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Molten Salt Chemistry - Part I

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Molten Salt Chemistry - Part I B.A.R.C.-670 GOVERNMENT OF INDIA ATOMIC ENERGY COMMISSION MOLTEN SALT CHEMISTRY - PART I. PREPARATION OF CeF3, PuF3, ThF4 AND PURIFICATION OF LiF-BeFz-ThF4 MIXTURE by P. N. Iyer, R. Prasad, V. N. Vaidya, K. Nag, Z. Singh, D. D. Sood and M. V. Ramaniah Radiochemistry Division BHABHA ATOMIC RESEARCH CENTRE BOMBAY, INDIA B.A.R.C.-670 GOVERNMENT OP INDIA ATOMIC ENERGY COMMISSION o • PI MOLTEN SALT CHEMISTRY - PART I. PREPARATION OP CeF , PuP3, Th?4 AND HJRIPICATION OF LiP-BeP2-ThP4 MIXTURE by P.N. Iyer, R. Praead, V.W. Vaidya, K. Nag, Z. Singh, D.D. Sood and M.V. Ramaniah Radiochemistry Division BHA3HA ATOMIC RESEARCH CENTRE BOMBAY, INDIA 1973 ABSIRACT Equipment set up for the preparation of CBJ-, PUP* and and purification of LiF-BeF.-ThF. mixture for the determination of solubility of CeF, and PuF, in ternary Mixtures of LiF BBF -ThF. and binary mixtuiea of LiF-ThF. are described. Using 144 these equipment the preparation of 75 g of CeF., tagged with Ce and 60 g of PuF_ and purification of 1*2 kg of LiF-BeF.-ThF^ mixture have been carried out. The analysis of compounds for fluoride by pyrohydrolyais and ths estimation of Li, Be and Th in the salt mixture has been reported. MOLTEN SALT CHEMISTRY . PART I. PRLPAHATIUN OF CeF PuF,, ThF. AND PURIFICATION OF LiF-BeF -ThF. MIXTURE o *» 2 4 by. P.N. Iyer, R. Prasad, V.N. Vaidya, K. Nag, Z. Singh, O.O. Sood and M.V. Ramaniah 1 , INTRODUCTION Indian NuclBar Pouio? Programme is well under way with the commissioning of Tarapur Atomic Power Station and thg construction of Rnjasthan Atomic PoigBr Project and Madras Atomic Power Project in progress. As the next step in the reactor programme, various breeder concepts have been evaluated from the point of vieiu of economy and fuel utilisation and fast breeder reactor is onB of them. The Depart- ment of Atomic Energy is building a Fast Breeder Test Reactor at Madras. Another concept uihich looks promising is that of Molten Salt Breeder Reactor* It uias decided in 1967 to investigate the feasibility of using such a reactor for Indian Nuclear Power Programme with particular Bmphasis on the use of plutonium. The molten salt reactors are attractive on Bccount of their good potential for the utilisation of thorium and plutonium provided 233 PuF, is sufficiently soluble in the snlt matrix. Enriched uranium, U o and plutonium are the possible start up fissile materials for thesB 233 reactors. A .nolten salt reactor utilising U as the fissile material can have a conversion ratio greater than unity and it la termed a Molten Salt Breeder Reactor (MSBR) . ThB reactors utilising enriched uranium or plutonium as the start-up materials act as converters (con- version ratio less than 1) but change over to breeders with the 233.. 232_. - , , U- Th fuel cycle. Conceptual Molten Salt Breeder Reactors are thermal spectrum reactors which use fluorides of fissile materials dissolved in e suitable molten salt mixture as fuel. Graphite is used as the moderator. For the salt matrix, fluorine, lithium-7 and beryllium have been chosen ae major constituents because of their low neutron • lbourpLian cross-sections and also because of the reasonable solubility of fissile and fertile Fluoridoa in mixtures of lithium and beryllium fluorides at temparaturas of practical interest. The fertilB thorium as the tetrafluorida is dissol- ved in the fluoride salt matrix to form a separate blanket stream or diroctly added to the core salt itself. Tha reactor is fcormi-id ag a tuio- fluld reactor or single-fluid reactor accordingly. Currently the single- fluid, two region concept is being preferred at OHNL because it leads to comparable breeding and eliminates the necessity of using graphite as ••• plumbing material to separate salt streams. These reactors operate at about 1300 f and at atmospheric pressure. Hoat is transferred from the fuel salt to coolant salt in an lntormeidi nte heat exchanger., A outactic mixture of sodium f1'uoroborate and sodium fluoride is selected as the coolant salt. The coolant salt passes through a stream generator uhnrti au» critical steam is generated for power gonuration. Another side stream of the cora fuel is continuously passed through a chemical processing system for the separation of protactinium nnd fission products. As there is no QXCGSS ro-'ictiuity built in the ro.u-.tar, provision for continuous on-line refuelling j c m th. MnJ. Li.i-i Salt Brt.'udar Hcnc tur.'j li.iuc ii nuinbiT of adv.wiL. yra o\/::r l.lir othur rtijetor syatoms. Thuao are I. Lou specific inventory (approximately 1 kg/MUe) as comparad to other breeder reactor designs, II. Absence of material handling problems such as fuel fabrication etc., III. Possibility of