Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors

Materials degradation phenomena, if not appropriately managed, have the potential to adversely impact the design functionality and safety margins of nuclear power plant (NPP) systems, structures and components (SSCs). Therefore, the U.S. Nuclear Regulatory Commission (NRC) has initiated an over-the-horizon multi-year research Proactive Management of Materials Degradation (PMMD) Research Program, which is presently evaluating longer time frames (i.e., 80 or more years) and including passive long-lived SSCs beyond the primary piping and core internals, such as concrete containment and cable insulation. This will allow the NRC to (1) identify significant knowledge gaps and new forms of degradation; (2) capture current knowledge base; and, (3) prioritize materials degradation research needs and directions for future efforts. This effort is being accomplished in collaboration with the U.S. Department of Energy’s (DOE) LWR Sustainability (LWRS) program. This presentation will discuss the activities to date, including results, and the path forward.

The Swedish reactor fleet consists of 7 BWRs and 3 PWRs which have been operating for up to 38 years and all have plans for long term operation (LTO).The Swedish Radiation Safety Authority (SSM) has carried out an investigation to identify possible improvements to assessments for safe LTO. The study covered ageing mechanisms for metallic materials and polymers, concrete structures, electrical and I&C equipment, as well as in-service inspection. The paper will concentrate on metallic materials.Most of the degradation mechanisms are controlled satisfactorily by the licensees through the existing inspection and ageing management programs. These should be intensified and reviewed with respect to LTO. The consequences of ageing degradation mechanisms for LTO should be analyzed and reported. Two ageing mechanisms are highlighted, low cycle fatigue in LWR environments and the embrittlement of RPV steels. However, analyses should be carried out for most of the ageing mechanisms to find early indications of degradation and thus ensure safety.

Materials currently used in nuclear power plants are reliable and are generally resistant to environmental degradation. However, occurrences of environmental degradation have been observed as the current fleet of reactors ages. Primary water stress corrosion cracking (PWSCC) is of particular interest to the US Nuclear Regulatory Commission (NRC). This paper provides a historical assessment of operating experience associated with PWSCC and welding issues associated with PWSCC resistant materials. The paper also considers the regulatory issues associated with PWSCC, especially those associated with gaps in the understanding of the behavior of PWSCC resistant material under actual reactor conditions.

Alloy 690 has been widely used in fabricating components of LWR plants as an alternative material to Alloy 600 which has exhibited a significant susceptibility to PWSCC. However, some authors have reported that Alloy 690 can suffer a significant susceptibility to SCC crack growth when highly cold worked. While most of the recent studies emphasize SCC propagation phase, EDF and its partners are focusing on the material’s resistance to SCC initiation. This paper summarizes the current work carried out at EDF MAI on the SCC initiation. By means of constant elongation rate tests (CERTs) and constant displacement tests, experimental investigation of the susceptibility to PWSCC were performed. No SCC was observed on either an extruded bar or on two plates, even after 24%-1D cold rolling, confirming the superior PWSCC resistance of Alloy 690 independent of a amount of intergranular precipitation of carbides, and also revealing that such cold rolling does not necessarily decrease the resistance to SCC. On the other hand, a experimental steam generator tube that has a degraded microstructure due to specific heat-treatment revealed its susceptibility to SCC, probably because of the interactive effect of microstructure with heavy intragranular carbide precipitations and the cold worked superficial layer. This phenomenon is in good agreement with results previously published. In this study, the maximal crack depth slightly increased when DH increased from 5 to 60 cc.kg−1H₂O. No significant prior ageing effect on the crack depth was observed, even when ageing was combined with high DH.

The role of cavity formation on crack growth of intergranular stress corrosion cracking (IGSCC) in high temperature water and creep cracking was examined in quantitatively for cold worked carbon steel (ASTM A106 (UNS K03006, CW carbon steel), Alloy 690 (UNSN06690, CW TT 690), and Alloy 600 (UNSN06600, CW MA 600). Three important patterns were observed: First, cavities were observed at grain boundaries just ahead crack tips and crack wall after tests in gas and high temperature water for all test materials. Second, population of cavities decreased with distance from crack tips and crack wall. This result seems to suggest that cavities will form at high stress region such as crack tips before crack advance as crack embryos. Third, good correlation was observed between the rate of cavity formation and crack growth of IGSCC and intergranular creep cracking not only carbon steel, but also TT690 and MA600. Finally, the formation of crack embryos from the collapse of vacancies induced by cold work and absorbed hydrogen play an important role on the process of crack growth both of SCC and creep for CW carbon steel, CW TT690, and CW MA600 in high temperature water.

While plant experience so far has shown excellent resistance of Alloy 690 to stress corrosion cracking in PWR primary water environments, laboratory tests have reported that susceptibility may be enhanced substantially by non-uniform cold working, particularly when the plane of crack growth is in the plane of rolling or forging. The Alloy 690 program aims to further the understanding of the mechanisms behind this susceptibility and the heat-to-heat variability reported for different materials.This paper contains results from crack growth tests and related metallography for different heats of Alloy 690.The four heats of Alloy 690 tested in this program possess a range of microstructural properties from heavily banded to highly homogeneous. All of the heats were susceptible to IGSCC propagation to some degree after being subjected to non-uniform cold working, with crack growth rates varying between approximately 1x10−11 to 5x–10−10 ms−1 after correcting for incomplete intergranular engagement of the crack front in some specimens. The results show that the level of inhomogeneity in the microstructure (i.e. banding) is not necessarily an indication of the level of susceptibility. The cause of higher susceptibility to IGSCC and the plant relevance of a highly cold worked material are subjects of much interest and debate.

Stress corrosion crack-growth experiments have been performed on cold-rolled alloy 690 materials in simulated PWR primary water at 360°C. Extruded alloy 690 CRDM tubing in two conditions, thermally treated (TT) and solution annealed (SA), was cold rolled (CR) in one direction to several reductions reaching a maximum of 31% and tested in the S-L orientation. High stress corrosion cracking (SCC) propagation rates (~8x10−8 mm/s) were observed for the 31%CR alloy 690TT material, while the 31%CR alloy 690SA exhibited 10X lower rates. The difference in intergranular SCC susceptibility appears to be related to grain boundary carbide distribution before cold rolling. SCC growth rates were found to depend on test temperature and hydrogen concentration. Tests were also performed on two alloy 690 plate heats, one CR to a reduction of 26% and the other to 20%. SCC growth rates at 360°C were similar to that measured for the 31%CR alloy 690TT CRDM tubing. Comparisons will be made to other results on CR alloy 690 materials.

The objective of this work is to determine the cyclic and stress corrosion cracking (SCC) crack growth rates (CGRs) in a simulated PWR water environment for Alloy 690 heat affected zone (HAZ). In order to meet the objective, an Alloy 152 J-weld was produced on a piece of Alloy 690 tubing, and the test specimens were aligned with the HAZ. The environmental enhancement of cyclic CGRs for Alloy 690 HAZ was comparable to that measured for the same alloy in the as-received condition. The two Alloy 690 HAZ samples tested exhibited maximum SCC CGR rates of 10−11 m/s in the simulated PWR environment at 320°C, however, on average, these rates are similar or only slightly higher than those for the as-received alloy.

Stress corrosion crack (SCC) arrest tests have been conducted on composite material specimens to study the SCC susceptibility of highly SCC resistant materials. The “composite arrest” test method entails fabricating composite material specimens consisting of a highly SCC susceptible material welded to a highly SCC resistant material. Results from Alloy 690, Alloy 690 weld metal and stainless steel composite arrest specimen tests showed that these materials are extremely resistant to SCC in deaerated water environments. Under aggressive 360°C (680°F) test conditions, cracks readily grew within the SCC susceptible starter material but did not transition into growth within the SCC resistant materials. In contrast, SCC readily grew within Alloy 600 control specimens. Results from extensive specimen analytical (microprobe, AEM, FIB, EBSD) characterization efforts and SCC mechanistic insight will also be discussed.

Environmentally assisted crack (EAC) growth in thermally treated Alloy 690 (Alloy 690TT) cold worked one-dimensionally to the thickness reduction of 25% was investigated by crack growth rate testing in the primary water at low and high temperatures. In water at 50 °C and a dissolved hydrogen (DH) of 2.7–2.9 ppm, the test results suggest that the crack growth rate is lower than 2.2×10−9 mm/s at a stress intensity factor of 35 MPa√m. Stress corrosion cracking (SCC) growth with a rate of <1×10−11 m/s was observed in the alloy in water at temperatures of 320 °C and 340 °C. Over a DH range of 0.37–2.6 ppm, the SCC growth in Alloy 690TT has a similar DH dependence as that in Alloy 600 and its weld metals. The magnitude of the DH dependence, however, is much smaller for Alloy 690TT. In water deaerated by N2 without DH, an extremely low SCC growth rate was observed. Increasing DH in water from 0 ppm to 0.38 ppm caused an active SCC growth, suggesting that hydrogen could promote SCC growth when increasing DH from a very low level. Results of the current work confirmed a general high resistance of 25% cold worked Alloy 690TT to EAC growth in the primary water at both low and high temperatures under moderate stress intensity factors.

Alloy 690 base metal, HAZ and weld metal were tested in representative PWR primary water at 290 to 360°C. Intergranular cracking was observed in all materials. Growth rates as high as 1.2 × 10−6 mm/s were observed in the S-L orientation with micro structural banded material after cold rolling or forging to align the planes of banding, rolling and cracking. However, not all banded material has exhibited such high growth rates. Growth rates on homogeneous Alloy 690, including extruded CRDM tubing, often showed growth rates in the range of 2 − 8 × 10−8 mm/s in cold worked condition and an S-L orientation. Crack growth rates in some Alloy 690 tests were in the range of 1 to 10 × 10−9 mm/s, primarily in orientations other than S-L. For cracks aligned along the HAZ, growth rates as high as 1.2 × 10−8 mm/s were observed. Alloy 152/52/52i weld metals always exhibited low growth rates, apart from a weld that was further cold worked by 20%, which grew at 7 × 10−9 mm/s.

The objective of this work is to determine the crack growth rates (CGRs) in a simulated PWR water environment for Alloy 152. In order to meet the objective, specimens made from a laboratory-prepared Alloy 152 double-J weld in the as-welded condition were tested. For the SCC CGR measurements, the specimens were pre-cracked under cyclic loading in a primary water environment, and the cyclic CGRs were monitored to determine the transition from the fatigue transgranular fracture mode to the intergranular SCC fracture mode. The environmental enhancement of cyclic CGRs for Alloy 152 was minimal; nevertheless, the transition from transgranular to intergranular cracking was successful. Weld samples tested from the single heat of Alloy 152 exhibited SCC CGR rates of 10−11 m/s in the simulated PWR environment at 320°C, which was about an order of magnitude lower than typical for Alloy 182.

The differences in the EAC susceptibility between different weld geometries and weld metals have been distinguished by the doped steam test method. Pure weld metals of Alloy 182 and 82 are clearly more susceptible to EAC than the pure weld metals of Alloy 152 and 52, which did not show any crack initiation. The dissimilar metal welds (DMW) with diluted microstructures are less susceptible than the pure weld metals of Alloy 182 and 82. No crack initiation/extension from hot cracks was observed in any of the studied weld metals. At the hot crack tips no crack growth was observed in any of the studied samples. This is related to the segregated microstructure of the hot crack tips. In accelerated doped steam tests selective dissolution takes place and metallic Ni or NiO forms a continuous layer in the middle of the cracks surrounded by the Cr-rich oxide layer. Selective dissolution typical for EAC was not observed inside the hot cracks or at their crack tips. EAC initiation occurred in the Alloy 600 base metal of the DMWs and selective dissolution inside the EAC cracks in Alloy 600 was extensive. The results are discussed based on the selective dissolution creep model of EAC.

Crack-growth experiments have been performed on five different alloy 52M welds in simulated PWR primary water at 350°C or 360°C. The alloy 52M test matrix included V-groove and narrow-gap welds, an overlay on alloy 182, and an inlay on alloy 82. For the overlay and inlay materials, crack growth rates are reported only on the alloy 52M weld well beyond the dilution zone. In one of the narrow gap welds, the crack path was oriented to pass through a distribution of pre-existing weld cracks and their influence on stress-corrosion behavior is evaluated. Intergranular stress corrosion cracking (IGSCC) is observed in several alloy 52M welds, however propagation rates remain below 5x10−9 mm/s in all cases. Comparisons will be made to our previous SCC measurements on alloy 152 and 52 welds.

The objective of this work is to investigate the behavior of a crack initiated in Alloy 182 as it approaches the Alloy 52M WOL interface. For this purpose, an Alloy 52M WOL was deposited on a double-J Alloy 182 weld. Compact tension specimens were fabricated with the notch in Alloy 182 and oriented towards the WOL, and tested in a simulated PWR environment. The first such test revealed that the SCC rates in Alloy 182 were found to decrease by an order of magnitude ahead of the interface, and that the crack advanced from Alloy 182 into Alloy 52M. The post test examination found that crack branching occurred at the interface between the two alloys. Growth in Alloy 52M along the interface appears severe, approx. 10−10 m/s. While for the most part (70%) the crack propagated along the interface, SCC cracking was also found to extend into Alloy 52M along the original direction. This cracking is substantial, yielding SCC rates of 10−11 m/s.

The current generation of weld metals used in BWRs, Alloy 182 and 82, have shown significant susceptibility to SCC in the laboratory and/or the field. With the laboratory data showing less than a factor of two difference in crack growth rate between the ~15% Cr Alloy 182 and the ~20% Cr Alloy 82, it seems clear that Alloy 82 will not be sufficiently resistant for long term SCC resistance. The higher Cr weld metals, Alloy 52i and 52, have dramatically greater (>100X) resistance to SCC than the current generation of weld metals, and indeed exhibit remarkable resistance to high corrosion potential, high water impurity levels, and high stress intensity factor.

Unidirectional cold rolling has been shown to promote intergranular stress corrosion cracking (IGSCC) in alloy 690 tested in PWR primary water. High-resolution scanning (SEM) and transmission electron microscopy (TEM) have been employed to investigate the microstructural reasons for this enhanced susceptibility in two stages, first examining grain boundary damage produced by cold rolling and second by characterization of stress corrosion crack tips. The degree of permanent grain boundary damage from cold rolling was found to depend directly on the initial IG precipitate distribution. Cold rolling to high levels of reduction was discovered to produce small IG voids and cracked carbides in alloys with a high density of grain boundary carbides. For the same degree of cold rolling, alloys with few IG carbides exhibited much less permanent damage. Although this difference in grain boundary damage appears to correlate with measured SCC growth rates, crack tip examinations reveal no interaction between the preexisting voids and cracked carbides with the propagation. In many cases, these features appeared to blunt propagation of IGSCC cracks. High-resolution characterizations are described for cold-rolled alloy 690 CRDM tubing and plate materials to gain insights into IGSCC mechanisms.

Many published results have shown that fracture resistance of Ni base alloys is remarkably reduced by intergranular cracking in low temperature (<150°C) hydrogenated water. This phenomenon is called low temperature crack propagation (LTCP). In order to study maintenance assessment of actual components, J-R tests for Ni base alloys (Alloy 690, 600, Alloy 52, 82 and 152 weld metal) were carried out in simulated PWR primary water to investigate susceptibility to LTCP.J-R tests were performed under the conditions of various temperatures and dissolved hydrogen content. As a result, Alloy 690, 600, Alloy 52 and 82 weld metal showed no significant susceptibility under the conditions evaluated to LTCP. Alloy 152 weld metal showed susceptibility to LTCP in water with dissolved hydrogen content of over 15ccH₂/kgH₂O at 50 °C.On the other hand, as result of investigation on operating condition of actual Japanese PWRs, it was confirmed that there are no possibility of plant operation in 50°C water with dissolved hydrogen content of over 15ccH₂/kgH₂O. Thus, it is evaluated that there is very little possibility that LTCP becomes a threat in actual Japanese PWRs.

