Document: NUREG-0800
Document ID: 3ea2f0ac-4d7e-464a-b1c4-390c3970f642
Document Type: srp
Title: provides specific thermal-hydraulic criteria.  The available radioactive fission product
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0707/ML070740002.pdf
Revision Date: 2023-06
Chapter: 4
Section ID: 4.4
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CFR Title: 

Content:
le model for predicting the effects of fuel densification in commercial LWRs. Application of acceptance criteria established in 10 CFR 50.46 significantly reduces the possibility of a violent chemical reaction between the Zircaloy cladding and the coolant, which would result, if it were to occur, 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 of the core. 2. 10 CFR Part 100 requires the calculation of the exposure to an individual caused by the release of fission products to the environment during a postulated reactor accident and consideration of the result when determining the acceptability of a reactor site. 10 CFR Part 100 and RG 1.195 and RG 1.196 apply to reactors with DC applications before January 10, 1997, unless the reactor has adopted the AST, as defined in 10 CFR 50.67 and RG 1.183. RG 1.195 and RG 1.196 can be used in place of RG 1.3, RG 1.4, RG 1.5, RG 1.25, and RG 1.77. RG 1.183 and the requirements of 10 CFR 50.34 apply to new reactors with DC applications after January 10, 1997; the source terms for both new reactors and the AST are based on total effective dose equivalent rather than whole body dose as used in 10 CFR Part 100 and RG 1.195 and RG 1.196. This section discusses acceptable fission gas release models to perform radiological dose calculations; these models ensure that doses are not underestimated. RG 1.3, RG 1.4, RG 1.183, and RG 1.195 provide acceptable assumptions that may be used to evaluate the radiological consequences associated with a LOCA for BWRs and PWRs. RG 1.25, RG 1.183, and RG 1.196 provide acceptable assumptions that may be used to evaluate the radiological consequences associated with a fuel-handling accident at a fuel handling and storage facility at reactor sites. RG 1.77, RG 1.183, and RG 1.195 identify acceptable analytical methods and assumptions that may be used to evaluate the