Light Water Reactor Environments Research Articles

Environmentally-assisted cracking of electropolished 316L stainless steel in molten FLiNaK salt

Environmentally-assisted cracking is a well-known and extensively studied phenomenon in light water reactor environments but remains relatively unexplored in proposed molten salt reactor systems where the mechanisms at play are expected to be quite different. Here, slow strain rate testing has been performed on 316L stainless steel tensile specimens during simultaneous exposure to a molten LiF-KF-NaF eutectic salt environment at 600°C with a strain rate between 1E-5 to 1E-7 s-1. Sample ductility and ultimate tensile strength were both observed to decrease for slower strain rates/longer salt exposures. Post-mortem characterization of fractured specimens revealed Fe deposits and Cr- and O-rich precipitates decorating surface grain boundaries and crack surfaces, suggesting a cracking mechanism driven by embrittlement of the crack tip due to oxide formation. Unlike previous studies, intergranular Cr corrosion into the bulk was not observed, and distinct differences in surface morphology between electropolished and non-electropolished sample surfaces are reported. Trace metal salt chemistry measurements showed accumulation of 316L alloying elements over the course of the tests. It appears likely that the surface oxide layer induced by electropolishing, in addition to oxygen and moisture impurities in the salt, led to the formation of LiCrO2 which had a net passivating effect that prevented significant bulk corrosion, though some interplay between mechanical and environmental stressors is still observed.

FeCrAl fuel/clad chemical interaction in light water reactor environments

This article investigates the fuel-cladding chemical interaction (FCCI) behavior of two commercial FeCrAl alloys, APMT composition (Fe-21Cr-5Al-3Mo wt.%) and C35M (Fe-13Cr-5Al-2Mo-0.2Si-0.03Y wt.%), after neutron irradiation. “H-cup” diffusion multiples of FeCrAl alloys and ceramic UO2 fuel were irradiated at a temperature of ∼300 °C to a total estimated burnup of 26 GWd/tHM. Post-irradiation Examination results demonstrate the excellent degradation resistance of FeCrAl alloys as accident tolerant fuel (ATF) cladding materials in light water reactor conditions. The study concludes that there was no irradiation-induced defects observed in either of the two commercial FeCrAl claddings. The formation of amorphous Al/U mixed oxide was observed at the fuel-clad interface, which can serve as a tritium permeation barrier and protect against potential chemical attack from the fuel. The study attributed the formation of amorphous Al/U mixed oxide to the low temperature and limited time of neutron irradiation. APMT forms more distinct Cr and Cr-Fe intermetallic at the FeCrAl-UO2 interface than C35M due to the higher bulk Cr:Al ratio.

Role of Zn injection on mitigating stress corrosion cracking initiation of Alloy 182 weld metal in simulated light water reactor environment

The stress corrosion cracking (SCC) behaviour of nickel-based weld metal Alloy 182 was studied in simulated light water reactor environments with or without zinc (Zn) addition. Accelerated SCC initiation tests with flat tapered tensile specimens, as well as SCC initiation and short crack growth tests with compact tension specimens revealed that Zn injection is able to effectively mitigate SCC. Even the lowest tested Zn concentration of 15 ppb seems to be enough to achieve a significant mitigation effect on SCC initiation and short crack growth. This beneficial effect may be related to the modified surface oxide film by Zn incorporation.

Effect of surface machining on the environmentally-assisted cracking of Alloy 182 and 316L stainless steel in light water reactor environments: results of the collaborative project MEACTOS

Abstract The main objective of the EU-funded project mitigating environmentally-assisted cracking through optimisation of surface condition (MEACTOS) was to gain knowledge on the ability of different surface machining procedures to mitigate environmentally-assisted cracking (EAC) in typical light water reactor structural materials and environments. Surfaces of cold-worked (CW) type 316L austenitic stainless steel and nickel-based weld metal Alloy 182 flat tapered tensile specimens were machined using different processes. EAC initiation susceptibility of these specimens was evaluated using constant extension rate tensile (CERT) tests under simulated boiling water reactor (BWR) and pressurized water reactor (PWR) conditions and assessed using constant load experiments. More than a hundred tests were performed covering about 10 years of autoclave testing time. Only minor or no measurable improvements in EAC initiation susceptibility as a function of surface treatments (grinding or advanced machining) compared to the standard industrial face milling were demonstrated. In most cases, the stress thresholds for EAC initiation determined in constant load tests confirmed the trend obtained from CERT tests. This paper summarises the most important results and conclusions concerning the EAC initiation behaviour for the CW 316L and Alloy 182 under reducing PWR and oxidizing BWR conditions.

