ISSN 1810-5408

Nuclear Science and Technology

Volume 4, Number 1, March 2014

Special issue for the 30th anniversary of the inauguration of the restoration and expansion of Dalat nuclear research reactor (20 March 1984 - 20 March 2014)

Published by VIETNAM ATOMIC ENERGY SOCIETY

NUCLEAR SCIENCE AND TECHNOLOGY

Volume 4, Number 1, March 2014

Editorial Board

Editor-in-chief Tran Huu Phat (VINATOM)

Executive Editors Vuong Huu Tan (VARANS) Le Van Hong (VINATOM) Cao Đình Thanh (VINATOM)

Editors Ngo Quang Huy (HUI) Phan Sy An (HMU) Le Hong Khiem (IOP) Cao Chi (VINATOM) Dao Tien Khoa (VINATOM) Nguyen Nhi Dien (VINATOM) Do Ngoc Lien (VINATOM) Bui Dieu (NCI) Dang Duc Nhan (VINATOM) Le Ngoc Ha (Tran Hung Dao Hospital) Nguyen Mong Sinh (VINATOM) Duong Ngoc Hai (IOM) Le Xuan Tham (DOST of Lamdong) Le Huy Ham (VAAS) Tran Duc Thiep (IOP) Nguyen Quoc Hien (VINATOM) Le Ba Thuan (VINATOM) Bui Hoc (HUMG) Huynh Van Trung (VINATOM) Nguyen Phuc (VINATOM) Dang Thanh Luong (VARANS) Nguyen Tuan Khai (VINATOM) Nguyen Thi Kim Dung (VINATOM) Hoang Anh Tuan (VAEA) Pham Dinh Khang (VINATOM)

Foreign Editors Hideki Namba (JAEA, Japan) Pierre Darriulat (INST) Philippe Quentin (CENBG, CNRS, France) Myung Chul Lee (WFNM) Yang (KAERI, Korea) IU.E.Ponionzkevich (DUBNA , Russia) Kato Yasuyoshi (TIT, Japan)

Managing Secretary Nguyen Trong Trang (VINATOM)

Science Secretary Hoang Sy Than (VINATOM) ...... Copyright: ©2008 by the Vietnam Atomic Energy Society (VAES), Vietnam Atomic Energy Institute (VINATOM). Pusblished by Vietnam Atomic Energy Society, 59 Ly Thuong Kiet, Hanoi, Vietnam Tel: 84-4-39420463 Fax: 84-4-39424133 Email: [email protected] Vietnam Atomic Energy Institute, 59 Ly Thuong Kiet, Hanoi, Vietnam Tel: 84-4-39420463 Fax: 84-4-39422625 Email: [email protected] ...... Contents

Results of Operation and Utilization of the Dalat Nuclear Research Reactor Nguyen Nhi Dien, Luong Ba Vien, Le Vinh Vinh, Duong Van Dong, Nguyen Xuan Hai, Pham Ngoc Son, Cao Dong Vu...... 1

Design Analyses for Full Core Conversion of The Dalat Nuclear Research Reactor Luong Ba Vien, Le Vinh Vinh, Huynh Ton Nghiem, Nguyen Kien Cuong...... 10

Conceptual Nuclear Designof a 20 MW Multipurpose Research Reactor Nguyen Nhi Dien, Huynh Ton Nghiem, Le Vinh Vinh, Vo Doan Hai Dang, Seo Chulgyo, Park Cheol, Kim Hak Sung...... 26

Some Main Results of Commissioning of The Dalat Research Reactor with Low Enriched Fuel Nguyen Nhi Dien, Luong Ba Vien, Pham Van Lam, Le Vinh Vinh, Huynh Ton Nghiem, Nguyen Kien Cuong, Nguyen Minh Tuan, Nguyen Manh Hung, Pham Quang Huy, Tran Quoc Duong, Vo Doan Hai Dang, Trang Cao Su, Tran Tri Vien...... 36

Production of Radioisotopes and Radiopharmaceuticals at the Dalat Nuclear Research Reactor Duong Van Dong, Pham Ngoc Dien, Bui Van Cuong, Mai Phuoc Tho, Nguyen Thi Thu, Vo Thi Cam Hoa...... 46

The gamma two-step cascade method at Dalat Nuclear Research Reactor Vuong Huu Tan, Pham Dinh Khang, Nguyen Nhi Dien, Nguyen Xuan Hai, Tran Tuan Anh, Ho Huu Thang, Pham Ngoc Son, Mangengo Lumengano...... 57

Progress of Filtered Neutron Beams Development and Applications at the Horizontal Channels No.2 and No.4 of Dalat Nuclear Research Reactor Vuong Huu Tan, Pham Ngoc Son, Nguyen Nhi Dien, Tran Tuan Anh, Nguyen Xuan Hai….62

Characterization of neutron spectrum parameters at irradiation channels for neutron activation analysis after full conversion of the Dalat nuclear research reactor to low enriched uranium fuel C.D. Vu, T.Q. Thien, H.V. Doanh, P.D. Quyet, T.T.T. Anh, N.N. Dien...... 70

Some results of NAA collaborative study in white rice performed at Dalat Nuclear Research Institute T.Q. Thien*, C.D. Vu, H.V. Doanh, N.T. Sy...... 76

A new rapid neutron activation analysis system at Dalat nuclear research reactor H.V. Doanh, C.D. Vu, T.Q. Thien, P.N. Son, N.T. Sy, N. Giang, N.N. Dien...... 84 Nuclear Science and Technology, Vol. 4, No. 1 (2014), pp. 01-09 Results of Operation and Utilization of the Dalat Nuclear Research Reactor

Nguyen Nhi Dien, Luong Ba Vien, Le Vinh Vinh, Duong Van Dong, Nguyen Xuan Hai, Pham Ngoc Son, Cao Dong Vu Nuclear Research Institute (NRI), Vietnam Atomic Energy Institute (VINATOM) 01 Nguyen Tu Luc, Dalat, Vietnam (Received 5 March 2014, accepted 26 March 2014)

Abstract: The Dalat Nuclear Research Reactor (DNRR) with the nominal power of 500 kW was reconstructed and upgraded from the USA 250-kW TRIGA Mark-II reactor built in early 1960s. The renovated reactor was put into operation on 20th March 1984. It was designed for the purposes of radioisotope production (RI), neutron activation analysis (NAA), basic and applied researches, and nuclear education and training. During the last 30 years of operation, the DNRR was efficiently utilized for producing many kinds of radioisotopes and radiopharmaceuticals used in nuclear medicine centers and other users in industry, agriculture, hydrology and scientific research; developing a combination of nuclear analysis techniques (INAA, RNAA, PGNAA) and physic-chemical methods for quantitative analysis of about 70 elements and constituents in various samples; carrying out experiments on the reactor horizontal beam tubes for nuclear data measurement, neutron radiography and nuclear structure study; and establishing nuclear training and education programs for human resource development. This paper presents the results of operation and utilization of the DNRR. In addition, some main reactor renovation projects carried out during the last 10 years are also mentioned in the paper. Keywords: DNRR, HEU, LEU, RRRFR, RERTR, WWR-M2, NAA, INAA, RNAA, PGNAA.

I. INTRODUCTION converted from HEU to Low Enriched Uranium (LEU) with 19.75% enrichment in The DNRR is a 500-kW pool-type September 2007. Then, the full core conversion reactor loaded with the Soviet WWR-M2 fuel of the reactor to LEU fuel was also performed assemblies. It was reconstructed and upgraded from 24th November 2011 to 13th January 2012. from the USA 250-kW TRIGA Mark-II reactor Recently, the DNRR has been operated with a built in early 1960s. The first criticality of the core configuration loaded with 92 WWR-M2 st renovated reactor was on the 1 November LEU fuel assemblies and 12 beryllium rods 1983 and its regular operation at nominal around the neutron trap. power of 500 kW has been since March 1984. The first fresh core was loaded with 88 fuel The reactor is used as a neutron source assemblies enriched to 36% (HEU- Highly for the purposes of radioisotopes production, Enriched Uranium). neutron activation analysis, basic and applied researches and training. As a unique research In the framework of the program on reactor in Vietnam, the DNRR has been Russian Research Reactor Fuel Return playing an important role in the research and (RRRFR) and the program on Reduced development of nuclear technique applications Enrichment for Research and Test Reactor as well as in nuclear power programme (RERTR), the DNRR core was partly development of the country. Safe operation and

©2014 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute

RESULTS OF OPERATION AND UTILIZATION OF THE DALAT NUCLEAR RESEARCH REACTOR effective utilization of the reactor expected at some main reactor renovation projects carried least to the year 2030 are a long-term objective out during the last 10 years are also of the DNRR. For this reason, so far the mentioned, too. Government has strongly supported for many specific projects in order to upgrade the facility II. BRIEF REACTOR DESCRIPTION and improve its operation and utilization. AND IT’S OPERATION The results of operation and utilization Main specifications of the DNRR are of the DNRR are presented in this paper and shown in Table I.

Table I. Specifications of the DNRR.

Reactor type Swimming pool TRIGA Mark II, modified to Russian type of IVV-9 Nominal thermal power 500 kW, steady state Coolant and moderator Light water Core cooling mechanism Natural convection Reflector Beryllium and graphite

Fuel types WWR-M2, dispersed UO2-Al with 19.75% enrichment, aluminium cladding Number of control rods 7 (2 safety rods, 4 shim rods, 1 regulating rod)

Materials of control rods B4C for safety and shim rods, stainless steel for automatic regulating rod Neutron measuring channels 6 combined in 3 housings with 1 CFC and 1 CIC each Vertical irradiation channels 4 (neutron trap, 1 wet channel, 2 dry channels) and 40 holes at the rotary rack Horizontal beam-ports 4 (1 tangential - No #3 and 3 radial - No #1, #2, #4)

Thermal column 1 Maximum thermal neutron 2.1x1013 n.cm-2.s-1 (in the neutron trap at core center) flux Main utilizations RI, NAA, PGNAA, NR, basic and applied researches, nuclear training

The reactor consists of a cylindrical increase the cooling efficiency for copping aluminum tank 6.26 m high and 1.98 m in with higher thermal power of the reactor. The diameter of the original TRIGA Mark II vertical section view of the reactor is shown in reactor. The reactor core, positioned inside the Fig. 1 and the cross-section view of the reactor graphite reflector, is suspended from above by core is shown in Fig. 2. an inner cylindrical extracting well so as to

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NGUYEN NHI DIEN et al.

~ 2000 mm

Rotating top lid

SR Pool tank Sh R Sh

Upper cylindrical ~ 6840 mm shell Extracting well RgR Concrete shielding

Spent fuel storage tank Thermal column door Sh Sh A Graphite Core Door plug (ex bulk-shielding SR experimental tank)

Fig. 1. Vertical section view of the DNRR reactor. Fig. 2. Cross-section view of the core with 92 fuel assemblies.

The reactor core has a cylindrical shape At present, the DNRR is operated with a height of 60 cm and a diameter of 44.2 mainly in continuous runs of 100 or 130 hrs, cm, that is constituted of 92 LEU fuel once every 3-4 weeks, for radioisotope assemblies, 7 control rods, a neutron trap at the production, neutron activation analyses, basic core center and 3 in-core irradiation facilities. and applied researches and training. The remaining time between two consecutive runs Type of fuel with a 235U enrichment of is devoted to maintenance activities and also to 19.75% of UO2+Al covered by aluminum physics experiments. From the first start-up to cladding is used. Each LEU fuel assembly the end of 2013, it totaled about 37,800 hrs of contains about 50.5 g of U-235, distributed on operation, namely a yearly average of 1300 three coaxial fuel tubes, of which the outermost hrs, and the total energy released was about one is hexagonal shaped and the two inner ones 760 MWd. Detailed yearly operation time of are circular. the DNRR is given in Fig. 3.

Fig. 3. Yearly operation time of the DNRR.

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RESULTS OF OPERATION AND UTILIZATION OF THE DALAT NUCLEAR RESEARCH REACTOR

So far, the reactor has proved to be safe disease therapeutics and 32P in injectable and reliable, as it has never suffered from any solution, 99mTc generator of gel type by 98Mo(n, incident, which significantly affected the )99Mo reaction have regularly been produced environment, and annual operation schedules and supplied once every 2 weeks. Other have been rigorously respected. The radioisotopes as 51Cr, 60Co, 65Zn, 64Cu, 24Na, unscheduled shutdowns were mainly due to etc. were also produced in a small amount unstable working of the city electric network. when requested. 53Sm in solution form was ready for labelling. Totally, about 5,500 Ci of III. MAIN RESULTS OF REACTOR radioisotopes have been produced and supplied UTILIZATION to medical uses so far with a yearly average in A. Radioisotopes and radiopharmaceuticals the last 5 years of about 400 Ci (Fig. 4) production correspondingly. Research on radioisotope and In order to support the application of radiopharmaceutical production serving 99mTc, 113mIn and 53Sm radioisotopes in clinical nuclear medicine and other users such as diagnosis and therapeutics, the preparation of industry, agriculture, hydrology, scientific radio-pharmaceuticals in Kit form for labelling research, etc. is oriented towards efficient use was carried out in parallel with the of the reactor. Via such research a variety of development of 99mTc generator systems. products including 131I, 32P applicators and About 17 labeled compounds kits have been solutions, 99mTc generators, etc. were produced. regularly prepared and supplied including Phytate, Gluconate, Pyrophosphate, Citrate, For medicine applications, radioisotopes DMSA, HIDA, DTPA, Macroaggregated HSA and radiopharmaceuticals have been delivered and EHDP, etc.. The annual production rate is to 25 hospitals throughout the country. The about 1000 bottles for each Kit which is main radioisotopes, such as 131I in NaI solution equivalent to 5000 diagnostic doses. and 131I capsule type, 32P applicators for skin

Fig. 4. Total radioactivity of RI produced annually at Dalat Nuclear Research Institute for medicine.

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Other applications of radioisotopes channel. An auto-pneumatic transfer system produced at the DNRR are radiotracer installed in 2012 at the DNRR can transfer a technique in sediment studies, oil exploitation, sample from irradiation position to measuring chemical industry, biology, agriculture and detector about 3 seconds. hydrology. Some main products are 46Sc, 192Ir, The k-zero method for INAA has been 198Au, 131I, 140La, etc. In addition, some small also developed to analyse airborne particulate sources of 192Ir and 60Co with low radioactivity samples for investigation of air pollution; crude have also been produced for industry oil samples and base rock samples for oil field applications. study. Based on developed k-zero-INAA B. Neutron activation analysis method, a multi-elements analysis procedures have been applied to simultaneously determine Research on analytical techniques based concentration for about 31 elements including on neutron activation and other related Al, As, Ba, Br, Ca, Cl, Cr, Cu, Dy, Eu, Fe, Ga, processes consists of the elaboration of Hf, Ho, K, La, Lu, Mg, Mn, Na, Sb, Sc, Sm, analytical processes and the design and Sr, Th, Ti, V, Yb, Zn. construction of analytical instruments. C. Neutron beam utilization Requests of many branches of the national economy for various types of samples The reactor has four horizontal beam have quickly been responded. NAA at the ports, which provide beams of neutron and DNRR has always been met the demand of gamma radiation for a variety of experiments. analytical services for geology exploration, oil They also provide irradiation facilities for prospecting, agriculture, biology, large specimens in a region close to the environmental studies, etc. reactor core. Besides, the reactor also has a large thermal column with outside dimensions The relatively high neutron flux in of 1.2m by 1.2m in cross section and 1.6m in irradiation channels of the reactor allows length (Fig. 5). elemental analysis using various neutron activation approaches, such as Instrumental Up to now, only three beam ports (No.2, NAA (INAA), Radiochemical NAA (RNAA), No.3 and No.4) and the thermal column have Delayed NAA (DNAA) and Prompt gamma been used for reseaches and applications. At NAA (PGNAA). By the end of 2013, a total the beam port No.2, a BGO-HPGe gamma-rays of about 60,000 samples have been irradiated Compton suppression spectrometer has been at the reactor with a yearly average of 2000 recently installed for PGNAA and samples. It can be estimated that those make experimental researches on neutron capture up 60% of geological samples, 10% of reactions. The filtered thermal neutron beams biological samples, 20% of environmental extracted from the tangential beam port No.3 samples, 5% of soil and agriculture materials, are used for nuclear structure studies, 3% of industrial materials. especially for experimental determination of nuclear energy levels and level density in In order to determine the elements regions below neutron binding energy. The having short-lived radionuclides, the method of filtered neutron beams at the piercing beam cyclic INAA with the alternation of irradiation port No.4 with quasi-monoenergies of 24keV, and measurement was implemented by using 54keV, 59keV, 133keV and 148keV are used the thermal column and vertical irradiation

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RESULTS OF OPERATION AND UTILIZATION OF THE DALAT NUCLEAR RESEARCH REACTOR for the mesurements of neutron total and study on radiation shielding design. Typical capture cross sections. In addition, these research activities using neutron beam of the neutron beams are also applied for practical DNRR are listed below.

Thermal column No. 2: Gamma spectrometry No. 3: Nuclear system with BGO detector for Colum n structure study PGNAA and neutron capture door reactions study Beam port # 2 Beam port # 3

Therm al Colum n

Stainless steel

Alum inum

Graphite reflector Core Pool tank wall

Beam port # 1 Bellow s Beam port # 4 assem bly No. 4: Nuclear data This port is closed Thermalizing column (closed) Concrete measurement shielding

Spent fuel storage tank

Fig. 5. Horizontal section view of the DNRR.

Neutron physics and nuclear data measurement - Measurement of isomeric ratio created 81 82 In the keV energy region, filtered in the reaction Br(n, ) Br on the 55 keV and 144 keV neutron beams; neutron beams are the most intense sources, which can be used to obtain neutron data for - And other investigations, such as reactors and other applications. The following average resonance capture measurements, experiments have been carried out at the using the - coincidence spectrometer for DNRR including: study on the (n, 2) reaction, etc.

- Total neutron cross section measurement Application of neutron capture gamma ray 238 spectroscopy for U, Fe, Al, Pb on filtered neutron beams at 144 keV, 55 keV, 25 keV and evaluation of - Development of PGNAA technique using average neutron resonance parameters from the filtered thermal neutron beam in experimental data; combination with the Compton-suppressed spectrometer for analyzing Fe, Co, Ni, C in - Gamma ray spectra measurement from steel samples; Si, Ca, Fe, Al in cement neutron capture reaction of some reactor samples; Gd, Sm, Nd in uranium ores, Sm, Gd materials (Al, Fe, Be, etc.) on filtered neutron in rare earth ores; etc.; beam at 55 keV and 144 keV; - Utilization of the PGNAA method for - Measurement of average neutron investigating the correlation between boron radioactive capture cross section of 238U, 98Mo, and tin concentrations in geological samples as 151Eu, 153Eu on the 55 keV and 144 keV a geochemical indication in exploration and neutron beams; assessment of natural mineral resources; analyzing boron in sediment and sand samples

6 NGUYEN NHI DIEN et al. to complement reference data for such samples Besides, the DNRR has been used as a from rivers; main tool for practical training, a set of equipment was supported under IAEA TC - Development of the spectrometer of project, bilateral projects with the Japan summation of amplitudes of coinciding pulses Atomic Energy Agency and Bhabha Atomic for (n, 2) reaction research and for measuring Research Center of India. The measuring activity of activated elements with high systems for practices at the Training Center possibility of cascade transitions. can meet the fast increasing demand and is D. Eduacation and training activities expected to move forward to the regional Training Center at Dalat Nuclear standard in the field of nuclear training. Research Institute which was established in E. Other applications 1999 is responsible for organizing training Research on sediment using radiotracer courses and training in reactor engineering, techniques was carried out to investigate bed nuclear and radiation safety, application of load layers displacement at estuaries nuclear techniques and radioisotopes in navigation channel region and to explain the industry, agriculture, biology and environment, sediment deposition phenomenon causing etc. Training courses on non-destructive frequent dredging activities. evaluation (NDE) including radiographic testing, ultrasonic testing as well as on security Research on radio-biology consists of of nuclear facilities and radiation sources have using gamma radiation associated with other also been done. The center also is the training factors for improving agricultural seeds and facility for expertise students from local applying radioactive tracers for studying universities and foreign postgraduate students. biological metabolism, especially nutrition Thereby, the human resource development is problems. These studies are to investigate conducted annually so that it can deal with phosphorus absorption and other nutritional scientific works of higher and higher quality problems during the growing processes of rice and meet a huge demand in the field of nuclear and other plants. Irradiation effects on some science and technology in Vietnam in the plants to gain higher yield or environment future. Thanks to the bilateral co-operation adapted varieties were also studied. with the Japan Atomic Energy Agency, US Gemstone colorizing experiments of Department of Energy, Bhabha Atomic topaz and sapphire in the reactor core, in the Research Center of India, and Korea Atomic rotary rack as well as in horizontal channels Energy Research Institute, we have conducted has been done. a variety of training courses in the four following key areas: As research purpose, silicon mono- crystals have been irradiated at the central - Reactor engineering for nuclear power neutron trap of the reactor. Irradiated products programme; of good quality, appropriate for fabrication of - Research and development activities; power diodes and thyristors have been created - State management in the field; thanks to proper neutron distribution in this - And University lecturer training program. irradiation facility and suitable cadmium ratio.

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RESULTS OF OPERATION AND UTILIZATION OF THE DALAT NUCLEAR RESEARCH REACTOR

IV. SOME MAIN REACTOR RENOVATION B. Reactor control and instrumentation system PROJECTS PERFORMED modification A. Reactor conversion from HEU to LEU fuels The Control and Instrumentation System In the framework of the program on (C&I) of the DNRR was designed and Russian Research Reactor Fuel Return manufactured by the former Soviet Union and (RRRFR) and the program on Reduced put into operation in November 1983. Due to Enrichment for Research and Test Reactor the spare part procurement problem was (RERTR), the DNRR core was partly suspected and using technology of the 1970’s converted from HEU to LEU in September with discrete and low-level integrated 2007. electronic components, the system technology was somewhat obsolete and un-adapted to After this success, the full core tropical climate. conversion study from HEU to LEU of the DNRR was also carried out during years 2008 The first renovation work was - 2010. The results of neutronics, thermal implemented during 1992-1993 period and the hydraulics and safety analysis showed that a renovated C&I system was commissioned at LEU core loaded with 92 fuel assemblies and the end of 1993. The most important 12 beryllium rods around the neutron trap renovation task was to redesign and construct a satisfies the safety requirements while number of electronic systems/blocks, which maintaining the utilization possibility similar play the key role in enhancing the reliability of to that of the previous HEU and recent mixed the system. This renovation work was focused fuel cores. mainly on the process and instrumentation Physics and energy start-up of the system, but not on the neutron measurement DNRR for full core conversion to low and data processing parts. Because of that, it enriched uranium (LEU) fuel were performed was necessary to fulfill the second renovation th th from November 24 , 2011 until January 13 , and modernization during the years of 2005- 2012 according to a planned program that 2007 to replace neutron measurement and was approved by Vietnam Atomic Energy signal processing parts of the existing C&I Institute (VINATOM). At 15:35 on system by the digital system named ASUZ- th November, 30 , 2011 the reactor reached 14R. The main items replaced under the criticality with core configuration including second modification are neutron detector 72 LEU FAs and neutron trap in center. Then channels; neutron flux control system the fuel loading for working core and power (NFCS), reactor protection system, control ascension test were also carried out from console and control panels, reactor protocol December, 6th, 2011 to January, 13th, 2012. and diagnostic system, etc. Experimental results of physical and thermal hydraulics parameters of the reactor during The commissioning of the new I&C start up stages and long operation cycles at system was finished in August 2007 and nominal power showed very good agreement operating license was approved in October with calculated results and met the safety 2007. requirements.

8 NGUYEN NHI DIEN et al.

V. CONCLUSIONS REFERENCES

The DNRR has been safely operated and [1] Nguyen Nhi Dien, Dalat Nuclear Research effectively utilized for 30 years. To achieve Reactor - Twenty five years of safe operation that, maintaining and upgrading the reactor and efficient exploitation, Dalat, (March 2009). technological facilities have been done with a [2] Duong Van Dong, Status of Radioisotope high quality. The reactor physics and thermal Production and Application in Vietnam, Dalat hydraulics studies have also provided the Sym. RR-PI-09, Dalat, (2009). important bases for safety evaluation and in- [3] V. V. Le, T. N. Huynh, B. V. Luong, V. L. core fuel management to ensure its safe Pham, J. R. Liaw, J. Matos, Comparative operation and effective exploitation. The Analyses for Loading LEU Instead of HEU safety and security for the reactor are one of Fuel Assemblies in the DNRR, RERTR Int’l the main issues that national and local Meeting, Boston, (2005). authorities are particularly interested in and [4] P.V. Lam, N.N. Dien, T.D. Hai, L.B. Vien, strongly support up. L.V. Vinh, H.T. Nghiem, N.M. Tuan and N.K. During 30 years of operation, the DNRR Cuong, Results of the reactor control system has been playing an important role in the use replacement and reactor core conversion at the th of atomic energy for peaceful purpose in Dalat nuclear research reactor, The 12 Annual Topical Meeting on Research Reactor Fuel Vietnam. The reactor has been used for Management, Hamburg, Germany, (2008). radioisotope production for medicine and industry purposes, NAA of geological, crude [5] P.V. Lam, N.N. Dien, L.V. Vinh, H.T. oil and environment samples, performance of Nghiem, L.B. Vien and N.K. Cuong, Neutronics and thermal hydraulics calculation fundamental and applied researches on for full core conversion from HEU to LEU fuel nuclear and reactor physics, as well as of the Dalat nuclear research reactor, RERTR creation of a large amount of human resource Int’l Meeting, Lisbon, Portugal, (2010). with high skills and experiences on [6] L.B. Vien, L.V. Vinh, H.T. Nghiem and N.K. application of nuclear techniques in the Cuong, Transient/ accident analyses for full country. A strategic plan and long-term core conversion from HEU to LEU fuel of the working plan for the DNRR has been set up in Dalat nuclear research reactor, RERTR Int’l order to continue its safe operation and Meeting, Lisbon, Portugal, (2010). effective utilization at least to 2025. [7] C.D. Vu, Study on application of k0-IAEA at It should be mentioned that a project for Dalat research reactor, Project report (code establishment of a new nuclear science and CS/09/01-01), (2010). technology center with a high power research [8] N.N. Dien, L.B. Vien, P.V. Lam, L.V. Vinh, reactor expected to put into operation between H.T. Nghiem, N.K. Cuong, N.M. Tuan, N.M. 2020-2022 is now under preparation and Hung, P.Q. Huy, T. Q. Duong, V.D.H. Dang, consideration. Therefore, the DNRR will be T.C. Su, T.T. Vien, Some main results of necessary and keep playing an important role commissioning of the Dalat Nuclear Research in scientific research, applications and human Reactor with low enriched fuel, Nuclear resource development for Vietnam in the Research Institute, (2012). coming time. [9] Safety Analysis Report (SAR) for the Dalat Research Reactor, Dalat, (2012).

