Document: NUREG-0800
Document ID: 7d56ab5c-90b1-42b4-a509-857fbfd674db
Document Type: srp
Title: review.10
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070724.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.4.9
CFR Part: 
CFR Title: 

Content:
vents following the accident to assure that the most severe case from the standpoint of release of fission products to the environment is analyzed. Unless unusual plant or site features are present or the applicant's calculation shows an unusually large amount of fuel damage, a specific calculation of the radiological consequences is not necessary. In this case a comparison of the pertinent plant and site features is sufficient to conclude that the consequences of this event meet the acceptance criteria given in subsection II. However, a specific evaluation of this accident should be performed for the first application involving a particular standardized design to establish a reference point for comparison of future applications incorporating the design. Where a specific calculation of the radiological consequences is to be performed, the core response aspects of the accident are reviewed by the CPBSRXB. Verification of the applicant's 20 calculation of the number of fuel rod failures and the amount of fuel reaching the fuel melting temperature is obtained from the CPBSRXB. The following assumptions regarding the plant 21 condition and release and transport of radioactivity are used in the independent AEBPERB22 calculations: 1. A coincident loss of offsite power is assumed at the time of the accident. 2. The integrity of the turbine and condensers is unaffected by the rod drop accident. 3. The combination of reactor operating mode, control rod positions, core burnup, etc., that results in the largest source term, is selected for evaluation. DRAFT Rev. 3 - April 1996 15.4.9-4 4. No allowance is made for activity decay prior to accident initiation, regardless of the reactor status for the selected case. 5. The amount of activity accumulated in the fuel-clad gap is assumed to be the same as that in Regulatory Guide 1.77 (Ref. 2).23 6. The nuclide inventory of the fraction of the fuel which reaches or exceeds the initiation temperature of fuel melting (typically 2842 C)