Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
, 
, and X-ray spectrum analysisOshima, Masumi*; Goto, Jun*; Hayakawa, Takehito*; Asai, Masato; Kin, Tadahiro*; Shinohara, Hirofumi*
Isotope News, (790), p.19 - 23, 2023/12
When analyzing samples that contain many radionuclides at various concentrations, such as radioactive waste or fuel debris, it is difficult to apply general spectrum analysis methods and is necessary to chemically separate each nuclide before quantifying it. The chemical separation is especially essential for analysis using a liquid scintillation counter (LSC). In this report, the authors explain the newly developed spectral determination method (SDM) in which the entire spectrum is fitted to quantify radioactivity of nuclides mixed in a sample. By applying the SDM to - and X-ray spectrum measured by LSC and -ray spectrum measured by Ge detector simultaneously, the authors demonstrated that radioactivity of 40 radionuclides mixed in a sample at concentrations varying by two orders could be quantified, which is useful to simplify chemical separation process in radionuclide quantification.
Tsuchiya, Harufumi; Toh, Yosuke; Ozu, Akira; Furutaka, Kazuyoshi; Kitatani, Fumito; Maeda, Makoto; Komeda, Masao
Journal of Nuclear Science and Technology, 60(11), p.1301 - 1312, 2023/11
Times Cited Count：1 Percentile：72.91(Nuclear Science & Technology)
Nb by activation method and half-life of 
Nb by mass analysisNakamura, Shoji; Shibahara, Yuji*; Endo, Shunsuke; Kimura, Atsushi
Journal of Nuclear Science and Technology, 60(11), p.1361 - 1371, 2023/11
Times Cited Count：1 Percentile：72.91(Nuclear Science & Technology)The thermal-neutron capture cross section (
) and resonance integral (I) for Nb among nuclides for decommissioning were measured by an activation method and the half-life of Nb by mass analysis. Niobium-93 samples were irradiated with a hydraulic conveyer installed in the research reactor in Institute for Integral Radiation and Nuclear Science, Kyoto University. Gold-aluminum, cobalt-aluminum alloy wires were used to monitor thermal-neutron fluxes and epi-thermal Westcott's indexes at an irradiation position. A 25-
m-thick gadolinium foil was used to sort out reactions ascribe to thermal-and epi-thermal neutrons. Its thickness provided a cut-off energy of 0.133 eV. In order to attenuate radioactivity of 
Ta due to impurities, the Nb samples were cooled for nearly 2 years. The induced radio activity in the monitors and Nb samples were measured by -ray spectroscopy. In analysis based on Westcott's convention, the and I values were derived as 1.11
0.04 barn and 10.50.6 barn, respectively. After the -ray measurements, mass analysis was applied to the Nb sample to obtain the reaction rate. By combining data obtained by both -ray spectroscopy and mass analysis, the half-life of Nb was derived as (2.000.15)
10
years.
Maruyama, Shuhei; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 60(11), p.1372 - 1385, 2023/11
Times Cited Count：1 Percentile：72.91(Nuclear Science & Technology)Nakamura, Izumi*; Otani, Akihito*; Okuda, Yukihiko; Watakabe, Tomoyoshi; Takito, Kiyotaka; Okuda, Takahiro; Shimazu, Ryuya*; Sakai, Michiya*; Shibutani, Tadahiro*; Shiratori, Masaki*
Dai-10-Kai Kozobutsu No Anzensei, Shinraisei Ni Kansuru Kokunai Shimpojiumu (JCOSSAR2023) Koen Rombunshu (Internet), p.143 - 149, 2023/10
In 2019, the JSME Code Case for seismic design of nuclear power plant piping systems was published. The Code Case provides the strain-based fatigue criteria and detailed inelastic response analysis procedure as an alternative design rule to the current seismic design, which is based on the stress evaluation by elastic response analysis. In 2022, it was approved to revise the Code Case with improving the cycle counting method for fatigue evaluation to the Rain flow method. In addition, the discussion to incorporate the elastic-plastic behavior of support structures is now in progress for the next revision of the Code Case. This paper discusses the contents and background of the 2022 revision, the progress of the next revision, and future tasks.
