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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 955 records · Page 53
All sourcesNASA NTRSDOE OSTI
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

Molecular dynamics simulation of thermal transport in UO 2 containing uranium, oxygen, and fission-product defects

Uranium dioxide (UO 2 ) is the most commonly used fuel in light-water nuclear reactors and thermal conductivity controls the removal of heat produced by fission, thereby governing fuel temperature during normal and accident conditions. The use of fuel performance codes by the industry to predict operational behavior is widespread. A primary source of uncertainty in these codes is thermal conductivity, and optimized fuel utilization may be possible if existing empirical models are replaced with models that incorporate explicit thermal-conductivity-degradation mechanisms during fuel burn up. This approach is able to represent the degradation of thermal conductivity due to each individual defect type, rather than the overall burn-up measure typically used, which is not an accurate representation of the chemical or microstructure state of the fuel that actually governs thermal conductivity and other properties. To generate a mechanistic thermal conductivity model, molecular dynamics (MD) simulations of UO 2 thermal conductivity including representative uranium and oxygen defects and fission products are carried out. These calculations employ a standard Buckingham-type interatomic potential and a potential that combines the many-body embedded-atom-method potential with Morse-Buckingham pair potentials. Potential parameters for UO 2+x and ZrO 2 are developed for the latter potential. Physical insights from the resonant phonon-spin-scattering mechanism due to spins on the magnetic uranium ions are introduced into the treatment of the MD results, with the corresponding relaxation time derived from existing experimental data. High defect scattering is predicted for Xe atoms compared to that of La and Zr ions. Uranium defects reduce the thermal conductivity more than oxygen defects. For each defect and fission product, scattering parameters are derived for application in both a Callaway model and the corresponding high-temperature model typically used in fuel-performance codes. The model is validated by comparison to low-temperature experimental measurements on single-crystal hyperstoichiometric UO 2+x samples and high-temperature literature data. Furthermore, this work will enable more accurate fuel-performance simulations and will extend to new fuel types and operating conditions, all of which improve the fuel economics of nuclear energy and maintain high fuel reliability and safety.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Summary report on UO 2 thermal conductivity model refinement and assessment studies

Uranium dioxide (UO 2 ) is the most commonly used fuel in light water nuclear reactors and thermal conductivity controls the removal of heat produced by fission, therefore, governing fuel temperature during normal and accident conditions. The use of fuel performance codes by the industry to predict operational behavior is widespread. A primary source of uncertainty in these codes is thermal conductivity, and optimized fuel utilization may be possible if existing empirical models were replaced with models that incorporate explicit thermal conductivity degradation mechanisms during fuel burn-up. This approach is able to represent the degradation of thermal conductivity due to each individual defect type, rather than the overall burn-up measure typically used which is not an accurate representation of the chemical or microstructure state of the fuel that actually governs thermal conductivity and other properties. To generate a mechanistic thermal conductivity model, molecular dynamics (MD) simulations of UO 2 thermal conductivity including representative uranium and oxygen defects and fission products are carried out. These calculations employ a standard Buckingham type interatomic potential and a potential that combines the many-body embedded atom method potential with Morse-Buckingham pair potentials. Potential parameters for UO 2+x and ZrO 2 are developed for the latter potential. Physical insights from the resonant phonon-spin scattering mechanism due to spins on the magnetic uranium ions have been introduced into the treatment of the MD results, with the corresponding relaxation time derived from existing experimental data. High defect scattering is predicted for Xe atoms compared to that of La and Zr ions. Uranium defects reduce the thermal conductivity more than oxygen defects. For each defect and fission product, scattering parameters are derived for application in both a Callaway model and the corresponding high-temperature model typically used in fuel performance codes. The model is validated by comparison to low-temperature experimental measurements on single crystal hyper-stoichiometric UO 2+x samples and high-temperature literature data. Ongoing works include investigation of the effect of phase separation to UO 2 +U 4 O 9 on the low temperature thermal conductivity of UO 2+x , and modeling of thermal conductivity using the Green-Kubo method. Ultimately, this work will enable more accurate fuel performance simulations as well as extension to new fuel types and operating conditions, all of which improve the fuel economics of nuclear energy and maintain high fuel reliability and safety.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Umpire: Integration with Other Software Technologies

