2020 ABSTRACTS

The goal of the EU’s Horizon 2020 MEACTOS (Mitigating Environmentally-Assisted Cracking Through Optimization of Surface Condition) collaborative project is to improve the safety and reliability of Gen II and III nuclear power plants by improving the resistance of critical locations, including welds, to environmentally-assisted cracking (EAC) through the application of optimized surface machining and improved surface treatments (e.g., peening). The effect of surface machining and treatment techniques on the material’s EAC initiation behavior will be quantified using accelerated testing methods developed in NUGENIA+ projects MICRIN+ and ASATAR, e.g., constant extension rate tensile (CERT) testing using tapered specimens, as well as constant load tests. The link between laboratory testing and component behavior will be examined in terms of EAC models. The project is approx. in its mid-term stage (2017-2022) and beside an interesting state-of-the-art workshop and a summer school on nuclear corrosion, the material procurement, specimen manufacture and characterization have been carried out, whereas the EAC initiation test program is ongoing. Two FCC alloys (Alloy 182 weld metal and cold-worked type 316L stainless steel) and three surface finishes are investigated for their EAC initiation susceptibility: conventional and advanced machining, and in case of Alloy 182 a peening treatment. The current presentation briefly introduces the project, presents the test program and shows very first PRELIMINARY results.

Stress corrosion cracking (SCC) in reactor internals and recirculation pipes is one of the major degradation issues in boiling water reactors (BWRs). Efforts to provide mitigation of this cracking have led to the development of noble metal additions (Pt) to the reactor feed water. Under moderate hydrogen water chemistry conditions the Pt nanoparticles, formed in-situ and deposited on the reactor component surfaces, work as catalysts for the efficient reduction of the oxidizing species formed by radiolysis. The catalytic property of the surface is now able to lower the electrochemical corrosion potential without the negative side-effects of the classical hydrogen water chemistry. Noble metal chemical addition (or online NobleChem, OLNC) is in use in a large number of BWRs worldwide. To verify and optimize this technology, a research project was started at PSI in 2010, investigating the deposition and distribution behavior of Pt in BWRs. Experiments in a sophisticated high-temperature water loop, simulating BWR conditions, were con-ducted to systematically study the effect of different parameters on the Pt deposition and distribution behavior on steel surfaces. These investigations were complemented by exposure of specimens to reactor water in the BWR plant Leibstadt (KKL, Switzerland). Specimens were analyzed by laser ablation-inductively coupled plasma-mass spectrometry, high-resolution scanning electron and transmission electron microscopy, as well as other methods. From a scientific point of view and based on the results gained, some recommendations for OLNC may be formulated: The Pt injection should be performed at a slow rate under reducing conditions for extended periods of time. Applications should be repeated on a regular basis to compensate for Pt erosion and to protect new surfaces and they should start as soon as feasible after plant start-up. Pre-oxidation of new components may contribute achieving higher Pt loadings early on. Even though extremely small amounts of Pt can be sufficient to mitigate SCC, it is important to note that the Pt surface loading value alone is not sufficient to fully judge on the catalytic properties of OLNC-treated component surfaces. The Pt inter-particle distance seems to be a more suitable parameter. Despite the fact that many questions have been answered in the framework of this 9.5 year project, the final verification of the Pt particle distribution on actual BWR plant component surfaces is still incomplete.

