Document: NUREG-0800
Document ID: c1d42ca1-cc58-40db-812c-918554bfa81b
Document Type: srp
Title: FEEDWATER SYSTEM PIPE BREAKS INSIDE AND OUTSIDE CONTAINMENT
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070704.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.2.8
CFR Part: 
CFR Title: 

Content:
bability events and below 120% of the design pressures for very 37 low probability events such as double-ended guillotine breaks. 15.2.8-7 DRAFT Rev. 2 - April 1996 2. The potential for core damage is evaluated on the basis that it is acceptable if the minimum DNBR remains above the 95/95 DNBR limit for PWRs based on acceptable correlations (see SRP Section 4.4). If the DNBR 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 must be such that the calculated doses at the site boundary are a small fraction of the 10 CFR Part 100 guidelines. 4. The integrity of the reactor coolant pumpsRCPs should be maintained, such that loss of ac power and containment isolation will not result in seal damage. 5. The auxiliary feedwater systemAFWS must be safety grade and automatically initiated when required. 6. Tripping of the reactor coolant pumpsRCPs should be consistent with the resolution to TMI Action Plan Item II.K.3.5. There are certain assumptions which should be used in the analysis regarding important parameters that describe initial plant conditions and postulated system failures. These are listed below. a. The power level assumed and number of loops operating at the initiation of the transient should correspond to the operating condition which maximizes the consequences of the accident. These assumed initial conditions will vary with the particular nuclear steam supply system and sensitivity studies will be required to determine the most conservative combination of power level and plant operating mode. These sensitivity studies may be