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DOE OSTI2016

VERA 3.5 Installation Guide

This guide describes the structure and setup of the standard VERA development environment (VERA Dev Env) and standard VERA Third Party Libraries (TPLs) that need to be in place before installing many of the VERA simulation components. It describes everything from the initial setup on a new machine to the final build, testing, and installation of VERA components. The goal of this document is to describe how to create the directories and contents outlined in Standard VERA Dev Env Directory Structure and then obtain the remaining VERA source and build, test, and install any of the necessary VERA components on a given system. This document describes the process both for a development version of VERA and for a released tarball of the VERA sources.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

VERA Installation Guide

This guide describes the structure and setup of the standard VERA development environment (VERA Dev Env) and standard VERA Third Party Libraries (TPLs) that need to be in place before installing many of the VERA simulation components. It describes everything from the initial setup on a new machine to the final build, testing, and installation of VERA components. The goal of this document is to describe how to create the directories and contents outlined in 'Standard VERA Dev Env Directory Structure' and then obtain the remaining VERA source and build, test, and install any of the necessary VERA components on a given system. This document describes the process both for a development version of VERA and for a released tarball of the VERA sources.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2015

VERA Installation Guide

This guide describes the structure and setup of the standard VERA development environment (VERA Dev Env) and standard VERA Third Party Libraries (TPLs) that need to be in place before installing many of the VERA simulation components. It describes everything from the initial setup on a new machine to the final build, testing, and installation of VERA components. The goal of this document is to describe how to create the directories and contents outlined in Standard VERA Dev Env Directory Structure and then obtain the remaining VERA source and build, test, and install any of the necessary VERA components on a given system. This document describes the process both for a development version of VERA and for a released tarball of the VERA sources.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2015

Illustration of VERA Capabilities to Model a Typical iPWR SMR

Small modular reactors (SMRs) are being considered by the commercial nuclear power industry as an option for more distributed generation and for replacement of older fossil fuel generating facilities. SMRs are more compact than operating pressurized water reactors (PWRs), producing about 200 MWe as compared to 1000 MWe or higher for their full-sized cousins, and are offered as “expandable” units; that is, their modular design allows the utility to add more units progressively at the same site. In Summer 2014, CASL reported on a progression of demonstrations illustrating VERA’s ability to simulate a typical SMR of the iPWR type, while developing a simple core cycle design for cycle economic studies. The work included development of fuel assembly lattices, 3D single assembly and full core (quarter symmetry) simulations of the proposed core cycle design using the CASL Virtual Environment for Reactor Applications (VERA). The report concluded that it would be difficult for the SMR to operate for the target 1400 EFPD cycle length and that the fuel cost would be on the order of 50% higher as compared with a PWR. Unfortunately, the Reference 4 core cycle design did not include control component management. The iPWR is intended to be operated using the control rods for reactivity control, as opposed to using soluble boron (like most currently operating PWRs), and the introduction of the control rods is expected to affect the length of the cycle. Therefore, this work continues the previous CASL demonstration of VERA capability to simulate a typical iPWR core, including the development of a control component management strategy using the Reference 4 core design and utilizing the soon-to-be released VERA3.3 tools. Note that this study is not intended as a validation study; rather it is merely an illustration of VERA’s capabilities to simulate a typical iPWR. Simulations included two variations on control rod banks – a two-bank scheme and a ten-bank scheme. Both schemes were modelled using the VERA Common Input, and more than six iterations on each control rod scheme were simulated to arrive at a combination of rod banks and insertion depths that provided a core reactivity near the target k eff of 1.000 ± 0.001 (more than 50 cases run) over two calendar months. In particular, VERA was used to determine the hot excess reactivity for the core design with all rods out (ARO), and then iterative simulations with the rod banks inserted to determine the critical control rod bank positions were completed. VERA options to simplify the process were exercised as possible, particularly restart functionality, coupled thermal-hydraulics using the internal T-H solver, xenon, and depletion capability. The fuel shuffling functionality was not exercised, since the iPWR SMR features a once-through cycle. The VERA method of characteristics (MOC) neutronics subcomponent was used to predict power and reactivity with control rod maneuvering, both with and without thermal-hydraulic feedback. The problem was scaled from 35 CPU cores to 3815, effectively demonstrating the solution time with respect to computing capacity. Computer clusters at Oak Ridge National Laboratory, Idaho National Laboratory and Tennessee Valley Authority were used. These studies provide an additional illustration of the application of VERA to a typical iPWR SMR. Reference 4 previously demonstrated that the codes could handle the differences in geometry for the iPWR, and this study expanded the illustration to include control rod management. Further, the work illustrates the need for in-depth simulations to support effective control rod maneuvering for an iPWR and highlights the need for control rod reactivity search capability within VERA. Additionally, the development of a control rod bank reactivity search capability will be extremely useful for not only this application, but also for BWRs as CASL develops BWR capability in Phase 2.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Executing BISON-CASL Fuel Performance Cases Using VERA-CS Output

