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Über dieses Buch

This two-volume set represents a collection of papers presented at the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors. The purpose of this conference series is to foster an exchange of ideas about problems and their remedies in water-cooled nuclear power plants of today and the future. Contributions cover problems facing nickel-based alloys, stainless steels, pressure vessel and piping steels, zirconium alloys, and other alloys in water environments of relevance. Components covered include pressure boundary components, reactor vessels and internals, steam generators, fuel cladding, irradiated components, fuel storage containers, and balance of plant components and systems.


Cables and Concrete Aging and Degradation–Cables

Simultaneous Thermal and Gamma Radiation Aging of Electrical Cable Polymers

Elevated temperature is the primary source of aging for nuclear power plant electrical cable insulation and jacketing, but gamma radiation is also a significant contributor to structural changes that result in loss of polymerPolymers mechanical and electrical properties in affected plant locations. Despite many years of research, the combined degradation effects of simultaneous exposure to thermal and radiation stresses are not well understood. As nuclearNuclear operators prepare for extended operation beyond initial license periods, a predictive understanding of exposure-based cable degradation is becoming an increasingly important input to safety, licensing, operations and economic decisions. We focus on carefully-controlled simultaneous thermal and gamma radiation aging and characterization of the most common nuclear cable polymers to understand relative contributions of temperature, time, dose and dose rate to changes in cable polymer material structure and properties. Improved understanding of cable performance in long term operation will help support continued sustainable nuclear power generation.

Principal Component Analysis (PCA) as a Statistical Tool for Identifying Key Indicators of Nuclear Power Plant Cable Insulation Degradation

This paper describes the use of Principal Component Analysis (PCA)Principal Component Analysis (PCA) as a statistical method to identify key indicators of degradation in nuclear power plant cable insulation. Seven kinds of single-point data and four kinds of spectral data were measured on cross-linked polyethylene (XLPE) that had undergone aging at various doses and dose rates of gamma radiation from a cobalt 60 source, and various elevated temperatures. To find the key indicators of degradation of aged cable insulation, PCA was used to reduce the dimensionality of the data set while retaining the variation present in the original data set. For example, PCA reveals that, for material aged at 90 °C, elastic modulus shows a positive correlation with total dose while mass loss, oxidation induction time and density show negative correlations with the same parameter.

How Can Material Characterization Support Cable Aging Management?

Low voltage (LV) power,Low voltage cables control and instrumentation cables are essential to the safe and reliable operations of nuclear power plants (NPPs). However, considering the huge consumption of cables in NPPs, it is impractical and cost prohibitive to replace cables when they reach the end of their design life. As a result, condition monitoringCondition monitoring and agingAging management of cables is critically important for the life extension of NPPs. The aging of LV cables is characterized by the degradation of the polymeric insulation and jacket materials leading to their mechanical failure or the loss of their ability to withstand critical conditions. This paper focuses on studying material characterization techniques (already existing and new) to monitor the change inPolymer characterization polymeric material propertiesMaterial properties and also to establish a general framework between material properties and ageing conditions that could assist in predicting the condition of cables and estimating their remaining life.

Aqueous Degradation in Harvested Medium Voltage Cables in Nuclear Power Plants

For medium voltage (MV) Medium voltage cablespower cables with voltages between 5 and 35 kV that provide power to emergency and safety support systems in nuclear power plants (NPPs), degradation and failure of cables due to water exposure has occurred. Systematic studies of current NPP MV cable systems are being carried out to determine and characterize cable-aging mechanisms and provide greater accuracy to models being developed to predict MV cable field performance to support existing cable aging management and monitoring programs. To validate models based on representative laboratory specimens, samples from harvested MV cable systems are being subjected to accelerated aging and degradation from humid conditions and submergence. Degradation will be characterized via partial discharge and voltage endurance testing of the cable and induced water tree growth in insulation. The technical approach to be used for the testing harvested MV cable samples is presented.

Frequency Domain Reflectometry Modeling and Measurement for Nondestructive Evaluation of Nuclear Power Plant Cables

Cable insulation polymers are among the more susceptible materials to age-related degradation within a nuclear power plant. This is recognized by both regulators and utilities, so all plants have developed cable aging management programs to detect damage before critical component failure in compliance with regulatory guidelines. Although a wide range of tools is available to evaluate cables and cable systems, cable aging management programs vary in how condition monitoring and nondestructive examination is conducted as utilities search for the most reliable and cost-effective ways to assess cable system condition. Frequency domain reflectometry (FDR)Frequency domain reflectometry is emerging as one valuable tool to locate and assess damaged portions of a cable system with minimal cost and only requires access in most cases to one of the cable terminal ends. This work examines a physics-based model of a cable system and relates it to FDR measurements for a better understanding of specific damage influences on defect detectability.

Aging Mechanisms and Nondestructive Aging Indicator of Filled Cross-Linked Polyethylene (XLPE) Exposed to Simultaneous Thermal and Gamma Radiation

Aging mechanismsAging mechanisms and a nondestructive aging indicator of filled cross-linked polyethylene (XLPE)Filled cross-linked polyethylene cable insulation material used in nuclear power plants (NPPs) are studied. Using various material characterization techniques, likely candidates and functions for the main additives in a commercial filled-XLPE insulation material have been identified. These include a mixture of brominated components such as decabromodiphenyl ether and Sb2O3 as flame retardants, ZnS as white pigment and polymerized 1,2-dihydro-2,2,4-trimethylquinoline as antioxidant. Gas chromatography-mass spectrometryGas chromatography-mass spectrometry, differential scanning calorimetryDifferential scanning calorimetry, oxidation induction timeOxidation induction time and measurements of dielectric loss tangentDielectric loss tangent are utilized to monitor property changes as a function of thermal and radiation exposure of the cable material. The level of antioxidant decreases with aging by volatilization and chemical reaction with free radicals. Thermal aging at 90 ℃ for 25 days or less causes no observable change to the cross-linked polymer structure. Gamma radiation causes damage to crystalline polymer regions and introduces defects. Dielectric loss tangent is shown to be an effective and reliable nondestructive indicator of the aging severity of the filled-XLPE insulation material.

Successful Detection of Insulation Degradation in Cables by Frequency Domain Reflectometry

We have succeeded in detecting the degradation of cable’sCablepolymeric insulationPolymeric insulation well before its continual use becomes risky. Degradation of organic polymers is mainly caused by oxidation if the ambience around the cable contains oxygen. When organic polymers are oxidized, polar carbonyl groups are formed, by which the permittivity is increased. This in turn decreases the characteristic impedanceCharacteristic impedance of a polymer-insulated cable. If we inject electromagnetic waves in a very wide frequency range into the cable and measure the ratio of reflected power to injected power, the information on the effects of the characteristic impedance changes is included in the frequency spectra of the ratio. If we do inverse Fourier transform, we can convert the data to a time domain. Therefore, we can know the degraded portion by multiplying the velocity of electromagnetic waves in the cable.

Capacitive Nondestructive Evaluation of Aged Cross-Linked Polyethylene (XLPE) Cable Insulation Material

Cross-linked polyethylene (XLPE) is used widely as insulation material in low-voltage instrumentation cables deployed in nuclear power plants (NPPs). Suffering from degradation due to exposure to heat and radiation, the insulating properties of the material gradually degrade, which may imperil the safe operation of NPPs. In this paper, a new capacitive sensing method is introduced for nondestructive evaluation of aged instrumentation cable with XLPE insulation. The capacitive sensor is capable of measuring capacitance of cable materials and may potentially be employed to extract the dielectric properties of insulation material from those of the entire cable. Interdigital capacitive sensor designs are targeted that maximize sensitivity to changes in the dielectric values of the cable polymers.

Tracking of Nuclear Cable Insulation Polymer Structural Changes Using the Gel Fraction and Uptake Factor Method

Cross-linked polyethylene (XLPE)XLPE cable insulation samples were exposed to heat and gamma radiation at a series of temperatures, dose rates, and exposure times to evaluate the effects of these variables on material degradation. The samples were tested using the solvent incubation method to collect gel fractionGel fraction and uptake factorUptake factor data in order to assess the crosslinkingCrosslinking and chain scission occurring in polymer samples with aging. Consistent with previous reports, gel fraction values were observed to increase and uptake factor values to decrease with radiation and thermal exposure. The trends seen were also more prominent as exposure time increased, suggesting this to be a viable method of tracking structural changes in the XLPE-insulated cable material over extended periods. For the conditions explored, the cable insulation material evaluated did not indicate signs of anomalous aging such as inverse temperature effectInverse temperature effect in which radiation-induced aging is more severe at lower temperature. Ongoing aging under identical radiation conditions and at lower temperature will further inform conclusions regarding the importance of inverse temperature effects for this material under these conditions.

Degradation of Silicone Rubber Analyzed by Instrumental Analyses and Dielectric Spectroscopy

Silicone rubber (SiR)SiR was gamma irradiated at 125, 145 and 185 °C or thermally aged at 220, 250 and 280 °C and the resultant changes in performance were evaluated. It has become clear from instrumental analyses that crosslinking via oxidation of silicon atoms and chain scission are induced by gamma raysGamma rays. Furthermore, from the temperature dependence of real relative permittivity at high frequencies, the thermal expansion coefficient was found to become smaller with the increase in dose. These results can be understood well by the chemical and structural changes in SiR induced by the degradationDegradation.

Cables and Concrete Aging and Degradation–Concrete

Automated Detection of Alkali-Silica Reaction in Concrete Using Linear Array Ultrasound Data

This paper documents the development of signal processing and machine learning techniques for the detection of Alkali-silica reaction (ASR). ASR is a chemical reaction in either concrete or mortar between hydroxyl ions of the alkalis from hydraulic cement, and certain siliceous minerals present in some aggregates. The reaction product, an alkali-silica gel, is hygroscopic having a tendency to absorb water and swell, which under certain circumstances, leads to abnormal expansion and cracking of the concrete. This phenomenon affects the durability and performance of concrete cause significant loss of mechanical properties. Developing reliable methods and tools that can evaluate the degree of the ASR damage in existing structures, so that informed decisions can be made toward mitigating ASR progression and damage, is important to the long-term operation of nuclear power plants especially if licenses are extended beyond 60 years. The paper examines the differences in the time-domain and frequency-domain signals of healthy and ASR-damaged specimens. More precisely, we explore the use of the Fast Fourier Transform to observe unique features of ASR damaged specimens and an automated method based on Neural Networks to determine the extent of ASR damage in laboratory concrete specimens.

