Izawa, Kazuhiko; Ishii, Junichi; Okubo, Takuya; Ogawa, Kazuhiko; Tonoike, Kotaro
Proceedings of 11th International Conference on Nuclear Criticality Safety (ICNC 2019) (Internet), 9 Pages, 2019/09
Japan Atomic Energy Agency, JAEA, is conducting the renewal program of the heterogeneous water moderated critical assembly STACY (Static Experiment Critical Facility) in order to verify the criticality calculation considering fuel debris which have been produced in the accident of Fukushima Daiichi Nuclear Power Station. The first criticality of the new STACY is scheduled at the beginning of 2021. After the first criticality, it is necessary to perform a series of critical experiments with a series of basic experimental core in order to gain a proficiency of operators and grasp the uncertainty that accompanies the result of critical experiments in STACY. Prior to the construction of the new STACY, a series of neutronic calculation was carried out for licensing and planning first series of critical experiment. In this paper, possible core configuration of the basic experimental core and their limitations are discussed and presented.
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.
Fukaya, Yuji; Nakagawa, Shigeaki; Goto, Minoru; Ishitsuka, Etsuo; Kawakami, Satoru; Uesaka, Takahiro; Morita, Keisuke; Sano, Tadafumi*
KURNS Progress Report 2018, P. 148, 2019/08
The Japan Atomic Energy Agency (JAEA) started the Research and Development (R&D) to improve nuclear prediction techniques for High Temperature Gas-cooled Reactors (HTGRs). The objectives are to introduce generalized bias factor method to avoid full mock-up experiment for the first commercial HTGR and to introduce reactor noise analysis to High Temperature Engineering Test Reactor (HTTR) experiment. To achieve the objectives, the reactor core of graphite moderation system named B7/4"G2/8"p8EUNU+3/8"p38EU(1) was newly composed in the B-rack of Kyoto University Critical Assembly (KUCA). The core plays a role of the reference core of the bias factor method, and the reactor noise was measured to develop the noise analysis scheme. In addition, training of operator of HTTR was also performed during the experiments.
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.
Yokoyama, Kenji; Sugino, Kazuteru; Ishikawa, Makoto; Maruyama, Shuhei; Nagaya, Yasunobu; Numata, Kazuyuki*; Jin, Tomoyuki*
JAEA-Research 2018-011, 556 Pages, 2019/03
We have developed a new unified cross-section set ADJ2017, which is an improved version of the unified cross-section set ADJ2010 for fast reactors. The unified cross-section set is used for reflecting information of C/E values (analysis / experiment values) obtained by integral experiment analyses, which are stored in the standard database for FBR core design via the cross-section adjustment methodology, which integrates with the information such as uncertainty (covariance) of nuclear data, uncertainty of integral experiment / analysis, sensitivity of integral experiment with respect to nuclear data. The ADJ2017 is based on Japan's latest nuclear data library JENDL-4.0 as in the previous version of ADJ2010, and it incorporates more information on integral experimental data related to minor actinides (MAs) and degraded plutonium (Pu). In the deveropment of ADJ2010, a total of 643 integral experimental data were analyzed and evaluated, and 488 of integral experimental data were finally selected to be used for the cross-section adjustment. In contrast, we have evaluated a total of 719 anlysis results, and eventually adopted 620 integral experimental data to create ADJ2017. ADJ2017 shows almost the same performance as ADJ2010 for the main neutronic characteristics of conventional sodium-cooled MOX-fuel fast reactors. In addition, for the neutrnic characteristics related to MA and degraded Pu, ADJ2017 improves the C/E values of the integral experimental data, and reduces the uncertainty induced by the nuclear data. ADJ2017 is expected to be widely used in the analysis and design research of fast reactors. Moreover, it is expected that the integral experimental data used for ADJ2017 can be utilized as a standard database of FBR core core design.
Tsujimura, Norio; Yoshida, Tadayoshi; Sanada, Yukihisa
JPS Conference Proceedings (Internet), 24, p.011013_1 - 011013_6, 2019/01
Tsujimura, Norio; Takahashi, Fumiaki; Takada, Chie
Progress in Nuclear Science and Technology (Internet), 6, p.148 - 151, 2019/01
Journal of Nuclear Science and Technology, 55(10), p.1180 - 1192, 2018/10
In Monte Carlo criticality analysis under material distribution uncertainty, it is necessary to evaluate the response of neutron effective multiplication factor () to the space-dependent random fluctuation of volume fractions within a prescribed bounded range. Normal random variables, however, cannot be used in a straightforward manner since the normal distribution has infinite tails. To overcome this issue, a methodology has been developed via forward-backward-superposed reflection Brownian motion (FBSRBM). Here, the forward-backward superposition makes the variance of fluctuation spatially constant and the reflection Brownian motion confines the fluctuation driven by normal noise in a bounded range. FBSRBM was implemented using Karhunen-Loeve expansion and applied to the fluctuation of volume fractions in a model of UO-concrete media with stainless steel.