As the Consortium for Advanced Simulation of Light Water Reactors (CASL) moves forward with more complex multiphysics simulations, there is increased focus on incorporating fuel performance analysis methods. The coupled neutronics/thermal-hydraulics capabilities within the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) have become relatively stable and major advances have been made in analysis efforts, including the simulations of twelve cycles of Watts Bar Nuclear Unit 1 (WBN1) operation. However, VERA-CS approaches for treating fuel pin heat transfer have well-known limitations that could be eliminated through better integration with the BISON-CASL fuel performance code. Several approaches are being taken to improve or replace the VERA-CS fuel models and improve integration with BISON-CASL. Tiamat is being developed to replace the fuel models with a direct coupling of BISON-CASL and the neutronics and coolant heat transfer of VERA-CS. Tiamat will execute in an inline mode that allows for in-memory transfer of VERA-CS data to BISON-CASL. This will provide fuel performance results during core simulation. However, fuel performance typically undergoes an independent analysis using a stand-alone fuel performance code with manually specified input defined from an independent core simulator solution or set of assumptions. This milestone covers initial efforts to facilitate using VERA for core simulation and fuel performance to operate in this mode of execution. This approach will be used for several CASL fuel performance challenges, such as RIAs and DNB. These issues require time-dependent full core results to establish the boundary conditions for the high-fidelity BISON-CASL simulations. The approach will also be used to improve fuel temperature models in VERA-CS, guide the development of Tiamat, and simplify the use of BISON-CASL within VERA for industry users.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

VUSAT - A VERA-CS Uncertainty and Sensitivity Analysis Toolkit

VERA-CS (the Virtual Environment for Reactor Applications, Core Simulator) is a high-fidelity multiphysics core simulator under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS includes three main components: MPACT for reactor physics and neutron transport, COBRA-TF for thermal-hydraulics, and ORIGEN for isotopic depletion. VERA-CS has been applied to an array of problems including core physics analysis, and the modeling for all of the fuel cycles of Watts Bar Nuclear Unit 1 (WBN1). MPACT is a 3D pin-resolved reactor transport code developed by Oak Ridge National Laboratory (ORNL) and the University of Michigan. COBRA-TF (CTF) is a three-field two-phase flow subchannel analysis code developed by ORNL and North Carolina State University. CTF solves the mass, momentum, and energy equations for the liquid, vapor, and droplet phases for each subchannel in the core and captures the axial flow within each channel as well as cross-flow between channels. Direct coupling between neutronics (MPACT) and thermal-hydraulics (CTF) in VERA-CS provides temperature and density feedback. CTF includes a number of models important for LWR safety and design analysis including flow regime dependent two-phase wall heat transfer, inter-phase heat transfer and drag, droplet breakup, and quench-front tracking. ORIGEN is developed by ORNL and released through the SCALE package. ORIGEN is directly integrated in MPACT through its API to calculate the changes in nuclides for every depletable cross-section region. Uncertainties exist in safety calculations with computer codes for nuclear reactor modeling and simulation, and they come from code limitations, scaling inaccuracies embedded in the experimental data, and uncertainties associated with the state of the reactor at the initiation of the transient. Consequently, for applications of high fidelity calculations using VERA-CS, proper uncertainty quantification (UQ) and sensitivity analysis (SA) of the computing models of reactor neutronics and thermal-hydraulics are important to comprehensively understand the results. New methods in UQ and SA have to be developed in multi-physics calculations using VERA-CS. Toward this end, a software package called VUSAT which stands for VERA-CS Uncertainty and Sensitivity Analysis Toolkit has been developed to facilitate the applications of VERA-CS within the context of best estimate plus uncertainty (BEPU) analysis framework. In this work, the VUSAT application to analyze departure from nucleate boiling (DNB) is presented.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Coupled Neutronic and Thermal Hydraulic Analysis of a Natural Circulation Based Small Modular Reactor (SMR) Using VERA-CS

