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DOE OSTI2017

The Significance of Strategic Foresight in Verification Technologies: A Case Study of the INF Treaty

Verification technologies can serve as important tools of statecraft in arms control efforts beyond providing monitoring capabilities and deterring noncompliance. The INF Treaty provides an insightful case study of the strategic side of verification technologies, illuminating how these capabilities can be used to shape negotiation positions and treaty content, provide domestic and international assurance, and manage tensions in the treaty implementation process. Operationalizing verification technologies in this way expands the tools available at the diplomatic and policy levels, but it requires foresight and preparedness to identify and develop the relevant technical capabilities. The most important finding of this case study is that technical preparedness had an important impact, but only after a period of latency. The verification approaches used in the INF Treaty were developed before the topic was raised in the negotiations, and thus the capabilities were available when certain political shifts had occurred

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2019

Progress Towards A New CTH Code Verification & Validation Test Suite

The CTH multiphysics hydrocode is used in a wide variety of important calculations. An essential part of ensuring hydrocode accuracy and credibility is thorough code verification and validation (V&V). In the past, CTH V&V work (particularly verification) has not been consistently well documented. In FY19, we have made substantial progress towards addressing this need. In this report, we present a new CTH V&V test suite composed of traditional hydrocode verification problems used by similar ASC codes as well as validation problems for some of the most frequently used materials models and capabilities in CTH. For the verification problems, we present not only results and computed errors, but also convergence rates. Validation problems include mesh refinement studies, providing evidence that results are converging.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2020

Assessment of the 3420 Building Filtered Exhaust Stack Sampling Probe Location: Stack Verification Following Fan and Air Blender Additions

The Pacific Northwest National Laboratory (PNNL) 3420 Building, which is part of the Physical Sciences Facility (PSF), houses radiological capabilities that results in the requirement that emissions monitoring must be conducted for potential radionuclides in the exhaust air discharge of this building. The air monitoring system is required to conform to Title 40 of the Code of Federal Regulations part 61 (40 CFR 61) Subpart H, which in turns requires a sampling probe in the exhaust stream to conform to the criteria of American National Standards Institute / Health Physics Society (ANSI/HPS) N13.1-2011, Sampling and Monitoring Releases of Airborne Radioactive Substances from the Stack and Ducts of Nuclear Facilities. To support the air emissions permit for the 3420 Building on the PNNL campus, stack testing that used computational fluid dynamics (CFD) modeling as a surrogate stack and verification tests of velocity uniformity and flow angle on the retrofitted facility stack was performed. The ANSI/HPS N13.1-2011 criteria for the air monitoring probe location are that velocity uniformity, gaseous tracer uniformity, and particulate tracer uniformity must be less than or equal to 20%COV. Furthermore, no point in the sampling location may have a gaseous tracer concentration that varies from the mean concentration by more than 30%. Additionally, the flow angle at the sampling location must not be more than 20°. The CFD modeling of the stack, as reported by Recknagle et al. (2018) demonstrated that the stack meets the criteria at the probe location. The velocity uniformity and flow angle results from the 3420 stack verification tests, performed in October 2020, demonstrated that the CFD model results may be used to support the qualification of the stack sampling location. The measured velocity uniformity verification test result was 1.4%COV. This value is well within the uniformity criterion, which is that the velocity uniformity be =20%COV. Additionally, this value is well within the criterion that the actual stack measurement must be within 5% of the surrogate stack result of 2.1%COV when all four fans were operating. Additionally, the measured average flow angle at the 3420 stack monitor location was 15.5 degrees. Although this is higher than expected based on the CFD model, the result is =20 degrees, so the criterion is met. Based on these stack verification test results, the reconfigured 3420 Building filtered exhaust stack meets the qualification criteria given in the ANSI/HPS N13.1-2011 standard. Further changes to the system configuration or operating conditions that are outside the bounds described in this and the CFD report (Recknagle et al., 2018) may require additional tests and additional analysis to determine compliance with the standard.

