Primary Water Stress Corrosion Cracking Research Articles

Estimation of residual stress in dissimilar metals welding using deep fuzzy neural networks with rule-dropout

Welding processes are used to connect several components in nuclear power plants. These welding processes can induce residual stress in welding joints, which has been identified as a significant factor in primary water stress corrosion cracking. Consequently, the assessment of welding residual stress plays a crucial role in determining the structural integrity of welded joints. In this study, a deep fuzzy neural networks (DFNN) with a rule-dropout method, which is an artificial intelligence (AI) method, was used to predict the residual stress of dissimilar metal welding. ABAQUS, a finite element analysis program, was used as the data collection tool to develop the AI model, and 6300 data instances were collected under 150 analysis conditions. A rule-dropout method and genetic algorithm were used to optimize the estimation performance of the DFNN model. DFNN with the rule-dropout model was compared to a deep neural network method, known as a general deep learning method, to evaluate the estimation performance of DFNN. In addition, a fuzzy neural network method and a cascaded support vector regression method conducted in previous studies were compared. Consequently, the estimation performance of the DFNN with the rule-dropout model was better than those of the comparison methods. The welding residual stress estimation results of this study are expected to contribute to the evaluation of the structural integrity of welded joints.

Characteristics of 3D Printed Functionally Graded Material for Replacement of Dissimilar Metal Weld in Nuclear Reactor

The dissimilar metal welds in the most of the reactors are connections between low alloy steel parts and stainless steel piping. There is a high possibility of primary water stress corrosion cracking (PWSCC) damage attributed to residual stress caused by the difference in material properties in the dissimilar metal weld joints. A number of accidents such as leakage of radioactive coolant due to PWSCC have been reported around the world, posing a great threat to nuclear safety. The objective of this study is to develop a technology that can fundamentally remove dissimilar metal welds by replacing the existing dissimilar metal parts with the functionally graded material (FGM) manufactured by metal 3D printing consisting of low alloy steel and austenitic stainless steel. A powder production, mixing ratio calculation, and metal 3D printing were performed to fabricate the low alloy steel-stainless steel FGM, and microstructure analysis, mechanical properties, and coefficient of thermal expansion (CTE) measurement of the FGM were performed. As a result, it is observed that CTE tended to increase as the austenite content increased in FGM. The gradual change of coefficient of thermal expansion in a FGM showed that the additive manufacturing using 3D printing was effective for preventing an abrupt change in thermal expansion properties throughout their layers.

Preferential intergranular oxidation as a potential degradation mechanism for Alloy X-750 CANDU spacers

Oxidation experiments on Alloy X-750 were performed in reducing CO-CO2 mixtures to assess the possibility of preferential intergranular oxidation (PIO) in relation to CANDU garter-spring spacers. The testing environments simulate hypothetical off-chemistry scenarios, promoting the oxidation of the alloying elements, but not Ni. Surface and cross-section microscopy corroborated by Auger electron spectroscopy confirmed the occurrence of internal and intergranular oxidation in Alloy X-750, accompanied by Ni nodule expulsion to the surface. Results are qualitatively similar to observations in Alloy 600 exposed to 480°C hydrogenated steam, used as an accelerated environment for primary water stress corrosion cracking (PWSCC) testing.

Optimization of dissolved hydrogen concentration for mitigating corrosive conditions of pressurized water reactor primary coolant under irradiation (1) evaluation of water radiolysis

ABSTRACT A major subject for evaluating the corrosive conditions in the pressurized water reactor (PWR) primary coolant is to determine the optimal hydrogen concentration for mitigating primary water stress corrosion cracking (PWSCC) without adverse effects on major structural materials. An analytical method combining water radiolysis and electrochemical corrosion potential (ECP) analyses was proposed for evaluating the corrosive conditions in PWR primary coolant. These procedures originally developed for boiling water reactors (BWRs) were successfully extended to PWRs with different water chemistry parameters, including pH, temperature, and radiation qualities, after minor changes in the original procedures and major input parameters, such as the inclusion of the effects of alpha radiolysis and Li+ (Na+) and H+ effects for the anodic polarization curve. This study discusses the results of water radiolysis analysis for PWR primary coolant, and the characteristic behavior of hydrogen peroxide (H2O2) as a function of hydrogen inlet concentrations in PWR primary coolant conditions (higher pH and α-ray irradiation). A possible reaction scheme involving eaq – was proposed for H2O2 suppression under alkaline conditions. The corrosive conditions were discussed in the following publication, using ECP as the major index for PWR corrosive conditions.

