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All sourcesNASA NTRSDOE OSTI
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↗
DOE OSTI2019

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↗
DOE OSTI2016

VERA Benchmarking Results for Watts Bar Nuclear Plant Unit 1 Cycles 1-12

The Consortium for Advanced Simulation of Light Water Reactors, the first US DOE Energy Innovation Hub, is developing and deploying a suite of directly coupled, high-fidelity physics methods known as VERA for performing engineering analyses of operating commercial nuclear power reactors. The centerpiece of VERA is the reactor core simulator, which employs the direct coupling of a 3D neutron transport code, a two-phase sub-channel thermal-hydraulic code, and a well-known isotopic depletion and decay code. Together these tools represent the most rigorous reactor simulator available today. Recently added capabilities include the depletion, shuffling, and discharge of fuel assemblies, enabling VERA to simulate multiple fuel cycles of operation. To demonstrate this capability, VERA has been used to simulate over 18 years of operation of TVA’s Watts Bar Nuclear Unit 1, and the results of these calculations have been used to quantify its performance in an industry-grade benchmarking activity. Comparisons are made with measured critical boron concentrations, control bank reactivity worths, isothermal temperature coefficients, and in-core neutron flux distributions. These results demonstrate that the advanced tools in VERA are capable for application to practical and common industrial problems.

Godfrey, Andrew T↗
DOE OSTI2026

The Vera C. Rubin Observatory Data Preview 1

We present Rubin Data Preview 1 (DP1), the first data from the National Science Foundation–Department of Energy Vera C. Rubin Observatory, comprising raw and calibrated single-epoch images, coadds, difference images, detection catalogs, and ancillary data products. DP1 is based on 1792 optical–near-infrared exposures acquired over 48 distinct nights by the Rubin Commissioning Camera (LSSTComCam) on the Simonyi Survey Telescope at the Summit Facility on Cerro Pachón, Chile in late 2024. DP1 covers ∼15 deg 2 distributed across seven roughly equal-sized noncontiguous fields, each independently observed in six broad photometric bands, ugrizy. The median FWHM of the point-spread function across all bands is approximately 1"14, with the sharpest images reaching about 0." 58. The 5σ point-source depths for coadded images in the deepest field, the Extended Chandra Deep Field South, are u = 24.55, g = 26.18, r = 25.96, i = 25.71, z = 25.07, and y = 23.1. Other fields are no more than 2.2 mag shallower in any band, where they have nonzero coverage. DP1 contains approximately 2.3 million distinct astrophysical objects, of which 1.6 million are extended in at least one band in coadds, and 431 solar system objects, of which 93 are new discoveries. DP1 is approximately 3.5 TB in size and is available to Vera C. Rubin Observatory data rights holders via the Rubin Science Platform, a cloud-based environment for the analysis of petascale astronomical data. While small compared to future LSST releases, its high quality and diversity of data support a broad range of early science investigations ahead of full operations in 2026.

Ground-based astronomy↗
DOE OSTI2021

Validation of Light Water Reactor Ex-Core Calculations with VERA

The Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) is a reactor simulation software. It offers unique capabilities by combining high-fidelity in-core radiation transport with temperature feedback by using MPACT (a deterministic neutron transport code) and COBRA-TF (a thermal-hydraulic code) with follow-on, fixed-source transport calculations using the Shift Monte Carlo code to calculate ex-core quantities of interest. In these coupled calculations, MPACT provides Shift with the fission source for follow-on ex-core calculations. These ex-core simulations can be set up to calculate detector responses, as well as the flux and fluence in ex-core regions of interest, such as the reactor pressure vessel, nozzle, and irradiated capsules. A Watts Bar Nuclear Plant Unit 1 (WBN1) ex-core model was developed, as described in this paper, and this model was used to perform coupon calculations. The results for the coupon flux calculations show close agreement with the reference values for cycle 1 produced by the two-dimensional Discrete Ordinates Transport (DORT) code and presented in a BWXT Services Inc. report. However, differences in the results (10%) seen in cycles 2 and 3 and the reasons for these differences are discussed in this paper. The VERA WBN1 model was also used to perform a vessel fluence calculation for cycle 1. Additionally, a collaboration between CASL and Duke Energy led to the first code-to-code validation of VERA for reactor ex-core applications that used a model for the Shearon Harris reactor. Results from this collaboration show excellent agreement between VERA and the Monte Carlo N-Particle Transport Code for the detector response calculations. The work performed under this collaboration is also detailed in this paper.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Demonstrate VERA Core Simulator Performance Improvements L2:PHI.P13.03