3420 Building↗

DOE OSTI2020

Verification of the CP2020 Library

For many years, “pencil beam” (aka “broomstick”) problems have been used in Monte Carlo neutronic code verification studies . At Los Alamos, they were used to verify the changes associated with the upgrade of NJOY / MCNP to allow continuous angular distributions (instead of discrete angular distributions) from S(α,β) scattering. Recently, “pencil beam” problems have been applied to the verification of charged particle data for MCNP. Another verification test for CP2020 was to compare the continuous energy reaction cross sections generated by ACER with multigroup reaction cross sections generated by GROUPR from the same evaluation file. This method was also used in the verification of the most recent S(α,β) neutron libraries at Los Alamos .

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2018

Argonne Reactor Code Software Verification and Validation Plan for the Versatile Test Reactor

This Argonne Reactor Code (ARC) software has taken a pivotal role in the design work of the Versatile Test Reactor. The ARC software itself consists of several independent or semiindependent pieces of software that are used in concert to carry out specific analysis and design objectives. In this report, we discuss the validation that is needed for the ARC software overall and discuss available, usable experimental data and how it relates to the VTR project. We also discuss the verification that is required for the ARC software with respect to its usage in the VTR project and give some concrete examples as to what it must accomplish. A list of goals for the verification and validation work are presented as the general procedure that will be implemented for every individual piece of software in ARC. The ARC software has been continuously developed for over 50 years and many of the features used for the VTR project have thus been rigorously tested for accuracy during that timeframe. The ARC software is presently under source code control and undergoes nightly regression testing of not only the individual pieces, but the interconnected aspects that are common to most analysis projects. Its greatest weakness, from a software verification and validation perspective, is both a lack of documentation and outdated documentation. As an example, much of the previous validation work will require contemporary effort to re-evaluate given the updated base data libraries and modifications to the software that have occurred since. This requires new documentation to be produced to support the usage of the software in the VTR project. The bulk of the ARC software is focused on steady state neutronics and TH behavior of a reactor system. In this regard, we have identified the ZPPR, FFTF, and EBR-II programs as the best candidates for sources of validation data for the VTR project. The ZPPR-15 series in the ZPPR program was chosen as it is the best representation of the VTR fuel form and reactor core layout of all of the ZPPR measurements. Additional support for it comes from the fact that most of the data has already been processed into usable ARC models for VTR validation. The FFTF and EBR-II reactors were both functional reactors that provide reactor scale measurements. FFTF is clearly a closer match to VTR and its startup measurements would be a very valuable validation basis. EBR-II has invaluable burnup related measurements that will play a key role in the ARC software validation work. Each piece of ARC software will require a report detailing the software verification. The validation work will be logically split into the results of modeling each identified experimental resource (ZPPR, EBR-II, FFTF). The intended timeline for the VTR project is to have the ARC software verified and validated by the end of fiscal year 2020. This is incumbent on sufficient funding allocated to V&V activities and no change in programmatic priorities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

A Verification and Validation Approach for COMSOL Multiphysics to Support High Flux Isotope Reactor (HFIR)

Over the last several decades, many reactors have successfully been converted from high enriched uranium (HEU) to low enriched uranium (LEU) fuels in United States in support of its global non-proliferation objectives. Of the reactors slated for conversion, five high-performance research reactors (HPRRs) remain. The high-flux isotope reactor (HFIR) at Oak Ridge National Laboratory (ORNL) is one of the five. Conversion of HFIR requires the qualification of a new fuel. To aid in the conversion process as well as improve safety margins, COMSOL Multiphysics was chosen to support and eventually supplement the steady-state heat-transfer code (SSHTC). In order to use COMSOL as a supplementary tool, i.e. one used for safety basis calculations, the code needs to undergo a verification and validation process. Verification and validation are important processes for the use of any software; however, physics-based solvers have a unique challenge, in that the validation is much more rigorous and its domain is only applicable to the general scope of the problem. This document details the verification and validation of COMOSL for analysis of HFIR. This report is intended for two different users: (1) current and future users of COMSOL for HFIR and (2) for individuals interested in assessing the scope of the validation of COMSOL for HFIR. This report is divided into six sections. The first of which details the process of verification and validation. Section one provides an overview of key concepts from ASME V&V 10, 20, and 40 and how these concepts are used to provide a validation of any multiphysics software. Section two covers the approach needed in order to validate multiphysics software for a given HFIR analysis. Parts three through five split up the key concepts highlighted in section two. The final section provides concluding remarks which demonstrate the scope and limitations of the validation study and how these may be expanded and improved.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Verification and Validation Activities for the Multi-Fidelity Toolkit

