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Shiotsu, Hiroyuki; Ito, Hiroto*; Sugiyama, Tomoyuki; Maruyama, Yu
Annals of Nuclear Energy, 163, p.108587_1 - 108587_9, 2021/12
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Shiotsu, Hiroyuki; Ito, Hiroto*; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11
Ito, Hiroto*; Shiotsu, Hiroyuki; Tanaka, Yoichi*; Nishihara, Satomichi*; Sugiyama, Tomoyuki; Maruyama, Yu
JAEA-Data/Code 2018-012, 42 Pages, 2018/10
JAEA-Data-Code-2018-012.pdf:4.93MB
Chemical composition of fission products transported in nuclear facilities in severe accidents is controlled by slower chemical reaction rates, therefore, it could be different from that evaluated on the chemical equilibrium assumption. Hence, it is necessary to evaluate the chemical composition with reaction kinetics. On the other hand, databases applicable to the analysis of nuclear facilities have not been constructed because knowledge of reaction rates of complex chemical reactions in severe accidents is currently limited. Accordingly, we have developed the CHEMKEq code based on a partial mixed model with chemical equilibrium and reaction kinetics to decrease uncertainties of the chemical composition caused by the reaction rate. The CHEMKEq code, under mass conservation law, firstly evaluates chemical species obeying the chemical equilibrium model, and then, relatively slow reactions are solved by the reaction kinetics model. Moreover, the CHEMKEq code has a multiplicity of use in evaluations of chemical composition because general chemical equilibrium and reaction kinetics models are also available and databases required to calculation are external file formats. This report is the user's guide of the CHEMKEq code, showing models, solution methods, structure of the code and calculation examples. And information to run the CHEMKEq code is summarized in appendixes.
Takamizawa, Hisashi; Ito, Hiroto; Nishiyama, Yutaka
Journal of Nuclear Materials, 479, p.533 - 541, 2016/10
Times Cited Count：6 Percentile：49.29(Materials Science, Multidisciplinary)To understand neutron irradiation embrittlement in high fluence regions, statistical analysis using the Bayesian nonparametric (BNP) method was performed for the Japanese surveillance and material test reactor irradiation database. The BNP method is essentially expressed as an infinite summation of normal distributions, with input data being subdivided into clusters with identical statistical parameters (such as mean and standard deviation) for each cluster to estimate shifts in ductile-to-brittle transition temperature (DBTT). Clusters typically depend on chemical compositions, irradiation conditions, and the irradiation embrittlement. Specific variables contributing to the irradiation embrittlement include the content of Cu, Ni, P, Si, and Mn in the pressure vessel, neutron flux, neutron fluence, and irradiation temperatures. It was found through numerous examinations that the measured shifts of DBTT correlated well with calculated ones. Data associated with the same materials were subdivided into the same clusters even if neutron fluences were significantly disparate among the results. This indicates that slowly developing or late-onset embrittlement mechanisms were not evident in the present study.
Zheng, X.; Ito, Hiroto; Tamaki, Hitoshi; Maruyama, Yu
Journal of Nuclear Science and Technology, 53(3), p.333 - 344, 2016/03
Times Cited Count：10 Percentile：68.36(Nuclear Science & Technology)Zheng, X.; Ito, Hiroto; Tamaki, Hitoshi; Maruyama, Yu
Mechanical Engineering Journal (Internet), 2(5), p.15-00032_1 - 15-00032_14, 2015/10
Zheng, X.; Ito, Hiroto; Kawaguchi, Kenji; Tamaki, Hitoshi; Maruyama, Yu
Reliability Engineering & System Safety, 138, p.253 - 262, 2015/06
Times Cited Count：9 Percentile：39.94(Engineering, Industrial)Ishikawa, Jun; Ito, Hiroto; Maruyama, Yu
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
Uchida, Kenichi*; Adachi, Hiroto; Kikuchi, Daisuke*; Ito, Shun*; Qiu, Z.*; Maekawa, Sadamichi; Saito, Eiji
Nature Communications (Internet), 6, p.5910_1 - 5910_8, 2015/01
Times Cited Count：47 Percentile：85.28(Multidisciplinary Sciences)Zheng, X.; Ito, Hiroto; Tamaki, Hitoshi; Maruyama, Yu
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 10 Pages, 2014/07
Integral severe accident code MELCOR 1.8.5 has been applied to estimating uncertainty of source term for the accident at Unit 2 of Fukushima Daiichi Nuclear Power Plant and to discussing important models or parameters influential to the source term. Forty-two parameters associated with models for the transportation of radioactive materials were chosen and narrowed down to 18 through a set of screening analysis. These 18 parameters in addition to 9 parameters relevant to in-vessel melt progression obtained by the preceded uncertainty study were inputted to the subsequent sensitivity analysis by Morris method. This one-factor-at-a-time approach can preliminarily identify inputs which have important effects on an output, and 17 important parameters were selected from the total of 27 parameters through this approach. The selected parameters have been integrated into uncertainty analysis by means of Latin Hypercube Sampling technique and Iman-Conover method, taking into account correlation between parameters. Cumulative distribution functions of representative source terms were obtained through the present uncertainty analysis assuming the failure of suppression chamber. Correlation coefficients between the outputs and uncertain input parameters have been calculated to identify parameters of great influences on source terms, which include parameters related to models on core components failure, models of aerosol dynamic process and pool scrubbing.
