Mano, Akihiro; Imai, Ryuta*; Miyamoto, Yuhei*; Lu, K.; Katsuyama, Jinya; Li, Y.
International Journal of Pressure Vessels and Piping, 199, p.104700_1 - 104700_13, 2022/10
Elastic-plastic analyses based on finite element methods are widely applied to simulate the nonlinear behaviors of materials. When the analysis is conducted by an implicit method, the stress values are generally updated with a time increment by using the so-called return mapping algorithm. This algorithm requires solving simultaneous nonlinear equations related to a constitutive model. In the present paper, we proposed a general method to reduce the number of equations in the return mapping algorithm based on the implicit function theorem. In addition, the proposed method was applied to the Gurson-Tvergaard-Needleman (GTN) model that considers the influence of damage due to nucleation and growth of microscopic void in materials in the simulation of the nonlinear behaviors. By using the GTN model with the proposed method, an elastic-plastic analysis was performed by the implicit method for a 4-point bending test of pipe with a through-wall crack. The numerical solution of the variation of the load-load line displacement from the analysis agreed with experimental result. Thus, we concluded that the proposed method is useful for simulating nonlinear behaviors, including void nucleation and growth in materials.
Nakamura, Hironori*; Hayakawa, Satoshi*; Shibata, Akihiro*; Sasa, Kyohei*; Yamano, Hidemasa; Kubo, Shigenobu
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 7 Pages, 2022/10
In order to evaluate long-term coolablity of the debris-bed with decay heat, a three-dimensional calculation method coupled with the debris bed module was developed in this study. The coupled code calculation results show that natural circulation of the coolant between the hot pool and the cold pool is established through the four intermediate heat exchangers after the activation of the dipped direct heat exchangers. The cold pool with the debris-bed is continually cooled not only by the natural circulation flow, but also by heat transfer to the hot pool through the plenum separation plate between the hot pool and the cold pool. The effect of the three-dimensional flow field around the core catcher on the temperature in the debris-bed is about 20K under the current calculation condition.
Yamano, Hidemasa; Kubo, Shigenobu; Kan, Taro*; Shibata, Akihiro*; Hourcade, E.*; Dirat, J. F.*
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 10 Pages, 2022/08
In this paper, the approach to event tree development and the scope of the event tree analysis were described with key points on core catcher loading. For the analytical conditions, two core catcher loading conditions were given as bounding and conservative cases. For important heading of the event tree, key important phenomena were included: strong back design, fuel-coolant interaction and quench in the sodium plenum design, jet attack, criticality and coolability on the core catcher. In this paper, preliminary trial quantification was attempted using a probability ranking table which is based on engineering judgement. This event tree analysis has identified the dominant sequence, and clarified the effect of the core catcher loading and effectiveness of design measures. This study suggests that the criticality measure is very important for the core catcher study.
Yamano, Hidemasa; Kubo, Shigenobu; Sasa, Kyohei*; Shibata, Akihiro*; Hourcade, E.*; Dirat, J. F.*
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 9 Pages, 2022/08
This paper describes coolability evaluations of a debris bed with a variety of decay heat removal system (DHRS) operating conditions with a whole vessel model assuming fuel accumulation on the core catcher in a short term. The evaluation tool is a one-dimensional plant dynamics code, Super-COPD, with a debris bed module. The coolability evaluations have indicated that the current core catcher design secures sufficient natural circulation flows around the core catcher to ensure the debris bed cooling when at least one circuit of DHRS was activated. Sensitivity analyses under a pessimistic condition have shown that the debris bed is coolable with at least one circuit of improved DHRS even if most of fuel accumulates on the core catcher in a short term.
Yamaguchi, Yoshihito; Mano, Akihiro; Li, Y.
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 10 Pages, 2022/07
The steam generator (SG) tube is one of the important components in pressurized water reactors. Flaws such as wall-thinning or stress corrosion cracking have been reported in SG tubes. The burst pressure where both the internal and external pressures from the primary and secondary coolant systems are considered must be predicted to assess the structural integrity of SG tubes. Burst tests were performed by various organizations. On the basis of the test results, failure estimation methods were proposed. In this study, previous burst test data and existing failure estimation methods for SG tubes with wall-thinning or crack were investigated. As a result, the coefficient of the existing estimation method for SG tube with uniform wall-thinning was updated. In addition, failure estimation methods that are suitable for SG tubes with crack or local wall-thinning were proposed by considering the effects of the flaw shape and size on the burst pressure. The applicability of the failure estimation methods was confirmed by comparing the predicted results with the burst test data in actual SG tubes.
