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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

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

Multiphysics analysis of reactivity changes due to solution flow in the past criticality accident at Windscale Works in 1970

Fukuda, Kodai; Yamane, Yuichi

Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 9 Pages, 2023/10

This study presents the results of multiphysics analysis, which investigates the change of reactivity caused by the motion of fluids, of Windscale Works criticality accident. The purpose of this study is to confirm previously reported trends of emulsion formation and increase in reactivity by the multi-physics analysis which takes the motion of fluids into account. Continuous energy Monte Carlo code MVP3 was used to calculate reactivity based on the material distribution obtained by CFD calculation using OpenFOAM. An interface program in python was developed to transfer data from OpenFOAM to MVP3. The change of reactivity caused by the motion of solutions was calculated without considering the generation of heat by fissions in a system that simulated the transfer vessel at Windscale Works. As a result, trends of emulsion formation and increase in reactivity were confirmed. The influence of the resolution of the calculation system on the results was also discussed.

Journal Articles

Study on criticality safety control of fuel debris for validation of methodology applied to the safety regulation

Suyama, Kenya; Ueki, Taro; Gunji, Satoshi; Watanabe, Tomoaki; Araki, Shohei; Fukuda, Kodai; Yamane, Yuichi; Izawa, Kazuhiko; Nagaya, Yasunobu; Kikuchi, Takeo; et al.

Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 6 Pages, 2023/10

To remove and store safely the fuel debris generated by the severe accident of the Fukushima Daiichi Nuclear Power Station in 2011 is one of the most important and challenging topics for decommissioning of the damaged reactors in Fukushima. To validate the adopted method for the evaluation of criticality safety control of the fuel debris through comparison with the experimental data obtained by the criticality experiments, the Nuclear Regulation Authority (NRA) of Japan funds a research and development project which was entrusted to the Nuclear Safety Research Center (NSRC) of Japan Atomic Energy Agency (JAEA) from 2014. In this project, JAEA has been conducting such activities as i) comprehensive computation of the criticality characteristics of the fuel debris and making database (criticality map of the fuel debris), ii) development of new continuous energy Monte Carlo code, iii) evaluation of criticality accident and iv) modification of the critical assembly STACY for the experiments for validation of criticality safety control methodology. After the last ICNC2019, the project has the substantial progress in the modification of STACY which will start officially operation from May 2024 and the development of the Monte Carlo Code "Solomon" suitable for the criticality calculation for materials having spatially random distribution complies with the power spectrum. We present the whole picture of this research and development project and status of each technical topics in the session.

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.

Journal Articles

Revision of the criticality safety handbook in light of the reality of the nuclear fuel cycle in Japan; With a view to transportation and storage of fuel debris

Suyama, Kenya; Ueki, Taro; Gunji, Satoshi; Watanabe, Tomoaki; Araki, Shohei; Fukuda, Kodai

Proceedings of 20th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM22) (Internet), 5 Pages, 2023/06

Since the 1990s, the importance of the handbook has changed significantly, as the computational power has improved and continuous energy Monte Carlo codes have become widely used, which enables highly accurate criticality calculations, when necessary, irrespective of the complexity of the system. Because the value of performing a large number of calculations in advance and summarizing the data has decreased, since the second edition was published publicly in 1999, there has been no revision of criticality safety handbooks in Japan for nearly a quarter of a century. In Japan, where the Fukushima Daiichi Nuclear Power Plant accident occurred in 2011, it became necessary to deal with criticality safety issues in the transport and storage of the fuel debris which contains complex constituent elements, and the summary the criticality safety management for such material is an urgent issue. In the area of burnup credit, the transport and storage of fuel assemblies with low achieved burnups due to the consequences of accidents might be the problem. In addition, nuclear data, which is the input for the continuous energy Monte Carlo code, has been improved several times, now JENDL-5 is available from the end of 2021, and its incorporation becomes a need in the field. This report provides an overview of the latest criticality safety research in Japan and the planned revision of the Criticality Safety Handbook, which could be applied to the transport and storage sectors.

JAEA Reports

Mesh effect around burnable poison rod of cell model for HTTR fuel block

Fujimoto, Nozomu*; Fukuda, Kodai*; Honda, Yuki*; Tochio, Daisuke; Ho, H. Q.; Nagasumi, Satoru; Ishii, Toshiaki; Hamamoto, Shimpei; Nakano, Yumi*; Ishitsuka, Etsuo

JAEA-Technology 2021-008, 23 Pages, 2021/06

JAEA-Technology-2021-008.pdf:2.62MB

The effect of mesh division around the burnable poison rod on the burnup calculation of the HTTR core was investigated using the SRAC code system. As a result, the mesh division inside the burnable poison rod does not have a large effect on the burnup calculation, and the effective multiplication factor is closer to the measured value than the conventional calculation by dividing the graphite region around the burnable poison rod into a mesh. It became clear that the mesh division of the graphite region around the burnable poison rod is important for more appropriately evaluating the burnup behavior of the HTTR core..