Cheung, Y. W.*; Hu, Y. J.*; Imai, Masaki*; Tanioku, Yasuaki*; Kanagawa, Hibiki*; Murakawa, Joichi*; Moriyama, Kodai*; Zhang, W.*; Lai, K. T.*; Yoshimura, Kazuyoshi*; et al.
Physical Review B, 98(16), p.161103_1 - 161103_5, 2018/10
Tada, Kenichi; Kikuchi, Takeo*; Sakino, Takao; Suyama, Kenya
Journal of Nuclear Science and Technology, 55(2), p.138 - 150, 2018/02
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.
Gunji, Satoshi; Tonoike, Kotaro; Izawa, Kazuhiko; Sono, Hiroki
Progress in Nuclear Energy, 101(Part C), p.321 - 328, 2017/11
Criticality safety of fuel debris, particularly MCCI (Molten-Core-Concrete-Interaction) products, is one of the major safety issues for decommissioning of Fukushima Daiichi Nuclear Power Station. Criticality or subcriticality condition of the fuel debris is still uncertain; its composition, location, neutron moderation, etc. are not yet confirmed. The effectiveness of neutron poison in cooling water is also uncertain for use as a criticality control of fuel debris. A database of computational models is being built by Japan Atomic Energy Agency (JAEA), covering a wide range of possible conditions of such composition, neutron moderation, etc., to facilitate assessing criticality characteristics once fuel debris samples are taken and their conditions are known. The computational models also include uncertainties which are to be clarified by critical experiments. These experiments are planned and will be conducted by JAEA with the modified STACY (STAtic experiment Critical facilitY) and samples to simulate fuel debris compositions. Each of the samples will be cladded by a zircalloy tube whose outer shape is compatible with the fuel rod of STACY and loaded into an array of the fuel rods. This report introduces a study of experimental core configurations to measure the reactivity worth of samples simulating MCCI products. Parameters to be varied in the computation models for the experimental series are:(1) Uranium dioxide with U enrichments of 3, 4, and 5 wt.%; (2) Concrete volume fraction in the samples of 0, 20, 40, 60, and 80%; and (3) Porosity of the samples filled from 0 to 80% where the sample void is filled with water. It is concluded that the measurement is feasible in both under- and over-moderated conditions. Additionally, the required amount of samples was estimated.
Sakai, Hironori; Ronning, F.*; Hattori, Takanori; Tokunaga, Yo; Kambe, Shinsaku; Zhu, J.-X.*; Wakeham, N.*; Yasuoka, Hiroshi; Bauer, E. D.*; Thompson, J. D.*
Journal of Physics; Conference Series, 807(3), p.032001_1 - 032001_6, 2017/04
We have used nuclear quadrupole resonance (NQR) to probe microscopically the response of a prototypical quantum critical metal CeCoIn to substitutions of small amounts of Cd for In. Approximately half of the Cd substituents induce local Ce moments in their close proximity, as observed by site-dependent longitudinal nuclear spin relaxation rates . To reaffirm that localized moments are induced around the Cd substituents, we find a Gaussian spin-echo decay rate of transverse nuclear spin relaxation. Further, for the NQR subpeak is found to be proportional to temperatures, again indicating local moments fluctuations around the Cd substituents, while that for the NQR main peak shows a -dependence. The latter temperature dependence is close to 0.75 in pure CeCoIn and indicates that the bulk electronic state is located close to a two dimensional quantum critical instability.
Proceedings of International Conference on Mathematics & Computational Methods Applied to Nuclear Science & Engineering (M&C 2017) (USB Flash Drive), 6 Pages, 2017/04
In Monte Carlo criticality calculation, the formation of a confidence interval is based on the central limit theorem for a series of tallies from generations in equilibrium. A fundamental assertion of the theorem is the convergence in distribution (CID) of an interpolated standardized time series (ISTS) of tallies. This article reports a spectral analysis approach to ISTS in order to assess the convergence of tallies in terms of CID. Numerical results are demonstrated for a preliminary model of uranium-concrete debris.