In 2014, VTT broke ground for a new radiological facility, the VTT Centre for Nuclear Safety. By 2016 the laboratories were ready for moving in equipment. A suite of new hot cells was installed in 2017, and by mid-2018 the facility received all necessary licensing from the authorities for normal operation. Besides the hot cells, the facility includes a range of instruments for microscopy, radiochemistry and radiation measurements, as well as laboratories for testing nuclear aerosols (severe accidents), radio-iodine (NPP stack filters), and bentonite (long-term nuclear waste deep-repository tunnel backfill). This presentation describes the hot-cells, primarily designed for irradiated structural materials testing and characterization, and then gives examples of IASCC studies carried out on Flux Thimble Tube (FTT) material irradiated in the Ringhals Unit 2. The FTT material was utilized to produce O-ring specimens for autoclave testing. The as-received condition of the materials was also examined by analytical TEM. The material sampled came from two different FTTs with different peak doses (~65 and ~100 dpa). Two test runs were done, and two specimens were tested in each autoclave run. In each run, one 100 dpa and one 65 dpa specimen was tested with equal stress levels of about 65% of YS. The first run was done in normal PWR conditions with DH of 2 ~20 cc/kg, and the second with DH of 2 ~50 cc/kg. Following autoclave testing, the specimens were then examined by light macroscopy to identify cracking locations. As expected, cracks initiated from the inside surface at the top and bottom of the O-ring, and from the outside surface at both sides of the O-ring, corresponding to the greatest tensile stresses during loading. The cracking times were shorter for the higher dpa materials and for tests in elevated hydrogen. The results of advanced analytical electron microscopy characterization of the as-irradiated material showed that both 65 and 100 dpa material showed strong Ni and Si co-segregation both at grain boundaries and as precipitates in the matrix, as well as a population of fine cavities throughout the matrix. The cavities were more prominent in the 100 dpa material, but in both materials they were on the order of 1-2 nm in diameter. As prior cold-worked material, no linear dislocations remained in the microstructure, having been replaced by Frank loops, but deformation bands containing twinned and epsilon martensite were still evident, typical of cold-worked austenitic stainless steels. However, no evidence was found for alpha-prime martensite in the as-irradiated condition, indicating that such deformation phase was not induced during TEM specimen preparation.

Tensile tests in high-temperature air with pre-charged hydrogen and elastic plastic fracture mechanics tests in hydrogenated high-temperature water (HTW) at 250 and 288 °C on low-alloy reactor pressure vessel (RPV) steels revealed a clear but moderate reduction of ductility and fracture resistance, respectively. The observed behavior is a consequence of synergistic effects between hydrogen embrittlement (HE) and the dynamic strain ageing (DSA), in which the HE was amplified by a high DSA susceptibility. The deformation microstructures in the vicinity of the crack tips in air and HTW of two RPV steels with high DSA susceptibility were characterized in detail. These investigations support the idea that the environmental degradation of fracture resistance in hydrogenated HTW was mainly due to the plasticity localization by the interaction between DSA and hydrogen in RPV steels. Synergistic effects of DSA and hydrogen lead to heterogeneous distribution of dislocations and formation of dislocation cells inside bainitic laths.

It is well known that the austenitic stainless steels had shorter low-cycle fatigue lives in borated and lithiated high-temperature water with 100 ppb or greater dissolved oxygen concentrations than that in borated and lithiated high-temperature water with dissolved oxygen concentration less than 100 ppb. This is inexplicable as the dissolved oxygen could increase the corrosion rate of the materials, facilitate the crack propagation, and therefore the increase in dissolved oxygen concentration should reduce the low-cycle fatigue lives. To understand why the increase in the dissolved oxygen concentration increased the low-cycle fatigue lives, we investigated the low-cycle fatigue behaviors of 316LN austenitic stainless steel in the water containing dissolved oxygen less than 5 ppb and in the water with 100 ppb dissolved oxygen. Results suggested that the less hydrogen enhanced decohesion and the more protective oxide film formed in the water with 100 ppb dissolved oxygen caused the increase in the low-cycle fatigue lives. In addition, the effects of dissolved hydrogen on the low-cycle fatigue behaviors of 316LN was also investigated. Results indicated that increase in the dissolved hydrogen in the borated and lithiated high temperature water had no effect on the low-cycle fatigue life of 316LN and the oxide behaviors of this material. The amount of hydrogen absorbed into the 316LN only slightly increased with increasing dissolved hydrogen concentration, indicating the main source of the absorbed hydrogen was the hydrogen produced by the corrosion reaction.

SCC test in simulated BWR environment was conducted in order to investigate the effect of thermal aging on SCC crack growth rate of type 630 (17-4PH). Thermal aging increased the hardness of type 630, and SCC growth rate tend to increase with the increase of hardness. SCC growth rate of the thermally aged specimen with 401HV (1.2×10E-11 m/s) showed approximately 10 times higher than that of the unaged specimen with 346HV (7.6×10E-13 m/s). No clear difference in SCC propagation path and crack tip oxidation behavior was observed between unaged and aged specimen. From this point, a major factor that accelerates SCC growth rate is assumed to be the hardness. Therefore, SCC crack growth rate of type 630 can be estimated by the hardness.

