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At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

RELAP-7 Software Verification and Validation Plan - Requirements Traceability Matrix (RTM) Part 2: Code Assessment Strategy, Procedure, and RTM Update

This document addresses two subjects involved with the RELAP-7 Software Verification and Validation Plan (SVVP): (i) the principles and plan to assure the independence of RELAP-7 assessment through the code development process, and (ii) the work performed to establish the RELAP-7 assessment plan, i.e., the assessment strategy, literature review, and identification of RELAP-7 requirements. Then, the Requirements Traceability Matrices (RTMs) proposed in previous document (INL-EXT-15-36684) are updated. These RTMs provide an efficient way to evaluate the RELAP-7 development status as well as the maturity of RELAP-7 assessment through the development process.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2025

Summary Report Of The FY25 Reactor Physics Verification And Validation Exercises In The Advanced Reactor Technologies - Gas-cooled Reactor Program

Valdiation and verification of numerical tools is critical for ensuring reasonable predictions for design scoping, licensing, and safety analsyis. In this report, two reactor physics verification and validation exercises are presented. The first of these exercises focuses on burnup analysis with data from the Advanced Gas Reactor (AGR) program. Simulations are performed with Monte Carlo N-Particle (MCNP) and are compared with the experimental measurements for the AGR 1 and 2 experiments that utilize both UCO and UO2 fuel. The second exercises utilizes data from the HTR-Proteus experiments to perform reactor physics validation. Specifications of the experimental facility are provdied, along with a demonstration of initial modeling efforts in Serpent for one of the determistic packing experiments. Both cases are part of the Generation-IV international forum (GIF) Very High-Temperature Reactor (VHTR) Computational Methods, Validation, and Benchmarking (CMVB) program, an international collaborative organization dedicated to the verification and validation of High-Temperature Gas-Cooled Reactor (HTGR) analysis. Participation in the CMVB allows the US Department of Energy (DOE) to leverage these existing validation activities to provide extra value through benchmarking activities with other CMVB members.

and Benchmarking (CMVB) program↗
DOE OSTI2018

Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors

This report documents the verification and validation (V&V) of the PLTEMP/ANL Version 4.3 code. A brief description of the code and its history are given in Section 1.1. Section 1.2 provides an overview of PLTEMP/ANL verification and validation work done to date. Section 1.3 lists some publications by other authors on PLTEMP/ANL verification and validation. Section 1.4 describes a UNIX script developed to routinely verify the code after a code modification. The V&V work documented in this report was performed over several years, as code development progressed. In each verification/validation testing discussion, the version of PLTEMP/ANL which was initially used for the V&V work is referenced, but all of the testing presented in this report is applicable to current release version 4.3 of the code.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2015

Survey of Verification and Validation Techniques for Small Satellite Software Development

The purpose of this paper is to provide an overview of the current trends and practices in small-satellite software verification and validation. This document is not intended to promote a specific software assurance method. Rather, it seeks to present an unbiased survey of software assurance methods used to verify and validate small satellite software and to make mention of the benefits and value of each approach. These methods include simulation and testing, verification and validation with model-based design, formal methods, and fault-tolerant software design with run-time monitoring. Although the literature reveals that simulation and testing has by far the longest legacy, model-based design methods are proving to be useful for software verification and validation. Some work in formal methods, though not widely used for any satellites, may offer new ways to improve small satellite software verification and validation. These methods need to be further advanced to deal with the state explosion problem and to make them more usable by small-satellite software engineers to be regularly applied to software verification. Last, it is explained how run-time monitoring, combined with fault-tolerant software design methods, provides an important means to detect and correct software errors that escape the verification process or those errors that are produced after launch through the effects of ionizing radiation.

Small Satellite Technology↗
DOE OSTI2011

Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC): Subcontinuum-scale Verification and Validation Strategy

The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V&V) is required throughout the system to establish evidence-based metrics for the level of confidence in M&S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V&V challenge at the subcontinuum scale, an approach to incorporate V&V concepts into subcontinuum scale modeling and simulation (M&S), and a plan to incrementally incorporate effective V&V into subcontinuum scale M&S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

White paper on Verification and Validation for Cyber Emulation Models

All disciplines that use models to predict the behavior of real-world systems need to determine the accuracy of the models’ results. Techniques for verification, validation, and uncertainty quantification (VVUQ) focus on improving the credibility of computational models and assessing their predictive capability. VVUQ emphasizes rigorous evaluation of models and how they are applied to improve understanding of model limitations and quantify the accuracy of model predictions.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2015

Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors

The document compiles in a single volume several verification and validation works done for the PLTEMP/ANL code during the years of its development and improvement. Some works that are available in the open literature are simply referenced at the outset, and are not included in the document. PLTEMP has been used in conversion safety analysis reports of several US and foreign research reactors that have been licensed and converted. A list of such reactors is given. Each chapter of the document deals with the verification or validation of a specific model. The model verification is usually done by comparing the code with hand calculation, Microsoft spreadsheet calculation, or Mathematica calculation. The model validation is done by comparing the code with experimental data or a more validated code like the RELAP5 code.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Human factors design, verification, and validation for two types of control room upgrades at a nuclear power plant

This paper describes the NUREG-0711 based human factors engineering (HFE) phases and associated elements required to support design, verification and validation (V&V), and implementation of a new plant process computer (PPC) and turbine control system (TCS) at a representative nuclear power plant. This paper reviews ways to take a human-system interface (HSI) specification and use it when migrating legacy PPC displays or designing displays with new functionality. These displays undergo iterative usability testing during the design phase and then undergo an integrated system validation (ISV) in a full scope control room training simulator. Following the successful demonstration of operator performance with the systems during the ISV, the new system is implemented at the plant, first in the training simulator and then in the main control room.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

RELAP-7 Software Verification and Validation Plan

This INL plan comprehensively describes the software for RELAP-7 and documents the software, interface, and software design requirements for the application. The plan also describes the testing-based software verification and validation (SV&V) process—a set of specially designed software models used to test RELAP-7. The RELAP-7 (Reactor Excursion and Leak Analysis Program) code is a nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework – MOOSE (Multi-Physics Object-Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s capability and extends the analysis capability for all reactor system simulation scenarios.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

Verification and Validation of Multisegmented Mooring Capabilities in FAST v8: Preprint

The quasi-static and dynamic mooring modules of the open-source aero-hydro-servo-elastic wind turbine simulation software, FAST v8, have previously been verified and validated, but only for mooring arrangements consisting of single lines connecting each fairlead and anchor. This paper extends the previous verification and validation efforts to focus on the multisegmented mooring capability of the FAST v8 modules: MAP++, MoorDyn, and the OrcaFlex interface. The OC3-Hywind spar buoy system tested by the DeepCwind consortium at the MARIN ocean basin, which includes a multisegmented bridle layout of the mooring system, was used for the verification and validation activities.

17 WIND ENERGY↗
DOE OSTI2016

Verification and Validation of Multisegmented Mooring Capabilities in FAST v8

The quasi-static and dynamic mooring modules of the open-source aero-hydro-servo-elastic wind turbine simulation software, FAST v8, have previously been verified and validated, but only for mooring arrangements consisting of single lines connecting each fairlead and anchor. This paper extends the previous verification and validation efforts to focus on the multisegmented mooring capability of the FAST v8 modules: MAP++, MoorDyn, and the OrcaFlex interface. The OC3-Hywind spar buoy system tested by the DeepCwind consortium at the MARIN ocean basin, which includes a multisegmented bridle layout of the mooring system, was used for the verification and validation activities.

17 WIND ENERGY↗
DOE OSTI2024

Verification and Validation Activities of Molten Salt Reactors Multiphysics Coupling Schemes at Idaho National Laboratory

This paper presents the latest verification and validation activities in molten salt reactor modeling and simulation performed at Idaho National Laboratory. Multiphysics solutions are obtained by coupling the neutronics code Griffin, the thermal hydraulics code Pronghorn, and the system analysis code SAM, under the MOOSE framework. We present various multiphysics coupling schemes with these codes for molten salt reactor problems and provide verification and validation results. First, we present verification test results of the Griffin-Pronghorn coupled scheme for the CNRS benchmark. Then validation test results are presented for the Griffin-SAM coupled scheme for the pump startup and coast down transients of the Molten Salt Reactor Experiment. Finally, the Griffin-Pronghorn-SAM coupled scheme is demonstrated for the Molten Salt Reactor Experiment reactivity insertion transient using a domain-overlapping coupling algorithm between Pronghorn and SAM. The results of these various coupling schemes demonstrate the ability to capture the effect of fuel flow and the various feedback mechanisms important to MSRs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

The Verification and Validation Strategy Within the Second Wind Forecast Improvement Project (WFIP 2)

The Wind Forecast Improvement Project 2 (WFIP2), in the complex terrain of the Columbia Gorge, focuses on a set of weather phenomena that poses particular challenges for wind and wind power forecasting. It further aims to understand and improve the skill of weather forecast models, particularly the National Oceanic and Atmospheric Administration's (NOAA's) High Resolution Rapid Refresh (HRRR), in complex terrain. The project also includes an extensive field campaign in the Columbia River Gorge, during which data from many different instruments were collected. WFIP2 is funded by the U.S. Department of Energy (DOE). Scientists from four U.S. national laboratories, NOAA, Vaisala, and universities contribute to the project with wide and varied interests and skill sets. Therefore, coordinated verification & validation (V&V) efforts across these members allows for the development of a clear picture of model improvements and scientific findings within WFIP2. The WFIP2 V&V team is tasked with providing tools, methods, and guidance to enable repeatable, metrics-based assessment of the Weather Research and Forecasting (WRF) model and associated modeling suites for analysis and forecasting of mesoscale weather phenomena that are important for wind energy in the Columbia River Gorge and other parts of the continental United States. This report summarizes what has been worked on and accomplished within the duration of the WFIP2 project, 4 years, by the V&V team.

