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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.
Gunji, Satoshi; Yoshikawa, Tomoki; Araki, Shohei; Izawa, Kazuhiko; Suyama, Kenya
Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 8 Pages, 2023/10
Since the compositions and properties of the fuel debris are uncertain, critical experiments are required to validate calculation codes and nuclear data used for the safety evaluation. For this purpose, JAEA has been modifying a critical assembly called "STACY". The first criticality of the new STACY is scheduled for spring 2024. This paper reports the consideration results of the core configurations of the new STACY at the first criticality. We prepared two sets of gird plates with different neutron moderation conditions (their intervals are 1.50 cm and 1.27 cm). However, there is a limitation on the number of available UO
fuel rods. In addition, we would like to set the critical water heights for the first criticality at around 95 cm. This is to avoid the reactive effect of the aluminum alloy middle grid plates (Approx. 98 cm high). The core configurations for the first criticality satisfying these conditions were constructed by computational analysis. A square core configuration with the 1.50 cm grid plate that is close to the optimum moderation condition needs 261 fuel rods to reach criticality. As to the 1.27 cm grid plate, we considered two core configurations with 1.80 cm intervals by using a checkerboard arrangement. One of them has two regions core configuration with 1.27 and 1.80 cm intervals, and the other has only 1.80 cm intervals. They need 341 and 201 fuel rods for the criticality, respectively. This paper shows these three core configurations and their calculation models.
Gunji, Satoshi; Araki, Shohei; Arakaki, Yu; Izawa, Kazuhiko; Suyama, Kenya
Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 9 Pages, 2023/10
JAEA has been modifying a critical assembly called STACY from a solution system to a light-water moderated heterogeneous system to validate computation results of criticality characteristics of fuel debris generated in the accident at TEPCO's Fukushima Daiichi Nuclear Power Station. To experimentally simulate the composition and characteristics of fuel debris, we will prepare several grid plates which make particular neutron moderation conditions and a number of rod-shaped concrete and stainless-steel materials. Experiments to evaluate fuel debris's criticality characteristics are scheduled using these devices and materials. This series of STACY experiments are planned to measure the reactivity of fuel debris-simulated samples, measure the critical mass of core configurations containing structural materials such as concrete and stainless steels, and the change in critical mass when their arrangement becomes non-uniform. Furthermore, two divided cores experiments are scheduled that statically simulate fuel debris falling, and also scheduled that subcriticality measurement experiments with partially different neutron moderation conditions. The experimental plans have been considered taking into account some experimental constraints. This paper shows the schedule of these experiments, as well as the computation results of the optimized core configurations and expected results for each experiment.
Gunji, Satoshi; Araki, Shohei; Watanabe, Tomoaki; Fernex, F.*; Leclaire, N.*; Bardelay, A.*; Suyama, Kenya
Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 9 Pages, 2023/10
Institut de radioprotection et de s
ret
nuclaire (IRSN) and Japan Atomic Energy Agency (JAEA) have a long-standing partnership in the field of criticality safety. In this collaboration, IRSN and JAEA are planning a joint experiment using the new STACY critical assembly, modified by JAEA. In order to compare the codes (MVP3, MORET6, etc.) and nuclear data (JENDL and JEFF) used by both institutes in the planning of the STACY experiment, benchmark calculations of the Apparatus B and TCA, which are critical assemblies once owned by both institutes, benchmarks from the ICSBEP handbook and the computational model of the new STACY were performed. Including the new STACY calculation model, the calculations include several different neutron moderation conditions and critical water heights. There were slight systematic differences in the calculation results, which may have originated from the processing and/or format of the nuclear data libraries. However, it was found that the calculated results, including the new codes and the new nuclear data, are in good agreement with the experimental values. Therefore, there are no issues to use them for the design of experiments for the new STACY. Furthermore, the impact of the new TSL data included in JENDL-5 on the effective multiplication factor was investigated. Experimental validation for them will be completed by critical experiments of the new STACY by both institutes.
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.
Hasegawa, Toshinari; Fukaya, Yuji; Ueta, Shohei; Goto, Minoru
Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 5 Pages, 2021/08
Burnable poison (BP) credit concept has been proposed as a criticality safety measure for commercial high temperature gas-cooled reactor (HTGR) fuel fabrication, so we estimated manufacturability of the BP-mixed UO
kernel for the practical use of the concept. As a BP, boron, gadolinium, erbium, and hafnium are investigated. Boron mixed fuel kernels are fabricated by mixing boric acid powder with U
O
powder. In the case of the other BPs, BP nitrate powder is mixed with UO powder. In order to confirm that BP remain in the kernels after the heat treatment processes, thermodynamic equilibrium analysis was performed. Above 450
C, boron would melt and vaporize during the heat treatment processes, so it was found that the boron mixed fuel kernel fabrication is difficult. On the other hand, it was found that gadolinium, erbium, and hafnium would change to solid oxides that do not melt and vaporize even at 2000C, and there was no problem with manufacturability of the BP-mixed fuel kernel.