Compositionally graded specimen made by laser additive manufacturing as a high-throughput method to study radiation damages and irradiation-assisted stress corrosion cracking

This study demonstrates the feasibility of using compositionally gradient specimens, fabricated by laser additive manufacturing (AM) and post-AM thermo-mechanical treatment, to accelerate alloy synthesis, radiation experiment, and the assessment of irradiation properties in light water reactor environments. The effects of minor Hafnium (Hf) doping in austenitic 316L stainless steel (SS) was selected as the topic of interest. By comparing to the data in literature, we confirmed that the compositionally graded specimen produces the same trend of void swelling, dislocation loops, radiation-induced segregation (RIS), radiation hardening as the wrought specimen produced by cast/forging process. Hf suppressed most radiation damages through strong interaction with point defects. The work also demonstrates the use of compositionally gradient specimens to study the irradiation-assisted stress corrosion cracking (IASCC) susceptibility of Hf-modified SS. While the suppression of radiation hardening and RIS are consistent with the IASCC mitigation by Hf, we emphasize Hf can alter the intrinsic deformation behavior of 316L SS, which reduces grain-boundary strain localization. The advantages and challenges of using compositionally gradient design for high-throughput nuclear alloy development and qualification are also discussed.

Gauge-Strain-Controlled Air and PWR Fatigue Life Data for 304 Stainless Steel—Some Effects of Surface Finish and Hold Time

We performed environmental fatigue testing in simulated primary water reactor (PWR) primary water and reference fatigue testing in air in the framework of an international, collaborative project (INCEFA-PLUS), where the effects of mean strain and stress, hold time, strain amplitude and surface finish on fatigue life of austenitic stainless steels in light water reactor environments are being studied. Our fatigue lives obtained on machined specimens in air at 300 °C lie close to the NUREG/CR6909 mean air fatigue curve and are in line with INCEFA-PLUS air fatigue lives. Our environmental fatigue lives obtained in simulated PWR primary water at 300 °C lie relatively close to the NUREG/CR6909 mean fatigue curve; derived from the NUREG/CR6909 mean air fatigue curve and the applicable environmental correction factor (Fen). The PWR results show that (1) a polished surface finish has a slightly higher and a ground surface finish a slightly lower fatigue life than the NUREG/CR6909 prediction; (2) the ratio of polished to ground specimen life is ~1.37 at 300 °C and ~1.47 at 230 °C; (3) holds—at zero strain after a positive strain-rate—have a slightly detrimental effect on fatigue life. These results are in line with the INCEFA-PLUS PWR fatigue lives. A novel gauge-strain extensometer was deployed in order to perform a true gauge-strain-controlled fatigue test in simulated PWR primary water.

Technical Basis Part 2 for Code Case N-889: Reference Stress Corrosion Crack Growth Rate Curves for Irradiated Austenitic Stainless Steels in Light Water Reactor Environments

Abstract A multiyear international data collection, data review, modeling, and implementation project was recently completed, producing stress corrosion crack growth rate (CGR) reference curves for irradiated austenitic stainless steels in light water reactor (LWR) environments that were adopted as ASME B&PV Section XI Code Case N-889. As described in a technical basis Part 1 paper, over 800 CGR data points were collected from six laboratories worldwide, an international expert panel reviewed and ranked the data, and the better-ranked data were used to calibrate empirical models for irradiation-assisted stress corrosion cracking (IASCC) CGR in boiling water reactor (BWR) normal water chemistry (NWC), BWR hydrogen water chemistry (HWC) and pressurized water reactor (PWR) primary water environments. Part 1 also describes the custom fitting process, quality of fit, and comparisons with related literature and data not used for fitting. This technical basis Part 2 paper describes shifting the mean models to the 75th percentile of the calibration data, simplifying to produce the N-889 curves, and comparing with previous reference curves and over 500 data points not used for developing the N-889 curves, including weld, cast, and heat-affected-zone (HAZ) materials, additional wrought laboratory data, and field data from repeated inspection of BWR core shrouds. Part 2 also describes the irradiated yield stress model in Case N-889, compares that model with its calibration data and other data not used for calibration, and presents example calculations using both yield stress and CGR equations.

