iSafe Presentation - Software Verification, Validation and Integration.
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Tran, Jenn↗
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Results for “software verification and validation”
Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
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Software Verification and Validation Guidelines for Non-Linear Soil-Structure Interaction Analysis
Seismic analysis of structures, systems, and components (SSCs), including the consideration of soil-structure interaction (SSI) effects, is an important and required step in the design and licensing of nuclear power plant SSCs important to safety. Historically, the SSI analysis of nuclear structures has been performed using equivalent linear methods. However, there has been considerable industry investment in alternative seismic design and analysis approaches to reduce the construction cost of new reactors. Toward that goal and in alignment with the Licensing Modernization Project (LMP) framework, the Nuclear Regulatory Commission (US NRC) has proposed a risk-informed, performance-based approach to seismic design that allows inelastic response in those nuclear plant structures not required for confinement. As a complementary effort, reactor designers are exploring nonlinear seismic analysis methods to optimize structural designs and reduce construction costs.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Package Testing Program Process for Validation and Verification of Software
The purpose of this document is to define the process for validation and verification (V&V) of software used to test the thermal data acquisition equipment in the Oak Ridge National Laboratory (ORNL) Package Testing Program (PTP) as part of the testing configuration(s). This process adheres to all applicable ORNL Standards Based Management System requirements and correlates with the PTP Quality Management System and the Software Quality Assurance Plan.
97 MATHEMATICS AND COMPUTING↗
Software Verification Toolkit (SVT): Survey on Available Software Verification Tools and Future Direction
Writing software is difficult. However, writing complex, well tested and designed, and functionally correct software is incredibly difficult. An entire field of study is devoted to the validation and verification of software to address this problem, and in this paper we analyze the landscape of currently available third party software. We have divided our analyses into three separate subsections with regards to software validation: formal methods, static analysis, and test generation. Formal verification is the most complex method in which to validate software correctness, but also the most thorough as it truly validates the mathematical validity of the source code. Static analysis generally is relegated to abstract syntax tree traversal techniques to find errors related to faulty software such as memory leaks or stack overflow issues. Automatic test generation is similar in implementation to static analysis, but pushes a bit further in verifying the boundedness of function inputs and outputs with regards to annotated or parsed criteria. The crux of this report is to analyze and describe the software tools that implement these techniques to validate and verify software. Pros and cons related to installation, utilization, and capabilities of the frameworks are described, and reproducible examples are provided with a focus on usability. The initial survey concluded that the most interesting tools of note are Z3, Isabelle/HOL, and TLA+ with regards to formal verification; and Infer, Frama-C, and SonarQube with regards to static analysis. With these tools in mind, a final conjecture is provided that describes future avenues of utilizing these tools for developing a verification framework to assist in validating existing software at Sandia National Laboratories.
97 MATHEMATICS AND COMPUTING↗
Verification and Validation Tests of the Gamma Library of the ARC Software Package (Rev.1)
As part of the verification and validation (V&V) efforts of the Argonne Reactor Code (ARC) software system to support the Versatile Test Reactor (VTR) project of the U.S. Department of Energy (DOE), work has been performed to verify and validate the gamma (or photon) library of the multigroup cross section generation code MC 2 -3 of the ARC system, which is the essential part of the coupled neutron and gamma heating calculations to determine the heat generation rate or power distribution in the core. This V&V study was carried out in three steps: 1) review of the procedures and utility programs for generating the MC 2 -3 gamma library, 2) verification tests of cross sections and KERMA factors by comparing the total heat generation rate of coupled neutron and gamma calculation with that obtained with reaction Q values under the assumption of local gamma energy deposition and by comparing the prompt heat generation rate with that calculated with the MCNP6.2 code, and 3) validation tests of cross sections and KERMA factors by analyzing the ZPPR-15D gamma dose data measured with thermoluminescent dosimeters (TLD).