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Nuclear Science and Technology, Vol. 4, No. 1 (2014), pp. 10-25 Design Analyses for Full Core Conversion of The Dalat Nuclear Research Reactor

Luong Ba Vien, Le Vinh Vinh, Huynh Ton Nghiem, Nguyen Kien Cuong Reactor Center – Nuclear Research Institute – Vietnam Atomic Energy Institute 01 Nguyen Tu Luc, Dalat, Lamdong Email: [email protected] (Received 5 March 2014, accepted 10 March 2014)

Abstract: The paper presents calculated results of neutronics, steady state thermal hydraulics and transient/accidents analyses for full core conversion from High Enriched Uranium (HEU) to Low Enriched Uranium (LEU) of the Dalat Nuclear Research Reactor (DNRR). In this work, the characteristics of working core using 92 LEU fuel assemblies and 12 beryllium rods were investigated by using many computer codes including MCNP, REBUS, VARI3D for neutronics, PLTEMP3.8 for steady state thermal hydraulics, RELAP/MOD3.2 for transient analyses and ORIGEN, MACCS2 for maximum hypothetical accident (MHA). Moreover, in neutronics calculation, neutron flux, power distribution, peaking factor, burn up distribution, feedback reactivity coefficients and kinetics parameters of the working core were calculated. In addition, cladding temperature, coolant temperature and ONB margin were estimated in steady state thermal hydraulics investigation. The working core was also analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and MHA. Obtained results show that DNRR loaded with LEU fuel has all safety features as HEU and mixed HEU-LEU fuel cores and meets requirements in utilization as well. Keywords: HEU, LEU, neutronics, thermal hydraulics, safety analyses

I. INTRODUCTION kinetics parameters. Because the higher content of 235U in a LEU FA compared to HEU FA, it In this full core conversion study, is needed to rearrange the fuel assemblies and neutronics, thermal hydraulics and safety berrylium rods with the different way to the analysis were carried out to investigate first HEU core to meet the safety requirements. characteristics of LEU working core fully loaded with LEU fuel. All computer codes Thermal hydraulics parameters at steady were validated with HEU and mixed cores. state condition were obtained by using PLTEMP3.8 code [11] introduced models and Using MCNP [6], REBUS-PC [5] and correlations that suitable for the concentric VARI3D computer codes, a series of static tube fuel type and natural convection regime of reactor physics calculation were performed to the DNRR. obtain neutronics parameters of the working core (see Fig. 1). Some parameters included in Based on the neutronics analysis the design of working core with shutdown parameters of the LEU core, the postulated margin, excess reactivity taking into account of transients and accidents selected for the DNRR irradiated Beryllium poisoning, control rod are analyzed. The RELAP5/MOD3.2 code [15] worths, detailed power peaking factors, was used for analysis of RIA (Reactivity neutron performance at the irradiation Initiated Accident), LOFA (Loss Of Flow positions, reactivity feedback coefficients, and Accident) transients.

©2014 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute

LUONG BA VIEN, LE VINH VINH, HUYNH TON NGHIEM, NGUYEN KIEN CUONG These study results showed that a LEU REBUS-MCNP Linkage [7] was used core loaded with 92 fuel assemblies and 12 to calculate burnup distribution using “two beryllium rods around the neutron trap satisfies way” linking option in which MCNP is used the safety requirements while maintaining the for calculating neutron flux and cross section utilization possibility similar to that of the in one group neutron energy and burn up previous HEU and recent mixed fuel cores. calculation is implemented by REBUS-PC. The MCNP5 code using an ENDF-B/VI cross section library was used to construct a detailed geometrical model of each reactor component and calculate control rod worths, multiplication coefficient, power distribution, neutron flux performance in irradiation positions, reactivity feedback coefficients, and kinetics parameters (prompt neutron life time

and delayed neutron fraction).

A detailed geometrical model of reactor components including all fuel assemblies, Fig. 1. The new designed working core loaded with control rods, irradiation positions, beryllium 92 LEU FA and 12 Beryllium rods. and graphite reflectors, horizontal beam tubes and thermal column was made in the MCNP II. CALCULATION MODELS AND model, except in the axial reflectors above and COMPUTER CODES below the fuel assembly where some materials were homogenized. Fig. 2a provides the radial A. Neutronics and Thermal Hydraulics and axial models of the reactor for Monte Calculation Carlo Calculations. The diffusion code REBUS-PC with The kinetics parameters were calculated finite difference flux solution method was used also by VARI3D code. The real and adjoint to perform core calculation for reactor physics fluxes which are required to compute these characteristics and operation cycle calculations parameters were provided by DIF3D-a main with micro neutron cross sections according to module of REBUS-PC code. 7 energy groups (collapsed from 69 energy In diffusion theory, the reactor was groups) that were generated by WIMS-ANL modeled in hexagonal geometry with a code [4]. The FA cross sections were heterogeneous representation of the fuelled and generated in a radial geometry with each fuel non-fueled portions (see Fig. 2a). Each element depleted based upon its unique neutron homogenized fuel assembly was modelled spectrum in the WIMS-ANL model. The using five equal volume axial depletion zones. REBUS-PC fuel depletion chains included The beam tubes were modeled using a production of six Pu isotopes, Am-241, Np- homogenized mixture of air or concrete, 237, and lumped fission product. Isotopic graphite and aluminum. precursors of Xe-135 and Sm-149 were also included in the depletion chains so that Xe and The reactor models for diffusion and Monte-Carlos computer codes were validated Sm transients during periods of shutdown and by comparing with good agreement not only to startup could be modelled.

11 DESIGN ANALYSES FOR FULL CORE CONVERSION OF … the fresh HEU configuration cores but also to for PLTEMP code. A fuel assembly was the HEU burnt cores. These models were then modelled as three concentric cylindrical tubes. applied for partial core conversion analyses of Before using PLTEMP code to DNRR [3]. The measured data collected during calculate for DNRR with fully LEU fuel the deployment of partial core conversion assemblies, the code was validated by project showed that the predicted calculation comparing analytical results with results are quite acceptable [8,9]. experimental results of mixed-core. The PLTEMP/ANL3.8 [15] thermal- B. Transient/Accidents analyses hydraulics code for plate and concentric-tube geometries with capability of calculating The DNRR has three barriers as other natural circulation flow was used for thermal- research reactors that prevent or limit the hydraulics analyses. A chimney model as well transport of fission products to the as Collier heat transfer correlation and CHF environment, which are fuels and cladding, Shah’s correlation have been recently reactor pool water and reactor confinement. implemented make the code suitable DNRR The safety system settings are showed in thermal-hydraulics calculation. Table I.

Fig. 2b shows the model of WWR-M2 fuel assembly, core and chimney of the DNRR Table I. Safety system settings.

Parameters Safety system settings Maximum thermal power (Pmax) 550 kW (110% FP) Minimum reactor period (Tmin) 20s Deficient level of pool water 60 cm Primary coolant flow rate 40 m3/h Secondary coolant flow rate 70 m3/h

In the Safety Analysis Report (SAR) for the DNRR [1], the possible initiating events were classified by groups. The initiating events in each group are then analyzed and justified in order to identify the limiting event that will be selected for further detail quantitative analysis. The limiting event in each group has potential consequences that exceed all others in that group. Limiting events were selected for detailed analyzed are as follows: (1) Uncontrolled withdrawal of a control rod; (2) Primary/Secondary Pump Failure; (3) Earthquake; (4) Fuel cladding failure. A summary of the core parameters used for the Fig. 2a. Radial and Axial models for Monte Carlo calculations (upper) and Radial model safety analysis is given in Table II. for Diffusion Theory calculations (under).

12 LUONG BA VIEN, LE VINH VINH, HUYNH TON NGHIEM, NGUYEN KIEN CUONG To ensure the fuel clad integrity in operational condition and to protect the public and the environment in case of accident, in the SAR for the DNRR, the following acceptance criteria were defined: 1) - For anticipated operational occurrences: 3) (1) Minimum margin to departure from nucleate boiling (DNB) shall be over 1.5; (2) Maximum temperature of fuel cladding 2) shall not exceed 400oC; Fig. 2b. DNRR model for PLTEMP (3) Fuel cladding integrity shall be (1-fuel assembly cross-section; 2-FA model for PLTEMP; 3-reactor coolant system model). assured. - For accident conditions: core corresponding to a cooling channel with (1) Core covering shall be maintained; maximum heat flux. The average channel (2) Core shall not be remarkably represents the rest of the cooling channels. damaged; Each channel was modelled as three fuel (3) Release of fission products into the element plates and four coolant flow gaps. The environment shall not be remarkable. nodding diagram of the DNRR for RELAP5/3.2 is presented in Fig. 2c. The RELAP5 code was used for analyzing the events of excess reactivity The MACCS2 code [19] was used to insertion by uncontrolled withdrawal of a estimate the radiological impact of the control rod and earthquake. The piping of the hypothetical accident on the surrounding primary cooling system and pool volume were public. The core radiation inventories were divided into nodes with similar dynamic calculated by ORIGEN2 code [20] using characteristics. The reactor core was divided neutron cross-sections of the actinides obtained into 2 channels with axial nodes. The hot from MCNP5 code. channel represents the hottest channel in the

Fig. 2c. Nodding diagram of DNRR for RELAP5/3.2.

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DESIGN ANALYSES FOR FULL CORE CONVERSION OF …

Table II. Core parameters used for safety analysis.

Parameters Values Power, kW 500 Coolant inlet temperature, oC 32 Peaking factor (shim rods at 300 mm) - Axial peaking factor 1.363 - Radial peaking factor 1.376 - Local peaking factor 1.411 Reactor kinetics - Prompt neutron life, s 8.92510-5 - Delayed neutron fraction (1$) 7.55110-3 Temperature reactivity coefficients - Moderator, %/K; (293-400oK) - 1.26410-2 - Fuel, %/oC; (293-400oK) - 1.8610-3 (400-500oK) - 1.9210-3 (500-600oK) - 1.5610-3 - Void, %/% of void (0-5%) -0.2432 (5-10%) -0.2731 (10-20%) -0.3097 Reactivity control - Shutdown worth, % (2 safety rods) 3.7 - Maximum withdrawal speed of one shim 3.4 rod, mm/s and of the regulating rod, mm/s 20 Reactor protection characteristics - Response time to overpower scram, s 0.16 - Response time to fast period scram, s Start-up range 9.1 Working range 6.7 - Drop time of control rods, s 0.67

III. RESULTS AND DISCUSSIONS results in large negative reactivities which alter flux and power distributions. A. Neutronics and Thermal Hydraulics Program Beryl [10] has been modified 3He and 6Li Poisoning of Irradiated to calculate the 3He, 6Li and 3H concentrations. Beryllium [10] The MCNP5 was then used to determine the Since 1984, the DNRR has been put into poisoning effect of 3He, 6Li and 3H operation with a considerable amount of concentrations on reactor core reactivity. The Beryllium used for neutron trap at the core comparison of reactivities between calculation center and periphery for improving neutron results and measured data of some beryllium reflection around. Because Beryllium has large blocks irradiated in DNRR (Table III) shows thermal neutron absorption cross sections, the that the negative reactivity of irradiated buildup of 3He, 6Li and 3H concentrations beryllium determined by above-mentioned

14 LUONG BA VIEN, LE VINH VINH, HUYNH TON NGHIEM, NGUYEN KIEN CUONG method is reliable. Six beryllium rods were position is estimated about 0.4 cent. used for measurement, two fresh beryllium Following calculation scheme for beryllium rods and four irradiated beryllium rods (two poisoning above, reactivity of the poisoning beryllium rods at the end 1994 and two at the process in new configuration cores about -1$. end 2002). 9-6 and 5-6 positions were chosen All calculation for design LEU cores, to measure reactivity of couple beryllium rods beryllium poisoning is included in the model through changing position of control rod for MCNP code. (Regulating Rod). The error of control rod

Table III. Comparison of calculated and measured of reactivities of irradiated beryllium rods in DNRR.

Measured reactivity Calculated reactivity Error (Cent) (Cent) (%)

2 Beryllium Rods at -3.89  0.4 -4.65  0.0038 16.34 the end 1994

2 Beryllium Rods at -6.28  0.4 -7.19  0.0039 12.66 the end 2002

The working core characteristics the Table IV. The shutdown margins of the core is met the safety requirement of -1.0%. From the calculation results of shutdown Calculated neutron flux at the neutron trap of margins, excess reactivities, power peaking the core is nearly the same as that of mixed factors, and neutron performance at the core (92HEU+12LEU). Table V shows the irradiation positions of 4 candidates cores, the control rod worths. Detailed neutron flux working core with the better features from the performance at the main irradiation positions safety and utilization point of view was chosen are presented in Table VI. for detailed analysis. The main calculated characteristics of working core is showed in

Table IV. Calculation results of working core compared with current mixed core.

Current Mixed Parameters LEU Core Core Excess Reactivity (%) – Fresh 6.63 Excess Reactivity (%) – After 600FPDs 3.79 Shutdown Margin (%) – Fresh -2.92 -4.56 Shutdown Margin (%) – After 600 FPDs -6.62 Radial Power Peaking Factor Control Rods Out 1.398 1.431 Control Rods In 1.434 Thermal Neutron Flux at Neutron Trap Center (n/cm2) Control Rods Out 2.22E+13 2.22E+13 Control Rods In 2.14E+13 Fast Neutron Flux at Neutron Trap Center (n/cm2) Control Rods Out 1.95E+12 3.15E+12 Control Rods In 1.92E+12

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DESIGN ANALYSES FOR FULL CORE CONVERSION OF …

Table V. Control Rods worths (%k/k).

Core1 MCNP Core1 MCNP Control Rods Fresh error Burnt error Shim rod 1 2.5896 0.000091 2.3539 0.000091 Shim rod 2 2.6100 0.000111 2.4033 0.000124 Shim rod 3 2.7784 0.000118 2.5381 0.000122 Shim rod 4 2.4687 0.000122 2.2604 0.000117 Regulating rod 0.4363 0.000126 0.3629 0.000119 Safety rod 1 2.1955 0.000106 2.3084 0.000115 Safety rod 2 2.2356 0.000119 2.3579 0.000105

Table VI. Neutron flux performance.

Thermal, <0.625eV Epithermal, Fast, <10MeV (n/cm2.s) <0.821MeV (n/cm2.s) Irradiation positions 2 (n/cm .s) Fresh Burnt Fresh Burnt Fresh Burnt Neutron Maximum 2.07E+13 2.20E+13 6.79E+12 7.12E+12 1.83E+12 1.92E+12 Trap Average 1.45E+13 1.49E+13 6.00E+12 6.04E+12 1.62E+12 1.63E+12 Channel Maximum 9.45E+12 9.86E+12 8.19E+12 8.42E+12 2.98E+12 3.02E+12 13-2 Average 7.00E+12 7.12E+12 6.53E+12 6.51E+12 2.46E+12 2.44E+12 Channel Maximum 5.41E+12 5.66E+12 9.63E+12 9.76E+12 4.22E+12 4.26E+12 7-1 Average 4.11E+12 4.18E+12 7.23E+12 7.15E+12 3.19E+12 3.15E+12 Channel Maximum 9.24E+12 9.71E+12 8.02E+12 8.22E+12 2.92E+12 2.99E+12 1-4 Average 6.85E+12 7.01E+12 6.41E+12 6.40E+12 2.42E+12 2.40E+12 Rotary Average 3.55E+12 3.56E+12 7.58E+11 7.56E+11 1.93E+11 1.93E+11 Specimen

Power Distribution and Power Peaking of control rods at 250 mm. Detailed axial Factors power distribution according to control rod Power peaking factors of the core with position was also calculated. Radial power different position of control rods were distributions at different control rod position calculated and presented in Table VII. The are showed in Fig. 3. maximum power peaking factor is in position Table VII. Power peaking factor according to control rod positions

Position Peaking Factor (mm) F.A. Radial Core Radial Axial Total 0 1.378 1.398 1.296 2.498 150 1.378 1.399 1.343 2.589 200 1.375 1.403 1.356 2.615 250 1.377 1.409 1.365 2.648 300 1.376 1.411 1.363 2.646 350 1.378 1.415 1.336 2.605 600 1.378 1.434 1.284 2.537

16 LUONG BA VIEN, LE VINH VINH, HUYNH TON NGHIEM, NGUYEN KIEN CUONG

0.973 0.980 1.165 1.164 0.947 0.957 1.126 1.129

0.913 0.849 0.858 0.979 1.107 1.006 0.885 0.963 0.901 0.851 0.858 0.964 1.085 1.001 0.895 0.958

0.877 0.762 0.825 1.004 SR 1.021 0.835 0.775 0.908 0.875 0.817 0.876 1.008 1.023 0.889 0.833 0.913

0.917 0.790 ShR 1.090 1.410 1.406 1.082 ShR 0.787 0.921 0.906 0.841 1.126 1.370 1.368 1.124 0.843 0.915

0.962 0.872 0.906 1.281 1.259 0.882 0.842 0.919 0.939 0.873 0.953 1.296 1.283 0.937 0.850 0.903

1.138 0.998 0.983 1.353 1.313 0.930 0.865 0.959 1.090 0.974 0.975 1.321 1.289 0.929 0.855 0.933

1.056 1.220 1.139 0.911 RgR 1.027 1.192 1.122 0.900

1.167 1.018 0.996 1.364 1.307 0.918 0.843 0.911 1.117 0.991 0.987 1.332 1.284 0.919 0.837 0.887

1.020 0.903 0.929 1.296 1.252 0.868 0.808 0.845 0.992 0.904 0.975 1.312 1.273 0.919 0.818 0.835

0.996 0.830 ShR 1.106 1.421 1.408 1.079 ShR 0.755 0.841 0.977 0.881 1.145 1.381 1.368 1.114 0.808 0.836

0.968 0.810 0.860 1.038 SR 1.005 0.816 0.745 0.831 0.959 0.863 0.910 1.038 1.006 0.866 0.803 0.836

1.016 0.918 1.031 1.124 0.986 0.858 0.843 0.895

0.994 0.917 1.019 1.097 0.970 0.857 0.846 0.889 1.198 1.180 0.985 0.973 1.137 0.960 1.156 0.949

Fig. 3. Radial power distribution (Upper values: Fresh Core; Under values: Burnt Core)

Reactivity Feedback Coefficients and kinetics parameters of the LEU cores Kinetics Parameters calculated using the VARI3D and MCNP5 Reactivity feedback coefficients codes. The calculated results from the two calculated with the MCNP5 are depicted in computer codes are in good agreement. Table VIII. The negative results of reactivity These data will be used in transient feedback coefficients show the inherent calculation for safety analysis of fully LEU safety of the LEU core. Table IX shows the core of DNRR.

Table VIII. Feedback reactivity coefficients.

Parameter DATA ±σ Moderator Temperature Reactivity Coefficient (%/oC) 293 oK to 400 oK -0.01317 0.00005 Fuel Temperature (Doppler) Reactivity Coefficient (%/oC) 293 oK to 400 oK -0.00192 0.00005 400 oK to 500 oK -0.00182 0.00003 500 oK to 600 oK -0.00154 0.00002 Moderator Density (Void) Reactivity Coefficient (%/% of void) 0 to 5 % -0.2514 0.0011 5% to 10 % -0.2784 0.0012 10 % to 20 % -0.3255 0.0006

17 DESIGN ANALYSES FOR FULL CORE CONVERSION OF …

Table IX. Calculated results of kinetics parameters for LEU core.

Decay Const. Relative Yield Fraction Family, i -1 λi (s ) ai βi

1 1.334E-02 3.507E-02 2.648E-04 2 3.273E-02 1.804E-01 1.363E-03 3 1.208E-01 1.742E-01 1.315E-03 4 3.030E-01 3.843E-01 2.902E-03 5 8.503E-01 1.594E-01 1.204E-03 6 2.856E+00 6.666E-02 5.033E-04 Total delayed neutron fraction, β VARI3D 7.551E-03 MCNP5 – Fresh 7.761E-03 MCNP5 – Burnt 7.762E-03 Prompt neutron life time, ℓ 8.925E-05

Burn up calculation extended about 11 years (calculated with 1300 hours per year) or 600 full power days (FPDs). The first cycle length was estimated by The burn up of U-235 reached average value of REBUS-MCNP Linkage system code. Burn up 8.2% and maximum value of 11.4%. In the calculations were performed by assuming that next cycle, about 8 fuel assemblies will be shim rods and regulating rod were in critical inserted so the reactor core will operate with position following each burn-up step. The 100 fuel assemblies. The Fig. 4 shows burn up value of reactivity for Xe-135 poisoning was distribution after 600 FPD operation. estimated about 1.2% k/k. The result of depletion shows that operating time may be

Fig. 4. Burn up distribution using REBUS-MCNP Linkage system after 600 FPD.

18 LUONG BA VIEN, LE VINH VINH, HUYNH TON NGHIEM, NGUYEN KIEN CUONG working core meets the requirements of The PLTEMP code was used for thermal hydraulics safety. At the power of calculating cladding temperature, coolant 500kW with systematic errors, maximum temperature and safety margins for the cladding temperatures are below the candidate cores. The calculated results are permissible value of 103oC [2] and far below presented in Table X and Fig. 5. At nominal the ONB temperature (estimated about 116oC power without uncertainties and maximum using Forster-Greif correlation). The maximum permissible inlet temperature (32oC), the outlet coolant temperature is calculated about maximum cladding temperature is 90.50oC. 60oC, much lower than saturated temperature Calculation was carried out for nominal power (108oC). with systematic errors (equivalent to 70kW power) and the maximum cladding temperature Fig. 6 shows the comparison of cladding is 95.69oC. In this case, by using Shah’s temperature of 92FA LEU cores and 89FA correlation, the obtained minimum DNBR is fresh HEU core. Compared to the 89FA fresh 9.9. The minimum flow instability power ratio HEU core established in 1984, cladding (MFIPR) is 2.04. From above-mentioned temperature of working core is about 2oC calculated results, it may conclude that the lower.

Table 10. Cladding temperature and ONB margin by PLTEMP Code.

500kW 550kW 600kW Distance without sys. error with sys. error with sys. error with sys. error (cm) T- T- T- T- Tc(oC) ONB(oC) Tc(oC) ONB(oC) Tc(oC) ONB(oC) Tc(oC) ONB(oC) 2.5 63.91 51.89 66.89 49.24 68.95 47.39 70.96 45.59 7.5 70.56 45.59 74.13 42.36 76.58 40.14 78.97 37.97 12.5 78.46 38.07 82.71 34.18 85.63 31.51 88.46 28.91 17.5 84.83 31.90 89.61 27.50 92.89 24.48 96.05 21.57 22.5 88.77 27.95 93.85 23.26 97.33 20.06 100.68 16.95 27.5 90.50 26.05 95.69 21.25 99.23 17.97 102.65 14.80 32.5 89.86 26.34 94.95 21.63 98.43 18.41 100.76 16.40 37.5 87.10 28.58 91.91 24.13 94.41 21.94 96.22 20.48 42.5 83.98 31.14 88.24 27.24 89.94 25.90 91.57 24.60 47.5 79.67 34.76 82.92 31.91 84.43 30.74 85.89 29.59 52.5 74.91 38.73 77.42 36.64 78.79 35.57 80.13 34.52 57.5 71.21 41.70 73.32 40.02 74.64 38.98 75.94 37.93

100.0

C) 100 o C) 92 LEU FA Core2 90.0 o 95 80.0 89 HEU FA Core T-clad 70.0 90

Temperature ( Temperature 92 LEU FA Core1 Temperature ( Temperature 60.0 85 50.0 80 40.0 T-coolant 75 30.0 70 20.0 DT-ONB 10.0 65 0.0 60 0 10 20 30 40 50 60 0 10 20 30 40 50 60 Distance from core bottom (cm) Distance from core bottom (cm) Fig . 6. Comparison of calculated cladding temperature Fig. 5. T/H parameters at 500kW without between 92FA LEU cores and HEU core. uncertainties.

19 DESIGN ANALYSES FOR FULL CORE CONVERSION OF …

The event of one shim rod inadvertently 2. Transient/Accidents analyses withdrawal with speed of 3.4 mm/s from stable Uncontrolled withdrawal of one shim operation of 100%FP (500 kW) are showed in rod or the regulating rod Fig. 7 and Table XI. In this case, the reactor In this event, it is assumed that one of power increases and reaches to the over-power the shim rods or the regulating rod is setting value (110%FP) within 3.39 seconds withdrawn in the most effective part from 200 generating a scram signal. After a delay time of mm to 400 mm at the speed for 3.4 mm/s of 0.16 seconds the reactor power is rapidly shim rod and 20 mm/s for regulating rod. The suppressed because of the control rods initial conditions are as follows: insertion. The peak power of the reactor is only attained 0.553 MW with a slight increase of the a) Start-up case: maximum fuel cladding temperature. With the (1) -1% k/k sub-critical; Power level: 10- assumption of no overpower scram signal 5%FP; Coolant inlet temperature: 32oC. appeared, a fast period scram signal is generated after 8.33 seconds from the initiation (2) Critical state; Power level: 10-3%FP; of transient event. The reactor will be Coolant inlet temperature: 32oC. shutdown after 6.7 second delay with a peak b) Steady-state operation: power of 0.957 MW. The maximum fuel cladding temperature is predicted to be 113.0oC Power level: 100%FP; Coolant inlet without any nucleate boiling occurrences. The temperature: 32oC. minimum DNBR (Departure from Nucleate In sub-critical status, when one shim rod Boiling Ratio) estimated about 6.5 is much is inadvertently withdrawn with the speed of higher than the acceptance criterion of 1.5. 3.4 mm/s, from the core, the reactor power With the same initial conditions, the only increases to the maximum value of calculated results for the event of withdrawal -7 2.7810 MW while the fuel cladding of the regulating rod are slightly different from temperature is unchanged. With initial those of above-mentioned event, when one conditions of criticality at the power level of shim rod is withdrawn. This can be explained -3 -6 10 %FP (510 MW) if there is no fast period by the similar insertion rate of reactivity in the signal and the overpower trip setting is two cases (about 0.02$/s). The regulating rod 110%FP, the fuel clad temperature reaches to has lower reactivity worth but higher 97.8oC, but still far below ONB (Onset of a withdrawal velocity compared to those of a Nucleate Boiling) temperature. shim rod.

Table XI. Transient results of one shim rod withdrawal from 100%FP.

Values Parameters 110%FP Scram Period Scram Time to Peak Power, s 3.6 15.1 Peak Power, MW 0.553 0.957 Time to Peak Clad Temperature, s 3.7 15.2 Peak Clad Temperature, oC 91.9 113.0 Minimum DNBR 6.5

20 LUONG BA VIEN, LE VINH VINH, HUYNH TON NGHIEM, NGUYEN KIEN CUONG

1.0 20 120 Period scram

0.9 18 DNBR 110 Period scram 0.8 16 Power (MW) Power 100 0.7 14

Temperature (oC) Temperature 90 0.6 12

0.5 10 80 DNBR

0.4 8 70 0.3 6 60 0.2 4 Overpower scram with set point 110% FP 50 Overpower scram with set point 110% FP 0.1 2

0.0 0 40 0 2 4 6 8 10 12 14 16 18 20 0 2 4 6 8 10 12 14 16 18 20 Time (s) Time (s) Fig. 7. Reactor power and cladding temperature transient of one shim rod withdrawal from a stable operation of 100%FP.