The Multi-Fidelity Toolkit (MFTK) is a simulation tool being developed at Sandia National Laboratories for aerodynamic predictions of compressible flows over a range of physics fidelities and computational speeds. These models include the Reynolds-Averaged-Navier-Stokes (RANS) equations, the Euler equations, and modified Newtonian aerodynamics (MNA) equations, and they can be invoked independently or coupled with hierarchical Kriging to interpolate between high-fidelity simulations using lower-fidelity data. However, as with any new simulation capability, verification and validation are necessary to gather credibility evidence. This work describes formal code- and solution-verification activities as well as model validation with uncertainty considerations. Code verification is performed on the MNA model by comparing with an analytical solution for flat-plate and inclined-plate geometries. Solution-verification activities include grid-refinement studies of HIFiRE-1 wind tunnel measurements, which are used for validation, for all model fidelities. A thorough treatment of the validation comparison with prediction error and validation uncertainty is also presented.

42 ENGINEERING↗

DOE OSTI2019

SAS4A/SASSYS-1 Verification and Validation Plan for the Versatile Test Reactor

Software verification and validation (V&V) hold an important role in the safety analysis for nuclear facilities. Codes and models utilized in calculations that demonstrate the safety basis as part of a license application or authorization must demonstrate an acceptable pedigree with regard to key critical characteristics that are comprised of the functional and performance requirements of the software. This document describes the methodology for verification and validation of the SAS4A/SASSYS-1 systems-level liquid-metal safety analysis software in support of DOE authorization of the Versatile Test Reactor (VTR). The overall goal of this V&V effort is to demonstrate the suitability of SAS4A/SASSYS-1 for evaluation of safety performance and characterization of safety margins in VTR. To that end, the V&V methodology will be used to demonstrate the adequacy of the component and physical models of SAS4A/SASSYS-1 for the corresponding VTR design elements. This methodology, which has been derived as a generalized approach to the U.S. NRC’s evaluation model development and assessment process (EMDAP) of Regulatory Guide 1.203, entails four high-level activities: software requirements identification, assessment base development, determination of acceptance, and gap closure. The methodology has been generalized to be suitable for both verification and validation of the software for the VTR project. Several high-level engineering analysis quality assurance measures as well as supporting V&V methodology documentation are also described in this document to help support demonstration of software acceptance. To support initial development of the SAS4A/SASSYS-1 V&V basis, existing verification test problems and systems benchmarks are also identified and briefly described in this document.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Continued Verification of MOOSE Structural Mechanics Tools for Modeling Core Bowing Phenomena in Fast Reactors

Under the U.S. Department of Energy Office of Nuclear Energy’s Advanced Modeling and Simulation (NEAMS) Program, an integrated multiphysics approach is being developed to model the core bowing phenomena important to liquid metal-cooled fast reactors. Core bowing is an important passive safety mechanism whereby increased power (which leads to temperature and flux gradients) influences the core to bow into less reactive configurations when the restraint system is properly designed. The phenomenon includes a complex interplay of radiation transport, duct temperature calculations involving fluid flow and heat transfer, and thermo-mechanical responses to the induced temperature and flux gradients. Structural material properties are also important to determining inelastic response to longer term flux gradients which cause irradiation creep and swelling. While core bowing provides a strong negative reactivity feedback when the restraint system is designed properly, it also results in additional forces between assemblies which increase the loads required to extricate them during refueling or control rod movement. Therefore, the restraint system must be designed with these tradeoffs in mind. The first stage of the work, which commenced in FY21 and continues through FY22, assesses thermo-mechanical modeling tools for producing core bowing predictions consistent with conventional tools. The Multiphysics Object Oriented Simulation Environment (MOOSE) Tensor Mechanics and Contact Modules are employed. This status report describes work on additional thermo-mechanical benchmark verification problems with increased complexity from the examples demonstrated in FY21. Several benchmark verification examples were selected from the IAEA verification and validation report. These examples involve clusters of ducts representative of a sector of a hexagonal reactor core which bow into each other and cause contact and load pad elevations, as well as single ducts subjected to irradiation fields undergoing swelling and subsequent bowing. The MOOSE-based results were compared to both IAEA benchmark participants’ results, analytic equations as available, and NUBOW-3D, a beam model code developed by Argonne National Laboratory. In every case, the MOOSE results agreed with other simulations results, providing additional verification basis of the tools for this particular physics application.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Office of Nuclear Verification FY 2023 Quarterly Report TSVT Q1

