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Journal Articles

Depletion calculation of subcritical system with consideration of spontaneous fission reaction

Riyana, E. S.; Okumura, Keisuke; Sakamoto, Masahiro; Matsumura, Taichi; Terashima, Kenichi

Journal of Nuclear Science and Technology, 59(4), p.424 - 430, 2022/04

JAEA Reports

Technology development of diamond-base neutron sensors and radiation-resistive integrated-circuits for shielding-free criticality approach monitoring system (Contract research); FY2020 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; High Energy Accelerator Research Organization*

JAEA-Review 2021-038, 65 Pages, 2022/01

JAEA-Review-2021-038.pdf:4.42MB

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, 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 FY2020, this report summarizes the research results of the "Technology development of diamond-base neutron sensors and radiation-resistive integrated-circuits for shielding-free criticality approach monitoring system" conducted in FY2020. The present study aims to develop key components of neutron detection system without a radiation shield for a criticality approach monitoring system. It is required high neutron detection efficiency for a few cps/nv under high gamma ray radiation environment (i.e. 1 kGy/h maximum) and compact-light-weight to fit constraints of the penetration size and the payload. In order to develop the monitoring system, the project aims to design and evaluate neutron detection devices based on diamond sensors and a high radiation resistive signal-processi

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

Journal Articles

Numerical investigations on the coolability and the re-criticality of a debris bed with the density-stratified configuration

Li, C.; Uchibori, Akihiro; Takata, Takashi; Pellegrini, M.*; Erkan, N.*; Okamoto, Koji*

Dai-25-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2021/07

The capability of stable cooling and avoiding re-criticality on the debris bed are the main issues for achieving IVR (In-Vessel Retention). In the actual situation, the debris bed is composed of mixed-density debris particles. Hence, when these mixed-density debris particles were launched to re-distribute, the debris bed would possibly form a density-stratified distribution. For the proper evaluation of this scenario, the multi-physics model of CFD-DEM-Monte-Carlo based neutronics is established to investigate the coolability and re-criticality on the heterogeneous density-stratified debris bed with considering the particle relocation. The CFD-DEM model has been verified by utilizing water injection experiments on the mixed-density particle bed in the first portion of this research. In the second portion, the coupled system of the CFD-DEM-Monte-Carlo based neutronics model is applied to reactor cases. Afterward, the debris particles' movement, debris particles' and coolant's temperature, and the k-eff eigenvalue are successfully tracked. Ultimately, the relocation and stratification effects on debris bed's coolability and re-criticality had been quantitatively confirmed.

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

Exploratory investigation for estimation of fuel debris criticality risk

Yamane, Yuichi; Numata, Yoshiaki*; Tonoike, Kotaro

Proceedings of 11th International Conference on Nuclear Criticality Safety (ICNC 2019) (Internet), 10 Pages, 2019/09

For the criticality safety of the operation treating the fuel debris in Fukushima Daiichi Nuclear Power Plant, the reactivity effect of its geometrical change has been investigated and the developed procedure has been applied to a trial analysis of a postulated scenario for the purpose of its verification.

Journal Articles

Study of the screening survey using an ambient dose equivalent rate survey meter in criticality accidentsy

Hoshi, Katsuya; Tsujimura, Norio; Yoshida, Tadayoshi; Kurihara, Osamu*; Kim, E.*; Yajima, Kazuaki*

Progress in Nuclear Science and Technology (Internet), 6, p.152 - 155, 2019/01

Journal Articles

Study on criticality in natural barrier for disposal of fuel debris from Fukushima Daiichi NPS

Shimada, Taro; Takubo, Kazuya*; Takeda, Seiji; Yamaguchi, Tetsuji

Progress in Nuclear Science and Technology (Internet), 5, p.183 - 187, 2018/11

After fuel debris is removed from the reactor containment vessel at Fukushima Daiichi NPS (1F) and collected in waste containers in the future, the waste containers will be disposed at a deep geological repository. The uranium inventory and uranium-235 ($$^{235}$$U) enrichment of the fuel debris are larger than those of high-level vitrified wastes which are produced from liquid waste during reprocessing of spent nuclear fuels. Therefore, there is a possibility not to be excluded that a criticality occurs in the geological media where the uranium precipitates at the far-field from the repository, after the uranium located in the repository is dissolved by groundwater. In this study, we calculated the quantity of uranium precipitated at the natural barrier, and studied dimension of uranium deposited in the natural barrier and carried out the criticality analysis.

