Document: NUREG-0800
Document ID: 1085af08-a4b7-4fe7-bfed-ca7a60f2022c
Document Type: srp
Title: LOSS-OF-COOLANT ACCIDENTS RESULTING FROM SPECTRUM OF
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070734.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.6.5
CFR Part: 
CFR Title: 

Content:
le axial shape as a function of break size. If the evaluation model did not discuss axial shapes, or the discussion is not applicable to a given case, sensitivity studies are requested. d. The initial stored energy was conservatively calculated by the applicant. The value used is checked against the applicant's steady-state temperatures, as given in SAR Section 4.4, similar calculations performed by the staff, or calculations done for similar plants by previous applicants. e. Appropriate analyses are presented to support any credit taken for control rod insertion. 15.6.5-9 DRAFT Rev. 3 - April 1996 f. The applicant's analysis conservatively addresses the operation of the reactor coolant pump including requirements for reactor coolant pump trip during small break loss-of-coolant accidents as required by Generic Letters 85-12, 86-05, and 86-06.76 5. Reactor protection system actions and safety injection actuation and delivery are consistent with the set points and the associated uncertainties and delay times listed in the SAR (OL, COL, or standard design certification review ). The ECCS flow rates should 77 be checked against the applicant's data on head-flow characteristics of the ECCS pumps given in Section 6.3 of the SAR and against typical safety injection tank discharge curves used for the analysis. The Regional Offices under the Office of Inspection and Enforcement may be requested to provide data of this type from the startup tests for 78 new designs and from periodic tests on duplicate designs. 6. The results of the applicant's calculations are consistent with those of staff calculations for typical plants and also with the results of calculations performed for similar systems by previous applicants. The following variables should be reviewed on a generic basis and spot-checked thereafter: power transients for various breaks; pressure transients at various system locations; flow transients near the break, in the core, and in the downcomer; reactor coolant