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JAEA Reports

Improvement of critical safety technology in fuel debris retrieval (Contract research); FY2020 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Tokyo Institute of Technology*

JAEA-Review 2021-037, 61 Pages, 2022/01

JAEA-Review-2021-037.pdf:4.24MB

The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2020. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station (1F), Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2019, this report summarizes the research results of the "Improvement of critical safety technology in fuel debris retrieval" conducted in FY2019 and FY2020. Since the final year of this proposal was FY2020, the results for two fiscal years were summarized. The purpose of research was to improve the criticality safety analysis methods in the case of fuel debris removal with the collaboration with Russian university, which has a lot of experiences in the criticality analysis. This research has been performed as two fiscal years project in FY 2019 and FY 2020 by Tokyo Institute of Technology (Tokyo Tech) and Tokyo City University (TCU) as the Japanese side, and National Research Nuclear University MEPhI as the Russian side. In FY2019, Tokyo Tech introduced a GPU server

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

JAEA-Technology-2021-023.pdf:5.25MB

Computational analyses on nuclear criticality characteristics were carried out for heterogeneous lattice systems composed of water moderator and fuel rods utilized in low-power 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.

Journal Articles

Melt impingement on a flat spreading surface under wet condition

Sahboun, N. F.; Matsumoto, Toshinori; Iwasawa, Yuzuru; Sugiyama, Tomoyuki

Proceedings of Asian Symposium on Risk Assessment and Management 2021 (ASRAM 2021) (Internet), 15 Pages, 2021/10

JAEA Reports

Improvement of critical safety technology in fuel debris retrieval (Contract research); FY2019 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Tokyo Institute of Technology*

JAEA-Review 2020-041, 30 Pages, 2020/12

JAEA-Review-2020-041.pdf:1.9MB

JAEA/CLADS had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project in FY2019. Among the adopted proposals in FY2019, this report summarizes the research results of the "Improvement of Critical Safety Technology in Fuel Debris Retrieval" conducted in FY2019.

Journal Articles

A Linear Equation of characteristic time profile of power in subcritical quasi-steady state

Yamane, Yuichi

Journal of Nuclear Science and Technology, 57(8), p.926 - 931, 2020/08

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

An equation of power in subcritical quasi-steady state has been derived based on one-point kinetics equations for the purpose of utilizing it for the development of timely reactivity estimation from complicated time profile of neutron count rate. It linearly relates power, $$P$$, to a new variable $$q$$, which is a function of time differential of the power. It has been confirmed by using one-point kinetics code, AGNES, that the calculated points ($$q, P$$) are perfectly in a line described by the new equation and that points ($$q, P$$) calculated from transient subcritical experiments by using TRACY made a line with a slope indicated by the new equation.

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

JAEA-Review-2020-007.pdf:6.63MB

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

Study on plutonium burner high temperature gas-cooled reactor in Japan; Introduction scenario, reactor safety and fabrication tests of the 3S-TRISO fuel

Ueta, Shohei; Mizuta, Naoki; Fukaya, Yuji; Goto, Minoru; Tachibana, Yukio; Honda, Masaki*; Saiki, Yohei*; Takahashi, Masashi*; Ohira, Koichi*; Nakano, Masaaki*; et al.

Nuclear Engineering and Design, 357, p.110419_1 - 110419_10, 2020/02

 Times Cited Count:1 Percentile:39.17(Nuclear Science & Technology)

The concept of a plutonium (Pu) burner HTGR is proposed to incarnate highly-effective Pu utilization by its inherent safety features. The security and safety fuel (3S-TRISO fuel) employs the coated fuel particle with a fuel kernel made of plutonium dioxide (PuO$$_{2}$$) and yttria stabilized zirconia (YSZ) as an inert matrix. This paper presents feasibility study of Pu burner HTGR and R&D on the 3S-TRISO fuel.

