Document: NUREG-0800
Document ID: ddaa3c7d-ce79-4a3f-aaae-4e4436ab7bc1
Document Type: srp
Title: NUCLEAR DESIGN
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0707/ML070740003.pdf
Revision Date: 2023-06
Chapter: 4
Section ID: 4.3
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Content:
f this SRP section is satisfied. If the equipment involved is subject to frequent downtime, the reviewer must determine if alternative measures should be provided or the extent of proposed outage time is acceptable. C. The reviewer will employ the same procedures as in item 5.A, above, to evaluate the scram reactivity information described in Subsection I.5 of this SRP section. The scram reactivity is a property of the reactor design and is not easily changed, but if restrictions are necessary the procedures in item 5.B, above, can be followed as applicable. D. The reviewer confirms the appropriateness of control rod reactivity worths used in the reactivity accident analyses reviewed under SRP Sections 15.4.8 and 15.4.9. Regulatory Guide 1.77 provides guidance for calculating maximum rod worths to be used in evaluating control rod ejection accidents for PWRs. RG 1.206 should also be consulted. Relevant experience and information from BWR situations and scenarios must be considered where applicable. 6. The information presented on criticality of fuel assemblies is reviewed in the context of the applicant’s physics calculations and the ability to calculate criticality of a small number of fuel assemblies. 7. The reviewer exercises professional judgment and experience to ascertain the following about the applicant’s analytical methods: A. The computer codes used in the nuclear design are described in sufficient detail to enable the reviewer to establish that the theoretical bases, assumptions, and numerical approximations for a given code reflect the current state of the art. B. The source of the neutron cross-sections used in fast and thermal spectrum calculations is described in sufficient detail so that the reviewer can confirm that the cross-sections are comparable to those in the current ENDF/B data files (i.e., ENDF/B-VII) and other sources of nuclear data, such as JENDL and JEFF3, etc. If modifications and normalization of the cross-section data have been