Tregoning, R.*; Wallace, J.*; Bouydo, A.*; Costa-Garrido, O.*; Dillstrm, P.*; Duan, X.*; Heckmann, K.*; Kim, Y.-B.*; Kim, Y.*; Kurth-Twombly, E.*; et al.
Transactions of 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 11 Pages, 2022/07
Fourteen organizations, representing eleven countries, participated in a leak-before-break (LBB) benchmark exercise that compared results from analyses among participating countries and identified the effects of weld residual stress (WRS) and crack morphology on crack opening displacement (COD), critical bending moment (CBM), and leak rate (LR) results. The participants determined whether the initial problem would meet their country's LBB acceptance criteria and then evaluated the effects of crack morphology and WRS for a prescribed crack size, geometry and loading. Six out of fourteen participants indicated that the initial problem met their LBB requirements. In the follow-on tasks, differences among the participant's CBM predictions were principally due to the material properties used in the analysis while the type of failure model chosen contributed much less. Most of the differences in the LR predictions were directly attributable to differences among the COD models, but a portion was attributable to the treatment of crack face pressure (CFP). The benchmark identified several aspects of an LBB analysis that could support a more realistic evaluation.
Kubo, Shigenobu; Payot, F.*; Yamano, Hidemasa; Bertrand, F.*; Bachrata, A.*; Saas, L.*; Journeau, C.*; Gosse, S.*; Quaini, A.*; Shibata, Akihiro*; et al.
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 8 Pages, 2022/04
Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.
Journal of Pressure Vessel Technology, 144(1), p.011506_1 - 011506_9, 2022/02
In the past few decades, the cracks because of stress corrosion cracking (SCC) have been detected in the dissimilar weld joints welded using nickel based alloy in piping system of boiling water reactors. Thus, the structural integrity assessment for such weld joints has become important. Nowadays, probabilistic fracture mechanics (PFM) analysis is recognized as a rational method for structural integrity assessment because it can consider inherent uncertainties of various influencing factors as probability distributions and quantitatively evaluate the failure probability of a cracked component. The Japan Atomic Energy Agency has developed a PFM analysis code PASCAL-SP for a probabilistic structural integrity assessment of weld joint in pipe in nuclear power plant. This study improves the analysis functions of PASCAL-SP for weld joint welded using nickel based alloy in boiling water reactor susceptible to SCC. As an analysis example of the improved version of PASCAL-SP, the failure probability of a weld joint is quantitatively evaluated. Furthermore, sensitivity analyses are conducted concerning the effect of leak detection and in-service inspection. From the analysis results, it is concluded that the improved version of PASCAL-SP is useful for structural integrity assessment.
Kumagai, Tomohisa*; Miura, Yasufumi*; Miura, Naoki*; Marie, S.*; Almahdi, R.*; Mano, Akihiro; Li, Y.; Katsuyama, Jinya; Wada, Yoshitaka*; Hwang, J.-H.*; et al.
Journal of Pressure Vessel Technology, 144(1), p.011509_1 - 011509_18, 2022/02
To predict fracture behavior for ductile materials, some ductile fracture simulation methods different from classical approaches have been investigated based on appropriate models of ductile fracture. For the future use of the methods to overcome restrictions of classical approaches, the applicability to the actual components is of concern. In this study, two benchmark problems on the fracture tests supposing actual components were provided to investigate prediction ability of simulation methods containing parameter decisions. One was the circumferentially through-wall and surface cracked pipes subjected to monotonic bending, and the other was the circumferentially through-wall cracked pipes subjected to cyclic bending. Participants predicted the ductile crack propagation behavior by their own approaches, including FEM employed GTN yielding function with void ratio criterion, are FEM employed GTN yielding function, FEM with fracture strain or energy criterion modified by stress triaxiality, XFEM with J or delta J criterion, FEM with stress triaxiality and plastic strain based ductile crack propagation using FEM, and elastic-plastic peridynamics. Both the deformation and the crack propagation behaviors for monotonic bending were well reproduced, while few participants reproduced those for cyclic bending. To reproduce pipe deformation and fracture behaviors, most of groups needed parameters which were determined toreproduce pipe deformation and fracture behaviors in benchmark problems themselves and it is still difficult to reproduce them by using parameters only from basic materials tests.
