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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 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Our Next Two Steps for Fukushima Daiichi Muon Tomography

After the vast disasters caused by the great earthquake and tsunami in eastern Japan, we proposed applying our Muon Tomography (MT) technique to help and improve the emergency situation at Fukushima Daiichi using cosmic-ray muons. A reactor-tomography team was formed at LANL which was supported by the Laboratory as a response to a request by the former Japanese Prime Minister, Naoto Kan. Our goal is to help the Japanese people and support remediation of the reactors. At LANL, we have carried out a proof-of-principle technical demonstration and simulation studies that established the feasibility of MT to image a reactor core. This proposal covers the next two critical steps for Fukushima Daiichi Muon Imaging: (1) undertake case study mock-up experiments of Fukushima Daiichi, and (2) system optimization. We requested funding to the US and Japanese government to assess damage of reactors at Fukushima Daiichi. The two steps will bring our project to the 'ready-to-go' level.

47 OTHER INSTRUMENTATION↗
DOE OSTI2013

Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO 2 particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Moving-mesh Simulations of Star-forming Cores in Magneto-gravo-turbulence

Star formation in our Galaxy occurs in molecular clouds that are self-gravitating, highly turbulent, and magnetized. We study the conditions under which cloud cores inherit large-scale magnetic field morphologies and how the field is governed by cloud turbulence. We present four moving-mesh simulations of supersonic, turbulent, isothermal, self-gravitating gas with a range of magnetic mean-field strengths characterized by the Alfvénic Mach number M{sub A,0}, resolving prestellar core formation from parsec to a few astronomical unit scales. In our simulations with the turbulent kinetic energy density dominating over magnetic pressure (M{sub A,0}>1), we find that the collapse is approximately isotropic with B ∝ ρ {sup 2/3}, core properties are similar regardless of initial mean-field strength, and the field direction on 100 au scales is uncorrelated with the mean field. However, in the case of a dominant large-scale magnetic field (M{sub A,0}=0.35), the collapse is anisotropic with B ∝ ρ {sup 1/2}. This transition at M{sub A,0}∼1 is not expected to be sharp, but clearly signifies two different paths for magnetic field evolution in star formation. Based on observations of different star-forming regions, we conclude that star formation in the interstellar medium may occur in both regimes. Magnetic field correlation with the mean field extends to smaller scales as M{sub A,0} decreases, making future Atacama Large Millimeter Array observations useful for constraining M{sub A,0} of the interstellar medium.

79 ASTRONOMY AND ASTROPHYSICS↗
NASA NTRS2007

Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores

A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.

Litchford, Ron J.↗
DOE OSTI2013

The behavior of ANGRA 2 nuclear power plant core for a small break LOCA simulated with RELAP5 code

This work discusses the behavior of Angra 2 nuclear power plant core, for a postulate Loss of Coolant Accident (LOCA) in the primary circuit for Small Break Loss Of Coolant Accident (SBLOCA). A pipe break of the hot leg Emergency Core Cooling System (ECCS) was simulated with RELAP 5 code. The considered rupture area is 380 cm{sup 2}, which represents 100% of the ECCS pipe flow area. Results showed that the cooling is enough to guarantee the integrity of the reactor core.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2008

Analysis of Lunar Highland Regolith Samples from Apollo 16 Drive Core 64001/2 and Lunar Regolith Simulants - An Expanding Comparative Database

