Hideo Machida

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).

Reliability Assessment of PLR Piping Based on Domestic SCC Data

This paper describes effects of a probability of detection (POD) of a stress corrosion crack on the reliability of the piping in a nuclear power plant (NPP). Various POD curves were proposed using the results of Japanese study on the detection for the stress corrosion crack which is frequently observed in austenitic stainless steel piping of boiling water reactor (BWR). Based on the proposed POD curves, the reliability of a flawed pipe was analyzed using probabilistic fracture mechanics (PFM) code. The results suggest that the detectable crack depth and oversight probability (e.g. human error) is important on the reliability of piping with the stress corrosion crack. The reliability of piping depends on the detectable crack depth rather than the oversight probability when the detectable crack depth is larger than 3mm. Meanwhile, it depends on the oversight probability, when the detectable crack depth is 3mm or less.Copyright

Structural integrity evaluation method for overheating rupture of FBR steam generator tube

Abstract This paper describes a structural integrity evaluation method for a SG tube of FBR in case of sodium–water reaction and creep rupture tests to obtain the strength of the tube material. In the SG of FBR, if intermediate size of water/steam leak (1–2 kg s −1 ) would occur from a tube, it could cause overheating rupture of the multiple tubes surrounding the initially failed tube due to generated sodium–water reaction heat. In the ultra-high temperature condition, the creep strength of the material is one of the dominant factors for failure behavior. Accordingly, we tried to apply the creep failure criterion for the overheating rupture of the SG tube. The creep rupture tests have been performed at ultra-high temperature conditions ranging from 1223.2 to 1323.2 K. The test material is ‘Mod.9Cr–1Mo steel’ which is one of the candidate materials for the tubes of the future SG of FBR. The test results have shown that tube rupture depends on the creep strength of the material; hence, instantaneous rupture does not occur even if the stress exceeds the design value of ultimate tensile strength. The test data have been suitably expressed using the Larson–Miller Parameter, and a structural integrity evaluation method based on the sum of the use-fraction associated with the creep damage has been proposed. Based on this method, the structural integrity of the tube in the sodium–water reaction flame has been evaluated. The results show that it is important to detect the initial leak of the tube within a short period and to reduce the steam pressure more rapidly by SG blowdown.

Probabilistic fracture mechanics analysis of nuclear piping considering variation of seismic loading

Abstract In conventional probabilistic fracture mechanics (PFM) analyses, seismic loading is considered as a large deterministic value, although there exists the variation of the seismic load as well as response of building and components. On the other hand, such stochastic behaviours have already been taken into account in the field of seismic probabilistic safety assessment. This paper proposes a new PFM model for nuclear piping that takes into account the variation of seismic loading. The distribution in ground acceleration is modelled with the seismic hazard curve. The distribution in piping response during a seismic event is modelled with a log–normal distribution. Since the seismic load has large variation, when not adopting an upper limit to the distribution in seismic stress, the break probability calculated from the present PFM analysis becomes equal to the probability that the seismic stress exceeds the collapse stress of a sound pipe. This implies that the existence of a crack has no effect in these PFM analyses, and this result does not satisfy the purpose of PFM analysis to evaluate the failure probability per crack. Therefore, the seismic stress was limited to the collapse stress of a sound pipe in the present PFM analysis to evaluate the conditional break probability per crack.

Development of LBB Assessment Method for Japanese Sodium Cooled Fast Reactor (JSFR) Pipes: 1—Study on the Premise for the Standardization of Assessment Procedure

This paper describes the premise for the standardization of Leak Before Break (LBB) assessment procedure applicable to Japanese Sodium cooled Fast Reactor (JSFR) pipes made of modified 9Cr-1Mo steel. For the early commercialization of the SFRs, economic competitiveness is one of the most essential requirements. Japan Atomic Energy Agency (JAEA) proposes an attractive plant concept and studies the applicability of some innovative technologies to the plant. One of the most practical means to reduce the construction costs is to reduce the total amount of structural materials. To meet the requirements, modified 9Cr-1Mo steel has attractive characteristics as a main structural material of SFRs, because the steel has both excellent thermal properties and high temperature strength. By employing the steel as the main structural material, remarkably compact plant design can be achieved. Since the ductility and toughness of the steel is relatively inferior to those of conventional austenitic stainless steels, a LBB assessment technique suitable for the pipes made of modified 9Cr-1Mo steel may be required. In addition, since the SFR pipes are mainly subjected to displacement controlled thermal loads, it is expected that fast unstable fracture is unlikely. Taking both material and structural features into account, the framework to establish a precise LBB assessment procedure for SFR pipes must be organized. For the standardization of the LBB procedure, the main investigation items were defined as follows: (1) Approval of the assessment flowchart eliminating uncertainty due to small scale leakage, e.g. self plugging phenomenon and influence of crack surface roughness on leak rate. (2) Proper selection of LBB assessment objects in JSFR. (3) Distinguishment between the matters covered by a design code and LBB, i.e. assumption of initial flaw(s). (4) Development of creep and/or fatigue crack extension assessment technique, including collection of necessary material data. (5) Development of unstable fracture assessment technique. (6) Development of leak rate evaluation technique. (7) Characterization of loads for LBB assessment. (8) Standardization of the procedure as the Japan Society of Mechanical Engineers (JSME) code.Copyright

