspirated in. Argon Plasma of 12 1/min 4. Atomic Absorption Spectrophotometry after Ag, Zn, Mg 1 +.5% solvent cxtration with ' TBP from 5 gm sample Fe, Si, Al, Ca 10 Air/Nitrous oxide & Acetylene Flame 5. Flame photometry after N .. solvent extraction sepa- - J7° ration with TBP-CCU Air/Acetylene Flame coru....2 - 748 - two Table VI P.2.of.2 6. Spectrophotometry 6.1 Pyrohydrolysis followed 10 + 5% • by bleaching of Zirconium ' Briochrome Xylenel complex 6.2 Pyrochydrolysis followed CI 5% by indirect determination with Mercuric-thiocynate - Fe HI 6.3 Separation by micro- kjeldhal distillation N 10 + 5% and determination with Nesseler's reagent 6.4 Anion exchange separation Th + 596 and estimation of Arsenazo III complex r. Direct potentiometry with F, CI, NH3' .10..- + 5% ^corresponding select ion. ^glSptrodes 3v • PiOlarography - after VI dissolution in phosphoric U .acid under argon atmos (for O/U) 500 5% phere 9. Low pressure measurement Moisture 20 0% By freezing the evolved moisture expanding in standard volume and measuring manometricaliy 10. Inert Gas Fusion Estimation by thermal H 0.01 + 10% conductivity after gas chromatographic sepa ration 11. Combustion-IR chromato graphy 10 • 10% Induction heating in oxygen stream and non- dispersive IR absorption - 749 - Table-Y.11 COMPARISON OP THE RESULTS OBTAINED BY THE .X R F .' METHOD WITH THAT OF CONVENTIONAL CHEMICAL METHODS FOR URANIUM Chemical Value XRF Value (URms/Litre) (URms/Litre) 1. Uranyl Nitrate Slurry . 385.31 381.61 2. Uranyl Nitrate slurry 297.54 300.89 3. Uranyl Nitrate Feed 252.91 253.66 4. Uranyl Nitrate Feed 254.40 259.00 5. Uranyl Nitrate Feed 154.72 155.35 6. Solution from tank 1201B 196.40 198.74 7. Uranyl Nitrate pure solution 106.80 106.66 8. -do- 65.46 66.11 9. Clear liquor 84.31 84.36 lO.Clear liquor 21.08 20.73 11,Uranyl Nitrate Raffinate 0.32 0.40 12. -do- 0.24 0.24 13. -do- 0.23 0.23 14. Lean solvent 10.71 11.39 15.Lean solvent 12.62 12.50 16.Lean solvent 11.74 11.15- 17.Lean solvent 11.45 11.30 SESSION VII URANIUM FUEL PREPARATION Chairman : Shri B. BHATTAWARJEE B A R C Reportsur: Shri D.K. BOSE B A R C - 750 - IMPROVEMENTS IN PROCESS TECHNOLOGY FOR URANIUM METAL PRODUCTION A.M. Meghal , H. Singh and K.S. Koppiker Uranium Extraction Division, BARC, Bombay The Cirus. and Dhruva research reactors in Trombay use metallic uranium -fuel. These requirements have been met by nuclear grade uranium ingots produced at Uranium Metal Plant. The plant operated continuously -from 1959 to 1980 when an expansion programme was undertaken. The experience gained since re-commissioning is discussed in the light of recent international developments in process technology. Short-term measures -for improvements in existing plant as well as long-term development trends are discussed. I. INTRODUCTION. The production o-f nuclear grade metallic uranium -for use as -fuel -for the research reactors in Trombay began at Uranium Metal Plant in 1959. The plant capacity was expanded by 1983 to meet the additional demand o-f -fuel -for Dhruva reactor. The plant has supplied the entire initial charge and periodic replenishments required, including the second charge for modified -fuel assemblies. Expansion o-f capacity is based on increase in number o-f process equipment o-f similar capacity as the original plant and increase in number o-f batch operations per equipment. Design assumes continuous operation, 24 hours a day and 300 days in a year. This strategy was adopted in the interests o-f high equipment reliability based on experience and due to constraints o-f time and space. The only major process change adopted in the expanded plant was the use o-f magnesio-thermic reduction o-f UF4 in place o-f cal c i o-thermi c reduction. This change was necessitated as the calcium metal was not indigenously available and supplies -from -foreign parties were uncertain. During expansion, a decision was also taken to replace pulse columns by mixer-settlers for solvent extraction. Both these changes were carried out based on the limited inhouse experience. The overall process flowsheet is shown in Figure 1 . The plant has over the years, been required to meet increasingly stringent standards of nuclear and industrial safety. The chemicals used are 'jery corrosive and toxic, similar- chemical plants elsewhere have been scrapped and re-bui 1 t <2>. - 751 - II. OPERATING EXPERIENCE The plant after completion o-f -expansion was commissioned in 1983. Since them continuous efforts were made -for optimisation and stream-lining, backed by laboratory and plant-scale R&D work. Experience has shown •foil ow i n g constraints. a. High level o-f labour intensive operations involving large number of low volume batches through long sequence o-f processing steps. b. Multiplicity o-f equipment increasing the maintenance load to an extent where existing resources are inadequate, with a consequent fall in equipment avai1abi1i ty. c. Irregular supply of critical inputs. d. Unexpected technical constraints in the process changes incorporated during expansion. e. Need to meet new safety standards laid dcwn by regulatory authorities. The details of these constraints and the improvesnonts in process technology for over-coming them by shor-v-term measures are described activity-wise below. 2.1 Pi ssolut i on: Charging of feed materials into dissolution tanks manually has been problematic . ' Scrip oxide powder dissolution can lead to NO* ewoiutinn. Mechanical addition by a hopper with vibratory feeder at reduced rate (30 kg/hr) has been found necessary. Scrubbing medium has been changed from water to \V. alkali to insure that effluent NOx level is below TLV limit of 15 ppm. i": pneumatic transfer system for yellow cake has been installed under -testing. No suitable equipment for safe' dissolution of turnings of metallic uranium has been found; Turnings dissolution been found unacceptable , as the incineration facilities at AFD are being augmented to meet the load from UMP . 2.2 Solvent Extraction; Major problem faced is build-up of siliceous crud in extraction mixer-settler. The feed solution after dissolution and filteration in presses contains over 0.5 g/l of Si02, while mixer-settlers need clear solution with less than 50 mg/l of suspended solids. Crud build-up necessitates frequent shut down for labour-intensive drainage and clean-up. A decision is now t vken to install a slurry-extractor system similar to N.F.C. di-sign<3). The efficiency of stripping mixer-settler and stl'.'ent processing units have been found to be lower due to solv*nt'aging'. Larger units are now under fabrication. 2.3 Raffinate Disposali With the installation: of slurry extractor, the load on raffinate .filteration will increase. Besides there is need to replace labour-intensive filter - 752 - presses. A rotary drum -filter OP a mechanised press wi 11 be installed, with drumming station to handle 500 1/hr of 10X soli ds. i 2.4 ADU Precipitation; Due to space constraints, the precipitation is carried out in 3 tanks in batches o-f 50 kg each. Filteration by vacuum nutsche -filter is 1abour-intensiwe time consuming <8 hrs.) operation. An automatic batching device with pH control needs development. Belt -filters o-f the Pannevis or Hindustan Dorr Oliver type and mechanised nutsche -filters have been evaluated. Tests are also planned on a centri-fuge also. Simultaneously with •filteration, a mechanised calcination equipment with closed drumming station is also being considered. 2.5 Conversion o-f U03 to UF4; Reduction o-f U03 to U02 and its subsequent conversion to UF4 is carried out in 125 mm rotary tubular -furnaces. At each stage powder transfer is manual, in batches of 50 kg, and -feed rate is low 5--6 kg/hr. At such low rates mechanical -feeders do not perform satisfactorily, especially since the powder is fine and cohesive - not grannular and free-flowing. The HF gas is highly corrosive and toxic. Seal life is low, about two weeks, when the original bellow-seal design is used with split bearing. Now an improved seal with integral bearing has been designed and life doubled. In addition repair time has been reduced. While the space does not permit installation of one bigger furnace, efforts are directed at better maintenance, more spares and operating control. 2.6 Maanesi o-thermic-Reduct ion : Experience has shown that cycle times are longer <48 hrs.>, break-down is more frequent, slag processing is morecomplex and ingot quality poorer than estimated. Efforts have been made and ingot size successfully increased from 60 kg to 200 kg. Other, measures include installation of integrated slag processing facility in a new location. provision of additional furnace, lining stand, blender cum dryer, and discharging stand. 2.8 Instrumentat i on 8 The need for instrumentation has grown with expansion. A centralised instrumentation station with automatic data logging, analysis and alarm indication is being set-rup. An automatic control system for the slurry extractor is:under development. This strategy of selective instrumentation and control has been adopted since - 753 - retro-fitting modern computer control system to entire plant will not only be exhorbitant, but also not -feasible. Some o-f the instrumentation, such as phase ratio indicator -for extraction and r-indicator -for magnesi o-thermy, needs R&D work. -Bulk o-f the instruments are however routine e.g. -flow monitors, temp, monitors and will be installed shortly. III. 'FUTURE DEVELOPMENTS Improvements in process technology in existing plant are limited by available space, existing lay-out, and the need to continue production without interruption. However, in the longrun, major improvements aught to be incorporated since -fuel -for the reactors is to be produced •for decade. Moreover alternative uses o-f metal , such as in AVLIS process, have to be borne in mind. Assessment o-f existing technology with re-ference to international developments, indicates -following areas o-f work: 3.1 Continuous dissolution; This is a widely accepted practice in plants abroad <4) (5) and ensures continuous uniform -feed to the wet plant. The existing slurry-extractor design has a solvent inventory 5-10 times that o-f mixer-settler. Development o-f a more compact extractor is needed. 3.2 U03 Production: The ADU route yields a powder o-f low tap density <2.3 g/cm) poor -flow ability and i neons: stant reactivity. General practice -for magnesi o—thermy is to use powder obtained by de-nitration, wheih is more dense <3-3.5 g/cc)and granular <4><5>. The ADU derived powder is not amenable to mechanised handling unless cake is extruded (4). The granular U03, by comparison, can be readily conveyed mechan i cal l,y. Hence denitration route -for U03 production wilt need investigation. 3.3 UF4 Production: A combined (or integrated) system for- reduction and hydrof1uorination needs to be developed to avoid intermediate trans-fer o-f pyro-phoric U02. This will also enable improved HF economy. Designing such a system will solve many o-f the production problems. 3.4 Metal ProrJuc t i on : High density pelletised charge with hard refractory liner in a top loading -furnace has been proven to be a safer and better method o-f metal production. While some work is started in UMP it needs considerable R&D input. 3.5 E-f-fluent Treatment: A wet process -for UF4 production needs to be developed for utilising the large amount of aqueous HF generated during hydrof 1uorination of U02. Denitration route of U03 production need be developed to reduce nitrate effluent. - 754 - IV. CONCLUSION The process technology used -for metal production is proven and well established. However technological improvements are required to reduce labour-intensive operations, increase scale of operation, reduce maintenance improve personnel and machine safety and increase productivity. Many of these changes can be incorporated only when better lay-out can be' made in new space. Meanwhile measures for improved production under existing constraints have been identified and action taken for itnpl ementat i on , REFERENCES 1. 6. Uirths and L. Ziehl, 'Special problems connected with production of uranium metal and uranium compounds', P1001, 2nd Int. UN Conf. on Peaceful Uses of Atomic Energy, Geneva 1958, Volume 4. 2. T.J. Heal, J.E. Littlehild and H. Page "Fuel Production - an advancing technology" Nuclear Engg. Int. April 1980, p 48-51 3. N. Swaminathan Recent Advances in Uranium Refining and Conversion Adv. Group Meeting of IAEA, Vienna 1986 4. H. Page 'Conversion of uranium ore concentrates to nuclear fuel at B.N.F.L.' -ibid- 5. A.W. Ashbrook 'Refining and conversion of yellow cake in Canada -ibid.- 6. R.P. Colborn 'Uranium refining in S. Africa' Proc. Advisory Group Meeting Panel 1979, IAEA, Vienna 1980 - 755 -r MAGNESIUM OIURANATE CONCENTRATE (YELLOW CAKE) 1 NITRIC DISSOLUTION A&B 32-5?. TBP SaVENT RAFF1NATE , 1NKER6SENE EXTRACTION FOR DISPOSAL RLRE URAW NITRATE Mu _^. AMMONIUM OIURAMATE a ^^ PRECIPITATION FILTRAT1DN FILTRATE FOR, DISPOSAL ADU CAKE CALCINATION -*-Nt3 JT REDUCTION NHi AMMONIA Ibltb- CRACKER r"' ** i»2 fflHYDPQUSHF HrOROFLUORJNATKW PROCESSED SLAG E CRUSHING >MgF2 SLAG MAGNESJO- THERMIC MAGNESIUM REDUCTON l.,^^^»v. GRINDING SLAG U'METAL INGOT "1PROCESSIN G T LEACH LIQUOf* FOR U'RECOVtfcY , Fig.1-FLOW SHhLT FOR URANIUM METAL' PRODUCTION ^M Drg.NoJJP/R/02 - 756 - IMPROVEMENTS IN EQUIPMENT DESIGN FOR HYDROFLUORINATION OF U02 TO UF4 A.C. C'edak, R.N. Kerkar , A.M. Meghal Uranium E\ trsrtion D i',' i ? i on , P.A.R.C. Uranium t e tr af 1 uor i de I . INTRODUCTION Uranium tetra -fluoride is a U02 + 4 HF UF4 + 2H20. A.H = - 42 KCal/mole a The reaction is carried out at. about 400 C. In the Uranium Metal Plant, this conversion is carried out in electrically heated rotary tubular convertors (Fig.'l). A controlled. quantity o-f U02 powder is -fed by a closed screw conveyor into the rotating tube. AHF gas a-fter metering is -fed at' the other end counter current to the U02 powder. Since HF gas, used in the conversion, is very hazardous and corrosive, it is utmost important that the system should be leak proof. To prevent the ingress o-f the HF from the rotating tube, seals are provided at both the ends. I I . CONSTRUCTION DETAILS OF THE EXISTING COWERTOR This consists of double tube, outer one of mild steel and inner of magnesium or inconel b> split brass ring. The seals and bearings are assemt in the same bearing housing. This, complicated arrangement of bearing was. -found to be problematic and time consuming -for the maintenance. The mean time between the two -failures was around two weeks. Moreover time required -for dismantling, replacing worn out parts and assembling was to long, thereby increasing the down time. This prompted us to modify the complicated seal arrangement. The choice was between mechanics! seal and -fabricated lip seal assembly. Since the mechanical seal o-f proven quality was not available -for this service, fabricated lip seal arrangement was selected. The objectives o-f modification were 1. to provide the seal with higher MTBF (operational rel i abi1 i ty> 2. to simpl ify the arrangement for easy and speedy ma i n tenance 3. to meet the leak tightness standard 4. to reduce the cost and eliminate imported components. III. DETAILS OF THE MODIFICATIONS 1. In—situ bearing arrangement uFig.3.> was replaced by single standard, bearing. 2. Instead of compl icated copper bellow seal and 1 ip seal , only.1ip seal arrangement is provided. 3. To reduce the load on the bearing, double tube assembly wa; changed to single inconel tube assembly. 4. Longitudinal baffles have been provided in the tube furnace to provide better mixing of U02 powder and HF. 1V. OPERATING EXPERIENCE OF MODIFIED CONl'ERTOR The furnace after- modification has been in operation for ivbou t t'.-'O .'ears. The^ per f ormance has been found to be - 758 - highly satisfactory. The- sdusntages accrued -from this modifiestion are : 1. The MTBF hs; been increased -from two weeks earlier to •four weeKs, 2. It has simpl i f i ed and reduced the maintenance. More over- it has brought down the inventory of spares. 3. Due to better heat transfer and higher exposure due to the internals the quality o-f the product K>. FUTURE DEVELOPMENT PLAN Encouraged by the performance o-f the in-hou=e modification of one of the convertor it has been decided to modify the remaining conuertors. However the following modifications are being considered for the future. 1. to provide mechanical seal of longer MTBF 2. to modify the internal flight and clamp rings and aux i11ar i es 3. to scale up for higher throughput by four to five times. Rotary Kiln type convertors are being preferred in comparison to'fluidised bed reactors for the production of uranium tetra fluoride. Our development work is in tune i'ii th the trend in the international scene. ^CKNOUJLEDGEMEHT The authors are vary grateful to Shri P.O. Sundaram, 3hr- i S.'J. MayeKsr foe their helpful suggestions and to "3hr i M.L.Sahu and Shri K.V. Ja i r-9 j for their help in making the mod i fi c a t ion s. i - 759 - C/« (•-' - eft£W"'S .K dr' VI.GA ., AfATVfS'"** f'.'tT A /t.s o'tnn le&t.. i i J<\ •"LU..I WEAK. WNG SOT";. 5PUT ffC*a WITH 6AI I. WEAK RIM; WW£R :IHCUP H6AD tocsins scsrw/ LOCWNS NUT / I5NOS HOUSNS • lp±wyf Art/PMC O/qS.MO. -UP/R/0& BEAMING HOUSING OIL SffAL \ .A . ILW.iS .VJCOUEL Tuae &td.6£ARIW6\ IKv I 8\\. 6lU» ' MNG BEARING ASSEMBLY HW TYPICHFQ T=««-3 ass. NO- UP/* /03 - 762 - MAGNESIO-THERMIC REDUCTION OF UF4 TO URANIUM METAL: PLANT OPERATING EXPERIENCE S.V. Mayekar, H. Singh, A.M. Meghal, K.S. Koppiker Uranium Extraction Division BARC, Trombay, Bombay - 85 Uranium Metal Plant has switched over from cal c i o-thermy to magnesi o-thermy -for production of uranium ingots. In this paper the plant experience -for magnesio-thermi c reduction is described. The operating parameters optimised include: green salt quality, magnesium quantity and quality and particle size. The effect of quality o-f lining has been discussed. Developmental work has also been carried out on use of graphite sleeves and on use of pelletised charge. Some experience in the machining of ingots for removal of surface slag is also discussed. Impurity problems occassioally encountered, have been investigated and results are discussed. Based on the experience gained, tfie specifications for operation have been laid down and areas for further improvement are identified. 1. INTRODUCTION Uranium Metal Plant has switched over from calcio-thermy to magnesio-thermy for the production of uranium ingots. Compared to calcium, magnesium is cheaper, consumption is less and can be easily purified to meet the stringent nuclear specifications. At U.M.P. magnesium granules obtained from Nuclear Fuel Complex, Hyderabad are used. Calcium needs to be imported. The process and equipment design for magnesio-thermy is however complex. Based on survey of international practice(l), trials were carried out at i).M.P.<2> and during plant expansion complete switch-over to magnesio-thermy was carried out. Currently 200 kg ingots are produced and plans for 300 kg ingots are now under consideration. II. MAGNESIO-THERMIC PROCESS Magnesio-thermic reduction is carried out in reactors, made of boiler quality mild-steel IS. 2002 Grade 2. Overall arrangement is shown in Figure I. The reactor is lined with magnesium fluoride powder by using a mould which is vibrated by an oscillating vibrator for 3 hours. Lining quality is confirmed by surface hardness and visual examination. Charge consists of 270 kg of 'green salt' (GS) and 43 kg of magnesium granules mixed in double cone blender. Strict quality control is exercised. Specifications are given in Table I. The charge is packed - 763 - inside lined reactor and cowered with 100-150 mm of magnesium -fluoride capping at the top. A plain reactor lid is tightened. No gasket is used. Sealed reactor is tested with argon gas to ensure a leakage rate of 150 lit/min. at 1 kgf/cm2 pressure. The tested reactor is loaded in electrically heated bogie hearth resistance furnace <100 kw) and heated gradually to 630-650°C over 16-18 hours and this final furnace temperature is maintained and reaction occurs with evolution of heat. This leads to rise in furnace temperature by 30-50*C. The furnace is then switched off, reactor is cooled inside furnace for 8 hour and then'cooled on bogie outside to room temperature. The lid is opened and contents discharged by inverting the reactor. The slag black is separated, crushed leached and ground for recycle as liner. The ingot is cleaned, pickled in acid, sampled and if required machined on outer surface for delivery to Atomic Fuels Division for fuel fabrication. HI. OPERATING EXPERIENCE 3.1 'Green Salt' Quality: Green salt (GS> contains upto 1.5% uranium oxides (called 'ammonium oxalate insoluble' or A0I>,.2.5% uranyl fluoride CUO^?) and traces of moisture <0.15O and HF acidity <0.1%>. Presence of H20 and HF leads to evolution of hydrogen by pre-reaction during heating at 1 cxoer temperatures itself <380-600°C>. Mg + H2O — MgO • + H2 Mg + 2HF — MgF2 + H2 U02F2 + H2 — . U02 + 2 HF UF4 + 2 H20 — U02 + 4 HF MgO + 2 HF MgF2 + H2 The net effect is formation of refractory magnesium fluoride film oh the surface of magnesium granules which retards the rate of reaction. Evolution of H2 and vaporisation of moisture can also lead to build-up of high pressure. Effect of uranyl' fluoride has been analysed. The heat of reaction of reduction is higher. U02F2 + 3 Mg — 2 MgO • MgF2 + U This leads to high temperatures while this yields better slag-metal separation, Jt also can create unsafe conditions due to pressure build-up and slag leakage. Low U02F2 content, on the other hand can lead to pre-mature reaction and poor quality of ingot. Optimum range is 1-2%, while the safety limit is 2.5%. Effect of AOI content is found to yield poor slag-metal separation probably due to incomplete reduction and high viscosity of oxide slag<3). A limit of 1.5% is fixed. Lower level's can be achieved only at the cost of reducing productivity of hydro-f1uorination furnaces which is undesirable. Storage of GS, especially in monsoon, requires special care. It rapidly picks-up moisture from humid air - 764 - ewers in cowered cans. This makes it necessary to adopt time-consuming and 1abour-intensiwe operation of moisture expulsion by heating under wacuum at 150°C -for 3 hours. 3.2 Magnesium Quality: Magnesium quality is -found to be a major wariable affecting operational ef f i c i ency. Particle size o-f granules is controlled to -10 + 40 mesh (Tyler). Presence o-f -fines can lead to rapid oxidation and poor slag-metal separation, as well as pressure build-up. Sur-face o-f granules is -found to wary -from shining bright to complete black. Black or oxidised magnesium yields poor results. The shape of granules is -found to wary -from rounded spherical granules to -flat sheets. This leads to wariation in process ewen for same batch o-f magnesium as shape is not constant. In addition it is -found that imported magnesium chips obtained by low-temperature mechanical chopping yield better results. The slag-metal separation is better, wariation is less, chipping weight is 1ower. Experience shows it is necessary to characterise magnesium surface and shape for better process control. 3.3 Maanesium Quality; Inwestigations were carried out by warying magnesium quantity as 40.5 kg (3.1% deficit), 42.5 kg (1.7% excess), 43 kq (2.9% excess), 43.5 kg (4. IX excess) and 44 kg (5.3% excess) for 270 kg of green salt. Stoichiometrical 1 y 41.8 kg of magnesium is required on 100%. UF4 basis. It was found that in all cases where magnesium was in deficit <43 kg.), the reduction to metal was incomplete and ingot consolidation was wery poor. Most of the metal was in fine: granules form which spontaneously caught fire on discharging. Similarly with 42.5 kg trials both recowery and consolidation were rather poor in ower half the batches. Satisfactory results were obtained with 43 kg and 43.5 kg. With 44 kg trials the frequency of slag spillage was higher and in one case eweh 'blow-out' occurred, possibly due to;high magnesium pressure. Based on the contradictory requirements of ingot quality and yield on one hand and unsafe blow-out in other, a compromise walue of 43 kg has been fixed. 3.4 Li ni no Quali ty: Quality of lining has been monitored by mould hardness tester used in foundry industry. It is found that a minimum surface hardness of 55 at bottom ingot portion is necessary for good results. Correspondingly the hardness at top is 45-70, higher walue being due to direction of transmission of wibrat ion from top. Two types of lining material hawe been tested - the unprocessed slag as discharged and ground, and the processed (chemically leached) slag. The former giwes better lining. Similar results are reported in US practice (4). However, the unprocessed slag contains higher amounts of impurities (iron) which lead to ingot contamination. As a compromise, 1:1.5 mixture of the two types of material is found to giwe satisfactory lining. - 765 - Vibration measurements during lining operation have been carried out. With the probe inserted in the powder mid-way amplitude" readings in the range of 19-22 mm and velocity in the range of 50-58 mm/second obtained. When there was a spring bolt failure the reading was 15 m and 62 mm/sec. 3.5 Graphite Liner: Graphite is the only material suitable for use with molten uranium without contamination*. While magnesium fluoride melts at liner—metal interface and leads to surface contamination, graphite yields a clean ingot^ surface. However, the thermal conduct i v i t:* is higher and backing refractory is necessary. Trials on dozen batches have been done by inserting a 10 mm thick cylinderical graphite sleeve inside pre-.formed MgF2 liner. In all cases ingot surface was clean, no fins were formed and the ingot could be melted without any need for surface machining. The carbon level increased from 75-100 ppm to 150ppm, which is however acceptable. The main disadvantage is breakage of graphite liner when discharged and its high fabrication cost. 3.6 Ingot Purity; Only in stray cases ingots obtained have been found to be impure. Boron impurity when found has been probably due to contamination from refractory liner. A limit of 5 ppm B in liner yields contamination within acceptable limits. Cadmium impurity often originates in magnesium or due to external contamination from plant scrappings. Iron impurity originates from magnesium', green salt and liner.' Other metallic impurities (Ni , Cr etc) derive from green salt and reflect state of corrossion in hydrof 1 uor i nat i on .. 3.7 lnoot Surface Quality; Majority of the ingots have a fairly regular. shape with some surface slag, a few ingots develop fins or highly irregular pan-cake shape. This occurs when lining failure occurs either due to soft lining or due to rough handling of lined/charged reactor. Ingots of irregular shape can not be mechanically cut for melting. High pressure water-jet cutting and plasma cutting are under investigation. Meanwhile strict supervisory control is exercised and frequency of irregular ingots has been reduced drast i cal ly. 3.8 Inoot Mach i n i no; About a third of the ingots produced are judged by A.F.D. to contain unacceptabl'y high level of surface slag or have a porous surface. These ingots have been machined in a horizontal lathe with copious quantity of coolant (for safety) at slow speed. Machining removes 15-25X of metal as turnings which need careful storage, incineration and recycle to dissolution stage. As an alternative, re-melting of ingots under positive argon pressure has been sugested for improved slag-metal separati on. i - 766 - 3.9 Operational Safety: Magnes i o-thermy is an inherently bomb reduction under pressure since the reaction temperature is higher than vaporisation temperature of magnesium. To avoid pressure build-up, ungasketed reactors with pre-determined leakage rate are used. The reactor vessels used are hydraulical1y tested periodically at 20 atm. The reactor -flange and lid are regularly examined, and if damaged, then they are repaired be-fore use. Bolts used have been tested in Phy. Met. Division, and -found to hold upto 8 atm. They serve as weak link for failure, as a substitute for pressure-rupture device. A few cases of lining failure with molten metal leakage through reactor have been observed. As a safety measure, a graphite crucible has been fitted inside reactor at bottom. Finally as a safety step, no entry into furnace area is allowed when the -Final temperature is reached. IK>. DEVELOPMENTAL ACTIVITIES 4.1 Pelletised UF4-Mo Charge: Recent developments in uranium magnesio-thermy include pelletised reactant charge(5). Six trials have been carried out at U.M.P. after pelletising the GS-Mg mixture in a manual hydraulic press to obtain 65mm x 50mm pellets or in pneumatic automatic press to obtain 20mm x 15mm pellets, with these high density pellets, the pre-heating time could be reduced by 50X, and the slag-metal separation was found to be better. However, regular use in production scale needs an automatic system for charge preparation, pelletising with lubrication, pellet baking for removal of lubricant, and charging into reactor wi th hard 1 i ner. 4.2 Firing Indication by "^-Monitor: Current method of detecting completion of reaction involves measurement of small temperature rise. Often due to thermo-couple positioning this is not significant, and indication is ambigous. As an alternative a i*-monitor is under testing by Electronics Division. This is based on change in T-attenuation consequent to density change from UF4 <7 gm/cc) to uranium (l?gm/cc>. 4.3 Improving Slag-Metal Separation: Slag-metal separation is reported to be better<4> with top-loading bottom heating furnaces of well-type. These also'enable safer operation with flange outside the furnaces. Design of such furnaces is being taken-up as a part of augmentation programme. In addition, efforts have been initiated with Reactor Engineering Division for computer simulation of thermal profiles with a view to process optimisation. Efforts are also b*ing made for in-situ temperature and pressure measurements with the assistance of Instrumentation Section. - 767 - ACKNOWLEDGEMENTS The authors are thankful to Shri S. Sen, Associate Director, CHEG -for his keen interest. Thanks are also due to colleagues in Central Workshop, Atomic Fuel Division, Metallurgy Division, Electronics Division and R.E.D. all o-f B.A.R.C. for co-operation. Assistance of Spectroscopy Division -for purity analysis is acknowledged. REFERENCES 1. T.K.S. Murthy and L.M. Mahajan 'Literature Survey Report on Magnesio-Thermic-Reduction o-f Uranium', Internal Repor t, /U. E.D. , B.A.R.C. 1982 2. K. Kumar and A.M. Meghal "Uranium Metal Production by Metallothermic Reduction at Uranium Metal Plant, BARC". Proceedings syrup on 'MetalIothermic Processes in Metal and Alloy Extraction', Nagpur, Dec. 83 p 200-215 3. CD. Harrington and A.E. Ruehle Uranium Production Technology D. Van Nostrand Company Inc. Princeton, NJ, 1959 4.. 6.6. Br i ggs 'Reduction o-f UF4 to Metal', in Metallurgy and Technology of Uranium and Uranium Alloys, M3, American Society of Metals, 1982. 5. H. Page "Conversion of Uranium Ore Concentrates to Nuclear Fuel B.N.F.L.' IAEA Techical Meeting on Advances in Uranium Refining and conversion, Vienna 1986 - 768 - Table I SPECIFICATIONS OF MATERIALS FDR M.T.R. GREEN SALT utx Moi sture < 0.1 Free Ac i di ty < O.l U02F2 < 2.5 A.O.I. < 1 .5 Tap density 2.5 -2.6 gm/cc Size Approx .95'/. -325 mesh MAGNESIUM ppm Chemical B < 0.5 Cd < 0.5 Fe < 200 Mn < 25 CI < 50 RE < 1 Physical mesh wtX + 10 0 10/20 45-75 20/40 25-55 - 40 0 LINING MATERIAL Moi sture < o.2y. L.O.I. < 0.2X. U < 0.5X Fe < 1000 pprn Tap densi ty 2.1 - 2.2 gm/cc mesh wtX. + 60 4-10 60/100 12-20 100/200 20-30 200/325 15-20 -325 40-50 - 769 - ^ ^f!l= ,^W"*^"« « P • "4 . •- •*>-" i -W-* ill |'WI inui I mi n •llJPIMrii I •.•! • II . *»»•.» mm \ Ft-p—Jr**1! " \***— ~-p-i—• .|-*»-t"— 'pT"'•""•*!" ••* —1 * - X * f— *-— p—J- —i—•*• •••|».i.t^^-|—. 1 I"-'*' 'I ' ••!•-! CAPflNO MAftKIAt •J>* I...J HE Arm a J\j£±±2*. KEACTO/j tlWINC. . :?fcu;-'.i- He AT IN a Hi* CUMtHTS ^P5 "I • ' !"vsl" *'M l^'r'P'H • 'I ''fff'B?'!'!"—V&~ i jpiii'" li' fW' 'ii'nM 'i.ii ijy|»T .'j, • I';'j>n'i':-.1—I—\r~ 1,8' lWw>"UiwuiJiniU!Uii4w),i|»|ii| j I It--.- Iiiiii ii n j^yiip.).^ .,.„.UJ. | .i|l|,,,||^„, .., ninm|l, ,i. jJ TMAGNESIO^THERMIC-REDUCTION - 770 - RECOVERY OF URANIUM AND LINING MATERIAL FROM MAGNESIUM FLUORIDE SLAG AT U.M.P. P.K. Bandyopadhyay, H.Singh, B.M. Shadakshari and A.M. Meghal Uranium Extraction Division, BARC, Bombay-85 Uranium metal is produced at U.M.P. by magnesi o-thermi c-reduct i on of uranium tetraf1uoride. During this process magnesium -fluoride slag is produced along with uranium • ingot. The slag contains 7.—AV. uranium which is processed to recover uranium and -finely ground refractory lining material. Process details and operational experience of slag processing are described in this paper. Processing includes both dry size reduction and wet operation. Size-reduction involves multi-stage crushing and screening. Wet operation, includes leaching with dilute nitric acid and uranium recovery -from leach liquor by solvent extraction. Leached cake is dried and ground to very fine size for use as lining material in magnesio-thermy. An outline of the scheme to treat large quantities of slag on a regular basis is presented. I. INTRODUCTION Magnesium fluoride slag is generated during the magnesio-thermic reduction UF4 + 2 Mg > U + 2 MgF2 Stoichiometrical1y 0.546 tonne of slag is generated for 1 tonne of uranium. .The slag contai ns. 2-4J< of uranium, traces of magnesium and is contaminated with iron. The slag has to be processed for recovery of uranium, and magnesium fluoride for re-use as refractory liner .in MTR reactors. Several process have been reported for slag processing I I . PROCESS FLOWSHEET The process flowsheet is shown schematically in Figure I. It includes both dry operations and wet operations. 2.1 Dry Operations: These include three stage size reduction by crushing and vibratory screening. The slag as obtained from MTR contains both massive fused slag block as well as fines from refractory lining. The fine material called 'soft lining material' SLM is separated for re-use in MTR directly. The coarse material is subjected to primary size reduction to reduce lump size to 100 mm. Secondary crushing is carried out in a single toggle jaw crusher to - 771 - reduce top size to 25 mm. At this stage preliminary screening is carried out to remove hard uranium-rich slag pieces or even unconsolidated metallic uranium. This material is taken -for leaching by percolation. The —25 mm material is ground in a 1000 rpm Hammer mill to reduce particle size to 100% -60 mesh, -for wet processing. As an alternative to Hammer mill, we t grinding in a bal1 mill was tested but was -found to generate excessive -fines which created problems in subsequent wet operation. 2.2 Ulet Operations: Based on the laboratory studies <4> <5), a nitric acid leaching process -followed by solvent extraction has been adopted -for plant scale. As shown in Figure 1, this involves leaching of the -fine material with 2.3 N nitric acid at 60 - 70°C for 3-4 hrs. The slurry is filtered in a vacuum nutsche filter. Three stage counter-current washing is carried out with water for complete removal of nitrate from cake. The washed cake is dried in electric oven and ground in a 3000 rpm Hammer mill to 50 - 55% -325 mesh for re-use as 1ining material. The coarse material is processed by percolation leaching. A bed of 0.5 tonne is made with graded slag pieces and dilute acid is allowed to percolate by gravity over-night. Uranium content of percolated acid is monitored, and when it is -found to be low, then water washing is given. The final cake is then disposed of as low-active solid waste through Waste Management Division. Since more MgF2 than is required gets generated in MTR, only about half is processed by agitation leaching while balance requires disposl by percolation leaching. Uranium is recovered from leach solution by solvent ex traction(6>. A 5 stage mixer-settler battery is used for extraction with 32X tri-n-butyl phosphate in kerosene at aqueous/organic ratio of 4-5. The ;oaded extract is mixed with extract'from main unit in the r»tio of 1 to 15. for scrubbing and strippig. The extraction raf-finate is partially re-used for leaching, while the bulk is neutralised with calcined magnesia for disposal. III. OPERATING EXPERIENCE 3.1 Crush i no; During crushing it is -found that hard slag pieces (containing upto 70'/.> and metallic uranium pieces lead to severe wear and tear. For tertiary crushing this problem is found to be more when Sturtveant disc grinder or Hammer Mill or Ball Mill are used. These machines do not permit passage of uncrushable pieces. Trials on roll-crusher have been successful. Dust generation during crushing can be suppressed by water sprinkling. It is essential to incorporate intermediate screening for removal of uncrushable pieces. 3.2 Leachino St Filteration; During leaching, the wear and tear of SS impeller is found to be high. Coatings such as of rubber and plastic materials, do not have stability in the aggressive conditions. Besides coatings can lead to cake- contamination. Steel impellers only cause - 772 - contamination in solution and are tolerable. Abrasive wear can be minimised by control over particle size. Based on experience, 60 mesh is -found to be satisfactory. Too -fine size, leads to problems in -f i 1 terat i on. Solution volumes needs careful control. Pulp density of 60-65'/. in leaching and counter-current cake washing are essential to avoid generation o-f large amounts o-f dilute solution. During •fi Iteration it is necessary to have high vacuum (600 mm Hg) •for high fi Iteration rate. 3.3 Typical Stream Analysis; Typical analysis of filtrate, wash, cake and extraction streams is given in Table I. The filtrate has a free acidity of less than 2 N. This needs to be increased to 2.5 N by acid addition before extraction. Fluoride level is high. Wash contains low uranium and can be recycled. The processed cake has K>^ry low uranium, iron and nitrate - an indicator of good washing. The material is coarse and non-hygroscopic, with moisture level being low. Extraction raffinate contains very low uranium and almost all the fluoride. Extract contains only traces of fluoride. 3.4 Material of Construction; Due to high fluoride concentration, the material of construction is a problem. Leaching vessel has severe conditions of higher temperature as well as abrasion in addition. In principal, nickel could be used. But its cost is very high. Stainless steel with generous corrosion allowance has been selected. The- only contamination it causes is in solution which is tolerated. Nutsche filter also is made of stainless steel. Solution storage tanks lined with polypropylene have been found to be satisfactory. Mixer-settlers made of perspex were found to leak due to TBP. Steel ; =, ruled out since slight leakage can lead to back-mixing across stages and loss in efficiency. Polypropylene has been • found to be satisfactory. But it has low mechanical strength and needs careful handling. PVDF samples have been tested and found suitable. Magnetic drive seal-less pumps of polypropylene have been used satisfactorily for 3 years. Their discharge head is low andare suitable for transfer application. Feed metering by gravity is best alternative. Pipe-lines made of polyprolylene tend to warp and crack. Teflon lined pipes are suitable but costly. Stainless steel pipes can be used for transfer application provided lay-out ensures complete draining after the operation. 3.5 Extraction efficiency; High A/0 ratio is needed to minimise solvent flow. However, this can lead to loss in efficiency of extraction if uranium loading of solvent exceeds 70 gpl and free acidity of feed is below 2 N. A ratio of 4 to 5 is found optimum. A recycle type mixer-settler design is adopted for maintaining A/0 ratio of 1:1.5 in mixer for better mass transfer. It is found by experience that when lean solvent contains more than 2.5 g U/l, raffinate concentration exceeds 0.5 g U/l. Hence efficient solvent processing is needed. Solvent 'age' affects performance, with used solvent giving lower capacity due to lower phase-separation rate. - 773 - 3.6 Li ner Pur i ty; Occass! onal 1 y processed slag is -found to be contaminated with boron. This generally occurs when •floor sweepings including dust and powder of bricks from furnaces are collected. Leaching does not remove boron. Hence such materials need to be rejected a-fter uranium recovery. Occasionally slag may contain higher uranium. This needs higher acid -for complete dissolution. Iron and nitrate impurities are controlled by wash operation. 3.7 Fine Gr i ndi no; Grinding of dried processed cake is being carried out in a high speed hammer mill provided with bucket elevator, screw conveyor and dust collector. The dust load studies by health physicist indicate a level of ? mg/m3, below the TLV of 10 mg/m3. However, in general, the dust generation is high. Hammer Mill is a high speed blower type machine where dust -flying is inevitable. Conventional ball milling is unsuitable for such -fine grinding. Trials on -fluid energy mill show high power consumption with efficient closed circuit operation at high capacity are needed. Our requirement is low, and tests on vibratory twin-tube machine have shown dust—free grinding can be e-f-f i c i en 11 y carried out. Belt type bucket elevators are -found to su-ffer from problem of belt slippage and wear. Gravity -fed hoppers do not yield uniform -feed-rate and vibratory -feeders are needed. Dust collection by mechanically shaken filter bag* is unsatisfactory. Wet scrubbing is essential. ' Operating experience at- U.M.P. has shown that existing facilities for processing slag are inadequate and nearly. 60 tonnes, of slag has accumulated. Technical details for a 100 tonnes/year slag processing plant have been worked out. Primary double toggle jaw crusher, secondary single toggle jaw crusher, tertiary roll crusher and double deck vibratory screens have been procured. Specifications for wet section have been made. Twin -tube mill is under procurement. All these machinery is proposed to be installed in the Phosphorous Section of Chemical Engineering Division. As an interim measure crushing facilities of Ore Dressing Section, BARC at Hyderabad are being used to meet the requirements. ACKNOWLEDGEMENTS The authors are grateful to Shri T.K.S. Murthy, ex-Director, Chemical Engineering Group, BARC under whose guidance the process operations were standardised. - 774 - REFERENCES 1. CD. Harrington and A.E. Ruchle Uranium Production Technology D. Can Nostrand Company Inc. Princeton, N.J. 195? 2. UI.E. Show, UI.C. Manser, R.6. Gier and S.H. Smiley "Natural, slightly enriched and depleted uranium chemistry', ch 8 in, 'Reactor Handbok' C2nd edition) Vol. 2, Fuel Reprocessing, S.M. Stoller and R.B. Richards Table I. TYPICAL STREAM COMPOSITIONS Raw SI act 1 limp size Upto 500 mm U : 0.5 — 4X Fe upto 2500 ppm Mg i 0.1—0.4/. Crushed SIaa mesh wtx + 60 6 60/100 15 100/150 23 150/200 19 200/270 30 -270 7 Slag Processed Cake (SPC) U 0.03 - 0.1/. Fe 100 - 500 ppm H20 8 - 12X N03 100 - 250 ppm Ground MQF2 1iner U < 0.5/. Fe 500 - 1000 ppm . B < 5ppm H20 0.04 - 0.V/. LOI 0.05 - 0.2"/ Slao Processing Filtrate (SPF) U 7-20 g/1 Fe 0.5-1 g/1 F 1 - 3 g/1 FA 1.5 - 2 N Slao Processing Ulash U : 0.5-4 g/1 FA : 0.05 - 0.5 N Extraction Ra-f-finate 0.05 - 0.5 g/1 F 1 - 3 g/1 FA 1.9 - 2.5 N Extract U 25 - 60 g/l F 0.05 - 0.1 g/1 FA 0.1 - 0.2 N Pur i ty nuclear pure - 776 - Figure. 1. SCHEMATIC PROCESS FLOWSHEET FOR SLAG PROCESSING SLAG FROM M.TR. 1 S.L.M. SEPARATION - _ SOFT LINING MATERIAL (SLM) TO MTD " 250 mm. LUMP BREAKING 110 0 mm. JAW CRUSHER 1 25 mm. SCREENING-1 • • FINE CRUSHING i SCREENING - 2 -60 Mesh 2 •3N ACID ' 1 1 | 'f AGITATION «i • 1 * PERCOLATION LEACHING • LEACHING 1 • , L FILTERATION S r 1 COUNTER CURRENT WASHING WASHING VAMCJ 4 1 1 DRYING UUAKbfc bLA(j (WASTE TOWMD) FINE GRINDING 1 -LININING MATERIAL J * IUMTR ' I EXTRACTION - RAFFINAE t WAQTC\ M " U-RECOVRY US/PNC Drg.N o-UP/R/ 04 - 777 - FUTURE TRENDS IN THE PROCESSING OF URANIUM SLAG GENERATED DURING PRODUCTION OF URANIUM METAX Keshava Chandra, Maheah Singh, H. Singh, A.M. Meghal, K.S. Koppiker and S. Sen Chemical Engineering Group Bhabha Atomic Research Centre, Trombay,Bonibay-85 The production of uranium ingot by magnesiothermic reduction generates uranium slag (U-alag). This U-slag is in the form of hard fused blocks with partly fused soft lining material. This contains 2 to % of uranium metal, 400 to 1500 ppm of iron and trace amounts of other impurities. Uranium and magnesium fluoride lining material from U-slag are recovered in uranium metal plant (UMP). Present process as followed at UMP is described in brief to identify the possible areas for future improvement' and innovations in the existing U-slag processing . technology. These areas pertain to improvements of system/equipment to minimise labour, remotisation of operations to reduce radiation exposure to the workers, reuse of nitric acid to reduce effluent generation etc. Zero-leak-oum-double containment concept for pollution control is described. An expanded facility for processing 120 Te of U-slag annually has been conceptualised. - 778 - 1.0 INTRODUCTION Uranium Metal Plant (UMP) by metallothermic reduction of UF4 with Mg metal produces U metal and MgF2 radioactive slag. The by product slag contains some U (2-3#) and lining material in loose powder form. Both of these components are recovered by processing this slag. The present facility for uranium slag treatment (USTF) at UMP is progressively becoming inadequate to cater the future demands. There is a need to have additional USTP, there fore it is obvious to give a fresh look for the improvement and upgrading of the existing technology before new facility is created. This paper scans present practice, future trends and possible areas for modifications. 2.0 PRESENT PRACTICE OF URANIUM SLAG PROCESSING AT UMP The flowsheet for uranium slag processing as followed by UMP ia given in Pig. 1. This process has two stages, Dry operations and Wet operations* Dry operations include grinding and screening of slag to -60 mesh and finer grinding of leached slag to -525.mesh after drying. Wet operations include leaching, solvent, extraction and concentration of uranium solutions. Existing USTP at UMP is integrated well for the management of safety,pollution and waste streams. 5.0 UPGRADING OP URANIUM SLAG TREATMENT TECHNOLOGY The existing technology was scanned for upgrading in important areas in present day oontext. Future trends in this technology for modification pertain to changes in systems/equipments for providing minimum exposure of radioactivity to minimum labour, inbuilt safety, cleaner (pollution free) environment inside aa well as outside - 779 - the facility and minimum discharge of waste streams. Modern concepts of 'Zero Leakage* of pollutants by design and discharge of 'As Low Aa Reasonably Achievable* (ALABA) waste streams are considered. 3»1 Modifications of Systems in the Process Dry operations in USTP consists of mainly crushing,grinding, screening and transportation of radioactive slag. It also includes transfer, drying and fine grinding (-325 mesh) of leached slag. These operations involve use of manual labour, generation of dust containing uranium and magnesium fluoride as pollutants. Wet operations in USTF- include acid leaching, solvent extraction and concentration of aqueous uranium solutions* Generation of HOX fumes during leaching, presence of fluorides in corrosive nitric acid, use of organic solvent and low values of uranium in raffinate are some of the problems which need review. Possible suggestions for upgrading the systems are given in Table-I for dry operations and Table-II for wet operations. 