The effects of inner surface finish condition of hollow specimens on the stress corrosion crack (SCC) initiation and small crack growth behavior of 316LN stainless steel were investigated in simulated pressurized water reactor (PWR) primary water at 325 °C with DH levels of 15 cc (STP) H₂/kg H2O by using slow strain rate tensile (SSRT) tests. The SSRT tests were interrupted at different strains (at 7.5 % strain and strain at the maximum load). The drilled specimen showed higher surface hardness and more enhanced SCC initiation and short crack growth behavior than honed specimens. Many cracks were observed along the gauge length direction even though the tests were interrupted at a small strain (about 7.5 %). It indicated that the crack initiated at the early stage of the specimen deformation. The maximum crack length of the drilled specimen was larger than those of honed specimens. The crack numbers of the drilled specimen are less than those of the honed specimens. Statistical results show that the length of most of the cracks was less than 0.2 mm for the specimen that was interrupted at the maximum load and at the strain of 7.5 %. Fractographic examinations showed mixed modes of transgranular and intergranular cracking.

Austenitic stainless steel is the regular material used in the canister of dry storage system, including Type 304, 304L, 316L stainless steels. When the spent fuel storage installations located nearby a coastal site, these types of austenitic stainless steel are prone to chloride-induced stress corrosion cracking (CISCC) in aggressive environment. Therefore, the purpose of this work is to evaluate the susceptibility to CISCC of candidate canister materials by using U-bend tests in simulated coastal atmospheric environments. The U-bend specimens were under periodic spraying with magnesium chloride solution (MgCl2), mixed chloride salt solution (NaCl and MgCl2) and synthetic seawater at different temperatures but constant relative humidity of 40 % for 1500 hours. Prior to the U-bend tests, specimens were prepared and underwent various pretreatments, including solution annealing and thermal sensitization. After the tests, specimens were examined with the scanning electron microscope (SEM) in order to observe the morphology and measure the length of cracks. According to the results of 40 °C, except for sensitized 304 and 304L specimens, no cracks longer than 500 μm were observed in the other U-bend specimens. The difference of resistance to CISCC between solution annealed and sensitized specimens are significant at low temperature. The number of cracks on the sensitized specimens was more than solution-annealed ones, mainly from the increasing of the micro-cracks and pits. The cracks observed on the sensitized specimens were longer than the solution annealed specimens because of the intergranular corrosion cracking. There are more cracks and pits observed at 60 °C under magnesium chloride periodic spraying than 40 °C. At higher temperature 80 °C, pitting coalescence is the major corrosion behavior instead of stress corrosion cracking. Furthermore, there is no significant influence of the test duration on the crack length and pitting morphology. Subsequently, the mixed chloride salt solution (NaCl and MgCl2) and synthetic seawater are going to be used for investigating the influence of different salt solutions on the tendency of CISCC. The cross section will be observed with the optical microscope to measure the depth of cracks in the future.

The SCC behavior of the 20 % CW Alloy 800H was investigated by crack growth rate (CGR) measurements in both supercritical water (SCW) and high temperature argon (HT-Ar). CGR of 20 % CW Alloy 800H are extremely high (over 2E-6 mm/s) in SCW environment at temperatures over 500 ℃. The apparent activation energy (Q) of SCC is 282.5 kJ/mol in SCW environment, which was slightly higher than the Q of creep crack in HT-Ar. Carbide was distributed continuously at the grain boundaries (GBs), and cavities with a diameter ~0.2 μm were observed on the GBs near the crack-tip. The TEM-EDS results shows that a Ni-enrichment and Cr-depletion zone occurred just ahead of crack-tips and a intergranular oxidation zone with a length of several micrometers was observed at the GBs ahead of crack-tips. Creep plays an important role in the crack growth of Alloy 800H in SCW environment. The presence of obvious GB oxidation beyond the crack-tip also indicates that corrosion-related GB oxidation might be a contributing factor.