in-lino fuol roproceocing, IU. High negative tempnr^iturQ coofficient or reactivity s-nd V. Absence of positiuo BXCODS rcuctiuity in the raactor. Somo of thG!3e features have boon brought out by the opnrat.lon of a 8 MU(t) Molten Snlh .Tmctor F.xpnrimf.nt at UiUlL. Thoro am of couroe a number of drawbacks c.lso in this reactor system Some of theoo are t 3 t 1) Need to pre-hsat tha equipment and to keep the equipment above ths malting point of tha salt mixture at all times, ii) Problem of long-term containment and recirculation of fuel, iii) Necessity for techniques of remote maintenance, it/} Problem of graphite damage at high power densities to be used in the reactor and v) Rigid requirement of leak tightness and component reliability. Extensive research and development work on all the aspects of Hoi ten Salt Reactor Technology has been in progress at Oak Ridge National Laboratory for more than 25 years and a recent review summarises the present status of this concept . The idea of using fusBd-fluorides as reactor fuel originated at OHNL and has been extensively studied since 1950* The basic feasibility of Buch a reactor system was demonstrated by the Aircraft Reactor Experiment (ARE) in 1954 whore NaF-ZrF.-UF. (53.1 - 40.7 - 6.2 mole pet) (2) 235 uias used as fuBlx '. Enriched uranium containing 93.4 pet U uas used. To test the types of fuels and equipment that mould be used in an MSBR the 0 Hlil(t) Molten Salt Reactor Experiment (MSRE) uas built and operated from Dune 1965* . This reactor operated at 1200 F and at atmospheric pressure. The initial fuel charge contained LiF-BBF_-ZrF^- UF. of composition 65 - 29 - 5 - 0.9 m/o and the uranium uas about 235 235 2Z% U. Tns molting point of the salt is 840°F. Operation with U as fuel uas terminated in March 1968. Uranium from the salt uas remove23d3 by fluorid^4B l volatility techniques and the reactor uas recharged with U as fuel* ' and operated from October 196B to December 1969. 233 The MSRE was the fir9t reactor in the world to operate on U fuel. tDurino demonstratg thB lase t thphase suitabilite of MSREy operatioof PuFn, aaboua a tmak 30a 0u gp ofuef PuFl fog r wa"SRs alsE o' added With the successful operation of MSRE the studies are nou directed towards development of MSBH. A number of designs for power reactor systems of this type have been proposed and evaluated at ORNL, Some of t 4 I tha designs are as follow9l A. Turn-region two-fluid MSBR 235^233^6) Iii thio design the fissile and fartile streams pass through asperate graphite tubes within the reactor core.. Same of *hs basic data fqr this reactor is as follows! Power 1000 Mlit(e) Fuel composition Blanket stream composition mole pet mol pet UF 63.60 LiF 71 BeF 32.1 BeF 2 ThF4 O.QO ThF4 - 27 UF 0.22 4 Breeding ratio - 1.05 Specific inventory 0.77 Kg/MU(o) B. Two-region two-fluid MS8R with protactinium removal Composition of fuel and blanket is Bame as in (A) Breeding ratio 1.07 Specific inuentory 0.68 KG/MW(e) C. One-fluid two-region W3BR In (A) and (B) tho use of graphite tubes in the reactor core posed serious problems and led to this concept. In this design the fissile and fertile materials are combined in one stream. Two regions are obtained by having different fuel- salt to moderator ratios in the inner and outer regions of the reactor. The central region is graph!to-moderated and is critical, whereas the outer region does not have any graphite and is sub- critical. ThB basic design data is as follows I Power 1000 MU(e) Fuel-cum-fartile salt composition mole pet LiF 71 .7 16 .0 ThF, 4 12 .0 t 5 i UF4 n.3 Breeding ratio 1.077 Specific inventory 1.6 Kg/NW(e) The concept (C) is the present favourite because It is technolo- gically simpler than the other two concepts. Though this design concept ie the best from the point of view of breeding gain and fuel inventory requirements! it envisages on-line reprocessing (which is not yet developed on an engineering scale) for the removal of protacti- nium and fission products and also requires replacement of core graphite every four years* As tha high performance MSBR»8 suffer from tha above dis- advantages, as an intermediate stage towards development of MSBR ( large HSRE type converters of 300 Mlil(e) and 1000 Wil(e) are being considered. These intermediate type of reactors envisage batch reprocessing for recovery of fissile material and discard of thB salt matrix with fission products and a core graphite life of 30 years* With addition of on- line reprcessing these reactors become breeders* The basic design data for a plutonium fuelled MSR with batch processing every 8 years and 30 year graphite life has fa sen published recently and is given b el out Pouer 300 MU)(e) Salt composition mole pet LiF 72 18 ThF4 10 Breeding ratio 0*84 Specific inventory 1.9 Kg/Hti/(e) ' .
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