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Journal Articles

Influence evaluation of loading conditions during pressurized thermal shock transients based on thermal-hydraulics and structural analyses

Katsuyama, Jinya; Uno, Shumpei*; Watanabe, Tadashi*; Li, Y.

Frontiers of Mechanical Engineering, 13(4), p.563 - 570, 2018/12

 Times Cited Count:1 Percentile:100(Engineering, Mechanical)

For the structural integrity assessments on reactor pressure vessels (RPVs) under pressurized thermal shock (PTS) events, thermal hydraulic (TH) behavior of coolant water is one of the most important influence factors. Configuration of plant equipment and their dimensions, and operator action have large influences on TH behavior. In this study, to investigate the influence of operator action on TH behavior during a PTS event, we developed an analysis model for a typical Japanese plant, and performed TH and structural analyses. Two different operator action times were assumed based on the Japanese and US' rules. From the analysis results, it was clarified that differences in operator action times have a significant effect on TH behavior and loading conditions, that is, following the Japanese rule may lead to lower stresses compared to that when following the US rule because earlier operator action caused lower pressure in the RPV.

JAEA Reports

User's manual and analysis methodology of probabilistic fracture mechanics analysis code PASCAL Ver.4 for reactor pressure vessel (Contract research)

Katsuyama, Jinya; Masaki, Koichi; Miyamoto, Yuhei*; Li, Y.

JAEA-Data/Code 2017-015, 229 Pages, 2018/03

JAEA-Data-Code-2017-015.pdf:5.8MB
JAEA-Data-Code-2017-015(errata).pdf:0.15MB

As a part of the structural integrity research for aging light water reactor components, a probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed in JAEA. The PASCAL code can evaluate the conditional failure probabilities and failure frequencies for core region in reactor pressure vessels under the pressurized thermal shock events. In this study, we improved many functions such as the stress intensity factor solutions, the fracture toughness models, or confidence level evaluation function by considering epistemic and aleatory uncertainties related to influence parameters in the structural integrity assessment. We also developed the analysis module PASCAL-Manager which calculates the failure frequency for the entire core region taking into consideration the failure probabilities obtained from PACAL-RV. Based on these improvements, the new analysis code is upgraded to PASCAL Ver.4. This report provides the user's manual and theoretical background of PASCAL Ver.4.

Journal Articles

Verification methodology and results of probabilistic fracture mechanics code PASCAL

Masaki, Koichi; Miyamoto, Yuhei*; Osakabe, Kazuya*; Uno, Shumpei*; Katsuyama, Jinya; Li, Y.

Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 7 Pages, 2017/07

A probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed by Japan Atomic Energy Agency (JAEA). PASCAL can evaluate failure frequencies of Japanese reactor pressure vessels (RPVs) during pressurized thermal shock (PTS) events based on domestic structural integrity assessment models and data of influence factors. In order to improve the engineering applicability of PFM to Japanese RPVs, we have performed verification of the PASCAL. In general, PFM code consists of many functions such as fracture mechanics evaluation functions, probabilistic evaluation functions including random variables sampling modules and probabilistic evaluation models, and so on. The verification of PFM code is basically difficult because it is impossible to confirm such functions through the comparison with experiments. When a PFM code is applied for evaluating failure frequencies of RPVs, verification methodology of the code should be clarified and it is important that verification results including the region and process of the verification of the code are indicated. In this paper, our activities of verification for PASCAL are presented. We firstly represent the overview and methodology of verification of PFM code, and then, some verification examples are provided. Through the verification activities, the applicability of PASCAL in structural integrity assessments for Japanese RPVs was confirmed with great confidence.

JAEA Reports

Guideline on a structural integrity assessment for reactor pressure vessel based on probabilistic fracture mechanics (Contract research)

Katsuyama, Jinya; Osakabe, Kazuya*; Uno, Shumpei; Li, Y.