Tsujimura, Norio; Yoshida, Tadayoshi; Yashima, Hiroshi*
JPS Conference Proceedings (Internet), 11, p.050005_1 - 050005_6, 2016/11
Toyooka, Junichi; Kamiyama, Kenji; Tobita, Yoshiharu; Suzuki, Toru
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11
Uesaka, Mitsuru*; Mineo, Hideaki
Nippon Genshiryoku Gakkai-Shi, 58(8), p.468 - 473, 2016/08
All the research reactors and critical assemblies (hereinafter RRCAs) in Japan are stopped in order to fulfil the new regulatory requirements, which were reinforced after the accident at the Tokyo Electric Power Company's Fukushima Daiichi Nuclear Power Station. These RRCAs have played important roles in the areas of human resource development, academic research, medical and industrial application of nuclear technology. Prolonged stoppage of RRCAs affects adversely those activities. Atomic Energy Society of Japan set up a group to discuss this issue. The group has shown a proposal that the roles of the RRCAs, which are indispensable facilities to nuclear human resource development, should be placed positively in the energy policy and the science and technology policy of the country.
Takahashi, Yoshiyuki*; Hori, Junichi*; Sano, Tadafumi*; Yagi, Takahiro*; Yashima, Hiroshi*; Pyeon, C. H.*; Nakamura, Shoji; Harada, Hideo
Proceedings of International Conference on the Physics of Reactors; Unifying Theory and Experiments in the 21st Century (PHYSOR 2016) (USB Flash Drive), p.645 - 652, 2016/05
For the reduction of radioactive toxicities, feasibility study of nuclear transmutation of minor actinides (MAs) and long-lived fission products (LLFPs) by utilizing innovative nuclear reactor system (i.e. fast breeder reactors and accelerator-driven systems) has been actively conducted. To design these nuclear reactor systems, the accurate nuclear data are required. Therefore, to obtain more accurate nuclear data, the project entitled as "Research and development for Accuracy Improvement of neutron nuclear data on Minor ACtinides(AIMAC)" has been started as one of the "Innovative Nuclear Research and Development Program". In a part of this project, the nuclear data of MAs are verified in the variable neutron spectra field at Kyoto University Research Reactor Institute-LINear ACcelerator (KURRI-LINAC) and Kyoto University Critical Assembly (KUCA). And the differential TOF data is cross-checked with an integral data for the validation of Np, Am, and Am. In this summary, the results of reaction rate of neutron capture cross section of Np are reported as an example in the study.
Gunji, Satoshi; Tonoike, Kotaro; Izawa, Kazuhiko; Sono, Hiroki
Proceedings of International Conference on the Physics of Reactors; Unifying Theory and Experiments in the 21st Century (PHYSOR 2016) (USB Flash Drive), p.3927 - 3936, 2016/05
Criticality safety of fuel debris including MCCI products is one of the major safety is-sues for decommissioning of Fukushima Daiichi Nuclear Power Station. Criticality or subcriticality condition of the fuel debris is still uncertain since its composition, location, neutron moderation, etc. are not confirmed. Also uncertain in criticality control of fuel debris is the effectiveness of neutron poison in cooling water. A database is being built by computation in JAEA, covering a wide range of possible conditions of such composition, neutron moderation, etc., to facilitate assessing criticality characteristics when fuel debris samples are taken and their conditions are known. The computation also has uncertainties to be clarified by critical experiments, which is planned by JAEA to be conducted with the modified STACY and samples simulating fuel debris compositions. This report introduces a study of experimental core configurations for reactivity worth measurements of samples simulating MCCI products. It is concluded that the measurement is feasible in both under- and over-moderated conditions. Additionally, required amount of samples was estimated.
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.
Abe, Hitoshi; Tashiro, Shinsuke; Miyoshi, Yoshinori
Nippon Genshiryoku Gakkai Wabun Rombunshi, 6(1), p.10 - 21, 2007/03
In MOX fuel fabrication facility, zinc stearate will be added into the MOX powder as addition material. If the material is added in large excess by miss operation, criticality characteristics of the MOX fuel would be influenced because it has neutron moderation effect. If criticality condition should be induced by the excess addition, physical variations, such as melting and pyrolysis of the material, must be caused by the fission energy and dynamic characteristics of the MOX fuel must be affected. To contribute quantitative evaluation of the dynamic characteristics, thermal properties data such as exo/endothermic calorific values, reaction rates, etc. with the respective physical variations and release behavior of pyrolysis gas were measured. It was found the exo/endothermic behavior with rinsing temperature of the material could be divided into six regions and rapid pressure rise caused by the pyrolysis reaction over about 400 C. Furthermore, on the basis of the results, evaluation model for the thermal properties under the criticality condition was also investigated.