MC21/CTF and VERA multiphysics solutions to VERA core physics benchmark progression problems 6 and 7
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Kelly, III, Daniel J.↗
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The Virtual Environment for Reactor Applications (VERA): Design and architecture
VERA, the Virtual Environment for Reactor Applications, is the system of physics capabilities being developed and deployed by the Consortium for Advanced Simulation of Light Water Reactors (CASL). CASL was established for the modeling and simulation of commercial nuclear reactors. VERA consists of integrating and interfacing software together with a suite of physics components adapted and/or refactored to simulate relevant physical phenomena in a coupled manner. VERA also includes the software development environment and computational infrastructure needed for these components to be effectively used. We describe the architecture of VERA from both software and numerical perspectives, along with the goals and constraints that drove major design decisions, and their implications. We explain why VERA is an environment rather than a framework or toolkit, why these distinctions are relevant (particularly for coupled physics applications), and provide an overview of results that demonstrate the use of VERA tools for a variety of challenging applications within the nuclear industry.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
VERA and VERA-EDU 3.5 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 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. MPACT uses the Method of Characteristics transport approach for 2D problems. For 3D problems, MPACT uses the 2D/1D method which uses 2D MOC in a radial plane and diffusion or SPn in the axial direction. MPACT includes integrated cross section capabilities that provide problem-specific cross sections generated using the subgroup methodology. The code can be executed both 2D and 3D problems in parallel to reduce overall run time. A thermal-hydraulics capability is provided with CTF (an updated version of COBRA-TF) that allows thermal-hydraulics analyses for single and multiple assemblies using the simplified VERA common input. This distribution also includes coupled neutronics/thermal-hydraulics capabilities to allow calculations using MPACT coupled with CTF. The VERA fuel rod performance component BISON calculates, on a 2D or 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. Input/Output capabilities include the VERA Common Input (VERAIn) script which converts the ASCII common input file to the intermediate XML used to drive all of the physics codes in the VERA Core Simulator (VERA-CS). VERA component codes either input the VERA XML format directly, or provide a preprocessor which can convert the XML into native input. VERAView is an interactive graphical interface for the visualization and engineering analyses of output data from VERA. The python-based software is easy to install and intuitive to use, and provides instantaneous 2D and 3D images, 1D plots, and alpha-numeric data from VERA multi-physics simulations. Testing within CASL has focused primarily on Westinghouse four-loop reactor geometries and conditions with example problems included in the distribution.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Determination of the spent fuel decay heat with the VERA core simulator
VERA core simulator capabilities to predict spent fuel decay heat have been explored. The comparison with the Serpent2 and NEWT/TRITON sequence from the SCALE package, performed on the NPP Krško fuel test cases, has enabled independent verification of the successful implementation of the VERA depletion/decay calculation sequence applied on the extended nuclide set. Since the VERA is a 3-D core simulator, the focus in the rest of the paper was on the investigation of the 3-D effects and non-linearity of the soluble boron concentration averaging process, which are difficult to assess with most of available computational tools. A difference in the decay heat prediction between the 3-D and 2-D model approaches was analyzed. In addition, the impact of the Inconel grids on the decay heat was determined. Averaging the concentration of soluble boron over the fuel irradiation period has been found to be a viable approach, since it results in slight decay heat overprediction, which is considered conservative for most applications. Finally, demonstration on the Watts Bar Unit 1 cycle 1 depletion has shown formidable VERA capabilities to accurately predict spent fuel decay heat.
