Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 23

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

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.

Journal Articles

Study on application of PFM analysis method to Japanese code for RPV integrity assessment under PTS events

Osakabe, Kazuya*; Masaki, Koichi*; Katsuyama, Jinya; Katsumata, Genshichiro; Onizawa, Kunio; Yoshimura, Shinobu*

Proceedings of 2015 ASME Pressure Vessels and Piping Conference (PVP 2015) (Internet), 8 Pages, 2015/07

A probabilistic fracture mechanics (PFM) analysis method for pressure boundary components is useful to evaluate the structural integrity in a quantitative way. This is because the uncertainties related to influence parameters can be rationally incorporated in PFM analysis. From this viewpoint, the probabilistic approach evaluating through-wall cracking frequencies (TWCFs) of reactor pressure vessels (RPVs) has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. As a study of applying PFM analysis to the integrity assessment of domestic RPVs, JAEA has been preparing input data and analysis models to calculate TWCFs using PFM analysis code PASCAL3. In this paper, activities have been introduced such as preparing input data and models for domestic RPVs, verification of PASCAL3, and formulating guideline on general procedures of PFM analysis for the purpose of utilizing PASCAL3. In addition, TWCFs for a model RPV evaluated by PASCAL3 are presented.

Journal Articles

Estimation of through-wall cracking frequency of RPV under PTS events using PFM analysis method for identifying conservatism included in current Japanese code

Osakabe, Kazuya*; Masaki, Koichi*; Katsuyama, Jinya; Katsumata, Genshichiro; Onizawa, Kunio

Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 7 Pages, 2014/07

The structural integrity of reactor pressure vessel (RPV) during pressurized thermal shock events is judged to be maintained unless the stress intensity factors at the crack tip is smaller than fracture toughness $$K$$$$_{Ic}$$ based on deterministic approach in the current Japanese code. Application of a probabilistic fracture mechanics (PFM) analysis method for the structural reliability assessment of RPVs has become attractive recently, because uncertainties of several parameters can be incorporated rationally. According to the PFM analysis method in the U.S., through-wall cracking frequencies (TWCFs) are estimated. In this study, in order to identify the conservatism in the current code, PFM analyses on TWCF have been performed for certain model of RPVs. The result shows that the current assumption in JEAC 4206-2007 is conservative as compared with realistic conditions. Effects of variation of PTS transients on crack initiation frequency and TWCF have been also discussed.

Journal Articles

Study of system safety evaluation on LTO of national project; Structural integrity assessment of reactor pressure vessels

Onizawa, Kunio; Masaki, Koichi; Osakabe, Kazuya*; Nishikawa, Hiroyuki*; Katsuyama, Jinya; Nishiyama, Yutaka

Nippon Hozen Gakkai Dai-9-Kai Gakujutsu Koenkai Yoshishu, p.374 - 379, 2012/07

To assure the structural integrity of a reactor pressure vessel (RPV) is known as one of the critical issues to maintain the safe long-term operation of a nuclear power plant. In Japan, the assessment methods for RPV integrity, stipulated in the codes and standards, have been endorsed by the regulatory body. Authors have initiated extensive research on the improvement of structural integrity assessment methods of RPVs. In this paper, we describe some research results obtained from the first-year activity. These include the study on revisiting the technical background of the methods, such as loading conditions, postulated crack definition, the other evaluation methods. In addition, studies on probabilistic methods for the applicability to the current rules and the standardization of the probabilistic analysis methods have been presented.

JAEA Reports

User's manual and analysis methodology of probabilistic fracture mechanics analysis code PASCAL3 for reactor pressure vessel (Contract research)

Masaki, Koichi; Nishikawa, Hiroyuki*; Osakabe, Kazuya*; Onizawa, Kunio

JAEA-Data/Code 2010-033, 350 Pages, 2011/03

JAEA-Data-Code-2010-033.pdf:5.32MB

The probabilistic fracture mechanics (PFM) analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed in JAEA. The PASCAL code evaluates the conditional probabilities of crack initiation and fracture of a reactor pressure vessel (RPV) under transient conditions such as pressurized thermal shock (PTS). Previous version of PASCAL (PASCAL ver.2) that was released in 2007 has many functions including the evaluation method for an embedded crack and conditional probabilities of crack initiation and fracture of a RPV, PTS transient database, inspection crack detection probability model and others. A generalized analysis method is available on the basis of the development of PASCAL ver.3 and sensitivity analysis results. Graphical user interface (GUI) including a generalized method and some functions of PFM have been also updated for PASCAL3. This report provides the user's manual, examples of analysis and theoretical background of PASCAL ver.3.

