Document: NUREG-0800
Document ID: a20ea81d-33c2-4a0c-a3b4-216d45d6fccd
Document Type: srp
Title: FEEDWATER SYSTEM PIPE.BREAKS INSIDE AND OUTSIDE CONTAINMENT (PWR)
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0523/ML052350148.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.2.8
CFR Part: 
CFR Title: 

Content:
REVIEW RESPONSIBILITIES Primary - Reactor Systems Branch (RSB) Secondary - Accident Evaluation Branch (AEB) I. AREAS OF REVIEW The transient that results from a postulated feedwater line break is sensitive to the break discharge rate; consequently, a range of break sizes should be evalu- ated both inside and outside containment to determine the acceptability of the response. Depending upon the size and location of the break and the plant operating conditions at the time of the break, the break could cause either a reactor coolant system cooldown (by excessive energy discharge through the break) or a reactor coolant system heatup (by reducing feedwater flow to the affected steam generator). Therefore, analyses of various postulated break sizes and locations are needed to identify the particular situation that is most limiting with respect to system effects. If a feedwater line rupture causes the water in the steam generator to be discharged through the break, the water will not be available for decay heat removal after reactor scram. The break location and size may be such to prevent addition of any feedwater to the affected steam generator. An auxiliary feed- water system is therefore provided to assure that feedwater is available to provide decay heat removal. The review includes evaluation of the applicant's postulated initial core and reactor conditions pertinent to the feedwater line break, the methods of thermal and hydraulic analysis, the postulated sequence of events including analyses to determine the time of reactor trip and time delays prior and subsequent to initi- ation of reactor protection system actions, the assumed response of the reactor coolant and auxiliary systems, the functional and operational characteristics of the reactor protection system in terms of its effects on the sequence of events, and all operator actions required to secure and maintain the reactor in a safe shutdown condition. The results of the analyses are reviewed to ensure that