Document: NUREG-0800
Document ID: 0b17303b-e5cc-4091-b22c-056b0c78eb34
Document Type: srp
Title: – 15.3.4
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070707.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.3.3
CFR Part: 
CFR Title: 

Content:
t relates to the calculated doses at the site boundary. DRAFT Rev. 3 - April 1996 15.3.3-4 The basic objectives of the review of the accident resulting from a rotor seizure or shaft break in a reactor coolant pump are: 1. To identify which of these accidents is the more limiting. 2. To verify that, for the accident, the plant responds in such a way that the criteria regarding fuel damage, radiological consequences, and system pressure are met. The specific criteria necessary to meet the relevant requirements of GDC General Design Criteria 27, 28, and 31 and 10 CFR Part 100 for the rotor seizure and shaft break event are: 26 1. Pressure in the reactor coolant and main steam systems should be maintained below acceptable design limits, considering potential brittle as well as ductile failures. 2. The potential for core damage is evaluated on the basis that it is acceptable if the minimum departure from nucleate boiling ratio (DNBR) remains above the 95/95 27 DNBR limit for PWRs and the critical power ratio (CPR) remains above the minimum 28 critical power ratio (MCPR) safety limit for BWRs based on acceptable correlations 29 (see SRP Section 4.4). If the DNBR or CPR falls below these values, fuel failure (rod perforation) must be assumed for all rods that do not meet these criteria unless it can be shown, based on an acceptable fuel damage model (see SRP Section 4.2), which includes the potential adverse effects of hydraulic instabilities, that fewer failures occur. Any fuel damage calculated to occur must be of sufficiently limited extent that the core will remain in place and intact with no loss of core cooling capability. 3. Any activity release of radioactive material must be such that the calculated doses at the 30 site boundary are a small fraction of the 10 CFR Part 100 guidelines. 4. The integrity of the reactor coolant pumps should be maintained such that loss of ac power and containment isolation will not result in pump seal damage. 5. The auxiliary feedwater