Search NASA⌕ Search

SEARCH · Search NASA

Results for “Code Assessment”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,693 records · Page 94
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Codes and Standards Update: February 2019

The goal of the DOE OE Energy Storage System Safety Roadmap is to foster confidence in the safety and reliability of energy storage systems. There are three interrelated objectives to support the realization of that goal: research, codes and standards(C/S)and communication/coordination. The objective focused on C/S is "To apply research and development to support efforts that refocused on ensuring that codes and standards are available to enable the safe implementation of energy storage systems in a comprehensive, non-discriminatory and science-based manner." The following activities support that objective and realization of the goal: (a.) Review and assess C/S which affect the design, installation, and operation of energy storage systems(ESS). (b.) Identify gaps in knowledge that require research and analysis that can serve as a basis for criteria in those C/S. (c.) Identify areas in C/S that are potentially in need of revision or enhancement and can benefit from activities conducted under research and development. (d.) Develop input for new or revisions to existing C/S through individual stakeholders, facilitated task forces, or through laboratory staff supporting these efforts. The purpose of this Codes and Standards Update is to support the above activities by providing information on efforts being conducted by U.S. standards developing organizations(SDOs)and other entities that are focused on ESS safety. The information is organized relative to the scope of each document in relation to ESS from the "macro to the micro"(e.g., from overarching covering considerable scope ,to installation specific, to ESS and then ESS components). Note that more macro documents are also likely to adopt by reference more micro documents.

25 ENERGY STORAGE↗
DOE OSTI2016

Early-time radiation flux symmetry optimization and its effect on gas-filled hohlraum ignition targets on the National Ignition Facility

Achieving ignition on the National Ignition Facility (NIF) is tied to our ability to control and minimize deviations from sphericity of the capsule implosion. Low-mode asymmetries of the hot spot result from the combined effect of radiation drive asymmetries throughout the laser pulse and initial roughness on the capsule surface. In this paper, we report on simulations and experiments designed to assess, measure, and correct the drive asymmetries produced by the early-time (≈first 2 ns or “picket”) period of the laser pulse. The drive asymmetry during the picket is commonly thought to introduce distortions in the hot-spot shape at ignition time. However, a more subtle effect not previously considered is that it also leads to an asymmetry in shock velocity and timing, thereby increasing the fuel adiabat and reducing the margin for ignition. It is shown via hydrodynamic simulations that minimizing this effect requires that the early-time asymmetry be kept below 7.5% in the second Legendre mode (P{sub 2}), thus keeping the loss of performance margin below ≈10% for a layered implosion. Asymmetries during the picket of the laser pulse are measured using the instantaneous self-emission of a high-Z re-emission sphere in place of an ignition capsule in a hohlraum with large azimuthal diagnostic windows. Three dimensional simulations using the code HYDRA (to capture the effect of non-azimuthal hohlraum features) coupled to a cross-beam energy transfer model [Michel et al., Phys. Plasmas 17, 056305 (2010)] are used to establish the surrogacy of the re-emit target and to assess the early-time drive symmetry. Calculations using this model exhibit the same sensitivity to variations in the relative input powers between the different cones of NIF beams as measured for the “Rev5” CH target [Haan et al., Phys Plasmas 18, 051001 (2011)] and reported by Dewald et al. [Phys. Rev. Lett. 111, 235001 (2013)]. The same methodology applied to recently improved implosions using different hohlraum geometries and picket powers show good agreement with experimental data.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

