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JAEA Reports

Evaluation of the minimum critical amount for heterogeneous lattice systems composed of fuel rods utilized in low-power water-moderated research and test reactors by using continuous-energy Monte Carlo code MVP with JENDL-4.0

Yanagisawa, Hiroshi

JAEA-Technology 2021-023, 190 Pages, 2021/11


Computational analyses on nuclear criticality characteristics were carried out for heterogeneous lattice systems composed of water moderator and fuel rods utilized in lowpower research and test reactors, in which the depletion of fuel due to burnup is relatively small, by using the continuous-energy Monte Carlo code MVP Version 2 with the evaluated nuclear data library JENDL-4.0. In the analyses, the minimum critical number of fuel rods was evaluated using calculated neutron multiplication factors for the heterogeneous systems of the uranium dioxide fuel rod in the Static Experiments Critical Facility (STACY) and the Tank-type Critical Assembly (TCA), and the uranium-zirconium hydride fuel rod in the Nuclear Safety Research Reactor (NSRR). In addition, six sorts of the ratio of reaction rates, which are components of neutron multiplication factors, were calculated in the analyses to explain the variation of neutron multiplication factors with the ratio of water moderator to fuel volume in a unit fuel rod cell. Those results of analyses are considered to be useful for the confirmation of reasonableness and validity of criticality safety measures as data showing criticality characteristics for water-moderated heterogeneous lattice systems composed of the existing fuel rods in research and test reactors, of which criticality data are not sufficiently provided by the Criticality Safety Handbook.

JAEA Reports

HAND: A Handy criticality analysis tool for fuel debris

Tada, Kenichi

JAEA-Data/Code 2020-014, 30 Pages, 2020/10


The decommissioning of TEPCO's Fukushima Daiichi Nuclear Power Plant accident is one of the most important issues in Japan. In the process of the decommissioning, preventing radiation exposure of workers is imperative originating in nuclear criticality of fuel debris. This study provides the handy tool enabling the analysis on nuclear criticality of fuel debris. The developed analysis tool named as HAND enables estimation of the criticality of fuel debris in short time. HAND deduces the range of parameters such as the size and composition, in which the criticality of fuel debris is specified. By selecting the range of the parameters using HAND in advance, total calculation time of the detail analysis will be reduced. Since the input data of HAND is designed to be simple and the output of HAND is to be user friendly, this tool is expected to be also an intuitive tool to study the criticality of fuel debris. This report explains the outline of the HAND and input instructions for HAND.

JAEA Reports

Basis for handling of nuclear fuel materials (Second edition)

Task Force on Writing Textbook of Nuclear Fuel Materials

JAEA-Review 2020-007, 165 Pages, 2020/07


The present textbook was written by Task Force on Writing Textbook of Nuclear Fuel Materials at the Nuclear Science Research Institute in order to improve technological abilities of engineers and researchers who handle nuclear fuel materials. The taskforce consists of young and middle class engineers each having certification for chief engineer of nuclear fuel. The present textbook mainly deals with uranium and plutonium, and shows their nuclear properties, physical and chemical properties, and radiation effects on materials and human body. It also presents basic matters for safety handling of nuclear fuel materials, such as handling of nuclear fuel materials with hood and glovebox, important points in storage and transportation of nuclear fuel materials, radioactive waste management, radiation safety management, and emergency management. Furthermore, incident cases at domestic and foreign nuclear fuel materials facilities are compiled to learn from the past.

Journal Articles

Development of a handy criticality analysis tool for fuel debris

Tada, Kenichi

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 4 Pages, 2019/05

The decommissioning of Fukushima Daiichi Nuclear Power Plant accident is one of the most important issues in Japan. The criticality safety of fuel debris is imperative to prevent exposure of workers. The investigating criticality monitoring system cannot detect the criticality of fuel debris quickly. The estimation of criticality of fuel debris is required for the fuel debris retrieval. Though the expert knowledge of reactor physics is necessary to estimate the criticality of fuel debris, many people who make a plan of fuel debris retrieval may not know well about criticality analysis. We developed a handy criticality analysis tool HAND to quickly estimate the criticality of fuel debris without expert knowledge of reactor physics. Since the input data of HAND is so simple and users can intuitively understand the calculation results, this tool is expected to be the effective tool to estimate the criticality of fuel debris.