This report describes the performance improvements made to the VERA Core Simulator (VERA-CS) during FY2016. The development of the VERA Core Simulator has focused on the capability needed to deplete physical reactors and help solve various problems; this capability required the accurate simulation of many operating cycles of a nuclear power plant. The first section of this report introduces two test problems used to assess the run-time performance of VERA-CS using a source dated February 2016. The next section provides a brief overview of the major modifications made to decrease the computational cost. Following the descriptions of the major improvements, the run-time for each improvement is shown. Conclusions on the work are presented, and further follow-on performance improvements are suggested.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

File-Based One-Way BISON Coupling Through VERA: User's Manual

Activities to incorporate fuel performance capabilities into the Virtual Environment for Reactor Applications (VERA) are receiving increasing attention [1–6]. The multiphysics emphasis is expanding as the neutronics (MPACT) and thermal-hydraulics (CTF) packages are becoming more mature. Capturing the finer details of fuel phenomena (swelling, densification, relocation, gap closure, etc.) is the natural next step in the VERA development process since these phenomena are currently not directly taken into account. While several codes could be used to accomplish this, the BISON fuel performance code [8,9] being developed by the Idaho National Laboratory (INL) is the focus of ongoing work in the Consortium for Advanced Simulation of Light Water Reactors (CASL). Built on INL’s MOOSE framework [10], BISON uses the finite element method for geometric representation and a Jacobian-free Newton-Krylov (JFNK) scheme to solve systems of partial differential equations for various fuel characteristic relationships. There are several modes of operation in BISON, but, this work uses a 2D azimuthally symmetric (R-Z) smeared-pellet model. This manual is intended to cover (1) the procedure pertaining to the standalone BISON one-way coupling from VERA and (2) the procedure to generate BISON fuel temperature tables that VERA can use.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

NEAMS Milestone Report: M2MS–20OR030102 FW–CADIS PWR Ex-Core Analysis with Shift Through VERA

This report presents the work completed for the NEAMS milestone M2MS–20OR030102 titled "FW–CADIS PWR Ex-Core Analysis with Shift through VERA." The work completed for this milestone includes the implementation, integration, and optimization of memory and performance improvement methods in Shift for fully coupled ex-core calculations through Virtual Environment for Reactor Applications (VERA). Fully coupled in this context means the transfer of moderator boron concentration, pin-wise fission source, depleted compositions, temperatures, and moderator densities from MPACT (with COBRA-TF (CTF)) to Shift. The ability to run ex-core calculations with VERA has been enabled and used for several years by Consortium for Advanced Simulation of Light Water Reactors (CASL) partners. However, this implementation was limited and potentially computationally burdensome. This work has enabled the ability to run higher-fidelity ex-core calculations on moderate computing clusters by focusing on multithreading, domain decomposition, and Forward-Weighted CADIS (FW-CADIS) variance reduction. Tests performed on small cores, a small modular reactor (SMR), and CASL progression problems show very promising memory reduction and computational performance. Recommendations for settings when running fully coupled high-fidelity ex-core calculations with VERA are documented. Without these optimization methods, many processors on a compute node would be left unused for the entire ex-core calculation. Therefore, these methods enable the user to better use the resources available and reduce computation time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