Cooling pump failure measures undertaken in design and construction, the removal of all control rods In the event of in-service primary or would not exceed 10 mm and insert a step secondary cooling pumps stopped working, the positive reactivity estimated of 0.3$. With this reactor is automatically shutdown by an reactivity insertion, the scram set-point of abnormal technological signal on low flow rate reactor overpower is attained almost (the setpoint is 40 m3/h for the primary flow, instantaneously. If the reactor scram is and 70 m3/h for the secondary flow). The initiated by overpower signal with a delay of residual heat after shutdown is about 6% FP 0.16 sec, the fuel surface temperature (30 kW) in maximum and the natural increases slightly before decreases with the convection process can itself assure the good power, the residual heat after shutdown is cooling of the core. sufficiently removed from the fuel by natural If the reactor is purposely maintained at convection of pool water without considerable full power operation, failure of cooling pumps increase of the temperature. leads to loss of heat removal from the pool water, and thus gradually increases of the pool Fig. 8 shows the analyzing results of water temperature. The results show that the the earthquake event assuming the clad temperature reaches the maximum protection system fails to shutdown the allowable operating clad temperature of 103 oC reactor, and Because of the loss of offsite at about 55 min; i.e. the reactor could continue power due to the earthquake, the primary and its operation for 55 minutes within the envelope secondary pumps stop operating. In this case, of the limiting conditions of operation. The the reactor power increases to the max value of results also show that even at the end of the 1.525 MW after 200 seconds from the simulation (7000 s) the clad temperature has initiation of this event. The reactor power then been well below that of the acceptance criterion rapidly decreases because the significant for anticipated operational occurrences. increasing of core water temperature so that the positive reactivity insertion is overtaken by the Earthquake negative reactivity feedback (about -0.44$). The postulated event of an earthquake of The reactor is then kept at subcritical state. The intensity grade VI is assumed to occur while cladding temperature reaches a maximum the reactor is at full power. Owing to the value of 118.2oC, then decreases with no

21 DESIGN ANALYSES FOR FULL CORE CONVERSION OF … significant overheating of the fuel. The value of about 1.12 MW. The cladding maximum outlet water reaches 89oC and temperature reaches to a maximum value of gradually decreases to a value at about 60oC, 118.38oC then gradually decreases to a stable which is still far below the saturation value of 115oC without nucleate boiling. The temperature. The minimum DNBR of 4.79 is maximum temperature of outlet water is 89oC much higher than the acceptance value. at the peak power then decreases and stabilizes at about 82oC, well below the In case the cooling pumps remain saturation point. The minimum DNBR in this working after the earthquake event (very case estimated about 4.74 is still far from the unlikely); the peak power reaches 1.57 MW acceptance criterion. within 300 seconds and decreases due to negative temperature feedback to a stable

120 1.6 16 Max. Cladding Temperature 110 1.4 14 DNBR 100 Power (MW) Power DNBR 1.2 12

Temperatute (oC) Temperatute 90 1.0 10 80 0.8 8 70 Max. Water Temperature at Outlet 0.6 6 60

0.4 4 50 Water Temperature at Inlet 0.2 2 40 Power 30 0.0 0 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 Time (s) Time (s)

Fig. 9. Power and Temperature responses to earthquake event while cooling pumps are stopped functioning.

Fuel cladding failure (MHA) 95% efficiency is not available, and there are no decay and deposition of radionuclides For the derivation of source term of this within the reactor building. event, it is assumed that no core melting occurs but cladding rupture of one fuel assembly is The evaluation of dose to a member of involved. It is also assumed that the damaged the public is calculated by code MACCS2 fuel assembly is irradiated at the maximum version 1.13.1, using the following neutron flux position in the core and the fuel assumptions: (1) The radionuclides are damage occurs immediately at the end of released to the environment through the 40 m operating cycle of 100 hrs with no decay. stack; (2) The Gaussian plume model is used to calculate air concentration of radioactivity; (3) From the damaged fuel assembly, 100% Tadmor and Gur parameterization is used for of noble gases (Xe, Kr), 25% halogens (I), and this analysis; (4) No building in the vicinity (an 1% of other radionuclides (Cs, Te) [21] are open area release), plume rise mechanics only released directly to the reactor building with due to momentum rise (non-buoyant plume) the assumption of no retention of volatile and no wet deposition are assumed; (5) The dry fission products in the pool water. During the deposition velocity is assumed to be 0.01 m/s, accident evolution, the emergency ventilation which corresponds to a particle with an system is not in place, the normal ventilation aerodynamic equivalent diameter of 2 m to 4 system V1 is in operation but HEPA filter with m (for unfiltered particulate releases) [15]; (6)

2211

LUONG BA VIEN, LE VINH VINH, HUYNH TON NGHIEM, NGUYEN KIEN CUONG Surface roughness length is specified as 50 cm; The effective equivalent doses, including (7) Mixing layer height is assumed to be 500 m cloudshine dose, inhalation dose and (see Table 36 in Appendix VII of Ref. 21); (8) groundshine dose, as a function of the distance -4 3 The breathing rate is 3.3x10 m /s; (9) No from the source are shown in Table XII and shielding and sheltering are assumed; (10) Fig. 10. It is seen that radiation exposure to the Doses at each downwind distance are general public with the maximum effective calculated for one year after the arrival of the dose of 0.64 mSv/year at distance from 400 m plume (11). The environmental release is to 500 m from the stack. This value is lower assumed to begin at the start of the weather than the annual dose limit of 1.0 mSv specified conditions: Pasquill class D2.0 (most frequent for the public [22]. stability class and most frequent wind speed).

Table XII. The annual effective dose to the public vs distance for the MHA.

Distance Effective Dose Distance Effective Dose (m) (mSv) (m) (mSv) 50 4.80E-02 1100 3.18E-01 150 1.43E-01 1300 2.59E-01 250 4.95E-01 1500 2.16E-01 350 6.42E-01 1700 1.83E-01 450 6.44E-01 1900 1.57E-01 550 5.94E-01 2250 1.23E-01 650 5.33E-01 2750 9.14E-02 750 4.74E-01 3250 7.08E-02 850 4.21E-01 3750 5.66E-02 950 3.75E-01 4250 4.64E-02

7.00E-01

6.00E-01

5.00E-01

4.00E-01

3.00E-01

EffectiveEquivalent mSv Dose, 2.00E-01

1.00E-01

0.00E+00 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 Downwind Distance, m Fig. 10. The annual effective dose to the public in MHA event within 5 km.

2113

DESIGN ANALYSES FOR FULL CORE CONVERSION OF …

- If one of the cooling pumps stopped IV. CONCLUSIONS working, the reactor is automatically shutdown Neutronics, steady-state thermal- by a scram signal on low flow rate. The decay hydraulic and transient/accidents analyses for heat is removed from the fuel by natural Dalat Nuclear Research Reactor show that with convection of pool water. In this event, if the a slight change in arrangement of Be rods, the reactor was purposely maintained at full power, main features of 92 LEU WWR-M2 FA cores it could be safely operated for 55 minutes when are equivalent to those of HEU and current maximum cladding temperature is still lower mixed fuel cores. than the permissible value of 103oC. The negative values of reactivity feedback - The postulated earthquake event of coefficients show the inherent safety feature and MSK intensity grade VI would cause a step shutdown margin of both candidate cores meets reactivity insertion of 0.3$. Even if the reactor the safety required value of -1% k/k. The fails to be scrammed, this positive reactivity working core with 92 fresh LEU fuel assemblies can be covered by negative temperature can be operated for 600FPDs or about 11 years feedback if the cooling pumps are stopped based on the current operating schedule without simultaneously, keeping the reactor sub-critical. shuffling. The neutron fluxes at the irradiation In case the cooling pumps continue operating positions are not much different from those of the after earthquake event, the negative temperature current mixed fuel core. feedbacks act to bring the reactor power to a stable level of about 1.12 MW without nucleate In thermal hydraulics aspect, the boiling. The minimum DNBR is much higher requirement of thermal-hydraulic safety margin than the acceptance criterion of 1.5. for two candidate cores in normal operational condition is satisfied. The calculated maximum - The maximum hypothetical accident cladding temperature in operational condition assumes 100% of noble gases (Xe, Kr), 25% is below the permissible value of 103oC. halogens (I), and 1% of other radio-nuclides (Cs, Te) in a most power fuel assembly after a long In transient/accidents aspect, some run are released into the environment through postulated initiating events and accident related 40m high stack. This event is considered to be to the conversion of the DNRR to full LEU very unlikely to occur for the DNRR. Even so, it core were selected and analyzed. Based on the would not cause undue radiological risk to the calculated results, conclusions might be environment or the public. withdrawn as following: - The excess reactivity insertions when ACKNOWLEADGMENTS inadvertent withdrawals of control rod from The authors would like to express their start-up or nominal power operation are gratitude to experts from the Reduced prevented by safety settings to initiate the Enrichment for Research and Test Reactors reactor scram at overpower and fast period. (RERTR) program of Argonne National None of these initiators would lead to the ONB Laboratory for financial support as well as very and DNB, ensuring the integrity of the fuel useful discussions during design calculation of cladding. The residual heat after shutdown is full core conversion for the Dalat Nuclear sufficiently removed from the fuel by natural Research Reactor. convection of pool water.

24 LUONG BA VIEN, LE VINH VINH, HUYNH TON NGHIEM, NGUYEN KIEN CUONG REFERENCES [13] Technical Design for Reconstruction and Enlargement of the Dalat Nuclear Research [1] Safety Analysis Report for the Dalat Nuclear Reactor - Volume 3. State Design Institute, Research Reactor, (2003). USSR State Committee for the Utilization of [2] VVR-M2 and VVR-M5 Fuel Assemblies- Atomic Energy, Moscow, (in Russian) (1979). Operation Manual, 0001.04.00.000 PЭ, (2006). [14] Additional Physics and Thermal-Hydraulic [3] Report on the 7th Core Loading for the Dalat Data for Reactor IVV-9, State Design Institute, Nuclear Research Reactor, NRI, Dalat, (in USSR State Committee for the Utilization of Vietnamese), (2009). Atomic Energy, Moscow, (in Russian) (1980). [4] J.R. Deen, W.L. Woodruff, C.I. Costescu, and [15] RELAP5/MOD3 Code Manual, SCIENTECH, L. S. Leopando, “WIMS-ANL User Manual Inc. Rockville, Maryland, (1999). Rev. 5”, ANL/RERTR/TM-99-07, Argonne [16] W. L. Woodruff, et al., A comparison of the National Laboratory, (February 2003). PARET/ANL and RELAP5/MOD3 Codes for [5] P. Olson, “A Users Guide for the REBUS-PC the Analysis of IAEA Benchmark Transients Code, Version 1.4,” ANL/RERTR/TM02-32, and the SPERT Experiments, RERTR (December 21, 2001). Program, ANL. [6] J. F. Briesmeister, Ed., “MCNP – A General [17] V. V. Le and T. N. Huynh, Application of Monte Carlo N-Particle Transport Code, RELAP5/MOD3.2 for the DNRR, Proceedings Version 4C”, LA-13709-M (April 2000). of JAEA Conf. 2006-001. [7] John G. Stevens, “The REBUS-MCNP Linkage”, [18] M. M. Shah, Improved General Correlation for Argonne National Laboratory, (2007). Critical Heat Flux during Upflow in Uniformly Heated Vertical Tubes, International Journal of [8] N.A. Hanan, J.R. Deen, J.E. Matos, “Analyses Heat and Fluid Flow, Vol. 8, No. 4, pp. 326- for Inserting Fresh LEU Fuel Assemblies 335 (1987). Instead of Fresh Fuel Assemblies in the DNRR in Vietnam, 2004 International Meeting on [19] MACCS2 Computer Code Application RERTR, Vienna, (2004). Guidance for Documented Safety Analysis, U.S. Department of Energy, (June 2004). [9] V. V. Le, T. N. Huynh, B. V. Luong, V. L. Pham, J. R. Liaw, J. Matos, “Comparative [20] A.G. Croff, A User Manual for the ORIGEN2 Analyses for Loading LEU Instead of HEU Computer Code, Oak Ridge National Fuel Assemblies in the DNRR”, RERTR Int’l Laboratory, (1980). Meeting, Boston, (2005). [21] INTERNATIONAL ATOMIC ENEGY AGENCY, Derivation of the Source Term and [10] Teresa Kulikowska et al., Raport IAE-40/A, Analysis of Radiological Consequences for (1999). Research Reactor Accidents, SAFETY [11] Arne P. Olson, M. Kalimullah, “A users guide REPORTS SERIES No. 53, VIENNA (2008). to the PLTEMP/ANL V3.8 Code”, [22] Governmental Decree for the Implementation ANL/RERTR, Argonne National Laboratory, of the Ordinance on Radiation Protection and (June, 2009). Control, Government of the Socialist Republic of Vietnam, No. 50/1998/ND-CP, Hanoi, (in [12] Le Vinh Vinh, Huynh Ton Nghiem and Vietnamese) (1998). Nguyen Kien Cuong, “Preliminary results of full core conversion from HEU to LEU fuel of the Dalat Nuclear Research Reactor”, RERTR Int’l Meeting, Beijing, (2009).

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Nuclear Science and Technology, Vol. 4, No. 1 (2014), pp. 26-35 Conceptual Nuclear Design of a 20 MW Multipurpose Research Reactor

Nguyen Nhi Dien, Huynh Ton Nghiem, Le Vinh Vinh, Vo Doan Hai Dang Reactor Center – Nuclear Research Institute – Vietnam Atomic Energy Institute 01 Nguyen Tu Luc, Dalat, Lamdong, Vietnam Seo Chulgyo, Park Cheol, Kim Hak Sung Korean Atomic Energy Institute, 150 Dukjin-dong, Yuseong-gu, Taejeon 305-353, Korea (Received 5 March 2014, accepted 10 March 2014)

Abstract: This paper presents some of studied results of a pre-feasibility project on a new research reactor for Vietnam. In this work, two conceptual nuclear designs of 20 MW multi-purpose research reactor have been done. The reference reactor is the light water cooled and heavy water reflected open-tank-in-pool type reactor. The reactor model is based on the experiences from the operation and

utilization of the HANARO. Two fuel types, rod and flat plate, with dispersed U3Si2-Al fuel meat are used in this study for comparison purpose. Analyses for the nuclear design parameters such as the neutron flux, power distribution, reactivity coefficients, control rod worth, etc. have been done and the equilibrium cores have been established to meet the requirements of nuclear safety and performance.

Keywords: HANARO, AHR, MTR, MCNP, MVP, HELIOS, dispersed U3Si2-Al, open-tank-in-pool, equilibrium core, BOC, EOC, shutdown margin.

I. INTRODUCTION has considerable experience in the research reactor technology through the design, Research reactor has been widely construction, operation and utilization of the utilized in various fields such as industry, High-flux Advanced Neutron Application engineering, medicine, life science, Reactor (HANARO) of 30 MWth. Therefore, environment, etc., and now its application in the framework of the joint study on the pre- fields are gradually being expanded together feasibility of MRR with KAERI, a model of with the development of its technology. The Advance HANARO Reactor (abbreviated as utilization of a research reactor is related to the AHR) has been developed to meet the necessary and essential technologies of requirements for use in the future [3,4]. Based information technology, nano-technology, on the model of AHR, a similar reactor model biotechnology, environmental technology and with plate fuel type MTR (abbreviated as space technology. Hence, R&D in the area of MTR) has been also developed for the purpose research reactor utilizations has a large effect of comparison between the two fuel types. on the growth of a national industry. II. NUCLEAR DESIGN REQUIREMENTS Vietnam has a plan to construct a high performance multipurpose research reactor A. General (MRR) to satisfy increasing utilization A research reactor should be designed in demands. So, the pre-feasibility studies to build conformity with user's requirements. The a new MRR have been set [1,2]. The Korea reactor type, power, and core configuration, Atomic Energy Research Institute (KAERI) systems and the installed experimental ©2014 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute

NGUYEN NHI DIEN, HUYNH TON NGHIEM, LE VINH VINH, VO DOAN HAI DANG facilities depend on the application purposes 1) The neutron flux variation at the irradiation and on the construction and operation costs as sites and the nose of the beam tubes should be well. Hence, a flexible design is an stable with a 5% variation regardless of a indispensable feature when considering a loading or unloading of samples. future expansion of its experimental facilities. 2) The axial neutron flux gradient in the The major basic principles to develop reflector region should be within ±20% over a models of the conceptual design are as follows. length of 50 cm. 1) Multipurpose research reactor with a 3) The maximum fast and thermal neutron medium power fluxes at an irradiation site inside the core 2) High ratio of flux to power should be greater than 1.3x1014 and 4.0x1014 3) High Safety and Economics n/cm2-s, respectively. The maximum thermal 4) Sufficient spaces and expandability of the neutron flux at the reflector region should be 14 2 facility for various experiments greater than 4.0x10 n/cm -s. Fundamentally, a research reactor should 4) The maximum local power peaking factor be designed to achieve the established safety should be less than 3.0. objectives such as the IAEA standards. The 5) The average discharge burn-up of the fuel nuclear design requirements for the AHR and assembly should be higher than 50% of the MTR are considered in two parts, functional initial fissile heavy material, U-235. and performance requirements. 6) The reactor operating cycle should be B. Funtional Requirements longer than 30 days. The functional requirements aim to ensure the safety of the reactor and ready to III. CORE CONCEPT operate in all conditions. 1) The power coefficient and temperature and void The basic concepts of the reactor are the coefficients of the reactivity should be negative for light water cooled and moderated, heavy water all operational and accident conditions. reflected, open-tank-in-pool type research reactor and 20 MW power cores loaded with 2) The shutdown margin should be at least 10 two typical geometric kinds of fuel elements as mk (1mk = k/k  1/1000) regardless of any rod or flat plate. changes in the reactor condition. A. Fuel 3) The second reactor shutdown system should be prepared to improve the reactor safety and Fuels selected for the design are its shutdown margin should be at least 10 mk commercial or commercial available. The fuel for all relevant design basis fault sequences. meat is fabricated by a dispersion of high 4) The excess reactivity should be at least 10 density U3Si2 particles into pure Al with its mk at the end of cycle for conducting uranium enrichment 19.75 wt%. Two kinds of experiments and 15 mk for the Xe override. fuel assemblies in the core are standard fuel assembly and control fuel assembly (including C. Performance requirements control rods inside fuel assembly). Some The performance requirements aim to specifications of the fuel elements and ensure meeting the requirements of use and assemblies are listed in Table I and their cross high economic efficiency. sectional views are showed in Figure 1. 27

CONCEPTUAL NUCLEAR DESIGN OF A 20 MW MULTIPURPOSE RESEARCH REACTOR

Table I. Specifications of the fuel element and assembly Fuel element 66.0w/%U, 5.2w/%Si, 72.8w/%U, 6.0w/%Si, Meat content 28.8w/%Al 21.2w/%Al Fuel length (mm) 700.0 700.0 Fuel diameter/widththickness (mm) 6.35/5.49 (In/Out) 64.0/51.4x0.61 (S/C)* Fuel density (g/cm3) 6.06 6.6 Cladding thickness (mm) 0.76/1.19 (In/Out) 0.37/0.445 (In/Out) Cladding material Al Al Fuel assembly Shape Hexagonal Square Element number 36/18 (S/C) 21/17 (S/C) * S/C: Standard fuel assembly / Control fuel assembly

a) AHR standard b) AHR Control c) MTR standard d) MTR control Fig. 1. Cross sectional view of AHR and MTR standard and control fuel assemblies

B. Core Arrangement core. The reactor regulating system shares control rods with the reactor protection system. The core has 23 lattices that consist of Fig. 2 shows the horizontal cross sectional fourteen standard assemblies, four control view of the AHR and MTR cores. Some assemblies and three in-core irradiation sites. specifications of the cores are listed in Table II. The heavy water reflector tank of 200 cm in diameter and 120 cm in height surrounds the

Fig. 2. The horizontal cross sectional view of the AHR and MTR cores

28 NGUYEN NHI DIEN, HUYNH TON NGHIEM, LE VINH VINH, VO DOAN HAI DANG

Table II. The specifications of the cores Reactor type AHR MTR Core volume (cm3) 1199.5 x 70 1527.7 x 70 Fuel assembly Number 16 S + 4 C 16 S + 4 C Control rod Number 4 4 Absorber material Hf Hf Total weight U-235 (kg) 9,87 10,12 In-core irradiation sites 3 3

Fig. 3. The layout of the experimental sites of the AHR and MTR

IV. NUCLEAR ANALYSIS distribution, the reactivity of the core and the reactivity worth of control rods were also To confirm that the conceptual cores assessed to meet the requirements. Two core satisfy the functional and performance configurations with one and three in-core requirements, nuclear analyses are performed irradiation sites were proposed. Although the for fresh core and equilibrium core with first configuration (with one irradiation site) is several code systems such as MCNP [5], MVP better in the fuel saving point of view, the [6], HELIOS [7], etc. configuration with three in-core irradiation A. Fresh Core sites was selected to meet predicted utilization The basic analysis of the core of in-core irradiation in the future. characteristics was performed for the fresh core As the ultimate goal of a research reactor with and without irradiation facilities. is its utilization, the irradiation facilities should The core configuration should be be designed in conformity with the user's designed to meet the functional and requirements. The required irradiation facilities performance requirements. The neutron flux at should be located at proper positions to the in-core irradiation sites and the reflector maximize neutron utilization and minimize region of the cores without irradiation facilities reactivity effect. Based on the neutron flux was calculated by the MCNPX code [8] using a distribution of the reflector region, the mesh tally. On the other hand, the power arrangement by their purposes has been studied to achieve the objectives above. Their 29

CONCEPTUAL NUCLEAR DESIGN OF A 20 MW MULTIPURPOSE RESEARCH REACTOR reactivity worth is considered as a priority The reactivity effect by the irradiation because of the influence to the reactor core. facilities was estimated to be 20.2 mk and 28.9 Various layouts of the irradiation facilities mk and the total control rods worth 182.4 mk were proposed, and one of them was selected. and 217.7 mk for AHR and MTR, respectively. To evaluate the stability of neutron flux at the Table III shows the neutron fluxes at the irradiation sites, their neutron fluxes were irradiation facilities. Figure 4 presents the calculated when the control rods are located at thermal and fast neutron distribution of the 300 mm and fully withdrawn. AHR fresh core.

Table III. Neutron fluxes at the experimental sites

Neutron flux [n/cm2/sec]( Thermal<0.625eV, Fast>1.0MeV) AHR MTR

Maximum Average Maximum Average Thermal Fast Thermal Fast Thermal Fast Thermal Fast CT 4.46E+14 1.46E+14 3.04E+14 9.80E+13 4.01E+14 1.13E+14 2.87E+14 8.06E+13 IR1 3.21E+14 1.18E+14 2.23E+14 8.29E+13 3.37E+14 9.31E+13 2.49E+14 6.76E+13 IR2 3.16E+14 1.20E+14 2.23E+14 8.26E+13 3.33E+14 9.16E+13 2.46E+14 6.65E+13 CNS 8.71E+13 1.15E+12 7.01E+13 8.76E+11 8.49E+13 1.69E+12 6.48E+13 1.24E+12 ST1 1.37E+14 1.96E+12 - - 1.40E+14 3.23E+12 - - ST2 2.40E+14 3.47E+12 - - 1.79E+14 1.01E+13 - - NR 1.28E+14 3.20E+11 - - 1.28E+14 1.32E+12 - - NTD1 4.74E+13 1.13E+11 4.31E+13 8.12E+10 4.93E+13 4.19E+11 4.26E+13 3.19E+11 NTD2 4.63E+13 9.91E+10 4.24E+13 7.60E+10 5.29E+13 4.84E+11 4.57E+13 3.56E+11 NTD3 5.16E+13 2.43E+11 4.70E+13 2.04E+11 4.64E+13 5.21E+11 3.93E+13 3.78E+11 HTS1 6.96E+13 3.42E+11 5.96E+13 2.71E+11 7.02E+13 6.30E+11 5.79E+13 5.03E+11 HTS2 2.23E+13 2.07E+10 1.93E+13 1.39E+10 2.25E+13 2.81E+10 1.97E+13 2.29E+10 NAA1 1.39E+14 4.96E+11 1.20E+14 3.88E+11 1.22E+14 8.15E+11 1.05E+14 6.27E+11 NAA2 4.11E+13 - 3.59E+13 - 4.00E+13 - 3.55E+13 - NAA3 1.74E+13 - 1.52E+13 - 1.53E+13 - 1.35E+13 - RI1 3.53E+14 1.47E+13 2.60E+14 9.05E+12 2.31E+14 1.49E+13 1.69E+14 9.28E+12 RI2 3.44E+14 1.42E+13 2.57E+14 8.91E+12 2.18E+14 1.47E+13 1.58E+14 9.13E+12 RI3 2.46E+14 4.03E+12 1.85E+14 2.54E+12 2.10E+14 1.53E+13 1.58E+14 9.56E+12 RI4 2.48E+14 4.23E+12 1.86E+14 2.82E+12 2.03E+14 1.45E+13 1.52E+14 8.92E+12 RI5 2.24E+14 3.12E+12 1.67E+14 2.10E+12 2.15E+14 1.55E+13 1.58E+14 9.51E+12

30

NGUYEN NHI DIEN, HUYNH TON NGHIEM, LE VINH VINH, VO DOAN HAI DANG

a) Thermal Neutron Flux b) Fast Neutron Flux

Fig. 4. Neutron flux profile at the AHR fresh core

B. Equilibrium Core 6-batch core) are assessed. The 9-batch cores show a high discharge burnup and a good fuel An equilibrium core is dependent on an economy, but the cycle lengths are less than 30 operation strategy, so there may be various days. They look proper for a low utilization equilibrium cores according to a reactor condition of the reactor. The 6-batch cores with operating strategy. In this report, an a cycle length greater than 30 days are suitable equilibrium core is proposed and analyzed to for the design requirements, so they are meet the established design requirements. selected for evaluating in detail. In the 6-batch Fuel Management core, three of the standard fuel assemblies or A candidate model for an equilibrium two of the standard fuel assemblies and two of core can be easily obtained by considering the control fuel assemblies are replaced for an target discharge burnup, cycle length and operation cycle, so the whole core will be excess reactivities at begin of cycle (BOC) and replaced for 6 cycles according to the loading end of cycle (EOC). There are many candidate strategy. There are many loading patterns that models according to the number of reloaded they depend on the fuel management strategy. fuel assemblies and the loading pattern. The The loading pattern showed in Table IV is equilibrium cores with 2 or 3 fuel assemblies evaluated in detail. reloaded for an operation cycle (the 9-batch or Table IV. Loading location of the fuel assemblies for 6-batch cores Cycle Assembly Number Loading Location (standard+control) AHR MTR 1 2+2 H14,H16,C1,C3 H9,H12,C1,C3 2 3+0 H8,H10,H12 H14,H15,H7 (move H14,H15,H7 to H2,H4,H6)