2021 Summer SPE Project

SPE is a project in order to develop new, more physics-based, seismic models of explosions (see Nelson et al. 2013). One key component of this effort is numerical modeling enabled by modern state-of-art code/software. Accurate modeling of the shape and amplitude of seismic waves from their generation to their propagation to remote monitoring seismic stations is important to our ability to determine the origin and strength of the source from remote recording. In this project, the modeling is performed by coupling two codes HOSS and SPECFEM3D. HOSS models the dynamic nonlinear processes happening near the explosion. SPECFEM3D computes the propagation of seismic waves as they travel through 3D complex Earth models. SPECFEM3D was modified in order to be driven by a set of time-series calculated by HOSS in lieu of a seismic source. The goal of this summer project is the investigation of several questions pertinent to the establishment of a full end-to-end modeling capability from the high-rate strain regime area to remote distances where seismic station record seismic waves generated by explosions. The investigated questions are: (1) How to perform proper filtering? Direct-solution modeling only sustains a limited range of frequency depending on the grid size. As we go from one modeling domain to the other via coupling, the mesh size is getting coarser to allow modeling at large scale but also to account for the fact that high-frequency waves do not physically travel to large distances. So filtering of the time-series generated by the near-field hydrodynamic modeling is a current practice often employed but its effects on the modeled waveforms has to be investigated. (2) Quantitative assessment of the efficiency of attenuation to remove high-frequency content of the wavefield. This assessment will allow to create meshes with a grid size appropriate to the actual physics of wave propagation for a given explosion. (3) Checking that the coupling process and the two codes respect the cylindrical symmetry that is expected in the case of a pure explosion in a half-space. (4) The effect of the state-of-stress in the near-source area on the modeled seismic waveforms. These questions will be investigated through the modeling of SPE-4P, the fourth explosion of this series because it was designed to have little interaction with the geologic surrounding and the free surface of the Earth so that it is the most explosion-like experiment, with the most symmetries to be verified.

58 GEOSCIENCES↗
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 OSTI2021

Modeling SARS-CoV-2 proteins in the CASP-commons experiment

Critical Assessment of Structure Prediction (CASP) is an organization aimed at advancing the state of the art in computing protein structure from sequence. In the spring of 2020, CASP launched a community project to compute the structures of the most structurally challenging proteins coded for in the SARS-CoV-2 genome. Forty-seven research groups submitted over 3000 three-dimensional models and 700 sets of accuracy estimates on 10 proteins. The resulting models were released to the public. CASP community members also worked together to provide estimates of local and global accuracy and identify structure-based domain boundaries for some proteins. Subsequently, two of these structures (ORF3a and ORF8) have been solved experimentally, allowing assessment of both model quality and the accuracy estimates. Models from the AlphaFold2 group were found to have good agreement with the experimental structures, with main chain GDT_TS accuracy scores ranging from 63 (a correct topology) to 87 (competitive with experiment).

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2017

Initial Implementation of Transient VERA-CS

In this milestone the capabilities of both CTF and MPACT were extended to perform coupled transient calculations. This required several small changes in MPACT to setup the problems correctly, perform the edits correctly, and call the appropriate CTF interfaces in the right order. For CTF, revisions and corrections to the transient timestepping algorithm were made, as well as the addition of a new interface subroutine to allow MPACT to drive CTF at each timestep. With the modifications completed, the initial coupled capability was demonstrated on some problems used for code verification, a hypothetical small mini-core, and a Watts Bar demonstration problem. For each of these cases the results showed good agreement with the previous MPACT internal TH feedback model that relied on a simplified fuel heat conduction model and simplified coolant treatment. After the pulse the results are notably different as expected, where the effects of convection of heat to the coolant can be observed. Areas for future work were discussed, including assessment and development of the CTF dynamic fuel deformation and gap conductance models, addition of suitable transient boiling and CHF models for the rapid heating and cooling rates seen in RIAs, additional validation and demonstration work, and areas for improvement to the code input and output capabilities.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2018

Validation of Statistical Uncertainties in Subcritical Benchmark Measurements: Part I - Theory and Simulations

Subcritical neutron noise measurements are often used to determine multiplication of a system with Special Nuclear Material. This work presents an uncertainty approach which incorporates the singles and doubles counting rates determined by the Hage-Cifarelli formalism. This is a moments-based approach which utilizes time correlations between prompt neutrons in order to assess the system multiplication. After the method is described, a validation is presented which utilizes simulated data generated using a 0-D point-kinetics Monte Carlo code.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Radiation Effects Related to Repaired BWR Core Shrouds

