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Li, L.*; Miyamoto, Goro*; Zhang, Y.*; Li, M.*; Morooka, Satoshi; Oikawa, Katsunari*; Tomota, Yo*; Furuhara, Tadashi*
Journal of Materials Science & Technology, 184, p.221 - 234, 2024/06
Times Cited Count：0 Percentile：0(Materials Science, Multidisciplinary)Ito, Takashi; Kadono, Ryosuke*
Journal of the Physical Society of Japan, 93(4), p.044602_1 - 044602_7, 2024/04
Hirota, Noriaki; Nakano, Hiroko; Fujita, Yoshitaka; Takeuchi, Tomoaki; Tsuchiya, Kunihiko; Demura, Masahiko*; Kobayashi, Yoshinao*
The IV International Scientific Forum "Nuclear Science and Technologies"; AIP Conference Proceedings 3020, p.030007_1 - 030007_6, 2024/01
Dynamic strain aging (DSA) and intergranular stress corrosion cracking (intragranular SCC) occur in high temperature pressurized water simulating a boiling water reactor environment due to changes in dissolved oxygen (DO) content, respectively. In order to clearly understand the difference between these phenomena, the mechanism of their occurrence was summarized. As a result, it was found that DSA due to intragranular cracking occurred in SUS304 stainless steel at low DO 
1 ppb, while DSA was suppressed at DO 100 to 8500 ppb due to the formation of oxide films on the surface. On the other hand, when DO was increased to 20000 ppb, the film was peeled from the matrix, O element diffused to the grain boundary of the matrix, resulting in intergranular SCC. These results are indicated that the optimum DO concentration must be adjusted to suppress crack initiation due to DSA and intergranular SCC.
Tada, Kenichi; Endo, Tomohiro*
Journal of Nuclear Science and Technology, 60(11), p.1397 - 1405, 2023/11
Times Cited Count：1 Percentile：72.91(Nuclear Science & Technology)The probability table method is a well-known method for addressing self-shielding effects in the unresolved resonance region. A long computational time is required to generate the probability table. The effective way to reduce the generation time of the probability table is the reduction of the number of ladders. The purpose of this study is the estimation of the optimal number of ladders using the statistical uncertainty in the probability table. To this end, the statistical uncertainty quantification method of the probability table was developed and the convergence behavior of the statistical uncertainty was investigated. The product of the probability table and the average cross section was considered the target of the statistical uncertainty. The convergence rate was affected by the average level spacing and reduced neutron width. The generation time of the probability table was less than half when the input parameter was changed from the number of ladders to the tolerance value.
Okita, Shoichiro; Goto, Minoru
Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 10 Pages, 2023/10
Tada, Kenichi
Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 8 Pages, 2023/10
The number of energy grids of the thermal neutron scattering law data has a large impact on the data size of a cross section file of continuous energy Monte Carlo calculation codes. The optimization of the number of energy grids is an effective way to reduce the data size. This study developed the linearization method of the thermal neutron scattering cross section to optimize the number of energy grids and the linearization function was implemented in the nuclear data processing code FRENDY. The linearization process which is used in the resonance reconstruction and the Doppler broadening was adopted. The criticality benchmarks which use ZrH as the moderator were calculated to estimate the impact of the difference of the energy grids on neutronics calculations. The calculation results indicate that the linearization of the thermal neutron scattering cross section improves the prediction accuracy of neutronics calculations.
Kubo, Kotaro; Zheng, X.; Tanaka, Yoichi; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Proceedings of the Institution of Mechanical Engineers, Part O; Journal of Risk and Reliability, 237(5), p.947 - 957, 2023/10
Times Cited Count：4 Percentile：69.72(Engineering, Multidisciplinary)Probabilistic risk assessment (PRA) is a method used to assess the risks associated with large and complex systems. However, the timing at which nuclear power plant structures, systems, and components are damaged is difficult to estimate if the risk of an external event is evaluated using conventional PRA based on event trees and fault trees. A methodology coupling thermal-hydraulic analysis with external event simulations using Risk Assessment with Plant Interactive Dynamics (RAPID) is therefore proposed to overcome this limitation. A flood propagation model based on Bernoulli's theorem was applied to represent internal flooding in the turbine building of the pressurized water reactor. Uncertainties were also taken into account, including the flow rate of the floodwater source and the failure criteria for the mitigation systems. The simulated recovery actions included the operator isolating the floodwater source and using a drainage pump; these actions were modeled using several simplifications. Overall, the results indicate that combining isolation and drainage can reduce the conditional core damage probability upon the occurrence of flooding by approximately 90%.