X-ray absorption study of oxide film formed on Alloy 600 in PWR primary water at 290°C with a decreased concentration of dissolved hydrogen

ABSTRACT Primary water stress corrosion cracking (PWSCC) of Alloy 600 remains an important degradation phenomenon in PWRs. In particular, the susceptibility to the PWSCC initiation at 290°C is of recent interest because that is one of the typical temperatures of the primary water in contact with Alloy 600 which remains in PWRs. In this paper, oxide film formed on Alloy 600 in simulated primary water at 290°C was analyzed by X-ray absorption fine structure (XAFS) measurements at the Fe K-edge. For the preparation of the oxide film, Alloy 600 was immersed in simulated primary water with dissolved hydrogen (DH) concentration of 5 or 30 ml/kg-H2O. The near edge region of the XAFS suggested that the oxide film for 5 ml/kg-H2O DH had a larger Fe3+/Fe2+ ratio than for 30 ml/kg-H2O DH. Additionally, the extended region of the XAFS revealed that Fe3+ mainly occupied the octahedral site of the spinel oxide. Previous reports indicate that an Fe3+-rich spinel oxide is less defective and protective against PWSCC initiation. Therefore, our results suggest that the lower DH concentration (5 ml/kg-H2O) could potentially mitigate the susceptibility of Alloy 600 to the PWSCC initiation at 290°C.

TARGET – Development of Submersible ROV System for BMN Inspection

Most PWRs have penetrations in the RPV lower heads for in-core nuclear instrumentation. These penetrations generally are made of nickel-based Inconel Alloy 600. Weld materials are typically Alloy 82/182. Operating conditions of PWR plants are causing nickel-based alloys cracking through a process called primary water stress corrosion cracking (PWSCC). In 2003, the licensee for the South Texas Project Unit 1 (STP-1) identified apparent boron deposits on the lower RPV head near two bottom mounted nozzles (BMNs). The NRC issued Bulletin 2003-02 to obtain information on licensee inspection activities and inspection plans for the RPV lower head. EPRI issued MRP-206 report that provides inspection and evaluation guidelines for BMNs for PWR plants, including guidelines for periodic bare metal visual examination for evidence of primary coolant leakage, or periodic non-visual nondestructive examinations for indications of service-induced cracking. The non-visual inspections (ultrasonic testing examination) may detect service-induced degradation before through-wall cracking, leakage, circumferential cracking below the bottom of the J-groove weld, release of loose parts, or incipient boric acid wastage of the low-alloy steel reactor vessel lower head material occurs. Therefore, periodic examinations will adequately manage potential for cracking by PWSCC and preserve structural integrity. INETEC developed TARGET system for BMNs inspection, consisted of submersible ROV and specially designed probe, composed of several UT probes. UT system and technique to detect, length and depth size the service-induced degradation in the BMN volume material is developed. The EPRI NDE Center performed a technical review and validated INETEC’s ultrasonic examination technique for BMNs. Aforementioned validation was done according to requirements defined by: 1) MRP-206, 2) MRP-411.

