Document: NUREG-0800
Document ID: 28c03fd2-3542-41ec-bfd7-5df2087c4ee6
Document Type: srp
Title: FUEL SYSTEM DESIGN
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070407.pdf
Revision Date: 2023-06
Chapter: 4
Section ID: 4.2
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Content:
Guide 1.157, or alternatively Appendix K to 10 CFR 50, to evaluate the performance of the ECCS. Regulatory Guide 1.126 provides an acceptable model for predicting the effects of fuel densification in commercial light water reactors. Application of acceptance criteria established in 10 CFR 50, §50.46 significantly reduces the possibility of a violent chemical reaction occurring between the Zircaloy cladding and the coolant, which would result in the production of explosive hydrogen gas following an accident. It also ensures that damage to the fuel system in the event of an accident is never so severe as to prevent cooling the core. 2. 10 CFR Part 100 requires that exposure to an individual caused by the release of fission products to the environment during a postulated reactor accident be calculated, and that the result be considered when determining the acceptability of a reactor site. Acceptable fission gas release models which are necessary for performing radiological dose DRAFT Rev. 3 - April 1996 4.2-16 calculations are discussed in this section and ensure that doses are not underestimated. Regulatory Guides 1.3 and 1.4 provide acceptable assumptions that may be used in evaluating the radiological consequences associated with a LOCA for BWRs and PWRs respectively. Regulatory Guide 1.25 provides acceptable assumptions that may be used in evaluating the radiological consequences associated with a fuel handling accident at a Fuel Handling and Storage Facility at reactor sites. And Regulatory Guide 1.77 identifies acceptable analytical methods and assumptions that may be used in evaluating the consequences of a rod ejection accident in PWRs. Evaluation of the radiological dose consequences associated with a postulated reactor accident, as prescribed in 10 CFR Part 100, provides assurance that nuclear reactors can be operated safely under worst case conditions. 3. GDC 10 requires the reactor core and associated coolant, control, and protection systems be designed