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Li, X.; Yamaji, Akifumi*; Sato, Ikken*; Yamashita, Takuya
Annals of Nuclear Energy, 214, p.111217_1 - 111217_13, 2025/05
Times Cited Count：0Nemoto, Yoshiyuki
JAEA-Research 2024-018, 16 Pages, 2025/03
JAEA-Research-2024-018.pdf:1.52MB
The author aimed to contribute to the analysis of the accident behavior of coated cladding with improved oxidation resistance by chromium (Cr) coating on the outer surface of conventional zirconium alloy fuel cladding, and investigated dependence of the oxidation behavior of Cr on the steam flow rate and on temperature. Coated cladding is being developed as one of the Accident Tolerant Fuel (ATF) claddings, and it is important to analyze the behavior of coated cladding under accidental conditions with high accuracy in the purpose of safety evaluation, for which the oxidation kinetics of Cr in high temperature steam is necessary. In this study, based on the results of oxidation tests in high temperature steam using a thermobalance, an oxidation kinetics for Cr in the temperature range and steam flow rate that encompass the conditions of considerable accident was proposed. The results can be used for future analyses in analysis codes such as SAMPSON, thereby contributing to the development of coated cladding.
Sono, Hiroki
Robutsuri No Kenkyu (Internet), (78), 12 Pages, 2025/03
The Static Experiment Critical Facility (STACY) was renovated from a "solution fuel reactor" to a "reactor using fuel rods and light water moderator", and restarted operation on August 2, 2024, after a hiatus of 13 years and 8 months. During that time, it took 8 years and 11 months to obtain its permission and approval, 3 years and 1 month for its construction, and 4 months for pre-operation inspections on the reactor performance. This article reports on the history of STACY from its birth to its restart of operation, as well as its future utilization.
Tanigawa, Masafumi; Seya, Kazuhito*; Asakawa, Naoya*; Hayashi, Hiroyuki*; Horigome, Kazushi; Mukai, Yasunobu; Kitao, Takahiko; Nakamura, Hironobu; Henzlova, D.*; Swinhoe, M. T.*; et al.
JAEA-Technology 2024-014, 63 Pages, 2025/02
JAEA-Technology-2024-014.pdf:3.02MB
The liquid waste treatment process generated sludge items at the plutonium conversion development facility. They are highly heterogeneous and contain large amounts of impurities (Na, Fe, Ni etc.). Therefore, the sludge items have very large sampling uncertainty and so the total measurement uncertainty is very large (approximately 24%). The plutonium scrap multiplicity counter (PSMC) measurement technique for sludge items was developed by joint research between the Japan Atomic Energy Agency (JAEA) and Los Alamos National Laboratory (LANL). The technical validity for sludge items using the PSMC was evaluated using various types of sample measurements and Monte Carlo N-Particle transport code calculations. The PSMC measurement parameters were found to be valid for use with sludge items and the validity of multiplicity analysis was confirmed and demonstrated through comparisons with standard MOX powder and a standard sludge. As a result, the PSMC measurement values were shown to be consistent and reasonable and the large amount of impurity (Fe, Ni etc.) did not impact the results. Therefore, the measurement uncertainty of the improved nuclear material accountancy (NMA) procedure by combined PSMC and high-resolution gamma spectrometry was shown to be 6.5%. In addition, an acceptance test was conducted using PSMC/HRGS and IAEA benchmark equipment. Measured Pu mass by both equipment agrees within the measurement uncertainty of each method, and so the validity for Pu mass measurement by PSMC/HRGS was confirmed. The above results confirm the applicability of PSMC/HRGS as an additional NMA method for sludge and a newly designed NDA procedure based on this study is applied to sludge for NMA in PCDF.
Aihara, Jun; Ueta, Shohei; Honda, Masaki*; Kasahara, Seiji; Okamoto, Koji*
JAEA-Research 2024-012, 98 Pages, 2025/02
JAEA-Research-2024-012.pdf:32.24MB
Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for the purpose of more safely reducing amount of recovered Pu. In Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO
(PuO-YSZ) small particle and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. ZrC layer is oxygen getter. In research project of Pu-burner HTGR carried out from fiscal year of 2014 to fiscal year of 2017, simulated CFPs were fabricated using Ce to simulate Pu. Moreover, simulated fuel compacts were fabricated using fabricated simulated CFPs. In this report, results of microstructural observation of CeO-YSZ and ZrC layer at each fabrication step are reported.
Naeem, M.*; Ma, Y.*; Tian, J.*; Kong, H.*; Romero-Resendiz, L.*; Fan, Z.*; Jiang, F.*; Gong, W.; Harjo, S.; Wu, Z.*; et al.