New displacement-based finite element analysis method for predicting the surface residual stress generated by ultrasonic nanocrystal surface modification

Peening techniques can be applied to relieve high tensile residual stress, which is one of the main causes of primary water stress corrosion cracking (PWSCC). Ultrasonic nanocrystal surface modification (UNSM) is a peening technique that generates a high compressive residual stress on the surface of a structure. Finite element (FE) analysis can be utilized for quantitative surface stress prediction according to the UNSM in various materials and loading conditions. In this paper, a new method to predict the surface residual stress generated by UNSM process through FE analysis is proposed. In order to accurately simulate the UNSM process through FE analysis, it is necessary to define an equivalent displacement calculation method that can replace the complex load states of UNSM. For this, we proposed the new equivalent displacement calculation method based on the contact mechanics theory and the energy conservation law. The new equivalent displacement calculation method was validated by comparing the indentation depth obtained by single-path UNSM FE analyses and experiments under various loading conditions. The actual UNSM process is carried out in multiple paths over a large area of the structure. By applying the new equivalent displacement calculation method, the compressive residual stresses were calculated through multi-path UNSM FE analyses and compared with the results obtained from the multi-path UNSM experiments. The surface residual stress results of the experiments and analyses showed good agreement, and it was confirmed that the UNSM process can be simulated well if the new equivalent displacement calculation method is used for the FE analysis.

Crack-Free 30% Chromium-Nickel Alloy Welding Products for Nuclear Service

Prior research in the development of 30% chromium-nickel alloy nuclear welding wires has resulted in the resolution of primary water stress corrosion cracking (PWSCC) and ductility dip cracking (DDC) as well as improvement in solidification cracking (SC) resistance. The resolution of DDC exhibits some Laves phase, which has a negative effect on SC resistance. In this study, the use of an alternate carbide former, tantalum (Ta), in combination with niobium (Nb) was researched. Three heats of recently designed Filler Metal 52MSS-Ta (i.e., HV1648, HV1673A, and VX131WXW) were melted, fabricated, and systematically studied. DDC and SC were evaluated with thermodynamic modeling using the Scheil solidification simulation model, two types of varestraint tests, and strain-to-fracture (STF) testing. The varestraint and STF test results showed an improved SC resistance with reduced Laves phase and concurrent excellent DDC resistance. Optimized compositions with low Laves phase also exhibited high threshold strain values (TSVs) in the STF test. VX131WXW — which contains 2.81 wt-% Ta, 0.6 wt-% Nb, and 6 wt-% iron (Fe) - exhibited a TSV of 24%. Thermo-Calc computed the Laves phase to be 0.24% for VX131WXW compared to 0.06% in HV1673A. This difference in Laves phase resulted in the lower SC resistance of VX131WXW compared to HV1673A when measured with longitudinal varestraint testing. The maximum crack distance for HV1673A was about 0.6 mm while that of heat VX131WXW was about 1.0 mm. The typical diluted weld deposit made with VX131WXW was also resistant to PWSCC due to the chromium content exceeding 24%. These simultaneous results mark progress toward crack-free welds and provide direction for further optimization of Ta-containing filler metals.

Influence of Dissolved Hydrogen Concentration on the Protective Property of Oxide Film of Alloy 600 Against Primary Water Stress Corrosion Cracking

Primary water stress corrosion cracking (PWSCC) of Alloy 600 has been a key issue in the field of nuclear energy. It is well known that dissolved hydrogen (DH) plays a crucial role in PWSCC initiation, and some studies showed that PWSCC initiation could be mitigated by decreasing the DH concentration. However, there is no consensus on the mechanism of the PWSCC mitigation in a low DH environment. Hence, to reveal the protective property of the passive film formed in various levels of DH, we analyzed the oxide film formed in simulated primary water at 345°C with the DH concentrations of 5 cc/kg and 30 cc/kg-H2O by electrochemical measurements (electrochemical impedance spectroscopy and Mott-Schottky test) and transmission electron microscopy (energy dispersive x-ray spectroscopy and electron energy loss spectroscopy) where the specimens of interest were tested by reverse U bend (RUB) tests, and it was confirmed that the PWSCC initiation time was longer for 5 cc/kg-H2O DH than for 30 cc/kg-H2O DH. Electrochemical measurements showed that the oxide film for the lower DH environment (5 cc/kg-H2O) had a higher electrical resistance and lower defect density than those for the higher DH (30 cc/kg- H2O). The microscopic observation indicated that the intergranular oxidation was relatively insignificant in the lower DH environment. The oxide film for both DH conditions consisted of outer oxides and inner Cr-rich barrier layer. The inner layer for 5 cc/kg-H2O DH had a higher concentration of Fe and a greater ratio of Fe3+/Fe2+ than that for 30 cc/kg-H2O DH. The results suggested that Fe3+ contributed to the formation of a less defective spinel-type structure in the inner oxide in the lower DH environment and thereby provided the alloy surface with corrosion protection. This protective oxide film prevented intergranular oxidation and led to the mitigation of PWSCC initiation in a low DH environment.