3.2 Management of Solid and -Effluent Streams Possibility of minimum discharge of pollutants in solid and effluent stream with 'As, Low As Seasonably Achievable'(ALAHA) has been studied. Reuse of radioactive effluents of the plant and regeneration of inactive pollutants such as fluoride are considered as given in Table-Ill. Conceptual scheme for zero discharge of solid and liquid streams containing pollutants in USTP is shown in Fig.2.' - 780 - 3.5 Pollution Control inside and outside the. Plant Management of radioactive uranium and inactive magnesium fluoride pollutants inside as well as outside the plant is studied in detail. Concept of 'Zero Leakage' by- design is blended with Bemi-remote operations (Table-I and Table-Il). In addition, pollutants are contained at the source of generation (syatemwise) separately in an enclosure. Polluted air from each enclosure is passed through cyclone separator and filter before discharge to the intake of ventillation system. Enclosed operational areas are appropriately ventillated through filter banks to atmosphere. Suggestions are given in Table-IV. Air. curtains can be deployed to seggregate active and inactive areas* 3.4 Safety Incorporation of safety through various safety committees right from inoeption Btage to operational stage is a desirable step. Training of staff is also considered under safety. Some of the suggestions for radiological and industrial safety are given in Table-V. 4.0 NEW URANIUM SLAG TREATMENT PACILITY(USTP) - A CONCEPT Based on the experience of existing UMP(BARC) and some of the suggestions put forward in this paper, a new uranium slag treatment facility (USTF) is conceptualised. Reuse of the solid and liquid waste streams is not included in new USTP. To meet the future demands, this facility is designed to treat 120Te of slag annually. Some of the - 781 - salient features of this facility are given in Table-VI. A conceptualised facility incorporating suggestions made in the paper may be a state of art technology in the treatment of slag for the recovery of uranium and lining material. ACKNOWLEDGEMENT The authors are thankful to Shri P.K. Bandopadhyay (UMP, BAHC) for hi8 valuable suggestions. - 782 - Table - I Suggested modifications in systems/aquipments of dry operations in uranium slag treatment facility (USTP) Sr. Systems Present practice Suggestions Remarks Ho. 1. Discharge On plain plateform Oyer enclosed grating Easy and of U-reactoi safe separa contents tion of powder and slag lumps Slag Ordinary 2001 Better designed containers for transport, Sturdy storage/ barrels with lid storage and repeated use container transport sealing with easy containers handling Crushing Manual breaking Mechanised lump breaking, single stage Dust free and of lumps, two jaw crushing, roll crushing and inter mechanised screening stage jaw stage screenings in a closed systems with and semi- cruahings, roll .ventillated area. Double containment of remotised crushing and each system separately ventillated with operation. interstage cleaning system. Semi-remote and semi Minimum screenings are automatic facilities for operation; labour with done in a Vacuum cleaning of surfaces (walls etc.) minimum ventillated area to remove charged dust exposure. Clean envi ronment inside the plant. Pine grinding Hammer mill Twin-tube mill grinding Less maintenance/ of liner grinding repair and dust material free semiautoma (minus 325 tic closed mesh} operation. - 783 - Table - ii suggested modifications in systems/equipments of wet operations in uranium slag treatnaat £acllity(usTF). sr. systene/equipnents Present pra.ct-.lce Suggestions Remarks Ho. 1.(a) Freleaehing slag Manual feeding of Pneumatic feeding Saving of labour and powder delivery system powder dust free envirotoaent lb) xgit *tion leaching Ventlllatlon without Ventlllatlon with No release. of gases •crabber scrubber for NOX In leaching area and to gases outside environment (c) Percolation leaching Acid Is percolated to Pug mixer Instead Increased uranium re- recover part of ura of percolator covery resulting into nium leaving behind ouch leas active slag. low active slag Filtration of leached Frew neutsche filters. The cake to be Manual handling of active •lurry and delivery or the slag Is manually sent directly to cake Is minimised. washed slag to calci transferred to trays, furnace In a nation furnace. which are loaded to tucket by meche- furnace manually, nlsed system 3. solvent extraction Mixer - settler made Mixer - settler More sturdy with higher of perspex. made of stainless capacity steel lined with PTFK 4. Concentration of stri 30 - 40 g U/l 45 - 50 g u/l by Necessary requirement pped uranluo solution evaporation for efflcaont ADO Pre before AOU precipitation cipitation 5. Llould level measurements Manually levels Liquid level Ind Minimise labour and and level controls in are measured icators and maximise accuracy of tanks. controllers liquid management. - 784 - gable - III Suggested Bsnageaent of effluent and solid streams or uranium slag treatment facility (USTP) (A.) Liquid Streams Sr.Ho. System Present practice Suggestions Remarks 1. Baffinate stream Keutralisation with Seleotive removal Minimise XgO/HaGB, filtration of activity in effluent through filter press, solid phase prior quantity, filtrate and oak* to to disposal or handling and MB*. removal of Pe and labour. reuBe. 2. imm. diuranate ppt. Direct disposal if U Be cycle for Minimise washings content is low washing after effluent (about 100 ppm) adjustment of pH quantity. (B) Solid Streaae 3. leached slag froii Disposal to WHD* Conversion to Kg Bause or Kg percolation tank and HP and HP. taste generation is minimum, ^. Part of leached -do- -do- -do- slag from agitation tank 5, Cake generated due -dc— Heutralisation ft Ho cake to neutralisation filter press ope formation of raffinate ration deleted. - 785 - table - IV" Suggested aanageaent of pollutants In uranium slog treataent faollity(USIP) (A) iBBlde the Plant Sr.Ho. Systeas PreBent practioe Suggest lona Beaarke 1. Dry operations Manually operated Double containment provision Ktnlfiising systens Inelde for each systea with semi- operators' processing rooa automatia operations ezposurs proTlded with controlled Xros outside. to dusty rentillatlon General area Tentillation operations. with filter bank. Contained Ho dust space is separately rentilla- leakage in ted through oyolone separator, the prooe- and filter. ssing area. 2. Vet operations leaching,filtration Sash operational unit to be Clean working extraction, preoi- proTided with fuse environment pltation,vaahing extraction faoility in operations are addition to rentlllatlon of done, in a rentl- the area. llatad area. (B) Outside the Plant 3. Solid and liquid Majc* part of these Zero discharge scheme as Better streans •txeaaa fo to WD* depioted in Pig. 2. pollution containing after preliminary oontrol pollutants treatment. Yantillation General area General area ventlllation Zero leakage oyaten *entillatlon with with filter tajik and to the filter tank. scrubber. surroundings. •WD - Vasts Kanageaent Division, BABO. - 786 - table - V Salient considerations for safety In uranlun.slag treatnent faoillty(U3T7) (i) Radiological Safety (i) Double cont:al-2Mnt (ii) Air curtain barriers between inactiTe and aotlTe areas (ill) Beaotisation of operations controlled froa separate rooa (IT) Llconoed entry into plant (B) Industrial Safety (I) Flaae proof components in an enclosed area for solvent eztraotion aretes (II) Corrosion resistant lining of equipments (ill) Aoid and abbrasive reeistant painting/lining of floors, walls ar,l supporting struoturea (IT) layout of plant & equipments for slnlaua handling of solids (T) SealantOBatlc/reaote operations controlled iron a separate roon (n) licenced entry into plant (T11) Xffielent training progreaaie .of staff (rill) Unlaws inventory of radio aotive and hazardous materials - 787 - Table - YI Soae ot the salient features of conceptualised uranium slag treatment facility (EST?) 1. Design Concepts t Seniautonatio ramotleed opaxationa for mlnlaux labour and. ninisua exposure | lero-lealcagetfdust/funeB by design ; collection of pollutant at the aour&» of feneration by double containment for batter pollution oontrol in inside and outside the facility}'is low as reasonably achievable' (ALAEA) discharge of effluents. 2. Oapaoity 120 Xe of slag per year 3. Area (a) Building 500 M2 (b) Open 200 M2 4. Cost of equipments b. 60.00 lakhs as Installed 5. Input (yearly) (a) Slag 120 Se (b) Hitrlo aoid $0,000 1 (50-55*) (o) fate? 2,10,000 1 6. Output (yearly) (a) Uraniua 3«0 Te (b) Lining materiel 60.0 Te 7. Vaate Streams (yearly) (a) Sollda 60.0 Te (oontaining 0.05 to 0.2* U) (b) Liquids 2,40,000 1 (oontaining 0.05 to 0.5* U) - 788 RAW. SLAG 110T/MONTH) PRIMARY SCREENING MANUAL BREAKING' :sc PRIMARY JAW CRUSHER | SECONDAR£Y =JAW CRUSlUn \ VIBRO SCREEN lOimii 1 ROLL CRUSHER MAKEUP +10 intsh ACID | DOUBLE PECK SCREEN 1- " * S0/10ina»h -GOnwsli 2M 1 tn AGITATION LEACHING PERCOLATION LEACHING UNO, FILTRATION CAKE | »[WASHIN(j WATER WAIER CAKE WASHING . 2,3 V/ASHINuG CAKF. CAKE I RECYCLE THP/KEROSENE f t tWFINATE SOLVENT EXTRACTION MgO/HoOH DEMORALISED WA r L!L-» ?rRippiNG NEUTRALISATION -»j PRECIPITATION (FILTRATION) •JFILFILTI H AU FILTRATION EFFLUENT DISPOSAL -*• WMD WATER JL CAKE WASHING I DRYING -CALCN. DRUMMING ORYER VJ03 TO UMP T LINER SLAG WASTE SLAG FINE GRINDINGsl— > TO UMP TO WMD FIG.l, URANIUM SLAG PROCESS FLOWSHEET - 789 - WATER 48,000lit. HN03 (90%) i L Mg METAL FOR REUSE 60 T« MgF2 CONVERSION TO 0-2 % U Me AND HF HF FOR REUSE URANIUM SLAG TREATMENT 120 T« MgF2 168000 llt.HN03< l«T H) | SLAO FACILITY ->1 4 - 10 g/i Mg j (USTF) I I - 3 g/l F 4- -H REMOVAL OP >FOR RESALE/ 009-0'9g/1 U I P AND U REUSE 500-lOOOppmF* j! OR -»{ REMOVAL OF 60 T« LINING MATERIAL Ft •FOR REUSE BASIS *. 120 Ta OF SLAQ PER YEAR. FIG. 2 CONCEPTUAL SCHEME FOR ZERO- DISCHARGE OF SOLID AND LIQUID STREAMS CONTAINING POLLUTANTS IN USTF - 790 - QUALITY ASSURANCE DURING URANIUM METAL PRODUCTION AT U.M.P. V.N. Krishnan, R.D. Shukla and M.S. Vi sweswara i ah Uranium Extraction Division, B.A.R.C. Uranium concentrates, supplied by UCIL., Jaduguda is processed at Uranium metal Plant through several steps comprising of dissolution, solvent extraction purification, precipitation, calcination,reduction , hydrof1uorination and magnesio-thermic reduction to uranium metal. At each of these steps, process control and purity control are absolutely essential. The outline of the Quality Assurance programme followed at UMP. and the important innovations / improvements introduced in the recent years are described. Most of the analyses of the impurities in nuclear grade uranium are now carried out at UMP. itself using Atomic absorption spectrophotometry At the Uranium Metal Plant in Trocnbay, magnesium diuranate (MDU) is processed to produce uranium metal, used as fuel in research reactors at Trombay, namely CIRUS and DHRUVA. Typical analysis of the MDU is given in Annexure I and the process flow sheet followed in UMP is shown in Figure I. It is essential to maintain close control over the process parameters and quality of ' the process material, as well ,as that of the metal produced. This is ensured by analysing samples, drawn from different stages of the process, to confirm that the process conditions are as per requirement. This is shown in Annexure II. Quality Assurance Programme at UMP involves analysis of a targe number of samples by Gravimetric, Volumetric, Spectrophotometric techniques and carrying out a few physical tests. However in the case of final product, Uranium metal, and a few process samples, the test for purity has to be carried out by Emission Spectrometric technique at Spectroscopy Division, since the tolerance limits for these impurities like boron, cadmium and rare earths are in sub ppm levels. Annexure III shows the specification for nuclear grade uranium metal . Recently, the process control Iaboratary has been equipped with Atomic Absorption Spectrophotometer to analyse trace metal impurities in samples. Atomic absorption spectrophotometry, with its selectivity and required - 791 - sensitivity, provides a quick method -for the analysis o-f all impurities, other than boron and rare earths. Analytical procedures -for the determination o-f all these impurities have been standardised. In principle, the procedure involves the selective- removal o-f bulk uranium by solvent extraction with TBP in CCl^ and the determination o-f impurities in the aqueous phase directly by AAS technique. Annexure IV gives a typical analysis o-f' an uranium ii.got sample produced at UMP. On an average about 1000 samples, involving about 1500 determinations, are analysed by the Quality Assurance Section, UMP. in a month. ANNEXURE I TYPICAL ANALYSIS OF MDU CONCENTRATE FROM UCIL Moisture - ~ 40/C Analysis on dry basis Acid insolubles 1 .0 Y. Si 02 4.0 V. u3o8 75-77 V. CaO + MgO 9.0 V. Hal ides 0.02 "A Iron 0.45 V. p 205 0.005 V. so4 0.15 V. Boron 20 ppm Th02 20 ppm Rare Earths 600- 1000 ppm (Gd - 30ppm) - 792 - ANNEXURE II DISSOLUTION MDU concentrate - U, H20, Fe, Th, RE, Si02 Dissolved slurry - Total U Crude uranyl nitrate solut i on - U, Free acid Uranyl nitrate cake - U, H>0, CQMPLEXING Complex slurry EXTRACTION Uranyl nitrate -feed solution - U, FA, RE, Th, Si02 TBP- Kerosene - TBP, DBP Organic Extract - U, '/. saturation, Density, saturation densi ty Scrubbed Organic - U Scrub raf-finate- - U Aqueous ra-f-finate - U Lean solvent - U, */. TBP, saturation, saturaton densi ty Pure uranyl nitrate solution - U, Impur i t i es as spec i -f i ed. Uranyl nitrate -from dosing - U tank. ADU i\1trate - U Urani um Ox i de - Tap density, Surface area, RE, B,Ni,Fe - 793 - HYDROFLUORINATION Green salt MAGNESIOTHERMIC REDUCTION Green salt - H20, FA, U02F2, AOI, Tap density, Fe,Ni Mg - Impurities as per Speci-ficat ion Lining material SLAG PROCESSING Slag - U • Slag processed slurry - U and Fe in residue Slag proceesed filtrate - U, FA, F Slag processed cake - U, H20, Fe MISCELLANEOUS Waste disposal.solution - U - 794 - ANNEXURE III SPECIFICATION FOR NUCLEAR PURE URANIUM METAL Maximum in ppm on Uranium basis A1 urn in i urn ANNEXURE IV TYPICAL ANALYSIS OF AN URANIUM METAL INGOT BY A.A.S. ppm on Uraniun basis A1 umin i um (Al ) 35 Cadmium (Cd) <0.1 Cobalt (Co) <1 .0 Chromi um (CD 10 Iron (F©) 120 Magnesi um (Mg) 12 Manganese (Mn) 5 Nickel (Ni ) 65 - 795 - FLOWSHEET FOR URANIUM METAL PRODUCTION MAGNESIUM DIURANATE 1 1 1 NITRIC DISSOLUTION 1 ^ (YELLOW CAKE) 1 1 1 ACID 1 1 RAFFINATE SOLVENT EXTRACTION 1 FOR > —— —^ I 1 1 DISPOSAL 1 PURE URANYL NITRATE 1 1 1 AMMONIUM DIURANATE 1 1 PRECIPITATION 1 1 1 1 1 1 FILTRATE FIG. 1. 1 FILTRATION 1 FOR —> 1 1 DISPOSAL 1 ADU CAKE 1 1 1 CALCINATION 1 1 1 URANIUM TRIOXIDE 1 1 -J REDUCTION 1 1 1 1 *t 1 I AMMONIA 1 1 r CRACKER 1 N2 • H2— URANIUM DIOXIDE 1 1 1 1 > 1 ANHYDROUS HF HYDROFLUORINATI ON 1 J' 1 1 PROCESSED SLAG URANIUM TETRAFLUORIDE U/ 1 1 1 MgF2 1 1 1 CRUSHING & GRINDING IMAGNESIO-THERMIC REDU CTION 1 1 1 1 SLAG 1 1 I -ii 1 II IN] TRIC K-l SLAG PROCESSING 1*— 1 1 1 1 ( *CID 1 'U' METAL INGOT 1 4t 1 1 LEACH LIQUOR FOR 'U' RECOVERY - 796 - HOVEL SURPACE CHEMICAL TREATMENT TO IMPROVE TUE QUALITY OF SINTERED U02 PELLETS B.Venkataramani and R.M.Iyer Chemistry Division, Bhabha Atomic Research Centre, Troinbay, Bombay 400 085. Sintering of U02 to form pellets is aided when U02 powder with high O/U ratios are used. A novel surface chemical treatment technique has been developed to deliberately oxidise a fraction of surface uranium in U02 powder to hexavalent state. Controlled oxidation is achieved by a solid state reaction using ethylene diamine carbamate (EDO and hydrogen peroxide vapour. EDC catalyses the oxidation of uranium by hydrogen peroxide and subsequently forms a surface complex with uranyl ions, which acts as binder. Using U02 powders of varying qualities (produced in Nuclear Fuel Complex, Hyderabad), it has been demonstrated both on laboratory- and pilot-plant-scale, that the newly developed surface chemical treatment improves the quality of the sintered U02 pellets and increases its grinding recovery. INTRODUCTION Sintering of U02 is controlled essentially by the diffusion of 0 and U. Therefore, the chemical nature of the presintered U02 powder, namely, the O/U ratio, affects significantly its sintering characteristics (1-5); U02 powder with high O/U ratio sintering more readily and to a higher density (1-5). In practice, it is difficult to produce U02 powders of required quality uniformly - 797 - in all batches. Hence, during pellet fabrication stages oxygen is introduced into U02 using moisture,steam or C02 (1,3,5,6) to improve its sinterability. However, this practice of controlled oxidation does not guarentee consistent and desired results during normal production. A surface chemical treatment techinque was developed with a view to oxidise uranium in U02 to hexavalent state uniformly and in a reproducible manner, so that U02 pellets with improved quality and with less rejection rates are produced, irrespective of the quality of the starting material. The basic chemistry underlying the surface chemical treatment is akin to the dissolution of U02, in that both the processes involve oxidation of uranium in U02 from the tetravalent to the hexavalent state. In the present surface chemical treatment the oxidation is sought to be achieved by a solid state route. It is known that U02 dissolves readily in presence of H202 and carbonate (7,8). Our investigations showed (9) that, inorganic carbonates, though effective in oxidising-surface uranium, are not good enough to improve the quality of sintered U02 pellets. More promising results were obtained with ethylene diamine carbamate (EDC) as an additive. EDC was chosen for the study based on the suggestion by Gaines Jr et al (10) and LeBlanc Jc (11) that, EDC id a fugitive binder for U02 powder. The salient results obtained while developing the surface chemical treatment with EDC are presented in this paper. - 798 - EXPERIMENTAL UOg powder: The U02 powder used for all the trials were part of the production lots o£ Nuclear Fuel Complex (NFC), Hyderabad and were obtained by the decomposition of ammonium diuranate. Preparation of EDC; Ethylene diamine carbamate, NH2(CH2>2 NHCOOH, was prepared by passing'CO2 through a 10% solution of ethylene diamine in methanol kept in ice cold water (12). EDC precipitates as a white solid, which is filtered, washed with ether or methanol and dried in vacuum at room temperature. Surface chemical treatment; A known amount of EDC was intimately mixed with UO2 powder and the mixture was kept in contact with H2O2 vapour emanating from 2-3% H202 solution. Assessment the treated powder; The extent of surface oxidation was tested by putting 1 g of the treated U02 powder in 20 ml of distilled water and estimating the concentration of uranyl ion in the supernatant solution. On an average 1 mg of soluble uranium was present in 1 g of the oxide when treated with 1% EDC. The quality and sintering characteristics of the treated powder were evaluated in NFC, Hyderabad. RESULTS AND DISCUSSION General Remarks U02 powder mixed with EDC absorbs more moisture than pure U02. Moist environment on U02 powder simulates a solution-like condition on the i - 799 - surface and facilitates the oxidation of surface uranium atoms to the hexavalent state. The uranyl ions so produced complex with EDC, as inferred by the optical spectra of the supernatant solution in contact with treated UO2 powder and also the IR spectra of the treated U02 powder (9). X-ray powder pattern (9) revealed that U02 was modified from U02 Q3 to U02 25 as a result of surface chemical treatment. When the treated powder is dried at 80 o C in vacuum, there is a considerable decrease in the amount of leacheable uranyl ion concentration, which shows that the uranyl-EDC complex has a tendency to adhere strongly on U02 surface. Results of Various Trials Tables 1,2 and 3 show the results of the trials with 160 g of Lot A, 5 kg each of Lot B,C and a and 100 kg of Lot E. Reliable data were available only for 50. kg of the treated U02 powder of Lot E, which are given in the Tables. surface area; As a result of surface chemical treatment, the surface area increases compared to original powder (Table 1). Surface area of U02 powder is one of. the physical parameters which affects the sintering characteristics; powders with larger surface area show better sinterability (1,3,4). Thus, the treatment improves the quality of the original powder. The increase in surface area could have come about as a result of the modification of U02 brought about by surface oxidation (1). Green density; Surface chemical treatment results in the fabrication of uniformly dense pellets, as is evident from the increase in the green density - ROO - of the pellets made from treated UO, powder (Table 2). The increase in the green density is due uranyl-EDC complex functioning as a chemical binder. Grinding recovery: As a consequence of the improvement in the quality of U02 powder, the surface chemical treated powders gave sintered U02 pellets having higher % recovery (Table 3). In all cases, the % rejection due to end caps and chips were much less compared bo original powders (Table 3), which reflects on the physical integrity of the pellets made out of treated powders. Major contribution to the % rejection of the pellets made from treated powders was in the form of pits and cracks (Table 3). This could be attributed to the decomposition of EDC during sintering and the escape of the gaseous products formed. Heating the green pellet at 250°-300° C prior to sintering could reduce the % re jection due to pits and cracks. Salient: Features of the surface Chemical Treatment Laboratory-scale experiments with U02 powders of varying qualities and limited data available on pilot-plant-scale trials have consistently shown that surface chemical treatment with EDC and H202 vapour does improve the quality of the U02 powder and results in the an increased % grinding recovery. The newly developed technique does not involve the use of toxic or corrosive chemicals and can be adopted without modifying the existing process flow-sheet. It should also be mentioned that the newly developed treatment is more elegant and much simpler to operate than the one* - 801 originally suggested by LeBlanc Jr (11), which is a 2-step process. First step involves oxidising the surface of U02 particles by exposing them to solutions containing oxidising reagents at pH 2. The second step involves adding the amine carbamate as an aqueous solution to the already treated U02 powder and blending the mixture with alumina. Uranyl-EDC complex formed during the treatment acts as a binder. But, unlike the conventional organic binders (which are just physically present in the green pellet), uranyl- EDC complex is a chemical binder, which is a part of the U02 matrix. Its binding characteritics are not governed by the quality of the original U02 powder and hence, will give uniform and consistent resuts, which is not true of conventional organic binders. ACKNOWLEDGEMENTS The authors wish to thank Dr. K. Balaramamoorthy, Chief Rxecutive, NFC, and Dr. T .S . Krishnan, Deputy Chief Executive, NFC, Hyderabad, for their cooperation in supplying 0O2 powder and evaluating its quality after surface chemical treatment. REFERENCES 1. J.Belle (ed), "Uranium dioxide: Properties and Nuclear Applications", USAEC, Washington DC,1961. 2. P.Balakrishna, T.R.Rainamohan and P.Ramakrishnan, Trans.Indian Ceram.Soc.,46, 153 (1987) - 802 - P.A.Haas, Nucl.Technol. ,£1, 393 (1988) F.Glodeanu, M.Spinzi and V.Balan, J.Nucl.Mater., JL53_, 156 (1988) G.H.Chalder, N.F.Y.Bright, D.L.Paterson and L.C.Watson, Proc. 2nd ON Int.Conf. on the Peaceful uses of At. Energy, Geneva, 1958, vol. 6, p.590. T.Terraza, J.Cerrolaza and E.Aparicio, Proc. 2nd UN Int.Conf. on the Peaceful uses of At. Energy, Geneva,1958, vol. 6,p.620 L.H.Johnson and D . W.Shoesmith , in "Radioactive Waste Forms for the Future" (eds.W.Lutze and R.C.Ewing), Elsevier Sci.Pub., Amsterdam, 1980, chapt. 11,p.635 C.A.Eligwe and A.E.Torma, Metall, 3^, 1 (1986) B.Venkataramani and R.M.Iyer, to be published G.L.Gaines Jr, T.J.Gallivan, H.H.Laska, P.A.Piacente, P.C.Smith and W.J.Ward, Patent US 4,427,579/A,24 Jan.1984 O.H.LeBlanc Jr, Patent US 4 , 572,810/A,25 Feb.1986; Patent GB 2,164,638/A,26 Nov.1986 (a) O.H.LeBlanc Jr, Private communication (b) E.Katchalski,C.Berliner-Klibanski and A.Berger, J.Am.Chem.Soc., T3., 1829 (1951) - 803 Table 1. Surface areas of U02 powder before and after surface chemical treatment uo EDC Surface Area, mVg 2 content original treated I Lot A 1% 2.69 3.71 Lot A 2 % 2.69 3.42 Lot B 1 % 3.03 3.31 LOt C 1.5 % 2.70 2.62 • Lot D 1 % 2.86 3.16 Lot A: 160 g of UOn powder +EDC exposed to H2O2 vapour for 5 days Lot B,C,D: 5 Kg of U02 powder + EDC exposed to H202 vapour for 5 days Table. 2. Green densities of U02 pellets made from original and surface chemical treated U02 powders uo? EDC Green density, g/cc content original treated Lot A 1 % 5.64- 5.69 5.64- 5.69 Lot A 2 % 5.64- 5.69 5.69- 5.75 Lot B 1 % 5.73- 5.78 5.76- 5.81 Lot C 1.5 % 5.70- 5.78 5.81- 5.87 Lot 0 2 % 5.69- 5.73 5.74- 5.77 Lot E 1 % 5.65- 5.70 5.71- 5.78 Lot E: 100 Kg of U02 powder + 1 % EDC exposed to H202 vapour for 16-17 h. Results are for 50 Kg of the treated U02 powder. - 804 - Table 3. Grinding recovery and rejection rates of sintered U02 . pellets made from original and surface chemical treated U(?2 powders. uo2 EDC % grinding % rejection content recovery origi- trea- pits & cracks chips Send caps . nal • ted % original treated original treated Lot B 1 • 27 65 56 33 17 2 Lot C 1.5 8 45 83 46 9 9 Lot D. 1 61 63 17 25 21 13 Lot E 1 " • 68 79 5 13 27 8 - 805 - PC BASED URANIUM ENRICHMENT ANALYSER V.K. Madan, K.R. Gopalakri3hnan and B.R. Bairi Electronics Division Bhabha Atomic Research Centre, Trombay, Bombay 400 085. It is important to measure the enrichment of unirradiated nuclear fuel " elements during production as a quality control measure. An IBM PC based Uranium enrichment analyser has been developed by the authors for Nuclear Fuel Complex (NFC), Hyderabad. The analyser has advantages over the analysers based on rate devices. It has flexibility, ease of calibration and automatically stores data and the results. An IBM PC plug in card has been used to acquire the nuclear data. An algorithm has been developed to ease the calibration, the data handling and the data storage and make the system more reliable. All the programs are written in BASIC. 1. INTRODUCTION It is important to assay fissile isotopes e.g. U-23S using nondestructive testing (NDT) methods during production, movement, • safeguards etc. Various techniques are used e.g. passive ganmui counting which, is a simple and effective method for determination of U-235 enrichment during production of fuel elements as a quality control measure [1- 3]. Electronics Division has supplied and upgraded enrichment analysers to nuclear fuel complex (NFC) Hyderabad using binary rate multipliers (SN 7497's), decimal rate multipliers (SN74167's) and a novel up down gating unit at various stages of the upgradation. The systems based on l'ate devices (2- 7) are tedious to calibrate. The flexibility of the system is also limited as one has to manually, recalibrate the' system from measuring the enrichment of the fuel elements to fuel pellets. Moreover the logging of the data is done manually. This paper describes nuclear assay equation, IBM PC based system and the salient features of the software. - 806 - 2. NUCLEAR ASSAY EQUATION The U-235 isotope decays by alpha-par-ticle emission to excited levels of its daughter nucleus Th-231, which in turn emits gamma rays of various energies. The energies and emission rates of the gamma radiation following the decay of U-235 are unique for this isotope and may thus be used for the quantitative and qualitative NDT assay of U-235 content in Uranium bearing materials. The most prominent gamma line, observed in the spectrum has the energy of 185.7 kev'. The method consi3t3 of viewing the Uranium sample with an appropriate detector e.g. Nal(Tl) •crystal, and fixing one single, channel analyser on the 185.7 keV peak and" a second single channel analyser on the. background spectrum some distance above 185.7 keV energy. If the output of these two single channel analysers are now counted and. the respective counts designated as Cll and C22, the following equation holds kj-Cll - k2.C22 E = ,(1) T where E is the enrichment as percentage of the total sample, T is the counting time, kj and k2 are calibration constants which depend upon detector parameter, window width etc. 3. PC PLUG IN CARD An IBM-PC Plug in Card PC-223 has been used [8]. It is well suited for real time applications of counting events and generating timing signal. It has got I/O base address selection with DIP switches. For counting/timing programmable interval timer 8253-5 have been used. The card has an 8 bit output port and an 8 bit input port as well as interrupt generating capability. 4. SOFTWARE DEVELOPMENT The software programs for datu acquisition, data storage and data bundling and process Lug have been written in BASIC [9], The 8253-5 have boen initialized by using them as I/O ports. The.base address chosen has been 512. For input port the base address 13 532 while - 807 - for output port . the address used is 524. For reading and writing in the I/O mapped I/O ports, • the instructions used have been INP and OUT. Fig.l shows the flow chart of the algorithm of the main section of the program written in BASIC. The program has been written in a conversational manner to ease inputting the parameters. After initialising and inputting ROD/PELLET to be assessed, the time, T, for acquisition of data is put in multiples of 5 seconds and the Cll and C22 counters are initialized. Since the 8253-5 are down counters, the data acquired is subtracted from XX. XX i3" the maximum capacity of the 16 bit counter. The YY is initialized for weightage of the higher byte. The count of Z indicates whether the data has been acquired for 5 seconds or not. If the value of Z is less or more "than 10% of variation of Z on each side of the nominal value, that . reading is rejected. Thus the effect of any transient is minimized and the system is made more reliable. After '5 seconds of counting the counters are read and the counts acquired are' calculated and designated as CI and C2. CI and C2 represent counting events in a 5 second slice. Agan a check is. made on CI and C2 and if they are within limits, they are processed. The counts of CI and C2 are added to the variable Cll and C22. The limitation of. the 16 bit counter are overcomed in the software by adding the counts collected every 5 seconds to Cll and C22. The value of coefficient s. k.^ and ko are read from a data file and value of the enrichment E i3 calculated and stored in another data file along with data and identification number of the fuel. 5. CALIBRATION The calibration of the unit is made simple and easy by calling a calibration routine. The prompt ROD/PELLET is replied and two standard samples are counted one after the other. The samples are automatically counted, the constants calculated and put In the data file. The calibration is so simple and easy that it can even be done each day. - 808 - 6. CONCLUSION An IBM PC based system for Uranium enrichment measurement has been developed for ease of data handling, data storage and make the system more reliable. The system is highly flexible and easy to calibrate. The software features overcome the bottleneck of 16 bit counting capacity of a 8553-5 counters by counting in slices oi 5 seconds and adding the counts to a numeric variable. The reliability of the system has been enhanced by incorporating a Z counter. 7. REFERENCES 1. P. Matussek, "Accurate Determination"of the U- 235 Isotope Abundance by Gamma Spectrometry", Report, Kernforschungszentrum Karlsruhe, Kfk 3752 (1985). 2. T.R. Canada, "An Introduction to Nondestructive Assay Instrumentation", IAEA (1934). 3. Martin E.R., "A Direct Reading Arithmetic Unit for Nondestructive Assay of Nuclear Materials", Nucl. Instr. and Meth. 104 (1973) 439. 4. P.L. Bhatia, V.K. Madan, S. G. Sule and V.A. Pethe, " Digital Enrichment Monitors", Proc. Symp. Nuclear Reactor . Instrumentation Vol. I, pp.53-58, Bombay, December 20-23, .(1976). 5. Instruction Manual for Instrumentation, "The SAM-2 system for Non-Destructive Analysis (NDA) of nuclear Materials", IMI No.4 (Rev.l), IAEA, Dept. of Safeguards and Inspection. IAEA (March 1974). 6. Instruction' Manual for Instrumentation, Attach- ement 1, "Use of SAM-2 system for Enrichment Measurements", IMI » 4, IAEA (Dec.1974). 7. V.K.. Madan, "A Fast and Economical Arithmetic Unit for Nuclear Data Processing", Proc. Symp. Indigenous Nuclear Equipments, Bombay, paper No. 37, pp 1-4 (1987). 8. PCL-223, Owner's Manual, Dynalog Micro Systems, Bombay (1980). 9. Gottfried, B. S., Theory and Problems of Programming is BASIC, Schaum's Outline Series, Mc-Graw (1988). - 809 - r moot *nt i I 1 T waa lYTt n ci^et-<».rrM> s CD KTIMJtC s •C33 - 9 s s s MJCII - let at c m»t mr MIX «8 ** r> BI D«T« n* a oi •i 3 IB H C~D a I L ID -i a3 —• • > FIG.l FLOW CHART OF A PROGRAM SECTION •0' it' L > a - 810- SeBBlon VII DISCUSSIONS Paper No. 1 L.M. GANTAYAT t Please elaborate on how you have controlled NOX during dissolution of uranium metal scrap. A.M. MEGHAL : One of the methods found effective was the addition of urea during dissolution of the scrap with nitric atid. L.M. GANTAYAT : What is the per centage of uranium recycled to UMP as scrap? A.M. WEGHAL : Most of the recycle scrap is sent from Atomio Fuels Division, where the scrap is incinerated to oxide:before returning to UMP. In normal operation, this scrap could be anywhere from 30 to 40 per cent of the input to UMP. SESSION VIII ENVIRONMENTAL AGF5CT3, HEALTH A7TD SAFETY Chairman : Shri R.C. FURI U CI L Heporteun Sratt S; ROY B A R C / - 811 - TREATMENT OF URANIUM TAILINGS VIS-A-VIS RADIUM CONTAINMENT P.M.MARKOSE, K.P.EAPPEN, M.RAGHAVAYYA AND K.C.PILLAI* Health Physics Unit, Oaduguda * Health Physics Division Bhabha Atomic Research Centre, Bombay One of the essential requirements of operations in a uranium mine-mill complex is the treatment and containment of tailings so that their discharge to the environment meets the accepted release standards. The treatment is directed towards fixation of radium, the potentially hazardous radio nuclide associated uith uranium mineral. Conventional lime neutralisation treatment itself has not been proved completely adequate in achieving this target. This paper revieus our experience in the uorks related to the containment e? radium in. tailings. Lime treated tailings effluents carry dissolved radium to the extent of 6000 to 8000 Bq.M"3, uhich is undesirable from the point of view of its environmental impact* A contact bed process for treating this effluent yea developed, achieving more than 90 % de contamination. The merits and demerits of the methods are discussed. Other options considered and trials carried out to reduce the dissolved radium in tailings are also discussed. A recent approach in the treatment of tailings is the extraction of radium from the tailings in an attempt to make it intrinsically safe. About 90 % of radium.could be extracted using ferric chloride in hydrochloric acid solution. Various implications of this approach to the containment of radium are discussed in the paper. The paper also indicates incidental containment of other-" toxins in the tailings, like manganese. - 812 - INTRODUCTION 226 Containment of Ra in the tailings is one of the most pressing problems faced by a uranium milling industry. This is duo to the fact that while conventional treatment of the solid-liquid uastes effectively contain the other pollutants present in the uastest it fails to produce an effluent having dissolved Ra belou the permissible limits. Additional control measures have been necessitated by this complexity. This paper reviews our experience in the uorks related to the release of Re to the environment and its containment methods. SCOPE Among the various options, "three possible stages uhere modifications or treatments additional to the existing ones that could control radium release from the tailings pond are outlined here,. The first one is at the present lime neutrali sation stage uhere suitable treatment is incorporated to stabilise 226Re in the tailings. Neither any successful method is known to have been practiced nor any literature in this regard could be traced. The second alternative is the treatment of effluent overflowing from the tailings, pond for containing the dissolved 22°Ra. It has bean.documented that treatment of the effluent with a solution of BaCl2 followed by adequate flocoulation can retain 22eRa effectively. Beverly (1974) in his review of waste management in mining and milling of uranium has mentioned the use of BaCl2 and Environmental Protection Agency (1979) has described the use of it in. various- mine effluents for containment of Ra. Even though ion exchange separations are possible to retain radium from effluents, no such feasible method has been known to be in successful operation anywhere. The third alternative would be to extract 226Ra quanti tatively from the tailings. A feu reports have been cited in literature towards this direction. Dissolution of 226Ra using - 813 - salts like NaCl, KC1 and BaCl2 havs been reported by Ryan and Levin (1979) and Haque and Ritcey (1983). Nixon et.al (1963) and Markose et al (1985) have used EDTA Par the dissolution of 2"Ra and observed that the dissolution is fast and independent of solid-liquid ratio* Approximately 85 % dissolution was observed by Nixon at al using DTPA solution- Sgeley (1977) observed that 94 35 226Ra could be extracted from conventional uranium mill tailings uith 3 PI HC1. Ryon et al (1979) extracted about 90 £ 226Ra u8ing 3 n HNOj. A pressure leaching system using CBC12 solution has been successfully employed by Demopoulose (1987) for simultaneous extraction of uranium and radium from uranium ore, Nirdosh et al (1983) found that among various salts triad, F9CI3 uas the most promising reagent for dissolution of 226Ra from ore* However, no uork could be traced relating to the us« of ferric chloride for ' extraction of tadium from the tailings. DISSOLUTION OF 226Ra IN TAILINGS magnitude of the problems regarding radium containmont can be assessed from the following behaviour of the radio nuclide in the tailings. Finer particles are known to contain major portion of the radioactivity ( Ma, 85 ) and consequently, though only 55 % of the solids ( fin as. ) are discharged into the taifcinga pond, about 80 % of the total radioactive nuclides are associated with them. Fractionation studies (Ea 86) have established that approximately 0,5 % of the total 226R8 dissolves during uranium leaching against approximately 95 % uranium and 60 % 230Th. After the conventional treatment of acidic waste uith lime to a pH 10, practically all the 230Th gets fixed in the solids while the dissolved 22oRa concentration in tha tailings remains to be of the order of 6000 to 8000 3q.W~3 ( approximately 0.3 % of total Ra ). As a result of the effluent decanted out of the tailings pond carries dissolved 226Ra to the extent - 814 - of 2000 to 3000 Bq.ra-3. This demands additional treatment of th8 waste or affluent for effective containment of the tailings, METHODS ATTEMPTED FOR 226Ra CONTAIN WENT BARREN LIQUOR An attempt uas made in the initial stages to treat the 226 barren liquor separately for control of Ra before it is mixed with the solids. The experiments uere based on the follouing assumptions. (i) The barren should hold practically all the dissolved radium present after the leaching stage and therefore if radium can be affectively precipitated, it should bring down the radium level in the final effluent, (ii) The pH of the discharged effluent after lime treatment falla rather sharply to about 7 from 10 partly due to crystallisation and partly due to conversion to carbonates and subsequent precipitation in the environment of the tailings pond.Tha pH of the effluent overflow uas found to be generally about 7, though the discharge is done at about pH 10 or above. This fall in pH brings manganese also into solution uhibh otheruise remains in the solids at higher pH range. Therefor* maintenance of alkalinity in the effluent uas to be ensured. Ue have conducted a series of pilot plan!) experiments uith a view to contain 226Ra at this stage. 600 litres of barren liquor from the mill uas treated uith lime to a pH of about 7, The pH uas further raised to 10 by the addition of washing soda or trisodium phosphates/It uas then mixed with 600 litres of solid slurry and additional washing soda or TSP was added to bring up any fall in pH and the resulting slurry was discharged to a--tank. Three batches, one treated with lime alone, another uith lime and washing sods and a third batch uith lime and TSP uere preserved in separate tanks. s# - 815 - Samples uera draun and analysed for a period of one month, of In all 16 batches^experimenttwore performed by changing various parameters. The results of a typical batch are presented in table-I. Results and discussions In all the batches the decontamination of the barren liquor was very satisfactory irrespective of whether treatment was done uith lime alone, lime and washing soda or lime and TSP. Progressive decontamination of the barren uas observed and 99 % of the radionuclides uere carried doun in the solids. Maintenance of pH uas stable in batches in which washing soda or TSP was used. This stability in higher pH helped in keeping manganese in the solids. However, when the solids were mixed tilth the treated barren ( Table -I, si.No.5, 6 ), the 226Ra valua in th9 dissolved state shot up to the order of 5000 - 8000 Bq.m"3 in all the cases. This may be due to the fact that the solids are carriers of the bulk of 226Ra. The total quantity of 226Ra in barren /ujuor is negligible compared to the radium present in the solids. The radium which has been removed from the ore during leaching may be existing in RaSO^ form in the solids. Uhen alkaline barren'as mixed with these solids, the loosely bound 226Ra (Be 61 ) gets dissolved again and the concentration of dissolved "°Ra in the mixture is rised. Thus this treatment does not result in achieving the control of ^2£>Ra in the tailings though the pH may remain high Cor longer duration, thus keeping manganese level lou. TREATMENT OF THE EFFLUENT OVERFLDU FOR 226Ra AND MANGANESE CONTAlNflEln Another alternative to the containment of Z2°Ra is the treatment of the effluent overflowing from the tailings pond. It has been well established that addition of BaCl2 precipitates Ba(Ra)S04 and the resulting liquid will have low radium - 816 - content (Op 85). Since this control method is selective for radium only, further treatment has to be adopted for limiting manganese released to the environment. A contact bed process uas developed by us ( Ma 81 ) and used in the tailings drain. Natural gradient of the location uas used for the flou of effluents into the beds. The system consisted of a series of silex bads, pyrolusite buds and barytas beds as shoun in figure 1. This uas aimed at adsorbing the dissolved fin also from the effluent, in addition to 226Ra. The results are shoun in table II, uhibh indicate that the contact bod is successful method for the effective control of 226R8 and Pin from the tailings. An overall 98 % decontamination efficiency uas achieved and the resulting affluent had neglibible 226Ra concentration; Moreover it adsorbs the manganese present in the water, thereby making the effluent suitable to be discharged safely into the enviornment. Since the contpinment of 226Ra and fin is achieved through surface adsorption, it is clear that the surface of the pyrolusite and barytas pebbles should be kept free ofsilt deposition. Silex pebbles remove the silt before the effluent comes in contact uith pyrolaite end barytas beds, moreover the pH of the effluent, should be above 7.5. Since pyrolusite itself is a source of manganese, any acidic effluent discharge through the bed would bB hazardous. Strict control over the tailings neutralisation step is essential for the contact bed process to be successful. EXTRACTION OF 226 Ra pRDM TH£ SCOP TAILINGS The third option for the safe disposal of solid uranium mill uaate to the environment is to extract the radium quantitatively from the solids. Even though the extraction of 226 Ra by EOTA or leaching by strong mineral acid are - 817 - offactive, thay can not be projected as practical reagents for radium removal because of economic and other consierations, A viable alternative was to try another relatively cheaper reagent which causes the extraction of 226fja to a significant level," FeCl3 solution in dilute HC1 uas tried as the leaching medium. Since it uas expected that HC1 itself uould dissolve a part of rom tailings, a control sample uas run along with every set of experiment uith HC1 of the same strength as that used in F0CI3 preparations Experiments uere run uith aged tailings as uell as fresh tailings uhich uere neutralised uibh line* In initial stages, 2 g of aged tailings uere treated uith 100 ml of leach solution and later on 30 g of fresh solid tailings uas legched in 100 ml of leach medium. The parameters extmined uere strengths of FeClg and HC1, effect of temperature and the effect of sulphate ion concentration on leachability of radium. Finally sample of repulped solid wastes prior to lime treatment uas leached uith the reagent for finding the efficiency of extraction. Results and discussion Results.of initial experiments to determine the strength of the leaching solution for extraction of ?26Ra efficiently have been given in table £11. It may be seen from the results that HC1 itself has leached radium to the extent of 75 % uhile addition of FeCl3 definitely increases the dissolution* Increase in the reaction temperature also had a positive effect on the extraction of 226Ra. At room temperature the maximum efficiency of extraction uith 0.5 M FeCl3 in 0.5 PI HC1 uas 90 % uhsreas at higher temperature ( 60 - 70 ° C ) 0.3 !