INTRODUCTION
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 age. Degradation of materials by mechanisms such as intergranular stress corrosion cracking (IGSCC) and primary water stress corrosion cracking (PWSCC) create both technical and regulatory issues. This paper provides an overview of the most significant technical and regulatory issues associated with materials degradation from the perspective of the Nuclear Regulatory Commission. It also addresses the need for additional research to support regulation.
Light water reactor components must be designed to withstand significant environmental challenges. Typically, the chemical environment to which these components are exposed includes high purity water and dissolved gasses including hydrogen peroxide, oxygen, and hydrogen. Additionally, components are exposed to high pressures, radiation, and temperatures over 300oC. To meet these conditions reactor components are fabricated from a variety of alloys in a variety of product forms. Typical materials include carbon steel and low alloy steel, stainless steel, and nickel alloys. Typical product forms include wrought, cast, and welded materials that may be in the cold worked, thermally treated, or annealed state. Both applied stresses, such as hoop stresses imparted by internal pressures, and residual stresses, such as those present in welded structures and cold worked materials, are likely to be present in the fabricated components. Stress corrosion cracking in the form of primary water stress corrosion cracking (PWSCC) has occurred in pressurized water reactors (PWRs). Stress corrosion cracking in the form of irradiation assisted stress corrosion cracking (IASCC) has been observed in both BWRs and PWRs. This paper will consider the three types of cracking described briefly above from a regulatory perspective. Its primary focus will be to address the environments in which components must operate, including potential changes to those environments, the sufficiency of materials currently in use for those environments, and the adequacy of current repair techniques. The paper will also consider gaps in the current knowledge base and the research necessary to fill those gaps. In addition to the three cracking phenomena listed above, the paper will consider welding issues associated the with Alloy 52/152 family of weld metals. This paper is not meant to be an exhaustive treatise on the subjects of stress corrosion cracking and weldability of Alloys 52/152 in light water reactors but rather it is meant to convey the comments of the Nuclear Regulatory Commission (NRC) regarding the status of these issues.
Intergranular Stress Corrosion Cracking Intergranular stress corrosion cracking (IGSCC) is a form of stress corrosion cracking which is intergranular in nature and which occurs in components exposed to BWR coolant. BWR coolant is oxidizing relative to nickel, chromium and iron. IGSCC is known to occur in wrought austenitic stainless steels and nickel alloys which have been sensitized or cold worked. Cracking is most common in sensitized stainless steel. Heat affected zones of welds are, therefore, of significant interest.
