Abstract
The use of non-austenitic materials in cask containment boundaries requires consideration of the potential for brittle fracture under severe loading conditions. In the USA, such guidance for service conditions which containment boundaries must withstand is provided in Title 10 of the Code of Federal Regulations, Part 71, ‘Packaging and Transportation of Radioactive Material’, paragraph 71.73, ‘Hypothetical accident conditions’. The hypothetical accident conditions include a ‘free drop’ of the package at a temperature of −29°C, ‘…onto a flat, essentially unyielding, horizontal surface…’ from a height of 9 m. Such an event could potentially result in brittle fracture of a non-austenitic material containment boundary. Nevertheless, motivation exists for utilising ferritic materials or titanium alloys for containment applications. US Nuclear Regulatory Commission Regulatory Guide 7.6, ‘Design Criteria for the Structural Analysis of Shipping Cask Containment Vessels’, specifically excludes consideration of brittle fracture in its design criteria. Regulatory Guides 7.11 and 7.12, ‘Fracture Toughness Criteria of Base Material for Ferritic Steel Shipping Cask Containment With a Maximum Wall Thickness of 4 Inches (0.1 m)’, and ‘Fracture Toughness Criteria of Base Material for Ferritic Steel Shipping Cask Containment With a Wall Thickness Greater Than 4 Inches (0.1 m) But Not Exceeding 12 Inches (0.3 m)’, respectively, provide highly conservative criteria for selection of ferritic steel for containment based upon empirical correlations and materials tests suitable only for ferritic steels. Brittle fracture prevention criteria for cask containment based upon fracture mechanics principles remain non-codified in the USA. Methodologies for brittle fracture prevention routinely used in industry have yet to be implemented for cask design due to regulatory reticence. The American Society of Mechanical Engineers (AS ME) has provided models for brittle fracture prevention in Section III (‘Rules for Construction of Nuclear Power Plant Components’) Appendix G (‘Protection Against Nonductile Failure’), and Section XI (‘Rules for Inservice Inspection of Nuclear Power Plant Components’) Appendix A (‘Analysis of Flaws’), of the Boiler and Pressure Vessel Code. The ASME Subgroup on Containment Systems for Spent Fuel and High Level Waste Transport Packagings (NUPACK) has specifically addressed the issue of developing brittle fracture criteria for packages based upon fracture mechanics methodologies in the existing ASME Code but, to date, without resolution. This paper provides a detailed discussion of existing brittle fracture criteria in the USA and a status of new standards development.
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