The material properties of the core shroud and supporting tie rod components are important factors for Boiling Water Reactor (BWR) life extension considerations. Given the variability of nickel and boron content found in 304 stainless steel shroud plates, and XM-19 and Inconel X-750 tie rod materials, this paper combines the latest spectral information in the vicinity of the core shroud to calculate hydrogen and helium gas production limits as well as atomic displacement damage spanning 80 full-power years (FPY) using the SPECTER code. Results indicate that, whereas boron has little effect on the atomic displacement damage, helium production from boron can be important and must be included in weldability assessments. Since the boron concentration is uncertain, helium generation is calculated assuming a composition range based on specifications and reported measurements. The results show that helium production caused by transmutations in the 304 SS shroud material are dominated by the concentration of boron and will have a significant impact on weldability after only a few years of service. For tie rods, the susceptibility to irradiation-assisted stress corrosion cracking (IASCC) has been assessed by determining helium production in Inconel X-750 IASCC specimens irradiated in the Advanced Test Reactor (ATR) reactor and correlating with the helium produced in a BWR. As a tie rod material in a high thermal neutron flux environment, Inconel X-750 is more likely to be IASCC sensitive during service in a BWR due to helium than the XM-19 stainless steel alloy because of a greater nickel content and boron concentration. Weldability of the core shroud plate material, IASCC of the tie rod materials, and the gamma photon contribution to radiation damage are discussed.

radiation damage in materials, BWR↗
DOE OSTI2017

Modelling Neutron-induced Reactions on 232–237 U from 10 keV up to 30 MeV

Comprehensive calculations of cross sections for neutron-induced reactions on 232–237 U targets are performed in this paper in the 10 keV–30 MeV incident energy range with the code EMPIRE–3.2 Malta. The advanced modelling and consistent calculation scheme are aimed at improving our knowledge of the neutron scattering and emission cross sections, and to assess the consistency of available evaluated libraries for light uranium isotopes. The reaction model considers a dispersive optical potential (RIPL 2408) that couples from five (even targets) to nine (odd targets) levels of the ground-state rotational band, and a triple-humped fission barrier with absorption in the wells described within the optical model for fission. A modified Lorentzian model (MLO) of the radiative strength function and Enhanced Generalized Superfluid Model nuclear level densities are used in Hauser-Feschbach calculations of the compound-nuclear decay that include width fluctuation corrections. The starting values for the model parameters are retrieved from RIPL. Excellent agreement with available experimental data for neutron emission and fission is achieved, giving confidence that the quantities for which there is no experimental information are also accurately predicted. Finally, deficiencies in existing evaluated libraries are highlighted.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Calculations versus measurements for residual dose rates from SNS spent structures

Here, residual dose rate measurements were conducted on target vessel #13 and proton beam window #5 after extraction from their service locations. These measurements are used to verify calculation methods of radionuclide inventory assessment that are typically performed for nuclear waste characterization and transportation of these structures. Neutronics analyses for predicting residual dose rates are carried out using the transport code MCNPX and the transmutation code CINDER90. For transport analyses a complex and rigorous geometry model of the structures and their surroundings are applied. The neutronics analyses are carried out using the Bertini and CEM high energy physics models for simulating particles interactions above the table-based cross section range. Finally, obtained calculational results are analyzed and compared to the measured dose rates and overall show good agreement within 25%, which shows applicability of the methods used in analyses.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

RISMC industry application #1 (ECCS/LOCA). Core characterization automation: Lattice codes interface for PHISICS/RELAP5-3D

The United States Nuclear Regulatory Commission (U.S. NRC) is currently proposing rule-making “10 CFR 50.46c” to revise the loss-of-coolant-accident (LOCA) and emergency core cooling system (ECCS) acceptance criteria to include the effects of higher burn-up on cladding performance. The key implications of this proposition are that the cladding performance cannot be evaluated anymore as separated from other disciplines, but needs to be evaluated in a coupled manner with other physics. The Risk-Informed Safety Margin Characterization (RISMC) Pathway initiated several programs at Idaho National Laboratory (INL) to support the industry in the transition to the new rule. One of these programs is the Industry Application #1 (IA1). IA1’s goal is to develop an Integrated Evaluation Model (IEM) called LOCA Toolkit for the US (LOTUS). This tool connects five major disciplines involved in LOCA analysis, among them is (automated) core design (CD-A). Here, this paper focuses solely on the design of LOTUS' CD-A module as well as a demonstration application of the CD-A module to a generic Pressurized Water Reactor (PWR). The demonstration of the other LOTUS disciplines together with a demonstration example of the final risk-informed safety margin analysis is published in a separate paper. LOTUS is meant to provide the methodology framework into which the user can plug his established tools and codes. With the CD-A, the plant owner/operator will characterize his core design with the tools he has. He will then integrate his tools/methods into the LOTUS framework for the downstream safety analysis. For the current CD-A demonstration, the tools used are the already coupled codes PHISICS and RELAP5-3D, while cross section generation is done using the Studsvik lattice code HELIOS-2.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2024