Maeda, Makoto; Segawa, Mariko; Toh, Yosuke; Endo, Shunsuke; Nakamura, Shoji; Kimura, Atsushi
Journal of Radioanalytical and Nuclear Chemistry, 332(8), p.2995 - 2999, 2023/08
Times Cited Count：0 Percentile：0.01(Chemistry, Analytical)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.
Kawamura, Sho; Kikuchi, Masanobu; Hosoya, Toshiaki
JAEA-Technology 2021-041, 103 Pages, 2023/02
JAEA-Technology-2021-041.pdf:8.7MB
In response to new regulatory standard for research and test reactor which is enforced December 2013, JRR-3 got license in November 2018 by formulate new design basis ground motion. After that we evaluated for insertion property of control rod using that new design basis ground motion, and that evaluation results were accepted as approval of the design and construction method by Nuclear Regulation Authority. Now, we re-evaluated to insertion property of control rod about neutron absorber and follower fuel element by time history response analysis method. In this report, it shows that new results have sufficiency of margin compared with the past results that are accepted as approval of the design and construction method.
Kubo, Kotaro; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Journal of Nuclear Science and Technology, 59(3), p.357 - 367, 2022/03
Times Cited Count：5 Percentile：65.59(Nuclear Science & Technology)Dynamic probabilistic risk assessment (PRA), which handles epistemic and aleatory uncertainties by coupling the thermal-hydraulics simulation and probabilistic sampling, enables a more realistic and detailed analysis than conventional PRA. However, enormous calculation costs are incurred by these improvements. One solution is to select an appropriate sampling method. In this paper, we applied the Monte Carlo, Latin hypercube, grid-point, and quasi-Monte Carlo sampling methods to the dynamic PRA of a station blackout sequence in a boiling water reactor and compared each method. The result indicated that quasi-Monte Carlo sampling method handles the uncertainties most effectively in the assumed scenario.
Ichihara, Yoshitaka*; Nakamura, Naohiro*; Moritani, Hiroshi*; Horiguchi, Tomohiro*; Choi, B.
Nihon Genshiryoku Gakkai Wabun Rombunshi, 21(1), p.1 - 14, 2022/03
In this study, we aim to approximately evaluate the effect of nonlinearity of reinforced concrete structures through seismic response analysis using the equivalent linear analysis method. A simulation analysis was performed for the ultimate response test of the shear wall of the reactor building used in an international competition by OECD/NEA in 1996. The equivalent stiffness and damping of the shear wall were obtained from the trilinear skeleton curves proposed by the Japan Electric Association and the hysteresis curves proposed by Cheng et al. The dominant frequency, maximum acceleration response, maximum displacement response, inertia force-displacement relationship, and acceleration response spectra of the top slab could be simulated well up to a shear strain of approximately =2.010
. The equivalent linear analysis used herein underestimates the maximum displacement response at the time of ultimate fracture of approximately =4.010. Moreover, the maximum shear strain of the shear wall could not capture the locally occurring shear strain compared with that of the nonlinear analysis. Therefore, when employing this method to evaluate the maximum shear strain and test results, including those during the sudden increase in displacement immediately before the fracture, sufficient attention must be paid to its applicability.
Gunji, Satoshi; Tonoike, Kotaro; Clavel, J.-B.*; Duhamel, I.*
Journal of Nuclear Science and Technology, 58(1), p.51 - 61, 2021/01
Times Cited Count：1 Percentile：12.16(Nuclear Science & Technology)The new critical assembly STACY will be able to contribute to the validation of criticality calculations related to the fuel debris. The experimental core designs are in progress in the frame of JAEA/IRSN collaboration. This paper presents the method applied to optimize the design of the new STACY core to measure the criticality characteristics of pseudo fuel debris that simulated Molten Core Concrete Interaction (MCCI) of the fuel debris. To ensure that a core configuration is relevant for code validation, it is important to evaluate the reactivity worth of the main isotopes of interest and their k
sensitivity to their cross sections. In the case of the fuel debris described in this study, especially for the concrete composition, silicon is the nucleus with the highest k sensitivity to the cross section. For this purpose, some parameters of the core configuration, as for example the lattice pitches or the core dimensions, were adjusted using optimization algorithm to find efficiently the optimal core configurations to obtain high sensitivity of silicon capture cross section. Based on these results, realistic series of experiments for fuel debris in the new STACY could be defined to obtain an interesting feedback for the MCCI. This methodology is useful to design other experimental conditions of the new STACY.
