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:
REVIEW RESPONSIBILITIES Primary - Reactor Systems Branch (RSB) Secondary - Accident Evaluation Branch I. AREAS OF REVIEW 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. Due to the negative moderator temperature coefficient this cooldown causes an increase in core reactivity. The core reactivity increase may cause a power level increase and a decrease in shutdown margin. 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 supplies makeup water to the unaffected steam generator(s). 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 evaluated both inside and outside containment to determine the acceptability of the system response. Past experience generally shows that the worst break is that which results in the maximum cooldown rate. The course the transient takes and its ultimate effects also depend on the assumed initial power level and mode of operation (i.e., hot shutdown, full power, one-, two-, or three-loop operation). Analyses with various assumed initial conditions are required to verify that the condition leading to the severest consequences has been identified. The topics reviewed include: postulated initial core and reactor conditions pertinent to the steam line break accident, methods of thermal and hydraulic analyses including the effects of hydraulic instabilities, postulated sequence of events including analyses to determine the time