We present modal data from QEMSCAN(registered TradeMark) beam analysis of Apollo 16 samples from drive core 64001/2. The analyzed lunar samples are thin sections 64002,6019 (5.0-8.0 cm depth) and 64001,6031 (50.0-53.1 cm depth) and sieved grain mounts 64002,262 and 64001,374 from depths corresponding to the thin sections, respectively. We also analyzed lunar highland regolith simulants NU-LHT-1M, -2M, and OB-1, low-Ti mare simulants JSC-1, -lA, -1AF, and FJS-1, and high-Ti mare simulant MLS-1. The preliminary results comprise the beginning of an internally consistent database of lunar regolith and regolith simulant mineral and glass information. This database, combined with previous and concurrent studies on phase chemistry, bulk chemistry, and with data on particle shape and size distribution, will serve to guide lunar scientists and engineers in choosing simulants for their applications. These results are modal% by phase rather than by particle type, so they are not directly comparable to most previously published lunar data that report lithic fragments, monomineralic particles, agglutinates, etc. Of the highland simulants, 08-1 has an integrated modal composition closer than NU-LHT-1M to that of the 64001/2 samples, However, this and other studies show that NU-LHT-1M and -2M have minor and trace mineral (e.g., Fe-Ti oxides and phosphates) populations and mineral and glass chemistry closer to these lunar samples. The finest fractions (0-20 microns) in the sieved lunar samples are enriched in glass relative to the integrated compositions by approx.30% for 64002,262 and approx.15% for 64001,374. Plagioclase, pyroxene, and olivine are depleted in these finest fractions. This could be important to lunar dust mitigation efforts and astronaut health - none of the analyzed simulants show this trend. Contrary to previously reported modal analyses of monomineralic grains in lunar regolith, these area% modal analyses do not show a systematic increase in plagiociase/pyroxene as size fraction decreases.

Schrader, Christian M.↗
DOE OSTI2019

Development of multi-objective core optimization framework and application to sodium-cooled fast test reactors

The optimization of a Sodium-cooled Fast Reactor (SFR) core is a challenging process, due to the large number of design parameters, the nonlinearities among inputs and outputs, and the complicated correlation among output parameters. This study attempts to develop a generalized framework for the SFR core optimization by coupling the sensitivity analysis, advanced optimization algorithm, and optionally the surrogate modeling. The framework is built based on the fast reactor modeling capability of the Argonne Reactor Computation (ARC) suite and the sensitivity analysis and optimization modules embedded in the DAKOTA code, both have been integrated within the NEAMS Workbench. The genetic algorithm is selected as the optimization method for its robustness, while the option of surrogate modeling is also explored to alleviate the computational burden caused by employing the ARC direct core physics simulation and thus enhance the efficiency of the optimization. Finally, the normalized deviations of performance parameters of the near-optimal solution from those of the ideal core are calculated and used as criteria to down select the final core design. The developed framework is applied to the Advanced Burner Test Reactor (ABTR) core, and optimal solutions are determined by balancing various objectives simultaneously, including peak fast flux, core volume, power, reactivity swing, plutonium mass feed, while at the same time satisfying the predefined constraints due to safety or economics considerations. The optimal ABTR core design obtained using the direct physical simulation and surrogated model are compared and discussed. It is found that using the accurately constructed surrogate models can significantly reduce the required computational time while maintaining satisfactory accuracy.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

IRSN working program status on tools for evaluation of SFR cores static neutronics safety parameters

As technical support of the French Nuclear Safety Authority, IRSN will be in charge of safety assessment of any future project of Sodium Fast Reactor (SFR) that could be built in France. One of the main safety topics will deal with reactivity control. Since the design and safety assessment of the last two SFR plants in France (Phenix and Superphenix, more than thirty years ago), methods, codes and safety objectives have evolved. That is why a working program on core neutronic simulations has been launched in order to be able to evaluate accuracy of future core characteristics computations. The first step consists in getting experienced with the ERANOS well-known deterministic code used in the past for Phenix and Superphenix. Then Monte-Carlo codes have been tested to help in the interpretation of ERANOS results and to define what place this kind of codes can have in a new SFR safety demonstration. This experience is based on open benchmark computations. Different cases are chosen to cover a wide range of configurations. The paper shows, as an example, criticality results obtained with ERANOS, SCALE and MORET, and the first conclusions based on these results. In the future, this work will be extended to other safety parameters such as sodium void and Doppler effects, kinetic parameters or flux distributions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

The Virtual Test Bed (VTB) repository: a library of multiphysics reference reactor models using NEAMS tools

With the next generation of nuclear reactors under development, modeling and simulation (MS) tools are being developed by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in order to support their design, licensing, and future operation. Mirroring the physical test beds currently under construction (i.e., EBR-II and ZPPR), the Virtual Test Bed (VTB) was launched by the National Reactor Innovation Center (NRIC) in collaboration with NEAMS to support the advanced reactor community. This collaborative effort, which involves multiple teams at both Idaho National Laboratory and Argonne National Laboratory aims to use NEAMS tools to model a wide range of reactor designs. Those models are automatically tested to ensure their continued functionality as the tools are further developed. Examples are extensively documented, each acting as a tutorial for applying the relevant NEAMS tools to that reactor design. Currently, five advanced reactor types (with a total of eight specific design variants) are simulated by a variety of different models. These models range from steady-state, core multiphysics simulations to integrated plant analysis during loss-of flow transients. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2006