«1 FeCl3 in 0.5 1*1 HC1 extracted 98 % radium. The initial results of leaching experiments on fresh tailings uith 0.2 PI FeCl3 in 0.2 ft HC1 uere, houaver, very discouraging ( Table IV ). The efficiency of extraction had reduced drastically. The reasons for this uers attributed - 018 - to the high sulphate ion concentration in thesolids, the louer HC1 strength and high solid to liquid ratio of 3:10 instead of 1:50 as in earlier casa. The afreet of sulphate ion on extraction of 226Ra from tailings uaa verified by repeat leaching of the solids with fresh leach medium so that sulphate ions are reduced in successive leaching steps. The results demonstrate that efficiency is increased as the sulphate ion concentration is reduced. However the maximum efficiency that could be achieved in 0.2 M FeCl3 in 0.2 (1 HC1 was only about 70 % even on a solid which uas washed to remove sulphates. Thus it was seen that strength of the leaching medium also affects the efficiency. Extraction of 226Ra from solids before neutralisation The loachability tests carried out so far uere from samples of lima treated tailings which had pH in the alkaline range. Since extraction of radium by FeCl3 is performed in acid medium, neutralisation steps prior to radium removal is not needed. Therefore it is advantageous if waste can be treated with FeCl3 before it goes for final neutralisation. In ordrir to verify this, a sample uas collected from the drum filter repulp slurry which represents the solid waste produced after the dissolution and separation of uranium from the ore in the mill. Initial pH of the collected sample uas 3.0. A portion was taken for leaching and the remainigg tailings uas washed in tap water, keeping the solid to liquid ratio as 1:2. The wash solution uaa taken for sulphate estimation and a sample of solid for leaching experiments. The rest of the sample was again washed as before end these stops uere repeated till the sulphate in the wash solution had reduced to very low value. The laat stages of washing had to be done with distilled water', since the tap water used for washing had relatively high salphate ions. Estimation - 819 - of the sulphate and radium leached ware carried out in samples of each step. The results are tabulated in table V. Five washings were dons through which the sulphate ion concentration was brought down from 2295 mg.l to 7.8 mg.l in washed solution. In this process the amount of sulphate leached along uith radium in each leaching stags decreased from 71.5 mg in the first step to 11.9 mg in the final stage. The pH ofthe uashed solution increased from 3 to 6.8 gradually. The liberation of radium shoued improvement at each step as the sulphate was removed. The final sample liberated 95 % "°Ra from the tailings which is a positive improvement from the previous experiments conducted uith fresh tailings. This significant increase in the extraction of radium i3 mainly due to the sulphate removal and also because of the HC1 strength being 0,5 M instead of 0.2 PI in the previous case. The least square fit drawn for radium liberated against sulphate gives the relations, as shoun in figures 2 and 3. D63 Y - 6.04 X ~°« • r . o.ge and Y - .7.99 X" 0,19° { r - 0.98 for sulphate in solution and sulphate leached respectively, this good correlation indicates the relation between radium and sulphate in the system. These results are comparable to the preliminary teats conducted with 2 g each of old tailings where 98 % efficiency was achieved. In the present case the solid used was 30 g instead of 2 g whereby changing the solid to liquid ratio from 1 : 50 to 3 : 10, Moreover, the total radium handled in this system was also marginally high. CoTiSCOTraiivn When HCl,was raised to 0.5 PI, the initial ..-teaching itself was as high as 65 %. But the comparison may not be justified completely, since earlier sample was a neutralised tailings where as the sample in this experiment was acidic. However it signifies the fact that radium can be successfully leabhed out of the tailings using FaCl3 in HC1. - 820 - CONTAINMENT OF LEACHED RADIUM It has been demonstrated that the bulk quantity of radium can be successfully leached out from tha tailings. The solid thua obtained may be safely disposed off as its radium content uould be about or less than 1.0 Bq.g""^ of the solids. The. liquid portion uhich contains most of the radium should be treated to remove tadium in a safely cont ainable form. In order to achieve this target, the extracted effluent uas treated as described belou. The solution was neutralised uith 1P1 NaOH to precipitate iron as ferric hydroxide. The bulk precipitate uas allowed to settle for half an hour and filtered through Whatman 42 using suction. The residue obtained uas rejected and the • effluent uas analysed for radium. The results of a number of experiments carried out are presented in table VI. The results in the table indicate that 84 % to 95 % of the extracted radium is adsorbed on ferric hydroxide and thus transferred to the solid phase. The liquid portion retains only about 5 to 16 % of radium, thus it can be discharged into the environment after suitable dilution, if required.The precipitate of the solid cake may be dried and packed separately for disposal at a suitable location. CONCLUSIONS Different methods at various stages of the plant operation have been tried for the primary objective of controlling the dispersion of 226Ra to the environment. Inplant treatment \W,uof of the ben-en^has bpen found to be unsuccessful in achieving the target due to the existing complexity of the situation. The second alternative of treating tha effluent from tailings pond has been dons uith the aim of controlling both Ra and Fin. Though the contact-bad mathod has advantage over BaCl2 treatment in the sense that it contains both "Ra and Pin from the effluents, the inherent diffacts in the system are tuo fold. One of them is the fact that the surface of the - 821 - pebbles have to be kept frBB of silt deposition. The second one is the possibility of acidic effluent entering the bed uhich not only defeats the purpose but also is hazardous doe to possible liberation of Mn from the bed itself. Extraction of 226Ra from the solid tailings appears to be a promising alternative to the above practices. Since solids contain more than 99 % of 226yja inventory of uranium ore, removal of 226Ra from it uould make the solid intrinsically safe for disposal with out any short term or long term envi ronmental considerations. Our experiments demonstrated that 226 more than 85 % of dissolved Ra may be adsorbed on FeJQH)3, Alternatively a HaS04 precipitation followed by flocoulation may be carried out to ensure stebility and better containment. Though experimental datas are not available, it may be assumed that since leaching is done in HC1 medium, a major portion of Mn also gets leached into solution along with Re uhich uould be fixed uith out any additional treatment. However, the . precipitate from radium extract haa to be carefully hadled and disposed off in such a u§y that it does not pose any operational risks as well as environmental hazards for the future.generations. ACKNOWLEDGEMENT Ue are grateful to Dr.I.S.Bhat, Head, E.S.Section, B.A.R.C for his keen interest and constant encouragement. Thanks are due to UCIL authorities for providing facilities for the work. The untiring services rendered by Shri P.N. Shyamaprasad and other members of Health Physics Unit, Oeduguda are gratefully remembered. - 822 - REFERENCES BB81 Benes P., Sedlacek 3., Sebasta F., Sandrink R, and Dohn 3., 1981, " Method of selective dissolution for characterisation of particulate forms of Ra anc.' Qa in natural and waste uatera ", Uater Research 15, 1294-1304. Be74 Beverly R.G.,1974, " Uaste management in mining and milling of uranium ", IAEA symposium of Radiation Protec tion in Mining and Hilling of Uranium and Thorium, Bordeaux, France. De87 Oemopoulose.G.P., 1987, "Pressure leaching of uranium ores in a Calcium chloride medium uith simultaneous solubilisetion of radium ", Trans.Institution of Mining and Metallurgy 96. Ea86 Eappen K.P., Markose P.M. and Pillai K.C.,1986, " Fractionation of radionuclides in acid-leach process of uranium ore ", Annual conference of IARP. Ep79 Environmental Protection Agency, U.S., 1979,° Potential Health and Environmental Hazards of uranium mine uastes " Ha83 Haque K.E..and Ritcey G.M., 1983, " Leaching of radio nuclides from uranium mill tailings and their flotation concentrates by hydrochloric acid and chloride salts ", Hydrometallurgy. 11, 91 - 103. Pla81 Markose P.M., Eappen K.P. and Raghavayya H., 1981, " Radium removal from tailings - a pilot study " , Syrap. Application of Nuclear and A]lied Techniques in Public Health and Pollution Control , Bombay. Ma85 Markose P.M., Raghavayya M. and Pillai K.C., 1985, " Leachability of radium from uranium mill tailings ", 3oisrn. Uater, Air and Soil Pollution 26, 95 - 105. NiB3 Nirdosh I., Baird M.H., Muthusuamy S.V. and Banerjee S., 1983, " Coextraction of uranium and radium from ore uith ferric chloride ", Dourn. Hydrometallury 10, 265 - 283. Nx83 Nixon A., Keller D., Fritze K.f Pidruezny A. and Corsini A., 1983, " Radium removal from Elliot Lake uranium mill solids by EOTA leaching '•, Journ. Hydrometa llurgy 10, 173 - 186. - 823 - 0p85 Optiz B.C., Shsruood R.E., Dodson M.E. and Seme R.3., 1985," Tailings neutralisation and other alternatives Por immobilising toxic materials in tailings ", Report NUREG/CR - 4259 PNL 5467., Pacific North West Laboratory, Richland, U.S. Ry79 Ryan R.K. and Levins O.N., 1979, n Extraction of radium from uranium tailings *, 9 th Annual Hydromet- allurgical Meeting, Toronto, Canada. Ro77 Ryon A.D., Hurst F.3. and SeelBy F.G., 1977, " Nitric acid leaching of radium end other significant radio nuclides from uranium ore tailings ", ORNL/TM. 5944. 9e77 Seeley F.G., 1977, " Problems in separation of radium from uranium ore tailings ", Journ. Hydrometallury 2, 249 - 264. - 024 - Table I A Typical Batch of Pilot Plant trial Experiments Si.No Code No, Sample pH Ra(alpha) fin (KBq.oj-3) (g.m-3) 1 pEI Barren Liquor(600L) 2.5 610.9 1341.0 2 pE2 Drum filter 3.2 0.4 46.0 alurry (600 L ) 3 pE3 . Lime added to pE 1 7.8 24.2 9.5 4 pE4. Uashing soda 10.0 0.8 <8.1 added to pE3 5 pE5 pE2 mixed to pE4 8.6 6.9 0.2 6 pE6 Uashing soda 10.0 6.8 <0.1 added to pE5 - 825 - .Tabla II Results of pilot plant studies SI, Flourate Oper. pH Eff.oP ad3orption(%) No. (lpm) Hra. Radium Manganese 1 15 1 9.9 76.3 52.0 2 15 3 9.4 94.5 84.4 3 20 7 9.7 94.2 90.0 4 20 10 10.6 97.0 • 97,2 5 15 16 9.1 96.5 98.2 6 30 20 - 88.5 94,3 7 20 26 B.3 . 87.4 96.8 8 20 33 8.7 85.9 96.9 9. 30 41 9.7 95.2 97.0 10 30 77 9.8 82.& 93.6 11 10 - 30 177 9.5 90.8 85.7 12 30 - 35 252 9.8 91.1 85.7 13 35 . 40 347 9.1 76.5 63.9 - 826 - 226 Table III Extraction of Ra from old tailings uith FeCl3 SI,No. Experimental conditions Ra extracted Control 0,2 PI HC1 - Room temperature 70,3 Control 0,5 ri HCI - ti 75.3 0.1 fl FeCl* in 0.2 PI HCI 11 80.0 0.2 n F9CI3 in 0.2 PI HCI 11 84.9 11 0,1 PI FeCl3 in 0.5 PI HCI 79.7 11 0.3 PI FBC13 in 0.5 PI HCI 86.8 0.5 PI Faci3 in 0.5 PI H61 11 90.7 0.2 M FeCl3 in 0.5 PI HCI - 60 to 70° C '98.0 0.3 in FeCl3 in 0.5 PI HCI - 60 to 70° C 98.0 Solid : Liquid ratio = 1:50 Occational shaking - 827 - Table IV Extraction of ^Z0Ra from fresh tailings SI.No. Experimental conditions SO4 in Ra Extracted solution {%) 1 0.2 M FeCL3 in 0.2 PI HC1 - 85° C 4680 .2.1 2 0.2 PI FeCl3 in 0.2 PI HC1 - 85° C 600 14.3' 3 0.2 PI FeCl3 in 0.2 PI HC1 - 85° C 180 22.1 4 0.2 PI FeCl, in 0.2 PI HC1 - 20° C, - 65.1 Uashed tailings. 5 0.2 PI FeCl, in. 0.2 PI HC1 - 80° Cf - 70.2 uashed tailings. Mechanical stirring Solid : Liquid ratio • 3:10 Contact time => 2 Hours - 028 - Table V Leaching of secondary cake in ferric chloride Sl.No. Experimental details pH So4 in Sulphate Extraction washings extracted efficiency (•9.1-*) in FeCl3 (ag/30 g) 1 Original slurry sample 3.0 2295 72 64.5 2 Uashed in tap water 3.7 743 43 74.5 3 Repeat uaahinga in tap uater 5.4 383 26 76.9 4 Repeat washings in tap water 6.5 K 14 86.9 5 Repeat waahing in tap water 6.9 26 13 88.4 6 Repeat washing in dist.water 6.8 8 12 95.0. Conditions Total slurry taken • 1 Kg Uaah volume » 2 Litres Temperature • 85° C 0.2 « FeCl3 in 0.5 d HC1 - 029 - Table VI Adsorption of radium on ferric hydroxide .No. Ra in solution Ra adso (KBq.m-3) (£) 1 3.2 + 0.16 96.8 2 17.7 + 0.53 82.3 3 26.0 + 0.78 84.0 4 2.1 i 0.18 97.9 5 8.2 + 0.41 91.8 - 830 - INLET FILTER BED (SILEX) •PYROLUSITE BED AERATION TANK -AERATION TANK -BARYTES BED FILTER BED J (SI LEX) OUTLET FIG 1. RADIUM & MANGANESE CONTAINMENT BED - 831 - »• 10 m 6 s < tu 3 _J y = f-<>4x - O-O 63 d r= 0-98 MJ OJ M _J t I 1 L I I -I I I I.I JO' -ud10" |0a SULPHATE IN SOLUTION (w^ i"0 FIG 2 EXTRACTION OF 22S Ro. AS A FUNCTION OF SULPHATE.IN SOLUTION 832 - O 3 a &4- < Ui > "" -J y*7-9SX -o.f9 d K JP * 9 . i 1 _l I I I I Ml J 1 i l t. ill 10 IPO — SULPHATE LEACHED (w^ FIG 3 EXTRACTION OF » RADON PROBLEM IH URANIUM INDUSTRY A.H. Khan and M. Raghavayya Health Physics Division Bbabha Atomic fioeearch Centre Health Physios Unit, Jaduguda ABSTRACT Radon (222Rn) emission 1B invariably associated with the mining and processing of uranium ores* Radon enters nine atmosphere through diffusion from exposed ore body, fractures and fissures in the rooks and is also brought in by ground water* Being the progenitor of a series of short lived radioisotopes it contributes over 70£ of the radiation dose to sine workers and thus accounts for nearly 30j6 of the total radiation doses received by workers in the whole nuolear industry* . This paper summarises the data on radon emanation from the ore body* backfilled sandu and nine,water* Radon and its progeny concentrations in different haulage levels and stopea of the Jaduguda uranium mine are presented to emphasise the need for a well planned ventilation. system to control :. radiation exposure of miners • Results of radon monitoring in' a few exploratory uranium mines are included to indicate jtbe need' for a good -ventilation system from the y%xy inception of the mining operations* Relative contribution of nine exhaust and tailings surfaces to the environmental radon is also given. Some instruments developed locally for oonitoring of radon and its progeny in mines and in the environment are briefly desoribed to' indicate the progress nde in this field. 'DiTROHJCTIOH Radon ( ^n) with its short-lived progeny is the main source of radiation exposure in the early stage, of the uranium fuel oyole* - 834 - Produced by decay of Ra in tba ore, radon being a gas diffuses in to the mine working areas* Inhalation of its abort lived progeny causes the major radiation dose to workers• Mining industry employ ing large work force thus emerges as one of the contributor to the per oaput and collective radiation dose in the nuclear energy sector* Radon enters the environment through mine air exhausts, crushing and grinding operations and by emanation from the tailings pond surfaces* RADON IN URANIUM MINE RADON INGRESS The exposed ore body, broken ore mucks and course mill tailings used as back fill material oollectively contribute a large proportion of the radon in mine workings* Though the emanation rate must prima rily depend on the radium oontent or the grade of the ore, other factors such as surface area, porosity fragmentation of ore pieces and atmospheric pressure, etc* of ten.have an over riding effect* Thus in the same mine wide variations in radon emanation rates' are commonly observed. Fraoturea and fissures in the zock and fine ore dusts deposited on mine surfaces Water percolating into the mine after travelling through mine ralised zones is yet another source of radon in mines*. Dissolved radon concentrations in mine water from Jaduguda and elsewhere are presented in Table II* In a study conduoted during an extended shut down period the water ingress, rate in Jaduguda mine was found to be 0.264 n .min.- The dissolved radon content of. this mine water varied from about 85.- 4000 KBq.m"5 with.a weighted mean concentra tion of about 500 KBq.n"'. Nearly 7% of this radon is: released from water during its flow in mine'workings (2). However, radon rioh water is only a source of oontaoination in the vioinity of entry points* - 835 - RAJON LEVELS AND LOSES The radon concentrations in air in the nine drifts and stopea result from the combined effeot of releasee from different sources outlined earlier* Ventilation air SIBO serves as a carrier of radon from one location to another* JThe varying quantities of broken ore piled in a stop©, changing ventilation conditions etc often result in large variationo in the radon concentrations in mines- The So levels obtained over a period of 2 years in different haulage levelb and Btopes of the Jaduguda mine are summarised in Table III* The equilibrium ratio between radon and its progeny BaA BaB (214Pb) and EaC (214Bi) averages around 1 « 0.60 » 0*30 1 0.15 in the relatively better ventilated areas of the mine* The equili brium factor (F • Wl 1 3.7/Hn) also varies widely with a mean value of about 0.4. The unattached fraotion of the radon progeny taken together ranges from 0.02 to 0*16 averaging at about 0*07* The radon concentrations obtained in some exploratory uranium mines are presented in Table IV to indicate the problem, and emphasise the need to adopt < control measures from the initial stages of the mining operations* The radiation dose to mine workers arises mainly from the inhalation of radon progeny followed by that from external exposure to gamma radiation and the inhalation of radioactive ore dust* The relative weigbtage given.to radon daughters, external gamma' radiation and ore dust in low grade mines in India is 0.75, 0*20 and 6*05,. respectively* It is customary to express radon progeny concentrations in units of working level (WL) which oan either be measured directly (3) or inferred from the radon concentrations and the F value prevailing in a particular, location. The WL is a measure of the combined concen trations of the short lived radon daughters such that potential alpha energy ultimately released will be 20.8/uJ.m"'. It is equivalent to 3.7 Bq.m" of radon in equilibrium with its each decay product. The cumulative exposure of individual mine woxkers is expressed in yet - 836 - another unit called working level aontb (ellf) which is equivalent to exposure to one IX over a poriod of 170 hours. The ICff-32 gives a done conversion factor of 10 nOv/eUl at unattached fractions around 0 .05 (4) • The external gaaaa radiation averages around 4/uGy.hr" • though alight variations can be oxpeoted depending on the ore grade in a particular location. The radiation dose to sine workers is presently eoopntcd froa the aabient radon progeny concentration* ganaa-radiatiOB and occupancy period. It la supplenented by-personal dosimetry of a selected group of workers representing different cate gories and locations* *be dose estinates of nine workers over a period of nearly 2 decades have been presented in earlier publica tions (5, 6). The frequency dlatributlon of effeotive dose equivalent estimates for nine workers daring 1988 Is shown in Flgore-1* BADOH BEI81S8 9DBIKC OHB FSOCESSIHG The processing In Bill largely coaprises of crashing and. grinding of the ore followed by leaching operations* The ore pieces get finely ground to very avail partiole sixes resulting In tzeaondous Increase in the exposed surface are*, ownslng larger snonstfnn rates* The radon locked in dosed pores also gats released. The leaching further enhances the process of radon emanation. In a study of Yugoslavian aianlnfo ore nearly, all radon contained or foraed in the ore la reported to be released during Billing operations.' Vhile radon release daring storage sad crushing operations is 7 and 1$ percent, the grinding and leeching operations account for 12 and 66 percent* respectively (7) • Bssnatloa coeffi cients* however* differ widely cue to different geological sad adaera- loglcal nature -of the ore • For exaaple, the c*—$*-i- ores are reported to release only about 20jC of the total radon oontained (8) • Thus* aasuaing conplete carnation* the radon released during aUllng opera tions would_ work out to be about 7-3 CBq#d~*. - 837 - Sinoe most milling operations are carried out in buildings having large openingo( tbe emanating radon gets diluted with the atmosphere. In view of -this, regular monitoring or radon is not considered necessary in mill. Recent measurements carried out using an integrating passive radon dosimeter are reported elsewhere in- . the8e proceedings (9)• RUXW FROM TAILINGS PONS Tbe ore grade being low, the bulk of tbe ore bandied emerges aa tailings which contain essentially all the radionuclides oontained originally in the ore except uranium. The coarse fraction of tailings containing nearly one third radium inventory is used as backfill in the mine. The finer fraction with relatively higher radium content is disposed of in tailings pond after.neutralization • . Tbe higher radium content,. larger effective surface area and inoreased porosity result in higher radon emanation rates. While small percentage of moisture (up to about 20$) ennanoes radon emana tion, the saturation or a water layer above tbe tailings surface inhibits tbe emanation rate* The distribution of radon emanation rates from the tailings pond at Jaduguda is shown in Figure-2. Tbe • —2 -1' radon emanation rates are found to vary from 0.10 to 5O.9 Bq.m .s —2 -1 with a geometrio mean of 1»53 Bq.a .a • Tbe large surface;area of the tailings pond makes it a potential sduroe of technologically enhanced environmental radon. The radon release estimates.from tbe uranium industry are given in Table V. MONITORING DEVICES Widely varying concentrations of radon with its progeny and resulting doses require a variety of monitoring devices and techni ques wbiob are not readily available* Some of the monitoring equip ment and teobniques developed at tbe Health Physios Unit, Jaduguda, are listed in Table 71. Descriptions of these devioes may be obtained from the references. - 038 - BISC0S310S It is evident froa the inforaation provided in this paper that radon and Its progeny are the sain scarce of radiation exposure in uranium industry- The concentrations generally prevailing in the working environment necessitate carefiil planning and adequate supply of fresh air in mines (lO). Sven exploratory uranlun mines need good ventilation* Though the radiation doses to mine workers are generally within the prescribed llaita for the individaalst the 5 mSv/y institatlonal average dose Unit appears rather difficult to maintain in uranium nines* The collective dose to uranium miners is about 30?6 of the radiation doeea eaticated for the entire nuclear energy sector* The estimated radon release from the uranium industry is about the save as that contributed by a circular land surface of radius about 5 *a with average ertanation rata of 0.02 Bq*n— 2* a— 1 froa the normal soil* The radon concentration in air at the periphery of the tailings pond Is of the order or 55 Bq.n"* whioh geta dilated with the ataospbera and redness to the background level of about 10 Bq.m J at 1 to. 2 Xm. froa the source. Vblle nine and •ill emanate radon during their operational life span, the tailings pond - " is a long tera source of environmental radon. Tbaa covering "the tailings with soil and growing vegetation Is considered necessary to inhibit radon emanation and prevent dust dispersal* .Keeping the tailings surface wet is also helpful. In some countries mining comparatively high grade ores* the current thinking is to scraps tbe top soil froa proposed disposal site and preserve for use as ' oovering material during decOBaisslonlng of .tbe tailings so that tbe abandoned taillnga sites aay nerge with tbe original landscape and radon emanation rates remain the same aa that from the original soil* - 839 - ACKNOWLEDGEMENT Authors wish to thank Sbri S-D. Soman, Director, Health & Safety Group, Dr- K.G. Pillai, Bead, Health Physios Division and Dr. 1.3* Bhat, Head, Environmental Studies Seotion, BARC, for their keen interest and support in this work. Thanks are duo to Sbri J.L. Bhasin, Chairman & Managing Director, and other members of tJCIL Management for extending the facilities for these studies• Assistance provided by Shri G.K. Srivastava, Sbri K.P. Eappen and other colleagues in preparation of the paper is gratefully acknow ledged' Sbri R.P. Singh is also thanked for typing the manuscript. - 040 REFERENCES 1. Kban, A.H. and Baghavayya, M., Badiologioal Characteristics of Uranium Mine Atmosphere, Bull, of Bad* Pxot. Vol. 9. Hos* 1 & 2 . (1986). 2. Khan, A.H., A Study on the Paotors Affecting the Buildup of 222Rn and its Progeny in Uranium Mines,.M.Sc. Thesis. University of Bonbay, 1979* 3. Kusnetz, HJ»., Badon Daughters in Mine atmospheres - A Field Method for Determining Concentration, Am. Ind- Hyg. Assoc J- 17, 85 (1956). 4. ICBP-32, Limits fox Inhalation of Badon Daughters by Workers, Annals of the ICB? 6, Ho. 1 (l98l). 5. Kban, A.H., Baghavayya, M. and Soman, S.D., Badiation Exposure of Indian Uranium Miners and Estimate of Associated Bisk, 6 th International Congress of IBP A, Berlin (West), Kay 7-12, 1984* - 6. Baghavayya, M., Khan, A.H. and J ha, C, Dose Estimates for Jaduguda Mine Wooers, Bull, of Bad. Prot. Vol. 9, Bos. 1 & 2 (1986). . • . 7. Maoek. J. and Strehoveo, P., paper in IAEA Proceedings on Moni toring of Radioactive Effluents from Huolear Facilities, IABA-SM-217/42 (1978). 8. Arobibald, J .P. and Hantel, J.H., Predication of Badiation Levels .and Ventilation Requirements in Underground Uranium Mines, CIM Bulletin, Vol. 77, No. 66 (1984). 9> Jna, G. and Bagfaavayya, M., Effective Dose Evaluation of Uranium Mill Workers at Jaduguda, National Symposium on Uranium Techno logy at BARC, Bombay, Deo. 13 - 15, 1989. 10. Baghavsyya, M. and Khan, A.H., Badon and Uranium Mine Ventilation, Bull, of Rad. Pxot. Vol. 12 No. 1 & 2 (1989). - 841 - 11. Raghavayya, M., An Inexpensive Radon Scintillation Call, Health Physios Vol. 40, N0.6 (1981). 12. Srivaetava, GJC., Raghavayya, II., Khan, A.H. and Kotrappa, P., A Low-level Radon Detootion System, Health Physics Vol. 46, Ho. 1 (1984) • 13• Jha, <*• and Raghavayya, U., Development of a Passive Radon Dosimeter, 5th Rational Symposium on Radiation Physics, Calcutta, Hov. 0983). 14* Khan, A.H., Ragbavayya, M., Sharma, DJI., Baman, H., Jethmalani,P.S.t Krishnamachari, G. and Iyer, M.H., Development of a Continuous Radon Monitor and InBtant Working Level Meter, Bull, of Bad. Prot. 1989 (in press)* 15. Kotrappa, P., Dua, SJC., Gupta, P.C., Pimpale, U.S. and Khan, A.H,,' Measurement of Potential Alpha Energy Concentration of En and Tn Daughters Using an Bleotret Doseoeter, Rad. Prot. Dosimetry, Vol. 5, »o. 1 (1983) • 16. Raghavayya, V; Iyengar, M.A.R. and Varicose, P«M.» Estimation of Radium-226 by Bnahometry,.Ball, of Radi Prot. Vol. 3, Ho. 4 (1980). 17. Srivastava, GJC.;. Estimation of Radium Body Burden or Uranium Miners by Measuring Rn in Exhaled Breath Using Sleotrostatio Technique, M.So. Theeis, University of Bombay ( 1983) • - 842 - Tablo Z t SAXON SANATION RATES Emanation xate Bq.m- 2 .8 -1 Souroe Range Mean Host rock 0.05 - 0.81 0.39 Backfilled tailings 0.49 - 1*53 1-01 Ulna stope (combined) 0.95 - 3*74 1-95 Table XI * RADON DISSOLVED IN 11IBB WATER Bn Concentration range Hloe lcBq.a.- 3 Jaduguda 85 - 4000 Bodal. . 17 Jajwal 983 - 3361 Asthotba 28-30 Bagjata 421 - 1105 - 843 Table III « RADON CONCENTRATION IN JADUGUDA MINES (operational areas) -~ Mean Rn levels (EBB) Location kBg .m~ ' Drives Stopsa Adits - 2 & 5 3.-30 - 230 ML 0-96 - 295 ML 0.82 O.64 370 ML 0.51 0.92 434 ML 0 .27 . .0.59 495 ML 0.14 0.35 555 ML 0.02 - 580 & 605 ML 0.11 - Overall mean oonc. ('EER) 1 0.58 kBq.m"* (-U.153 *!•) Derived Air Concentration 1 1-3 kfiq.m.- 3 Equilibrium Equivalent Radon (HER) - Rn (obs.) x F - 044 - Table IT t Bn COHC. IS EXPL0H1T0ET 1QHES „. Badoh concentration rang* Bodal (before ventilation) L48 - 6.62 (after ventilation) OJOO4 - 0.01 Jajwal (before ventilation) 9-47 - 14*25 (after ventilation) 0.22 - 2-55 Astotha 0.02 - 0.23 Bagjata O.69 - 8.57 Sable 7 1 BAXOV BEXSASB HATE BSXU4TES Source Holea., rate C Bq.d"1 Mine 59 Mill 7 Tailings Fond 56 Total 122 - 845 - Table VI i MONITORING DEVICES Device/Technique Application Reference Scintillation Cell En nonitoring in mines 11 Low Level Radon Environmental Bn moaitoriog 12 Detection System using electrostatic teobaifju© Passive En dosimeter Integrating En monitoring sod 13 dosimetry using SSHTD Continuous Bn Continuous hourly evaluation of 14 Monitor Rn cone* using electragtat£.o Rn-dtr removal and sclat* count Instant Working Rapid Rn-dtr WL measurement 14 Level Meter Alpha Sleotret Potential alpha energy oonc. 15 Dosimeter and dose evaluation Radon Bubbler Evaluation of Rn & Ra in water 16 Radon in Breath Evaluation of Ra body burden 17 Measurement from Rn in exhaled breath - 846 36 1928 Per Caput dose : 16-1 mSv/V 30 Collecti^F dose : 176 man. Sv/y Mean attendance; 22'9 days/y 24 - .18 c S 12 10 15; 20, 25 30 .35 40 Dose Equivalent (mSv/year) FIG. 1.Frequency distribution of dose equivalent •a* ^ l , > - "\ \ \ \, Vi 4 X s \ \ I \ \ \o s CO I \ ^* \ CUMULATIV E PERCENTAG a CD \ ' *0 ». m 03 N . * | £ °? S a • II '• W> - - - — •'*••*• - ••••- - 1 ? £ * • FIG.2 . RADO N EMANATIO FRO M TAILING S PON D 0 1 . 0- 5 2 4 6 8 9 99. 3 « , 9 0 0 o 0 ~c > I r u v>1 f 01IX WW 11 N()av y 848 - EFFECTIVE DOSE EVALUATION OF URANIUM HILL WORKERS AT 3ADUGUDA GIRIOHAR 3HA AND PI. RAGHAVAYYA Health Physics Division; BARC H.P.Unit; Jaduguda nines PQ Singhbhum (Bihar) Routine monitoring of areas (within and outside the ore processing facility) for external as uell aa internal dose assessment is carried out in order to evaluate the affect of uranium milling operations on radiation exposure of occupational workers at Jaduguda. Method used for the dose measurement was deployment of personal radon dosimeter as area monitor in different units of. the facility. Detectors used were CaSO<(Dy) TLD's for the measurement of external gamma exposure and' Cellulose Nitrate track detector for iter- nal alpha exposure arising due to .the inhalation of radon and its daughter products. This paper presents averages of the dose obtained over a measurement period of 3 years. Based on the average annual dose, work areas have been categorised into three groups viz 0-5 mSv, 5-"5 5 mSv and 15 mSv. Personnel doses received in each • category were 3.67 +, 0,74, 8.42 +_ 2.48 and 25.93 +, 10.76 mSv.a" respectively. Above values were also used for calculating institution dose. INTRODUCTION Routine monitoring of work areae for assessment of radiation exposure was carried out inorder to evaluate the radiological impact of discharges from uranium mine and mill operations on.the health of the occupational radiation worker* at Jaduguda, i - 849 - Sources of radiation exposure comprise of external gamma radiation, radon and its progeny and the longlived alpha emitters originating from pulverisation of uranium ors to extract uranium concentrate. In order to quantify the radiation exposure, it uas felt necessary to measure the dose on integrated bases. The personal radon dosi meter, developed here, was deployed 88 area monitor at different locations of the complex to measure cumulatibs gamma and integrated alpha dose. Contribution of dose from inhalation of airborne ore dust had been measured using tuo stags HASL cyclone sampler and were reported elseuhere£l). This paper presents only the averages of the annual exposure measured over a period of three years from 1986 to 1988. MATERIALS AND F1ETH0DS materials Detector used in the personal radon dosimeter (PRD) for registration of' alpha tracks uas Cellulose Nitrate, a Solid State Nuclear Track Detector (SSNTD), manufactured by M/s Kodak Ltd. under the trade name LR 115, Type II, palliculable. The sensitive layer of the film is 12/(jm thick on a 100/tlm plastic base. The CN material is coloured red for better contrast during microscopic vieuing. External gamma dose uas measured using thermoluminis- cent (Tl) detector pelette. Tuo such tablets uere mounted on an aluminium card separated by the CN chip of size 23mm x 15 mm. This arrangement uas found adequate for keeping the CN detector in position during exposure. Exploded vieu of ths dosimeter assembly is di&played in Fig. 1. Othar relevant information pertaining to the dosi meter are shown in table-1^2). - 050 - EXPOSURE, PROCESSING AND DOSE CALCULATION Exposure Dosimeters uere hung in tha place of interest at a convenient height of about 2.5 m above ground level, preferably auay from the ualls. Twenty four (24) locations uere covered in this program. Uork places uere selected in the manner which uould yield representative' exposure data for the rlspactive operations. PRD monitor uere deployed in areas such as crusher house, mill house, deuatering and filtration units, chemical house, tailings treetment plant and tailings pond. Apart from the main operating plants, areas monitored included control laboratory, site offices, time office, central stores, administration buildings, estate office.etc. Exposure uas continued for a period of about two months. After completion of an exposure, detector cards uere excha nged with fresh ones to continue exposure and the retrieved datectora were cleaned and processed. Detector Processing The TLD's uere read for gamma.dose using a Thermo luminescent reader. The CN detectors uere mounted on an etching stand and etched in one batch using 10$ NaOH solution at 60?c for tuo hours in an incubator to maintain a constant temperature throughout the etching period.' Tracks thus developed wars cleaned and counted using a spark counter (3). Dose Calculation Exposure E (Bq.h.l" ) uas claculated from the observed net track density (T.Cm ) using the calibration aquation : E - 0.554 T (1) c , u 3 The observed radon concentration, C 0a^ ' ^~ ' * obtained by taking.the ratio of.the exposure (E) to the - 851 - expdsure-duration measured in hours (h). The passive radon dosimeter primarily provides the time integrated radon concentration for the period of exposures, but for the purpose of dosimetry, the parameter of interest is the integrated exposure in terms of Working Level Month (VLM) units. This was derived from the observed radon concentration first by transforming it into the Equilibrium Equivalent Radon (EER) concentration using the equilibrium radlon (F) and then by applying the appropriate conversion co-efficient as shown in table I. The value of the P factor considered in the present treatment was 0.5, following the suggestions of the International Commission on Radiological Protection (ICRP). The ICRP has also recommended a dose conversion factor of 10 raSv. (WLM)~ , which has been adopted in these calculations (4). RESULTS :AND DISCUSSION The computed annual exposure level (raSv.a ) to indivi duals occupying the respective area for a duration of 8h.d"* from. 270 d.a" working period in respect of all the 24 loca tions monitored is presented in table II. The reported exposure level (table II, column 3) is the mean of three years of. continuous measurements. From the data it is evident that individual doses :vary from 2.64 to 41.12 mSv.a"* . Considering program and excercising oontrol measures, work places were categorised into three groups viz. 0-5 mSv, 5-15 mSy and _/> 15 mSv. Institution dose, calculated on the basis of the respective mean values for operational areas worked out to 7.44 mSv.a" I. Among the work plaoes monitored, it may be seen that doses aire well within limits. o!n analysis oi' the dose - 852 - profile we find that there are work places where the inhala tion dose due to the alpha activity from radon and radon daughter products are predominent dose contributors while at others it is the external exposure due to gamma radiation which is leading. While ventilation is an effective means of reducing exposure due to the radon and its progeny, there is hardly any ready made means for controlling the external gamma exposure, particularly so when unsealed sources like uranium ore is encountered. Hence for such exposures, good house keeping and personal hygiene assumed immense importance in controlling the radiation hazard. ACKNOWLEDGEMENTS The keen interest of Shri S.D. Soman, Director, Health and Safety Oroup, Dr. K.C. Tillai, Head, Health Physics Division and Dr. I.S. Bhat,Head, Environmental Studies Section, BAUC, in this work is gratefully acknow ledged. Authors are thankful to the authorities in UCIL for extending the required facility during the.course of this study. Assistance rendered by S/Shri A.K. Karua and A.K. Dwivedi of 1IPU (UCIL) Jaduguda in carrying out this work is remembered. - 853 - REFERENCES (1) Giridhar 3ha, A.H. Khan, PI. Raghavayya and P. Kotrappa ( 1976) •Measurement of API AD and MM AD of airborne particulates in uranium mining and milling industry1. 3rd annual conference of Indian Association for Radiation Protection (IARP). (2) Giridhar 3ha (1987) 'Development of a passive radon dosimeter for applications in radiation protection and uranium exploration'. A thesis for the degree of Doctor of' Philosophy, submitted to the university of Bombay. (3) Giridhar. 3ha, M. Raghavayya and N. Padmanabhan (1984) •A spark counter for counting of alpha tracks in SSNTD films'. IARP Bulletin, Vol. 7, No. 1. pp 39 - 42. (4) ICRP - 32 (1982) 'Limits for inhalation of radon daughters by workers'. Pergamon press, Oxford. - 354 - Table-1 Technical specifications of-passive radon dosimeter Detectors (Passive) a) For radon-HH2 Cellulose Nitrate film (LR-115, Type II, Kodak) b) For gamma radiation.Thermoluminescent dosimeter CaSCL(Dy) in teflon discs Both the detectors are mounted on an aluminium card. Detector packaging Cylindrical aluminium chamber (59mm dia x 35' mm depth) with matching top and bottom cover. Loop provided on the top cover for fastening. Total wt. - 100g Membrane Latex, 50 urn thick, permeability co-efficient 6.34 x 10"** cm1 /s allowing 98.8. 7. of radon-222 and eliminating 99.97. of radon 220. • Equilibibration time-42 min • Efficiency of etchable '41.9 '/•• track-, registration in the C.N. film used Lower limit of detection -.15.6 Bq.h/1 Exposure period .6—8 weeks Track . counting Spark counter for track density less than 5000; microscope for track density over .5000 Calibration equation • E =!0.554 T. Ei=^ exposure (Bq.h/1) T • = track / cm*. Response 2.06\'track. cm .1/Bq.h Equation- for estimating WLM 4 8.81 :< I0" .F.-T radon daughter exposure F ijs equilibrium factor in terms of working 8.8(1 x lO"** is the conversion level month ' (WLM) t rw.t. ico-ef f i c ient radon exposure ; • - 855 - Tc^le-II AVERAGE ANNUAL EXPOSURE QF OCCUPATIONAL WORKERS (SURFACE) UCIL JADUGUDA Cate of Locat i on MEAN EXPOSURE (rr.Sv/a) work place Total exposure Contribution (alpha + gamma) o'' - • • °~:a Mean S.D. Mean S.D. GROUP 'A' AREA E.S. Laboratory 62 0.43 ,26 0. 12 (0-5 mSv/a) Hospital 04 1.80 .41 O. 02 Estate Office 53 1.43 .69 1. 10 Adm. Building 06 .99 .92 O. 41 R.M. Shop 58 .61 .86 o,3 3 Time Office 07 .95 .78 34 Garage 93 .08 .97 0.3 2 Central Stores 99 1.00 .72 o.4 1 Water Treatment PI. 09 0.80 .53 o.0 2 Site Office 16 0, 60 .70 0.21. 3 C.R. & D. Lab. 93 0. 21 .71 0.30 Byproduct Plant 64. 0. 27 .63 .37 R.U.R.P. 3.38 0. 38 .00 .50 S.U..R.P. 2.22 0..4 3 .97 .51 M.U.R.P. 2.89- 1. 45 .19 .90 GROUP "B'AREA Crusher House 8.02 3.11 1.39 0.30 (5 - 15 mSv/a> Mill House 12.21 2.80" 5.51 2.24 Drum Filter Area 11.35 .4.34 3.09 0.57 Ion-Exchange and Precipitation .Area 6.54: 0.55 1.60 .0.46 Pyrolucite and Lime Plant . • 5.80 2.62 1.77 1.51 Tailings Treatment Plant • 6.58 0.29 1.82 0.40 GROUP 'C AREA Disc-filter Station 41.12 13.36' 34.87 15.1 ( > 15 mSv/a) Product Recovery Area . • 19.11 4.97 13.43 3.27 Tailings Pond(Pump House ) 17.56 3.16 1.27 0.09 - 856 - 0:1 2 cm i ^_i i PERFORATED PROTECTIVE CAP tEv m w nr w n M kg CT m KI ^nrrr RUBBER "L MEMBRANE DOSIMETER /d( CHAMBER TLD LR-115 TLD AL CARD ..UfrtJ/nx^^. SEC-AA <£ZIZZZZIZ A H LA ki is wi M WM M n HI n M WJ J I AL- CARD ^iSiTOti'n'WwffBaJ ASSEMBLY Flq.1. PERSONNEL RADON DOSIMETER - 857 - RADIOLOGICAL AND ENVIRONMENTAL SAFETY ASPECTS OF URANIUM FUEL FABRICATION PLANTS AT NUCLEAR FUEL COMPLEX, HYDERABAD S.Viswanathan, B.Surya Rao, A.R.Lakshmanan and T.Krishna Rao Health Physics Unit Nuclear Fuel Complex Hyderabad Nuclear Fuel Complex, Hyderabad manufactures uranium dioxide (U02) fuel assemblies for PHWR and BWR operating in India. Starting materials are magnesium diuranate received from UCIL, Jaduguda and imported UF6. Both the products are first converted to ammonium ov^uranate, calcined, and reduced to U02. U02 powder i9p*precompacted, granulated and finally palletised and sintered. Sintered pellets are machined, cleaned and loaded into zircaloy tubes which are then sealed. Sines tha uranium processed here is free from its daughter product activities, external radiation from gamma activity is negligible. Inhalation of airborne uranium compounds is one of the main sources of exposure. The critieality hazard in handling slightly, enriched uranium (Max. 2.66% enrichment) is very low and is taken care of by built-in control based an safe geometry and concentration. The workers are provided with protective respirators wherever required. They are regularly monitored for axternal exposure and also for internal contamination. Soil, water and air from NFC environment ar» routinely monitored to control the release of pollutants and asses'fe the impact on the environment'. RADIOLOGICAL HAZARD CONTROL VENTILATION Ten air changes are maintained by supply air and exhaust fan^ in wet section , filtered and washed air supply in all dust proof areas. All the equipment: - 858 - which generate dust are enclosed and kept at negative pressure by local exhausts. Manual charging of uranium, oxide powder .was replaced.by pneumatic conveyers to avoid the release of uranium to:work atmosphere. To control the "release of uranium into the environment the dust laden air is filtered through wet rotoclones. Recently, prefliters and electrostatic precipitators with trapping efficiency of more than 957. are installed in the exhaust systems of UOP & EFFP respectively. CONTAMINATION CONTROL The plants are divided into (X) Controlled area which comprises of production of U02 and assembly sections (2) Supervised areas which comprise of change house, office and H.P. rooms. Controlled areas are subdivided into contaminated area where uranium compounds are handled in dispersible form and non- contaminated areas where sealed fuel pencils/bundles are handled; The shop floor and walls in controlled area are made smooth and non—porous for easy deccntaminatirn. Vacuum cleaning and wet mopping are. regularly carried out to prevent accumulation/spread of uranium contamination. EXTERNAL RADIATION EXPOSURE" • All," the personnel in the controlled • area are monitored for external exposure by TLDs. The1 external dose received, by "the" personnel during 1985-1788 is given it table—I . : The." average annual dose equivalent per person during the last" 4 years varied from 1.07 mSv to 1.53 mSv and |the.corresponding plant dose varied .from- 0.49 manSv to 0.66-manSv. The average plant dose" per'ton'b.f fuel produced during the last 4 years\ works out to be 8 rnilli manSv. AIRBORNE URANIUM Periodic-monitoring of airborne uranium in work atmosphere was conducted, to evaluate the occupational exposure and.to identify sources of release of uranium far effecting engineering controls. Average airborne activity in different sections is given in table-II. The data reveals that airborne uranium levels in all the sections except in grinding and blending sections are within derived'air concentration limit. Leakage of U02 powder frcm the grindei and discharges.of U02 from the blende*' give rise to high 'air activity. The abovR operations arc- carried out in closed•roams and personnel ent.r\ is restricted' during the operations - Plant personnel 'are regularly .monitored for urinary uranium excretion and uranium lung, burden.- - 859 - The results are given in Table III and IV. Urinary uranium values showed the body burden less.than-0.2 tolerance value. -'''T{ie'data on uranium lung" burden- shows, that the average burden is 10 mg which is only 257. of the maximum permissible value thereby indicating the contamination control is satisfactory. Measures have been taken to bring down the release of uranium to the atmosphere by introduction of electrostatic precipitators and absolute filters. INDUSTRIAL HYGIENE SURVEILLANCE . Personnel engaged in uranium processing, in addition to radiation hazards,- are expos.ed to various^ chemicals used in. the processes.; Regularjinidnitoring -or' the work atmosphere 'reveals' that the levels"'- of " chemical pollutants (viz NOx, NH3 8c HND3 mist) in different active"plants are well within TLVs. WASTE MANAGEMENT Sizeable quantities of liquid and.solid wastes are generated in the various processes.: These wastes are monitored for uranium cpntent prior' to disposal. SOLID • WASTE^ Uranyl nitrate raf.finate cake (UNRC) resulting from the neutralisation of raffinatB with NaOH contains uranium ranging from 0.1% to 0.87. (average 0.4X). In'the, beginning UNRC was packed in steel !driims and buried ,in ear.thern. pits. Sirice J.986 UNRC is sent to UCIL for recovery of-uranium.' , LIQUID WASTE Total.quantities of liquid effluents generated from uranium plants during 1984 to 19B8 era given in the Table-V. Liquid effluents from natural uranium plants mainly consist of NH4N03 and NaN03^ They are chemically.treated to bring down uranium level below 50 mg/1, the value arrived on the basis that the salts once used as fertiliser (NH4N03) for a crop should not- raise the soil uranium level to more than l/30th of the background level. These effluents are mostly sold to private industries for production of salts for use in glass industries/explosives. The analytical data reveals ihat the uranium level on an average within .10 m.g/1. The-; quantity of effluents produced per ton of fuel varied from 23 m3 to 33 m3j N.H4T- from enriched uranium plant, is .treated with lime to immobilise fluoricia and the slurry is taken to - 860 - Solar Evaporation Ponds (SEP) for concentration and storage. ENVIRONMENTAL SURVEILLANCE STORM DRAIN WATER Besides the process' effluents lean effluents consisting of floor washings, accidental spillages etc., generated from different plants carry chemical pollutants like chloride,.nitrate and uranium. This water is constantly used for garden within NFC after removal of suspended materials and insoluble materials in the settling tanks. Before pumping to the garden • system, the water is analysed to assess its suitability for garden use (ISs 2940). Mostly the water was found suitable for gardening. WELL WATERS Wells in and around NFC are monitored at regular intervals for their chemical quality to assess the impact of the constant use"of lean effluents for gardening. The levels of uranium.in well waters are given in Table—VI. Uranium levels in wells within NFC as well as in the neighbourhood are observed to be below 15 ug/1, the limit proposed for drinking water by EPA. Further no increase in uranium level in well waters was observed over the past 17 years. Background nitrate levels in well waters from this region ranged from "io.to 250 ppm. 50JC of the wells in the. region outside the passible influence-of NFC showed nitrate levels in water above the WHO recommended permissible level of 45 ppm. Use of solar evaporation ponds for storing and recovery of salts from effluents and reuse of storm drain waters for insite gardening prevent any possibility of nitrate pollution in the neighbourhood. SOIL Uranium level in soils within NFC ranges from 1.0 to 3.9 ppm, the highest being from the Area very close to uranium plants. By controlling the emission from the exhaust and prohibiting the staring of contaminated materials out side the plants' premises, contamination of the sail can be avoided. AIR Analysis of air samples collected from different zones in NFC were analysed for uranium content. The uranium level ranges from 2.3 to 13.0 ng/m3 which are far below the permissible level of S00 ng/m3. - 861 - CONCLUSION Surveillance of work atmosphere and environment indicates that the occupational radiation exposure is insignificant at NFC and high quality of safety is observed in the fuel fabrication plants. ACKNOWLEDGEMENTS The authors are grateful to Sri S.D. Soman, Director, Health and Safety'Group for his keen interest in this work. The authors thank Dr. K.C. Pillai, Head, HPD and Dr. I.S. Bhat, Head, ESS for their guidance in. this study. Thanks are also due to NFC management for their support. REFERENCES 1. Viswanathan,S; Surva Rao,B; Lakshmanan,A.R and - Krishna Rao,T. Health and Safety aspects of • Nuclear Fuel Complex for the years 19B4, 1985, . 