Journal Articles

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 1; Overview

Kaji, Yoshiyuki; Nemoto, Yoshiyuki; Nagatake, Taku; Yoshida, Hiroyuki; Tojo, Masayuki*; Goto, Daisuke*; Nishimura, Satoshi*; Suzuki, Hiroaki*; Yamato, Masaaki*; Watanabe, Satoshi*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

In this research program, cladding oxidation model in SFP accident condition, and numerical simulation method to evaluate capability of spray cooling system which was deployed for spent fuel cooling during SFP accident, have been developed. These were introduced into the severe accident codes such as MAAP and SAMPSON, and SFP accident analyses were conducted. Analyses using Computational Fluid Dynamics (CFD) code were conducted as well for the comparison with SA code analyses and investigation of detail in the SFP accident. In addition, three-dimensional criticality analysis method was developed as well, and safer loading pattern of spent fuels in pool was investigated.

JAEA Reports

Progress report on Nuclear Safety Research Center (JFY 2015 - 2017)

Nuclear Safety Research Center, Sector of Nuclear Safety Research and Emergency Preparedness

JAEA-Review 2018-022, 201 Pages, 2019/01

JAEA-Review-2018-022.pdf:20.61MB

Nuclear Safety Research Center (NSRC), Sector of Nuclear Safety Research and Emergency Preparedness, Japan Atomic Energy Agency (JAEA) is conducting technical support to nuclear safety regulation and safety research based on the Mid-Long Term Target determined by Japanese government. This report summarizes the research structure of NSRC and the cooperative research activities with domestic and international organizations as well as the nuclear safety research activities and results in the period from JFY 2015 to 2017 on the nine research fields in NSRC; (1) severe accident analysis, (2) radiation risk analysis, (3) safety of nuclear fuels in light water reactors (LWRs), (4) thermohydraulic behavior under severe accident in LWRs, (5) materials degradation and structural integrity, (6) safety of nuclear fuel cycle facilities, (7) safety management on criticality, (8) safety of radioactive waste management, and (9) nuclear safeguards.

JAEA Reports

A Guide to introducing burnup credit, preliminary version (English translation)

Okuno, Hiroshi; Suyama, Kenya; Ryufuku, Susumu*

JAEA-Review 2017-010, 93 Pages, 2017/06

JAEA-Review-2017-010.pdf:2.47MB

There is an ongoing discussion on the application of burnup credit to the criticality safety controls of facilities that treat spent fuels. With regard to such application of burnup credit in Japan, this document summarizes the current technical status of the prediction of the isotopic composition and criticality of spent fuels, as well as safety evaluation concerns and the current status of legal affairs. This report is an English translation of A Guide to Introducing Burnup Credit, Preliminary Version, originally published in Japanese as JAERI-Tech 2001-055 by the Nuclear Fuel Cycle Facility Safety Research Committee.

JAEA Reports

Solvent extraction and release behavior of ruthenium and europium in fire accident conditions in reprocessing plants (Contract research)

Amano, Yuki; Watanabe, Koji; Masaki, Tomoo; Tashiro, Shinsuke; Abe, Hitoshi

JAEA-Technology 2016-012, 21 Pages, 2016/06

JAEA-Technology-2016-012.pdf:1.81MB

To contribute to safety evaluation of fire accident in fuel reprocessing plants, solvent extraction behavior of ruthenium, which could form volatile species, was investigated. Distribution ratios of ruthenium at fire accident conditions were obtained by extraction experiments with several solvent composition at different temperature as parameters. In order to investigate release behavior of ruthenium and europium at fire accident, release ratios of ruthenium and europium were also obtained by solvent combustion experiments.

Journal Articles

Effects of $$alpha$$-radiation on a direct disposal system for spent nuclear fuel, 1 Review of research into the effects of $$alpha$$-radiation on the spent nuclear fuel, canisters and outside canisters

Kitamura, Akira; Takase, Hiroyasu*

Journal of Nuclear Science and Technology, 53(1), p.1 - 18, 2016/01

 Times Cited Count:3 Percentile:14.58(Nuclear Science & Technology)

Not only geological disposal of vitrified waste generated by spent fuel (SF) reprocessing, but also the possibility of disposing of SF itself in deep geological strata (hereinafter "direct disposal of SF") may be considered in the Japanese geological disposal program. In the case of direct disposal of SF, the radioactivity of the waste is higher and the potential effects of the radiation are greater. Specific examples of the possible effects of radiation include: increased amounts of canister corrosion; generation of oxidizing chemical species in conjunction with radiation degradation of groundwater and accompanying oxidation of reducing groundwater; and increase in the dissolution rate and the solubility of SF. Focusing especially on the effects of $$alpha$$-radiation in safety assessment, this study has reviewed research into the effects of $$alpha$$-radiation on the spent nuclear fuel, canisters and outside canisters.

