Kunio Onizawa

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.

Development of a reconstitution technique for Charpy impact specimens by surface-activated joining for reactor pressure vessel surveillance

Abstract The reconstitution technique by surface-activated joining (SAJ) has been investigated for the reuse of undeformed portions of tested Charpy impact specimens. SAJ can be achieved by the removal of surface contamination by rotating one of the specimens in a vacuum while applying modest frictional force. In this study, the temperature distribution during joining and hardness distribution after joining were measured. Charpy impact tests were then performed to evaluate the thermal and mechanical effects on Charpy specimens reconstituted by SAJ. The results showed that SAJ can be achieved with a hardened width of less than 1·5 mm, and on a width of less than 3 mm heating up above reactor operation temperatures, produced in either side of the joined interface. Charpy transition temperatures could be evaluated from the reconstituted Charpy specimens. It is concluded from comparison with other welding methods that the SAJ method is most suited for reconstituting irradiated surveillance specimens.

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.

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

Effects of Neutron Flux and Irradiation Temperature on Irradiation Embrittlement of A533B Steels

Irradiation embrittlement of A533B steels with low copper contents were investigated from the point of dose rate and irradiation temperature effects. Change of neutron flux in the range from {minus}10{sup 12} to {minus}10{sup 13} n/cm{sup 2}/s (E > 1 MeV) did not have a significant effect on the embrittlement. Irradiation temperature change of 1 C resulted in the transition temperature shift ({Delta}T{sub 41J}) of about 1 C and yield stress change ({Delta}{sigma}{sub y}) of about 0.8 MPa. Factors that might affect the embrittlement of low copper steels are also discussed.

Thermal and stress analyses of the reactor pressure vessel lower head of the three mile island unit 2

Abstract Thermal and stress analyses were performed using the finite element analysis code ABAQUS to clarify the factors which caused tears in the stainless steel liner of the reactor pressure vessel lower head of the Three Mile Island Unit 2 (TMI-2) reactor pressure vessel during the accident on 28 March 1979. The present analyses covered the events which occurred after approximately 20 tons of molten core material were relocated to the lower head of the reactor pressure vessel. They showed that the tensile stress was highest in the case where the relocated core material consisting of homogeneous UO2 debris was assumed to attack the lower head and the debris was then quenched. The peak tensile stress was in the vicinity of the welded zone of the penetration nozzle. This result agrees with the findings from the examination of the TMI-2 reactor pressure vessel that major tears in the stainless steel liner were observed around two penetration nozzles of the lower head.

Development of Stress Intensity Factor Coefficients Database for a Surface Crack of an RPV Considering the Stress Discontinuity Between Cladding and Base Metal

Under a transient loading like pressurized thermal shock (PTS), the stress discontinuity near the interface between cladding and base metal of a reactor pressure vessel (RPV) is caused by the difference in their thermal expansion coefficients. So the stress intensity factor (SIF) of a surface crack close to the interface should be calculated taking account of the stress discontinuity. Many SIF calculations have to be performed many times in Monte Carlo simulation of the probabilistic fracture mechanics (PFM) analysis. To avoid the time consuming process from the SIF calculation in the PFM analysis, the influence coefficients were developed to calculate the SIF easily and accurately corresponding to the stress distributions in the cladding and base metal. Stress distributions in cladding and base metal are modeled to linear and third-order polynomial expressions, respectively. The non-dimensional SIF coefficients were obtained from FEM analyses. The SIF value at the surface was determined by linear extrapolation of SIF value near the surface. Using the SIF coefficients, the SIF values at the crack tips at both surface and deepest points of a surface crack are evaluated accurately and in a reasonable time.Copyright

Reconstitution of Charpy Impact Specimens by Surface Activated Joining

The method of surface activated joining (SAJ) was applied to reconstitution of Charpy impact specimens. SAJ makes use of friction at material surfaces in a vacuum to achieve joining without melting. This paper describes verification to apply SAJ to reconstitution of Charpy impact specimens with unirradiated reactor pressure vessel plate materials, JRQ and HSST-03. By optimizing joining parameters, heat affected zones induced by friction had a width of about 1 mm to either side of the joint. The original Charpy impact properties in the transition region were reproduced from reconstituted impact specimens with the insert length of 10 mm. The maximum temperature during joining was appreciably low at a given distance from the joining interface, compared with other conventional welding techniques.

Study on Flaw Acceptance Standard of ASME Code Sec. XI Based on Failure Probability

Based on the failure probability, the flaw acceptance standard of ASME Code Sec. XI is examined with some concerns weather the failure probability is uniform for flaws with various aspect ratios and failure frequencies are small enough. In this paper, the results of preliminary case studies are described on the failure probability of reactor pressure vessels (RPVs) with a surface flaw specified in Sec. XI. PFM code PASCAL was used for case studies. A PTS (Pressurized Thermal Shock) transient prescribed by NRC/EPRI PTS Benchmark Study was used as an applied load. Analysis results showed that the conditional failure probability of a RPV with an initial flaw of acceptable depth depends on the aspect ratio. In the case flaw shapes are close to semi-circular, the failure probability are higher than that of the cases aspect ration are less than 0.6 by one order of magnitude due to the difference of fracture behavior at the surface point. A case study for determining the acceptable flaws based on failure probability was also carried out.Copyright

Research and development related to PFM for aged nuclear components

Abstract At the Japan Atomic Energy Research Institute (JAERI), research activities related to probabilistic fracture mechanics (PFM) have been conducted as a part of the research program on aging and structural integrity of LWR components. This paper describes the outline of two activities related to PFM, i.e. the development of a PFM code and a contract research on ‘Application of PFM Methodology to Reliability Assessment of Nuclear Components’ implemented by the Japan Welding Engineering Society (JWES). In the former research, a new PFM code PASCAL (PFM Analysis of Structural Components in Aging LWR) was developed. This code has some new functions in models of semi-elliptical crack extension, elastic–plastic fracture analysis based on R6 method and options for the evaluation of overlay cladding and warm pre-stress (WPS) effect. Besides, the code has the function to evaluate the effect of irradiation embrittlement recovery by thermal annealing of a reactor pressure vessel and re-irradiation embrittlement. Based on the analyses on benchmark problem conducted by USNRC/EPRI, performance and functions introduced in the code were examined. Some case studies were also carried out to investigate the influence of various parameters. On the other hand, JAERI has been sponsoring the PFM related activities in relation to the structural integrity of LWR components. These activities have been conducted at JSME and JWES. The objective of this activity has been to provide for the future need of PFM methodology.