Sensitivity Analysis of Crack Shape on Screening Parameter in JSME FFS Rules for Nuclear Power Plants

2017 ◽  
Author(s):  
Naoki Miura ◽  
Kiminobu Hojo
Keyword(s):  
Sensitivity Analysis ◽  
Fracture Mechanics ◽  
Surface Crack ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Limit Load ◽  
Evaluation Procedure ◽  
Crack Shape ◽  
Screening Parameter

In the 2012 version of the JSME Fitness-for-Service Rules for Nuclear Power Plants, the procedure to calculate screening parameter, SC, which is used for selecting the analysis method (limit load controlled by plastic collapse, elastic-plastic fracture mechanics, or linear elastic fracture mechanics), has been revised to reflect a semi-elliptical surface crack. Both limit load solution and stress intensity factor solution are needed to calculate SC, and the solutions for a semi-elliptical surface crack are different from those for a fan-shape surface crack. In this study, the effect of the difference in crack shape on SC is investigated. Through the results on the sensitivity analysis, the adequacy of the evaluation procedure of SC is ascertained.

Safety Assessment of Disposal Container for Higher Activity Low Level Waste

2010 ◽  
Author(s):  
Motonori Nakagami ◽  
Seiji Komatsuki ◽  
Kyosuke Fujisawa ◽  
Takashi Nishio ◽  
Thomas Quercetti ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fracture Strain ◽  
The Body ◽  
Unstable Fracture ◽  
Maximum Weight ◽  
Drop Tests

As one of the studies on “yoyushindo disposal” whose concept is similar to an intermediate disposal, the development of a disposal container has been conducted by the Federation of Electric Power Companies of Japan. To assess a drop event of a waste package in which stored the radioactive wastes from nuclear power plants, the toughness of the disposal container was evaluated by drop tests using three specimens which have actual dimensions, drop analysis, fracture mechanics assessment and macroscopic tests. The three specimens for drop tests were manufactured in consideration of the design specifications and the manufacture operations in nuclear power plants. The lid plates of the specimens were welded to the body plates without pre- and post-weld heat treatment by using a remote automated welding machine. The drop tests showed that no penetration cracks or splash of its content occurred in the disposal container under conservative conditions such as the maximum weight and height in the handling. Drop analysis and the fracture mechanics assessment indicate that the strain induced by the drop impact did not exceed the fracture strain and an unstable fracture did not occur. And macroscopic tests showed that penetration cracks did not occur at 8m drop events. These tests and evaluations confirmed that the disposal container had sufficient toughness.

Sensitivity Analysis of Major Parameters on Piping Failure Probabilities

2004 ◽  
Vol 261-263 ◽  
pp. 821-826
Author(s):  
Sung Gyu Jung ◽  
Chang Soon Lee ◽  
In Gyu Park ◽  
Se Hwan Lee ◽  
Tae Eun Jin
Keyword(s):  
Sensitivity Analysis ◽  
Size Distribution ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Crack Size ◽  
Deterministic Approach ◽  
Inspection Interval ◽  
Different Types ◽  
Failure Probabilities

In-service inspections (ISI) of pipes in the nuclear power plants are currently performed based on mandated requirements in the ASME Section XI, which is based on deterministic approach of the critical welds. The 20 years of ISI experience in U.S.A. has revealed less correlation between the critical welds and actual failures, and much conservatism in current ISI requirements. To reduce those problems, risk-informed ISI technology has been developed and proved to be useful. This paper presented a method for predicting piping failure probabilities in an application of risk-informed ISI, and analyzed the effect of input parameters on piping failure probabilities. Results generated using this approach revealed that the calculated failure probabilities can be sensitive to the different types of stressors, crack size distribution, inspection interval, etc..