This project encompasses the continued development and training of a U.S. operational team, the Test Site Verification Team (TSVT), supporting verification of nuclear testing activities. TSVT builds on decades of U.S. nuclear testing history and nuclear explosion monitoring experience. The Team maintains readiness to deploy internationally on short notice to provide field-based support of verification of declared or undeclared nuclear testing and associated activities, as well as follow-on activities including monitoring and capability disablement and dismantlement, as established by negotiated agreement or treaty. The roles and structure of the TSVT are integrated with other NA-243 deployable verification teams and the interagency. FY23 TSVT activities focus on continued capability buildup within the team, including Team trainings and exercises with a focus on missions in confined spaces (e.g., tunnels, mines, other underground facilities), increased familiarity with foreign nuclear weapons testing programs, demonstration of Team capacity to deploy, train, and practice sustained OPSEC in non-western locations, establishment of sustainable storage and maintenance of equipment, specification and procurement of additional equipment to support field observations, further evolution of concepts of operation documents (CONOPs), and mission coordination with Headquarters and associated Stakeholders. Activities will culminate with a full-scale domestic team exercise at the end of FY23 focusing on underground activities signatures/observations and safety including advanced outdoor safety and familiarity in working around explosive test environments. In addition, we will be further articulating approaches and capacity relevant to the identification of nuclear tests, as well as monitoring of nuclear testing activities and/or dismantlement of nuclear test sites and anticipate developing additional equipment requests in support of this evolution. The TSVT Team Leads will also coordinate with its Senior Advisor, the TSVT Logistics and Readiness (L&R) Training Lead, and NNSA Headquarters to draft a five-year TSVT training and exercise plan, that strategically and incrementally builds capacity and expertise in key areas of significance for the continuum of nuclear and nuclear-related testing activities that fall within the team’s mission space.

42 ENGINEERING↗

DOE OSTI2025

Adopting Code Verification Methodology Based on Model Form

Code verification is an essential part of credibility analysis for computational models. It assesses whether the mathematical model is implemented correctly into the code and whether the numerical methods behave consistently, and is done before solution verification and validation. Robust guidance for code verification exists in the literature. However, there is no known, concise guide for selecting the approach based on the model form that also presents an overview of the common elements. This document was written to address this gap as an accessible reference for beginning a code-verification effort.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2025

Adapting Code Verification Methodology to Model Form

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2025

Verification of the PERSENT Software

Ongoing commercial design activities require a thorough verification of the Argonne Reactor Computation codes be performed. DIF3D is central to this system and substantial work has been done to verify its accuracy on several identified commercial needs. This manuscript details the verification work done on PERSENT which relies upon the DIF3D code for its forward and adjoint flux solution. Previous work identified the PERSENT features required to be verified to support commercial design activities, features of which are generally applicable to hexagonal-Z fast reactor designs. The scope of this verification effort includes verifying PERSENT’s ability to correctly calculate four key quantities: perturbation worth distributions, kinetics parameters, sensitivity coefficients, and cross section uncertainty quantification. This manuscript provides the verification tasks and their results with respect to these quantities needed for commercial design activities. For the perturbation worth distributions, hand calculations are deployed to verify the PERSENT calculated results. Similarly, hand calculation of the PERSENT computed kinetics parameters is also used to verify the PERSENT results. In both of these, the input to PERSENT is manipulated to ensure the hand calculation exactly matches the equations PERSENT is calculating. The sensitivity coefficients involve calculating the derivatives of a parameter (such as reactivity worth), with respect to the cross section data. Direct finite difference calculations with DIF3D are used to verify the PERSENT calculated results. For the uncertainty quantification, manufactured input to PERSENT is used to allow an exact hand calculation to reproduce the PERSENT calculated results. The work detailed in this report verified that significant issues were identified for earlier versions of PERSENT for sensitivity coefficients which were corrected in this work and thus version 12.1.0 of PERSENT must be used to reproduce all of the verified work in this report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

