Light Water Reactor Components Research Articles

Probabilistic fracture mechanics analyses of a reactor pressure vessel using the irradiation embrittlement evaluation based on the Bayesian nonparametric method

In the structural integrity assessment of an embrittled reactor pressure vessel (RPV) based on probabilistic fracture mechanics (PFM), it is crucial to consider both of the mean value and the uncertainty of the embrittlement prediction. Typically, the embrittlement prediction uncertainty is given by a normal distribution, and its standard deviation is determined from the residuals of the measured and predicted values of all data used to develop the embrittlement prediction method. Therefore, the same uncertainty is assumed regardless of neutron fluence and the available data. To overcome this issue, the Japan Atomic Energy Agency recently developed the embrittlement prediction method based on machine learning and Bayesian statistics, i.e., the Bayesian nonparametric (BNP) method, and introduced it to the PFM analysis code. The BNP method can predict the probability distribution according to data scarcity and provide more reasonable uncertainty because it estimates significant uncertainties where the scatter of actual measured data is large and the number of data is small. In this study, PFM analysis for a Japanese model RPV in a pressurized water reactor is conducted using the PFM analysis of structural components in aging light water reactors, where the embrittlement probability distribution calculated using the BNP method is used. Furthermore, the effects of different embrittlement prediction methods and uncertainties on the failure frequency of RPVs are investigated in the PFM analyses, and the results are presented.

Irradiation of ultrasonic sensors and adhesive couplants for application in light water reactor primary loop piping and components

The Electric Power Research Institute (EPRI) Nuclear Sector and US Department of Energy Light Water Reactor Sustainability Program are committed to engaging in research and development endeavors to address materials aging issues specific to long term operation of light water power reactors. To this effect, EPRI launched an industry initiative to develop nondestructive evaluation systems for online monitoring of existing cracks in light water reactor primary coolant loop piping and components. One of the goals of this initiative is to develop a sensor system (or systems) that can determine nondestructively if cracks are growing or arrested and, in the case of the former, to characterize their growth rates. A missing component of this initiative is an experimental assessment of how sensors and adhesive couplants will perform in service when exposed to chronic energetic neutron radiation, particularly at the primary coolant loop hot and cold leg dissimilar metal welds, which join the primary loop piping to the reactor pressure vessel and reside in the vicinity of the reactor core. The objective of this experimental study was to determine how ultrasonic transducers and adhesive couplants perform when exposed to irradiation in a test reactor to simulate and accelerate in-service exposure. To achieve this objective, the signal stability of piezoelectric transducers and performance of adhesive couplants as a function of accumulated fast neutron fluence were characterized by collecting ultrasonic data in-situ during irradiation. Of particular interest were the ultrasonic signal quality and time decay of the amplitude of acoustic reflections as a function of fast neutron fluence. The results of the study showed that, of the 8 transducer/substrate sample assemblies tested, only 3 generated usable ultrasonic signals through the conclusion of the irradiation campaign. It was found that high temperature epoxy tends to ultrasonically couple the sensors to the substrates better than three types of refractory ceramic cements studied, as is supported by post irradiation examination. The results obtained through this experimental study will be utilized in the achievement of the overall goal of development of a sensor system to perform online monitoring of primary loop components.This manuscript has been authored in part by UT-Battelle, LLC, under contract DE AC05-00OR22725 with the US Department of Energy (DOE). The publisher acknowledges the US government license to provide public access under the DOE Public Access Plan (https://energy.gov/downloads/doe-public-access-plan [energy.gov]).

Investigation of surface treatment effects on the environmentally-assisted cracking behaviour of Alloy 182 in boiling water reactor environment

ABSTRACT Surface treatments of primary circuit components in light water reactors are regarded as possible ways to mitigate environmentally-assisted cracking (EAC). To date, it is not fully conclusive which surface condition is suitable to reduce the EAC initiation susceptibility. Constant extension rate tensile (CERT) tests were performed by several labs using flat tapered tensile specimens with different surface conditions (ground, industrial face milled, advanced face milled and shot peened), exposed to a boiling water reactor normal water chemistry environment at 288°C. Despite some scatter in the results, the CERT tests revealed that the EAC initiation susceptibility seems lowest for the advanced face milled surface and highest for the shot peened surface. However, it must be emphasised that the differences were moderate and that the surprising behaviour of the shot peened surface can be explained. The mechanical grinding of the surface did not significantly retard EAC initiation compared to industrial face milling.