Benchmark Analysis on Probabilistic Fracture Mechanics Analysis Codes Considering Multiple Cracks and Crack Initiation in Aged Piping of Nuclear Power Plants

2014 ◽  
Author(s):  
Yinsheng Li ◽  
Kazuya Osakabe ◽  
Genshichiro Katsumata ◽  
Jinya Katsuyama ◽  
Kunio Onizawa ◽  
...  
Keyword(s):  
Risk Assessment ◽  
Fracture Mechanics ◽  
Crack Initiation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Multiple Cracks ◽  
Benchmark Analysis ◽  
Probabilistic Fracture Mechanics

In recent years, cracks have been detected in piping systems of nuclear power plants. Many of them are multiple cracks in the same welded joints. Therefore, structural integrity evaluation and risk assessment considering multiple cracks and crack initiation in aged piping have become increasingly important. Probabilistic fracture mechanics (PFM) is a rational methodology in structural integrity evaluation and risk assessment of aged piping in nuclear power plants. Two PFM codes, PASCAL-SP and PRAISE-JNES, have been improved or developed in Japan for the structural integrity evaluation and risk assessment considering the age related degradation mechanisms of pipes. Although the purposes to develop these two codes are different, both have almost the same basic functions to obtain the failure probabilities of pipes. In this paper, a benchmark analysis was conducted considering multiple cracks and crack initiation, in order to confirm their reliability and applicability. Based on the numerical investigation in consideration of important influence factors such as crack number, crack location, crack distribution and crack detection probability of in-service inspection, it was concluded that the analysis results of these two codes are in good agreement.

Response Analysis of a Seismic Isolated Structure Considering a Probabilistic Approach

2014 ◽  
Author(s):  
Naoki Akamatsu ◽  
Satoshi Fujita ◽  
Keisuke Minagawa
Keyword(s):  
Sensitivity Analysis ◽  
Nuclear Power ◽  
Seismic Isolation ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Large Scale ◽  
Probabilistic Approach ◽  
Nonlinear Behavior ◽  
Response Analysis ◽  
Isolation Systems

Japan is one of the most advanced countries in earthquake technology. Isolation systems are widely used in large-scale structures such as hospitals and communication centers. For example, an isolated office building has been used as a hub of recovery from accident by Great East Japan Earthquake in Fukushima nuclear power plant. In the meantime, application of probabilistic risk assessment is used for structure of nuclear power plants. In 2006, Regulatory Guide for Reviewing Seismic Design was revised and according to guideline, it is necessary to consider the residual risk1. In addition, seismic isolation systems are expected to be used for nuclear power plants. Recently, the risk of isolation system’s failure needs to be assessed in case of large ground motion. This paper deals with probabilistic approach on seismic response of an isolated structure. Consequently, sensitivity analysis is carried out. Then, as nonlinear behavior in rubber bearings occurs during huge earthquake, it has to be considered in the sensitivity analysis.

Allowable Flaw Size of Japanese Cast Stainless Steel Pipe Using Probabilistic Fracture Mechanics Method

2017 ◽  
Author(s):  
Shotaro Hayashi ◽  
Mayumi Ochi ◽  
Kiminobu Hojo ◽  
Takahisa Yamane ◽  
Wataru Nishi
Keyword(s):  
Stainless Steel ◽  
Fracture Mechanics ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Thermal Ageing ◽  
Probabilistic Fracture Mechanics ◽  
Stainless Steel Pipe ◽  
Technical Basis ◽  
Cast Stainless Steel

The cast austenitic stainless steel (CASS) that is used for the primary loop pipes of nuclear power plants is susceptible to thermal ageing during plant operation. The Japanese JSME rules on fitness-for-service (JSME rules on FFS)[1] for nuclear power plants specify the allowable flaw depths. However, some of these allowable flaw sizes are small compared with the smallest flaw sizes, which can be detected by nondestructive testing. ASME Section XI Code Case N-838[2] recently specified the maximum tolerable flaw depths for CASS pipes determined by probabilistic fracture mechanics (PFM). In a similar way, the allowable flaw depths of CASS pipes were calculated by PFM analysis code “PREFACE”[3] which considers uncertainty of the mechanical properties of Japanese PWR CASS materials. In order to confirm the validity of PREFACE, the allowable flaw depths calculated by PREFACE were compared with the maximum tolerable flaw depths in the technical basis of Code Case N-838. As a result, although the J calculation method and the embrittlement prediction model of CASS are different, these were qualitatively consistent. In addition, the sensitivity of ferrite content to the allowable flaw depths was investigated.