The scratching repassivation in room temperature such as rotating disc electrode are often used to evaluate corrosion and stress corrosion cracking. However, for in-situ scratching repassivation test in high-temperature pressurized water it is very difficult to obtain valuable data. There are quite few papers related to the study of repassivation kinetics of nuclear materials using scratch electrode technique in high-temperature pressurized water. In order to evaluation the corrosion and stress corrosion cracking behavior of various materials rapidly, it is necessary to develop a in-situ scratching testing method in high-temperature pressurized water. The new in-situ scratching repassivation test system in high-temperature pressurized water has been built. The design temperature and design pressure of this system are 350 °C and 20 MPa, respectively, and the maximum scratch speed is 3.3 m/s. By using this scratch electrode system, the in-situ scratching repassivation kinetics of materials including Alloy 800, Alloy 690 and three high entropy alloys (Co1.5CrFeNi1.5Ti0.5Mo0.1, AlCoCrFeNiSi0.1 and TaNbHfZrTi ) are studied in high-temperature pressurized water with different dissolved hydrogen (DH) and dissolved oxygen (DO) contents at 300 °C for nuclear power plants. The results demonstrate that under DO condition the cBV value increases with the rise of DO concentration; and under DH condition, the cBV value passes through a local maximum at DH = 1.0 ppm. In addition, the repassivation results in the present work confirm the applicability of the slip-dissolution model for explaining stress corrosion cracking by means of electrochemical measurement. Compared to other methods for evaluating the stress corrosion cracking susceptibility of engineering materials, studying the repassivation kinetics by rapid scratch technique is a very economical and time-efficient evaluation method for nuclear materials in high-temperature pressurized water. The repassivation rates of the two alloys in high-temperature pressurized water were ranked in the following sequence: TaNbHfZrTi＞Co1.5CrFeNi1.5Ti0.5Mo0.1＞690TT＞AlCoCrFeNiSi0.1, which means that the proper high entropy alloy could have better corrosion and stress corrosion cracking behavior compared to traditional corrosion resistant alloys.

Some cracking-leakage failures happened in 316L and 304L stainless steel tubes of a heavy water collection system in a CANDU nuclear power plant, where the environment was high-temperature vapor at atmospheric pressure. To verify the failure cause and explore countermeasures, two parts of experimental research on SCC in simulated service environments were carried out. One part was conventional tests on SCC behavior of stainless steels 316L, 304L and Alloy 690 in boiling 42% MgCl2 at 155 °C, according to ASTM G30-97, in both the liquid and the vapor above. Severe SCC was observed in all the specimens of 304L and 316L, but it was mainly intergranular, different from the transgranular SCC which happened in the plant. No SCC but general corrosion was observed on the Alloy 690, especially in the vapor. Another part was tests in simulated service environments, i.e., in both the solution containing 0.7 ppm Li+ + 100 ppm Cl- and its vapor plus air or N2 at 250 °C. Two loading methods were used, one was slow strain rate testing as dynamic loading and the other was U-bend testing as static loading. Results showed that both, in the solution and in the vapor plus air, 316L and 304L were highly susceptible to SCC. The fracture morphology was all transgranular quasi-cleavage, similar to that of failed tubes in service. When pure nitrogen was used to replace the air above and to purge the solution, SCC extent was significantly decreased. Alloy 690 always exhibited excellent resistance to SCC. The results showed that the vapor was also a severe environment like the liquid to cause SCC. The main failure cause of the tubing in plants should be SCC in the vapor of leaked heavy water coolant. Countermeasures were suggested.

The role of hydrogen (H) in the modification process of the oxide film of Alloy 600 was investigated through a dual-exposed oxidation experiment with an in-situ H-charging method in a high-temperature high-pressure water environment. The selective dissolution of the NiO modified the oxide film into a defective and porous, single-layered Cr₂O₃ skeleton after oxidation with H-charging. The modification process depended on three different chemical states of H with a varied distribution: (i) H₂ molecule in water, (ii) neutral H in the grain boundary of the oxide film, and, (iii) proton (H+) in the lattice of the oxide film.

The stress corrosion cracking behavior of Alloy 600 was studied in high-temperature water at 288 to 360 °C. The effects of dissolved oxygen (DO) and dissolved hydrogen (DH) on crack growth rate (CGR) are discussed. Results show that the CGR of Alloy 600 in hydrogenated water (at the Ni/NiO phase boundary) is higher than in oxygenated water at 325 and 360 °C. High-resolution characterization of crack-tips revealed that the effects of DO and DH were partially correlated to the grain boundary oxidation. The higher CGRs in hydrogenated water might be attributed to more severe grain boundary oxidation ahead of the crack-tips.