Fukaya, Yuji; Ueta, Shohei; Goto, Minoru; Ohashi, Hirofumi
Annals of Nuclear Energy, 151, p.107937_1 - 107937_9, 2021/02
Times Cited Count：2 Percentile：31.78(Nuclear Science & Technology)Feasibility study on Burnable Poison (BP) credit concept to High Temperature Gas-cooled Reactor (HTGR) fuel fabrication has been performed. By mixing BP into fuel material in the first place of fuel fabrication, criticality safety is ensured in the all fuel fabrication process even with high enrichment fuel such as 14 wt% used in commercial HTGR. However, the poison effect also prevents the criticality even in the HTGR core, and it may shorten cycle length and achievable burn-up of the core. Therefore, the effect is evaluated by whole core burn-up calculation. As a BP, boron, gadolinium, erbium, and hafnium are investigated. As a result, it is found that boron and gadolinium suit this concept and the 14 wt% fuel can be fabricated in the plant fabricating 9.9 wt% High Temperature engineering Test Reactor (HTTR) fuel. With the boron and gadolinium, the commercial HTGR fuel can be fabricated with the safety measure as same as Light Water Reactor (LWR) fuel facility to treat the fuel with the enrichment up to 5 wt%. Especially, gadolinium is significantly suitable to this concept due to the dependency to spectrum, and more enhanced safety measure is feasible as well.
Araki, Shohei; Gunji, Satoshi; Tonoike, Kotaro; Kobayashi, Fuyumi; Izawa, Kazuhiko; Ogawa, Kazuhiko
Proceedings of European Research Reactor Conference 2020 (RRFM 2020) (Internet), 7 Pages, 2020/10
Critical experiments of thermal neutron system are still expected to be playing an important role for wide technical issues. The Japan Atomic Energy Agency (JAEA) is renovating the Static Experimental Critical Facility (STACY) to maintain the experimental capability. The new STACY is designed as a general-purpose criticality facility. Its core mainly consists of low enriched UO fuel rods, grid plates, and light water moderator. The first experiment campaign in the new STACY aims to obtain criticality characteristics of fuel debris, which will be used in validation of criticality analysis methods. The designs of the experimental core configurations are in progress.
on effective multiplication factor in 1/f
spectrum random systemAraki, Shohei; Yamane, Yuichi; Ueki, Taro; Tonoike, Kotaro
Proceedings of International Conference on the Physics of Reactors; Transition To A Scalable Nuclear Future (PHYSOR 2020) (USB Flash Drive), 8 Pages, 2020/03
We investigated the dependence of the effective multiplication factor (k
) in the 1/f noise model. We conducted the two-group Monte Carlo calculations. We found that the standard deviation of the k distribution showed the positive correlation with the value because the spatial distribution of the fuel became less uniform as the value increased.
Gunji, Satoshi; Clavel, J.-B.*; Tonoike, Kotaro; Duhamel, I.*
Proceedings of 11th International Conference on Nuclear Criticality Safety (ICNC 2019) (Internet), 11 Pages, 2019/09
The new criticality experiments facility STACY will be able to contribute to the validation of criticality calculations related to the fuel debris. The experimental core design is in progress in the frame of JAEA/IRSN collaboration. This paper presents the method applied to optimize the design of core configurations of the new STACY to measure the criticality characteristics of pseudo fuel debris focused on Molten Core Concrete Interaction (MCCI) debris. To ensure that a core configuration is relevant for code validation, it is important to evaluate the reactivity worth of the main isotopes and the keff sensitivity to their cross sections. To obtain maximum sensitivity of 
Si capture reaction, some parameters of the core configuration, as for example the lattice pitch or the core dimensions, were adjusted using optimization algorithm to research efficiently the optimal core configurations.
Tada, Kenichi
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 4 Pages, 2019/05
The decommissioning of Fukushima Daiichi Nuclear Power Plant accident is one of the most important issues in Japan. The criticality safety of fuel debris is imperative to prevent exposure of workers. The investigating criticality monitoring system cannot detect the criticality of fuel debris quickly. The estimation of criticality of fuel debris is required for the fuel debris retrieval. Though the expert knowledge of reactor physics is necessary to estimate the criticality of fuel debris, many people who make a plan of fuel debris retrieval may not know well about criticality analysis. We developed a handy criticality analysis tool HAND to quickly estimate the criticality of fuel debris without expert knowledge of reactor physics. Since the input data of HAND is so simple and users can intuitively understand the calculation results, this tool is expected to be the effective tool to estimate the criticality of fuel debris.
Tada, Kenichi; Kikuchi, Takeo*; Sakino, Takao; Suyama, Kenya
Journal of Nuclear Science and Technology, 55(2), p.138 - 150, 2018/02
Times Cited Count：3 Percentile：30.05(Nuclear Science & Technology)The criticality safety of the fuel debris in Fukushima Daiichi Nuclear Power Plant is one of the most important issues and the adoption of the burnup credit is desired for the criticality analysis. The assay data of used nuclear fuel irradiated in 2F2 is evaluated to validate SWAT4.0 for BWR fuel burnup problem. The calculation results revealed that number density of many heavy nuclides and FPs showed good agreement with the experimental data except for 
U, 
Np, 
Pu and Sm isotopes. The cause of the difference is assumption of the initial number density and void ratio and overestimation of the capture cross section of Np. The C/E-1 values do not depend on the types of fuel rods (UO or UO-GdO
) and it is similar to that for the PWR fuel. These results indicate that SWAT4.0 appropriately analyzes the isotopic composition of the BWR fuel and it has sufficient accuracy to be adopted in the burnup credit evaluation of the fuel debris.
