ASME/ANS RA-S-1.2

Description / Abstract: The scope of a PRA covered by this standard is limited to analyzing the progression of severe accidents from the onset of core damage through radionuclide release to the environment or a determination that a release to the environment will not occur. It includes the analysis of the various phenomena that occur inside the reactor vessel, the containment structure, and neighboring structures that might participate in the radiological release pathway to the environment. This analysis involves carrying the postulated accident sequences through a probabilistic logic structure such as a containment event tree (CET) (or equivalent) and determining the radionuclide release characteristics (e.g., magnitude and timing) for the various pathways through the CET.

This scope includes postulated accident sequences initiated from all modes of reactor operation (at-power, shutdown, and transition states). It also includes accident sequences initiated by internal events and/or external hazards addressed in ASME/ANS RA-Sa-2009 [1].

The assessment of radiological releases is restricted to radionuclides that originate in fuel located within the reactor pressure vessel. It does not address spent fuel pool radionuclide release nor releases related to purposeful human-induced security threats (e.g., sabotage); this limited scope is consistent with that of ASME/ANS RA-Sa-2009 [1]. This standard is limited in scope to single reactor accidents and does not address accident sequences involving releases and interactions among multi-reactor units and fuel storage facilities such as the occurrence at Fukushima Daiichi during March, 2011.

The requirements described in this standard address commercial LWRs (currently operating nuclear plants and so-called evolutionary or advanced LWRs with sufficiently detailed design information to evaluate plant response to accident sequences involving substantial core damage). Revisions may be necessary so that it can be applied to next generation designs. This standard is applicable throughout the life cycle of a plant. Of course, this applicability must recognize that some supporting requirements (SRs) cannot be met during the early phases of design and operation when data procedures, training, etc. are not available for evaluation.

The applicability to other LWR designs would have to be evaluated on a case-by-case basis. Caution must be exercised when applying these requirements to reactor and containment designs that are substantially different from operating LWR designs or current evolutionary LWR designs.