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Validation and Verification of Python based Neutron Spectrum Unfolding Software
To validate and verify the python-based code (PySL), designed to replicate the programs used by STAYSL for Beam Correction Factor (BCF) and Self-Shielding Factor (SHIELD), a series of tests were performed. To test BCF a python script was written to generate a random flux history file and both versions of the code processed the data. The test verified matching values up to at least one decimal place, approximately 10,000 tests where run and each one passed. Isotopes began to fail the tests once neutron saturation was reached. To verify this the total time of exposure was varied the isotopes that failed were compared to a list of their half-lives. The test process for SHIELD was very similar but, in this case, the code began by producing an input file with varying thickness and device type/environment for the SHIELD input. The failure condition for this test was if any of the data points for an isotope had a difference above 3%. Approximately 40 of these tests were run and there were only 3 isotopes that had reoccurring failures but only 2% of their points were above the 3% difference. A visual comparison was conducted by plotting the results from both programs. Although the test failed, the differences between their values were minuscule, and the self-shielding factor’s shape was preserved when plotted. Next steps for this project will be validating and verifying the python-based SigPhi code and then reproducing and testing the least squares unfolding performed by STAYSL.
73 - NUCLEAR PHYSICS AND RADIATION PHYSICS↗
SAM: A Modern System Code for Advanced Non-LWR Safety Analysis
The System Analysis Module (SAM), developed at Argonne National Laboratory and by collaborators at other organizations, is for advanced non–light water reactor safety analysis. SAM aims to provide fast-running, modest-fidelity, whole-plant transient analysis capabilities that are essential for fast-turnaround design scoping and engineering analyses of advanced reactor concepts. To facilitate code development, SAM utilizes the MOOSE object-oriented application framework, its underlying finite element library, and linear and nonlinear solvers to leverage modern advanced software environments and numerical methods. SAM aims to solve tightly coupled physical phenomena, including fission reaction, heat transfer, fluid dynamics, and thermal-mechanical responses in advanced reactor structures, systems, and components with high accuracy and efficiency. Finally, this paper gives an overview of the SAM code development, including goals and functional requirements, physical models, current capabilities, verification and validation, software quality assurance, and examples of simulations for advanced nuclear reactor applications.
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From Sim to Real: A Pipeline for Training and Deploying Traffic Smoothing Cruise Controllers
Designing and validating controllers for connected and automated vehicles to enhance traffic flow presents significant challenges, from the complexity of replicating real-world stop-and-go traffic dynamics in simulation, to the intricacies involved in transitioning from simulation to actual deployment. In this work, we present a full pipeline from data collection to controller deployment. Specifically, we collect 772 km of driving data from the I-24 in Tennessee, and use it to build a one-lane simulator, placing simulated vehicles behind real-world trajectories. Using policy-gradient methods with an asymmetric critic, we improve fuel efficiency by over 10% when simulating congested scenarios. Our comprehensive approach includes reinforcement learning for controller training, software verification, hardware validation and setup, and navigating various sim-to-real challenges. Furthermore, we analyze the controller's behavior and wave-smoothing properties, and deploy it on four Toyota Rav4’s in a real-world validation experiment on the I-24. Lastly, we release the driving dataset, the simulator and the trained controller, to enable future benchmarking and controller design.
42 ENGINEERING↗
An open-access simulated earthquake ground-motion database for an M7 Hayward Fault earthquake in the San Francisco Bay Region
Comprehensive understanding of earthquake ground motions, particularly in the near-fault region of large-magnitude events, is limited by gaps in strong-motion data. This challenge is prominent in areas with high seismic hazard but infrequent large earthquakes where data is sparse and difficult to interpret. These data limitations lead to uncertainties in the development of site-specific ground motions, which are crucial for engineering risk assessments. To address these challenges, physics-based regional-scale ground-motion simulations have been developed. With the emergence of exaflop-scale computing ecosystems, it is now possible to simulate regional earthquake processes at unprecedented fidelity and generate the large number of fault rupture realizations necessary to characterize both intra- and inter-event ground-motion variability. This article introduces a new database of simulated earthquake ground motions, created for applications in earthquake engineering, earthquake planning, and emergency response. The inaugural version of the database features simulated ground motions for a magnitude 7 Hayward Fault earthquake in the San Francisco Bay Region (SFBR), using the EarthQuake SIMulation (EQSIM) simulation framework and the Graves–Pitarka kinematic rupture model. The aim is to provide high-fidelity, spatially dense, three-component motions generated on the Department of Energy’s (DOE) newest generation of graphics processing unit (GPU)-accelerated supercomputers. These motions are being made openly available to the engineering, scientific, and disaster planning communities. In addition, this work develops protocols for the efficient dissemination of these large data sets and emphasizes community engagement to build confidence in their application. This article discusses the methodology behind the data, underlying software verification and validation, scalable data management, and a user interface for data access. The goal is to facilitate widespread use and elicit expert feedback to maximize the utility and exploitation of simulated motions. While the initial focus is on the San Francisco Region, simulations for additional regions will be added as the DOE program progresses.