Coupled Physics Simulation of Expansive Reactions in Concrete with the Grizzly Code

The Grizzly code is being developed under the US Department of Energy’s Light Water Sustainability program as a tool to model aging mechanisms and their effects on the integrity of critical nuclear power plant components. An important application for Grizzly is the modeling of aging in concrete structures, which can be due to a number of mechanisms. Initial focus in this area has been on modeling expansive reactions due to alkali-silica reactions or radiation-induced volumetric expansion. Grizzly is an inherently multiphysics modeling platform that naturally permits including the effects of multiple coupled physics in a simulation. Models have been developed for transport of heat and moisture in concrete, and these have been coupled and used as inputs to models for expansive reactions. This paper summarizes this capability, and demonstrates it on a representative structure.

Overview of EPRI Long Term Operations Work on Nuclear Power Plant Concrete Structures

The Electric Power Research Institute (EPRI) has been engaged in collaborative research and development activities related to concrete in nuclear applications over the past several years in concert with the nuclear generation industry, foreign and domestic national laboratories and regulatory bodies and universities. The EPRI Long Term Operations program is focused on performing research activities that will help the industry extend operation beyond the first period of license renewal, which for US plants means operation beyond 60 years. In this overview talk, three subjects will be addressed—radiation damage in boiling and pressurized water reactor concrete biological shields, boric acid attack of pressurized water reactor spent fuel pool concrete substructures and alkali-silica reaction degradation of concrete structures. The results of these and other studies are expected to support utilities as they demonstrate technical bases to regulatory bodies for long term operation of commercial nuclear plants.

The Effects of Neutron Irradiation on the Mechanical Properties of Mineral Analogues of Concrete Aggregates

Plans for extended operation of US nuclear power plants (NPPs) beyond 60 years have resulted in a renewed focus on the long-term aging of materials in NPPs, and specifically on reactor cavity concrete. To better understand the effects of neutron irradiationNeutron irradiation on reactor cavity concrete, a select group of mineral analoguesMineral analogues of concrete aggregatesAggregates were irradiated at the Oak Ridge National Laboratory High Flux Isotope Reactor at three different fluence levels and at two temperatures. The purpose was to investigate the degradation of mechanical properties at neutron doses above the levels expected in US NPPs under extended operation. Preliminary findings using nanoindentation clearly show that changes in the mechanical properties of these minerals can be observed and correlated to the neutron-induced damage. Scanning electron microscopy reveals changes in deformation and fracture mechanisms in the irradiated mineral analogies. Results for the nanohardnessNanohardness as a function of dose and temperature are presented and discussed for quartz, calcite, and dolomite.

Accident Tolerant Fuel Cladding

Accident Tolerant FeCrAl Fuel Cladding: Current Status Towards Commercialization

FeCrAl alloys are rapidly becoming mature candidate alloys for accident tolerant fuel applications. The FeCrAl material class has shown excellent oxidation resistance in high-temperature steam environments, a key aspect of any accident tolerant cladding concept, while also being corrosion resistant, stress corrosion cracking (SCC) resistant, irradiation-induced swelling resistant, weldable, and formable. Current research efforts are focused on design, development and commercial scaling of advanced FeCrAl alloys including large-scale, thin-walled seamless tube production followed by a broad spectrum of degradation evaluations in both normal and off-normal conditions. Included in this discussion is the theoretical analysis of the alloying principles and rules, alloy composition design, and overview of the most recent empirical database on possible degradation phenomena for FeCrAl alloys. The results are derived from extensive in-pile and out-of-pile experiments and form the basis for near-term deployment of a lead-test rod and/or assembly within a commercially operating nuclear power plant.

Interdiffusion Behavior of FeCrAl with U3Si2

Advanced steels, including FeCrAl are being considered as an alternative to the standard light water fuel (LWR) cladding, Zircalloy. FeCrAl has superior mechanical and thermal properties and oxidation resistance relative to the Zircalloy standard . Uranium Silicide (U3Si2) is a candidate to replace uranium oxide (UO2) as LWR fuel because of its higher thermal conductivity and higher fissile density relative to the current standard, UO2. The interdiffusion behavior between FeCrAl and U3Si2 is investigated in this study. Commercially available FeCrAl, along with pellets fabricated at the Idaho National Laboratory were placed in diffusion couples. Individual tests have been run at temperatures ranging from 500 ℃ to 1000 ℃ for 30 h and 100 h. The interdiffusion is analyzed with an optical microscope and scanning electron microscope (SEM). Uniform and planar diffusion regions along the material interface are illustrated with backscatter electron micrographs and energy-dispersive X-ray spectroscopy (EDS).

Mechanical Behavior of FeCrAl and Other Alloys Following Exposure to LOCA Conditions Plus Quenching

The US Department of Energy is working with commercial fuel vendors to develop advanced technology or accident tolerant fuels (ATF) for the current fleet of light water power reactors. General Electric and Oak Ridge National Laboratory are evaluating the concept of using iron-chrome-aluminum (FeCrAl) alloys as cladding for the current fuel of uranium dioxide pellets. In the case of a loss of coolant accident, the reactor may need to be flooded with fresh water when the cladding could be in the temperature range above 1000 ℃. It is important to determine the integrity of the cladding material after being quenched in water. Tests were performed for six alloys of interest which were exposed for 2 h at 1200 ℃ in air, argon or steam and then quenched in ambient temperature water. The resulting mechanical properties were evaluated and compared with the mechanical properties of the as received material. The FeCrAl alloy retains its yield strength after the high temperature excursions, with minimal oxidation but with some loss of ductility.

Mechanical Behavior and Structure of Advanced Fe-Cr-Al Alloy Weldments

FeCrAl alloys are promising for developing accident tolerant nuclear fuel claddings. These alloys showed good environmental compatibility and oxidation resistance in elevated-temperature water and steam, as well as low radiation-induced swelling. However, FeCrAl alloys may suffer from several degradation mechanisms, one of which may be a susceptibility to cracking during welding. Here, a set of advanced modified FeCrAl alloys were designed and produced by varying Al-content and employing additions of Nb and TiC. Strength, ductility, and deformation hardening behavior of the advanced FeCrAl alloys and their weldments are discussed.

Investigating Potential Accident Tolerant Fuel Cladding Materials and Coatings

Thermal energy release and hydrogen generation due to breakaway oxidation of Zr fuel cladding materials are of concern in accident scenarios involving extreme temperature increase (up to 1200 °C). As a result, potential accident tolerant fuel cladding (ATFC) materials and coatings are being investigated. Physical vapor deposited CrN coatings are considered as possible protective barrier materials for Zircaloys. In addition, Fe–Cr-Al alloys are considered potential candidate materials for ATFC due the formation of protective alumina at high temperatures which maintains resistance by preventing oxide breakdown. Both CrN-coated Zircaloys and a Fe–Cr-Al model alloy were exposed to 300 °C water and steam environments up to 1200 °C to evaluate their resistance to corrosion under normal reactor operating conditions and to high temperature steam oxidation. Surface analytical techniques are used to evaluate the effectiveness of oxides and/or coatings over the 300 °C water to 1000 °C steam temperature regime.

Steam Oxidation Behavior of FeCrAl Cladding

In order to better understand the high temperature steam oxidation behavior ofFeCrAl FeCrAl alloys, this study addressed two topics. The first is continuing to evaluate the effect of alloy composition on performance of commercial and laboratory-made candidate FeCrAl alloys. For a few optimized compositions, this includes the performance of commercially-made tubing where it is clear that dropping the Cr content from 20% to 10–13% reduces the maximum operating temperature in steamSteam by ~50 °C. The second addresses more realistic accident conditions. Model FeCrAl compositions that were exposed in ~300 °C water for 1 year were subsequently “ramp” tested in steam at 5 °C/min to 1500 °C to assess the effect of the Fe-rich oxide formed in water on the subsequent steam oxidation resistance. For Fe-18Cr-3Al+Y, the 1 year exposures in three different LWR water chemistries did not affect the ability to form alumina to 1500 °C. However, for marginal alloys Fe-13Cr-4Al and Fe-10Cr-5Al, some specimens began forming voluminous Fe-rich oxide at lower temperatures.

In-Situ Proton Irradiation-Corrosion Study of ATF Candidate Alloys in Simulated PWR Primary Water

Irradiation enhanced corrosionCorrosion behavior of Accident Tolerant Fuel candidate alloys T91 and Fe15Cr4Al were evaluated using in-situ proton irradiation-corrosion experiments in hydrogenated pure water (at 320 °C, 3 wppm H2) with a 5.4 MeV protonProton beam. The thin sample acted as a “window” to allow protons to fully penetrate the sample while maintaining system pressure. The area of the samples exposed to the proton beam experienced effects from displacement damage and radiolysis products. The aim of the study was to characterize the effect of radiation on the kinetics and character of oxidation caused by accelerated waterside corrosion under irradiationIrradiation. Samples irradiated with protons for total displacement damage of ~0.1 dpa (dose rate in water, 400 kGy/s) at an exposure time of 24 h were compared. Oxide morphology, phase structure, and composition of the oxide, metal and the metal/oxide interface were investigated using TEM and EDS and are related to the test conditions. The oxidation rate or resulting oxide thickness is dependent on the alloy Cr content; the oxidation rate increased as the Cr content decreased. The resulting oxide consists of an inner layer of Cr-rich spinel oxide and outer magnetite crystals in the unirradiated region; while the irradiated region consists of Cr-rich inner oxide spinel that was partially dissolved and coverage of outer non-faceted hematite precipitates.

Hydrothermal Corrosion of SiC Materials for Accident Tolerant Fuel Cladding with and Without Mitigation Coatings

As a candidate material for accident-tolerant fuelAccident-tolerant fuel cladding for light water reactors (LWR), SiCf–SiC composite materials possess many attractive properties. However, prior work has shown that SiC is susceptible to aqueous dissolution in LWR coolant environments. To address this issue, candidate coatings have been developed to inhibit dissolution. For this study, CVD SiC samples were prepared with Cr, CrN, TiN, ZrN, NiCr, and Ni coatings. Uncoated SiC and SiCf–SiC samples were also prepared. The samples were exposed for 400 h in 288 ℃ water with 2 wppm DO in a constantly-refreshing autoclave to simulate BWR–NWC. Cr and Ni coated samples lost less mass than the uncoated SiC sample, indicating an improvement in performance. The CrN coating resisted oxidation, but some of the coating was lost due to poor adhesion. The TiN coated sample gained significant mass due to oxidation of the coating. ZrN and NiCr coatings showed significant corrosion attack. SiCf–SiC ceramic matrix composite materials dissolved much faster than the CVD SiC sample, demonstrating the need for mitigation coatings if CMCs are to be used in LWRs. This work demonstrates the promise of Cr, Ni and CrN coatings for corrosion mitigation in LWRs, and shows that NiCr and ZrN are not promising coating materials.

