PWR Environment Research Articles

Effects of irradiation and stress on crack initiation behavior of 316SS flux thimble tubes in PWR environment

Irradiation assisted stress corrosion cracking (IASCC) behavior and irradiated microstructural features of cold-worked 316 stainless steel flux thimble tube specimens were investigated in 340 °C simulated PWR primary water. It covers three dose levels (∼40, ∼60 and ∼100 dpa) and five stress levels (35%, 40%, 50%, 60% and 70% of irradiated yield strength at high temperature. Crack initiation in highly irradiated materials occurred rapidly in 33 failed specimens at high-stress level. A clear dose dependence for IASCC initiation was observed. The susceptibility to IASCC initiation reached saturation around 40–60 dpa. Fracture surface of different dose level O-ring specimens showed that intergranular cracking was the main fracture mode. Discrete slip steps were observed on intergranular fractured regions of each dose level specimens, which is associated as evidence of dislocation channeling. A threshold stress level which was close to 30% yield strength was determined on various testing results from literature and this study. Data scatter was observed for those results from different studies but all agreed on that the failure rate increases with stress level and neutron dose.

Unveiling the dual role of ion irradiation on the corrosion behavior of lean-Cr FeCrAl alloys in a simulated PWR environment

This study focuses on the synergistic effect of ion irradiation and corrosion on FeCrAl alloys with lean-Cr content (7–10 wt%) and with 13 wt% Cr. Upon subjecting unirradiated, 1-dpa, and 5-dpa samples to a simulated PWR environment for 15 days, it is found that irradiation alters the distribution of outer oxide particles and the composition of inner oxide layer. Our results emphasize the dual role of irradiation, depending on the perspective considered: It suppresses overall corrosion while promoting localized corrosion. Cr content effect on the resistance to irradiation-corrosion coupling is also discussed.

Short communication: “Effect of dissolved hydrogen on the early corrosion behavior of 316 stainless steel in simulated PWR environments”

In simulated pressurized water reactor environments, the corrosion test on 316 stainless steel was conducted for 40 to 4000 s, with 25 and 50 cc/kg-H2O dissolved hydrogen concentrations (NH and HH). It was found that the external Fe-rich oxide particles and the inner Cr-rich layer were simultaneously formed at the beginning of corrosion. However, the first formed Cr-rich layer was mainly Cr(OH)3 in the NH condition, while Cr2O3 in the HH condition. As exposure time increased, fraction of Cr2O3 increased due to transformation from Cr(OH)3 in both conditions. In addition, for HH condition, partial transformation to FeCr2O4 was observed.

Mechanistic study of incipient corrosion for nuclear grade lean-Cr FeCrAl alloys in a simulated PWR environment

This study systematically investigated the incipient corrosion behavior of Fe-13Cr-4.5Al (wt%) and lean-Cr (7–10 wt%) with 5Al (wt%) FeCrAl alloy with a timescale of 4 months under simulated PWR conditions (18.6 MPa and 360 °C) in an autoclave system. The corrosion induced two types of structure of oxide scale. FeCrAl alloys corroded for 3 months exhibited a duplex structure containing outer oxide particles and an inner Cr-rich spinel. During the corrosion test lasting 3–4 months, a triplex structure containing outer oxide particles, an inner compact Cr-rich spinel layer, and a porous spinel layer was observed. Moreover, Cr enriched at the periphery of outer oxides, indicating the formation of FeCr2O4 spinel oxide.