As part of the work supported by a US Department of Energy (DOE) Office of Nuclear Energy Gateway for Accelerated Innovation in Nuclear FY 2020 Voucher, Holtec International subsidiary SMR LLC (hereinafter referred to as Holtec) and Oak Ridge National Laboratory entered into a cooperative research and development agreement (CRADA) to develop coupled multiphysics core models of the Holtec developed small modular reactor (SMR), SMR-160. The scope of the 1 year project was to use the DOE–developed tool set, VERA, to analyze several aspects of a representative SMR-160 core design. The goals of the project were to perform the code-to-code benchmarking and to provide Holtec with a confirmatory calculations to compare against the results with the codes used in the SMR-160 design, identify and resolve areas of discrepancy between the two models to give Holtec a basis for selecting certain tool and model options, and develop a workflow so that the VERA model could be adapted easily to any future changes in the design of the SMR-160 core. The work proceeded in four phases: 1. Developing the initial VERA model based on Holtec-provided core design information; 2. Comparing VERA results with Holtec-provided calculations for the first five cycles; 3. Developing a VERA transient model based on the prior model; and 4. Training Holtec personnel on VERA and hand-off the developed models. This report summarizes at a high level the completed tasks. Details of the tasks and results are reserved for the full report, which will not be publicly released for 5 years per the terms of the CRADA. The report is titled “Coupled Neutronic and Thermal Hydraulic Analysis of a Natural Circulation Based Small Modular Reactor (SMR) using VERA-CS” and has the ORNL technical report number ORNL/TM 2021/2303 (NFE-20-08305)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Multiphysics Coupling Methods for Molten Salt Reactor Modeling and Simulation in VERA

New multiphysics coupling capabilities for molten salt reactor (MSR) analysis have been developed in the Virtual Environment for Reactor Applications (VERA). This development consisted of two main efforts. First, a generic species transport module was added in the CTF code, which is the thermal-hydraulics (TH) code for VERA. This module uses the velocity fields for which CTF solves during the TH calculation to transport species through the core and around the primary loop. Additionally, a gas sparging model has been added to CTF to model the movement of certain species, namely, fission products such as xenon gas, to transport between the molten salt and gas bubbles present in the salt. The second effort in this development was coupling this capability to VERA’s neutron transport code MPACT. This effort focused on coupling the detailed TH transport models in CTF to MPACT to account for feedback effects in the neutron transport calculations. Finally, the thermochemistry code Thermochimica has also been coupled to VERA. Thermochimica performs pointwise calculations for chemical potential and Gibbs free energy and determines what phases are produced by the temperature, pressure, and elemental concentrations at different locations in the primary loop. These capabilities are demonstrated using a model of the Molten Salt Reactor Experiment (MSRE). Furthermore, this reactor operated at Oak Ridge National Laboratory in the 1960s, providing sources of experimental data that were used to develop the model. Various combinations of species were modeled using VERA’s new multiphysics coupling capabilities. Species distributions and reactivity effects behaved as expected for the MSRE model, demonstrating that the coupling is behaving correctly and causing appropriate feedback. The results of these calculations show the potential for VERA to be used for a wide variety of MSR analyses.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Assessment of CASL VERA for BWR analysis and application to accident tolerant SiC/SiC channel box