Journal Articles

29th report of ITPA topical group meeting

Isayama, Akihiko; Sakakibara, Satoru*; Furukawa, Masaru*; Matsunaga, Go; Yamazaki, Kozo*; Watanabe, Kiyomasa*; Idomura, Yasuhiro; Sakamoto, Yoshiteru; Tanaka, Kenji*; Tamura, Naoki*; et al.

Purazuma, Kaku Yugo Gakkai-Shi, 86(6), p.374 - 377, 2010/06

no abstracts in English

Journal Articles

27th report of ITPA topical group meeting

Osakabe, Masaki*; Shinohara, Koji; Toi, Kazuo*; Todo, Yasushi*; Hamamatsu, Kiyotaka; Murakami, Sadayoshi*; Yamamoto, Satoshi*; Idomura, Yasuhiro; Sakamoto, Yoshiteru; Tanaka, Kenji*; et al.

Purazuma, Kaku Yugo Gakkai-Shi, 85(12), p.839 - 842, 2009/12

no abstracts in English

Journal Articles

Dynamics of ion internal transport barrier in LHD heliotron and JT-60U tokamak plasmas

Ida, Katsumi*; Sakamoto, Yoshiteru; Yoshinuma, Mikiro*; Takenaga, Hidenobu; Nagaoka, Kenichi*; Hayashi, Nobuhiko; Oyama, Naoyuki; Osakabe, Masaki*; Yokoyama, Masayuki*; Funaba, Hisamichi*; et al.

Nuclear Fusion, 49(9), p.095024_1 - 095024_9, 2009/09

 Times Cited Count:26 Percentile:26.42(Physics, Fluids & Plasmas)

Dynamics of ion internal transport barrier (ITB) formation and impurity transport both in the Large Helical Device (LHD) heliotron and JT-60U tokamak are described. Significant differences between heliotron and tokamak plasmas are observed. The location of the ITB moves outward during the ITB formation regardless of the sign of magnetic shear in JT-60U and the ITB becomes more localized in the plasma with negative magnetic shear. In LHD, the low Te/Ti ratio ($$<$$ 1) of the target plasma for the high power heating is found to be necessary condition to achieve the ITB plasma and the ITB location tends to expand outward or inward depending on the condition of the target plasmas. Associated with the formation of ITB, the carbon density tends to be peaked due to inward convection in JT-60U, while the carbon density becomes hollow due to outward convection in LHD. The outward convection observed in LHD contradicts the prediction by neoclassical theory.

Journal Articles

Report on ITPA meetings, 24

Idomura, Yasuhiro; Yoshida, Maiko; Yagi, Masatoshi*; Tanaka, Kenji*; Hayashi, Nobuhiko; Sakamoto, Yoshiteru; Tamura, Naoki*; Oyama, Naoyuki; Urano, Hajime; Aiba, Nobuyuki; et al.

Purazuma, Kaku Yugo Gakkai-Shi, 84(12), p.952 - 955, 2008/12

no abstracts in English

Journal Articles

Spectroscopic observations of beam and source plasma light and testing Cs-deposition monitor in the large area negative ion source for LHD-NBI

Oka, Yoshihide*; Tsumori, Katsuyoshi*; Ikeda, Katsunori*; Kaneko, Osamu*; Nagaoka, Kenichi*; Osakabe, Masaki*; Takeiri, Yasuhiko*; Asano, Eiji*; Komada, Seiji*; Kondo, Tomoki*; et al.

Review of Scientific Instruments, 79(2), p.02C105_1 - 02C105_4, 2008/02

 Times Cited Count:0 Percentile:100(Instruments & Instrumentation)

In the present studies, we studied the cesium lines in the source plasma during beam shots on the LND MN-NBI system. It was found for the first time in the LHD-source 2, that both the amount of Cs I (neutral Cs) and Cs II (Cs$$^{+}$$) in the source plasma light rose sharply when beam acceleration began, and continued rising during a 10 s pulse. We think that this was because the cesium was evaporated/sputtered from the source backplate by the back-streaming positive ions.

