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Imagawa, Yuya; Toyota, Kodai; Onizawa, Takashi; Kato, Shoichi
JAEA-Testing 2023-004, 76 Pages, 2024/03
JAEA-Testing-2023-004.pdf:2.08MB
This manual describes the methods for conducting material tests in air, argon gas, and sodium, and for organizing the data obtained, as a part of the development of high-temperature structural design technology for fast reactors. This manual reflects the revision of test methods in Japanese Industrial Standards (JIS) to the "FBR Metallic Materials Test Manual, PNC TN241 77-03" published in 1977 and the "FBR Metallic Materials Test Manual (Revised Edition), JNC TN9520 2001-001" published in 2001. Also, it was written with reference to the recommended room temperature / elevated temperature tensile test method by the Japan Society of Mechanical Engineers (JSME) and the test standard for the elevated-temperature low-cycle fatigue test method by the Society of Materials Science, Japan (JSMS), which are the standard for material test methods in the domestic academic society.
Fukaya, Yuji; Goto, Minoru; Shibata, Taiju
IAEA-TECDOC-2040, p.133 - 136, 2023/12
Japan has developed back-end technologies to establish a multi-recycling fuel cycle with fast breeder reactors (FBRs) to ensure energy resources. Even though the development of FBR has been retreated to one of fundamental research, the reprocessing technologies for uranium fuel and disposal technologies had been completed for Light Water Reactor (LWR) fuel cycle on the process. These technologies were inherited to utilities and are about to be practical. Now, Japan had been completed High Temperature Engineering Test Reactor (HTTR) a prototype and research reactor, a commercial High Temperature Gas-cooled Reactor (HTGR) design Gas Turbine High Temperature Reactor 300 (GTHTR300) with related reprocessing technologies, and is planning domestic demonstration reactor project. In this context, a representative fuel cycle policy is reprocessing in Japan. However, Japan has investigated various fuel cycle scenarios to expand the usage of the commercial HTGR. Then, we would like to introduce the scenarios and development status of related technologies in the present study.
Sato, Nobuaki*; Kirishima, Akira*; Sasaki, Takayuki*; Takano, Masahide; Kumagai, Yuta; Sato, Soichi; Tanaka, Kosuke
Current Location of Fuel Debris Chemistry, 178 Pages, 2023/11
Considerable efforts have been devoted to the decommissioning of the TEPCO's Fukushima Daiichi Nuclear Power Station (1F) and now the retrieval of fuel debris is being proceeded on a trial basis. It can be said that the succession of science and technology related to debris, that is, human resource development, is important and indispensable. For that reason, we thought that a specific textbook on decommissioning is necessary. Regarding the 1F fuel debris, we still do not know enough, and it would be difficult to describe the details. However, 12 years have passed since the accident, and we have come to understand the situation of 1F to a certain extent. At this stage, it is essential for future development to organize the current situation by combining examples of past severe accidents. Therefore, we presented in this book the current state of fuel debris chemistry research from the perspectives of solid chemistry, solution chemistry, analytical chemistry, radiochemistry, and radiation chemistry.
Yokoyama, Keisuke; Uwaba, Tomoyuki
Journal of Nuclear Science and Technology, 60(10), p.1219 - 1227, 2023/10
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
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.
Araki, Shohei; Gunji, Satoshi; Arakaki, Yu; Yoshikawa, Tomoki; Murakami, Takahiko; Kobayashi, Fuyumi; Izawa, Kazuhiko; Suyama, Kenya
Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 8 Pages, 2023/10
New experiments simulating fuel debris in the new criticality assembly, STACY, are designed to contribute to the validation of criticality calculations for criticality control of the fuel debris in the Fukushima Daiichi Nuclear Power Plant accident. In the new STACY experiment, a two-region core consisting of a driver region and a test region was investigated in order to configure a debris-simulated core with under-moderation condition (lattice pitch 1.27-cm) having the constraint of available fuel rod number. The test region with a 1.27-cm lattice pitch is surrounded by the driver region, in which fuel rods are arranged in a checkerboard pattern on a 1.27-cm lattice plate, with a 1.80-cm lattice pitch. Neutron spectra and sensitivity were calculated by using MCNP6 and ENDF/B-VII. The core which has a 17
17 test region with 373 fuel rods is the largest two-region core under the constraint. It was found that the core which has a 1717 test region can simulate the neutron spectra of under-moderation condition in a 1313 region inside the test region with the root-mean square percentage error of less than 5%. It was also confirmed that the sensitivity of 
Si and 
Ca (n,
) reactions when the concrete simulant, was loaded could be simulated.
