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DOE OSTI2022

VERA neutronics high-fidelity benchmark for a modern PWR core design

Nearly 20 years since the origin of the infamous 'Kord Smith Challenge', and after 10 years and hundreds of millions of dollars invested by the US Department of Energy into the development of the Virtual Environment for Reactor Applications (VERA), the capability to accurately simulate a modern pressurized water reactor (PWR) fuel cycle at the fuel rod level with high-fidelity Monte Carlo (MC) stochastic transport methods still evades us. The deterministic methods in VERA make approximations that allow it to be successfully benchmarked against measured data from hundreds of PWR fuel cycles, but none of these benchmarks provided information at the local fuel rod level or allowed quantification of these approximations' impact for modern PWR reload designs. Using MPACT to establish the estimated isotopic distributions and thermal hydraulic conditions in a recent and challenging reactor design, a 2D quarter-core benchmark problem was established and solved consistently with both MPACT, the 51-energy group deterministic transport solver in VERA, and Shift, the continuous-energy MC transport solver in VERA. Impacts of energy groups, ray spacing, and scattering treatment were evaluated. Though MPACT has been previously shown to perform well compared to MC methods for initial Cycle 1 cores with only fresh fuel, this is the first benchmark of MPACT for a challenging modern reload core with depleted fuel and burnable absorbers. This work establishes additional credibility for the methods in MPACT and highlights the efficiency of these methods relative to those in use by MC-based tools. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity multiphysics simulations of light water nuclear reactors. To accomplish this, CASL is developing the Virtual Environment for Reactor Applications (VERA), which is a suite of code packages for thermal hydraulics, neutron transport, fuel performance, and coolant chemistry. As VERA continues to grow and expand, there has been an increased focus on incorporating fuel performance analysis methods. One of the primary goals of CASL is to estimate local cladding failure probability through pellet-clad interaction, which consists of both pellet-clad mechanical interaction (PCMI) and stress corrosion cracking. Estimating clad failure is important to preventing release of fission products to the primary system and accurate estimates could prove useful in establishing less conservative power ramp rates or when considering load-follow operations.While this capability is being pursued through several different approaches, the procedure presented in this article focuses on running independent fuel performance calculations with BISON using a file-based one-way coupling based on multicycle output data from high fidelity, pin-resolved coupled neutron transport–thermal hydraulics simulations. This type of approach is consistent with traditional fuel performance analysis methods, which are typically separate from core simulation analyses. A more tightly coupled approach is currently being developed, which is the ultimate target application in CASL.Recent work simulating 12 cycles of Watts Bar Unit 1 with VERA core simulator are capitalized upon, and quarter-core BISON results for parameters of interest to PCMI (maximum centerline fuel temperature, maximum clad hoop stress, and minimum gap size) are presented for Cycles 1–3. In conclusion, based on these results, this capability demonstrates its value and how it could be used as a screening tool for gathering insight into PCMI, singling out limiting rods for further, more detailed analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

A High-Fidelity Model of the Peach Bottom 2 Turbine-Trip Benchmark Using VERA

This work presents a high-fidelity simulation of the Peach Bottom turbine trip (PBTT) benchmark using the Virtual Environment for Reactor Applications (VERA), a multiphysics reactor modeling tool developed by the U.S. Department of Energy’s Consortium for Advanced Simulation of Light Water Reactors energy innovation hub. The PBTT benchmark, based on a 1977 transient event at the end of cycle 2 in a General Electric Type-4 boiling water reactor (BWR), is a critical test case for validating core physics models with thermal feedback during rapid reactivity events. VERA was employed to perform end-to-end, pin-resolved simulations from conditions at the beginning of cycle 1 through the turbine-trip transient, incorporating detailed neutron transport, fuel depletion, and subchannel thermal hydraulics. The simulation reproduced key benchmark observables with high accuracy: the peak power excursion occurred at 0.75 s, matching the scram time and closely aligning with the benchmark average of 0.742 s; the simulated maximum power spike was approximately 7600 MW, which is within 3% of the benchmark average of 7400 MW; and void-collapse dynamics were consistent with benchmark expectations. Reactivity predictions during cycles 1 and 2 remained within 1500 pcm and 400 pcm of criticality, respectively. These results confirm VERA’s ability to model complex coupled neutronic and thermal hydraulic behavior in a BWR turbine-trip transient, which will support its use in future studies of modeling dryout, fuel performance, and uncertainty quantification for transients of this type.

