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Togawa, Orihiko; Okuno, Hiroshi
JAEA-Review 2023-043, 94 Pages, 2024/03
JAEA-Review-2023-043.pdf:1.53MB
In order to translate nuclear disaster prevention documents written in Japanese into English, the Basic Act on Disaster Management, the Act on Special Measures Concerning Nuclear Emergency Preparedness, and the Convention on Nuclear Safety were surveyed for corresponding terms in Japanese and English. The survey results were integrated and unified English translations were selected. As a result, a Japanese-English correspondence table of technical terms in the field of nuclear disaster prevention was prepared and proposed.
Togawa, Orihiko; Okura, Takehisa; Kimura, Masanori
JAEA-Review 2022-049, 76 Pages, 2023/01
JAEA-Review-2022-049.pdf:3.74MB
Before construction and after operation of nuclear facilities, environmental consequence assessments are conducted for normal operation and an emergency. These assessments mainly aim at confirming safety for the public around the facilities and producing relief for them. Environmental consequence assessments are carried out using observations/ measurements by environmental monitoring and/or model predictions by calculation models, sometimes using either of which and at other times using both them, according to the situations and necessities. First, this report investigates methods, roles, merits/demerits and relationship between observations/measurements and model predictions which are used for environmental consequence assessments of nuclear facilities, especially holding up a spent nuclear fuel reprocessing plant at Rokkasho, Aomori as an example. Next, it explains representative examples of utilization of data on observations/measurements and results on model predictions, and considers points of attention at using them. Finally, the report describes future direction, for example, improvements of observations/measurements and model predictions, and fusion of both them.
Collaborative Laboratories for Advanced Decommissioning Science; Tokyo Institute of Technology*
JAEA-Review 2022-043, 52 Pages, 2023/01
JAEA-Review-2022-043.pdf:3.48MB
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 FY2021. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, 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 FY2021, this report summarizes the research results of the "Fuel debris criticality analysis technology using non-contact measurement method" conducted in FY2021. The purpose of research was to improve the fuel debris criticality analysis technology using non-contact measurement method by the development of the fuel debris criticality characteristics measurement system and the multi-region integral kinetic analysis code. It was performed by Tokyo Institute of Technology (Tokyo Tech), National Institute of Advanced Industrial Science and Technology (AIST), and National Research Nuclear University (MEPhI) as the first year of four years research project. For the criticality characteristic measurement systems to be developed by the Japanese and Russian sides, …
Yamane, Yuichi
Journal of Nuclear Science and Technology, 59(11), p.1331 - 1344, 2022/11
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)The reactivity was estimated from a time profile of neutron count rate or a simulated data in a quasi-steady state after sudden change of reactivity or external neutron source strength. The estimation was based on the equation of power in subcritical quasi-steady state. The purpose of the study is to develop the method of timely reactivity estimation from complicated time profile of neutron count rate. The developed method was applied to the data simulating neutron count rate created by using one-point kinetics code, AGNES, and Poisson-distributed random noise and to the transient subcritical experiment data measured by using TRACY. The result shows that the difference of the estimated and reference value was within about 5% or less for (
-1) for simulated data and within about 7% or less for 
-1.4 and -3.1 for the experimental data. It was also shown that the possibility of the reactivity estimation several ten seconds after the status change.
Onoda, Yuichi; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa
Proceedings of Asian Symposium on Risk Assessment and Management 2021 (ASRAM 2021) (Internet), 11 Pages, 2021/10
The effectiveness evaluations technology of the measures for improving resilience by applying a fracture control concept under ultra-high temperature conditions has developed for prototype sodium-cooled fast reactor Monju as a model plant, and the trial evaluation has conducted using this technology in this paper. The important accident sequences to which the fracture control concept is expected to be applied under ultra-high temperature condition are identified by investigating the results of the existing researches of level-2 probabilistic risk assessment for Monju. Accident sequences categorized in protected loss of heat sink and loss of reactor level are both identified as such important accident sequences which has the potential to prevent core damage. This study has developed the technology to evaluate the effectiveness of improving resilience, where the headings which stand for success or failure of the measures to improve resilience are introduced into the event tree, the branch probability of them is set, and the effectiveness of improving resilience is expressed as the reduction of core damage frequency. As a result of the trial evaluation of the effectiveness for the measures to improve resilience, it is confirmed that core damage frequency can be reduced by applying fracture control concept. The branch probability of the measures to improve resilience proposed in this study is tentatively assigned based on the assumption. This value is expected to be quantified by the forthcoming analyses of the integrity for the reactor vessel structure at ultra-high temperature. The technology developed in this study will be applied for the evaluation of improving resilience of the next generation sodium-cooled fast reactor.
Yamane, Yuichi
Journal of Nuclear Science and Technology, 57(8), p.926 - 931, 2020/08
Times Cited Count：1 Percentile：11.54(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, 
, to a new variable 
, 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 (
) are perfectly in a line described by the new equation and that points () calculated from transient subcritical experiments by using TRACY made a line with a slope indicated by the new equation.
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.
Itoi, Tatsuya*; Nakamura, Hideo; Nakanishi, Nobuhiro*
Nihon Genshiryoku Gakkai-Shi ATOMO
, 58(5), p.318 - 323, 2016/05
no abstracts in English
Kunugi, Tomoaki; Takase, Kazuyuki; Ogawa, Masuro; Seki, Yasushi
16th IEEE/NPSS Symp. on Fusion Engineering (SOFE '95), 1, p.313 - 316, 1996/00
no abstracts in English
; *;
JAERI-M 9825, 37 Pages, 1981/12
no abstracts in English
JAERI-M 6985, 36 Pages, 1977/03
no abstracts in English