31 CONCEPTUAL NUCLEAR DESIGN OF A 20 MW MULTIPURPOSE RESEARCH REACTOR

3 3+0 H7,H9,H11 H13,H10,H16 (move H13,H10,H16 to H3,H5,H1) 4 2+2 H13,H15,C2,C4 H8,H11,C2,C4 5 3+0 H2,H4,H6 H14,H15,H7 (move H14,H15,H7 to H2,H4,H6) 6 3+0 H1,H3,H5 H13,H10,H16 (move H13,H10,H16 to H3,H5,H1)

Once a cycle length and a loading presents the calculated results of the average pattern are determined, an equilibrium core is burnup and reactivity of 6 cycles for different obtained by numerical iterations. The initial cycle lengths. From these results, it can be core is loaded with the new FAs then the concluded that the 36 days cycle for AHR and burnup calculations are iterated by the loading 34 days cycle for MTR meet the performance pattern until the parameters of burnup and requirements. reactivity are stable over 6 cycles. Table V

Table V. Burnup and reactivity of the equilibirum cores

Reactor type AHR MTR Cycle Length (days) 35 36 37 33 34 35 Average Burnup (%U-235) - BOC 23.43 24.02 24.61 22.38 23.04 23.70 - EOC 31.82 32.65 33.47 29.08 29.94 30.81 - Discharge 50.35 51.77 53.18 48.65 49.91 51.17 Reactivity (mk) - BOC (no Xe) 111.9 109.9 107.8 87.8 85.8 83.6 - Fuel Depletion 37.5 38.7 39.9 15.1 16.7 18.3 - Xenon Buildup 38.1 38.1 38.0 36.2 36.3 36.3 - Power Defect 3.0 3.0 3.0 3.0 3.0 3.0 - EOC (eq. Xe) 33.4 30.1 26.9 33.5 29.8 26.0 - Shutdown Margin 15.0 17.1 19.6 22.2 24.2 26.4

Power Distribution 6 cycles at a 300 mm position of the control The power distribution is strongly rods was calculated. Table VI shows maximum dependent on the positions of the control rods total peaking factors for the 6 cycles equilibrium and it was checked for all possible positions at cores and Table VII shows the power 5 cm intervals. The largest maximum linear distributions and peaking factors at the cycle power of the equilibrium cores was observed at that total peaking factor reaches the maximum a 300 mm position of the control rods. The value. The maximum local power peaking factor power distribution for the equilibrium cores of for AHR and MTR are 2.56 and 2.79 respectively. 3230

NGUYEN NHI DIEN, HUYNH TON NGHIEM, LE VINH VINH, VO DOAN HAI DANG

Table VI. Maximum total peaking factor for the equilibrium cycles

Reactor Parameter Cycle type 1 2 3 4 5 6 Position of FA H02 C2 H01 H03 C3 H04 AHR Fq(peaking factor) 2.47 2.56 2.5 2.46 2.56 2.49 Vị trí FA H09 H04 H01 H11 H04 H01 MTR Fq 2.69 2.77 2.74 2.76 2.76 2.79

Table VII. Power distribution and peaking factor for equilibrium cores (cycle 5 for AHR, cycle 6 for MTR) AHR MTR Location Total Power (kW) Fq Total Power (kW) Fq H01 1117 1.93 1167 2.79 H02 1061 1.72 1158 1.82 H03 1314 2.21 1229 1.96 H04 897 1.58 1092 2.51 H05 1064 1.73 1223 2.02 H06 1305 2.18 1143 1.82 H07 811 1.38 1033 2.12 H08 1071 1.64 1179 2.21 H09 1329 2.37 985 1.67 H10 1063 1.76 1066 2.31 H11 1088 1.65 1213 2.26 H12 1298 2.29 987 1.66 H13 1224 1.78 1137 2.01 H14 991 1.4 1093 1.91 H15 1184 1.57 1170 2.05 H16 1050 1.42 1174 1.96 C1 577 2.44 445 1.28 C2 491 2.04 527 1.55 C3 589 2.56 449 1.29 C4 476 1.95 531 1.56

Reactivity Coefficients region is to cool the fuel assemblies, and so called a ‘coolant’ and the light water in the To affirm the inherent safety, the gaps of the flow tubes is called a ‘moderator’. reactivity coefficients should be determined. Nuclear characteristics of these two light water They include temperature coefficients of fuel, regions are somewhat different, and a heat light water and heavy water. Physical changes transfer between them is small. Therefore, their of water due to a temperature change could be temperature variations following a power considered in two ways: one is a density change are also different, thus the respective change, and the other is a cross section change temperature coefficients were computed for a nuclear reaction. There are the gaps of the separately. The effect of a spectrum hardening flow tubes for AHR. The light water in the fuel

33 CONCEPTUAL NUCLEAR DESIGN OF A 20 MW MULTIPURPOSE RESEARCH REACTOR of neutrons following a temperature increase coefficient of moderator. where almost of for heavy water is so small that it can be arriving neutrons are slowed down) and meet negligible. Table VIII presents the result of the functional requirements. The temperature temperature and void coefficients. From this variation of moderator is so small, therefore its result, they are negative (except temperature contribution to power coefficient is small.

Table VIII. Reactivity coefficients of temperature and void Parameter AHR MTR Fuel temperature coefficient (mk/K) <-0.002 <-0.02 Light water temperature coefficient (mk/K) - Coolant -0.059 -0.11 - Moderator 0.06 Light water void coefficient (mk/%) 0 - 5 % -1.23 -1.79 5 - 10 % -1.37 -1.97 10 - 20 % -1.48 -2.25 Heavy water void coefficient (mk/%) 0 - 5 % -1.26 -0.79

V. CONCLUDING REMARKS the temperature coefficients are negative showed the inherent safety feature. The From the functional and performance parameters for utilization and for the safety requirements, two reactor models AHR and aspects of the reactor respectively meet the MTR were proposed and investigated. The performance and functional requirements. reference reactors are the light water cooled and moderated, heavy water reflected and The comparison of cores loaded with 2 open-tank-in-pool type research reactors with a different fuel types, AHR and MTR, shows that 20 MW power. the AHR fuel type core has a little longer operation cycle and higher discharge burn up as a The maximum fast and thermal neutron result. In the safety point of view, the MTR core 14 flux in the core region are greater than 1.0×10 has an advantage because of shutdown margin, 2 14 2 n/cm s and 4.0×10 n/cm s, respectively. In temperature and coolant void coefficients are the reflector region, the thermal neutron peak higher compared to those of AHR core. occurs about 28 cm far from the core center and the maximum flux is estimated to be REFERENCES 4.0×1014 n/cm2s. [1] Luong Ba Vien and C. Park et.al., Joint For the equilibrium cores, the cycle KAERI/VAEC pre-possibility study on a new length is greater than 30 days, the whole core research reactor for Vietnam, KAERI/TR- will be replaced for 6 cycles, and the assembly 2756/2004, (May, 2004). average discharge burnup is greater than 50%. [2] Nguyen Nhi Dien et al., Report on Study Project For the proposed fuel management scheme, the No BO/06/01-04, (in vietnamese), (2008). maximum peaking factor Fq is less than 3. The [3] Seo Chul Gyo, Huynh Ton Nghiem et al., shutdown margins by the 1st and 2nd Conceptual Nuclear Design of a 20 MW shutdown systems are greater than 10 mk and 34

NGUYEN NHI DIEN, HUYNH TON NGHIEM, LE VINH VINH, VO DOAN HAI DANG

Multipurpose Research Reactor - KAERI/TR- [7] E. A. Villarino, R. J. J. Stamm'ler, A. A. Ferri, 3444/(2007). J. J. Casal, HELIOS: Angularly Dependent Collision Probabilities, Nucl. Sci. & Eng., 112, [4] Hee TaekChae, Le Vinh Vinh et al., Conceptual 16, (1992). Thermal Hydraulic Design of a 20MW Multipurpose Research Reactor - KAERI /TR- [8] Denise B. Pelowitz (Editor), MCNPX User's 3443/(2007). Manual, LA-CP-05-0369, Los Alamos National Lab, (2005). [5] J. F. Briesmeister (Editor), MCNP-A General Monte Carlo N-Particle Transport Code, LA- 12625-M, Los Alamos National Lab, (1993). [6] Yasunobu NGAYA et al., MVP/GMVP II: General Purpose Monte Carlo Codes for Neutron and Photon Transport Calculations based on Continuous Energy and Multigroup Methods, JAERI 1348, (2005).

35

Nuclear Science and Technology, Vol. 4, No. 1 (2014), pp. 36-45 Some Main Results of Commissioning of The Dalat Research Reactor with Low Enriched Fuel

Nguyen Nhi Dien, Luong Ba Vien, Pham Van Lam, Le Vinh Vinh, Huynh Ton Nghiem, Nguyen Kien Cuong, Nguyen Minh Tuan, Nguyen Manh Hung, Pham Quang Huy, Tran Quoc Duong, Vo Doan Hai Dang, Trang Cao Su, Tran Tri Vien Nuclear Research Institute –Vietnam Atomic Energy Institute 01-Nguyen Tu Luc, Dalat, Vietnam (Received 5 March 2014, accepted 13 April 2014)

Abstract: After completion of design calculation of the Dalat Nuclear Research Reactor (DNRR) for conversion from high-enriched uranium fuel (HEU) to low-enriched uranium (LEU) fuel, the commissioning programme for DNRR with entire core loaded with LEU fuel was successfully carried out from 24 November 2011 to 13 January 2012. The experimental results obtained during the implementation of commissioning programme showed a good agreement with design calculations and affirmed that the DNRR with LEU core have met all safety and exploiting requirements. Keywords: HEU, LEU, physics start up, energy start up, effective worth, Xenon poisoning, Iodine pit.

I. INTRODUCTION hours testing operation without loading at nominal power (December, 13rd, 2011). Physics and energy start-up of the Dalat Nuclear Research Reactor (DNRR) for full II. PHYSICS START UP core conversion to low enriched uranium (LEU) fuel were performed from November Physics startup of reactor is the first 24th, 2011 until January 13th, 2012 according to phase of carrying out experiments to confirm an approved program by Vietnam Atomic the accuracy of design calculated results, Energy Institute (VINATOM). The program important physical parameters of the reactor provides specific instructions for manipulating core to meet safety requirements. Physics fuel assemblies (FAs) loading in the reactor startup includes fuel loading gradually until to core and denotes about procedures for carrying approach criticality, loading for working core out measurements and experiments during and implementing experiments to measure physics and energy start-up stages to guarantee parameters of the core at low power such as that loaded LEU FAs in the reactor core are in control rods worth, shutdown margin, accordance with calculated loading diagram temperature effect,… and implementation necessary measurements A. Fuel loading to approach criticality to ensure for safety operation of DNRR. The loading of LEU FAs to the reactor Main content of the report is a brief core was started on November, 24th, 2011 presentation of performed works and achieved following a predetermined order in which each results in the physics and energy start up stages step loaded one or group LEU FAs to the for DNRR using LEU fuel assemblies, that is reactor core. After each step, the ratio of from starting loading LEU fuel to the reactor th N0 core (November, 24 , 2011) until finishing 72 (N0 is initial number of neutron count rate, Ni

©2014 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute

LUONG BA VIEN et al.

Ni is that to be obtained after step ith) was for working core, effective worth of loaded evaluated to estimate critical mass. At 15h35 fuel assemblies and shutdown margin were on November, 30th, 2011 the reactor reached preliminarily evaluated to ensure shutdown critical status with core configuration including margin limit not be violated. Fig. 3 shows the 72 LEU FAs and neutron trap in center (see current working core of DNRR, including 92 Fig. 1 and 2). LEU FAs (80 fresh LEU FAs and 12 partial burnt LEU FAs, the burn up about 1.5 to 3.5 Established critical core configuration %) and neutron trap at the center. Total mass of with 72 LEU FAs having neutron trap is in U-235 that was loaded to the reactor core is good agreement with design calculated about 4246.26 g. Shutdown margin (or results. With 72 LEU FAs, by changing subcriticality when 2 safety rods are fully position of some fuel assemblies, all new criticality conditions were achieved with withdrawn) is 2.5 $ (about 2% k/k), smaller lesser inserting position of regulating rod. It is than calculated value (3.65 $) but still concluded that the above critical configuration completely satisfy the requirement >1% for (Fig. 1) is the minimum one among the DNRR. Excess reactivity of the core established configurations. The critical mass configuration is about 9.5 $, higher than of Uranium is 15964.12 g in which Uranium- calculated value (8.29 $), ensuring operation 235 is 3156.04 g. time of the reactor more than 10 years with recent exploiting condition. So, it can be said B. Fuel loading for the working core that the current working core meets not only After completion of fuel loading to safety requirements but reactor utilization approach criticality, fuel loading for working also (ensure about shutdown margin and core was carried out from December, 6th, 2011 sufficient excess reactivity for reactor to December, 14th, 2011. During fuel loading operation and utilization).

Fig. 1. Critical core configuration and Fig. 2. N0/Ni ratio versus number of FAs order of loaded fuel assemblies loading to the core

37 SOME MAIN RESULTS OF COMMISSIONING OF …

LEU Fuel Wet channel

Berrylium Dry channel

Aluminum Empty cell

Neutron trap

Fig. 3. Working core configuratiom with 92 LEU FAs

C. Performed experiments in the working core working core in configuration with 82 fresh configuration LEU FAs and 92 LEU FAs. Determination of control rod worth Control rods worths and integral characteristics in core configuration with 92 To calibrate control rod worth, doubling LEU FAs are presented in Table I, Fig. 4 time method was applied for regulating rod and 5. Measured results were smaller than while reactivity compensation method was design calculated results about 12% in used for shim rods and safety rods. The average. calibration of control rods of DNRR were implemented two time during fuel loading for

Table I. Effective worth of regulating rod, 4 shim rods and 2 safety rods in core configuration with 92 LEU FAs.

Effective reactivity ($) Control Rod Measured value Calculated value Regulating rod 0.495 0.545 Shim rod 1 2.966 3.237 Shim rod 2 3.219 3.263 Shim rod 3 2.817 3.473 Shim rod 4 2.531 3.086 Safety rod 1 2.487 2.744 Safety rod 2 2.195 2.795

38

LUONG BA VIEN et al.

Reactivity ($) Reactivity ($)

Position (mm) Position (mm) Fig. 4. Integral characteristics of regulating rod Fig. 5. Integral characteristics of 4 shim rods

Thermal neutron flux distribution From the measured results, it can be measurement in the reactor core seen that the maximum peaking factor of 1.49 is achieved at outer corner of hexagonal Measurement of thermal neutron flux tube of the fuel assembly in cell 6-4. Neutron distribution following axial and radial in the distribution of working core has large reactor core was carried out by Lu metal foils deviation from North (thermal column) to neutron activation. A number of positions in South (thermalizing column). Neutron flux the reactor core were chosen to measure in southern region of the core (cell 12-1 and thermal neutron flux distribution including 12-7) is about 28 % smaller than those in neutron trap, irradiation channels 1-4 and 13-2, Northern region (cell 2-1 and 2-7). The and 10 FAs at the cells: 1-1, 2-2, 2-3, 2-7, 3-3, asymmetry of the reactor core has reason 3-4, 4-5, 6-4, 12-2 and 12-7. Figs 6 to 9 present from the not identical reflector that was noted the measured results of axial and radial neutron from the former HEU fuel core. flux distributions of the reactor core.

1.1 1.1

1 1

0.9 0.9 FA cell 2-3 0.8 0.8 FA cell 3-3 0.7 0.7 FA cell 4-5

0.6 0.6

0.5 0.5 RelativeUnit

RelativeUnit 0.4 0.4

0.3 0.3

0.2 0.2

0.1 0.1

0 0 -35 -30 -25 -20 -15 -10 -5 0 5 10 15 20 25 30 35 0 5 10 15 20 25 30 35 40 45 50 55 60 Position from the bottom to the top (cm) Position from the bottom to the top (cm) Fig. 6. Axial thermal neutron flux distribution in Fig. 7. Axial thermal neutron flux distribution in the the fuel assemblies neutron trap

39 SOME MAIN RESULTS OF COMMISSIONING OF …

Fig. 8. Thermal neutron flux distribution of FAs Fig. 9. Thermal neutron flux distribution of the and irradiation positions in comparison with FA’s corners in comparison neutron trap. with neutron trap

Determination of effective worth of FAs, determined by comparing position change of beryllium rods and void effect control rods before and after withdrawing FA or beryllium rod or before and after inserting The measurements of effective worth of watertight aluminum tube. Reactivity worth FAs, beryllium rods and void effect (by values were obtained using integral inserting an empty aluminum tube with characteristics curves of control rods. diameter of 30 mm) were also performed. These are important parameters related to Figs 10÷12 show the measured results safety of the reactor. Positions for of effective worth of 14 FAs in the reactor measurement of effective reactivity of FAs, core at different positions; effective worth of Be rods and void effect were chosen to beryllium rods around neutron trap and a examine the distribution, symmetry of the new beryllium rod at irradiation channel 1-4; core and the interference effects at some void effect at neutron trap, irradiation special positions. Effective reactivity of FAs, channel 1-4 and cell 6-3, which surrounded beryllium rods and void effect were by other FAs.

Fig. 10. Effective worth of FAs in the reactor core Fig. 11. Effective worth of Be rods in the reactor core

40 LUONG BA VIEN et al.

.

Reactivity ($) Reactivity

Temperature (0C)

Fig. 12. Measured results of void effect at Fig. 13. Negative reactivity insertion some positions in the reactor core dependent on pool water temperature temperature The most effective worth of fuel established after each increased step of pool assembly measured at cell 4-5 is 0.53 $. water temperature about 2.50C. Basing on the Measured results of effective reactivity of fuel change of regulating rod position (due to assemblies and Be rods show a quite large change of temperature in the reactor core) the tilting of reactor power from North to South temperature coefficient of moderator was direction. Void effect has negative value in the determined. reactor core (cell 1-4 and 6-3) while positive in Heating process of water in reactor pool the neutron trap. Void effect in neutron trap by operating primary cooling pump took long has positive value because almost neutrons time so water in neutron trap also heated up coming in neutron trap are thermalized, that is and inserted positive reactivity (as explanation absorption effect of water in neutron trap is in measurement of void effect), as opposed to dominant compared to moderation effect. The temperature effect in the reactor core. So, a replacement of water by air or decreasing of hollow stainless steel tube 60 mm diameter water density when increasing steadily of was inserted in neutron trap to eliminate temperature introduces a positive reactivity. positive temperature effect of neutron trap. With the core using HEU fuel also has positive reactivity of void in neutron trap. Fig. 13 shows measured results of temperature coefficient of moderator with Determination of temperature coefficient initial temperature of 17.7 oC. Based on these of moderator results, the temperature coefficient of Temperature coefficient of moderator is moderator is determined about -9.110-3 $/oC. the most important parameter, demonstrating Measured result without steel pipe containing inherent safety of reactor. To carry out air at neutron trap was about -5.210-3 $/oC. experiment, the temperature inside reactor pool Thus, temperature coefficient of moderator 0 was raised about 10 C by operating primary including neutron trap still has negative value. cooling pump without secondary cooling Temperature coefficient of moderator of the pump. To measure temperature coefficient of core loaded with 88 HEU FAs measured in moderator, criticality of the reactor was 1984 was -8.010-3 $/oC.

41 SOME MAIN RESULTS OF COMMISSIONING OF …

III. ENERGY START-UP 13-2 and rotary specimen was measured by using Au foil activation method. Also, on A. Power ascension test January 17th, 2012 thermal neutron flux of On January 6th, 2012 reactor power has positions mentioned above was measured at been increased at levels of 0.5% nominal power level 100%. Measured results of thermal power, 10% nominal power and 20% nominal neutron flux at several irradiation positions in power. At each power level, thermal neutron the reactor core with different power levels are flux in neutron trap, irradiation channels 1-4, presented in Table II.

Table II. Measured results of thermal neutron flux at several irradiation positions at different reactor power levels Irradiation Power (% Nominal power) positions 0,5 10 20 100 Neutron trap 1.143E+11 2.063E+12 4.174E+12 2.122E+13 Channel 1-4 5.288E+10 9.719E+11 1.965E+12 8.967E+12 Channel 13-2 4.749E+10 8.542E+11 1.682E+12 N/A Rotary Specimen N/A N/A N/A 4.225E+12

Based on the reactor power determined system parameters was about 372 kW. This by thermal neutron flux measurements at low value enables us to raise the reactor power to power levels, on January 9th, 2012 the reactor full power level. 15h32 on January, 9th, 2012 the was ascended power: 0.5%, 20%, 50%, 80% reactor was raised to 100% nominal power and and then operated at 80% nominal power maintained at this power about 65 hours before during 5 hours for determination of thermal decreasing to 0.5% nominal power to measure power and examination of technological Xenon poisoning transient. Table III presents parameters and gamma dose before raising the the values of thermal power of the reactor reactor power to nominal level. during the first 8 hours after the reactor reached 100% nominal power. The data indicate that Thermal power of the reactor thermal power is just only about 460 kW, lower corresponding to 80% nominal power level than design nominal power about 10%. (based on indication of control system) after 5 hours calculated based on primary cooling

Table III. Thermal power of the reactor with operation time after the reactor reached 100% nominal power T (1) T (1) G P Time in out I I [oC] [oC] [m3/h] [kW] 15h30 29,2 22,4 49,4 390 16h00 30,3 22,9 49,3 423 17h00 31,0 23,1 49,8 456 18h00 31,0 23,0 49,8 462 1h00 30,9 22,9 49,8 462 20h00 30,8 22,9 49,6 454 21h00 30,8 22,9 49,8 456 22h00 30,7 22,8 50,5 457 23h00 30,6 22,7 50,1 459 24h00 30,5 22,6 49,6 455

42 LUONG BA VIEN et al.

B. Xenon poisoning transient and Iodine hole C. Power adjustment The experiment to determine the curve In the process of gradually raising power built up of Xenon poisoning and then in energy start-up, although power indication calculating its equilibrium poisoning was on control system was 100% but calculated conducted from January 9th, 2012 to January thermal power of the reactor through flow rate 12th, 2012 when the reactor was in 100% of primary cooling system and difference nominal power (indicating of control system between inlet and outlet temperatures of the without adjusting power) . Next, Iodine hole heat exchanger was only 460 kW, smaller than was also determined from 12 to January 13th, nominal power about 10%. The reason was 2012 after reducing power of the reactor from mainly due to power density of the core using 100% to 0.5% nominal power by monitoring 92 LEU FAs were higher than the mixed core the shift position of regulating rod. using 104 FAs before. The adjustment to increase thermal power of the reactor was Fig. 14 presents measured results of performed by changing the coefficients on the Xenon poisoning curve and Iodine pit of the control panel. After adjusting, the reactor was above experiment. Xenon equilibrium operated to determine thermal power at power poisoning and other effects is totally about -1.1 setting 100%. The results of thermal power eff and the maximum depth of Iodine pit obtained from the next long operation was determined about -0.15  after 3.5 hours eff about 510.5 kW. This value includes 500 kW since the reactor was down to 0.5% nominal thermal power of the reactor and about 10 kW power. After adjusting thermal power up to generated by primary cooling pump. 500 kW, during the long operation from March, 12-16, 2012, after the reactor was D. Measurement of neutron flux and neutron operated 68 hours at nominal power, total spectrum after power adjustment value of poisoning and temperature effects is After carrying out reactor power

about -1.32 eff. adjustment, thermal neutron flux at some

Negative ($) reactivity

Xe poisonning Iodine Pit PpppPithole Time (hour) Fig. 14. Negative reactivity insertion by Xenon poisoning with operation time and Iodine pit

43 SOME MAIN RESULTS OF COMMISSIONING OF … irradiation positions in the reactor core and start up were carried out successfully. DNRR neutron spectrum in neutron trap were was reached criticality at 15:35 on November, measured again by neutron activation foils. 30th, 2011 with 72 LEU FAs, consistent with Measured maximum neutron flux at neutron calculated results. Then, the working core with trap was 2.23  1013 n/cm2.s (compared with 92 LEU FAs has been operating 72 hours for calculated result was 2.14÷2.22  1013 n/cm2, testing at nominal power during from January, th th depending on shim rods position). Those in 9 , 2012 to January, 13 , 2012. 13 2 channel 1-4 and 13-2 were 1.07  10 n/cm .s Experimental results of physical and 12 2 and 8.6110 n/cm .s, respectively. The thermal hydraulics parameters of the reactor experimental error of neutron flux was during start up stages and long operation cycles estimated about 7%. at nominal power showed very good agreement From reaction rate measured by foils with calculated results. On the other hand, irradiation method in neutron trap, neutron experimental results of parameters related to spectrum obtained by SAND-BP computer safety such as peaking factor, axial and radial code. Obtained results of neutron spectrum in neutron flux distribution of reactor core, neutron trap (Fig. 15) showed that comparing negative temperature coefficient, temperature with mixed-core HEU-LEU fuel, when of the reactor tank, temperature at inlet/outlet neutron trap having thinner Beryllium layer, of primary cooling system and secondary thermal neutron flux increased while epi- cooling system,…it could be confirmed that thermal and fast neutron flux decreased with a current core configuration with 92 LEU FAs significant percentage. meets the safety and exploiting requirements. Measured neutron flux at irradiation IV. CONCLUSIONS positions and actual utilization of the After completing design calculation and reactor after full core conversion also preparation, start up of DNRR with entire LEU showed that the reactor core using LEU fuel FAs core was implemented following a is not much different than previous core

detailed plan. As a result, physics and energy using HEU fuel.

.sec

2

Neutron flux/Lethagy, n/cm

Fig. 15. Measured neutron spectrum in neutron trap before and after conversion

44 LUONG BA VIEN et al.

ACKNOWLEDGMENTS REFERENCE

The NRI’s staffs that performed start up [1] P. V. Lam, N. N. Dien, L. V. Vinh, H. T. work of DNRR with entire LEU fuel core Nghiem, L. B. Vien, N. K. Cuong, “Neutronics would like to express sincere gratitude to the and Thermal Hydraulics Calculation for Full leadership of Ministry of Science and Core Conversion from HEU to LEU of the Dalat Nuclear Research Reactor”, RERTR Int’l Technology, Vietnam Atomic Energy Institute, Meeting, Lisbon, Portugal, 2010. Vietnam Agency for Radiation and Nuclear Safety, who have regularly regarded, guided [2] L. B. Vien, L. V. Vinh, H. T. Nghiem, N. K. and created the best condition for us to Cuong, “Transient Analyses for Full Core Conversion from HEU to LEU of the Dalat implement our works. We also express our Nuclear Research Reactor”, RERTR Int’l thanks to Argonne National Laboratory and Meeting, Lisbon, Portugal, 2010. experts from RERTR program (Reduced Enrichment for Research and Test Reactors) [3] “Process of physics and energy start up for full and specialists, professionals in program core conversion using LEU fuel of the Dalat Nuclear Research Reactor”, Nuclear Research RRRFR (Russian Research Reactor Fuel Institute, 2011. Return) has supported in finance as well as useful discussions during design calculation of [4] “Operation logbook of DNRR”, 2011-2012 full core conversion, upgrading equipments and carrying out start up of DNRR.

