Primary Water Stress Corrosion Cracking Research Articles

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.

PWSCC growth rate model of alloy 690 for head penetration nozzles of Korean PWRs

Abstract This work aims to establish a model of a primary water stress corrosion crack growth rate of Alloy 690 material for the head penetration nozzles of Korean pressurized water reactors. The test material had an inhomogeneous microstructure with bands of fine-grains and intragranular carbides in the matrix of coarse-grains, which was similar to the archive materials of the head penetration nozzles. The crack growth rate was measured from the strain-hardened materials as a function of the stress intensity factor in simulated primary water at various temperatures and dissolved hydrogen contents. The effects of strain-hardening, temperature, and dissolved hydrogen on the crack growth rate were analyzed independently, and were then introduced as normalizing factors in the crack growth rate model. The crack growth rate model proposed in this work provides a key element of the tools needed to assess the progress of a stress corrosion crack when detected in thick-wall Alloy 690 components in Korean reactors.

An insight into PWR primary water SCC mechanisms by comparing surface and crack oxidation

Oxidation and stress corrosion cracking (SCC) of 316L stainless steel were studied in simulated pressurized water reactor primary water. Surface, crack flank, and crack tip oxides were analyzed and compared by high-resolution characterization, including oxidation state mapping. All oxides were found to have a triplex structure, although of different dimensions and composition, revealing the effects of local water chemistry and applied stress. The higher oxidation rate at the crack tip could be explained due to the existence of a higher dislocation density, higher level of stress and cation unavailability from the environment. The implications to SCC mechanisms are discussed.

PWSCC growth analyses by boundary element method considering weld residual stress distributions

Primary water stress corrosion cracking (PWSCC) is an important issue for dissimilar metal welds in pressurized water reactors due to mainly high weld residual stress (WRS). With regard to the PWSCC evaluation, while systematic stationary crack analyses ranging from initial to final lengths have been extensively adopted, they are repetitive and may not realistic. In the present study, growing crack analyses by boundary element method as well as a WRS were carried out for a typical pressurizer surge nozzle, and compared with those of conventional stationary crack analyses by finite element method and ASME code equations. Subsequently, sensitivity analyses were performed in use of three crack growth criteria having different equivalent stress intensity factor (SIF) definitions and their results were quantified to recommend practical one. As anticipated, ASME code provided the highest value due to inherent conservatism of the influence function method. Discrepancies in the SIF values between the growing and stationary crack analyses increased as the cracks became deeper and longer because their shapes were significantly distorted due to complex WRS distributions.

Effects of corrosion-inhibiting surface treatments on irradiated microstructure development in Ni-base alloy 718

The objective of this study is to understand the influence of surface treatments, namely laser shock peening (LSP) and ultrasonic nanocrystal surface modification (UNSM), on the irradiation tolerance of a Ni-base alloy. Surface treatments such as LSP and UNSM have been shown to mitigate the potential for primary water stress corrosion cracking (PWSCC). As such, there is emerging interest in the development and implementation of LSP and UNSM for nuclear reactor components. The LSP process utilizes a laser-generated plasma, while UNSM utilizes mechanical contact at ultrasonic vibration speeds, to create near-surface compressive residual stresses, high dislocation densities, twins and subgrains, and nanostructuring of the workpiece. These resultant microstructural changes can substantially affect the creation and evolution of irradiation damage. Herein, we study precipitation hardened Ni-base Alloy 718+, which is utilized in reactor components exposed to PWSCC-inciting environments. Specimens are treated with LSP or UNSM, then irradiated with 2.0 MeV protons to 7 displacements per atom (dpa) at 500°C. The dislocation line density is roughly an order of magnitude larger in the unirradiated LSP and UNSM specimens than in the baseline (untreated) specimen. Irradiation-induced dislocation loop nucleation results in an increase in the areal density of dislocation-type defects. Irradiation also induces disordering of the γ′ precipitates; this disordering appears more extensive in the baseline than in the LSP and UNSM specimens. Results are considered in the context of overall sink strength. Finally, irradiation-induced softening is observed in all specimens through nanoindentation, and is ascribed to the overall change in sink strength, resulting from the competition between γ′ disordering and dislocation loop nucleation.

