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:
eration 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 regulations are as follows: 1. Pressure in the reactor coolant and main steam systems should be-main- tained below acceptable design limits, considering potential brittle as well as ductile failures. 2. The potential for core damage is evaluated on the basis that it is acceptable if the minimum DNBR remains above the 95/95 DNBR limit for PWRs based on acceptable correlations (see SRP Section 4.4). If the DNBR falls below these values, fuel failure (rod perforation) must be assumed for all rods that do not meet these criteria unless it can be shown, based on an acceptable fuel damage model (see SRP Section 4.2), which includes the potential adverse effects of hydraulic instabilities, that fewer failures occur. Any fuel damage calculated to occur must be of sufficiently 15.1.5-3 Rev. 2 - July 1981 limited extent that the core will remain in place and intact with no loss of core cooling capability. 3. The radiological criteria used in the evaluation of steam system pipe break accidents (PWRs only) appear in the appendix to this SRP section. 4. The integrity of the reactor coolant pumps should be maintained, such that loss of a-c power and containment isolation will not result in pump seal damage. 5. The auxiliary feedwater system must be safety grade and, when required, automatically initiated. 6. Tripping of the reactor coolant pumps should be consistent with the resolution to Task Action Plan item II.K.3.5. There are certain assumptions regarding important parameters used to describe the initial plant conditions and postulated system failures which should be used. These are listed below: a. The reactor power level and number of operating loops assumed at the initiation of the transient should correspond to the operating condition which maximizes the consequences of the