Assessing the effects of switching from multi-channel to three-dimensional nodalisations of the core in TRACE

Modern thermal-hydraulics (TH) systems analysis codes are increasingly being updated to include three-dimensional (3D) TH components towards capturing the complex flow phenomena in large components of light water reactors (LWRs). These are becoming increasingly popular for modelling reactor pressure vessels and cores since they potentially offer a more realistic representation of, for example, the effects of asymmetric loop flows and injection, which cannot be captured using traditional multi-channel models. Experience suggests, however, there may be unexpected side effects of upgrading from traditional multi-channel models. In this paper, we present comparative results demonstrating our experiences in switching to 3D components using the US-NRC system code TRACE. We present a series of studies in which 1D and 3D nodalisations are compared for representative PWR loss-of-coolant-accident (LOCA) scenarios. We see a clear trend for many of the LOCA scenarios considered. In particular, the peak cladding temperature (PCT) for 3D nodalisations is significantly lower than for traditional multi-channel models in many cases. Counter current flow limitation (CCFL) at the core upper tie plate appears to be one of the driving phenomena for many of the observed differences. This trend is, however, not universal; For a surge line break in the large scale test facility (LSTF), heterogeneities in the core upper tie plate lead to hot regions in the core, which are not captured in the multi-channel model. For large PWR cores, 3D nodalisations appear to represent the reality more closely. For smaller cross-sections, however, the validity of the TRACE’s CCFL prediction for 3D nodalisations is questionable. Specific guidelines are therefore needed to help TRACE users select the most appropriate 3D nodalisation for different geometries.

Probing the Radial Chemistry of Getter Components in Light Water Reactors via Controlled Electrochemical Dissolution.

Getters are among the key functional components in the tritium-producing burnable absorber rods (TPBARs) of light water reactors (LWRs) and are used to capture the released tritium gas. They are nickel-plated zircaloy-4 tubes that, upon exposure to irradiation or tritium in the light water reactors, undergo alteration in structure, chemical composition, and chemistry. Understanding the radial tritium distribution is key to gaining insight into the evolution of new chemistry upon irradiation to predict getter performance. The holy grail is to develop a method akin to selectively peeling off the layers of an onion in an effort to get a radial map of elements and particularly tritium across the getter. Toward this goal, the overall aim of this work is to establish a correlative technique that can be used to determine radial tritium distribution across getters. To this end, this work specifically focuses on the validation of a correlative method for controlled radial dissolution of nickel-plated getters. Here, pristine getters as well as getters loaded with different mass ratios of hydrogen and deuterium are used as the nonradioactive surrogates of tritium, the idea being that the methodology can be readily extended to tritiated getter components. Here, the surface nickel layers as well as the bulk zirconium layers are sequentially dissolved in a controlled, uniform way using voltage-assisted electrochemical dissolution techniques. The dissolution is complemented by periodic elemental analysis of the electrolyte solution during and post dissolution. This is complemented by microscopic analyses on the exposed surfaces to provide a correlative technique for a complete picture of the radial distribution of various elements across the getter

An electrochemical technique for controlled dissolution of zirconium based components of light water reactors

Zircaloy-4 (Zr-4) based liners and getters are the principle functional components of Tritium-Producing Burnable Absorber Rods (TPBARs) in light water nuclear reactors where they reduce tritiated water into tritium gas. Upon tritium exposure, zirconium tritide is formed, which changes the chemical composition, structure and morphology of these materials. Their thermodynamic properties are affected by (i) the hydride phase identity, (ii) radial and spatial tritide/hydride (T/H) distribution, and (iii) the changes in structure and morphology of the material upon T/H-migration, and their comprehensive knowledge is needed to predict performance of these materials. This work demonstrates that controlled potential electrochemistry techniques to be highly efficient for controlled oxidative radial dissolution of Zr-4 based liners (both unloaded and loaded with hydride/deuteride as chemical surrogates for tritium). The electrodissolution is further combined with microscopic techniques to accurately determine the distribution of hydride phases. This work demonstrates a reliable technique for radially etching the liners after irradiation to provide insight into the radial and spatial distribution of tritium within the TPBAR, improving the fundamental understanding of tritium transport and providing a basis for validating predictive models.