Design, manufacturing and repair of tube-to-tubesheet welds of steam generators of CPR1000 units

This paper focuses on tube-to-tubesheet welds of Steam Generators by examining aspects of their design, materials and manufacturing to find solutions for restoring welds that had suffered from internal impact damage by loose parts. After looking into three designs of the weld, a flush-type design was adopted since it optimizes properties of the weld. Also the minimum size of the weld required in RCC-M was verified to be sufficient to sustain all of the loads and cycles involved in operation for 40 years. Then the materials and the manufacturing procedures were examined with regard to their effect on the susceptibility of the welds to Primary Water Stress Corrosion Cracking, and suitable resistance to such cracking was demonstrated. In addition, general welding defects which often existed during manufacturing were evaluated, and their repair methods by welding, including welding repairs after hydraulic test, were introduced. Ultimately, a repair method was developed to make the internal-impact-damaged SGs satisfy all structural requirements and resistant to Primary Water Stress Corrosion Cracking.

Measurement of the residual stresses in a PWR Control Rod Drive Mechanism nozzle

Residual stress in the welds that attach Control Rod Drive Mechanism nozzles into the upper head of a PWR reactor vessel can influence the vessel’s structural integrity and initiate Primary Water Stress Corrosion Cracking. PWSCC at Alloy 600 CRDM nozzles has caused primary coolant leakage in operating PWRs. We have used Deep Hole Drilling to characterise residual stresses in a PWR vessel head. Measurements of the internal cladding and nozzle attachment weld showed that although modest tensile stresses occur in the cladding, the attachment weld contains tensile residual stresses of yield magnitude. Despite the large dispersion of residual stress data for nozzle attachments of this type, all available data suggest that assuming a residual stress profile bounded by the weld material’s yield stress would be conservative for assessment purposes.

Virtual reality development and simulation of BMI nozzle inspection system for use during regular refueling outage of APR1400 family of reactors

Reactor vessel bottom mount instrumentation nozzle is J-groove welded to reactor vessel inside. This location has shown leaks in some NPPs in the past such as pinhole leaks at the South Texas Project in 2003 and a similar trace of leak at Palo Verde Unit 3 in 2013. Inspection of reactor vessel bottom mounted instrumentation nozzle weld crack requires removal of reactor internals in the case of APR1400 family of reactors without special tools. In Korea, the removal of reactor internals is prohibited during regular refueling outage following regulatory guidelines, hence the inspection of defects in reactor vessel bottom mounted instrumentation nozzle from inside of reactor vessel has not been performed during a regular refueling outage. In this study, a remote weld inspection system for reactor vessel bottom mounted instrumentation nozzles is developed in a virtual environment, and the operation of the inspection system was simulated to prove the validity of the system. The aim of the reactor vessel bottom mounted instrumentation nozzle inspection system was to inspect reactor vessel bottom mounted instrumentation nozzle during regular refueling outage without removing RI. This allows inspection of reactor vessel bottom mounted instrumentation nozzle weld for every refueling outage and allows finding defect at an early stage of defect formation and prevent reactor vessel crack caused by primary water stress corrosion cracking which lead to the bottom mounted instrumentation nozzle leak.

Evaluation of Primary Water Stress Corrosion Cracking Resistance of Three Heats of Alloy 600 in 400 °C Hydrogenated Steam Condition

For Alloy 600, primary water stress corrosion cracking (PWSCC) is one of the key material degradation mechanisms in pressurized water reactors (PWRs). To identify the governing factors of PWSCC resistance, a systematic investigation into the role of each factor was performed. A PWSCC initiation test was performed for 3 heats of Alloy 600 in the 400 °C hydrogenated steam condition. Based on the test results, the effects of known factors like chemical composition, mechanical strength, grain boundary carbide coverage, grain boundary character, and surface cold work on PWSCC resistance were discussed. In addition, surface oxide morphology and penetrative oxide depth was compared. From this study, grain boundary character was considered to be the most dominant factor affecting the PWSCC resistance.