VERA-Grizzly Ex-Core Calculations: Watts Bar Unit 1 Cycles 1-2

The critical structures that comprise light-water reactor (LWR) nuclear power plants are subjected to operating environments that can challenge their integrity. Structures in close proximity to the reactor core, such as the reactor pressure vessel (RPV) and the biological shield wall, are subjected to high levels of radiation emanating from the core, as well as elevated temperatures. As the US fleet of operating LWRs ages, the effects of these operating environments on the integrity of these structures must be considered to ensure their continued safe operation. Extending the lifetime of commercial reactors and maintaining the aging reactor fleet require accurate prediction of the exposure of ex-core components to neutron and photon radiation. In particular, concrete degradation studies must be performed to evaluate the safety and long-term operation of reactors with lifetime extensions. The concrete reactor bioshield is important for providing radiological protection during operation and must last for the entire lifetime of the reactor. Recent interest in lifetime extensions furthers the need to accurately simulate concrete material degradation in the reactor bioshield. As a result of this need, the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has funded this study to couple its tools, Virtual Environment for Reactor Applications (VERA) and Grizzly. VERA allows users to set up models to calculate time-dependent and fully coupled solutions (with thermal feedback) for ex-core quantities of interest such as vessel and coupon fluence and detector responses for multiple statepoints and cycles. Grizzly is a finite-element application based on the Multiphysics Object Oriented Simulation Environment (MOOSE) framework that is used to enable aging materials calculations. This report highlights the work performed to calculate the fluence in the vessel and concrete for Watts Bar Nuclear Plant Unit 1 (WBN1) Cycles 1 and 2. The fluences obtained from VERA were successfully transferred to Grizzly using a Python script. Four simulations were run with Grizzly: (1) the Mazars model with the initial Young’s modulus being the instantaneous modulus, (2) the Mazars model with the initial Young’s modulus being the delayed modulus, (3) the Mazars model with the initial Young’s modulus being the delayed modulus with the addition of the effects of micro-damage caused by irradiation, and (4) the Mazars model with the initial Young’s modulus being the instantaneous modulus, and with the addition of micro-damage and creep. Details regarding the methods used to obtain the fluence and the statistical errors associated with the VERA Monte Carlo Shift calculations are discussed in greater detail in this report. The results obtained from the four Grizzly models are also presented in this report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Integrating BISON into VERA-CS and Fuel Temperature Calculations

Approaches to integrating BISON into VERA-CS were identified as: 1) Coupled TIAMAT calculations in which BISON is solved simultaneously with MPACT and CTF via direct coupling, 2) Inline TIAMAT in which BISON calculations are performed as part of the same VERA-CS calculation sequence as coupled MPACT-CTF calculations but there is no feedback from the BISON calculations into the MPACT-CTF calculations, and 3) Stand-alone (“Offline”) BISON calculations in which desired data is parsed from the results of an already-completed MPACT-CTF VERA-CS analysis and BISON calculations are performed afterwards. The current maximum capabilities of these approaches in the latest VERA build are a single ramp-to-power for Coupled TIAMAT, analysis of two subsequent operating cycles for a single pin in Inline TIAMAT, and BISON analysis for every fuel pin in a PWR quarter-core for multiple operating cycles in Stand-alone BISON.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

VERA Configure, Build, Test, and Install Quick Reference Guide

This document contains quick reference information on how to configure, build, test, and install VERA using the TriBITS CMake build system. The primary audience are developers of VERA that need to configure and build the software. The secondary audience are power users of VERA that will need to configure and build VERA on their own.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2015

VERA Benchmarking Results for Watts Bar Nuclear Plant Unit 1 Cycles 1-12

Among the most important accomplishments during Phase 1 of the Consortium for Advanced Simulation of Light Water Reactors (CASL) is the development and deployment of what is known as VERA-CS, the steady-state nuclear reactor core simulation capability of the Virtual Environment for Reactor Applications, VERA. With this product CASL now has the ability to perform and support analyses relating to common evolutions of operating commercial pressurized water reactors (PWRs), including startup testing, power escalation, fuel cycle depletion, and fuel assembly discharge, reinsert, and shuffling. VERA-CS is a high-fidelity, multi-physics engineering tool that utilizes modest high-performance computing (HPC) systems and engineering-scale clusters to simultaneously simulate the local fuel rod neutronics and coolant channel thermal-hydraulics over the life of the reactor. Though it is still under development, the results demonstrated in this report provide encouraging evidence that VERA-CS will become highly acclaimed for industrial, scientific, and educational applications alike.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Operational Reactor Model Demonstrations with VERA: Watts Bar Unit 1 Cycle 1 Zero Power Physics Tests (Revision 1)

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a collection of methods and software products known as VERA, the Virtual Environment for Reactor Applications. VERA will provide coupled, high fidelity nuclear reactor analysis capabilities in multiple areas of physics such as neutronics, thermal-hydraulics, fuel mechanics, and others. A key milestone for this endeavor is to validate VERA against measurements from operating nuclear power reactors. This document demonstrates the capabilities of VERA to accurately simulate the initial startup physics tests of Watts Bar Nuclear Power Station, Unit 1 (WBN1), demonstrating the many achievements that have been made by the CASL partners over the past year.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

VERA-CS Modeling and Simulation of PWR Main Steam Line Break Core Response to DNB

The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time step of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.