Characterization of the Hydrothermal Corrosion Behavior of Ceramics for Accident Tolerant Fuel Cladding

Accident-tolerant fuel (ATF)Accident-Tolerant Fuel (ATF) is an increasingly important research topic for the nuclear industry, and ceramics such as SiCSiC are strong contenders for deployment as ATF cladding. The hydrothermal corrosion Hydrothermal corrosioncharacteristics of SiC and Al2O3Al2O3 were investigated via constantly-refreshing autoclave corrosion and post exposure characterization. Four different types of chemical vapor deposited (CVD) SiC specimens were examined (two with high electrical resistance, one with low electrical resistance, and a single crystal 4H structural variant). Al2O3 specimens were prepared in single crystal and polycrystalline states. PWR primary water, BWR–HWC, and BWR–NWC environments were maintained throughout the experiments. Characterization conducted using SEM and EDS was used to determine factors affecting corrosion rates and susceptibility to grain boundary attack in each water chemistry condition. Raman spectroscopy was also used to determine chemical variation of the surface with corrosion. Grain boundary attack was found to be significant for both alumina and SiC polycrystalline variants.

Corrosion of Multilayer Ceramic-Coated ZIRLO Exposed to High Temperature Water

The corrosion behavior of ceramic coated ZIRLO tubing was evaluated in supercritical water to determine its behavior in high temperature water. The coating architecture consisted of a 4 bilayer TiAlN/TiN coating with Ti bond coat on Zirlo tubes using cathodic arc physical vapor deposition (CA-PVD) technique. On exposure to deaerated supercritical water at 542 °C for 48 h coated tubes exhibited significantly higher weight gain compared to uncoated Zirlo. Examination revealed formation of a uniform ZrO2 layer beneath the coating and of a thickness similar to that on the uncoated tube inner surface. The defects generated during the coating process acted as preferential paths for diffusion of oxygen resulting in the oxidation of substrate Zirlo. However, there was no delamination of the coating.

General SCC and SCC Modeling

Calibration of the Local IGSCC Engineering Model for Alloy 600

Many Stress Corrosion Cracking (SCC) models have been developed so far. Quantitative empirical models, trying to predict initiation and crack growth rate of nickel alloys exposed to pressurized water reactor primary water do not describe physical mechanism and suffer a lack of accuracy. By contrast, models describing the possible involved physical mechanisms responsible for degradation (selective oxidation of grain boundaries in the case of Alloy 600Alloy 600 exposed to PWR primary water) are usually qualitative. In the current paper, a ‘local’ model is proposed to better predict SCC. In order to succeed, behaviors assumed to be involved in the SCC process were calibrated and coupled: intergranular oxidation rate, intergranular stresses, resistance to cracking of oxidized grain boundaries. The output of the model is the time to reach a given crack depth. This paper introduces the first calibration of parameters for Alloy 600 exposed to primary water.

Prediction of IGSCC as a Finite Element Modeling Post-analysis

A numerical approach was developed to predict Stress Corrosion Cracking (SCC) kinetics and location, in 3D. The calculation is a fast post finite element modeling analysis chaining IGSCC initiation and crack growth models. Firstly, the proposed methodology offers the possibility to optimize the calibration of models, when SCC tests are simulated. Secondly, calibrated models can be used to predict SCC in large structures. In the proposed paper, benefits and limitations of the methodology will be introduced.

Monte Carlo Simulation Based on SCC Test Results in Hydrogenated Steam Environment for Alloy 600

We investigated the applicability of a stress corrosion cracking (SCC) engineering model and simulation method developed on the basis of the SCC of sensitized 304 stainless steel in a simulated BWR environment to the primary water stress corrosion cracking (PWSCC). We conducted a uniaxial constant loading test on Alloy 600 in a 400 °C hydrogenated steam environment and found that the number of cracks observed on a specimen surface after every passage of 450 h could be approximated to Poisson distribution, indicating that a Poisson random process model is applicable to the SCC in this system. By applying the engineering model, we statistically processed experimental data by assuming that the time distribution of occurrence of microcracks follows exponential distribution, and then obtained input data for the SCC simulation. Using coalescence coefficient, k, as a fitting parameter to obtain a reasonable k-value, it was found that the best agreement between the experimental and simulation results for the number of microcracks and the maximum crack length at k = 0.15. This is about one third the k-value of 0.5 found in sensitized 304 stainless steel in the BWR environment, indicating that coalescence is more subdued in PWSCC than in SCC in the BWR environment.

Protection of the Steel Used for Dry Cask Storage System from Atmospheric Corrosion by Tio2 Coating

Austenitic 304 stainless steels (SS) and carbon steels (CS) are widely used as structural materials for components and pipe assemblies in nuclear power plants. Steels also act as the important canister materials in the dry storage of spent nuclear fuels. However, it is well known that stainless steels and carbon steels are susceptible to stress corrosion cracking (SCC) in certain environments induced by sea salt particles and chlorides. It is known that the TiO2 coating acts as a non-sacrificial anode and protects the steel cathodically under UV illumination. In this study, the photoelectrochemical behavior of the steel with TiO2 coating by sol-gel method was investigated to mitigate atmospheric SCC. In addition, the slow decline of photovoltage of the samples coated with TiO2 after UV illumination have been studied. These results indicate that the TiO2 treatment with or after UV illumination would effectively reduce the steel corrosion rate in atmospheric environments.

Predictive Modeling of Baffle-Former Bolt Failures in Pressurized Water Reactors

Baffle-former boltBaffle-former bolt failures have been observed in recent inspections of pressurized water reactors (PWRs). These failures are understood to be primarily caused by irradiation-assisted stress corrosion cracking (IASCC). A prognostic method has been developed for simulating the degradation of baffle-former bolts due to IASCC. The method characterizes the evolution of stress in a reactor environment, as well as the redistribution of stress amidst neighboring bolt failures. Empirically-validated Weibull parameters are utilized in a stochasticStochastic framework, and the model is exercised as a Monte Carlo simulation to evaluate a range of plausible scenarios from which trends of bolt failure rates and patterns can be determined. This semi-empiricalSemi-empirical methodology is informed by both operating experience and a more detailed, predictive finite element analysis model that encompasses the full range of phenomenological effects common to the operating environment within a PWR.

Technical Basis and SCC Growth Rate Data to Develop an SCC Disposition Curve for Alloy 82 in BWR Environments

Alloy 82 weld metal shows higher Stress Corrosion Cracking (SCC) resistance in BWR environments than Alloys 182 and 132. To define the relative factor of improvement in SCC resistance and its impact on plant management properly, it is appropriate to establish a SCC growth rate disposition curve for Alloy 82 in BWR environments. In this study, several factors that influence SCC growth behavior of Alloy 82 are evaluated based on the latest Crack Growth Rate (CGR) data collected in Japan BWR Owners Group projects. The goal is to provide a technical basis on which the validity of the data will be evaluated prior to proposing a new disposition curve. The factors evaluated include effects of type of Alloy 82 weld, specimen size, post weld heat treatment (PWHT) and sulfate addition on SCC growth rate.

BWR SCC and Water Chemistry

SCC and Fracture Toughness of XM-19

The effect of stress intensity factor, cold workCold work, corrosion potential and water purity on the stress corrosion crack (SCC) growth rate behavior of as-received and as-received plus 19.3% cold worked XM-19 was investigated in 288 °C BWR water. For 19.3% cold rolled XM-19, high to very high crack growth ratesCrack growth rate were consistently observed at high corrosion potential, largely independent of heat or orientation. As received XM-19 exhibited SCC growth rates ~5–10X slower than cold worked XM-19, but these rates are still considered high. For all materials and conditions, low corrosion potential conditions reduced the growth rates by about an order of magnitude, and somewhat more if impurities were present in the water. The SCC growth rates for both conditions of XM-19 were somewhat higher than the equivalent conditions of 18-8 stainless steels, such as Types 304/304L and 316/316L. Higher growth rates tend to be observed at higher yield strength, and XM-19XM-19 stainless steel has an elevated yield strength from nitrogen-strengthening; incomplete annealing in the as-received material can also increase the yield strength. The J-R fracture resistanceJ-R fracture resistance of 19.3% cold rolled XM-19 and as-received XM-19 in multiple orientations and with replicates was evaluated in 288 °C air. The data show a significant effect of crack orientation in the plate (the rolling plane coincides with the plane of the plate), consistent with the inhomogeneous nature of the microstructure. The fracture resistance of as-received XM-19 was good, but the 19.3% cold rolled XM-19 specimens exhibited low toughness, to the extent that many tests were invalid. Fracture resistance in 80–288 °C water environments was not evaluated, but is relevant to LWR components. Irradiation of this heat of XM-19 is in progress at the Idaho National Laboratory Advanced Test Reactor.

On the Effect of Preoxidation of Nickel Alloy X-750

Nickel AlloyNickel based alloysX-750X-750 is a Ni-Cr–Fe alloy with good corrosionCorrosion properties and high strength at elevated temperature. It is commonly used for spacer grids in Boiling Water Reactors (BWR)BWR. In this environment, the material can suffer from significant corrosion, leading to weight loss by metal dissolution. To further improve the characteristics of this material, a process called preoxidation is often performed. This results in the formation of strengthening γ’-Ni3(Ti, Al) precipitates and a thin oxideOxide formation on the surface. In this paper, preoxidized and non-preoxidized specimens are compared with respect to their oxidation properties. We report about microstructural studies made on specimens exposed in simulated BWR environment for 24 h and 840 h. Electron microscopy techniques have been used to investigate the oxide microstructures. A comparison between these specimens shows the complexity of the corrosion process and the impact of preoxidation. Preoxidized specimens show thinner and more homogenous oxides than non-preoxidized ones. They lose less mass and build thinner oxides. The preoxidation layer consists of a bilayer oxide of NiFe2O4NiFe2O4 and Cr2O3 that is preserved during the long exposure. NiFe2O4 spinel crystals are present on the surface of all exposed specimens, a result of re-precipitation of dissolved metal ions.

Microstructures of Oxide Films Formed in Alloy 182 BWR Core Shroud Support Leg Cracks

This paper contributes to a TEM examination on the oxide films formed at three locations along a crack path in Alloy 182 weld from a BWR core shroud support leg, namely, the crack mouth, the midway between the mouth and the crack tip, and the crack tip. In the crack mouth the oxide film was approximately 1.6 μm in thickness and consisted of relatively pure NiO. The midway oxide film was mainly a nickel chromium oxide with a film thickness of 0.3 μm. At the crack tip the oxide film was a nickel chromium iron oxide with a film thickness of 30 nm. In all studied locations the main oxides had the similar rocksalt structure and the cracks were much wider than the thicknesses of the oxide films. It probably suggests that the corroded metal was largely dissolved into the coolant. The different dissolution rates of nickel, chromium and iron cations in the oxide films are clearly displayed with the compositions of the residual oxides. The oxide stability under different redox potentials along the crack path is briefly discussed.

