Stephen M. Bruemmer

Radiation-induced material changes and susceptibility to intergranular failure of light-water-reactor core internals

Abstract Current understanding of radiation-induced material changes that occur in light-water-reactor (LWR) core components is critically reviewed and linked to intergranular failure processes. Although the basic science of radiation damage processes in metals is reasonably well established, accurate prediction of microstructures, microchemistries and mechanical property changes in complex stainless alloys during irradiation at LWR temperatures is not possible at present. Mechanistic understanding of these radiation-induced changes in commercial alloys is considered to be of paramount importance for the mitigation of the intergranular environmental cracking that occurs in service. Fundamental research is needed to define defect–solute interactions and microstructural evolution at intermediate temperatures and dose rates pertinent to LWRs where transient effects often dominate behavior. In addition, it is essential that radiation effects on matrix microstructure and microchemistry and grain boundary microchemistry be understood. Finally, a stronger emphasis on accurately quantifying radiation effects on environmental cracking mechanisms and kinetics is needed.

Microstructural and microchemical mechanisms controlling intergranular stress corrosion cracking in light-water-reactor systems

This review paper examines mechanisms controlling IGSCC in selected LWR components. Emphasis is placed on identifying material microstructures and microchemistries which promote susceptibility to premature failure. Two important examples are evaluated in some detail: stainless steel pipe cracking and primary-side SCC of alloy 600 steam generator tubing. In each case, grain boundary segregation and precipitation phenomena in these materials are reviewed and assessed relative to the mechanisms of IGSCC. This paper summarizes materials presented at the 1993 International Summer School on the Fundamentals of Radiation Damage held at the University of Illinois. A more comprehensive overview of SCC mechanisms and LWR examples was provided at the school, but will not be included in this article. Microstructural and microchemical aspects controlling IGSCC described here serve as a lead-in to the following paper focussing on how irradiation influences SCC resistance of reactor core components.

High-Resolution Characterization of Intergranular Attack and Stress Corrosion Cracking of Alloy 600 in High-Temperature Primary Water

Abstract Intergranular (IG) attack regions and stress corrosion cracks in Alloy 600 (UNS N06600) U-bend samples tested in 330°C pressurized water reactor water were characterized by analytical tran...

Influence of irradiation temperature and dose gradients on the microstructural evolution in neutron-irradiated 316SS

A cold worked 316SS baffle bolt was extracted from the Tihange pressurized water reactor and sectioned at three different positions. The temperature and dose at the 1-mm bolt head position were 593 K and 19.5 dpa respectively, whereas at two shank positions the temperature and dose was 616 K and 12.2 dpa at the 25-mm position and 606 K and 7.5 dpa at the 55-mm position. Microstructural characterization revealed that small faulted dislocation loops and cavities were visible at each position, but the cavities were most prominent at the two shank positions. Measurable swelling exists in the shank portions of this particular bolt, and accompanying this swelling is the retention of very high levels of hydrogen absorbed from the environment. The observation of cavities in the CW 316SS at temperatures and doses relevant to LWR conditions has important implications for pressurized water reactors since SA 304SS plates surround the bolts, a steel that usually swells earlier due to its lower incubation period for swelling.

Effects of irradiation on intergranular stress corrosion cracking

Abstract Intergranular stress corrosion cracking (IASCC) is a pervasive and generic problem in current light water reactor and advanced reactor designs that can lead to widespread component failure. IASCC is believed to be due to either to changes in the grain boundary composition, the microstructure or the water chemistry and corrosion potential. Of greatest interest are the changes in composition and microstructure since IASCC exhibits a well-defined, although not invariant, dose threshold. Changes in grain boundary composition are a result of radiation-induced segregation (RIS) and result in enrichment of nickel, depletion of chromium as well as changes in the impurity element compositions at the grain boundary. Although the basic theory of RIS is believed to be understood, quantitative descriptions of observed changes are not yet possible and hinder the correlation between RIS and IASCC. Changes in the microstructure are intimately linked to the strength and ductility of the irradiated alloy and strong correlations between IASCC and irradiated yield strength have been found. However, a fundamental understanding of the deformation mechanisms and the way in which deformation is coupled to IG cracking in alloys irradiated under LWR conditions (250–360°C, 1–5 dpa) is lacking. Finally, although radiation is known to affect IGSCC through changes in water chemistry and corrosion potential, it is not a necessary condition. Overshadowing and slowing progress on this important problem is a lack of well-defined-data from properly irradiated and properly characterized materials, due principally to inherent experimental and financial difficulties. As such, the specific mechanism(s) of IASCC remain unknown.