Kawaguchi, Maho*; Shiba, Shigeki*; Iwahashi, Daiki*; Okawa, Tsuyoshi*; Gunji, Satoshi; Izawa, Kazuhiko; Suyama, Kenya
Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 8 Pages, 2023/10
The Nuclear Regulation Authority (NRA) has been working on an experimental approach for evaluating the criticality of fuel debris produced by the Fukushima Daiichi Nuclear Power Plant (FDNP) accident since 2014, collaborating with the Japan Atomic Energy Agency (JAEA). As part of the approach, JAEA has modified the STAtic experiment Critical facilitY (STACY) for critical experiments to evaluate characteriscs of pseudo-fuel debris. As the preliminary analyses, we verified critical characteristics with major nuclear data libraries for the proposed core configuration patterns. The three-dimensional continuous-energy Monte Carlo neutron and photon transport code, SERPENT-V2.2.0 was used with the latest JENDL, JENDL-5. As a result, larger multiplication factors of JENDL-5 across the modified STACY core configuration patterns were evaluated in comparison to the other libraries. And, 
H scattering and 
U fission sensitivity coefficients of JENDL-5 were different from those of the other libraries. Comparing among analyses with those libraries, the updated S(
, 
) of JENDL-5 might affect the result of critical characteristics in the critical analyses for the modified STACY core configuration.
Araki, Shohei; Gunji, Satoshi; Arakaki, Yu; Murakami, Takahiko; Yoshikawa, Tomoki; Hasegawa, Kenta; Tada, Yuta; Izawa, Kazuhiko; Suyama, Kenya
Proceedings of 4th Reactor Physics Asia Conference (RPHA2023) (Internet), 4 Pages, 2023/10
To conduct integrated thermal power measurements for the performance test of the modified STACY, we re-analyzed the experimental data measured in the solution fuel STACY using the activation method. We validated its feasibility under the limited number of activation detectors. The re-analyzed results of the activation method by using MVP and PHITS with JENDL-4.0 indicated that the effect of the difference of the position between activation detectors was small enough, and the results agreed with that of the fission product analysis within almost 10%. It is conceivable that the activation method could be adopted instead of the fission product analysis.
Sato, Takumi; Otobe, Haruyoshi; Morishita, Kazuki; Marufuji, Takato; Ishikawa, Takashi; Fujishima, Tadatsune; Nakano, Tomoyuki
JAEA-Technology 2023-016, 41 Pages, 2023/09
JAEA-Technology-2023-016.pdf:2.74MB
This report summarizes the results of the stabilization treatments of post-experiment nuclear materials in Plutonium Fuel Research Facility (PFRF) from August 2018 to March 2021. Based on the management standards for nuclear materials enacted after the contamination accident that occurred at PFRF on June 6, 2017, the post-experiment nuclear materials containing plutonium (Pu): samples mixed with organic substances that cause an increase in internal pressure due to radiolysis (including X-ray diffraction samples mixed with epoxy resin and plutonium powder which caused contamination accidents), carbides and nitrides samples which is reactive in air, and chloride samples which may cause corrosion of storage containers, were selected as targets of the stabilization. The samples containing organic materials, carbides and nitrides were heated in an air flow at 650 
C and 950 C for 2 hours respectively to remove organic materials and convert uranium (U) and Pu into oxides. U and Pu chlorides in LiCl-KCl eutectic melt were reduced and extracted into liquid Cd metal by a reaction with lithium (Li) -cadmium (Cd) alloy and converted to U-Pu-Cd alloy at 500 C or higher. All of the samples were stabilized and stored at PFRF. We hope that the contents of this report will be utilized to consider methods for stabilizing post experiment nuclear materials at other nuclear fuel material usage facilities.