Impact of LWR assembly structural features on cladding burst behavior under LOCA conditions

This provides an initial scoping study on clad balloon and burst behavior for burnup extension of reactor fuel. The associated issues with burnup extension are fuel fragmentation, relocation, and dispersal in the event of cladding failure. The general finding of this work is that the structural features, spacer grids and mixing vanes, locally suppress cladding deformation but have little impact on the overall clad performance during loss-of-coolant accidents. The work detailed in a previous report by Capps et al. focused on core optimization via neutronics, thermal hydraulics and thermomechanical analysis for burnups beyond 62 GWD/tU and enrichments above 5%. Uncertainty of fuel fragmentation relocation and dispersal in high-burnup rods during accident conditions was also investigated. The dispersal aspect of fuel fragmentation depends on cladding rupture. Thus, assessing uncertainties in the rupture behavior is helpful in estimating the dispersal of high-burnup fuel. This study builds on the previous work by assessing the impact of assembly structural features on cladding balloon and burst behavior in a full-length fuel rod. In this work, the BISON fuel performance code was used to generate 2D radial and height meshes containing structural features commonly used in nuclear fuel assemblies. First, meshes were generated with spacer grids. Results were then compared to the cladding burst temperature and balloon strain results from the previous work. A mesh sensitivity study was performed to ensure that mixing vanes and spacer grid effects were appropriately considered, resulting in a more refined mesh than the previous study. The balloon deformation and burst times of the cladding were compared to the original case. Consideration was also given to the effect of rod initial pressure. In conclusion, 3D quarter rod simulations were also performed and found good agreement with the 2D simulations in clad deformation and reasonable agreement in burst times.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Coarse grained simulation of convectively driven turbulent mixing, transition, and turbulence decay

Accurate predictions with quantifiable uncertainties are essential to many practical turbulent flow applications exhibiting extreme geometrical complexity and broad ranges of length and time scales. Under-resolved computer simulations are typically unavoidable in such applications, and implicit large-eddy simulation (ILES) often becomes the effective strategy. In this work, we focus on assessing ILES initialized with well-characterized 25 6 3 homogeneous isotropic turbulence datasets generated with direct numerical simulation (DNS). ILES solutions based on the LANL xRAGE code are studied as function of resolution for 6 4 3 , 12 8 3 , 25 6 3 , and 51 2 3 grids. ILES performance of new directionally-unsplit high-order numerical hydrodynamics algorithms in xRAGE is examined in this context. Compared to the initial 25 6 3 DNS, we find longer inertial subranges and higher turbulence Reynolds number (Re) for 25 6 3 and 51 2 3 xRAGE — attributed to having linked DNS (Navier–Stokes based) solutions to nominally inviscid (higher Re) Euler based ILES solutions. For fixed 25 6 3 resolution, we find that significantly larger simulated turbulence Re can be achieved with the higher-order unsplit (vs. split) discretizations.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2023

Ray-based cross-beam energy transfer modeling for broadband lasers

Broadband lasers have the potential to mitigate cross-beam energy transfer (CBET) in direct-drive inertial confinement fusion (ICF) experiments. Furthermore, a quantitative assessment of the bandwidth required for CBET mitigation necessitates the development of broadband ray-based CBET models that can be implemented in the radiation-hydrodynamic codes that are used to design ICF experiments. Two different approaches to broadband ray-based CBET modeling (discrete and fixed spectrum) are developed and compared to wave-based calculations. Both approaches give good agreement with wave-based calculations in ICF-relevant configurations. Full-scale 3D calculations show that the bandwidth required for adequate CBET mitigation increases with increasing scale and drive intensity.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Calculations vs. measurements of remnant dose rates for SNS spent structures

Residual dose rate measurements were conducted on target vessel #13 and proton beam window #5 after extraction from their service locations. These measurements were used to verify calculation methods of radionuclide inventory assessment that are typically performed for nuclear waste characterization and transportation of these structures. Neutronics analyses for predicting residual dose rates were carried out using the transport code MCNPX and the transmutation code CINDER90. For transport analyses complex and rigorous geometry model of the structures and their surrounding are applied. The neutronics analyses were carried out using Bertini and CEM high energy physics models for simulating particles interaction. Obtained preliminary calculational results were analysed and compared to the measured dose rates and overall are showing good agreement with in 40% in average.