Comparing CrN and TiN Coatings for Accident-Tolerant Fuels in PWR and BWR Autoclaves

The development of coatings for accident-tolerant fuels (ATFs) for light water reactor (LWR) applications promises improved corrosion resistance under accident conditions and better performances during operation. CrN and TiN coatings are characterized by high wear resistance coupled with good corrosion resistance properties. They are generally used to protect materials in applications where extreme conditions are involved and represent promising candidates for ATF. Zr cladding tubes coated with 5 µm-thick CrN or TiN, exposed in an autoclave to simulated PWR chemistry and BWR chemistry, were characterized with SEM, EDS, and STEM. The investigation focused on the performance and oxidation mechanisms of the coated claddings under simulated reactor chemistry. Both coatings provided improved oxidation resistance in a simulated PWR environment, where passivating films of Cr2O3 and TiO2, less than 1 µm-thick, formed on the CrN and TiN outer surfaces, respectively. Under the more challenging BWR conditions, any formed Cr2O3 dissolved into the oxidizing water, resulting in the complete dissolution of the CrN coating. For the TiN coating, the formation of a stable TiO2 film was observed under BWR conditions, but the developed oxide film was unable to stop the flux of oxygen to the substrate, causing the oxidation of the substrate.

Study of high burnup effect on steam oxidation of Zircaloy and its regulatory implications via the developement of a pre-transient oxide model of TRANOX

A mechanisitc model that gives oxygen distribution of Zircaloy cladding with pre-transient oxide and absorbed hydrogen under steam oxidation environments has been developed. The model has been incorporated into TRANOX. The model has been validated against steam oxidation experiments with Zircaloy-4 specimens that were pre-oxidized in simulated PWR environments and past experimental data available in open literature. The high burnup effects on cladding oxidation owing to pre-transient oxide and hydrogen can be categorized into: thinning of initial Zr-matrix by the formation of pre-transient ZrO2, increasing diffusion resistance of oxygen with pre-transient ZrO2, and increasing oxygen diffusion rate with a thicker β-phase by β-stabilizing hydrogen. Relative magnitudes of these effects with temperature, pre-transient oxide thickness, and hydrogen content determine the dynamics of high burnup fuel oxidation. Mechanistically modeling the high burnup effect, the model comprehensively simulates the integrated effects of pre-transient oxide and absorbed hydrogen. This study confirms that the corrected Cathcart Pawel correlation-based Equivalent Cladding Reacted evaluation method employed by the U.S Nuclear Regulatory Commission is an effective, yet accurate, simplification of the high burnup cladding oxidation modeling for safety analysis of representative LBLOCA of current PWR systems. The application of the developed model to safety analysis revealed allowable discharge burnup limits of 59 and 76 MWd/kgU for Zircaloy-4 and Zr-Nb alloy, respectively from the view point of fuel accident safety.

Characterization of oxide layers formed on type 316 stainless steel exposed to the simulated PWR primary water environment with varying dissolved hydrogen and zinc concentrations

The effect of dissolved hydrogen (DH) content and zinc addition on the corrosion behavior of type 316 stainless steel was investigated in the primary water environment of pressurized water reactors. The coupled effect, as well as the individual effect of increasing DH concentration and zinc addition, was investigated. Increasing DH from 25 (normal) to 50 (high) cc/kg-H2O resulted in 3 times thicker oxide layer. With the addition of 30 ppb zinc, the oxide layer became substantially thinner in both normal and high DH conditions. The measured electrochemical properties of oxide layers were in agreement with the observed corrosion behavior such that the zinc incorporated oxide film showed significant increase in reaction resistance and bode impedance value. Meanwhile, both reaction resistance and bode impedance slightly decreased with increasing DH concentration. Overall, the best corrosion resistance was observed in the simulated PWR environment with normal DH concentration and zinc addition.

TEM investigation of SCC crack tips in high Si stainless steel tapered specimens

ABSTRACT The stress corrosion cracking (SCC) mechanism is investigated in high Si duplex stainless steel in a simulated PWR environment based on TEM analysis of FIB-extracted SCC crack tips. The microstructural investigation in the near vicinity of SCC crack tips illustrates a strain-rate dependence in SCC mechanisms. Detailed analysis of the crack tip morphology, that includes crack tip oxidation and surrounding deformation field, indicates the existence of an interplay between corrosion- and deformation-driven failure as a function of the strain rate. Slow strain-rate crack tips exhibit a narrow cleavage failure which can be linked to the film-induced failure mechanism, while rounded shaped crack tips for faster strain rates could be related to the strain-induced failure. As a result, two nominal strain-rate-dependent failure regimes dominated either by corrosion or deformation-driven cracking mechanisms can be distinguished.