BWR↗

DOE OSTI2025

RTN-095: The Vera C. Rubin Observatory Data Preview 1

We present Rubin Data Preview 1 (DP1), the first release of data from the NSF-DOE Vera C. Rubin Observatory, consisting of raw and calibrated single-epoch images, coadds, difference images, detection catalogs, and other derived data products. DP1 is based on 1792 science-grade 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 during the first on-sky commissioning campaign in late 2024. DP1 covers a total of ~15 sq. deg. over seven roughly equally-sized non-contiguous fields, each independently observed in six broad photometric bands, ugrizy, spanning a range of stellar densities and latitudes and overlapping with external reference datasets. The median image quality across all bands, measured by the FWHM of the point-spread function, is approximately 1.13 arcseconds, with the sharpest images reaching about 0.65 arcseconds. DP1 contains approximately 2.3 million distinct astrophysical objects, of which 1.6 million are extended in at least one band, and 431 solar system objects, of which 93 are new discoveries. DP1 is approximately 3.5 TB in size and available to Rubin 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 across all four LSST themes, providing a valuable opportunity to engage with Rubin data ahead of the start of full operations in late 2025.

79 ASTRONOMY AND ASTROPHYSICS↗

DOE OSTI2020

Boron and Lithium in Calcium Sulfate Veins: Tracking Precipitation of Diagenetic Materials in Vera Rubin Ridge, Gale Crater

The NASA Curiosity rover's ChemCam instrument suite has detected boron in calcium-sulfate-filled fractures throughout the sedimentary strata of Gale crater including Vera Rubin ridge. The presence of elevated B concentration provides insights into Martian subsurface aqueous processes. Here, we extend the data set of B in Ca-sulfate veins across Gale crater, comparing the detection frequency and relative abundances with Li. We report 33 new detections of B within veins analyzed between Sols 1548 and 2311 where detections increase in Pettegrove Point and Jura members, which form Vera Rubin ridge. The presence of B and Li in the Ca-sulfate veins is possibly due to dissolution of preexisting B in clays of the bedrock by acids or neutral water and redistribution of the elements into the veins. Elevated frequency of B detection in veins of Gale crater correlates with presence of dehydration features such as desiccation cracks, altered clay minerals and detections of evaporites such as Mg-sulfates and chloride salts in the host rocks. The increased observations of B also coincide with decreased Li concentration in the veins (average Li concentration of veins drops by ~15 ppm). Boron and Li have varying solubilities, and Li does not form salts as readily upon dehydration as B, causing it to remain in the solution. So the weak negative correlation between B and Li may reflect the crystallization sequence during dehydration on Vera Rubin ridge.

58 GEOSCIENCES↗

DOE OSTI2020

Development of the MPACT 69-group Library for Magnox Reactor Analysis using CASL VERA

The Consortium for Advanced Simulation of Light Water Reactors (CASL) has developed the CASL toolset, Virtual Environment for Reactor Analysis (VERA), for pressurized water reactor (PWR) analysis. Recently the CASL VERA was improved for Magnox reactor analysis, which required the development of a new cross section library and new geometrical and thermal feedback capabilities for graphite-moderated Magnoxreactors. The MPACT neutronics module of the CASL core simulator is a 3D whole core transport code, which requires a new cross section library with a different energy group structure due to the different neutronic characteristics of Magnox compared with PWR. A new 69-group structure was developed based on the MPACT 51-group structure to have more thermal energy groups and to be a subset of the SCALE 252-group structure. The ENDF/B-VII.1 MPACT 69-group library was developed for Magnox reactor analysis using the SCALE/AMPX and VERA-XSTools for which a super-homogenization method was applied, and transport cross sections were generated for graphite using a neutron leakage conservation method. Benchmark results show that new MPACT 69-group library works reasonably well for Magnox reactor analysis.

Kim, Kang Seog↗

DOE OSTI2017

VERA Core Simulator methodology for pressurized water reactor cycle depletion

This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclide transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. Lastly, these results provide confidence in VERA-CS’s capability to perform high-fidelity calculations for practical PWR reactor problems.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Standalone BISON Through VERA: User's Manual

Activities to incorporate fuel performance capabilities into the Virtual Environment for Reactor Applications (VERA) are receiving increasing attention. 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 Core Simulator (VERA-CS) 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 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, 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, for this work, it uses a 2D azimuthally symmetric (R-Z) smeared-pellet model.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Watts Bar Unit 2 Startup Results with VERA