Suyama, Kenya; Kashima, Takao
Proceedings of International Conference on Nuclear Criticality Safety (ICNC 2015) (DVD-ROM), p.273 - 282, 2015/09
In the technical development of the criticality safety control of the fuel debris of Fukushima accident in Japan, there have been a discussion on a possibility of adopting BUC with FP. The Expert Group on Burnup Credit Criticality Safety (EGBUC) under the Working Party on Nuclear Criticality Safety (WPNCS) in OECD/NEA Nuclear Science Committee had carried out an international burnup calculation benchmark "Phase-IIIB" and "Phase-IIIC" for BWR fuel assemblies. In these benchmarks the difference of the calculation results of 
Gd among the participants obtained keen interests because it showed rather larger difference among the participants. Authors has been carried out additional analyses on the accumulation of the gadolinium isotopes in the used nuclear fuel during the burnup. Without cooling time, the assembly-averaged amount of Gd against the burnup value depends on the burnout property of gadolinium in the burnable poison rods. However, after few year cooling time, Gd increase drastically by the decay of Eu. In this case, the amount of gadolinium isotopes in the burnable poison rods has less importance. It means that the adopted parameters and data concerning the Eu generation have much more importance than the burnup treatment of the burnable poison rods for better prediction of Gd.
Nomura, Yasushi*; Okuno, Hiroshi; Miyoshi, Yoshinori
JAERI-Tech 2004-030, 64 Pages, 2004/03
JAERI-Tech-2004-030.pdf:4.59MB
no abstracts in English
ICNC2003 Technical Program Committee
JAERI-Conf 2003-019, 896 Pages, 2003/10
JAERI-Conf-2003-019-Part1.pdf:39.68MBJAERI-Conf-2003-019-Part2.pdf:44.42MB
These proceedings contain technical papers at the 7th International Conference on Nuclear Criticality Safety ICNC2003 held on 20-24 October 2003, in Tokai, Ibaraki, Japan, following ICNC'99 in Versailles, France. The theme of this conference is "Challenges in the Pursuit of Global Nuclear Criticality Safety". These proceedings represent the current status of nuclear criticality safety research throughout the world.
6th NUCEF Seminar Working Group
JAERI-Conf 2003-018, 132 Pages, 2003/10
JAERI-Conf-2003-018.pdf:10.41MB
no abstracts in English
Sakurai, Kiyoshi; Nojiri, Ichiro*
JAERI-Conf 2003-019, p.855 - 857, 2003/10
This paper provides overview of sub-criticality safety analysis seminar (July 2000-July 2003, JAERI, total 40 engineers from universities, research institutes and enterprises) for nuclear fuel cycle facility with the Monte Carlo method in Japan. MCNP-4C2 system (MS-DOS version) was installed in each note-type personal computer. Fundamental theory of reactor physics and Monte Carlo simulation including MCNP-4C manual was lectured. Effective neutron multiplication factor and neutron spectrum were calculated for JCO deposit tank, JNC uranium solution storage tank, JNC plutonium solution storage tank and JAERI TCA core. In the seminar, methodology of safety management for nuclear fuel cycle facility was discussed in order to prevent criticality accident.
Komuro, Yuichi
Nihon Genshiryoku Gakkai-Shi, 45(4), p.265 - 269, 2003/04
no abstracts in English
Nomura, Yasushi; Mochizuki, Hiroki*
JAERI-Tech 2002-068, 131 Pages, 2002/11
JAERI-Tech-2002-068.pdf:5.59MB
no abstracts in English
Yamamoto, Toshihiro; Miyoshi, Yoshinori
Nuclear Science and Engineering, 142(3), p.305 - 314, 2002/11
Times Cited Count：2 Percentile：10.15(Nuclear Science & Technology)Mechanisms of a positive temperature reactivity coefficient that occurs in a dilute plutonium solution are investigated based on the perturbation theory and the four-factor formula. The temperature coefficient of a solution fuel is positive if the adjoint flux or 
f between 0.05eV and 0.2eV increases with neutron energy. As compared to Pu-239, Pu-241 has a tendency to make the temperature coefficient of a plutonium solution. As Pu-241 in a plutonium solution decays into Am-241 with time, the temperature coefficient becomes more positive. Since the capture cross sections of most neutron absorbers such as boron and gadolinium decreases with increasing neutron energy, soluble absorbers make the temperature coefficient positive for higher concentration plutonium solution. Cadmium and samarium solved in a dilute plutonium solution exceptionally can keep the temperature coefficient. A fixed neutron absorber generally makes the temperature coefficient negative regardless of the property of absorber materials.