Contribution of research to the understanding and mitigation of environmentally assisted cracking in structural components of light water reactors

ABSTRACTEnvironmentally assisted cracking (EAC) is a potential threat to the safety and integrity of water-wetted components in operating water-cooled nuclear power plants. Two forms of EAC are commonly distinguished, depending on the form of loading contributing to damage: stress corrosion cracking (SCC) and corrosion fatigue. A number of instances of in-service degradation due to EAC have occurred in operating plants worldwide, often leading to unplanned plant outages. Understanding the causes of EAC is essential to minimise the loss of plant availability due to its occurrence and to avoid the possibility of catastrophic failure, for example, if a crack grew to a critical size in a major pressure boundary component. This paper will describe some examples of these phenomena in the main materials of construction of pressure boundary and other critical components in pressurised and boiling water reactors (BWRs). Over the last several decades, substantial research programmes have been carried out in a number of laboratories worldwide, aimed at further understanding of the processes leading to EAC to manage occurrences in plant and minimise future failures. Selected areas of research on EAC in light water reactor environments are discussed. Corrosion fatigue in low-alloy pressure vessel steels was the subject of considerable attention in the 1980s and early 1990s because of its potential threat to pressure vessel integrity and the publication of data, suggesting that there is a major influence of environment on fatigue crack growth in some laboratory tests. The author’s research provided insight into the conditions under which the major environmental effects occur and contributed to the development of an ASME Code Case for pressurised water reactor (PWR) conditions which provided a means of screening based on steel sulphur content and loading conditions. More recently, the research focus in this area has moved to austenitic stainless steels, again providing support to Code Case development and furthering mechanistic understanding. A recent review of knowledge gaps for EPRI provides a basis for future research on environmentally assisted fatigue and will inform the development of new assessment methodologies. A key area of the current study concerns differences in loading conditions between specimens in laboratory tests and plant components subject to transient loading. In the case of SCC, stainless steels have shown the greatest propensity to cracking in BWRs, while Alloy 600 has been a major cause of in-service failures in PWRs, both on the primary side, as recognised by Coriou in the early 1960s, and in secondary environments where a number of different corrosion-related failure processes have been identified. High-strength alloys, such as Alloy X-750 used for fastener applications, have also caused failures in both reactor types. For austenitic materials, SCC susceptibility is enhanced by irradiation, resulting in failures in core internals components. Ferritic stainless steels also undergo SCC under some specific circumstances but are generally more resistant than the lower chromium austenitic materials.

Modeling the Cycle-Dependent Material Hardening Behavior of 508 low Alloy Steel

Chaboche-based material hardening models were used to investigate the cycle/time dependency of the elastic-plastic behavior of 508 low alloy steel (LAS), a commonly used material for light water reactor components. Strain-versus-stress curves were derived from uniaxial tensile and fatigue experiments conducted under various conditions, such as in-air at room temperature, in-air at 300 °C, and primary loop water conditions for a pressurized water reactor (PWR). These data were then used to estimate material parameters such as the elastic modulus, yield stress, and variables related to Chaboche-based linear and nonlinear kinematic hardening models. The results estimated from the fatigue test data suggest that the material behavior is highly sensitive to fatigue cycles, in addition to temperature. The authors anticipate that the findings of this research will help develop computer-based advanced modeling tools to extrapolate stress-strain evolution of reactor components under multi-axial stress states and, hence, will help predict their fatigue life more accurately.

Preliminary study for a PWR steam generator with CUPID/MARS multi-scale thermal-hydraulics simulation

For the thermal hydraulics analysis of nuclear reactor components of LWRs (Light Water Reactors) such as reactor vessel, steam generator, containment, a multi-dimensional two-phase flow code, named CUPID, has been being developed. The CUPID code pursues a capability of multi-physics and multi-scale thermal hydraulics analysis. In the present study, multi-scale simulation was performed by coupling with system-scale code, MARS. The coupled code was assessed to visualize the flow behavior of the steam generator of the Advanced Power Reactor (APR1400). The primary side of the steam generator and remaining Reactor Coolant System (RCS) is modeled by MARS and secondary side is by CUPID. For the secondary side simulation by the CUPID part, a porous media approach was adopted to two-fluid model and conductor model to simplify the complicated geometry of the steam generator. In order to obtain a porosity of a given computing cell, a special algorithm was employed to directly calculate volume ratio by mapping the 3D CAD file onto the grid system. Besides, the proper constitutive relationships for U-tubes are considered further. To treat the complex thermo-hydraulic phenomena on the shell side of a steam generator, a set of constitutive models available in the literature for a two-phase flow map, interfacial heat and mass transfer, interfacial drag, wall friction, wall heating, and heat partitioning in flows over tube bundles were applied to close the numerical model. This paper presents the description of the coupling method, porous media approach to simplify the steam generator, and the simulation results using the coupled code.