MPACT Verification and Validation Manual (rev 1)

The MPACT code provides the capability to calculate the steady-state and transient reactor core neutron distribution for operating nuclear power plant (PWR) conditions across multiple fuel cycles. As required in the CASL SQA plan, it is the responsibility of UM and ORNL as joint owners of MPACT to ensure that verification and validation activities are performed and documented in a V&V manual with supporting publications and CASL technical reports which can be distributed for reference and distribution within VERA. This document provides the first revision of the MPACT verification and validation (V&V) manual and describes the current state of MPACT V&V and updates the framework for future MPACT V&V activities. During FY16, MPACT V&V activities in the areas of code verification have continued to mature and work has progressed on code error evaluation using the Method of Manufactured Solutions (MMS). Several specific tasks have also been completed to improve MPACT code verification to include tasks to improve the unit and regression test coverage. A joint UM/ORNL task was completed to develop standardize coding standards and workflow for the continued collaborative development of MPACT. During the past year, MPACT validation work has continued in both the areas of measured data from critical experiments as well as measured data from operating nuclear power plants. Preliminary work was also performed in FY16 on the use of post-irradiation examinations for evaluation and validation of the isotopic depletion capability in MPACT. Based on the results of V&V efforts this past year, the confidence level has increased in the ability of MPACT to model an operational Pressurized Water Reactor. Areas were identified which require increased emphasis during FY17 to include updating all the V&V problems with the new 51-group library, the completion of cycle 2 of the BEAVRS benchmark, initial validation for BWR applications, the addition of Uncertainty Quantification results into the benchmark results, and the completion of specific validation tasks for the CASL Challenge problems.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS1975

General verification description

A brief general description of the ASTP flight program verification was presented. The total program verification effort assures the accuracy and adequacy of the LVDC flight program and verifies that the final program meets mission requirements and conforms to program documentation. The flight program's functional requirements to integrate the guidance and control system with the launch vehicle sequencing system are verified directly by analysis of many special logic checks designed for this purpose and indirectly by the correct overall program response to nominal and numerous perturbed conditions. Verification of the interaction of function requirements is accomplished on every case run during the verification effort.

Source record↗

NASA NTRS1978

Integrated guidance, navigation and control verification plan primary flight system

The verification process and requirements for the ascent guidance interfaces and the ascent integrated guidance, navigation and control system for the space shuttle orbiter are defined as well as portions of supporting systems which directly interface with the system. The ascent phase of verification covers the normal and ATO ascent through the final OMS-2 circularization burn (all of OPS-1), the AOA ascent through the OMS-1 burn, and the RTLS ascent through ET separation (all of MM 601). In addition, OPS translation verification is defined. Verification trees and roadmaps are given.

Source record↗

NASA NTRS1982

Geometric verification

Present LANDSAT data formats are reviewed to clarify how the geodetic location and registration capabilities were defined for P-tape products and RBV data. Since there is only one geometric model used in the master data processor, geometric location accuracy of P-tape products depends on the absolute accuracy of the model and registration accuracy is determined by the stability of the model. Due primarily to inaccuracies in data provided by the LANDSAT attitude management system, desired accuracies are obtained only by using ground control points and a correlation process. The verification of system performance with regards to geodetic location requires the capability to determine pixel positions of map points in a P-tape array. Verification of registration performance requires the capability to determine pixel positions of common points (not necessarily map points) in 2 or more P-tape arrays for a given world reference system scene. Techniques for registration verification can be more varied and automated since map data are not required. The verification of LACIE extractions is used as an example.

Grebowsky, G. J.↗

NASA NTRS1982

Automated verification of flight software. User's manual

(Automated Verification of Flight Software), a collection of tools for analyzing source programs written in FORTRAN and AED is documented. The quality and the reliability of flight software are improved by: (1) indented listings of source programs, (2) static analysis to detect inconsistencies in the use of variables and parameters, (3) automated documentation, (4) instrumentation of source code, (5) retesting guidance, (6) analysis of assertions, (7) symbolic execution, (8) generation of verification conditions, and (9) simplification of verification conditions. Use of AVFS in the verification of flight software is described.

Saib, S. H.↗