JAEA-Research 2016-022, 40 Pages, 2017/02

JAEA-Research-2016-022.pdf:4.04MB

For reactor pressure vessels (RPVs) in the light water reactors, the fracture toughness decreases due to the neutron irradiation embrittlement with operating years. In Japan, to prevent RPVs from a nil-ductile fracture, deterministic fracture mechanics methods in accordance with the codes provided by the Japan Electric Association are performed for assessing the structural integrity of RPVs under the pressurized thermal shock (PTS) events by taking the neutron irradiation embrittlement into account. On the other hand, in recent years, probabilistic methodologies for PTS evaluation are introduced into regulations in Europe and the United States. For example, in the United States, a PTS screening criterion related to the reference temperature derived by the probabilistic method is stipulated. If the screening criterion is not satisfied, it is approved to perform the evaluation based on the probabilistic method by calculating numerical index such as through-wall crack frequency (TWCF). To reach the objectives that persons who have knowledge on the fracture mechanics can carry out the PFM analyses and obtain TWCF for a domestic RPVs by referring to this report, we develop the guideline on a structural integrity assessment method based on PFM by reflecting the latest knowledge and expertise.

Journal Articles

Statistical analysis using the Bayesian nonparametric method for irradiation embrittlement of reactor pressure vessels

Takamizawa, Hisashi; Ito, Hiroto; Nishiyama, Yutaka

Journal of Nuclear Materials, 479, p.533 - 541, 2016/10

 Times Cited Count:0 Percentile:100(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.

Journal Articles

Bayesian statistical analysis on chemical composition contributing to irradiation embrittlement at high fluence region

Takamizawa, Hisashi; Nishiyama, Yutaka

Proceedings of 2016 ASME Pressure Vessels and Piping Conference (PVP 2016) (Internet), 5 Pages, 2016/07

It has been accepted that neutron irradiation embrittlement of reactor pressure vessel is caused by irradiation-induced formation of solute clusters (SCs) and matrix damages (MDs). In the present study, to analyze the contribution of chemical composition contained in SCs to irradiation embrittlement at high fluence region, statistical analysis using the Bayesian nonparametric (BNP) method was performed for Japanese PWR surveillance data. The significance of P, Si and Mn contents, which are not necessarily included in embrittlement correlations unlike the Cu and Ni content, was evaluated. The BNP method can learn the complexity of the statistical model itself from the input data and infer the predicted data with individual probability distribution of predict condition. The result suggested that irradiation embrittlement was most affected by the Si content at high fluence region.

Journal Articles

Development of failure evaluation method for BWR lower head in severe accident, 2; Applicability evaluation of the FEM using uni-axial material data for multi-axial deformation analysis

Nemoto, Yoshiyuki; Kato, Hitoshi; Kaji, Yoshiyuki; Yoshida, Hiroyuki

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05

For the evaluation of reactor pressure vessel (RPV) lower head rupture probably occurred during the severe accident in Fukushima Daiichi Nuclear Power Plants, JAEA is conducting the thermal-hydraulics / mechanical coupling analysis. In the mechanical analysis based on the finite element method (FEM), material property data previously obtained from uni-axial material tests are applied. The lower head of BWR such as Fukushima NPP, has complicated structure compared to PWR, with control rod guide tubes, stub tubes, etc., therefore the mechanical analyses need to treat multi-axial deformation of the materials. To perform such mechanical analysis, the applicability of the analytical model using uni-axial data for multi-axial deformation analysis must be validated. In this study, the internal pressure creep tests were performed because which can realize the multi-axial deformation condition. In addition, mechanical analyses were conducted and the analytical results were compared with the experimental data.

Journal Articles

New reactor cavity cooling system using novel shape for HTGRs and VHTRs

Takamatsu, Kuniyoshi; Hu, R.*

Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 12 Pages, 2014/12

continuous closed regions; one is an ex-reactor pressure vessel (RPV) region and another is a cooling region having heat transfer area to ambient air assumed at 40 ($$^{circ}$$C). The RCCS uses novel shape so that the heat released from the RPV can be removed efficiently with radiation and natural convection. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design greatly reduces the possibility of losing the heat sink for decay heat removal. Therefore, HTGRs and VHTRs adopting the new RCCS design can avoid core melting owing to overheating the fuels.

