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Title: Development of a Standard for Verification and Validation of Software Used to Calculate Nuclear System Thermal Fluids Behavior

Abstract

With the resurgence of nuclear power and increased interest in advanced nuclear reactors as an option to supply abundant energy without the associated greenhouse gas emissions of the more conventional fossil fuel energy sources, there is a need to establish internationally recognized standards for the verification and validation (V&V) of software used to calculate the thermal-hydraulic behavior of advanced reactor designs for both normal operation and hypothetical accident conditions. To address this need, ASME (American Society of Mechanical Engineers) Standards and Certification has established the V&V 30 Committee, under the responsibility of the V&V Standards Committee, to develop a consensus Standard for verification and validation of software used for design and analysis of advanced reactor systems. The initial focus of this committee will be on the V&V of system analysis and computational fluid dynamics (CFD) software for nuclear applications. To limit the scope of the effort, the committee will further limit its focus to software to be used in the licensing of High-Temperature Gas-Cooled Reactors. In this framework, the standard should conform to Nuclear Regulatory Commission (NRC) practices, procedures and methods for licensing of nuclear power plants as embodied in the United States (U.S.) Code of Federal Regulations and othermore » pertinent documents such as Regulatory Guide 1.203, “Transient and Accident Analysis Methods” and NUREG-0800, “NRC Standard Review Plan”. In addition, the standard should be consistent with applicable sections of ASME Standard NQA-1 (“Quality Assurance Requirements for Nuclear Facility Applications (QA)”). This paper describes the general requirements for the V&V Standard, which includes; (a) the definition of the operational and accident domain of a nuclear system that must be considered if the system is to licensed, (b) the corresponding calculational domain of the software that should encompass the nuclear operational and accident domain to be used to study the system behavior for licensing purposes, (c) the definition of the scaled experimental data set required to provide the basis for validating the software, (d) the ensemble of experimental data sets required to populate the validation matrix for the software in question, and (e) the practices and procedures to be used when applying a validation standard. Although this initial effort will focus on software for licensing of High-Temperature Gas Reactors, it is anticipated that the practices and procedures developed for this standard can eventually be extended to other nuclear and non-nuclear applications.« less

Authors:
; ;
Publication Date:
Research Org.:
Idaho National Laboratory (INL)
Sponsoring Org.:
DOE - NE
OSTI Identifier:
983356
Report Number(s):
INL/CON-09-17411
TRN: US1004456
DOE Contract Number:  
DE-AC07-05ID14517
Resource Type:
Conference
Resource Relation:
Conference: 18th International Conference on Nuclear Engineering (ICONE 18),Xi'an, China,05/17/2010,05/21/2010
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; ACCIDENTS; COMPUTERIZED SIMULATION; DESIGN; ENERGY SOURCES; ENGINEERS; FLUID MECHANICS; FOSSIL FUELS; GREENHOUSE GASES; HYPOTHETICAL ACCIDENTS; LICENSING; NUCLEAR ENGINEERING; NUCLEAR POWER; NUCLEAR POWER PLANTS; REACTORS; REGULATIONS; REGULATORY GUIDES; VALIDATION; VERIFICATION; High-Temperature Gas Reactor; Software Verification and Validation; Standard

Citation Formats

Richard R. Schultz, Edwin A. Harvego, and Ryan L. Crane. Development of a Standard for Verification and Validation of Software Used to Calculate Nuclear System Thermal Fluids Behavior. United States: N. p., 2010. Web.
Richard R. Schultz, Edwin A. Harvego, & Ryan L. Crane. Development of a Standard for Verification and Validation of Software Used to Calculate Nuclear System Thermal Fluids Behavior. United States.
Richard R. Schultz, Edwin A. Harvego, and Ryan L. Crane. Sat . "Development of a Standard for Verification and Validation of Software Used to Calculate Nuclear System Thermal Fluids Behavior". United States. https://www.osti.gov/servlets/purl/983356.
@article{osti_983356,
title = {Development of a Standard for Verification and Validation of Software Used to Calculate Nuclear System Thermal Fluids Behavior},
author = {Richard R. Schultz and Edwin A. Harvego and Ryan L. Crane},
abstractNote = {With the resurgence of nuclear power and increased interest in advanced nuclear reactors as an option to supply abundant energy without the associated greenhouse gas emissions of the more conventional fossil fuel energy sources, there is a need to establish internationally recognized standards for the verification and validation (V&V) of software used to calculate the thermal-hydraulic behavior of advanced reactor designs for both normal operation and hypothetical accident conditions. To address this need, ASME (American Society of Mechanical Engineers) Standards and Certification has established the V&V 30 Committee, under the responsibility of the V&V Standards Committee, to develop a consensus Standard for verification and validation of software used for design and analysis of advanced reactor systems. The initial focus of this committee will be on the V&V of system analysis and computational fluid dynamics (CFD) software for nuclear applications. To limit the scope of the effort, the committee will further limit its focus to software to be used in the licensing of High-Temperature Gas-Cooled Reactors. In this framework, the standard should conform to Nuclear Regulatory Commission (NRC) practices, procedures and methods for licensing of nuclear power plants as embodied in the United States (U.S.) Code of Federal Regulations and other pertinent documents such as Regulatory Guide 1.203, “Transient and Accident Analysis Methods” and NUREG-0800, “NRC Standard Review Plan”. In addition, the standard should be consistent with applicable sections of ASME Standard NQA-1 (“Quality Assurance Requirements for Nuclear Facility Applications (QA)”). This paper describes the general requirements for the V&V Standard, which includes; (a) the definition of the operational and accident domain of a nuclear system that must be considered if the system is to licensed, (b) the corresponding calculational domain of the software that should encompass the nuclear operational and accident domain to be used to study the system behavior for licensing purposes, (c) the definition of the scaled experimental data set required to provide the basis for validating the software, (d) the ensemble of experimental data sets required to populate the validation matrix for the software in question, and (e) the practices and procedures to be used when applying a validation standard. Although this initial effort will focus on software for licensing of High-Temperature Gas Reactors, it is anticipated that the practices and procedures developed for this standard can eventually be extended to other nuclear and non-nuclear applications.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {2010},
month = {5}
}

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