Technical Basis Part 1 for Code Case N-889: Data Selection and Modeling of Stress Corrosion Crack Growth Rates for Irradiated Austenitic Stainless Steels in Light Water Reactor Environments

Abstract A multiyear international data collection, data review, modeling, and implementation project was recently completed, producing stress corrosion crack growth rate (CGR) models and reference curves for irradiated austenitic stainless steels in light water reactor (LWR) environments. The curves have been approved as ASME B&PV Section XI Code Case N-889, and this paper is the technical basis Part 1 for Case N-889, summarizing collection, review, and selection of calibration data, modeling, and analysis of fit. Over 800 CGR data points were collected from six laboratories worldwide, an international expert panel reviewed and ranked the data, and the better-ranked data were used to calibrate empirical models for irradiation-assisted stress corrosion cracking (IASCC) CGR in boiling water reactor (BWR) normal water chemistry (NWC), BWR hydrogen water chemistry (HWC), and pressurized water reactor (PWR) primary water environments. This Part 1 paper also discusses related literature and compares with other data not used for mean model calibration. A technical basis Part 2 paper describes shifting the mean models to the 75th percentile, simplifying to produce Case N-889 reference curves, and comparing with previous reference curves and over 500 data points not used for reference curve development, including weld, cast, and heat affected zone (HAZ) materials, additional wrought laboratory data, and field data from repeated inspection of BWR core shrouds. Part 2 also describes the irradiated yield stress model in Case N-889, compares that model with its calibration data and other data not used for calibration, and presents example calculations.

Stress corrosion crack initiation testing with tapered specimens in high-temperature water – results of a collaborative research project

ABSTRACT The applicability of an accelerated test technique using tapered tensile specimens for investigating the stress corrosion crack (SCC) initiation behaviour of structural materials in high-temperature water was assessed in the framework of a European collaborative research project (MICRIN – MItigation of CRack INitiation). The main advantage of using a tapered geometry is, that in a single test a stress gradient is obtained through the gauge length, and therefore a stress threshold for SCC initiation can be determined in a reasonable timeframe. This method was used to investigate two different materials that were known to be susceptible to SCC in light water reactor environment: a high-Si stainless steel and a Ni-base weld metal (Alloy 182). The results of the international test programme confirmed that the tapered specimen test methodology could be used to identify a SCC initiation stress threshold, albeit that significant scatter was present in the data.

Laser glazing of cold sprayed coatings for the mitigation of stress corrosion cracking in light water reactor (LWR) applications

The effect of laser power, traverse speed, and cold spray coating thickness were examined with the goal of mitigating and repairing stress corrosion cracking (SCC) in light water reactor (LWR) environment. For this purpose, SCC-susceptible Alloy 600 substrate material was coated with SCC-resistant Alloy 690 via the cold spray technique. The cold spray coated substrate was then laser-glazed using various laser parameters. Single pass and multiple pass laser glazed regions were created to determine the effects. For this purpose, the area of the fusion zone, depth of the fusion zone, and elemental composition of the cross section were examined by optical microscopy, scanning electron microscopy, and energy dispersive spectroscopy. It was shown that chromium dilution from the laser glazed region into the substrate is not a significant factor in laser glazed cold spray coated samples. Separate laser glazed samples were subjected to ASTM 633C to determine coating adhesion. Finally, by using interrupted SCC crack growth rate tests, it was shown that low traverse speeds and high laser powers produce deep fusion zones within the cold sprayed coating and substrate that can be used to seal pre-existing cracks, thus stopping SCC growth.

Effect of deformation level and orientation on SCC of 316L stainless steel in simulated light water environments

The stress corrosion cracking (SCC) behavior of 316L stainless steel (SS) forged to 0%, 12.6%, 21.4%, and 39.8% reduction in thickness was investigated at constant K in light water reactor environments. The yield strength, specimen orientation, and water chemistry were correlated with crack growth rate, and their dependencies are discussed. The crack growth rate (CGR) of 316L SS increased monotonously with yield strength irrespective of the specimen orientation or water chemistry. Higher CGRs were observed when cracks propagate along the plane parallel to forging plane than normal to forging plane. The effect of local deformation on the anisotropic cracking behavior for different orientations, crack paths and CGRs are also discussed.