45 Nuclear Science and Technology, Vol. 4, No. 1 (2014), pp. 46-56 Production of Radioisotopes and Radiopharmaceuticals at the Dalat Nuclear Research Reactor

Duong Van Dong, Pham Ngoc Dien, Bui Van Cuong, Mai Phuoc Tho, Nguyen Thi Thu, Vo Thi Cam Hoa Nuclear Research Institute, 01 Nguyen Tu Luc Street, Dalat City, Vietnam (Received 5 March 2014, accepted 8 March 2014)

Abstract: After reconstruction, the Dalat Nuclear Research Reactor (DNRR) was inaugurated on March 20th, 1984 with the nominal power of 500 kW. Since then the production of radioisotopes and labelled compounds for medical use was started. Up to now, DNRR is still the unique one in Vietnam. The reactor has been operated safely and effectively with the total of about 37,800 hrs (approximately 1,300 hours per year). More than 90% of its operation time and over 80% of its irradiation capacity have been exploited for research and production of radioisotopes. This paper gives an outline of the radioisotope production programme using the DNRR. The production laboratory and facilities including the nuclear reactor with its irradiation positions and characteristics, hot cells, production lines and equipment for the production of Kits for labelling with 99mTc and for quality control, as well as the production rate are mentioned. The methods used for production of 131I, 99mTc, 51Cr, 32P, etc. and the procedures for preparation of radiopharmaceuticals are described briefly. Status of utilization of domestic radioisotopes and radiopharmaceuticals in Vietnam is also reported. Keywords: Radioisotope; Radiopharmaceutical; Labelled KIT, Nuclear Medicine.

I. INTRODUCTION put forward a limited radioisotope production programme to support radioisotope application During the last 30 years of operation, the in medicine, agriculture and industry. For this DNRR has been successfully used for producing objective the core of the present 500-kW many kinds of radioisotopes and reactor reconstructed from the previous 250- radiopharmaceuticals used in medicine and other kW TRIGA MARK II reactor is equipped with economic and technical fields. Providing about more neutron irradiation channels and with a 400Ci per year of radioisotopes including I-131, neutron trap for improving thermal neutron P-32, Tc-99m generator, Kit in-vivo and in-vitro, flux. In addition, the reactor characteristics are Sr-46, Cr-51, etc. Each year, about 300,000 more useful as far as radioisotope production is patients have been diagnosed and treated by concerned, i.e. of higher excess reactivity, the radioisotopes produced at DNRR that contributed cadmium ratio in neutron irradiation channels to push forward the development of nuclear being rather high in the thermal neutron trap medicine in Vietnam. and rather low in the fast neutron channels. In a developing country of low economic The establishment of a laboratory for routine level, the benefit of establishment of a nuclear production of radioisotopes was carefully research center with a research reactor of low considered by balancing the investment power will be recognized by society only when requirement and the production technology of its contributions to social progress become choice, as well as the radioactive waste evident. This point of view has oriented us to treatment problem and radiation protection. ©2014 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute

DUONG VAN DONG et al.

II. PRODUCTION OF RADIOISOTOPE effective irradiation volume of trap will AND RADIOPHARMACEUTICALS increase a factor of 1.5. A. Production laboratory and facilities Construction of neutron irradiation positions Due to the low power of DNRR we must take into consideration of the irradiation position construction, core management and reactor operation mode in order to improve the neutron flux, to maximize the volume available for target irradiation and to balance the formation-decay of activated radionuclides.

For the purpose of augmentation of thermal neutron flux, the central irradiation channel (so called neutron trap) was surrounded by beryllium metal block of thickness of 1.7 cm and height of 60 cm. Fig. 1. Neutron plux depletion in target.

The effect of beryllium gave an improving in flux and quality of thermal neutron. As cited in Figs. 1 & 2, this neutron trap has a diameter of 64 mm originally and has only one guiding tube of diameter of 38 mm in the centre for holding the target 65cm 65cm containers. This construction of neutron trap has been found inconvenient in exhaustive exploitation of irradiation volume. So it has been proposed for reconstruction. The design work is based on the fact of self shielding- effect of targets and cooling water circulation, Old trap New trap as an example of this, neutron flux depletion in TeO and MoO target under reactor irradiation Fig. 2. Neutron trap construction for the 2 3 optimization of effective irradiation was noted in Fig. 1. volume exploitation.

As shown in the case of target sample of Irradiation techniques diameter of 2 cm, the neutron flux in its center The targets held in the quartz ampoule dropped about 10 percent. This fact leads us to were irradiated with thermal neutron either in design a neutron trap which is composed of the neutron trap at the center of the reactor core two channels of 24-mm diameter. The or in the rotary specimen rack. For fast neutron sectional cut of old and new neutron trap was irradiation, it was carried out in a dry channel shown in Fig. 2. With this new construction, inserted between fuel elements of the reactor

47 PRODUCTION OF RADIOISOTOPES AND RADIOPHARMACEUTICALS AT … core. Before irradiation, the targets were radioisotopes of higher specific radioactivity, purified to remove traces of impurities. such as 131I and 99Mo. 99Mo with high specific radioactivity used for 99Mo-99Tcm generator.

99Mo was produced by neutron irradiation of

MoO target at the centre of neutron trap, where 3 thermal neutron flux is of highest value. The

distribution of neutron flux in an irradiation

position is a very important parameter for the

management of target irradiation.

Reactor operation schedule

The schedule of reactor operation mainly Fig. 3. Quartz ampoule and aluminum container depends on the kinds of radionuclide produced for containing target. and their role. The formation rate of these Reactor core management for the irradiation of kinds of radionuclides and the required targets minimal specific radioactivity of radioisotopes The core management plays an are indispensable factors to decide on the important role in the optimization of research option of reactor operation mode. The DNRR reactor utilization for production of was offered to produce some important radioisotopes. The core management is based radionuclides for nuclear medicine application. 131 99 m on the nuclear reaction applied to produce a Among these radioisotopes, I and Tc predescribed radionuclides, the neutron isotopes are most highly evaluated. So the activation cross section and/or requested reasonable schedule of reactor operation must specific radioactivity of a specified radioactive be chosen, taking into consideration of the 131 99m products. Besides, the neutron flux and production yield and quality of I and Tc characteristics of irradiation position such as radioisotope products. Basing on the formation R , neutron flux distribution were also taken rate under reactor activation and half life of cd 99 131 into consideration. Mo and I radionuclides a reactor operation At the DNRR, the irradiation channel of schedule of 130-150 hrs of continuous run every three weeks has been applied. lowest cadmium ratio, Rcd =1.90 is used for fast neutron irradiation to produce the radionuclide with (n, p) nuclear reaction, such as 32S(n, p)32P. 32P isotope produced in this channel is of high specific radioactivity and is used for preparation of injectable 32P solution. Meanwhile the rotary specimen rack of highest cadmium ratio, Rcd = 4.5 is used for production of 32P of low specific radioactivity with 31P(n, γ)32P reaction. This 32P product was used to prepare the 32P applicators for skin disease treatment. In the neutron trap of highest thermal neutron flux and of Rcd = 2.93, the (n, Fig. 4. Annual operation time at DNRR γ) nuclear reaction was applied to produce the since 1984 to 2013.

48 DUONG VAN DONG et al.

B. Production laboratory in 1990 with 2 shielded cells ball-joint The main utilization of the DNRR is the manipulators (Fig. 7). production of radioisotopes for nuclear medicine, agriculture, sedimentology and other scientific research. About 90 percent of time is used for radioisotope production. An area of 200 sq.m is reserved for a rather limited programme of isotope production. The facilities available for the isotope production consist of one hot cell with master slave manipulator (Fig. 5).

Fig. 6. 131I isotope production line equipped in 2008 with 2 shielded cells.

Fig. 5. Hot cell with master slave manipulator

One 131I isotope production line equipped by the IAEA TC Project VIE/0/002 131 in 1987 with 4 shielded cells, one I isotope 99m Fig. 7. Tc generator production line. production line equipped in 2008 by the National Project with 2 shielded cells ball-joint All these facilities are connected with manipulators (Fig. 6), and five shielded fume the existing ventilation system of the reactor. hoods for isotope labelling and -emitted Equipment for the production of Kits to isotope processing. 99m be labeled with Tc isotope and for the One 99mTc generator production line quality control of radioisotopes and (using fission 99Mo solution) equipped radiopharmaceuticals was also supplied by the under the IAEA TC Project No. VIE/6/016 National Projects (Figs. 8 and 9).

49 PRODUCTION OF RADIOISOTOPES AND RADIOPHARMACEUTICALS AT …

- Other radioisotopes such as 60Co, 65Zn, 64Cu, 24Na, 86Rb, 46Sc, 71Ge, 55Fe, etc., were also produced in a small amount when requested. C. Radiochemical processing of activated targets Iodine-131: Iodine-131 is produced from the irradiated tellurium dioxide in neutron trap. The target of tellurium dioxide contained in a welded aluminum capsule, according to the nuclear reaction as follows:

Fig. 8. Sterile hot cell. The irradiated tellurium dioxide powder is transferred to a Vycor distillation vessel and connected to the iodine-131 tellurium processing system. The processing furnace is heated up to 750oC in order to distill the iodine-131 over to a charcoal column trap connected in-line of the distillation system. The charcoal column trap is rinsed with the de- ionized water then eluted with sodium hydroxide 0.05N to form the final product of iodine-131 solution. The scheme in Fig. 10 Fig. 9. Clean room. shows the flow chart of the operation and procedures. Since the beginning of 1984 (the year of reactor inauguration) up to now the The target used in the production is an radioisotope production at the DNRR has analytical grade material of natural tellurium concentrated on the following radionuclides: as tellurium dioxide obtained from Fluka Inc. The chemical purity of the target as TeO2 is - 32P in injectable orthophosphate >95%. The specification of the target before solution and 32P applicator for skin disease being fired in a muffle furnace through therapeutics. analysis by emission spectrograph should - 131 I in NaI solution. contain of selenium less than 0.05% and heavy metals less than 0.1%. After being fired - 99Mo-99Tcm generator. in the muffle furnace the analysis should give 51 - Cr in injectable sodium chromate selenium less than 0.005% and heavy metal solution and Cr-EDTA. less than 0.1%.

50 DUONG VAN DONG et al.

Fig. 10. The flow chart of the operation and procedures of I-131.

Final product specification for use 99mTc generator: The final product as sodium iodide, 131I Among the two reactions of choice for solution in NaOH, without reducing agents will production of 99Mo parent isotope, the large be used as 131I bulk solution for investment for use of 235U(n, fission)99Mo radiopharmaceuticals production. The reaction let us to opt for the 98Mo(n, ) 99Mo specification of the final product is as follows: reaction to produce 99mTc generator. Physical appearance: Colorless solution. In order to separate 99mTc from its parent Radioactivity of 131I: more than 11.1 99Mo we first used the MEK extraction method. GBq (300 mCi) I-131/mL. The inherent disadvantages of this 133I content: less than 0.80% of the 131I method compelled us to start our studies on the content at assay time. preparation of gel type generators in late 1984 pH: more than 11 in the framework of the IAEA-CRP on the “Development of 99mTc generators using low 131 Radionuclidic purity: I content more power research reactor”. This represents the than 99.9%. state-of-the-art for generator technology and Radiochemical purity: Iodide more than 95%. promises opportunities for both developed and

51 PRODUCTION OF RADIOISOTOPES AND RADIOPHARMACEUTICALS AT … developing countries particularly with respect 99Mo could be used to produce portable, to eliminating the need for fission 99Mo. Two chromatographic type 99mTc- generators which directions of preparation of gel type generators have a good performance for application in were studied: clinical investigations. Among the established procedures the column loading procedure was - Preparation of chromatographic highly evaluated, because it proved to be generators using zirconium molybdate (ZrMo) or prominent figures for easy and safe operation, titanium molybdate (TiMo) column packing for low cost of technology facilities, equipment materials synthesized from the neutron irradiated and for the capability to match the traditional molybdenum trioxide and the zirconium chloride technology of the fission 99Mo based 99mTc- and/or titanium chloride, respectively. generator production. - Preparation of chromatographic DNRI had been proposed attending in generators using TiMo column packing material these studies program. The commercial (preformed TiMo) synthesized from the inactive production of PZC generator through the molybdenum compound and TiCl4 and establishment of national project stage 2006- subsequently neutron activated in the reactor. 2008 for the routine production of 99Mo/99mTc In both modes of preparation we have generator. In this project the 99mTc-generator carried out studies on three different options of used PZC coming locally synthesis and from generators: KAKEN - Japan as the column material, 99Mo - The chromatographic generator using formed from MoO3 irradiated at DNRR, the 0.9% NaCl solution as eluant. semi-automatic loading and adsorption machine had studied, designed and installed in the hot - The chromatographic generator using cells available. The generator assembly had also organic solvent as eluant Solid-Solvent- been designed and fabricated, (Figs. 11, and 12). extraction). - The chromatographic generator using dilute saline as eluant and 99mTc concentration column. In the other hand, under the framework of Forum for Nuclear Cooperation in Asia (FNCA) program, the PZC based technology for production of 99mTc- generator has been studied at DNRI as well as FNCA member countries in the past several years. PZC adsorbent of high performance for 99Mo adsorption was easy to synthesize from isopropyl alcohol (iPrOH) and ZrCl4. The procedures and relevant 99mTc- generator designs for the preparation of PZC based 99mTc- generators were successfully set up. The columns of from 1.0 gram to 4.0 gram Fig. 11. Schema of 99mTc – Generator weight of PZC and from 100 mCi to 500 mCi Design of commercial PZC-99mTc generator

52 DUONG VAN DONG et al.

sulfur. Our glass apparatus for this production process is shown in Fig. 13. It can be used for distillation either in the vacuum

or in the N2 gas flow by changing the upper stopper of the distillation vessel. The distillation parameters and post-distillation purification of 32P solution were adopted as described in literature.

Fig. 12. The semi-automatic loading machine

In conclusion, it is strongly believed that ZrMo, TiMo and PZC based generator play an importance role as alternative technology for production of 99Mo/99mTc generator from reaction 98Mo(n, γ)99Mo. However these Fig. 13. The glass apparatus for 32P production methods were not very appropriate for the low process using nuclear reaction 32S(n, p)32P power research reactor as DNRR. Because of those reasons, it is necessary to build a new The 32P applicators for skin disease research reactor with power at least of 10 MW, treatment were produced by neutron irradiation and the neutron flux is high enough to research of a soft plate preformed from cloth binder and and produce radioisotopes. a covering mixture of red phosphorus and glue. Phosphorus-32: After irradiation in the reactor, the radioactive plate was impregnated with plastic and covered 32P isotope was produced according to two with Scotch adhesive. The mechanical nuclear reactions: 32S(n, p)32P and 31P(n, )32P. strength of the preformed plate was not lost The first reaction was used for the under 75-hour irradiation in a thermal neutron 32 production of injectable carrier-free P flux of 5x1012 n.cm-2.s-1. Under this irradiation 32 solution, the second for that of P –isotope a plate containing 65 mg P per square applicators for skin disease treatment. centimeter gives a radioactivity of 15 mCi 32P. First the injectable 32P solution of The absorbed dose rate on the surface of the 2 radioactivity of ten mCi scale was produced plate of size 50 x 40 mm was measured as 110 -1 -1 from irradiated MgSO4 target using magnesia Rad.min at the center and 75 Rad.min on the as absorbent to separate 32P isotope from edge. Medical doctors’ experience over ten

MgSO4 solution. In the case of Ci scale years showed that with repeated treatment of production, the large amount of waste three or five 15-minute applications the produced from this technology caused following diseases will be cured: Eczema, skin storage problems. Recently, we have cancer, bump scar, etc. At present more than introduced the distillation technique to 75 Ci 32P in applicator form are used annually separate 32P from reactor irradiated elemental in the country.

53 PRODUCTION OF RADIOISOTOPES AND RADIOPHARMACEUTICALS AT …

Cr-51 isotope: chromatography techniques for chemical The production of 51Cr isotope was purity, and the spectrometry and neutron carried out based on the Szilard-Charmel activation analysis for chemical purity. Biodistribution assay, biological tests reaction using reagent grade K2CrO4 target. The chemical separation of recoiled 51Cr (apyrogenity, sterility, toxicity) and nuclide was based on the selective adsorption physicochemical tests (pH, turbidity) are also of this isotope on an inorganic ion exchanger carried out regularly. Si-ZrP (Silica gel supported zirconium - phosphate) synthesized by us. Other isotopes were also produced when requested in small amounts for industrial and agricultural applications. The methods for production of these isotopes were selected from investigation results or different reference sources. D. Production of Kits for labelling with 99mTc: In furthering the application of 99mTc isotope, the local availability of Kits for labelling with 99mTc plays an important role. Fig. 14. HPLC system for QC. With IAEA support the basic equipment for the III. LOCAL PRODUCTION VOLUME production of Kits has been installed in our AND DEMAND laboratory. At present many kinds of in-vivo Kits have been successfully prepared and put These types of radioisotopes have to use in the country, they are Phytate, regularly been supplied to more than 25 Gluconate, Pyrophosphate, Citrate, DMSA, hospitals in Vietnam two times per month. The HIDA, DTPA, Maccroaggregated HSA and 1311 radioisotope labelled radiopharmaceuticals EHDP (1-hydroxy ethylidene-1, 1-disodium- such as 131I-Hippuran; 131IMIBG have also phosphate). been regularly supplied to hospitals. The studies on the preparation of Radioisotope production rate is shown in Fig. Radioimmuno-assay Kits and therapeutic 15 and Table I. agents and/or radionuclides were also carried In order to support the application of out. The future production of the above 99mTc, 113mIn, 177Lu and 153Sm radioisotopes in mentioned items is foreseen and planned. clinical diagnosis and therapeutics, the E. Quality control preparation of radiopharmaceuticals in Kit Radioisotope and radiopharmaceutical forms has been carried out. The following Kits quality control was carried out for all batches have regularly been manufactured in DNRI: of our products. The gamma spectrum analysis Phytate, Gluconate, Pyrophosphate, Citrate, using Ge-Li detector coupled with a DMSA, EHIDA, DTPA, HSA multichannel analyser is used for radionuclide macroaggregated, HEDP, HmPAO, MIBI, purity control, the TLC, HPLC and gel- MDP.

54 DUONG VAN DONG et al.

Radioimmunoassay Kits: The RIA Kit domestic market. Other RIA and IRMA Kits production and distribution programme have can be supplied to end-users by dispensing also started. T3 and T4 Kits have been selected process based on the contract. locally by end-users with a share of 50% of

Fig. 15. Total activity of radioisotopes produced Fig. 16. Radioisotopes and Radiopharmaceuticals at DNRI produced at the DNRI Table I. The supply/demand for radioisotopes and diagnostic Kits in Vietnam.

Product Supply Demand at present 131I- Diagnostic and 20-30 Ci/month 40-50 Ci/month therapeutic capsule/solution 99mTc-Generator 10 generators (200- 40 generators (200- 500mCi/each)/month 500mCi/each)/month

32P-Solution/ 50 Ci/month 50 Ci/month Applicator Kits for 99mTc-Labelling - MDP 400 Kit/ month 500 Kit/ month - DTPA 200 Kit/ month 300 Kit/ month - DMSA 200 Kit/ month 300 Kit/ month - PHYTATE 200 Kit/ month 300 Kit/ month - Orther 200 Kit/ month 300 Kit/ month

IV. THE APPLICATION OF LOCAL - There are six centres of PET-CT and PRODUCTS IN THE COUNTRY cyclotron in Hanoi Capital and Ho Chi Minh City. - Number of nuclear medicine departments in Vietnam: 25 - Radiopharmaceuticals used in these centres: Na131I solution and capsule, Sodium- These departments almost are located in (99mTc) pertechnetate (99mTc-Generator) 131I- the big cities of the country (Fig. 17). Hippuran, Sodium-(32P) orthophosphate, 131I- - Number of gamma cameras (planar MIBG, In-vivo Kits (MDP, DTPA, DMSA, and SPECT): 22

55 PRODUCTION OF RADIOISOTOPES AND RADIOPHARMACEUTICALS AT …

Phosphon, Glucon, Phytate, MAHSA, EHIDA, ordination Meeting, Bandung, Indonesia, HMPAO, MIBI, MAG-3, etc.). (October 1987). - Locally manufactured products take [2] Radioisotope production and quality control. about 50% of total market. In order to get a Technical Reports Series No. 128. IAFA, Vienna, (1971). higher market share now we increase the production by loading generations with [3] Le Van So, Investigation on the silica gel importing raw materials such as 99Mo and 131I supported form of micro crystalline zirconium- solutions. phosphate ion exchanger and its applications in chemical separation. I.- Preparation, ion exchange properties and stability of Si-ZrP, J. Radioanal. Nucl. Chem., (Articles) 9 (1) 17-30 (1986). [4] Le Van So, Richard M. Lambrecht, Development of alternative technologies for a gel-type chromatographic 99mTc generator. J. Labelled Compd. Radiopharm. 35:270 (1994). [5] Ngo Quang Huy et al, Reactor physics experimental studies on Dalat nuclear research reactor, 50A-01-04 Research Project Final Report, (1990) (in Vietnamese). [6] Tran Ha Anh et al, Studies on Dalat Nuclear Reactor Physics and Technique and on Measures to ensure the safety and efficiency of the reactor, KC-09-15 Research Project Final Report, (1995). [7] Nguyen Nhi Dien, Dalat nuclear research reactor - status of operation and utilization, Dalat Sym. -RR-PI-05, Dalat, (2005). [8] Duong Van Dong, Status of Radioisotope Fig. 17. Location of Nuclear Medicine Departments Production and Application in Vietnam, Dalat in Vietnam. Sym. -RR-PI-09, Dalat, (2009).

REFERENCES [9] Duong Van Dong, Status of the study on PZC based Tc-99m generator and potential of its 99m [1] Le Van So, Production of Tc isotope from the commercial production in Vietnam, Nihon chromatographic generator using zirconium- Genshiryoku Kenkyu Kaihatsu Kiko JAEA- molydate and titanium-molybdate targets as Conf, Journal Code: L2150A, page 25-29 column packing materials. Research Co- (2007).

56 Nuclear Science and Technology, Vol. 4, No. 1 (2014), pp. 57-61 The gamma two-step cascade method at Dalat Nuclear Research Reactor

Vuong Huu Tan1, Pham Dinh Khang2, Nguyen Nhi Dien3, Nguyen Xuan Hai3, Tran Tuan Anh3*, Ho Huu Thang3, Pham Ngoc Son3, Mangengo Lumengano4 1)Vietnam Agency for Radiation and Nuclear Safety, 113 Tran Duy Hung, Hanoi, Vietnam 2) Vietnam Atomic Energy Institute, 59 Ly Thuong Kiet, Hanoi, Vietnam 3) Nuclear Research Institute, 01 Nguyen Tu Luc, Dalat, Vietnam 4) Agostinho Neto University, Av, 4 Fevereiro, 71 Ingombotas, Luanda, Angola *Email: [email protected] (Received 7 March 2014, accepted 13 March 2014)

Abstract: The event-event coincidence spectroscopy system was successfully established and operated

on thermal neutron beam of channel N0. 3 at Dalat Nuclear Research Reactor (DNRR) with resolving time value of about 10 ns. The studies on level density, gamma strength function and decay scheme of intermediate-mass and heavy nuclei have been performed on this system. The achieved results are opening a new research of nuclear structure based on (n, 2) reaction. Keywords: event-event coincidence, thermal neutron beam, nuclear structure.

I. INTRODUCTION In this work, the gamma two-step cascade (TSC) method has been developed to The nuclear parameters obtained from optimize solution and to reduce Compton intensities of two-step cascades have scatter and pair-production phenomena in the considerably higher reliability than those gamma spectra of nuclei decay gamma obtained within known methods due to cascades. This is allowed to determine unsuccessful relation between the experimental precisely gamma cascade intensities and to find spectra and desired parameters of the gamma- intermediate levels in an energy region near a decay process. For excited levels below 2 MeV, binding energy. Since, the transition their spectroscopic information in detail were probabilities and quantum characteristics of known very well from investigations of (n, ), intermediate levels are split. The characteristics (n, e), (d, p)... reactions. However, for higher allow comparing transition probabilities excited levels, the information is not enough between theory and empirical results [2]. because of low intensity of transitions and bad resolution of detectors [1]. II. TSC METHOD The traditional gamma spectrometer The method is based on event-event allows getting more information about nuclear coincidence measurements of two γ-rays from data and nuclear structure from their spectra. The the cascade decay of a compound nucleus background, however, is high due to Compton following thermal neutron capture. The total scattering. In order to reduce the background, it energies of the γ-rays and their time is necessary to develop advanced spectrometers differences are measured by two germanium such as Compton suppression, pair production, detectors. Coincidence events are selected or coincidence systems. which have a sum energy given by the energy

©2014 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute

THE GAMMA TWO STEP CASCADE METHOD AT DALAT RESEARCH REACTOR difference between the capture state and the spectrometer. The detectors were shielded by pre-selected low-lying state. The detected lead blocks of 10 cm in thickness. The distance spectrum then contains information on two between the source and the detectors’ surfaces types of transitions. The 1st type includes first is 4 cm. In order to decrease the back scattered transitions populated in the intermediate gamma rays and filter out X-ray, two lead region of excited energy. Because of large plates of 2 mm in thickness were placed in number of levels in this region, no front of the detectors and sample. The spectrometer is available for data acquisition. background count rate was less than 600 counts The 2nd one includes transitions that the per second (cps) in 0.2 ÷ 8 MeV range [5]. intermediate levels dominate low energy Data acquisition system levels [2, 3, 4]. In this case, the event-event coincidence spectroscopy can be used in The electronics configuration used in advance for level densities determination. those - coincidence experiments is shown in Figure 1. III. TSC GAMMA MEASUREMENT The detector signals are amplified with Neutron beam and detectors 572 amplifier (AMP) modules with a shaping arrangement time of 3.0 µs and about 1 keV per channel. The experiment system has been The output signals of the amplifiers are installed at the tangential beam port of the digitized by 7072 analog-to-digital converter DNRR. The thermal neutron beam was (ADC) modules. The timing signals of both moderated by Si filter. The neutron flux, the detectors are put through 474 timing filter cadmium ratio and the neutron beam diameter amplifier (TFA) modules. 5 -2 -1 at the sample position were 2.410 n.cm .s , The shaped and amplified timing 230 and 1.5 cm respectively. signals by 474 TFA are plugged into 584 CFD Two horizontal GMX35 detectors modules, which are used in slow rise time manufactured by ORTEC with the energy rejection option (SRT) mode. The CFD output resolutions of 1.9 keV at 1332 keV (60Co) signal of the first channel is used as 556 time- have been used in the - coincidence to-amplitude converter (TAC) start signal.

Fig. 1. The - coincidence electronics.

58

NGUYEN XUAN HAI et al.