Sugita, Yutaka; Taniguchi, Naoki; Makino, Hitoshi; Kanamaru, Shinichiro*; Okumura, Taisei*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(3), p.121 - 135, 2020/09
A series of structural analysis of disposal containers for direct disposal of spent fuel was carried out to provide preliminary estimates of the required pressure resistance thickness of the disposal container. Disposal containers were designed to contain either 2, 3 or 4 spent fuel assemblies in linear, triangular or square arrangements, respectively. The required pressure resistance thickness was evaluated using separation distance of the housing space for each spent fuel assembly as a key model parameter to obtain the required thickness of the body and then the lid of the disposal container. This work also provides additional analytical technical knowledge, such as the validity of the setting of the stress evaluation line and the effect of the model length on the analysis. These can then be referred to and used again in the future as a basis for conducting similar evaluations under different conditions or proceeding with more detailed evaluations.
Fukushima, Masahiro; Tojo, Masayuki*
Journal of Nuclear Science and Technology, 56(12), p.1061 - 1062, 2019/12
Times Cited Count：0 Percentile：0.32(Nuclear Science & Technology)Reactor Physics that treat the essentials of how fission nuclear reactors work fundamentally has important roles on safe operations and design studies of various types of nuclear reactors. From the latest activities in the field of reactor physics, this report summarizes some outstanding researches and developments published in scientific journals including the Journal of Nuclear Science and Technology.
Hashimoto, Shintaro; Sato, Tatsuhiko
Journal of Nuclear Science and Technology, 56(4), p.345 - 354, 2019/04
Times Cited Count：5 Percentile：48.99(Nuclear Science & Technology)Particle transport simulations based on the Monte Carlo method have been applied to shielding calculations. Estimation of not only statistical uncertainty related to the number of trials but also systematic one induced by unclear physical quantities is required to confirm the reliability of calculated results. In this study, we applied a method based on analysis of variance to shielding calculations. We proposed random- and three-condition methods. The first one determines randomly the value of the unclear quantity, while the second one uses only three values: the default value, upper and lower limits. The systematic uncertainty can be estimated adequately by the random-condition method, though it needs the large computational cost. The three-condition method can provide almost the same estimate as the random-condition method when the effect of the variation is monotonic. We found criterion to confirm convergence of the systematic uncertainty as the number of trials increases.
Choi, B.; Nishida, Akemi; Li, Y.; Muramatsu, Ken*; Takada, Tsuyoshi*
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07
After the 2011 Fukushima accident, nuclear power plants are required to take countermeasures against accidents beyond design basis conditions. In seismic probabilistic risk assessment (SPRA), uncertainty can be classified as either aleatory uncertainty, which cannot be reduced, or epistemic uncertainty, which can be reduced with additional knowledge and/or information. To improve the reliability of SPRA, efforts should be made to identify and reduce the epistemic uncertainty caused by the lack of knowledge. In this study, we focused on the difference in seismic response by modeling methods, which is related epistemic uncertainty. We conducted a seismic response analysis with two kinds of modeling methods; a three-dimensional finite-element model and a conventional sway-rocking stick model, by using simulated various input ground motions, which is related to aleatory uncertainty. And then we quantified the seismic floor response results of the various input ground motions of each modeling methods. For the uncertainty quantification related to different modeling methods, we further perform a statistical analysis of the floor response results of the nuclear reactor building. Finally, we discussed how to utilize the results from these calculations for the quantification of uncertainty in fragility analysis for SPRA.
Li, X.; Sato, Ikken; Yamaji, Akifumi*; Duan, G.*
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07
Molten corium-concrete interaction (MCCI) is an important ex-vessel phenomenon that could happen during the late phase of a hypothetical severe accident in a light water reactor. In the present study, a three-dimensional (3-D) numerical study has been performed to simulate COMET-L3 test carried out by KIT with a stratified molten pool configuration of simulant materials with improved MPS method. The heat transfer between corium/crust/concrete was modeled with heat conduction between particles. Moreover, the potential influence of the siliceous aggregates was also investigated by setting up two different case studies since there was previous study indicating that siliceous aggregates in siliceous concrete might contribute to different axial and radial concrete ablation rates. The simulation results have indicated that metal melt as corium in MCCI can have completely different characteristics regarding concrete ablation pattern from that of oxidic corium, which needs to be taken into consideration when assessing the containment melt-through time in severe accident management.
