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Development of a Standard for Verification and Validation of Software Used to Calculate Nuclear System Thermal Fluids Behavior. PUBLIC ACCESS

[+] Author Notes
Richard Schultz, Edwin Harvego

1Idaho National Laboratory

Ryan Crane

2ASME

Mechanical Engineering 132(05), 56-57 (May 01, 2010) (2 pages) doi:10.1115/1.2010-May-6

Abstract

This article focuses on the need for development of a standard for verification and validation (V&V) of software used to calculate nuclear system thermal fluids behavior. The V&V 30 Committee has been established to develop an ASME standard for verification and validation of computational fluid dynamics and system analysis software that will be used in the design and analysis of advanced nuclear reactor systems, with an initial focus on high-temperature gas-cooled reactors. The processes and procedures that will be addressed in the new standard will be used in the design and analysis of advanced reactor systems to be licensed in the United States. Recently, the V&V20 standard was released: Standard for Verification and Validation (V&V) in Computational Fluid Dynamics and Heat Transfer. Because of similarities in the standards being developed by the V&V20 and V&V30 Committees, it is important to define the relationship between the work embodied in the V&V20 Standard versus the work that will be forthcoming in the V&V30 Standard, as noted in the V&V20 Standard.

Article

To address the need for internationally recognized standards for verification and validation (V&V) of software used in the thermal-hydraulic analyses of advanced nuclear power plants, the V&V 30 Committee has been established to develop an ASME standard for verification and validation of computational fluid dynamics and system analysis software that will be used in the design and analysis of advanced nuclear reactor systems, with an initial focus on High-Temperature Gas-Cooled Reactors. The committee reports to the Verification and Validations Standards Committee, which falls under the Board on Standardization and Certification as depicted in the organizational structure shown in Figure 1. The title of the committee is “Verification and Validation in Computational Nuclear System Thermal Fluids Behavior.” As defined in its charter, the committee … “Provides the practices and procedures for verification and validation of software used to calculate nuclear system thermal fluids behavior. The software includes system analysis and computational fluid dynamics, including the coupling of this software.”

Figure 1 Relationship of Verification and Validation Standards Committee and Subcommittees to Board on Standards and Testing.

Grahic Jump LocationFigure 1 Relationship of Verification and Validation Standards Committee and Subcommittees to Board on Standards and Testing.

The processes and procedures that will be addressed in the new standard will be used in the design and analysis of advanced reactor systems to be licensed in the U.S. As such, the standard should conform to Nuclear Regulatory Commission (NRC) practices, procedures and methods for the licensing of nuclear power plants as embodied in the Code of Federal Regulations and other pertinent documents (such as Regulatory Guide 1.203, “Transient and Accident Analysis Methods”1 and NUREG-0800, “NRC Standard Review Plan”2). In addition, the standard should be consistent with applicable sections of ASME Standard NQA-13 (“Quality Assurance Requirements for Nuclear Facility Applications (QA)”).

Recently the V&V20 standard was released: Standard for Verification and Validation (V&V) in Computational Fluid Dynamics and Heat Transfer.4 Because of similarities in the standards being developed by the V&V20 and V&V30 Committees, it is important to define the relationship between the work embodied in the V&V20 Standard versus the work that will be forthcoming in the V&V30 Standard. As noted in the V&V20 Standard: “The scope of this Standard is the quantification of the degree of accuracy of simulation of specified validation variables at a specified validation point for cases in which the conditions of the actual experiment are simulated. Consideration of solution accuracy at points within a domain other than the validation points, i.e., a domain of validation, is a matter of engineering judgment specific to each family of problems and is beyond the scope of this Standard.” This statement clearly limits the applicability of the V&V20 standard to the domain defined by the validation points.

In contrast, the aim of the V&V30 Standard is to expand the domain of validation to encompass points beyond the range defined by the V&V 20 Standard. In other words, the V&V 30 Standard complements the V&V 20 Standard by defining a methodology for experimental validation of an expanded calculation envelope that encompasses the operational and accident domain of the nuclear system. Therefore V&V30 is expected to address: (a) applicable NRC requirements for defining the operational and accident domain of a nuclear system that must be considered if the system is to be licensed, (b) the corresponding calculation 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, such as the V&V20 Standard, to demonstrate that the validated software is capable of performing the needed licensing calculations. Each of the above areas to be addressed by the V&V30 Standard is discussed in an in-depth paper given in the Proceedings of the International Conference on Nuclear Engineering-18.

The verification and validation requirements for software intended for the design and analysis of advanced nuclear power systems are determined by the operational and accident envelopes of the reactor plant being considered. Specifically, as depicted in Figure 2, the V&V requirements can only be satisfied if the calculation envelope of the thermal-hydraulic software is demonstrated to either match or encompass the system operation and accident envelopes.

Figure 2 Venn Diagram of System and Calculation Envelope.

Grahic Jump LocationFigure 2 Venn Diagram of System and Calculation Envelope.

This formation of the new ASME V&V30 Committee, its objectives, and the approach that will be taken are described in greater detail in an ICONE-18 paper with the same title and authors as this short article. The initial focus of the standard will be on advanced High-Temperature Gas Reactors, but it is anticipated that this standard or additional standards will be developed in the future to include other reactor concepts as well as potential non-nuclear applications.

The organizational structure established by ASME for developing this and related V&V standards is described, along with the processes that will be employed to ensure consistency among the related standards and the nuclear regulatory environment.

Copyright © 2010 by ASME
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