The CFD output signal of the second appearing in the corresponding coincidence channel is delayed 100 ns and served as a TAC data file. The coincidence spectrum of one stop signal. detector with the chosen peak in another detector can be created by the same procedure. The full scale of TAC is set at 100 ns, They are coincidence spectra between high- and output signal is digitized in 8713 ADC energy primary and low-energy secondary with selection of 1024 channels for a 10 V transitions or among the low-energy secondary input pulse. The TAC “Valid Convert” signal transitions as obtained in the work [3, 4]. is used to gate 7072 ADCs, and the delay or Besides, the summation spectrum of amplitudes synchronizing with AMP output signal is of coincidence pulses can be created by implemented by interface software. Recorded summation of pairs of coincidence data. Every coincident events have three values, including full-peak in the summation spectrum is coincidence gamma-ray energies from detector 1, detector 2 and time interval between two γ- corresponding to the - cascade decays from rays in a pair event [5]. The resolving time for the capture state to the determined low-lying this configuration is about 10 ns with 60Co excited level. The TSC spectrum of one source measurement (see Figure 2). detector associated with the defined energy (E) summation peak will be taken by choosing Coincidence Data Processing pairs of coincidence data having summation in In the experiment, the data, which the range of E ± E (with E/E ≤ 0.005) (see contains all pairs of - coincidence data from Figure 3). The TSC spectrum gives information two HPGe-detectors, were stored in the on levels in the region between the capture state memory of computer. Indeed, that is pairs of and the defined E low-lying level. From all channel numbers associated with energies of obtained TSC spectra we can build up the decay - coincidence pairs. The coincidence scheme of the investigated nucleus on the base spectrum of each detector can be created from of methods and the criteria given in Ref. [5]. the corresponding data file by the procedure The measured values of gamma two-step 35 that the count number of each channel of the cascade energies and intensities of Cl(nth, 36 spectrum is equal to times of that channel 2γ) Cl reaction were shown in Table 1.

500

5000 E1+E2 = 8579 keV 1164.87 keV 1164.87

4000 400 keV

3000 300

788.43

1959.36 keV 1959.36

Counts 2000 10ns Counts

200 keV 6627.75

7413.95 keV 7413.95 2863.82 keV 2863.82 1000 keV 7790.32

100 keV 5715.19

1601.08 keV 1601.08

3061.86 keV 3061.86

5517.2 keV 5517.2

5902.7 keV 5902.7

2676.30 keV 2676.30 6977.85 keV 6977.85 0

0 0 10 20 30 40 2000 4000 6000 8000 Resolving time (ns) Energy keV Fig. 2. The resolving timing spectrum Fig. 3. The TSC spectrum of 36Cl belongs to final level from 8579 keV.

59 THE GAMMA TWO STEP CASCADE METHOD AT DALAT RESEARCH REACTOR

35 36 Table 1. The gamma two-step cascade energies and intensities of Cl(nth, 2γ) Cl reaction. Measured values XCI 6/18/013

Eγ Up level Low level Eγ Up level Low level I- (keV) (keV) (keV) (keV) (keV) (keV) 787.03 1952.98 1164.01 786.30 1951.20 1164.89 10.520 1164.01 1952.98 787.03 1162.78 1951.20 788.44 2.290 1370.00 3331.99 1958.98 1372.86 3332.32 1959.41 0.384 1958.98 1958.98 0.00 1959.36 1959.41 0.00 12.560 1164.01 1164.01 0.00 1164.87 1164.89 0.00 27.20 3723.00 4886.09 1164.01 3723.00 N/A 517.05 517.05 0.00 517.08 2468.28 1951.20 24.300 1950.98 2465.97 517.05 1951.14 1951.20 0.00 19.390 789.03 789.03 0.00 788.43 788.44 0.00 16.320 1164.60 1164.60 0.00 1164.87 1164.89 0.00 27.20 1601.49 1601.49 0.00 1601.08 1601.12 0.00 3.484 1958.48 1958.48 0.00 1959.36 1959.41 0.00 12.560 2864.28 2864.28 0.00 2863.82 2863.96 0.00 5.770 7413.06 8579.71 1165.01 7413.95 8579.70 1164.89 10.520 6979.37 8579.71 1602.99 6977.85 8579.70 1601.12 2.290 3062.98 8579.71 5518.16 3061.86 8579.70 5517.76 3.521 5518.16 5518.16 0.00 5517.2 5517.76 0.00 1.689 6621.31 8579.71 1957.98 6619.64 8579.70 1959.41 7.830 5716.18 8579.71 2863.98 5715.19 8579.70 2863.96 5.310 788.23 788.23 0.00 788.43 788.44 0.00 16.32 1950.17 1950.17 0.00 1951.14 1951.20 0.00 19.39 6629.20 8579.71 1950.17 6627.75 8579.70 1951.20 4.690 7792.32 8579.71 788.23 7790.32 8579.70 788.44 8.310

IV. RESULTS - Determining the lifetime level, width level and gamma transition strength from the Within the framework of this research experimental data of gamma intensity and project, the obtained results are as follows: electromagnetic transfer selection. - Setting up successfully the event- - Providing methods and experimental event coincidence spectrometer with for facilities for basic researches, education and measuring nuclear structure data on thermal training. neutron beam. - Measuring and analyzing the V. CONCLUSION gamma cascade transition data for nuclei of The γ-γ coincidence spectrometer is a 239U, 182Ta, 153Sm, 172Yb, 59Ni, 55Fe and 49Ti. useful tool in research on nuclear spectroscopy in The experimental data are to evaluate excited states in the intermediate energy DNRR. Besides, the spectrometer can also be below the neutron binding energy. used in research on the lifetime of some excited states and γ-γ angular correlations that are - Evaluating nuclear structure for completely new research fields. For some those nuclei based on analyzed data and elements in the deformed nuclei region with high theoretical models. possibility of cascade transitions, this

60 NGUYEN XUAN HAI et al. spectrometer can be used for the neutron REFERENCES activation analysis because of very low [1] A. A. Vankov et al. In Proc. Conf. on Nuclear gamma backgrounds. Data for Reactors. Helsinki 1970, IAEA, Vienna, The research method and facilities Vol.1, p.559 (1970). for TSC measurements will play a [2] H.H. Bolotin. Thermal-neutron capture gamma- significant role in carrying out R&D gamma coincidence studies and techniques, programs of nuclear technique applications Proceedings of the 1981 International Symposium so far, as well as in preparing human on Neutron Capture Gamma Ray Spectroscopy and Related Topics, Grenoble, France, p.15-34 resources for the nuclear data program in (1981). Vietnam in the near future. [3] S.T. Boneva et al. Two-step cascades of neutron radiative capture: 1. The spectroscopy of excited ACKNOWLEDGMENTS states of complex nuclei in the range of the The authors would like to express neutron binding energy, Physics of Elementary Particles and Atomic Nuclei, Vol.22, Part.2, their sincere thanks to the researchers of p.479-511 (1991). DNRR for their cooperation concerning to [4] S.T. Boneva et al. Two-step cascades of neutron neutron irradiations. This research is funded radiative capture: 2. Main parameters and by Ministry of Science and Technology, peculiarities complex nuclei compound-states - Vietnam Atomic Energy Institute and decay, Physics of Elementary Particles and Nuclear Research Institute. Atomic Nuclei, Vol.22, Part.6, p.1431-1475 (1991).

[5] Vuong Huu Tan et al. Investigation of gamma cascade transition of 153Sm, 182Ta, 59Ni and 239U

using the gamma two step cascade method, Final report of the research project, Ministry of

Sciences and Technology, Code BO/05/01/05, (2005-2006).

61

Nuclear Science and Technology, Vol. 4, No. 1 (2014), pp. 62-69 Progress of Filtered Neutron Beams Development and Applications at the Horizontal Channels No.2 and No.4 of Dalat Nuclear Research Reactor

Vuong Huu Tan1, Pham Ngoc Son2*, Nguyen Nhi Dien2, Tran Tuan Anh2, Nguyen Xuan Hai2 1Vietnam Agency for Radiation and Nuclear Safety, 113-Tran Duy Hung, Hanoi 2Nuclear Research Institute, 01 Nguyen Tu Luc, Dalat *E-mail: [email protected] (Received 5 March 2014, accepted 12 March 2014)

Abstract: The neutron filter technique has been applied to create mono-energetic neutron beams with high intensity, at the horizontal channels No.2 and No.4 of the Dalat nuclear research reactor. The mono-energetic neutron beams that have been developed for researches and applications are thermal (0.025eV), 24keV, 54keV, 59keV, 133keV and 148keV. The relative intensities of main peak in filtered neutron energy spectra and the collimated neutron fluxes at the sample irradiation positions are 90  96% and 2.8×105  7.8×106 n/cm2.s, respectively. Monte Carlo simulations and transmission calculations were performed to each neutron energy beam for optimal design of geometrical structure and neutron filter materials. These filtered neutron beams have been applied efficiently for experimental researches on neutron total and capture cross sections measurements, and elemental analysis in various kinds of samples based on the prompt gamma neutron activation analysis method. This paper reviews the progress of filtered neutron beams development and its applications for past many years at the Dalat nuclear research reactor. Keywords: Filtered neutron beam, nuclear data measurement, Dalat nuclear research reactor

I. INTRODUCTION neutron induces nuclear reaction data measurements since 1991 [2]. For efficient and The Dalat nuclear research reactor extensive uses of the neutron channel, the (DNRR), located in campus of the Nuclear neutron filter technique has been also applied Research Institute, VINATOM, was originally to create high intensity neutron beams with a TRIGA MARK II reactor with a nominal quasi-monoenergies of 24keV, 59keV and power of 250kW completed construction and 133keV at the channel No.4 in 2008 [3]. reached critical state in 1963. The reactor then has been upgraded to nominal power of 500 kW since 1984. There are three radial and one tangential beam ports at DNRR, each of which penetrates the concrete shield structure and the reactor water to provide external beams of neutron originated from reactor core [1]. The cross section view of horizontal channels of DNRR is shown in Fig.1. The radial beam port No.4 has been used to develop mono-energetic neutron beams of thermal, 54keV and 148keV (previous reported as 55 and 144keV) by the Fig. 1. Structure of horizontal neutron channels of neutron filter technique for basic research on the Dalat nuclear research reactor

©2014 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute

VUONG HUU TAN, PHAM NGOC SON, NGUYEN NHI DIEN, TRAN TUAN ANH, NGUYEN XUAN HAI In order to enhance the utilizations of II. FILTERED NEUTRON BEAMS DNRR for neutron capture experiments and The optimal design structure for prompt gamma-rays neutron activation insertion of filters into the horizontal channel, analysis (PGNAA) applications, the beam port beam collimators and radiation shielding No.2 of the reactor have been opened in chamber at the Dalat nuclear research reactor is advance since 2010 [4] for development of a shown in Fig.2 [4]. The neutron energy spectra modern prompt gamma-ray Compton after filtered through a suitable composition suppression facility used a high efficient materials used as filters can be calculated as the HPGe-BGO detectors system. In the works of following expression: neutron filters development, the Monte-Carlo , (1) simulation used MCNP5 code, and (E ) 0, ( E )*exp( k d k  t k ( E )) transmission calculation by CFNB code [5] k have been performed for each neutron beam where 0 (E) ,  (E) are energy for optimal design of geometrical structure, distributions of the neutron spectra before and neutron filter materials and radiation after transmitted through the filters; k, dk and shielding. t,k(E) are the mass density, length of filter and th The applications of these filtered total cross section of k filter material, neutron beams were mainly focus on nuclear respectively. The filter information and data measurements and PGNAA elemental physical parameter of each energy beam is analysis, although these beam lines have presented in the following sub-sections. possibility for many other researches and applications such as nuclear level density and isomer ratio determination, Boron neutron capture therapy (BNCT) research, neutron dosimeter calibrations,... On the nuclear data measurements respects, the channels provide essential neutron beams for precise experimental reaction data of neutron total and capture reaction cross sections. On the Fig. 2. The design structure of filtered neutron beam PGNAA application subject, the assessment of facility at the channel No. 2 of DNRR analytical sensitivity for elements of B, H, Hg, Si, Ca, C, S, Al, Fe, Cl, Ti,... has been carried The thermal neutron beams: out, and shown that the new PGNAA The thermal neutron beam at the channel spectroscopy installed at the channel No.2 is a No.4 was developed in 1991 [2]. The material good facility supplemented to the neutron compositions of filters are 98cm Si, 1cm Ti and activation analysis (NAA) method at the Dalat 35g/cm2 S. The measured thermal neutron flux 6 2 nuclear research reactor. The detail is 1.710 n/cm .s, and Cadmium ratio Rcd(Au) characteristics of filtered neutron beams = 112 [3]. In order to enhance the utilizations development and results of its applications are of the Dalat research reactor for researches and presented in the next sections. applications based on the neutron capture

63 PROGRESS OF FILTERED NEUTRON BEAMS DEVELOPMENT AND APPLICATIONS AT … reactions, the well thermal neutron beam at the nuclear research reactor from 1990s [2, 3]. channel No. 2 has been developed and serviced Firstly, the two neutron beams with mono- since 2011 [4]. The neutron filters for this energies of 54keV and 184keV were created at 0.0253eV neutron beam line are single crystals the channel No.4, and provided a good of 80cm Si and 6cm Bi. The measured thermal experimental station for basis researches on neutron flux at outer position of the beam line reactions of neutron with material in keV 6 2 is 1.610 n/cm .s, and the value of Cadmium energy region. The filter information and ratio Rcd(Au) is 420. physical parameters of these neutron beam The neutron beams of 54keV and 148keV: lines are introduced in Table I [3], and the corresponding neutron spectra are shown in The neutron filter technique has been Figs. 3-4. applied at the horizontal channels of Dalat

Table I. Physical parameters of the 54keV and 148keV neutron beams at the channel No.4

Parameters 54keV 148keV

Neutron flux (n/cm2.s) 6.7x105 3.9x106

Energy resolution (keV) 1.5 14.8

Peak relative intensity (%) 78.05 95.78

Beam collimated diameter 3 cm 3 cm

B 0.2g/cm2 B 0.2g/cm2 Filter compositions Si 98cm Si 98cm S 35g/cm2 Ti 1cm

200 4000 Exp. data Unfolding spectrum 14000 180 Fitted line Transport calculation Intensity 160 12000 3000 140 54keV 10000

Intensity (a.u) Intensity 120 8000 100 2000

80 Counts 6000

60 Relativeintensity 4000 1000 40 148keV

20 2000 0 0 0 8.0E+04 1.2E+05 1.6E+05 2.0E+05 0 200 400 600 800 1000 1200 Neutron energy (eV) Channel

Fig. 3. Energy spectrum of the 148keV neutron beam at Fig. 4. Measured energy spectrum of the 54keV neutron the channel No.4 of DNRR beam at the channel No. 4 of DNRR

64

VUONG HUU TAN, PHAM NGOC SON, NGUYEN NHI DIEN, TRAN TUAN ANH, NGUYEN XUAN HAI The neutron beams of 24keV, 59keV and 133keV: [3] based on the neutron source from the channel No. 4 of DRR. The characteristics of As progressive necessary of reactor these neutron beam lines are introduced in based mono-energetic neutron beam lines for the references [6, 7], and summarized in experimental researches on neutron Table II. The calculated and measured energy interaction with mater, the new three filtered spectra of these neutron beam lines are neutron beam of 24keV, 59keV and 133keV shown in Figs. 5-7. have been developed and applied from 2008

Fig. 5. Measured neutron spectrum for 24keV beam by Fig. 6. Calculated neutron spectrum for the 59keV proton recoil proportional counter filtered neutron beam

Table II. Characteristics of the 24keV, 59keV and 133keV neutron beams at the channel No.4

Parameters 24keV 59keV 133keV Neutron flux (n/cm2.s) 6.1x105 5.3x105 3.2x105 Energy resolution (keV) 1.8 2.7 3.0 Peak relative intensity (%) 96.72 92.28 92.89 Beam collimated diameter (cm) 3 3 3 B 0.2g/cm2 B 0.2g/cm2 B 0.2g/cm2 Fe 20cm Ni 10cm Cr 50g/cm2 Composition of Filters Al 30cm V 15cm Ni 10cm S 35g/cm2 Al 5cm Si 60cm S 35g/cm2

1.0x103 CFNB MCNP

8.0x102

6.0x102

2

Intensity (a.u) 4.0x10

2.0x102

0.0 5.0x10-2 1.0x10-1 1.5x10-1 2.0x10-1 E (MeV) n

Fig. 7. Calculated neutron spectrum for the 133keV filtered neutron beam

65 PROGRESS OF FILTERED NEUTRON BEAMS DEVELOPMENT AND APPLICATIONS AT …

III. NUCLEAR DATA MESUREMENTS 193Ir [9]; 185Re and 187Re [10, 11]. In addition, the horizontal thermal Column of DNRR, a Neutron capture cross section measurements: well thermalized neutron channel, has been The measurements of neutron capture also used for measurement of thermal neutron cross sections for a number of nuclides have capture cross section and resonance integral of been performed on the filtered neutron beams 69Ga and 71Ga [12]. A typical result of our with mono-energies of 24, 54, 59, 133 and 148 measurements in comparison with data from keV, at the Dalat nuclear research reactor. The other laboratories is shown in Fig. 7 [10]. measured neutron capture cross sections data 10 were obtained relative to the standard capture ENDF/BVII 185 186 197 198 Re(n,γ) Re YU.N.TROFIMOV cross sections of the Au(n,) Au reaction M.LINDNER S.J.FRIESENHAHN by the activation method. An abridged A.K.CHAUBEY R.P.ANAND description of data analysis procedure is A.A.BERGMAN This w ork presented as follows: 1 The average capture cross sections,

x Cross section (barn) <a> , for nuclide x at average neutron spectrum <> can be determined relative to 197 that of Au standard by the following 0.1 1.E+04 6.E+04 1.E+05 2.E+05 relations: Neutron energy (eV)

x x x Au Au Au Au 185 C f (,t) fc I  N   a  (2) Fig. 7. Neutron capture cross section of Re [10]   a  Au Au Au x x x ; C f (,t) fc I  N Measurements of neutron total cross sections:  , (3) The total neutron cross section f (,t)  t t t (1 e 1 )e 2 (1 e 3 ) measurements are being carried out by the transmission method for natural elements of U, where the superscript „x‟ denotes sample C, Fe and Al, at the filtered neutron energies of nucleus, and „Au‟ denotes the reference 24keV, 54keV, 59keV, 133keV and 148keV. nucleus 197Au. „C‟ stands for net counts of The experimental value of neutron total cross the corresponding gamma peak. „t ‟, „t ‟ and 1 2 section,  , can be exactly determined from the „t ‟ are irradiating, cooling and measuring t 3 following expression: times, respectively. „λ‟ is decay constant of the 1 1 1  product nucleus; „εγ‟ is the detection efficiency  0 , (4)  t ln ln  of detector; „Iγ‟ is the intensity of interesting γ- d T d  ray, and „f ‟ is the correction factor for self- c where „T‟ is transmission coefficient of shielding multiple scattering effects that can be the collimated neutron beam that transmitted exactly calculated by the Monte Carlo method. through a purity sample with thickness d In recent years, we have conducted a (cm); „‟ denotes density of the sample series of cross section measurements for 3 (Atom/cm ). „0‟ and „‟ are measured neutron capture (n, ) reactions in different neutron fluxes at before and after positions of nuclides, and reported in scientific papers such the irradiating sample, respectively. A 109 186 158 139 152 191 as: Ag, W, Gd [8]; La, Sm, Ir, measurement of the transmission spectrum for

66 VUONG HUU TAN, PHAM NGOC SON, NGUYEN NHI DIEN, TRAN TUAN ANH, NGUYEN XUAN HAI 54keV neutron beam at the channel No.4 of of BGO-HPGe detectors was completely DNRR is shown in Fig.8. developed and installed at the experimental space of this beam port, from 2012. A preliminary study on calibration and analytical sensitivity for several domination elements has been conducted. A prompt gamma-ray spectrum of geometrical sample measured in single and Compton-suppression modes is shown in Fig.9. The results in this study allows us to estimate that the new PGNAA facility installed at the channel No.2 of DNRR is qualified to participate in analytical services at

the Institute. A calibration curve for Boron Fig. 8. Measured neutron spectrum of 54keV analysis is presented in Fig.10, and the results neutron beam transmitted through different of comparison analysis used standard soil thickness of C sample. sample (NIST-2711a) is given in Table III. IV. PGNAA APLICATIONS

From 1998, the 148keV filtered neutron beam at the channel No.4 has been applied for possibility studies the method of in-vivo prompt gamma neutron activation analysis (IVPGNAA) that involves the exposure of the living human organs to a small dose of neutrons. At that time, IVPGNAA is a new technique for directly determination of toxic elements accommodated in a specific living Fig. 9. Prompt gamma-rays spectrum measured at human organ such as concentrations of Hg in the thermal neutron beam No.2 for soil sample, in kidney and Cd in liver. The research was single and Compton-suppression modes. carried out on a physical phantom installed at Model Line Exp data Equation y = A + B*x 25 Reduced 0 Linear fitting the channel No.4 [13]. The results given from Chi-Sqr Adj. R-Square 1 Value Standard Error this investigation introduced a high effective cps A -2.799 0.158620 20 cps B 0.520 0.00544162 new experiment with 148keV neutron beam 15

instead of thermal neutrons [13].

cps The low background and well thermal 10 filtered neutron beam from the channel No.2 5 [4] of Dalat nuclear research reactor is an 0 advantage neutron source for prompt gamma- ray neutron activation analysis (PGNAA). 10 20 30 40 50 g B Accordingly, a modern Compton suppression Fig. 10. The calibration curve for Boron analysis by PGNAA spectroscopy used a compact system using the PGNAA facility at the channel No.2

67 PROGRESS OF FILTERED NEUTRON BEAMS DEVELOPMENT AND APPLICATIONS AT …

Table III. The results of comparison analysis used the standard soil sample (NIST-2711a), by using the PGNAA facility at the channel No.2 NIST-2711a Sample (Standard sample: Montana soil)

Elements Measured values Reference values

B (g/g) 50.5 ± 2.9 50 Gd (g/g) 7.59 ± 3.34 5 Sm (g/g) 6.96 ± 1.07 5.93 ± 0.28 Ca (%) 2.43 ± 0.59 2.42 ± 0.06 Al (%) 7.1 ± 0.3 6.72 ± 0.06 Si (%) 31.66 ± 3.93 31.4 ± 0.7 K (%) 2.39 ± 0.28 2.53 ± 0.10 Ti (%) 0.29 ± 0.06 0.32 ± 0.01 Na (%) 2.02 ± 0.48 1.20 ± 0.01 Fe (%) 3.01 ± 0.35 2.82 ± 0.04

V. CONCLUSIONS supplementation to the neutron activation analysis (NAA) method at the Dalat research The accomplishment of research reactor. activities on the topics of filtered neutron beams development and it‟s applications based The new development of neutron beam on the neutron sources from the horizontal with possible mono-energy of 2keV, and channel No.2 and No.4 of Dalat nuclear extension of application studies such as Boron research reactor is reviewed in this report. The neutron capture therapy (BNCT) and neutron neutron filter technique has been effectively dosimeter calibration are proposed. applied to provide mono-energetic neutron beam lines with qualified characteristics for ACKNOWLEDGEMENTS related applications at the Nuclear Research This research is partly funded by Institute, VINATOM. The basis researches on Vietnam National Foundation for Science experimental neutron induce nuclear reaction and Technology Development cross sections conducted by using these (NAFOSTED) under grant number “103.04- neutron beams have been performed with 2012.59”. The authors are immensely interesting results, and this research activity is grateful to Mr. Luong Ba Vien, Deputy proposed to be continued, in order to Director of the Nuclear Research Institute, participate in providing of precise experimental VINATOM, for his great encouragement nuclear reaction data and educational and critical reading of the manuscript. experiments. The new PGNAA facility installed coupling with the well thermal neutron beam at the channel No.2 plays as an important application of this channel for studies on neutron capture experiments and elemental analysis. This will be an important

68 VUONG HUU TAN, PHAM NGOC SON, NGUYEN NHI DIEN, TRAN TUAN ANH, NGUYEN XUAN HAI REFERENCES [8] Vuong Huu Tan, Pham Ngoc Son, et al. Neutron Capture Cross Section Measurements [1] General Atomic, Triga Mark II Reactor – of 109Ag, 186W and 158Gd on Filtered Neutron General Specifications and Description. GA- Beams of 55keV and 144keV. Nuclear Science 2627 (1961). and Technology, Vol.3 No.1, pp.1-7; IAEA, [2] Vuong Huu Tan. Application study on the Nuclear data section, INDC-VN-011, (2004). reactions induced by neutron, gamma and [9] Vuong Huu Tan, Pham Ngoc Son, et al. charge particles based on the available nuclear Measurement of Neutron Capture Cross facilities in Vietnam. State scientific project 139 152 191,193 Section of La, Sm and Ir at 55keV report, code: KC-09-08A (1994). and 144keV. Proc. of Symposium on Nuclear [3] Vuong Huu Tan. Study on development of Data, Tokai, Ibaraki, Japan, SND2006-V.02-1 nuclear spectroscopy to be used at the neutron (2007). beams for cascade gamma transitions and [10] Vuong Huu Tan, Pham Ngoc Son, et al. nuclear data measurements. Ministry 185, Capture Cross Section Measurements of scientific project report, code: BT12-07-09- 187 Re with Filtered Neutron Beams at the Dalat NLNT, (2009) (in Vietnamese). Research Reactor. Journal of the Korean [4] Pham Ngoc Son. Development of filtered Physical Society, Vol.59, No.2, pp. 1757-1760 neutron beam based on the horizontal channel (2011). No.2 of the Datal nuclear research reactor. [11] Pham Ngoc Son, Vuong Huu Tan. Filtered Ministry scientific project report, code: Neutron Capture Cross Section of ĐT.08/09/NLNT, (2012) (in Vietnamese). 186 187 W(n,γ) W reaction at 24 keV. Proceedings [5] Vuong Huu Tan, Pham Ngoc Son, et al. of the 4th Asian Nuclear Reaction Database Development of filtered neutron beams of 24, Development Workshop, al-Farabi Kazakh 59, and 133 keV at Dalat research reactor. National University, Almaty, Kazakhstan, 23 – Nuclear Science and Technology, ISSN: 1810- 25 October 2013, IAEA Nuclear Data Section, 5408, No.3, pp.8-15 (2009). INDC(KAS)-001, (2014). [6] Pham Ngoc Son, Vuong Huu Tan, Phu Chi [12] Pham Ngoc Son, Vuong Huu Tan, et al Hoa, Tran Tuan Anh. Development of Filtered Measurement of Thermal Neutron Cross- Neutron Beams of 24keV and 59keV at Dalat section and Resonance Integrals of the Research Reactor. Accepted to be published in 69Ga(n,)70Ga and 71Ga(n,)72Ga Reactions at World Journal of Nuclear Science and Dalat Research Reactor. Journal of the Korean Technology, Vol.4 (2014). Physical Society, Vol.59, No.2, pp. 1761-1764, ISSN: 0374-4884 (2011). [7] Tran Tuan Anh, Pham Ngoc Son, Vuong Huu Tan, Pham Dinh Khang, Phu Chi Hoa. [13] V. H. Tan, et al. Development of In-Vivo Characteristics of Filtered Neutron Beam Prompt Gamma Activation Analysis Using Energy Spectra at Dalat Reactor. Accepted to The Filtered Neutron Beam at The Dalat be published in World Journal of Nuclear Reactor. Proceeding of 11th Pacific Basin Nucl. Science and Technology (2014). Conf., Canada, (May 1998).