Journal Articles

Research and development on passive cooling system

Takada, Shoji

Nuclear Engineering and Design, 233(1-3), p.185 - 195, 2004/10

 Times Cited Count:5 Percentile:60.33(Nuclear Science & Technology)

Experiments are carried out to investigate the effects of the natural convection of superheated gas as well as of the stand pipes on the temperature distributions of the components and the heat removal performance in the water-cooling panel system for the MHTGR for decay heat removal, and to verify the design and evaluation methods. The numerical results of the code THANPACST2 are compared with the experimental data to verify the numerical methods and axi-symmetric model proposed, which can simulate the three-dimensional configuration of the stand pipes on the upper head of the pressure vessel by using porous body cells. The experiments revealed that temperatures increased with elevation on the upper head, because the stand pipes restrict radiation heat transfer to the upper cooling panel and reduce the heat transfer area on the upper head which was superheated by natural convection of helium gas in the pressure vessel. The numerical methods were able to closely duplicate the pattern of the rising temperature profile with elevation around the top of the upper head.

Journal Articles

Grain boundary embrittlement of light water reactor pressure vessel steels

Nishiyama, Yutaka; Suzuki, Masahide

Kinzoku, 73(8), p.48 - 52, 2003/08

no abstracts in English

Journal Articles

Development of PFM code with a function of ductile crack extension analysis

Li, Y.*; Kato, Daisuke*; Shibata, Katsuyuki; Onizawa, Kunio

Nippon Kikai Gakkai Rombunshu, A, 69(678), p.239 - 245, 2003/02

no abstracts in English

Journal Articles

Design study on Gas Turbine High Temperature Reactor (GTHTR300)

Kunitomi, Kazuhiko; Katanishi, Shoji; Takada, Shoji; Takizuka, Takakazu; Nakata, Tetsuo; Yan, X.; Takei, Masanobu; Kosugiyama, Shinichi; Shiozawa, Shusaku

Nippon Genshiryoku Gakkai Wabun Rombunshi, 1(4), p.352 - 360, 2002/12

no abstracts in English

Journal Articles

Irradiation behavior of low-copper reactor pressure vessel steels

Suzuki, Masahide; Nishiyama, Yutaka

Kinzoku, 71(8), p.42 - 45, 2001/08

no abstracts in English

Journal Articles

Development of a diagnostic technique using ultrasonic wave for evaluation of irradiation embrittlement in reactor pressure vessel materials

Ishii, Toshimitsu; Ooka, Norikazu; Niimi, Motoji; Kobayashi, Hideo*

Kinzoku, 71(8), p.20 - 24, 2001/08

no abstracts in English

Journal Articles

Comparison of transition temperature shifts between static fracture toughness and Charpy-V impact properties due to irradiation and post-irradiation annealing for Japanese A533B-1 steels

Onizawa, Kunio; Suzuki, Masahide

Effects of Radiation on Materials: 20th International Symposium (ASTM STP 1405), p.79 - 96, 2001/07

no abstracts in English

Journal Articles

Aging degradation of light water reactor materials; Reactor internal and pressure vessel materials

Tsukada, Takashi; Ebine, Noriya

Nippon AEM Gakkai-Shi, 9(2), p.171 - 177, 2001/06

no abstracts in English

Journal Articles

Development of a RPV reliability analysis code based on probabilistic fracture mechanics methodology

Shibata, Katsuyuki; Kato, Daisuke*; Li, Y.*

Nippon Genshiryoku Gakkai-Shi, 43(4), p.387 - 396, 2001/04

 Times Cited Count:2 Percentile:77.82(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Evaluation of the magnetic and mechanical properties of reactor pressure vessel steels by incremental permeability change curve measurements

Ebine, Noriya; Suzuki, Masahide

Nippon Oyo Jiki Gakkai-Shi, 25(4-2), p.1051 - 1054, 2001/04

no abstracts in English

Journal Articles

Evaluation of nuclear reactor pressure vessel integrity by magnetic interrogation and problems of relevant magnetic field analysis

Ara, Katsuyuki*; Ebine, Noriya

Denki Gakkai Magunetikkusu Kenkyukai Shiryo (MAG-01-55), p.1 - 6, 2001/03

no abstracts in English

164 (Records 1-20 displayed on this page)