PWSCC initiation and propagation in a CRDM penetration nozzle

In the winter of 2012, several cracks were detected in the penetration nozzles of the reactor pressure vessel closure head in a pressurized water reactor in Korea. Primary water stress corrosion cracking in the CRDM penetration nozzle is usually initiated due to a combination of a particular alloy, temperature, boric acid and weld residual stress. Garud et al. suggested an empirical equation for PWSCC initiation time based on a strain rate damage model. In this study, the authors introduce a user subroutine to apply Garud’s simplified SRDM model and XFEM to simulate PWSCC crack initiation location and growth. The suggested model includes a new algorithm to track cracks and update crack domains by simulating primary water contact conditions. A finite element model of the CRDM penetration nozzle were constructed to simulate such phenomena. Only one quarter of the nozzle were modeled because of its geometrical symmetry. In order to calculate the welding residual stress, a heat transfer analysis with a model change function followed by a thermal stress analysis were performed. Adopting the calculated welding residual stress as initial condition, the PWSCC crack initiation and propagation analysis were done for over 11 cycles of normal operating conditions. After 11 cycles of normal operation, the crack depth and length were measured and compared with the reported cracks from operating nuclear power plants in Korea. A comparison of PWSCC driving force among hoop stress, von Mises stress and maximum principal stress were investigated. Also a comparison of PWSCC dimension between two different Alloy 600 material conditions (MA, CW) were investigated.

Identification of nonregular indication according to change of grain size/surface geometry in nuclear power plant (NPP) reactor vessel (RV)-upper head alloy 690 penetration

Abstract During the fabrication process of reactor vessel head penetration (RVHP), the grain size of the tube material can be changed by hot or cold work and the inner side of the tube can also be shrunk due to welding outside of the tube. Several nonregular time-of-flight diffraction (TOFD) signals were found because of deformed grains. In this paper, an investigation of nonregular TOFD indications acquired from RVHP tubes using experiments and computer simulation was performed in order to identify and distinguish TOFD signals by coarse grains from those by Primary Water Stress Corrosion Crack (PWSCC). For proper understanding of the nonregular TOFD indications, microstructural analysis of the RVHP tubes and prediction of signals scattered from the grains using Finite Element Method (FEM) simulation were performed. Prediction of ultrasonic signals from the various sizes of side drilled holes to find equivalent flaws, determination of the size of the nonregular TOFD indications from the coarse grains, and experimental investigation of TOFD signals from coarse grain and shrinkage geometry to identify PWSCC signals were performed. From the computer simulation and experimental investigation results, it was possible to obtain the nonregular TOFD indications from the coarse grains in the alloy 690 penetration tube of RVHP; these nonregular indications may be classified as PWSCC. By comparing the computer simulation and experimental results, we were able to confirm a clear difference between the coarse grain signal and the PWSCC signal.

An extended finite element method‐based representative model for primary water stress corrosion cracking of a control rod driving mechanism penetration nozzle

AbstractPrimary water stress corrosion cracking incidents have been reported in nuclear reactors over the past several decades. Garud et al developed an empirical equation to express primary water stress corrosion cracking (PWSCC) initiation time by using experimental data. This strain rate damage model has been used in multiple simulation studies. Some of these studies used the extended finite element method (XFEM) to simulate the PWSCC propagation in Alloy 600. However, several studies showed that the accuracy of XFEM depends on the mesh quality. Different mesh qualities can change the heat flux of a welding procedure, leading to different weld residual stresses. We performed a parametric study on PWSCC initiation and propagation of a control rod driving mechanism by using different mesh qualities. The major variables explored here are number of elements per bead, number of circumferential elements, and number of weld beads. Finally, an XFEM‐based representative model was suggested for PWSCC initiation and propagation simulation.

Interacting sensitivities of alloy 600 PWSCC to stress intensity factor, yield stress, temperature, carbon concentration, and crack growth orientation Alloy 600

Analyses of Alloy 600 primary water stress corrosion cracking data show there are interacting sensitivities of crack growth rate to applied stress intensity factor, yield stress, temperature, carbon concentration and crack growth orientation. The effects of these variables are expressed quantitatively using a crack advance model that is rate-limited by crack-tip-strain rate. Crystallographic texture and dynamic strain aging are considered to account for contrasting behavior observed for ST and LT orientation crack growth. Crack growth is conjectured to be rate-limited by “junction” pinning of mobile dislocations strengthened by interstitial carbon (ST) and “lineal” pinning by substitutional Cr and Fe (LT).

