Document: NRC Regulatory Guide
Document ID: e16da529-b6b4-4fdf-bc3f-7490180363f3
Document Type: regulatory_guide
Title: Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants (Rev. 2)
Source: NRC Regulatory Guide Division 1
Source URL: https://www.nrc.gov/docs/ML2018/ML20183A423.pdf
Revision Date: 2023-06
Chapter: 
Section ID: RG-1.89
CFR Part: 
CFR Title: 

Content:
ty period for the EQ equipment being evaluated. The survivability period is the maximum duration, post-accident, that the particular EQ component is expected to operate and perform its intended safety function. D-2.2 Fission Product Concentrations The radioactivity released from the core during a design-basis loss-of-coolant accident should be based on the assumptions provided in Regulatory Position 3 and Appendix A to Regulatory Guide (RG) 1.183, “Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,” or approved alternative assumptions. Although the design-basis LOCA is generally limiting for radiological EQ purposes, there may be certain components for which another design-basis accident may be limiting. In these cases, the assumptions in Appendices B through H to RG 1.183, as applicable, should be used, or approved alternative assumptions. EQ calculations may assume applicable features and mechanisms, provided that any prerequisites and limitations identified about their use are met. Additional considerations include the following: • For pressurized-water reactor ice condenser containments, the source should be assumed to be initially released to the lower containment compartment. The distribution of the activity should be based on the forced recirculation fan flow rates and the transfer rates through the icebeds as functions of time. • For boiling-water reactor Mark III designs, it should be assumed that all the activity initially is released to the drywell area and the transfer of activity from these regions by containment leakage to the surrounding reactor building volume should be used to predict the qualification levels within the reactor building (secondary containment). DG-1361, Appendix D, Page D-3 D-2.3 Dose Model for Containment Atmosphere The beta and gamma dose rates and integrated doses from the airborne activity within the containment atmosphere and from the plateout of aerosols on