Tanaka, Masaaki; Uchibori, Akihiro; Okano, Yasushi; Yokoyama, Kenji; Uwaba, Tomoyuki; Enuma, Yasuhiro; Wakai, Takashi; Asayama, Tai
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
The book, JSME Series in Thermal and Nuclear Power Generation Vol.3 Sodium-cooled Fast Reactor, was published as a 30th anniversary memorial project of Power & Energy Systems Division. This paper describes an introduction of the book on a part of key technologies regarding safety assessment, thermal-hydraulics, neutronics, and fuel and material development. This introductory paper also provides an overview of an integrated evaluation system named ARKADIA to offer the best possible solutions for challenges arising during the design process, safety assessment, and operation of a nuclear plant over its life cycle, in active use of the R&D efforts and knowledges on thermal-hydraulics and safety assessment with state-of-the-art numerical analysis technologies.
Hayashi, Hirokazu; Tsubata, Yasuhiro; Sato, Takumi
Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 22(3), p.97 - 107, 2023/08
The Japan Atomic Energy Agency has chosen nitride fuel as the first candidate for the transmutation of long-lived minor actinides (MA) using accelerator-driven systems (ADS). The pyrochemical method has been considered for reprocessing spent MA nitride fuels, because their decay heat should be very large for aqueous reprocessing. This study was conducted to investigate the effect of decay heat on the pyrochemical reprocessing of MA nitride fuels. On the basis of the estimated decay heats and the temperature limits of the materials that are to be handled in pyrochemical reprocessing, quantities adequate for handling in argon gas atmosphere were evaluated. From these considerations, we proposed that an electrorefiner with a diameter of 26 cm comprising 12 cadmium (Cd) cathodes with a diameter of 4 cm is suitable. On the basis of the size of the electrorefiner, the number necessary to reprocess spent MA fuels from 1 ADS in 200 days was evaluated to be 25. Furthermore, the amount of Cd-actinides (An) alloy to produce An nitrides by the nitridation-distillation combined reaction process was proposed to be about one-quarter that of Cd-An cathode material. The evaluated sizes and required numbers of equipment support the feasibility of pyrochemical reprocessing for MA nitride fuels.
variations by nonuniform core configurationsGunji, Satoshi; Araki, Shohei; Suyama, Kenya
Nuclear Science and Engineering, 197(8), p.2017 - 2029, 2023/08
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)The fuel debris generated by the accident at the Tokyo Electric Power Company's Fukushima Daiichi Nuclear Power Plant is expected to have not only heterogeneous but also nonuniform compositions. Similarly, damaged fuel assemblies remaining in the reactor vessels also have nonuniform configurations due to some missing fuel rods. This non-uniformity may cause changing neutron multiplication factors. The effect of non-uniformity on the neutron multiplication factor is clarified by computations, and the possibility of experimentally validating the computations used for criticality management is being investigated. For this purpose, in this study the criticality effects of several core configurations of a new critical assembly, STACY, of the Japan Atomic Energy Agency with nonuniform arrangements of uranium oxide fuel rods, concrete rods, and stainless-steel rods were studied to confirm benchmarking potential. The difference in these arrangements changed the neutron multiplication factor by more than 1 $. We confirmed that changes in local neutron moderation conditions and the clustering of specific components caused this effect. In addition, the feasibility of benchmark experimental cores with nonuniform arrangements is evaluated. If benchmarking of such experiments could be realized, it would help to validate calculation codes and to develop criticality management methods by machine learning.
Ikusawa, Yoshihisa; Nagayama, Masahiro*
JAEA-Data/Code 2023-006, 24 Pages, 2023/07
JAEA-Data-Code-2023-006.pdf:1.42MB
Core fuels with stainless steel cladding and high plutonium content mixed oxide (MOX) fuel in a water-cooled environment, such as supercritical water-cooled reactors (SCWR) and reduced-moderation water reactors (RMWR), have been studied. In order to contribute to the research and development of such a core fuel concept, the fuel performance code "FEMAXI-8" was verified based on the results of post irradiation examinations of MOX fuel irradiated in the experimental fast reactor "JOYO". FEMAXI-8 is the latest version of the behavior analysis code developed by JAEA to analyze the behavior of light water reactor fuels under normal operation and transient conditions. This latest code has been improved and developed to allow the selection of stainless steel cladding property models to analyze improved fuels such as accident tolerant fuels. The purpose of this report is to confirm the prediction accuracy of FEMAXI-8 for the irradiation behavior of the new type of core fuel that is currently being developed. As a result of the verification, it was confirmed that FEMAXI-8 has sufficient analysis accuracy for the irradiation behavior of sodium-cooled fast reactor MOX fuel with stainless steel cladding, which exceeds the plutonium content and irradiation conditions of light water reactors. In the future, the analysis accuracy of FEMAXI-8 could be improved by adopting the O/M ratio dependence of MOX fuel thermal conductivity and the irradiation behavior evaluation model at high temperature.