Effect of Local Strain Distribution of Cold-Rolled Alloy 690 on Primary Water Stress Corrosion Crack Growth Behavior

Abstract This work aims to study the stress corrosion crack growth behavior of cold-rolled Alloy 690 in the primary water of a pressurized water reactor. Compared with Alloy 600, which shows typical intergranular cracking along high angle grain boundaries, the cold-rolled Alloy 690, with its heterogeneous microstructure, revealed an abnormal crack growth behavior in mixed mode, that is, in transgranular cracking near a banded region, and in intergranular cracking in a matrix region. From local strain distribution analysis based on local mis-orientation, measured along the crack path using the electron back scattered diffraction method, it was suggested that the abnormal behavior was attributable to a heterogeneity of local strain distribution. In the cold-rolled Alloy 690, the stress corrosion crack grew through a highly strained area formed by a prior cold-rolling process in a direction perpendicular to the maximum principal stress applied during a subsequent stress corrosion cracking test.

Estimation of residual stress in welding of dissimilar metals at nuclear power plants using cascaded support vector regression

Abstract Residual stress is a critical element in determining the integrity of parts and the lifetime of welded structures. It is necessary to estimate the residual stress of a welding zone because residual stress is a major reason for the generation of primary water stress corrosion cracking in nuclear power plants. That is, it is necessary to estimate the distribution of the residual stress in welding of dissimilar metals under manifold welding conditions. In this study, a cascaded support vector regression (CSVR) model was presented to estimate the residual stress of a welding zone. The CSVR model was serially and consecutively structured in terms of SVR modules. Using numerical data obtained from finite element analysis by a subtractive clustering method, learning data that explained the characteristic behavior of the residual stress of a welding zone were selected to optimize the proposed model. The results suggest that the CSVR model yielded a better estimation performance when compared with a classic SVR model.

Development and testing of a high-chromium, Ni-based filler metal resistant to ductility dip cracking and solidification cracking

Ni-based filler metals for repair and construction of nuclear reactor components require 30 wt% Cr to insure resistance to primary water stress corrosion cracking (PWSCC). Current generation alloys can be susceptible to solidification cracking and ductility dip cracking, depending on Nb content and the amount of eutectic that forms during weld solidification. Computational thermodynamic modeling was used to identify alternative eutectic-forming elements to Nb, such as Hf and Ta. Previous work has utilized a design of experiment methodology (DOE) in conjunction with computational and experimental techniques to optimize a new filler metal composition that should be resistant to weld cracking. In this paper, the optimized compositions were subjected to further weldability testing, namely the cast pin tear test (CPTT) and the strain-to-fracture (STF) test. An optimized Ta-bearing composition was found to be crack resistant, while an optimized Hf-bearing composition was highly susceptible to solidification cracking. Differences may be related to the nature of the eutectic constituents that form at the end of solidification. The Hf-bearing composition did not form Laves phase as its final eutectic constituent, while Laves is present in both the Nb- and Ta-bearing systems.

Thermal ageing and short-range ordering of Alloy 690 between 350 and 550 °C

Thermal ageing of Alloy 690 triggers an intergranular (IG) carbide precipitation and is known to promote an ordering reaction causing lattice contraction. It may affect the long-term primary water stress corrosion cracking (PWSCC) resistance of pressurized water reactor (PWR) components. Four conditions of Alloy 690 (solution annealed, cold-rolled and/or heat-treated) were aged between 350 and 550 °C for 10 000 h and characterized. Although no direct observation of ordering was made, variations in hardness and lattice parameter were attributed to the formation of short-range ordering (SRO) in all conditions with a peak level at 420 °C, consistent with the literature. Prior heat treatment induced ordering before thermal ageing. At higher temperatures, stress relaxation, recrystallization and α-Cr precipitation were observed in the cold-worked samples, while a disordering reaction was inferred in all samples based on a decrease in hardness. IG precipitation of M23C6 carbides increased with increasing ageing temperature in all conditions, as well as diffusion-induced grain boundary migration (DIGM).