Effect of Chloride Transients on Crack Growth Rates in Low Alloy Steels in BWR Environments

The objective of this study was to quantify the effect of chloride transients on stress corrosion cracking of pressure vessel low alloy steels. Two heats of reactor pressure vessel steel were evaluated at various chloride concentrations in both NWC and HWC environments. The tests showed that low alloy steels can exhibit a delayed cracking response during a chloride transient. The delayed response is attributed to the concentrating and dilution processes of anionic impurities inside the crack. The crack can maintain its original growth rate for a certain period after the chemistry change. The effects of chloride concentration, stress intensity factor (K) and periodic load cycling on the crack incubation and growth in low alloy steel will be discussed. The results from this work will provide a direct input to development of a crack growth model for low alloy steels, water chemistry guidelines and effects of chloride transients on crack growth.

Electrochemical Behavior of Platinum Treated Type 304 Stainless Steels in Simulated BWR Environments Under Startup Conditions

As reactor startup begins, the ECP is initially high in the oxygenated water environment established during a cold shutdown. Consequently, the components would exhibit the higher crack initiation and propagation rates of IGSCC during startup period than in the remainder of the cycle. The corrosion current density response of stainless steel exposed to H2O2 was larger than that of those exposed to O2, and it remained at a higher value even at the low level of several ppb. As noble metal was applied in the BWRs to catalyze the chemical reactions of H2O2 and O2, this study evaluated the corrosion behaviors of both oxidants on the components of stainless steel. The corrosion potentials and corrosion current densities of 304SS with Pt coating were investigated in pure water with dissolved oxygen or hydrogen peroxide concentrations at various temperatures.

Investigations of the Dual Benefits of Zinc Injection on Cobalt-60 Uptake and Oxide Film Formation Under Boiling Water Reactor Conditions

Zinc injection in reactor feed water is a well-known mitigation strategy for prevention of radioactive 60Co deposition in both Boiling and Pressurised water reactors. Furthermore, zinc leads to the formation of a thinner, more stable oxides arising from the thermodynamically driven replacement of Ni and Fe in the characteristic spinel type oxide formed on stainless steel. However, the interaction of zinc with the oxide formation under different water chemistries is not fully understood. Oxidation tests on type 316 stainless steel were performed under two hydrogen water chemistry conditions (HWC), with and without zinc injection and the resultant oxides analysed using analytical electron microscopy (AEM), field emission gun scanning electron microscopy (FEG-SEM), and energy dispersive X-ray spectroscopy (EDXS). The present work identifies and quantifies the positive microstructural changes that Zn has on the oxide formation on a #600 grit surface and an OPS polished 316 SS surface under boiling water reactor (BWR) conditions.

SCC Mitigation in Boiling Water Reactors: Platinum Deposition and Durability on Structural Materials

Noble metalNoble metal injection is widely used to mitigate stress corrosion cracking (SCC) of reactor components. Despite its wide use, there are still open questions regarding the parameters affecting the application process and possible improvements to it. Laboratory experiments in a high-temperature water loop at PSI were complemented by exposure of specimens in the mitigation monitoring system (MMS) at KKL. The influence of parameters such as flow conditions, structural material composition, surface roughness and geometry on the deposition behavior of the platinum (Pt)PlatinumnanoparticlesNanoparticles was investigated. Furthermore, the long-term stability of the coverage of surfaces by Pt particles was analyzed. The composition of the underlying alloy was found to have an effect on the deposition behavior, whereas surface roughness has no measurable impact. Pt showed a limited durability on steel surfaces and, after the end of the application, the remobilized Pt seems to re-deposit only minimally on nearby surfaces.

Confirmation of On-Line NobleChem™ (OLNC) Mitigation Effectiveness in Operating Boiling Water Reactors (BWRs)

The development and implementation of On-Line NobleChemTM (OLNC) has now occurred for many years with almost the entire US BWR fleet, including Mexico, and some European BWRs using this process to achieve mitigation of IGSCC. The data confirming the effectiveness of OLNC has indirectly shown significant reductions in incidences of crack initiation as well as the suppression of growth of any previously existing cracks. However, the robustness of the process and its ability to protect all core and lower plenum regions requires on-going confirmation of the presence of nanometer sized platinum particles as well as verifying the adequacy of the coverage on surfaces of the core structural materials. Efforts by GE Hitachi Nuclear Energy (GEH) have continued in cooperation with the BWR utilities to both properly inject the platinum solution into the coolant as well as to monitor the surface characteristics of actual core components. This paper will show the results of on-going efforts to confirm platinum coverage of stainless steel surfaces using Field Emission Scanning Electron Microscopy (FE-SEM) methods. Additional efforts to tie all the process methods together to establish and confirm OLNC performance will also be presented. Finally, the paper will review continuing plans moving forward to validate OLNC performance across the BWR fleet.

Development of the Fundamental Multiphysics Analysis Model for Crevice Corrosion Using a Finite Element Method

It is necessary to study crevice structures which can mitigate crevice corrosionCrevice corrosion as the origin of SCC of materials used in BWRBWR core region. A fundamental crevice corrosion simulation model has been developed to design corrosion control structures for these materials. Effects of the width and the depth on the corrosive environment in a crevice were studied based on that model. Calculated pH in the crevice decreased with time for all crevice geometries. The lowest pH was found at the deepest position in the crevice for all the cases. It seemed there was only a negligible difference in pH where the crevice depth was deeper than the specific depth which depended on the crevice width.

In Situ Electrochemical Study on Crevice Environment of Stainless Steel in High Temperature Water

In situ electrochemical impedance spectroscopy measurement within crevice of stainless steel in 288 °C water has been conducted to analyze crevice water chemistry. Small sensors ($$ {\varphi} {\sim} 250\,\upmu{\text{m}}) $$ measured local solution electrical conductivity κcrev, polarization resistance and electrochemical corrosion potential. Real-time response of the κcrev as functions of bulk water conductivity and dissolved oxygen (DO) concentration has been quantitatively analyzed. The κcrev differ more than an order of magnitude depending on the oxygen potential inside the crevice. The κcrev increased with addition of small amount of bulk DO (e.g. 30 ppb). The maximum κcrev was observed with DO of 32,000 ppb and became more than 100 times higher than that of bulk water. The effect of geometrical factors on the crevice environment was also found to play an important role in the water chemistry inside.

Zirconium and Fuel Cladding

Corrosion Fatigue Crack Initiation in Zr-2.5Nb

In-service inspections of Zr-2.5NbZr-2.5Nb pressure tubes may reveal blunt flaws such as fretting wear or crevice corrosion marks. These flaws pose no immediate threat to the integrity of the pressure tube but may be potential fatigue crack initiationCrack initiation sites. An understanding of the effect of the coolant environment, specifically on fatigueFatigue crack initiation, is important in this context. Tests were conducted on notched transverse tensile specimens at 275 and 300 °C with a load rise time between 50 and 3600 s. Current tests investigated the effects of applied loading frequency and hydrogenHydrogen on fatigue crack initiation. Results have indicated that long rise time and a water environment reduce the time to fatigue crack initiation in non-hydrided and pre-hydrided specimens as compared to tests conducted in air. If enough hydrogen is able to diffuse to the notch during the test, it may also be possible to reach conditions where there is an interaction between corrosionCorrosion, fatigue and hydride cracking.

Cluster Dynamics Model for the Hydride Precipitation Kinetics in Zirconium Cladding

Hydride precipitation in zirconium claddingZirconium cladding is known to cause severe loss of toughness and greatly increase the risk of mechanical failure and fuel leakage. Modeling hydride formation kinetics is critical to the safety assessment of the fuel-cladding system and the entire reactor system. Existing reduced order models do not provide such details as number density and size distribution of hydride precipitates. We have recently developed a cross-scale cluster dynamics modelCluster dynamics modeling with increased physical details and enhanced predictive capability for the hydride formation kinetics in zirconium. Our model takes information from atomistic simulations, such as migration energy of interstitial hydrogenHydrogen and formation/binding energy of hydride embryos/clusters, as input, and establishes and solves a system of rate equations that describe the evolution of concentrations of freely migrating hydrogen as well as sessile hydride clusters of all different sizes. Used here to simulate an in situ hydride growth experiment on a TEM, our model is able to reproduce the linear growth behavior of pre-existing hydrides under hydrogen ion implantation and provide possible explanations for the estimated growth rate.

Modeling Corrosion Kinetics of Zirconium Alloys in Loss-of-Coolant Accident (LOCA)

Correctly predicting the mechanical behavior of zirconium fuel cladding during a LOCA transient is critical for nuclear safety analysis as the fuel rod needs to maintain its coolable geometry throughout the LOCALOCA sequence. A physically-based zirconium alloy corrosion model called the Coupled Current Charge Compensation (C4) is developed. The model calculates the coupling of oxygen, electron and hydrogen currents and predicts the oxide, oxygen-stabilized $$ \alpha $$-Zr and prior-$$ \beta $$-Zr layers kinetics as well as the oxygen concentration profiles during a LOCA scenario. The results obtained during isothermal conditions are compared to experimental data for validation. Future developments of the C4 model include an implementation into the nuclear performance code BISON, which currently does not provide a physical description of the oxygen and hydrogen concentration profiles in the cladding. Thanks to the C4 implementation into BISON, structural integrity of the fuel cladding following a LOCA event can be assessed.

Progressing Zirconium-Alloy Corrosion Models Using Synchrotron XANES

The corrosion and hydrogen pickup of in-reactor zirconium-based cladding is currently limiting the maximum fuel burnup in light-water reactors. Since the oxidation rate and hydrogen pickup fraction of zirconium alloys vary significantly as a function of exposure time, burnup, and alloy composition, it is critical to better understand the underlying mechanisms to model and predict corrosion behavior. Following the analysis of ~500 autoclave coupons, a physically based zirconium-alloy corrosion model founded on first principles, named “Coupled Current Charge Compensation (C4)”, has been developed. The model reproduces the differences in oxidation kinetics and hydrogen pickup between different zirconium alloys, such as Zr-Nb and Zircaloy-4. Since oxidized solute elements affect the corrosion process through a space-charge compensation mechanism, synchrotron nano-beam X-ray Absorption Near-Edge Spectroscopy has been performed on multiple oxidized Zr-Nb alloys to determine the oxidation-state profile of niobium in the oxide layer. The results inform the C4 model and the associated hydrogen pickup fraction.

Advanced Characterization of Hydrides in Zirconium Alloys

The mechanical properties of zirconiumZirconium alloys are affected by the presence of hydrides. The strain fields around hydridesHydrides, which are affected by the size, orientation, and hydride phase, are believed to influence the apparent hysteresis between solubility limits on heating and cooling. TEM characterization of dislocation fields near hydrides in Zircaloy-4 specimens, which were exposed to 300 °C primary-water conditions for 600 h, was performed both before and after a heating and cooling cycle. In addition, EELS characterization is provided before heating. In situ TEM imaging/recording and nano-diffraction allowed monitoring of the morphology of dissolving hydrides throughout the temperature cycling. No dislocations in the matrix surrounding the hydrides were visible prior to heating; however, when the hydrides dissolved, dislocations were visible in the space the hydrides had previously occupied, providing a map of the original hydride distribution. These dislocation ‘nests’ are likely the preferential sites for subsequent hydride precipitation and elucidate the so-called ‘memory effect’. Advancing the understanding of hydride formation kinetics, hydride morphology, and hydrogen solid solubility limits can help to reduce uncertainties and conservatism when addressing the risks of hydrogen embrittlementHydrogen embrittlement and hydride cracking in zirconium components.