Materials Science & Engineering A, 924, p.147819_1 - 147819_10, 2025/02
Times Cited Count：0 Percentile：0.00(Nanoscience & Nanotechnology)Sono, Hiroki
Genshiryoku Kiko, Genken OB Kai Kaiho, (86), P. 2, 2025/01
The Static Experiment Critical Facility (STACY) was renovated from a "solution fuel reactor" to a "reactor using fuel rods and light water moderator", and restarted operation on August 2, 2024, after a hiatus of 13 years and 8 months. During that time, it took 8 years and 11 months to obtain its permission and approval, 3 years and 1 month for its construction, and 4 months for pre-operation inspections on the reactor performance. This article reports on the history of STACY from its birth to its restart of operation, as well as its future utilization.
Johansen, M. P.*; Gwynn, J. P.*; Carpenter, J. G.*; Charmasson, S.*; McGinnity, P.*; Mori, Airi; Orr, B.*; Simon-Cornu, M.*; Osvath, I.*
Critical Reviews in Environmental Science and Technology, 55(6), p.422 - 445, 2025/00
Times Cited Count：0 Percentile：0.00(Environmental Sciences)
Er(n,
)
Er and 
Hf(n,)
Hf reactionsNakamura, Shoji; Shibahara, Yuji*; Endo, Shunsuke; Rovira Leveroni, G.; Kimura, Atsushi
Journal of Nuclear Science and Technology, 14 Pages, 2025/00
Times Cited Count：0Ito, Tatsuya; Xu, S.*; Xu, X.*; Omori, Toshihiro*; Kainuma, Ryosuke*
Shape Memory and Superelasticity, 9 Pages, 2025/00
Fukuda, Kodai
Annals of Nuclear Energy, 208(1), p.110748_1 - 110748_10, 2024/12
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Endo, Akira
Annals of the ICRP, 52(4), p.5 - 7, 2024/12
In its Publication 155, International Commission on Radiological Protection (ICRP) has developed data on the Specific Absorbed Fraction (SAF) for reference males and females at ages of newborn, 1 year, 5 years, 10 years, and 15 years. The SAF represents the fraction of energy emitted within a source region which is absorbed in a target region per mass of the target region and is essential for calculating absorbed doses in organs or tissues for internal exposure. By combining the data of Publication 155 with the SAF data for reference adult males and females already published as Publication 133, an SAF dataset for the calculation of age-dependent dose coefficients for members of the public for environmental intakes of radionuclides has been completed. This, together with revised biokinetic models and nuclear decay data, means that the key building blocks for calculating new dose coefficients are in place. The outcome will soon be available in a series of ICRP Publications of Dose Coefficients for Intakes of Radionuclides by Members of the Public.
Wakui, Takashi; Saito, Shigeru; Futakawa, Masatoshi
Materials, 17(23), p.5925_1 - 5925_14, 2024/12
Times Cited Count：0 Percentile：0.00(Chemistry, Physical)The ductile properties of irradiated materials are one of the important indicators related to their structural integrity. Indentation tests are used for evaluating the ductile properties easily and rapidly. Constants in the material constitutive equation were identified via inverse analysis using the Kalman filter, such that the numerical experimental results reproduced the indentation test results. Numerical tensile experiments were conducted using stress-strain curves with the identified constants to obtain nominal stress-strain curves. Furthermore, two methods were proposed for evaluating the total elongation. Evaluated minimum total elongation was 10 %. The evaluation results of ion-irradiated materials were similar to the tensile test results of irradiated materials.
Metcalfe, R.*; Tachi, Yukio; Sasao, Eiji; Kawama, Daisuke*
Science of the Total Environment, 957, p.177375_1 - 177375_17, 2024/12
A safety case for an underground radioactive waste repository must show that groundwater will not in future transport radionuclides from the repository to the near-surface environment (the biosphere) in harmful quantities. Safety cases are developed step-wise throughout a programme to site and develop a repository. At early stages, before a site is selected, safety cases are generic and based on simplified safety assessment models of the disposal system that have conservative parameter values. Later, when site-specific conditions are known, more realistic models are needed for the long-term geo-environmental evolution and their impacts on radionuclide migration/retention. Uplift is one such environmental change, which may be particularly important in countries near active tectonic plate boundaries, such as Japan. Here we review the state of knowledge about how the properties of fractured granitic rocks evolve during uplift, based on studies in Japan. Hence, we present conceptual models and a generic scenario for mass transport and retardation processes in uplifting granitic rocks as a basis for realistic numerical models to underpin safety assessment.