Simulated ground-motion database↗
Computational Analysis of Coupled Geoscience Processes in Fractured and Deformable Media
Prediction of flow, transport, and deformation in fractured and porous media is critical to improving our scientific understanding of coupled thermal-hydrological-mechanical processes related to subsurface energy storage and recovery, nonproliferation, and nuclear waste storage. Especially, earth rock response to changes in pressure and stress has remained a critically challenging task. In this work, we advance computational capabilities for coupled processes in fractured and porous media using Sandia Sierra Multiphysics software through verification and validation problems such as poro-elasticity, elasto-plasticity and thermo-poroelasticity. We apply Sierra software for geologic carbon storage, fluid injection/extraction, and enhanced geothermal systems. We also significantly improve machine learning approaches through latent space and self-supervised learning. Additionally, we develop new experimental technique for evaluating dynamics of compacted soils at an intermediate scale. Overall, this project will enable us to systematically measure and control the earth system response to changes in stress and pressure due to subsurface energy activities.
58 GEOSCIENCES↗
SAS4A/SASSYS-1 VERIFICATION AND VALIDATION TEST SUITE
SAS4A/SASSYS-1 is a safety analysis software package already under copyright by Argonne National Laboratory. There is interest from third-party users of the software to gain access to our internal verification and validation test suite. Therefore we are seeking to establish copyright over the test suite, which includes input files and reference results for hundreds of test cases that demonstrate functionality of the software. Making the test suite available to licensed users allows them to utilize the broad range of tests for commercial grade dedication activities needed to qualify the SAS4A/SASSYS-1 software under their own Software Quality Assurance program for use in a regulatory environment (e.g. with the Nuclear Regulatory Commission).
FANNING, THOMAS↗
CI/CD Efforts for Validation, Verification and Benchmarking OpenMP Implementations
Software developers must adapt to keep up with the changing capabilities of platforms so that they can utilize the power of High-Performance Computers (HPC), including exascale systems. OpenMP, a directive-based parallel programming model, allows developers to include directives to existing C, C++, or Fortran code to allow node level parallelism without compromising performance. This paper describes our CI/CD efforts to provide easy evaluation of the support of OpenMP across different compilers using existing testsuites and benchmark suites on HPC platforms. Our main contributions include (1) the set of a Continuous Integration (CI) and Continuous Development (CD) workflow that captures bugs and provides faster feedback to compiler developers, (2) an evaluation of OpenMP (offloading) implementations supported by AMD, HPE, GNU, LLVM, and Intel, and (3) evaluation of the quality of compilers across different heterogeneous HPC platforms. With the comprehensive testing through the CI/CD workflow, we aim to provide a comprehensive understanding of the current state of OpenMP (offloading) support in different compilers and heterogeneous platforms consisting of CPUs and GPUs from NVIDIA, AMD, and Intel.