Analytical electron microscopy and three-dimensional atom probe tomography (ATP) examinations of surface and near-surface oxidation have been performed on Ni-30%Cr alloy 690 materials after exposure to high-temperature, simulated PWR primary water. The oxidation nanostructures have been characterized at crack walls after stress-corrosion crack growth tests and at polished surfaces of unstressed specimens for the same alloys. Localized oxidation was discovered for both crack walls and surfaces as continuous filaments (typically <10 nm in diameter) extending from the water interface into the alloy 690 matrix reaching depths of ~500 nm. These filaments consisted of discrete, plate-shaped Cr₂O₃ particles surrounded by a distribution of nanocrystalline, rock-salt (Ni-Cr-Fe) oxide. The oxide-containing filament depth was found to increase with exposure time and, at longer times, the filaments became very dense at the surface leaving only isolated islands of metal. Individual dislocations were oxidized in non-deformed materials, while the oxidation path appeared to be along more complex dislocation substructures in heavily deformed materials. This paper will highlight the use of high resolution scanning and transmission electron microscopy in combination with APT to better elucidate the microstructure and microchemistry of the filamentary oxidation.

There are two so-called grain boundary chromium concentrations: the grain boundary chromium concentration at carbide free grain boundary sections and the interfacial chromium concentration at the interface between carbides and the matrix. Existing models for predicting grain boundary chromium depletion either predicts the former or the latter. A Monte Carlo based precipitation kinetics simulation approach has been developed to simulate the evolution of both the grain boundary and interfacial chromium concentration. Application of the method to Alloy 690 yields good agreement with experimental observations with regard to both the chromium depletion evolution and carbide precipitation kinetics. The effect of grain boundary character will also be discussed.

In-situ and ex-situ oxide characterization was performed using synchrotron X-rays (SPring-8), to investigate the mechanism of stress corrosion cracking (SCC) of a non-sensitized cold worked 316 stainless steel specimen in high temperature water.The refreshed type auto-clave with two diamond windows was originally designed and fabricated to conduct the in-situ measurement in high temperature water using synchrotron X-rays. Chevron notched compact tension type specimens of a non-sensitized cold worked 316 stainless steel were loaded, exposed in simulated BWR water and analyzed using synchrotron X-rays under in-situ and ex-situ conditions.The oxide films and their layered structure near exposed surface were identified by X-rays diffraction and the local stress / strain beneath the oxide films were measured by X-rays diffraction.From these experimental results, the oxidation process for the non-sensitized cold worked 316 stainless steel in high temperature water was characterized, and a new model for the corrosion process of non-sensitized cold worked 316 stainless steel in high temperature water was proposed.

This work contributes to characterization of the oxide films formed on nickel-base alloys (Alloy X-750, Alloy 82 and Alloy 182) under simulated BWR water environments at ~10 or 18 m/s with or without iron injection. HR SEM/TEM and FIB techniques were applied. The oxide thicknesses on different alloys were substantially different, ranging from 50 nm to 8 μm. For Alloy X-750 and Alloy 182 exposed without iron injection, similar oxide phase compositions consisting of sub-micron Fe₂O₃ and NiFe₂O₄ grains as well as NiO were formed but with substantially different microstructures. For the corroded Alloy X-750 there was an additional dense layer of possibly Ni_1.5Cr_0.5O₃ in between the NiFe₂O₄ and NiO layers. On Alloy 82 which contained a relatively low Fe-content only a thin but dense film of Cr_1.3Fe_0.7O₃ was seen. With iron injection the oxide films formed on Alloy 82 were similar to that on the Alloy 182 without iron injection, suggesting that iron injection may play a similar role as if the alloy had an elevated iron content. The implication of the observations for material corrosion behavior in BWRs is elaborated.

The oxide films that formed along stress corrosion crack (SCC) paths in both high temperature aerated and deaerated water (AW and DW) were characterized through extensive analytical transmission electron microscopy (ATEM), energy dispersive spectroscopy (EDS), Auger, X-ray diffraction (XRD), and focused ion beam/scanning electron microscopy (FIB/SEM) analyses. SCC growth rate experiments were conducted in AW and DW environments at temperatures as high as 360°C, with anion additions during portions of testing. Rapid growth rates occurred in anion faulted AW and were as much as two orders of magnitude faster than those measured in the DW environments. The oxides generally consisted of a dual-layered structure with an inner layer of chromium enriched oxide (relative to the metal substrate) and an outer layer of iron rich oxide. Nickel enrichment was often observed at and ahead of crack tips and along crack flanks in the DW environment and was infrequently observed in the AW environment. Localized corrosion, here referred to as “oxide bulbs” were solely detected in the AW environment and occurred periodically along crack paths and at some crack tips. The development of a nickel enriched zone ahead of the crack tip in the DW environment is speculated to be a signature of a slow growing crack rather than an essential signature of the crack propagation mechanism.

Initiation process of stress corrosion cracking (SCC) usually takes long precursor and incubation periods. It can be assumed that SCC initiation is a form of localized oxidation or oxide penetration into the metal substrate. In this investigation, apart from fundamental mechanistic studies, the oxide / metal interface morphology after the oxidation of austenitic steels were analyzed. Oxidation tests were performed using circumferential notched round bar specimens in order to study the oxidation under a high stress state. Based on the oxide / metal interface profiles obtained from cross-sectional images of the specimens, characteristic curves for oxidation localization were drawn and parameters that describe the features of oxidation localization in laboratory specimens were obtained. The trends in these localized parameters with oxidation time are compared with samples removed from operated components and a correlation to SCC initiation is discussed.

To reveal the effect of cold work on SCC susceptibility in simulated BWR environment for low carbon austenitic stainless steels, a CBB test was performed where the microstructures were analyzed by EBSD. In this study, two of type 316NG and one of type 304NG stainless steel were tested and analyzed. Type 316NG #1 which has relatively higher Ni and Mo content than type 316NG #2 showed higher SCC susceptibility and type 316NG #2 having lower Ni and Mo content than type 316 #1 showed lower susceptibility. But the number of cracks doesn’t change dramatically, then it implies that the cold work prior to CBB test doesn’t increase the SCC initiation sites. Type 304NG has no cracks by CBB test. EBSD parameters concerned with plastic strain within grains or random grain boundaries have no significant difference. ∑3 ratios were also analyzed by EBSD and it was found that the deviation of ∑3 boundaries increases and ∑3 ratio decreases with the applied strain. ∑3 boundaries show the good resistance to intergranular corrosion and cracking so that cold work might accelerate the micro-cracks propagation and coalescence.

Stress corrosion cracking (SCC) of Type 316L stainless steels with controlled surface hardness distribution was investigated under simulated boiling water reactor (BWR) condition by creviced bent beam (CBB) tests. Intergranular microcracks having a depth of less than 50 μm are observed in 50 hours. Until 250 hours, depths of the microcracks are not increased. After 500 hours, cracks with depths of more than 50 μm are observed. SCC is initiated preferentially in a steep hardness gradient area.

The influence of surface cold work (CW) strongly affects the crack initiation behavior of austenitic stainless steels under simulated light water reactor (LWR) environment. Within a parametric study crack initiation and crack growth rate (CGR) experiments were performed under simulated boiling water reactor (BWR) environment to determine critical conditions for plant components which might undergo crack formation and subsequent crack growth. Within this project AISI 347 was tested in solution annealed and/or several CW conditions.Comprehensive characterization in initial CW condition in comparison to as-tested material condition after e.g. exposure tests for crack initiation studies and CGR-experiments, clearly indicate the influence of localized plastic deformation within the grains of the material on the processes of crack initiation and propagation. With increasing amount of CW the addiction to intergranular cracking seems to increase due to high cumulative strains on active slip paths. Increased strength and reduced plasticity of the material in CW material condition superimpose the localized strengthening effects.

Intergranular Stress Corrosion Cracking (IGSCC) in austenitic Stainless Steels (SS) in Boiling Water Reactor (BWR) operating conditions have been reported worldwide. Nitrogen containing Stainless Steel is used in BWRs and it can affect IGSCC behavior. In this investigation type 304L stainless steel with two different levels of nitrogen was evaluated in the sensitized and non-sensitised strain-hardened condition. Experiments were carried out in high temperature water with controlled dissolved oxygen. In the sensitised condition, the Crack Growth Rate (CGR) reduced and in the non-sensitised strain-hardened condition the CGR increased with increase in nitrogen level in SS. Transmission electron microscopic (TEM) investigations of the as-rolled SS and the SS after tensile testing at 288 °C indicated that rolling resulted in higher grain boundary strain which is a possible cause for higher CGR in the SS with higher nitrogen. Nitrogen did not have a noticeable effect on the deformation mechanism, for the SS after tensile testing at 288 °C, and the dislocation structures observed were similar for both the SS.

The ASTM A262 Practice A procedure is frequently used to assess whether Type 304/304L austenitic stainless steels are “sensitized”. In this study, Type 304L steel containing 18 ppm boron examined in the as-received (mill solution-annealed) condition exhibited a “dual” structure after the Practice A test despite its low C content. Detailed characterization of this alloy was performed to assess the precipitation behavior in this steel. Intergranular, Cr-rich M₂B- borides, identified by electron diffraction, were observed in the as-received condition. Samples aged at 700°C (nominally “sensitized”) produced fully “ditched” grain boundaries having Cr levels in excess of 14 wt.% and concomitant Cr-depleted zones less than 50 nm in extent. Despite exhibiting fully “ditched” grain boundaries, Cr levels of ∼18–20 wt.% with no Cr-depleted zones were detected in specimens aged at 900°C (nominally “stabilized”). These results show the Practice A test can be mis-used/mis-interpreted for assessing sensitization.

Disposition lines are used for flaw tolerance analysis, structural integrity assessment and life prediction of reactor components. One parameter normally not accounted for in the development of disposition lines is the grain size of the material. However, in actual reactor systems there can be variations in grain size within one component, as well as between different components. The objective of this project was therefore to study if the susceptibility to intergranular stress corrosion cracking (IGSCC) in one heat of Type 316L stainless steel is affected by the grain size.CGR measurements were conducted on sixteen 25 mm CT specimens in simulated BWR environments. Eight specimens were loaded by pull rods (active load), and the remaining specimens were bolt loaded. Three different grain sizes were studied on the same heat of material: 26, 347 and 590 µm.Intergranular cracking was observed in 13 out of 16 specimens. No obvious difference in crack growth rate (CGR) was observed between the three grain sizes. Further, a clear effect of hydrogen water chemistry (HWC) was observed in all material conditions, where the CGR was considerably lower in HWC compared to normal water chemistry (NWC).

For a better understanding toward stress corrosion cracking (SCC) in dissimilar metal welds with 304L stainless steel and Alloy 82, the SCC growth behavior in the transition regions of weld joints was investigated via slow strain rate tensile (SSRT) tests in 280 oC pure water with a dissolve oxygen level of 300 ppb. Prior to the SSRT tests, samples with dissimilar metal welds were prepared and underwent various pretreatments, including post-weld heat treatment (PWHT), shot peening, solution annealing, and mechanical grinding. In addition to the SSRT tests, measurements of degree of sensitization and micro-hardness on the transition regions of the metal welds were also conducted. According to the test results, the samples having undergone PWHTs exhibited relatively high degrees of sensitization. Distinct decreases in hardness were observed in the heat-affected zones of the base metals in all samples. Furthermore, the fracture planes of all samples after the SSRT tests were located at the stainless steel sides and were in parallel with the fusion lines. Among the treating conditions investigated in this study, a PWHT would pose a detrimental effect on the samples in the aspects of mechanical property and degree of SCC. Solution annealing would lead to the greatest improvement in ductility and SCC retardation, and shot peening would provide the treated samples with a positive improvement in ductility and corrosion retardation, but not to a great extent.

Stress corrosion cracking susceptibility of austenitic alloys subjected to various kinds of prior deformation and weld-shrinkage was investigated by material and microstructural characterization, crack growth rate tests in high temperature water, and crack tip characterization. The active cracking paths and cracking kinetics as functions of types and ratios of grain boundaries, microstructural anisotropy, strength or hardness, local strain in terms of the distribution of misorientation are quantitatively studied.

The effect of static load hold times of 6 to 744 h on corrosion fatigue life of low- and highcarbon and stabilized austenitic stainless steels was investigated with both sharply notched and pre-cracked fracture mechanics specimens in simulated boiling water reactor (BWR) hydrogen (HWC) and normal water chemistry (NWC) at 288 °C. With regard to continuous cyclic saw tooth loading with a load ratio R close to zero in HWC environment, an increase of the genuine corrosion fatigue initiation life was observed with increasing static load hold periods at maximum or mean load of the applied load range, which seemed to saturate for long hold periods above 12 to 24 h. On the other hand, static hold periods at minimum load, where potential microcracks are closed, had no effect on genuine fatigue initiation life. Furthermore, the static load hold times had very little effect on the subsequent stationary short corrosion fatigue crack growth rates. Static hold times of up to 744 h had no effect on the corrosion fatigue crack growth rates in pre-cracked solution annealed stainless steel specimens in NWC and HWC environment. No significant effect of static load hold times on the technical corrosion fatigue initiation life are thus expected based on these preliminary results and the current US NRC regulatory guide 1.207 seems to be adequate from this point of view. A credit for a mitigating effect of long static load hold periods in the field cannot be derived from this work.

Laboratory studies on austenitic stainless steels in PWR primary coolant environments have shown that the ASME XI procedures used to assess fatigue crack growth of reactor components may not always be conservative. Recent work has shown that significant environmental enhancement of growth rates can occur in this environment, especially for some long rise time loading cycles. Although enhancements up to eighty times relative to air data have been observed, under some conditions retardation of the enhanced growth rates can also occur, leading to rates close to the ASME XI air line. Several factors appear to influence retardation, including temperature, water flow rate and material composition. The current study addresses the influence of material composition and it is shown that steels of high sulfur content (>0.02%) are much more prone to retardation than low sulfur (<0.01%) steels. Work aimed at elucidating possible mechanisms for this effect is described.

This is a study of the 107 cycle fatigue limit of Type 304L Stainless Steel, as measured in fully reversed (R=-1) load-controlled tests, at 150°C and 300°C, in air and PWR water. The staircase method was used to determine the fatigue limit. The tests run here utilized a cycle frequency of 1.818Hz and are compared to other tests from the literature that were run at 30Hz. The fatigue limit measured in the tests run at the high frequency was higher than that measured here. This is explained by measurements of the strain developed during cycling, using the different cycle frequencies. The tests run at the higher frequencies yielded lower strains for a given stress and, as expected, this resulted in higher fatigue limits. Using 107 cycles to define a run-out also led to a lower fatigue limit. These results are important as most previous fatigue limit measurements utilized 106 cycles or less to define a run-out, and when lives as long as 107 cycles are used the tests are generally run at high cycle frequencies, thus leading to higher fatigue limits than those measured here.This study utilized hysteresis loops measured during load controlled cycling, which illustrate the complex nature of the cyclic deformation that can develop. At 300°C, and at high stresses, ratcheting (an increase in the average strain) was observed, but at low stresses retrograde motion (a decrease in the average strain) was observed. The development of these strains should be an important consideration in any calculation of the behavior of actual components. These strains were also used to develop cyclic stress-strain curves, which can be used for design calculations and to correlate strain and load-control tests.

Over the past ten years, evaluation for license renewal and new reactors has provided significant experience and insight on the use of the environmental fatigue multiplier (F_en) approach, and recognized the need for further refinement of this methodology, as well as its application to other areas. Hence, the NRC has initiated further research work on environmentally assisted fatigue (EAF). The objectives of these research activities are as follows:1.Develop a transient stress evaluation software tool for rapidly determining thermal transient stresses in reactor components.2.Develop an American Society of Mechanical Engineers (ASME) Boiler & Pressure Vessel Code (Code) fatigue calculation software tool for estimating fatigue usage factors in reactor components.3.Develop revised fatigue cumulative usage factor (CUF) limit criteria for postulated high energy line break (HELB) locations.4.Update existing EAF methodology, develop application techniques for applying the methodology, and revise U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.207 accordingly, if appropriate.5.Participate in a round-robin study, conducted by the Electric Power Research Institute (EPRI) Advisory Panel on Environmental Fatigue, to solve a fatigue sample problem Each of these activities is briefly described in this paper.