비등수형 원자로 발전소에의 레이저 피닝 적용기술

Toshiba has developed a laser peening system for PWRs(pressurized water reactors) as well after the one for BWRs(boiling water reactors), and applied it for BMI(bottom-mounted instrumentation) nozzles, core deluge line nozzles and primary water inlet nozzles of Ikata Unit 1 and 2 of Shikoku Electric Power Company since 2004, which are Japanese operating PWR power plants. Laser pulses were delivered through twin optical fibers and irradiated on two portions in parallel to reduce operation time. For BMI nozzles, we developed a tiny irradiation head for small tubes and we peened the inner surface around J-groove welds after laser ultrasonic testing (LUT) as the remote inspection, and we peened the outer surface and the weld for Ikata Unit 2 supplementary. For core deluge line nozzles and primary water inlet nozzles, we peened the inner surface of the dissimilar metal welding, which is of nickel base alloy, joining a safe end and a low alloy metal nozzle. In this paper, the development and the actual application of the laser peening system for PWR power plants will be described. Key words: Fiber delivered laser peening, Pressurized water reactors, Primary water stress corrosion cracking, Reactor vessel, Laser ultrasonic testing

Structural Integrity Analysis of Nuclear Power Plant Pressure Vessel Penetration Nozzle Repaired

Primary water stress corrosion cracking (PWSCC) has been observed in CRDM nozzles, BMI nozzles and other penetration nozzles. The industry has used the repair method of replacement of nozzles fabricated of Alloy 690. After the replacement of the nozzle, the structural integrity analysis of new nozzle and welds should be performed to ensure the pressure boundary compliance with the original design requirement. In this paper, the pressurizer top head instrument nozzle of PWR nuclear power plant is evaluated as a typical pressure vessel penetration nozzle. The results showed that the repaired nozzle satisfies the ASME Code design requirement and the crack growth of the postulated flaw in 40 years of the nuclear plant life is acceptable.PWSCC degradation mechanism has been observed in CRDM nozzles, BMI nozzles and other penetration nozzles [1]. In some nuclear power plants built in China earlier, such as DAYABAY nuclear power plant and QINSHAN nuclear power plant, PWSCC degradation mechanism has been found in CRDM nozzle welds which manufactured of Alloy 600 and welded of Alloy 82/182[2]. The repair of the degraded nozzles is the popular choice for the nuclear power plant owners. After the replacement of the nozzle, the structural integrity analysis of new nozzle and welds should be performed to ensure the pressure boundary compliance with the original design requirement. In this paper, the pressurizer top head nozzle of PWR nuclear power plant is evaluated as a typical pressure vessel penetration nozzle. Stress intensities were conservatively determined for pressure and applicable thermal transients and compared to the allowable values of the ASME Code, Section III. Thermal stress of the transients was obtained from 3D finite element model (FEM). Residual stress of J-groove weld was obtained from 2D FEM analysis and used for fracture mechanics analysis. All of the analysis showed that the repaired nozzle satisfies the ASME Code design requirement and the crack growth of the postulated flaw in 40 years of the nuclear plant life is acceptable.

Effect of Heat Treatment and Surface Condition on Inter-Granular Corrosion of Alloy 800

Abstract The alloy 800 is being used for steam generator tubings for pressurized heavy-water reactors (PHWRs). This material is susceptible to degradation because of severe operating conditions, like high temperature, stress, and corrosive environment. These degradation mechanisms include primary water stress corrosion cracking (PWSCC), secondary side or outer diameter stress corrosion cracking (ODSCC), inter-granular corrosion (IGC), fretting, wear, denting, high cycle fatigue, corrosion fatigue, etc. The present study was conducted to analyze the effect of solution annealing temperatures, sensitization treatments, and surface conditions on the corrosion rate. The alloy 800 tubular samples were investigated by means of a conventional corrosion test (according to ASTM G28-02) and an electrochemical potentiokinetic reactivation (EPR) test. Susceptibility to inter-granular corrosion under various experimental parameters was examined by using both the test methods and the results are compared. The observed trends in corrosion rate obtained by using conventional method and EPR test were found to be similar. These results were used to obtain most optimum heat-treatment parameters and surface condition, which will yield best corrosion resistance.