Ito, Hiroto; Zheng, X.; Tamaki, Hitoshi; Maruyama, Yu
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 8 Pages, 2014/07
Zheng, X.; Ishikawa, Jun; Ito, Hiroto; Tamaki, Hitoshi; Maruyama, Yu
JAEA-Review 2014-016, 32 Pages, 2014/06
JAEA-Review-2014-016.pdf:1.98MB
Boron carbide (B
C) is one kind of neutron absorbing control rod/blade materials used in light water reactors (LWRs). In Fukushima Daiichi Nuclear Power Station (NPS), all units used BC as absorber material. The degradation of control rod/blade will affect the early phase of in-vessel core melt progression. Furthermore, the release of carbon compound gases including carbon dioxide (CO
) as well as boron compounds due to the oxidation of BC with steam is possible to affect source terms of radioactive materials. Past experiments related to BC degradation and oxidation and numerical modeling in severe accidents codes are investigated in the current report with a main view to apply the acquired knowledge into the modification of THALES-2 developed at JAEA as an integral severe accident analysis code. The eutectic interactions of BC with other materials such as stainless steel and Zircaloy will lower the melting point of control rod/blade. The Nagase's correlations for eutectic interaction are selected as one of candidates to be applied into THALES-2. The oxidation reaction of BC with steam will release considerable amount of thermal energy, and form CO, boric acids and boron oxide, which could make an impact onto source terms by changing the pH value of a water pool where those dissolve. The IRSN correlation is chosen to be used in the modeling of oxidation reaction in THLAES-2.
Katsuyama, Jinya; Ito, Hiroto*; Li, Y.*; Osakabe, Kazuya*; Onizawa, Kunio; Yoshimura, Shinobu*
International Journal of Pressure Vessels and Piping, 117-118, p.56 - 63, 2014/05
Times Cited Count：9 Percentile：58.71(Engineering, Multidisciplinary)Several probabilistic fracture mechanical (PFM) analysis codes have been improved or developed in Japan, such as PASCAL-SP developed at JAEA, and PRAISE-JNES developed at JNES for structural integrity assessment of aged piping in nuclear power plants. Although they were developed for different purposes, they have similar functions. In this paper, in order to confirm the reliability and applicability of two PFM analysis codes, PASCAL-SP and PRAISE-JNES, benchmark analyses on piping failure probability have been carried out considering typical aging mechanisms, such as fatigue crack growth for piping materials in BWR plants Moreover, a criterion is proposed to judge whether the differences between the analysis results from two codes can be acceptable. Based on the proposed criterion, it is concluded that the analysis results of these two codes are in good agreements.
Li, Y.*; Ito, Hiroto*; Osakabe, Kazuya*; Onizawa, Kunio; Yoshimura, Shinobu*
International Journal of Pressure Vessels and Piping, 99-100, p.61 - 68, 2012/11
Times Cited Count：7 Percentile：50.73(Engineering, Multidisciplinary)A benchmark analysis was conducted using two probabilistic fracture mechanics analysis codes for aged piping in nuclear power plants, in order to confirm their reliability and applicability. These analysis codes have been improved or developed in Japan for the structural integrity evaluation and risk assessment considering the age related degradation mechanisms. In the benchmark analysis, the primary loop recirculation system piping in the boiling water reactor was selected as the typical piping system and stress corrosion cracking and fatigue were taken into account as the typical aging mechanisms. Moreover, a criterion was proposed for judging whether the differences between analysis results from the two codes are acceptable. Based on the benchmark analysis results and numerical investigation, it was concluded that the analysis results of these two codes agree very well.
Ito, Hiroto*; Li, Y.*; Osakabe, Kazuya*; Onizawa, Kunio; Yoshimura, Shinobu*
Journal of Mechanical Science and Technology, 26(7), p.2055 - 2058, 2012/07
Times Cited Count：2 Percentile：15.24(Engineering, Mechanical)Probabilistic fracture mechanics is a rational methodology in the structural integrity evaluation and risk assessment for aged piping in nuclear power plants. Several probabilistic fracture mechanical analysis codes have been improved or developed in Japan. In this paper, to verify the reliability and applicability of two of these codes, a benchmark analysis was conducted using their basic functions in consideration of representative piping systems in nuclear power plants and typical aging mechanisms. Based on the analysis results, we concluded that the analysis results of these two codes are in good agreement.