Calculated Shoulder to Gauge Ratio of Fatigue Specimens in PWR Environment

A ratio of shoulder to gauge displacements (S2G) is calculated for three different fatigue specimens in a pressurized water environment. This ratio needs to be known beforehand to determine the applied shoulder displacements during the experiment that would result in the desired strain amplitude in the gauge section. Significant impact of both the applied constitutive law and specimen geometry on the S2G is observed. The calculation using the fully elastic constitutive law results in the highest S2G values and compares very well with the analytical values. However, this approach disregards the plastic deformation within the specimens that mostly develops in the gauge section. Using the constitutive laws derived from actual fatigue curves captures the material behaviour under cyclic loading better and results in lower S2G values compared to the ones obtained with the fully elastic constitutive law. Calculating S2G values using elastic–plastic constitutive law based on the monotonic uniaxial tensile test should be avoided as they are significantly lower compared to the ones computed with elastic–plastic laws derived from hysteresis loops at half-life.

Correlative Microscopy: Elucidating the Mechanisms of SCC in Structural Alloys in PWR Environments

Correlative Microscopy: Elucidating the Mechanisms of SCC in Structural Alloys in PWR Environments

Pitting and Electrochemical corrosion behaviour of 316L austenitic stainless steel subjected to warm deformation

Electrochemical and pitting corrosion behaviour of undeformed, deformed and annealed after deformation of the 316L stainless steel have been investigated by using electrochemical corrosion test methods. The results indicated that pitting occurred in undeformed sample, while the small pits were observed in deformed and annealed sample after immersion test in 10% FeCl3 solution. Microstructural characterization of undeformed, deformed and corroded sample were analysed by optical microscope and SEM. A Potentiodynamic polarization test in PWR environment was performed to evaluate the electrochemical behaviour of samples after deformation and heat treatment. The results showed that the corrosion rate decrease after heat treatment of deformed samples because heat treatment after deformation form a more protective oxide film on the surface of samples as compared to oxide film form on the surface of undeformed sample. The Ecorr shifted toward the passive direction with increase in annealing temperature after deformation. The shift of Ecorr towards the passive direction indicates a more protective passive film formed on the surface of heat treated samples after deformation.

Role of irradiation and irradiation defects on the oxidation first stages of a 316L austenitic stainless steel

The role of irradiation and irradiation defects on the oxidation first stages of 316 L alloy was investigated. A sample with both a proton pre-irradiated and an unirradiated area was exposed to a simulated PWR environment during 24 hours. Irradiation defects and Radiation Induced Segregation at grain boundary and on irradiation defects were characterized and quantified and their effect on the oxidation was evaluated. Irradiation affects the morphology, thickness and chemistry of the oxide layers formed. It enhances the oxidation kinetic and induces the formation of an inner oxide richer in chromium. Defects induced by irradiation act as preferential nucleation sites.

Stress corrosion cracking characteristics of CF8A austenitic stainless steels and interactions between multiple cracks in a simulated PWR environment

The stress corrosion cracking behaviors of CF8A austenitic stainless steels in a simulated pressurized water reactor environment are systematically investigated by slow-strain-rate tensile tests. Numerous transgranular stress corrosion cracking (TGSCC) cracks are identified upon the fracture of CF8A austenitic stainless steel surfaces. The specimens tested at lower strain rates showed higher fractions of TGSCC, a larger number of cracks, and smaller crack opening displacement. These phenomena are caused by anodic dissolution and crack interactions. The anodic dissolution is dominant at lower strain rates, and thus the crack-tip morphologies are changed by the diffusion of metal atoms and micro-pores. The crack interactions increase rapidly with increasing number of cracks and decrease of the crack-tip radius and crack spacing ratio. In addition, the inhomogeneity of the growth of interaction cracks results in the small cracks being shielded by the preferred growth cracks. On the contrary, the main cracks are accelerated by the stagnation of small cracks, eventually becoming critical cracks that lead to specimen fracture.