A high-temperature recirculation loop was developed to simulate relevant secondary side boiling conditions in nuclear power plant steam generators under both normal operating conditions and during impurity transients. In-situ electrochemical impedance spectroscopy was used to study the corrosion behavior and impurity enrichment in stainless steel 316L and Alloy 690.

Up to now, no SCC failure of Alloy 690 HTT in NPPs has been reported publicly. However, structure wear has already become the main reason for the failure of SG tubes, especially fretting wear between SG tubes and the supporting structures. To carry out the laboratory investigation of fretting wear behavior of Alloy 690, a high-temperature high-pressure (HTHP) water fretting wear test equipment was firstly developed. In brief, the equipment was consisted of an autoclave, a vibration actuator, a normal load control system, a water chemistry measuring and adjusting system, a heating system and a water circulation system. The three key mechanical parameters (normal load, displacement and frequency) can be controlled with very high accuracy, e.g. the normal force can be applied with an accuracy of ±0.3 N. In addition, the tangential friction force and the displacement can be quickly (100 μs) measured and recorded to form the fretting logs. Then, the effects of normal force, displacement and temperature on the fretting wear behavior between Alloy 690 HTTs and their supporting structures in HTHP water that simulating the secondary water condition are studied, with detailed characterization of the worn scars. The following conclusions can be drawn: (1) The fretting wear mechanism between 690 and 405 changes with increasing temperature in pure water, from adhesive wear at room temperature to abrasive wear and delamination at 100 °C to abrasive wear at 200 °C. The severest fretting wear happens when the test temperature is 100 °C. (2) The fretting wear mechanism between 690 and 405 in HTHP water also changes with different normal force and displacement combinations. When the displacement keeps constant (±40 μm), the wear mechanism changes from abrasive wear, delamination and oxidation wear to adhesive as the increasing of normal force. However, wear mechanism changes from adhesive to abrasive wear, oxidation wear and delamination as the increasing of the displacement when the normal force keeps steady (110 N). Surface abrasion (or material loose) is the main damage for abrasive wear and delamination, while no obvious damage is found when Alloy 690 and 405 SS keeps adhesive during the whole test. In addition, the coefficient of friction, wear volume and maximum wear depth decrease with the increasing of normal force, while they increase with the increasing of displacement. As the limitation of fretting wear data from HTHP water, the results of this study are helpful for the safe operation of the NPPs.

Alloy 82 and 182 materials are used for various attachment welds to the reactor pressure vessel (RPV) and other dissimilar welds. Typically, the root of a Ni-base weld is Alloy 82 and remainder is Alloy 182. Although Alloy 82 is deemed to be more resistant to cracking, if indications were detected in Ni-base welds, Alloy 182 is considered to be more limiting for flaw evaluation because of its material properties. However, if it can be demonstrated that Alloy 82 has a significant factor of improvement (FOI) in crack growth rate (CGR) compared to Alloy 182, the by-product of this work could significantly extend inspection intervals for certain welds and lead to changes in the relevant Codes. This presentation provides results to-date from an EPRI project that has convened a panel of worldwide experts in the field of CGR testing. The Expert Panel is screening all available CGR data and is developing recommendations for stress corrosion crack (SCC) CGR disposition curves for Alloy 82 in low electrochemical potential (ECP) environments. The project is also assessing additional test data that are under development by the Japanese BWR Owner’s Group (JBOG). This project uses EPRI-accepted protocol for creating ad-hoc committees/expert panels to review existing field data/experience, laboratory data and other relevant experience, as well as separately developed contractor information, to create a consensus-based CGR database and associated technical reports.

The SCC growth rate of several heats and welds of Alloy 82 was evaluated in high-temperature water as a function of corrosion potential, chloride/sulfate, post weld heat treatment, stress intensity factor and temperature spanning over 100,000 hours of testing. A database of these results was created and evaluated to show the overall effects of key parameters.