Zhang, T.; Lu, K.; Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.
Fatigue & Fracture of Engineering Materials & Structures, 44(12), p.3399 - 3415, 2021/12
The Gurson-Tvergaard-Needleman (GTN) model is considered a promising approach in failure prediction as it takes the micromechanical behavior of ductile metals into consideration and its function exhibits a relatively clear physical meaning. Although the GTN model has been widely investigated in the past decades, its engineering applications have scarcely progressed due to the difficulty in determining the eight strongly coupled parameters. Based on the physical background of GTN model, a set of methods was established to determine the parameters in the GTN model. The knowledge of continuum damage mechanics was used to experimentally determine the development of void volume fraction through the variation of effective Young's modulus in a uniaxial tensile test, and three parameters regarding void nucleation were analytically derived using a newly established method. Other parameters in the GTN model were also uniquely determined through a joint use of the chemical composition analysis (for the initial void volume fraction), the cell model analyses (for the two constitutive parameters), and the inverse finite element method (for the two failure parameters). The reliability of this novel parameter determination method was verified through the failure prediction of both cracked and uncracked specimens of carbon steel STPT410.
Mano, Akihiro; Katsuyama, Jinya; Li, Y.
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 7 Pages, 2021/07
Probabilistic fracture mechanics (PFM) is expected as a more rational methodology for the structural integrity assessments of nuclear power components because it can consider the inherent probabilistic distributions of various influencing factors and quantitatively evaluate the failure probabilities of the components. The Japan Atomic Energy Agency (JAEA) has developed a PFM analysis code, PASCAL-SP, to evaluate the failure probabilities of piping caused by aging degradation mechanisms, such as fatigue and stress corrosion cracking in the environments of both pressurized water and boiling water reactors. To improve confidence in the analysis results obtained from PASCAL-SP, a benchmarking study was conducted together with the PFM analysis code, xLPR, which was developed jointly by the U.S. Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute. The benchmarking study was composed of deterministic and probabilistic analyses related to primary water stress corrosion cracking in a dissimilar metal weld joint in a pressurized water reactor surge line. The analyses were conducted independently by NRC staff and JAEA using their own codes and under common analysis conditions. In the present paper, the analysis conditions for the deterministic and probabilistic analyses are described in detail, and the analysis results obtained from the xLPR and PASCAL-SP codes are presented. It was confirmed that the analysis results obtained from the two codes were in good agreement.
Yamaguchi, Yoshihito; Mano, Akihiro; Katsuyama, Jinya; Masaki, Koichi*; Miyamoto, Yuhei*; Li, Y.
JAEA-Data/Code 2020-021, 176 Pages, 2021/02
In Japan Atomic Energy Agency, as a part of researches on the structural integrity assessment and seismic safety assessment of aged components in nuclear power plants, a probabilistic fracture mechanics (PFM) analysis code PASCAL-SP (PFM Analysis of Structural Components in Aging LWR - Stress Corrosion Cracking at Welded Joints of Piping) has been developed to evaluate failure probability of piping. The initial version was released in 2010, and after that, the evaluation targets have been expanded and analysis functions have been improved based on the state-of-the art technology. Now, it is released as Ver. 2.0. In the latest version, primary water stress corrosion cracking in the environment of Pressurized Water Reactor, nickel based alloy stress corrosion cracking in the environment of Boiling Water Reactor, and thermal embrittlement can be taken into account as target age-related degradation. Also, many analysis functions have been improved such as incorporations of the latest stress intensity factor solutions and uncertainty evaluation model of weld residual stress. Moreover, seismic fragility evaluation function has been developed by introducing evaluation methods including crack growth analysis model considering excessive cyclic loading due to large earthquake. Furthermore, confidence level evaluation function has been incorporated by considering the epistemic and aleatory uncertainties related to influence parameters in the probabilistic evaluation. This report provides the user's manual and analysis methodology of PASCAL-SP Ver. 2.0.