CASL members TVA, Westinghouse, and Oak Ridge National Laboratory have successfully completed a detailed simulation of the initial startup of Watts Bar Nuclear Unit 2 (WBN2) using the advanced reactor simulation tools known as VERA. WBN2 is the first commercial power reactor to join the nation’s electrical grid in over two decades, and the modern core design and availability of data make it an excellent benchmark for CASL. Calculations were performed three months prior to the startup, and in the first blind application of VERA to a new reactor, predicted criticality and physics parameters very close to those later measured by TVA. Subsequent calculations with the latest version of VERA and using exact measurement conditions improved the results even further.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Summary of VERA Core Simulator Performance Improvements

This report describes the performance improvements made to the VERA Core Simulator (VERA CS) during FY2016. At the beginning of the year, a timing study was performed to identify areas of the code where the run-times could be improved. This study identified six broad areas for improvement: • Cross section lookup and CMFD stabilization, • CTF parallel partitioning, • MOC methodology, • Subgroup methodology, • MOC parallel partitioning, and • CMFD Krylov solution techniques. The goal of this work was to reduce the VERA-CS run-times so a complete cycle depletion could be run with 1,000 cores overnight (i.e., the “1,000 core depletion” criterion). This goal was set by our industry partners so they could achieve reasonable turnaround on medium-sized computer clusters. The results presented in this report show a code speed-up by a factor of 5, which successfully satisfies the goal set by industry.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2015

Validate VERA-CS using the industry standard BEAVRS Cycle Depletion Benchmark Problem

VERA-CS is the core simulator component of VERA which includes 3 main components; MPACT for neutron transport/reactor physics, COBRA-TF for thermal-hydraulics, and ORIGEN for time dependent isotopics. These three components have each undergone individual validation but there is also a need to validate the coupled codes. This work aims to build on the validation basis of VERA-CS using the BEAVRS Benchmark which was released by MIT in 2013.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2014

VERAOUT - VERA HDF5 Output Specification (Revision 1)

This document describes the revised specification (Revision 1) for the format of the common VERA Output file (VERAOUT). VERAOUT is a binary HDF5 file that all of the physics codes in VERA-CS produce. Having all physics codes write to a single file simplifies post-processing and data comparison. The edits and format written to the VERAOUT file is described below. This specification covers common scalar and array data typically produced by neutronic and fuel performance codes. This document will be periodically revised to add additional data as more physics codes and/or more data is added to VERA-CS. A version number is written to the HDF file to keep track of different versions of the VERAOUT so that backwards compatibility may be maintained. A sample VERAOUT reader has been written to serve as an example for users wishing to do data postprocessing. Information on this program is included below.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2014

VERA-CS Validation Plan

The VERA core simulator, VERA-CS, provides the capability for steady-state reactor core simulation for operating nuclear power plant (PWR) conditions across multiple fuel cycles. This capability parallels that of industry core simulators which are typically licensed by the NRC for core design, core surveillance/monitoring activities, and safety-related calculations that support accident analyses. The licensing basis for these codes for these applications is a rigorous process of validating the software against measured data and quantifying the resulting biases and uncertainties. This document provides an initial plan for performing similar activities with VERA-CS, in a manner consistent with benchmarking activities performed in the nuclear industry.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2014

Coupled Neutronics Thermal-Hydraulic Solution of a Full-Core PWR Using VERA-CS

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a core simulator called VERA-CS to model operating PWR reactors with high resolution. This paper describes how the development of VERA-CS is being driven by a set of progression benchmark problems that specify the delivery of useful capability in discrete steps. As part of this development, this paper will describe the current capability of VERA-CS to perform a multiphysics simulation of an operating PWR at Hot Full Power (HFP) conditions using a set of existing computer codes coupled together in a novel method. Results for several single-assembly cases are shown that demonstrate coupling for different boron concentrations and power levels. Finally, high-resolution results are shown for a full-core PWR reactor modeled in quarter-symmetry.