Kromar, Marjan↗
Simulation of the BEAVRS benchmark using VERA
The Virtual Environment for Reactor Applications (VERA) is being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL). Here, VERA is used to perform the Benchmark for Evaluation and Validation of Reactor Simulations (BEAVRS), which provides two cycles’ worth of operating power history, along with a full, detailed description of the geometry, and measured data. Cycle 1 and 2 are simulated with VERA and the results are compared to the measured zero power physics tests, critical boron concentration, and flux maps.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
VERA BWR progression problems
During the first phase of the Consortium for Advanced Simulation of Light Water Reactors (CASL) program, the Virtual Environment for Reactor Applications (VERA) was developed with a focus on capabilities for high-fidelity, multiphysics simulation of pressurized water reactors (PWRs). During this development effort, a set of progression problems was created ranging from smaller pin cell calculations to larger 3D full-core calculations. These progression problems helped to guide the development of the software and served as benchmarks against which to test VERA. Since 2019, efforts have been made to extend the capabilities of VERA to model boiling water reactors (BWRs). Because BWR simulations come with many unique challenges, a set of BWR progression problems was developed to aid in this new effort. The BWR progression problems range from 2D lattice calculations to 3D mini-core problems, and reference neutronic solutions were computed using continuous-energy Monte Carlo codes. MPACT, one of the neutronics code in VERA, was benchmarked using the BWR progression problems. The code is capable of computing solutions to all problems. The eigenvalues computed by MPACT agree well with the Monte Carlo reference solutions. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
VERA Core Physics Benchmark Progression Problem Results for VERA 3.6
This report contains a comparison of results obtained with VERA 3.6 Release Candidate 0 for the Core Physics Benchmark Progression Problems 1 through 9 to both computational tools and measured data from Watts Bar Unit 1 (WBN1). (Ref. 1 and Ref. 2) These problems progress from single pin cell analyses to full core depletion calculations. VERA results obtained for Problems 1 through 5 are compared against results obtained using the SCALE 6.2 Beta code KENO-VI. Results for Problem 6 and 7 are compared against those obtained using a coupled neutronic and thermal-hydraulic combination of the KAPL software MC21 and the CASL tool CTF. (Ref. 3, Ref. 4 and Ref. 5) VERA results from problems 5 and 9 are compared to data from WBN1. For all of the analyses contained in this report, VERA provided similar results to other state-of-the-art software programs and real reactor data. This indicates that the program is capable of modeling a full-scale PWR in a more discretized and computationally efficient manner than any other reactor physics program. Since the descriptions of the Core Physics Benchmark Progression Problems and the logic behind their selection is discussed in-depth elsewhere it is not covered in the scope of this document. Additionally, in-depth discussions of the software programs used in this analysis are not provided for the same rationale.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
MC21/CTF and VERA-CS Mutiphysics Solutions to VERA Core Physics Benchmark Problems 6 and 7
The continuous energy Monte Carlo neutron transport code, MC21, was coupled to the CTF subchannel thermal-hydraulics code using a combination of CASL tools and in-house Python scripts. An MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 6 demonstrated excellent agreement with MC21/COBRA-IE and VERA solutions. The MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 7, Watts Bar Unit 1 at beginning of cycle (BOC) hot full power (HFP) equilibrium xenon conditions, is the first published coupled Monte Carlo neutronics / subchannel T-H solution for this problem. MC21/CTF predicted a critical boron concentration of 854.5 ppm, yielding a critical eigenvalue of 0.99994 ± 6.8E-6 (95% CI). Excellent agreement with a VERA solution of Problem 7 was also demonstrated for integral and local power and temperature parameters.
Aviles, Brian↗
Coupled fuel performance calculations in VERA and demonstration on Watts Bar unit 1, cycle 1
As the core simulator capabilities in the Virtual Environment for Reactor Applications (VERA) have become more mature and stable, increased attention has been focused on coupling Bison to provide fuel performance simulations. This technique has been a very important driver for the pellet-clad interaction challenge problem being addressed by the Consortium for Advanced Simulation of Light Water Reactors (CASL). Here, two coupling approaches are demonstrated on quarter core problems based on Watts Bar Nuclear Plant Unit 1, Cycle 1: (1) an inline approach in which a one-way coupling is used between neutronics/thermal hydraulics (through MPACT/CTF) and fuel performance (through Bison) but no fuel temperature information is passed back to MPACT/CTF, and (2) a two-way approach (coupled) in which the fuel temperature is passed from Bison to MPACT/CTF. In both approaches, power and temperature distributions from MPACT/CTF are used to inform the Bison simulations for each rod in the core.The demonstrations presented here are the first integrated fuel performance simulations in VERA, which opens many possibilities for future work, including applications to accident-tolerant fuel efforts and transient simulations, which are of critical importance to CASL. These demonstrations also highlight the potential to move away from the current Bison-informed fuel temperature lookup table approach, which is the default in MPACT/CTF simulations, if performance improvements are made in the near future.