The paper presents the crud analysis of a twice-burned fuel deposit using the AREVA crud sampling method in a domestic B&W PWR. The patented AREVA sampling method allowed the separation of deposit flakes that retain all the characteristics of unperturbed crud deposition.The twice-burned fuel assembly sampled was positioned in its second cycle of irradiation immediately against a fresh fuel assembly where the thermal-hydraulic (T/H) calculations indicated it as the first burn fuel assembly exhibiting the highest boiling for the core.Radiochemistry data will be presented that suggest that intense boiling was present at the location of the sample collection, producing a crud accumulation during the second cycle of operation considerably larger then during the first cycle of operation.The unperturbed 3D analysis of the deposition will be presented, allowing an in depth understanding of the processes of deposition on PWR fuel in conditions of intense localized boiling.

In order to mitigate PWSCC initiation and propagation in Ni base alloys, Japanese PWR utilities desire to employ optimized dissolved hydrogen (DH) control operation in the near future. Prior to the application of the optimized DH control operation to PWR, the effect of DH concentration on the fuel crud deposition should be clarified.Crud deposition tests were carried out in boric acid (1200ppm) + lithium hydroxide (2.2ppm) + hydrogen (7 to 25cm3-STP/kg-H₂O) solutions at 325°C under sub-cooled boiling and non-irradiated conditions. The corrosion resistance of zircaloy-4 was also investigated.From the test results, it was revealed that a crud layer composed of NiFe₂O₄ and NiO was formed on zircaloy-4 fuel cladding. NiO was easy to form in the crud layer under the 7cm3-STP/kg-H₂O condition. The amounts of deposited crud layer, boron incorporated into the crud layer and the corrosion resistance of the zircaloy-4 cladding were not affected by DH concentration.

The Paper presents the results of applying the AREVA NP Inc. BWR Crud Model to determine the effect a high zinc injection operational program at a US Plant has on the fuel.The paper explains the AREVA NP Inc. BWR Crud Model capabilities and its role in the AREVA Fuel Risk Assessment Strategy.Simulated crud characteristics are analyzed along the 6 years of the fuel’s operation and then compared with actual crud analysis results. Also a number of “What If” scenarios are explored and their results are placed in the context of the fuel operation risk assessment.

Corrosion properties of Zr-Sn and Zr-Nb zirconium alloys were studied under simulated PWR conditions (or, more exactly, VVER conditions — boric acid, potassium hydroxide, lithium hydroxide) at temperatures up to 340°C and 15MPa using in-situ electrochemical impedance spectroscopy (EIS) and polarization measurements. EIS spectra were obtained in a wide range of frequencies (typically 100kHz — 100μHz). It enabled to gain information of both dielectric properties of oxide layers developing on the Zr-alloys surface and of the kinetics of the corrosion process and the associated charge and mass transfer phenomena. Experiments were run for more than 380 days; thus, the study of all the corrosion stages (pre-transition, transition, post-transition) was possible.Experimental impedance spectra were approximated by equivalent circuits based models. Jonscher-type analysis was applied to estimate frequency independent oxide capacitance. A model based on the integrated Stern-Geary equation was developed to correlate instantaneous and integral corrosion rates of zirconium alloys.

Three factors are considered paramount in fuel performance; these are the heat flux, crud layer and oxide thickness. Both the crud layer and the oxide thickness may be affected by plant chemistry.AREVA NP has developed a Fuel Condition Index (FCI) for PWR Fuel that provides a method to assign a single numerical value based on both limiting thermal-hydraulic factors and plant operating chemistry conditions to assess observed or expected fuel performance. FCI is an AREVA NP fuel damage risk assessment tool to be used in the Level II crud risk analysis work.. FCI, considering Heat Flux, boiling conditions, crud layer and pin oxide, extends the capabilities of the Industry’s currently used High Duty Core Index (HDCI), considered by AREVA to also be a Level II tool.The chemistry parameters and operating conditions selected for the example calculations presented in the paper are based on AREVA NP knowledge and Industry consensus.This paper describes the FCI developed by AREVA NP (patent pending) and the results of its use at PWR plants compared with the results of applying at the same plants the High Duty Core Index (HDCI).

Zirconium alloys are used as nuclear fuel cladding material due to their mechanical and corrosion resistant properties together with their favorable cross-section for neutron scattering. At running conditions, however, there will be an increase of hydrogen in the vicinity of the cladding surface at the water side of the fuel. The hydrogen will diffuse into the cladding material and at certain conditions, such as lower temperatures and external load, hydrides will precipitate out in the material and cause well known embrittlement, blistering and other unwanted effects. Using phase-field methods it is now possible to model precipitation buildup in metals, for example as a function of hydrogen concentration, temperature and external load, but the technique relies on input of parameters, such as the formation energy of the hydrides and matrix. To that end, we have computed, using the density functional theory (DFT) code GPAW, the latent heat of fusion as well as solved the crystal structure for three zirconium hydride polymorphs: δ-ZrH_1.6, γ-ZrH, and Є-ZrH₂.

Hydrogen ingress into zirconium alloy fuel cladding during operation in nuclear reactors can degrade cladding performance, both during operation and under dry storage, due to formation of brittle hydrides. At temperature and under stress, hydrogen redistribution and reorientation can occur, reducing cladding resistance to failure. Thus, it is crucial to understand the kinetics of hydride dissolution and re-orientation under stress and at temperature. High-energy and micro-beam synchrotron diffraction are used to study the kinetics of hydride reorientation and hydride distribution near a crack tip in previously hydrided Zircaloy sheet. Reorientation of hydrides in bulk samples is studied in situ (at temperature and under applied tensile stress). In-situ transmission diffraction data provides unique strain and orientation information on the hydrides. Micro-beam diffraction has been performed on previously cracked compact tension specimens under load. Measurement of the hydride distribution and associated strains can be performed with the micro-beam to determine hydrogen response to an applied strain field.

A method based on the exposure of notched tensile specimens to iodized methanol is investigated to study the influence of stress state on the iodine-induced stress corrosion cracking (I-SCC) of zirconium alloys. Based on uniaxial tensile tests at different strain rates, in air and at room temperature, an anisotropic viscoplastic behavior model was established. This model was successfully compared to a notched specimen tensile test where local strains were measured by digital image correlation. Moreover, two I-SCC experiments were designed, on a tensile test machine and with an autonomous bending device. Examinations of failure surfaces of smooth specimens and investigations of iodine location were undertaken. The intergranular aspect of the cracking was observed and no iodine was found outside open cracks. Model calculations with different specimen geometries confirmed the possibility of having different stress states and will be used to determine local mechanical fields at crack initiation sites after I-SCC tests.

During root cause of fuel failure examinations at pressurized water reactor (PWR) unit A in 2004, it was discovered that fuel assemblies located on the core baffle exhibited excessive and varying indications of grid strap wear, and in one case fuel rod failure due to wear. The grid strap wear was due to interaction of the peripheral fuel assemblies with the reactor vessel (RV) internals baffle plate. Subsequent inspections at this unit in 2005 and other similarly designed units revealed that the wear had been occurring in the past and appeared to be progressing with time based on limited inspections. Extent of condition visual inspections were conducted from 2006 to 2009 at similarly designed units and revealed some level of wear had occurred at each of the units, but to different degrees. It was also established that grid strap wear has been observed since initial unit operation and has increased significantly with the switch from a nickel-based alloy to Zircaloy-4 grids in the 1980s.Several investigations were made into the cause of the interaction. One such investigation included samples of worn grid straps obtained from unit A in early 2009 that were evaluated in a hot cell facility. The investigations performed at the hot cell facility and the investigation results are discussed in this paper.

The effect of microstructure, stress intensity factor, corrosion potential and water purity on stress corrosion crack growth rate behavior of Alloy X-750 was investigated in 288 °C water. This material was provided by a utility in the form of an unused stabilizer support bracket. Alloy X-750 has exhibited SCC in field components, and this study was designed to examine its microstructure and SCC response in some detail to determine the suitability of Alloy X-750 for long term reliable service in BWRs.

The stress corrosion cracking susceptibility of Inconel alloy X-750 in the HTH heat-treated condition has been evaluated in high-purity water @ 93 and 288°C under Normal Water Chemistry (NWC) and Hydrogen Water Chemistry (HWC) conditions. SCC crack growth rates of approximately 1.2x10-7 mm/s (K=28 MPa√m) under NWC conditions and 1.4x10-8 mm/s (K=28 MPa√m) under HWC conditions in 288°C high-purity water were observed. The alloy was also tested in HWC at 93 °C. No SCC crack growth was observed at K= 35 MPa√m for the 525 h of testing at 93°C. At 288°C the fracture mode transitioned from predominantly transgranular cracking under fatigue conditions to intergranular cracking under constant stress intensity for both environments. Atom Probe Tomography (APT) was used to characterize the crack tip and ahead of the crack tip. The results suggest that oxygen diffuses ahead of the crack tip along specific crystallographic directions.

A modified alloy 718 has been developed as an alternative material to nickel-based alloy X-750 used in the boiling water reactors (BWR). Stress corrosion cracking (SCC) of a jet pump beam made of alloy X-750, occurred in a BWR plant. In order to apply modified alloy 718 to the jet pump beam, an evaluation of resistance to SCC initiation based on a uni-axial constant load (UCL) test in high temperature water was carried out. All UCL test specimens of modified alloy 718 at an applied stress beyond the yield strength in 288˚C showed no failure after 10,124 hours testing. On the other hand, UCL specimens of alloy X-750 at an applied stress of 900MPa failed within 10,000 hours. Therefore, it was confirmed that modified alloy 718 had longer SCC initiation time in high temperature water than alloy X-750. Moreover, fatigue and spring properties of modified alloy 718 were also evaluated, and the characteristics of both modified alloy 718 and alloy X-750 were compared.

The electrochemical behavior of alloy 718 in a chloride-containing boric acid solution was studied to determine the influence of chloride ions as contaminants of pool water of nuclear power plants on the corrosion behavior of the alloy. Experiments were performed at 20°C and 60°C with chloride concentrations from 1.5 to 15 000 ppm, using stationary measurements i.e. OCP versus time measurements and plotting of current-potential curves. After the electrochemical tests, the samples were observed using optical microscopy. Immersion tests in chloride-containing boric acid solutions were also carried out: samples were immersed for a time as long as 17 weeks at open circuit potential and their residual mechanical properties were measured. Results showed that, whatever the chloride concentration, there was no corrosion for samples immersed at open circuit potential. However, when the samples were polarized at high potentials, intergranular corrosion might be observed in occluded zones.

The stress corrosion cracking (SCC) behavior near the fusion boundary (FB) of a dissimilar weld joint with Alloy 182-A533B low alloy steel (LAS) in high-temperature oxygenated water doped with sulfate has been investigated, with a focus on the relationship between the SCC crack arrest/reinitiation behavior and the microstructural characteristics of the heat-affected zone (HAZ) in LAS adjacent to the FB. Cracks propagated perpendicular to the FB along the dendrite grain boundary in the dilution zone (DZ) of Alloy 182, and then spherical or crack-like oxides were formed in the HAZ of LAS adjacent to the FB; no obvious SCC susceptibility was observed in the non-HAZ region of LAS. Crack arrest occurred when spherical oxides formed in the LAS. Crack-like oxides tended to propagate along the prior austenite grain boundary in the coarsegrained HAZ (CGHAZ) of LAS. It has been suggested that the microstructure and continuity between the dendritic grain boundary in the DZ of Alloy 182 and the prior austenite grain boundary in the CGHAZ of LAS across the FB, that is, the continuity of the potential crack path played an important role in the arrest/reinitiation of SCC crack in the FB region. Above microstructural characteristics varied depending on the multiple heat cycles of the welding process.

Recent investigations at different laboratories have shown a strong effect of chloride contaminations on the crack growth rate of low-alloy steel (LAS) in oxygenated high-temperature water (HTW). Therefore, a research project was launched to systematically investigate the effects of chloride contaminations on the environmental degradation of materials under BWR-relevant water conditions.This project focused on investigations of the general corrosion and crack initiation behavior of LAS (German RPV steel 22NiMoCr3 7) in oxygenated HTW without chloride and at different chloride contamination levels up to 50 ppb. Chloride was added either permanently or temporarily to simulate a chloride transient during plant operation.During these tests, Electrochemical Noise (EN) and Electrochemical Impedance Spectroscopy (EIS) measurements were performed to monitor the electrochemical behavior. After the tests, the specimens were examined macroscopically and microscopically. In addition, the oxide layer thickness was determined using the Focused Ion Beam (FIB) technique and different surface analysis techniques were performed to analyse the composition of the oxide layer.The applied tests clearly revealed a decrease in oxide layer thickness during permanent chloride contamination. Temporary transients, however, did not cause a long-term memory effect.

The effect of temperature on the crack growth rate (CGR) in BWR normal water chemistry (NWC) has been investigated in various studies over the years. However, the effect has not been clarified unambiguously, since some studies report a maximum in CGR at a temperature of about 200 °C, while others have reported a monotonic increase of CGR with temperature. To clarify the effect of temperature, testing has been performed in oxygenated high-purity water (NWC with 500 ppb O₂ and conductivity ranging from 0.1 to 0.2 μS/cm).Crack growth rate measurements were conducted on six 25 mm compact tension (CT) specimens in simulated NWC environment at ~30 MPa√m. Two heats of sensitized stainless steel and one heat of alloy 182 were tested, and the temperature ranged from 288 to 100 °C.Intergranular or interdendritic cracking was observed in all specimens. The DCPD response clearly showed that the crack growth rate in stainless steel increased monotonically with increasing temperature. Alloy 182 showed no temperature dependence in this work. A possible reason for the different conclusions in the literature was identified. For short testing times on each temperature (about 50 h) a maximum in CGR at ~200 °C was observed, while a monotonic increase was observed if longer durations were considered.

This study reinforces the expectation that a consistent benefit of low corrosion potential is achievable at intermediate temperatures associated with BWR start up. Such low corrosion potentials can probably only be achieved using NobleChem™ and injection of H₂ or other reductants such as hydrazine or carbohydrazide because very low residual levels of O₂ can elevate the corrosion potential. The high growth rates that occur during start up merit mitigation, although this study did not find growth rates that were orders of magnitude higher than at 288 °C. However, this study did not attempt to simulate all aspects of start up, especially the sources of dynamic strain such as differential thermal expansion, which can be estimated by are not known.

An experimental program has been conducted in order to understand how the spinodal decomposition may affect material properties changes in Type 316L BWR pipe weld metals. The program includeed Charpy-V, tensile, SCC crack growth and in-situ fracture toughness testing as a function of aging time and temperature. In this paper we report results of fracture toughness, SCC crack growth rate and fracture morphology studies of Type 316L stainless steel weld metals under simulated BWR conditions, consisting of 288°C, high purity water containing 300 ppb dissolved oxygen (defined for purposes of this paper as “In-Situ”). SCC crack growth results show an approximately 2X increase in crack growth rate over that of the unaged material. In-situ fracture toughness measurements indicate that environmental exposure can result in a reduction of toughness by up to 40% over the corresponding at-temperature air values. Detailed analysis of the results strongly suggest that spinodal decomposition is responsible for the degradation in properties measured ex-environment. Analysis of the results also strongly suggests that the in-situ properties degradation is the result of hydrogen absorbed by the material during exposure to the high temperature aqueous environment.

