Katsuyuki Shibata

Development of a PFM code for evaluating reliability of pressure components subject to transient loading

In order to improve the PFM methodology and to evaluate the reliability and integrity of an aged RPV, the PFM code PASCAL is being developed. This code evaluates the conditional probability of crack initiation and fracture of pressure components subject to a transient loading based on Monte Carlo simulation. In addition to the common functions established in existing codes, the code has some original functions and features in the elasto-plastic fracture criterion based R6 method, the simulation models for the semi-elliptical crack extension and the effect of thermal annealing, improvement in Monte Carlo simulation and so on. This paper describes the main features of the code, the results of verification analysis and case studies on influence parameters by using above functions. The verification analysis and case studies are carried based on NRC/EPRI PTS benchmark problem. The basic performance of the code was verified by comparing the results with those by existing codes. From the result of case studies, the effectiveness and performance of main functions are examined and the influence of some parameters, such as fracture criterion, WPS, semi-elliptical crack extension models, existence of overlay clad, initial aspect ratio on the failure probability are also discussed.

Results of reliability test program on light water reactor piping

Abstract The Japan Atomic Energy Research Institute has conducted a piping reliability test program to demonstrate the safety and reliability of light water reactor primary piping. In this program, pipe fatigue test, leak-before-break (LBB) verification test and pipe rupture test were carried out to examine the integrity of piping, to verify the LBB and to demonstrate the effectiveness of protective measures against jet impingement and pipe whip loads under a pipe rupture event. In the pipe fatigue test, a procedure to predict the fatigue crack growth was developed, and the integrity of piping during the plant service life was evaluated. In the LBB verification test, the pipe fracture test and the leak rate test were performed to verify the LBB in the primary piping. In the pipe rupture test, the influence of jet impingement on the target disk and the deformation behavior of whipping pipe and restraint were investigated. Using the test results, the jet impingement behavior and the effectiveness of pipe whip restraint were demonstrated.

Probabilistic fracture mechanics analysis of nuclear structural components: a review of recent Japanese activities

This paper describes a review of recent Japanese activities on probabilistic fracture mechanics (PFM) analyses. Japan Atomic Energy Research Institute (JAERI) has sponsored research committees on PFM organized by Japan Society of Mechanical Engineers (JSME) and Japan Welding Engineering Society (JWES) for more than 10 years. The purpose of the continuous activity is to establish standard procedures for evaluating failure probabilities of Japanese nuclear structural components such as PV&P and steam generator tube, combining the state-of-the-art knowledge on structural integrity of nuclear structural components and modern computer technology such as parallel processing. This paper shows two topics of the newest results of JWES committee, PFM analysis of aged reactor pressure vessel considering embedded cracks and PFM analysis of piping considering seismic loading, and one topic by JAERI itself, development of PTS analysis code for transient loading (PASCAL).

Progress of a research program on seismic base isolation of nuclear components

Abstract Development of an evaluation code and related test program have been conducted to provide the technical base of the seismic base isolation of nuclear components. In the Phase I (FY1991–FY1995) of the research, a methodology and a computer code Ver.1 for evaluating the effect of seismic base isolation of nuclear components were developed. Case study was carried out on the effectiveness of base isolation of emergency transformer. Difference of input earthquake motion, type of isolation device and influence of the soil property were studied. Case study of a cost/benefit analysis in introducing the base isolation to emergency transformer was tried as an application of the computer code. As the Phase II (FY1996–FY2000), in order to obtain the test data of component base isolation systems, a verification test program, in which the test utilizing the real earthquake and the test by a shaking table are to be carried out, has been initiated since FY1996. In the tests, dynamic response and failure mode of base isolation systems will be examined. This paper overviews the progress of Phase I and II researches.

Ductile fracture behavior of circumferentially cracked type 304 stainless steel piping under bending load

Abstract A ductile pipe fracture test program has been conducted in Japan Atomic Energy Research Institute (JAERI) to investigate the ductile fracture behavior of circumferentially cracked pipes and to demonstrate the validity of the leak before break concept in LWR pipings. In the paper are described the scope of the pipe test program and current test results for 6-inch diameter type 304 stainless steel pipes. Test pipes with a through-wall or a part-through crack in the circumferential direction were bent under low or high compliance condition, and stable or unstable pipe fracture behavior was investigated. J based tearing instability criterion and the net section collapse criterion are compared with the pipe test results, and the validity of these fracture criteria is discussed. Furthermore, geometries of acceptable flaws in pipes are evaluated considering the leak before break condition.