1986, 1987 and 1988. 2. Lakshmanan,A.R; Krishna Rao,T and Viswanathan,S Nitrate and Fluoride levels in Drinking waters in the Twin cities of Hyderabad and Secunderabad. Indian J. Envirgn. Hlth, Vol. 28, No. 1,1986 3. Nitrates, Nitrites and N-nitroso compounds Environmental Health Criteria - 3. World Health Organisation, Geneva, 1978. 4. (Richard Cothern, C, William L. Lappenbusch and Joseph A. Cotruvo. Health Effects Guidance for Uranium in Drinking water. Health Physics Vol. 44 No. 1, 1983. - 862 - TABLE-1 EXTERNAL RADIATION EXPOSURE 19B6 1987 1988 Total dose 0.69 0.66 . 0.52 (man Sv) Average 1.49 1.53 1.13 dose(mSv) Dose per ton 8.57 9.58 7.66 of production (man mSv/t) • • T A B L E - 2 AIRBORNE U ACTIVITY AT DIFFERENT SECTIONS (Bq/m3) Section 1985 1986 1987 1988 1989 (a) Natural Uranium Plants • DP area" 1.21 1.60 0.92 1.05 0.99 Grinding* 2.80 3.05 1.74 0.55 1.88 Blending* 2.45 3.32 2.80 1.59 2.33 Compaction 2.11 2.07 1.98 1.66 1.26 (b) Enriched Uranium "Rlants DP area 0.33 0.30 . 0.35 0.36 0.45 Srinding* 0.84 0.98 2.16 2.34 0.82 Blending* Compaction 1.38 0.92 1.61 1.41 1.31 * Respirator areas DAC : 2.0 Bq/m3 - 063 - T A B a E - 3 URANIUM THORAX BURDEN (mg) No. of persons having U thorax burden Range 19BB 19B9 (mg) Nat.U Enr.U Nat.U Enr.U 0-10 38 10. 18 21 10 - 20 11 1 1 1 20 - 30 - 1 1 - 30 - 40 2 — - - > 40 "" *"• — — Maximum permissible thorax burden Nat. U: 40 mg Enr. U: 28 mg . T A B L E - 4 BIOASSAY FOR URANIUM Range No. of persons having U body burden (tolerance) 1985 . 1986 1987 19SB 0.J3 - 0.2 119 : .127 191 123 0.2 - 0.5 •1 1 3 0.5 - 1.0 r- - - > 1.0 — - - One tolerance = 50 Ug U/dm3 of urine. T A B L E - 5. GENERATION OF LIQUID EFFLUENTS YEAR Fuel Effluent Effluent Uranium Production " (m3) (m3)/(t) in. the (tones) effluent (g/m3) (a) Natural Uranium Plants 1984-85 58.15 . :1Q33 31.5 7.8 1985-86 62.48 •2174 34.8 4.0 1986-87 64.60 :2154 33.4 4.5 1987-88 56.21 !1411 25.1 7.9 . 1988-89 51.74 ;1302 25.2 6.1 (b; Enriched lUraniuin Plants- 1.984.-85 16.16 899 55.-6 25.0 19BS-8o 14.56 854 58.6 16.2 1986-87 15-36• 1050 68.4 • 17.6 1987-Qt 12,80 852 66.6 19.9 1988-8c;' 16.16 703 '43.5 15.9 - 864- - T A B L E - 6 LEVELS OF URANIUM IN GROUND WATER AROUND N.F.C Dire Approx. Uranium(ug/1) — ction dista Description from nce 1985 19B6 19B7 1988 1989 NFC from NFC (Km) NFC Openwell-1 6 4 5 5 8 NFC Openwell-2 3 4 3 4 5 NFC OpenweI1-3 3 4 — 6 6 NFC Openwe11-4 3 3 4 • 5 8 NFC BoreweU-1 - 5 9 . 9 . 9 NFC Borewell-2 4 4 5 5 . 9 NFC Borewell-3 4 5 4 6 7 NFC Borewell-4 . 3 3 3 3 - NFC Borewell-5 2 - — 3 5 NFC Borewell-8 6 6 7 5 7 DAE Colony BW'C N 2 3 2 — 3 3 DAE Colony BW'D' N 2 2 2 - 2 3 Kamala Nagar N 3 3 3 — ' — 4 H.C.L BW 3 E 1 2 3 — — 4 H.C.L BW 4 E 1 2 2 — •' - 3 Ashok Nagar ' E 1 4 3 3 3 2. Kushaiguda NE 4 2 5 - - 8 Charlapally . NE 4 3 8 — - 2 Mallapuram (R) S 1 2 2 — - 3 Railway Bridge s 1 3 7, 5 5 4 Meerpet w 2 4' 6 — - 2 - 865 - LIMITS OF PLUTONIUM CONTAMINATION IN REPROCESSED URANIUM FOR HANDLING IN URANIUM PLANTS . V.K.Sundararo and M.R.Iyer Health Physics Division Bhabha Atomic Research Centre Bombay-400085 The use of depleted uranium in plants handling natural uranium calls for setting up limits of plutonium and trans-plutonics, since their presence would pose additional handling h/izards.. The limits should be such that the health physics precautions adopted for (natural uranium should be adequate'for handling the reprocessed uranium. For this purpose Derived Air Concentration (DAC) values for mixtures of Plutonium and transplutonics in reprocessed fuel from PHWR reactors of various burn-ups and cooling periods have been arrived at. All the alpha emitters that build up in the fuel are taken into account. The activity of the various radionuclides are computed using the isotope generation code, ORIGEN 2 and the DAC values are taken from ICRP Publication 30. The DAC of the mixture is found to be more or less constant with burn-up for cooling times - 866 - more than 2 years. At shorter times the DAG of the mixture is found to increase with burn-up as Cra-242 with a half-life of 163 days and a relatively higher DAC of 4 Bq/ro3 forms a significant fraction of the activity. Based on the DAC for the mixture, limits of Plutonium contamination in reprocessed uranium are arrived at. One of the posssible criteria that can be used for obtaining the limit is that the additional radio-toxicity due to the presence of Plutonium and transplutonics is equal to the toxicity of uranium. On this basis the limit is found to be lower by a factor of 2 for uranium with a burn-up of 6500 Mwd/TeU than that with a burn-up of 1000 Mwd/TeU. ' - 1. Introducticr. Handling of depleted reprocessed uranium, in uranium; metal and fuel fabrication" plants calls for setting up; limits of Plutonium and transplutonics 3ihce their, presence would po3e additional handling hazards. The• limits for these have to be set such that the health physics precautions adopted for natural uranium should\ • be adequate in handling the reprocessed uranium of various types received from fuel reprocessing plants. It' has become necessary to study the variation'of DAC.of! the mixture as a.i function of. the burn-up of the reprocessed fuel to provide guidelines for setting, the limits of plutonium in the uranium. - 867 - 2. Method of calculation. In order to quantify the relative toxicity of the transplutonics,the basis of computation.followed is to arrive at the alpha activity of the mixture and the equivalent plutonium that contribute just one additional DAC when uranium is. present to the level of one DAC. It i3 assumed that there is no preferntial separation of plutonium or trans-plutonics in the uranium processing operations. In the following the term 'mixture' refers to all the alpha emitting nuclides other than the fuel actinides U-234.U-235 and'0-238. The DAC for the mixture of trans-plutonics is arrived at using the rule for mixture DACmix .=; 1/2 [fi/DACi] Bq/m3 where . fi is~ the fraction of activity of the ith component of the 'mixture with DACi. DAC values corresponding to the more restrictive W class are taken from ICRP-30. The activity inventories of the various nuclides for different burn-ups and cooling periods are computed using:the isotope generation code 0RIGEN-2 /!/ with the cross-section set appropriate to the reactor type. The gross alpha activity, Amix per gram of uranium IK given by - 868 - Amix = Su / [DACu/DACmix] Bq/g of uranium where Su = specific activity of uranium, Bq/g (2.5x104 Bq/g) \ DACv = DAC of uranium, Bq/m*' From the gross alpha activity, the total plutonium equivalent QPU is given by QPU = Ami* x £239 / [Sz39 x E233] g/g of uranium where f23 9,S239 and E2 3 3 are the activity fraction . of Pu-239 in the mixture, specific activity of Pu-239 (2.3x109 Bq/g) and isotopic fraction of Pu-239 appropriate to the burn-up and cooling time as obtained from ORIGEN 2 code. 3. Results and Discussion Calculations have been done for PHWR fuel with burn- ups varying from 3000 to 12000 Mwd/TeU and for 1000 Mwd/TeU of CIRUS fuel each cooled for different periods from 1 to 10 years. All the alpha emitting nuclides are taken into account. Variation of DAC of the alpha emitting mixture as a function of irradiation and cooling time for PHWR fuel is shown in Fig.l. The increase in DAC for short cooled mixture is due to the presence of larger fraction of Cm-242 with a half-life of 163 days and a DAC of 4 Bq/m3 compared to a DAC of 869 - 0.08 to 0.09 Bq/m3 for all the other major components viz. Pu-238,Pu-239,Pu-240 and Am-241. At longer cooling times the DAC of the mixture is seen to remain more or less constant. The alpha activity, Ami* to give one DAC, worked out from this shows a similar trend. Plutonium concentration, QPU decreases with burnrup and cooling time as illustrated in Fig.2. This is due to the build up of Am-241 which forms a significant fraction of the alpha activity of the mixture. Pu-239,Pu-240 and Cm-242 are 'ihe other major contributors to the alpha activity of the mixture". In Table I is presented the composition of .alpha activity, of the mixture in PHWR fuel with a typical exposure of 6500 Mwd/TeU and in CIBUS fuel with a burn-up of 1000 Mwd/TeU. Since the computations take into account the growth of Am-241 activity, the equivalent Plutonium to give:one DAC shows a decreasing trend in the high burn-up fuel. For the average burn-up of .6500 Mwd/TeU and 5 years cooling period, the quantity of Plutonium works out to 38.nano-grams/g of uranium. In the.case of reprocessed CIRUS fuel with 1000 Mwd/TeU exposure, the growth of Am-241 is not significant and the value turns out to be 80 nano-grams/g of uranium even though the DAC of the mixture is the same as that of the reprocessed fuel from PHWR. The Plutonium quantity to give one DAC is found to be lower by a factor of 2 in the case of high burn-up fuel. Plutonium equivalent to give one DAC thus arrived at can form the basis for working out the limit of Plutonium that can be allowed in reprocessed uranium. „ - 870 - f. 4. Acknowledgement The authors would like to thank Dr.K.C.Pillai for his keen interest in the work. 5. References • /l/ ORIGEN 2 : A versatile computer code for calculating the nuclide compositions and characteristics of nuclear materials A.G.Croff.Nuclear Technology, Vol.62,Sept.1983. - 071 - Table 1 : Composition of major components of alpha emitters in PHWR and CIRUS fuels Nuclide DAC Reactor Fraction of alpha activity Bq/m3 type after a cooling time (years) 0 1 3 5 10 Pu-238 0.09 PHWR 2.60-2 6.61-2 9.05-2 8.37-2 6.90-2 CIRUS - - - Pu-239 0.08 PHWR 1.05-1 2.40-1 3.25-1 3.05-1 2.62-1 CIRUS 8.63-1 8.69-1 8.67-1 8.65-1 8.59-1 Pu-240 0.08 PHWR 1.29-1 2.88-1 3.90-1 3.66-1 3.14-1 CIRUS 1.12-1 1.11-1 l.llrl 1.10-1 1.10-1 Am-241 0.08 PHWR 4.14-3 4.75-2 1.61-1 2.33-1 3.47-1 CIRUS 2.14-3 5.49-3 8.51-2 1.48-2 Cm-242 4.0 PHWR 7.30-1 3.49-1 2.14-2 1.03-3 CIRUS 8.63-3 1.86-3 - Utrt- :<'h\ 0.:>- PHWR .'3.67-3 7.85-3 9.86-3 8.58-3 6.07-3 CIIUIS Ho to Rcasl :'..G0-2 as 2.60*10-2 (.'j.n,,i-; on.- i.onsi than 0.1 %' not (.-resented above - 872 - 0.3 47 a 0.2 -O 3 CO < 0.0 0.0 5.0 10.0 15.0 Burn-up m 1000 Mwd/ton of U Ftg.l s DAC as a function of irradiation Note ! Indicated years are cooling times - 873 - 70.0 i r e " 50.0 o> o c 10 c 1 year o 30.0 3 years 5 years c 10 years ID o 10.0 0.0 5.0 10.0 15.0 Burn-up in 1000 Mwd/ton of U Fig.2 : Pu concentration to give one DAC of transplutonics as a function,, of irradiation Note : Indicated yea" 3 are cooling times - 874 - JlO.j. •.vi'l'iOC; Of U.A.iU.. SL iSAJt A. A. .••.A^ius, n. i-*,.«i.UiiiUi:Ljii»\;i, /. iC.xia;-u<.M-.iJR?;iY and ft.K. oAriiOvRAJ Atonic ilin-.rals Division Department of Atomic Energy Hyderabad - 500 016. jaccharom.ycos Ceroviaiae, species of yf«i«t HK8 isolated in yur L'lv? e.'feot of bivalent iomi like .-nan aii^u.-, uia.'jne.jium and trivalent ions.i like ferric iron :ind "5+ ulunaniuin on bio-jorptioti waa studied. Fe >10 ppru, '5+ 2+ '-'+ A1^^25 !>*MII, Mil >500 n:«a and r./:'"^pO p.jrn oonc-n Ovation in'ilbi-trtd t-»e biosr-orption capacity. The iron^nt ytudics r-y-ul l. INTRODUCTION The application of microorganisms as biosorbants for heavy metals have a great potential as an alternative to the methods currently used for decontamination of the industrial waste waters or effluents from uranium mill. The use of algae, fungi and bacteria in scavenging", the metals from municipal waste water have earlier been reported. Rothstein and Meier (1951) reported that the surface of yeast cells contain reactive groups which are chemically similar to high molecular weight pyrophosphates and capable of adsorbing heavy metal ions. It has also been reported earlier that Car^oxyl (-C00~), sulphydryl (-SH~) and phosphoryl (-P04 ) groups present as reactive sites on the cell walls have metal binding capacity. (Dounce and Flagg, 1949) . Norris and Kelly (1979) described the use of yeast for accumulating nickel and copper and found it to be better than alagae. Galun .£±. al., (1983) successfully used Penicillium precultured mass for recovering uranium from solutions. Beyeridge and Murrey (1978) studied the cell wall of Bacillus subtilis and reported that the elements have affinity towards the cell wall for attachment. Shumate et al., (1978) and Strangberg et al., (1981) tested the Saccharomyces Cerevisiae and Pseuodomonas aeruginosa as biosorbants for uranium. Both microorganisms accumulated uranium to the extent of 10-15/4 of the dry weight. - 876 - Stricter controls art Trradually being enforced on the discharge of metals into the environment. Since conventional metal recovery techniques are becoming increasingly expensive, potential use of microorganisms need proper investigation. Our investigations were directed towards the removal of uranium by biosorption from eluates rich in uranium and the acidic leach liquors and ion exchange barren solutions. The use of microbial cells as biosorbant for heavy metals has been proved to be a rapid method since in this method pre-cultured biomass is used for the metal recovery. PiATSillAL AND H.ii'.-iUi)3 Species of yeast Saccharomyces Cerevisiae used in this study was isolated from a Hyderabad variety of grapes. Glucose yeast extract medium having compositioni glucose » .10 g/i, yeast extract » 3.0 g/l# malt extract a 3.0 g/1 (Winge and Roberts, 1988) was used for isolating and maintaining'Saccharomyces Cerevisiae in pure form. It was identified by adopting established procedure (Lodder, J. 1984). The culture was regularly subcuiturod on glucose yeast extract agar slants. To get rich ceXLjnass of yeast, 3 1 of the glucose yeast extract broth was inoculated in Haffklne's flask and Incubated on the rotary shaker for 7 days. The cell mass was obtained by centrifugation at 10,000-12,000 rpm for 15 minutes. It was observed that the average yeast cell yield (on dry weight basis) from 3 1 of culture was 0.5 g. Dry cell weight was 3 taken by air drying the cell suspension at 80°C in an oven. During the biosorption experiment roughly 0.5 g (dry weight) of yeast cell mass in suspension form was ta"ken for 100 ml of the uranium solution. The cell mass concentration was increased correspondingly, where high concentration o£ uranium in solution was taken. Before inoculating, the yeast cell mass was washed twice with sterile distilled water to remove traces of nutrients and extraneous•- matter. Solutions with known concentration of uranium were taken in duplicate flasks. The initial pH and volume were measured. One flask was inoculated with 0.5 g of yeast cell suspension and shaken for 1-2 hours on rotary shaker. The second flask was kept as blank for comparison purpose. After the required time of incubation the liquor was centrifuged and the solution was filtered through bacterial filter Whatman GF/C to obtain cell free solution. The pH of the liquor was observed and the uranium value In the liquor was estimated. The uranium loaded yeast cell mass was treated with 50 ml of 0.1 M ammoniumcarbonate solution for about 1-2 hours for desorption of uranium. The cell mass was separated by centrifugation which could again be used for bios'orption. To study the effect of bivalent and trivalent ions, salts of ferric sulphate, aliminium sulphate, manganese sulphate and magnesium sulphate were taken in known concentration along with the experimental solution and the uranium Absorption capacity of the cell mass was determined. - 878 - RESULTS AND DISCUSSION During the adsorption cycle the uranium content of the liquor decreased by 80-94% (Table 1) . Pig. 1 shows the comparative biosorption values with three different species of microorganisms. In the desorption stage about 80% of uranium could be recovered from the yeast cell mass. The desorbed yeast cell mass could be used again for biosorption from fresh solution. Thus the yeast cell was found to behave like an ion exchanger, The accumulation of uranium by Saccharomyces Cerevi3iae is a surface phenomena. Yeast cell walls have pyrophosphate reactive groups which are capable of adsorbing metal ions. Uranium biosorption occurs by complexation of uranyl ion (UO_2+ ) with negatively charged reactive sites (-P04~ or -COO ) on the cell surface of yeast. The use of various algae for this purpose has been reported (Sterriff and Lester, 1979}, for the removal of Se, Mo, Ra and U from lead mine effluents. Bhurat et al., (1979) demonstrated the use of filamenteous Aspergillus fumiqatus in the treatment of uranium mill effluents. The present experiments have demonstrated that uranium absorption by yeast cell mass may prove to be highly suitable for the removal of uranium from dilute solutions and is applicable for waste water treatment purpose. However, it has been-noticed that - 879 - bivalent ions like manganese (>500 ppm ); magnesium (>50 ppm ); and trivalent iona Fe5 ( 10 ppm ) and A1^ (>25 ppm ) inhibited the uranium sorption capacity by the cell mass (Fig. 2)* Conclusion She use of microbial cells as biosorbants may prove to be an alternative to existing methods of recovery of metals from a variety of industrial discharges or as a tool for pollution control. Shis ie particularly pertinent to effluents containing radioactive nuclides. ACKNOWLEDGEMENTS She authors are thankful to Mr. A.C. Saraswat, Director, Atomic Minerals Division, (AMD) for permission to present the data and Dr. 3. Vievanathan, AMD, for the encouragement and to Mr. K.K. Dvivedy, AMD, for his invaluable guidance during the course of this investigation. Thanks are also due to Dr.R.K. Malhotra and his oolleagues for the uranium analytical data. - 880 - REFERENCES Beveridge, T#J«# Williams, F.M.R., and Koval, jr.J., (1978) . . The effect of chemical fixatives on cell wall of Bacillus aubtili3. Canadian J. Microbiol,v.24, p,l43$-j451. Bhurat, M.C., .Jayaram, K.M.V., Mathur, A.K., and Dwivedy, K.K. (1979). Studies on microbial precipitation of uranium. Ind. Jr. Microbiol, v.12, p.158-162. Dounce, A.L., and Flagg. J.F., (1949). The chemistry of uranium compounds. McGrawhill, Mew York, p.55-145. Galun, M.B., Keller, P., Fieldstem, H, Siege1, S., and Seigal, B. (1983). Recovery of uranium from solution using fungi.II. Release of uranium loaded PeniciIlium bioma8s. Water, Air, Soil. Pollu. v.20, p.277-285. Lodder, J. (1985). Yeast classification (Ascomycetes) North Holland Pub. Amesterdam. Norris, P.R., Kelly, D.p., (1979) Accumulation of metals by bacterid and yeast. Dev.Ind.Microbial, v.20, p.299-308. Rothastein, A and Meier:, R. (1951). The relationship of the cell surface metabolism. The chemical nature of uranium complexing groups of the cell surface. J. Cellphysioi.v. 38, p.245-270. Shumate S.E., Rancher, C.W., Strandberg, G.W.in& parrot;'/ J.R, (1978). Biological removal of metal ions from aqueous process streams. Biotechnol. Bioeng. syntp. v.8, p.13-20. Steritt, R.M., and Lester, J.N., (1979) Mineral Environ. v.l, p.45-47. - 881 Strandberg, G.W., Shumate, S.E., and Parrot:, J.R. (1981). Microbial cells as blosorbants for heavy mfttals. Appl. Environ. Microbial, v.41, p.237-245. Winge, 0., and Roberts, C,, (1958). The chemistry and biology of yeast, Ed., A.H. Cook. Acad. Press, New York. - 882 - Table 1 BIOSORPTION OF URANIUM BY YEAST SI.No. Experimental Uranium conc.ppm Contact Biosorption time hrs. solution Initial Final % 1. Eluate 64.5 9.8 1 88.0 2. Eluate 159.0 30 2 81.0 3. Eluate 68.5 0.7 2 99.0 4. Leach liquor 129 68.5 2 47.0 5. Uranyl sulphate 14.9 1.5 1 90.0 solution 6. Uranyl sulphate. 16.4 0.7 2 96.0 solution - 883 - CONTROL ^ EXPERIMENTAL Vs 14 (A PSEUDOMONAS Sp. SACCHAROMYCES ASPERGILLUS CEREVISIAE FUMIGATUS Fig.1 DEPLETION IN URANIUM CONCENTRATION BY DIFFERENT MICROORGANISMS < >- CQ Tx *fc "k, •a. Z X u- < o I I o. < o B O KB S CD "-in fc 5S •Sis! s o r-4 en i r 'Noiidtosoia HniNvm v. - 885 - Session VIII DISCUSSIONS Paper No. 1 R.P. VERMA : What is the contact time allowed for adsorption of radium values,in the pyrolusite - barytes beds? K.P. EAPPEN t The flow rate was on an average 20Lpm. The contact time would be around 5 - 10 min. R.P. VERMA : In actual operation, silt will come along-witb the tailings effluent water and this silt will deposit over the bed and may make it ineffective. Paper Ho. 2 M.C. SUBBA RAMU J Radon daughters attach to aerosols and the deposition of these aerosols leads to inhalation hazard. For evaluating the activity to do»e conversion factor, AMAD of radon daughter aerosols must be known.. What is the AMAD value in Jaduguda uranium mine? G. JHA : Although the radiation exposure rlsh are mainly due to radon progeny, baslo potential for exposure under various conditions has been characterised from the measured concentration of radon and equilibrium status of the daughters by formulating a mathematical equation WIM - 7.77 x 10~4 ?.T. Where WIM la the unit of exposure to short-lived daughter products of Rn in working level month. F is the measured equilibrium factor. T is the track density and 7.77x10~4 is the proportionality constant. This prodedure is as suggested by ICRP in its publication. No. 32. 886 - Paper No. 4 G. S. MURTHY : V/bat ia the method of estimation adopted by you to determine the concentration of uranium in effluents and in the atmosphere. 3. VISWANATHAN J In the effluent uranium estimation is done by fluorimetric method after initially separating it by solvent extraction leehhique. In air, aftei- allowing for the decay of short lived radon and thoron daughter product, uranium is estimated by counting, (1.5 dpm = 1ug of U (Natural)) R.P. VERMA : Have you studied the migration of uranium in underground water? What is its range? Io it absorbed by the soil ai d migration is retarded? S. VISWANATHAN : Migration of uranium in ground water is very low and is generally retained in the soil. L.M. GANTAYAT : The limit for uranium in liquid effluents of EUOP is around 25g/M5 which is higher than 7g/M5 for natural UOP. why? 3. VISWANATHAN i This could be due to the presence of fluoride ion in the effluents from EUOP. Pape r No. 5 L.M. GAMTAYAT » Have these values on DAC for reprocessed uranium been found prtctical in reprocessing plants? what are the limits? V.K. 3UNDARAM i This study is only for developing a mathematical method of calculations. We do not have any practical data with us. - 887 - SRIVASTAVA : When the natural UOP of NPC processes depleted uranium what should be the limit for plutonium? K.C. PILLAI : I may be able to give some information. What has been reviewed isthe upper limit for Pu contamination in depleted uranium which is processed in plants like NPC. On the basis of the calculations shown the upper limit has been fixed at 0.4 ug Pu/g of U. S. VISWANATHAN : Before processing the depleted U the. Pu values are determined for each drum of concentrate. Paper No. 6 B.V. Shah : What is the specific capacity of the bio-organism tested by you? A.K. MATHUR : The yeast has a capacity to adsorb uranium to 10—15% of its dry weight from uraniferous liquors . e.g. 0.5g of dry yeast used in our experiment could adsorb 80-90 mg U,0fl from ion exchange eluates. K.P. HAPPEN : What are the parameters like acidity or temperature gradient etc. to keep the reactanta effective for adsorption. A.K. MATHUR : The experiments were done on acidified leach liquors, eluates and barren liquors at pH,2.0-3.0, and at ambient temperature (30° - 35°C). L.M. GANTAYAT : Do you expect the bio-aorption process to be cost-effective in concentrating the low pride ores. A.K. MATHUR J Since yeast can be produced on industrial scale it can be obtained cheaply. Hence, thf process may be co3t. effective. SESSION IX 1. HEALTH a SAFETY ASPECTS 2. GENERAL CHEMISTRY OP URANIUM Chairman : Dr. D.2). SOOD B A Eu Reporteurx Dr. .P..K. KAPOOR A E R B - 888 - OPERATIONAL HEALTH PHYSICS EXPERIENCE AT URANIUM METAL PLATO, TBOMBAY P.P.VJ. Nambiar, Pusbparaja and J.V. Abraham Hoaltb Physics Division Bbabba Atomic Research Centre Bombay 400 085 INTRODUCTION Hataxal uranium Is handled in large quantities, in the form of concentrated eolutionst dry powder and metal ingot at Uranium Metal Plant, Trombay. This poses potential radiation as well as chemical hazards to workers and the environment• During the xeoent expansion of the plant capacity, considerable changee have been introduced in the equipment and processing. Magnesio Thermio Beduotion (MIR) replaces caldothermic redaction to produce metal ingot* A separate slag processing section was introduced to treat the magnesium fluo ride slag. Mixer settlers replace columns in the purification stage. This paper present* the operational health physios experience during the year 1988. Brief accounts of two fire incidents that occurred are also given. HAZARDS Handling hazards of natural uranium have been discussed in an earlier report* ' • The potential hazards associated with the produc tion of uranium, metal at the plant are » l) Chemical and xadlotoxioity of uranium ii) toxicity of various ohemloals used or generated in the prooess (e.g. bydxofluorlo aoid, ammonia and N0X fumes) an* ill) fire hasard due to pyrophorio nature of uranium metal in turnings ox powder form. - 889 RADIOLOGICAL SURVEY Area Monitoring! The beta-gamma radiation field In various aeotlona of the plant la measured regularly a sing appropriate radiation survey meters* The gamma radiation level in noat of the areas lies b«low 3 mr/h. Beta-gamma doae rates are comparatively higher (8 - 40 nr/b), particularly In diBobarge room and ingot storage areas* A separate enolosure hag been recommended for chipping and storage of ingots so that exposure to workers can be reduced* Personnel exposure * About 145 persons working in the plant are issued TLB badgeB on a quarterly basis* Total doae received by the Institution during 1988 was 29*9 rems giving an average individual annual dose of 206 mrea* Maximum external exposure at the plant during the year V&B 865 mren. Air aotivity levela » To estimate air activity levels of natural uranium, air samples were taken regularly at various locations using portable pumps* The looglived alpha aotivity levels were deter mined and the results are presented in Table 1* Solubility classi fication as applicable to the different areas is also given in the last column of the Table* During the actual powder transfer operations the air aotivity values are near about the DAC values of 10 Bq/nr (w class) and 0*6 Bq/nr (Y class)* However, respirators are used during such transfer operations* Particle size studies were carried out in the plant in various sections using 12 mm cyclone sampler as well as Anderson •ampler* AMAD values were calculated using long-lived alpha activity oount rate on the filters* 'xbe values are reported elsewhere and a typical set of data is given in Table-2. An average AHAD of 5 p was established for the plant for the purpose of modification of DAC values* - 890 - Determination of HP/K0T concentration : Concentration} of HP in nydrofluorination area and HO fumes in dissolution eeotion and oateide the plant were determined using bubbler samples. The samples were oolleoted in 0.1 H NaOH at tbe rate of 1 lpm and the concentration of HP/W0X was estimated by ion-seleotive eleotrode/spectrophotometrio methods The concentration levels during normal operations varied from 0.2 to 1 mg/m* while TLV for.HP and NO are 2.5 ng/a' and 9 mg/ir respectively. However XL valnes were exceeded at times during maintenance of HP furnaoe or opening of HP line in hydro- fluorination area. Surface contamination » The floor of reduotion and by defluorina tion furnaoe areas is painted with epoxy for easy decontamination. Powder spillages oocur during the changing of bunkers and transfer of VO, from tbe trays into oans« Alpha contamination levels in the / 2 plant vary fron 10-15 dpa/ca in Ammonium Di»ranate (ADU) and Caloination section to about 15 - JO dpm/om in Dissolution platform. Alpha contamination levels in laboratory flooring are in the range 2-8 dpm/cm . V B&2 slag powder gets spilled in 11TB aeotion during lining, charging, discharging and hammer mill operations. Begular cleaning is neoessary to avoid accumulation of this slag powder whioh also contains 2 - 3 £ uranium (Y class) • A new plant Is being commi ssioned for treating this slag to recover uranium and magusium fluoride. BioaBsay i Workers who are actively engaged in handling uranium at various stages of operation are issued urine sample bottles every year. The analysis of their urine samples Indicates any intake of D & W olasses of uranium. The results are expressed as fraotion of recommended limit (PBL). Few cases where inhalation of oxides are - 891 - suspected are eent to whole body counting unit for thorax oounting. Rirty nine samples were collected for bioaseay during the year* Maximum value of PEL reported was 0.09. However, urine samples collected following a fire incident on 26.12-88 showed some intake of nat. uranium, giving a Committed Dose Equivalent (CDg) of maximum 980 mrema to one worker from UMP • Keeping in view the recent ICHP recommendations (ICH'-M) and the emphasis on chemical toxioity of uranium the frequency of urine analysis for ol&ss D & W compounds needs review* Waste disposal * Processing of uranium oonoeotrate involves many precipitation and separation stages which generate sizable quantity of solid and liquid effluents* Solid wastes oonsist mainly of repulp cake (fiffc) and xeffinate cake (UHHC) and the liquid waste is made up of ammonium diuxnate filtrate (ADO?) and raffinate filtrate after neutralisation* Two hundred and thirty four drums weighing about 300 Kg eaob were monitored and tagged for disposal to Waste Management Division (WD) • The maximum beta-gamma done rate measured outside the drams was 19 mr/h. About one million litres of liquid waste was generated with a total of 28 mCl of gross alpha aotivity and 53*5 mCi of gross beta activity* DffiLETED URAHIUM PE0CB33IHC About 8 tonnes of depleted uranium in 2.6 x 10* litres was also processed in the plant .ntaring Sept. 87 - March 88. Prooedure and necessary precautions for safe handling of depleted uranium were discussed and detailed out before starting the processing. The plutonium-239 (Pu-239) and fission produot (iP) activity levels in the solution varied from batch to batob and was In the range 0.02 - 0.13 ug Pu/g of uranium and 0.06 - 0.3 uCi of fission produot/g of uranium. The permissible levels of Pu and IP in depleted uranium for handling in UUP were earlier fixed at 0.4 ugPu/g - 892 - of uranium and 0.4 uCi IP/g of uranium* The gamma speotrometrio studies of the samples indicated that the major IP contaminants in the solution were Zr-95. Nb-95t Bu-106 and Hb-106 isotopes. Air samples taken during the processing did not have any unusual high activity levels* Cose rate on the feed tanks and mixer settlers were about twice the dose rates normally encountered (2-3 mr/h) while processing natural uranium. The UNB cake drums showed dose rate of 23 mr/h while one of the drums showed 190 mr/h. The higher dose rate in this case is mainly because most of the fission produots in the raffinate is precipitated in the neutralisa tion stage* URANIUM MACHINING During 1938, maohining of 12 ingots was taken up in the workshop area of the plant. Heoessary safety precautions were recommended to reduce radiation exposure as well as fire hazard. This included perspex hood arrangement to the lathe to proteot the operator from flying turnings without obstructing the view of the ingot and copious flow of cooling oil during lathe operation to reduoe fire hazard. Protective gears inoluded goggles* rubber gloves and respirators. The uranium ohips generated daring machining were stored Immersed in coolant under oonstant supervision • Air activity in the vicinity was of the order of 1 - 1.8 Bq/nr. Beta dose on the controls of the lathe was 4 mr/h. Urine samples of the operator did not show any abnormal aotivity (< 0.05 PEL). Thorax counting of the operator shoved thorax burden wbion is one tenth of the thorax burden limit. UNUSUAL INCIDEST3 1. There was a fire inoident in the Main Ball on 24-10-1988 involving uranium metal kept in a drum to be taken up for dissolu tion. The fire was put out by the Plant personnel using dry ponder and CO. fire extinguishers, fire brigade was also called. Power - 893 - supply was cat off for some time as a precautionarr measure Air sasplea collected from the main ball ahowed long-lived alpha activity in the ranga 2-2 - 5>7 Bq/"r • *** eaaploa ware also colleoted on the next day while the affeoted area was taken up f0r decontamination. The airborne alpha activity levels were in tbe range 5*4 to 6 Bq/a . Transferable alpha surface contamination levels in tbe area varied from 10 - 15 dpm/om • Ho personal contamination was deteoted. Brine eaople oolleoted from the personnel involved in tbe fire fighting and olean up operation did not show any aigaifioant intake of uranium. 2- Another fire incident occurred in dissolution sootion (26-12-88) during dissolution of uranium turnings' About 9 metres of PEP process exhaust duot burned out damaging also electrical oablee and fittings in tbe area. Plant power supply was put off and personnel were evacuated as a precautionary measure. lire brigade was called in to put out the fire* t Air samples were colleoted after restoration of power and counted for long-lived alpha activity. It -»ae found to be in the range 3-10 Bq/m5. The dissolution platform showed alpha surface contamination in the range 12 - 240 dpm/om and the area around showed alpha surface contamination of 10 - 60 dpm/om . Tbe Fire Brigade personnel and their equipment sere monitored till dean. Exhaustive monitoring and decontamination of the plant was conducted after establishing rubber station and tbe area was oleared for free access after a few days* Urine aasples were oolleo ted from all the porsonnel involved in the fire fighting and olean up operations. Tbe analysis of the samples indicated Intake of uranium by two persons involving a committed dose equivalent of 9.8 mSv (980 mrem) and 12.2 mSv (1220 area). - 894 - The cause of the fixe «as Investigated by a epeoially conotitutoa committea, and also by Safety Coamittee. It is ouspeo- ted that the nitric acid taken for dissolution was not adequate for completion of the reaction. This resulted In the exposure of undissolved uranium on draining out the solution from the tank. This might have caught fire initially, spreading the fire to the FHP duoting. The turnings dissolution has been suspended till a proper equipment was designed for the purpose* CONCLUSIONS The general radiation field and tbe air aotivity.levels are low ;ln:tb»6different areas;o£/tbe planet* However, air aotivity is significant in areas such as ingot discbarge rood where respirators are in use* The ventilation in these areas as well as in Dissolution Seotion needs to be augmented* The job recently taken up by tbe plant, viz* machining of uranium.is giving rise to accumulation of uranium turnings in tbe plant. This poses a potential fire hazard and steps need to be evolved for speedy conversion of these metal turnings into its non-pyrophorio compounds* ACKNOWLEDGEMENTS Thanks are due to Dr. K.C* Pillai, Head, Health Physics DiviglonKand Dr. H.B. Iyer, Head, Radiation Hazards Control Seotion. for their continued interest and useful suggestions • REFERENCES 1. PUSBPARAJA, 3* SOMASUNDARAM, P.P.V.J. HAMBIAR, S-B. WATAMWAR and R* RANGARAJAN, 'Safe Handling of Natural Uranium', BARC/l-879, Bombay, 1986. 2. PUSHPARAJA, P.B. SAW.ANT, J*V* ABRAHAM and P.P.V.J. NAMBIAR, ipartiole Size Distribution of Aerosols in Uranium Metal Plant' Bulletin of Radiation Protection, Vol. 11 No* 3 & 4, July-Dec. 1988. - 895 - Table 1 Air aotLvitr levels in various iBeotion g Air activity levels Solubi si. Location (Bq/m3) lity Ho. Average Maximum class 1. Reduction furnace 1.9 4-56 W 2. JITB 1.53 7.73 W 3« HP furnace 2-36 10.00* w 4^ Discharge Boom 0.41 0.72 Y 5- ADJj/Calcination , 3.36 10.70* W 6. Dissolution 0.45 0.60 w 7. Slag processing 0.88 1.80 . Y "8. Charging 1.90 3.6 W • Samples were collected during the actual powder transfer operation. DAC for W class conpounds 1 10 Bq/n' DAC for Y olass compounds > 0.6 Bq/or - 896 - Table 2 AHAD values at different locations (Anderson Saqpler) Sr«Ho. . Looat.'o^ of saipling AKAD Geoaetrlo etd. (H deviation 1. ADO preeipitatioa and 5*5 2.0 calcination 2. Heduotion Furnace 6.3 2.7 3' HP Furnace 9,0 3.2 4. Charging & MTE 5.3 2-4 5- Slag proeeaaing 5.6 2.2 - 897 - RADIOACTIVITY LEVELS IS THE PROCESS STREAMS OF URAHIUM METAL PLAHT AT TBOMBAY Pushparaja, S.G. Sahasrabudhe, J«V. Abraham, P.P.V.J. Nambiar and M.R. Iyer Health Physios Division Bhabha Atomic Research Centra Bombay 400 085 IHTRODUCTION Magnesium diuranate (MDU) concentrate received from Jaduguda is processed at Uranium Metal Plant (UMP) to obtain nuolear pure uranium metal for use in research reactors* The concentrate contain ing about 70 percent uranium is dissolved in nitrio aoid and the filtrate forms the feed solution for purification by solvent extrac- tion. The extraction raffinate containing impurities is neutralized and filtered* The two liquid waste streams generated are* (i) the raffinate filtrate whioh is sent to Waste Management Division for disposal, and (11) ammonium diuranate filtrate (ADUj) which Is directly drained to sea after gross alpha beta analysis* A gamma speotrometrio study has been oonduoted in order to Identify the important radionuclides present in the effluents* This paper presents the results of the analysis carried out* RADIOASSAY Aotivity content > The daughter proouots of uranium gets eliminated in the initial processing of uranium at Jaduguda. Depending upon the time elapsed after separation of uranium and degree of purity, the beta emitting immediate daughter products of uranium, via. Tborium-234 and Protaotiniura-234i exist in different levels of equilibrium at various stages of metal production. Another radionuclide of importance from environmental safety point of view is Radium-226 whioh is present in MDU In traoe quantities* - 898 - Method » Ton millilitre samples in 25 ml precalibrated glass bottles were taken for analysis. Gamma spectra of liquid samples were taken using 54 co 4K Coaxial Germanium intrinaio detector with resolution of 2 K«V at 1332 KeV. The activity in the samples wae determined using any of the energy valued ' given in Table 1. As eeen from the Table, Uranium-238 haa two low-energy gamma ray emissions which have interferences' from other radionuclides- Hence the total uranium in the samples was determined by the Chemical methods^2'5'. tfranium-235 and Radium-226 both emit gamma rays of energy of.about 185 KeV. Uranium-235 activity can be estimated using the alpha speoifio activity of natural uranium of O.67 uci/g and the known composition of uranium isotopes in natural uranium (vis. 99*28 j£ of Dranium-238, 0.72 % of Uranium-235 and 0.0055 $> of Uranium-234) • Eadium-226 activity was determined by subtracting the estimated contribution from Uranium-235. BESOMS AND DISCUSSION The concentration of the radionuclides in the process streams is given in Table 2« AB seen from the Table, the activity of the two daughter produots (Tb-234 & Pa-234) in the feed solution as well as in a standard uranium solution is in equilibrium with the parent uranium aotivity. However, in the effluent streams (Sr.No. 3, 4 & 5) the daughter produots are in super equilibrium with the parent. This indicates that tbe daughter produots are preferentially removed at the various stages of processing and tbe uranium concentration cannot be estimated using tbe measured activities of. Tborium-234 and Protaotinium-234 in the effluent streams. Chemioal analysis (by B.S. Seotion, HJ>. Division) of the solid rafflnate oako showed significant amount of Hadium-226 (.—115 BqAg) • The concentration of Hadium-226 in the raffinate filtrate after treat ment is 2*2 Bq/ml. - 899 - The studies indicated that Radius-226 in liquid effluents does not poae any significant radiation hazard. Howevert the beta activity levels due to TborluB-234 andProtaotiniua-234 are signi ficant in ADO filtrate and needs to be controlled before its di&osai. to the sea* ACKHOWLEDGBfflSTS Thanks are due to Br. K.C. Pillai, Head, Health Physios Division for his continued interest in the vork. BESEHBICES 1. B. Browne and H.B. Firestone, 'Table of Badioaotive Isotopes'* Bd. V.S. Shirley* wlley-Interaoience, 1986. 2. W. Davies and W. Gray, Talanta, 11 (1964) 1203. 3. J.C. Ingleo in 'Uannual of Analytical Methods for tb© Cranium Concentrating Plants', Method U-3» Mines Branoh, Sept. of Mines and Technical Surveys, Canada, 1959. - 900 Table 1 Gamma Energies of the Radionuclides Energy Branching Huolide Half-life EeV intensity 0-238 4.7 x 109 y 13*0 8.7 • 484) 0.07 Ib-234 24*1 d 13-3 9.8 63.29 3*9 92.38 2.6 92.80 3«9. Pa-234 1-17 » 1001 0.59 U-254 2.484x10s y 13.6 10.4 Ba-226 1600 y 186 3*28 8 U-235 7.038X10 y 143.7 11.5 163.8 4.7 185.7 57 -00 205.3 4.7 - 901 - Table 2 Activity levels (Bq/al) in the process streams of HUP 22 35 255 226 Sr. Type of . Tb-234 PB-234 W 0 U Ha Ho. sample Uraalua (92 Ktfl) (fOOl Ke7) (l85KeV) (calcu lated) 1. D-atd 190.7 189.7 185-7 194.5 8.9 185-6 2. XT-feed 2214 2363 1942 1781 103 1678 3- Haffi- 1.85 344-2 390 3-65 0.09 3.56 nate 4« AHJF 1.17 170 "" 164 0.75 0.03 0.72 5- Baffinate 0.87 filtrate L31 1.63 2.6 0.02 2-58 902 - RADIOLOGICAL AND CONVENTIONAL BAFETY ASPECTS OF MACHINING OPERATIONS OF URANIUM INGOTS V.B. JOSHI, I.K. OOMMEN, S.SENGUPTA AND T.B. IYENGAR HEALTH PHYSICS DIVISION BHABHA ATOMIC RESEARCH CENTRE BOMBAY - 400 085 This paper describes the radiological and conventional aspects and basic criteria adopted in the planning and operation of machining of Uranium Ingots at the Bhabha Atomic Research Centre, Bombay during 19B8-1989. The machining was carried out to remove the hard coating of magnesium fluoride slag from the outer surface of the Uranium Ingot, which was formed during the reduction process. Each ingot was of cylindrical shape <~25 cm dia x "Z0 cm ht.,> and weighed about 180 kg. A total number of 112 ingots, weighing 2.2 metric tonnes were taken up for the special machining operation. As part of feasibility studies a preliminary test run was carried out and various parameters governing the airborne activity and fire hazard were generated. By optimising the operational, safety facilities like use of appropriate protective equipment, ventilation facilities and sufficient and proper coolant flow, the airborne activity could be brought down to very low levels (fractions of DAC) and achieve ALARA principle in a practical way. The entire operation of machining, of 2.2 te of the material was spread over a period of about six months^ Tho collective dose for - 903 - the whole operation was 1.85 person mSv (0.185 manrem). The contribution from internal exposure as measured by bio assay techniques, was negligible. The operation was concluded with decontamination and waste disposal by following appropriate safety procedures. INTRODUCTION t The uranium ingots obtained during the reduction process have the hard coating of magnesium fluoride slag on the outer surface . This slag is normally removed by machining the outersurface of ingots mechanically. The machining of such hard coating requires heavy duty lathe machine and carbon indexable tools. As a test case adoption of facilities available in a non-radioactive area was attempted. This paper describes the radiological and conventional safety aspects and basic criteria adopted in planning and execution of such a Jab at a nonradioactive area like central workshop, BARC and the relevant optimisation t procedures. The major safety aspects covered are i) .fire hazard 11) chemical toxicity and ill) radiological health'hazard. URANIUM INBOTB AND JOB DETAILS I Weight per ingot i 80 - 200 kg Shape and size i Cylindrical, 25 cm dia and 30 cm height Surface dose rate t fl, T i 170 mR/h (1.7 mBy/h) Nature of job t O.D. turning and facing both sides Machine used i Heavy duty lathe machine (H-22) Tools used i Carbide indexable tools. - 904' - Coolant : Water soluble oil No. of operators Involved. : 14 Period of operation : 6 months No. of Ingots t 112 Total weight of uranium t 2.2 te. S. RADZ0LOBICAL AND CONVENTIONAL SAFETY MEASURES TAKEN i) Place of work i An isolated room (radiography room behind the main buildings of ii) Fire hazard i Any fabrication work on uranium metal produces finely divided forma of metal like turnings, filings etc. which ara pyrophoric in nature. They not only pose firs and explosion hazard but also pose inhalation hazard due to formation of oxidation product, U02» after burning. To generates various parameters governing firs hazard and airborne activity an initial test run was carried out with two Ingots. During the trial run the air activity detected was of the order of 26 Bq/m3 which was 37 times that of one DAC<2> (1 DAC - 0.7 Dq/m for UC-2> - This necessitated a proper planning and optimisation procedures. The reason fortha incroasad airactivity was found to be due to insufficient supply of coolant over the scraps which were getting generated. In subsequent test runs this supply was made adequate by using good quality of - 905 - circulating pump. There after the airactivity level shown was only a fraction of DAC Other, measures for fire hazard Hera collection of scrap in big metallic tray filled with coolant which was kept below the lathe machine, and periodic removal of the scraps from metallic trays to the drums filled with coolant. An appropriate type of fire extinguisher, 'Anul Powder', mixture of chlorides of Na, Ba, & K, was provided at the work place for meeting any emergency. iii) Radiation hazard : Before the start of work, all operators were given proper instructions for handling the radioactive material in general and uranium metal in particular. They were provided with chest and wrist TLD badges and protective equipements like handgloves, overshoes, dust respirators etc. and a temporary barrier was installed at the work place to avoid the spread of contamination. . Airactivity was monitored by using staplex air sampling pump during the entire period of operation. The general radiation dose rate level was "2 mR/h (0.02 mGy/h) near the lathe machine area. The evaluation of TLD badges (reported by Division of Radiological Protection) showed maximum whole body and skin doses as 0.65 mSV and 7.1 mSv respectively (Fig. 1). The collective dose works out to be 1.85 person mSv (0.185 manrem). Contamination was periodically monitored. The -lathe machine part3 such as control panels, check wheel etc. were showing alpha contamination of the order of 1400 to 4400 2 Bq/cm (Table II). The operators wore chocked for personal contamination. Although there was no contamination detected - 906 - on hands or clothes of operators, their handgloves and overshoes did show contamination of the order of 140 to 160 2 Bq/cm . The contamination was controlled by periodic cleaning of lathe machine parts, floors etc. and changing the handgloves and overshoes of the operators. The urine samples for assessing the internal contamination showed no activity (man. permi. excretion level - 50 pg/litre for Nat. U>. DISCUSSION j Since entire operation of machining of 2.2 te of uranium was carried out at a nonradioactive area like central workshop, every effort was made to keep the exposures to the operators as low as achievable and also the possibility of fire hazard to the minimum. Following steps (3) were effective In achieving the ALARA Principle j i) Sufficient flow of coolant was kept over the contact point of carbide tool and moving ingot which helped in reducing the temperature of hot scrap and theraby minimising fire and inhalation hazard. ii> Apporpriate steps were taken for storage and disposal of scraps to avoid"-fire and explosion hazard. Hi) Instructions were given to the'operators to avoid close proximity with the ingots to keep the beta-gamma exposure to the minimum. The use of thick gloves also was suggested while handling the ingot* to avoid dose to the fingers. iv) Air was continuously monitorrad to avoid the possibility of inhalation hazard by the operators. - 907 - Y> Contamination was periodicaliy checked. Decontamina tion operations were carried out to keep the spread of activity to the minimum. ACKNOWLEDGEMENT t The Authors are thank-ful to Shri S. Chellappa, 'formerly Assoc. Director. Eng. Serv. Group and Shri S.D. Soman, Director, Health and Safety Group, who have initiated the programme and provided guidance. -Thanks are" also due to Dr. K.C. Pillai, Head, Health Physics Division, 8c Dr. M.R. Iyer, Head, RHCS. for their keen interest during the course of this work. REFERENCES i 1) Safe handling of Uranium, Pushparaja, S. Somasundaram, P.P.V.J. Nambiar, S.B. Watmwar and R. Rangarajanj BARC/ 1-879 U986>. 