Small-Scale Mechanical Testing on Proton Beam-Irradiated 304 SS from Room Temperature to Reactor Operation Temperature

Austenitic stainless steels are common structural components in light water reactors. Because reactor components are subjected to harsh conditions such as high operating temperatures and neutron radiation, they can undergo irradiation-induced embrittlement and related failure, which compromises reliable operation. Small-scale mechanical testing has seen widespread use as a testing method for both ion- and reactor-irradiated materials because it allows access to the mechanical properties of the ion beam-irradiated region, and for safe handling of a small amount of activated material. In this study, nanoindentation and microcompression testing were performed on unirradiated and 10 dpa proton-irradiated 304 SS, from 25°C to 300°C. Increases in yield stress (YS), critical resolved shear stress (CRSS) and hardness (H) were seen in the irradiated region relative to the unirradiated region. Relationships between H, YS, and CRSS of irradiated and unirradiated materials are discussed over this temperature range.

ENVIRONMENTAL FATIGUE OF METALLIC MATERIALS IN NUCLEAR POWER PLANTS – A REVIEW OF KOREAN TEST PROGRAMS

Environmental fatigue of the metallic components in light water reactors has been the subject of extensive research and regulatory interest in Korea and abroad. Especially, it was one of the key domestic issues for the license renewal of operating reactors and licensing of advanced reactors during the early 2000s. To deal with the environmental fatigue issue domestically, a systematic test program has been initiated and is still underway. The materials tested were SA508 Gr.1a low alloy steels, 316LN stainless steels, cast stainless steels, and an Alloy 690 and 52M weld. Through tests and subsequent analysis, the mechanisms of reduced low cycle fatigue life have been investigated for those alloys. In addition, the effects of temperature, dissolved oxygen level, and dissolved hydrogen level on low cycle fatigue behaviors have been investigated. In this paper, the test results and key analysis results are briefly summarized. Finally, an on-going test program for hot-bending of 347 stainless steel is introduced.

Interatomic potential for studying ageing under irradiation in stainless steels: the FeNiCr model alloy

The degradation of austenitic stainless steels in a radiation environment is a known problem for the in-core components of nuclear light water reactors. For a better understanding of the prevailing mechanisms responsible for the materials' degradation, large-scale atomistic simulations are desirable. In this framework and as a follow-up on Bonny et al (2011 Modelling Simul. Mater. Sci. Eng. 19 085008), we developed an embedded atom method type interatomic potential for the ternary FeNiCr system to model the production and evolution of radiation defects. Special attention has been drawn to the Fe10Ni20Cr alloy, whose properties were ensured to be close to those of 316L austenitic stainless steels. The potential is extensively benchmarked against density functional theory calculations and the potential developed in our earlier work. As a first validation, the potential is used in AKMC simulations to simulate thermal annealing experiments in order to determine the self-diffusion coefficients of the components in FeNiCr alloys around the Fe10Ni20Cr composition. The results from these simulations are consistent with experiments, i.e., DCr > DNi > DFe.

Interatomic potential to study plasticity in stainless steels: the FeNiCr model alloy

Austenitic stainless steels are commonly used materials for in-core components of nuclear light water reactors. In service, such components are exposed to harsh conditions: intense neutron irradiation, mechanical and thermal stresses, and aggressive corrosion environment which all contribute to the components' degradation. For a better understanding of the prevailing mechanisms responsible for the materials degradation, large-scale atomistic simulations are desirable. In this framework we developed an embedded atom method type interatomic potential for the ternary FeNiCr system to model movement of dislocations and their interaction with radiation defects. Special attention has been drawn to the Fe–10Ni–20Cr alloy, whose properties were ensured to be close to those of 316L austenitic stainless steel. In particular, the stacking fault energy and elastic constants are well reproduced. The fcc phase for the Fe–10Ni–20Cr random alloy was proven to be stable in the temperature range 0–900 K and under shear strain up to 5%. For the same alloy the stable glide of screw dislocations and stability of Frank loops was confirmed.