Mano, Akihiro; Katsuyama, Jinya; Li, Y.
Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 7 Pages, 2020/08
A probabilistic fracture mechanics (PFM) analysis code PASCAL-SP has been developed by Japan Atomic Energy Agency (JAEA) for evaluating the failure probability of piping in nuclear power plant considering aged-related degradations such as primary water stress corrosion cracking (PWSCC) in pressurized water reactor environments and fatigue. To strengthen the confidence of analysis results, benchmarking study is being performed with PFM analysis code xLPR which has been developed by U.S.NRC in collaboration with EPRI. The benchmarking study consists of deterministic and probabilistic analyses on PWSCC under the common analysis conditions. In addition, deterministic sensitivity analysis on weld residual stress distributions is also included in the benchmarking study. These analyses are carried out by U.S.NRC and JAEA independently using their own codes. At current stage, the deterministic analyses by both xLPR and PASCAL-SP codes have been finished and probabilistic analyses are underway. This paper presents the details of conditions and comparisons of the results between the two codes in the deterministic analyses. In the deterministic analyses, both codes provided almost the same results including the values of stress intensity factor. In addition, probabilistic analysis conditions and results obtained from PASCAL-SP are presented.
Katsuyama, Jinya; Miyamoto, Yuhei*; Lu, K.; Mano, Akihiro; Li, Y.
Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 8 Pages, 2020/08
We have developed a probabilistic fracture mechanics (PFM) analysis code PASCAL4 for evaluating failure frequency of reactor pressure vessels (RPVs). It is known that flaw distributions have an important role in failure frequency calculation in PFM analysis. Previously, we proposed likelihood function to obtain more realistic flaw distributions applicable for both case when flaws are detected and when there is no flaw indication as the inspection results based on Bayesian update methodology. Here, it can be applied to independently obtain posterior distributions of flaw depth and density. In this study, we improve the likelihood function to enable them to update flaw depth and density simultaneously. Based on the improved likelihood function, an example is presented in which flaw distributions are estimated by reflecting NDI results through Bayesian update and PFM analysis. The results indicate that the improved likelihood functions are useful for estimating flaw distributions.
Mano, Akihiro; Katsuyama, Jinya; Li, Y.
Mechanical Engineering Journal (Internet), 7(3), p.19-00567_1 - 19-00567_11, 2020/06
Non-destructive examinations (NDEs) have an important role in assurance of the structural integrity of nuclear components including pipe lines. In Japanese nuclear power plants, NDEs are performed for welds in piping in accordance with the rules such as the Rules on Fitness-for-Service for Nuclear Power Plants of the Japan Society of Mechanical Engineers. For the welds where stress corrosion cracking (SCC) is not postulated, NDEs are performed in each 10-year interval. For each interval, the extent of examination is specified in the rules. In general, there are two kinds of sampling method for selecting welds to be examined in each interval considering the specified extent of examination. The first method is the fixed location sampling method, in which welds for NDEs are same as those examined in the last interval. The second method is the random location sampling method, in which welds for NDEs are selected from those not examined in the last interval. The selection of the sampling method is important to assure the structural integrity of piping. Probabilistic fracture mechanics (PFM) analysis which is one of rational structural integrity assessment methods can quantitatively calculate failure probability of welds in piping considering aging degradation mechanisms such as SCC and fatigue as well as crack detections and repair of cracked welds through NDE. In this study, to clarify the influence of the sampling methods on structural integrity of piping, we evaluated the failure probability of a typical nuclear piping considering NDEs based on the two sampling methods through PFM analysis. From the results, we clarified the quantitative influence of two sampling methods on failure probability of piping.
Mano, Akihiro; Katsuyama, Jinya; Miyamoto, Yuhei*; Yamaguchi, Yoshihito; Li, Y.