Replicating PWR Primary Water Conditions in Low Pressure H2-Steam Environment to Study Alloy 600 Oxidation Processes

The effects of thermodynamic parameters associated with low pressure H2-steam on the oxidation behavior of Nickel have been investigated to simulate PWR primary water conditions. An Y2O3-ZrO2 solid state reference electrode operated between 372 and 480°C, and exposed to variable hydrogen partial pressures, was used to obtain in-situ potential measurements to identify the Ni/NiO transition. Microstructural analyses on Alloy 600 coupons exposed to H2-steam under different oxidizing conditions showed a marked dependence of the intergranular oxidation processes, as expected for a PWR environment. Overall, the H2-steam simulated an environment that can be considered to be representative of PWR primary water.

PWR effect on crack initiation under equi-biaxial loading development of the experiment

The lifetime extension of the nuclear power stations is considered as an energy challenge worldwide. That is why, the risk analysis and the study of various effects of different factors that could potentially represent a hazard to a safe long term operation are necessary. The methodology for fatigue dimensions of the Pressurized Water Reactor components (PWR) is based on the use of design curves established from test carried out in air at 20 °C on smooth specimens by integrating safety coefficient that covers the dispersion of tests associated with the effects of structures. To formally integrate these effects, some international codes have already proposed and suggested a modification of the austenitic stainless steels fatigue curve combined with a calculation of an environmental penalty factor, namely Fen, which has to be multiplied by the usual fatigue usage factor. The aim of this paper is to present a new device “FABIME2E” developed in the CEA-LISN in collaboration with EDF and AREVA. These new tests allow quantifying accurately the effect of PWR environment on semi-structure specimen. This new device combines the structural effect like equi-biaxiality and mean strain and the environmental penalty effect with the use of PWR environment during the fatigue tests.

EBSD Characterization of IGSCC Crack Initiated on Alloy 600 in Simulated PWR Primary Water Environment

In the present work, we crystallographically examined intergranular cracks initiated on Alloy 600 in a simulated PWR environment by using EBSD. The cracks initiated on a flat tensile specimen of the alloy in a short period of slow strain rate tests (SSRTs), that is, the initiation of cracks was accelerated due to the asperity of the specimen surface induced by mechano-chemical polishing with colloidal silica suspension prior to the SSRTs. As thick oxide films were formed on the specimen surface during the SSRTs, the sputtering of the films was carried out for the removal of the film. After the sputtering, clear crystallographic information such as inverse pole figures (IPFs) on the specimen surface could be obtained. Based on the information, the intergranular cracks were crystallographically analyzed. Furthermore, three-dimensional characterization of the cracks was also performed.

Customization of the coupled environment fracture model for predicting stress corrosion cracking in Alloy 600 in PWR environment

The coupled environment fracture model (CEFM) has been modified and calibrated to predict crack growth rate (CGR) in Alloy 600 under typical PWR primary coolant conditions. The customized CEFM provides quantitative predictions of the effects of stress intensity factor, hydrogen concentration, yield strength, and temperature on CGR in Alloy 600 in PWR primary coolant environments, as well as explaining the dominating mechanism of stress corrosion cracking (SCC) in this alloy. The importance of the mechanical properties, such as yield strength and stress intensity factor, as identified previously from Artificial Neural Network (ANN) analysis of CGR data in the literature, has been confirmed theoretically. The success in explaining the intergranular stress corrosion cracking (IGSCC) of Alloy 600, argues that the CEFM, which was originally developed to describe IGSCC in sensitized stainless steels, is equally applicable for describing IGSCC in Alloy 600.

