Document: NUREG-0800
Document ID: 4b58df3f-1f93-4271-aa73-22de04083915
Document Type: srp
Title: LOSS OF NORMAL FEEDWATER FLOW
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0703/ML070300709.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.2.7
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Content:
by the organization responsible for instrumentation ans control systems confirms that their design is consistent with the requirements for safety systems actions for these events. To the extent deemed necessary, the reactor systems reviewer evaluates the effect of single active failures of systems and components which may alter the course of the transient. For new applications, LOOP should not be considered a single failure; loss of feedwater should be analyzed with and without a LOOP in combination with a single active failure. This part of the review uses the procedures described in the SRP sections for Chapters 4, 5, 6, 7, 8, and 9 of the SAR (or DCD). The mathematical models used by the applicant to evaluate core performance and to predict system pressure in the reactor coolant system and main steam line are reviewed by the organization responsible for reactor systems to determine if these models have been previously reviewed and found acceptable by the staff. If not, a generic review of the model proposed by the applicant is initiated. 15.2.7-9 Revision 2 - March 2007 The values of system parameters and initial core and system conditions used as input to the model are reviewed by the organization responsible for reactor systems. Of particular importance are the reactivity coefficients and control rod worths used in the applicant's analysis, and the variation of moderator temperature, void, and Doppler coefficients of reactivity with core life. The justification provided by the applicant to show that he has selected the core burnup that yields the minimum margins is evaluated. The results of the analysis are reviewed, including the effects of the LOOP and the possibility of the event developing into a more serious event (e.g., a stuck open PORV on the pressurizer that could lead to a SBLOCA if not isolated), and compared with the acceptance criteria presented in subsection II of this SRP section regarding maximum pressure in the reactor coolant and main steam