Document: NUREG-0800
Document ID: 3b714304-f91e-4999-917a-7d27866735e3
Document Type: srp
Title: STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE OF CONTAINMENT (PWR)
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0523/ML052350117.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.1.5
CFR Part: 
CFR Title: 

Content:
e listed below: a. The reactor power level and number of operating loops assumed 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 (NSSS) design, and sensitivity studies will be required to determine the most conservative combination of power level and plant operating mode. These sensitivity studies may be presented in a generic report and referenced in the SAR. b. Assumptions as to the loss of offsite power and the time of loss should be made to study their effects on the consequences of the accident. A loss of offsite power may occur simultaneously with the pipe break, or during the accident, or offsite power may not be lost. Analyses should be made to determineithe most conservative assumption appropriate to the particular plant design. The analyses should take account of the effect that loss of offsite power has on reactor coolant pump and main feedwater pump trips and on the initiation of auxiliary feedwater flow, and the effects on the sequence of events for these accidents. c. The effects (pipe whip, Jet impingement, reaction forces, temperature, humidity, etc.) of postulated steam line breaks on other systems should be considered in a manner consistent with the intent of Branch Technical Positions ASB 3-1 and MEB 3-1 (Ref. 1). d. The worst single active component failure should be assumed to occur. The assumed single failure may cause more than one steam generator to blow down, or may be in any of the systems required to control the transient. e. The maximum-worth rod should be assumed to be held in the fully withdrawn position. An appropriate rod reactivity worth versus rod position curve should be used. f. The core burnup (time in core life) should be selected to yield the most limiting combination of moderator temperature coefficient, void coeffi- cient, Doppler coefficient, axial power