Core Simulator↗

DOE OSTI2015

Assessment of Pellet-Clad Interaction Indicators in Watts Bar Unit 1 using the VERA Framework

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity, multiphysics simulations of light water reactors (LWRs) by coupling a variety of codes within the Virtual Environment for Reactor Analysis (VERA). One of the primary goals of CASL is to predict local cladding failure through pellet-clad interaction (PCI). This capability is currently being pursued through several different approaches, such as with Tiamat, which is a simulation tool within VERA that more tightly couples the MPACT neutron transport solver, the CTF thermal hydraulics solver, and the MOOSE-based Bison-CASL fuel performance code. However, the process in this paper focuses on running fuel performance calculations with Bison-CASL to predict PCI using the multicycle output data from coupled neutron transport/thermal hydraulics simulations. In recent work within CASL, Watts Bar Unit 1 has been simulated over 12 cycles using the VERA core simulator capability based on MPACT and CTF. Using the output from these simulations, Bison-CASL results can be obtained without rerunning all 12 cycles, while providing some insight into PCI indicators. Multi-cycle Bison-CASL results are presented and compared against results from the FRAPCON fuel performance code. There are several quantities of interest in considering PCI and subsequent fuel rod failures, such as the clad hoop stress and maximum centerline fuel temperature, particularly as a function of time. Bison-CASL performs single-rod simulations using representative power and temperature distributions, providing high-resolution results for these and a number of other quantities. This will assist in identifying fuels rods as potential failure locations for use in further analyses.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Uncertainty quantification and sensitivity analysis with CASL Core Simulator VERA-CS

Uncertainty quantification and sensitivity analysis are important for nuclear reactor safety design and analysis. A 2x2 fuel assembly core design was developed and simulated by the Virtual Environment for Reactor Applications, Core Simulator (VERA-CS) coupled neutronics and thermal-hydraulics code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). An approach to uncertainty quantification and sensitivity analysis with VERA-CS was developed and a new toolkit was created to perform uncertainty quantification and sensitivity analysis with fourteen uncertain input parameters. Furthermore, the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel center-line temperature, and maximum outer clad surface temperature were chosen as the selected figures of merit. Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis and coolant inlet temperature was consistently the most influential parameter. We used parameters as inputs to the critical heat flux calculation with the W-3 correlation were shown to be the most influential on the MDNBR, maximum fuel center-line temperature, and maximum outer clad surface temperature.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by a system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2013

Nanocrystalline spinel ferrite (MFe{sub 2}O{sub 4}, M = Ni, Co, Mn, Mg, Zn) powders prepared by a simple aloe vera plant-extracted solution hydrothermal route

Graphical abstract: This figure shows the specific magnetization curves of the as-prepared MFe{sub 2}O{sub 4} (M = Ni, Co, Mn, Mg, Zn) powders obtained from room temperature VSM measurement. These curves are typical for a soft magnetic material and indicate hysteresis ferromagnetism in the field ranges of ±500 Oe, ±1000 Oe, and ±2000 Oe for the CoFe{sub 2}O{sub 4}, MgFe{sub 2}O{sub 4} and MnFe{sub 2}O{sub 4} respectively, whereas the samples of NiFe{sub 2}O{sub 4} and ZnFe{sub 2}O{sub 4} show a superparamagnetic behavior. Highlights: ► Nanocrystalline MFe{sub 2}O{sub 4} powders were synthesized by a novel hydrothermal method. ► Metal acetylacetonates and aloe vera plant-extracted solution are used. ► This biosynthetic route is very simple and provides high-yield oxide nanomaterials. ► XRD and TEM results indicate that the prepared samples have only spinel structure. ► The maximum M{sub s} of 68.9 emu/g at 10 kOe were observed for the samples of MnFe{sub 2}O{sub 4}. - Abstract: Nanocrystalline spinel ferrite MFe{sub 2}O{sub 4} (M = Ni, Co, Mn, Mg, Zn) powders were synthesized by a novel hydrothermal method using Fe(acac){sub 3}, M(acac){sub 3} (M = Ni, Co, Mn, Mg, Zn) and aloe vera plant extracted solution. The X-ray diffraction and selected-area electron diffraction results indicate that the synthesized nanocrystalline have only spinel structure without the presence of other phase impurities. The crystal structure and morphology of the spinel ferrite powders, as revealed by TEM, show that the NiFe{sub 2}O{sub 4} and CoFe{sub 2}O{sub 4} samples contain nanoparticles, whereas the MnFe{sub 2}O{sub 4} and MgFe{sub 2}O{sub 4} samples consist of many nanoplatelets and nanoparticles. Interestingly, the ZnFe{sub 2}O{sub 4} sample contains plate-like structure of networked nanocrystalline particles. Room temperature magnetization results show a ferromagnetic behavior of the CoFe{sub 2}O{sub 4}, MnFe{sub 2}O{sub 4} and MgFe{sub 2}O{sub 4} samples, whereas the samples of NiFe{sub 2}O{sub 4} and ZnFe{sub 2}O{sub 4} exhibit a superparamagnetic behavior.

36 MATERIALS SCIENCE↗