E-mail :firstname.lastname@example.org Abstract JRR-3 (Japan Research Reactor No.3) at Tokai site of JAEA is a light water cooled and moderated swimming pool type research reactor with nominal thermal power of 20MW. RR-3 is utilized from academic research to industrial use as neutron beam experiments, irradiation tests of the reactor material, and manufacturing radioisotopes for medicine use and the silicon semiconductor by neutron transmutation doping (NTD). Reactor reached first critical state in 1990 and it was continued operating for 20 years till the Great East Japan Earthquake on 2011.
JRR-3 was in this regular maintenance period, when the Great East Japan Earthquake took place on 11th March 2011. The Great East Japan earthquake's acceleration was measured larger seismic acceleration than that of seismic design of JRR-3. The ground around the buildings was sunk about 40cm. Therefore, some facilities were damaged such as exhaust duct led to a stack and electric transformers in secondary cooling tower. However, the reactor building with the solid foundations and the safety-related facilities survived the earthquake without serious damage, and no radioactive leakage has been occurred. To confirm soundness, checking and testing of the integrity for every components of JRR-3 have been carried out without delay.
Current status of Kazakhstan research reactors I.E.Kenzhina1, E.Ishitsuka2, K.Okumura2,
N.Takemoto2, A.O.Mukanova1 and Y.V.Chikhray1
1Institute of Experimental and Theoretical Physics, Al-Farabi Kazakh National University
2Sector of Nuclear Science Research, Japan Atomic Energy Agency (JAEA)
The overview of research reactors and works related to in-pile experiments in Kazakhstan has been shown in the report. Mainly, such works are activities on investigation of functional and structural materials for reactors, as well as, for fusion blankets. The key schemes and principles of experiments are also introduced.
Current Status and Strategic Plan of
the Dalat Nuclear Research Reactor in the Next Decade Luong Ba Vien*, Nguyen Nhi Dien
Dalat Nuclear Research Institute, VINATOM, 01 Nguyen Tu Luc, Dalat, Vietnam
*Corresponding author:email@example.com Abstract
Dalat Nuclear Research Reactor (DNRR) with the nominal power of 500 kW is today the unique one in Vietnam. The reactor was reconstructed and upgraded from 250-kW TRIGA Mark II research reactor, and restarted operation on March 20, 1984. During the last 32 years of operation, the DNRR was efficiently utilized for reactor physics and neutron beam researches; radioisotope production for medical and other uses; sample irradiation for neutron activation analysis (NAA); and education and training for nuclear human resource development in the country.
In the next ten years, together with ensuring for safe operation, the reactor will be considered to implement the operation regime of 150 hrs/cycle in order to increase the quantity of radioisotope production as well as the number of irradiation samples for NAA. The use of the reactor for nuclear education and training in order to support the human resource development for the national nuclear power program will be paid much attention. Besides, in preparation for the effective utilization of a new research reactor in future, the broadening of researches on production of various radioisotopes, development of neutron scattering and neutron radiography facilities will also be planned to carry out at the DNRR.
This paper presents the current status of operation and utilization of the DNRR. In addition, the strategic plan of the reactor in the next decade is also given and discussed.
Current status of JRR-3 (2): Maintenance in long-term shutdown period and Work for Re-operation Eigo KAMIISHI, Shigeru WADA and Yoji MURAYAMA
Department of Research Reactors and Tandem Accelerator
Nuclear Science Research Institute, Japan Atomic Energy Agency
JRR-3 has been in shutdown for approximately five years after the last reactor operation. The so long-term reactor shutdown period is our first experience after first criticality. During the shutdown period, main facilities such as cooling system and safety system have been kept in good operating condition by checking operation regularly once in a month. In addition to the hardware side, we feel the importance of keeping and training skills of operators. Therefore, we carried out training using reactor simulator and operation of the most of reactor facilities (the cooling system, instrumentation and control system, etc.) for 10 days a year.