Journal Articles

Effects of $$alpha$$-radiation on a direct disposal system for spent nuclear fuel, 2; Review of research into safety assessments of direct disposal of spent nuclear fuel in Europe and North America

Kitamura, Akira; Takase, Hiroyasu*; Metcalfe, R.*; Penfold, J.*

Journal of Nuclear Science and Technology, 53(1), p.19 - 33, 2016/01

 Times Cited Count:1 Percentile:8.33(Nuclear Science & Technology)

Not only geological disposal of vitrified waste generated by spent fuel (SF) reprocessing, but also the possibility of disposing of SF itself in deep geological strata (hereinafter "direct disposal of SF") may be considered in the Japanese geological disposal program. In the case of direct disposal of SF, the radioactivity of the waste is higher and the potential effects of the radiation are greater. Specific examples of the possible effects of radiation include: increased amounts of canister corrosion; generation of oxidizing chemical species in conjunction with radiation degradation of groundwater and accompanying oxidation of reducing groundwater; and increase in the dissolution rate and the solubility of SF. Therefore, the influences of radiation, which are not expected to be significant in the case of geological disposal of vitrified waste, must be considered in safety assessments for direct disposal of SF. Focusing especially on the effects of $$alpha$$-radiation in safety assessment, this study has reviewed safety assessments in countries other than Japan that are planning direct disposal of SF. The review has identified issues relevant to safety assessment for the direct disposal of SF in Japan.

Journal Articles

Conceptual study of a plutonium burner high temperature gas-cooled reactor with high nuclear proliferation resistance

Goto, Minoru; Demachi, Kazuyuki*; Ueta, Shohei; Nakano, Masaaki*; Honda, Masaki*; Tachibana, Yukio; Inaba, Yoshitomo; Aihara, Jun; Fukaya, Yuji; Tsuji, Nobumasa*; et al.

Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.507 - 513, 2015/09

A concept of a plutonium burner HTGR named as Clean Burn, which has a high nuclear proliferation resistance, had been proposed by Japan Atomic Energy Agency. In addition to the high nuclear proliferation resistance, in order to enhance the safety, we propose to introduce PuO$$_{2}$$-YSZ TRISO fuel with ZrC coating to the Clean Burn. In this study, we conduct fabrication tests aiming to establish the basic technologies for fabrication of PuO$$_{2}$$-YSZ TRISO fuel with ZrC coating. Additionally, we conduct a quantitative evaluation of the security for the safety, a design of the fuel and the reactor core, and a safety evaluation for the Clean Burn to confirm the feasibility. This study is conducted by The University of Tokyo, Japan Atomic Energy Agency, Fuji Electric Co., Ltd., and Nuclear Fuel Industries, Ltd. It was started in FY2014 and will be completed in FY2017, and the first year of the implementation was on schedule.

Journal Articles

Nuclear criticality safety standard for a fuel reprocessing plant assuming burnup credit published by the Atomic Energy Society of Japan

Nakajima, Ken*; Itahara, Kuniyuki*; Okuno, Hiroshi

Proceedings of International Conference on Nuclear Criticality Safety (ICNC 2015) (DVD-ROM), p.496 - 502, 2015/09

An outline of the standard "Procedures for Applying Burnup Credit to Criticality Safety Control of a Reprocessing Facility: 2014" (AESJ-SC-F025: 2014) published in April 2015 by the Atomic Energy Society of Japan (AESJ) is presented. The AESJ published more than 60 Standards. However, many of them were in the field of nuclear power reactors or radioactive wastes. Ten years ago the AESJ published "Basic Items of Criticality Safety Control: 2004" (AESJ-SC-F004:2004), which prescribed basic ideas, requirements and methods on nuclear criticality safety controls of facilities handling with nuclear fuel materials in general for preventing a nuclear criticality accident. However, it did not include any specific procedures for adopting burnup credit. Therefore, a new standard was envisaged as the first Standard for fuel reprocessing plants, which clarified the specific procedures to apply burnup credit to designers, operators, maintenance persons and administrators.

