Kazuya Tsutsumi

Fatigue Crack Initiation of 304L Stainless Steel in Simulated PWR Primary Environment: Relative Effect of Strain Rate

The French Regulatory Commission insisted on a survey justifying the assumed mechanical behavior of components exposed to Pressurized Water Reactor (PWR) water under cyclic loading without taking into account its effect. In the US and Japan, the fatigue life correlation factors, so called Fen, are formulated and standardized on the basis of laboratory data to take into account the effect on fatigue life evaluation.However, the current fatigue codification, suffers from a lack of understanding of environmental effects on the fatigue lives of stainless steels in simulated hydrogenated PWR environments. Samples tested in a recent study were analyzed to highlight the strain rate effect (within a range 0.4%/s to 0.004%/s) at the early stage of fatigue life in PWR primary environment for a 304L stainless steel. The deleterious effect of PWR primary environment on fatigue crack initiation was observed with a quantitative microscopic approach. Multi scale observations of oxide morphology and microstructure were carried out from common optical microscopy using recent technologies such as 3D oxide reconstruction, and DualBeam observations.Copyright

Fatigue Crack Growth Curve for Austenitic Stainless Steels in PWR Environment

Although reference fatigue crack growth curves for austenitic stainless steels in air environments and boiling water reactor (BWR) environments were prescribed in JSME S NA1-2002, similar curves for pressurized water reactors (PWR) were not prescribed. In order to propose the reference curve in PWR environment, fatigue tests of austenitic stainless steels in simulated PWR primary water environment were carried out. According to the procedure to determine the reference fatigue crack growth curve of BWR, which of PWR is proposed. The reference fatigue crack growth curve in PWR environment have been determines as a function of stress intensity factor range, Temperature, load rising time and stress ratio.Copyright

A proposal of fatigue life correction factor Fen for austenitic stainless steels in LWR water environments

The fatigue life of austenitic stainless steel has recently been shown to undergo remarkable reduction with decrease in strain rate and increase in temperature in water. Either of these parameters as a factor of this reduction has been examined quantitatively and methods for predicting the fatigue life reduction factor F en in any given set of conditions have been proposed. All these methods are based primarily on fatigue data in simulated PWR water owing to the few data available in simulated BWR water. Recent Japanese fatigue data in simulated BWR water clearly indicated the effects of the environment on fatigue degradation to be milder than under actual PWR conditions. A new method for determining F en in BWR water was developed in the present study and a revised F en in PWR water is also proposed based on new data. These new models differ from those previously used primarily with regard to the manner in which strain amplitude is considered to affect F en in the environment.

Evaluation of the Susceptibility to SCC Initiation of Alloy 690 in Simulated PWR Primary Water

Alloy 690 has been widely used in fabricating components of LWR plants as an alternative material to Alloy 600 which has exhibited a significant susceptibility to PWSCC. However, some authors have reported that Alloy 690 can suffer a significant susceptibility to SCC crack growth when highly cold worked. While most of the recent studies emphasize SCC propagation phase, EDF and its partners are focusing on the material’s resistance to SCC initiation. This paper summarizes the current work carried out at EDF MAI on the SCC initiation. By means of constant elongation rate tests (CERTs) and constant displacement tests, experimental investigation of the susceptibility to PWSCC were performed. No SCC was observed on either an extruded bar or on two plates, even after 24%-1D cold rolling, confirming the superior PWSCC resistance of Alloy 690 independent of a amount of intergranular precipitation of carbides, and also revealing that such cold rolling does not necessarily decrease the resistance to SCC. On the other hand, a experimental steam generator tube that has a degraded microstructure due to specific heat-treatment revealed its susceptibility to SCC, probably because of the interactive effect of microstructure with heavy intragranular carbide precipitations and the cold worked superficial layer. This phenomenon is in good agreement with results previously published. In this study, the maximal crack depth slightly increased when DH increased from 5 to 60 cc.kg−1H2O. No significant prior ageing effect on the crack depth was observed, even when ageing was combined with high DH.

PWSCC Susceptibility of Alloy 690, 52 and 152

Long-term constant load stress corrosion cracking (SCC) testing for alloys 690/152/52 at 360 ℃ is ongoing, showing no rupture for more than 105 h, suggesting immunity to primary water (PW) SCC initiation under stress level assumed for primary circuit components in pressurized water reactor (PWR) plants. Since the mechanical plug for steam generators has the largest cold work strain, a mock-up PWSCC test, using a mechanical plug of alloy 690, was also performed for evaluation of time to failure under stress and cold work conditions assumed for operating plan. As a result, it was proven that no crack initiated up to approximately 4 × 104 h at 360 ℃. PWSCC susceptibility was also evaluated in terms of crack growth rate (CGR). The CGR of alloy 690 increased after cold working, and the degree of increment is significantly affected by the nature of carbide precipitate along grain boundaries. It was found that increase in CGR caused by cold working remained relatively low when grain boundary carbides precipitated continuously along grain boundaries and coherently with the matrix. Contrarily, CGR grew higher in the materials with lower coherency. It was also revealed that alloy 690 with no grain boundary carbides (solution annealed alloy) showed a small increase of CGR after cold working.