Environmental degradation of fracture resistance in high-temperature water environments of low-alloy reactor pressure vessel steels with high sulphur or phosphorus contents

The fracture behaviour of low-alloy reactor pressure vessel steels with various sulphur and phosphorus contents, different environmentally-assisted cracking (EAC) and temper embrittlement (TE) susceptibilities was evaluated by elastic plastic fracture mechanics tests in air and various simulated light water reactors environments. A moderate but clear reduction of fracture initiation resistance occurred in a) high-sulphur steel with high EAC susceptibility in oxygenated high-temperature water with aggressive occluded crevice environment and preceding EAC growth, and in b) high-phosphorus steel with high TE susceptibility, where the reduction of fracture resistance was most pronounced in hydrogenated high-temperature water.

Effect of dynamic strain ageing on environmental degradation of fracture resistance of low-alloy RPV steels in high-temperature water environments

Elastic plastic fracture mechanics tests on different low-alloy reactor pressure vessel (RPV) steels in air and various simulated light water reactor environments revealed a clear, but moderate reduction of the fracture initiation resistance of RPV steels with high susceptibility to dynamic strain ageing (DSA) in high-temperature water (HTW). In HTW, the reduction of fracture initiation resistance increased with decreasing loading rate and was most pronounced in hydrogenated HTW. The stronger reduction of fracture resistance in hydrogenated HTW was mainly due to the localization of plastic deformation by interaction between DSA and hydrogen.

Study of Steel Corrosion Cracking Mechanisms for VVER RPV Internals Based on Simulation Tests

Different hypotheses of irradiation assisted stress corrosion cracking for materials of internals in the light water reactor environment were analyzed. Irradiation-induced grain boundary chromium depletion, localized deformation, and material hardening were considered. Heat treatment and cold working of 18Cr–10Ni–Ti steel were made to simulate the influence of neutron irradiation. Autoclave tests in the light water reactor environment at a temperature 290°C and at a strain rate of 3 × 10–7 s–1 were carried out. 18Cr–10Ni–Ti steel specimens in the initial state, after heat treatment, and after heat treatment followed by cold working were used. Fracture surfaces and lateral surfaces of specimens were examined by scanning electron microscopy. The microstructure of the material was also examined using transmission electron microscopy. It was shown that the grain boundary chromium depletion and the material hardening are not the main causes of corrosion cracking of 18Cr–10Ni–Ti steel in the primary coolant deoxygenated environment. The localization of deformation has a profound effect on corrosion cracking in the absence of the grain boundary sliding. The conditions determining the crack propagation by the corrosion cracking mechanism are defined.

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.

Influence of mean stress and light water reactor environment on fatigue life and dislocation microstructures of 316L austenitic steel

Influence of mean stress on fatigue life of the austenitic stainless steel 316 L in air and light water environments (boiling water reactor/hydrogen water chemistry) at 288 °C was determined with a series of tests carried out in load-control mode. Fatigue life was found to increase with application of compressive and tensile mean stress in air and light water reactor environments. Secondary hardening was regarded as the main reason for this behavior. A modified Smith-Watson-Topper (SWT) model was considered to account for mean stress and was shown to predict fatigue life accurately in air and water environments. The reduction of fatigue life in water environment, determined with the SWT curves, was about 2.5. Observations of the end-of-life dislocation arrangements by transmission electron microscopy showed that the dislocation microstructure depends essentially on plastic strain amplitude, which in turn is strongly correlated to stress amplitude and mean stress. The microstructures were found consistent with those usually observed after strain-controlled experiments. At rather low plastic strain amplitudes, corduroy structure consisting of small dislocation loops was observed. Acting as significant obstacle to dislocation motion, corduroy structure affects overall dislocation mobility therefore contributing to notable secondary cyclic hardening.

Fatigue behavior of 316L austenitic stainless steel in air and LWR environment with and without mean stress

The fatigue life design curves in nuclear codes are generally derived from uniaxial straincontrolled fatigue test results. Evidently, the test conditions are very different from the actual components loading context, which involves much more complex thermo-mechanical loading including mean stress, static load holding time and variation in water chemistry, etc. In this work, the mean stress and environmental effects on fatigue life of 316L austenitic stainless steel in air and light water reactor (LWR) environment were studied using hollow fatigue specimens and testing under load-controlled condition. Both positive (+50 MPa) and negative (-20 MPa) mean stresses showed beneficial effect on fatigue life in LWR environment and in air. This is tentatively attributed to mean stress enhanced cyclic hardening, which leads to smaller strain response at the same loading force. -20 MPa mean stress was found to increase fatigue limit, whereas the effect of +50 MPa mean stress on fatigue limit is still unclear. The preliminary results illustrate that the environmental reduction of fatigue life is amplified in load-controlled fatigue tests with tensile mean stress.