JAEA Reports

Study on release behavior of radioiodine from fuel solution under criticality accident condition

Tashiro, Shinsuke; Abe, Hitoshi

JAEA-Technology 2015-044, 20 Pages, 2016/03

JAEA-Technology-2015-044.pdf:1.26MB

In order to estimate public dose under a criticality accident in fuel solution of a fuel reprocessing plant, release behavior of radioiodine from the fuel solution to atmosphere is very important. In this report, time evolution of $$^{133}$$I concentration in gas phase of TRACY core tank was measured until the concentration in the solution decreased. Furthermore, cumulative release ratio (CRR) and release rate (RR) from the solution to the atmosphere of radioiodine were evaluated by applying previously-reported evaluation model. As a result, for the case of short transient criticality, RR of $$^{133}$$I became maximum at 1 hour later from the ending and almost constant after 8 hour later. Furthermore, relationship of each elapsed time between total fission number and release rate of $$^{133}$$I could be derived. On the other hand, for the case of long criticality excursion, such as JCO criticality accident, the CRR and RR of radioiodine increased monotonously with time.

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

Assessment of human body surface and internal dose estimations in criticality accidents based on experimental and computational simulations

Sono, Hiroki; Ono, Akio*; Kojima, Takuji; Takahashi, Fumiaki; Yamane, Yoshihiro*

Journal of Nuclear Science and Technology, 43(3), p.276 - 284, 2006/03

 Times Cited Count:1 Percentile:10.64(Nuclear Science & Technology)

For a study on the applicability of a personal dosimetry method to criticality accident dosimetry, an assessment of the human body surface and internal dose estimations was performed by experimental and computational simulations. The experimental simulation was carried out in a criticality accident situation at the TRACY facility. The neutron and $$gamma$$-ray absorbed doses in muscle tissue were separately estimated by a dosimeter set of an alanine dosimeter and a thermoluminescence dosimeter made of enriched lithium tetra borate with a phantom. The computational simulation was conducted with a Monte Carlo code taking account of dose components of neutrons, prompt $$gamma$$-rays and delayed $$gamma$$-rays. The computational simulation was ascertained to be valid by comparison between the calculated dose distributions in the phantom and the measured ones. The assessment based on the experimental and computational simulations confirmed that the personal dosimetry using the dosimeter set provided a first estimation of the body surface and internal doses with precision.

Journal Articles

Measurement of neutron and $$gamma$$-ray absorbed doses inside human body in criticality accident situations using phantom and tissue-equivalent dosimeters

Sono, Hiroki; Kojima, Takuji; Soramasu, Noboru*; Takahashi, Fumiaki

JAERI-Conf 2005-007, p.315 - 320, 2005/08

Personal dosimeters provide a fundamental evaluation of external exposures to human bodies in radiation accidents. The dose distribution inside the body, which is needed to estimate the exposures from a result of personal dosimetry, has been evaluated mostly by computational simulations, while experimental data to verify the simulations are not sufficiently supplied, in particular, in criticality accident situations. For the purpose of obtaining the experimental data on external exposures inside the body, a preliminary experiment on criticality accident dosimetry was carried out at the Transient Experiment Critical Facility (TRACY) using a human phantom and tissue-equivalent dosimeters. The neutron and $$gamma$$-ray absorbed doses inside the phantom could be satisfactorily measured by the combined use of an alanine dosimeter and a thermoluminescent dosimeter made of enriched lithium tetra borate. The doses measured in and on the phantom were regarded as reasonable in dose level and distribution by comparison with the doses measured in the free air.

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.

Journal Articles

Evaluation of $$gamma$$-ray dose components in criticality accident situations

Sono, Hiroki; Yanagisawa, Hiroshi*; Ono, Akio*; Kojima, Takuji; Soramasu, Noboru*

Journal of Nuclear Science and Technology, 42(8), p.678 - 687, 2005/08

 Times Cited Count:4 Percentile:32.08(Nuclear Science & Technology)

Component analysis of $$gamma$$-ray doses in criticality accident situations is indispensable for further understanding on emission behavior of $$gamma$$-rays and accurate evaluation of external exposure to human bodies. Such dose components were evaluated, categorizing $$gamma$$-rays into four components: prompt, delayed, pseudo components in the period of criticality, and a residual component in the period after the termination of criticality. This evaluation was performed by the combination of dosimetry experiments at the TRACY facility using a thermoluminescent dosimeter (TLD) made of lithium tetra borate and computational analyses using a Monte Carlo code. The evaluation confirmed that the dose proportions of the above components varied with the distance from the TRACY core tank. This variation was due to the difference in attenuation of the individual components with the distance from the core tank. The evaluated dose proportions quantitatively clarified the contribution of the pseudo and the residual components to be excluded for accurate evaluation of $$gamma$$-ray exposure.