Measurement of leak-rate through fatigue-cracks in pipes under four-point bending and BWR conditions

Abstract Leak-rate tests were performed using 114 mm and 165 mm (4 and 6 in) diameter, schedule 80 pipes made of austenitic stainless steel SUS304 and carbon steel STS42. Each pipe contained a through-wall fatigue crack and was mounted on a four-point bending machine of 400 kN maximum loading. Tests were done under a pressure of 7 MPa, with a subcooling temperature. The leak rate was measured by a Venturi flow meter and a differential pressure transducer attached to the pressure vessel. Comparisons of the effect of pipe material, diameter and crack angle were made. This paper shows that from a Leak-Before-Break viewpoint, the stainless-steel pipe is superior to the carbon-steel one, and that the pipe with the larger diameter is better than the one with the smaller diameter. No unstable fracture was observed in the tests.

Stress intensity factor analyses of surface cracks in three-dimensional structures—Comparison of the finite element solutions with the results obtained by the simplified estimation methods

Abstract The stress intensity factor analyses of surface cracks in various three-dimensional structures were performed using the finite element computer program EPAS-J1. The results obtained by EPAS-J1 were compared with other finite element solutions or the results obtained by the simplified estimation methods. Among the simplified estimation methods, the equations proposed by Newman and Raju give the distributions of the stress intensity factors along a crack front, which were compared with the results obtained by EPAS-J1. It was confirmed by comparing the results that the finite element program EPAS-J1 gives the reasonable stress intensity factors of surface cracks in three-dimensional structures.

Improvements to a probabilistic fracture mechanics code for evaluating the integrity of a RPV under transient loading

Probabilistic fracture mechanics, which can evaluate the failure probability considering uncertainties in defect size, material properties, chemical compositions and non-destructive inspection, is a promising and rational methodology for assessing the reliability and integrity of structural components. In this paper, a description is given of a new probabilistic fracture mechanics analysis code which has been developed for evaluating the conditional probability of crack initiation and failure of a reactor pressure vessel under transient conditions such as a pressurized thermal shock. In addition, some improvements in the reliability and efficiency of probabilistic fracture mechanics analysis are reported and some results are presented to show their effectiveness.

Recent Japanese Probabilistic Fracture Mechanics Researches Related to Failure Probability of Aged RPV

In order to prepare for the need of probabilistic methodology in design, inspection and maintenance of nuclear components, JAERI (The Japan Atomic Energy Research Institute) has conducted PFM (Probabilistic Fracture Mechanics) researches of Phase 1 and Phase 2 since late 1980s. In order to establish the standard procedure, Phase 1 had been conducted from 1988 to 1994 by entrusting contract researches to JWES (The Japan Welding Engineering Society), MRI (Mitsubishi Research Institute Incorporation) and JSME (The Japan Society of Mechanical Engineers). Subsequently, JAERI initiated Phase 2 in 1996 aiming at more practical application. JAERI had entrusted contract researches from 1996 to 2000 to JWES. JAERI also initiated a development of a PFM Code PASCAL (PFM Analysis of Structural Components in Aging LWR) as well as the contract research. The development of PASCAL-Ver.1 was completed in 2000. PASCAL-Ver.1 was released in 2001. Using PASCAL-Ver. 1, round robin analyses on the usability of the code, the effect of annealing on the failure probability of an RPV, the probabilistic evaluation of the flaw acceptance standard in ASME ( or JSME) Code have been performed within the contract research. This paper presents the overview of activities related to RPV and some results of round robin analyses conducted in PFM Sub-Committee in JWES. In addition, the outline of PASCAL-Ver.1 is also introduced.

Embedded Crack Treatments and Fracture Toughness Evaluation Methods in Probabilistic Fracture Mechanics Analysis Code for the PTS Analysis of RPV

Using the probabilistic fracture mechanics analysis code PASCAL, we studied the treatment method of an embedded crack and the fracture toughness evaluation methods on the probability of crack initiation and fracture of a reactor pressure vessel (RPV). For calculating the stress intensity factor (SIF) of an embedded crack, the ASME and CRIEPI procedures were introduced into the PASCAL code. The CRIEPI method enables us to calculate the SIF values at three points on the crack tip. Under a severe pressurized thermal shock (PTS) condition, the crack growth analysis methods with different SIF calculation points and crack growth directions are compared. To evaluate precisely the fracture toughness after neutron irradiation, the new fracture toughness curves based on the Weibull distribution were incorporated into the PASCAL code. The calculated results with these new curves showed little difference in the conditional probabilities of RPV fracture as compared to the curve currently used in the U.S.Copyright