Document: NUREG-0800
Document ID: 3b714304-f91e-4999-917a-7d27866735e3
Document Type: srp
Title: STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE OF CONTAINMENT (PWR)
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0523/ML052350117.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.1.5
CFR Part: 
CFR Title: 

Content:
evaluation of the new analytical model. A secondary review is performed by the Accident Evaluation Branch and the results are used by RSB to complete the overall evaluation of the break analysis. The Accident Evaluation Branch (AEB) evaluates the fission product release and verifies that the radiological consequences resulting from a steam line break are within acceptable limits. This evaluation is performed for the design basis case as described in the appendix to this SRP section. The results of AEB's analysis is transmitted to RSB for use in the SER writeup. II. ACCterANCE CRITERIA The general objective of the review of steam line rupture events is to verify that short-term and long-term coolability has been achieved by confirming that the primary reactor coolant system is maintained in a safe status for a break equivalent in area to the double-ended rupture of the largest steam line. RSB acceptance criteria are based on meeting the relevant requirements of the following regulations: A. General Design Criteria 27 and 28, as they relate to the reactor coolant system being designed with appropriate margin to assure that acceptable fuel design limits are not exceeded, and that the capability to cool the core is maintained. B. General Design Criterion 31, as it relates to the reactor coolant system being designed with sufficient margin to assure that the boundary behaves in a nonbrittle manner and that the probability of propagating fracture is minimized. C. General Design Critierion 35, as it relates to the reactor cooling system and associated auxiliaries being designed to provide abundant emergency core cooling. In addition, task action plan items necessary to meet the requirements to maintain adequate decay heat removal and reactor coolant pump integrity and operation are Items II.E.1.2, lI.K.2.1, II.K.2.8, II.K.3.5, II.K.2.16, II.K.3.25, and II.K.3.40 of NUREGs 0694, 0718, and 0737. Specific criteria necessary to meet the relevant requirements of the above