Journal Articles

Energetic ion measurements using a directional probe

Nagaoka, Kenichi*; Isobe, Mitsutaka*; Shinohara, Koji; Osakabe, Masaki*; Shimizu, Akihiro*; Okamura, Shoichi*

Plasma and Fusion Research (Internet), 1(1), p.005_1 - 005_2, 2006/01

A directional Langmuir probe (DLP) method has been applied for energetic particles (ions) measurement in a magnetically confined plasma. Two experimental demonstrations have been performed in the compact helical system (CHS). One is neutral beam modulation experiment and the other is the measurement of energetic ion loss induced by MHD bursts. The results of the DLP are consistent with that of a neutral particle analyzer (NPA) and a lost ion probe (LIP). These experiments show that this method is applicable outside and also inside of the last closed flux surface.

Journal Articles

Positron lifetime measurement on centrifuged Bi$$_{3}$$Pb$$_{7}$$ intermetallic compound

Ono, Masao; Huang, X. S.*; Shibata, Yasuhiro*; Iguchi, Yusuke*; Sakai, Seiji; Maekawa, Masaki; Chen, Z. Q.*; Osakabe, Toyotaka; Kawasuso, Atsuo; Naramoto, Hiroshi*; et al.

Proceedings of 1st International Conference on Diffusion in Solids and Liquids (DSL 2005), p.531 - 533, 2005/07

Recently, we formed atomic-scale graded structures in some miscible alloys and observed the decomposition in Bi$$_{3}$$Pb$$_{7}$$ intermetallic compound by sedimentation of atoms under strong gravitational field. In this study, we measured positron lifetime of centrifuged Bi$$_{3}$$Pb$$_{7}$$, to which the composition change was very small as it was treated at low temperature. It was found that the positron lifetime became longer than that of starting state. This indicated that the point defects (vacancy or divacancy) increased in the sample by centrifugal treatment. We are now investigating the relationship between increase in point defects and sedimentation of atoms.

Journal Articles

Neutron diagnostics for the energetic ion transport analysis

Nishitani, Takeo; Osakabe, Masaki*; Shinohara, Koji; Ishikawa, Masao

Purazuma, Kaku Yugo Gakkai-Shi, 80(10), p.860 - 869, 2004/10

no abstracts in English

Journal Articles

Heating and current drive by N-NBI in JT-60U and LHD

Kaneko, Osamu*; Yamamoto, Takumi; Akiba, Masato; Hanada, Masaya; Ikeda, Katsunori*; Inoue, Takashi; Nagaoka, Kenichi*; Oka, Yoshihide*; Osakabe, Masaki*; Takeiri, Yasuhiko*; et al.

Fusion Science and Technology, 44(2), p.503 - 507, 2003/09

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

High energy negative-ion-based neutral beam injection (N-NBI) is expected as an efficient and reliable tool of heating and current driving for reactor plasmas such as ITER. A world wide activity on developing technology of negative ion production and beam formation started in 1980ユs and the great progress has been achieved up to now. In particular, Japan has two large projects that planned adopting N-NBI for real plasma experiments; the JT-60U tokamak and the LHD heliotron, which further motivated the R&D activity. These R&D programs were carried out at JAERI and NIFS separately in Japan, and both were successfully done. The first beam injection experiment was made on the JT-60U in 1996, followed by the LHD in 1998. They were the first experiments on heating plasma by high energy beam in tokamaks and in stellerators, and the obtained results were very promising.

Oral presentation

Technical review of structural integrity assessment procedure on reactor pressure vessel, 1; Comparison of procedures in Japan and U.S. and study on the postulated flaw

Masaki, Koichi; Nishikawa, Hiroyuki*; Osakabe, Kazuya*; Katsuyama, Jinya; Onizawa, Kunio

no journal, , 

The structural integrity of a reactor pressure vessel (RPV) should be maintained to ensure the safe long-term operation of a nuclear power plant. The assessment methods for RPV integrity are stipulated in the codes and standards. We compared the assessment methods for RPV integrity under pressurized thermal shock events in Japan and the United States. After reviewing the technical background of the assessment method, we clarified the technical issues based on the latest knowledge in this field. We also identified based on sensitivity analysis that a postulated crack which is provided in the domestic code was more conservative than that of the United States.

