Document: NUREG-0800
Document ID: a28650b3-fa03-428d-a2bd-4a27f583710b
Document Type: srp
Title: NUREG-0800
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070484.pdf
Revision Date: 2023-06
Chapter: 6
Section ID: 6
CFR Part: 
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Content:
as carbon dioxide, during a hydrogen releasing accident. If inadvertently actuated during normal operation, containment could potentially be pressurized by the inerting system. In accordance with this regulation, the containment must be designed to withstand this potential inadvertent pressurization to ensure that its integrity is maintained, thus precluding the release of radioactivity to the environment. III. REVIEW PROCEDURES The procedures described below are followed for the review of BWR pressure- suppression containments. The reviewer selects and emphasizes material from these procedures as may be appropriate for a particular case. Portions of the review may be carried out on a generic basis for aspects of functional design common to a class of BWR pressure-suppression type containments or by adopting the results of previous reviews of plants with essentially the same containment functional design. Upon request from the primary reviewer, otherthe secondary review branches will provide 61 input for the areas of review stated in subsection I of this SRP section. The primary reviewer obtains and uses such input as required to assure that this review procedure is complete. 1. The SCSB reviews the analyses of the drywell and wetwell temperature and pressure response for BWR pressure-suppressionMark I, II and III containments. The SCSB 62 63 performs confirmatory analyses, when necessary, using the CONTEMPT-LT computer code (References 22 and 23) . Input data for the code, including mass and energy 64 release data, are generally taken from the safety analysis report. 65 The SCSB normally analyzes only the design basis loss-of-coolant accident, which has 66 been found from previous reviews to be the recirculation line break for Mark I and II DRAFT Rev. 7 - April 1996 6.2.1.1.C-10 plants. For Mark III plants, the steam line break has been determined to be the design basis loss-of-coolant accident. However, mass and energy releases from the recirculation line break