In response to the accident at Fukushima Daiichi NPS, the Reactor Regulation Act was revised in June 2012, for the purpose of introducing new regulations based on lessons learned from the accident and the latest technical analysis. The Act requires the application of ‘back-fit system' that is a system for adopting the latest technological findings and obligating approved nuclear facilities to conform to the new requirements. We have conducted the necessary checks and reassessments of JRR-3, and completed the preparations for the relevant applications. We then submitted the application document to change the current reactor license to the Nuclear Regulation Authority (NRA) for their review to verify conformity to new regulatory requirements on September 26, 2014.
Capabilities and Capacities of JMTR towards Deployment
of Innovative Nuclear Energy Systems and Technologies T. Kusunoki, M.Tanimoto, M. Kaminaga, M.Ishihara and M. Araki
Japan Atomic Energy Agency, 4002 Narita, Oarai, Higashiibaraki, Ibaraki 311-1393, Japan
Tel : +81-29-266-7001, Fax. : +81-29-266-7471,
The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a testing reactor dedicated to the irradiation tests of materials, fuels and RI productions. It achieved first criticality in March 1968, and is being operated at thermal power of 50MW at about seven operation cycles a year, with about 30 operation days a cycle. In August 2006, operation of the JMTR was terminated due to assessment for the operation continuation decision. As a result of the national discussion, the JMTR was decided to continue operation after necessary refurbishment works. The refurbishment was finished after four years on schedule in March 2011 (JFY2010).
However, at the end of the JFY 2010 just after the completion of the refurbishment, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart were delayed. On the other hand, based on the safety assessments considering the 2011 earthquake new regulatory requirements for research and test reactors have established on Dec.18, 2013 by the NRA (Nuclear Regulation Authority). After evaluation and analyses to the new regulatory requirements were carried out, an application to the NRA was submitted on March 27, 2015.
After taking measures for the safety requirements and the permission by the NRA, the renewed JMTR will be operated for a safety research of LWRs, basic research for nuclear engineering such as HTGR and nuclear fusion research, industrial use such as high burn-up experiment of the current LWR fuel, production of Mo-99, and education & training of nuclear scientists and engineers.
In this report, current status and technologies catalog of the JMTR is described. The technologies catalog of the JMTR can be viewed in the technologies catalogue of research reactors in IAEA website. The users of the catalogue will be governmental and private sector organizations responsible for the development and/or deployment of innovative nuclear energy systems including designers, manufacturers, vendors, research institutions, academia and other organizations directly involved in the development of materials, fuels and RI productions for nuclear energy industry.
A Proposal For Material Testing Study Using Multipurpose Research Reactor Tran Chi Thanh, Nguyen Nhi Dien
VINATOM, 59 Ly Thuong Kiet, Hanoi, Vietnam
A multipurpose research reactor with nominal power of about 15 MW has been planned for a new Centre for Nuclear Energy Science and Technology (CNEST) of Vietnam in the next ten-year period. Besides traditional applications of low power research reactors for nuclear science research, education and training, and services, this high flux reactor has been planned for study on the material testing and other applications. Preliminary proposal for effective utilization of the new multipurpose research reactor of CNEST is given in this report for discussion and getting ideas from experienced experts in related areas.
Session2: Facilities and Utilization
Utilization of research reactors in Vietnam for
research, training and service activities Ho Manh Dung*, Luong Ba Vien and Nguyen Nhi Dien
Dalat Nuclear Research Institute, VINATOM
01 Nguyen Tu Luc, Dalat, Vietnam
*Corresponding author: firstname.lastname@example.org Abstract
The Dalat research reactor in Nuclear Research Institute (NRI) of Vietnam is a pool type reactor with 500 kW normial power. In addition, a new research reactor with about 15 MW power has been planned for construction in the next ten-year period. The research reactors in Vietnam are mainly utilized for nuclear science and engineering research, education and training, and service activities. Utilization of the Dalat research reactor is illustrated by examples which have been carried out within the last thirty years as well as the preparation of human resources for the utilization of the new research reactor in Vietnam may be needed in the future will be anticipated and discussed in this presentation.