Katsuyama, Jinya; Tobita, Toru; Ito, Hiroto*; Onizawa, Kunio
Journal of Pressure Vessel Technology, 134(2), p.021403_1 - 021403_9, 2012/04
Times Cited Count：10 Percentile：45.01(Engineering, Mechanical)Stress corrosion crackings (SCCs) in recirculation pipes of Type 316L stainless steel (SS) have been observed near butt-welding joints. These SCCs in Type 316L SS grow near the welding zone mainly due to high tensile residual stress by welding. In present work, scatters of welding conditions such as heat input and welding speed were measured experimentally by fabricating a series of butt-welded pipes. Residual stress distributions were measured by stress relief and X-ray diffraction methods. The effects of welding conditions on residual stress have been evaluated by parametric FEM analyses considering the variation of some welding parameters based on the experiments. The effects of welding conditions on crack growth behavior have been also evaluated by using calculated residual stress distributions. It was clearly shown that the uncertainties on welding heat input and speed have strong influences on SCC growth behavior.
Katsuyama, Jinya; Ito, Hiroto; Tobita, Toru; Onizawa, Kunio
Yosetsu Gakkai Rombunshu (Internet), 28(2), p.193 - 202, 2010/06
Weld residual stress near the welded joint of pipe is one of the most important factors to assess the structural integrity of piping because the tensile residual stress becomes a driving force of a stress corrosion cracking (SCC). In this study, a weld simulation method has been developed and verified. An effect of uncertainty of welding conditions, such as scatters of heat input and welding speed during welding, on weld residual stress at the piping butt-welds was evaluated using the simulation method by varying the welding conditions. Probabilistic fracture mechanics analysis using PASCAL-SP was also performed to evaluate the effect of uncertainty of weld residual stress on the break probability of piping. It was clarified that the break probability increased with increasing the uncertainties of residual stress.
Ito, Hiroto; Kato, Daisuke*; Osakabe, Kazuya*; Nishikawa, Hiroyuki; Onizawa, Kunio
JAEA-Data/Code 2009-025, 135 Pages, 2010/03
JAEA-Data-Code-2009-025.pdf:17.49MB
As a part of the aging and structural integrity research for LWR components, new PFM (Probabilistic Fracture Mechanics) analysis code PASCAL-SP (PFM Analysis of Structural Components in Aging LWR - Stress Corrosion Cracking at Welded Joints of Piping) has been developed. This code evaluates the failure probabilities at welding lines of aged piping by a Monte Carlo method. PASCAL-SP treats stress corrosion cracking (SCC) in piping, including approaches of NISA and JSME FFS Code. The development of the code has been aimed to improve the accuracy and reliability of analysis by introducing new analysis methodologies and algorithms considering the latest knowledge in the SCC assessment and fracture criteria of piping. In addition, the accuracy of flaw detection and sizing at in-service inspection and residual stress distribution were modeled based on experimental data and introduced into PASCAL-SP. This report provides the user's manual and theoretical background of the code.
Onizawa, Kunio; Nishikawa, Hiroyuki; Ito, Hiroto
International Journal of Pressure Vessels and Piping, 87(1), p.2 - 10, 2010/01
Times Cited Count：46 Percentile：90.31(Engineering, Multidisciplinary)Probabilistic fracture mechanics (PFM) analysis codes for reactor pressure vessels (RPVs) and piping, called as PASCAL series has been developed at JAEA. The PASCAL2 code can evaluate the conditional probability of flaw initiation and fracture of an RPV under transient conditions such as pressurized thermal shock considering neutron irradiation embrittlement of the materials surrounding the reactor core of the RPV. Recent improvements of PASCAL2 are related to the treatment of weld-overlay cladding. The results of PFM analysis using the improved code have indicated that the residual stress by weld-overlay cladding affects the fracture probability of RPV. The results also indicate that the fracture toughness of cladding should be taken into consideration for the evaluation of realistic through-wall cracking probability. The PASCAL-SP code has been developed recently. The PASCAL-SP evaluates the probabilities of failures such as leakage and break of safety-related piping complying with Japanese regulation and rules. Effects of welding residual stress distribution as well as inspection accuracy are focused in this study. Residual stress distributions, which affect SCC behavior, have been determined by parametric FEM analyses verified through comparisons with welding experiments and incorporated to the code.
Onizawa, Kunio; Ito, Hiroto
NEA/CSNI/R(2009)2, p.275 - 285, 2009/00
Since probabilistic fracture mechanics (PFM) analysis method treats the scatter and uncertainties of data related to aging degradation, it is, therefore, useful to apply the PFM analysis to the piping reliability evaluation. Stress corrosion cracking (SCC) has been observed at some piping joints made by Austenitic stainless steel in BWR plants. In JAEA, we have been developing PFM analysis code, PASCAL-SP, for aged piping based on the latest knowledge on SCC. The PASCAL-SP evaluates the failure probability of piping at aged welded joints under SCC by a Monte Carlo method. Although many factors have some influence on failure probabilities in the PFM analysis, we are mainly focusing on the effects of in-service inspection and welding residual stress distribution. Through parametric PFM analyses by the PASCAL-SP, the effects of in-service inspection and the uncertainties of residual stress distribution on the failure probability are studied.
