§ 53.450 Analysis requirements.
(a) Requirement to have a probabilistic risk assessment (PRA), or other systematic risk evaluations (SREs), or a combination thereof. A PRA, other SREs, or a combination thereof for each commercial nuclear plant must be performed and used together with other generally accepted approaches for systematically evaluating engineered systems to identify potential failures, susceptibility to internal and external hazards, and other contributing factors to event sequences that might challenge the safety functions identified in § 53.230 and to support demonstrating that each commercial nuclear plant meets the safety criteria of § 53.220.
(b) Specific uses of analyses. The PRA, other SREs, or a combination thereof, together with other generally accepted approaches for systematically evaluating engineered systems must be used to—
(1) Inform the selection of the LBEs, as described in § 53.240, which must be considered in the design to determine compliance with the safety criteria in subpart B of this part.
(2) Inform the classification of SSCs according to their safety significance in accordance with § 53.460 and to identify the environmental conditions under which the SSCs and operating staff must perform their safety functions.
(3) Evaluate the adequacy of defense-in-depth measures required in accordance with § 53.250.
(4) Identify and assess all plant operating states where there is the potential for the uncontrolled release of radioactive material to the environment.
(5) Identify and assess events that challenge plant control and safety systems whose failure could lead to the uncontrolled release of radioactive material to the environment. These include internal events, such as human errors and equipment failures, and external events identified in accordance with subpart D of this part.
(6) Inform the establishment and updating of appropriate measures for plant operations, including availability controls, to ensure that the configurations and special treatments for SR SSCs and NSRSS SSCs provide the capabilities, availability, and reliability consistent with satisfying the safety criteria under §§ 53.220 and the analyses of licensing-basis events other than design-basis accidents (DBAs) under § 53.450(e).
(c) Maintenance and upgrade of analyses. The PRA, other SREs, or a combination thereof must be maintained (e.g., updated to reflect plant changes such as modifications, procedure changes, or plant performance data) at least every 5 years until the permanent cessation of operations under § 53.1070 and upgraded (e.g., changed in scope or use of new methods) in conformance with generally accepted methods, standards, and practices that have been endorsed or otherwise found acceptable by the NRC.
(d) Qualification of analytical codes. The analytical codes used in modeling the physical behavior of plant systems in the analyses of licensing-basis events (including but not limited to thermodynamics, reactor physics, fuel performance, and mechanistic source term codes) must be qualified for the range of conditions for which they are to be used.
(e) Analyses of licensing-basis events other than design-basis accidents. (1) Analyses must be performed for LBEs other than design-basis accidents (DBAs). These LBEs must be identified using insights from a PRA, other SREs, or a combination thereof with other generally accepted approaches for systematically evaluating engineered systems to identify and analyze equipment failures and human errors.
(2) The analysis of LBEs other than DBAs must include definitions of evaluation criteria for each event or specific categories of LBEs to determine the acceptability of the plant response to the challenges posed by internal and external hazards to provide an appropriate level of safety.
(3) The analyses of LBEs other than DBAs must address event sequences from initiation to a defined end state and be used in combination with other engineering analyses to demonstrate that the functional design criteria required by § 53.420 provide sufficient barriers to the unplanned release of radionuclides to satisfy the evaluation criteria defined for each LBE other than DBAs, to satisfy the safety criteria specified in accordance with § 53.220 and provide defense in depth as required by § 53.250.
(4) The methodology used to identify, categorize, and analyze LBEs must include a means to identify event sequences deemed significant for controlling the risks posed to public health and safety.
(f) Analysis of design-basis accidents. (1) The analysis of LBEs required by § 53.240 must include analysis of DBAs that address possible challenges to the safety functions identified under § 53.230. The events selected as DBAs must be those that, if not terminated, have the potential for exceeding the safety criteria in § 53.210.
(2) The DBAs selected must be analyzed using deterministic methods that address event sequences from initiation to a safe stable end state and assume only the SR SSCs identified under § 53.460 and human actions addressed by the requirements of subpart F of this part are available to perform the safety functions identified in accordance with § 53.230.
(3) The analysis must conservatively demonstrate compliance with the safety criteria in § 53.210.
(g) Other required analyses. Analyses must be performed to assess—
(1) Fire protection. Fire protection measures to demonstrate, through inclusion of fires in the analysis of LBEs or by separate analyses, that a fire or explosion in any plant area would not—
(i) Prevent equipment from fulfilling the safety functions identified in accordance with § 53.230; or
(ii) Challenge the safety criteria in §§ 53.210 and 53.220.
(2) [Reserved]
(3) Dose to members of the public. Measures taken under § 53.425, including estimating—
(i) The quantity of each of the principal radionuclides expected to be released annually to unrestricted areas in liquid effluents produced during normal reactor operations and the dose to the maximally exposed member of the public in unrestricted areas.
(ii) The quantities of each of the principal radionuclides of the gases, halides, and particulates expected to be released annually to unrestricted areas in gaseous effluents produced during normal reactor operations and the dose to the maximally exposed member of the public in unrestricted areas.
(iii) The annual external radiation dose in unrestricted areas and the maximally exposed member of the public in unrestricted areas due to direct radiation from contained radiation sources from the commercial nuclear plant during normal reactor operations.