Utilization of RSG-GAS Reactor for Power Ramp Experiment
Yusi Eko Yulianto
RSG-GAS reactor BATAN has been operated since 29 years a go has an experimental Facility which called as Power Ramp Test Facility (PRTF). That PRTF is a close loop high pressure and heat removal process, which consists of primary loop and secondary loop. The primary loop is a high pressure of water circulation system which has an order to produce a high pressure for target Capsule. The secondary loop is a water cooling system which circulate the reactor cooling system to remove the heat from the capsule. The PRTF has been installed with the distance of 420 mm from the reactor core. The installation can be moved by controlling near and go away from the core. The Capsule is put on the capsule carrier which is installed at the trolley. The trolley can be move remotely go in and out of the reactor core. During the reactor operation target capsule has the opportunity to get the level of neutron exposure depending on the adjusted position from the reactor core. The experiment on PRTF is purposed to get differences characteristic of mechanical and nuclear reaction behavior of the nuclear fuel pellet during the experiment by different position and time exposure. The physical and nuclear characteristic of the pellet is very important in research and development of the technology of the fuel element fabrication. Safe operation of PRTF is a priority in the reactor operation. So, the safe operation should be achieved to avoid the risk to the reactor operation. Facility has important 4 parameters to protect the system against parameter expansions. Specific number of Operating Limit Condition (OLC) for Pressure expansion in the primary loop is 155< and >158 Bar, for temperature is 23 oC, for cooling flow in secondary loop is 600< and >900 liters/hour and radioactivity in primary loop not exceed 2,08x104 counts/second. Action of the protection if system across the OLC is REEX (Return Experiment) are: automatic execution by the system to send the capsule back to the save position from the reactor core without any risk of stable nuclear reactions.
Utilization of irradiation facility at reactor triga puspati
Muhammad Rawi Mohamed Zin, Zarina Masood, Abdul Aziz Ramli,
Mohd Ashaar Khalid, Mohd Fairuz Farid
Reactor Technology Center, Malaysian Nuclear Agency. Abstract
Reactor TRIGA PUSPATI (RTP) has been utilized since its first criticality for various purposes such as irradiation of samples for Neutron Activation Analysis, Radioisotope production for both industrial and radiopharmaceutical tracer applications and neutron beam port applications. Depending on the purpose, neutron irradiation could be made in dry tube located almost within the reactor centre or in the rotary rack at the outer perimeter of reactor core, or in the few places in reactor core via Pneumatic Transfer System (PTS). These irradiation facilities are also available for researching some materials for making an electronic component or nuclear detector system. In addition to in core irradiation facilities, there are four horizontal irradiation channels or beam ports available at RTP. Two beam ports respectively utilized for Small Angle Neutron Scattering and Neutron Radiography Imaging (NURI). NURI instrument has been frequently used for getting internal images of various materials for example, woods and fossilized objects. Beside the utilization of existing facility, there are some studies have been conducted to improve thermal neutron flux at the reactor core with modelling new design of reactor core. Therefore, this article presents the status of irradiation facility and the plan to improve the capability of the facility.
Development of New Instruments for Irradiation Tests
with Research/Testing Reactors Yoshinori MATSUI and Kunihiko TSUCHIYA
Japan Atomic Energy Agency, 4002 Narita, Oarai, Higashiibaraki, Ibaraki 311-1393, Japan
Tel. : +81-29-266-7035, Fax. : +81-29-266-7071,
The JMTR has been widely used for the irradiation tests to meet user requirements, and many types of irradiation techniques have been developed since started JMTR irradiation service from 1971. Recent irradiation research needs more accurate control and evaluation of environmental parameters such as temperature, neutron flux/fluence, surrounded conditions of sample and so on. After re-operation, the JMTR is planned to contribute the worldwide research fields and industrial fields. Expected roles of the JMTR are the safety research for LWRs (aging management, development of next generation LWR and safety measure development for LWRs), progress of science and technologies (contribution to fusion reactor and HTGR, basic research), expansion of industrial uses (production of 99Mo/99mTc and Si semiconductor) and education and training.
