Design of Hazardous Mechanical Structures, Systems and Components for Extreme Loads
11 Quality Assurance and Control in Construction and Procurement of Safety-Related Structures, Systems and Components in Hazardous Facilities
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Hazardous facilities are defined as those facilities that process, store, handle or transport detonable, toxic or pathogenic materials and waste in a form and quantity such that a life-threatening or serious injury hazard potentially exists for collocated workers, the public, or significant damage to the environment. These facilities are divided into three hazard categories as described in Chapter 1. For safety-related structures, systems and components (SSC) in the more hazardous facilities (hazard category 1 and SSC performance categories 4 and 3) characteristic of high-hazard nuclear facilities, a formal written QA program plan should be required. Hazard category 2 and performance category 2 SSC characteristic of petrochemical and biomedical and moderate hazard nuclear facilities also should have a written, somewhat less formal, QA plan. For the lower and PC-1 and PC-0 SSC, a formal written QA plan is not usually required. This is typically referred to as the graded approach to QA. It should be understood the QA usually is the procedures and documentation that ensure that the QC examination and tests contained in applicable industry standards or owner specifications are properly implemented and evaluated.
Section 11.2 discusses Safety Classification and a graded and defense-in-depth approach to QA. Section 11.3 discusses quality in the constructed project, Section 11.4 provides observations, conclusions and recommendations, and Section 11.7 provides References. Appendix 11.A provides a detailed comparison of ISO-9001-94  with NQA-1-2000-10CFR50  Appendix B and 10CFR830.122  which define the basic nuclear QA requirements in the U.S. Appendix 11.B of this chapter provides details concerning the existing Nuclear Procurement Issues Committee (NUPIC) and the role this organization plays in auditing nuclear safety-related product, material and services supplied to nuclear power plants in the U.S. Appendix 11.C is a comparison of ISO-9001-94 requirements and ASME NQA-1 requirement prepared by ASME. Appendix 11.D contains a summary of IAEA QA requirements in comparison to British nuclear QA requirements. Appendix 11.E provides a discussion of QA excesses that have provided the basis for less emphasis on document QA and more emphasis on project and facility performance responsibility for QA. Finally, Appendix 11.F is the Forward to a typical QA program that identifies the different levels of QA that are part of a graded approach to QA. It also distinguishes between the format and content references for a QA program manual applicable to a professional services organization which typically follow the 10CFR832.122  or IAEA QA  and that which is meant to guide a vendor of nuclear safety related products and services  which generally follows the QA format of the USNRC.
By far the most formal application of QA has been its application in the nuclear industry.