Umpire is an open-source C++ library that provides memory and resource management functions for use with heterogeneous memory spaces. The Umpire library is being developed to simplify coordination and sharing of limited memory resources for complex integrated applications developed in the ECP/ATDM and ASC programs, and other multi-physics codes that combine many physics packages and libraries. The overarching goal of Umpire is to provide a single application-facing API for memory management on all platforms by decoupling memory operations from platform-specific memory spaces. It provides operations for: memory space discovery, memory provisioning, allocation and transfers, and introspection. Umpire development began in the Center for Applied Scientific Computing at Lawrence Livermore National Laboratory (LLNL). Umpire is funded primarily by LLNL ASC and ATMD portfolios and receives a small amount of support as part of ECP ST Project 2.3.1.06 STMP08-ISC. The source code is hosted and developed on GitHub. In this report, we explore possible integrations with other memory hierarchy management software. We examine both projects supported by ECP, as well as third-party and vendor provided tools.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2019

The HANDE-QMC Project: Open-Source Stochastic Quantum Chemistry from the Ground State Up

Building on the success of Quantum Monte Carlo techniques such as diffusion Monte Carlo, alternative stochastic approaches to solve electronic structure problems have emerged over the past decade. The full configuration interaction quantum Monte Carlo (FCIQMC) method allows one to systematically approach the exact solution of such problems, for cases where very high accuracy is desired. The introduction of FCIQMC has subsequently led to the development of coupled cluster Monte Carlo (CCMC) and density matrix quantum Monte Carlo (DMQMC), allowing stochastic sampling of the coupled cluster wave function and the exact thermal density matrix, respectively. In this Article, we describe the HANDE-QMC code, an open-source implementation of FCIQMC, CCMC and DMQMC, including initiator and semistochastic adaptations. We describe our code and demonstrate its use on three example systems; a molecule (nitric oxide), a model solid (the uniform electron gas), and a real solid (diamond). An illustrative tutorial is also included.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

SU-F-T-54: Determination of the AAPM TG-43 Brachytherapy Dosimetry Parameters for A New Titanium-Encapsulated Yb-169 Source by Monte Carlo Calculations

Purpose: To determine the AAPM TG-43 brachytherapy dosimetry parameters of a new titanium-encapsulated Yb-169 source designed to maximize the dose enhancement during gold nanoparticle-aided radiation therapy (GNRT). Methods: An existing Monte Carlo (MC) model of the titanium-encapsulated Yb-169 source, which was described in the current investigators’ published MC optimization study, was modified based on the source manufacturer’s detailed specifications, resulting in an accurate model of the titanium-encapsulated Yb-169 source that was actually manufactured. MC calculations were then performed using the MCNP5 code system and the modified source model, in order to obtain a complete set of the AAPM TG-43 parameters for the new Yb-169 source. Results: The MC-calculated dose rate constant for the new titanium-encapsulated Yb-169 source was 1.05 ± 0.03 cGy per hr U, indicating about 10% decrease from the values reported for the conventional stainless steel-encapsulated Yb-169 sources. The source anisotropy and radial dose function for the new source were found similar to those reported for the conventional Yb-169 sources. Conclusion: In this study, the AAPM TG-43 brachytherapy dosimetry parameters of a new titanium-encapsulated Yb-169 source were determined by MC calculations. The current results suggested that the use of titanium, instead of stainless steel, to encapsulate the Yb-169 core would not lead to any major change in the dosimetric characteristics of the Yb-169 source, while it would allow more low energy photons being transmitted through the source filter thereby leading to an increased dose enhancement during GNRT. Supported by DOD/PCRP grant W81XWH-12-1-0198 This investigation was supported by DOD/PCRP grant W81XWH-12-1- 0198.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2014