Main mechanical factors affecting creep strain at crack tip of stress corrosion cracking in full life cycle

Stress Corrosion Cracking (SCC) is a failure mechanism that is caused by environment, susceptible material and tensile strain at crack tip. The mechanical state at crack tip is one of the main factors affecting stress corrosion crack propagation rate in structural materials of nuclear power plants. To understand the effect of mechanical factor on creep strain on SCC crack growth rate in the light water reactor environments, the creep strain at crack tip in full life cycle is studied by elastic-plastic finite element method (EPFEM) in this paper. Study results indicate that it is suitable to use the creep strain in front of crack tip as a mechanical factor of SCC behaviors, and also show that wedging stress is the main mechanical factor affecting creep strain in micro crack stage, while external load are gradually becoming the main mechanical factor in long crack stage. Crack propagation rate is very slow in micro crack stage, and it will expand rapidly under the combined effect of residual stress and working load if there is an initial crack l.

INCEFA-PLUS (Increasing safety in NPPs by covering gaps in environmental fatigue assessment)

INCEFA-PLUS is a European Project (funded by the EC HORIZON2020 program) started in July 2015 and lasts five years (until June 2020). The project involves 16 members from all over Europe and its main goal is to guarantee safety in operations of Nuclear Power Plants (NPPs) by delivering experimental data to support the development of improved environmental fatigue assessment guidelines. The issues of common interest are being studied: the effects of mean strain/stress, hold time, strain amplitude and surface finish on the fatigue performance of austenitic stainless steels in the LWR (Light Water Reactor) environment. Within the framework of the CEN (European Committee for Standardization) workshop FATEDA, the consortium developed a fatigue data format, used in an online environmental fatigue database, available for all participants. The paper describes phase one (of three) testing, including conditions such as surface finish and hold time. The plans that are being made for phase two testing will be detailed. The materials used in the tests, the surface finish data and some experimental results for phase one are presented, together with a summary of the INCEFA-PLUS testing protocol. Finally, a review of existing assessment methodologies and a summary of dissemination activities are provided.

Environmental Degradation Effect of High-Temperature Water and Hydrogen on the Fracture Behavior of Low-Alloy Reactor Pressure Vessel Steels

Structural integrity of reactor pressure vessel (RPV) in light water reactors (LWR) is of highest importance regarding operation safety and lifetime. The fracture behaviour of low-alloy RPV steels with different dynamic strain aging (DSA) & environmental assisted cracking (EAC) susceptibilities in simulated LWR environments was evaluated by elastic plastic fracture mechanics tests (EPFM) and by metallo- and fractographic post-test analysis. Exposure to high temperature water (HTW) environments at LWR temperatures revealed only moderated reductions in the fracture initiation and tearing resistance of low alloy RPV steels with high DSA or EAC susceptibility, accompanied with a moderate, but clear change in fracture morphology, which indicates the potential synergies of hydrogen/HTW embrittlement with DSA and EAC under suitable conditions. The most pronounced degradation effects occurred in a) RPV steels with high DSA susceptibility, where the fracture initiation and tearing resistance reduction increased with decreasing loading rate and were most pronounced in hydrogenated HTW and b) high sulphur steels with high EAC susceptibility in aggressive occluded crevice environment and with preceding fast EAC crack growth in oxygenated HTW. The moderate effects are due to the low hydrogen availability in HTW together with high density of fine-dispersed hydrogen traps in RPV steels. Stable ductile transgranular tearing by microvoid coalescence was the dominant failure mechanism in all environments with additional varying few % of secondary cracks, macrovoids and quasi-cleavage in HTW. The observed behavior suggests a combination of plastic strain localisation by the Hydrogen-enhanced Local Plasticity (HELP) mechanism, in synergy with DSA, and Hydrogen-enhanced Strain-induced Vacancies (HESIV) mechanism with additional minor contributions of Hydrogen-enhanced Decohesion Embrittlement (HEDE) mechanism.