Document: NUREG-0800
Document ID: 16afcc26-09c8-4627-8906-8c56ffe8434b
Document Type: srp
Title: LOSS OF NORMAL FEEDWATER FLOW
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070683.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.2.7
CFR Part: 
CFR Title: 

Content:
h. The value of parameters used in the analytical model should be suitably conservative. The following values are considered acceptable for use in the model. a. The initial power level is taken as the licensed core thermal power for the number of loops initially assumed to be operating plus an allowance of 2% to account for power measurement uncertainties, unless a lower power level can be justified by the applicant. The number of loops operating at the initiation of the event should correspond to the operating condition which maximizes the consequences of the event. b. Conservative scram characteristics are assumed, i.e., for a PWR – maximum time delay with the most reactive rod held out of the core and for a BWR – a design conservatism factor of 0.8 times the calculated negative reactivity insertion rate. c. The core burnup is selected to yield the most limiting combination of moderator temperature coefficient, void coefficient, Doppler coefficient, power profile and radial power distribution. d. Mitigating systems should be assumed to be actuated in the analyses at setpoints with allowance for instrument inaccuracy in accordance with Regulatory Guide 1.105. Compliance with Regulatory Guide 1.105 is determined by ICSB.27 Technical Rationale28 The technical rationale for application of the above acceptance criteria to reviewing analyses of transients initiated by steam system piping failures is discussed in the following paragraphs:29 1. Compliance with GDC 10 requires that the reactor core and associated coolant, control, and protection systems be designed with appropriate margin to ensure that specified 15.2.7-7 DRAFT Rev. 2 - April 1996 acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. GDC 10 is applicable to SRP Section 15.2.7 because this section evaluates the loss of normal feedwater flow transient. A part of the evaluation relates to the reactor coolant