CTF↗
DOE OSTI2018

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

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing multiphysics core-simulator software for light water reactors (LWRs) known as VERA-CS in order to improve the state of the art in modeling and simulation of challenge problems that are limiting to the nuclear industry. One of these challenge problems includes fuel rod crud deposition, which can lead to crud-induced power shift (CIPS) and crud-induced localized corrosion (CILC). This paper documents work that was performed to develop a preliminary CILC-modeling capability in VERA-CS in support of the crud challenge problem. The CILC capabilities were developed by coupling VERA-CS to the CASL-developed Cicada code, which provides 1D and 3D clad conduction and oxide growth modeling tools, as well as coupling to the CASL-developed MAMBA code, which is used for modeling clad crud deposition. An approach called rod thermal-hydraulic reconstruction (ROTHCON) was developed and integrated into VERA-CS. This allows the modeler to capture spacer-grid turbulence and heat transfer effects in the CTF subchannel code so that the spatial resolution of crud and oxide rod surface growth could be better resolved. After implementing these capabilities, several assessments were performed to ensure that the capabilities function as expected, and a pin-resolved quarter-core simulation was run as a demonstration.

Salko Jr, Robert↗
DOE OSTI2019

Development of Molten Salt Reactor Modeling and Simulation Capabilities in VERA

In recent years, the nuclear community has seen a resurgence of interest in molten salt reactors (MSRs). Because of their molten, flowing fuel, MSRs present several new challenges not present in many other reactors, necessitating tightly coupled multiphysics modeling and simulation for accurate analysis. The Virtual Environment for Reactor Applications (VERA) has been developed to provide a high-fidelity multiphysics framework for reactor analysis. Though it was originally developed for light water reactor applications, VERA’s multiphysics capabilities make for a natural extension to MSR analysis. Recent efforts have focused on capturing the effects of moving fuel. A lumped depletion capability has been added that accounts for the moving and mixing of the fuel, as well as time spent outside the active core. A general multi-phase species transport module has also been added to track the movement and evolution of salt constituents as the salt flows through the core and around the primary loop. This paper presents a discussion of the new MSR capabilities in VERA, along with results for the lumped depletion and species transport capabilities for the Molten Salt Reactor Experiment, demonstrating VERA’s effectiveness in multiphysics simulations of MSRs.

Graham, Aaron↗
DOE OSTI2017

Demonstration of Load-Follow Operations with Standalone BISON through VERA

As part of the Consortium for Advanced Simulation of Light Water Reactors (CASL), the Virtual Environment for Reactor Applications (VERA) is being developed to provide high-fidelity multiphysics simulations of nuclear reactor cores. VERA contains a number of single physics packages that are coupled together. MPACT is the primary deterministic neutron transport solver, CTF provides thermal hydraulics solutions, and BISON - which is built off of the MOOSE framework - is used for fuel performance simulations. Over the past year, a standalone BISON capability has been developed using a one-way coupling between MPACT/CTF coupling calculations and BISON. This has allowed CASL contributors to gain insight into the fuel performance characteristics while a more tightly coupled methodology between MPACT, CTF, and BISON has been under development. The standalone capability has been used to simulate Watts Bar Unit 1 (WBN1) Cycles 1-3 and to screen for quantities of interest to pellet-clad interaction in Cycles 6-7. The simulation and screening results are presented here. CASL recently partnered with Exelon and the University of Illinois at Urbana Champaign (UIUC) to begin simulating some of Exelon's plants, with particular focus on load-follow operations. In such operations, the power output from the plant will vary based on the demand or anticipated demand at the time. For example, demand is much lower late at night than during the day when people are awake, so it operating the plant at lower power in the evenings could be advantageous. Historically, nuclear power generation has been among the cheapest sources of energy, so most plants in the United States have been operated at full power except for outages and various small scale events. However, in the current economic climate, priorities have changed. A number of questions arise when load-follow operations are considered, particularly with respect to VERA's tools and capabilities. Can the coupled neutronics and thermal hydraulics simulations with MPACT and CTF perform well without substantial convergence issues? Can BISON handle the somewhat rapid power changes present with load-follow? Will clad hoop stresses be alarmingly high as a result of these power changes? This work addresses these questions and demonstrates operations to assess feasibility and robustness before Exelon and UIUC proceed with their simulations using VERA. To demonstrate load-follow capability, two test problems were developed. The first is a single rod that experiences nominal operations during Cycle 1, and then it simulates 18 months of representative load-follow operations in Cycle 2. This single-rod case was used to test the robustness of MPACT, CTF, and BISON on a typical rod. The second problem is a quarter-core model based on WBN1, Cycle 3, with a single month of load-follow operations inserted into the middle of the cycle. This provides better understanding of the stresses to the clad during the numerous power changes. Sections of this article are adapted from the original, full technical report on this capability.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