Sakamoto, Yoshiaki
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 30(1), p.15 - 18, 2023/06
The entire process of nuclear power generation is called the nuclear fuel cycle, and each process generates various types of radioactive waste. These radioactive wastes are generated from the operation and decommissioning of these facilities, and are treated and disposed of appropriately according to their radioactivity concentrations and properties. This paper describes the basic outline of the nuclear fuel cycle and the fundamentals of the treatment and disposal of radioactive waste (including radioactive waste from the use of radioactive materials in facilities other than the nuclear fuel cycle), called the back end of the nuclear fuel cycle.
and Zr at precisely controlled high temperaturesShirasu, Noriko; Sato, Takumi; Suzuki, Akihiro*; Nagae, Yuji; Kurata, Masaki
Journal of Nuclear Science and Technology, 60(6), p.697 - 714, 2023/06
Times Cited Count：1 Percentile：72.91(Nuclear Science & Technology)Interaction tests between UO and Zr were performed at precisely controlled high temperatures between 1840 and 2000 C to understand the interaction mechanism in detail. A Zr rod was inserted in a UO crucible and then heat-treated at a fixed temperature in Ar-gas flow for 10 min. After heating in the range of 1890 to 1930 C, the Zr rod was deformed to a round shape, in which the post-analysis detected the significant diffusion of U into the Zr region and the formation of a dominant -Zr(O) matrix and a small amount of U-Zr-O precipitates. The abrupt progress of liquefaction was observed in the sample heated at around 1940 C or higher. The higher oxygen concentration in the -Zr(O) matrix suppressed the liquefaction progress, due to the variation in the equilibrium state. The U-Zr-O melt formation progressed by the selective dissolution of Zr from the matrix, and the selective diffusion of U could occur via the U-Zr-O melt.
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.
Kusaka, Ryoji; Kumagai, Yuta; Watanabe, Masayuki; Sasaki, Takayuki*; Akiyama, Daisuke*; Sato, Nobuaki*; Kirishima, Akira*
Journal of Nuclear Science and Technology, 60(5), p.603 - 613, 2023/05
Times Cited Count：1 Percentile：31.61(Nuclear Science & Technology)
methodNauchi, Yasushi*; Sato, Shunsuke*; Hayakawa, Takehito*; Kimura, Yasuhiko; Suyama, Kenya; Kashima, Takao*; Futakami, Kazuhiro*
Nuclear Instruments and Methods in Physics Research A, 1050, p.168109_1 - 168109_9, 2023/05
Times Cited Count：0 Percentile：0.02(Instruments & Instrumentation)Measurement of neutrons from spent nuclear fuel is performed in this study using the H method, which detects 2.223 MeV rays from neutron capture reaction of hydrogen using a highly pure germanium (HPGe) detector. The detection of the 2.223 MeV ray is affected by intense ray emission from fission products (FPs) because the emission rate of rays from the FP is seven orders of magnitude higher than the emission rate of neutrons. To shield the intense ray from the FP, the HPGe detector is placed off the axis of a collimator, whereas a polyethylene block is placed on the axis. In this geometry, the detector is shielded from the intense rays from the FP, but the detector can measure 2.223 MeV rays from the H reactions in the polyethylene block. The measured count rate of the 2.223 MeV rays is consistent with the expected rate within the statistical error, which is calculated based on the nuclide composition, which is primary 
Cm, estimated via depletion and decay calculations. Accordingly, the H method is considered feasible to quantify the number of neutron leakage from spent nuclear fuel assembly, which is applicable to certify burn up of the assembly.