2) ICRP U97B) Limits for intakes of radionuclides by workers, Publlcatian-30 and supplements. 3) ICRP (1977) Recommendations of the ICRP, Publication-26. - 908 - FIG i EXTERNAL EXPOSURE DATA (WHOLE BODY It SKIN 1 *) 0.7 0.66 0.6- / / 0.6- \ 0.4- / Ui 03 0.3- / Ut 0.2 ' 02 0.2 '1 ' Y 0.1 0.07 0.06 / 036 0.029 fcjoie? y£01 ka Efa p °p •• »pSLi U £a Eg •BE o o oo 10 11 12 13 14 PERSONS - 909 - Table-II Place Contamination lavel Remsrks Bq/cro < ot ) i) Uranium ingot surface 480 to 800 (before machining) ii> On operators hand gloves 160 sent to active Iaundry. On operators overshoes 140 Floors i) Adjacent to lathe machine 720 to 4200 { Periodic decon- < tamination done H> Near barrier 240 to 360 ( to keep the ( levels undei— iii) before barrier 20O < control. < Lathe Machine parts < ( i) Check wheel 140O to 4400 ( ( ii) controls 1200 to 3200 ( - 910 - RADIATION RISKS,MEDICAL SURVEILLANCE PROGRAMME AND RADIATION PROTECTION IN MINING & MILLING OF URANIUM ORES DR. A.K.RAKSHIT URANIUM CORPORATION OF INDIA USD P.;O.Jaduguda mines,Singhbhum, Bihar. Mining and Milling of Uranium ores comprise multiple operations such as development,drilling,blasting*handling crushing,grinding,leaching of the ore and concentration, drying, packaging and storing of the concentrate product*; Apart from the hazards of any metal mining & milling operations due to dust* Noise,Chemicals, accidents etc there are radiation risks also resulting from exposure to airborne radioactivity and external radiation.; The inhalation risk is of more concern in underground mines than in open pit minG3» The objective of a Medical Surveillance Programmo ( an occupational Health Programmed is ,to ensure a healthy work force. It should ultimately lead %o health maintenance and improvement, less absenteeism,increased productivity and the achievement of worker and corporate goals*' The progrsinnre Includes - 911 - prevention,acute care, counselling and rehabilitation. Radiological workers require special monitoring for their workrelated radiation exposure effect by film monitoring service,whole Body counting and Bioassay. Radiation protection in the mining and milling of Uranium* ores includes the use of personal protective Equipment,workstation protection,personal Hygiene and House keeping. INTRODUCTION Medical Surveillance Programme and Radiation protection of individuals working in Uranium mines and mills are concerned with the prevention of the occurance of non-stochastic effects, and limitation of stochastic risks to levels deemed to be acceptable. In addition to basic occupational health and safety procedures there are special risks for ion using radiation. Chemical toxicants,dusts,noise etc. needing programme to identify#mon.itor and control them. OCCUPYIONAL RISKS Radiation risks*- continuing activity in the mining and milling has resulted in the exposure of large numbers of workers to the radioactive elements. - 912 - Radiation risks result from exposure to airborne .• radioactivity and external radiations The inhalation risk is of more concern in underground mines than in Open Pit mines. The contaminants may exist in parent gas of as unattached daughter atoms, as daughter bearing dust particles in the mine atmosphere. In mills the risk is both due to external exposures and internal contamination depending on the areas of activities,; Non-radiation Risks*- Besides the radiation risks the workmen are exposed to the hazards of other Chemicals and work process e.g. Silica Dust in drilling*blasting and crushing. Dust in (Pyrolusite) manganese ore,Dicomal Dust in pre-coat area,Sulphuric Acid Leaching,Acid-.; fumes in Drum filter are a. Nitrous fumes in mines, mineral oil, Noise and humidity. HEALTH EFFECTS* The types of physical ailments encountered by the workmen may be divided as follows*- (A) Physical ailmont encountered by Radiation workers (B) General Occupational Health Problems. (C) Psycho-Social problems - Medical complications. - 913 - 2.1 The physical ailment encountered in Radiation workers include Lung Cancer* non-malignant Respiratory Disease ( Emphysema,Fibrosis/Silicosis), Damage to the Bone marrow and Kidney. 2.2 The General Occupational Health Problems include- Accidents including death,musculo-skeletal injuries, skin rashes,disorder of senses - vision, hearing,Heart Disease. 2.3 The psycho-social problems are substance abuse, depression etc. 3. MBDJCAL SURVEILLANCE PROGRAMME I "A Surveillance Programme is a strategy to detect a problem - it is never a solution. Medical Surveillance and Radiation Protection Programme consisted of two types of monitoring which are compli mentary to each others Health monitoring is being done by occupational physicians whereas Radiological Surveillance is. being carried out by the Health Physicists or Radiation Protection > specialists*' - 914 - 3o!l HBAIffH MONITORING includes (a) Medical Examination and (b) Analysis and management of working post.) 3.1.1 MEDICAL EXAMINATION is aimed at finding individual suitability for a particular type of work before joining any type of work and monitor the health of the worker throughout the individual's professional life at periodic intervals.Medical examinations are done as followsa- (A) Examination prior to assignment (Pre-Employment) •• (B) Periodical medical checks (C) Surveillance after actual or suspected over-exposure (D) Monitoring after suspension of work*! 3*1*2 The miners and other workmen have been under nwdical supervision since the operation of our mine in 1967-08. Although there were at that time no official regulati- ons laying down the type and frequency of the medical examinations performed,Chest X-Ray,haemogram,urine analysis and general clinical examinations were carried out up-on recruitment of workers and at intervals of 5 years subsequently*' - 915 - 3«11,3 MINIMUM EXAMINATION PROTOCOL which is being followed are - weight, height>blood pressure, pulse, cardiovascular* locomotor, respiratory Hematoganglions,Digestive,Endocrine-nutrition^ Abdomen, Neurology, Genito-urinary, skin, pregnancy^ Ear, Hose and Throat. 3.1.4 Other examinations inc.lude visual acuity.,jEye test. Chest X-Ray , spirometry,E.c.G,E0E.G,Audiogram and. psychotechnical tests which are carried out a3 and when required. 3.1.5 frequency of the examinational- The clinical examina tion including Chest X-Ray, Haematological and urine analysis are carried out once in 5 years according to the regulation of mines safety Directorate. The selective category of workers working in the areas of potentially hazardous activities are being examined more frequently, i.e. at 2 years interval. 3.1.6 The occasion had not arisen to effect detailed medical surveillance after actual or suspected over exposure or monitoring after suspension of work due to Radiation sickness. - 916 - 3.'2 ANALYSIS .AND MANAGEMENT OF WORKING POST:- maintenance of medical records of workman occupationally exposed along with their personal exposure and risk record is importanto Very little is known about the long range effects of low-level exposure to radiation in employment situation. At each clinical examination the medical findings are recorded on a special medical sheet.; 4.- RADIOLOGICAL SURVEILLANCE Radiological surveillance programme is carried out by Health physicists.: Regular Dust,.Noise , Chemical toxins 4 and Radiation surveys are being carried out by them,} Individual Dosimetry, TLD> urine analysis, who.le body count are being studied for the programme0f 5. RADIATION PROTECTION 5o'l Protective Ecmfoment. - This should not be considered as a substitute for good engineering control measure. Primary consideration should always be given to good ventilation and other measures, thereby preventing the air in mines and mills from becoming contaminated. The objective of the use of equipment is to protect the worker from radiation,dust,noise, ©ye injury at his work place or workstation. - 917 - PERSONAL PROTECTIVE EQUIPMENT*- The most used equipments are respirators,protective clothing, footwear,gloves and safety glasses. Respirators are not recommended for the control of exposure for routine operational conditions, but may have to be used in special situations where environmental" control is not practicable. While using respirators safety may be affected by visibility or mobility restrictions. All workers should receive training in the proper wearing and maintenance of respirators. PROTECT IVS CLOTH, FOOTWEAR AND GLOVES Work clothing protects the worker from exposure of the skin to contamination. Clothing from the point of view of radiation protection is generally not so important in Uranium mining and milling as in other sections of the nuclear industry. Normally the mildly radioactive uranium ore dust is not a significant source of external exposure. However,such clothing does not provide sufficiant protection against toxic meaterials.Personal protective equipment has to be maintained in clean and good condition to provide the desired protection. - 918 - 1.2 WORKS!'AT ION PROTECT ION t - The operation like crushing, rock breaking and transporting the ores etc. generate significant radioactive contamination in the form of ore or concentrate dust or radon and radon daughter releases. Most of the other operations are mostly mechanised. In the former type of operation specially in the case of high grade operations shielding the cab or remote controlled operators cab are being considered for underground mining in some places. 2 RAD3^I0N_HrGISNS/PERS0NAIi HYG33BNB A good personal hygiene programme requires that workers arriving at work change to work clothing. Handling of ora or concentrate with bare hands is not advisable. Eating in the work place is prohibited. Workers should change and shower after work. Keeping of long hair, beard3 and moustaches is to be discouragedo 3 HOUSEKEEPING - A good housekeeping is important and is a mu3t. Cleaning up of spilled ore and use of water sprays in dust creating area's are being done. There is a programme of periodical cleaning operations in our mines and due recognition is given for best housekeeping in the form of prizes annually.. - 919 5.4 EPIC AT ION - Practical experience Indicates that making workers fully aware of the occupational risk is a difficult task.The Implementation of a thorough, well designed find executed educational programme concentrating on the radiation hazards in the work place supported by good personal examples and good supervision is crucial to the success of utilisation of protective equipment and radiation hygiene programme. In our mines both the Vocational Training Centre and Health Physics Unit arrange regular worker education programme in batches in this regard. 6 COMMENTS* 6,1 RESUUTS OF ENVIRONMENTAL AND INDIVIDUAL SURVEYS AND MEDICAIi EXAMINATIONS!- The results of the Dust concentration/exposure to acid and other Chemicals and gases and radiation surveys in the different areas of mining and milling activities by the health physicists has established that the findings are well within the TI«V and maximum permissible levels at Jadugoda. The periodical medical examination and laboratory findings are in keeping with the absence of any disease or injury in the workmen which could be attributed to occupational exposures. - 920 - FREQUENCY OF MEDICAL EXAMINATIONS AND MINIMUM EXAM. PROTOCOL! - The medical examinations which are being carried out at 5 years interval,in our opinion, are adequate for the workmen,except the high risk groups of workers.For the latter category instead of a frequent complete medical examination,specific examinations selected according to the potentially contributory factors of hazards may be done at more frequent intervals e.g. 2 or 3 years. The idvtas are godning ground all over the world to.reduce the programme of medical examination to the minimal tests and the examination needed based on the concepts that workers should only undergo tests of direct and evaluable benefit to them. MEDICAL RECORD KEEPING I- We would like to emphasise the importance of maintaining Medical Records of workman. Because of bur lack of Knowledge on the effect in health of long-> terra lowlevel radiation exposure in employment situation the Importance of continued observation and statistical evaluation of medical records becomes Imperative. - 921 - 6.4 OCCUPATIONAL HBAUrH UNIT AND COMPUTERISATION. «- In the present system all the data related to personal (service.) records. Individual's occupational history and exposure, and the medical examination records are not centralised which is a drawback. In the course of the years;a considerable quantity of data is amassed in the medical file which becomes increasingly difficult to classify*store and utilise. An occupational health unit with the facilities of the use of computers is a basic rieed in such a situation and is highly recommended. B iBlilOGRAPHY 1. Cqrkill, Dave (1986) - Personal Hygiene and Protective, equipment. Interregional Trg.Course on Radiation Protection in the mining and milling of Radioactive • ores,Poco3 de Caldas^Bro. .zll. ' 2. Ghosh,D.K. (1979) - Evaluation of dust hazards during mining and milling of Uranium ores.M.Sc.Thesis, Bombay University. 3« Goyal J.P.- Medical Examination of Radiation worker - 922 - IAEA (1964) - Radiological Health & safety in mining and milling of Nuclear materials*. P^cn-eodings of a symposium jointly organised by XAB.&» X&C,WHO IABA(l9Q3) - Safety standards,Radiation Protection of workers in the mining and millim.T of Radioactive ores* IASA(l986) - What the General Practitioner(MD) should know about Medical Handling of over-exposed Individuals. Technical Document 366. IAEA(1987) - Application of Doze Limitation System to wn'iM) and milling of Radioactive Ores#Safety series No.82 IAEA(1987) - Radiation Protection in Occupational Health, Safety series No.83 Marcose P.M. etal(l978) - Distribution of Radium and Chemical Toxins in the enviroiment of a Uranium Complex. Mildon, C.Anne(1986) - Medical Surveillance of workers in mine and mill "'' Mildon,C.Anne(1986) - Practical operation of a Medical Surveillance Programme. - 923 - 12. Muir,David C.F.and Mastromatteo Emest(l986J - Medical Screening in Industry,Canadian Medical Assocn.Journal,Vol 135,Sept 15,1986 13. Raghavayya,M and Saha S.C. - Health Hazards associated with Uranium Mining in India. 14. Saha, S.C.(l97l) - Du3t Problem in Uranium Mining operation at Jaduguda. 15. Saha,S.C.,Ghosh,D.K.and Somayaji,K.:S. (1971) - Dust evaluation studies during mining and milling of Uranium*. - 924 - SEPARATION OF URANIUM(VI), THORIUM AND ZIRCONIUM BY SOLVENT EXTRACTION WITH CROWN liTHER N.V. DcORKAR AND S.M. KHOPKAR' Department of Chemistry Indian Institute of Technology, Bombay-400 076 A rapid and selective method was developed for the separation of uranium (VI), thorium and zirconium. Uranium (VI) was extracted with 0.02 M DC18C6 from 6.0 - 8.5 M hydrochloric acid. Zirconium was extracted with 0.025 M DC18C6 from 8.5 - 10 M hydrochloric acid while thorium was extracted with 0.065 M 18C6 in methylene chloride from 0.04 M picric acid. Hydrochloric, nitric, perchloric, sulphuric and /CXCGt/c ' acid were used for the stripping of the elements from organic phase. They were also separated from large number of binary mixtures. Uranium (VI), thorium and zirconium were also separated from multicomponent mixture containing cerium (III), scandium,- yttrium, hafnium, titanium and vanadium. The method was also applied for the analysis of uranium, thorium in monazite and zirconium in zircon sand. - 925 - Introduction Many procedures for the solvent extraction separation of zirconium (1,2), thorium and uranium (VI) are available however very few methods involve the use of crown ethers as the extractants. An attempt was unsuccessful to extract zirconium with DC1SC6 from hydrofluoric acid containing alkali fluoride (3). DC18C6 and DC24C8 were used for the extraction of thorium from picrate and nitrate media but such studies were confined to the study of the nature of species and type of bonding between metal and oxygen atom of the crown ether (4). The extraction of uranium from chloride and nitrate solution vith DC18C6 in dichloromethane indicated the formation of 1:2 complex with metal (5). l-phenyl-3 methyl-4-trifluoroacetate (6) and 2-thenoyl- trifluoroacetone (7) in combination with DC18C6 and 15C5 were used to extract uranium. This paper presents the systematic investigations on the solvent extraction separation of zirconium, thorium and uranium (VI) from picrate as well as hydrochloric acid media. EXPERIMENTAL Apparatus and neagents : A model 866C spectrophotometer (ECIL, India) with 10 mm, matched corex cuvetts, wrist action flask shaker were used. Stock solution of zirconium, thorium and uranium (VI); 0.500 gm of zirconium (IV) nitrate penta- hydrate (B.D.H. England) was dissolved in 100 ml deionized water containing !•/. nitric acid. On standardisation it was found to contain 1 mg/ml (8). 1.0 gm thorium nitrate tetrahydrate (8.D.H. England) was dissolved in 250 ml deionised water containing !•/. nitric acid. This solution on standardisation contained 1.68 mg/ml of thorium (9). A stock solution of uranium (VI) v/as prepared by dissolving 1.083 gm of uranium (VI) nitrate hexahydrate (B.U.H. England) in 500 ml of deionized water containing lyi nitric acid. - 926 - On standardisation (9) it was found to contain 1 mg/ml of uranium (VI). 18crown6, Dibenzo-18-crown6, Dicyclohexyl- 18crown6 and Dicyclohexyl l8crown6 (Merck, Germany) were used without further purification. General Procedure : An aliquot of solution containing either zirconium (25 |ig), or uranium (50 \ig) was taken. The hydrochloric was added so as to have its concentration as 6 - 9 M in a total volume of 10 ml. Then 10 ml of 0.0*2 - 0.25 M DC18C6 in dichloromethane was added the solution was shaken vigorously on a wrist action flask shaker.for.10 minutes. After separating the organic phase, zirconium or uranium was stripped with 0.5 M hydrochloric acid. Similarly to an aliquot of thorium solution (25 ug) picric acid was added to have its concentration 0.04 M. Then the pH of the solution was adjusted to 2.0 - 3.5 either with 0.05 M picric acid or lithium hydroxide solution; The solution was then shaken with 0.065 M 18C6 on wrist action flask shaker for 15 minutes. The organic phase was once again equilibrated with 10 ml of 0.5 M nitric acid to backstrip . thorium. Zirconium, thorium and uranium (VI) were determined in aqueous phase as their complex with arsenazo III at 665 (10), 660 (ll) and 6b5 nm (12) respectively. rtESULTS AND DISCUSSION Extraction of Zirconium and Uranium as Function of Hydrochloric Acid Concentration : The optimum acid concentration fqr the quantitative extraction of zirconium-and uranium (VI) was ascertained by extracting them with 0.02 and 0.025 M different crown ethers at various concentration of hydrochloric acid (Table 1). zirconium was extracted quantitatively, from 8.5 - 10.0 M hydrochloric acid with 0.025 M DC18C61 while the optimum acid concentration for extraction of uranium(VI) was 6 - 8 M hydrochloric acid with 0.02 M LC18C6. The other crown ethers were riot effective.. Thorium was. not : extracted with any of the crown ether frtyn 1-10 m hydrochloric ac id. - 927 - Extraction of Thorium as a Function of pH : The optimum pH for the quantitative extraction of thorium from 0.04 M picric acid was ascertained by extracting it at different pH with 0.065 M solution of various crown ethers in dichloromethane. The best pH was 2.0 - 3.5 for complete extraction with 18C6 (Fig. l). Effect of Crown Ether Concentration : The ideal concentration of DC18C6 for quantitative extraction of zirconium and uranium (VI) was found out by extracting them with varying concentration of crown ether in dichloromethane from 6.0 and 8.5 M hydrochloric acid respectively. While the optimum concentration for the extraction of thorium was ascertained by extracting it with varying concentration of 18C6 in dichloromethane from. 0.04 M picric'acid at pH 2.0 - 3.5 (Table 2). The best concentration was 0.025 M DC18C6 for zirconium 0.02 M LC18C6 for uranium (VI) and 0.065 M 18C6 for thorium. Effect of Picric Acid Concentration : The optimum concentration of picric acid for the extraction of thorium was ascertained by extracting it • varying concentration of. picric acid with 0.065 M 18C6. The best conceatration of picric acid was 0.04 M at 2.0 - 3.5 (Table 3). • effect of Diluents : Benzene, toluene, xylene, carbontetrachloride, chl'oroform, dichloromethane, dichloroethane and nitro benzene were tested for DC18C6 and 18C6 (Table 4). With exception of chloroform, dichloromethane, dichloroethane jnd nitrobenzene other solvents were poor diluents. ,;ffect of Stripping A'jents : Zirconium, uranium (VI) and thorium were stripped with mineral and organic acids in the concentration range of U.05 to 8 M (Table-£ ). Hydrochloric acid (0.5 - 6.0), nitric acid (0.5 - 5.0), sulphuric acid (l.O - 8.0) and - 928 - perchloric acid (1 - 8.0 M) were effective for the stripping. Uranium and zirconium were not stripped after 3 W and 6.0 M of hydrochloric acids respectively. Hydrochloric acio was preferred for the above purpose as the determinations of zirconium and uranium (VI) were carried out in chloride media. While thorium was stripped with 0.5 M nitric acid. Nature of Extracted Species : The composition of extracted species were ascertained from the plot of log D vs log [DC18C6] at fixed acidity. The corresponding slopes were for zirconium and uranium. The slopes of other plot of log D vs log [. .Cl~l at fixed DC18C6 concentration was 6 and 2 for zirconium and uranium. Hence the probable composition was [ITC^C^; 2DC18C6] and- + [ZrCl6.2H DC18C6]. Similarly composition of extracted species was ascertained by log-log plot of thorium. The probable composition of extracted species is 1:1:4 [Th(picrate)^,; 18C6], The findings are in agreement with earlier workers (4,5). Effect of Diverse Ions : Zirconium, uranium (VI) and thorium were extracted in the presence of various diverse ions. The tolerance limit was set as the amount of diverse ions: required to cause a variation of + 0.5/. in the recovery of elements. Alkali and alkaline were tolerated in ratios exceeding 1:50 and the other ions in the ratios of 1:20 for zirconium and uranium. Sodium, potassium, barium and scandium were tolerated in the ratio of 1:7 for thorium. Separation of Zirconium, Thorium and Uranium (VI) from Multicomponent Mixtures : A mixture of cerium (III), thorium and uranium (VI) was separated by extracting cerium (ill) from L5C5 (0.05 M) from U.04 M picric acid at pH 6.0 followed by extraction of thorium with 18C6 (c.06'3 M) from 0.04 M picric acid at - 929 - pH 2.0. The extracted cerium and thorium were stripped with 1 M perchloric acici and 0.5 M nitric acid respectively. A mixture of zirconium, uranium (VI) and scandium/ yttrium was separated by extracting zirconium and uranium from 8.5 M hydrochloric acid with CC18C6 (0.03 M) scandium remains in aqueous phase. From the organic phase zirconium was first stripped with 6 M hydrochloric acid followed by stripping of uranium (VI) with 0.5 M hydrochloric acid. The thorium, zirconium, hafnium and titanium were separated by selective extraction of thorium with 18C6 (0.065 M) from 0.04 M picric acid at pH 2.0 followed by extraction of zirconium with DC13C6 (0.03 M) from 8.5 M hydrochloric acid. Finally hafnium was extracted with DC18C6 (0.07 M) from 9.0 M hydrochloric acid when titanium was not extracted and remains in aqueous phase. A mixture of thorium, zirconium, uranium and vanadium (IV) was separated by selective extraction of thorium with 18C6 (0.03 M) from 0.04 M picric acid at pH 2.0. Zirconium and uranium (VI) were extracted with DCltiCo (0.03 M) from 3.5 M hydrochloric acid and stripped selectively with 6 M and 0.5 M hydrochloric acid respectively. Vanadium (IV) was remained behind in the aqueous solution. All the elements were determined by the spectrophotometrically as their complex with arsenazo III, xylenol orange, Alizarin reds, and tiron (Table 6). Analysis of Uranium, Thorium and Zirconium : Uranium and thorium was brought in solution by dissolving monazite (1.0 gm)- as per procedure described earlier (13). An aliquot of solution containino uranium was extracted with 0.02 M DC1SC6 from 6 M hydrochloric acid was stripped with 0.5 M hydrochloric acid and oetermined spectrophotometrically.. Under these conditions - 930 - cerium, thorium, yttrium and calcium were not extracted. The amount of uranium was 0.28/ against the standard value of 0.30/. Amount of thorium was ascertained by extracting it with 0.065 M 18C6 from 0.04 M picric acid at pH 2.0 and after stripping with 0.5 M nitric acid it was determined spectrophotometrically with arsenazo III. 0.10 gm zircon sample was fused with 2.0 gm borex. The cooled mass was extracted with 2 M hydrochloric acid and made upto 100 ml with deionized water. An aliquot of solution was taken and zirconium was extracted with 0.03 M DC18crown6 from 8.5 M hydrochloric acid. It was stripped with 6 M hydrochloric acid and determined spectrophotometrically as its complex with arsenazo-III at 660 nm. The amount of zirconium found was 65.6/ against 65.0/ . Conclusion The proposed method is simple, rapid and selective. The total period required for separation and determination is two hours. The separation of zirconium, uranium, thorium, hafnium, cerium, scandium is significant as they are.associated in fission products and minerals. The method is applicable to analysis of uranium, thorium and zirconium in real samples. - 931 - References B. Nandi, N.ft. Das and S.N. Ohattacharyya, Solvent extraction and ion exchange, ,1, 141 (1983). C.P. Vibhute and S.M. Khopkar, Analytical Letters, 16, 1037 (1983). R. Caletka, R. Hausbeck and V. Krivan, Talanta, 33, 219 (1986). W. Wang, C. 3ozhong, A. Wang, Yu Ming and Xiannian, He Huaxue Yu Fongshe Huaxe, 4, 139 (1982). W. Wang, C. Bozhong, J. Zhong-Kao, A. Wahg> J. Radioanal. Chem., 76, 49 (1983). S.K.. Mundra, S.A. Pai and M. 5. Subramanian, J. Radioanal. Nucl. Che.m., .116, 203 (1987). A.G. Godbole, N.V. Thakur, ft. Swarup and S.K. Patil, J. Radioanal. Nucl. Chem., 108, 89 (1987). F.J. Welcher, 'Analytical uses of Uthylene Diamine Tetracetic Acid1, D Van Nostrand Co. In Princeton, New Jersey (1958),, p' 184. A.I. Vogel, 'Textbook of Quantitative Inorganic Analysis', Longmans, London (1978), 4th ed., p. 483. S.V. Hlinson and K.I. Petrov, •Analytical Chemistry of Zirconium and Hafnium', Ann Arbor, Humpherey Science (1964), p 138. S.S. Sawin, Talanta, lj., 1 (1964).. J. Borak, L. Slovak, and J. Fischer, Talan.ta, 17, 215 (1970). N.V. Dsorkar and S.M. Khopkar, Analyst, 114, 105 (1989). - 932 - Table 1 : Effect of Hydrochloric and Concentration on Extraction of Uranium and Zirconium HC1 •/. Extraction Concentra 18crown6 DB18cxown6 DC18crown6 DC24i:rown 8 tion [M] U(VI) Zr(lV) U(VI) Zr(lV) U(VI) Zr(IV) U(VI) Zr(lV) 4.0 0%O0 0.0 0.0 0.0 16.0 0.0 0.0 0.0 4.5 20.00 0.0 10.0 0.0 27.2 0.0 14.7 0.0 5.0 20.00 0,0 16.9 0.0 45.4 0.0 23.0 0.0 5.5 26.00 0.0 16.0 0.0 90.9 0.0 51.2 0.0 6.0 29.10 0.0 21.0 0.0 100.0 0.0 79.0 0.0 7.0 29.10 0.0 20.1 0.0 100.0 10.0 83.0 0.0 7.* 29.10 0.0 20.1 0.0 100.0 15.7 83.0 7.0 8.0 29.10 8.7 20.1 7.6 100.0 67.8 83.0 20.7 8.5 11.30 16.1 7.9 13.7 98.6 99.9 76.1 69.0. 9.0 5.40 19.3 0.0 20.5 96.3 100.0 62.0 86.0 10.0 0.00 15.0 0.0 20.5 79.7 100.0 48.0 86.3 - 933 - Table 2 : Effect of Crown Eth Concentration on extraction of Uranium (VI), Zi onium and Thorium Crown ether . Uranium (VI) Zirconium Thorium concentra- DC13C6 DC18C6 18C6 tion 1.0 x 10"2M '/. E D /. E D •/. E D 0.5 67.0 2.1 10.1 0.11 20 0.25 0.8 75.0 3.0 33.3 0.5O 1.0 85.7 6.0 32.3 0.49 40.7 0.68 2.0 99.9 ' 999 66.6 1.99 61.5 1.60 3.0 100.0 oo 1OO.0 oo 75.7 3.01 4.0 100.0 oo 100.0 OO 81.8 4.50 5.0 100.0 oo 100.6 OO 85.7 6.00 6.0 100.0 oo 100.0 OO 88.2 7.50 6.5 . 100.0 oo 100.0 OO 100.0 oo 7.0 100.0 oo - - 100.0 oo 7.5 100.0 oo — — 100.0 00 - 934 - TaDle 3 : Effect of Picric Acid Concentration on Extraction of Thorium Picric acid -2 * E D 1.0 x 10 ^ M 0.6 11.5 0.12 0.7 16.6 0.20 0.8 27.5 0.38 1.0 44.4 0.80 1.5 77.7 3.5 2.0 88.0 8.0 2.5 94.0 15.9 3.0 96.4 27.0 3.5 97.6 41.0 3.7-4.0 99.9 999 - 935 - Table 4 : Effect of Diluents •/. Extraction Dielectric Diluent constant Uranium(VI) Zirconium Thorium DC18C6 DC18C6 18C6 s Benzene 2.28 20.0 22.0 25.0 Toluene 2.30 22-. 0 20.1 11.8 Xylene 2.38 46.0 12.0 11.8 Carbontetrachloride 2.24 50.0 20.9 0.00 Chloroforon 4.8 • 100.0 97.6 • 77.3 Dichloromethane 9.08 100.0 100.0 100.0 Dichloroethane 10.5 100.0 100.0 100.0 Nitrobenzene 34.8 100.0 100.0 100.0 - 936 - Table 5 : Effect of Stripping Agents „ + _* 4__ Ele- •/. Backstrioped otrippxng mont J^_ agents 0.05 0.1 0.5 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 : #~ U 21.3 50 100 100 100 100 710 15.0 0.0 0.0 0.0 HC1 Zr 63.0 79.7 100 100 100 100 100 100 100 98.0 68.1 Th 31.3 56.6 77 85.3 100 100 100 100 100 100 100 U 10.1 23 80 100 100 100 85 92.0 79.7 - - HN0? Zr 72.3 88.2 100 100 100 100 100 100 96.3 95 81.0 Th 39 56.7 100 100 100 100 100 98.9 96 92 81.0 U 26. 56 100 100 100 100 100 100 100 ICO ICO H2S04 Zr 57.1 81 99.6 100 100 100 ICO ICO 100 100 100 Th 67 92 99.8 100 100 100 100 lOO ICO 100 100 U 14 40 71 100 100 100 100 97 92 — — HC104 Zr 47.6 76 88.9 100 100 1U0 100 100 lOO 100 lOO Th 47.8 89.7 100 100 100 100 100 100 100 100 100 b 13.1 27 52 80 100 100 100 100 1U0 100 loo ,C00H Zr 47.0 58.2 79.7 10010 0 100 100 100 100 100 100 Th 25.7 60 80.1 10010 0 100 100 100 100 100 100 - 937 - Table 6 : Separation of Zirconium, Thorium and Urnnium from Multicomponent Mixtures Mix Extraetant Counter PH Amt. Stripping •/> \*ea g ent ture crown ether ion taken agents Reco for deter M eg very i•nin aitio n mg 3+ Ce 15C5, 0.05 0.04 ,Vi P.A. 6.0 25 1.0 M HC104 99.0 A, 655 4+ Th 18C6, 0.07 0.04 M P.A. 2.0 25 0.5 M HN03 98.1 A, 660 2+ uo2 Unextracted 100 Aqueous 100.0 A, 665 phase Zr4+ DC16C6, 03 8.5 M HC1 — 25 6.0 M HC1 99.0 A, 660 U02+ LC18C6, 03 8.5 W HC1 50 0.5 M HC1 98.0 A, 655 Sc/Y Unextracted - - 50 Aqueous ICO c, 525 phase 4+ Th 18C6, 0.065 0.04 ,V, i'.A. 2.0 0.5 M HN03 99.3 A, 660 Zr4+ L/C18C6, 0.03 8.5 iV. HC1 .'25 0.5 M HC1 99.6 A, 665 4+ Hf DC10C6, 0.07 9.0 m HC1 25 0.5 M HC104 99. u 4, 540 Ti4+ Unextracted - 5U Aciueous 55.1 390 phase 4+ Th 18C6, 0.065 0.04 M P.A. 2.0 50 U.5 M HN03 99.1 A, 660 Zr4+ DC18C6, 0.03 8.5 M HC1 25 6.0 M HC1 96.0 660 uof+ LC18C6, 0.03 8.5 M HC1 50 0.5 M HC1 98.7 A, 655 V02+ Unextracted _ _ 50 Aqueous 99.0 d, 540 phase A - ,wsenaz o III - a - Xylenol orange c - Alizarin Red-S, b - riron - 938 - 100- °/oE Concentration of Counter anion, x 10 M Fig.JiEffect of Counter Anion Concentration - 939 - URANYL ION TRANSPORT ACROSS TRI-n-BUTYL PHOSPHATE / n-DODECAME LIQUID MEMBRANES J.P.Shukla and S.K.Misra* Radiochemistry Division Bhabha Atomic Research Centre Trombay, Bombay - 400 085 Carrier - facilitated transport of uranium(VI) against its concentration gradient from aqueous nitrate acidic solutions across organic bulk liquid membrane (BLM) and supported liquid membranes (SLM) containing TBP as the aobile carrier and n- dodecane as the membrane solvent was investigated.Extremely dilute uranyl nitrate solutions in about 2.5M nitric acid generally constituted as the source phase.Uranyl transport appreciably increased with both .stirring of the receiving phase and the carrier concentration in the organic xembrane ,while enhanced acidity of the strip side adversely affected the portioning of the cation into this phase-Among the several reagents tested, dilute ammonium carbonate (r'lM) solution served efficiently as the strippant. INTRODUCTION Cation transport across various liquid sembrami configurations viz.,bulk liquid,emulsion liquid and supported liquid membrane (SLM) affording relatively- higher i'lui'.&s it: drawing increasing attention particularly for processing dilute reetal solutions for the separation and recovery of metals of critical and strategic importance, and for the decontamination of low and medium level radioactive .wastes.Potential practical applications of .such membrane have also been envisaged for the recovery of metals from hydroraetallurgical leach solu.tions ; plutonium and americium removal from nitric acid waste streams generated by plutonium recovery operations in Purex Process(1-4).Of them,SLMs consist of thin polymeric microporous or gel like films impregnated with liquids that enhance the transport of a desired permeant across the film and separation by this technique offer several advantages over classical solvent extraction processes,e.g.,(i) loss of extractant (membrane carrier) by entrainment is eliminated,(ii) high concentration factors can be accomplished (iii) extractant inventory is extremely low, and (iv) small amount of energy and limited space are required. Concerning the most troublesome nuclear ' waste management,partial neutralization of the acid waste followed by utilizing solid SLMs have been found useful in transfering and concentrating both Pu and Am from nitrate solutions(5).Likewise, transport of uranyl ions through SLM consisting of TBP in kerosene oil supported in Celgard 2400 polypropylene microporous film has also been recently reported (6).Using a * Fuel Reprocessing Division, - 940 - uranium flux of 12 lb per square ft per year, a cost of about 15c/lb of uranium recovered has been reported as against, estimated costs from ion exchange and solvent extraction processes ranging from 20 to 40 c/lb(7) .These estimates are thus found to be attractive for uranium recovery particularly from low - grade sources.Guided by these considerations, a comprehensive work programme has been initiated by vis to explore varied analytical applications of "liquid membranes for treating dilute actinide solutions emmploying some commonly available carriers. Of them,tri-n-butyl phosphate(TBP), has been commercially exploited for the recovery and purification of U(VI) and Pu(IV) from various matrices in acidic solutions using solvent extraction technique.In the present study,TBP dissolved in n- dodecane ,was selected as the mobile carrier in the facilitated transport of U(VI) across an organic bulk liquid membrane (BLM) as well as SLM. Effects of important parameters that affect cation flux in liquid membranes such as the activity of the permeant salt, feed acidity, TBP concentration in the organic membrane phase, nature and concentration of strippants in the receiving phase were evaluated. *Enka' Accurel polypropylene (PP) films were tested as the solid supports for SLM. Some silicon flat-sheet membranes with different inorganic fillers indigenously made by us were also tested for their possible applications. EXPERIMENTAL All reagents were, analytical grade products.U0a.(N03)a. 6H^.O was dissolved in desired nitric acid molarity. U-233 tracer was used throughout this study. To characterize the equilibrium properties of the carrier, uranium distribution coefficients (Du) were obtained by simply equilibrating an aqueous acidic uranyl nitrate solution with TBP/n-dodecane in an equilibration tube at room temperature.Details of glass BLM and SLM transport cells are described elsewhere(8).BLM cell of Shulman Bridge type consisted of a bulk TBP/n-dodecane phase separating two aqueous phases ,namly the source phase and the receiving .phase.Single - stage SLM measurements were carried out with a simple two compartment, permeation cell which consisted of a feed solution (12cm' )separated by a product solution chamber (2 cm )by a liquid membrane having an effective membrane area of 1.13 cma.The feed and product samples were mechanically stirred well at room temperature to avoid concentration polarisation conditions at the membrane interfaces and bulk of solution. Membrane permeabilities were determined by monitoring the uranium concentration radiometrically primarily in the receiving phase as a function of time. Uranium flux , Jm, was computed following the equation : Jm - Cu,feed x V / A x t where Cu « initial U concentration in the source phase,mol/drtT V " volume of receiving phase,dm3 A » effective area of the membrane,m t » time elapsed,s - 941 - The uranium solution used in the present study was deliberately kept very dilute, being in the range of 15 - 150 mg dm*"^ Membrane Supports Besides, silicon-rubber thin supports with different inorganic fillers like silica, calcium Bilicate ,calcium carbonate and zinc oxide indigenously fabricated in UED,BARC,two'Enka'Accurel PP flat- sheet type hydrophobic microporous polymeric membranes ,coded as 2E HF-PP and 2E-PP were also used .These membranes were about 130 - 180 jim thick,and had a nominal porosity of about 40%, with a pore diameter of the order of 0.2 urn.Filling the pores of these supportB with the carrier was accomplished by immersing the membrane in the organic phase.The pores were immediately and apparently quantitatively filled with the carrier simply by capillary forces.Such a SLM support disallowed the transport of water through the membrane and wore free from osmotic effects.The U (VI) - TBP complex could diffuse , across the membrane supported in the microporous structure. Sample membranes taken from different portions of the film gave permeation results reproducible to approximately ± 10%. RESULTS AND DISCUSSION TBP dissolved in n-dodecane served as the organic membrane in both BLM and SLM permeability measurements.Uranium (VI) is highly extractable by TBP from nitric acid media and therefore it permeated easily across these liquid membranes.Carrying out control experiments with no TBP in membrane solvent, clearly indicated that uranium flux through such membranes was generally negligible ruling out any possibility of leakage or transport by the membrane solvent itself* Of the different aqueous strripants such as ammonium carbonate, oxalic acid,and acetic acid tested, more than 70% of uranium could be recovered with 1M ammonium carbonate or more allowing a concentration of nearly six times. With rest of the strripants, maximum recovery achieved never exceeded 20%. Feed : Strip Ratio With a view to demonstrate that during the removal of Hctinides sufficiently high concentration factors could also be attained,flux measurements were carried out by keeping the volume of source phase (12 cm5)almost six times larger than that of the receiving phase (2 cm*). More than 70% of the initial uranium could be removed from the feed solution and was found,six times more concentrated in .the strip solution. In all permeation experiments, uranium flux always exceeded 10~7 mol/m 2 /» . Experiments thus performed with different feed to strip volume ratios clearly demonstrated that concentration enrichments of six or more could be achieved during the single-stage uranium removal process. - 942 - Permeation Through BLM Results for the single-ion transport of nearly 15-150 mg U dm from an aqueous feed adjusted to different nitric acid molarities through a TBP/n-dodecane BLM membrane into an ammonium carbonate strip solution are summarized in Table I.Maximum uranium permeation nearly over 70% took place from relatively lower feed acidity upto 2.5 M HNOjwhile its enhanced acidity to about 5M adversely affected the uranium transport and plummeted to low as 20% or even lesser.This is partly in accord with the expected trend since the flux of a cation varies with H*** ion concentration following the relationship: Jm - A I- [ H+3^ [ TBP ^ Cu, feed and hence there would be an increase in permeability with increase in proton or nitric acid concentration.Evidently this is true upto 2.5 mol/dm^nitric acid concentration. U0j+ions form UOz'NOsW^TBP type of complexes upto moderate acidity but above 6 m'ol / dm»or more, the nitric acid starts reacting with TBP to form complexes of the type TBP.m HNOJJO).Considerable decrease in uranium transport at higher acidity may be attributed to the fact that the flux depends upon the concentration gradient of diffusing species inside the membrane just adjacent to the aqueous solutions interfacing it. This depends on the distribution coefficient of UO^ions between the membrane and aqueous phases. Uranyl ion flux is associated with free nitrate and uranyl ions present in the source phase. The ratio of dissociated nitrate ions to undissociated HNO3 decreases with increase in HNO3 concentration. Moreoverfat higher nitric acid molarity uranyl ions may exist in the form of HUOatNOaJjj' which are unextractable by TBP causing fall in solute flux.In all the cases the cation permeability gradually increased upto about 4 - 6 hr of operation, but showed a slight decrease thereafter .Such membranes also transport nitric acid .Hence the nitric acid concentration of the strip solution increases with time and the driving force of the transfer,the nitrate ion concentration gradient1 is neutralized,When this occurs , equilibrium is reached and no further changes in uranium concentration can be expected.Interestingly, "the trend for transport of uranyl ions through the BLM is seen to be the same as that of this cation in liquid-liquid extraction systems. Composition of the organic solution! has a marked effect on a cation flux. When transport across a membrane occurs via a carrier, as in facilitated transport, the flux is generally expected to increase with increasing carrier concentration.However in facilitated transport of uranium with TBP, a more complex behaviour is seen. Table II summarizes data on uranium flux vs. TBP concentration in the membrane.With increasing carrier concentration in an inerc diluent like n-dodecane,the flux gradually increased reaching a maximum at about 30%. Here, both the amount of uranium chat could be extracted into the membrane and the viscosity of the organic solution" increased.An increase in viscosity of TBP solution may lead to decrease of the diffusion coefficient and hence - 943 - permeability of the diffusing species.These opposing effects resulted in a maximum permeation at about 30 % TBP.Above thi^ concentration, the permeation decreased with increasing carrier concentration. Studying the effect of uranium concentration in the teed solution ranging from 15 - 150 ag dm^Ln which the product side contained a negligible concentration of uranium, revealed that cation flux increased sharply with increasing uranium molarity tested upto this ranye.Thus the uranium flowed down to concentration gradient.However, as predicted by the theory, uranium will flow up its concentration gradient under appropriate conditions.Experiments were also performed in which the feed and product solutions initially contained oqual uranium concentration ( 50 mg dm~^) .The uranium concentration of the feed solution declined to less than 5mg dm"3 affording a final enrichment factor defined as the ratio of the concentration in strip to that in the feed solution.greater than 10. The difference in behaviour between experiments with varying feed acidity ,carrier concentration and uranium molarity can be understood by considering the probable expression for the rate of formation of the diffusing species at the feed interface: d [UOx( NO3 >2. • 2TBP ] / dt= k« [ TBP ]* ( NO 3 f~ { VO^] where k" is rate constant for the formation of the UO^" -TBP complex.This equation is based on the stoichiometry of uranium extraction,detailed elsewhere (6,7).Rate of diffusion of metal species will thus depend upon any changes in NO3 ,TBP and U02£+ concentration in the feed side. Permeation Through SLM Based on our above findings with BLM on the uranium transport , the SLM membranes were operated only at the optimum concentrations of feed acidity (2.5 M HNO3) ,carrier concentration (30% TBP) and the strippantdM ammonium carbonate). Under these conditions, more than 70% of uranium could be easily recovered employing Accurel 2E-PP membrane support.Relativly lesser recovery of uranium was accomplished with 2E HF-PP type support for which cause is still not clear. Though 50% or sometimes greater recovery of uranium could be achieved using indigenously fabricated silicon-rubber films , but often they showed poor reproducibility.Attempts are underway to characterize these membranes and also to improve their performances. The life-time of the Accurel 2E-PP membranes, found relatively superior to other supports with respect to uranium permeability, was evaluated by periodically measuring the uranium flux under the optimum conditions.Before performing each new flux measurment,the feed and strip so lu'-!:'•#** were replaced with fresh ones.At the end of the permeation sxp'iriment, the SLM was left in contact with the depleted feed?. anoS uranium loaded strip solution.Mo appreciable reduction of Che uranium flux was obtained when continuously operated even 7«?r seven days and thus r.his polymeric support can be used .without any deterioration for atleast over a week's time. 944 - CONCLUSION It is established chat maximum uranium flux through both the BLM and SLiM membranes is attained at the room temparature with feed acidity around 2.5M HNOTJ and 30% TBP in n- dodecane.Dilute ammonium carbonate has been found to ' be the best strippant.Enrichment factors of the order of 10 or more are not difficult to achieve by properly manipulating the feed:strip volume ratios.Applicability of both BLM and SLM separation techniques for the processing of quite dilute (mg dm~^) uranyl nitrate solutions is convincingly demonstrated. / Acknowledgements The authors wish vto thank Sri J.S.Gill.U.E.D..B.A.R.C. for kindly providing us some of the silicon-rubber supports.Thanks are also due.to Dr.P.R.Matarajan,Head,Radiochemistry Div.and Dr. R.K.Dhumwad,Head,Lab. Section, F.R.D. for their keen interest in this work. REFERENCES l.P.R.Danesi.Sep.Sci.S Tech.,19,857(1984-85) . 2.P.R.Danesi,R.Chiarizia,P.Rickert and E.P,Horwitz,Sol.Ext.Ion Exch.,3,lll(1985). 3.A.C.Muscatello,J.D.Navratil. and M.Y.Price, in Liquid Membranes: Theory and Applications, R.D.Noble and J.D.Way (Eds), ACS Sym. Series No. 347, Washington (1987). 4.R.D.Noble, C.A.Koval and J.J.Pellegrino, Chem. Engg. Progr.,58 (1989).; 5.A.C.Muscatello and J.D.Navratil, Chemical Separations, Litarvan, Denver, Vol II, p.439 (1986). 6.M.A.Chaudry, N.Islam and D. Mohammad, J. Radioanal.Nucl. Chem., 109, 11 (1987). 7.W. C. Babcock, R. W. Baker, D. J. Kelly and E. D. Lachapelle, ISEC - 80 (1980). 8.J,P.Shukla and S.K.Misra, J.Membrane Sci.(Communicated). 