Flow Accelerated Corrosion (FAC) of carbon steel is a phenomenon that has been studied for many years. However, to date, the specific behavior of welds and weld assemblies of carbon steel towards this phenomenon has been scarcely examined. An experimental program of FAC of welds and weld assemblies is being conducted by EDF and CRIEPI. This paper describes the results obtained on the behavior of weld metal independently of its behavior in a weld assembly as well as the sensitivity to FAC of various weld assembly configurations. Tests are performed, at EDF, in the CIROCO loop which permits to follow the FAC rate by gammametry measurements, and at CRIEPI, in the PRINTEMPS loop where FAC is measured by laser displacement sensor. Welds are performed by two different methods: Submerged Arc Welding (SAW) and Gas Tungsten Arc Welding (GTAW). The influence of several parameters on FAC of welds is examined: welding method, chromium content and temperature. For weld assemblies, only the impact of chromium content is studied. All the tests are conducted in ammonia medium at pH 9.0 and oxygen concentration lower then 1 ppb. Chemical parameters, as the pH, the conductivity and oxygen concentration, are measured in situ during the test and surface characterizations are performed after the test. The results show that, with more than 0.15% chromium, no FAC is detected on the weld metal, which is similar to the base metal behaviour. For the same and lower chromium content, the two types of metal have the same FAC rate. Concerning the temperature effect, for both metals FAC rate decreases with temperature increase above 150°C. Below 150 °C, their behaviour seems to be different. For weld assemblies, the study of different configurations shows that the chromium content is the main parameter affecting the behaviour of the specimens. Additional tests and modeling studies will be conducted in order to complete the results.

The mitigating effects of chromium on flow-accelerated corrosion of carbon steel occur at concentrations in the metal as low as 0.02% and, in some coolant environments, are seen very soon after exposure and remain virtually constant. We have modeled such effects by including a diffusion barrier at the metal-oxide interface, below the mainly magnetite layer which forms the conventional barrier. This extra barrier is a fixed layer that forms almost instantaneously with a composition depending upon the chromium concentration. It is very thin, so would be undetectable by normal surface analysis techniques, but has its own properties of porosity, density, etc. The secondary barrier of magnetite behaves as modeled before, in time achieving a steady-state thickness that depends upon its dissolution characteristics and the fluid dynamics of the coolant. By adjusting the properties of the chromium-dependent layer, we have been able to predict the FAC of carbon steel of different chromium contents in typical reactor feed-water environments.

Cold-worked 304 stainless steels (SS) are known to be susceptible to stress corrosion cracking (SCC). This study employs atom-probe tomography (APT) for local chemical analysis of the oxides formed. Autoclave experiments on a set of samples with/without cold-work prior to oxidation, and with/without stress applied during oxidation, were carried out under simulated pressurised water reactor (PWR) primary conditions. APT and analytical transmission electron microscopy (ATEM) were combined to investigate chemical and structural implications of surface and grain boundary oxidation in 304 SS. Focussed ion beam (FIB) milling was used to prepare specimens containing the same grain boundary for every analysis technique. Grain boundary and deformation band oxidation were observed in all but the unstressed and non-cold worked sample. Cavities were found ahead of the Cr-rich oxide in some of the samples. APT data suggests the presence of hydrogen in Nickel-rich regions.

Photoelectrochemical ex-situ technique has been used to investigate the effects of dissolved hydrogen, surface conditions and chromium content on the semiconducting properties of the oxide films formed on nickel-base alloys in Pressurized Water Reactors (PWR) primary water. For this purpose, two hydrogen contents were used, the first one a low content of ≤ 1 cm3.kg–1 and the second one a high content of 43 cm3.kg–1. Also two nickel-base alloys were chosen, alloy 600 and alloy 690, and three kinds of surface preparations were applied 1200 abrasive paper, 1 µm diamond paste and colloidal silica. The main result concerns the effects of dissolved hydrogen on the semiconducting type of the oxide phase at the higher energies. This oxide phase, associated to the protective internal subscale, shifts from n-type at lower content of hydrogen to insulating behaviour while the content of hydrogen increases. The two other parameters chromium content and surface preparations have no influence on the electronic properties of the oxide scale.

The results of in situ SERS investigations of Alloys’ 600 and 690 surface films were combined with the results of a number of ex situ studies conducted by other researchers who used a variety of experimental techniques. Comparing the results of different investigations revealed the surface films’ composition and microstructure were most sensitive to alloy composition and the concentrations of aqueous metal cations (especially Fe+2).Earlier studies established that the films typically consist of two layers and that saturation concentrations of (Ni++)_aq and (Fe+z)_aq affect the composition and crystal structure of the films’ outer layers. The current investigation indicates that aqueous Ni++ and Fe+2 also affect the composition and structure of the surface films’ inner layers.The faces of intergranular stress corrosion cracks exhibit oxides that are different from the oxides formed on free surfaces. The crack faces are covered by M_A600O and M_A690O. It is proposed that nickel is oxidized inside intergranular stress corrosion cracks (but not on free surfaces) because of (1) alkalization of the crack’s water due to the removal of aqueous boron species by their adsorption/precipitation on oxide, and (2) galvanic coupling of cracks to free surfaces, which is made possible by the water’s electrical conductivity (largely due to Li+).Diffusion Path Analyses is able to qualitatively explain the effects of alloy composition and water chemistry on film microstructure. Earlier studies of the adsorption of aqueous, heavy metal cations onto colloidal oxide particles provide insight into the mechanism by which (Fe+z)_aq, (Ni++)_aq, and aqueous boron species change the inner layer of the oxide films formed on Alloy 600 and Alloy 690. The present results are relevant to SCC of Alloys 600 and 690 and to cation release from Alloys 600 and 690.

Modern dual-column FIB-SEMs are capable of automatically milling and acquiring images that can be used to reconstruct sample volumes in 3D. We will demonstrate that this technique, applied to the environmental degradation of materials in nuclear reactors, is capable of revealing features, phases and/or defects in 3D with nm resolution. In this paper, we have used this technique to characterize surface oxidation and cracking in Zr alloys, the effect of cold work on the oxidation resistance of austenitic steels and crack growth in welded 316L stainless steel.

Secondary side corrosion cracking of steam generator tubes in Mill Annealed Alloy 600 occurs in flow-restricted areas where impurities get concentrated under heat flux. During spring 2009, eddy current test showed a circumferential indication (tube support plate elevation) in Bugey-3 unit for the first time on French nuclear plants in which very little PbO was detected in deposits in the late 80’s after 67000h of service. The corresponding tube was therefore removed. IGSCC cracks and outer oxides layers formed on these pulled tubes were examined by ATEM. The results are compared to previous ones obtained on a tube pulled out from another unit (Dampierre-4) where Pb was detected during ATEM observations and not suspected to be at the origin of IGSCC. Then, results are compared to those obtained on specimen tested in laboratory environment all leading to SCC rates comparable to secondary side corrosion cracking rates observed in the field. The oxides formed were compared to identify the typical environment responsible for the degradation observed on the pulled tubes. It appears that the best laboratory environment reproducing oxides morphology observed on the selected pulled tubes was a NaOH with Pb environment even thought Si pollution was sometimes detected in outside oxides layer.

Balance of plant systems in nuclear plants, such as service water systems, are a critical part of the facility’s infrastructure. System integrity and performance are vital for plant reliability and essential to achieving a plant life of 40 years and beyond. The low temperature and pressure service water piping systems are primarily degraded by corrosion in untreated waters. Corrosion allowances, based upon very simplistic considerations of general corrosion in untreated raw water, were a part of the original design. However, long term service in many such systems has shown that localized corrosion phenomena, from microbiologically influenced corrosion (MIC), pitting, and underdeposit effects, have compromised system integrity. Because of the complexity and random nature of corrosion processes, it is nearly impossible to develop a mathematically deterministic model (like the typically used corrosion allowances) that accurately predicts pipe wall loss. However, when statistical distributions are used to describe the various corrosion processes, mathematical algorithms that incorporate all of the distributions, iterated a statistically significant number of times, can be used to forecast the most probable number of leaks. This approach was used to predict the condition of service water piping at a US Nuclear Power Plant, comparing results to service experience and inspection results. The results were ultimately used by the plant for targeting inspections and for long term planning of replacements and replacement schedules.

Of the 104 currently operating nuclear power plants in the U.S, there are 66 plants that have containment buildings constructed with an inner steel liner plate in contact with a thick concrete shell. The steel liner is nominally 6 to 10 mm [0.25 to 0.375 in] thick and is designed in conjunction with the concrete containment building to function as an essentially leak tight barrier against the release of radiation under accident conditions. Corrosion of the containment liner has been observed and corrosion penetration of the liner associated with foreign materials embedded in the concrete from original construction has occurred in a few U.S. plants. This paper reviews plant operating experience, evaluates factors that can affect containment liner corrosion susceptibility, and discusses the mechanisms for through-wall corrosion initiated at the concrete/liner interface.

Sulphur species in oxidation states between S2− and S6+ are known to interfere with the protective oxide films that form on steam generator tubing materials. By assisting in the breakdown of passive films, intermediate oxidation state sulphur species can cause intergranular attack and pitting of steam generator tubing over a wide pH range. Intermediate oxidation state sulphur species have also been observed to induce stress corrosion cracking of sensitized Alloy 600 at low temperatures.This work employed electrochemical methods to investigate the effect of thiosulphate on the degradation of steam generator alloys (Alloy 600, Alloy 690, and Alloy 800). The effect of thiosulphate on steam generator tubing degradation was investigated at 150°C in crevice chemistries containing 0.0015 M sodium thiosulphate simulating the local chemistry environments developed in steam generator crevices or under sludge. The detrimental effect of thiosulphate on the boundary conditions of the recommended ECP/pH zone of Alloy 800 was discussed. Accelerated corrosion tests were also performed at selected potentials to confirm the detrimental effects of thiosulphate on the ECP/pH domain.

The corrosion products produced on polycrystalline Ni metal and Ni-Cr (20%) (NiCr) alloy surfaces exposed to aqueous environments chosen to emulate possible solution conditions in the steam generator (SG) tubing of pressurized water reactors (PWR) were studied using XPS. Additional measurements modelling the distribution of oxidized Ni and Cr species on select alloy specimens were carried out using ToF SIMS. Exposure of Ni metal and NiCr alloy samples to mildly oxidizing potentials in basic solutions resulted in the preferential growth of a β-Ni(OH)₂ phase; driven by the dissolution of metallic Ni at both 25°C and 150°C. The presence of β-Ni(OH)₂, Cr(OH)₃ and small amounts of a Cr6+-containing oxide on NiCr specimens oxidized under mildly oxidizing conditions at 150°C in neutral solutions suggested that the dissolution of both metallic Ni and Cr followed by the back deposition of the corresponding corrosion products was responsible for oxide growth under these conditions. In acidic media oxide nucleation at 150°C under mildly oxidizing potentials was determined to occur via the dissolution of both Ni and Cr species on NiCr specimens as well. The increased stability of Ni2+ in acidic solution led to a limited precipitation of β-Ni(OH)₂ resulting in the formation of very thin oxides containing higher levels of Cr(OH)₃. Reactions on metallic Ni and NiCr surfaces under highly oxidizing potentials resulted in an increase in the NiO content of these films compared to similar exposures carried out at milder oxidation conditions attributed to accelerated dehydration of the β-Ni(OH)₂ phase. In addition, an increase in the Cr(OH)₃ contribution on the alloy surface oxidized at a more oxidative potential suggested a more rapid dissolution of Cr under these conditions; overall, uneven films were formed from these conditions. The composition of the corrosion product formed after an exposure to a highly oxidizing potential was found to be unchanged following a subsequent reaction of equivalent length at a much lower oxidizing potential in basic solution.

Chromium concentration, weld-induced residual stress, and cold work are factors that affect the stress corrosion cracking (SCC) of both wrought Alloy 600 and Alloy 82 weld metal. However, chemical inhomogeneities and plastic strains that result from weld residual stresses that arise during solidification in welds makes it difficult to separate the beneficial influence of elevated chromium in Alloy 82 and the negative impact of weld-induced plastic strains. SCC growth rate tests on a single set of Alloy 82 weld cradles in four conditions were examined, including: aswelded, as-welded and cold worked, fully annealed, and fully annealed and cold worked. Annealing welds resulted in significant improvement in their SCC performance relative to the aswelded materials, but not significantly better than annealed Alloy 600 with similar chromium levels. This work shows that weld-induced plastic strains and cold work are similar and both are key drivers for the SCC performance for these materials.

The stress corrosion cracking behavior of dissimilar metal (DM) welds, including Alloy 52-A 508 and Alloy 82-A 508, under simulated BWR coolant conditions was studied. Effects of postweld heat treatment and specimen size on the corrosion fatigue and SCC growth rates of DM welds were evaluated. The crack growth rates of the DM weld heat-treated at 621 °C for 24 h was observed to be faster than those for the as-welded. But the DM weld heat-treated at 621 °C for 8 h + 400 °C for 200 h showed better SCC resistance than the as-welded. The longer the heat treatment at 621 °C, the higher the chromium carbides density along the grain boundary was observed. The SCC growth rates observed for the 1/2T CT specimens were faster than those for 1T CT specimens. It could be accounted for by the shorter distance for oxygen and anions to diffuse to the crack tip in the thinner specimen.

Stress corrosion cracking (SCC) of wrought Alloy 600 and parent weld metals (Alloys 182/82) is a significant cause of failure in pressurized water reactors (PWR). Only a small number of welds fabricated from Alloy 82 are affected by PWSCC. Most of these welds were not thermally heat treated opposite to what is done in the French industrial practice. This paper describes constant load crack growth rate (CGR) tests on Alloy 82 with and without post weld heat treatment. Metallurgical examination of Alloy 82 was carried out using mainly Transmission Electron Microscopy. The heat treatment seems to be highly beneficial by decreasing the CGR. This result can be explained by the effect of thermal treatment on the precipitation in Alloy 82.

Even if Alloy 182 is usually exhibiting a high susceptibility to cracking in the laboratory, the field experience did not reveal to date a significant susceptibility to PWSCC in PWRs, when the welds have been perfectly stress relieved. However, recently, an increasing number of cracks was reported in USA, Sweden and Japan on Alloy 182 and a few cases on Alloy 82. This paper addresses the work on initiation of PWSCC engaged at EDF R&D (MAI). The main objective is to calibrate an engineering model to predict the time to initiate IGSCC vs. temperature and loading for a weld 182 having a high susceptibility to SCC crack growth. The effect of cyclic loading, strain path and dendrite orientation on initiation was also partly evaluated. Under static loading, initiation was observed down to 350 MPa at 360°C. A limited effect of cyclic loading (R = 0.9, f = 2.8 10−4 Hz) was observed at 360°C for a maximal stress of 350 MPa. The effect of partial periodic loading increased when the temperature decreased and when the stress increased. An empirical model predicting the time to SCC initiation was calibrated.

The US Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute (EPRI) are working cooperatively under a memorandum of understanding to validate welding residual stress predictions in pressurized water reactor primary cooling loop components containing dissimilar metal (DM) welds. These stresses are of interest as DM welds in pressurized water reactors are susceptible to primary water stress corrosion cracking (PWSCC) and tensile weld residual stresses are one of the primary drivers of this stress corrosion cracking mechanism. The NRC/EPRI welding residual stress (WRS) program currently consists of four phases, with each phase increasing in complexity from lab size specimens to component mock-ups and ex-plant material.This paper discusses Phase III of the WRS characterization program, comparing measured and predicted weld residual stresses profiles through the dissimilar metal weld region of pressurizer safety and relief nozzles removed from a cancelled plant in the United States. The DM weld had already been completed on all of the plant nozzles before use in the mock-up program. One of the nozzles was completed with the application of the stainless steel safe-end weld to a section of stainless steel pipe. Measurements were taken on the nozzles with and without the welded pipe section. Several independent finite element analysis predictions were made of the stress state in the DM weld. This paper compares the predicted stresses to those found by through-thickness measurement techniques (Deep Hole Drilling and Contour Method). Comparisons of analysis results with experimental data will allow the NRC staff to develop unbiased measures of uncertainties in weld residual stress predictions with the goal of developing assurances that the analysis predictions are defensible through the blind validation provided using well controlled mock-ups and ex-plant material in this program.