Journal Articles

Improvement of neutron startup source handling work by developing new transportation container for High-Temperature engineering Test Reactor (HTTR)

Shimazaki, Yosuke; Sawahata, Hiroaki; Shinohara, Masanori; Yanagida, Yoshinori; Kawamoto, Taiki; Takada, Shoji

Journal of Nuclear Science and Technology, 54(2), p.260 - 266, 2017/02

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

The High-Temperature engineering Test Reactor (HTTR) has three neutron startup sources (NSs) in the reactor core, each of which consists of $$^{252}$$Cf with 3.7 GBq and is contained in a small capsule, installed in NS holder and subsequently in a control guide block (CR block). The NSs are exchanged at the interval of approximately 7 years. The NS holders are transported from the dealer's hot cell to the reactor facility of HTTR using a transportation container. The loading work of NS holders to the CR blocks is subsequently carried out in the fuel handling machine maintenance pit of HTTR. Technical issues, which are the reduction and prevention of radiation exposure of workers and the exclusion of falling of NS holder, were extracted from the experiences in past two exchange works of NSs to develop a safety handling procedure. Then, a new transportation container special to the NSs of HTTR was developed to solve the technical issues while keeping the cost as low as that for overhaul of conventional container. As the results, the NS handling work using the new transportation container was safely accomplished by developing the new transportation container which can reduce the risks of radiation exposure dose of workers and exclude the falling of NS holder.

JAEA Reports

Shielding calculation by PHITS code during replacement works of startup neutron sources for HTTR operation

Shinohara, Masanori; Ishitsuka, Etsuo; Shimazaki, Yosuke; Sawahata, Hiroaki

JAEA-Technology 2016-033, 65 Pages, 2017/01


To reduce the neutron exposure dose for workers during the replacement works of the startup neutron sources of the High Temperature Engineering Test Reactor, calculations of the exposure dose in case of temporary neutron shielding at the bottom of fuels handling machine were carried out by the PHITS code. As a result, it is clear that the dose equivalent rate due to neutron radiation can be reduced to about an order of magnitude by setting a temporary neutron shielding at the bottom of shielding cask for the fuel handling machine. In the actual replacement works, by setting temporary neutron shielding, it was achieved that the cumulative equivalent dose of the workers was reduced to 0.3 man mSv which is less than half of cumulative equivalent dose for the previous replacement works; 0.7 man mSv.

JAEA Reports

Fabrication and test results of testing equipment for remote-handling of MA fuel, 3; Testing equipment for fuel loading

Tazawa, Yujiro; Nishihara, Kenji; Sugawara, Takanori; Tsujimoto, Kazufumi; Sasa, Toshinobu; Eguchi, Yuta; Kikuchi, Masashi*; Inoue, Akira*

JAEA-Technology 2016-029, 52 Pages, 2016/12


Transmutation Physics Experimental Facility (TEF-P) planned in the J-PARC project uses minor actinide (MA) fuels in the experiments. These MA fuels are highly-radioactive, so the fuel handling equipment in TEF-P is necessary to be designed as remote-handling system. This report summarizes fabrication and test results of the testing equipment for fuel loading that is one of components of the testing equipment for remote-handling of MA fuels. The testing equipment which had a remote-handling system for fuel loading was fabricated. And the test in combination with the mock-up core was performed. Through the test, it was confirmed to load/take the dummy fuel pin to/from the mock-up core without failure. It was shown that the concept design of the fuel loading equipment of TEF-P was reasonable.

Journal Articles

Dissolution behavior of lithium compounds in ethanol

Furukawa, Tomohiro; Hirakawa, Yasushi; Kondo, Hiroo; Kanemura, Takuji

Nuclear Materials and Energy (Internet), 9, p.286 - 291, 2016/12


 Times Cited Count:2 Percentile:25.63(Nuclear Science & Technology)

In order to exchange the components which received irradiation damage during the operation at the International Fusion Materials Irradiation Facility, the adhered lithium, which is partially converted to lithium compounds such as lithium oxide and lithium hydroxide, should be removed from the components. In this study, the dissolution experiments of lithium compounds (lithium nitride, lithium hydroxide, and lithium oxide) were performed in a candidate solvent, allowing the clarification of time and temperature dependence. Based on the results, a cleaning procedure for adhered lithium on the inner surface of the components was proposed.