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Application of regression, variance, and density based global sensitivity methods to integrated VERA-CS and BISON simulations [Application of regression, variance, and density based global sensitivity methods to integrated VERA-CS and BISON simulations]
Uncertainty quantification (UQ) and sensitivity analyses (SA) are performed for coupled simulations between VERA-CS, a coupled pin resolved neutron transport and subchannel thermal hydraulics code, and the fuel performance code BISON. The interface between VERA-CS and BISON is performed in a multiphysics environment known as the LOCA Toolkit for U.S. light water reactors (LOTUS) currently under development at Idaho National Laboratory (INL). A focus is placed on using a variety of SA measures, including two regression based (Pearson and Spearman), one variance based (Sobol indices), and three moment independent measures (Delta moment independent measures with L 1 , L 2 , and L ∞ norms). The problem under inspection is a single assembly depletion case for three fuel cycles. The figures of merit are the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel centerline temperature (MFCT), and gap conductance at peak power (GCPP). SA results show MDNBR to be linear with consistent rankings throughout the fuel cycles. MFCT is linear, but with a change in rankings at the switch from open gap to closed gap models. GCPP is nonlinear at intermediate states that coincide with the onset of contact between fuel and cladding. Furthermore, these nonlinear states allow for the showcasing of higher order SA measures over first order methods.
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Integrating the PanDA Workload Management System with the Vera C. Rubin Observatory
The Vera C. Rubin Observatory will produce an unprecedented astronomical data set for studies of the deep and dynamic universe. Its Legacy Survey of Space and Time (LSST) will image the entire southern sky every three to four days and produce tens of petabytes of raw image data and associated calibration data over the course of the experiment’s run. More than 20 terabytes of data must be stored every night, and annual campaigns to reprocess the entire dataset since the beginning of the survey will be conducted over ten years. The Production and Distributed Analysis (PanDA) system was evaluated by the Rubin Observatory Data Management team and selected to serve the Observatory’s needs due to its demonstrated scalability and flexibility over the years, for its Directed Acyclic Graph (DAG) support, its support for multi-site processing, and its highly scalable complex workflows via the intelligent Data Delivery Service (iDDS). PanDA is also being evaluated for prompt processing where data must be processed within 60 seconds after image capture. This paper will briefly describe the Rubin Data Management system and its Data Facilities (DFs). Finally, it will describe in depth the work performed in order to integrate the PanDA system with the Rubin Observatory to be able to run the Rubin Science Pipelines using PanDA.
79 ASTRONOMY AND ASTROPHYSICS↗
Dancoff-based Wigner-Seitz approximation for the subgroup resonance self-shielding in the VERA neutronic simulator MPACT
The MPACT neutronics module of the Virtual Environment for Reactor Analysis (VERA) has used the subgroup method for resonance self-shielding calculation, for which two-dimensional (2D) fixed-source transport calculations are performed using the method of characteristics for resonance energy groups. When considering thermal feedbacks, the subgroup calculation must be performed at each outer iteration. Therefore, the computing time for cross section processing is a significant burden for computational efficiency. The Dancoff-based Wigner-Seitz approximation (DWA) capability has been implemented into MPACT in conjunction with the subgroup method, which has been used in SCALE/XSProc since SCALE version 6.0 and has recently been called an equivalent Dancoff-factor cell (EDC) method. The issue of relatively large reactivity bias in DWA for the gadolinia rods was resolved by introducing multiple Dancoff factors. Benchmark results for the VERA pressurized and boiling water reactor benchmark suites show that the DWA capability would significantly enhance computational efficiency with comparable accuracy. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
MC21/COBRA-IE and VERA-CS multiphysics solutions to VERA core physics benchmark problem #6
Not Available
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
VERA-Workbench: A Unified Multi-Physics Interface for the VERA Suite of Tools
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73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Comparative Studies of Sensitivity Analysis for Coupled VERA-CS/FRAPCON and VERA-CS/BISON Simulations
The abstract is contained in the attached document.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