Fatigue Life of the Strain Hardened Austenitic Stainless Steel in Simulated PWR Primary Water

Over the last 20 years or so, many studies have revealed the deleterious effect of the environment on fatigue life of austenitic stainless steels in pressurized water reactor (PWR) primary water. The fatigue life correlation factor, so-called Fen, has been standardized to consider the effect on fatigue life evaluation. The formulations are function of strain rate and temperature due to their noticeable negative effect compared with other factors [1,2]. However, mechanism causing fatigue life reduction remains to be cleared.As one of possible approaches to examine underlying mechanism of environmental effect, the authors focused on the effect of plastic strain, because it could lead microstructural evolution on the material. In addition, in the case of stress corrosion cracking (SCC), it is well known that the strain-hardening prior to exposure to the primary water can lead to remarkable increase of the susceptibility to cracking [3,4]. However, its effect on fatigue life has not explicitly been investigated yet.The main effort in this study addressed the effect of the prior strain-hardening on low cycle fatigue life of 304L stainless steel (SS) exposed to the PWR primary water. A plate of 304LSS was strain hardened by cold rolling or tension prior to fatigue testing. The tests were performed under axial strain-controlled at 300 °C in primary water including B/Li and dissolved hydrogen, and in air. The effect on environmental fatigue life was investigated through a comparison of the Fen in experiments and in regulations, and also the effect on the fatigue limit defined at 106 cycles was discussed.Copyright

Fatigue Crack Growth Threshold of Austenitic Stainless Steels in Simulated PWR Primary Water

Fatigue crack propagation behaviors in PWR environment can be evaluated by using crack growth rate (CGR) curves which are given in JSME code on Fitness-for-Service for nuclear power plant. The CGR curves, however, are only defined in crack growth region and crack growth thresholds are not considered. Since there is a case that stresses in low ΔK region is applied to the components in case of fatigue, it is needed to investigate near-threshold fatigue CGR to establish fatigue assessment. In this study, CGR tests for stainless steels were carried out, and CGR in the region and ΔKth were obtained in simulated PWR primary water. It was found that CGR was accelerated in high temperature water compared to that in air and ΔKth existed even in water environment. ΔKth was influenced by temperature, stress ratio and frequency, independently of materials. Oxide-induced-crack-closure has an important role in high temperature water. ΔKth was formulated and ΔKth evaluation method, whose accuracy were ±25% between experimental data and evaluation value, was proposed.© 2007 ASME

Effect of Factors on Fatigue Life in PWR Water Environment

It is known that the fatigue life in elevated temperature water is substantially reduced compared with that in the air (1–4) . Although the key parameters that have an effect on fatigue lives are strain rate and temperature in PWR water environment, it is necessary to consider the other factors on fatigue life for accurate evaluation. The effects of many factors on fatigue life have been investigated experimentally in the EFT project of Japan Nuclear Energy Safety Organization (JNES). Many tests have been done for carbon, low alloy, stainless steel and nickel-based alloy, and the environmental fatigue life equation that evaluates quantitative factor influencing the fatigue life has been proposed. In this paper, in order to evaluate effects of material structure difference between base metal and weld metal, strain amplitude, strain rate, strain ratio, temperature, sulfur content in steel, aging, water flow rate and strain holding, fatigue tests were performed in simulated PWR water environment.Copyright

Evaluation of Fatigue Damage in LWR Water With and Without Threshold and Moderation Factor

During the past twenty years, the fatigue initiation life of LWR structural materials, carbon, low alloy and stainless steels has been shown to decrease remarkably in the simulated LWR (light water reactor) coolant environments. Several models for evaluating the effects of environment on fatigue life reduction have been developed based on published environmental fatigue data. Initially, based on Japanese fatigue data, Higuchi and Iida proposed a model for evaluating such effects quantitatively for carbon and low alloy steels in 1991. Thereafter, Chopra et al. proposed other models for carbon, low alloy and stainless steels by adding American fatigue data in 1993. Mehta developed a new model which features the threshold concept and moderation factor in Chopra’s model in 1995. All these models have undergone various revisions. In Japan, the MITI (Ministry of International Trade and Industry) guideline on environmental fatigue life reduction for carbon, low alloy and stainless steels was issued in September 2000, for evaluating of aged light water reactor power plants. The MITI guideline provide equations for calculations applicable only to stainless steel in PWR water and consequently Higuchi et al. proposed in 2002 a revised model for stainless steel which incorporates new equations for evaluation of environmental fatigue reduction in BWR water. The paper compares the latest versions of these models and discusses the conservativeness of the models by comparison of the models with available test data.Copyright

The Modified Rate Approach Method to Evaluate Fatigue Life Under Synchronously Changing Temperature and Strain Rate in Elevated Temperature Water

The fatigue life of steel in elevated temperature water is strongly affected by the composition of the environmental water, temperature and strain rate. The effects of these parameters on fatigue life reduction have been investigated experimentally in these years. One problem to be discussed is the fact that the previous experiments which leaded main conclusions on the environmental effects were generally executed by keeping a set of experimental parameters constant. In the transient condition in an actual plant, however, such parameters as temperature and strain rate are not constant. In order to evaluate fatigue damage in an actual plant on the basis of experimental results under constant temperature and constant strain rate conditions, the modified rate approach method was developed. The method can be applicable to changing temperature condition and strain rate condition separately. In the present study, an additional model is proposed with considering that both temperature and strain rate change simultaneously in an actual plant. The applicability of this method is discussed and verified experimentally. The fatigue lives predicted by this method are scattered within the factor of 2 or 3 bands against test results even when several parameters changed synchronously.Copyright