Journal Articles

Improvement of exchanging method of neutron startup source of high temperature engineering test reactor

Sawahata, Hiroaki; Shimazaki, Yosuke; Ishitsuka, Etsuo; Yamazaki, Kazunori; Yanagida, Yoshinori; Fujiwara, Yusuke; Takada, Shoji; Shinozaki, Masayuki; Hamamoto, Shimpei; Tochio, Daisuke

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 8 Pages, 2016/06

In the HTTR, $$^{252}$$Cf is loaded in the reactor core as a neutron startup source and changed at frequency. In this exchange work, there were two technical issues; slightly higher radiation exposure of workers by neutron leakage and reliability of neutron source transportation container in handling. To reduce the radiation dose by neutron leakage, detail numerical evaluations using PHITS code were carried out, the effective shielding method for fuel handling machine was proposed. Easily removable polyethylene blocks and particles were used as the neutron shielding, and installed in the cooling paths of the fuel handling machine. As a result, the collective effective dose by neutron was reduced from about 700 man-microSv to about 300 man-microSv. As to the neutron source transportation container, the handling performance was improved and the handling work was safety accomplished by downsizing.

Journal Articles

Development of transportation container for the neutron startup source of High Temperature engineering Test Reactor (HTTR)

Shimazaki, Yosuke; Ono, Masato; Tochio, Daisuke; Takada, Shoji; Sawahata, Hiroaki; Kawamoto, Taiki; Hamamoto, Shimpei; Shinohara, Masanori

Proceedings of International Topical Meeting on Research Reactor Fuel Management and Meeting of the International Group on Reactor Research (RRFM/IGORR 2016) (Internet), p.1034 - 1042, 2016/03

In High Temperature Engineering Test Reactor (HTTR), three neutron holders containing $$^{252}$$Cf with 3.7 GBq for each are loaded in the graphite blocks and inserted into the reactor core as a neutron startup source which is changed at the interval of approximately ten years. These neutron holders containing the neutron sources are transported from the dealer's hot cell to HTTR using the transportation container. The holders loading to the graphite block are carried out in the fuel handling machine maintenance pit of HTTR. There were two technical issues for the safety handling work of the neutron holder. The one is the radiation exposure caused by significant movement of the container due to an earthquake, because the conventional transportation container was so large ($$phi$$1240 mm, h1855 mm) that it can not be fixed on the top floor of maintenance pit by bolts. The other is the falling of the neutron holder caused by the difficult remote handling work, because the neutron holder capsule was also so long ($$phi$$155 mm, h1285 mm) that it can not be pulled into the adequate working space in the maintenance pit. Therefore, a new and low cost transportation container, which can solve the issues, was developed. To avoid the neutron and $$gamma$$ ray exposure, smaller transportation container ($$phi$$820mm, h1150 mm) which can be fixed on the top floor of maintenance pit by bolts was developed. In addition, to avoid the falling of the neutron holder, smaller neutron holder capsule ($$phi$$75 mm, h135 mm) with simple handling mechanism which can be treated easily by manipulator was also developed. As the result of development, the neutron holder handling work was safely accomplished. Moreover, a cost reduction for manufacturing was also achieved by simplifying the mechanism of neutron holder capsule and downsizing.

Journal Articles

JSFR design progress related to development of safety design criteria for generation IV sodium-cooled fast reactors, 4; Balance of plant

Chikazawa, Yoshitaka; Kato, Atsushi; Nabeshima, Kunihiko; Otaka, Masahiko; Uzawa, Masayuki*; Ikari, Risako*; Iwasaki, Mikinori*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05

Design study and evaluation for SDC and safety SDG on the BOP of the demonstration JSFR including fuel handling system, power supply system, component cooling water system, building arrangement are reported. For the fuel handling system, enhancement of storage cooling system has been investigated adding diversified cooling systems. For the power supply, existing emergency power supply system has been reinforced and alternative emergency power supply system is added. For the component cooling system and air conditioning, requirements and relation between safety grade components are investigated. Additionally for the component cooling system, design impact when adding decay heat removal system by sea water has been investigated. For reactor building, over view of evaluation on the external events and design policy for distributed arrangement is reported. Those design study and evaluation provides background information of SDC and SDG.

