Document: NUREG-0800
Document ID: 2717fe7f-71fd-4f1e-bd08-7685b24763ba
Document Type: srp
Title: Standard Review Plans are prepared for the guidance of the Office of Nuclear Reactor Regulation staff responsible for th
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0301/ML030160606.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15
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experimental data without recourse to any physical modeling. In most applications, especially those with a large number of processes and parameters, it is difficult, if not impossible, to design test facilities that preserve total similitude between the experiment and the nuclear power plant. Therefore, optimum similarity criteria are identified and scaling rationales developed for selecting existing data or designing and operating experimental facilities. The reviewers should confirm that the similarity criteria and scaling rationales are based on the important phenomena and processes identified by the accident scenario identification process and appropriate scaling analyses. The reviewers should confirm that scaling analyses were conducted to ensure that the data and the models will be applicable to the full scale analysis of the plant transient. Scaling compromises that are identified must be addressed in the bias and uncertainty evaluation. The experimental data base must be demonstrated to be sufficiently diverse so that the expected plant specific response is bounded and that the evaluation model calculations are comparable to the corresponding tests in non-dimensional space. This demonstration allows extending the conclusions relating to code capabilities, drawn from assessments comparing calculated and measured test data to the prediction of plant specific transient behavior. c. Accident Scenario Identification Process The accident scenario identification process is required in order to determine the needed modeling and assessment requirements for the code. The accident scenario identification process is also needed to identify and rank the reactor component and physical phenomena modeling requirements based on their importance to acceptable modeling of the scenario and their impact on the figures of merit for the calculation. This process is highly dependent on the type of reactor and the accident scenario of interest. Often a single computer code is used