Document: NUREG-0800
Document ID: ed948426-dd77-4047-be5e-2b7ab22de3f5
Document Type: srp
Title: FEEDWATER SYSTEM PIPE BREAK INSIDE AND OUTSIDE CONTAINMENT (PWR)
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0705/ML070550009.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.2.8
CFR Part: 
CFR Title: 

Content:
e means), rod drop (boiling-water reactor (BWR)), steam line rupture, reactor temperature and pressure changes, and cold water addition. GDCs 27 and 28 apply because this SRP section is for the review of feedwater system pipe breaks inside and outside containment that can result in transient conditions affecting reactor coolant temperature and pressure with consequent changes in core reactivity. The SAR analyses of these transients must demonstrate that reactivity, pressure, and temperature changes will not be severe enough for an unacceptable impact on the reactor coolant pressure boundary or on core cooling capability. The analyses must be reviewed by the staff independently in accordance with this SRP section. 5. GDC 31 requires reactor pressure boundary design with sufficient margin to ensure that, when stressed under operation, maintenance, test, and postulated accident conditions, the boundary is nonbrittle and the probability of rapidly propagating fracture is minimal. The design must reflect consideration of service temperatures and other conditions of the boundary material under operation, maintenance, test, and postulated accident conditions and the uncertainties in determining material properties; effects of irradiation on material properties; residual, steady state, and transient stresses; and flaw sizes. GDC 31 applies because this SRP section is for the review of feedwater system pipe breaks inside and outside containment that could result in transient reactor coolant temperature and pressure conditions that could affect the reactor coolant pressure boundary adversely. A feedwater system pipe break could result in either an RCS cool-down by excessive energy discharge through the break or an RCS heat-up by reduced feedwater flow to the affected steam generator. Heat-up of the reactor coolant by reduced feedwater flow to the affected steam generator and by the subsequent 15.2.8-9 Revision 2 - March 2007 addition of decay heat could result in undue stress