Fracture assessment for a dissimilar metal weld of low alloy steel and Ni-base alloy

Recently, instances of SCC in Ni-base alloy weld metal of light water reactor components have been reported. Despite the possibility of propagation of SCC crack to the fusion line between low alloy steel (LAS) of pressure vessel and Ni-base alloy of internal structure, a fracture assessment method of dissimilar metal welded joint has not been established. The objective of this study is to investigate a fracture mode of dissimilar metal weld of LAS and Ni-base alloy for development of a fracture assessment method for dissimilar metal weld. Fracture tests were conducted using two types of dissimilar metal weld test plates with semi-elliptical surface crack. In one of the test plates, the fusion line lies around the surface points of the surface crack and the crack tips at the surface points have intruded into LAS. Material ahead of the crack tip at the deepest point is Ni-base alloy. In the other, the fusion line lies around the deepest point of the surface crack and the crack tip at the deepest point has intruded into LAS. Material ahead of the crack tip at the deepest point is LAS. The results of fracture tests using the former type of test plate reveal that the collapse load considering the proportion of ligament area of each material gives a good estimation for fracture load. That is, fracture assessment based on plastic collapse mode is applicable to the former type of test plate. It is also understood that a fracture assessment method based on the elastic-plastic fracture mode is suitable for the latter type of test plate.

Ductile Fracture Simulation of Pipe under Bending with Initial Defects

Crack propagating evaluation is needed to predict and prevent structural damages since many structure defects from numerous crack propagating. It is widely known that ductile fracture occurs when external load is exerted to the material, these load include strong and unpredictable load such as earthquakes and collision of objects. Ductile material fractures via nucleation of void, growth of void and coalescence of voids. Many studies have been conducted; Kikuchi and Sannoumaru have published papers on the studies of ductile fracture. In the first paper [1], the thickness effect on the microscopic fracture process and fracture toughness is studied experimentally. In the second paper [2], dimple fracture tests were performed using three point bending specimens. In the test, loading condition is changed from mode I to mixed mode condition to study the effect of the mixed mode loading. Numerical simulation is conducted using Gurson’s constitutive equation. It is found that crack growth direction is affected significantly by the loading condition. Ductile fracture of a pipe used in Light Water Reactor components is researched in this paper. Four point bending of a pipe experiment had been done by the Central Research Institute of Electric Power Industry [3]. They were experimented in two conditions; one at room temperature (23 ) and second at high temperature (300 ). As a result, crack propagation behavior differs largely form each other. At room temperature, crack propagates parallel from the pre crack, and at high temperature, crack propagates in a slanting direction from the pre crack. Results show that that difference from the two temperature distinction of a tensile test in a stress strain curve is very little. In this paper, this problem is studied at first by experiments, observation of fracture surface and numerical simulation.

Stress Intensity Effect on Solid State Oxidation of Ni-Cr Alloy with Different Chromium Concentrates

The life extension of aging nuclear power plant is a challenge due to the influence of various factors on the materials properties. Ni-base alloys are widely used in light water reactor component and stress corrosion cracking (SCC) has been found in pressurized water reactor (PWR) components such as vessel head penetrations and bottom mounted instrument nozzles fabricated with Alloy 600 and its weld metals [1,2]. Ni-base alloys composed of several elements among them nickel and chromium are the major ingredients. X-ray photoelectron spectroscopy (XPS) study of Ni-Cr alloy examined that Cr is oxidized preferentially to Cr2O3 initially but that Ni in the alloy is oxidized more rapidly to NiO than Ni metal [3]. Alloy 600 XPS study found the chromite-rich inner layer due to inward oxygen penetration and simultaneously with outward migration of nickel [4]. However, some theoretical and experimental studies have been carried out on Ni/H2O and oxygen preadsorbed Ni/H2O to understand water dissociation and adsorption phenomena [5,6].Over the years various attempts have been made to obtain mechanism of SCC but it still require more fundamental study to interpret clearly. This study presents an approch based on the multiscale modeling, to assess the influence of alloy composition and stress intensity (K) on the initial stage solid state oxidation of the Ni-Cr binary alloy. The multiscale modeling considers different length scales such as finite element method (FEM) – quasicontinuum (QC)-quantum chemical molecular dynamics (QCMD), for analyzing crack tip molecular domain.