Application of the Virtual Environment for Reactor Applications (VERA) to BWR analysis is assessed in this paper by comparing results to those calculated using other widely-used modeling tools, namely the U.S. Nuclear Regulatory Commission’s PARCS/PATHS and the Serpent Monte Carlo particle transport code. Additionally, VERA is used to calculate 3-D temperature and fast neutron flux distributions in silicon carbide (SiC) fiber-reinforced, SiC matrix composite (SiC/SiC) BWR channel boxes, which are being studied as an Accident Tolerant Fuel core structural material concept. We found that the code-to-code comparisons were favorable, and the SiC/SiC channel box evaluation demonstrates the many advanced modeling features VERA provides while also highlighting the non-uniformity in fast neutron flux distributions that can play a role in potential SiC/SiC channel box deformation. Traditional BWR analysis tools do not have the calculation fidelity necessary for coupled assessment of flux and temperature gradients in a SiC/SiC channel box. VERA is a state-of-the-art modeling environment that was developed to increase the safety and economic competitiveness of nuclear power through improved modeling accuracy. While VERA has already been deployed in the nuclear industry for PWR applications, the current study is a vital initial step in the extensive development, validation, and verification that VERA must go through to be useful for BWR applications.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Development of Preliminary VERA-CS Crud Induced Localized Corrosion Modeling Capability

The objective of this milestone was to develop a crud-induced localized corrosion (CILC) screening capability in VERA-CS using existing codes and capabilities that had been developed in the Consortium for Advanced Simulation of Light Water Reactors (CASL) as part of the efforts to resolve the crud challenge problem. The milestone work focused on improving existing CASL software and implementing clad corrosion modeling into VERA-CS. Clad corrosion modeling was implemented by creating an application interface (API) in the Cicada software so that its clad heat transfer and corrosion models could be used in VERA-CS. Cicada was coupled to the subchannel code CTF for solving crud and corrosion growth. Additionally, a clad corrosion model was added directly to CTF to serve as an alternate approach. The rod thermal hydraulic reconstruction (ROTHCON) process, which is used for capturing computational fluid dynamics (CFD)-predicted rod-surface heat and turbulence behavior that can impact crud-induced power shift (CIPS) and CILC prediction, was also implemented into VERA-CS. Usability improvements were made to this feature, which was developed over several previous milestones, so that the VERA-CS user can link spacer grids in a VERA-CS model to an applicable pregenerated spacer grid data file if available. Several assessments were performed to demonstrate the proper functioning of the new capability. Sensitivity studies were performed to demonstrate the effect of crud lithium content, oxide thermal conductivity, Cicada mesh refinement, and ROTHCON coupling mesh refinement on predicted amounts of crud and oxide growth. A full-scale demonstration was performed by running Seabrook Cycle 5 with the CILC capability. Some general observations were made from the assessments. It was observed that predicted corrosion thickness is less than anticipated for CILC cases, so further work is needed to better understand and improve the corrosion modeling. Refining the coupling mesh in the ROTHCON process leads to increased corrosion and crud thickness, which can affect where crud and corrosion grow, how thick it gets, and how many rods experience crud and corrosion growth. Results indicate that applying the ROTHCON process and refining the coupling mesh leads to only a slight increase in total core maximum crud and corrosion thickness, but it leads to a significant increase in the number of rods with a maximum thickness that surpasses a set threshold value (i.e., more rods grow thicker corrosion and crud). Computational performance of the simulation is degraded by a factor of 5{6 with high coupling mesh refinement. The impact of the ROTHCON coupling mesh refinement on total core crud mass is less significant; however, using the ROTHCON process with CFD-generated data over the Yao{Hochreiter{Leech (YHL) model in CTF leads to significantly different amounts of total core crud mass. Future work on the CILC screening capability will include model improvement and validation, development and testing of a CILC-risk screening criteria, and performance improvements.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

VeraShift I/O Optimization for Ex-core Calculations with VERA

The initial release of Shift in the Virtual Environment for Reactor Applications (VERA) provided users with the ability to perform ex-core calculations. After this release, a task was undertaken to optimize input parameters and output of desired quantities for ex-core calculations. As requested by analysts, this report presents the detailed input and output changes made to VERA to enhance usability. Specifically, some of the major optimizations added were detector and bioshield setup using the VERA common input, updated Shift input defaults, consolidated output, and output of pin importance's. Close interaction between developers, analysts, and users enabled these useful optimization updates, which will be included in the release of VERA 4.1, and some are included in VERA 4.0.1. Under these auspices, this work will continue under the VERA Users Group (VUG).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