69 Nuclear Science and Technology, Vol. 4, No. 1 (2014), pp. 70-75 Characterization of neutron spectrum parameters at irradiation channels for neutron activation analysis after full conversion of the Dalat nuclear research reactor to low enriched uranium fuel

C.D. Vu1*, T.Q. Thien1, H.V. Doanh1, P.D. Quyet2, T.T.T. Anh3, and N.N. Dien1 1 Nuclear Research Institute, 01 Nguyen Tu Luc St., Dalat, Lamdong 2 Chu Van Anhigh school, Ductrong, Lamdong 3 The University of Dalat, 01, Phu Dong Thien Vuong St., Dalat, Lamdong *Email: [email protected] (Received 12 March 2014, accepted 9 May 2014)

Abstract: In the framework of the program on Russian Research Reactor Fuel Return (RRRFR) and the program on Reduced Enrichment for Research and Test Reactor (RERTR), the full core conversion of the Dalat Nuclear Research Reactor (DNRR) to low enriched uranium (LEU, 19.75% 235U) fuel was performed from November 24, 2011 to January 13, 2012. The reactor is now operated with a working core consisting of 92 WWR-M2 LEU. After the full core conversion, the neutron

spectrum parameters which are used in k0-NAA such as thermal neutron flux (th), fast neutron flux

(fast), f factor, alpha factor (), and neutron temperature (Tn) have been re-characterized at four different irradiated channels in the core. Based on the experimental results, it can be seen that the thermal neutron flux decreases by 6÷9% whereas fast neutron flux increases by 2÷6%. The neutron spectrum becomes‘harder’ at most of irradiated positions. The obtained neutron spectrum parameters from this research are used to re-establish the procedures for Neutron Activation Analysis (NAA) according to ISO/IEC 17025:2005 standard at NuclearResearch Institute.

Keywords: Neutron Activation Analysis (NAA), k-zero method, neutron flux, HEU, LEU.

I. INTRODUCTION core configuration consisting of 92 WWR-M2 LEU fuel assemblies [1, 2]. Since March 2012, Dalat nuclear research reactor was the reactor has been continuously operated upgraded from the TRIGA Mark-II designed about 100÷130 hours per month at nominal and constructed by the United States. The power of 500 kW for radioisotopes production, project of reconstruction and upgrade of the activation analysis and other researches. reactor was started in March 1982. The criticality was reached at 19:50 on November At the DNRR, there are four irradiated 01, 1983 and its regular operation at nominal channels used for NAA (Fig. 1): (1) the fast power of 500 kW was started from March 1984 pneumatic transfer system for short irradiation with the core loaded with 88 WWR-M2 fuel at the channel 13-2 and thermal column (Ti<45 assemblies enriched to 36% (HEU- Highly sec); (2) another pneumatic transfer system for Enriched Uranium) [1]. short and medium irradiation at the 7-1 channel (Ti: 45÷1200 sec); (3) the rotary rack with 40 Through the full core conversion project irradiated holes placed inside the graphite performed from November 24, 2011 to January reflector for long irradiation (Ti>20 min). 13, 2012, the DNRR now is operated with a

©2014 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute

CAO DONG VU et al.

II. EXPERIMENTAL

To standardize the irradiation channel, it is necessary to identify three basic parameters

such as thermal neutron flux (th), the ratio of thermal to epi-thermal neutron flux (f), and the coefficient describing the deviation of neutron spectrum distribution from the 1/E shape (α). In addition, two other parameters, fast neutron

flux (fast), and neutron temperature (Tn) are also considered to be the characteristic Fig. 1. Dalat research reactor cross-section. parameters of the neutron spectra in the irradiation channel [3]. The neutron spectrum parameters used In this study, bare multi-monitor method in k0-NAA including thermal neutron flux, fast neutron flux, and the factors of f, alpha, and [3, 4, and 5] using set of four monitors (Al- neutron temperature have been re-characterized 0.1%Au, Al-0.1%Lu, 99.98%Ni and 99.8%Zr) at four irradiated channels after full core was applied to determine the parameters of the conversion to LEU fuel. The obtained neutron neutron spectra at four irradiated positions of spectrum parameters from this research are the reactor. The experimental conditions are used to re-establish the procedures for Neutron described in Table I.

Activation Analysis using k0-IAEA software. Table I. The irradiation, decay and counting times for the monitors with Au, Lu, Ni, and Zr.

Irradiation time Decay Products (Ti)/position + Monitors time Counting time (Tc) [T1/2, E (keV)] (weight) (Td) - 15 m/channel 7-1 and 13- - 1200s for Ni and Lu monitors 65Ni [2.5h, 366.3, 1115.5, 4÷6 h 2 - 1800s for combination 1481.8], 176mLu [3.6h, 88.4] - 3 h/Thermal column 97 - 7200s for Zr monitor and Zr [16.7h, 743.4] ~1 d - 1h/Rotary rack combination 97Nb [16.7h, 657.9] +Al-0,1%Au wire (~5mg) - 900s for Au monitor 198Au [2.7d, 411,8]; 177Lu +Al-0,1%Lu wire (~5mg) - 7200s for Zr monitor [6.7d, 112.9, 208.4]; 95Zr + 99.8%Zr foil (~10mg) ~3 d 95 - 10800s for Ni and [64d, 756.7]; Nb [64d, 58 +99.98%Ni foil (~30mg) Lumonitors, and combination 765.8]; Co [70.8d, 810.8]

After an appropriate decay time for each parameters simultaneously, monitors were isotope, the samples were measured with the combined and measured at 0.5 hours, 1 hour, gamma spectrometry using HPGe detector and 3 hours with the decay time of 6 hours, 1

(FWHM ~ 2.2 keV at 1332 keV). The samples day and 3 days, respectively. The k0-IAEA were placed at 14 cm from the detector surface. software was employed for the treatment of In order to determine the neutron spectrum experimental data. For the purpose of quality

71 CHARACTERIZATION OF NEUTRON SPECTRUM PARAMETERS AT … control of the analytical procedure, 30 mg, 70 assigned value; and (5) 3.28<|Uscore|, the mg and 100 mg samples of the standard laboratory result is significantly different from reference material named NIST-679 (Brick the assigned value [4]. Clay) were irradiated at 45 sec, 1 hour, and 10 hours, respectively. The U-score is calculated III. RESULTS AND DISCUSSION according to the following equation: 푈푠푐표푟푒 = A. Neutron spectrum parameters at the 2 2 irradiated channels of the DNRR after full core (푋퐴푛푎 − 푋퐶푒푟푡 )/ 휎퐴푛푎 + 휎퐶푒푟푡 , where: conversion to LEU fuel XAna, and XCert are the analytical results, and certificated values, Ana,and Cert are the Neutron spectrum parameters at the uncertainty of XAna, and XCert. The results of channel 13-2, thermal column, channel 7-1, the laboratory are interpreted according to the and rotary rack of the DNRR after full core 5 possible evaluation classes as follows: (1) conversion to LEU fuel are given in Table II,

|Uscore|1.64, the laboratory result does not III, IV and V. In order to study the stability of differ significantly from the assigned value; (2) the neutron field at irradiated channels, the

1.64<|Uscore|<1.96, the laboratory result experiments at channels 13-2, 7-1, and probably does not differ significantly from the thermal column (Table II, III and IV) were assigned value; (3) 1.96<|Uscore|<2.58, it is not repeated three times in three different clear whether the laboratory result differs operation cycles of the reactor. However, at significantly from the assigned value; (4) the rotary rack (Table V), the parameters were

2.58<|Uscore|<3.28, the laboratory result is obtained only from two experiments (in probably significantly different from the March, and April 2012). Table II. Neutron spectrum parameters at the channel 13-2 after core coversion of the DNRR.

Parameters Experimental period Average ± SD Aug. 2012 Feb. 2013 Mar. 2013 12 2 th(× 10 n/cm /s) 4.21 0.17 4.34 0.17 4.070.09 4.21 0.14 12 2 fast(× 10 n/cm /s) 6.220.39 7.610.75 6.010.39 6.610.87  -0.073  0.009 -0.068  0.019 -0.067  0.004 -0.069  0.003 f 13.1  0.3 10.8  0.2 8.3 0.7 10.7  2.4

Tn (K) 317 5 307  9 312  11 312  5

Table III. Neutron spectrum parameters at the thermal column after core coversion of the DNRR.

Parameters Experimental period Average ± SD Jul. 2012 Mar. 2013 Apr. 2013 11 2 th(× 10 n/cm /s) 1.26  0.54 1.24 0.03 1.27  0.09 1.21  0.27 8 2 fast(× 10 n/cm /s) 8.99 0.06 8.440.49 8.03 0.06 8.29 0.11 - -0.117  0.032 -0.094 0.167 -0.140  0.015 190  8 195  4 198  2 197  4

Tn (K) 306  6 298  7 291  8 297  3

72 CAO DONG VU et al.

Table IV. Neutron spectrum parameters at the channel 7-1 after core coversion of the DNRR.

Parameters Experimental period Average ± SD Mar. 2012 Apr. 2012 May2012

12 2 th(× 10 n/cm /s) 4.30 0.14 4.12 0.18 4.24 0.12 4.22 0.04 12 2 fast(× 10 n/cm /s) 3.860.35 3.690.10 4.140.23 3.900.23  -0.022  0.032 -0.041  0.025 -0.031  0.028 -0.031  0.009 f 9.6  0.9 10.2  0.4 9.3  0.7 9.7  0.5

Tn (K) 300  5 300  5 301  5 300  0.6

Table V. Neutron spectrum parameters at the rotary rack after core coversion of the DNRR.

Parameters Experimental period Average ± SD Mar. 2012 Apr. 2012

12 2 th(× 10 n/cm /s) 3.68 0.04 3.84 0.15 3.760.11 12 2 fast(× 10 n/cm /s) 0.31 0.05 0.32 0.04 0.32 0.01 0.099  0.010 0.104  0.010 0.102  0.003 30.1  2.5 30.0  1.0 30.1  0.4

Tn (K) 294  6 297  6 295  2

B. Comparison of the neutron spectrum results in Table VI the absolute value of α at parameters before and after full core conversion to channel 7-1 increases by approximately 1.7 LEU fuel times at negative side after full core conversion. This means that the neutron Table VI shows the thermal (th) and spectrum at the channel 7-1 becomes 'harder' fast (fast) neutron fluxes,  coefficient, and f rather than that of before conversion. At the at the channel 7-1, and rotary rack measured rotary rack, the α factor significantly increases before [3] and after the full core conversion. by 2.5 times at positive sign. This means that The obtained results in Table VI show that epi-thermal neutron spectrum at this position after full core conversion, thermal neutron flux tends to deviate below the 1/E distribution [5]. reduces 8% at channel 7-1, and 6% at rotary rack whereas the fast neutron flux at channel As the old channel 13-2 was removed 7-1 and rotary rack increases by 2% and 6%, from the core in November 2006, and a new respectively. This means that epi-thermal pneumatic transfer system together with the neutron flux also increases (f decreases) channel 13-2 was reinstalled in June 2012, leading to the occurrence of the interference therefore, there are no data for neutron reactions in k0-NAA such as (n, p), (n, n') spectrum at channel 13-2 during etc.[4]. On the other hand, also from the 2006÷2011 period.

73 CHARACTERIZATION OF NEUTRON SPECTRUM PARAMETERS AT …

Table VI. The thermal and fast neutron flux at channel 13-2 and Rotary rack before and after full core conversion of the DNRR.

2 2 th (n/cm /s) fast(n/cm /s) α f [3], measured in 2010 with HEU-LEU fuel Channel 7-1 4.59 × 1012 3.81 × 1012 -0.019 11.09 Rotary rack 4.01 × 1012 0.30 × 1012 0.040 42.28 This work, measured in 2012 with LEU fuel Channel 7-1 4.22 × 1012 3.90 × 1012 -0.031 9.70 Rotary rack 3.76 × 1012 0.32 × 1012 0.102 30.10 This work/[3], 7-1 0.92 1.02 1.67 0.87 This work/[3], Rotary rack 0.94 1.06 2.54 0.71

Table VII presents the thermal neutron conversion, but mainly relates to the flux values at the channel 13-2 and thermal modification of structure of the thermal column column measured in 2003 [5] and after full which was installed together with a new core conversion. The results from Table VII pneumatic transfer system in 2012. The new show that after the core conversion, the thermal facility for thermal column was put close to the neutron fluxes at channel 13-2 reduce 9% and graphite reflector in which the sample was increase approximately 21 times at thermal placed 10.8 cm deeper in contrast to the old column. This unusual change at the thermal irradiated position. column does not result from the core

Table VII. Themal neutron flux before and after full core conversion.

Thermal column Channel 13-2 HEU (2003) [5] 5.80 × 109 4.62 × 1012 LEU (2012) 1.24 × 1011 4.21 × 1012 LEU/HEU 21.38 0.91

C. Analysing of SRM NIST-679 (Brick clay) Tables VIIIa and VIIIb show that the using obtained neutron parameters |Uscore| for all analytical values are less than To assess the quality of the neutron 1.64, which means that all results are spectrum data set obtained through this study, acceptable. This analysis also shows that it is necessary to re-characterize the neutron the SRM named NIST-679 was analyzed by k0- NAA. The analytical results obtained before spectrum parameters after the core conversion. and after the core conversion are given in Nevertheless, the data obtained from this study Table 8a and Table 8b, respectively. are reliable and can be used to calibrate the irradiated channels for k0-NAA at the DNRR.

74 CAO DONG VU et al.

Table VIIIa. Analytical results of SRM NIST-679 before the core conversion.

Analyzed value Certified value No. Element U Position Conc. Unc. Conc. Unc. score 1 Al 103500 5208 110100 3400 -1.06 7-1 2 Dy 6.95 1.98 7.15 0.27 -0.10 7-1 3 Mn 1764 436 1852 45 -0.20 7-1 4 As 8.9 3.1 9.5 0.2 -0.19 RR 5 La 50.0 12.4 49.9 0.5 0.01 RR 6 Fe 92133 6168 90500 2100 0.25 RR 7 Sc 22.1 2.3 22.8 0.2 -0.30 RR 8 Th 13.47 1.63 13.46 0.12 0.01 RR Table VIIIb. Analytical results of SRM NIST-679 after the core conversion.

Analyzed value Certified value No. Element U Position Conc. Unc. Conc. Unc. score 1 Al 106500 8758 110100 3400 -0.38 7-1 2 Dy 6.4 1.3 7.15 0.27 -0.56 7-1 3 Mn 1742 116 1852 45 -0.88 7-1 4 As 8.3 1.42 9.5 0.2 -0.84 RR 5 La 45.5 2.77 49.9 0.5 -1.56 RR 6 Fe 92880 3001 90500 2100 0.65 RR 7 Sc 21.9 2.5 22.8 0.2 -0.36 RR 8 Th 13.2 0.2 13.46 0.12 -1.11 RR

IV. CONCLUSION REFERENCES

Re-establishment of the neutron spectrum [1] N.N. Dien, Project of fuel conversion at Dalat parameters including th, fast, , f, and Tn at research reactor, Dalat Nuclear Research four irradiated channels for NAA at the DNRR Institute (2011). after full core conversion to LEU fuel was [2] N.N. Dien, Report on the physics start-up for carried out. conversion to LEU fuel at Dalat research reactor, Dalat Nuclear Research Institute, (2012). After replacement of the core with LEU [3] C.D. Vu, Project report (code CS/09/01-01) fuel assemblies, the thermal neutron flux in Study on application of k0-IAEA at Dalat most of irradiated channels decreases by 6÷9% research reactor, Vietnam Atomic Energy while the epi-thermal neutron flux and fast Institute (2010). neutron increase by 2÷6%; neutron spectrum [4] H.M. Dung*, M.C. Freitas, J.P. Santos, J.G. becomes‘harder’ in most of the investigated Marques, Re-characterization of irradiation positions. facilities for k0-NAA at RPI after conversion to LEU fuel and re-arrangement of core New neutron spectrum parameters configuration, Nuclear Instruments and Methods obtained through this study will be useful for in Physics Research A 622, 438–442 (2010). characterization of the irradiation channels in [5] H.M. Dung, Study for development of k-zero k0-NAA analytical procedure at the DNRR Neutron Activation Analysis for multi-element after full core conversion to LEU fuel. characterization, PhD thesis, the Natural Science University, Hochiminh city (2003).

75 Nuclear Science and Technology, Vol. 4, No. 1 (2014), pp. 76-83 Some results of NAA collaborative study in white rice performed at Dalat Nuclear Research Institute

T.Q. Thien*, C.D. Vu, H.V. Doanh, N.T. Sy Dalat Nuclear Research Institute 01 Nguyen Tu Luc St., Dalat, Lam Dong * Email: [email protected] (Received 5 March 2014, accepted 14 March 2014)

Abstract: White rice is a main food for Asian people. In the framework of Forum for Nuclear Cooperation in Asia (FNCA), therefore, the eight Asian countries: China, Indonesia, Japan, Korea, Malaysia, the Philippines, Thailand and Vietnam selected white rice as a common target sample for a collaboration study since 2008. Accordingly, rice samples were purchased and prepared by following a protocol that had been proposed for this study. The groups of elements that were analyzed by using neutron activation analysis in the white rice samples were toxic elements and nutrient elements, including: Al, As, Br, Ca, Cl, Co, Cr, Cs, Fe, K, Mg, Mn, Na, Rb and Zn. The analytical results were compared between the different countries and evaluated by using the Tolerable Intake Level of World Health Organization (WHO) and Recommended Dietary Allowance or Adequate Intake (AI) of the U.S. Institute of Medicine (IOM) guideline values. These data will be very useful in the monitoring of the levels of food contamination and in the evaluation of the nutritional status for people living in Vietnam and other Asian countries. Keywords:White rice, neutron activation analysis, FNCA,tolerable intake level, dietary reference intakes, adequate intake.

I. INTRODUCTION Utilization in the framework of the forum FNCA. Vietnam has participated in the FNCA FNCA (Forum for Nuclear Cooperation since 2000. in Asia) was formally established in March 1999 at the 10th session of the International In the FNCA workshop held in Dalat, Conference on Nuclear Cooperation in Asia Vietnam, in 2008, the eight among twelve region ICNCA (International Conference for member countries of the FNCA which are Nuclear Cooperation in Asia) initiated and China, Indonesia, Malaysia, Japan, Korea, the funded by the Japanese government. FNCA is Philippines, Thailand and Vietnam, agreed to supposed to enhance mutual understanding, participate in a collaborative study on the exchange of information and experience to social analysis of food samples as a sub-project and economic development in Asia through thematic in NAA. White rice has been selected research, collaboration, technology applications as research subjects for this work because of its initiatives for peaceful purposes. Up to 2012, importance as the basic staple food for people FNCA has 12 member countries, including: wholives in Asia. Specifically, the major rice Australia, Bangladesh, China, Indonesia, producing countries in Asia are China, India, Malaysia, Japan, Kazakhstan, Korea, Mongolia, Indonesia, Malaysia, Bangladesh, Thailand, the Philippines, Thailand and Vietnam. Vietnam, etc. These countries accounts for over 80% of production and consumption of rice in NAA (Neutron Activation Analysis) is the world. This highlights the importance of the one of the projects under the ResearchReactor

©2014 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute

TRAN QUANG THIEN et al. information gained from the study because rice II. EXPERIMENTS is the staple food as well as providing a large A. Sample collection and preparation portion of the calories in the Asian diet [1]. Eighteen samples were collected from the The objective of this study was to Department of Agriculture and Rural determine the inorganic elements in the white Development Centre Tiengiang province rice of Vietnam and compared it with seven agricultural seed wherein rice is the most Asian countries by NAA method, these results common type on the market which are are preliminary by the level of nutrients and presented in Table I. toxic elements in rice for safety. Table I.The information sampling of Vietnam’s rice samples at Department of Agriculture and Rural Development Centre Tiengiang province

No. Type 1 Ham Chau Rice 2 IR 50404 Rice 3 Japan 504 Rice 4 Jasmine 85 Rice 5 Jasmine Rice 6 OM 4218 Rice 7 OM 4900 Rice 8 OM 5451 Rice 9 OM 5472 Rice 10 OM 5976 Rice 11 OM 6162 Rice 12 OM 6377 Rice 13 OM 6976 Rice 14 Otim Rice 15 Seri Rice 16 Tai Nguyen Rice 17 Taiwan Fragrant Rice 18 Thom Lai Rice

The collected rice samples were brought B. Analysis to the lab and washed with distilled water and The rice samples were analyzed by then dried in a drying oven at a temperature of INAA in Dalat Nuclear Research Institute. The 60 0C for 4 hours, then ground into fine analytical procedures are followed with particles using an agate mortar in order to prevent contamination. Rice samples were ISO/IEC 17025 [2]. A concurrent analysis of repeatedly ground until a particle size of 60 reference standard samples for quality control meshes. Finally, the samples were subdivided was made for each batch of analysis. Analyses into subsamples weighing from 100-300 mg were made using a combination of both short prior to analysis by INAA.[1] and long irradiations. The HPGe detector with

77 SOME RESULTS OF NAA COLLABORATIVE STUDY IN WHITE RICE ... multichannel analysis system was used to The result of standard reference material measure the gamma rays from the sample after was shown in Table II and the result of fifteen irradiation. The concentration of the elements elements concentrations in eighteen samples of was calculated using the relative method and/or white rice are in Table III. k-zero method. In Table II, The average result are caculated through 3 times analysis, it's not III. RESULTS AND DISCUSSION much different to the value of certificate. The Z-score of all elements is lower than 2, that A. The content of elements in white rice mean this results are satisfactory. samples Table II. Result of standard reference material IAEA-V-10

No. Ele. Aver. Sd. Cert. Z-score Ana/Cert 1 Mg 1579 121 1360 1.81 1.161 2 Ca 20865 2892 21600 -0.25 0.966 3 Cl 7360 100 - - - 4 Mn 46 5 47 -0.20 0.979 5 Na 507 9 500 0.78 1.014 6 K 20119 3530 21000 -0.25 0.958 7 Br 7.3 0.5 8 -1.40 0.913 8 Sc 0.016 0.002 0.014 1.00 1.143 9 Cr 6.6 0.5 6.5 0.20 1.015 10 Fe 196 21 186 0.48 1.054 11 Co 0.15 0.03 0.13 0.67 1.154 12 Zn 25.5 2.2 24 0.68 1.063 13 Rb 7.7 0.5 7.6 0.20 1.013 Aver: Average result; Sd: Standart deviation Cert: Certificate In Table III, the concentration of Mg Zn in rice samples determined by eight element are not analyzed in all samples, elements participating countries are summarized in Table concentration of Al, Ca and Fe are not obtained 4. The results of quality control analysis for and reported limit of detection, the result of other fifteen elements are summarized as a relative elements are included concentration and error (%) with absolute value and are shown in uncertainty. The highest concentration are K Fig. 1. The relative error of most of the element, the lowest come from Co and Cs. The elements evaluated in Fig. 1 were less than other elements have no significant differences 15%, except for some few elements such as Al between all samples except Rb. of Malaysia; Co of Vietnam; Mg of China, B. Comparing the elements concentration in Korea and Vietnam, Mn and Na of Korea rice white rice of 8 countries samples. Results of fifteen elements: Al, As, Br, Ca, Cl, Co, Cr, Cs, Fe, K, Mg, Mn, Na, Rb and

78 TRAN QUANG THIEN et al.

Table III.The analytical results of eighteenwhite rice samples in Vietnam

Al As Br Ca Cl Co Cr Cs Fe K Mg Mn Na Rb Zn No. Type C. U. C. U. C. U. C. U. C. U. C. U. C. U. C. U. C. U. C. U. C. U. C. U. C. U. C. U. C. U. Ham Chau 1 <3 0.12 0.02 0.43 0.05 <165 294 11 0.031 0.006 <0.4 0.026 0.007 <14 518 10 NA. 8.0 0.2 22.6 0.3 1.5 0.3 23.0 0.5 Rice 2 IR 50404 Rice <8 0.16 0.04 0.17 0.05 <160 242 11 0.026 0.008 <0.5 0.061 0.010 <10 1649 17 NA. 15.9 0.2 11.5 0.2 11.4 0.6 21.3 0.6

Japan 504 3 <4 0.06 0.02 0.78 0.07 <120 407 14 0.022 0.008 <0.4 0.019 0.008 <14 527 11 NA. 5.1 0.1 48.5 0.3 1.0 0.3 21.7 0.8 Rice Jasmine 85 4 <5 0.13 0.03 0.31 0.07 <185 235 30 0.028 0.008 <0.5 0.056 0.012 <18 1543 17 NA. 19.1 0.1 18.2 0.3 10.3 0.7 22.8 0.8 Rice 5 Jasmine Rice <4 0.12 0.02 0.22 0.04 <100 192 10 0.030 0.008 <0.5 0.052 0.007 <20 453 9 NA. 3.5 0.1 9.0 0.2 1.9 0.4 19.2 0.6

6 OM 4218 Rice <5 0.15 0.04 0.22 0.06 <130 282 25 0.031 0.008 <0.6 0.045 0.009 <17 1548 16 NA. 12.8 0.1 17.1 0.3 8.3 0.6 22.2 0.7

7 OM 4900 Rice <7 0.10 0.03 0.34 0.07 <140 346 31 0.043 0.008 <0.5 0.044 0.010 <19 1780 18 NA. 16.0 0.1 14.8 0.3 5.9 0.5 26.2 0.7

8 OM 5451 Rice <7 0.15 0.03 0.29 0.06 <120 199 23 0.036 0.007 <0.6 0.046 0.008 <16 1254 16 NA. 11.8 0.1 18.8 0.3 11.4 0.7 22.6 0.7

9 OM 5472 Rice <4 0.11 0.03 0.26 0.06 <100 293 22 0.034 0.007 <0.5 0.047 0.011 <22 1414 15 NA. 11.9 0.1 14.9 0.2 9.7 0.6 24.8 0.7

10 OM 5976 Rice <5 0.11 0.03 0.18 0.05 <110 186 11 0.031 0.009 <0.6 0.043 0.011 <17 1227 14 NA. 12.7 0.1 11.3 0.2 10.7 0.6 24.0 0.7

<1 11 OM 6162 Rice 0.08 0.03 0.17 0.05 <120 307 13 0.043 0.009 <0.4 0.039 0.011 <16 1509 16 NA. 13.9 0.2 12.6 0.2 8.3 0.6 23.7 0.7 0 12 OM 6377 Rice <7 0.14 0.03 0.23 0.06 <120 266 27 0.058 0.009 <0.6 0.063 0.011 <13 1636 16 NA. 16.8 0.1 13.1 0.2 15.7 0.7 23.4 0.8

13 OM 6976 Rice <6 0.11 0.03 0.34 0.05 <100 382 14 0.027 0.007 <0.5 0.038 0.009 <19 1487 16 NA. 6.7 0.3 13.5 0.2 4.8 0.5 25.0 0.7

14 Otim Rice <4 0.17 0.03 0.47 0.04 <100 225 11 0.044 0.009 <0.3 0.051 0.013 <28 568 11 NA. 7.5 0.1 14.4 0.2 2.8 0.5 21.4 0.7

15 Seri Rice <4 0.22 0.02 0.18 0.04 <135 236 11 0.041 0.009 <0.4 0.074 0.011 <21 501 9 NA. 5.4 0.1 13.5 0.2 8.2 0.7 21.1 0.7

