Document: NUREG-0800
Document ID: 621763b9-bd48-4ded-8c42-b942a23a06d4
Document Type: srp
Title: – 15.2.5
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070680.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.2.1
CFR Part: 
CFR Title: 

Content:
ate a more serious plant condition without other faults occurring independently. d. An incident of moderate frequency in combination with any single active component failure, or single operator error, shall be considered an event for which an estimate of the number of potential fuel failures shall be provided for radiological dose calculations. For such accidents, fuel failure must be assumed for all rods for which the DNBR or CPR falls below those values cited above for cladding integrity unless it can be shown, based on an acceptable fuel damage model (see SRP Section 4.2), that fewer failures occur. There shall be no loss of function of any fission product barrier other than the fuel cladding. 3. The applicant should analyze these transients events using an acceptable analytical model. The equations, sensitivity studies, and models described in References 4 through 7 are acceptable. (Refs. 8 and 9 describe acceptable transient analysis computer codes used for design analysis of the Advanced Boiling Water Reactor, or ABWR.) If other 30 analytical methods are proposed by the applicant, these methods are evaluated by the staff for acceptability. For new generic methods, the reviewer requests an evaluation by RSBSRXB. The values of the parameters used in the analytical model should be suitably conservative. The following values are considered acceptable for use in the model: a. The reactor is initially at 102% of the rated (licensed) core thermal power (to account for a 2% power measurement uncertainty), and primary loop of flow is at the nominal design flow, less the flow measurement uncertainty. 31 b. Conservative scram characteristics are assumed, i.e., maximum time delay with the most reactive rod held out of the core for PWRs and a 0.8 multiplier on the predicted reactivity insertion rate for BWRs. DRAFT Rev. 2 - April 1996 15.2.1-6 c. The core burnup is selected to yield the most limiting combination of moderator temperature coefficient, void coefficient,