Microstructure and Microdeformation Effects on IGSCC of Alloy 600 Steam Generator Tubing

Abstract Microdeformation characteristics in alloy 600 tubing have been examined after various tensile deformations. Microstructure developed during processing was found to control subsequent microdeformation behavior. Grain boundary carbides were the most effective source of dislocations, activating at lower macrostrains and continuing to operate at higher macrostrains than other sources. Ledges within grain boundaries, twin boundaries, and matrix carbides also acted as dislocation sources. Most dislocation activity at low strains was confined to planar arrays. A conceptual model is presented to explain the effects of interfacial and matrix microstructure on microdeformation and primary side stress corrosion cracking (SCC) of alloy 600 tubing. Microstructure is linked to intergranular stress corrosion cracking (IGSCC) resistance through its influence on microdeformation behavior and the resultant crack tip stress state. Dislocation source activity at grain interfaces is proposed to be critical in control...

Influence of sulfur, phosphorus, and antimony segregation on the intergranular hydrogen embrittlement of nickel

The effectiveness of sulfur, phosphorus, and antimony in promoting the intergranular embrittlement of nickel was investigated using straining electrode tests in IN H2SO4 at cathodic potentials. Sulfur was found to be the critical grain boundary segregant due to its large enrichment at grain boundaries (104 to 105 times the bulk content) and the direct relationship between sulfur coverage and hydrogen-induced intergranular failure. Phosphorus was shown to be significantly less effective than sulfur or antimony in inducing the intergranular hydrogen embrittlement of nickel. The addition of phosphorus to nickel reduced the tendency for intergranular fracture and improved ductility because phosphorus segregated strongly to grain interfaces and limited sulfur enrichment. The hydrogen embrittling potency of antimony was also less than that of sulfur while its segregation propensity was considerably less. It was found that the effectiveness of segregated phosphorus and antimony in prompting intergranular embrittlementvs that of sulfur could be expressed in terms of an equivalent grain boundary sulfur coverage. The relative hydrogen embrittling potencies of sulfur, phosphorus, and antimony are discussed in reference to general mechanisms for the effect of impurity segregation on hydrogen-induced intergranular fracture.

Post-irradiation deformation characteristics of heavy-ion irradiated 304L SS

Abstract Post-irradiation deformation behavior in Ni-ion-irradiated 304L stainless steel (SS) is examined as a function of radiation dose and deformation temperature. For similar strain levels, specimens exhibit a transition from dislocation slip to deformation-induced twinning at 25°C with increasing radiation dose. At 288°C twinning is no longer observed and highly localized slip occurs by the formation of narrow “channels” containing a reduced defect density. The observations are discussed in terms of radiation-induced defect character and expected deformation mechanisms.

Superplastic behavior in a commercial 5083 aluminum alloy

When considering the forming and post-forming properties required of a superplastic material, attractive candidates are commercial Al-Mg-Mn weldable alloys such as AA5083. There have been several investigations of hot deformation of 5083-type alloys in the literature. Only two studies evaluated commercial-purity 5083 and they achieved tensile elongations of 150% and 200%. Alloy modification has produced improved behavior in three 5083-type alloys developed specifically for SPF. Two were deemed high-purity 5083 (low Fe and Si) and achieved elongations of 450% and 630%. Engineering strains up to 700% were measured by Watanabe et al. in a 5083-based alloy with the addition of 0.6% Cu as a grain refiner. These results suggest that alloy modifications such as reduced Fe and Si contents or Cu additions may be required to improve superplastic response. Unfortunately, specific SPF-grade 5083 alloys are substantially more expensive than the commercial grade, and the addition of Cu decreases the corrosion resistance of the base material. The purpose of this work is to examine the effect of prior degrees of cold work on the SPF behavior of a standard-grade 5083 alloy. Superplastic behavior of this material at 510[degree]C is assessed and compared to published results for the SPF-grade alloys.

Examinations of oxidation and sulfidation of grain boundaries in alloy 600 exposed to simulated pressurized water reactor primary water

High-resolution characterizations of intergranular attack in alloy 600 (Ni-17Cr-9Fe) exposed to 325°C simulated pressurized water reactor primary water have been conducted using a combination of scanning electron microscopy, NanoSIMS, analytical transmission electron microscopy, and atom probe tomography. The intergranular attack exhibited a two-stage microstructure that consisted of continuous corrosion/oxidation to a depth of ~200 nm from the surface followed by discrete Cr-rich sulfides to a further depth of ~500 nm. The continuous oxidation region contained primarily nanocrystalline MO-structure oxide particles and ended at Ni-rich, Cr-depleted grain boundaries with spaced CrS precipitates. Three-dimensional characterization of the sulfidized region using site-specific atom probe tomography revealed extraordinary grain boundary composition changes, including total depletion of Cr across a several nm wide dealloyed zone as a result of grain boundary migration.