Journal Articles

Incommensurate structures of intermediate phase and martensite phase in Ni$$_{2}$$MnGa

Kushida, Hiroaki*; Fukuda, Kodai*; Terai, Tomoyuki*; Fukuda, Takashi*; Kakeshita, Tomoyuki*; Oba, Takuya*; Osakabe, Toyotaka; Kakurai, Kazuhisa; Kato, Kenichi*

Journal of Physics; Conference Series, 165, p.012054_1 - 012054_4, 2009/06

Journal Articles

Crystal structure of martensite and intermediate phases in Ni$$_{2}$$MnGa studied by neutron diffraction

Kushida, Hiroaki*; Fukuda, Kodai*; Terai, Tomoyuki*; Fukuda, Takashi*; Kakeshita, Tomoyuki*; Oba, Takuya*; Osakabe, Toyotaka; Kakurai, Kazuhisa; Kato, Kenichi*

European Physical Journal; Special Topics, 158(1), p.87 - 92, 2008/05

 Times Cited Count:22 Percentile:72.08(Physics, Multidisciplinary)

Oral presentation

Neutron diffraction studies on the martensite phase and intermediate phase of Ni$$_{2}$$MnGa alloy crystal

Kushida, Hiroaki*; Fukuda, Kodai*; Terai, Tomoyuki*; Fukuda, Takashi*; Kakeshita, Tomoyuki*; Oba, Takuya*; Osakabe, Toyotaka; Kakurai, Kazuhisa

no journal, , 

no abstracts in English

Oral presentation

Effect of BP cell model for HTTR burnup characteristics

Fujimoto, Nozomu*; Honda, Yuki; Fukuda, Kodai*; Goto, Minoru; Tochio, Daisuke; Takada, Shoji

no journal, , 

In the HTTR, BP has been used for keeping excess reactivity change small by burn-up. We will present the analytical results of critical control rod position changing in order to estimate excess reactivity changing at zero power. The effect of mesh division around BP for critical control rod position is cleared by sensitivity analysis.

Oral presentation

Research outline of the Criticality Safety Research Group

Gunji, Satoshi; Watanabe, Tomoaki; Suyama, Kenya; Araki, Shohei; Fukuda, Kodai

no journal, , 

The criticality safety research group had been conducted research using deterministic methods to ensure the criticality safety. However, the retrieval work of fuel debris in the Fukushima Dai-ichi Nuclear Power Station cannot be evaluated by the conventional criticality management methods, therefore it is necessary to develop a risk-informed control method. In this presentation, among the research activities undertaken to address these research issues, the following research topics are presented. (1) Criticality Map Database, (2) Support for the new STACY installation and certification, (3) Study of the new STACY experiments, (4) Post-Irradiation Examination (PIE) / Combustion Analysis

Oral presentation

Analysis of criticality incident at Windscale Works

Fukuda, Kodai; Yamane, Yuichi

no journal, , 

A preliminary analysis of the criticality incident at Windscale Works in 1970 was performed. The open-source CFD tool box OpenFOAM and the Monte Carlo neutron transport calculation code MVP3.0 were used for the analysis. As the result, the progress of the incident and the change of reactivity during the incident were confirmed.

Oral presentation

Analysis on the past criticality accident at Windscale Works

Fukuda, Kodai

no journal, , 

no abstracts in English

Oral presentation

Kinetics analysis to investigate effects of fuel debris particle size on criticality behavior in terms of thermal transfer

Fukuda, Kodai; Yamane, Yuichi

no journal, , 

A kinetic analysis was performed to clarify the effect of particle size on the total number of fissions in the case of criticality accidents in which fuel debris particles are dispersed in water. As a result, it was clarified that the total number of fissions can be 10 times higher if the particle radius is reduced by one order of magnitude, depending on the conditions such as the quantity of water around fuel debris.

Oral presentation

Neutronics - thermal hydraulics behavior of inclined offshore floating BWR

Fukuda, Kodai; Suyama, Kenya; Obara, Toru*

no journal, , 

In recent years, efforts have been made in Japan to develop an offshore nuclear power plant with BWR. The effects of the marine environment, such as an inclination, to the reactor behavior are considered to be clarified. In this presentation, the result of the preliminary neutronics - thermal hydraulics coupled analysis, which aims for the clarification of the reactor behavior when BWR is inclined, is reported. The reactor system code TRACE and neutronics code PARCS are used for the analysis. The Peach Bottom 2 is modeled and analyzed for the preliminary analysis.

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