Cracking susceptibility of austenitic stainless steels is known to be affected by dissolved oxygen (DO) or corrosion potential. In low-DO environments, crack growth rate (CGR) is significantly lower than that in high-DO environments. A strong dependence of CGR on corrosion potential has also been seen in irradiated stainless steels. While it has been shown that reducing the potential can reduce the CGRs of irradiated SSs, some high-dose specimens have shown elevated CGRs even in low potential environments. Thus, it is not clear how irradiation affects the dependence of CGR on corrosion potential. In the present study, a disk-shaped compact tension specimen of Type 316 SS was tested in low-DO environment. The specimen had been irradiated in the BOR-60 reactor to 5 dpa at 320°C. Post-irradiation CGR tests were performed in a low-DO environment. The effect of unloading on crack growth behavior in low-DO environment is discussed.

The susceptibility of neutron irradiated austenitic stainless steels to the initiation of irradiation-assisted stress corrosion cracking (IASCC) was assessed. Solution annealed (SA), high purity (HP) type 304 stainless steel with and without additions of Mo and Si, and HP type 316L +Hf were strained by constant extension rate testing (CERT) in simulated 288°C BWR NWC at a rate of 3.5 × 10−7/s. CERT test data and fracture analysis showed that IASCC susceptibility increased in order of HP304, HP304+Mo, HP316L+Hf, and HP304+Si. This trend was also observed when comparing fracture surfaces of the same alloys tested by CERT in BWR NWC after proton irradiation. Differences were insignificant among reported crack growth rate (CGR) values for the same neutron irradiated alloys, and no connection between crack initiation and CGR was confirmed from the alloys tested.

In order to investigate the effect of neutron dose rate on tensile properly and irradiation stress corrosion cracking (IASCC) behavior, crack growth rate (CGR) and, tensile tests and microstructure observation have been conducted with type 304 stainless steel specimens. The specimens were irradiated in high temperature water simulating boiling water reactor (BWR) environments up to about 1dpa with two different dose rates at the Japan Materials Testing Reactor (JMTR). While radiation hardening increased with the dose rate, CGR was not affected by the dose rate. Increase of the yield strength of the low dose rate specimens was caused by the increase of number density of Frank loops. Little difference of radiation-induced segregation at grain boundaries was observed in specimens irradiated by different dose rates. Furthermore, no dose rate effect on local plastic deformation behavior was found near crack tip in the crystal plasticity simulation

The Japan Atomic Energy Agency (JAEA) has an in-pile irradiation test plan to evaluate in-situ effects of neutron/γ-ray irradiation on stress corrosion crack (SCC) growth of irradiated stainless steels using the Japan Materials Testing Reactor (JMTR). SCC growth rate and its dependence on electrochemical corrosion potential (ECP) are different between in-pile test and post irradiation examination (PIE). These differences are not fully understood because of a lack of in-pile data. This paper presents a systematic review on SCC growth data of irradiated stainless steels, an in-pile test plan for crack growth of irradiated SUS316L stainless steel under simulated BWR conditions in the JMTR, and the development of the in-pile test techniques.

Crack growth rate testing was performed on CT specimens with doses in the range ~10 – 47.5 dpa. Two specimens of Type 304L (same heat) were tested in BWR and PWR environments with the objective to compare the CGR behavior of fast reactor irradiation with BWR irradiation. Three specimens of heats tested previously, but at other doses, were tested for assessment of dose and K on IASCC. One specimen of Type 304L was tested in BWR NWC and HWC at two different K levels, while two specimens of cold worked Type 316 were tested at various K levels and temperatures in PWR primary water. To assess the effect of temperature on IASCC, two specimens were tested in either BWR NWC and HWC or PWR primary water at different temperatures. The paper will discuss the effects of fast reactor versus light water reactor irradiation, K, ECP, dose and temperature on the CGR.

The key factors affecting crack growth behavior of neutron irradiated stainless steel were investigated in this study. A crack growth rate (CGR) test was conducted under accurate constant K control on a neutron-irradiated 8 mm RCT specimen (9.6 dpa) of high purity 316L stainless steel with Hf addition in simulated BWR (NWC, HWC) and PWR environments at 320°C and 288°C.The effects of water chemistry, electrochemical corrosion potential (ECP), stress intensity factor (K), and temperature on CGR were examined. In addition, the CGR results from this test were integrated with those reported previously in the Cooperative IASCC Research (CIR) program to compared data and determine the effect of the addition of Hf on crack growth rate.

This paper aims to review new results regarding Irradiation Assisted Stress Corrosion Cracking (IASCC) of neutron irradiated Ti-stabilized austenitic stainless steel 08Ch18N10T (chemically similar to AISI 321) from WWER 440 reactor’s core internals of NPP Greifswald decommissioned after 15 years in service. Two components (core barrel and core shroud basket) irradiated in the LWR conditions (5×10−9 — 4×10−8 dpa/s, 260–330°C) to doses about 2–5 dpa were used for the testing.IASCC was investigated by Slow Strain Rate Tensile (SSRT) and Crack Growth Rate (CGR) tests in simulated WWER water environment at 320°C. The IASCC presence been demonstrated if detected the presence of areas of mixed intergranular (IG) and transgranular (TG) fracture on fracture surface. The two tests represent different stress strain conditions for IASCC development, namely for the crack initiation. The test results showed that plane stress condition facilitates IASCC initiation in the thick components. The fracture surface observations indicate that IASCC crack grows based on strain-controlled fracture mechanism.The results are compared with other data obtained by SSRT tests on the steel irradiated in fast reactor. Relation between the results on fast and in-service irradiated materials is mentioned, but disparateness in data not allowed any conclusions.

Irradiation-assisted stress corrosion cracking is of concern for the safe and economic operation of light water reactors. In this study, cracking susceptibility of austenitic stainless steels was investigated by using slow strain rate tensile (SSRT) tests in a simulated pressurized water reactor (PWR) environment. The specimens were irradiated to 5, 10, and 48 dpa in the BOR60 reactor at 320°C. The SSRT results showed that yield strength was increased significantly in irradiated specimens while ductility and strain hardening capability were decreased. Irradiation hardening was found to be saturated below 10 dpa. The irradiated yield strength of cold-worked specimens was higher than that of solution-annealed specimens. Fractographic examinations were also performed on the tested specimens, and the dominant fracture morphology was ductile dimples. Intergranular cracking was rarely seen on the fracture surface. Transgranular cleavage cracking, however, was found more frequently on the specimen tested in simulated PWR environment.

This work deals with the study of the Irradiation Assisted Stress Corrosion Cracking and presents a methodology coupling microstructure,, strain field and crack’s initiation analysis. Its purpose is to improve the understanding of irradiation effects on the mechanical behaviour of internal stainless steel structures in a Pressurized Water Reactor environment.Proton-irradiations were performed at the Michigan Ion Beam Laboratory on specimen that underwent Slow Strain Rate Tensile tests performed in a Pressurized Water Reactor (PWR) environment. In order to correlate cracks initiation to microstructure, full field analysis is performed at a microscale thanks to a scanning electron microscopy (SEM) digital imaging correlation technique coupled with a crystallographic grain orientation mapping performed on the same area after irradiation. Moreover, cracking features were characterized using SEM.

The hybrid model of PWR primary water stress corrosion cracking (PWSCC) in unirradiated Ni alloys presented at the 2009 Environmental Degradation meeting has been extended to irradiated stainless steels in PWR primary environments in a pilot effort reported here. The preliminary IASCC model is an empirical/theoretical hybrid strain rate model that combines submodels developed by various investigators. The major differences from the PWSCC model include using the Rice-Drugan-Sham (RDS) theoretical expression for strain rate near a growing crack in elastic-perfectly plastic materials and including an empirical dose function. The RDS strain rate expression is appropriate for highly-irradiated materials that show little or no strain hardening. The dose function incorporates many investigators’ observations that irradiation has little effect on SCC below some low dose, then an increasing effect as dose increases until there is no further effect above a high-dose saturation level.

A proton-irradiated SUS316 stainless steel was exposed to the simulated BWR NWC environment for 70 hours during a constant extension rate tensile test and the resulted oxide film was examined using transmission electron microscopy. The oxide film on both the unirradiated and irradiated parts of the sample consists of an outer layer of hematite particles and an inner layer of (Fe, Cr, Ni)₃O₄ spinel. Formation of hematite under BWR NWC condition is consistent with the predication by the potential-pH diagram. Both the outer layer and the inner layer of the oxide film show a strong dependence on grain orientation. Some grains exhibit an inner layer thickness of 40–100 nm while some others have barely any oxidation. Persistent damage induced by proton irradiation did not show a strong influence on the oxidation process as the thickness structure and compositions of the oxide film on both the unirradiated and irradiated parts of the sample were very similar.

Irradiation creep and irradiation stress relaxation data have been obtained for CW 316 SS and SA 304L SS in the Halden reactor. The measurements were performed on-line during irradiation. The irradiation creep in this thermal neutron spectrum reactor was observed to be very different than values measured in fast neutron spectrum reactors. The steady state irradiation creep rate was higher in this and other thermal neutron spectrum reactors than in fast neutron spectrum reactors. On the other hand, the transient irradiation creep component was lower in thermal neutron spectrum reactors than fast neutron spectrum reactors. Therefore, fast reactor irradiation creep data are not recommended for application to light water reactors.

This paper is the fourth in a series that reexamines older data on irradiation creep of austenitic stainless steels and combines it with data from more recent experiments to develop a comprehensive but relatively simple understanding of irradiation creep and its dependence on fabrication and environmental variables. Of particular importance are the roles of void or bubble-induced swelling to accelerate creep and of carbide densification to obscure early swelling and creep strains. This paper particularly focuses on one experiment to explore the effect of compressive stress states and another experiment to examine the impact of changes in irradiation temperature and stress level.

A cluster dynamics model is being developed for the prediction of the entire microstructure evolution of austenitic stainless steel (SS) under irradiation. The particular features of the model are the description of (i) the creation and evolution of Stacking Fault Tetrahedra (SFT), (ii) the evolution of the Frank loops and network dislocation microstructure, (iii) the influence of helium considering the mean number of helium atoms in cavities. This model was calibrated against 304L austenitic SS data on Frank loops and SFTs after irradiation in the Bor60 fast reactor at 320°C and 9.4×10−7 dpa/s. The predictions of the model were then compared to experimental data (i) from the Ringhals PWR thimble tube CW316 SS irradiated at 315°C, 10−7 dpa/s, and 20 appm He/dpa, and (ii) EBR-II fast breeder reactor SA316 SS irradiated at 470°C and 6×10−7 dpa/s. Results were found to be in good agreement with the experimental data at 315°C, and in poor agreement at 470°C. The heterogeneous nucleation of cavities at the interface of precipitates, which is known to occur at high temperature and is not described in the model, is believed to be responsible for this discrepancy at 470°C. The effect of helium on the cavity microstructure evolution at (i) 315°C and 10−7 dpa/s and (ii) 470°C and 6x10−7 dpa/s was investigated by comparing predictions of the model at 0 and 20 appm He/dpa. Helium was predicted to have no significant influence on the cavity microstructure at 315°C, and a significant influence at 470°C.

Irradiation assisted stress corrosion cracking appears to be linked to the localization of slip into dislocation channels. Three austenitic steels with varying degrees of cracking susceptibility were irradiated with 2 MeV protons at 360°C to 5 dpa and strained in 288°C simulated BWR conditions. Deformation behavior was characterized by Schmid factors, resolved shear stresses, slip continuity across grain boundaries, and the angle between dislocation channels and the cracked boundaries. Cracking susceptibility was found to correlate with the dislocation channel properties, such as the resolved shear stress and slip continuity at grain boundaries. Higher cracking susceptibility was found at grain boundaries perpendicular to the tensile axis and adjacent to low Schmid factor grains, which have high normal stresses acting on the boundary. Localized deformation and high normal stress have significant roles in IASCC, though they do not fully describe cracking susceptibility.

The effect of environment and prestrain on IASCC of austenitic stainless steels was investigated in post-irradiation CERT tests in Ar at 288°C and simulated BWR NWC. Two alloys susceptible to IASCC were selected for this study, which are commercial alloy 304 and high purity Fe-15Crl2Ni. Samples were irradiated to 5 dpa with 2 MeV protons at 360°C and strained to about 2% in Ar at 288°C. The results showed that the water environment is the key to inducing IASCC at 288°C. No intergranular cracking was observed in either alloy following straining in Ar up to ~3.5%. However, cracking occurred once the samples were subsequently strained to an additional 1% in simulated BWR NWC. Cracks tended to be long and extend over many grain boundaries rather than single grain facets, perhaps due to the prestrain in Ar. Cracking induced by 0.2% plastic strain in simulated BWR NWC in CP304 with 3.4% prestrain in Ar was also observed. Normal stress is critical in determining the crack initiation location in CP304 and Fe-15Cr12Ni when strained in simulated BWR NWC because the cracked grain boundaries were preferentially aligned perpendicular to the tensile direction.

The sensibility of precipitation-strengthened A286 austenitic stainless steel to Stress Corrosion Cracking (SCC) is studied by means of Slow Strain Rate Tests (SSRT). First, alloy cold working by Low Cycle Fatigue (LCF) is investigated. Fatigue tests under plastic strain control are performed at different strain levels (Δ ε_p/2=0.2%, 0.5% and 0.8%) in order to establish correlation between stress softening and deformation microstructure resulting from LCF tests. Deformed microstructures have been identified through TEM investigations. Three states of cyclic behaviour for precipitation-strengthened A286 have been identified: hardening, cyclic softening and finally saturation of softening. It is shown that the A286 alloy cyclic softening is due to microstructural features such as defects — free deformation bands resulting from dislocations motion along family plans <111>, that swept defects or γ’ precipitates and lead to deformation localization. In order to quantify effects of plastic localized deformation on intergranular stress corrosion cracking (IGSCC) of the A286 alloy in PWR primary water, slow strain rate tests are conducted. For each cycling conditions, two specimens at a similar stress level are tested: the first containing free precipitate deformation bands, the other not significant of a localized deformation state. SSRT tests are still in progress.

Irradiated AISI 304 stainless steel extracted from the Chooz A center filler assembly has been the subject of a number of studies. Previously the results of slow strain rate tensile and constant load autoclave tests of 30 dpa material have been reported. They showed an influence of temperature, strain rate and environment on the fracture behavior of the material. The irradiated microstructure and deformation micro structures of those materials following testing have now been examined by TEM. The findings suggest that substantial channel deformation was associated with the purely ductile fracture following SSRT testing in argon, while the intergranular fractures following SSRT in simulated PWR environment and constant load testing in both simulated PWR and argon environments were associated with very localized deformation primarily exhibiting alpha martensite. This possibility is discussed in the light of literature.

The degradation of grain boundary strength induced by corrosion is one of the causes of intergranular stress corrosion cracking. The micro tensile testing method for measuring the strength of an individual grain boundary was applied to alloy 600 specimens exposed to simulated PWR primary water. The exposure caused grain boundary oxidation that progressed about 0.2 µm perpendicular to the boundary; and depending on the exposure time, the depth can be over 2 µm. Specimens of dimensions 1×2×4 µm containing one grain boundary were made by focused ion beam (FIB) micro-processing and were tested in tension in an FIB system. Intergranular fracture occurred at 180–300 MPa for an oxidized grain boundary while it did not occur even at 1000 MPa for a non-oxidized grain boundary. It was confirmed by transmission electron microscopy (TEM) that the cracking propagated on the interface between the metal matrix and the intergranular oxide.

Nickel-base alloys such as alloy 600 (Ni-16Cr-9Fe) are known to exhibit intergranular stress corrosion cracking (IGSCC) at pressurized water reactor (PWR) primary water environments. From the microscopic observations, it was found that oxygen plays a role in primary water stress corrosion cracking (PWSCC) of nickel-base alloys and Scott suggests an internal oxidation model. However, it was found that needed oxygen diffusivity to explain the internal oxidation model should be several orders greater than the measured oxygen diffusivity. In this study, oxygen diffusion coefficients in the nickel-base alloys were evaluated by atomistic modeling of oxygen diffusion process based on the proposed vacancy-mediated diffusion model. Density functional theory is used to calculate the energy of a system. Activation barrier energy of diffusion of atomic oxygen is quantified by finding minimum energy path through the most favorable path. Phonon analysis is performed using the direct force-constant method.