MCNP6 Elevation Scaling of Cosmic Ray Backgrounds - Paper 29

The galactic cosmic-ray (GCR) source option was implemented in the all-energy, all-particle transport code MCNP62 in 2010. This source option along with other newly added features to MCNP6 have been used to produce increasingly accurate global cosmic and terrestrial background flux data files (background.dat).3 This data file is produced on a grid and is read and sampled by MCNP6 whenever a user invokes the background source option. The most recent adaptation of this data file, Version 4, will be completed in 2014. This paper reports on current progress toward enhancements that have been made to the modeling and simulation of these background spectra, in particular the elevation scaling used to produce more accurate spectra at off-grid locations. Cosmic radiation continuously bombards Earth with various particles, such as protons, alpha particles, and heavier nuclei, some of which are deflected by the Earth's shielding magnetic field. Particles that carry sufficient momentum can overcome the deflection and penetrate into the atmosphere. The sufficient momentum is dependent on the terrestrial coordinates due to the shape of the Earth's magnetic field and the Lorentz force's proportionality to the sine of the angle between the velocity vector of the incoming particle and the magnetic field direction. As the particles propagate through the atmosphere, collisions with atmospheric molecules generate secondary particles such as neutrons, protons, photons, muons, pions, and other exotic particles. These secondary particles often have sufficient energy to undergo additional nuclear interactions, and so on, forming what is known as a cascade shower. The particles in the cascade shower are constantly undergoing energy losses due to interactions, especially charged particles whose paths are bent into helical trajectories around magnetic field lines. These energy losses make the integral flux of the cosmic background at the ground level highly dependent on the atmospheric depth traveled by the incident particle and its progeny, and therefore dependent on the elevation of the ground where the cascade deposits its energy. For relatively small depth changes of approximately 100 g/cm{sup 2} or less, the ground-level cosmic flux follows an exponential model. The tabulation of background particle fluxes on the surface of the earth is important for a variety of reasons, one of which is the design of nuclear material detection systems. This method is also useful in producing user-specified effective attenuation lengths for particles. Attenuation lengths are an essential parameter used to compare particle spectra from different locations, elevations, and energy ranges. Traditionally these attenuation lengths are inferred from measured data. Now with the ability to simulate an accurate galactic cosmic ray source and transport it and its shower through a modeled atmosphere, these attenuation lengths can be calculated for any set of conditions. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

Development of a BlueCRAB/MELCOR Framework for Supporting Realistic Mechanistic Source Term Calculations in Microreactors

Efforts are currently underway to deploy microreactor modeling and simulation tools to better support vendors and regulatory authorities in submitting and reviewing licensing applications. In particular, the Nuclear Regulatory Commission is expected to rely on the Comprehensive Reactor Analysis Bundle (BlueCRAB) multiphysics toolset in performing design- and beyond-design-basis accident analyses. In addition, the Nuclear Regulatory Commission has been using the MELCOR code to estimate mechanistic source terms during accidents. As MELCOR relies on isotopic inventory and reactor temperature/power evolution profiles during accident conditions—all of which can theoretically be obtained from BlueCRAB—the ultimate goal of this activity is to establish a common BlueCRAB-MELCOR framework. However, prior to the present research, BlueCRAB had never been used to calculate such quantities of interest at the full-core level. While there are many Monte Carlo (MC) codes capable of computing such quantities of interest, they are unable to readily account for multiphysics feedback. BlueCRAB allows for the coupling of different physics codes together to perform multiphysics-informed calculations. Therefore, the purpose of this fiscal year 2023 work is to investigate the feasibility and challenges of performing such calculations within BlueCRAB so as to generate the data that MELCOR relies on. To demonstrate the methodology, the proposed workflow was applied to a prototypical heat pipe-cooled microreactor model. To predict isotopic concentrations (taking into account the ac- cumulation of fission products during operation), the necessary microscopic cross sections were generated via OpenMC and tabulated with respect to temperature and burnup. Next, a recently developed capability in Griffin (the reactor physics application in BlueCRAB) was used to convert the OpenMC output format into the ISOXML format used by Griffin. A multiphysics microscopic depletion calculation that involved performing a coupled full-core, heterogeneous neutron trans- port and thermal calculation at each depletion step was conducted to deplete the core to end of life (EOL) conditions so as to provide both isotopics and the initial condition for the transient calculation. Following a brief null-transient to verify that the initial condition had been properly restarted and was indeed in thermal equilibrium, a heat pipe failure transient was simulated. Thus, the entire workflow of using BlueCRAB to generate MELCOR inputs, from cross-section generation to producing isotopic inventory and power/temperature evolution profiles during transients, is demonstrated. This report also details the identified gaps in the workflow and how they were (for the most part) addressed. Future work should focus on directly including MEL- COR into the workflow by performing a MELCOR calculation using the BlueCRAB-generated input data. In addition, the heat pipe reactor design should be improved so as to reflect more prototypical burnup characteristics at EOL.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