VERA Depletion Benchmarks by CASL VERA, SERPENT and McCARD with ENDF/B-VII.0 and VII.1
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Kim, Kang Seog↗
VERA-CS Verification & Validation Plan
This report summarizes the current status of VERA-CS Verification and Validation for PWR Core Follow operation and proposes a multi-phase plan for continuing VERA-CS V&V in FY17 and FY18. The proposed plan recognizes the hierarchical nature of a multi-physics code system such as VERA-CS and the importance of first achieving an acceptable level of V&V on each of the single physics codes before focusing on the V&V of the coupled physics solution. The report summarizes the V&V of each of the single physics codes systems currently used for core follow analysis (ie MPACT, CTF, Multigroup Cross Section Generation, and BISON / Fuel Temperature Tables) and proposes specific actions to achieve a uniformly acceptable level of V&V in FY17. The report also recognizes the ongoing development of other codes important for PWR Core Follow (e.g. TIAMAT, MAMBA3D) and proposes Phase II (FY18) VERA-CS V&V activities in which those codes will also reach an acceptable level of V&V. The report then summarizes the current status of VERA-CS multi-physics V&V for PWR Core Follow and the ongoing PWR Core Follow V&V activities for FY17. An automated procedure and output data format is proposed for standardizing the output for core follow calculations and automatically generating tables and figures for the VERA-CS Latex file. A set of acceptance metrics is also proposed for the evaluation and assessment of core follow results that would be used within the script to automatically flag any results which require further analysis or more detailed explanation prior to being added to the VERA-CS validation base. After the Automation Scripts have been completed and tested using BEAVRS, the VERA-CS plan proposes the Watts Bar cycle depletion cases should be performed with the new cross section library and be included in the first draft of the new VERA-CS manual for release at the end of PoR15. Also, within the constraints imposed by the proprietary nature of plant data, as many as possible of the FY17 AMA Plant Core Follow cases should also be included in the VERA-CS manual at the end of PoR15. After completion of the ongoing development of TIAMAT for fully coupled, full core calculations with VERA-CS / BISON 1.5D, and after the completion of the refactoring of MAMBA3D for CIPS analysis in FY17, selected cases from the VERA-CS validation based should be performed, beginning with the legacy cases of Watts Bar and BEAVRS in PoR16. Finally, as potential Phase III future work some additional considerations are identified for extending the VERA-CS V&V to other reactor types such as the BWR.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Ex-Core Radiation Transport Modeling with VERA User Manual
It is important to be able to accurately predict the neutron flux outside the immediate reactor core for a variety of safety and material analyses. Monte Carlo radiation transport calculations are required to produce these high–fidelity ex-core responses. The Virtual Environment for Reactor Applications (VERA) provides the automated capability to launch independent Shift fixed-source and eigenvalue Monte Carlo (MC) calculations for user-specified state points during a standard Virtual Environment for Reactor Applications (VERA) calculation. VERA couples MPACT with COBRA-TF (CTF) to Shift to perform ex-core tallies for multiple state points concurrently, with each component capable of parallel execution on independent processor domains. In these ex-core calculations, MPACT is coupled to CTF and performs the in-core depletion and heat transfer calculation, followed by a fixed-source Shift transport calculation including ex-core regions to produce ex-core responses. The fission source, fuel pin temperatures, moderator temperature and density, boron concentration, and fuel pin depleted isotopic compositions can be transferred to Shift from the MPACT calculation. Specifically, VERA performs fluence calculations in the core barrel outward to the end of the pressure vessel and detector response calculations in ex-core detectors. It also performs the requested tallies in any user-defined ex-core regions. VERA takes advantage of the General Geometry (GG) package in Shift. This gives VERA the flexibility to explicitly model features outside the core barrel, including detailed vessel models, detectors, and power plant details. A very limited set of experimental and numerical benchmarks is available for ex-core simulation comparison. The Consortium for Advanced Simulation of Light Water Reactors has developed a set of ex-core benchmark problems to include as part of the VERA verification and validation set of problems. The ex-core capability in VERA has been tested on small representative assembly problems, multi-assembly problems, as well as quarter-core and full-core problems. VERA View has also been extended to visualize these vessel fluence results from VERA. This manual serves to present a guide to VERA users about the methodology behind ex-core calculations and the details of input, output, and analysis of results from these calculations. Details in this version of the manual are based on features in VERA 4.0.1.
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