Alloy 600, a nickel base alloy containing 15 % chromium, is used in the primary circuit of Pressurized Water Reactor (PWR). This alloy is well-known to be susceptible to Stress Corrosion Cracking (SCC) in PWR primary water. Despite the fact that many laboratory studies have been performed and that many models are proposed in the literature, the mechanisms involved are still not well-known. In some models, the transport of species (oxygen, hydrogen and chromium) has a key role. Therefore, experiments and calculations have been performed to study the transport of chromium, hydrogen and oxygen in Alloy 600 and in model microstructures. The results lead to the conclusion that the transport of oxygen and hydrogen cannot be considered as the rate-controlling steps. The asymmetric aspect of the crack tip and of the chromium depletion ahead of the crack lead to the conclusion that chromium diffusion could play a significant role in the mechanism.

The influence of corrosion potential and time of exposure on intergranular oxidation and embrittlement in Alloy 600 has been investigated in simulated primary water. The corrosion potential was observed to have a very significant influence on the proportion of embrittled grain boundaries. A maximum in grain boundary embrittlement was observed at 30 kPa of hydrogen partial pressure while little and almost no embritlement was observed at, respectively, less than 1 kPa and 650 kPa, in good agreement with the known influence of dissolved hydrogen concentration on PWSCC initiation and growth. A significant increase in the depths of intergranular oxide penetration and in grain boundary embritlement was observed with exposure time, the maximum depth of oxide penetrations increasing from 0.6 µm after 500 h, to 2.2 ¡am after 1500h, and up to 5 μm after 4500h of exposure. Intergranular oxide penetrations were also characterized by TEM in order to get insights into the mechanism(s). Cr-depletion was observed ahead of one of the intergranular oxide penetrations, suggesting that accelerated mass transport may play an important role in the oxide penetration process. Accelerated mass transport may also explain the presence of a hexagonal Cr₇S₈ particle observed just beyond the tip of an intergranular oxide penetration ending at a Cr carbide.

Detailed examinations of intergranular attack (IGA) in alloy 600 were performed after exposure to simulated PWR primary water at 325°C for 500 h. High-resolution analyses of IGA characteristics were conducted on specimens with either a 1 µm diamond or 1200-grit SiC surface finish using scanning electron microscopy, transmission electron microscopy and atom probe tomography techniques. The diamond-polish finish with very little preexisting subsurface damage revealed attack of high-energy grain boundaries that intersected the exposed surface to depths approaching ~2 µm. In all cases, IGA from the surface is localized oxidation consisting of porous, nanocrystalline MO-structure and spinel particles along with regions of faceted wall oxidation. Surprisingly, this continuous IG oxidation transitions to discontinuous, discrete Cr-rich sulfide particles up to ~50 nm in diameter. In the vicinity of the sulfides, the grain boundaries were severely Cr depleted (to <1 at%) and enriched in S. The 1200 grit SiC finish surface exhibited a preexisting highly strained recrystallized layer of elongated nanocrystalline matrix grains. Similar IG oxidation and leading sulfide particles were found, but the IGA depth was typically confined to the near-surface (~400 nm) recrystallized region. Difference in IGA for the two surface finishes indicates that the formation of grain boundary sulfides occurs during the exposure to PWR primary water. The source of S remains unclear, however it is not present as sulfides in the bulk alloy nor is it segregated to bulk grain boundaries.

The objective of this study is to link grain boundary oxidation to grain boundary cracking in nickel-base alloys using the stability of Ni and NiO as a frame of reference. Accelerated stress corrosion cracking tests and exposures were conducted on alloy 600, Ni-9Fe, and Ni-9Fe-5Cr (LCr) in constant extension rate mode in supercritical water (SCW) at 400°C using dissolved hydrogen concentrations of 47 cc/kg and 200 cc/kg to control the stability of NiO and Ni respectively. Unstressed samples of Ni-9Fe exposed in the NiO stable regime and LCr exposed in both the Ni and NiO stable regimes show grain boundary oxides extending several microns below the sample surface. Constant extension rate tensile test results showed that cracking was more pronounced in samples where intergranular oxides were able to form, except in alloy 600 where no intergranular oxides formed. Comparison with oxide penetration from 400°C hydrogenated steam revealed that the supercritical water environment was more aggressive, but does not suggest a different mechanism of cracking is operating.

The purpose of this discussion is to describe the Quantitative Micro-Nano (QMN) project. The QMN approach is directed toward predicting SCC in components of the primary and secondary systems of water-cooled nuclear plants. Two week-long meetings, QMN-1 and QMN-2 have been held in the last two years, 2010 and 2011. A third meeting is being planned for 2012 in June. The QMN program is based on the five segments of initiation and propagation: initial condition, precursors, incubation, proto-cracks, and propagation of larger cracks. Each of these segments is the subject of specific meetings where the detailed atomic processes are described and quantified. These mechanistic elements will then be synthesized into overall models. This same approach is being taken in structural biology and electronic materials. QMN benefits from presently available methods for characterizing structure and chemistry at an atomic level and then manipulating the results with arrays of multi-particle computer models. Participation in the QMN meetings is organized by a scientific advisory committee that recommends topics and presenters.

A detailed understanding of strain path effect on IGSCC and oxidation is of great importance for the prediction of the initiation of IGSCC of Ni base alloys 600, 182 and 690 exposed to the primary water of PWRs. In this study, a cross specimen was used in order to quantify the effect of a change of strain path on the strain localization and to increase the understanding of the contribution of the strain hardening and the strain localization at the grain boundaries on the precursors for initiation of SCC. Oxides were characterized using TEM and correlated with the local deformation.

An experimental study of short stress corrosion crack initiation, growth and coalescence has been carried out on two Alloy 600 heats with different susceptibilities to PWSCC.The key result is that maximum crack length and depth were much lower on the heat that exhibited the higher crack initiation rate. In addition, surface crack growth rates were observed to be much lower than expected from measurements on CT specimens and many cracks tended to become dormant after an initial short crack propagation stage. The coalescence criterion proposed by Parkins was found to be appropriate for the description of PWSC crack coalescence.These results indicate that predicting the minimum lifetime of a component based on the minimum time for crack initiation and on crack growth rates measured on pre-cracked specimens may lead both to very over-conservative predictions and to a misleading ranking of lifetimes.The development of a predictive model similar to the Parkins’ model for pipeline steels is not yet possible because it requires:More experimental work to obtain data and understanding of the behavior of short cracks. The experimental methodology developed for this study can be used for this purpose;A 3-D modeling of the local stresses in presence of an array of short cracks. Such work has been undertaken by Kamaya et al., mainly for stainless steels in high-temperature water.

Hump-SSRT tests of alloy 600 have been carried out under simulated primary water and dry hydrogen gas at the temperature ranging from 360 to 320°C and rapid straining electrode tests of nickel based alloys have been carried out under simulated primary water at 320°C with and without dissolved hydrogen (DH).The followings are clarified in this study.1)Intergranular (IG) cracking is observed not only under simulated primary water but also under dry hydrogen gas. Constant loading condition are needed to generate IG cracking under hydrogen gas and the activation energy of PWSCC is almost the same as activation energy of IG cracking under dry hydrogen gas.2)Rapid repassivation of alloy 690 and slow repassivation of alloy 600 are observed under the condition without DH. Repassivation was not observed and only cathodic current is observed under the condition with DH. These cathodic current reaches a plateau value in a few seconds and the cathodic current of alloy 690 is about 1 /5 of alloy 600 and 132.According to these results, it is strongly suggested that the basic mechanism of PWSCC is one kind of hydrogen embrittlement. However, more details of the role of corrosion, surface condition and loading condition should be clarified in future.

Susceptibility of intergranular stress corrosion cracking of Alloy 600 in simulated primary water environment of pressurized water reactor with dissolved hydrogen of 2.75 ppm at 360°C was evaluated by means of slow strain rate testing (SSRT) and numerical analysis to predict the microscopic stress distribution. Mill annealed and 10 or 20 % cold rolled Alloy 600 flat tensile specimens were mechanically polished using SiC papers, then finished with colloidal silica. SSRT was terminated at the tensile strain of 10 %. Many IGSCC cracks were observed on the surface of the tensile specimen. The surface of tensile specimens were characterized by EBSD to investigate the crystallographical feature of the intergranular crack initiation. Alloy 600 tends to exhibit IGSCC at around a 30 degree of misorientation angle. Most crack initiate at random grain boundaries. On the other hand, coincidence boundaries, such as Σ3, Σ9, exhibit no cracks. Susceptibility to IGSCC was also evaluated by numerical simulation in which the crystalline structure of actual test specimen was FEM modeled. In the numerical analysis microscopic stress distribution in each crystals revealed the preferential grain boundaries of IGSCC. The numerical analysis is in fair agreement with the experimental results. Therefore, the numerical calculation proposed is an important method to predict SCC and also the combination of numerical analysis and SSRT with EBSD observation provides basic insight on the mechanism of SCC in high temperature and high pressure water environment.

New stress corrosion studies are presented from initiation tests and crack growth measurements on Alloy 600 in primary PWR environment. The effects of hydrogen, lithium and boron have been studied. Initiation tests have been performed at a high hydrogen content of 70 cc/kg. A longer initiation time was measured compared to intermediate levels but shorter than for low hydrogen concentration of 5 ml/kg. Initiation tests at a Li content of 6 ppm have also been performed. The time to initiation was longer than for 3.5 ppm and about the same as for 2.2 ppm. In addition crack growth measurements using different Li/B contents to simulate a reactor cycle and study any variations with the different composition has been performed. The results show a weak influence of the primary environment. A correlation with the boron concentration was noticed and for high boron contents also a weak dependence on the lithium concentration. The results show that the beginning of cycle conditions usually used for PWSCC studies creates conservative crack growth rates.

Environmentally assisted cracking(EAC) has been studied intensively for several decades, but its mechanisms have not been understood due to the difficulty in crack detection. Therefore it is necessary to predict the cracking behavior using probabilistic estimation. In this study, Probabilistic EAC(PEAC) model is developed and the model is applied to the primary water stress corrosion cracking(PWSCC) of Alloy 600, which is one of the severe EAC problems. In the PEAC modeling, Bayesian parameter updating technique is applied to decrease the uncertainty of parameters by considering the probability of detection. According to this model, it is possible to predict crack growth rate and crack distribution and its results can be comparable with real data. The probability of failure is estimated from developed failure criteria. And the results from this study are applied to estimate the risk reduction by adopting RI-ISI in a PWR.

This study investigates the link between microstructure and stress corrosion cracking susceptibility for three heats of alloy 600 representative of plant components (forged control rod drive mechanism nozzle, rolled divider plate, and rolled divider plate stub). The experimental approach was designed to determine the effect of the manufacturing process (forged versus rolled materials) and the pre-strain (as-received versus pre-strained materials) on stress corrosion cracking of alloy 600. Results showed the carbide distribution to be the main microstructural parameter influencing SCC but a lattice curvature parameter, in synergy with the carbide distribution, has proven to give a better representation of the materials SCC susceptibilities.

The recent field experience and several experimental results have shown the possible deleterious effect of a heat affected zone (HAZ) induced by welding on the susceptibility to the stress corrosion cracking (SCC) of Alloy 600 of bottom penetrations exposed to primary water of PWRs. This work tried to quantify the increasing susceptibility to initiation and crack propagation in 600/182 HAZ. The rolled plate did not exhibit any susceptibility to SCC except for a cold work higher than 10% typically. By contrast, the weld metal was well known for its high susceptibility to SCC. Metallurgical and mechanical characterizations of the HAZ indicated a slight gradient of Vickers micro hardness close to the fusion line (up to few mm) and a lack of intergranular precipitates up to 500 µm from the fusion line. SCC tests clearly demonstrated that a non-susceptible plate may exhibit a significant susceptibility to SCC propagation in the HAZ. Results of initiation tests did not allow to observe any SCC in the base metal, due to the high susceptibility to SCC of the weld.

Many papers have reported about the effects of strain hardening by cold rolling, grinding, welding, etc. on stress corrosion cracking susceptibility of nickel based alloys and austenitic stainless steels for LWR pipings and components. But, the residual strain value due to cold rolling, grinding, welding, etc. is not so quantitatively evaluated.Therefore, authors quantitatively measured and evaluated the residual strain value for strain hardened tensile or torsion specimens of nickel based alloys by applied strain, FWHM of gamma X-ray diffraction, EBSD and Vickers’ hardness measurement to make a calibration curve, at first. Using these curves, the residual strain in the heat affected zone of weld joint and cold worked plates, etc., was quantitatively evaluated from the measuring data of Vickers’ hardness or EBSD or FWHM of gamma peak in X-ray diffraction. As a result, residual strain value in the heat affected zone of weld joint was evaluated as about 10% of von Mises equivalent strain from FWHM, EBSD and Vickers’ hardness measurement. But the residual strain of the 20% and 30 % cold rolled Alloy 600 or TT690 were evaluated about 40% and 60 % of von Mises equivalent strain, respectively.

Microscopic strains associated with stress corrosion cracks have been investigated in stressed C-rings of Alloy 600 boiler tubing. Polychromatic X-ray Microdiffraction (PXM) was used to measure deviatoric strain tensors and the distribution of dislocations near cracks that had been propagated in electrochemically-accelerated corrosion tests. Stress corrosion cracking (SCC)-generated intergranular cracks were produced in two Alloy 600 specimens after 6h and 18h tests. The diffraction patterns and resultant strain tensors were mapped around the cracked area to a one micron spatial resolution. The strain tensor transverse to the crack growth direction showed tensile strain at the intergranular region just ahead of the crack tip for both specimens. Both cracks were found to follow grain boundary pathways that had the lowest angle of misorientation. Dislocation distributions within each grain were qualitatively obtained from the shapes of the diffraction spots and the effect of “hard” and “soft” grains on the crack pathway was explored for both 6h and 18h specimens.

Irradiated reactor pressure vessel (RPV) surveillance data is used to predict decreases in RPV fracture toughness due to irradiation embrittlement. A limited amount of data at fluences that many U.S. PWR RPVs will reach in 60 or more years of operation exists today. However, there is a significant amount of test reactor data available at high fluences, which shows higher embrittlement shifts than the power reactor data-based correlations. This has significant implications for plant operation to 60 years. A coordinated program for withdrawal and testing of the U.S. PWR RPV surveillance capsules is being developed, with the intent of filling high fluence gaps in existing PWR data. This paper summarizes the methodology, optimization strategy, and current results of this coordinated U.S. PWR reactor vessel surveillance program (RVSP). The coordinated RVSP has been optimized to maximize the quantity and quality of high fluence data while minimizing the burden on the industry.

Managing materials’ aging degradation issues is of high importance to the long-term safety and reliability of major components as current PWRs age. Many U.S. utilities have completed the process of renewing their operating license for an additional twenty years. While doing so, they committed to develop aging-management programs and inspection plans.The U.S. PWR industry is proactively developing generic inspection requirements and standards for reactor vessel (RV) internals. This paper describes AREVA NP’s efforts — specifically for Babcock & Wilcox (B&W)-designed units — during the last twenty years, to assist in developing a comprehensive aging-management program for RV internals to fulfill previously made regulatory commitments.

10 CFR 50 Appendix A General Design Criteria (GDC) 4 requires that primary piping systems exhibit an extremely low probability of rupture in order to exclude dynamic effects associated with postulated primary pipe ruptures. The Leak-Before-Break (LBB) methodology, as described in NRC Standard Review Plan (SRP) 3.6.3, was developed to meet this goal. Per SRP 3.6.3, active degradation mechanisms are not permitted in systems approved for LBB and Pressurized Water Reactors (PWRs) are currently experiencing Primary Water Stress Corrosion Cracking (PWSCC). For the long term, NRC began a cooperative research program with the Electric Power Research Institute to develop a probabilistic assessment tool that quantitatively assesses the probability of primary piping system rupture. This paper provides an overview of the xLPR program, focusing on the cooperative structure used for model development and results from the proof-of-concept pilot study.