9.J.Korkisch, Modern Methods for che Separation of Rare Metal Ions, Univ. Of Vienna, Austria, p.144 (1969). - 945 - TABLE I Flux and Permeation of Uranium as a function of Source Phase Nitric Acid Molarity. Uranium concentration 27.03 mg dm~^ Carrier (TBP) concentration 30% (v/v) TBP/ n-dodecane Strippant 1M Ammonium carbonate Source phase Time Uranium flux Uranium acidity elapsed Jm permeation HNO3 _, .. {mol dm ) ( hr ) ( x 107 mol/m /B) (%) 1.0 2 3.8 24.2 3 3.7 34.9 4 4.3 55.2 6 3.4 65.3 8 1.8 46.3 2.5 2 3.3 20.7 3 4.0 38.4 4 3.7 50.4 6 3.5 66.8 8 1.5 38.1 5.0 2 2.8 17.8 3 2.0 * 19.1 4 0.8 9.7 8 0.2 3.9 946 - TABLE II Flux and Permeation of Uranium as a function of Carrier (TBP) Concentration in the Organic Membrane Source phase acidity : 2.5M HNOa - Uranium concen tration : 23.6E> mg dm Strippant : 1M Ammonium carbonate. TBP Time Uranium flux Uranium concentration elapsed Jm permeation in n-dodecane (v/v) -7 (hr) (xlO mo]L/m^s ) ( % ) io 1.5 0.2. 0.6 . 3.0 0.3 3.7 4.5 1.7 27.5 6.0 2.4 52.9 7.5 2.2 50.4 20 1.5 5.7 31.4 3.0 5.1 55.7 4.5 • 3.6 50.4 6.0 2.4 52.9 7.5 1.1 29.7 30 '1.5 3.8 20.7 •3-P >^- 3.5 38.4 3.1 50.4 6.'i\W~0 3.1 66.8 7.5 1.4 38.6 40 1.5 0.8 4.5 3.0 1.7 18.9 4.5 1.3 21.9 6.0 0.9 20.1 7.5 0.3 7.4 PROJECT MANAGEMENT Chairman :, Sfcrl J.L. BHASIN CMD, U C;I I Reporteur: Shri A.M.. HHJHAL B [A R C - 947 - lUNnill/l'ANt.'Y ANU UttXMttrt' l-JMl 11NISICHINlI M|'3(V|CK FDR THE URANIUM INDUSTRY A. K. BHATTACHARYYA VICE CHAIRMAN DEVELOPMENT CONSULTANTS LIMITED INTRODUCTION Amongst the third world countries, India can be termed as the pioneer not only in the field of exploitation of Nuclear Energy for peaceful purpose but also for formulating a comprehensive and balanced plan for supporting the nuclear programme. With a number of operating tests and power reactors by early 60's exploitation by indigenous sources of uranium began. With this in view, the mining in Jaduguda area in Bihar was initiated. Due to foresight of Dr. Homi J. Bhabha and Dr. Bikhram Sarabhai, India, not only embarked upon a programme aiming at self-sufficiency in the'production of uranium, but also initiated UU:[JH to ensure that the Ixitfic technology and detail engineering capability for installation of uranium mining and processing is absorbed and the country becomes self-sufficient in the field as fax- as possible. - 948 - INDIAN APPROACH lianically, world-wide three alternate ways have been tried out by countries which are presently active in thi3 field. (a) sponsoring agency with turnkey responsibility coupled with a promise to support their financing actively so as to ensure that they invest in R&D for developing technology, detail engineering and manufacturing capability. This method has been widely practised in West Europe, both in the field of mining and processing. In this alternate, the agencies are wholly responsible for performance guarantee, not only for the process buL also for the individual equipment. (b) The second alternate, which has been practised by socialist countries of EaBt Europe is to have state control for develop ment of basic technology, process and manufacture. In this alternate, responsibilities are not very clearly defined, neither time scale nor performance is guaranteed, as both the owner and the executing agencies are under the contton umbrella. As far as can be noted from published literature, this process has not been' very successful in producing techno- economically justifiable systems. In this method, adoption of latest state-of-art for system and equipnent ia difficult. The use of new high technology in the field of control, instrument ation, data processing, etc. was not introduced and relatively out-ncxjed equipments were continued to be used. - 949 - (c) The third alternate is being practised widely in America and to mmi oxLcnt. in U.K. In this alternate, services of irKlepurtdent consultants was encouraged, who vere expected to review the status of industry world-wide, develop a system which is most applicable under the condition in which the project was expected to operate and develop or modify process design etc. and select equipment to meet the condition peculiar to a particular project. These consultants, as and when required, acquired the services of very specialised organisations where such technolofv , in the highest form, was available only from very limited sources and with advantage, which the earlier two alternates did not enjoy. This method is also particularly suitable for countries where industrial base of manufacture of equipment and design of some of the sophisticated chemical processes are yet to develop to the degree desired. The Apex Body of the Indian Atomic Energy Commission assessed the Indian resource and rightly came to the conclusion that the firBt two alternates will not produce the desired result. The primary factor which was against the adoption of the first alternate was that- .the state-of-art of equipment and proceasesu available with Indian manufacturers coupled with expected low off-take of their products made it imperative that the industry would.not respond by investing requisite amount. Similar experience with Jndian parties were faced in other fields, particularly heavy industry, such as steel etc. Here the first alternate method was tried for moat of the earlier steel mills and the responsibility for their construction was entrusted to foreign organisations due to which there was a poor growth of indigenous technology. - 950 - As regards second alternate, whatever success has been achieved in this field could not have been achieved in a country where mixed economy is practised and State has no direct control over all the . agencies. In view of the above-mentioned facts, the authorities concerned wisely decided that the entire field of activity from exploitation to final product, such as fuel bundles can be successfully conducted with the maximum utilisation of the Indian resources by adopting the third alternate. The path, basically followed, was that of BARC and at a ' later date, IGCAR was entrusted with basic R&D work not only in the field of process but in development of components, control, etc. The Atomic Mineral Division (AM)) was entrusted with the exploration and identification of the source of raw material not only for the uranium, but also for other minerals required for Nuclear Industry. UCIL waa entrusted with exploitation and milling of the uranium ore to produce uranium concentrate. The subsequent process of manufacturing fuel bundle was entrusted to Nuclear Fuel Complex • (NFC). Another important agency, Indian Rare Earths, was entrusted with the techno logies involved in production of minerals required for Nuclear Industry. The services of consultants were utilised not only to crystalise the project concerned, but also to ensure infrastructure and utility services required. Their responsibility also included consultancy and associated engineering services that are normally required for project of this nature which included development of schematics, procurement services, detail engineering, field uu[*:rviuio'n and wherever necessary, pre-commissioning test and assist- . ance during comnissioning. By adopting this method, effort has been made to make this industry independent from foreign control as far as - 951 - possible. It has also ensured that, if quality demanded, foreign equipment and technology could be imported piecemeal without blanket dependence on foreign source. This method aiao keeps the door open for absorbing the imported technology and its exploitation in future projects. By this method, some of the foreign parties can be induced to undertake collaboration with Indian counterpart to maximise use of Indian resources. PROGRAMS UPTO TOE TURN OF THE CENTURY By mid 80s, the planning body of the Government of India, after consultation with Atomic Energy Cocnnission came to the decision that it is desirable to target the installed capacity of atomic energy to the tune of 10 million KW by utilising reactor of 'CANDU' type. This ambitious progranuie prompted ABC to formulate a plan to meet yellow cake requirement of power generation target. This effort was to be supplemented by the manufacturing facilities of the Nuclear Fuel Complex (NFC) and the Indian Rare Earth (IRE) and other associated industries which were operating under the guidance of AEC. Some of the major projects taken up are ennumerated below :- (a) Uranium Oxide Fuel Projects at Turamdih and Hyderabad (b) Zirconium Sponge Project near Tuticorin. (c) Zirconium Alloy Project at Hyderabad. (d) Fuel Assembly Project at Hyderabad. - 952 - II Urn IM.-4-II Uijcidcti lUil. Umbo pruJecLa will tusti Indian Lculutology to the extent possible and the primary responsibility will be entrusted with the Indian Consultancy Organisations. To ensure the latest technology utilisation, it has been further decided that as and when required, packages of technology and equipment will be imported, but as far as possible, this import will be channelised through respective Indian counterparts. In view of the attitude of the industrially advanced countries, towards third world together with basic requirement of strict surveillance deBired by the Government, this can only be assured by indirect import of technology and equipment, as and when required. ' In the mining sector, the magnitude of the problem, particularly because of the fact that Indian deposits are of poor grade and limited nature, can be understood from the fact that 10 million KW installed capacity target in nuclear fuel may require exploitation of 15 or more deposits. The exact number will be very much dependent upon the characteristics of the deposits. As against these numbers, only two mines are operating and only two more are in the process of being developed. i The assignment assumes bigger dimension because of the locations of the production source which are mostly located in the underdeveloped region of the country without any available infrastructure. The other ships of fuel fabrication and fabrication of other aophiuticalod metal may also require augmentation and necessary steps are being taken to achieve this. - 953 - . ol'hnr cnunl'.rioH. DcsriHity of population, uuuHonal rvit'-urw of Uho rain require comprehensive planning of waste management and suitable agencies should be entrusted with these works at an early date. This takes even a bigger dimension in thin industry where disaster condition is to be taken into consideration. (C) PROCESS The selection of process should take into account the available infrastructure, manufacture base of the country and the time schedule for implementation of the project. Conventional method may be found wanting if the stringent time schedule has to be maintained. Faster methods of shaft shinking, use of decline/incline/adit, wherever necessary, may have to be adopted. For transportation of ore from mine face to surface applicable equipment should be reviewed carefully and use of mobile trackless equipment should be considered. In the chemical process in some countries, such as Canada, it has been found out that the adoption of a particular process for a particular* ore composition has given better result. Instead of blindly following process repeatedly, a research body should be developed to analyse and find out most suitable chemical process to which such ore-' would respond. These studies should involve not only projects which are either operative or under implementation, but also samples from future possible sources, as unless the data is available well in advance, it will be difficult to adopt most applicable process. - 954 - -From the stand point of environment, though absolute zero discharge i3 desirable, it may not be possible to achieve the same, particularly because of seasonal high rainfall. Effort should be made to make the system as close to zero discharge as possible during dry season and during rainy season all possible efforts should be made to keep the dilution to the acceptable limit. System adopted should take care of varying quantity and quality of the waste handled. v During implementation of the project, congnisance should be taken of the basic structure of the Indian industry, particularly while floating tender packages. On the one hand, multiplicity of packages will tend tc complicate management problem with adverse effect on time schedule, on the other hand, the package containing large number of items, of which prime contractor has limited knowledge, will tend to have sub standard or non-harmonious components in otherwise attractive packages. In this respect, the role of the consultant is very vital. LONG TERM STRATBGY FOR DEVELOPMENT OF CONSULTANCY Planning and executing authority of this industry may have to consider now a strategy by which they could not only utilise presently available consultancy services of the country, but also ensure its healthy growth. There are a few'distinct fields to be considered. Each field has its own peculiarity and requires special attention if it has to be nurtured :- - 955 - EXPLORATION AND IDENTIFICATION OF DEPOSITS The activity under this item will, of necessity, be spearheaded by AMD with close cooperation with Geological Survey of India (GSI). Involvement of consultant in specialised field should be looked into. STUDIES LEADING TO IDENTIFICATION OF PROBABLE SITES AND FIXING THE PRIORITY OF IMPLEMENTATION Activity under this item has to be sponsored by project authorities with assistance of consultants who may be more acquainted with infrastructure and its development, project implementation techniques, etc. This activity should lead to a project report having authentic data and pragmatic approach. In this activity, involvement of Environment and Meteriological Department will have tc be ensured from the very early stage, even if! it means establishment of weather station at all probable sites. 'Consulting Engineers may be • asked to coordinate all the work and prepare the required data base and reports. BASIC R&D FOR PROCESS AND EQUIPMENT Work under this item is expected to be guided by the scientific bodies of Atomic Energy Commission and its implementing agencies. Actual implementation may be entrusted to suitable organisation. - 956 - D. PROJECT CONSULTANCY FOR BASIC ENGINEERING, PROCUREMENT, DETAILED ENGINEERING AND PROJECT MANAGEMENT INCLUDING CCWflSSIONING " For activity under this item, the leading role has to be played by implementing agencies, such as UCIL, NEC, etc. It is expected that in this field, the role of Consulting Engineer will be considerable. In other countries, where exploitation of strategic material has been under the umbrella of the defence, the cost factors are difficult to isolate. Users of peaceful energy have been beneficiary of the results of R&D and engineering work done for which the cost was borne under classified head of defence. In India, rigid and conventional procedure now being followed for this type of activity, where neither the quantum of work nor the duration can be clearly defined, particularly where infrastructure and state-of-art are far from the adequate, is not condusive to healthy development. It is, therefore, felt unless a more flexible type of arrange ment is developed, somewhat similar to what is being followed by Nuclear Power Corporation, for remunerating the consultancy organisation for the effort, it will be difficult to make any headway in this field. In case of projects like power plants, water treatment, roads and bridges, ' etc. where the technology is well established and store house of experience is considerable for both the systems and the details of the project and the duration of the project can be estimated with reasonable accuracy and as such cost of consultancy can be established with clause of variation due to price escalation only. It may be noted that the data base available is also far from adequate even for enviornmental studies. For more complicated problem of de-coxmdBBioning of unit and identification of the characteristics of the deposits, etc., this drawback is even more extensive. - 957 - Before closing the paper. I would like to bring out that under very adverse international and national conditions, Atomic Energy Commission alongwith its autonomous and semi-autonomous bodies with the assistance of their consultant, has achieved a standard of technological excellence far above the average achievements of similar bodies in the country. Even when compared with the international standard of achievement, if due consideration is given to the nature of difficulties that the third world countries face, the achievement is praise-worthy. Now that the country has to augment the effort many-fold, scaling up of the effort is essential if the standing that India has in this field is to be maintained. I am confident that the Atomic Energy Commission and their associate bodies together with their consultants will be able to rise to the occassion. * * * * - 958 - PROJECT MANAGEMENT - PROBLEMS IN EXECUTION, REASONS FOR DELAYS, LESSONS LEARNED AND SOME SUGGESTIONS D.G.Nair, PACT, Udyogamandal Introduction Project Management involves both specialist and non-specialist skills. While application of scienti fic Project Management tools serve as art aid to the Project Manager, he needs something more to fulfil ' his tasks in time. It is the ability to correctly assess the situation, to foresee events and his total understanding of the Project which provide him alterna tives to complete the task in time. Successful imple mentation of the Project calls for judicious drawing up of basic activities, proper and timely interaction among the various agencies, followed by a systematic and regular 'follow up to minimise delay and cost over runs are essential factors for.the timely completion. There are many factors affecting the timely execution of a project both technical and non-technical in the Initial, Engineering, Procurement, Construction and Testing & Commissioning phases of a project. In this paper an attempt has been made to point, out various reasons for delay and problems faced during the various phases of execution of projects and some suggestions to overcome these. An elaborate examination of all phases of a project cannot be covered by a paper/short presentation. Scope of this presentation has been done by restricting to detailing certain areas by way of suggestions, while Annexure-1 covers points that require consideration during the different phases of a project. - -J"3? - 1. Project Execution Philisophy Project execution strategies depend upon (i) Nature and magnitude of the project, (ii) Type of orga nisation, (iii) Work environment and culture of the organisation and (iv) Time and resources availability. 1.1 Project lmplementation_Plan For large scale projects, a comprehensive execu tion plan covering the organisation structure, mode of impr-amentation, procedures, control system etc. shall be prepared so that all associated with the project are clear about their role and can effectively •perform their functions. These- guidelies which properly interpret the owner/client's objectives and priorities must be established in the initial stage itself. 1.2 consultantclient interaction For the smooth progress of work, good rapport between the client and consultant and full involvement of client is essential. Finalisation of purchase and construction contracts are areas where delay generally occurs and it comes difficult to make up this slip page. To avoid/minimise delay, client should get fully involved in all activities pertaining to purchase which may include visit to some of the vendors' offices, association during preparation of recommendation. In practice, it is found that by such arrangement ordering of all major equipment could be comopleted between 4 to 8 months from zero date of the project by deploying personnel to •Consultant's office or -Con- -sultant should be asked to" nominate a task force to perform these functions from client's office which shall include vendor data/drawings review. When the construction aptivities pickup and design activities taper down the venue is shifted to site. Client and Consultant working in offices adjacent to each other helps in smooth day to day interaction. - 960 - Maintaining close client and consultant intera ction will help in freezing the design concepts/para meters, P&ID, equipment layout etc, and also expediting approval of client. i 1.3 It is often the experience that the operating personnel of. OWNER group does not get involved.in the initial stages. This association at a stage when the construction activity is very much in an advanced stage, creates delays while accommodating various operating and maintenance requirement. It is con sidered essential that these personnel are fully asso ciated in the design stage itself not only to ensure adequate provisions are made from the operation and maintenance point of view, but also to avoid delay consequent to the provisions to be made when the project is in an advanced stage execution. 1.4 Planning the Time Many contractors/vendors are involved in the planning work. It is important not to impse on them the time schedule made by the Project planners, but rather obtain first from the manufacturers and con tractors their own schedules which they have made based on desired completion time given to them. These sche dules normally reflect, their resources and possibili ties of achieving the timely comopletion. For overall planning work, these schedules have to be analysed and adjusted in consultation with the vendors/contrac tors so that the final planning/schedule is agreed to by all the parties involved. The ability to adhere to overall time schedule is entirely dependent on the ability to execute parallel activities. The more overlap one can get between design, pruocurement and manufacturing activities, the better are the chances of. timely completion. • For example, civil engineering can commence based on preliminary information on foundation loads and outline of founda- - 961 tion and site work can start on this basis. When more information becomes available, changes might be necessary and modification work to some extent must be accepted. This is a part of expenses to be incurred in order to achieve time schedule and this fact should be accepted by the owner. The art of making parallel activities at small extra cost as possible is the art of experienced project management. 2. FAST TRACT CONTRACTING This does not require a complete design before inviting bids for portions of the work. This process has the advantage of getting the work started sooner than the traditional method of completing the design, getting bids and selecting a contractor. This is possible in a project where the basic data for tendering can be derived from the details available from a similar project already executed. The con tract can be a combination of lump-sum, unit price, material and. labour. For example, lump sum pricing for equipment installation, unit price for piping erection and material and labour pricing for system checkout and testing. Fast tracking when 'done properly, achieves a shorter overall schedule for a lower overall cost. Though-this system has the associated risk it has been generally found to be of bene fit to the owner, if the risk involved are ably tackled and handled keeping the overall project goals. From my experience in the project exe cution this has been found to be very effective for the following reasons: (i) It forces sound and expeditious decisions because participants know that to change a decision means to a cause delay. (ii) Increases motivation because results are visible sooner. Accomplishment demonstrated will have greater influence on other participants. (iii) Fast-tracking also creates a situation in which the progress in the field forces the design effort to be expedited which in turn can lead to more schedule improve ment. •3. Interface Management A large project is essentially a joint venture of the owner, Government, financing institutions, Consultants, Vendors and contrac tors, implementation of a project needs co-ordinated working of these agencies and' their inter-relati'onships require monitoring. The relative importance of each interface may vary during different 'stages of a project. Interactions across no two interface will have identical characteristics. special monitoring efforts are called for, if the benefits of an effective cost/schedule monitoring systems are to be fully reaped by a project. It will be worthwhile to conduct a real case study to see how much of the time and effort at top management levels will be expended in solving problems related to these areas. Many of such problems could be identified and solved at an early stage itself if a good interface monitoring system is established. 4• Sub-contract and Management / Construction • Philosophy 4.1 Sub-contract terms, conditions and general technical specifications may be forwarded to pro-qualified contractors in the beginning of the project for thicr study, 'i'heir comments/ - 963 - observations can be discussed across the table to understand their difficulties in sticking to the conditions of enquiry documents. Based on this revisions can be made in the documents, .to the extent possible. This pre-bid conference with contractors also help to explain the pro ject' concept, the relevant' site information, availability of local labour, the monitoring systems etc. This will help in finalisation once bids are received as the deviations are likely to be the minimum. 4.2 A detailed work plan in the form of a network shall form a part of tender enquiry documents. This is necessary to enable the bidders to get familiarised with the quantum of work, sequence, relationship with the total system of activities. At the time of award of contract de tailed break-up of various activities with physical quantities scheduled for a month clearly indicating the responsibilities of owner and contractor, is to be drawn up and agreed upon. This should also cover the pro gramme of deployment of requisite machinery for construction/erection work, sequence and method of erection. An effort in this direc tion clearly identifies how at difficult points of time the contractor will be operating at various fronts simultaneously and also exer cises a check on -adequacy of his resources in terms of materials, machinery and labour. 4.3 construction equipment/Erection reserves. During execution, very often need' arises for crashing of activities which calls for - 964 - additional tools like crane, winches etc. for which an average contractor may take time to mobilise. Similarly, it is a common experience that during erection and trial runs certain items may need replacement such as bearing, packing, seal, pressure gauge, indicationlamps etc. The possibility of maintaining sufficient erection reserve at site by respective supply-cum ~ erection contractors to be discussed and agreed upon during the pre-contract discus sions . 4.4 The philosophy , of adopting multiple fronts and multi-contractor strategy has certain advantages. Funds requirement of individual contracts would be taxed to a lessor extent. There would be less loading on one agency, for meeting peak load period requirement. Problems of co-ordination that can arise can largely be avoided by a judicious distribution of work among different contractors ie. the division may be based on area instead of type of .job which minimises interdependence between various agencies and this ensures multiple agencies for similar type of work. 4.5 Management of Labour It is a common feature in a project site that work is affected due to multiplicity of unions, interiinion rivalries, unreasonable demands, the tendency of contractors to adopt short cut and piecemeal approach to get over the difficulties temporarily. Special attention has to be given in matters like provisions in line with statutory regulations, decent wages, housing, medical facilities, safety etc. While prime resonsibillity rests with the Contractors, the project authority too has to be constantly - 965 - alive to these aspects and play a vital role to maintain a harmonious' working climate. It is desirable that by the active involvement of OWNER, Government Labour Dept. and Unions, a uniform policy/agreement is evolved in respect "of uniform wages, period of work, selection of workers, amenities to be provided etc. This will avoid difficult labour situation at later stage due to difficult wage policies adopted by different contractors. 5. Learning from Experience for Project Planning & Control 5 .1 Endeavour to review the history of the project to extract the useful experience for other pro jects is rarely made. Mistakes are not fully documented and only -those involved had a complete picture of the cases. Project planning and control to a large extent could be made more effective on the historical data available from past experience/feed back data from already executed projects. During the execution of a project, various hypes of problems are faced and solved; many modifica- tions carried out for reasons such as non-avail ability of specified materials and equipment, non-suitability of a particular'design for site conditions etc. Availability of.«such data means that one is better equipped to avoid such even tualities and to improve upon the performance. Similarly operational data are of vital information which will help in areas like selec tion of supplier (vendor rating) with respect to adherence to time schedules, selection o.t - 966 - .equipment and design of.optimum -system, import substitution if any made of any system/equipment, problems during operation etc., in the execution of future projects.. Maintenance of such a data bank will help the owner/consultants to avoid repetition of mistakes and facilitate utilisation of time and resources to provide improve>ddesigns. Job Specification Book For every project after completion,- compil ing and maintaining a job specification book, will be of immense value for future reference. An.alysis of Project Implementation Apart from the usual completion report on a project, it is considered essential that success and failures in the implementation are identified and lessons learned for the future projects. In-depth analysis by way of review/ discussion by members of project implementation team from all disciplines pinpointing the crucial factors relating to success and failures shall be done immediately after the completion of project. This should be ' :viewed by each member as a healthy measure to improve the performance and not as session to pinpoint errors/omissions on the part of others. - 5.4 suggested measures to speed up Construction/Erection 5.4.lEarly and sufficient attention towards drainage, approach road, construction water arid power suply and also demarcation - 967 - of storage and yard for construction sub-contractors. A monsoon protection programme for taking protection measures againsb monsoon and making available adequate numbers of pumps for dewatering, augmentation of drainage etc. Completing underground piping well before the arrival of equipment at site. .Completion of foundations in a phased manner taking into account the delivery position, to facilitate placement of equipment directly on foundations thus avoiding storage and double handling. Civil works, g.et affected due to delay in the receipt of embedded parts like anchorv plates, bolts etc. This point should receive appropriate priority and tied up durin'g the pre-award discussion with equipment/plant suppliers and closely followed up. Modifications/substitutions are a part of construction activity for many reasons such as site conditions, expediting a particular activity. Towards expedi tious and satisfactory decisions, a field design cell at site is considered an advantage for field design services. Release of construction drawings is held up for want of data from vendor, and this affects the progress. In such cases a - 968 suitable and safe information has to be assumed on the basis of past experience, extent of risk assessed and work continued with the concurrence of- owner. As soon as the data are available, corrective action, if required, should be carried out. Generally it is found that reworking • to be done due to this was insignificant. It is worthwhile to. go ahead with cal culated risk and be prepared for rectifi cation at a later stage, if required, rather than waiting and delaying the project. 5.4./? There are many ways of designing a system /equipment to fulfil a function. Changing the entire design,which involves rework and affecting progress of engineering, for marginal improvement is to be avoided since this is only wastage of time and delay. 5.4.9 It is often found that rate of. progress of civil works is behind schedule for shortage of construction materials like steel scaffold ings, shuttering materials etc. The procurement of these items takes time. If these items are procurred on a centralised basis and issued to contracting agencies, with necessary commer cial adjustment made, better progress could be achieved. b.4.10 Structural fabricators usually take long time in preparing the fabrication drawings and time is' lost on this count. Whenever the contractors do not have inhouse facility for this, it is desirable that the OWNER/Consult- ant supply the detailed fabrication drawings foe speedy fabrication work. - 969 - 5,4,il .Initiate procurement action, if not available locally for construction equipment like Heavy duty cranes, welding sets, DG sets etc. Contractors* resources are normally limited. Availability of these will help expedi tious construction. 5..4.12 Investigate quality of bricks, sand etc. locally available for thier suitability, or else plan action needed to meet requirement of such items. 5.4.13 Recheck and reorder necessary quantities of packing materials. These are placed in posi tion just before start up time. Any shortage can result in emergency purchase at higher cost or material of inferior quality, apart from holding up of start up activities. • 5.4,14 Avoid dependence on a single contract particularly towards the end of project as the contractors have tendency bo cut down the work force at this stage when we are racin-g against time. ,5.4.15 Erection plan and construction equipment /are finalised needed .for erection of heavy lift/ before the award of contract and get the commitment of erection agency. This plan shall cover detailed erection sequence and methodology for erection of equipment over certain pre-determined weight/ length. 5.4.16 Identify and develop local contractors, who will prove useful, for taking up jobs of less technical nature'. This is apart from meet ing the expectation of local people will also help in expediting the miscellaneous jobs at the fag end of the project. - 970 - 5.4.17 To the extent possible plan a balanced work loading of a contractor for any month eg. take a proper mix of large size and small dia piping to avoid the normal tendency of the con tractor to take up large dia or heavy thickness piping in prefernece to small dia piping. 5.4.18 Ensure each contractor has work front for next two months by timely review of the materials required including construction aids resources, drawings needed so that he can plan and/ensure inputs. 5.4.19 Expediting task force shall advise site management on the likely receipt of equipment to plan in advance unloading, handling, storage, preparation of fouondations and mobilising resources etc. so that equipment as far as possible are taken to the installation site. —ooOoo - 971 - ANNEXURE - I INITIAL PHASE Development of Infrastructure facilities Site Devel^nent - levelling, approach roads, drainage. Construction power, general illumination. & water. Site office and identifying the type and area requirement for storages/ construction contractors. Accommodation facilities and transport. Soil Investigation. Traffic Survey. Availability of construction materials for civil works. Construction and Testing equipment. Systems & Procedures Project Co-ordination Procedure - shall contain: Scope of the document. Definitions. General description of project, battery limit, interface activity. Agencies for implementation. Division of responsibility. Job centres. Communication procedure. Meetings. - 972.- 1.2 6.9 Expatriate services. 6.10 s t a tu to r y approvals. Engineering Co-ordination procedure Table of contents (Typical) . 7.1 Scope of document. 7.2 Scope.of project. 7.3 Battery limits 7.4 Design basis. 7,5 Split of work. 7.6 List of documents. 7.7 Document description. 7.8 Equipment numbering procedure. 7.9 Document numbering .Procedure. 7.10 List of plant units. 7.11 List of equipments. 7.12 List of vendor packages. 7.13 Design procedure - sequence/approval. 7.14 Distribution of documents. 7.15 Change order. 7.16 Design conferences. 7.17 Communication procedure. 8. Procurement co-ordination procedme -shall contain: (Typical) 8.1 Scope of document. 8.2 Procurement prackages. 8.3 Procurement philosophy. 973 - 8.4 Pre-qualification procedure. • 8.5 Vendor list. 8.6 Standard commercial conditions. 8.6.1 Conditions of tender. 8.6.2 Conditions of contract - supply. 8.6.3 ' Conditions of contract - supply & erection. 8.7 Enquiry document. 8.8 Evaluation procedure. 8.9 Approval procedure. 8.10 Order document. .8.11 Distribution of document. 8.12 Vendor data control. 8.13 Expediting procedure, 8.. 14 Inspection procedure. 8.15 ' Shipment 8.16 Traffic management. 8.17 Numbering of do jmehts. 8.18 Communication. 8.19 Billing & Payment 8.20 Reporting procedure. 9. . Planning, Monitoring & Reporting procedure. 10. cost control & Reporting procedure. 11. Stores management procedure. 12. construction management procedure. 13. • start up & Commissioning procedure. 14. Finance Management procedure. 15. .Training procedure. 16. Advance:planning* Steel'& Cement. - 974 - • i ENGINEERING PHASE Know-how and basic engineering i. Scope of work/division of responsibility to be clearly defined, ii. List of documents to be provided,, content/adequacy of each and programme of delivery to be specified in the sequence required. iii. Adoption of specifications to s-uit our conditions. iv. Test method, inspection procedure. v. Tolerance - shall be capable of being achieved practically. « Detailed Engineering' i. Inadequate specifications - suppliers seek clarifica tions or take cover or seek alternative specifica- ' tio'n - leads to delay in purchase commitment. ii. Specific tolerance different from normal - shall be highlighted while preparing specification - should be brought to the attention during negotia tions. iii. Testing method, Inspection procedure shall be clearly specified and agreed upon. iv. Specifications in respect of method of packing for certain materials like shaped bricks, cast iron pipes etc., the rejection is as hogh as 30 to 40% in the caso of bricks and ' 20-25% in the case of castings. 975 - 2.2 v. Go-ahead with calculated risk. Progress of various disciplines of engineering depends- on the' informa tion from vendors. Delay in getting the details affects the timely release of construction drawings. Assume a safe information on the basis of previous experience,extent of risk assessed and continue with the work. Corrective action is taken if required when details are received. Many a time it is found • reworking'to be done is insignificant. It is worth while to take bold decision and go-ahead with certain amount of risk rather than delay the work. vi. ODC - Impact on design to be considered and also other facilities for unloading at site and assembly at site. . vii. 'Engineering audit and preparing punch list when 90% of engineering is over. viii Off-loatfing of design responsibility to vendor ' - avoid wherever possible. ix. Changes introduced by statutory bodies during the implementation of project. x. Detailed engineering to accommodate materials already procured on the basis of,preliminary designs. xi. Well defined packa.'ges with responsibility from design to commissioning on a single agency - Ensure that, equipment and services in the scope of the package should-be compatible for engineering with least coordination problem/ contents of package should attract sufficient number of -established bidders, Battery limits are clearly specified. - 976 - 2.3 xii. Sequence and method of erection - Hold on construc tion drawing for want of vendor data may call for postponement of constru .ction of foundations. This has to be clearly marked in the drawings rele ased for construction. - 977 - III. PROCUREMENT PKXSE REASONS FOR DEL^Y a. Delays occure in ordering cue to following reasons: 1. Enquiry specifications are not clear and complete thereby r.ecesrltating lengthy prertender and post- tender ccrres?:ndence. ii. The bidders sie not prequalified and may include agencies not c:mpeter.t. iii. Commercial eruditions are vague, contradictory or unreasonably difficult. iv. The enquiry foes not specify information to be given in the bid. v. Opening cf public bid prior to conclu sion of tech nical and connercial acceptability of bids leading to retencer. vi. Lack -of pri:rity list for tendering/ordering of packages. vii. Lack of-advanci preparatory action on import licence b. Delay in the executi:n of order: i. Vendor, vitbotc studying in detail, gives an unwork able schedule. ii. Executio:i pla; has not been agreed to. in order and therefore effective monitoring is not possible. iii. Interaction w:-.h vendor is not carried out constant ly and tlmeLy r.elp to and solve his proiblems where- ever necessary is not given. - 978 - 3.2 iv. More important is given to contractual provisions rather than practical requirements to attain sche dule. v. Choosing a third party by us not acceptable to the vendor. PREQUALIFICATION PRCC5DURE for each category of equipment, materials & service. 1. Criteria for pregualification - Guidelines. 1.1 Prequcliffication is done vis-a-vis procurement packages identified. 1.2 Parties prequalifiec shall be competent for, the respective packages and the rompetence is to be assessed in the following manner: i. Must be equipad for and experienced in the manufac ture of items covered by the packages. ii. Must have •. experience in manufacture/fabrication- of equipment of size at least similar to the largest piece covered by the package. iii. Shall have annual capacity of executing orders of 2 times th= approximate value of package. iv. Order book position should allow spare capacity for atleast IH times the value of package during the fabricati:n. period. v. Reputation fo: timely execution of orders, preference to earlier cunomcers). vi. Requisite in-:.ouse engineering team, for .interpreta tion of drawings, designs, and cooes as well as for preparation of shop-designs and drawings (shop visit). - 979 3.3 vii. Adequate testing facilities (shop visit). viii. Acceptable arrangements with acceptable third party inspection agencies like Lloyds, Bureau Varitas, EIL etc. ix. Where applicable, should have requisite aftersales service organisation. 980'.'- C0NSTRUCTI0N PHASE Problems' usually arise due to lack of clarity. Ensure before awarding a contract that all relevant points regard ing execution philosophy, technical details, quality control, payment terms etc. are defined very clearly in the contract document to minimise disputes arising out of interpretation. Labour relations in the environment is not sensed in advance.and precautionary measures not taken. Detailed work programme specifying schedule for' issue of drawings and departmental materials and stage-wise completion dates are not agreed to in advance. Number and type of construction equipment, testing facili ties etc. to be mobilised by the contractor are not speci fied and agreed upon before the award of contract. Lack of detailed planning on method sequence for erection of heavy equipment du ring pre-award stage. Monsoon protection programme - Adequate drains/augmenta tion of drains, clearance of drains, pumps for dewatering, upkeep of roads to be planned and arrangements made. Buffer stock of materials like brick, metal, sand etc. to tide over period of short supply. Tendering of the labour to prolong the work towards the end of the project. Labour unrest by way of stoppages, go-slow for e .tra wage and incentive etc. result in inter ruptions. Capability of Contractor in tactfully handling' such eventualities shall be one of the considerations while prequalifying. ' ' - 981,-- 4.2 In certain areas unskilled labour for construction work move in from agriculture sector. They are motivated to take up construction to tide over the lean period in agriculture operation. Planning adequate DG sets or making the Contractor to provide during the pre-award stage to meet the poroblems of shortage of power.; Availability of welding gas to be assessed to meet the peak requirement and if required this is to be arranged well in--advance and supplied to the contractors. Availabililty of frontage, drawings'and material in proper sequence and proper timing of start to be-assessed periodi cally rescheduling of erection programme if necessary. Sufficient input sh ould be available atleast for three months. Delay in ciVil works due to non-availabililty of embedded parts, anchor bolts/ temp plate etc. should be tied up during pre-award discussions and followed up. i Project site-must have adequate support by way of •• field design service to attend to changes in the design'necessi tated during execution- eg. converting steel design to RCC, vice versa, changing cable routing, changing the reinforcement etc. Adequate spares by way of Erection reserve for replacement during testing and no-doad trial runs of equipment. Lack of timely action in getting statutory approvals. Quality control system and methods are not made clear in the tendering stage. -982 - 4.3 18. Proper safety precautions are not taken. 19. Delay in payment. 20. Auditing the installations by the Contnissioning team is not done when 80% of the erection work is completed. • - 983 *" TESTING AND COMMISSIONING PHASE The problems and consequent delays experienced in this phase are more due to indirect reasons consequent to lap,ses in the earlier phases of the project. Meticulous care in the earlier phases of work with several levels of check and stage-wise inspection during manufacture and erection will reduce the delay'-. As and when sub systems are .ready'they-shall be tested in sections without waiting for the whole plant to be completed. Problems /delay&'occur due to following reasons: Raw materials arranged are of quality different from design basis'. ' Operating staff are not adequately trained. Identifying training needs and programme for training are not imple mented well.in advance. Operating personnel shall, as far as possible, be asso ciated from the erection stage itself. Utility supplies are not stady and adequate. Proper and timely arrangements are not made for supplies/ services for third parties and for peripheral facilities. 