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Journal Articles

Effect of fuel particle size on consequences of criticality accidents in water-moderated solid fuel particle dispersion system

Fukuda, Kodai; Yamane, Yuichi

Journal of Nuclear Science and Technology, 60(12), p.1514 - 1525, 2023/12

https://doi.org/10.1080/00223131.2023.2224327

Times Cited Count：1 Percentile：68.31(Nuclear Science & Technology)

This study aims to clarify the effect of fuel particle radius on the criticality transient behavior and the total number of fissions in water-moderated solid fuel dispersion systems. Neutronics/thermal hydraulics-coupled kinetics analysis was performed in a hypothetical fuel debris system, where small fuel particles aggregate in water and become supercritical. Results showed that the number of fissions is 10 times larger when the fuel particle radius is reduced by one order of magnitude under conditions where heat transfer, i.e. from fuel to water, is emphasized. Moreover, there is a possibility that lower reactivity could give a larger number of fissions when the fuel particle size is very small. In addition, the number of fissions may be overestimated or underestimated to an unexpected extent unless appropriate fuel particle size is set on the analysis.

Journal Articles

Evaluation of thermal expansion reactivity feedback effect in water-moderated fuel-particle-dispersion system

Fukuda, Kodai

Proceedings of 4th Reactor Physics Asia Conference (RPHA2023) (Internet), 4 Pages, 2023/10

Brief evaluations were performed using the N-F model to quantitatively clarify the effect of thermal expansion on the consequences of criticality accidents in the water-moderated fuel-particle-dispersion system. The analysis clarified that ignoring thermal expansion can lead to underestimation or overestimation of the consequences by several tens of percent. It is concluded that evaluators can ignore the thermal expansion when they evaluate the consequences of the prompt supercritical transient in water-moderated solid fuel-dispersion systems, such as fuel debris systems. Only the Doppler effect can be considered when the fuel-temperature-feedback coefficient is prepared. However, depending on the required accuracy, the evaluators should take care of the error caused by ignoring thermal expansion.

JAEA Reports

Fuel debris criticality analysis technology using non-contact measurement method (Contract research); FY2021 Nuclear Energy Science & Technology and Human Resource Development Project

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, …

Journal Articles

Reactivity estimation based on the linear equation of characteristic time profile of power in subcritical quasi-steady state

Yamane, Yuichi

Journal of Nuclear Science and Technology, 59(11), p.1331 - 1344, 2022/11

https://doi.org/10.1080/00223131.2022.2053221

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.

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.

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.

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

https://doi.org/10.1080/00223131.2020.1736682

Times Cited Count：1 Percentile：11.8(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.

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

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

https://doi.org/10.3327/jnst.41.481

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.

Journal Articles

Revised data for 2nd version of Nuclear Criticality Safety Handbook/Data Collection

Okuno, Hiroshi; Ryufuku, Susumu*; Suyama, Kenya; Nomura, Yasushi; Tonoike, Kotaro; Miyoshi, Yoshinori

JAERI-Conf 2003-019, p.116 - 121, 2003/10

https://doi.org/10.11484/jaeri-conf-2003-019

This paper outlines the data prepared for the 2nd version of Data Collection of the Nuclear Criticality Safety Handbook. These data are discussed in the order of its preliminary table of contents. The nuclear characteristic parameters (k $_{rm inf}$ , M $^{2}$ , D) were derived, and subcriticality judgment graphs were drawn for eleven kinds of fuels which were often encountered in criticality safety evaluation of fuel cycle facilities. For calculation of criticality data, benchmark calculations using the combination of the continuous energy Monte Carlo criticality code MVP and the Japanese Evaluated Nuclear Data Library JENDL-3.2 were made. The calculation errors were evaluated for this combination. The implementation of the experimental results obtained by using NUCEF facilities into the 2nd version of the Data Collection is under discussion. Therefore, related data were just mentioned. A database is being prepared to retrieve revised data easily.

JAEA Reports

Proceedings of the International Symposium NUCEF 2001; Scientific Bases for Criticality Safety, Separation Process and Waste Disposal

NUCEF 2001 Symposium Working Group

JAERI-Conf 2002-004, 714 Pages, 2002/03

JAERI-Conf-2002-004.pdf:69.13MB

This volume contains 94 papers presented at the 3rd NUCEF International Symposium NUCEF 2001 held on October 31 - November 2, 2001, in Tokai, Japan, following the 1st symposium NUCEF'95 (Proceedings: JAERI-Conf 96-003) and the 2nd symposium NUCEF'98 (Proceedings: JAERI-Conf 99-004). The theme of this symposium was " Scientific Basis for Criticality Safety, Separation Process and Waste Disposal". The papers were presented in oral and poster sessions on following research fields: (1) Separation Process, (2) TRU Chemistry, (3) Radioactive Waste Disposal, (4) Criticality Safety.

Journal Articles