Document: NUREG-0800
Document ID: a7fafe76-5570-446b-96dd-3e53ec9bccb3
Document Type: srp
Title: NUREG-0800
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0523/ML052340663.pdf
Revision Date: 2023-06
Chapter: 4
Section ID: 4
CFR Part: 
CFR Title: 

Content:
on may be supplemented with examinations of related topical reports from reactor vendors, generic studies by staff consultants, and startup reports from operating reactors which contain information on measured power distributions. 3. The reviewer determines from the applicant's presentations that suitably conservative reactivity coefficients have been developed for use in reactor analyses such as those for control requirements, stability, and transients and accidents. The reviewer examines: a. The applicability and accuracy of methods used for calculations including the use of more accurate check calculations. b. The models involved in the calculations such as the model used for effective fuel temperature in Doppler coefficient analyses. c. The reactor state conditions assumed in determining values of the coefficients. For example, the pressurized water reactor (PWR) moderator temperature coefficient to be used in the steam line break analysis is usually based on the reactor condition at end of cycle with all control rods inserted except the most reactive rod, and the moderator temperature in the hot standby range. d. The applicability and accuracy of experimental data from critical experiments and operating reactors used to determine or justify uncertainty allowances. Measurements during startup and during the cycle of moderator temperature coefficients and full power Doppler coefficients in the case of PWRs, and results of measurements of transients during startup in the case of boiling water reactors (BWRs), should be examined.. As part of the review, comparisons are made between the values and uncertainty allowances for reactivity coeffi- cients for the reactor under review and those for similar reactors previously reviewed and approved. Generally, many essential areas will have been covered during earlier reviews of similar reactors. The reviewer notes any differences in results for essentially identical 4.3-8 Rev. 2 - July 1981 reactors and any lack of