Plastic Collapse Evaluation for Multiple Circumferential Flaws in a Pipe

Multiple stress corrosion cracks initiate in a weld joint of primary loop recirculation system (PLR) piping in many cases. To prepare a stability assessment method for a pipe with these flaws is one of the major interests for Fitness-for-Service (FFS) Codes of Japan Society of Mechanical Engineers (JSME). This paper presents plastic collapse assessment method of a pipe with multiple circumferential flaws and proposals to revise FFS Codes. Plastic collapse strength of a pipe with multiple circumferential flaws is evaluated with respect to the weakest axis where the minimum plastic collapse moment is obtained, and a program assessing the weakest axis is presented in this study. A ratio of collapse strength corresponding to the weakest axis to that putting all cracks together is defined as strength correction factor, FMC . The strength correction factors when the number of flaws is two and three are summarized in the assessment diagrams, and proposals to revise FFS Codes are reported based on these results.© 2008 ASME

Effect of Crack Detection Performance and Sizing Accuracy on Reliability of Piping With Stress Corrosion Cracks

Integrity assessment of primary recirculation system (PLR) piping with stress corrosion cracks is one of the recent interests in Japanese BWR plants. Crack detection performance and a sizing error in non-destructive test (NDT) have a great influence on the integrity assessment of piping. This paper presents effects of crack detection performance and sizing accuracy on the reliability of PLR piping with stress corrosion cracks. Failure probability (leak and break) of PLR piping maintained according to the rules in Fitness-For-Service (FFS) Codes of JSME was analyzed using a probabilistic fracture mechanics (PFM) code which could analyze initiation and growth behavior of stress corrosion cracks. For the reliability of PLR piping, reducing oversight of cracks and crack sizing error are effective, and detectable crack size has not much influence on reliability of piping. For the higher reliability, reducing oversight of cracks is more important than the improvement of the detectable crack size in NDT, and the qualified inspectors should be employed for sizing of detected cracks in order to be suitable measurement in NDT.© 2008 ASME

Crack opening displacement of a through-wall crack in a plate subjected to bending load

This study was performed in order to clarify crack opening displacement (COD) of through-wall cracks in a plate subjected to bending load. The former COD evaluation methods were mainly developed corresponding to tensile load, but there has been nothing that has been developed corresponding to bending load. Therefore, the authors evaluated CODs of the through-wall cracks in plates which were subjected to a bending load using finite element method (FEM) analyses, and proposed a simplified COD evaluation method accounting for both tensile and bending loads. The proposed method is useful for leakage evaluation at a crack opening of an elbow crown or in the vicinity of the coolant surface of a vessel in which the bending stress is relatively large.

Revision of Flaw Evaluation Methods of Pipes Having a Circumferential Flaw in JSME Fitness-for-Service Codes

This paper shows the technical basis of revision to the JSME Fitness-for-Service Code (the FFS code) for flaw evaluation methods of pipes which have a very shallow circumferential flaw. When a flaw in a pipe is very small, the allowable stress between the FFS code and the JSME Design and Construction Code (the design code) is mismatched. Fracture strength of a pipe with small flaw depends on the tensile strength accompanied by large deformation. Therefore, fracture mechanics is not applicable in such a case. This mismatch has been resolved for an axial crack assessment by improving definition of flow stress for shallow crack. In this study, the authors investigated this mismatch in the allowable stress in the flaw assessment for a pipe with a circumferential crack. Some past fracture test results of pipes showed that flawed pipes did not fracture at the flaw section when the circumferential flaw size was small and they failed by oval deformation or plastic buckling. Allowable stress for such behavior has been incorporated in some existing design codes as a restriction for plastic collapse. Through the reevaluation of the existing piping fracture test results, the applicability of fracture evaluation methods defined in the FFS code was examined for the case that flaw size was very small. As a result, the fracture evaluation method based on flow stress was found not to be applicable when flaw size was very small, and the failure criterion in this case depended on the collapse strength accompanying with ovalization. Revisions of the FFS code reflecting these examination results were proposed in this paper.Copyright

Research Plan on Failure Modes by Extreme Loadings Under Design Extension Conditions

As the important lessons learned from Fukushima-nuclear power plant accident, preparation based on Probabilistic Risk Assessment (PRA) with adequate scenario is strongly recognized as essential countermeasures against severe accidents, which are possible in nuclear plants. IAEA requires design extension conditions (DEC) for considering severe accidents. From a view point of structural design, the strength evaluation approach for DEC is somewhat different from conventional one for design basis accident (DBA). There are additional failure modes by extreme loadings under DEC. Best estimation with possible scenarios is necessary for PRA and planning of accident management (AM).This paper introduces study plan on failure modes and their mechanisms by extreme loadings under DEC.First step is the list-up of possible failure modes which should be assumed for extreme loadings such as very high temperature, high pressure and great earthquakes. Next is clarification of failure mechanism and relevant limit strength.One of examples is the failure modes of structural discontinuities under high pressure such as ductile fracture and local failure. Another example is ones of pipes under severe earthquake such as collapse and low cycle fatigue.To clarify above questions, such different scale tests were planned and conducted as the fundamental tests with simulated materials, structural element tests and structural tests. Preliminary results of above tests and next plans are explained.Copyright