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:
(PWR) REVIEW RESPONSIBILITIES Primary - Reactor Systems Branch (RSB)(SRXB)1 Secondary - Accident Evaluation Branch (AEB)Emergency Preparedness and Radiation Protection Branch (PERB)2 Instrumentation and Controls Branch (HICB)3 I. AREAS OF REVIEW The transientanticipated operational occurrence (AOO) that results from a postulated feedwater 4 line break is sensitive to the break discharge rate; consequently, a range of break sizes should be evaluated 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 (RCS) cooldown (by 5 excessive energy discharge through the break or a reactor coolant systemRCS 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 feedwater system (AFWS) is therefore provided to assureensure that 6 7 feedwater is available to provide decay heat removal. DRAFT Rev. 2 - April 1996 15.2.8-2 The review includes: 1. evaluation of the applicant's postulated initial core and reactor conditions pertinent to the feedwater line break, 2. 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 initiation of reactor protection system (RPS) actions, 8 3. the assumed response of the reactor coolant and auxiliary systems, the functional 9 and operational characteristics