Bayesian approach for prediction of primary water stress corrosion cracking in Alloy 690 steam generator tubing

Alloy 690 tubing has been shown to be highly resistant to primary water stress corrosion cracking (PWSCC). Nevertheless, predicting the failure by PWSCC in Alloy 690 SG tubes is indispensable. In this work, a Bayesian-based statistical approach is proposed to predict the occurrence of failure by PWSCC in Alloy 690 SG tubing. The prior distributions of the model parameters are developed based on the prior knowledge or information regarding the parameters. Since Alloy 690 is a replacement for Alloy 600, the parameter distributions of Alloy 600 tubing are used to gain prior information about the parameters of Alloy 690 tubing. In addition to estimating the model parameters, analysis of tubing reliability is also performed. Since no PWSCC has been observed in Alloy 690 tubing, only right-censored free-failure life of the tubing are available. Apparently the inference is sensitive to the choice of prior distribution when only right-censored data exist. Thus, one must be careful in choosing the prior distributions for the model parameters. It is found that the use of non-informative prior distribution yields unsatisfactory results, and strongly informative prior distribution will greatly influence the inference, especially when it is considerably optimistic relative to the observed data.

Evaluating the reliability of PWSCC resistance of TT Alloy 690 and associated welds to the end of anticipated PWR plant life

In order to evaluate the primary water stress corrosion cracking (PWSCC) resistance and possible SCC growth rate (SCCGR) for thermally treated (TT) Alloy 690 to the end of anticipated PWR plant life, including various levels of cold work and associated welds, long-term PWSCC initiation tests using reverse U-bend specimens and SCCGR measurement tests were conducted at 360 °C in simulated PWR primary water. In addition, the possibility of long-range ordering (LRO) and/or short-range ordering (SRO) was evaluated by micro-Vickers hardness measurements, electron diffraction analysis and nano-scale EDS by high resolution TEM. The studied alloys were TT Alloy 690 and a Ni-33at%Cr model alloy after long-term ageing in air at 360, 400, 420 or 475 °C. No SCC was detected on any of the reverse U-bend specimens after testing for up to 45,000 h at 360 °C in simulated PWR primary water. Moreover, SCCGRs for the as-received TT Alloy 690 and with various levels of cold work or its associated welds did not increase when tested at 360 °C in simulated PWR primary water after thermal ageing at 360, 400, 420 or 475 °C in air for up to ∼20,000 h. In the case of the Ni-33at%Cr model alloy, LRO/SRO was clearly identified by electron diffraction analysis after thermal ageing at 475 °C in air. However, no evidence of LRO/SRO was observed in the case of the as-received TT Alloy 690 or after various levels of cold work or in its associated welds by any of the test methods after long-term thermal ageing in air at any temperature. From these test results, it is concluded that TT Alloy 690 and its associated welds in real components of PWRs have extremely high SCC resistance and extremely low SCCGRs should PWSCC be initiated, up to the end of any currently anticipated PWR plant life time.

Analysis of environmentally assisted crack initiation of Alloy 182 in simulated light water reactor primary water with the engineering initiation model EngInit

ABSTRACT Environmentally assisted cracking of nickel-base alloys has been observed in the primary coolant of light water reactors and is known as primary water stress corrosion cracking. In respect of crack growth, disposition curves are available for integrity analysis. However, in respect of initiation, there is no generally accepted approach. Here, a methodology is proposed to derive a usage factor towards initiation. Reaching a usage factor of 1 means that initiation is incipient. How far one is away from a usage factor of 1 tells us about the margin towards initiation. The key function in the approach is the rate at which the usage factor increases. It is proposed to be a function of strain rate and stress. Data from constant extension rate tests with flat, tapered tensile specimens have been used to parameterise the model. The parameterised model is subsequently used to predict initiation in a tensile-loaded, large-notch specimen.