Kitamura, Yasunori*; Fukushima, Masahiro
Nuclear Science and Engineering, 186(2), p.168 - 179, 2017/05
Times Cited Count：1 Percentile：10.62(Nuclear Science & Technology)An inconsistency between the reactivity worth of short-size samples measured by the critical-water-level (CWL) method and that conventionally analysed for validating the nuclear data and the nuclear calculation methods has been known. The present study investigated this inconsistency in terms of a simple theoretical framework and proposed a simple and practical technique for correcting the measured sample reactivity worth without making supplementary experiments. A series of Monte Carlo calculations that simulated typical sample reactivity worth measurement by the CWL method showed that this inconsistency is effectively reduced by the present correction technique.
Maeda, Makoto; Komeda, Masao; Tobita, Hiroshi; Ozu, Akira; Kureta, Masatoshi; Bogucarska, T.*; Crochemore, J. M.*; Varasano, G.*; Pedersen, B.*
Dai-37-Kai Kaku Busshitsu Kanri Gakkai Nihon Shibu Nenji Taikai Rombunshu (CD-ROM), 7 Pages, 2017/02
JAEA and EC/JRC are carrying out collaborative research to develop NDA techniques that can be utilized for quantification of high radioactive special nuclear materials such as spent fuel and next generation minor actinide fuels. In the research, reliability of neutron transport codes is important because it is utilized for design and development of a demonstration system of next-generation Differential Die-away (DDA) technique in JAEA. In order to evaluate the reliability, actual neutron flux distribution in a sample cavity was examined in PUNITA device using JRC type DDA technique and JAWAS-T device using JAEA type DDA technique, and then the measurement results were compared with the simulation results obtained by the neutron transport codes. The neutron flux distribution in the target matrix was also examined in the PUNITA and compared with the simulation results. We report on the measurement and simulation results of the neutron flux distribution and evaluation results of the reliability of the neutron transport codes.
Nakajima, Norihiro; Nishida, Akemi; Miyamura, Hiroko; Iigaki, Kazuhiko; Sawa, Kazuhiro
Kashika Joho Gakkai-Shi (USB Flash Drive), 36(Suppl.2), 4 Pages, 2016/10
Since nuclear power plants have dimensions approximately 100m
and their structures are an assembly made up of over 10 million components, it is not convenient to experimentally analyze its behavior under strong loads of earthquakes, due to the complexity and hugeness of plants. The proposed system performs numerical simulations to evaluate the behaviors of an assembly like a nuclear facility. The paper discusses how to carry out visual analysis for assembly such as nuclear power plants. In a result discussion, a numerical experiment was carried out with a numerical model of High Temperature engineering Test Reactor of Japan Atomic Energy Agency and its result was compared with observed data. A good corresponding among them was obtained as a structural analysis of an assembly by using visualization. As a conclusion, a visual analytics methodology for assembly is discussed.
Chimi, Yasuhiro; Takamizawa, Hisashi; Kasahara, Shigeki*; Iwata, Keiko; Nishiyama, Yutaka
Nuclear Engineering and Design, 307, p.411 - 417, 2016/10
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)To investigate influential parameters for irradiation-assisted stress corrosion cracking (IASCC) growth behavior, we attempt to analyze statistically existing data on the crack growth rate (CGR) in irradiated austenitic stainless steels (SSs) in boiling water reactor (BWR) environments using the Bayesian nonparametric (BNP) method. From the probability distribution of CGR and some input parameters, such as yield stress of irradiated material (
), stress intensity factor (
), electrochemical corrosion potential (ECP), and fast neutron fluence, the mean CGR is estimated and compared with the measured CGR. The analytical results show good reproducibility of the measured CGR. The results also indicate the possible neutron fluence effects on CGR in high CGR region (i.e., high neutron fluence condition) by radiation-induced segregation (RIS), localized deformation, and/or other mechanisms than radiation hardening.
