Document: NRC Regulatory Guide
Document ID: 5f799693-27fd-4e13-a5e1-4c02f393d90a
Document Type: regulatory_guide
Title: Best-Estimate Calculations of Emergency Core Cooling System Performance + HISTORY –HISTORY 04/2013 – Periodic Review of Revision 0 – Reviewed with issues identified for future consideration 03/1987 – Draft RS 701-4, Proposed Revision 0
Source: NRC Regulatory Guide Division 1
Source URL: https://www.nrc.gov/docs/ML0037/ML003739584.pdf
Revision Date: 2023-06
Chapter: 
Section ID: RG-1.157
CFR Part: 
CFR Title: 

Content:
uld be performed for other important parameters for the transient of interest to evaluate compensating errors. For small-break loss-of-coolant accidents, the clad ding temperature response is the most important parameter; however, the ability of the codes to pre dict overall system mass and reactor vessel inventory distribution should also be statistically examined. In evaluating the code uncertainty, it will be nec essary to evaluate the code's predictive ability over several time intervals, since different processes and phenomena occur at different intervals. For example, in large-break loss-of-coolant accident evaluations, separate code uncertainties may be required for the peak cladding temperature during the blowdown and post-blowdown phases. Justification for treating these uncertainties individually or methods for combining them should be provided. The experimental information used to determine code uncertainty will usually be obtained from facili ties that are much smaller than nuclear power reac tors. Applicability of these results should be justified for larger scales. The effects of scale can be assessed through comparisons to available large-scale separate effects tests and through comparison to integral tests from various sized facilities. If there are scaling prob lems, particularly if predictions are nonconservative, the code should be improved for large-scale plants (i.e., nuclear reactors). Codes not having scaling ca pability will not be acceptable if their predictions are nonconservative. 4.3 Other Sources of Uncertainty When a best-estimate methodology is used to predict reactor transients, sources of uncertainty other than the limitations in the individual models and numerical methods (i.e., code uncertainty) are introduced. The following contributors to the overall calculational uncertainty should also be considered in the uncertainty analysis. 4.3.1 Initial and Boundary Conditions and Equipment Availability When a plant input model is