SU-E-T-380: Evaluation of BEBIG HDR 60Co System for AccuBoost Therapy

Purpose: In this project, the possibility of utilizing the BEBIG 60Co HDR system for AccuBoostTM treatment has been evaluated. Methods: Dose distributions in various breast sizes have been calculated for both Co-60 and Ir-192 sources using the MCNP5 code. These calculations were performed in breast tissues with thicknesses of 4cm, 6cm, and 8cm. The initial calculations were performed with the same applicator dimensions as the existing applicators used with the HDR Ir-192 system. The activity of the Co-60 source was selected such that the dose at the breast center was the same as the values from 192Ir. Then, the applicator thicknesses were increased to twice of those used with HDR Ir-192 system, for reducing skin and chest doses by Co-60 system. Dose to breast skin and chest wall were compared for both applicators types, with and without inclusion of a focusing cone at the applicator center. Results: The results showed that loading HDR Co-60 source inside the thin applicators impose higher doses to breast skin and chest wall compared to the 192Ir source. The area of the chest wall covered by 10Gy when treated by Co-60 with the thin and thick applicators, or treated by Ir-192 with thin applicator are 79cm2, 39cm2, and 3.8cm2, respectively. These values are reduced to 34cm2, 0cm2, and 0cm2 by using the focusing cone. It is worth noting that the breast skin areas covered by the 60Gy isodose line are 9.9cm2 and 7.8cm2 for Co-60 with the thin and thick applicators, respectively, while it is 20cm2 for Ir-192 when no focusing cone is present. These values are 0cm2, 0cm2, and 11cm2 in the presence of the focusing cone. Conclusion: The results indicate that using Co-60 with the thicker applicators is beneficial because of the higher half-life of Co-60, and the reduced maximum skin dose when compared with Ir-192.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2024

Current State of Neutron Source Term Calculations

SOURCES-4C has some significant deficiencies that are mostly due to problems with the default data libraries that it uses. Some research groups have modified these libraries to fix most of the problems, but not all of them. Adding modern data libraries to SOURCES-4C is one of the most promising approaches to resolving these issues. Other promising options include using either NEDIS or a GEANT4 tool. Both of these codes approximately match the accuracy of the modified SOURCES-4C code. The NEDIS code is Russian. The GEANT4 tool might be computationally slow and it might also no longer be actively maintained. There are a handful of other (α,n) codes, but they are not accurate.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

MetaboDirect: an analytical pipeline for the processing of FT-ICR MS-based metabolomic data

Background: Microbiomes are now recognized as the main drivers of ecosystem function ranging from the oceans and soils to humans and bioreactors. However, a grand challenge in microbiome science is to characterize and quantify the chemical currencies of organic matter (i.e., metabolites) that microbes respond to and alter. Critical to this has been the development of Fourier transform ion cyclotron resonance mass spectrometry (FT-ICR MS), which has drastically increased molecular characterization of complex organic matter samples, but challenges users with hundreds of millions of data points where readily available, user-friendly, and customizable software tools are lacking. Results: Here, we build on years of analytical experience with diverse sample types to develop MetaboDirect, an open-source, command-line-based pipeline for the analysis (e.g., chemodiversity analysis, multivariate statistics), visualization (e.g., Van Krevelen diagrams, elemental and molecular class composition plots), and presentation of direct injection high-resolution FT-ICR MS data sets after molecular formula assignment has been performed. When compared to other available FT-ICR MS software, MetaboDirect is superior in that it requires a single line of code to launch a fully automated framework for the generation and visualization of a wide range of plots, with minimal coding experience required. Among the tools evaluated, MetaboDirect is also uniquely able to automatically generate biochemical transformation networks (ab initio) based on mass differences (mass difference network-based approach) that provide an experimental assessment of metabolite connections within a given sample or a complex metabolic system, thereby providing important information about the nature of the samples and the set of microbial reactions or pathways that gave rise to them. Finally, for more experienced users, MetaboDirect allows users to customize plots, outputs, and analyses. Conclusion: Application of MetaboDirect to FT-ICR MS-based metabolomic data sets from a marine phage-bacterial infection experiment and a Sphagnum leachate microbiome incubation experiment showcase the exploration capabilities of the pipeline that will enable the research community to evaluate and interpret their data in greater depth and in less time. It will further advance our knowledge of how microbial communities influence and are influenced by the chemical makeup of the surrounding system. The source code and User’s guide of MetaboDirect are freely available through (https://github.com/Coayala/MetaboDirect) and (https://metabodirect.readthedocs.io/en/latest/), respectively.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2019

Exploiting fission chain reaction dynamics to image fissile materials

Radiation imaging is one potential method to verify nuclear weapons dismantlement. In this work, we present a method to discriminate between non-multiplying and multiplying neutron sources using a neutron coded aperture imaging system. This method applies time-correlated pulse-height (TCPH) analysis to identify neutrons in fission chain reactions and recreate images using only these events. This analysis was applied to measurements of weapons-grade plutonium (WGPu) metal and 252 Cf performed at the Nevada National Security Site (NNSS) Device Assembly Facility in July 2015. The results demonstrate it is possible to eliminate the non-fissile 252 Cf source from the image while preserving the fissile WGPu source. TCPH analysis was also applied to additional scenes in which the WGPu and 252 Cf sources were measured individually. The results of these separate measurements further demonstrate the ability to remove the non-fissile 252 Cf source and retain the fissile WGPu source. Simulations performed using MCNPX-PoliMi indicate that in a one hour measurement, hollow WGPu spheres are retained at a 1σ level for neutron multiplications M ≃ 2:7 and above. This work has potential application to dismantlement verification where it may be valuable to verify that all neutron sources in an image are multiplying.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