Surface Deformation Nanostructures and Stress Corrosion Crack Precursors Originating from Surface Grinding

AbstractStress corrosion cracking (SCC) in light water reactor components has long been studied from a post mortem perspective, yielding insights into water chemistries and effects on crack propagation. Analysis of a cracked component does not effectively provide information on the corrosion events or on SCC initiation. It is important that microstructures of these early stages be understood because the original surface of the component formed during fabrication is often not the final surface condition that is exposed to reactor water. Pre-service grinding of reactor components and welds is performed for a variety of reasons, from aesthetics to preparation for non-destructive testing. It is this final surface microstructure that often controls SCC initiation. Surface and near-surface characteristics have been investigated in 304SS metal coupons on which controlled grinding was performed. These examinations indicate the extent of subsurface microstructural damage before high-temperature water exposure. Analytical electron microscopy techniques have been used to gain insights into possible surface precursors to corrosion damage and SCC initiation. Nanocrystalline grains were commonly found at the surface in lightly ground to heavily abraded materials within the first ˜0.5-10 ųm along with high dislocation densities, twinning and lath structures.

Erratum to: Microstructural and Stress Corrosion Cracking Characteristics of Austenitic Stainless Steels Containing Silicon

Austenitic stainless steels (SSs) core internal components in nuclear light water reactors (LWRs) are susceptible to irradiation-assisted stress corrosion cracking (IASCC). One of the effects of irradiation is the hardening of the SS and a change in the dislocation distribution in the alloy. Irradiation may also alter the local chemistry of the austenitic alloys; for example, silicon may segregate and chromium may deplete at the grain boundaries. The segregation or depletion phenomena at near-grain boundaries may enhance the susceptibility of these alloys to environmentally assisted cracking (EAC). The objective of the present work was to perform laboratory tests in order to better understand the role of Si in the microstructure, properties, electrochemical behavior, and susceptibility to EAC of austenitic SSs. Type 304 SS can dissolve up to 2 pct Si in the bulk while maintaining a single austenite microstructure. Stainless steels containing 12 pct Cr can dissolve up to 5 pct bulk Si while maintaining an austenite structure. The crack growth rate (CGR) results are not conclusive about the effect of the bulk concentration of Si on the EAC behavior of SSs.

Microstructural and Stress Corrosion Cracking Characteristics of Austenitic Stainless Steels Containing Silicon

Austenitic stainless steels (SSs) core internal components in nuclear light water reactors (LWRs) are susceptible to irradiation-assisted stress corrosion cracking (IASCC). One of the effects of irradiation is the hardening of the SS and a change in the dislocation distribution in the alloy. Irradiation may also alter the local chemistry of the austenitic alloys; for example, silicon may segregate and chromium may deplete at the grain boundaries. The segregation or depletion phenomena at near-grain boundaries may enhance the susceptibility of these alloys to environmentally assisted cracking (EAC). The objective of the present work was to perform laboratory tests in order to better understand the role of Si in the microstructure, properties, electrochemical behavior, and susceptibility to EAC of austenitic SSs. Type 304 SS can dissolve up to 2 pct Si in the bulk while maintaining a single austenite microstructure. Stainless steels containing 12 pct Cr can dissolve up to 5 pct bulk Si while maintaining an austenite structure. The crack growth rate (CGR) results are not conclusive about the effect of the bulk concentration of Si on the EAC behavior of SSs.

Quantitative Estimation of the Growth of Environmentally Assisted Cracks at Flaws in Light Water Reactor Components

Since environmentally assisted cracking (EAC) is an important degradation mechanism affecting the structural materials of nuclear power plants, numerous EAC experiments have been performed in the past three decades using standard specimens in simulated high temperature water environments to evaluate the various core materials used in light water reactors (LWRs). However, the environment, the condition of the material, and the mechanical properties near flaws in LWR components are not absolutely equivalent to those near the crack tip in standard specimens; thus, more research needs to be done before EAC growth in an actual LWR component can be accurately estimated using existing experimental EAC data. By combining the film slip-dissolution/oxidation model with the elastic-plastic finite element method and existing experimental EAC data, we have derived a method by which an estimation of EAC growth at flaws in actual LWR components can be made. In this paper we propose and discuss the use of this method. The results show that this new method basically concurs with the Fracture Research Institute (FRI) model in evaluating EAC growth across a semi-elliptic crack front under a simple tensile load and is also in approximate agreement with the experimental results obtained in evaluating EAC growth along a semi-elliptic crack front under complex loading conditions. The approach is expected to form a bridge between predicting EAC growth rate in core materials and evaluating EAC growth in key structural components in LWRs, and it is also expected that it can be used as a pre-analytical tool for EAC experiments using nonstandard specimens.