Ni-base Alloy 82 weld metal is widely used in structural components in BWRs. With its higher SCC resistance in BWR environments than Alloys 182 and 132, fewer crack incidents have been experienced in operating BWR units. However, recent crack incidents led to a common recognition: it is appropriate to develop an SCC growth rate disposition curve for Alloy 82 in BWR environments in order to properly evaluate the impact of possible cracks in Alloy 82 on plant safety. Upon the request from Japan BWR owner’s group (JBOG), the international expert panel organized by EPRI has been evaluating the crack growth rate database of Alloy 82 in BWR environments that consists of 331 data available to date. Only approx. 45 % of data have passed the experts’ screening and few data remained at lower K range of <30 MPa√m. A collaborative research project among six international laboratories was conducted by JBOG since 2018 to support the disposition curve development. This presentation describes experimental evaluations by TEPCO to accumulate high-quality CGR data at lower (<30 MPa√m) and higher (>65 MPa√m) K ranges for the accountable determination of the K dependency (the slope) of the disposition curve.

Ni-base Alloy 82, which is used as weld metal for boiling water reactor (BWR) components, is known to have better resistance to stress corrosion cracking (SCC) than Alloy 182. In recent years, it was reported that SCC was detected on Alloy 82 weld metal of access hole cover at the Shimane nuclear power plant unit 2 in Japan. It is concerned that SCC damage of components made of Alloy 82 may occur in operating plants. Therefore, in order to evaluate structural integrity and inspection interval of the components correctly, it is urgent to develop the disposition curve of SCC growth rate for Alloy 82. In Japan, a lot of SCC growth rate data for Alloy 82 in BWR environments have been obtained by the studies sponsored by the Japanese government and those of the BWR owners group; however, a disposition curve of SCC growth rate for Alloy 82 in BWR environments has not been established in the “Rules on Fitness-for-Service for Nuclear Power Plants” of the Japan Society of Mechanical Engineers (JSME FFS Code). In order to apply the disposition curve in Japan, it is necessary to be endorsed not only in Japan but also in overseas. Based on the above background, currently an Expert Panel (EP) organized by the Electric Power Research Institute (EPRI) has been reviewing the database of SCC growth rate for Alloy 82 obtained in simulated BWR reactor water environments. The database includes parameters such as K value, water chemistry, potential drop method data and fracture surface information. All the data of the database has been reviewed and scored in the viewpoint of data quality by the EP, and the EP plans to establish a disposition curve of SCC growth rate for Alloy 82 in BWR environments based on high score data. While scoring work was progressed, it was found that not enough data with high score at low and high K regions remained. Therefore, in this study, SCC growth tests were conducted to obtain enough SCC growth rate data on Alloy 82 at low and high K regions. As a result, most of the SCC growth rate data, that was obtained in this study, were located near the high score data group, but the data in that group were widely scattered. Charge of the test material, weld type, and heat treatment condition for these data were not necessarily the same. In this study, the data were categorized based on yield strength, which is one of the representative material factor. As a result, it was found that SCC growth rate was increased and gradually saturated with the increase in yield strength. Moreover, SCC growth rate data for the group with high yield strength were less scattered than those of low yield strength.

Stress corrosion cracking (SCC) propagation tests were conducted on Alloy 82 in BWR NWC and HWC conditions for the development of SCC crack growth rate (CGR) disposition curve, which is currently being reviewed by the members of a EPRI expert panel (EP). As one of the possible factors that affect CGR, the effect of welding methods was discussed in this presentation. Compact tension (CT) specimens were fabricated from two types of weld joints: TIG (Tungsten Inert Gas) and MIG (Metal Inert Gas). The SCC propagation test results of Alloy 82 showed that the CGR of MIG tended to be slightly lower than those of TIG. In order to investigate the reasons for this difference, Vickers hardness measurements, microscopic observation and electron backscatter diffraction (EBSD) analysis were conducted on the same butt-weld plates from which the CT specimens were fabricated. No obvious difference in Vickers hardness distribution was confirmed in the weld metal between TIG and MIG specimens. However, the growth direction of columnar grains or dendrites tended to change more at weld pass boundary in MIG than in TIG specimens. The grain shape analysis performed by EBSD indicated that more columnar grains or dendrites growing in directions different to the SCC propagation in MIG than in TIG specimens. It was also indicated by fracture surface observation that SCC propagation direction might have changed in MIG specimen due to its specific microscopic features. Based on these results, we presume that the lower CGR in MIG could be relevant to (1) grain or dendrite boundaries growing in directions different to SCC growth and (2) the discontinuity of their growth direction at weld pass boundary. The specific features of microstructure in MIG are considered relevant to the spray transfer behavior of droplets. However, TIG-like microstructure can also be expected if changing some parameters such as heat input, etc.

