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

摘要

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 aim of the V&V 30 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. 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.