17 WIND ENERGY↗
DOE OSTI2016

Contingency Contractor Optimization Phase 3 Sustainment Verification and Validation V&V) Report - Contingency Contractor Optimization Engineering - Prototype

The reports and test plans contained within this document serve as supporting materials to the activities listed within the “Contingency Contractor Optimization Tool – Prototype (CCOT-P) Verification & Validation Plan” [1]. The activities included test development, testing, peer reviews, and expert reviews. The engineering prototype reviews were done for both the software and the mathematical model used in CCOT-P. Section 2 includes the peer and expert review reports, which summarize the findings from each of the reviews and document the resolution of any issues. Section 3 details the test plans that were followed for functional testing of the application through the interface. Section 4 describes the unit tests that were run on the code.

97 MATHEMATICS AND COMPUTING↗
NASA NTRS2017

Verification and Validation Process for Progressive Damage and Failure Analysis Methods in the NASA Advanced Composites Consortium

The Advanced Composites Consortium is a US Government/Industry partnership supporting technologies to enable timeline and cost reduction in the development of certified composite aerospace structures. A key component of the consortium's approach is the development and validation of improved progressive damage and failure analysis methods for composite structures. These methods will enable increased use of simulations in design trade studies and detailed design development, and thereby enable more targeted physical test programs to validate designs. To accomplish this goal with confidence, a rigorous verification and validation process was developed. The process was used to evaluate analysis methods and associated implementation requirements to ensure calculation accuracy and to gage predictability for composite failure modes of interest. This paper introduces the verification and validation process developed by the consortium during the Phase I effort of the Advanced Composites Project. Specific structural failure modes of interest are first identified, and a subset of standard composite test articles are proposed to interrogate a progressive damage analysis method's ability to predict each failure mode of interest. Test articles are designed to capture the underlying composite material constitutive response as well as the interaction of failure modes representing typical failure patterns observed in aerospace structures.

Wanthal, Steven↗
DOE OSTI2018

Verification and Validation of the STAT7 Code

This document presents the verification and validation of the STAT7 Version 1.0 code which has been developed to perform the steady-state thermal hydraulic safety calculation of the MIT Research Reactor required upon conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel. The key capabilities of the code including the statistical processing, water property generation, and thermal hydraulic solution are verified independently. While the first two capabilities are checked using statistical tests and NIST data tables, each component of the thermal hydraulic solution is verified using simple geometry test problems. The comparison between results obtained by the code and hand calculations on these problems shows that each of these components work correctly. Additionally, a comprehensive comparison between STAT7 Version 1.0 and PLTEMP/ANL_4.2 calculations for a five-channel test problem is performed for the validation. The good agreement between the obtained results for all thermal hydraulic parameters belonging to the fuel core region is achieved. An issue in bypass flowrate calculation was detected when both fin-friction and bypass flow calculations are involved. This problem can be corrected in a later version of the code, however, does not affect the application of STAT7 code as it has been used to date where the flow distribution in the core region and bypass channel is specified as an input parameter.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Verification and Validation of the STAT7 Code

This document presents the verification and validation of the STAT7 Version 1.0 code which has been developed to perform the steady-state thermal hydraulic safety calculation of the MIT Research Reactor required upon conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel. The key capabilites of the code including the statistical processing, water property generation, and thermal hydraulic solution are verified independently. While the first two capabilites are checked using statistical tests and NIST data tables, each component of the thermal hydraulic solution is verified using simple geometry test problems. The comparison between results obtained by the code and hand calculations on these problems shows that each of these components work correctly. Additionally, a comprehensive comparison between STAT7 Version 1.0 and PLTEMP/ANL Version 4.2 calculations for a five-channel test problem is performed for the validation. The good agreement between the obtained results for all thermal hydraulic parameters belonging to the fuel core region is achieved. An issue in bypass flowrate calculation was detected when both fin-friction and bypass flow claculations are involved. This problem can be corrected in a later version of the code, however, does not affect the application of STAT7 code as it has been used to date where the flow distribution in the core region and bypass channel is specified as an input parameter.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