Document: NUREG-0800
Document ID: c1d42ca1-cc58-40db-812c-918554bfa81b
Document Type: srp
Title: FEEDWATER SYSTEM PIPE BREAKS INSIDE AND OUTSIDE CONTAINMENT
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070704.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.2.8
CFR Part: 
CFR Title: 

Content:
r GDC 25. An appropriate rod reactivity worth versus rod position curve should be assumed. f. The core burnup (time in core life) should be selected to yield the most limiting combination of moderator temperature coefficient, void coefficient, Doppler coefficient, axial power profile, and radial power distribution. g. The initial core flow assumed for the analysis of the feedwater line rupture accident should be chosen conservatively. If the minimum core flow allowed by the technical specifications is assumed, the minimum DNBR margin results for the case of a feedwater line rupture inside containment. However, this may not be the most conservative assumption. For example, maximum initial core flow results in increased reactor coolant systemRCS cooldown and depressurization, decreased shutdown margin, and an increased possibility that the core will become critical and return to power. Since it is not clear what initial core flow is most conservative, the applicant's assumption should be justified by appropriate sensitivity studies. h. During the initial 10 minutes of the transient, should credit for t for operator action be required (i.e., RCP trip), an assessment for the limiting consequence must be performed in order to account for operator delay and/or error. Technical Rationale40 The technical rationale for application of these acceptance criteria to reviewing the feedwater system pipe breaks inside and outside containment (PWR) is discussed in the following paragraphs:41 1. The requirements of 10 CFR Part 100 specify how the exclusion area, low population zone, and population center distance should be determined. Further, radiation exposure criteria stipulated in 10 CFR Part 100 provide reference values to be used in the site suitability determination based on postulated fission product releases associated with accidental events. 10 CFR Part 100 is applicable to this section because it specifies the methodology for calculating radiation exposures at the site