984 ,* HOW TO AVOID DELAY - SOME GUIDELINES Do not under-estimate importance of-functions and always engage competent professional agencies where inhouse facility is inadequate. Where team-work is' required engage a competent task force with proper representation and assign them definite time targets right from the.initial phase of project activity. A one-man task force drawing help from other personnel- with on-going responsiblities never work. For every phase there should be clear definition of inputs and out-puts and definite time objectives. Higher management should closely monitor the progress and check the quality and direction of achievements. • For services and supplies from external agencies, never invite bids from a party who is not pre-gualified for competence, capability and experience. Never assume information - it shall be verified for accu racy and dependability of source. Where repetition of an existing, proven system is possible better g Reliability and delivery assurance should be the main- criteria for selection of vendors and contractors. ***»** - 985 - CONDITIONS REQUIRED FOR OPENING OF A COWWERCIAL MINERAL DEPOSIT 5 SHASTRY URANIUM CORPORATION OF INDIA LTD OADUGUDA MINES It has boon observed that once a mineral deposit is discovered and ore reserves are estimated, it presumed that the deposit is commercially exploitable. Estimation of ore reserves, alone is not sufficient to consider a deposit exploitable. There ere many more investigation necessary to make a deposit commercially mineable. Data regarding rock characteristics, behaviour of the ore body, hydrological conditions extraction properties of ore disposal of mine water and waste rock and suitable sties for mill tailings disposal, are rsauired to be collected for assessing the opening of e new deposit. In this paper all these conditions are discussed. It has been observed that whenever a mineral deposit is disco vered, end ore reserves are estimated, it is generally believed that the deposit is ready for mining; It is presumed that with estimation of ore reserves, there is nothing more rewired for proving the deposit. In reality it is not so. Even after the estimation of ore reserves, the data required is quite considerable to pronounce a deposit economi cally workable. A number of tests have to be carried out before a decision can be token oh out the mlneability of a deposit. These points are discussed in this paper with special reference to an underground mining operations. I GEOLOGICAL CRITERIA 1) Ore Reserves> , Ore reservaa ia the first most important factor in the deter mination of minsabillty of a mineral deposit. Before discussing this point, it is necessary to know the definition of ore. Ore is a natural aggregate of one or more economic minerals, from which one or more metals can be recovered with profit. The emphasis on profit is very significant, and therefore all mineral a cannot bo called orB and all miners! .. 986 occurrences do not mske ore deposits. It may be possible to extreet metals from any mineral'deposit but if the extraction is not" economi cal the mineral deposit cannot be termed as a commercial deposit. ii) Category of ore reserves; Ore reserves are knoun mineral assets, readily available for exploitetion. Reserve implies quantity and quality of ore available within certain dimensions. Category of reserves indicates the confidence level of the reserve estimations. Tor determining mineability of a deposit, it ia very necessary to know the category of ore reserves. Generally, the ore reserves are placed in 3 categories with decre sinn level of confidence. There is no unifor mity in the use of the terminology, indicating the catenory of reserves. The different categories of reserves generally adcpted ore as follows. 1st category and category 3rd category i) Measured reserves - Indicated reserves Inferred reserves ii) Proved reserves Possible reserves Probable reserves iii.) Reasonable assured Estimated additional-I Estimated additional XI. In the first cataoory of reserves.measured, proved or reasonably assured, the percentage of error tn estimation could be upto + 20 "?. In the second category? Indicated, possible and ostimnted additinnal-I, the error could be upto + 30 %. In the last category of reserves, such as inferred, probable or estimated adriitional-II, the error could be higher than £ 30 %. Tor the purpose of mine planning and dasInn work, it is necessary that the reserves should have high confidence levels and therefore the reserves of first two cataqories only are considered. For mineability of a deposit the cntBrjory of ore reserve becomes vary Important. - 987 Hi) Mineable reaervBa: Once the ore reserves are estimated, the question that is asked, now much or these are mineable or recoverable. Generally one feels that all the estimated reserves should be mineable. In actual practice, all these reserves cnn never be mined. In an underground mining operation, maximum percentage of ore th.-' c~n be mined, does not exceed 85 percent. 70 to 75 percent mining extr ction is consi dered to be good mining recover/. In an underground mine, a number or men are renuired to work, for the srfety of men and enuipment main haulage levels, travel ways and ventilation airways uave to be kept intact by lenvlng SOTB pillars in ore. In these pillars about 15 to 25 percent ore is left inside, which may not be recoverable. iv) Average grade of a depositt When ore reserves are estimated,-iloori with 'onn"'ge, the Bversg? gr-ide of deposit is also estimated. Average gr:de of thB ore, is an indicator, of the quality of the ore ev.il.'ble in tie depor.it. Estimation of overage grade of the ore is important not only for mining but for processing also, as proceas plants "re desigred to *-.reat ore of certain gradna only. Any large scalp fluctuations of feed grade, adversely affect the extraction process of the plant. Estimation of accurate average grade is the most important, sinqle requirement, for determination of mineobility of mineral deposit. Average grade is over-sensitive to the economies of operations. When the grade of mine ore fsill3, the' production of final product also f >ll8 and cost of product risen steeply History of mintnq industry is replete wi'h examples, where diappointments and closures of mines have been c-msad by over estimation of aweraqe grade. In recent years to got fnirly accurate picture of the averane grade of the deposit, geoetatisticul techniques are applied. Geostat V) Dilution of ore In mining Estimated average grade of the deposit is not thn average grade or run or the mine (Rum) ore. ROM grsrtes re always loiiier than the estimated aver-ge n,r"de of the deposit, because while mining, the ore gets mixed up with surrounding nan ore portion or waste rock. Some times ore is present in the highly crushed rocks. When this ore is •lined, there 1B generally an ovarbre:'k(which result in dilution of grade. Over break also results whan 'here are structure deformities in the ore-bedy and consequently there is reduction of grade of ore. Hou much overbreak would take pl*ce ia generally difficult bo predict but it has been observed that dilution of grade is of the order of 15 to 20 f>, the is the ROC grades are less than calculated average gmdBO by about 15 to 20 %. vi) minimum reserves for deposit j It la r ther difficult to define as to what should be the minimum quantity of ore available for opening up a mineable commer cial deposit. A deposit must have that much reserves which can sust'-'in production at lean- for 15 years. The mine must have a minimum life of 15 veara, 'so tht the invnstmnnts cm be amortised at a reasonable rots. Underground mine construction in a slow process. It tnk«is 5-7 yeers for conntruction end cam'"iaBioninn of an underground minn, depnncling upon the size of the operation. It is no*- justifiable to h»va a mine, of life less than 15 years when the construction period itBslf la 5-7 years. It is believed that for an independent underground r.-.ine to be workable «#. the production tsrnet should not be less than 350 tonnes of ora per day. Thin would mean thai- the deposit should have at le^st 2 mi]J ion tonnes of gross reserves for economic mining. This is a very approximate figure and would vary from mineral to mineral, depending upon the overage gr de of the deposit. t 989 - II MNING CRITERIA Once ore reserves are estimated and the parameters of the a orebody known the deposit is reedy for exploitation. But before a decision or exploitation! is taken, mining conditions of the deposit need to be examined; i) Characteristics of the ore body: When the ore reserves ere estimated, the different parameters such BS length, breadth, depth averAge width, average grrde, incli nation of thB body are known. Hfter cnreful study of *hese parameters it Is possible to arrive at a possible rate of ore production par day; whether the mine is to be en underground or open pi I- etc. It has been observed that ore bodies with inclinations more than 60° and less than 10° with horizontal are comparatively eosy to nine. Ore bodies ' with inclination 25° - 40° are difficult to mine. The type or the minemlieation also matters a lot in mining of ore minerals. If the ore body is patchy, discontinuous and contains a number of waste zones, it may become difficult to mine such deposit economically. H) Rock conditions t Tor successful mining operation!) the rock conditions of the mineralised horizon must be known before the mine design work is undertaken. Oiamond drilled core samples from all, over the mine, at certain intervals should be collected from ore bearing and adjoining horizons for laboratory tests. Amongst many tests that are reouired uniaxial compressive strength, Young's modulus, abrasim hardness index, total hardness index etc fro some, Th» results of these teBts are very important in design of bits for the boring machines (Tunnel boring *nd rise boring mn-hines). These data of rock mechanics, greatly help to design, stopir Some deposits ere located in thick fnult zones, hlohly weathered rocks or greatly disturbed areas. Hlninn in such zones can be very difficult, tricky and in none c oaa mny he very expensive also. In Bona extreme cases it may not bn possible to mine the ore. iii) Hydroloolcal conditions I > It is very essential th*± oil details nf hydrologies condi tions are known Tor the area whet*the deposit is going to be mined. r""jor fault zones, water scculfera, can pone eeriouo problems for mining. Many of t'-e surface boreholes wv,ich gat exposed during mining also tranaoort large quantities of water, rendering mining operation very difficult. IT the mineralisation is in the sedimen tary rocks such an aendstonne limestone etc. the doteils of hydro- logical conditions must be known uiell before design of mine, as acquifer sandstones are known to hold very large amounts of water. iu^ Radon concentration! For desinninn Vfintilatica system of an underground urnnium mine, it is necessary to know the concentration end release of radic nuclides into the work environment. Estimnton of these can be made from the information of ors> grade, porosity and frneturee of tiie ore bearing rock A. Ground water analysis nives the data on the content, and pocn'blo release of rfdlo-n' c 1 ldr>, HI ORE EXTRACTION CRITERI)1 Before it is decided that t a ore ia fit for processing o number of tests ere required to be carried out. Theso tests ore required to bn done on laboratory ncelo nnri lt>ter on pilot plant scoio. Snmn of tl'o tests nre as fot'owoi i) Li".or-'tion 3lze> crunhlnci ond qrindlnn testa: For processing, the ore hns to be crushed nnd ground nnd it it neconsury to know to what size the ore hns to be ground. In orrtor to know to which size tho ore hns to bn around, the liberation size of t'>o oro minor".! must be known. T'lle i>> doternlnod in laboratories. Liberation nizo verier from minor-il to miner'7! nnd deposit to deposit. - 991 - Once the liberation size one particular org mineral is known, then the crushing and grinding teat9 need to be carried out. In ths9 entire millino operationa, crushing and grinding are highly enerny intensive operations and therefore require very careful study and design. A major sHnre of operational cosf la attributed to crushing and grinding. Crushing and grinding of ore bearing rock is related to the Hardness and abrasive characteristics of the ore. Conventional grinding system consists of rod mill, ball mill combinations. In recent Mmea, "Autogeneoja and semi Autogenous" grinding systems, which are operationally more cost effective are being adopted wherever it ia possible to adopt. ii) Laachlnq testa : Extraction of metals from ores depends upon the mineralogy of the ore. Leeching characteristics of the ore '•hough dependent upon the mineralogy, veries from ore to ore. It io necessary to carry out mineralogical and leachinn studies before a suitable flow sheet is evolved. Extraction, of metals from refractory ores becomes very difficult. If leaching efficiencies are low and percentage of metal extraction is small, then ore deposit ceases to be mineable even though the reserves of the deposit mey large and grade very high. In case of,sulphide ore deposits, the sulphide minerole should be amenable to flotation. It has been observed that in procession by hydrometsllurnical route, the extraction loesrn are of the order of 15-20 peroent. The losses are in leDchin':, filtration and ion exchange concentrator stages. About 15 i> of the metal values cannot be recovered during bhsse operations. IV) ENVIRONMENTAL CONDITIONS When s mine and a mill are in operation, considerable nuantlfcitlas of solid masted tailings), affluent.? unri fumes or gnuee urn gener ten1. All those contain 9ome amount of harmful constituents which npnd to be ramovsd before discharging. - 992 1) Tailings disposal! After mining and processing of ore solid waste In the fonr of tailings are generated. The tailings are separated into sand and sli<-a fractions. Sand goes to the mine BS a fill in the stopBs, while slices need to be discharged in a pond. -Availability of a proper place for discharge of sll ?e has now become a critical factor for construction of mines and mills. Siloes are generally disposed off in a valley. Such a valley should be available uithin reasonable distance from the "'111. The valley should be in the geologically favourable formation, free from major structural deformities such as faults etc. The formation should have vary lou permeability, porosity and hydraulic conductivity. Normally most of the valleys are the courses of streams or rivalets. In ouch cases, the selected area of the valley for storing tailings should be at the head of stream. In rscont ye^rs, antipollution laws have become very stringent and it Is no i ore possible to discharge, mine waters, mill effluents and tailings into rivers, for disposal of the effluents and tailings.proper disposal systems will have to constructed to contain all wastes. Availability of proper sites for disposal of the tellings, has now become a very important conditions in obtaining clearance8 for oper/tinn a mine and a mill. In mineral industry a new system called rsro discharge system for liquid effluents has been evolved. In thl? system nil the liquid effluents foe mine, mill and tailings are treated and recycled for use in the plant or stored nnd hardly any liquids effl'ients ere let off. For planning such a system lot of data on geology, hydrology, tectonics and climatic conditions are required. In recent years the doses levels of radiation have benn brought down from what thoy werfi,say.20 yenrs ngo. To minimi**'• the axposure to radon, It ls.helnn felt that the uranium mines should be more mochanlaed. - 993 In this context, it may be mentioned that m.->.jor number or urnnlun deposit in India, are small with total reserves of U308 being of the order of 2000 — 3000 tonnes. In such small mines it is difficult to achieve a high degree of mechanisation. Large scale investments required for high mechanisation may not be economical in operating e wary small mine. V) LOCATION AND INFRASTRUCTURE. FACILITIES Location of the deposit and presence of good infrastructural facilities go a long way in setting un a mine and a plant without o'uch problems. - , Location of a mineral deposit hss a teamendous bearing on the economics of the ore minerals produced. Unfortunately one C'.nnot choose the location of a mineral deposit, as the mine has to be constructed at a place inhere the deposit exists. If e deposit'is located in deep interior area's, inside thick forests or in hilly regions like Himalayeyon or sub Himalayan regions, this lnfreetruc- tursl facilities mill hawe to be created before undertaking Bine and plant construction work. The1infrestructure facilities inc'urte, roddp.rail and telecommunications, availability of rau materials, chemicals, building m< teriais, fuel,steal and cement etc. Availabi lity of sic*lled manpower locally will be an asset. In mines and prncean plants, the requirements of water is very larne. Water is required for lndustri'il" s'a well as for drinking pumoses. The water source should be near by and u>°'ter should be available in sufficient quantities, all through the yeor., Same is true for pouer,fining and processing industry needs l'rqe quantity of power. The source of pouer should also he close by. if good inf restructure-1 facilities :ire pre ant nf the •ine locutionj 1> will not only help reduce the necessity 6f investments to create these facilities but eleo help to start production in much shorter time. 994 - The author wiahae to exprcsa his gratitude to Shri O.L.Bhaain, Chairmen & Panagina Director, UCIL. and Shri M.K.Bntra, Adviaor, UCK. Tor their encouragement f-o publish this paper. - 995 - Annexure-1 Effect of dilution or ore on cost Dilution of ore and consequent lowering of average grade ore hns gerious economic consequences on the cost of the product. Calcula tions on a fictitious deposit of uranium,of 0.1* average grille, 5 million tonnes reserves, with mine production of 300,000 tonnes of ore per year, have been worked out in the following t^ble. The table shows that when dilution increases frcm 10 * to 40 % the correspon ding increase in cost of the product is 11 * to 66 %. Table-1 Dilution * 0 10 20 30 40 Ore Reserves, million tonnes 5.0 S.O 5.0 5.0 5.0 Reserves after dilution 5.0 5.5 6.0 6.5 7.0 Av.grade U308 % 0.1 0.09 o.oa ".07 0.06 Mine production Tonnes per year 300, onn 300,nno 3no,nno 300,000 300,000 Mill recovery B5* • 85* 05* 85* 85* U30B production tonnes per year 255 .229.5 204 176.5 153 Total Expenditure per year. Rj. in crores 54 54 54 54 54 Cost of U308/kg. "a,,211 7 **.2352 Rs.2647 ta.3025 It. 3529 Increase in cost with roapect to 0 dilution 0 11* 2B< 42* 66^. - 996 - FIGURE-1. ' CATEGORY OF ORE RESERVES VERTICAL LONGITUDINAL SECTION OF MINE INDEX BORE - HOLE. ni LEVEL & UNDERGROUND WORKINGS. MEASURED CATEGORY, ORE RESERVES. ORE RESERVES INDICATED CATEGORY. l',>t.","-t-i L ORE RESERVES INFERRED CATEGORY. -997 — FIGURE-2. MINING DILUTION WASTE ROCK MINE EXCAVATION (TUNNEL) WASTE ROCK 10 TONNES OF ORE OF 0.10 '/ U308 + 1 TONNE OF WASTE OFO-01 % U30a 11 TONNES OF ORE OF 0-09 I U30a DILUTION: 10 % - 998 MANAGEMENT OF URANIUM MINING AND PROCESSING WASTES AT TURAMDIH PROJECT R.C.PURI-CHIEF SUPDT.(PROJ),UCIL R.P.VERMA-ADDL .SUPDT.(MILL),UCIL Based on environmental impact assessment, comprehensive plan for management of wastes has been drawn up* No solid waste from the mine is being disposed off outside the project area. The quantity of waste generated after processing of ore is large because of low content of ura nium in the ore* A big tailings pond has been planned in specially selected suitable valley near the plant* No liquid effluents are to be discharged into general surr ounding environment* Mine water is to be fed to the pr ocess plant* Effluents from tailings pond will be coll ected in a storage cum evaporation pond. All water from different zones of the project shall be collected in zo nal ponds and then pumped to tailings effluent storage pond. All the ponds will be provided with requisite im pervious liners. The effluents of the storage pond will be treated for removal of Radium and Manganese and dis charged into monitoring pond* Large surface areas for various ponds are envisaged to take advantage of evapo ration with aim for zero discharge* To reduce Impact -from gaseous emissons. high efficiency dust suppression and extraction systems shall be provided. High stacks have been incorporated for DG Sets, Boiler Plants, Sul phuric Acid Plant and Dust Extraction Systems for Cru shing and Grinding section and the quality of discharges will be very much within the prescribed limits*. The pa per describes the management plan in detail* - 999 - INTRODUCTION Turamdih Project of UCIL is located about 5 KM south of Jamshedpur town on Tatanagar-Hata PWD road near Sunder- nagar suburb. The drainage from the area leads via dra ins and nallas to Kharkhal river lying about 6 KM away on the north-west. An area of 682,40 acres has already been acquired for the project for construction of mine, process plant, tailings disposal, common services and to wnship* Another area of : about 400 acres is proposed to be acquired for construction of various ponds downstream of tailings pond, Figure-X shows the layout plan of the project. The project area generally consists of plain land with a few mounds and small hills. The location of the project is in village sites outside municiple limits. The land is mostly covered by alluvial soil containing no harmful constituents. The area is surrounded by small viT llages all around covering population of few hundreds. The villagers use well water for drinking and pond water for bathing/washing* No factories or chemical plants are lo cated within 5 KM radius. The local atmosphere is clear and free from obnoxious or harmful gases. No dust gene rating sources are located nearbye. Figure 2 shows the location of the industries around Turamdih Project, Turamdih Mine is rattd to produce 1500 tonnes of ore per day. The process plant located at this plant will also receive-ore by aerial ropeway from Narwapahar Mine which is also rated to produce 1500 tonnes of ore par day. So the plant will receive 3000 tonnes of ore per day. The uranium content of the ores is very low and almost all the quantity of rock treated will report as tailings, Turamdih Mine and Process Plant have been cleared envir- - 1000 - nroentally by Deptt, of Environment, Govt, of India and by Bihar State Pollution Board. Very high standards of pollution control have been kept even in a location wh ere the base level pollution is already low. MANAGEMENT OF MINING WASTES Mine Waste Rock The waste rock genera'ted in the mines will be in lumps and it wilX be disposed off in the lew lying areas about 1 Km from the mine entry within the project premises it self thus preventing its dispersal in the surrounding ar eas* Locations with adequate areas have been identified for'this purpose. Phased reclamation programme at five year intervals has also been prepared. The height o'» the dumps will not exceed 5 metres. The sides of tha dumps will be suitably sloped and land scaped by covering with earth and turfing with grass. Top of the dumps will be covered with suitable soil and trees will be planted over it. The land will -thus be reclaimed as forest area. A detailed note on the subject is given in annexure-I. Mine Water The mine water in total will be utilised in the Process Plant. During monsoon when the mine pumping will be heavy, excess water will be pumped to the storage pond downstream of the tailings pond. Mine Air In the underground environment of the mine the maximum permissible limit for Radon is prescribed as 35 pCi (EER) per liter of air for a mine working 8 hours per shift and - 1001 - six days per week. The prescribed limits for concentration of dust is 170 ppcc or in terms of weight 0.5mgm/M of respirable air. For diesel engine enmissions in case of diesel equipment used underground, air quantities required to be circula- ted are 0.07&A /sec per KW of engine power. Blasting also prodivies obnoxious fumes which are required to be brought down to the value of 50 ppm in case of CO, and 5 ppm for oxides of Nitrogen within 5 minutes by adequate ventila- tlon. Large quantities of air of the order 150M /sec will be circulated in mines to ensure compliance with these standards. This comes to 0.091 M /sec of air per tonnes of ore raised which is equivalent to quantities circulat ed In high grade uranium mines abroad. In general,mining operation will be wet. However, bag filter type dust co llection system has been provided for underground crusher in mine. The exhaust air from the mines will not create any1hazard as is borne out by the measurements carried at Jaduguda Mines of the Company. Moreover, the ventila tion exhaust shafts are located inside the: project premi ses, away from the'township and neighbouring villages. Ore Handling System s The ore is wet and contains 5# moisture. The primary cr ushing on surface at the mine site is from 300mm size to 150 mm size. The crushers will have in-built dust ext raction systems. The handling of otfe from stock piles, by conveyors and ropeway is not expected to generate dust as is experienced at Jaduguda Mines & Mill of the Company However, provision of spraying high pressure water on the stock pile has been kept to suppress the dust, if any,is experienced during peak summer season and provision of water sprinklers during flow on the conveyors has also been kept if dust i3 experienced any time!. - 1002 - PROCESSING The wastes generation during processing of uranium ore aro conventional ones but presence of radionuclides impo se extra precaution^steps to have control on their re lease and exposure to human beings. The wastes generati on can be classified into; 1. Fine dust and particulate matters 2, Liquids 3. Solids 4, Airborne wastes, DUST AND PARTICULATE MATTERS The or© contains high percentage of silica and very small quantity of uranium and other radioactive constituents, hence hazards posed by inhalation of fine dust particles will be conventional ones like chances of silicosis di sease among occupational workers. The control measures adopted are mainly for the containment of silica bearing dust of very fine size produced during crushing & grind ing of tho ore. However, at the final stage of produc tion, where UgOg Content In the 'Yellow Cake* reaches to high percentage; the dust poses radiation hazards. Hence, control measures envisaged take into account the above problems in different sections of the process plant. Dust Extraction System To check escape of dust particles in working places and in the environment, dust extraction systems will bo pro vided in potentially dusty sections mentioned below ; 1, Primary Jaw Crusher, - 1003 2. Secondary and Tertiary crushers 3. Control screen section 4. Yellow cake drying and packing section. Primary Jaw Crusher The mined ore to be received in this section, will con tain about 5# moisture and will be reduced to 150mm lu mps size. As the crushing is coarse and ore surface wet, dust generation is expected to be less, hence bag type dust filters have been selected for this section which will have connection to crusher's feed and discharge points as well as conveyor's duct and hood assemblies. Dust laden air sucked from these points will be filter ed and particulates collected in a hopper for further disposal. The collection efficiency of the equipment would be 99% and permissible emission limit has been kept as 25rag/m in the exhaust air. For efficient di spersal of dust particulates, the stack night shall be kept to 35 meters as per the stipulation of the Depart ment of Environment; . econdary.Tertiary Crushing Sections and the Control fcreen. The primary crushers' product of 150mm size will be fur ther reduced to 20mm in these sections. The reduction ratio being higher, dust generation range would be in range of 10 microns and below. The dust extraction sy stem selected for these sections will be water based high efficiency medium pressure venturi. The collection efficiency of the system would be about 99# for dust pa rticles, aimed to collect particles of 1.4 micron G M D. For efficient dispersal of particulate matters, the sta ck shall be 35 mts high and emission limit in exhaust would not exceed 25mg/rn as stipulated by D.O.E. 1004 Grinding Section The crushed ore of minus 20mm size will form feed to this section and shall be ground to 150 microns in a combination of Rod and Ball mill following wet grinding process. Dust generation in this section is expected to be less and Bag filters shall be provided to control dust emission from different equipments and for subse quent disposal. Yellow Cake Drying 8. Packing Section Emphasis is to go for a completely dust leak-proof sys tem for drying and packing operations of yellow cake. The concentration of Uranium in the product will be hi gher hence inhalation of fine M D U dust, even in small quantity will pose grave radiological hazards to occu pational ..workers. Hence, dust extraction system for this section shall be predominantly conditioned by rec overy of high valued product emitted in sub-micron size particles. This will automatically ensure much less em ission than conventional systems. To make system more reliable, particulate ladden air shall1, be passed through water based high efficiency medium pressure venturi scr ubber in series with mist eliminators working at 99.5^ efficiency for scavenging out. trace of;mist and particu lates. Liquid shall be recycled within the process and air discharged through a 27 meter high chimeny. Emission, targetted in the exhaust, is not to exceed 0.3 mg/M3 of particulate matters. LIQUID WASTES The liquid wastes to be handled will comprise of 1. Domestic effluent. - 1005 - 2. Mine Water 3. Decant Water from Tailings Pond Domestic Wastes Sewarage from residential quarters shall be treated in a sewage treatment plant. Following treatment process shall be adopted to make it absolutely safe :- 1. Extended aeration system to provide excellent oxy gen effluent mixing for achieving BOD below 20mg/lts. 2. Chlorination of the treated and clarified liquor be fore discharge, to make it free from harmful microbes. 3. Provision of"Polishing*pond to collect chlorinated liquor for further impoundment and clarification in case of Narwapahar Project and collection of treated effluent in a big pond (P^) of 10 million M capacity down stream of Tailings Dam for confinement and loss by evaporation during dry seasons in case of Turamdih Project. This collected water may be utilised for dillution of other effluent streams befor discharge to nalla. Mine Water , O Abqut 1800 M /day of Mine Water generated from seepage and backfilling of sand underground,, shall be pumped out at jthe surface apd will be stored In, a settling pond. An elaborate scheme has been prepared to treat it by clari fication v/ith floculant dosing and further removal of suspended particles by filtering through sand bed fil ters. The filtered water will be used in different se ctions of the process plant and the mines as industrial water. The treatment steps envisaged are shown in figure No.3. - 1006 - Tailings Pond Decant Water About 725,000 M3 of water and 600000 tons of solids ann ually shall be pumped to the Tailings pond on achieving full targeted production of ore. Additional liquid due to rainfall within the pond area itself contribute to the liquor volume to be handled. To restrict additional rainwater inflow from adjacent areas, peripheral cut off drains have been envisaged on the hills around the tail ings pond. This will reduce liquid load to be treated as overflow from the dam. The rain water through drains will collect in the rain water pond P^, A detail scheme for tailings effluent water management has been prepared as shown in figure No,4, The tailings pond proper for confinement of solid waste cover an area of 175 acres, and additional 400 acres of land shall be acquired for construction of four water storage ponds, down stream to tailings pond as shown in figure No.4 for collection of decanted water overflowing from it. Sizes of ponds would be as under ; PA - Tailings decant water pond 406tO00M capacity V>2 - Preventive monitoring pond 1,30,000 M 3 P3 - Treated water pond 12,20,000 M 3 P4 - Rain water storage pond 1000000 M The total hold up capacity of P^, P2, P3 ponds will be about 3 years hold up of the total effluents likely to over flow from the tailings pond. All these ponds and the tailings pond, except P. will be provided with impervious lining so that seepage and por- colation of contaminated effluent to ground water does not take place. Capacities of these ponds have been se lected based on maximum rainfall encountered in the area - 1007 - in a particular year (once in 100 years) and average ye arly evaporation rate of water, and considering no loss due to seepage because of water proofing etc. Ponds will provide large surface area for loss of stored water by solar evaporation route. Treatment of Tailings Pond Effluents; q A Barium Chloride treatment Plant of 250 M/hr. capacity will be constructed to treat stored effluent so that treated water can be discharged after sufficient dillu- tion during wet seasons. The treatment system envisa ges Manganese and Radium removal by Lime and Barriura Chloride with necessary clarifloculation and filtration of decanted effluent overflowing to pond P, from the Tailings Dam, The treated water from the Plant will collect in a twin chamber pond (P2) for monitoring of the effluent. If it confirms to standards prescribed then it will be stored in P3 pond and if not would be recycled to P, for further Bad, treatment. The pond P, being of 1.2 mem. capacity will hold expected effluent for one year duration and during detentionf maximum water will evaporate by solar heat and a part would be utilised in the Barium Chloride treatment Plant. The leftover treated effluent stored in pond P3 will be discharged to public streams during high flood seasons when excess flow is available in the river. The effluent at the outfall from the Project site will, not exceed 5 pCi/litrs. for Radium and 1,0 mg./lit. for maganase. If necessary, when the concentration of these substances does not meet the stipulated requirement the rain and treated sewage water available in Pond P. would be utilised for dillutlon purpose. The discharged effluents will be led through close routing and would join the mid stream of the Kharkai river through pipe 1008 line and diffusers and will meet the standard IS:2490 and IS;2296 as stipulated by the Department of Enviro nment, New Delhi* AIR BORNE WASTES Air borne wastes other than the main process plant are from the following places :- 1; Sulphuric Acid Plant ,2. Boiler House y3«- Diesel Generators 4. Reagent Plant (Pyrolusite & lime) 5, Tailings Pond. Sulphuric Acid Plant A 100 tonnes/day Sulphuric Acid Plant would be installed to meet the requirement of.acid leaching of ore. The plant will be based on elemental sulphur to produce 98# pure ac id by following DC-DA process (double conversion and double absorption) and sophisticated equipments to control tempe rature, flow and.concentration of different streams to achieve high conversion efficiency. The plant stack shall be made 50 meter tall which will help in efficient disper sal of pollutants like So2, and Acid mist contained in tail gases. Sufficient number of demisters shall be prov ided to keep the acid mist level in stack gases below 50 micro gras/NM . Continuous stack air monitoring systems sh all be provided to record emission of So, & So3 in the exhaust as well as three permanent monitoring stations at 120° apart aat the boundary of the plant will be establi shed as per guideline of Bihar State Pollution Control Bo ard, to monitor So2 concentration in ambient air and it shall be maintained below 2 ppra. - 1009 - Boiler Plant The boilers will run mainly in winter season to augment supply of steam for uranium leaching circuit* L.S.H.S Oil shall be used for firing the boiler. This oil con tains only 1% sulpher. Consequently So2 concentration in exhaust gas would be around 0.22$ only. Height of stack of 35 meters for efficient dispersal of gases will reduce their impact on ambient air quality to negligible limits. Diesel Generators Diesel Generators will run during power failure to keep emergency equipments running. The fuel used would be High Speed Diesel of grade IS-1460-1974. This oil con tains i%» Sulpher. the exhaust gas composition is exp ected to be Co2 = 13.4#, N2 = 74,14# So2 a 0.05#, S P M - insignificant Tailings Pond The tailings product comprising of ground ore will be subjected to. classification by hydrocyclones. The slimy fraction would be pumped to tailings pond for impound ment. When the surface of the piles dry up and fine particles are• subjected to air erosion dispersal of radioactive wastes to public area may occur. Along with particulate matters radon gas emanated from the tailings surface may get air borne and escape to resi- dential areas in settlements located on down wind di rection. Three major conditions which contribute to wind eros ion are (1) Loose finely divided dry tailings (2) Sm ooth and bare tailings surface (3) Strong winds. To counteract these, following measures have been envis aged; - 1010 - 1. Submergence of tailings pile under water. This will prevent drying up of piles surface and control emana- *•• tion of radon gas as well as reduce volume of liquor, by solar evaporation, to be treated in radium removal plant. 2. Sectoral discharge of tailings in the pond if adopted may provide economy in the liners application and be tter control of piles submergence and effluent manage ment, but requirement of land and construction materi als like soil, sand etc. will increase. Check on lin ers failure and seepage of radioactive effluent under ground would be detected easily in this method. ACKNOWLEDGEMENT The authors are thankful to the Chairman and Managing Director, Uranium Corporation of India Limited for his guidance and permission to present/publish this paper. REFERENCES Detailed Project Report, Uranium Cre Mining and Proce ssing Project at Narwapahar and Turamdih. - 1011 - Annexure-I PHASED LAND BECLAIMATION PLAN FOR WASTE DUMP (TURAMDIH MINE) EXPECTED QUANTITY OF WASTE BOCK Waste rock from Turamdih Mine will be generated while constructing the various entries viz. access decline, conveyor incline, vertical shaft and excavating various chamber/bins and the three main development levels. In addition, waste would come from exploration development which may be undertaken during production phase of the nine and from the waste faces in the orebody itself while operating the orebody. The total estimated quantity of waste rock is as follows; a) From Access Decline - 126644 MT b) From Conveyor Incline - 77,655 MT c) From Men & Material Vertical Shaft - 20,442 MT d) From Mine Development Ex-Shaft - 86,016 MT e) From Miscellaneous Waste Faces - 92,273 MT 4,03,000 MT The rate of waste production per day would be about 150 tonnes during construction & development phase and would fall thereafter, EXPECTED UTILISATION OF WASTE ROCK Waste rock would be utilised in various civil constru ctions underground like supporting arches, pump founda tions, ore chutes, ventilation doors, miscellaneous eq- upment foundation etc. It would also be utilised for 1012 construction of tailings pond dams. These dumps have to be raised in phased manner and waste rock would be utilised as per schedule. As per experience at Jadugu- da Mines, it is expected that about 30# of the waste rock generated would be utilised in the initial years and 20# of the waste rock generated would be used in the later years. QUANTITY OF WASTE ROCK FOR DISPOSAL After utilisation of the waste rock as stated in SI. No.2 above, the position with regard to waste rock ge nerated and to be disposed at 5 years intervals comes as follows :- Project years Quantity expected Quantity expec- • to be generated ted for disposal 0-5 years 1,38,000 MT 96,600 MT 5-10 years 1,55,000 MT 1,08,500 MT 10-15 years 75,000 MT 60,000 MT 15-18 years 35,000 MT 28,000 MT Total : 4,04,000 MT 2,93,100 MT MODE OF DISPOSAL OF WASTE ROCK Initial waste-rock from mining the access decline and conveyor incline will be brought to surface by the load haul dump vehicle and then loaded into dumpers which would transport it to the waste dump. As the declines advance to distance more th3n 500m, low profile dumpers would be loaded by LHD at specially prepared turn outs in underground workings itsolf. These would take the rock to waste dumps. When the first lift of conveyor in the conveyor incline is commissioned, waste rock will be loaded in 10 tonnes trucks from head pulley - ion rock bin for transport to waste dump. This practice will work during first 5 years. When the service shaft is commissioned, waste rock from development and ex-shaft development will be brought to the surface in 1.13M3 side tipping cars. These will be taken out from the cage at the unloading • ' "gantry above the shaft bank and waste rock would be tipped into a 20 tonne surge bin from where it will be taken by 10 tonne trucks for dumping at the waste tips. This prac'tice will continue till the end of the life of the mine, LOCATION OF WASTE DUMPS & MANNER OF STACKING A low lying land area near* the north boundary and close to forest land has been selected for filling the waste rock. The filling of the area upto normal ground level of the Project (170 M level) is so much that maximum qu antity of waste will be used to level it up. Very litt le quantity would be dumped above this level. The height of the benches will be restricted to '3m only with 1:2 slope for stability. During first five years ground will be levelled upto 166m level, during 2nd 5 years duration filling will be done upto 169 M level, during 3rd 5 'years • duration, the dumping will be done upto 171 M level and during the last phase from 15 to 18years, when the quan tity is very less, the dumping will be upto level of 172 M i.e. 2 M above general ground'''level. The pattern of dumping will be made in such a way that rain water outlet from the area is kept from the western side. Figure 5, sheets 1 & 2 give details of the location of waste dump and manner of disposal of waste. - 10H - RECLAIMATION PLAN As the dumps for each phase are completed, the exposed sides and the top surface of the stacks will be covered with thick layer of alluvial soil which would be dozed and transported from inside the project area as per gra ding plan. Grass turfing would be done on this surface and trees would be grown over it /as per the social aff orestation scheme drawn up. Ultimately all the dumping area measuring 'about-10 acres will be covered into a forest belt. This belt would be contiguous to the exis ting forest land in the area. In this fashion, the area covered by waste dumps gets reclaimed for forest purposes. 1015 - - '9'S MAP SHOWING LOCATION OF INDUSTRIES AROUND TURAMDIH PROJECT LEGEND 1. ISAHiyX PPCJECT- t- » ' i. KSVSII . -.CE523 3. GAIUU!.*.-.A 4. DOOU »WOI 5. HXnGHti.HlWVKi—_ 6. PWOBOAO 7. DISTRICT BOARD B0AD--GS3 B. RAILWAY LIME 1 +*•* Z MINE WATER TREATMENT , flLf EREPWATElf FILTERED WATER MAKEUP O/H TANK MINE CLARIFI- wnre «O00M> EO WATER .^.FILTERED WA- (JJVER HEAD, 3 ^TERfflVEIHlAO TANK) 300M TANK} 500M3 PROCESS TO PROCESS FOR WINKING TO REAGENT-, f ILTEREO FILTERE0 SAND FILTEFILT R BACK MI&EWATER , WSeWATEWATER WASH PUMPS PUMPS SOMVhr TANK LIME FLOCCULANT HOPPER CLARIFIER SMJOVA, _C\. TURAMDIH MINE WATER CLARI 1 WAmm' r— l(WlMJ/hIO0MJ/hr I [ t t t MINE WATER STORAGE PONO 4500 M3 MMVVR 160-OOHLVL 18 8 as! SECTION-'A-A 172M.LVL f-l/IPULVU-17JM.LVL*- TURflHS w r17000M.LVL. GROUND PROFILE >^f" •160-OOMLVL 8 f * SECTION-BB r-)7000M. LVL 1021 - SESSION XI PANEL DISCUSSION Panel Members: . Dr. .P.K. Iyengar Director BARC - CHAIRMAN Shri N. Vittal Additional Secy. DAE Shri R.K. Garg CMD, IRE Ltd. 'Shri J.L. Bhaain CMD, nciL Shri A.C. Saraswat Director, AMD Shri N. Swaminathan DCE, NEC Shri N. Ramakrishnan Hindustan Dorr Oliver Shri S. Sen 'Associate Director,. ChEG, BARC, SECRETARY. -1022- PANEL DISCUSSION ON PRESENT STATUS AND FUTURE STRATEGIES ON URANIUM TECHNOLOGY S.SEN, (Chairman, Organising Committee): We are now coming to the last session, i.e. panel discussion on present status and future strategies on uranium technology. Me are happy to have Dr. P.K. Iyengar. Director, BARC as the chairman of the Panel. The other members are Shri. R.K. Garg, CMD, IRE, Shri. J.L. Bhasin, CMD, UCIL, Shri. A.C. Saraswat, Director, AMD, and in place of Shri. K. Balaramamoorthy, CE, NFC, we have with us Shri. N. Swaminathan, DCE, NFC. Unfortunately, Shri. S.L. Rati, MD, NPCIL is not afile to attend. In his place, we have a person from industry Shri. N. Ramakrishnan, Marketing Manager, Hindustan Dorr-Oliver, Shri. N. Vittal, Additional Secretary, DAE and myself will also be in the panel. I request Dr. Iyengar to start the discussions. Dr. P.K. Iyengar: Well, all of you know that uranium is radioactive but I did not know that the number of people attending a conference on uranium will also have an exponential decay towards the end of the session. Since the half life is very long, the decay al3o has been slow, so we still have enough number to participate in this panel discussion. A casual observation I would like to make, because I am not an expert on uranium technology, sometimes I wonder whether all the technology that we develop using a lot of man made engineering has not been copied by nature or we are copying nature. If we look at the shell oil the egg that the hen lays it looks like a very pure form of composite substance made of calcium. Sometimes,we wonder what do we feed the hen, perhaps a little cereals, some stones and eventually it produces a beautiful coating of calcium compound on the egg which seems to be very pure and does the job of protecting the egg itself.The question is of extracting uranium, doing chemical engineering on it, making it into the form that is suitable for introducing into the reactor. So we should ask ourselves. Have we to learn something from nature in trying to produce a fuel which is useful, what is 10,000 MW days per ton, or whatever you want to call it and that is what the chemical engineers and uranium technologists have to do. Of course, if you extend this example of hen and the egg we have other examples like coconut tree on the sea-front, but giving sweet coconut water. I do not know what it does to convert sea water into very sweet water. Or you take the rose plant, it is able to accumulate all the colouring matter .in such a way that it produces lovely colours. This shows that there are processes in nature which can concentrate certain "chemicals and all the chemistry we are trying to do is"being done probably more efficiently by biological systems including plants and animals. The animal which can produce the uranium fuel in the form you want obviously must be a big animal like dinosaur. Unfortunately, dinosaur is long extinct and hence we have to do it ourselves. This involves various technologies - for example, the technology associated with Shri. Saraswat i.e. geology. Perhaps we should be able to smell and find out where uranium is located. Since we cannot smell it properly we should find out plants and shrubs which thrive on uranium and concentrate it so that we can identify from photographs. The area where such plants thrive must be rich in uranium, can this be true? If even that fails, we have magnetometers which tell you whether there could be uranium, radiation monitors which will give you the location of uranium. -1023- Next we come to ore dressing of the ores found. Then I believe you have bacteria which can concentrate the uranium - first they eat uranium and next excrete it and we can then collect the right quantity of uranium. I am told we are also trying that at AMD. So Shri. Saraswat will now talk about smelling uranium and recovering it by bacterial leaching. Then of course comes chemical processing and we have a number of experts in this area. In place of Shri. Balaramamoorthy, we have Shri. Swaminathan who has worked at BARC for several years and chemical engineers always talk about contactors and what is known as floatation in ore dressing or solvent extraction etc. which is very essential for producing uranium. And finally of course, all this is possible only if you have the money and if you have the right management structure and here on my right is Shri. Vittal who has new ideas on how atomic energy should be managed. X think all the scientists have grown old and have old traditional techniques which can be imported if need be but the government has changed unfortunately so that they may not allow import of management technology. So with this introduction now let me turn to the subject of this panel discussion. This is on present status and future strategies. He already had detailed discussion on almost every aspect of uranium technology in the technical sessions during the last three days and hence there is not much need to discuss our present status, but we could discuss future strategies. The future strategies would mean that how much of uranium we need, is it under the earth in India, and if yes, what are the problems associated with locating it, mining and processing it. Then comes technology associated with the safety whether it is good for health or not, whether the economics of it will allow it to be done. We have problems with fabrication .of fuel. Can these problems be overcome with new technology and if so at what costs? Can we have new technology which we can use to cut down the cost of processing so that nuclear fuel will be economical and produce energy in a satisfactory manner. So what I propose to do is since the time is short, I would like to highlight the future strategies. For this purpose, what is more important is to have some searching questions from the audience so that the experts here can reply to that particular point. I take this is acceptable to all of you. So we start if there is anybody who wants to ask questions regarding locating the ore. I see great experts in the audience also. Any questions on that. I am sure Shri. Phadke has some questions. Shri. ! Phadke (Former Director, AMD) : May I ask Shri. Saraswat what new methods have been adopted for searching out uranium in recent years.