VERA Integration in the NEAMS Workbench

The Virtual Environment for Reactor Applications (VERA) is a cutting edge simulation code suite used to solve complex multiphysics problems for nuclear reactors. VERA makes use of high performance computing (HPC) systems to solve large problems that were not historically tractable. While extensive work has been put into the usability of VERA, it still takes significant effort to build reactor models, ensure their correctness, properly execute the calculations, and conduct the necessary analysis of the results. This difficulty is common for many complex codes such as VERA. As part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the NEAMS Workbench has been developed and maintained. The goal of the NEAMS Workbench is to provide a single interface with which to execute a wide variety of physics codes, from building the input models to analyzing the results. This document serves as a manual detailing the current integration of VERA with the NEAMS Workbench and a guide for configuring Workbench to allow remote execution of VERA.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

VERA 3.6 Release

CASL has developed a formalized process to release the Virtual Environment for Reactor Applications (VERA) to non-CASL users through the Radiation Shielding Information Computational Center (RSICC). This milestone represents a major release of VERA with updates that significantly reduce computational resources required and introduce new capabilities related to fuel performance integration. This release will also be used by CASL Test Stands and training activities. VERA 3.6 represents an incremental release, with improvements to core component capabilities, significant performance improvements, and updated documentation. This release followed a significantly revised release process, with formal documentation, testing, readiness reviews, and release authorization. The major tasks that were completed as part of this milestone were: 1) Revision of the VERA RSICC Release Process and Procedures; 2) Identification of the VERA release components; 3) Creation of a Release Candidate; 4) Independent evaluation of the Release Candidate by the Advanced Modeling Applications (AMA) Focus Area; 5) A Release Candidate Readiness Review; 6) Development and execution of a System Test Plan; 7) Updates of manuals and other required documentation; 8) Export Control reviews; 9) A Release Readiness Review; and 10) Delivery of the release package to RSICC. Synopses of the work done to complete each of these major tasks are summarized in Section 2 below. Supporting details are provided in the PHI Kanban ticket #4424 Support VERA 3.6 Release Activities and its child tickets.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

VERA 3.3 Release Notes

The Virtual Environment for Reactor Applications components included in this distribution include selected computational tools and supporting infrastructure that solve neutronics, thermal-hydraulics, fuel performance, and coupled neutronics-thermal hydraulics-fuel performance problems. The infrastructure components provide a simplified common user input capability and provide for the physics integration with data transfer and coupled-physics iterative solution algorithms. Neutronics analysis can be performed for 2D lattices, 2D core and 3D core problems for pressurized water reactor geometries that can be used to calculate criticality and fission rate distributions by pin for input fuel compositions. Two neutronics capabilities are provided in this distribution including MPACT (from the University of Michigan) that supports 2D lattices and Insilico (from Oak Ridge National Laboratory) that supports 2D lattices, 2D core planes, and 3D core geometries. Both codes include integrated cross section capabilities that provide problem-specific cross sections for the problems defined. A thermal-hydraulics capability is provided with CTF (an updated version of the COBRA-TF code developed by the Pennsylvania State University) that allows thermal-hydraulics analyses for single and multiple assemblies using the simplified VERA common input. This distribution also includes an initial coupled neutronics/thermal-hydraulics capability to allow calculations of single assemblies using Insilico or MPACT coupled with CTF. The VERA fuel rod performance subcomponent BISON-CASL calculates, on a 3D basis, fuel rod temperature, fuel rod internal pressure, free gas volume, clad integrity and fuel rod waterside diameter. These capabilities allow simulation of power cycling, fuel conditioning and deconditioning, high burnup performance, power uprate scoping studies, and accident performance. Specifically targeted at VERA’s core simulator functionality (VERA-CS), the existing VERA subchannel-neutronics capability based on CTF and Insilico has been extended to use the axisymmetric 2D R-Z fuel rod modeling functionality of BISON-CASL to produce a three-way coupled neutronics, thermal-hydraulics and fuel performance capability known as Tiamat. This distribution is intended to be used for testing and evaluation of the released components of VERA. Testing within CASL has focused specifically on Westinghouse four-loop reactor geometries and conditions with example problems included in the distribution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2013