Journal Articles

Examination for neutron dose assessment method from induced sodium-24 in human body in criticality accidents

Takahashi, Fumiaki; Endo, Akira; Yamaguchi, Yasuhiro

Journal of Nuclear Science and Technology, 42(4), p.378 - 383, 2005/04

 Times Cited Count:2 Percentile:18.81(Nuclear Science & Technology)

Experiments were made to verify a dose assessment method from activated sodium in body in criticality accidents. A phantom containing sodium chloride solution was irradiated in the Transient Experiment Critical Facility to simulate activation of sodium. Monte Carlo calculations were performed to obtain quantitative relation between the activity of induced Na-24 and neutron dose in the phantom. In the previous work, conversion coefficients from specific activity of induced Na-24 to neutron dose had been analyzed with the MCNP-4B code concerning neutron spectra at some hypothesized configurations. One of the prepared coefficients was applied to evaluate neutron dose from the measured activity. The estimated dose agreed with the dose analyzed by the Monte Carlo calculation in the present study within an acceptable uncertainty, which is indicated by the IAEA. In addition, the dose calculated with the prepared coefficient was close to the result measured with dosimeters. These results suggest that the prepared coefficients can be applied to dose assessments from induced Na-24 in body.

Journal Articles

Characteristics of severe accidents of Reduced-Moderation Water Reactor (RMWR)

Yonomoto, Taisuke; Akie, Hiroshi; Kobayashi, Noboru; Okubo, Tsutomu; Uchikawa, Sadao; Iwamura, Takamichi

Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 11 Pages, 2004/10

Reduced-Moderation Water Reactor (RMWR) is a light-water cooled high-conversion reactor that is being developed by JAERI with collaboration from the Japanese industries. Since RMWR utilizes the highly enriched plutonium, the safety concern for RMWR includes the possibility of recriticality during severe accidents as is the case with the liquid metal cooled fast breeder reactor. In order to clarify this concern, characteristics of severe accidents of RMWR are analyzed in this study. The results obtained so far indicate that (1) the mechanical impact of recriticality in the core, if occurs, is supposed to be insignificant due to the absence of water, (2) the mixture of the fuel and cladding debris in the lower plenum does not cause recriticality when they are well mixed and distributed flatly, and (3) if requires, the installation of neutron-absorption material with realistic geometry can effectively prevent recriticality in the lower plenum even for the conservatively-assumed spherical accumulation of core debris.

JAEA Reports

Investigation of evaluation model of mist release behavior from burst of bubble on the solution surface

Abe, Hitoshi; Tashiro, Shinsuke; Morita, Yasuji

JAERI-Research 2004-014, 19 Pages, 2004/09

JAERI-Research-2004-014.pdf:1.33MB

no abstracts in English

Journal Articles

Dose evaluation in criticality accident conditions using transient critical facilities fueled with a fissile solution

Nakamura, Takemi*; Tonoike, Kotaro; Miyoshi, Yoshinori

Radiation Protection Dosimetry, 110(1-4), p.483 - 486, 2004/09

 Times Cited Count:2 Percentile:17.98(Environmental Sciences)

Dose measurement and evaluation technique in criticality accident conditions with a thermo luminescence dosimeter (TLD) was studied at the Transient Experiment Critical Facility (TRACY) of Japan Atomic Energy Research Institute (JAERI). The tissue absorbed dose can be derived from the ambient dose equivalent given by measurement with a TLD using the conversion factor given by computation. Using this technique, the neutron dose around the SILENE reactor of the Institute for Radioprotection and Nuclear Safety (IRSN) of France was measured in the Accident Dosimetry Intercomparison Exercise (June 10-21, 2002) organized by OECD/NEA and IRSN. In this exercise, the $$gamma$$ dose was also measured with a TLD. In this report, measurements and evaluation results at TRACY and SILENE are presented.

Journal Articles

Development of remote surveillance squads for nuclear emergency

Yanagihara, Satoshi; Kobayashi, Tadayoshi; Miyajima, Kazutoshi; Tanaka, Mitsugu

International Journal of Robotics and Automation, 18(4), p.160 - 165, 2003/12

no abstracts in English

Journal Articles

Power profile evaluation of the JCO precipitation vessel based on the record of the $$gamma$$-ray monitor

Tonoike, Kotaro; Nakamura, Takemi*; Yamane, Yuichi; Miyoshi, Yoshinori

Nuclear Technology, 143(3), p.364 - 372, 2003/09

 Times Cited Count:2 Percentile:19.75(Nuclear Science & Technology)

no abstracts in English

43 (Records 1-20 displayed on this page)