The Swedish Radiation Safety Authority set up its database of operationally induced damage of mechanical components in the early 1990s. The paper provides an analysis of the database illustrating amongst things system and time dependence of degradation in Swedish nuclear power plants.The Swedish model has been the basis of two similar international databases.

In reactor water chemistry, there are three major chemical parameters: pH, redox potential, and electrical conductivity. The pH is conventionally evaluated by measuring Li and B concentrations. Instead of redox potential, the concentration of dissolved hydrogen is generally measured to evaluate the redox condition. The conductivity data provide the information on the concentration of impurities. Until now, these methods have been used for measuring chemical parameters in nuclear power plants (NPPs). However, these methods have inherent limitations as real-time monitoring means, because they are normally applied at room temperature. In order to overcome this limitation, we developed a water chemistry measurement system for enhancing the accuracy of chemistry monitoring. We adopted a ceramic-based pH electrode as a pH sensor. A Pt electrode was used to measure the redox condition. Also, using a potential transient technique, the Pt electrode was also used for measuring the concentration of corrosive species.

Objectives of this investigation were to compare the effects of advanced amines 1,8-diazabicyclo [5.4.0] undec-7-ene (DBU), ammonia (NH3), dimethyl amine (DMA), Ethanolamine (ETA) and morpholine on solution pH in the presence of FeCl2 and to characterize oxide morphology on steel exposed to amine containing amine solution. Aerated solutions containing100 ppm of DMA or ETA showed the highest room temperature solution pH. DMA containing solution showed the highest resistance to flocculation as indicated by the longer time required to floc. A solution containing 100 ppm DBU did not show any flocculation. Oxide morphology on steel samples exposed to an acidic 100 ppm DBU containing solution showed fine round (about 0.1 µm in diameter) oxide particles.

Characterization of the microstructure and chemical composition of oxide films formed on Alloy 600 in high temperature water with various dissolved hydrogen (DH) levels that allowed the Ni/NiO phase transition to occur were conducted. The results showed that increasing DH in water decreased the thickness of the oxide film, in conjunction with the increase of Cr- but decrease of Ni-concentrations in the oxide film. Further, it was found that the DH deteriorated the stability and protectiveness of the oxide film for values of DH up to and slightly above the Ni/NiO transition value. The DH effect was attributed to an oxidative dissolution process of Ni following the reduction of Ni in the oxide by hydrogen, which suggests the corrosion of Ni-base alloys in high temperature hydrogenated water is dominated by a synergic process of dissolution and oxidation./

This EPRI study quantified the benefits of zinc addition and hydrogen optimization to facilitate modification of inspection intervals for PWR pressure boundary components susceptible to PWSCC. Available experimental and plant data on the effects of such chemical mitigation on PWSCC initiation and crack growth were reviewed. After assessing the statistical confidence in these results, the benefits were quantified. Zinc addition was demonstrated to have a strong mitigative effect on PWSCC initiation of Alloy 600. There is some evidence for a benefit of zinc for crack growth of Alloy 600 at low stress intensity factors. Hydrogen optimization was demonstrated to have a strong mitigative effect on crack growth, particularly in Alloy 82 and 182. The study developed expressions for calculating factors of improvement in initiation times and growth rates, and defined equations and parameters necessary for probabilistic modeling of the benefit of chemical mitigation in the context of relevant regulatory frameworks/

The Supercritical water-cooled reactor (SCWR) is one of the advanced nuclear reactors being developed to meet the soaring energy demand. The corrosion resistance of structural materials used in the SCWR becomes one of the major concerns as the operation conditions are raised up to ~600°C and ~25 MPa as compared to pressurized water reactors (PWRs) at ~315°C and ~15.5 MPa. Oxidation has been observed as the major corrosion behavior. To mitigate the oxidation corrosion, stabilities of metals and oxides need to be understood with respect to environmental temperature and oxygen partial pressure. Computational thermodynamics provides a practical approach to assess phase stabilities of such multicomponent multi-variable systems. In this study, calculated phase stability diagrams of alloys and corresponding oxides were used to guide the interpretation of oxidation behavior of SCW-exposed structural materials. Examples include ferritic-martensitic steel, austenitic steels and Ni-base alloy, e.g., HCM12A (Fe-11Cr), D9 (Fe-15Cr-15Ni), 800H (Fe-21Cr-32Ni), and 690 (Ni-30Cr-10Fe). Calculated results are in good overall consistence with the experimental data./

Tapered uniaxial tensile samples under constant load and bent ring samples under constant strain were used to investigate the stress corrosion cracking (SCC) of Alloy 600 and 690 in supercritical water. To verify the effectiveness of the tapered tensile sample design, the SCC of two 304 stainless steels with high and low carbon content were tested in supercritical water at 400°C, 25MPa and 10–15ppb dissolved oxygen. Both testing methods are effective to initiate SCC in 304SS, Alloy 600 and 690. Based on the bent ring SCC testing, the SCC resistance of Alloy 690 in supercritical water is lower than Alloy 600. This unexpected lower SCC resistance is probably due to the much lower tensile ductility (less than 10%) of the tested Alloy 690 as compared with the standard 40–50% ductility and the severe plastic deformation and strain in the bent ring samples. The large TiN particles and their inhomogeneous distribution in the alloy may promote the SCC initiation in this specific Alloy 690 in the bent ring samples. SCC is more likely to initiate from the Electrical Discharge Machined (EDM) or end mill machined surfaces than from ground and polished surfaces.

The 14CrODS alloy was corroded in both steam and supercritical water (SCW) at 500°C. The weight gain observed in SCW was consistently about 1.5 times higher than that in steam and the corrosion rate was higher in SCW compared to steam. To explain this discrepancy, the oxide microstructure of samples corroded in both environments was analyzed using microbeam synchrotron diffraction and fluorescence, and scanning electron microscopy (SEM). In both environments, the oxide formed a three-layer structure, similar to previous observations on ferritic-martensitic alloys. Compared to the SCW samples, the steam samples exhibited thinner and denser oxide layers, which contained more Cr₂O₃ especially at specific interfaces. These results are discussed and an explanation for the divergence in oxidation behavior between steam and SCW is proposed.

Grain boundary engineering (GBE) was performed on nickel-based alloy 690 by thermomechanical processing (TMP) to alter the grain boundary character distribution (GBCD). It was found that 5% and 35% thickness reduction in single and multiple steps followed by solution annealing and water quench yielded a high fraction of special boundaries. The total length fraction of the low ∑ CSL (coincidence site lattice) was as high as 87.2%. The grain boundary network was disrupted after the TMP treatment, and the average grain size calculated after exclusion of special twin boundaries can be as much as 5 times larger than the as-received (AR) sample. The GBE sample showed better oxidation resistance compared to the AR sample during the long term air oxidation. In the cyclic oxidation test, both AR and GBE samples showed a mass gain at the beginning of the test which was then followed by a mass loss. The mass change of GBE samples oscillated after the first couple cycles, while the AR sample became relatively stable. The oxide film most likely consists of duplex structures with one stable layer that was formed inside and one unstable layer that was formed outside. The stable inner layer was the protective layer and prevented alloy 690 from further oxidation.

SCC is strongly influenced by water chemistry parameters, especially when crack chemistry can be concentrated from differential aeration or thermal gradients or boiling. Mitigation of the effects of the high corrosion potential associated with oxidants is markedly and efficiently accomplished by electrocatalysis, which requires that there be a stoichiometric excess of reductants over oxidants. Mechanisms and criteria for effective SCC mitigation are summarized, with particular focus on the critical location for the catalyst in a crack and experimental support for these concepts. Optimization of electrocatalysis by OnLine NobleChem- is described, for example where Pt is injected at levels of 0.002 to 0.05 ppb in the reactor water.

Boiling water nuclear reactors (BWRs) throughout the world have applied the NobleChem™ (or noble metal chemical addition: NMCA) or Online NMCA (OLNC) process just before end-of-cycle shutdown or during an operation to mitigate the stress corrosion cracking (SCC) of structural materials in BWRs. When injected into BWR environments, the noble metal particles deposit on Type 304 stainless steel surfaces and reduce the corrosion potential, which decreases the propensity for SCC. Very fine noble metal particles are formed and able to potentially deposit inside a crack and maintain catalytic surfaces in the critical regions inside the crack. However, the current NMCA or OLNC process also introduces the unnecessary ionic species to the reactor water.

The effect of minor additions of platinum group metals (PGMs) on the electrochemical behaviour and stress corrosion cracking (SCC) resistance of 304 stainless steels was investigated under different environmental conditions relevant to light water reactors. Corrosion studies were performed in simulated PWR chemistry at ambient temperature in order to evaluate the catalytic effects of the PGM additions. High temperature SCC tests were carried out within autoclave environments using slow strain rate tensile (SSRT) and constant loading methods. The pattern of results obtained in this study provide strong evidence that minor additions of ruthenium to 304SS are beneficial in terms of SCC resistance and general macroscopic corrosion behaviour. The improved resistance to intergranular SCC may be associated with enrichment of ruthenium and molybdenum species within the dual oxide surface layers, as revealed by high resolution transmission electron microscopy.

On-line noble metal chemical addition (OLNC) is the third generation of hydrogen water chemistry developed to maintain the ECP of boiling water reactor structural materials in a range that mitigates intergranular stress corrosion cracking. The method utilizes the on-line injection of dilute Na₂Pt(OH)₆ into the feedwater over a period of approximately 10 days. The first application of OLNC occurred in July of 2005 and a total of 17 BWRs have applied the technology to date, with many more applications scheduled. It is expected that OLNC will become the de facto standard because it eliminates 60 hours of outage application time and it addresses the crack flanking concerns that can arise under certain conditions. Shut down dose rate data are now available for 7 plants, several with multiple cycles. This paper examines this behavior and it’s implication for optimizing future OLNC applications and post application operations./

In boiling water reactors (BWRs), stress corrosion cracking (SCC) has occurred in Alloy 600 and Alloy 182, 132 and 82 weld metals. Most BWRs mitigated SCC by injecting H₂ into the feed water, although the levels vary by ~8X depending on whether they use electrocatalysis and low H2, or only moderate H₂. Under moderate hydrogen water chemistry conditions, the reactor water has ~250 ppb H2, which is very close to the Ni/NiO phase boundary at 274 ºC, the temperature at which most BWR structural materials operate. A peak in growth rate has been observed at the Ni/NiO phase boundary, and extensive work has quantified and modeled its effect in PWR primary water. It was proposed that the same effects of dissolved H₂ that have been reported in PWR primary water applies to pure BWR water, primarily because in deaerated water, B/Li have no discernible effects on crack growth. This study evaluates the effect of dissolved H₂ on Alloy 182 weld metal at temperatures down to 274 ºC on SCC growth rate.

Boiling water reactors (BWRs) operate with very high purity water. However, even the utilization of near theoretical conductivity water cannot prevent intergranular stress corrosion cracking (IGSCC) of sensitized stainless steel, wrought nickel alloys and nickel weld metals under oxygenated conditions. IGSCC can be further accelerated by the presence of certain impurities dissolved in the coolant. The goal of this paper is to present the technical basis for controlling various impurities under both oxygenated, i.e., normal water chemistry (NWC) and deoxygenated, i.e., hydrogen water chemistry (HWC) conditions for mitigation of IGSCC. More specifically, the effects of typical BWR ionic impurities (e.g., sulfate, chloride, nitrate, borate, phosphate, etc.) on IGSCC propensities in both NWC and HWC environments will be discussed. The technical basis for zinc addition to the BWR coolant will also provided along with an in-plant example of the most severe water chemistry transient to date.

The coolant in a boiling water reactor (BWR) usually contains a relatively high level of dissolved oxygen from intrusion of atmospheric air during a cold shutdown. Accordingly, the structural materials in the primary coolant circuit (PCC) of a BWR could be exposed to a strongly oxidizing environment for a short period of time during a subsequent startup operation. Due to limitedmeasurable water chemistry data, a well-developedcomputer code DEMACE was used in the current studyto investigate the variations in redox species concentration and in electrochemical corrosion potential (ECP) of components in the PCC of a domestic BWR during startup operations. Our analyses indicated that the dissolved hydrogen level in the reactor coolant at a low power level without steam generation in the core was higher than that at a power level with a minor amount of steam generated in the core. The dissolved oxygen concentrations in the reactor coolant would be relatively high and more than 1000 ppb during startup operations at power levels greater than 2.5%. In the meantime, the concentrations of hydrogen peroxide could be more than 2000 ppb at the core outlet region during startup operations, which rendered a strongly oxidizing coolant environment. The electrochemical corrosion potentials of structural componentsin the PCC of the analyzed BWR generally followed the concentration trend of H2O2. It was predicted that the coolant environment in a BWR during a plant startup could be highly oxidizing, and the structural components would therefore suffer from a more serious corrosion problem than that under operations at the rated power level.

In this study, water radiolysis occurring in nuclear power plants was simulated by sonochemistry. Generated hydroxyl radicals can recombine in others species such as H₂O₂ and H₂. It is shown that solution conductivity is an important parameter on the evolution of open circuit potential due to the thickness variation of the diffusion layer which may contain sonolysed species (OH, H₂, H₂O₂) in different concentrations. Current densities increase under ultrasonic irradiation due to an increase of mass transport and charge transfer and to the presence of sonolysed species. The oxide film formed under ultrasonic for 1 h at a passive potential of +0.2 V_SCE shows early stage of passivation and higher disordered state which imply a great decrease of the corrosion resistance behavior of the sample. The polarisation resistance R_p, of the stainless steel is divided by a value of 4.5 under ultrasonic irradiation.

NobleChem™ has been successfully applied to commercial BWRs worldwide just before end-of-cycle shutdown, during a mid-cycle shutdown, or normal plant operation. This process is a proven technology for mitigating the intergranular stress corrosion cracking (IGSCC) of structural materials by catalyzing the oxygen and hydrogen recombination and eventually lowering the electrochemical corrosion potential (ECP) below the IGSCC protection potential (-230mV_SHE). In order to achieve the catalytic benefit of noble metal at metal surfaces, excess hydrogen is essential. However, there may be certain locations where neither noble metal nor hydrogen is present.

Incidents of intergranular stress corrosion cracking (IGSCC) have occurred in boiling water reactors (BWRs) for decades. The electrochemical corrosion potential (ECP) is currently a major indicator for the IGSCC susceptibility of stainless steel (SS) components in BWR environments. This study proposes a novel technique of titanium dioxide (TiO₂) treatment to mitigate the IGSCC problems in BWRs that could eventually lead to a lower demand of dissolved hydrogen for hydrogen water chemistry (HWC). Electrochemical polarization analyses and ECP measurements were conducted to investigate the impact of ultraviolet (UV) radiation on the electrochemical behavior of oxygen and TiO₂ treated specimens in 288 °C pure water. Prior to the electrochemical tests, all specimens were thermally sensitized and pre-oxidized in high temperature pure water containing 300 ppb dissolved O₂. Afterwards, 38 nm TiO₂ nanoparticles were deposited on the specimens by hydrothermal deposition at 150 °C for 96 hrs. The surface morphologies of the specimens were examined by scanning electron microscopy (SEM), energy dispersive X-ray spectroscopy (EDX) and Laser Raman Spectra (LRS). SEM and EDX results showed that the distribution of TiO₂ deposited on the oxides with both hematite (α -Fe₂O₃) and magnetite (Fe₃O₄) structures was not uniform and continuous. LRS results showed that the TiO₂ particles on the treated specimens had an anatase-type structure. In addition, the ECPs of the TiO₂ treated specimens with UV radiation were 100 mV lower than those without UV in high temperature water containing various levels of dissolved O₂. The results of electrochemical potentiodynamic polarization analysis revealed that the corrosion current densities of the treated specimens and the exchange current densities of the O₂ reduction reactions were comparatively lower in the presence of UV radiation. Without UV radiation, however, no significant differences were observed between the TiO₂ treated and untreated specimens. These results indicate that the TiO₂ treatment in combination with UV radiation would effectively reduce the corrosion rate of Type 304 stainless steels in high temperature oxygenated environments.

