Document: NUREG-0800
Document ID: a20ea81d-33c2-4a0c-a3b4-216d45d6fccd
Document Type: srp
Title: FEEDWATER SYSTEM PIPE.BREAKS INSIDE AND OUTSIDE CONTAINMENT (PWR)
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0523/ML052350148.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.2.8
CFR Part: 
CFR Title: 

Content:
he amount of secondary coolant expelled from the system has been calculated conservatively by evaluating the applicant's methods and assumptions, by comparison with an acceptable analysis performed on another plant of similar design, or by comparison with staff calculations for typical plants which will be-available from RSB on request. The reviewer confirms that a commitment has been made in the SAR to conduct preoperational tests to verify that valve discharge rates and response times including, for example, opening and closing times (delay times) for main feedwater, auxiliary feedwater, turbine and main steam isolation valves, and steam generator and pressurizer relief and safety valves, has been conserva- tively modeled in the accident analyses. In addition, preoperational testing should include verification of reactor trip delay times, startup delay times for actuation of the auxiliary feedwater system, safety injection signal delay time, and delay times for delivery of any high concentration boron injection required to bring the plant to a safe shutdown condition. Using the information developed in the review, the AEB reviewer evaluates the radiological consequences of the design basis feedwater line break. This evaluation based on a qualitative comparison with the results of the design. basis steam line break, or on a detailed analysis using the approach described in the appendix to SRP Section 15.1.5. The reliability and operability of the auxiliary feedwater systems (AFWS) are reviewed to assure conformance to the following TMI Action Plan Items (References 6 and 7) as they relate to auxiliary feedwater system performance requirements following feedwater piping failures. (a) Items II.E.1 and II.K.2.1 (b) Items II.E.1.2 and II.K.2.8 The influence of reactor coolant pump trip during ECCS initiation is reviewed to assure conformance to the TMI Action Item II.K.3.5 (References 6 and 7). Should tripping of the reactor coolant pumps require manual action, delays