Influence of Alloying Elements and Effect of Stress on Anisotropic Hydrogen Diffusion in Zr-Based Alloys Predicted by Accelerated Kinetic Monte Carlo Simulations

The presence of hydrogen (H) can detrimentally affect the mechanical properties of many metals and alloys. To mitigate these detrimental effects requires fundamental understanding of the thermodynamics and kinetics governing H pickup and hydride formation. In this work, we focus on H diffusion in Zr-based alloys by studying the effects of alloying elements and stress, factors that have been shown to strongly affect H pickup and hydride formation in nuclear fuel claddings. A recently developed accelerated kinetic Monte Carlo method is used for the study. It is found that for the alloys considered here, H diffusivity depends weakly on composition, with negligible effect at high temperatures in the range of 600–1200 K. Therefore, the small variation in H diffusivity caused by variations in compositions of these alloys is likely not a major cause of the very different H pickup rates. In contrast, stress strongly affects H diffusivity. This effect needs to be considered for studying hydride formation and delayed hydride cracking.

Stainless Steel Aging and CASS

Influence of δ-Ferrite Content on Thermal Aging Induced Mechanical Property Degradation in Cast Stainless Steels

Thermal degradation of cast stainless steels was studied to provide an extensive knowledgebase for the assessment of structural integrity during extended operations of reactor coolant systems. The CF3 and CF8 series cast stainless steels with relatively low (5–12%) δ-ferrite contents were thermally aged at 290–400 °C for up to 10,000 h and tested to measure changes in tensile and impact properties. The aging treatments caused significant reduction of tensile ductility, but only slight softening or negligible strength change. The thermal agingThermal aging also caused significant reduction of upper shelf energy and large shift of ductile-brittle transition temperature (ΔDBTT). The most influential factor in thermal degradation was ferrite content because of the major degradation mechanism occurring in the phase, while the nitrogen and carbon contents caused only weak effects. An integrated model is being developed to correlate the mechanical propertyMechanical properties changes with microstructural and compositional parameters.

Microstructure and Deformation Behavior of Thermally Aged Cast Austenitic Stainless Steels

Cast austenitic stainless steels (CASS)Cast Austenitic Stainless Steels (CASS) consist of a dual-phase microstructure of delta ferrite and austenite. The ferrite phase is critical for the service performance of CASS alloys, but can also undergo significant microstructural changes at elevated temperatures, leading to severe embrittlement. To understand thermal aging embrittlementThermal aging embrittlement, fracture toughnessFracture toughness J-R curve tests were performed on unaged and aged CF8 specimens at 315 ℃. The microstructure of CF8 was also examined before and after thermal aging with transmission electron microscopy and atom probe tomography. While no microstructural change was observed in the austenite after thermal aging, a high density of G-phase precipitates and a phase separation of alpha/alpha prime were detected in ferrite. To study the deformation behavior, tensile tests were performed at room temperature with in situ wide-angle X-ray scatteringWide-Angle X-ray Scattering (WAXS) measurements. The differences in lattice strains between ferrite and austenite were much higher in the aged than in the unaged samples, suggesting a higher degree of incompatible deformation between ferrite and austenite in the aged samples.

Microstructural Evolution of Cast Austenitic Stainless Steels Under Accelerated Thermal Aging

Thermal aging degradationThermal aging degradation of cast austenitic stainless steels (CASS) was studied by electron microscopy to understand the mechanisms for thermal embrittlement potentially experienced during extended operations of light water reactor coolant systems. Four CASS alloys—CF3, CF3M, CF8, and CF8M—were thermally aged up 1500 h at 330 and 400 °C, and the microstructural evolution of the material was characterized by analytical aberration-corrected scanning transmission electron microscopy. The primary microstructural and compositional changes during thermal aging were spinodal decompositionSpinodal decomposition of the δ-ferrite into α/α′, precipitation of G-phaseG-phase precipitation in the δ-ferrite, segregation of soluteSolute segregation to the austenite/ferrite interphase boundary, and growth of M23C6 carbides on the austenite/ferrite interphase boundary. These changes were shown to be highly dependent on aging temperature and chemical composition, particularly the amount of C and Mo. A comprehensive model is being developed to correlate the microstructural evolution with mechanical behavior and simulation.

Electrochemical Characteristics of Delta Ferrite in Thermally Aged Austenitic Stainless Steel Weld

An austenitic stainless steel Type 316L weld was thermally aged for 20,000 h at 400 °C and electrochemical characterization was performed to measure corrosion resistance in δ–ferrite phase. It is well known that a severe thermal aging causes decrease of fracture resistance and increase of the hardness of δ–ferrite, which was related to the spinodal decomposition. After thermal aging, the DL-EPR response of 316L weld was dominated by parent austenite matrix without reactivation peak. To characterize the δ–ferrite only, austenite phase was selectively dissolved from the matrix by electrochemical etching method. The double–loop electrochemical potentiokinetic reactivation (DL-EPR) analysis of the δ–ferrite phase showed degradation in corrosion resistance after thermal aging with the appearance of a cathodic loop and reactivation peak during the reverse scan. The degradation in corrosion resistance of δ–ferrite phase could be attributed to the localized Cr-depletion due to spinodal decomposition and precipitation of intermetallic phases during thermal aging.

Effect of Long-Term Thermal Aging on SCC Initiation Susceptibility in Low Carbon Austenitic Stainless Steels

The objective of this study was to clarify the effect of long-term thermal aging on SCC initiation susceptibility in low carbon austenitic stainless steels. Specimens used were Type 304L and 316L austenitic stainless steels. Both steels were cold worked to 20% thickness reduction (CW) followed by long-term thermal aging at 288 °C for 14,000 h (LTA). Creviced Bent Beam (CBB) testing was carried out to estimate the SCC initiation susceptibility under BWR simulated water condition at high temperature. The results of the CBB tests showed that Type 304L specimens with CW and LTA treatment exhibited no SCC susceptibility. In contrast, the SCC initiation susceptibility of Type 316L increased by the combination of cold work and long-term thermal aging. To understand these results, evaluations on the changes of microchemistry, microstructure and mechanical properties induced by the CW and LTA treatment have been performed, and their correlation with the SCC initiation susceptibility was discussed.

Crack Growth Rate and Fracture Toughness of CF3 Cast Stainless Steels at ~3 DPA

Cast austenitic stainless steels (CASS)Cast Austenitic Stainless Steels (CASS) used in reactor core internals are subject to high-temperature coolant and energetic neutron irradiationNeutron irradiation during power operations. Due to both thermal agingThermal aging and irradiation embrittlementIrradiation embrittlement, the long-term performance of CASS materials is of concern. To assess the cracking behavior of irradiated CASS alloys, crack growth rate (CGR) and fracture toughness J-R curve tests were performed on two CF3 alloys. Miniature compact tension specimens were irradiated to ~3 dpa, and were tested at ~315 °C in simulated LWR coolant environments with low corrosion potentials. No elevated cracking susceptibility was observed at this dose in the test environments. The power exponents of the 3 dpa J-R curves were much lower than that of unirradiated or irradiated specimens at lower doses, indicating a significant decline in fracture resistance. A preliminary microstructural study revealed irradiation-induced microstructural changes in both austenite and ferrite, suggesting an embrittlement mechanism involving both phases at this dose level.

Effects of Thermal Aging and Low Dose Neutron Irradiation on the Ferrite Phase in a 308L Weld

The integrity of reactor internal components made of austenitic stainless steel weldsAustenitic stainless steel weld with a duplex structure can potentially be affected by thermal agingThermal aging and/or neutron irradiationNeutron irradiation induced embrittlement. There have not been sufficient studies on the long-term service performance of SS welds in light water reactors. In this study, thermal aging was performed at 400 °C for up to 2220 h on a 308L weld, and the irradiation was conducted in the Halden reactor at ~315 °C to 0.08 dpa (5.6 × 1019 n/cm2, E > 1 meV). The microstructural evolution of the ferrite phase was characterized using atom probe tomography (APT) and auxiliary transmission electron microscope studies. Spinodal decomposition and Ni-Mn-Si solute clusters were observed in both the thermally aged and neutron irradiated 308L welds. As compared with thermal aging, low dose neutron irradiation induced similar spinodal decomposition with slightly larger concentration wavelength and amplitude. The solute clusters in irradiated ferrite phase also show a larger mean size, a wider size distribution, but a lower number density as compared with those in thermally aged ferrite phase. In addition, the neutron irradiation significantly promotes segregation of trace elements, particularly phosphorus, at the Ni-Mn-Si solute clusters.

Microstructural Evolution of Welded Stainless Steels on Integrated Effect of Thermal Aging and Low Flux Irradiation

The combined effect of thermal agingThermal aging and irradiation on cast and welded stainless steelStainless steel solidification structures is not sufficiently investigated. From theory and consecutive aging and irradiation experiments, the effect of simultaneous low rate irradiation and thermal aging is expected to accelerate and modify the aging processes of the ferriteWeld ferrite phase. Here, a detailed analysis of long-term aged material at very low fast neutron flux at LWR operating temperatures using Atom Probe Tomography is presented. Samples of weld material from various positions in the core barrel of the Zorita PWR are examined. The welds have been exposed to 280–285 °C for 38 years at three different neutron fluxes between 1 × 10−5 and 7 × 10−7 dpa/h to a total dose of 0.15–2 dpa. The aging of the ferrite phase occurs by spinodal decompositionSpinodal decomposition, clustering and precipitation of e.g. G-phase. These phenomena are characterized and quantitatively analyzed in order to understand the effect of flux in combination with thermal aging.

Welds, Weld Metals, and Weld Assessments

The Use of Tapered Specimens to Evaluate the SCC Initiation Susceptibility in Alloy 182 in BWR and PWR Environments

A better understanding of stress corrosion cracking (SCC) initiation is one of the keys towards developing proactive mitigation techniques for the safe and economic operation of nuclear power plants. However, SCC initiation laboratory studies are very time consuming and require multiple specimens, Hence, in the framework of the European “MICRIN+” research project, an accelerated test method was evaluated for screening the SCC susceptibility in a relatively short time frame. The effects of surface roughness and strain rate on SCC initiation susceptibility in Alloy 182 weld metal were evaluated in simulated BWR and PWR environments. Constant extension rate tensile tests were performed using flat tapered tensile specimens with different surface finishes (ground and polished) in hydrogenated water at 288 and 340 °C. The surface crack distribution and crack length as well as stress thresholds for SCC initiation were analyzed by detailed post-test quantitative characterization. Some test data were analyzed by the EngInit SCC initiation model. The accelerated test technique was successfully applied and revealed very promising results. The highest crack density and lowest stress thresholds for crack initiation were found on the ground surfaces and at the lowest strain rates. A test’s load response can be fed to the EngInit model; with parameters to be determined by comparing EngInit’s damage to the experimental surface crack distributions. EngInit can potentially be used to link laboratory test results on flat tapered specimens to SCC initiation in components in the field.