Journal Articles

A Possible breakthrough of power handling by plasma shaping in tokamak

Kikuchi, Mitsuru; Fasoli, A.*; Takizuka, Tomonori*; Diamond, P. H.*; Medvedev, S.*; Wu, Y.*; Duan, X.*; Kishimoto, Yasuaki*; Hanada, Kazuaki*; Pueschel, M. J.*; et al.

Proceedings of 8th IAEA Technical Meeting on Steady State Operation of Magnetic Fusion Devices (CD-ROM), 20 Pages, 2015/05

The standard D shaped H-mode operation showed excellent plasma confinement ut has important issues of transient and steady state heat flux. To solbe this issues, we proposed new scenario using plasma shaping as one of possible scenario of future tokamak reactor.

Journal Articles

Simplified calculation method for radiation streaming

Matsuda, Norihiro

Hoshasen Shahei Handobukku; Kisohen, p.229 - 288, 2015/03

no abstracts in English

Journal Articles

Current status of CrossRef. revisited

Nagaya, Shun

Karento Aweanesu, (322), p.13 - 17, 2014/12

no abstracts in English

Journal Articles

Study of SOL/divertor plasmas in JFT-2M

Kawashima, Hisato; Sengoku, Seio; Uehara, Kazuya; Tamai, Hiroshi; Shoji, Teruaki*; Ogawa, Hiroaki; Shibata, Takatoshi; Yamamoto, Masahiro*; Miura, Yukitoshi; Kusama, Yoshinori; et al.

Fusion Science and Technology, 49(2), p.168 - 186, 2006/02

 Times Cited Count:2 Percentile:18.25(Nuclear Science & Technology)

Experimental efforts on JFT-2M have been devoted to understand SOL/Divertor plasmas and to investigate power and particle controllability. Open divertor configuration was used for the first decade of JFT-2M started in 1984. We found out the SOL/Divertor plasma properties such as in/out asymmetry, heat and particle diffusivities, and SOL current at ELMs. Handling of power and particle was demonstrated by active control methods such as local pumping, edge ergodization, divertor biasing, and edge heating. For improvement of power and particle control capability of divertor, it was modified to closed configuration in 1995, which demonstrated the baffling effects with narrower divertor throat. Dense and cold divertor state (n$$_{e}$$$$^{div}$$ = 4$$times$$10$$^{19}$$ m$$^{-3}$$ and T$$_{e}$$$$^{div}$$ = 4 eV), compatible with the improved confinement modes (e.g. H-mode), was realized by strong gas puffing. Being related with the core confinement at H-mode, the edge plasma fluctuations were identified by an electrostatic probe. These are reviewed in this paper.

JAEA Reports

Remote handling design for moderator-reflector maintenance in JSNS

Teshigawara, Makoto; Aizawa, Hideyuki; Harada, Masahide; Kinoshita, Hidetaka; Meigo, Shinichiro; Maekawa, Fujio; Kaminaga, Masanori; Kato, Takashi; Ikeda, Yujiro

JAERI-Tech 2005-029, 24 Pages, 2005/05


This report introduces the present design status of remote-handling devices for activated and used components such as moderator and reflector in a spallation neutron source of the Material and Life Science Facility (MLF) at J-PARC. The design concept and maintenance scenario are also mentioned. A key maintenance scenario adopts that the used components should be taken out from the MLF to the other storage facility after the volume reduction of them. Almost full remote handling is available to the maintenance work except for the connection/disconnection pipes of the cooling water. Total six remote handling devices are used for moderator-reflector maintenance. They are also available to the proton beam window and muon target maintenance. Maintenance scenario is separated into two works. One is to replace used components to new ones during beam-stop and the other is dispose used components during beam operation. Required period of replacement work is estimated to be $$sim$$15 days, on the other hand, the disposal work is $$sim$$26 days after dry up work ($$sim$$30 days), respectively.