It is necessary to measure various mechanical/thermal data under neutron irradiation conditions by the in-situ instruments. In this study, the attractive irradiation tests using (1) new instruments as multi-paired thermocouples, (2) radiation monitors of SPND (Self-Powered Neutron Detector) and SPGD (Self-Powered Gamma Detector), (3) chemical sensors of ECP and hydrogen gas sensors and (4) Linear Voltage Differential Transformer (LVDT) under neutron irradiation have been developed for the application at JMTR re-operation and standardization of irradiation tests in the research reactors.
Firstly, it is difficult to install many thermocouples in the irradiation capsule because of small diameter of the capsule. The K and N type multi-paired thermocouples were developed to evaluate the axial temperature distribution of the irradiation species. The multi-paired thermocouple is 1.8mm in diameter and has maximum 7 hot junctions. Secondly, the SPGD and SPND with only a few millimeters in diameter have been developed as the radiation monitors. These monitors are composed of three parts, i.e. emitter, collector, and insulator. The emitters of SPND and SPGD were selected to high cross-section of neutron capture and Compton scattering materials, respectively. Thirdly, in the IASCC irradiation tests, it is important to develop the ECP sensor for measurement water chemistry. Joining method between ceramics and metal was selected to the brazing and the durability was evaluated under the high-pressure and temperature water conditions. This joining method was also applicable to hydrogen gas sensor. Fourthly, the FP gas pressure gauges with LVDT have been developed to evaluate the fuel behavior in the fuel element. Various in-situ instruments have been developed for advanced neutron irradiation tests and the basic properties of these instruments have been accumulated to obtain for high-precision data in the out-pile and in-pile tests.
Yoshinori MATSUI 1, Kuniaki MIURA 2 and Kunihiko TSUCHIYA 1
1: Japan Atomic Energy Agency, 4002 Narita, Oarai, Higashiibaraki, Ibaraki 311-1393, Japan
2: SUKEGAWA ELECTRIC CO.,LTD, Kamitezuna 3333-23 Takahagi, Ibaraki, 318-0004, Japan
Tel : +81-29-266-7034, Fax. : +81-29-266-7071,
In June 2011, the Japanese government referred to the instruction of the accident at the Fukushima Dai-ichi (1F) Nuclear Power Plant (NPP), in the report of Japanese government to the IAEA ministerial conference. In accordance with such situation, we have begun a research and development which corresponds to the provisions so as to monitor the NPPs situations during a severe accident from 2012. Considering that reactor pressure vessel is broken under severe accidents, we realized that a development of the measurement data line is necessary. Thus, it is important for reactor monitoring system to develop the metal sheathed mineral insulated cable (MI cable) which can transmit measurement data even in severe accident conditions.
In this study, we modified sheath material and insulating material for metal sheathed K-type thermocouples (T/Cs). These test samples of T/Cs were combinations of type 316 stainless steel (316SS) or Inconel alloy (NCF600) as the sheath material, and high purity MgO or Al2O3 as the insulating material were selected as the insulating material for the high heat resistance T/Cs. We tested four types of T/Cs, which were prepared. The electric properties such as insulation resistance and conduction resistance of these T/Cs were evaluated in the PWR condition (325°C×15MPa) and the conditions under vacuum or air up to 1000°C simulated the severe accident.
In the PWR condition, conduction resistance and insulation resistance of these T/Cs were constantly, regardless of difference of insulating material or sheath material. In the condition under vacuum, the insulation resistance decreased with temperature despite the difference in insulating material. In the condition under air at 1000°C, the insulation resistance of T/Cs with 316SS was not able to be measured after about 100 h. After the experiments, the sheath of each T/C with 316SS was broken and the insulating material was observed from the crack of sheath material. It is considered that the oxidation of 316SS occurred during the experimental condition. On the other hand, the insulation resistance of T/Cs with NCF600 was constant during the experimental condition (about 100 h) and the sheath material of the T/Cs with NCF600 was not broken. From the results, it was found that the T/Cs with NCF600 as the sheath material, is useful under the severe accident.
The project is supported by R&D program for Plant Safety Enhancement of the Agency for Natural Resources and Energy, Ministry of Economy, Trade and Industry (METI), Japan.