Development of an Integrity Evaluation System for Steam Generator Tubes in a Nuclear Power Plant

2007 ◽  
Vol 120 ◽  
pp. 157-162
Author(s):  
J.C. Kim ◽  
Sang Min Lee ◽  
Yoon Suk Chang ◽  
Jae Boong Choi ◽  
Young Jin Kim ◽  
...  
Keyword(s):  
Steam Generator ◽  
Evaluation System ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Limit Load ◽  
Evaluation Procedure ◽  
Efficient System ◽  
Reference Pressure ◽  
Steam Generator Tubes

Steam generators working in nuclear power plants convert water into steam from heat produced in the reactor core and each of them contains from 3,000 to 16,000 tubes. Since these tubes constitute one of primary barriers under radioactive and high pressure condition, the integrity should be maintained carefully during the operation. The objective of this research is to introduce an integrity evaluation system for steam generator tubes as a substitute of well-trained engineers or experts. For this purpose, a couplet examination has been carried out on the complicated evaluation procedure and an efficient system named as STiES was developed employing three representative integrity evaluation methods: fracture mechanics analysis (crack driving force diagram and J-integral/Tearing modulus method) and limit load method. Exemplary analyses for steam generator tubes with various types of flaws showed good applicability of the proposed integrity evaluation system. So, it is anticipated that the system can be used for the calculation of reference pressure to decide either the continued operation or repair until next outage.

Initiation and Growth of Small Fatigue Cracks of Steels Used for Nuclear Power Plants Under Low Cycle Regime

2014 ◽  
Author(s):  
Shota Hasunuma ◽  
Takeshi Ogawa
Keyword(s):  
Fatigue Crack ◽  
Fracture Mechanics ◽  
Crack Initiation ◽  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Low Cycle Fatigue ◽  
Fatigue Cracks ◽  
Cycle Fatigue

Low cycle fatigue tests were conducted for carbon steel, STS410, low alloy steel, SFVQ1A, and austenitic stainless steel, SUS316NG, which were used for nuclear power plants, in order to investigate the mechanism of fatigue damage when the plants were subjected to huge seismic loads. In these tests, the surface behavior of fatigue crack initiation and growth was observed in detail using cellulose acetate replicas, while the interior behavior was detected in terms of fracture surface morphology developed by multiple two-step strain amplitude variations with periodical surface removals. Fatigue crack growth rates were evaluated by elasto-plastic fracture mechanics approach. For SFVQ1A and SUS316NG, the fracture mechanics approach is available in order to predict the crack growth life from the metallurgical crack initiation size to the final crack length of the specimens. For STS410, numerous small cracks initiated, grew and coalesced each other on the specimen surface under low cycle fatigue regime.

Development of Load Multipliers (Z-Factors) in Elastic-Plastic Fracture Mechanics Evaluation in Rules on Fitness-for-Service for Nuclear Power Plants

2012 ◽  
Author(s):  
Seiji Asada ◽  
Masao Itatani ◽  
Naoki Miura ◽  
Hideo Machida
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Fracture Mechanics ◽  
Ferritic Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Elastic Plastic ◽  
Surface Flaws ◽  
Steel Piping

Not only nonmandatory Appendix C, “Evaluation of Flaws in Piping,” in ASME Boiler & Pressure Vessel Code Section XI but also Appendix E-9, “Elastic-Plastic Fracture Mechanics Evaluation,” in the JSME Rules on Fitness-for-Service for Nuclear Power Plants use the load multiplier Z-factor that is applied to elastic-plastic fracture mechanics evaluation for a circumferential flaw of austenitic stainless steel piping and ferritic steel piping. The Z-factor is defined as the ratio of the limit load to the load at fracture load. Basically, the Z-factor equations were conservatively formulated by using the Z-factors for circumferential through-wall flaws. However, the Codes require flaw evaluation for circumferential surface flaws. Accordingly, Z-factors for circumferential surface flaws should be developed to have the consistency. Therefore Z-factor equations of austenitic stainless steel piping and ferritic steel piping have been developed for circumferential surface flaws.

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