Effect of heat treatment on grain boundary carbides and primary water stress corrosion cracking resistance of Alloy 182 weld

Alloy 182 welds were heat-treated at 600, 620, and 650 °C for 10 h. The primary water stress corrosion cracking initiation test showed significant improvement in the cracking resistance of the heat-treated specimens over the as-welded specimens. Microstructure analysis demonstrated that Cr23C6 grain boundary carbides dominated in the as-welded specimen, while Cr7C3 carbides were dominant after heat treatment. The depth of grain boundary oxidation ahead of the crack tip was large for the as-welded specimen, but less for the heat-treated specimens. Therefore, the presence of Cr7C3 carbides, in part, contributed to the improved cracking resistance of the Alloy 182 weld.

Analysis of dislocation density in strain-hardened alloy 690 using scanning transmission electron microscopy and its effect on the PWSCC growth behavior

The dislocation density in strain-hardened Alloy 690 was analyzed using scanning transmission electron microscopy (STEM) to study the relationship between the local plastic strain and susceptibility to primary water stress corrosion cracking (PWSCC) in nuclear power plants. The test material was cold-rolled at various thickness reduction ratios from 10% to 40% to simulate the strain-hardening condition of plant components. The dislocation densities were measured at grain boundaries (GB) and in grain interiors of strain-hardened specimens from STEM images. The dislocation density in the grain interior monotonically increased as the strain-hardening proceeded, while the dislocation density at the GB increased with strain-hardening up to 20% but slightly decreases upon further deformation to 40%. The decreased dislocation density at the GB was attributed to the formation of deformation twins. After the PWSCC growth test of strain-hardened Alloy 690, the fraction of intergranular (IG) fracture was obtained from fractography. In contrast to the change in the dislocation density with strain-hardening, the fraction of IG fracture increased remarkably when strain-hardened over 20%. From the results, it was suggested that the PWSCC growth behavior of strain-hardened Alloy 690 not only depends on the dislocation density, but also on the microstructural defects at the GB.

Development of probabilistic primary water stress corrosion cracking initiation model for alloy 182 welds considering thermal aging and cold work effects

We experimentally investigated the effects of thermal aging and cold work on the microstructure, mechanical properties, and primary water stress corrosion cracking (PWSCC) initiation time for Alloy 182 welds. The effects of thermal aging and cold work on the PWSCC initiation time of Alloy 182 were modeled based on the plastic energy concept and the PWSCC initiation data of this study and previous reports by considering censored data. Based on the results, it is estimated that the PWSCC resistance of the Alloy 182 weld firstly increases and then decreases with thermal aging time when the applied stress is kept constant.

Susceptibility to Severe PWSCC (primary water stress corrosion cracking) of LTMA (low temperature mill anneal) Alloy 600

It has been proposed that a primary water stress corrosion cracking (PWSCC) in pressurized water reactor (PWR) is governed by a lattice contraction due to a short range ordering reaction in Alloy 600. This leads researcher to think that the kinetics of lattice contraction may control a susceptibility of PWSCC in Alloy 600. A lattice variation with ordering treatment at 400 <sup>o</sup>C was systematically investigated using high resolution neutron diffraction(HRPD) in high temperature mill anneal (HTMA), low temperature mill anneal (LTMA), and sensitized (SEN) Alloy 600. The results showed that ordering treatment caused an isotropic lattice contraction due to short range ordering (SRO) reaction. The lattice contractions of (111) plane are saturated to be 0.04% in 4 to 1500 hours at 400 <sup>o</sup>C according to prior treatment condition. The lattice contraction in the magnitude of 0.03% of (111) plane in LTMA Alloy 600 is faster by 8 times and 66 times than that of SEN and HTMA, respectively. This fact may explain why the LTMA is most susceptible to PWSCC through of kinetics of lattice contraction in Alloy 600. Thus, it is possible to conclude that the susceptibility of Alloy 600 to PWSCC is governed by the kinetics of (111) lattice contraction.