Crack growth rate (CGR) testing on specimens of Alloy 82 weld metal in BWR environments has been performed. The objective of the study was to generate high-quality stress corrosion CGR data at stress intensity factors in the range of 30 to 40 MPa√m, and at very high K (80 MPa√m) in BWR normal water chemistry (NWC) and hydrogen water chemistry (HWC). Two TIG welds and one MIG weld of Alloy 82 in the as-welded condition were tested. One of the TIG welds was also tested in a post weld heat treated (PWHT) condition. The scope of work covered testing of ten compact tension (CT) specimens of three different sizes; two 0.5TCT-, six 1TCT- and two 1.5TCT-specimens. The tests showed that the as-welded conditions cracked faster than Alloy 82 in the PWHT condition. All data in the as-welded condition related to NWC fall above the Swedish disposition line, and this set of data suggests a K1.0 dependency. Regarding the PWHT condition and NWC, most data are enveloped by the disposition line, and in this case a K2.5 dependency is observed. In addition, HWC was effective in mitigating cracking, especially in the PWHT condition, and the factor of improvement relative to NWC ranges from 2 to 15. Sulfate was added to enhance the degree of intergranular cracking and this specie had a minor effect on the CGR relative to clean environments. The data also indicated that the specimen size has no effect on the CGR data.

In recent years several stress corrosion cracking (SCC) incidents occurred in welds of boiling water reactors (BWRs) made of Ni-base Alloy 182. This material is widely used in light water reactors as weld filler metal to join the low-alloy steel reactor pressure vessel to both wrought Ni-based alloys (e.g., Alloy 600) or austenitic stainless steels (e.g., AISI 304L or 316L) by manual shielded metal arc welding. Therefore, a systematic parameter study is performed at PSI to investigate the SCC initiation behavior of such weld metals. In BWRs a rather wide range of temperatures is present, which generally can have a strong effect on the SCC behavior in Ni-base alloys. Thus, temperature effects on the SCC initiation behavior of an Alloy 182 weld metal are studied. Because SCC mitigation methods often involve surface modifications, different surface treatments are also examined on their SCC initiation properties. The effect of advanced surface treatments like industrial surface treatment, advanced machining and peening is investigated additionally in the framework of a collaborative EU project (MEACTOS) with a second Alloy 182 weld. A comparison of the microstructure of both Alloy 182 weld metals (the one used in the MEACTOS project and in the PSI internal project) is expected to improve the understanding of the SCC initiation behavior. First preliminary results on the influence of temperature and surface condition on the SCC initiation behavior of Alloy 182 weld metal in high-purity, high-temperature water using accelerated constant extension rate tensile (CERT) tests with flat tapered specimens are presented. A clear trend for higher SCC initiation susceptibility in Alloy 182 towards higher temperatures could be identified, whereas the influence of the surface treatment is not fully conclusive yet.

The weld filler metal Alloy 182/82 has been used to join the low-alloy metal nozzle of the reactor pressure vessel and stainless steel safe end. In modern PWR plants, it often has been replaced by Alloy 52M or Alloy 52M has been used as repair weld metal due to its high SCC resistance with higher chromium content. Unlike bulk metals, deposited weld metals exhibit heterogeneous properties such as dilution of chromium, residual stress, special type of dendrite boundary, precipitation of carbides and depletion of chromium on dendrite boundary due to the welding and the post weld heat treatment process, which may affect the PWSCC behavior of weld metal. In the dilution zone of Alloy 52M weld metal near the fusion boundary, obvious depletion of chromium was observed at the dendrite boundary. The oxidation rate and PWSCC crack growth rate in the Alloy 52M dilution zone were found to be higher than that in the undiluted region. With the increase of post weld heat treatment time, the depletion of chromium was alleviated and the corresponding SCC crack growth rate decreased. The PWSCC behavior of Alloy 52M weld metal in the dilution zone was related to the microchemistry at the dendrite boundary and the resultant oxidation kinetics.