Tai Nguyen 16 <3 0.06 0.02 0.65 0.07 <100 378 12 0.039 0.009 <0.5 0.016 0.007 <16 470 9 NA. 5.2 0.1 40.9 0.3 1.0 0.3 17.5 0.7 Rice 17 Taiwan Rice <5 0.11 0.03 0.23 0.04 <110 330 12 0.024 0.010 HL 0.066 0.010 <22 842 13 NA. 5.9 0.1 10.5 0.2 7.5 0.6 24.8 0.9

18 Thom Lai Rice <3 0.08 0.02 0.40 0.04 <110 208 9 0.031 0.008 <0.5 0.029 0.007 <14 463 8 NA. 3.9 0.1 7.8 0.2 1.2 0.4 19.8 0.6

Average <5 0.12 0.33 <124 278 0.034 <0.5 0.045 <18 1077 NA. 10.1 17.4 6.8 22.5 Unit: mg/kg; C. : Concentration; U. : Uncertainty; NA. : Not Applicable

79 SOME RESULTS OF NAA COLLABORATIVE STUDY IN WHITE RICE ...

Table IV. The analytical results of white rice (unit: mg/kg) [1]

eVietnam Ele. aChina Indonesia bJapan Korea cMalaysia dPhilippine Thailand (This work) Al <4.46 <20.52 <1.66 <1.38 <2 <2.82 <2.33 <5 As 0.55 0.08 0.1 0.13 0.11 0.07 0.09 0.12 Br 0.35 0.45 0.5 0.19 13.6 5.35 0.43 0.33 Ca N.A <4.53 49.5 53.9 <10 39.1 <15 <124 Cl 264 210 239 193 225 236 239 278 Co <0.3 0.77 N.A 0.005 0.026 N.A 0.022 0.034 Cr 0.25 0.38 N.A <0.01 <0.08 N.A <0.4 <0.5 Cs <0.07 0.09 N.A 0.009 0.016 N.A N.A 0.045 Fe N.A 4.65 N.A 1.58 <5 N.A <16 <18 K 977 739 611 660 573 637 620 1077 Mg 379 131 149 241 <150 90 59 N.A Mn 9.25 9.95 7.66 9.06 6.19 7.89 9.23 10.1 Na 10.3 7.7 5.69 4.1 13.7 5.17 4.58 17.4 Rb <3.35 7.64 N.A 1.39 2.1 3.24 1.34 6.8 Zn 15.3 24.2 18.5 15.3 10.1 15.4 21.4 22.5

N.A: not applicable; (a) Mean values are derived from four different samples; (b) Mean values from a sample of known origin and two samples of unknown origin; (c) Mean values are derived from two different samples; (d) Mean values from four samples of unknown origin; (e) This work, average value of eighteen samples from known origins. As can be seen from Table IV, the Al Indonesia and the Philippines were more than a concentrations in rice saples from all dozen times higher than those of other participating countries were below the countries. Five elements Cl, K, Mn, Na and Zn detection limit, hence only the limit of did not differ significantly and the average detection (LOD) were reported. Indonesia had content of the standard deviation were 236±27, an LOD value of 20.52 for Al, highest 737±178, 8.67±1.32, 8.58±4.84 and 17.8±4.7 compared to other countries. Korea and Japan mg/kg respectively. Concentrations of Mg had the lowest LOD values in the eight were reported by six countries excluding countries. K concentration range is from 553 to Malaysia and Vietnam. Thailand showed the 1077 mg/kg. K in Vietnam rice samples had lowest levels of Mg, 59 mg/kg, while the Mg the highest value which is 1077 mg/kg. Cl and content of China was the highest at 379 mg/kg. Mg have similarly eminent values. Seven Only three countries namely Japan, South elements of As, Br, Cl, K, Mn, Na and Zn were Korea and the Philippines reported Ca data determined by all participating countries, but which were 49.5, 53.9 and 39.1 mg/kg LODs were not reported. As content of China respectively. In addition, the levels of Cr, Cs had the highest value, 0.55 mg/kg and the other and Fe in Indonesian rice were higher countries have equivalent levels of As, 0.1 compared to those of other countries. mg/kg. Br concentrations of Malaysia,

80 TRAN QUANG THIEN et al.

C. Dietary intake level of the toxic elements rice consumption varies in different countries, and nutrition elements of 8 countries and therefore a consensus value of 300 To estimate the dietary intake level of grams/day was set, to be able to compare the inorganic constituents on consumption of white intake of As, Cl, K, Mn, Na and Zn from rice rice, it was necessary to conduct a survey of consumption in all participating countries. This daily consumption of rice. For example, the is to assess whether or not, the ingested levels amount of the average daily consumption of of the elements can be considered as harmful rice in Korea in 2000 was 256 grams, or in or beneficial to human health. Data are shown Vietnam in 2010 is 360 gram [3, 4]. However, in Table V.

China Indonesia Japan/Philippines 25 20

, % , 15 10 Error 5 0 Al As Br Ca Cl Co Cr Cs Fe K Mg Mn Na Rb Zn

Fig. 1. The absolute value of the relative error (%) of the value analysis to value certification/reference.

Table V. The RDA value of 6 elements each day through white rice, assuming consumption of 300 grams /day for adults[1]

Vietnam Ele. China Indonesia Japan Korea Malaysia Philippine Thailand (This work) As (µg) 165 24 30 39 33 21 27 36 Cl (mg) 79.2 63 71.7 57.9 67.5 70.8 71.7 83.4 K (mg) 293 222 183 198 172 191 186 323 Mn 2.78 2.99 2.30 2.72 1.86 2.37 2.77 3.03 (mg) Na (mg) 3.09 2.31 1.71 1.23 4.11 1.55 1.37 5.22 Zn (mg) 4.59 7.26 5.55 4.59 3.03 4.62 6.42 6.75

The WHO has established a Tolerable Recommended Dietary Allowance (RDA) or Intake Level for weekly consumption, which is adequate intake (AI) for the necessary elements 15 mg/kg of body weight for As [5]. Assuming [6, 7]. Zn has the highest RDA of 11 mg/day a body weight of 70 kg of an adult, the for men. AI highest values for Cl, Mn defined Tolerable Intake Level for As daily by the IOM is 2.3 g/day for all adults, for Na consumption will be 150 microgram As. In and K, the highest AI values are respectively addition, the Institute of Medicine (IOM) in the 1.5 and 4.7 g/day. United States has established the value of the

81 SOME RESULTS OF NAA COLLABORATIVE STUDY IN WHITE RICE ...

Calculations for the RDA or AI for the countries namely China, Indonesia, Japan, elements As, Cl, K, Mn, Na, Zn are shown in Korea, Malaysia, the Philippines, Thailand and Figure 2.Tolerable Intake Level of As in China Vietnam. A total of fifteen elements in thirty is higher than Tolerable Intake Level of WHO five samples of white rice collected from eight which was about 10%, for the other countries. countries were determined by INAA method. The level of Mn is almost equal to the value of Within the framework of project participants the RDA of IOM. This shows just rice FNCA/NAA, NAA laboratory of Vietnam has consumption of 300 g/day may provide collected and analyzed fifteen elements in sufficient Mn necessary for the human body. eighteen samples of white rice types. Results of The intake level for the remaining elements Vietnam’s rice has been compared with the (Cl, K, Na and Zn) were below the RDA or AI. results of the seven countries participating In the case of Zn, the range of daily members. consumption from 21.6% (Malaysia, The analytical data were compared Indonesia) to 51.9% (Indonesia) can only between the participating countries and supply approximately 21.6% to 51.9% Zn assessed according to the daily intake using the necessary for the human body. Similarly, guideline values set by the WHO and IOM. consumption of Cl at 2.5% to 3.6%, K at 3.7% The results showed an elevated amount of As to 6.2% and 0.3% Na were below the in Chinese rice which exceeded by recommended values. These essential elements approximately 10%, the RDA recommended can be obtained anyway, from other foods such by WHO. In addition the research gave an as meat, fish, vegetables, eggs, milk, etc. which overview of the levels of nutritional elements are eaten together with the rice. Na, Mn, Cl, K and Zn in rice consumed in the eight countries. Information on the intakes of IV. CONCLUSIONS Mn (of approximately 100%), Zn, Na, Cl A collaborative study on the (21.6÷51.9) % and K (lower than 10%) in determination of elemental abundance in rice comparison to the requirements of IOM was using NAA was participated in by eight obtained from the study.

China Indonesia Japan Korea Malaysia Philippines Thailand Vietnam 1000.0

100.0

%RDA 10.0

1.0

0.1 As Cl K Mn Na Zn Fig. 2. Assess daily nutrient consumption (%) for the six elements through white rice.

82 TRAN QUANG THIEN et al.

In future, FNCA will carry on to expand [3] Ministry of Agriculture and Forestry, the scope of research in elemental abundance in Agricultural and forestry statistical yearbook food samples to strengthen the collaboration 2003. Ministry of Agriculture and Forestry, between Asian countries for the continued Seoul, (2003). application of NAA in the assessment for [4] National Institute of Nutrition, A review of the contamination and mineral potentiality in the nutrition situation in Vietnam 2009-2010, basic foodstuffs. Medical Publishing House, Hanoi, (2011). [5] World Health Organization, Evaluation of ACKNOWLEDGEMENTS certain food additives and contaminants, (Thirty-third report of the Joint FAO/WHO We would like to thank the MEXT of Expert Committee on Food Additives). WHO Japan for support of this research. Technical Report Series, No. 776, (1989).

REFERENCES [6] Institute of Medicine, Food and Nutrition Board, Dietary reference intakes for vitamin A, [1] J. H. Moon et. al, A NAA collaborative study vitamin K, arsenic, boron, chromium, copper, in white rice performed in seven Asian iodine, iron, manganese, molybdenum, nickel, countries, Journal of Radio- analytical silicon, vanadium, and zinc, National Academy Chemistry, Volume 291, Issue 1, pp 217-221 of Sciences, Washington DC, (2001). (January 2012). [7] Institute of Medicine, Food and Nutrition [2] Center for Analytical Techniques (CATech), Board, Dietary reference intakes for water, Dalat Nuclear Research Institute (NRI), potassium, sodium, chloride and sulfate. “TCCS-MSH from 01 to 03”, Dalat, (2011). National Academy of Sciences, Washington DC, (2004).

83 Nuclear Science and Technology, Vol. 4, No. 1 (2014), pp. 84-91 A new rapid neutron activation analysis system at Dalat nuclear research reactor

H.V. Doanh*, C.D. Vu, T.Q. Thien, P.N. Son, N.T. Sy, N. Giang and N.N. Dien Nuclear Research Institute, 01 Nguyen Tu Luc Street, Dalat, Vietnam *E-mail: [email protected] (Received 5 March 2014, accepted 12 March 2014)

Abstract: An auto-pneumatic transfer system has been installed at the Dalat research reactor for rapid instrument neutron activation analysis based on very short-lived nuclides. This system can be used to perform short irradiations in seconds either in the vertical channel 13-2 or in the horizontal thermal column of the reactor. The transferring time of sample from irradiation to measurement position is approximately 3.2 seconds. A loss-free counting system using HPGE detector has been also setup in compacting with the pneumatic transfer system for measurement of sample’s activity, automatically starting for data acquisition at irradiated sample’s arrival. This new facility was tested and shown to have high potential for the determination of short-lived nuclides with half-lives from 10  100 seconds. This work presents the results of timing parameter measurements, characterization of irradiation facilities, and application of this system to determining Selenium concentration in several biological reference materials. Keywords: Auto-pneumatic transfer system, neutron activation analysis, short-lived nuclides.

I. INTRODUCTION In the recent years, through the IAEA Instrumental neutron activation analysis TC Project RER/4/028, a new automatic PTS (INAA) has been developed and applied at the for rapid neutron activation analysis based on short-lived nuclides has been developed. This 500 kW Dalat research reactor (DNRR) since facility consists of three main parts introduced 1984. Until now, it is capable of analyzing in reference [1]. The first part, consisting of more than 40 elements based on radionuclides two aluminum irradiation tubes, which are with short, medium and long-lived time. For inserted into the vertical channel No.13-2 and short-lived nuclides with half-lives from 2 the horizontal thermal column (TC) of the minutes to 2.6 hours, samples are often reactor. The second part is a digital signal irradiated at the neutron channel No.7-1 of processing spectrometer connected to a 40% Dalat research reactor through a semi-auto relative efficiency HPGe detector coupled with pneumatic transfer system (PTS) with valid a transistor reset preamplifier. The third part is irradiation time from 45 seconds to 20 composed of pneumatic chambers, loading and minutes. Measurements are often performed sliding devices in Cabin-1 which facilitates the using a gamma spectrometer coupled with a fully automatic irradiation-counting HPGe (GMX-30190), but with manual procedures. It has also a sample automatic manipulation between loading and counting loader for the sequential routing of the samples procedures. Therefore, the shortest-lived in multi-samples operation mode: when the 28 nuclides that could be detected are Al (T1/2 = measurement for one sample is finished, the 52 51 2.24 min), V (T1/2 = 3.75 min), and Ti (T1/2 next sample is loaded and sent to the = 5.76 min). irradiation and then counting positions. In this

©2014 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute

HO VAN DOANH et al. system, there is also has optical sensors for 77mSe, 179mHf, 46mSc, and 110Ag can be used for controlling the transport of the capsule as a INAA at Dalat reactor, which the former sample carrier, and for accurately measuring system can not detect. the capsule flight time from irradiation position to detector. The installation diagram is shown The main purpose of this work is to test in Fig. 1. the system for both mechanical and analytical reliability. A systematic study has been carried This PTS system can be used to perform out including measurements for timing short irradiations in seconds. The return time of parameters of the system and neutron flux at sample from irradiation position to counting irradiation positions, and the application of this position is about 3.2 s. Timing information for system to determining of Selenium in a number both irradiation and counting will be instantly of biological reference materials for validation delivered to the activation analysis workstation purpose. computer. The digital gamma spectrometer is selected and tuned for accurate measurement at high and varying counting rates, using loss-free counting technology. Accordingly, shorter- lived nuclides (half-life < 1 min) such as 20F,

Fig. 1. Diagram of the auto-pneumatic transfer system installed at DNRR.

85 A NEW RAPID NEUTRON ACTIVATION ANALYSIS SYSTEM AT …

II. EXPERIMENT recorded at the computer via a signal from the fast solenoid providing the return gas (START A. Timing measurements of return time) and from the optical sensor at The accuracy of irradiation time of an the detector station recording the arrival of irradiation facility should be checked and capsule (END of return time). calibrated as a type of analytical qualify B. Neutron spectrum parameters of irradiation control [2]. In this experiment, the absolute position irradiation time, the sample transferring time from entrance to irradiation position of The neutron spectrum parameters aluminum tube (Tin) and the reverse movement including thermal neutron flux th, fast neutron

(Tout) were determined. The experiments were flux f, the thermal flux to epithermal neutron done outside the reactor before installation of flux (epi) ratio were measured at sample irradiation facilities inside the reactor. The irradiation positions in channel No.13-2 and arrangement of timing measurements is shown thermal column using Au, Zr, and Ni monitors. in Fig. 2. The setup includes two NaI(Tl) Monitors were inserted into a high purity detectors placed at the top and bottom sides of polyethylene vial and loaded into rabbit the irradiation tube. The detectors detect (capsule) for irradiation. The activity gamma radiation pulses from a 131I source measurements were carried out by a calibrated inserted in a capsule while moving inside. The gamma-ray spectrometer combined with HPGe counters are set for running in Multi-Channel detector (GMX-30190). The measured Scaler (MCS) mode; MCS mode records the spectrum were analyzed by using the k0-IAEA counting rate of events as a function of time. program. The irradiation, decay and counting The return time from the irradiation to times for each monitor are shown in Table I. the measurement positions was determined by Typically monitors with masses of 4 mg for a series of irradiation (50 replicates) for a total Al-0.1%Au foil (IRMM-530R), 30 mg for pure weight (capsule, vial and sample) of about 4.4 Ni (wire), 10 mg for Zr (foil) were irradiated gram and air pressure of 3.1 bars over a for 10 min at 13-2 channel (2 h at thermal 97 distance of 40 m for 13-2 channel and 36 m for column), and the decay time is 1 day for Zr 198 95 58 thermal column. The return times were and 3 days for Au, Zr and Co.

Fig 2. Arrangement for experiment of timing measurements. 86 HO VAN DOANH et al.

Table I. The irradiation, decay and counting times for the monitors.

Time/position Decay Counting time Measured irradiation (monitor, time (combination) radionuclides mass) (T1/2, -rays in keV)

10 min/ 13-2 channel ~ 1 d 1  2 h 97mNb (60 s, 743.4)*; 2 hours/ thermal column 97Nb (16.7 h, 657.9) (Al-0.1% Au, ~ 4 mg) ~ 3 d 0.5  3 h (5 h) 198Au (2.7 d, 411.8); (99.8% Zr, ~ 10 mg) 95Zr (64 d, 765.8); (99.98% Ni, ~ 30 mg) 58Co (70.8 d, 810.8) * Nuclide 97mNb is decayed from nuclide 97Zr with half-life of 16.7h.

C. Determination of Selenium were irradiated for 25 s, allowed 20 s delay time to eliminate interference of 116mIn with a A variety of reference materials (Tuna half-life of 2.18 s [3, 4]) and counted for 25 s Fish IAEA-436, Oyster tissue NIST 1566b, at a distance of 10 cm from the detector Bovine Liver NIST 1577, Bovine Liver NIST (GMX40-76-PL). The concentrations of 1577b) were selected to assess reliability of Selenium were determined by both k-zero and this system on the short-time activation relative methods. application. All of the samples were irradiated 12 -2 -1 at a neutron flux of 4.210 n.cm .s in the III. RESULTS AND DISSCUSION 13-2 channel and counted on the calibrated HPGe gamma-ray spectrometer (GMX40-76- A. Timing measurements PL). The results for average transferring time In order to evaluate the limit of detection of sample from the top to bottom of the of Se in biological samples, two 200mg aluminum irradiation tube (Tin) is (0.628  replicates of each material (IAEA 436 and 0.021) s for the channel No.13-2 irradiation

NIST 1566b) were weighed and packed in high tube (a length of 6 m) and Tout is (0.323  purity polyethylene bags. The samples were 0.030) s (averaged for 90 runs over the three irradiated for 5, 10, 15, 20, 25, 30, 35 and 40 s. days). For thermal column irradiation tube (a

After a delay of 3.2 s (including both length of 2.8 m), Tin is (0.248  0.019) s and transferring time of sample from irradiation Tout is (0.146  0.004) s, as shown in Table II. position to detector and the time required to The result obtained for measuring the return start the detector). Each sample were counted time from the irradiation position to the for 20 s at a distance of 10 cm from detector. measurement position was found to be (3.165 ± To test accuracy for the analysis of the 0.002) s for channel No.13-2. That for thermal Se concentration in biological reference column was (3.025  0.013) s. It should be materials, four 200 mg replicates of each noted that this timing parameters are included material (IAEA 436, NIST 1566b, NIST 1577 in the time required to start the detector after and NIST 1577b) were weighed. The samples receiving the start signal.

87 A NEW RAPID NEUTRON ACTIVATION ANALYSIS SYSTEM AT …

Table II. The result of time measurements.

Irradiation The transferring time throughout The return time from irradiation position aluminum irradiation tube (second) position to detector position (second)

Tin Tout This word Manufacturer* 13-2 channel 0.628  0.021 0.323  0.030 3.165 ± 0.002 3.301  0.013

Thermal column 0.248  0.019 0.146  0.004 3.025  0.013 3.261  0.022

* Sample weight:  8 g for thermal column tube and  6 g for 13-2 channel tube, operation air pressure:  3.1 bars, distance: 30 meters.

There are significant differences channel No.13-2 and 4.91% for thermal between this work and that of the manufacturer column. The relative error is less than 1% at in capsule sample weight and distance from irradiation time of 5 s for channel No.13-2, and irradiation position to measurement position. 10 s for thermal column. The large error for the Hence, there are differences ( 7%) in the first second is due to delay of the system in result of the return time from irradiation starting the irradiation timer and in ejecting the position to detector position. However, it is not capsule once the “end of irradiation” signal has a problem for analytical measurements. been received. Results for absolute irradiation time at This timing delay problem can be channel No.13-2 and thermal column were adjusted through the control unit and the determined by a series of irradiations ranging software package for managing optimal from 1 to 30 s (3 replicates), as shown in Fig. 3 operation and the analytical procedures. and Fig. 4. The relative error of irradiation time However, it is not a problem for INAA because in the first second is 16.02% for channel the time parameters remain unchanged for all No.13-2 and 26.43% for thermal column, and samples, standards, and control material those for irradiation time of 2 s is 1.5% for

18 28

16 24 14 20 12

10 16

8 12

Relative error, % error, Relative 6 Relative error, % error, Relative 8 4

2 4

0 0 0 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30 0 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30

Irradiation time, second Irradiation time, second

Fig. 3. The relative error of irradiation time Fig. 4. The relative error of irradiation time for for 13-2 channel. thermal column.

88 HO VAN DOANH et al.

B. Neutron spectrum parameters of irradiation positions energy [5], measured using the 58Ni(n,p)58Co nuclear reaction. The thermal neutron flux at The results of the determination of the irradiation position in the thermal column neutron spectra parameters are shown in Table is 1.25E+11 n.cm2.s-1, associated with much III. This table includes data obtained for the lower fast and epithermal neutron flux. thermal, fast neutron flux, the ratio of thermal Hence, thermal column is a useful irradiation to epithermal neutron flux (th/epi). The channel for eliminating interference reactions thermal neutron flux at the irradiation position induced by fast neutron, in which sample is in the channel No. 13-2 is 4.2E+12 n.cm2.s-1, irradiated in an extremely well thermalized and associated with 0.5 times of epithermal. neutron field [6]. The integral fast neutron flux is 6.61E+12 n.cm-2.s-1 for all neutrons above 2.9MeV in

Table III. The results of neutron spectra parameters at irradiation positions in the channel No.13-2 and thermal column of DNRR.

Irradiation position 2 2 th (n/cm /s) F (n/cm /s) th /epi 13-2 channel (4.2  0.1) x 1012 (6.6  0.9) x 1012 10.7  2.4

Thermal column (1.24  0.03) x 1011 (8.4  0.5) x 108 195  4

C. Determination of Selenium sensitivities for Se rapid determination in a variety of biological matrices. Finally, measurements of detection limits of Se in IAEA 436 and NIST 1566b The accuracy for determination of samples were performed. The results for these Selenium using the short-lived nuclide 77mSe measurements are presented in Fig 4. The was evaluated by analyzing a number of obtained results confirm that in irradiation certified reference materials with different from 15 s to 25 s at irradiation position of the levels of Se (IAEA 436, NIST 1566b, NIST channel No.13-2 coupled with counting for 1577 and NIST 1577b). The agreement roughly 20 s at 10 cm distance from detector, between measured and certified values was the detection limits for Se is within the range generally very good with u-score < 1.64, as 0.5  0.7 ppm, depending on the sample shown in Table IV. composition. It provides adequate analytical

1.8 NIST 1566b 1.6 IAEA 436 1.4 1.2 1.0 0.8

detection limit, ppm limit, detection 0.6 0.4 0.2 0 5 10 15 20 25 30 35 40 45

Irradiation time, second

Fig. 4. The detection limits of Se in IEAE 436 and 1566b.

89

A NEW RAPID NEUTRON ACTIVATION ANALYSIS SYSTEM AT

For the determination of the Selenium lived nuclide, not only completion times are a by the instrumental neutron activation analysis, distinct advantage but analytical sensitivities the long-lived nuclide 75Se or the short-lived are also improved. The data for procedures are nuclide 77mSe can be used [7]. With the short- listed in Table V.

Table IV. The results of concentration analysis for Se in biological reference materials.

Certificated k-zero method The relative method Reference value material This work This work (in ppm) u-score u-score (in ppm) (in ppm) IAEA 463 4.63  0.48 4.55  0.50 0.12 4.19  0.46 0.66

NIST 1566b 2.06  0.15 2.48  0.57 0.71 2.18  0.42 0.27

NIST 1577 1.10  0.10 1.24  0.31 0.43 1.17  0.22 0.29

NIST 1577b 0.73  0.06 0.70  0.11 0.24 0.80  0.17 0.39

Table V. Parameters were used for INAA analysis of Selenium in biological sample by using 77mSe and 75Se isotopes.

Radionuclide 75Se 77mSe Half-life 120 d 17.4 s Activation 20 h at 3.5 x 1525 s at 4.2 1012 (n/cm2/s) x 1012 (n/cm2/s) Decay time 20 d 20 s Counting time 23 h 25 s Detection limit 1.4 ppm 0.6 ppm Sample: IAEA 4.63 ± 0.48 4.63 ± 0.48 436 ppm ppm The results 4.35 ± 1.1 ppm 4.19 ± 0.46 ppm

column and channel No.13-2 were also IV. CONCLUSION determined in order to establish analytical A fast pneumatic sample transfer system procedures using the k0-NAA method. The for analyzing of extremely short-lived nuclides system was applied to determine the by neutron activation analysis has been concentration of Se in the biological sample by installed and operated at Dalat nuclear research using the short-lived nuclide 77mSe. The results reactor. In this study, time parameters of the obtained through this research have opened a system were calibrated, thereby reducing new possibility on using INAA technique for irradiation time to seconds with precision. measurement of extremely short-lived nuclides Neutron spectra parameters of the thermal at Nuclear Research Institute.

90

HO VAN DOANH et al.

ACKNOWLEDGEMENTS [4] L.S. McDowell, et al., Determination of Selenium in individual food items using the This project was carried out under the short-lived nuclide 77mSe, Journal of nuclear research and development program of Radioanalytical and Nuclear Chemistry, Vol. the Ministry of Science and Technology, 110, No. 2, p. 519 (1987). Vietnam. [5] A. D. Becker, Characterization and use of the new NIST rapid pneumatic tube irradiation REFERENCES facility, Journal of Radioanalytical Chemistry, Vol. 233, No. 1-2, p. 155 (1998). [1] S.S. Ismail, A new automated sample transfer system for instrumental neutron activation [6] R. Gwozdz, F. Grass, J. Dorner, Fluorine analysis, journal of Automated Methods and analysis of standard materials by short-time 20 Management in Chemistry, Vol. 2010, (2010). activation analysis using F, Journal of Radioanalytical and Nuclear Chemistry, Vol. [2] Yong-Sam Chung, et al., Characteristics of a 169, No.1, p. 57 (1993). new pneumatic transfer system for a neutron activation analysis at the HANARO research [7] D. Behni, et al., Combination of Neutron reactor, Nuclear Engineering and Technology, Activation Analysis, Tracer Techniques, and Vol. 41, No. 6, p. 813 (2009) Biochemical Methods in the Investigation of Selenium Metabolism, Journal of [3] U.M. El-Ghawi, et al., Determination of Radioanalytical and Nuclear Chemistry, p.439 Selenium in Libyan Food Items Using (1989). Pseudocyclic Instrumental Neutron Activation Analysis, Biological Trace Element Research, Vol. 107, p. 61 (2004).

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