Popova, Irina I.↗
DOE OSTI2015

Neutron-Irradiated Samples as Test Materials for MPEX

Plasma Material Interaction (PMI) is a major concern in fusion reactor design and analysis. The Material-Plasma Exposure eXperiment (MPEX) will explore PMI under fusion reactor plasma conditions. Samples with accumulated displacements per atom (DPA) damage produced by fast neutron irradiations in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) will be studied in the MPEX facility. This paper presents assessments of the calculated induced radioactivity and resulting radiation dose rates of a variety of potential fusion reactor plasma-facing materials (such as tungsten). The scientific code packages MCNP and SCALE were used to simulate irradiation of the samples in HFIR including the generation and depletion of nuclides in the material and the subsequent composition, activity levels, gamma radiation fields, and resultant dose rates as a function of cooling time. A challenge of the MPEX project is to minimize the radioactive inventory in the preparation of the samples and the sample dose rates for inclusion in the MPEX facility.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Thermocatalytic Heat Pipes for Geothermal Resource Recovery

Heat pipes are an important technology that allow orders of magnitude faster heat transfer than simple conduction. However, operating principles in heat pipes place fundamental bounds on their performance (critical heat flux and efficiency). Conventional heat pipe functionality is inherently tied to vaporization and condensation of the working fluid charged in the heat pipe. These fluids each have different operating temperature ranges based on the capillary, entrainment, sonic, and boiling limits of the heat pipe design. These limits, typically the capillary limit, dictate the maximum heat flux a heat pipe can carry, and most importantly for geothermal systems, the distance over which the pipes can operate (100 to 200 m maximum under optimum conditions). A thermocatalytic heat pipe breaks the inherent limitations of phase change thermo- and hydrodynamics and can transform heat pipe technology as a potentially more efficient means of extracting heat from a geothermal resource. The thermocatalytic heat pipe uses a working fluid to transport both sensible and chemical heat. An endothermic chemical reaction at depth removes heat from the reservoir and produces reactive intermediates, which are transported to the surface and used to run a reverse exothermic reaction that releases heat for use in power generation or other useful purposes. This technology offers two distinct advantages over conventional geothermal heat recovery technologies: (1) lower heat loss to the rock outside of the geothermal reservoir, and (2) higher heat transfer rates to the well field within the geothermal reservoir. Both advantages offer opportunity to reduce risks and lower costs of geothermal energy recovery. In this report, we discuss an initial effort to assess the efficacy and limitations of this technology for extracting heat from both porous/permeable and nominally impermeable geothermal reservoirs. Numerical simulation capabilities of the STOMP-GT code were enhanced to enable simulations of thermochemical heat pipes traversing geothermal reservoirs. An array of potential thermochemical reaction systems was evaluated and screened. Of these, an ethanol dehydration reaction was most promising in the vapor-liquid reaction set. A solid-phase dehydration reaction (CuSO4·5H2O) showed the highest reaction enthalpy per unit volume but would require development of a nonaqueous carrier fluid to implement it in a heat pipe. Subsurface reservoir simulations predicted long-term performance of the heat pipes for each geothermal reservoir type. The performance of U-shaped wells and coaxial wells was evaluated for a suite of reactions for both hydrothermal and hot dry rock reservoirs and was compared with a baseline case of simply pumping water through the wells. The heat pipe technology was additionally evaluated for an enhanced geothermal system (EGS) with an injection borehole, production borehole, and intervening hydraulically conductive fracture. All reservoir types showed significant improvement in heat recovered over a 20-year operating period ranging from a 1.8X increase for the hot dry rock case to more than 2.5X more energy recovered for the EGS case.

15 GEOTHERMAL ENERGY↗
DOE OSTI2020

FAST-1.0: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients: Developed under NQA-1-2017

Fuel Analysis under Steady-state and Transients (FAST) is the US Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include: heat conduction through the fuel and other materials, heat transfer from the cladding-to-coolant, cladding elastic and plastic deformation, including creep, fuel-cladding mechanical interaction, fission gas release from the fuel, rod internal pressure and void volume, and cladding oxidation. The code contains necessary material and coolant properties, as well as clad-to-coolant heattransfer correlations, for normal operation through postulated accidents for today’s US-based light water reactor (LWR) fuel designs. FAST-1.0 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.0, which is the first official version of this code. This document is one of a series of documents on FAST; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