Journal Articles

Short design descriptions of other systems of the HTTR

Sakaba, Nariaki; Furusawa, Takayuki; Kawamoto, Taiki; Ishii, Yoshiki; Ota, Yukimaru

Nuclear Engineering and Design, 233(1-3), p.147 - 154, 2004/10

 Times Cited Count:10 Percentile:58.07(Nuclear Science & Technology)

The HTTR mainly consists of the core components, reactor pressure vessel, cooling systems, instrumentation and control systems, and containment structures. The design of remaining utility systems is described in this paper. They are: auxiliary helium systems which include the helium purification system, the helium sampling system, and the helium storage and supply system; fuel handling and storage system. The helium purification systems are installed in the primary and secondary helium cooling systems in order to reduce the quantity of chemical impurities. The helium sampling systems monitor the concentration of impurities. The helium storage and supply systems keep the steady pressure of the helium system during the normal operation. The fuel handling and storage system is utilised to handle the new and spent fuels safely and reliably.

Journal Articles

Present status of the liquid lithium target facility in the international fusion materials irradiation facility (IFMIF)

Nakamura, Hiroo; Riccardi, B.*; Loginov, N.*; Ara, Kuniaki*; Burgazzi, L.*; Cevolani, S.*; Dell'Ocro, G.*; Fazio, C.*; Giusti, D.*; Horiike, Hiroshi*; et al.

Journal of Nuclear Materials, 329-333(1), p.202 - 207, 2004/08

 Times Cited Count:14 Percentile:68.23(Materials Science, Multidisciplinary)

International Fusion Materials Irradiation Facility (IFMIF), being developed by EU, JA, RF and US, is a deuteron-lithium (Li) reaction neutron source for fusion materials testing. In the end of 2002, 3 year Key Element technology Phase (KEP) to reduce the key technology risk factors has been completed. This paper describes these KEP tasks results. To evaluate Li flow characteristics, a water and Li flow experiments have been done. To develop Li purification system, evaluation of nitrogen and tritium gettering materials have been done. Conceptual design of remote handling and basic experiment have been donde. In addition, safety analysis and diganostics design have been done. In the presentation, the latest design and future prospects will be also summarized.

JAEA Reports

Study on system integration of robots Operated in nuclear fusion facility and nuclear power plant facilities

Oka, Kiyoshi

JAERI-Research 2004-009, 225 Pages, 2004/07


In the present paper, I describe the necessary approaches and elemental technologies to solve the issues on the system integration of the typical robot systems for maintenance in the nuclear fusion facility and rescue in the accident of the nuclear power plant facilities. These robots work under the intense radiation condition and restricted space in place of human. In particular, I propose a new approach to realize the system integration of the robot for actual use from the viewpoints of not only the environmental and working conditions but also the restructure and optimization of the required elemental technologies with a well balance in the robot system. Based on the above approach, I have a contribution to realize the robot systems working under the actual conditions for maintenance in the nuclear fusion facility and rescue in the accident of the nuclear power plant facilities.

Journal Articles

Calculation of nuclear characteristic parameters and drawing subcriticality judgment graphs of infinite fuel systems for typical nuclear fuels

Okuno, Hiroshi; Takada, Tomoyuki

Journal of Nuclear Science and Technology, 41(4), p.481 - 492, 2004/04

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Nuclear characteristic parameters were calculated and subcriticality judgement graphs were drawn for revision purposes of the Data Collection for the Nuclear Criticality Safety Handbook. The nuclear characteristic parameters were the neutron multiplication factor in infinite media, migration area and diffusion constants for 11 kinds of typical fuels encountered in criticality safety evaluation of nuclear fuel cycle facilities. These fuels included ADU-H$$_{2}$$O, UF6-HF and Pu(NO$$_{3}$$)$$_{4}$$-UO$$_{2}$$(NO$$_{3}$$)$$_{2}$$ solution, of which data were not cited in the Data Collection. The calculation was made with the Japanese evaluated nuclear data library JENDL-3.2 and a sequence of criticality calculation codes, SRAC, POST and SIMCRI. The subcriticality judgement graphs that depict the region satisfying the inequality relation of the neutron multiplication factor less than 0.98 between the two variables (a) uranium enrichment, 239Pu/Pu ratio or plutonium enrichment and (b) H/(Pu+U) ratio were drawn for the same kinds of fuels except UF6-HF in infinite media.

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