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:
ent 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 of reactor trip and time delays prior to and subsequent to initiation of the reactor protection system, assumed Rev. 2 - July 1981 USNRC STANDARD REVIEW PLAN Standard review plans are prepared for the guidance of the Office of Nuclear Reactor Regulation staff responsible for the review of applications to construct and operate nuclear power plants. These documents are made available to the public as part of the Commission's policy to Inform the nuclear industry and the general public of regulatory procedures and policies. Standard review plans are not substitutes for regulatory guides or the Commission's regulations and compliance with them is not required. The standard review plan sections are keyed to the Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants. Not all sections of the Standard Format have a corresponding review plan. Published standard review plans will be revised periodically, as appropriate, to accommodate comments and to reflect new informa- tion and experience. Comments and suggestions for Improvement will be considered and should be sent to the U.S. Nuclear Regulatory Commisslon. Office of Nuclear Reactor Regulation. Washington. D.C. responses of the reactor coolant and auxiliary systems, functional and operational characteristics of the reactor protection system in terms of its effects on the sequence of events, operator actions required to secure and maintain the reactor in a safe shutdown condition, core power excursion due to power demand created by excessive steam flow out the break, and variables influencing neutronics. The results of the analyses are reviewed to ensure that pertinent system parameters are within expected ranges. The parameters of importance for these transients include reactor coolant system (RCS)