PWR Effect On Crack Initiation Under Equi-biaxial Loading

The lifetime extension of the nuclear power stations is considered as an energy challenge worldwide. That is why, the risk analysis and the study of various effects of different factors that could potentially prevent safe long term operation are necessary. These structures, often of great dimensions, are subjected during their life to complex loading combining varying mechanical loads, multiaxial, with nonzero mean values associated with temperature fluctuations and also PWR environment. Based on more recent fatigue data (including tests at 300°C in air and PWR environment, etc…), some international codes (RCC-M, ASME and others) have proposed and suggested a modification of the austenitic stainless steels fatigue curve combined with a calculation of an environmental penalty factor, namely Fen, which has to be multiplied by the usual fatigue usage factor. The aim of this paper is to present a new device "FABIME2E" developed in the LISN in collaboration with EDF and AREVA. These new tests allow quantifying the effect of PWR environment on disk specimen. This new device combines the structural effect like equi-biaxiality and mean strain and the environmental penalty effect with the use of PWR environment during the fatigue tests.

PWR EFFECT ON CRACK INITIATION UNDER EQUI-BIAXIAL LOADING: DEVELOPMENT OF THE EXPERIMENT

The lifetime extension of the nuclear power stations is considered as an energy challenge worldwide. That is why, the risk analysis and the study of various effects of different factors that could potentially represent a hazard to a safe long term operation are necessary. These structures, often of great dimensions, are subjected during their life to complex loading combining varying mechanical loads, multiaxial, with non-zero mean values associated with temperature fluctuations and also PWR environment.Historically, the methodology for fatigue dimensions of the Pressurized Water Reactor components (PWR) (ASME, RCC-M, KTA, …) is based on the use of design curves established from tests carried out in air at 20°C on smooth specimens by integrating safety coefficients that cover, among other parameters, the dispersion of tests associated with the effects of structures.Based on more recent fatigue data (including tests at 300°C in air and PWR environment, etc…), some international codes (RCC-M, ASME and others) have proposed and suggested a modification of the austenitic stainless steels fatigue curve combined with a calculation of an environmental penalty factor, namely Fen, which has to be multiplied by the usual fatigue usage factor.The aim of this paper is to present a new device "FABIME2E" developed in the LISN in collaboration with EDF and AREVA. These new tests allow quantifying accurately the effect of PWR environment on semi-structure specimen. This new device combines the structural effect like equi-biaxiality and mean strain and the environmental penalty effect with the use of PWR environment during the fatigue tests.

The increase in fatigue crack growth rates observed for Zircaloy-4 in a PWR environment

Cyclic stresses produced during the operation of nuclear reactors can result in the extension of cracks by processes of fatigue. Although fatigue crack growth rate (FCGR) data for Zircaloy-4 in air are available, little testing has been performed in a PWR primary water environment. Test programs have been performed by Gee et al., in 1989 and Picker and Pickles in 1984 by the UK Atomic Energy Authority, and by Wisner et al., in 1994, that have shown an enhancement in FCGR for Zircaloy-2 and Zircaloy-4 in high-temperature water. In this work, FCGR testing is performed on Zircaloy-4 in a PWR environment in the hydrided and non-hydrided condition over a range of stress-intensity. Measurements of crack extension are performed using a direct current potential drop (DCPD) method. The cyclic rate in the PWR primary water environment is varied between 1 cycle per minute to 0.1 cycle per minute. Faster FCGR rates are observed in water in comparison to FCGR testing performed in air for the hydrided material. Hydrided and non-hydrided materials had similar FCGR values in air, but the non-hydrided material exhibited much lower rates of FCGR in a PWR primary water environment than for hydrided material. Hydrides are shown to exhibit an increased tendency for cracking or decohesion in a PWR primary water environment that results in an enhancement in FCGR values. The FCGR in the PWR primary water only increased slightly with decreasing cycle frequency in the range of 1 cycle per minute to 0.1 cycle per minute. Comparisons between the FCGR in water and air show the enhancement from the PWR environment is affected by the applied stress intensity.