Session 4: Radioisotope
Evaluation of 99Mo/99mTc Production with Irradiated MoO3 Pellets Takuya ISHIDA1, Takayuki SHIINA2, Akio OHTA2, Yoshitaka SUZUKI1
3:FUJIFILM RI Pharma Co. Ltd, 453-1 Shimo-okura, Matsuo, Sammu, Chiba 289-1592, Japan
4: Kyoto University Research Reactor, 2, Asashiro-Nishi, Kumatori, Sennan, Osaka 590-0494, Japan
Tel.: +81-29-266-7033, Fax. : +81-29-266-7071,
Technetium-99m (99mTc) is one of the radioisotopes, which are used most as radiopharmaceuticals, and it is obtained from the parent nuclide of Molybdenum-99（99Mo）. 99Mo is commercially produced in nuclear reactors by nuclear fission ((n,f) method) of 235U. To minimize the radio-active wastes, neutron activations ((n, γ) method) of 98Mo or (n, 2n) method) of 100Mo are intensively performed. Especially the research and development (R&D) on the production of 99Mo/99mTc by (n, γ) method have been carried out in the Japan Materials Testing Reactor (JMTR). However, the specific radioactivity of 99Mo by (n, γ) method is extremely low compared with that by fission method ((n,f) method), and as a result, the radioactive concentration of the extracted 99mTc solution is also low. Thus, it is necessary for the high radioactive concentration of the 99mTc solution to develop the 99Mo/99mTc separation/extraction/concentration method.
In this study, the experiments of 99Mo/99mTc production were performed to enhance recovery yields of 99mTc and to get a high quality of 99mTc product. The procedures are described as follows. (1) High-density MoO3 pellets were irradiated in the Kyoto University Reactor (KUR). (2) 99mTc was extracted with MEK. (3) 99mTc extracted in MEK was purified and concentrated with acidic alumina column. (4) Product of 99mTc solution was checked in several factors such as radionuclidic and radiochemical purities.
At the first step, the 99Mo/99mTc separation/extraction/concentration devices were improved in the cold tests with rhenium (Re) and the recovery rate achieved about 100%.
At the second step, the irradiated MoO3 pellets were dissolved in 6M-NaOH and the 99Mo/99mTc solution was treated with the devices. From the results, the 99mTc recovery rates achieved 80±5% of our goal. Finally, the extracted 99mTc solution was evaluated as the commercial product and the amounts of Al, MEK and endotoxin, pH, radionuclide purity, radiochemical impurity, and the osmotic pressure of the 99mTc solution were suitable for regulation of raw materials of radiopharmaceuticals written in the guidance.
Experiences on the irradiation for radioisotope production at Dalat Nuclear Research Reactor Duong Van Dong
Dalat Nuclear Research Institute, VINATOM, 01 Nguyen Tu Luc, Dalat, Vietnam
This is an outline of the experiences on the irradiation for radioisotope production programme using a 500-kW low power research reactor in Vietnam.
The production laboratories and facilities including the nuclear reactor with its irradiation positions and characteristics, hot cells, production lines and equipment for the production and 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.
Keyword: Irradiation, Radioisotope production, research reactor, radiopharmaceuticals preparation Beryllium Production and Applications Overview Including Fission Test Reactor-Related Activities
C.K. Dorn1, T. Ong2, E.E. Vidal1, And K.J. Smith1
1Materion Beryllium & Composites 14710 West Portage River South Road, Elmore, Ohio 43416-9500 U.S.A.
Materion Corporation is the world’s largest producer of beryllium metal. Beryllium (Be) metal and other beryllium-containing materials are known for their unusual combination of properties. This accounts for the long-standing interest in these materials since the 1930s. Important fields of application for beryllium metal, beryllium-containing composites, and beryllium compounds include acoustics, aerospace structures, x-ray transmission, motion control, nuclear test reactors (both fission and fusion), laser-based optical systems, high-energy particle physics research, high-performance automotive applications, and thermal management. Beryllium-containing alloys, which typically contain less than 2% Be, are used extensively in commercial electronics, telecommunications infrastructure, automotive electronics, oil and gas equipment, tooling for plastic molding, and medical equipment applications. This paper will describe the processes used at Materion to make beryllium as well as highlighting important end- use applications, including reflectors in nuclear fission materials test reactors.