RVA. 3-D Visualization and Analysis Software to Support Management of Oil and Gas Resources

A free software application, RVA, has been developed as a plugin to the US DOE-funded ParaView visualization package, to provide support in the visualization and analysis of complex reservoirs being managed using multi-fluid EOR techniques. RVA, for Reservoir Visualization and Analysis, was developed as an open-source plugin to the 64 bit Windows version of ParaView 3.14. RVA was developed at the University of Illinois at Urbana-Champaign, with contributions from the Illinois State Geological Survey, Department of Computer Science and National Center for Supercomputing Applications. RVA was designed to utilize and enhance the state-of-the-art visualization capabilities within ParaView, readily allowing joint visualization of geologic framework and reservoir fluid simulation model results. Particular emphasis was placed on enabling visualization and analysis of simulation results highlighting multiple fluid phases, multiple properties for each fluid phase (including flow lines), multiple geologic models and multiple time steps. Additional advanced functionality was provided through the development of custom code to implement data mining capabilities. The built-in functionality of ParaView provides the capacity to process and visualize data sets ranging from small models on local desktop systems to extremely large models created and stored on remote supercomputers. The RVA plugin that we developed and the associated User Manual provide improved functionality through new software tools, and instruction in the use of ParaView-RVA, targeted to petroleum engineers and geologists in industry and research. The RVA web site (http://rva.cs.illinois.edu) provides an overview of functions, and the development web site (https://github.com/shaffer1/RVA) provides ready access to the source code, compiled binaries, user manual, and a suite of demonstration data sets. Key functionality has been included to support a range of reservoirs visualization and analysis needs, including: sophisticated connectivity analysis, cross sections through simulation results between selected wells, simplified volumetric calculations, global vertical exaggeration adjustments, ingestion of UTChem simulation results, ingestion of Isatis geostatistical framework models, interrogation of joint geologic and reservoir modeling results, joint visualization and analysis of well history files, location-targeted visualization, advanced correlation analysis, visualization of flow paths, and creation of static images and animations highlighting targeted reservoir features.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2026

VHClass

The code is used to predict the taxonomic source of an antibody heavy chain sequence. The code assigns a binary label to the input set of sequences - camelid or human. This prediction is generated using a random-forest based classification algorithm which is the backbone of the code. A complementary code splits the antibody sequence into antibody features - framework regions and CDR regions.

Davis, Anastasiia↗
DOE OSTI2025

Status of the MARS code

This report describes major features of the most recent version of the MARS code as well as ongoing developments. The list of features includes various options for geometry models, a beam line builder based on MADX code, import of geometry models in GDML format, use of structured and unstructured meshes for scoring purposes, an update to the recent TENDL library for a number of projectiles at low energies (up to 250 MeV), and a recently implemented method to calculate spatial distribution of residual dose in a single computer run without an intermediate source. Examples of the code application to various projects are presented as well.

Rakhno, Igor [Fermilab] (ORCID:0000000265828058)↗
DOE OSTI2022

Linear source method of characteristics in HELIOS2

The Studsvik Scandpower code HELIOS2 is a fast lattice transport code that has gained wide use for reactor fuel cycle analysis in commercial, regulatory, and research environments. It contains two transport solvers - Collision Probabilities (CP) and Method of Characteristics (MoC) - with a general geometry ability and generalized depletion for a variety of fuel types, experimental facilities, and isotope production applications. The HELIOS2 lattice code has been extended with the development of a linear source MoC option. The linear source equations are derived within the scope of the HELIOS2 general geometry and non-cyclic ray tracing. The implementation was tested for fine and coarse spatial mesh models of the 2D C5G7 MOX computational benchmark, which provides a pure and extreme test of a transport solver. The linear source yields at least comparable accuracy with reduced computational resource requirements relative to the flat-source approximation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Advanced Pellet Cladding Interaction Modeling Using the US DOE CASL Fuel Performance Code: Peregrine

The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermomechanical- chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale code that is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