Oxidation behaviour of solution-annealed and thermally-treated Alloy 690 in low pressure H2-Steam

Oxidation experiments were performed on Alloy 690 in a low pressure H2-steam environment at 480 °C to investigate the effects of the microstructure and thermal treatment on the “precursors events” associated with Primary Water Stress Corrosion Cracking. Advanced analytical electron microscopy characterisation revealed that the Alloy 690 did not exhibit preferential intergranular oxidation but rather only diffusion-induced grain boundary migration promoted by the formation of an external Cr2O3 oxide layer for both material conditions (solution-annealed vs. thermally-treated). These data demonstrate that for Alloy 690 the diffusion-induced grain boundary migration is a necessary but not sufficient condition for preferential intergranular oxidation.

Correlation between Grain Boundary Migration and Stress Corrosion Cracking of Alloy 600 in Hydrogenated Steam

Solution-annealed Alloy 600 samples were tested under active load conditions in a superheated low pressure H2-steam environment over a range of oxidising potentials relevant to Primary Water Stress Corrosion Cracking (PWSCC). Increased SCC susceptibility was observed for potentials the more reducing than the Ni/NiO transition where a deeper Preferential Intergranular Oxidation (PIO) occurred. Advanced microstructural analysis clearly showed that SCC initiated along the grain boundaries that exhibited Diffusion-Induced Grain Boundary Migration (DIGM), as well as enrichments in Al and Ti. Rather than the result of a single dominant mechanism, the initiation of SCC in Ni-base alloys involves the synergistic interactions between DIGM and the formation of Al- and Ti-enriched oxides, promoting PIO that can fracture with applied stress.

Improvement of Probabilistic Fracture Mechanics Analysis Code PASCAL-SP With Regardto PWSCC

Probabilistic fracture mechanics (PFM) analysis is expected to be a rational method for structural integrity assessment because it can consider the uncertainties of various influence factors and evaluate the quantitative values such as failure probability of a cracked component as the solution. In the Japan Atomic Energy Agency, a PFM analysis code PASCAL-SP has been developed for structural integrity assessment of piping welds in nuclear power plants (NPP). In the past few decades, a number of cracks due to primary water stress corrosion cracking (PWSCC) have been detected in nickel-based alloy welds in the primary piping of pressurized water reactors (PWRs). Thus, structural integrity assessments considering PWSCC have become important. In this study, PASCAL-SP was improved considering PWSCC by introducing several analytical functions such as the models for evaluation of crack initiation time, crack growth rate (CGR), and probability of crack detection. By using the improved version of PASCAL-SP, the failure probabilities of pipes with a circumferential crack or an axial crack due to PWSCC were numerically evaluated. Moreover, the influence of leak detection and nondestructive examination (NDE) on failure probabilities was detected. Based on the obtained numerical results, it was concluded that the improved version of PASCAL-SP is useful for evaluating the failure probability of a pipe considering PWSCC.

Effects of heat treatment on mechanical properties and sensitization behavior of materials in dissimilar metal weld

In order to mitigate primary water stress corrosion cracking (PWSCC) of dissimilar metal weld (DMW) in the primary coolant system of pressurized water reactor (PWR), a heat treatment has been considered. However, the heat treatment conditions should be carefully determined to prevent detrimental effect on the materials in the DMW region. In this study, changes in the mechanical property of materials in DMW and degree of sensitization of SS were evaluated for various heat treatment conditions. In addition, material properties variation and sensitization within the heat affected zone (HAZ) were investigated after the application of the heat treatment condition of 600 °C for 10 h, which is considered as the representative heat treatment condition. Based on the test results, it was found that the application of the representative heat treatment condition (600 °C for 10 h) would result in negligible detrimental effect on the mechanical properties of materials in DMW and sensitization of SS.