Light Water Reactor Coolant Environments Research Articles

Effects of pressurized water reactor environment and cyclic loading parameters on the low cycle fatigue behavior of 304L stainless steel

Austenitic stainless steels used in light water reactor coolant environments are susceptible to environmentally assisted fatigue due to non-monotonic loading conditions, primarily associated with load-follow, thermal transients, or intermittent plant shutdowns and start-ups. This study investigates the effects of a high-temperature pressurized water reactor (PWR) water environment and cyclic loading parameters on the low cycle fatigue behavior of austenitic 304L stainless steel. Prolonged exposure to a PWR environment and cyclic loading conditions such as a lower strain rate or a higher fraction of slow strain rate enhances the initiation and accelerates the crack growth rate of fatigue cracks, resulting in decreased fatigue life. The deformation-induced α'-martensite is observed in proximity to fatigue crack tips primarily in specimens tested in simulated PWR primary water, while cellular dislocation structures are more frequently observed near crack tips in specimens tested in high-temperature air. The deformation-induced martensitic transformation from γ-austenite to α'-martensite, occurring via the precursor ε-martensite phase, contributes to the accelerated fatigue crack growth rate in a PWR environment with hydrogen.

Research on the Application of Fen for Environmentally Assisted Fatigue Evaluation of the Austenitic SS Pipe Under Combined Transient Loads

Accumulative test data indicate that the effects of the light water reactor (LWR) environment could cause the fatigue resistance of primary pressure boundary components materials to be significantly reduced. Environmentally assisted fatigue (EAF) is the abbreviation of the environmentally assisted fatigue. In 2007, Nuclear Regulatory Commission (NRC) issued RG. 1.207. It was updated in 2014. And, it requires that the effects of LWR environment on the fatigue life reduction of metal components should be considered for new design plants. And it suggests to use environmental correction factor, Fen, to account for EAF. NRC regulation (NUREG), NUREG/CR-6909 (NRC, 2013, “Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials,” U.S. Nuclear Regulatory Commission, Argonne, IL, Standard no. NUREG/CR-6909), presents the detail Fen calculation formula. Fen is a function of temperature, strain rate, dissolved oxygen level in water, and sulfur content of the steel. Accordingly, Fen calculation will present a comparatively conservative result. Depending on the experience of the primary pressure boundary piping transient operation, Fen varies during each transient. More uncertainty and confusion are raised during the application of the Fen method. The research work in this paper includes: first, the typical character of piping thermal transient is derived based on the existing experience. Second, small specimen EAF tests are conducted depending on the above derived combined loading characters. Then effort is taken to improve the application of the Fen method for the combined multitransient loading conditions. And the results are compared with those of the lowest instantaneous Fen method and equalization of the weighted Fen method. Finally, a designed test plan is presented.

Hydrothermal corrosion of SiC in LWR coolant environments in the absence of irradiation

Assessment of the thermodynamics of SiC corrosion under light water reactor coolant environments suggests that silica formation is always expected in the range of applicable pH and potential. Autoclave testing of SiC-based materials in the absence of ionizing radiation was performed. The kinetics data from these tests, when compared with kinetics of silica dissolution in water and post-exposure characterization of SiC samples, suggest that oxidation of SiC to form silica is the rate-limiting step for recession of SiC in high temperature water. Oxygen activity in water was determined to play an important role in SiC recession kinetics. A simplified model of a power loop shows the effect of silica dissolution from the hot region (resembling fuel) and deposition in the cold regions.

Measurement methods for surface oxides on SUS 316L in simulated light water reactor coolant environments using synchrotron XRD and XRF

Synchrotron X-ray diffraction (XRD) and X-ray fluorescent (XRF) measurement techniques have been used for non-destructive characterization of surface oxide films on Type 316L austenitic stainless steels that were exposed to simulated primary water environments of pressurized water reactors (PWR) and boiling water reactors (BWR). The layer structures of the surface spinel oxides were revealed ex situ after oxidation by measurements made as a function of depth. The layer structure of spinel oxides formed in simulated PWR primary water should normally be different from that formed in simulated BWR water. After oxidation in the simulated BWR environment, the spinel oxide was observed to contain NiFe2O4 at shallow depths, and FeCr2O4 and Fe3O4 at deeper depths. By contrast, after oxidation in the simulated PWR primary water environment, a Fe3O4 type spinel was observed near the surface and FeCr2O4 type spinel near the interface with the metal substrate. Furthermore, by in situ measurements during oxidation in the simulated BWR environment, it was also demonstrated that the ratio between spinel and hematite Fe2O3 can be changed depending on the water condition such as BWR normal water chemistry or BWR hydrogen water chemistry.

A Review of the Effects of Coolant Environments on the Fatigue Life of LWR Structural Materials

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code specifies design curves for the fatigue life of structural materials in nuclear power plants. However, the effects of light water reactor (LWR) coolant environments were not explicitly considered in the development of the design curves. The existing fatigue-strain-versus-life (ε-N) data indicate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. Under certain environmental and loading conditions, fatigue lives in water relative to those in air can be a factor of 15 lower for austenitic stainless steels and a factor of ≈30 lower for carbon and low-alloy steels. This paper reviews the current technical basis for the understanding of the fatigue of piping and pressure vessel steels in LWR environments. The existing fatigue ε-N data have been evaluated to identify the various material, environmental, and loading parameters that influence fatigue crack initiation and to establish the effects of key parameters on the fatigue life of these steels. Statistical models are presented for estimating fatigue life as a function of material, loading, and environmental conditions. An environmental fatigue correction factor for incorporating the effects of LWR environments into ASME Code fatigue evaluations is described. This paper also presents a critical review of the ASME Code fatigue design margins of 2 on stress (or strain) and 20 on life and assesses the possible conservatism in the current choice of design margins.

Technical Basis for Revised Reference Crack Growth Rate Curves for Pressure Boundary Steels in LWR Environment

Reference fatigue crack growth rate curves have been contained in Appendix A of Section XI since its inception. The curves have been designed to be applicable to carbon and low alloy pressure vessel steels exposed to either air or light water reactor coolant environments. Data obtained over the past several years have shown a different behavior of these steels in the LWR environment than that predicted by the present reference curve. A revised set of reference curves have been formulated, incorporating a new curve shape as well as a dependency of growth rate on R ratio (minimum load/maximum load). This article provides the background and justification for such a revision, details the methodology used to develop the revised curves, and includes an evaluation of the adequacy and impact of the revised curves as compared with the single curve which they replace.