Reactor internal components are subjected to neutron irradiation in light water reactors, and with the aging of nuclear power plants around the world, irradiation-induced material degradations are of concern for reactor internals. Irradiation-induced defects resulting from displacement damage are critical for understanding degradation in structural materials. In the present work, microstructural changes due to irradiation in austenitic stainless steels and cast steels were characterized using transmission electron microscopy. The specimens were irradiated in the BOR-60 reactor, a fast breeder reactor, up to ~40 dpa at ~320°C. The dose rate was approximately 9.4x10-7 dpa/s. Void swelling and irradiation defects were analyzed for these specimens. A high density of faulted loops dominated the irradiated-altered microstructures. Along with previous TEM results, a dose dependence of the defect structure was established at ~320°C.

Steam Generator Divider Plates (SGDPs) of EDF 900 MWe PWRs have encountered short SCC cracks (depth less than 2 mm). The objective of the present investigation was to demonstrate that a crack initiated in a superficial cold worked layer (< 2 mm) in a SGDP would stop before it reached the bulk material.Archive materials have proved enhanced susceptibility to SCC in laboratory when the pre-strain was higher than 0.07. Additionally, constant displacement tests were performed on notched shot-peened tensile specimens in primary water (up to 5000 h, 360°C) using a crack monitoring. Cracks initiated at the notch and in the smooth parts of the specimen, close to the weld spots used for the monitoring. Crack depths reached 450–650 μm and hardness at the crack tips were 235–245 HV0.1, corresponding to a tensile strain of 0.07–0.08, in accordance with examinations performed on a removed SGDP.It seems reasonable to assume that short cracks in plants are arrested.

Stress Corrosion Cracking of nickel alloys has been a major concern for all the Nuclear Power Plants over the last forty years. Since 2002, some cases of Stress Corrosion Cracking (SCC) have been reported on Steam Generator (SG) Divider Plates. However, evidence of propagation following the first detection has never been observed (based on nearly one hundred in-service inspections).EDF has been conducting an extensive program of investigations on two divider plates from decommissioned Steam Generators concerned by SCC. This program allowed us to reach two main objectives: Confirm the high reliability of non-destructive examinations to detect and characterize inservice SCC flaws,Improve our knowledge on SCC occurring on SG divider plates (relationship between morphology and microstructure).All these elements lead to a better understanding of the SCC damage of the SG divider plates and contributed to strengthen the maintenance policy.

NDE investigations using ultrasonic inspections were performed on core basket bolts at a WER 440 unit. Bolts with indications were replaced. Destructive investigations were performed on three bolts with indications. These investigations comprised of micro structural investigations using optical microscopy, hardness measurements and scanning electron microscopy of fracture surfaces. The bolts are M12 bolt manufactured from solution annealed Ti-stabilized stainless steel. Only one bolt had suffered from cracking. The washer of this bolt had been welded to the shielding plate, which had resulted in unintentionally high stresses. The observed cracking is regarded to be a result of intergranular stress corrosion cracking, enhanced by irradiation. The reason for the NDE indications in the other bolts were incomplete filling of the flat slot of the bolt head when the bolt was welded to the washer to prevent unscrewing of the bolt during operation.

This paper describes the results of a laboratory analysis performed on a reactor coolant pump (RCP) seal that experienced elevated leak-off temperatures after three months in service. Analyses included visual examinations, dimensional measurements, scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), metallography, Vickers microhardness, gamma spectroscopy, and x-ray diffraction. It was determined that the elevated leak-off temperatures were caused by a buildup of orange-brown deposits on the seal surfaces. These deposits contained primarily iron and oxygen along with 8–10% chromium. Sequential gamma spectrographic analysis through the deposit layer indicated the deposits were all approximately the same age, 300+ days, based on the Co-58/Co-60 ratios. The laboratory data indicated deposition of corrosion products, not corrosion/degradation of the seal, was the most plausible mechanism for the deposit buildup. The source of the deposits appeared to be a Type 410 stainless steel component in the system external to the RCP seal.

Few cases of SCC have been observed in French PWRs, in high strain hardened and non-sensitized austenitic stainless steels exposed to nominal primary environment. In particular, Intergranular SCC has been observed on several pressurizer heaters. Thus, a R&D program using hot laboratory investigations has been conducted in order to identify the root causes of the degradation, to understand the mechanisms responsible for SCC in nominal primary water, and to improve the resistance of heaters.A surface annealing process based on induction heating was used, in collaboration with heater supplier Thermocoax, to reduce the superficial cold work and the residual stresses to avoid any SCC initiation while preserving electrical properties of mineral insulating material.The combination of optimized parameters and process industrialization has produced positive results for the prevention of crack initiation.

This paper presents results of residual stress measurements and modelling within the cladding and J-groove weld of Control Rod Drive (CRD) specimens in the as-welded and Post Weld Heat Treated (PWHT) states. Knowledge of the residual stresses present in CRD nozzles is critical when modelling the fracture mechanics of failures of nuclear power plant components to dictate inspections intervals and optimise plant downtime. The specimens comprised of ferritic steel blocks with 309L stainless steel cladding and a single J-groove weld attaching the 304 stainless steel nozzles. Multiple measurements were made through the thickness of the specimens in order to give biaxial residual stress profiles through all the different fusion boundaries. The results show the effect of PWHT in reducing residual stresses both in the weld and ferritic material. The beneficial use of measurements is highlighted to provide confidence in the modelled results and prevent over conservatism in integrity calculations, costing unnecessary time and money.

Biblis Unit A is a Siemens/KWU type four loop PWR equipped with SG tubing made from Stainless Steel Alloy 800, operating from 1975 until the shutdown in 2011 due to the German political decision of a quick nuclear phase out. During the regular SG tubing inspections using eddy current testing in 2005 and 2006, a few SG tube indications were detected within the tube sheet between upper and lower mechanical tube expansion. These indications were limited to the outer tube bundle periphery. As a consequence two tubes with EC indications were pulled in 2006 for further investigations. The destructive examinations revealed axially oriented cracks starting from the outer tube surface (ODSCC). The analysis of the debris from the upper expansion area clearly indicated the presence of secondary side water in the volume between tube sheet and SG tube. Penetration of secondary side water is only possible by a penetration path along the upper expansion.The ODSCC found in this area within the tube sheet is of no safety relevance because it could at maximum lead to a limited leakage from primary to secondary side.Nevertheless, a comprehensive root cause analysis was initiated by the NPP operator to clarify in which way such a penetration path can develop. To identify the source of circumferential tensile stresses as precursor of SCC at the tube OD, investigations on the stress level of the heating tubes were carried out. The study includes an assessment of the manufacturing and operation of the SGs and the analysis of the deformation behavior in the upper expansion areas well as a consideration of the secondary side water chemistry data. One target of the root cause analysis was to focus future SG tube NDE. The results are summarized in this paper.

A leak was discovered in a Type 316 stainless steel socket weld in the sampling line for the boron injection tank. A section of the pipeline containing the leaking weld was removed for laboratory investigation that included visual and Stereovisual inspections, liquid penetrant (PT) testing, metallography, scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), and ferrite content determinations. The leak path was a through-wall transgranular crack in the socket weld. Cracking initiated along the weld-metal-to-base-metal interface at the tip of the crevice between the socket and pipe. The crevice was exposed to oxygenated boron solution at <180°F. Shallow intergranular attack (IGA) was found in the exposed base metal inside the crevice. Based on the investigation results, it was concluded that transgranular stress corrosion cracking (TGSCC) is the primary cracking mechanism.

In 2009, 12 tubes were pulled at the tube support plate level of a Steam Generator at one of France’s oldest nuclear power plants. In order to identify the involved mechanisms and to characterise the defects, metallurgical examinations were carried out on these tubes.In addition, an objective was to evaluate the pressure that the tubes could withstand without leakage. Leakage/burst pressure tests are analysed and an attempt is made to correlate these results to the shape of the defect. Results suggest that the leakage pressure is mainly influenced by one parameter — the initial depth of the damage.This is carried out firstly using results from previous investigations into pulled tubes, of which several came from a decommissioned Steam Generator and included a part of their tube support plate and, secondly, using analytical modelling.

Fouling of steam generators has a significant negative impact on the material and thermal performance the steam generators of pressurized water reactors. Corrosion products that originate from various components in the steam cycle of a nuclear power plant get pumped forward with the feed water to steam generators where they deposit on the tube bundle, tube support structure and the tube sheet. Heavy accumulation of deposit within the steam generator has led to some serious operational problems, including loss of thermal performance, under deposit corrosion, steam generator level oscillations, flow accelerated corrosion of carbon steel tube support plates and the failure of steam generator tubes due to high cycle fatigue.This paper will review the factors affecting steam generator fouling, examine the relationship between fouling and degradation of the thermal and material performance of steam generators and investigate the effectiveness of remedial measures to mitigate fouling.

Stainless steel components in cooling water systems, using brackish- or seawater as cooling medium, are potentially vulnerable to corrosion. In many cases such corrosion issues are mainly ascribed to Microbially Influenced Corrosion (MIC). Within the scope of a study, field tests were performed in different environments (Swedish East Coast, Swedish West Coast, mouth of the river Elbe in Germany). In some cases even such materials, that are assumed to be resistant, have been affected by corrosion. However, based on the outcomes of the field tests, it can be concluded, that all results, where corrosion occurred, are explainable by environmentally assisted chloride induced corrosion. Crevice conditions produced from consolidated biofilms on component surfaces or from sludge deposits, most probably support the initiation and the progress of this kind of corrosion. Based on the test results, practical recommendations are derived, as how to prevent corrosion in brackish- or seawater environments.

Cold working can significantly increase the susceptibility of metals to environmentally assisted cracking. However, the reasons for this increase susceptibility are still unclear. This is due in part to the difficulty in quantifying and modelling plastic deformation at the required scale. Here, we use a new experimental procedure to study the local microstructural distribution of strain in 304L stainless steel. Digital image correlation was used to map strain at the microstructural level with sub-micron resolution. The results clearly show that a high degree of strain localization develops within individual grains, in the form of highly localized shear bands and micro-twinning. Electron backscatter diffraction was used to quantify the lattice orientation changes in the same area. Analysis of this data included the calculation of kernel average misorientation and of intragranular orientation spread following grain reconstruction. Comparisons of results clearly show that, in most cases, there is no evidence in the lattice orientation data analysis of the high levels of strain measured. This has important implications in the use of lattice orientation data in the study of the effects of plastic deformation on environment-assisted cracking.

Increased susceptibility to stress corrosion cracking (SCC) of stainless steels in good quality PWR primary coolant has been linked to prior deformation in the material. Deformation heterogeneity at the micro structural level is believed to induce susceptibility, however only limited research has been undertaken to quantify the link between microstructure and susceptibility. Analysis of electron backscattered diffraction data has been used to provide a unique description of the cold rolled micro structure of a grade of cold worked 304H stainless steel, which has been found to show high SCC propagation rates. The deformation twin lamellae within the bulk cold rolled material have been found to show preferential alignment with respect to the macroscopic direction of rolling. Additionally, deformation twins are found to interact with the crack significantly. The potential influences of this preferential micro structure alignment on the SCC behavior of the material are discussed.

In AISI 304 stainless steel pipe welds weld shrinkage causes large variations in residual plastic strain in different parts of the weld metal and heat-affected zone (HAZ). The amount of strain was analyzed by EBSD quantitatively by comparing the intra-grain misorientations to the calibration curve. Highest degrees of plastic strain (10...20%) were detected in the HAZ close to the root area of a prototypical BWR plant weld. Strain in the weld metal varies in the different directions of solidification, being high in the weld bead boundaries and near the fusion lines. Preliminary studies of the effects of mechanical and elastic anisotropy of the weld metal microstructure on the grain size level were performed by EBSD and nanoindentation. The residual stress distribution in the same weld cross-section was determined by a contour method. The residual strain and stress distributions are superimposed and EAC susceptibility of various areas of the pipe weld is evaluated and discussed.

This research assesses the thermal stability of lean grades of duplex stainless steel relative to standard grade alloys. Both hot rolled plate and gas-tungsten-arc weld deposits of selected alloys were isothermally aged in the temperature regime of α-α’ phase separation (T = 800°F / 427°C) for times up to 1,000 hours. The degree of embrittlement was assessed via changes in hardness and Charpy impact energy. Additionally, the materials were characterized by light optical microscopy, electron microprobe analysis (EMPA), analytical electron microscopy (AEM), and x-ray diffraction (XRD) to better understand the factors that affect embrittlement. Impact testing shows that at equivalent thermal exposure, small changes in alloy composition have a significant effect on the degree of embrittlement. Microscopy reveals that spinodal decomposition of the ferrite occurs in all the alloys tested at 800°F (427°C). Additionally, transmission electron microscopy shows complex intermetallic formation, likely G phase, in the standard grade weld metal. Relative to their standard grade counterparts, both the lean grade plate and the lean grade weld deposit display ~10X slower embrittlement kinetics for the conditions studied. Despite the relatively complex metallurgy of these alloys, this research indicates that lean grade duplex stainless steels could have broad applicability to lower temperature components in nuclear power systems.

Field experience of 300 series stainless steels in the primary circuit of PWR plant has been good. Stress Corrosion Cracking of components has been infrequent and mainly associated with contamination by impurities/oxygen in occluded locations. However, some instances of failures have been observed which cannot necessarily be attributed to deviations in the water chemistry. These failures appear to be associated with the presence of cold-work produced by surface finishing and/or by welding-induced shrinkage. Recent data indicate that some heats of SS show an increased susceptibility to SCC; relatively high crack growth rates were observed even when the crack growth direction is orthogonal to the cold-work direction. SCC of cold-worked SS in PWR coolant is therefore determined by a complex interaction of material composition, microstructure, prior cold-work and heat treatment. This paper will focus on the interactions between these parameters on crack propagation in simulated PWR conditions.

Stress corrosion cracking (SCC) initiation tests have been conducted on Alloy 600 at temperatures from 304 to 367°C. Tests were conducted with in-situ monitored smooth tensile specimens under a constant load in hydrogenated environments. A reversing direct current electric potential drop (EPD) system was used for all of the tests to detect SCC initiation. Tests were conducted to examine the effects of stress (and strain), coolant hydrogen, and temperature on SCC initiation time. The thermal activation energy of SCC initiation was measured as 103 ± 18 kJ/mol in hydrogenated water, which is similar to the thermal activation energy for SCC growth. Results suggest that the fundamental mechanical parameter which controls SCC initiation is plastic strain not stress. SCC initiation was shown to have a different sensitivity than SCC growth to dissolved hydrogen level. Specifically, SCC initiation time appears to be relatively insensitive to hydrogen level in the nickel stability region.

The fracture toughness of nickel based alloys can be degraded when exposed to low temperature deaerated water. This paper uses elastic-plastic fracture mechanics testing to explore some fundamental environmental and metallurgical factors that influence this form of environmentally assisted cracking. Elastic-plastic fracture toughness tests were performed on 0.6T compact tension samples of EN82H weld metal as a function of temperature, hydrogen content in the water, and hydrogen content in the weld metal. These results are used in conjunction with literature data and fundamental understanding to outline the environmental and material conditions that promote susceptibility to this form of hydrogen embrittlement.

Wrought Alloy 690 is well established for corrosion resistant nuclear applications but development continues to improve the weldability of a filler metal that retains the corrosion resistance and phase stability of the base metal. High alloy Ni-Cr filler metals are prone to several types of welding defects and new alloys are emerging for commercial use. This paper uses experimental and computational methods to illustrate key differences among welding consumables. Results show that solidification segregation is critical to understanding the weldability and environmentally-assisted cracking resistance of these alloys. Primary water stress corrosion cracking tests show a marked decrease in crack growth rates near 21 wt. % Cr at the grain boundary. While filler metals with 21–29 wt.% grain boundary Cr show similar PWSCC resistance, the higher alloyed grades are more prone to solidification cracking. Modeling and aging studies indicate that in some filler metals minor phase formation (e.g., Laves and σ) and long range order (LRO) must be assessed to ensure adequate weldability and inservice performance.