Effect of Thermal Aging on Fracture Mechanical Properties and Crack Propagation Behavior of Alloy 52 Narrow-Gap Dissimilar Metal Weld

Determination of the fracture toughness properties and thermal aging behavior ofDissimilar Metal Weld (DMW) dissimilar metal weld (DMW) joints is of utmost importance for successful structural integrity and lifetime analyses. This paper presents results from fracture resistance (J-R), fracture toughness (T0) and Charpy-V impact toughness tests as well as fractography performed for an industrially manufactured narrow-gap DMW mock-up (SA508-Alloy 52-AISI 316L). Tests were performed on post-weld heat treated, 5000 h aged and 10,000 h aged material. The results show that this DMW is tough at the SA 508-Alloy 52 interface, which typically is the weakest zone of a DMW. The DMW joint maintains its high fracture resistance also after thermal agingAging. Crack propagates for a large part in the carbon-depleted zone (CDZ) of SA 508 but deflects occasionally to the Alloy 52 side due to small weld defects in µm scale. Ductile-to-brittle transition temperature determined from Charpy-V impact toughness tests increases due to thermal aging, but only to a minor extent. No significant change is observed for the T0 transition temperature due to aging.

Distribution and Characteristics of Oxide Films Formed on Stainless Steel Cladding on Low Alloy Steel in Simulated PWR Primary Water Environments

The properties of oxide filmOxide film formed on stainless steel (SS) cladding on low alloy steel (LAS)Low Alloy Steel (LAS) after immersion in simulated PWR primary water environments with different dissolved oxygen contents are investigated. The HAZ in the LAS consist of overheated crystal region, complete recrystallized region and incompletely recrystallized region, while SS cladding consist of austenite zone and austenite and ferrite mixing zone. Pitting appeared on 309L SS after immersion in high temperature waterHigh temperature water due to the dissolution of inclusions existed previously on 309L SS which contain higher ferrite content. Raman spectra and TEM results show that the outer layer is mainly Fe-rich spinel oxides while the inner layer is mainly Cr-rich oxides. Ni is mainly concentrate at the oxide/substrate interface due to the low oxygen affinity. The inner oxide layer on 308L SS is thinner than that on 309L SS, implying that ferrite distributed on austenite is not favorable for the growth of oxides. Reducing the oxygen content in PWR primary water favored the formation of spinel oxides.

Microstructural Characterization of Alloy 52 Narrow-Gap Dissimilar Metal Weld After Aging

The safe-end dissimilar metal weld (DMW)Dissimilar Metal Weld (DMW) joining the reactor pressure vessel to the main coolant piping is one of the most critical DMWs in a nuclear power plant (NPP). DMWs have varying microstructures at a short distance across the ferritic-austenitic fusion boundary (FB) region. This microstructural variation affects the mechanical properties and fracture behavior and may evolve as a result of thermal aging during long-term operation of an NPP. This paper presents microstructural characterizationMicrostructural characterization performed for as-manufactured and 5000 h and 10,000 h thermally aged narrow-gap DMW representing a safe-end DMW of a modern pressurized water reactor (PWR) NPP. The most significant result of the study is that the thermal agingAging leads to a significant decrease in a hardness gradient observed across the ferritic-austenitic FB of the as-manufactured DMW.

A Statistical Analysis on Modeling Uncertainty Through Crack Initiation Tests

Because a large time spread in most crack initiation tests makes it a daunting task to predict the initiation time of cracking, a probabilistic model, such as the Weibull distribution, has been usually employed to model it. In this case, although it might be anticipated to develop a more reliable cracking model under ideal cracking test conditions (e.g., large number of specimen, narrow censoring interval, etc.), it is not straightforward to quantitatively assess the effects of these experimental conditions on model estimation uncertainty . Therefore, we studied the effects of some key experimental conditions on estimation uncertainties of the Weibull parameters through the Monte Carlo simulations. Simulation results suggested that the estimated scale parameter would be more reliable than the estimated shape parameter from the tests. It was also shown that increasing the number of specimen would be more efficient to reduce the uncertainty of estimators than the more frequent censoring.

Plant Operating Experience

Laboratory Analysis of a Leaking Letdown Cooler from Oconee Unit 3

This paper covers the results of laboratory examinations performed on a leaking letdown coolerLetdown cooler from Oconee Unit 3. The laboratory scope included dewatering, pressure testing, visual inspections, metallography, Vickers micro-hardness, scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), X-Ray Diffraction (XRD), and Optical Emission Spectroscopy (OES). The laboratory examinations identified one tube containing a through-wall crack. The most likely cause of the crack appeared to be OD-initiated caustic stress corrosion cracking (SCC)Caustic stress corrosion cracking. The presence of heavy deposits on the tube OD surface and heat tinting on the primary and secondary flow seals indicated boiling occurred near the tight radius region of the bundle. Once boiling occurred, caustic-forming species such as calcium phosphate deposited and concentrated on the tube OD surface. The literature indicates as caustic concentrations approach ~20%, the conditions become favorable for caustic SCC to occur in austenitic stainless steels such as Type 316L.

Root Cause Analysis of Cracking in Alloy 182 BWR Core Shroud Support Leg Cracks

Cracks Crackwere found in the Alloy 182Alloy 182 weld and buttering of two core shroudCore shroudsupport legsSupport leg in the Forsmark BWRBWR Unit 1. This paper presents the root cause analysis based on mechanical and metallurgical analysis by Microscopy, Electron Backscatter Diffraction, and micro- and nano-indentation analysis along the crack path. The High Resolution Analytical Transmission Electron Microscopy examinations on the oxides are presented in detail in another paper at this conference. The morphology of the cracks, local hardness, crystallographic orientations and compositions along the crack path and crack tip regions were examined. The crack penetrated the boat keel surface into the A182 Low Alloy Steel mixing zone. Cracking was identified as IGSCC and signs of grinding and elevated micro-hardness near the boat top surface region were detected. Plastic deformation was observed in a crack tip region in a sample section with the largest crack depth, while not at the crack periphery.

Microbially Induced Corrosion in Firefighting Systems—Experience and Remedies

Firefighting water systems are important safety systems in all industries, including nuclear power plants (NPPs). However, they are susceptible to microbially induced corrosion, which is a degradation mode needing special attention. Leakages were observed in a fire fighting system made from stainless steel at a nuclear power plant shortly after maintenance and modernization work, which included replacement of part of the old carbon steel pipelines with stainless steel pipelines, as well as exchange of some Type 304 stainless steel pipes with Type 316 pipes due to relining parts of the system. The failure analysis revealed sub-surface corrosion cavities with pinholes at the inner surface and finally penetrating the whole pipe wall thickness. It was concluded that the reason for the leaks was due to microbially induced corrosion, (MIC). The paper will present the results from failure analyses, explain the remedial actions taken at the power plant, and discuss the implication of these findings on new similar systems, including the importance of avoiding iron deposits and optimization of water quality.

Managing the Ageing Degradation of Concealed Safety Relevant Cooling Water Piping in European S/KWU LWRs

Safety relevant cooling water pipingCooling water piping is designed to transport excess heat both from the reactor core and the fuel pool under all operational and emergency conditions. Plain carbon steels are used, with or without different types of protective coating systems. Outside of the reactor building, such piping is often buried or concealed, which also gives additional protection against damage. Relevant ageing mechanisms are shallow pit corrosion and microbiological induced corrosion (MIC). While these ageing mechanisms can lead to localized leaks and limited loss of cooling water, neither the mechanical integrity nor the cooling ability of the systems are compromised. The extent and ageing management of the mechanisms are described, based on more than 30 years of operating experience. This includes NDT results and post service examinations after retrofits. Based on this positive operational experience, two different NDT concepts are recommended to manage the long time operational ageing.

Identification of PWR Stainless Steel Piping Safety Significant Locations Susceptible to Stress Corrosion Cracking

Stress corrosion cracking (SCC) of stainless steelStainless steel was originally considered only an issue with boiling water reactors (BWRs), but operating experience has shown that this phenomenon also occurs in pressurized water reactors (PWRs), such as in off-chemistry locations of stagnant branch connection piping. In this paper, the safety significant stainless steel piping locations susceptible to SCC are identified for three representative PWR plants (Plant A [Babcock and Wilcox-designed], Plant B [Westinghouse-designed], and Plant C [Combustion Engineering-designed]). For the purpose of this paper, “safety significant” is defined as having a high consequence of failure as determined by the plant’s risk-informed in-service inspection (RI-ISI) program. Weld locations are considered susceptible to SCC when the water is stagnant and ≥200 °F during steady-state reactor operation. The results of this work will be used to develop guidance for selection of welds to inspect when addressing currently existing inspection requirements.

IASCC Testing—Characterization

On the Use of Density-Based Algorithms for the Analysis of Solute Clustering in Atom Probe Tomography Data

Because atom probe tomography (APT)Atom Probe Tomography (APT) provides three-dimensional reconstructions of small volumes by resolving atomic chemical identities and positions, it is uniquely suited to analyze solute clustering phenomena in materials. A number of approaches have been developed to extract clustering information from the 3D reconstructed dataset, and numerous reports can be found applying these methods to a wide variety of materials questions. However, results from clustering analyses can differ significantly from one report to another, even when performed on similar microstructures, raising questions about the reliability of APT to quantitatively describe solute clustering. In addition, analysis details are often not provided, preventing independent confirmation of the results. With the number of APT research groups growing quickly, the APT community recognizes the need for educating new users about common methods and artefacts, and for developing analysis and data reporting protocols that address issues of reproducibility, errors, and variability. To this end, a round robin experiment was organized among ten different international institutions. The goal is to provide a consistent framework for the analysis of irradiated stainless steels using APT. Through the development of more reliable and reproducible data analysis and through communication, this project also aims to advance the understanding between irradiated microstructure and materials performance by providing more complete quantitative microstructural input for modeling. The results, methods, and findings of this round robin will also apply to other clustering phenomena studied using APT, beyond the theme of radiation damage.