| I Dr. ?.K. Iyengar : Which Shri. Phadke did not know when he was in service. Shri. A.C. Saraswat: There are no new methods. What we have achieved is based on the know how which we already had. We have only made a maximum utilisation of manpower. The initial thrust of uranium exploration was in Singbhum district of Bihar. The exploration in Meghalaya, Cuddappah and Bastar regions are on conceptual basis. Interaction with other outside agencies was also resorted to. Gradually R & D efforts are increasing. Availability of drilling rigs -1024- have increased gradually and in the 8th Five Year Plan it will increase to about 1 lakh metres per year. In the rest of the world, the average drilling rate per ton of ore is 200 metres whereas we are having 50 metres per ton. In Meghalaya, it is one metre per ton. Geochronology and "airborne survey and bacterial leaching are some of the techniques that have been followed. And various workers have been allotted to handle work for at least 5 to 6 years so that there is more commitment. We are committed to a programme of 10,000 MWe by 2000 A.D and hence the role of AMD in DAE is going to be the key factor. P.R. Roy (Director, Materials Group, BARC): There seems to be some disagreement in the uranium availability data between the geologists and the miners who are unwilling to take this data at its face value. A.C. Saraswat: During the last 4 1/2 years, we had over 20 meetings of uranium profile committee to define our uranium resource position. The figures provided to the committee have been accepted by UCIL. The figures given were 47,500 T in July 87, 62,800 T by July 1988 and 71,000 T of uranium oxide by November 1989. About 23,000T of uranium have bean added during the period 1987-1989. This increase in availability has come mostly from Turamdih (East), Turamdih (W), South and other places in Singbhum Dt. Thus, the mill at Jaduguda and the mill proposed at Turamdih can be fed without interuption for years to come. The problem is that mining people are reluctant to go to places away from established places. J.L. Bhasin: If Chairman permits, I would like to interrupt here. Dr. P.K. Iyengar: I think this is an interdepartmental affair. Anyway, UCIL is supposed to have accepted certain figures. What should they accept? Is it a figure derived from multiplying certain volume by concentration or is any defined procedure followed? Is UCIL satisfied with this procedure? A.C. Saraswat: AMD collects data on the extent of the deposit, grade etc and presents it to UCIL for their study. Dr. P.K. Iyengar: This data is all collected by AMD and you give it to UCIL. Is it accepted by UCIL? J.L. Bhasin: Problem regarding this matter is this. When we have to take a decision regarding opening a mine or setting up a mill, we have to take into account certain facts like how long they can be kept working and at what throughput. Por example, at Jaduguda we have been keeping tho mill working at over 85% capacity for over 20 years. So much infrastructure has to be built up to organise these that any unexpected shortening of its life-span will be disastrous. There are certain well-established procedures for arriving at this information so that full data is available to us. People should not believe that as soon as geologists say that there is a good uranium deposit . , we should go ahead to open a mine and a mill there. Let us take the case of Bodal deposit about which a lot was said a decade back. It was a flash news at that time. What is the situation today. We are not fighting shy of opening new mill or mine, but we want to be sure that our decision is correct. Let us take Domiasiat deposit. We do not - 1025 - have even a sample to teat the metallurgical flow sheet for extractina uranium. If we have to open this mine, it is to process the ore and we would like to have complete information before plunqing ahead. Dr. P.K. Iyengar: Do you want to be certain that the sample they produce should have the right concentration or how do you ao about it? J.L. Bhasin: Sir, it has to be satisfactory both in respect of quality and quantity. We want to have conclusive data that so many millions of tons of ore of such grade is available in the deposit. We have to test a representative sample for metallurgical characteristics of the ore before we can justify opening a new mine, a mill and build necessary infrastructure at the site. A.C. Saraswat: I would like to react to the statement regarding Bodal. Perhaps this is a stray example. It is not that there is no uranium even a gram of it there. Let us take Narwapahad, some 20 years back. UCIL was requested to take up this deposit for regular mining but it was rejected. But today the same mine is being developed by UCIL. On Domiasiat we are confident about its grade and the availability of uranium which lies within 40 metres from the surface. The figure proved today is more than the figure given when Jaduguda mine was opened. Unless we take up this work now, certain procedural factors like environmental clearance/forest conservation clearance etc cannot be taken up in hand. And these procedures can be expected to tie you down for some years. And during this period further reserves would be proved. The deposit as proved today can sustain a production of 300T U^Ogper year even assuming 40% loss during mining and milling steps. Dr. P.K. Iyengar: To ensure that UCIL takes your fiqures to be correct, you have to take larger samples and at more frequent intervals. Is there any technological improvement in your methodology to improve the confidence level of your data and do it faster? Now that you have over 30 years of experience. A.C. Saraswat: Each deposit has its own peculiar characteristics and problem. The best person available here is Mr. M.K. Batra, previous CMD of UCIL who is in the audience. He has visited the Domiasiat deposit. Mr. Bhasin is yet to visit it. Dr. P.K. Iyengar: You have not answered my query about improvement in the technology used by AMD for proving a deposit. A.C. Saraswat: During the last two years our experienced geologists and physicists are camping at the site and have systematically collected the data using the best available equipment, and I see no chance of making any mistake. J.L. Bhasin: Unless UCIL has some authentic samples of ore representing the ore body, on which we can carry out leaching studies etc, we cannot commit funds to establish a mine and mill and other infrastructure. He want to produce uranium but we want our decision to be sound. N. Vittal (Additional Secretary, DAE): Thoaa discussions bring out the importance of management aspects also. This point has also been -1026- stressed earlier by Dr. P.K. Iyengar. Now we seem to have come to some such situation. I am reminded of an old story. There was a rat which was being harassed by a cat. The rat went to an old owl which was known to be very wise, seeking advice. The owl thought for a while and suggested that the best way would be for the rat to become a cat. The rat went back thinking deeply and returned after a few days and requested the owl to advise on how to become a cat. The owl replied that it is there only to give policy advice and its implementation is to be done by the rat. The situation is identical. He see on one side AMD and UCIL on the other. AMD is stressing that it has a deposit and UCIL is in the role of a rat trying to implement the decision. In the context of uranium technology, I would like to give an overall picture because I happen to deal with nuclear power programme also. We are talking of 10,000 MWe of nuclear power. .When we prepared the first profile in 1984, we had visualised an investment of Rg.14,000 crores. When we revised it last year, it came to Rs.22,000 crores. This year we are talking of Rs.29,000 crores. Like the story of Jack and the Beanstalk, this figure seems to be going up higher and higher. One of the reasons is that the uranium requirements are going up. Originally, the fuel burn up was 6,700 MW days per ton of uranium. Now it has come down and now our uranium requirement has gone up by 16 per cent. This is really a technological challenge for people from Nuclear Power Corporation. I have an assurance from Mr. Saraswat that enough uranium is available. But most of it is underground. How to get it to the fuel fabrication plant? I want to point out that it takes tremendous amount of money and time to bring out uranium from the day it is discovered. Generally, the time required is minimum 7 to 8 years. We had Narwa and Turamdih in view way back in 1970 and a budget of Rs.200 crores was provided early in 1981. It took years for getting environmental clearance and the "avatar" in which the project is approved now costs about Rs.495 crores. That too for Part 1 only and the work has just started. Thus uranium technology is a field where there is a greater need to reduce expenditure. Chairman pointed out earlier, and his question regarding availability of new technique/technology has not been answered. I think we have to speed up our inputs in developing a deposit, in mining/extraction of uranium etc. Delay means more money, value of time hao to be realised at all levels of decision making. T.K.S. Murthy (Director, Chemical Engg. Group-Retired, BARC): Can I ask some questions to Director AMD as a matter of speculation? From the accumulated knowledge of geology, can you say that there are or there are no chances of finding a much better grade or much bigger deposits of uranium in India or due to some geological reasons such a possibility is ruled out. A.C. Saraswat: Over the years, we have only found small deposits of higher grade and bigger deposits of low grades 0.04 to 0.06% U— Og • Only in recent years, we have made a breakthrough in discovering a better grade deposit of sandstone type. We have the know-how but we have a management problem that the inputs so far have not been sufficient for a country of our size. Technically and scientifically, there are chances of finding better grade and bigger deposits in India. -1027- Dr. P.K. Iyengar: Thank you, I would like to make a short comment on what has been said. It was mentioned that for the nuclear power profile we require so much of uranium. Everybody involved has been cautious and conservative, with the result I have a feeling that the need for uranium might have been bloated to a higher figure and we have asked for huge funds. Let us start from the beginning. Is there no possibility of increasing the burn up of fuel in the power reactor? Is it possible for us to buy at least part of this uranium from outside by trading technology in return for uranium. I don't believe that uranium is going to be a big problem. Most important factor is to have a proper attitude to it. He know that we don't have everything. We should start to think how we can optimise our requirements. This should set right the situation under which our requirements go up year by year. It is a question to bti set right by reactor physicists. Our need for funds are going up every year. It is likely that the Planning Commission nay direct DAE to borrow money from the market. I think there is a need for greater co-operation between AMD and UCIL in reducing the time required for making uranium available from a known deposit. I think we can spend a few more minutes on this topic of ore. D.V. Bhatnagar (Supdt. C.R&D - Retired, UCIL): I would like to suggest that when we test ore samples for metallurgical trials, we should insist on testing the bore-hole cores. This is very important. Before finalising the leaching conditions core samples which represent the entire ore body should be carefully tested. Dr. P.K. Iyengar: Do you mean to say that we should have one bore hole core sample large enough to carry out all tests? D.V. Bhatnagar: No, we should take as many samples as possible. Each to represent certain areas of the deposit. R.K. Garg (CMD, Indian Rare Earths Ltd.): What we have been doing so far and what we need to do in the future has to be clearly defined. For example, let us start with AMD's drilling operation. Mr. Saraswat has mentioned that our inputs in drilling operations are small in comparison to other countries. At the same time, we should realise that drilling is not the first thing to be done. After a thorough geological survey etc. initial drilling should be on scouting basis aimed at proving the availability of say 3000 T of uranium. Then only one should go for extensive drilling. This change in strategy which will save time may have to be adopted by AMD. Deposits of Jaduguda and Narwapahad involved several agencies in testing ores etc. This strategy of involving other agencies of DAE especially in carrying out leaching tests and preparation of metallurgical flow sheet etc. is sound. If work is done in parallel, it would help reduce the time factor for taking investment decision and this is very essential. Dr. P.K.Iyengar: Now we move over to uranium ore processing. Is it true that the chemistry of uranium is rather difficult? During the first half of this century, uranium was not considered to be important. I remember a VIP visitor to Apsara in 1959 who saw the large quantity of hydrofluoric acid used for produciny UR). . He mentioned that we are using one of the costliest chemical reagent. - 1028 - How nan you thrive on it, ha wanted to know, since every gram of it is imported. Uranium metal requires an elaborato process. X know to-day KF is made in the country. X would like to know whether we have made any significant change in uranium procasa chemistry during the last 25 years. Any new method to extract uranium from ore involving lesser number of steps or at lower cost and to yield higher grade concentrate? T.K.S. Hurthy: X would like to make some comments. We are processing uranium ore at Jaduguda for the past 20 years. Here we do not use HP but a common and cheap reagent H^SQ^. we are lucky in this respect that our ores though low grade consume less acid than in most of the other countries. For recovery we have two specific routes - ion exchange and solvent extraction. Today the ion exchange resins are mads in the country and the chemistry of these two routes are well understood and remain unchanged. We have no problem in this regard. Use of HP comes only when we come to final conversion of uranium concentrate that too after purification to nuclear grade. And the consumption of HF is also low. Except for the risk involved, we have no problem. Today HF is made in the country. Thus today from the ore to the reactor fuel there is no step where we are dependent on imported chemical. Dr.P.K. Iyengar: Let me ask some questions. What is the efficiency when you mine the ore? And what is the efficiency when you process this ore? T.K.S. Hurthy: X would like to answer the second question. In general, when we treat the ore at Jaduguda, about 93-95% of uranium is leached. About 5% being bound with refractory minerals and hence not easy to dissolve. Out of the dissolved uranium, about 90-95% is finally recovered. Our soluble losses are on.the higher side. This is a problsm needing some improvement. When we treat ore from Narwapahad the leaching efficiency is lower to an appreciable extent due to mineral characteristics, to give a final recovery of about 80%. Dr. P.K. Iyengar: You say that there are two stops each having a yield of 90-92% which works out to a net yield of 80-82% only. Is this correct. T.K.S. Murthy: This is very close to the figure. In the Mill, the average overall recovery is said to be around 90 per cent. Perhaps Mr. Bhaain may like to comment on this. J.L. Bhasin: The overall recovery in the Mill at Jaduguda is between 85-90%. We havo optimised at a figure of 90% beyond which improvements will require extra investment which is not justifiable. The overall mining recovery is tlso around 85% sinco about 10-15% is left as it is locked up in the pillars of the mine. At the stage of the concentrate, it should be minimum 85-86% of the mill feed. Dr. P.K. Iyengar: bat ma get this clear. You say that there ia o recovery of about 90-92% at certain stage and about 85% at the final stage. - 1029 - T.K.S. Murthy: Yes, to recover the remaining uranium, we will have to invest in more equipment and facilities which may work out to be expensive. Dr. P.K. Iyengar: I am concerned with 5 per cent increase in availability of uranium which we are supposed to- organise by augmenting production facilities and here I find we are losing as much as 10-20% of uranium in our process steps. J.L. Bhasin: This 20% is not something which can be extracted and recovered. Dr. P.K. Iyengar: I am of the opinion that once you have done the. mining of ore, you should extract all uranium available in the ore. With uranium we must be having some other valuable constituents which can be recovered as by-product. J.L. Bhasin: At Jaduguda, the operations are in two parts - mining and milling. The overall recovery from ore is around 85-90%. Capacity of the plant is 1000 T ore/day. This plant has to be operated at certain minimum capacity and this is 70 per cent. Our ores contain by-products which are recovered - like magnetite, molybdenite and chalcopyrite concentrates. Dr. P.K. Iyengar: There was a suggestion on improving the overall recovery by 5 per cent in milling. Would you like to comment on that? J.L. Bhasin: Yes, we have been looking into this problem and have made some plans. S. Sen: I think earlier the cost of production was mentioned. I would like to point out that mining cost itself is over 50% of the overall cost. Hence any cost reduction study should include mining operation also. Dr. P.K. Iyengar: How do you propose to reduce mining costs. You must be leaving lot of ore behind. S. Sen: Earlier, it was mentioned that about 15% of ore is lost. A.C. Saraswat: If I may interrupt, the losses during mining are more. Sometimes touching almost 30%. During milling, we lose another 15- 20%. Hence, during calculation of uranium reserves we have assumed an overall loss of 40 per cent before arriving at the figures. Of course, the entire picture changes if we would reduce this losa by introducing new technologies like solution mining in geologically favourable localities. J.L. Bhasin: Lot us be on the ground level. We can mine the ore and process it leaving all the overall losses as mentioned earlier. Talkino of solution mining etc are impractical since I do not see any deposit in India where it can be applied and our endeavour should be to improve our mining methods to reduce ore losses. Dr. P.K. Iyengar: Do you leave large pillars behind? - 1030 - J.b. Bhaain: Ha do mining by scientific method. Xbout 40% of the ground ore (couraa sand) ia sent back to the nine after extraction of uranium. Dr. P.K. Iyengar: What ia the overall loaa ore in the mine? i7,li. Bhasin: Xt is around 10-15 per cent only. S. Sen: Solution mining has been carried out in US etc in areaa where population density is very low. Hence, problems like ground water pollution etc is not given much importance. In our country, this factor may be critical. The other point raiaed by Shri. T.K.S. Murthy was reducing the soluble loss. Our ground ore especially from Jaduguda contains too much fines which clog the filters. One of the ways to improve the filtration is to employ belt filters. Mr. Ramakriahnan who is with us may like to say something about it. N. Ramakriahnan: Earlier at Jaduguda, we had supplied 2 stage rotary drum filters. Those days we have highly efficient belt filters with multistage washing facility as a single unit thus avoiding 2 stage filtration. S. Sun: X understand we can bring down the soluble loss to as low 2 per cent. That is the figure given elsewhere. 2 N. Ramakrishnan: We have belt filters of upto 150 M area with upto 3M width. Dr. P.Kl Iyengar: Are you introducing this in India? N. Ramakrishnan: Yes, we have already done that. P.K. Shah (FRD, BARC): What is the status of uranium recovery from sea water using inorganic ion exchangers. This involves only single step. Dr. P.K. Iyengar: Lot of work has been done on this subject in BARC and abroad. And X understand everywhere they have abandoned the work since it is uneconomical. Price of uranium is going down and hence there is no incentive for developing this technology. P.K. Shah: Xt should* be possible to just drop the ion exchange adsorber.in sea and then collect it later on after certain duration of time when it would have adsorbed uranium. This does not require any elaborate sat up. Dr. P.K. Iyengar: X think the Japanese and Germans have done lot of work on this topic and have left it. If Mr. Shah wants to carry out soma pilot tests in his Division he should do so and collect some data. T.K.S. Murthy: X want to refer to the solution mining technique discussed a little while ago. Moat of our deposits are hard rock formation and hence the leaching solution is unlikely to percolate through the ore body and the recoveries will be low. -1031- Dr. P.K. Iyengar: There is a suggestion from one of the audience that we should recover upto 50 per cent by this method and then do regular mining. T.K.S. Murthy: I see no logic in that since the cost of mining and processing per ton oC ore are not sensitive to the grade. Dr.P.K. Iyengar: Now let us concentrate on uranium processing. Is there any suggestion on this? K.S. Koppikar: I want to add a couple of points. Whether we do heap leaching or solution mining, the uranium recoveries are going to be lower by about 20 per cent. On one side, we are talking of our uranium reserves being only 50 or 60 thousand tons and hence we want to get the maximum out of it. We should just forget about heap leaching or solution mining if we do not want to leave 20 per cent of uranium in the ore. Another point that I want to stress ia the process route followed at Jaduguda. It ends up with magnesium diuranate (MDU). This concentrate gives headache at refining stage since it contains lot of silica. Even at Uranium Metal Plant we have a large number of drums filled.with silica cake which we now propose to send to Jaduguda. Is it not possible for us, after experiencing this difficulty for 25 years, to change over to an improvised technology? Instead of MDU why not get ammonium diuranate so that this silica problem can be solved once for all. Even at NFC, Hyderabad we expect to have the dissolution step for another 25 years and now they are planning to expand this plant, capacity to 500 TPY. We may have another plant at Turamdih whore this problem may not be present. But how do we tackle this silica problem at Hyderabad? We have to improve our technology at Jaduguda. Way back in early 60s we opted for a simple technology and chose magnesia which is available in plenty in our country. Though we were aware of the new technology of solvent extraction, we decided not to go in for it at that time in order to make the process simple. But after 25 years, we have to improve our technology. Now we are setting up a new mill at Turamdih where we will be producing about 350 T of U^Ogas MDU which may have to be sent for refining elsewhere in the country. In this plant, we would be eluting uranium from ion exchange column using dilute sulphuric acid. One step after it, if we introduce amine extraction, we can get ammonium diuranate (ADU). Even though late, we should still consider its introduction seriously. Another point is, we propose to have all our future fuel plants near uranium mines. This is not a sound practice. Because the working culture for purification/fuel fabrication plant is different from mining. If these two groups are located at the same place, we are likely to face problems. Dr. P.K. Iyengar: Mr. Bhasin, do you feel that this process can be introduced in your plant? J.L. Bhasin: Near our new plant at Turamdih, the NFC plant is also coming and we have thought of introducing this step there. -1032- >: c :ir- K;S. 'Koppikar: ' As a matter of fact, one of the reasons for/setting up new 'NFC'plant "at Turamdih was to avoid transport of over 1000 T/Y of silica coke from Hyderabad. Dr. P.K. Iyengar: What is your objection to changing this process at Jaduguda front MDU to Ammonium diuranate? K.S. Koppikar: One of the reasons given for not accepting this modification to the process was that the cost of processing would go up by about Rs.20 per Kg UsOgr. And this is when we expect the cost of U^Og from Turamdih to go up to Rs.4000/Kg. For transporting the silica waste from NFC, DAE is incurring heavy expenses. S. Sen: Even for Jaduguda, ADU was considered as the end product at the project design stage. But availability of ammonia and cost led us to choose MDU route. Me have enough experience of producing ADU and hence we should take an over all picture now and reconsider this matter. R.P. Verma: Supply position for ammonia at Jaduguda is even now not attractive. Moreover, we have completed all formalities for getting clearance from environmental agencies etc for a process as defined today and any change will lead to inordinate delay again. Hence, when NFC set up its plant, they can take up this matter. UCIL will give the Sulphuric acid eluate to them and they can produce ADU. N. Swaminathan: This entire exercise of setting up the plant at Turamdih is with a view to integrating it with UCIL plant process. This is a typical management exercise to overcome the problem of silica. Subsequently, due to some other compulsions, with a view to providing the large number of fuel bundles to NFC, decision was taken to set up the second expansion plant at NFC, Hyderabad itself before setting up the Turamdih plant. Now we are left with the problem of facing MDU and this problem of silica will continue. As per our present plans, we may process up to about 1000 T of U/year at Hyderabad itself. The integrated flow sheet developed at BARC has to be co-ordinated by BARC people. I do not think that NFC is geared to develop it and take it to fruition. NFC can co-ordinate the work and give managerial inputs and supply of equipment. As far as the process development or design of equipment is concerned, a separate group has to be created. For this BARC is more suited. For a production plant like NFC geared to solve day-to-day production problem, such developmental work calls for a tremendous effort. Dr. P.K. Iyengar: I would summarise this discussion by saying that the integration of the process for NFC and UCIL would benefit both and the work has to be carried out at BARC. Now we will consider NFC technology - What is the present status and what are the future needs. -1033- N. Swaminathan: The situation as it is today is bright and hopeful. Even though we are in a hand-to-mouth situation regarding supply of power reactor demands, last six months we have been consistently producing over 500 bundles per month. We are not looking back. Perhaps this is one of the best consistent performance over a 6 month period for a long time. May be we will touch 100 T mark this year, which would probably be the highest production in recent years. Ironically when this plant was set up in 1970, the expected capacity was 100T only. Today, the recovery is about 65% from powder to the finished assembly. Perhaps this is also one of the best efforts to be sustained over a long period of time. We had a lot of problems and bulk of the headache has been solved by the process people at NFC itself. Much of the problem is at the stage of end capping. All expansion plants are based on 60% recovery 'which has been well demonstrated. NFC is to be expanded. Though the first plant is yet to be sanctioned, it is expected to be ready 3 1/2 years from the sanction date. May be by 1993, this expanded plant nay be ready as per present estimate . We have studied various aspects of uranium precipitation like ADU and AUC. In my opinion we should persist with ADU. As far as oxide production is concerned, we have no problem since all equipment is of indigenojs ori-din. No foreign exchange payment is involved and hence the cost of production can be expected to be low. If the overall vecevery ia improved, the cost of production can be brought down. The cost of fuel in a nuclear power plant is only about 8 per cent which includes the cost of uranium concentrate and all process costs uptu £U'3l bundle. In our expanded plant, we are planning for improved affluent treatment, avoiding past mistakes on nitrate effluent handling etc. We propose to have better ventilation system for uranium plant. We will have a system of double containment and instead of rotor-blown system, we will have electrostatic precipitator for the first time, with absolute filters. These changes "may take another 6-8 months, and technology wise, there need not be any concern «£« far as NFC is concerned. Dr. P.K. Iyengar: These are very encouraging statements. To put up a new plant in 3 1/2 years and expect the same type of recovery is commendable. Any comments on this. S. Sen: It is only recently that the overall recovery has gone up from 45% to 60%. At Eldorado plant in Canada, they were getting a recovery of upto 80%. Dr. P.K. Iyengar: Where does the rest of uranium go? S. Sen: It is being recycled to dissolution stage and processed once again. It is not lost. Dr. P.K. Iyengar: Can you have any other process to improve the recovery? The sol-gel process for example. Have you any experience? P.R. Roy: since the process of sintering is involved where diametric control is not yet very good, percentage of loss during grinding would be substantial. This problem is more in BWR as compared to_PHWR where collapsible cladding is used. From German reports, we know that the overall recovery is about 85%. -1034- Because there are various stages of operations and during compaction, lot of powder is lost and all depends on the method for collecting it. It should be easy and in fact faster provided you do not allow any chemical impurity to come in contact. And we are still not very happy as far 83 the dry powder is concerned. If there is any chemical contamination, it means a lot as far as the process is concerned. Or. P.K. Iyengar: So it should be a process untouched by hand as far as possible. N. Swaminathan: I will explain it to you, Sir. There is a question on where is this 35 per cent loss. Out of this, it may be noted 10 per cent goes straight to assembly section including defective sampling. We have no NDT method of evaluating end cap bend. Even though we process well with bare tubes, we still lose a lot in assembly section. Perhaps it is not the powder technology which is influencing the 10 per cent loss, it is something else. Upto pellet stage, if you ignore this part, the recovery is 65% plus, let us say 75 per cent. There is one point, that is as rightly pointed out, the handling losses are very heavy to the extent of 10-12 per cent. That is from beginning to precompacting stage and the final compacting stage. One problem, and that is where the big debate is going on,is how to get a free flowing UO powder for which all are struggling. In such a case, we can have better mechanical transfer system and a closed circuit and we* will reduce all handling losses. Hence, they are not real loss in the sense that the material is lost. But it does account in once through recovery calculations. Hence this 10 per cent loss is a notional loss than a technological loss and it can be interpreted that way. That way, .upto pellet stage we can say that we have almost 75-80 per cent recovery. We have no answer right now for the 10-12 per cent loss at assembly stage. We have to look into this problem a little bit. S. Sen: One thing I wish to poin.t out here is that for better recovery of pellets, consistency of powder is very important. So far as the precipitation is concerned, they are now equipped to get good ADU powder. The quality of powder is very consistent now. But drying, reduction and stabilisation is still done in small capacity furnaces. They are planning to go in for bigger furnaces. Improvisations can be done if we go in for fluidisation and a moving bed for drying end reduction. The handling losses would be less and the product also would be better. They are thinking along these lines. Perhaps Shri. SwJfminathan may wish to make some comments. N. Swaminathan: Yes, there is a RBU process where they use AUC. In 12 seconds residence time, they are able to calcine, reduce and get the powder. From AUC to UO it hardly takes 12 seconds. That typa of furnace we have to develop. This will take time. It will not be in our present plant. There is scope and we are working on it. We have to have a fluidised bed furnace which ia perhaps the best way if you really want to have a control on the powder. But then we have to get a suitable powder for such a furnace. That is what we are trying to get. At least at dried ADU stage, we should get a homogeneous and uniform powder. Then we can think of fluidised bed. It is my -1035- personal view that the fate of the powder is sealed at calcination and reduction stage mostly,and for that we must have a feed in proper form. AUC trials on batch precipitation are now in hand in that direction. Dr. P.K. Iyengar: I think it is a very involved problem because we have spent about 20 years since AFD started producing UO^pellets, perhaps from 1972-73. S. Sen: I think this may be of some relevance. Canadians are still using ADU process, but they are using moving bed furnace. France is using fluidised furnace. Dr. P.K. Iyengar: That is all right. , It means that this has to be looked into in every way and changes have to be made at every step. Even perhaps the way in which material is handled by two people for moving from one place to another. Flowability of the powder is another point. J.P. Srivastava (NFC): ADU powder is not the same in all cases, even though we say ADU. It can be very flowable and can be fluidised. Other type of ADU is one which cannot be fluidised. For UO^. what we have been doing by continuous precipitation method is a gelatinous precipitate which cannot be fluidised. Recently what we have got by equilibrium precipitation method can be fluidised. Now we can definitely say we have something. As Shri. Swaminathan has said, yes, it is possible to fluidise ADU. But we have to produce ADU of that type. And we are already working in that direction. Already a moving type drier has been made and this work has been presented in a paper read at this symposium. Dr.'P.K. Iyengar: Thank you S.K. Mehta: Apart from process development, I think we should also stress on properly using them. I think a beginning has been made for carrying out a lot of mechanisation and I think this has to be accelerated. I am told that sometimes bolts are stuck in the premises and that recovery takes a lot of time and we have to open the entirei thing. If it could be mechanised, recovery can be much better. I have seen that in various units, production is very high in one month and very low in another month. There is lack of optimisation and co ordination of production. Reasons could be irregular power supplies. I think we have not succeeded in providing stable power supply to NFC, though they have done a big job. We have been talking for almost 12 years that the power supply should be steady. I think it discourages them a lot when a plant is shut down due to power supply. And when they want to restart, the time taken is much more. Dr. P.K. Iyengar: That means we have to build NFC along with a nuclear power station so that they can have a steady power supply? N. Swaminathan: I would liko to inform that we are going for augmentation of power supply. Three units each of 2.5 MW are going to be installed. We ore getting a captive power plant. 1036- Dr. P.K. Iyengar: My God ! That is expensive electricity. N. Swaminathan: There is no other way. Sometimes our production plants are idle for 3 to 4 months due to power shortage. Andhra is heavily dependent on hydro-power. And if we have 2 years drought we are in for a heavy power cut. Dr. P.K. Iyengar: Is there any other point to be covered? Now we go co other subjects. There is a proposal to augment our uranium availability to make nuclear fuel -one suggestion is to improve the process for recovering uranium from monazite. The second is to recover uranium from phosphoric acid plants. Let us have some comments and then we can close our response. S. Sen: How much uranium is available from these sources? R.K. Garg: I have mentioned about this in my talk in the inaugural session. You can get about 10 MT UxOgf per year if all the monazite is processed. At present all available uranium is not recovered from this source. We are planning to do that. We have plans to process monazite from Orissa and if we include that also we may be able to obtain about 20 MT UjjOg" per year from monazite, depending of course on the availability of monazite. N. Vittal: The whole discussion has been concentrated on uranium technology - the present status of this technology and its future implications. And before we leave, I would like to state our position on nuclear power and our uranium reserves. For our 10,000 MW nuclear power programme we are depending mainly on our uranium ores - what AMD explores and what UCIL mines and there is need to supplement our resources, which are mainly from the mines at Jaduguda, and the copper tailings,where by physical process we are getting,if I remember right, 27 MT out of 167 MT comes from copper tailings. In the future, there is tho possibility of recovering by chemical methods an additional tonnage of about 46-49 MT UjCyper annum. There are other problems also involved in this effort which are being studied by UCIL and also by Hindustan Copper Limited. In our thrust to additive recovery of uranium, there is one area being seriously looked into to substitute physical method with chemical method of uranium recovery. Now uranium from Phosphoric acid - the work has been carried out for a long time at 3ARC and now we are working on two proposals - one is to put up a pilot plant at FACT, Cochin at a cost of about Rs.2.5 crores or so and as our recent profile indicates we are thinking of recovering somewhere about 1750 MT V3O& over the whole period. On an average per annum we can get over 100 MT of UjOg for every million tonnes of P20^-produced in phosphoric acid plants. This is another source we are looking at. Third was monazite which has been already discussed. I wish to make a few points. We should constantly look ovor our processes for recovering uranium - both at mining stage, or recovery from copper taulings or at NFC, especially on how to improve our processes. If Canadians are achieving 85% recovery there is no reason why we cannot also achieve this. In all areas, we have to have a bench mark - we have to take ths best out of that. -1037- This is for NPC. As Dr. P.K. Iyengar said, they are very conservative, and this approach may build in so much of cushion that it can cover like grandmothers gown a multitude of technological inefficiencies. We should try to 3ee that we improve our performance and get the best out of our performance, and get the best out of the resources we have, and the best of ourselves. Vising our three resources, we should constantly improve ourselves. Dr. P.K. Iyengar: Thank you very much. I think we have got only five minutes since we are scheduled to close at 17.30 hrs. S. Sen: I think on Meghalaya uranium deposit nobody has thought of Project Management part. Each one is looking from his own angle. AMD is only interested in establishing the resources, UCIL is interested in mining and we are doing extractability studies, the local, self government will look after transport, roadways and communication. Dams are built by civil engineers to provide water supply, but the total discharge from this area has to go down towards Bangladesh border which is very close. Hence, water management will require a thorough study. This will have to be looked in a bigger concept. If we dispose off wastes containing manganese, radium or any other pollutant, it will go to Bangladesh. For this type of project, a total approach is necessary. Simply giving it to AMD or UCIL will not do. It has to be done on a total basis. Any approval has to come as a package. We should not have piecemeal approvals as at present. This is the approach to be made and all things have to be considered at the same time. If the Mill has to be erected within a few years, piecemeal approach is impossible. On Turamdih I am not making any comments since it is at an advanced stage of planning. But for a project like the one at Meghalaya, communication is very difficult. We have to take construction material, raw materials etc from a long distance and hence this project requires s total approach by a Nodal Body to look into every aspect right from the beginning. Otherwise, it will end up like a blind man describing an elephant. R.Dhanaraju: I feel that some of the basic differences between the Jaduguda ore and the ore from the new deposit presented by AMD now is not well appreciated. These have a bearing on the future programme of uranium extraction. About the accessibility of the area, Shri. Sen has already pointed out. UCIti people are familiar with the STB ore for over 20 years but have not been able' to see the difference between the metamorphic rocks at Jaduguda and the soft sandstone type ore proposed by AMD. For example, at Jaduguda ore is available at a depth of 400-500 M whereas at Meghalaya, it is confined to 32-35 M levels. Thus. in Meghalaya open cast mining is feasible. The recovery from ore at Meghalaya is likely to be less than 80% and uranium is present in che form of uraninite and also in other refractory components. Hence some extra effort may be needed to develop a leaching process. Similarly, the ore available in Cuddappah is carbonaceous containing about 0.03% Mo and studies have to be carried out on this ore also. -1038- J.L. Bhasin: It is a typical problem with uranium deposits. Had it been a copper deposit, we could have mined the ore, put a concentrator close by and then transported the copper concentrate to the smelter somewhere else. But, for uranium, the metal has to be extracted and recovered right at the mine head. This involves setting up a mill and a host of other facilities. At Meghalaya, the additional problem of roads and infratructure has to be solved. And this involves a huge investment. No doubt if the ore is available at a shallow depth, it would be an advantage and we will be happy. But we have to look at the problem in complete totality before we take the investment decision. Dr.P.K. Iyengar: I think we have had enough discussion on this topic. I feel that we had a wide ranging discussions on several topics-from mining, ore processing to refining etc. From what we have heard over the last one hour, we believe that there is enough ore, but the time required to produce uranium at economic lost consistent with ecological problems, and other social problems is still very large. Each ore has a different characteristic and should be looked at scientifically and chemistry must be made use of to recover uranium preferentially to determine the quantum of investment required. We can isolate these problems and attend to them. But one question that comes to my mind is why the basic problems connected with uranium technology are not getting solved at the right time. Is it the lack of R & D effort in the sub-units like UCIL, AMD, HFC, etc. At NFC, everybody seems to be so busy with fire fighting of problems that arise every day that they do not seem to have the time or the mechanism to try new ideas. They have the managerial experience and management has to intervene to net the thing right. In the last remark, the difference in characteristics of ore found at Jaduguda and Domiaaiat have been mentioned. He should take all these into account and finalise a process which in spite of large investment on roads etc as mentioned by Shri. Sen, can still give uranium at an acceptable price. And at the same time, we have the responsibility to encourage wider participation of people from all parts of the country and a li.tle investment in Meghalaya for sociological reasons may not be out of place. From my point of view, I feel that what is required is a greater effort and more emphasis on problem solving at the individual level in each of the constituent units and a co-ordination of such effort in a meaningful way at conferences like the present one. I am happy that tfie first such meeting has taken place now. I am sure we are prepared to look at our own problem and express ourselves clearly how we propose to solve them. Management can then take a decision on how to go about it in the future. It is true that fuel cost is only about 8% of the nuclear power generation cost. May be it may go upto 20%. Due to increasing.overall costs, we should economise on each and every aspect of nuclear fuel cycle. I am sure some of our efforts in reprocessing and recycling of uranium can help to reduce this cost and also make significant change in our total requirement of uranium for the entire programme. I believe that this -1039- will be looked into in depth. There are probably alternate methods of fuelling PHWRs. I have been told that initially we based our calculations on 6,700 MWD per ton of fuel and now we havs come down to 5000 MWD/T. Some factors influencing this decision are associated with the operation of the fuelling machine. Some are associated with our reluctance to try out alternatives to the existing systems. I would only like to conclude that it is a good experiment to put our heads together. Conferences like this will encourage all of you 'to put down your thoughts in a more systematic manner so that it can be taken 'note of by the management. I thank all of you for participating in this symposium. Vpf