Hirota, Noriaki; Nakano, Hiroko; Takeda, Ryoma; Ide, Hiroshi; Tsuchiya, Kunihiko; Kobayashi, Yoshinao*
Zairyo No Kagaku To Kogaku, 61(6), p.248 - 252, 2024/12
A comparative analysis of the 0.2 % yield stress in SUS304L stainless steel revealed that lower strain rates and higher temperatures significantly reduce yield stress. Grain refinement from 68.6 
m to 0.59 m minimally impacted the rate of yield stress reduction at slower strain rates. However, finer grains showed a decrease in yield stress at reactor operating temperature compared to room temperature. In slow strain rate tests under conditions promoting intragranular stress corrosion cracking (SCC), SUS304L with grain sizes of 28.4 m or smaller exhibited similar fracture strains comparable to those at reactor operating temperatures, whereas coarse-grained SUS304L showed reduced fracture strain. Microstructural analysis showed that in smaller grains, over 87 % of the fracture surface was ductile. In particular, SUS304L with 0.59 m grains exhibited a higher presence of {111} / 
3 boundaries, which decreased with grain growth. These results indicate that grain refinement will suppress intragranular SCC by slowing corrosion progression through increased {111} / 3 boundaries.
Sc, 
Cu, 
Zn, 
Ag, and 
InNakamura, Shoji; Shibahara, Yuji*; Endo, Shunsuke; Rovira Leveroni, G.; Kimura, Atsushi
Journal of Nuclear Science and Technology, 61(11), p.1415 - 1430, 2024/11
Times Cited Count：1 Percentile：0.00(Nuclear Science & Technology)Neutron capture cross-sections of nuclides targeted for decommissioning are necessary to contribute to the evaluation of radioactivity produced. The present study, Sc, Cu, 
Zn, Ag and In nuclides were selected as target ones, and their thermal-neutron capture cross-sections were measured by an activation method at Kyoto University Research Reactor. The thermal-neutron capture cross-sections were obtained as follows: 27.18
0.28 barn for Sc(n, )
Sc, 4.340.06 barn for Cu(n, )Cu, 0.7190.011 barn for Zn(n, )Zn, 4.050.05 barn for Ag(n, )
Ag and 8.530.27 barn for In(n, ) 
In
. The results for Sc and Zn nuclides supported evaluated values within the limits of uncertainties, while those for the other nuclides were slightly different from evaluated ones. The obtained results are useful not only for the evaluation of production amount, but also for the monitor selection other than Au and Co by considering those nuclides as flux monitors.
neutron diffractionNaeem, M.*; Ma, Y.*; Knowles, A. J.*; Gong, W.; Harjo, S.; Wang, X.-L.*; Romero-Resendiz, L.*; 6 of others*
Materials Science & Engineering A, 916, p.147374_1 - 147374_8, 2024/11
Times Cited Count：1 Percentile：59.42(Nanoscience & Nanotechnology)Okita, Shoichiro; Abe, Yutaka*; Tasaki, Seiji*; Fukaya, Yuji
Radioisotopes, 73(3), p.233 - 240, 2024/11
Miyazawa, Takeshi; Uwaba, Tomoyuki; Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Onizawa, Takashi; Ando, Masanori; Kaito, Takeji
JAEA-Technology 2024-009, 140 Pages, 2024/10
JAEA-Technology-2024-009.pdf:8.03MB
For the purpose of enhancing the reliability of fast reactor fuel designing using modified type 316 steel, the out-of-pile and in-pile mechanical data of modified type 316 steel cladding tubes and wrapper tubes were statistically analyzed with the knowledge on material science and engineering; the high-temperature strength equations of modified type 316 steel, which can be applied to high-dose neutron irradiation environment, were derived. The out-of-pile high-temperature tensile and creep data of modified type 316 steel cladding tubes and wrapper tubes were derived up to 900
C, which is higher than the upper limit temperature of anticipated transient event of fast reactor. Using the extended database, the best-fit equation and the lower limit equation were derived for out-of-pile 0.2% proof strength, ultimate tensile strength and creep rupture strength while the best-fit equation and the upper and lower limit equations for creep strain. Furthermore, the degradation factors for tensile and creep strength, which will be produced by in-reactor environment (i.e., neutron irradiation in liquid sodium), were evaluated using the existing neutron irradiation data of modified type 316 steel, which were derived using the experimental fast reactor Joyo, the French proto-type fast reactor Phenix, the American experimental fast reactor FFTF. The derived equations were validated by the comparison with the experimental data.
Nishihara, Kenji; Sugawara, Takanori; Fukushima, Masahiro; Iwamoto, Hiroki; Katano, Ryota; Abe, Takumi
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
A pilot plant for the accelerator-driven system is proposed as a scaled-down version of a lead-bismuth cooled ADS with 800 MW thermal output for transmutation of minor actinides. In this presentation, the design policy of the pilot plant is presented.