Comparative Study on Short Time Oxidation of Un-Irradiated and Protons Pre-Irradiated 316L Stainless Steel in Simulated PWR Water

Achieving a better understanding of the Irradiation Assisted Stress Corrosion Cracking resistance is one of the issues to improve the durability of Pressurized Water Reactors. To do so, assessing the interaction of irradiation defects with oxidation of internal vessel bolts, made of 316L alloy, is crucial. In this work we studied the effect of protons pre-irradiations at 1 dpa on the very first steps of oxidation (1 min < t < 24 h) in simulated PWR environment. The morphology of the oxide layer was investigated using optical microscopy and Scanning Electron Microscopy. The oxidation kinetics for short term oxidation is discussed based on the obtained results. It was observed that crystallographic orientation has an effect on the oxidation process. The level of cold-work and the presence of precipitates were taken into account and both seemed to accelerate the oxidation kinetic. Finally, irradiation also tended to speed-up the oxidation phenomenon.

Hydrogen Trapping by Irradiation-Induced Defects in 316L Stainless Steel

The irradiation-induced defects in stainless steel internal components of pressurized water reactors combined with hydrogen uptake during the oxidation process could be a key parameter in the mechanism for Irradiation-Assisted Stress Corrosion Cracking (IASCC). The ultimate aim of this study is to characterize the effects of irradiation defects on hydrogen uptake during the oxidation of an austenitic stainless steel (SS) in primary water. The focus was made on the interactions between hydrogen and these defects. A heat-treated 316L SS containing a low amount of defects is compared with ion implanted samples. Both materials were characterized by Transmission Electron Microscopy (TEM). Hydrogen uptake was then promoted by cathodic charging using deuterium as isotopic tracer for hydrogen. The deuterium distribution was first characterized by SIMS (Secondary Ion Mass Spectrometry) profiles. This technique highlighted some deuterium segregation in link with the localization of implantation-induced defects, i.e. dislocation loops and cavities. Using TDS (Thermal Desorption Spectrometry) experimental results and literature data, a numerical model was used to simulate the deuterium profiles, providing diffusion and trapping/detrapping information associated with irradiation defects in the 316L SS.

Grain Boundary Oxidation of Neutron Irradiated Stainless Steels in Simulated PWR Water

To elucidate the mechanisms of irradiation assisted stress corrosion cracking (IASCC), stress corrosion cracking (SCC) tests on 3 dpa, 19 dpa and 73 dpa neutron-irradiated 316 stainless steel were performed and the effects of irradiation on grain boundary (GB) oxidation were investigated. O-ring specimens were prepared from irradiated flux thimble tubes and a constant load SCC test was performed in a simulated pressurized water reactor primary water at 320 °C. After the SCC test, the oxidation condition of GBs was examined by transmission electron microscopy. Evidence of GB oxidation was found in all examined GBs, even at the relatively low dose of 3 dpa. The morphology of GB oxidation was sharp wedge-shaped. The average GB oxidation length at 3 dpa, 19 dpa and 73 dpa were 100 nm, 340 nm and 400 nm, respectively, indicating the promotion of GB oxidation due to irradiation. In the GB oxide, Fe and Ni depletion and Cr enrichment were observed. Also, Ni enrichment on GB was observed in front of the GB oxidation.

Irradiation Assisted Stress Corrosion Cracking (IASCC) of Nickel-Base Alloys in Light Water Reactors Environments—Part I: Microstructure Characterization

Nickel-base alloys, 625DA and 625Plus have received renewed interest as potential structural materials in nuclear reactors to replace the austenitic stainless steels, which show high susceptibility of irradiation-assisted stress corrosion cracking (IASCC). We investigated the microstructural response of both alloys after 2 MeV protons irradiated to 5dpa at 360 °C in the Michigan Ion Beam Laboratory (MIBL). Transmission electron microscopy was performed on plan-viewed samples with a depth range 9–12 μm prepared by jet-polishing. Detailed analysis included changes in phases, dislocation loops, voids swelling, and radiation induced segregation (RIS). Nano-scaled irradiation-induced precipitates and dislocation loops were pervasive. Voids were absent in these alloys. RIS occurred at random high angle grain boundaries examined. A complete characterization of the irradiated microstructure is required to understand their mechanical and IASCC behavior.

Irradiation Assisted Stress Corrosion Cracking (IASCC) of Nickel-Base Alloys in Light Water Reactors Environments Part II: Stress Corrosion Cracking

Radiation-induced microstructural changes control the Irradiation Assisted Stress Corrosion Cracking (IASCC)Irradiation-assisted stress corrosion cracking of core materials, which is a key factor in the extension of the operating lifetime of Light Water Reactors (LWRs). Nickel-base alloys are considered as potential structural materials to replace highly IASCC susceptible austenitic stainless steels. Constant extension rate tensile (CERT) tests were conducted on proton irradiated high strength nickel-base alloy 625 with two different heat treatment conditions (625Plus and 625DA) in both simulated BWR NWC and PWR primary water. Crack length per unit area and fraction of grain boundaries that cracked were used to assess the IASCC susceptibility. Both 625Plus and 625DA showed a very high IASCC susceptibility. 625DA also exhibited greater changes in all microstructure features than 625Plus.

Solute Clustering in As-irradiated and Post-irradiation-Annealed 304 Stainless Steel

A commercial purity 304SS was irradiated to 5 dpa Kinchin-Pease (10 dpa full-cascade) using 2 meV protons at 360 °C. Post-irradiation annealing (PIA) was applied to reduce or remove IASCCIrradiation-Assisted Stress Corrosion Cracking (IASCC) susceptibility. This paper focuses on the links between irradiation-induced hardening and irradiated microstructures of the as-irradiated and PIA conditions; the irradiated microstructure is assessed by transmission electron microscopy (TEM) and atom probe tomography (APT)Atom Probe Tomography (APT). Dislocation loops, Ni–Si clusters, and Cu-enriched clusters are present in the as-irradiated condition. When the dislocation loopsDislocation loops are removed by PIA, ~40% of the as-irradiated hardnessHardness remains and can be rationally attributed to the solute clustersSolute clusters still present in the PIA microstructure. The observations indicate that hardening in the as-irradiated condition is controlled by both dislocation loops and solute clusters and suggest that radiation-induced solute clusters may be important to detailed understanding of IASCC (irradiation-assisted stress corrosion cracking)Irradiation-Assisted Stress Corrosion Cracking (IASCC).

IASCC Testing—Initiation and Growth

Irradiation-Assisted Stress Corrosion Cracking Initiation Screening Criteria for Stainless Steels in PWR Systems

The Irradiation-Assisted Stress Corrosion Cracking (IASCC)Irradiation-Assisted Stress Corrosion Cracking (IASCC)initiationInitiation data for austenitic stainless steelsStainless steel in Pressurized Water Reactor (PWR)Pressurized Water Reactor (PWR) primary water environments were collected from available research programs and evaluated by the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP). The objective was to determine the relationship between applied tensile stress, neutron fluence, and initiation of IASCC at nominally constant load. Analysis of the available data shows that the applied tensile stress level for initiation of IASCC decreases with increasing neutron dose in PWR environments above the PWR threshold for IASCC of three displacements per atom (dpa). An apparent asymptotic value between approximately 30 and 35% of irradiated yield strength has been observed for neutron dose levels between approximately 10 and 100 dpa. Maximum testing times up to approximately 5000 h are now available, but these still are several orders of magnitude less than 60–80 year operating times. However, the results from this study can currently be used by the nuclear industry to assess the effects of irradiation on austenitic stainless steels in PWR systems as an indicator of the combination of stress and neutron dose at which IASCC becomes possible, particularly for subsequent license renewal (SLR) evaluations.

Novel Technique for Quantitative Measurement of Localized Stresses Near Dislocation Channel—Grain Boundary Interaction Sites in Irradiated Stainless Steel

A process for quantitatively measuring the residual stress near dislocation channel—grain boundary interaction sites has been developed especially for irradiated stainless steel using High Resolution Electron Backscatter Diffraction (HREBSD)High resolution electron backscatter diffraction. Tensile stress acting normal to the grain boundary at 15 different discontinuous channel—grain boundary sites were observed to be highly elevated, with peak stresses reaching ~2 GPa, which is roughly an order of magnitude greater than the stresses observed at sites where slip transfer occurred at the grain boundary. A clearly observable difference can be made between the stress profiles present at discontinuous and continuous channel—grain boundary interaction sites. This difference is consistent with the theory that high tensile stress at the grain boundary may be a key driving factor for the initiation of irradiation assisted stress corrosion cracksIrradiation assisted stress corrosion cracking.

IASCC Susceptibility of 304L Stainless Steel Irradiated in a BWR and Subjected to Post Irradiation Annealing

Post-irradiation annealing (PIA)Post-irradiation annealing was conducted to investigate the cause of irradiation-assisted stress corrosion cracking (IASCC)IASCC. The effects of PIA on irradiation hardeningHardening, dislocation channelDislocation Channels formation, and IASCC susceptibility were examined for a 304L stainless steelStainless steel irradiated to 5.9 dpa in the Barsebäck-1 reactor (Sweden). The annealing treatments were performed at temperatures in the range 450–600 °C and times ranging from 1–20 h. Longer annealing times and higher temperatures, as represented by iron diffusion distance, resulted in a significant reduction in irradiation hardening. IASCC susceptibility was measured for the as-irradiated and two PIA conditions (500 °C: 1 h and 550 °C: 20 h) via interrupted CERT tests under simulated BWR-NWC conditions. The annealing treatments progressively reduced IASCC susceptibility (as measured by the final intergranular fracture fraction) and dislocation channel density.

Irradiation Assisted Stress Corrosion Cracking Susceptibility of Alloy X-750 Exposed to BWR Environments

The effect of irradiation on stress corrosion cracking susceptibility and fracture toughness of alloy X-750 has been investigated. The material has been irradiated at a target temperature of 288 ℃ in the Advanced Test Reactor at Idaho National Laboratory to a fluence of approximately 1.93 × 1020 n/cm2 (E >1 meV). Stress corrosion cracking crack growth rates were determined in both unirradiated and irradiated materials in normal water chemistry and hydrogen water chemistry environments. Although the effect of irradiation on tensile properties and fracture toughness was observed, there was no significant effect of irradiation observed on the propagation rate of stress corrosion cracks.

Evaluation of Crack Growth Rates and Microstructures Near the Crack Tip of Neutron-Irradiated Austenitic Stainless Steels in Simulated BWR Environment

In order to understand irradiation-assisted stress corrosion cracking (IASCC) growth behavior, crack growth rates (CGRs), locally deformed microstructure, and oxide film properties for neutron-irradiated austenitic stainless steels (SSs) are investigated. Crack growth rate tests have been performed in simulated Boiling Water Reactor (BWR) water conditions (at ~288 ℃) on a neutron-irradiated 316L SS at ~12–14 dpa. After the crack growth rate tests, the microstructures near the crack tip of the CT specimens are examined with scanning transmission electron microscope (FE-STEM). In comparison with a previous study at <~2 dpa, the irradiated 316L SS at ~12 dpa shows a less benefit of low electrochemical corrosion potential (ECP) conditions on CGR. An inactive crack tip immersed over 1000 h was filled with oxides, while almost no oxide film was observed near the active crack front in the low-ECP conditions. In addition, a high density of deformation twins and dislocations were found near the fracture surface of the crack front. It is considered that both localized deformation and oxidation are possible dominant factors for the SCC crack growth in highly irradiated SSs.