Session 5: Human resource development/
Human Resource Development Of Nuclear Researcher/Engineer Using JMTR And Related Facilities
M. Tanimoto, S.Eguchi, N.Takemoto, T. Kusunoki, and M. Araki
Neutron Irradiation and Testing Reactor Center, Japan Atomic Energy Agency
4002 Narita, Oarai, Higashiibaraki, Ibaraki, 311-1393, Japan affiliation
The JMTR (the Japan Materials Testing Reactor) is expected to be a key infrastructure with related facilities to contribute the nuclear Human Resource Development (HRD) by a research and On-the-Job-Training (OJT) in order to support global expansion of nuclear industry. In particular the training program for foreign young researchers and engineers were started from JFY 2011.
Using JMTR and its facilities, various trainings such as an operations, simulation and examinations are performed for engineers, students of domestics and foreigner including non-nuclear personnel.
This program includes lectures on the bases of nuclear energy, the irradiation research and safety management at the JMTR, and the practical trainings about neutronic and thermal design, the reactor operating of the JMTR facility using simulation system, the handling irradiated specimen at the JMTR Hot Laboratory. Having completed an appropriate course in a couple of weeks, the participants also experience various technical tour such as, other research reactor, JRR-3 JAEAs other Research reactor, RI production facility of private company and commercial BWR reactor. Contents of activities are follows:
In these programs, basic understanding on irradiation test and post irradiation examination is aimed to achieve by overall and practical training such as the neutronic/thermal designs of irradiation capsule, evaluation of neutron fluence and post irradiation examination, etc.
(b) Training of neutronic calculation
An analysis procedure using Monte Carlo method has been carried out in irradiation tests of JMTR to evaluate irradiation field at each specimen by using MCNP code.
(c) Training of Reactor Operation using Simulator
This program can simulate events and actions on normal and the accident conditions in the reactor and the irradiation facility of JMTR.
This practical training course using the JMTR and related facilities has been provided by Neutron Irradiation Testing Reactor Center for foreign young researchers and engineers in Asian and other countries which are planning to introduce power reactors. The aim of this course is to contribute to the human resource development in nuclear research field and to increase the future use the JMTR.
Evaluation of Curve for Tritium Release Rate into Primary Coolant
for Research and Testing Reactors
I.E.Kenzhina1, E.Ishitsuka2, K.Okumura2,
N.Takemoto2, A.O.Mukanova1 and Y.V.Chikhray1 1Institute of Experimental and Theoretical Physics, Al-Farabi Kazakh National University
2Sector of Nuclear Science Research, Japan Atomic Energy Agency (JAEA)
Increase of tritium concentration in the primary coolant for research and testing reactors during reactor operation had been reported. To clarify the tritium sources, a curve of tritium release rate into the primary coolant for the JMTR and JRR-3M are evaluated. As a result, the tritium release rate is related with produced 6Li by (n,α) reaction from 9Be, and evaluation results of tritium release curve are shown as the dominant source of tritium release into the primary coolant for the JMTR and JRR-3M are beryllium components. Scattering of the tritium release rate with irradiation time were observed, and this phenomena in the JMTR occurred in earlier time than that of the JRR-3M.
Neutronic Calculation and Design of Irradiation Containers for
In-core Irradiation at the Dalat Research Reactor Nguyen Minh Tuan*, Tran Quoc Duong, Tran Quang Thien
The Dalat reactor is a 500 kW, tank-type, beryllium and graphite reflected, light-water cooled and moderated nuclear research reactor which has been designed mainly for radioisotope production and neutron activation analysis (NAA). The design requirements of in-core irradiation should ensure nuclear and radiation safety, radioactive waste reduction and high efficiency of using thermal neutron.
This report presents neutronic calculation and design of irradiation containers for in-core irradiation that help significantly increase production of radioisotope and minimize radiation exposure for the worker involved and allow us to apply NAA technique using epithermal neutron.