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:
rator actions required to secure and maintain the reactor in a safe shutdown condition. The sequence of events described in the applicant's safety analysis report (SAR) is reviewed for the performance of the RPS, the engineered safety systems, and operator actions to secure and maintain the reactor in a safe conditions. 4. The auxiliary feedwater system is reviewed for whether the flow is acceptable for transient control following a feedwater line break. 5. COL Action Items and Certification Requirements and Restrictions. For a DC application, the review will also address COL action items and requirements and restrictions (e.g., interface requirements and site parameters). For a COL application referencing a DC, a COL applicant must address COL action items (referred to as COL license information in certain DCs) included in the referenced DC. Additionally, a COL applicant must address requirements and restrictions (e.g., interface requirements and site parameters) included in the referenced DC. Review Interfaces Other SRP sections interface with this section as follows: 1. General information on transient and accident analyses is provided in SRP Section 15.0. 2. Design basis radiological consequence analyses associated with design basis accidents are reviewed under SRP Section 15.0.3. 3. Effects of blow-down loads, including jet propulsion piping and component supports and the design bases for safety and relief valves are reviewed under SRP Sections 3.6.2 and 3.9.1 through 3.9.3. Design bases for safety and relief valves is also reviewed under SRP Section 3.9.3. 15.2.8-4 Revision 2 - March 2007 4. Values of the parameters in the analytical models of the reactor core are reviewed for compliance with plant design and specified operating conditions, acceptance criteria for fuel cladding damage limits are determined, and the core physics, fuel design, and core thermal-hydraulics data in the SAR analysis are reviewed under SRP Sections 4.2, 4.3, and 4.4. 5. Fracture