(cr) at 318 K in the presence of ironYoshida, Yasushi*; Kitamura, Akira; Shibutani, Sanae*
Journal of Nuclear Science and Technology, 60(8), p.900 - 910, 2023/08
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Takahatake, Yoko; Watanabe, So; Arai, Tsuyoshi*; Sato, Takahiro*; Shibata, Atsuhiro
Applied Radiation and Isotopes, 196, p.110783_1 - 110783_5, 2023/06
Times Cited Count：0 Percentile：0.01(Chemistry, Inorganic & Nuclear)Tada, Kenichi; Kondo, Ryoichi; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 60(6), p.624 - 631, 2023/06
Times Cited Count：2 Percentile：53.91(Nuclear Science & Technology)The sensitivity analysis and the uncertainty quantification have an important role in improving the evaluated nuclear data library. The current computational performance enables us to the sensitivity analysis and uncertainty quantification using the continuous energy Monte Carlo calculation code. The ACE file perturbation tool was developed for these calculations using modules of FRENDY. This tool perturbs the microscopic cross section, the number of neutrons per fission, and the fission spectrum. The uncertainty quantification using the random sampling method is also available if the user prepares the covariance matrix. The uncertainty of the k-effective using the perturbation tool was compared to the current sensitivity analysis codes SCALE/TSUNAMI and MCNP/KSEN. The comparison results indicated that the random sampling method using this tool accurately estimates the uncertainty of k-effective.
Tada, Kenichi
Kaku Deta Nyusu (Internet), (135), p.1 - 10, 2023/06
This article summarizes presentations at the IAEA technical meeting on nuclear data processing. In this technical meeting, the current development status of nuclear data processing codes and comparisons of the processing results using these codes were presented.
Kubo, Kotaro
Science and Technology of Nuclear Installations, 2023, p.7402217_1 - 7402217_12, 2023/06
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Tada, Kenichi; Endo, Tomohiro*
EPJ Web of Conferences, 284, p.14013_1 - 14013_4, 2023/05
Times Cited Count：0 Percentile：0.21(Nuclear Science & Technology)The self-shielding effect in the unresolved resonance region has a large impact on the fast- and intermediate-spectrum reactors. The probability table method is widely used for continuous-energy Monte Carlo calculation codes to treat the effect. In this method, a table provides the probability distribution of the cross-section for a nuclide in the given energy grid points. The table is generated by averaging with a lot of "ladders" which represent pseudo resonance structures. Though many nuclear data processing codes require the number of ladders as an input parameter to generate the probability table, an optimal number of ladders has not been investigated. Our previous study revealed that the suitable number of ladders depends on the nuclide and its resonance parameters. This result indicates that it is very difficult for users to find the optimal number of ladders. We developed the calculation method of the statistical uncertainty for the probability table generation.
Abe, Satoshi; Shibamoto, Yasuteru
Nuclear Engineering and Technology, 55(5), p.1742 - 1756, 2023/05
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
Kubo, Kotaro; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Journal of Nuclear Science and Technology, 60(4), p.359 - 373, 2023/04
Times Cited Count：5 Percentile：78.52(Nuclear Science & Technology)Probabilistic risk assessment (PRA) is an essential approach to improving the safety of nuclear power plants. However, this method includes certain difficulties, such as modeling of combinations of multiple hazards. Seismic-induced flooding scenario includes several core damage sequences, i.e., core damage caused by earthquake, flooding, and combination of earthquake and flooding. The flooding fragility is time-dependent as the flooding water propagates from the water source such as a tank to compartments. Therefore, dynamic PRA should be used to perform a realistic risk analysis and quantification. This study analyzed the risk of seismic-induced flooding events by coupling seismic, flooding, and thermal-hydraulics simulations, considering the dependency between multiple hazards explicitly. For requirements of safety improvement, especially in light of the Fukushima Daiichi Nuclear Power Plant accident, sensitivity analysis was performed on the seismic capacity of systems, and the effectiveness of alternative steam generator injection by a portable pump was estimated. We demonstrate the use of this simulation-based dynamic PRA methodology to evaluate the risk induced by a combination of hazards.
Abe, Satoshi; Okagaki, Yuria
Nuclear Engineering and Design, 404, p.112165_1 - 112165_14, 2023/04
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Okumura, Takuma*; Hashimoto, Tadashi; 40 of others*
Physical Review Letters, 130(17), p.173001_1 - 173001_7, 2023/04
Times Cited Count：1 Percentile：83.24(Physics, Multidisciplinary)Tada, Kenichi
Robutsuri No Kenkyu (Internet), (75), 13 Pages, 2023/03
In addition to nuclear data processing, FRENDY has various functions such as editing nuclear data and plotting cross section data. This document introduces these functions.
Tada, Kenichi
Shahei Kaiseki No V&V Gaidorain Sakutei Ni Mukete, p.11 - 16, 2023/03
An overview of the nuclear data processing code FRENDY is introduced for shielding calculation code users who are not familiar with FRENDY. This paper explains the nuclear data processing flow in FRENDY, the purpose of use, input examples, verification and validation reports, and so on.
