ASME NTB-5-2022 pdf download
ASME NTB-5-2022 pdf download.Guidance for Determination of Risk-Informed Safety Classification for Light Water Reactor Nuclear Facility Pressure Retaining Components.
3 RISK-INFORMED SAFETY CLASSIFICATIONS (a) The RISC process is described in Appendix I of this document. Pressure-retaining items should be classified high-safety significant (HSS) or low-safety significant (LSS), except as noted in 3(b) below. (b) (1) Class 1 portions of the reactor coolant pressure boundary (RCPB) that do not meet (i) or (ii) below: (i) in the event of postulated failure of the component during normal reactor operation, the reactor can be shut down and cooled down in an orderly manner, assuming makeup is provided by the reactor coolant makeup system. (ii) the component is or can be isolated from the reactor coolant system by two valves in series (both closed, both open, or one closed and the other open). Each open valve must be capable of automatic actuation and, assuming the other valve is open, its closure time must be such that, in the event of postulated failure of the component during normal reactor operation, each valve remains operable and the reactor can be shut down and cooled down in an orderly manner, assuming makeup is provided by the reactor coolant makeup system only. (2) ASME Boiler and Pressure Vessel Code, Section III, Class 1 items, that portion of the Class 2 feedwater system [> NPS 4 (DN 100)] of Pressurized Water Reactors (PWRs) from the steam generator to the outer containment isolation valve, and (3) items that are within a break exclusion region 3 [> NPS 4 (DN 100)] for high-energy piping systems and their associated supports (NB, NC and NF) should be classified high-safety significant (HSS).
5 APPLICABLE DISCIPLINES Personnel with expertise in the following disciplines should be included in addressing I-3.2.2, I-3.2.3 and I-4.2 of the RISC classification process. (a) probabilistic risk assessment (PRA) (b) plant operations (c) system design (d) safety or accident analysis Other disciplines may be added, such as materials engineering, chemistry, or nondestructive examination, relevant to the specific system or equipment issues. Personnel may be experts in more than one discipline, but are not required to be experts in all disciplines. For new system or equipment designs, where there is limited service experience, personnel with expertise from similar plant designs (e.g., earlier or same versions or models) should be used. To qualify as an expert, personnel should be experienced in the applicable discipline and related nuclear power plant requirements, and in the application of the requirements of the Code relating to the applicable discipline. Personnel selected for their expertise should have a minimum of four years of varied nuclear application experience, including 2 years in the applicable discipline for which they are serving as an expert. This experience should indicate that the expert has sufficient knowledge of anticipated plant and system operating and test conditions and their relationship to Code design criteria pertinent to the applicable Code item. In addition, the expert should be knowledgeable of the specific Code requirements pertaining to his specialty field. Guidelines reflecting the appropriate degree of Code knowledge are contained in ASME Section III, Division 1, Appendix XXIII “Guide B – Nonmandatory Guidelines for Establishing Code Knowledge.”
6 PRA SCOPE AND TECHNICAL ADEQUACY The PRA should be of sufficient scope and level of detail to support the RISC process, including verification of assumptions on equipment reliability from equipment not within the scope of this document. The PRA should be subjected to a review process where it is assessed against a standard 4 or set of acceptance criteria that is accepted by the regulatory agency having jurisdiction over the plant site. All deficiencies identified that impact the RISC process should be reconciled during the analysis to support the RISC process. The resolution of all PRA issues that impact the RISC process should be documented. EPRI report 1021467-A (Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk- informed In-Service Inspection Programs, Palo Alto, CA: 2011. 1021467) provides one example of addressing PRA completeness.
