Document: NUREG-0800
Document ID: b8e8a93f-5bc3-4f5b-9d72-fdf90a3c2e45
Document Type: srp
Title: RADIOLOGICAL CONSEQUENCES OF CONTROL ROD DROP ACCIDENT (BWR)
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0523/ML052350427.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.4.9
CFR Part: 
CFR Title: 

Content:
accident for the plant in question and compare these with the corresponding features and resulting doses for previously reviewed plants to ascertain whether a specific calculation of the radiological consequences should be performed. The reviewer should examine the applicant's description of the control rod drop accident, in particular,'the sequence of events 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 accept- ance 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 CPB.- Verification of the applicant's calculation of the number of fuel rod failures and the amount of fuel reaching the fuel melting temperature is obtained from the CPB. The following assumptions regarding the plant condition and release and transport of radioactivity are used in the' independent AEB 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. 4. No allowance is made for activity decay prior to accident initiation, regardless of the