Oral presentation

Technical review of structural integrity assessment procedure on reactor pressure vessel, 2; Effects of transient loads on brittle fracture initiation

Masaki, Koichi; Osakabe, Kazuya*; Nishikawa, Hiroyuki*; Katsuyama, Jinya; Onizawa, Kunio

no journal, , 

The structural integrity of a reactor pressure vessel (RPV) should be maintained to ensure the safe long-term operation of a nuclear power plant. Probabilistic fracture mechanics (PFM) code is required to assess the structural integrity of a RPV during pressurized thermal shock (PTS) events quantitatively. The PASCAL3 for probabilistic analysis have been developed in JAEA. In PTS reevaluation project in the United States, PFM analysis using FAVOR has been conducted to obtain screening criteria for PTS. To consider the future application of the probabilistic method to the code in Japan, we have conducted deterministic and probabilistic analyses using PASCAL3 and evaluated the effects of transient loads on brittle crack initiation and fracture. The results showed that the structural integrity of a RPV could be assessed quantitatively through the evaluation of fracture probabilities.

Oral presentation

Study of system safety evaluation on LTO of National project structural integrity assessment of reactor pressure vessels

Katsumata, Genshichiro; Masaki, Koichi*; Osakabe, Kazuya*; Nishikawa, Hiroyuki*; Katsuyama, Jinya; Nishiyama, Yutaka; Onizawa, Kunio

no journal, , 

To assure the structural integrity of a reactor pressure vessel (RPV) is known as one of the critical issues to maintain the safe long-term operation of a nuclear power plant. In Japan, the assessment methods for RPV integrity, stipulated in the codes and standards, have been endorsed by the regulatory body. Authors have initiated extensive research on the improvement of structural integrity assessment methods of RPVs. In this paper, we describe some research results obtained from the second-year activity. These include the study on revisiting the technical background of the methods, such as loading conditions, postulated crack definition, the other evaluation methods. In addition, studies on probabilistic methods for the applicability to the current rules and the standardization of the probabilistic analysis methods have been presented.

Oral presentation

Evaluation of through-wall cracking frequency of reactor pressure vessel under pressurized thermal shock events using probabilistic fracture mechanics analysis method

Nishikawa, Hiroyuki*; Masaki, Koichi*; Osakabe, Kazuya*; Katsuyama, Jinya; Onizawa, Kunio

no journal, , 

To assure the structural integrity of a reactor pressure vessel (RPV) is one of the most critical issues to maintain the safe long-term operation of a nuclear power plant. In Japan, the assessment methods for RPV integrity using deterministic fracture mechanics, provided in the codes and standards, have been endorsed by the regulatory body. Meanwhile, a regulation on the fracture toughness requirements against PTS events with a probabilistic approach has been established in the U.S. In this paper, referring to the approach in the U.S., the through-wall cracking frequency (TWCF) for aging RPV in Japan against PTS events was calculated using a probabilistic fracture mechanics analysis code PASCAL3. As the results, we clarified that existing methodology on structural integrity assessment of RPV considering postulated axial flaw has a certain amount of conservativeness, and the value of TWCF is useful to assess impacts on structural integrity of RPV quantitatively.

Oral presentation

Crack growth analysis for stainless steel piping weld considering material boundary

Masaki, Koichi*; Osakabe, Kazuya*; Katsuyama, Jinya; Kikuchi, Masanori*

no journal, , 

S-version finite element method (S-FEM) is an evaluation method for fully automatic crack growth simulation system on the basis of auto-mesh technique. S-FEM has been applied to simulate the crack growth considering the material boundary. Crack growth analyses have been conducted for a plate consisted of two different materials by tensile membrane stress and a stainless steel piping under weld residual stress field. Effects of the crack growth rates in the different materials on the behavior of the crack growth have been evaluated by S-FEM and the influence function method. Results show that the crack sizes obtained from each method are in a good agreement.

Oral presentation

Effects of transient type and flaw density on through-wall cracking frequency of reactor pressure vessel under pressurized thermal shock events

Masaki, Koichi*; Osakabe, Kazuya*; Katsuyama, Jinya; Katsumata, Genshichiro; Onizawa, Kunio

no journal, , 

To assure the structural integrity of a reactor pressure vessel (RPV) is one of the most critical issues to maintain the safe long-term operation of a nuclear power plant. In Japan, the assessment methods for RPV integrity using deterministic fracture mechanics are provided in Japan Electric Association Code (JEAC). Meanwhile, a regulation on the fracture toughness requirements against PTS events based on a probabilistic fracture mechanics (PFM) analysis has been established in the U.S. In this paper, in order to apply probabilistic approach to domestic regulation, sensitivity analyses for flaw density or transient by reference to the data in the U.S. were performed using a PFM analysis code PASCAL3. We evaluated the effect of the flaw density or transient on through-wall cracking frequency (TWCF) and showed the specific example as a practical use of PFM.

23 (Records 1-20 displayed on this page)