Journal Articles

Study on criticality control of fuel debris by Japan Atomic Energy Agency to support Nuclear Regulation Authority of Japan

Tonoike, Kotaro; Yamane, Yuichi; Umeda, Miki; Izawa, Kazuhiko; Sono, Hiroki

Proceedings of International Conference on Nuclear Criticality Safety (ICNC 2015) (DVD-ROM), p.20 - 27, 2015/09

From the viewpoint of safety regulation, criticality control of the fuel debris in the Fukushima Daiichi Nuclear Power Station would be a risk-informed control to mitigate consequences of criticality events, instead of a deterministic control to prevent such events. The Nuclear Regulation Authority of Japan has set up a research and development program to tackle this challenge. The Nuclear Safety Research Center of Japan Atomic Energy Agency, commissioned by the authority, has launched activities such as computations of criticality characteristics of the fuel debris, development of criticality risk assessment method, and preparation of criticality experiments to support them.

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.

JAEA Reports

Investigation of release behavior of volatile ruthenium species from thermal decomposition of ruthenium nitrosylnitrate

Abe, Hitoshi; Masaki, Tomoo; Amano, Yuki; Uchiyama, Gunzo

JAEA-Research 2014-022, 12 Pages, 2014/11

JAEA-Research-2014-022.pdf:1.03MB

To contribute safety evaluation of boiling and drying accident of high active liquid waste (HALW) in fuel reprocessing plant, release behavior of Ru, which was considered as an important nuclide for evaluating public dose from the volatile viewpoint, has been investigated. It has been reported that release of Ru becomes conspicuously after HALW is dried up. In this work, to grasp the release behavior of Ru, release ratio of Ru with thermal decomposition of Ru nitrate, which would be in the dried HALW, was measured and release rate constant of Ru from the nitrate was estimated. It was found that the calculation result of release rate of Ru from the nitrate with rise of temperature by using the constant could well simulate the result acquired from the beaker-scale experiment.

JAEA Reports

Proceedings of the International symposium NUCEF 2005; February 9-10, 2005, Techno Community Square RICOTTI, Tokai-mura, Ibaraki-ken, Japan

International Symposium NUCEF 2005 Working Group

JAERI-Conf 2005-007, 359 Pages, 2005/08

JAERI-Conf-2005-007.pdf:24.79MB

Japan Atomic Energy Research Institute (JAERI) held the international symposium NUCEF2005 at Techno Community Square RICOTTI in Tokai-mura on February 9 and 10, 2005. This symposium was co-organized by Japan Nuclear Cycle Development Institute (JNC), and Nuclear Fuel Cycle Safety Research Committee authorized the program. Two hundred thirty-nine participants from 11 countries presented fifty-nine papers, and discussed recent research activities and its outputs on waste disposal safety, fuel cycle facility safety including criticality safety, and separation process development. The presented papers are compiled in the proceedings.

Journal Articles

Study on safety evaluation for nuclear fuel cycle facility under accident conditions

Abe, Hitoshi; Tashiro, Shinsuke; Morita, Yasuji

JAERI-Conf 2005-007, p.199 - 204, 2005/08

Source term data for estimating release behavior of radioactive nuclides is necessary to evaluate synthetic safety of nuclear fuel cycle facility under accident conditions, such as fire and criticality. In JAERI, the data has been obtained by performing some demonstration tests. In this paper, the data for the criticality accident in fuel solution obtained from the TRACY experiment, will be mainly reviewed. At 4.5 h after the transient criticality, the release ratio of the iodine were about 0.2% for re-insertion of transient rod at just after transient criticality and about 0.9% for not re-insertion. Similarly the release coefficient and release ratio for Xe were estimated. It was proved that the release ratio of Xe-141 from the solution was over 90% in case that the inverse period was over about 100 (1/s). Furthermore, outline of the study on the fire accident as future plan will be also mentioned.

164 (Records 1-20 displayed on this page)