Jarmusch, Aaron↗
PARET/ANL (V.7.7) Verification and Validation Report
This report documents the software testing which has been performed for the PARET/ANL version 7.7 software. The software testing is based on code capabilities identified by research reactor analysts as frequently used in their safety analyses. The verification and validation procedures have been performed and documented to address the steady-state capabilities of the software, as described in Chapter 2, and the transient capabilities, as described in Chapter 3. Testing based on the comparison between PARET/ANL calculations and analytical solutions, hand calculations, or other code calculations of the test cases confirms that all the identified capabilities of the software were implemented correctly. In addition, results from code comparisons against SPERT-I and SPERT-IV experiments for various flow rates are reported for the peak power, energy release and cladding surface temperature. The comparisons showed overall good agreement for the peak power and conservative predictions of the cladding surface temperature
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
PARET/ANL v7.7 Verification and Validation Report
This report documents the software testing which has been performed for the PARET/ANL version 7.7 software. The software testing is based on code capabilities identified by research reactor analysts as frequently used in their safety analyses. The verification and validation procedures have been performed and documented to address the steady-state capabilities of the software, as described in Chapter 2, and the transient capabilities, as described in Chapter 3. Testing based on the comparison between PARET/ANL calculations and analytical solutions, hand calculations, or other code calculations of the test cases confirms that all the identified capabilities of the software were implemented correctly. In addition, results from code comparisons against SPERT-I and SPERT-IV experiments for various flow rates are reported for the peak power, energy release and cladding surface temperature. The comparisons showed overall good agreement for the peak power and conservative predictions of the cladding surface temperature.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Qualification of SAS4A/SASSYS-1 for Sodium-Cooled Fast Reactor Authorization and Licensing (Final CRADA Report)
This project focuses on assessment and qualification of the SAS4A/SASSYS-1 advanced reactor safety analysis code and its capacity to fulfill requirements associated with license/authorization of a sodium-cooled fast reactor (SFR). Specific objectives of this proposal include: development of a formal documentation requirements matrix that complies with a vendor’s NQA-1 program to identify needs for qualification documentation; completion of software methods qualification gap analysis to identify and close gaps in the verification and validation basis of the software; generation of an Acceptance Test Report to confirm compliance with the software requirements specification; and closure of gaps in Software Design Description (SDD) to assist with the eventual generation of a Modeling Licensing Topic Report (LTR).
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Transition of SRT Source Term Software to Modern Version Control System to Enable More Efficient Verification, Validation, and User Engagement
The Simplified Radionuclide Transport (SRT) code assists in reactor design and licensing by providing a detailed analysis of radionuclide transport under specific transient scenarios. This analysis is essential for ensuring that reactors meet safety and regulatory standards before they are authorized to operate. Mechanistic source term analysis, which involves a realistic evaluation of radionuclide behavior from the source to the environment, is a key component of this process. It provides a comprehensive understanding of potential release scenarios, supporting the development of robust safety measures.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
STAT7 v1.2 Verification and Validation Report
This report documents the software testing which has been performed for the STAT7 Version 1.2 (hereafter STAT7 v1.2) software, a code for the statistical propagation of uncertainties in steady-state thermal hydraulics (TH) analysis of plate-fueled research and test reactors. The key capabilities addressed in this testing were selected based on the use of the software for analysis work supporting the conversion to low-enriched uranium fuel of U.S. High-Performance Research Reactors (USHPRR) such as the Massachusetts Institute of Technology Research Reactor (MITR-II). This testing scope includes comparison of the code’s statistical processing capabilities using statistical tests, comparison of coolant property evaluations to NIST data tables, and systematic comparison of key TH solver capabilities to the results of hand calculations and to predictions from PLTEMP/ANL Version 4.4. The good agreement achieved in this testing campaign, within the acceptance criteria of a 1% relative difference, also demonstrates that these capabilities have been implemented in STAT7 v1.2 correctly.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
STAT7 v2.0 Verification and Validation Report
This report documents the software testing which has been performed for STAT7 Version 2.0 (hereafter STAT7 v2.0), a code for the statistical propagation of uncertainties in steady-state thermal hydraulics (TH) analysis of plate-fueled research and test reactors. The key capabilities addressed in this testing were selected based on the use of the software for analysis work supporting the conversion to low-enriched uranium fuel of U.S. High-Performance Research Reactors (USHPRR) such as the Massachusetts Institute of Technology Research Reactor (MITR-II). This testing scope includes comparison of the code’s statistical processing capabilities using statistical tests, comparison of coolant property evaluations to NIST data tables, and systematic comparison of key TH solver capabilities to the results of hand calculations and to predictions from PLTEMP/ANL Version 4.4. The
22 GENERAL STUDIES OF NUCLEAR REACTORS↗