Effect of Specimen Size on the Crack Growth Rate Behavior of Irradiated Type 304 Stainless Steel

Crack growth rate (CGR)Crack Growth Rate (CGR) testing in BWR normal water chemistry was performed on compact tension (CT) specimens of two different sizes (B = 8 mm and B = 19 mm), machined from a Type 304 SS core shroud at a dose of ~1 dpa. The objectives were to study the effect of specimen sizeSpecimen size on the CGR, and to determine the K validityK validity limit for a CT specimen dimension used in previous studies. The results show that for materials with significant strain hardening capacity remaining, there is no effect of specimen size on the CGR when testing is conducted at stress intensity factors valid according to ASTM E399 using the flow strength. For materials at higher dose in which the strain hardening capacity is lost or greatly reduced, a different K validity criterion might be applicable.

Plastic Deformation Processes Accompanying Stress Corrosion Crack Propagation in Irradiated Austenitic Steels

During stress corrosion crack propagation, stress values in the crack tip vicinity often exceed material yield stress limit. Plastic deformation processes may accompany and influence cracking. Here, stress corrosion crack propagation and deformation mechanisms were investigated using EBSD analysis. The investigated material was 304L Ti-enriched austenitic stainless steel irradiated to 10.4 dpa at 320°C in the BOR-60 fast reactor. Crack growth tests were conducted in a simulated Normal Water Chemistry (NWC) environment in the temperature range 288–320 °C using compact tension specimens. By analyzing crack trajectory and grain structure in the crack vicinity, it was established that grain orientation with respect to the acting stress direction was not a key factor controlling crack propagation. No crystallographic orientation susceptible to cracking was identified. Also, EBSD analysis revealed strong inhomogeneity in plastic strain distribution along the crack path. Most crack-adjacent grains remained virtually strain-free whereas few grains experienced strong plastic strain. These areas were presumed to be “plastic bridges” or “ductile ligaments.”

PWR Oxides and Deposits

Effect of Grain Orientation on Irradiation Assisted Corrosion of 316L Stainless Steel in Simulated PWR Primary Water

Simultaneous exposure of 316L stainless steel to a proton beam and high purity water containing 3 wppm H2 was used to study the effect of radiation on corrosion. Protons create displacement damage in the solid and radiolysis products in the water. Irradiations lasted 24 h at a damage rate of 7 × 10−7 dpa/s. The 316L was solution annealed at 1050 °C for 30 min, 5% cold worked, and heat treated at 1100 °C for 10 min resulting in 23 μm grains. Samples were pre-characterized by EBSD to correlate grain orientation with oxide properties measured by Raman spectroscopy and TEM. Following irradiation, hematite was identified exclusively in areas exposed to radiolyzed water, both under the beam and downstream. Inner oxide layers in the unirradiated region had a strong dependence on grain orientation, whereas the irradiated region has little to no grain orientation dependence.

Finite Element Modelling to Investigate the Mechanisms of CRUD Deposition in PWR

Corrosion Related Unidentified Deposition (CRUD)Corrosion Related Unidentified Deposition (CRUD) in PWR may cause severe issues, such as Tube Support Plate (TSP) blockage, fuel cladding cracking, and subsequently increased radiation doses for workers. The primary objective of this work is to develop an all-inclusive deposition model, which will reproduce the morphology and elucidate the contributing electrokinetic mechanisms. In this paper the development and verification of a model of the streaming current linking the potential distribution and the fluid flow behaviour using the Finite Element Method (FEM)Finite Element Method (FEM) is presented. In the model, coupled anodic and cathodic regions were found at the inlet of a pipe restriction, associated with a region of recirculating flow following the front facing step (FFS). The corresponding current densities and overpotential at the metal/solution interface were calculated. The coupled anode and cathode may explain the observed deposition process—generating deposits at the front facing step first, followed by a region free of deposits and then repeating ripples of deposited material. At the restriction outlet, a cathode was found which balances the current loops. In this paper, the simulated initiation and propagation processes of the electrokinetic depositionElectrokinetic deposition are presented.

Properties of Oxide Films on Ni–Cr–xFe Alloys in a Simulated PWR Water Environment

The iron contentIron content in Ni–Cr–xFe (x = 0–9 at.%) alloys strongly affected the properties of oxide films Oxide filmformed in a simulated PWR primary water environment at 310 °C. Increasing the iron content in the alloys increased the amount of iron-bearing polyhedral spinel oxide particles in the outer oxide layer and facilitated the local oxidation penetrationLocal oxidation penetration into the alloy matrix from the chromium-rich inner oxide layer. The local oxidation penetration was caused by the pile-up of the cation vacancies.

PWR Secondary Side

Effect of Applied Potential and Inhibitors on PbSCC of Alloy 690TT

Alloy 690TTAlloy 690TT has been shown to be susceptible to leadLead stress corrosion cracking (PbSCC)PbSCC at pHT values of 9.1 and above and with no cracks occurring at lower pHT values like 8.5. Previous work on PbSCC has been completed at the open circuit potential (OCP). A test program has been completed at applied electrochemical potentials for Alloy 690TT at pHT values of 8.5 and 9.5. Testing has shown sporadic PbSCC occurrence at pHT 8.5 which previously showed no cracking although an exact causal factor was not identified. At pHT 9.5, applying a potential did not stop PbSCC from occurring, although at +75 mV applied potential there was the inclusion of an apparent incubation time where none had previously been observed. To minimize to PbSCC, scoping testing was completed on four candidate inhibitorsInhibitor at pHT 9.5. Three of the four inhibitors tested (TiO2, H3BO3, and CeB6) showed reductions in maximum crack depths compared to testing without an inhibitor. Although promising, inhibitors still require additional testing before widespread use can be recommended.

Corrosion of SG Tube Alloys in Typical Secondary Side Local Chemistries Derived from Operating Experience

In spite of strong industry improvements (new alloys, more stringent chemistry control…), Outer Diameter Stress Corrosion Cracking (ODSCC) still occurs in modern SG alloys, although at a lesser extent than it used to affect 600MA tubesTubes . Currently, alloy 600TT, and even 800NG, do indeed suffer from ODSCC. Since the chemistry of the secondary sideSecondary side of PWR NPPs improved, past tests do not enable to assess simply the risks for modern alloys (too extreme pH for instance). Thus, IRSN reviewed what typical chemical conditions could be met in actual steam generatorsSteam generators. Based on this review, IRSN performed corrosionCorrosion tests in these conditions to assess the potential risks for alloys 600TT and 690TT in “typical” top-of-tubesheet environments. This paper will present the safety, technical and industrial frames of these tests, as well as the first results. An emphasis will also be put on the scientific implications of these results.

Investigation on the Effect of Lead (Pb) on the Degradation Behavior of Passive Films on Alloy 800

Alloy 800 has been demonstrated to be susceptible to Pb-induced degradation such as stress corrosion cracking (SCC) in laboratory experiments. PbPb has been proposed to cause such degradation by affecting the formation of protective oxides comprising the passive filmsPassive film on the alloys. To understand the detrimental effects of Pb, Alloy 800 samples pre-passivated under all volatile treatment (AVT) conditions were exposed to alkaline environment of pH280 °C 9.5 at 280 °C in the absence and presence of Pb. The Pb-free and Pb-containing passivated surfaces, along with bare surfaces as control samples, were characterized using electrochemical and surface analytical techniques. The results contrast the susceptibility of Alloy 800Alloy 800 to Pb-induced degradation reported in literature, where experiments are usually performed on bare surfaces with excess Pb, typically as PbO, present in the aqueous solutions; suggesting that the effect of Pb depends on whether it is present in solution or incorporated in the film.

Influence of Alloying on α-αʹ Phase Separation in Duplex Stainless Steels

Thermal embrittlement caused by phase transformations in the temperature range of 204–538 °C limits the service temperature of duplex stainless steels. The present study investigates a set of wrought (2003, 2101, and 2205) and weld (2209-w and 2101-w) alloys in order to better understand how alloying elements affect thermal embrittlement. Samples were aged at 427 °C for up to 10,000 h. The embrittlement and thermal instability were assessed via nanoindentation, impact toughness testing, and atom probe tomography (APT). Results demonstrate that the spinodal amplitude is not an accurate predictor of mechanical degradation, and that nanoindentation within the ferrite grains served as a reasonable approximate for the embrittlement behavior. Compositionally, alloys with a lower concentration of Cr, Mo, and Ni were found to exhibit superior mechanical properties following aging.

Stress Corrosion Cracking of Alloy 800 in Secondary Side Crevice Environment

Alloy 800 nuclear grade (NG) is a material of choice for replacement steam generators (SG) due to its inherent resistance to primary water Stress Corrosion Cracking (SCC)stress corrosion cracking (SCC). However, the long term performance of SGs depends on the performance of the material in upset conditions. Various degradation modes have been observed in Alloy 800NG under simulated secondary crevice environments (SCE) in C-ring and CERT experiments. Furthermore, the first incidences of SCE SCC have been observed in Alloy 800NG SG tubes in nuclear power plants and may be the sentinel events at the onset of more extensive cracking in the future. Understanding the parametric dependencies of SCC obtained under representative SCE and plausible transient conditions are keys to predicting future SG performance, validating mitigation strategies, and addressing life extension issues. The results of SCE crack growth rate (CGR) testing of Alloy 800Alloy 800NG in conditions representative of an acid-sulfate chemistry upset condition will be presented.

Using Modern Microscopy to “Fingerprint” Secondary Side SCC in Ni–Fe Alloys

Aggressive aqueous environments (Pb, S, pH extremes) used in laboratory tests have been shown to induce stress corrosion cracking (SCC) in Ni–Fe–Cr alloys. These conditions are used to simulate the extremes of secondary side crevice environments that are unlikely to occur under normal operating conditions but laboratory testing can still be used to establish sensitivities to abnormal chemistry conditions. Advances in modern microscopy have enabled the characterization of these secondary-side SCC systems at near-atomic resolution, helping to reveal mechanistic characteristics unique to each SCC mode. International progress investigating secondary-side SCC phenomena using analytical transmission electron microscopy (TEM)Transmission Electron Microscopy (TEM) is reviewed in this paper. The unique chemistry and degradation associated with different modes of SCC are identified and compared among Ni–Fe–Cr steam generator tube alloys of interest (Alloy 690 and Alloy 800). It is revealed that each SCC mode exhibits distinctive characteristics, or a “fingerprint”, which can be used to identify the aggressive environment responsible for inducing SCC.

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