Document: NUREG-0800
Document ID: 7ed8e5d3-fcfd-49cc-84ad-3b25caba06af
Document Type: srp
Title: STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE OF CONTAINMENT
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0705/ML070550006.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.1.5
CFR Part: 
CFR Title: 

Content:
(PWR) REVIEW RESPONSIBILITIES Primary - Organization responsible for the review of transient and accident analyses for PWRs Secondary - None I. AREAS OF REVIEW 1. The steam release resulting from a rupture of a main steam pipe will cause an increase in steam flow which decreases with time as the steam pressure decreases. The increased steam flow causes increased energy removal from the reactor coolant system and results in a reduction of coolant temperature and pressure. The negative moderator temperature coefficient and the cooldown of the reactor system causes an increase in core reactivity. The core reactivity increase may cause a loss of reactor core shutdown margin and a resulting increase in reactor power. If the plant is at power, the reactor is automatically tripped and the main steam and feedwater line isolation valves are automatically closed. Decay heat is removed as necessary through the unaffected steam generators by venting steam from the secondary system safety and relief valves. The auxiliary feedwater system (AFWS) supplies makeup water to the unaffected steam generator(s). For AP1000 the passive RHR (PRHR) provides the safety related means of decay heat removal. 15.1.5-2 Revision 3 - March 2007 Analysis of the transient following a steam line break is sensitive to the fluid discharge rate at the break so that a range of break sizes must be considered both inside and outside containment to determine the acceptability of the system response. The course that the transient takes and its ultimate effects also depend on the assumed initial power level and mode of operation (e.g., hot shutdown; full power; one-, two-, or three-loop operation). Evaluation with various assumed initial conditions is required to verify that the condition leading to the severest consequences has been identified. The specific areas of review are as follows: – Postulated initial core and reactor conditions pertinent to the steam line break accident; – Methods of thermal and