International Journal of Pressure Vessels and Piping, 179, p.103945_1 - 103945_6, 2020/01
Weld residual stress (WRS) is one of the most important factors in the structural integrity assessment of piping welds, and it is considered a driving force for crack growth. It is characterized by large uncertainty. For more rational assessment, it is important to consider the uncertainty of WRS for evaluating crack growth behavior in probabilistic fracture mechanics (PFM) analysis. In existing PFM analysis codes, WRS uncertainty is set by statistically processing the results of multiple finite element analyses. This process depends on the individual performing PFM analysis, which may lead to uncertainties whose sources would be different from the original WRS. In this study, we developed a new WRS evaluation model based on Fourier transformation, and the model was incorporated into PASCAL-SP, which has been developed by Japan Atomic Energy Agency. Through improvements to the code, WRS uncertainty can be considered automatically and appropriately by inputting multiple WRS analysis results directly as input data for PFM analysis.
Watanabe, Tadashi*; Katsuyama, Jinya; Mano, Akihiro
International Journal of Nuclear and Quantum Engineering (Internet), 13(11), p.516 - 519, 2019/10
The estimation of leak flow rates through narrow cracks in structures is of importance for nuclear reactor safety, since the leak flow could be detected before occurrence of loss-of-coolant accidents. The two-phase critical leak flow rates are calculated using the system analysis code, and two representative non-homogeneous critical flow models, Henry-Fauske model and Ransom-Trapp model, are compared. The pressure decrease and vapor generation in the crack, and the leak flow rates are found to be larger for the Henry-Fauske model. It is shown that the leak flow rates are not affected by the structural temperature, but affected largely by the roughness of crack surface.
Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.
Journal of Nuclear Engineering and Radiation Science, 5(3), p.031505_1 - 031505_8, 2019/07
Probabilistic fracture mechanics (PFM) analysis is expected as a rational method for the structural integrity assessment because it can consider the uncertainties of various influence factors and can evaluate the quantitative value such as failure probability of a cracked component as the solution. In the Japan Atomic Energy Agency, a PFM analysis code PASCAL-SP has been developed for the structural integrity assessment of piping welds in nuclear power plants. In the latest few decades, a number of cracks due to primary water stress corrosion cracking (PWSCC) have been detected in the nickel-based alloy welds in the primary piping of pressurized water reactors (PWRs). Thus the structural integrity assessment taking account of PWSCC has become important. In this paper, we improved PASCAL-SP for the assessment considering PWSCC by introducing the several analytical functions such as the evaluation models of crack initiation time, crack growth rate and probability of crack detection. By using improved PASCAL-SP, the failure probabilities of pipes with a circumferential crack or an axial crack due to PWSCC were evaluated as numerical examples. We also evaluated the influence of a leak detection and a non-destructive examination on the failure probabilities. On the basis of the numerical results, we concluded that the improved PASCAL-SP is useful for evaluating the failure probability of pipe taking PWSCC into account.
Mano, Akihiro; Katsuyama, Jinya; Li, Y.
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05
In Japanese nuclear power plants, non-destructive examinations (NDEs) are performed for welds in piping in accordance with the rules such as Rules on Fitness-for-Service for Nuclear Power Plant of the Japan Society of Mechanical Engineers (JSME FFS). A set of NDEs is performed in each 10-year interval, and the extent of examination in each interval is specified in the rules. Welding lines to be examined are selected considering the extent of examination based on two sampling methods. One is the fixed location sampling method that welds to be examined are selected from welds examined in the last interval. The other is the random location sampling method that welds to be examined are selected from other than welds examined in the last interval. The selection of the sampling methods is considered to be one of the important factors in in-service inspection. Probabilistic fracture mechanics (PFM) analysis is expected to be more rational method for the structural integrity assessment because it can consider the uncertainties of various influence factors and evaluate the quantitative values such as failure probability of a cracked component as the solution. In this study, we investigated the influence of the sampling methods related to the NDE on failure probability of typical nuclear piping based on PFM analyses. Through sensitivity PFM analyses, we confirmed that failure probability value obtained from PFM analysis is useful as a quantitative numerical index for selecting the sampling method in an in-service inspection.
Lu, K.; Mano, Akihiro; Katsuyama, Jinya; Li, Y.; Iwamatsu, Fuminori*
Journal of Pressure Vessel Technology, 140(3), p.031201_1 - 031201_11, 2018/06