VERA Geodetic Activities

This report briefly describes the geodetic activities of VERA in the year 2012. The regular geodetic observations are carried out both in K- and S/X-bands. The frequency of regular observations is three times a month-twice for the VERA internal observations in K-band. The networks of the S/X sessions are JADE of GSI and IVS-T2. The raw data of the T2 and JADE sessions are electronically transferred to the Bonn, Haystack, and GSI correlators via Internet. Gravimetric observations are carried out at the VERA stations. An SG was installed at Mizusawa and placed in the vicinity of the VERA antenna in order to monitor vertical displacement at the end of 2008, and the observations continued throughout the year. Also at the VERA-Ishigakijima station, continuous operation of the SG started in 2012. The crustal movements generated by the 2011 earthquake off the Pacific coast of Tohoku continued during 2012, and displacement of the VERA-Mizusawa position by post-seismic creeping continued.

Jike, Takaaki↗
DOE OSTI2018

VERA Infrastructure

The Virtual Environment for Reactor Applications (VERA) Infrastructure provides the infrastructure to build, configure, test, install, and package for release the VERAS suite of codes that is being developed by the Consortium for the Advanced Simulation of Light-water reactors (CASL). It includes the list of codes that can be built within VERA, the location on the configuration-controlled CASL development computer, the version number used in any particular software installation, the location of the third-party libraries used for testing VERA on a suite of the CASL testing and development computers, and the scripts to configure, build, test, install, and package for release a particular version of VERA. The VERA Infrastructure contains a list of scripts, configuration options, and utilities that are specific to CASL, but contain no physics or numeric, nuclear technology, or encryption technology.

Baird, Mark↗
DOE OSTI2017

Initial Verification and Validation Assessment for VERA

The Virtual Environment for Reactor Applications (VERA) code suite is assessed in terms of capability and credibility against the Consortium for Advanced Simulation of Light Water Reactors (CASL) Verification and Validation Plan (presented herein) in the context of three selected challenge problems: CRUD-Induced Power Shift (CIPS), Departure from Nucleate Boiling (DNB), and Pellet-Clad Interaction (PCI). Capability refers to evidence of required functionality for capturing phenomena of interest while capability refers to the evidence that provides confidence in the calculated results. For this assessment, each challenge problem defines a set of phenomenological requirements against which the VERA software is assessed. This approach, in turn, enables the focused assessment of only those capabilities relevant to the challenge problem. The evaluation of VERA against the challenge problem requirements represents a capability assessment. The mechanism for assessment is the Sandia-developed Predictive Capability Maturity Model (PCMM) that, for this assessment, evaluates VERA on 8 major criteria: (1) Representation and Geometric Fidelity, (2) Physics and Material Model Fidelity, (3) Software Quality Assurance and Engineering, (4) Code Verification, (5) Solution Verification, (6) Separate Effects Model Validation, (7) Integral Effects Model Validation, and (8) Uncertainty Quantification. For each attribute, a maturity score from zero to three is assigned in the context of each challenge problem. The evaluation of these eight elements constitutes the credibility assessment for VERA.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Updated Verification and Validation Assessment for VERA

The Virtual Environment for Reactor Applications (VERA) code suite is assessed in terms of capability and credibility against the Consortium for Advanced Simulation of Light Water Reactors (CASL) Verification and Validation Plan (presented herein) in the context of three selected challenge problems: CRUD-Induced Power Shift (CIPS), Departure from Nucleate Boiling (DNB), and Pellet-Clad Interaction (PCI). Capability refers to evidence of required functionality for capturing phenomena of interest while credibility refers to the evidence that provides confidence in the calculated results. For this assessment, each challenge problem defines a set of phenomenological requirements against which the VERA software is assessed. This approach, in turn, enables the focused assessment of only those capabilities relevant to the challenge problem. The evaluation of VERA against the challenge problem requirements represents a capability assessment. The mechanism for assessment is the Sandia-developed Predictive Capability Maturity Model (PCMM) that, for this assessment, evaluates VERA on 8 major criteria: (1) Representation and Geometric Fidelity, (2) Physics and Material Model Fidelity, (3) Software Quality Assurance and Engineering, (4) Code Verification, (5) Solution Verification, (6) Separate Effects Model Validation, (7) Integral Effects Model Validation, and (8) Uncertainty Quantification. For each attribute, a maturity score from zero to three is assigned in the context of each challenge problem. The evaluation of these eight elements constitutes the credibility assessment for VERA.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