A Model-based Interpretation of Low-frequency Changes in the Carbon Cycle during the Last 120,000 years and its Implications for the Reconstruction of Atmospheric (delta) 14-C

A main caveat in the interpretation of observed changes in atmospheric (Delta)C-l4 during the last 50,000 years is the unknown variability of the carbon cycle, which together with changes in the C-14 production rates determines the C-14 dynamics. A plausible scenario explaining glacial/interglacial dynamics seen in atmospheric CO2 and (delta)C-13 was proposed recently (Kohler et al., 2005a). A similar approach that expands its interpretation to the C-14 cycle is an important step toward a deeper understanding of (Delta)C-14 variability. This approach is based on an ocean/atmosphere/biosphere box model of the global carbon cycle (BICYCLE) to reproduce low-frequency changes in atmospheric CO2 as seen in Antarctic ice cores. The model is forced forward in time by various paleoclimatic records derived from ice and sediment cores. The simulation results of our proposed scenario match a compiled CO2 record from various ice cores during the last 120,000 years with high accuracy (r(sup 2) = 0.89). We analyze scenarios with different C-14 production rates, which are either constant or based on Be-10 measured in Greenland ice cores or the recent high-resolution geomagnetic field reconstruction GLOPIS-75 and compare them with the available (Delta)C-14 data covering the last 50,000 years. Our results suggest that during the last glacial cycle in general less than 110%0o f the increased atmospheric (Delta)C-14 is based on variations in the carbon cycle, while the largest part (5/6) of the variations has to be explained by other factors. Glacial atmospheric (Delta)C-14 larger than 700% cannot not be explained within our framework, neither through carbon cycle-based changes nor through variable C-14 production. Superimposed on these general trends might lie positive anomalies in atmospheric (Delta)C-14 of approx. 50% caused by millennial-scale variability of the northern deep water production during Heinrich events and Dansgaard/Oeschger climate fluctuations. According to our model, the dominant processes that increase glacial (Delta)C-14 are a reduced glacial ocean circulation (+ approx.40%0), a restricted glacial gas exchange between the atmosphere and the surface ocean through sea ice coverage (+ approx. 20%), and the enrichment of dissolved inorganic carbon with C-14 in the surface waters through isotopic fractionation during higher glacial marine export production caused by iron fertilization (+ approx.10%).

Koehler, Peter↗
DOE OSTI2022

The Virtual Test Bed (VTB) Repository: A Library of Multiphysics Reference Reactor Models using NEAMS Tools

With the next generation of nuclear reactors under development, modeling and simulation (M&S) tools are being developed by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in order to support their design, licensing, and future operation. Mirroring the physical test beds currently under construction (i.e., EBR-II and ZPPR), the Virtual Test Bed (VTB) was launched by the National Reactor Innovation Center (NRIC) in collaboration with NEAMS to support the advanced reactor community. This collaborative effort, which involves multiple teams at both Idaho National Laboratory and Argonne National Laboratory aims to use NEAMS tools to model a wide range of reactor designs. Those models are automatically tested to ensure their continued functionality as the tools are further developed. Examples are extensively documented, each acting as a tutorial for applying the relevant NEAMS tools to that reactor design. Currently, five advanced reactor types (with a total of eight specific design variants) are simulated by a variety of different models. These models range from steady-state, core multiphysics simulations to integrated plant analysis during loss-of-flow transients.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Initial Comparison of Direct and Legacy Modeling Approaches for Radial Core Expansion Analysis

Radial core expansion in sodium-cooled fast reactors provides an important reactivity feedback effect. As the reactor power increases due to normal start up conditions or accident scenarios, the core and surrounding materials heat up, causing both grid plate expansion and bowing of the assembly ducts. When the core restraint system is designed correctly, the resulting structural deformations introduce negative reactivity which decreases the reactor power. Historically, an indirect procedure has been used to estimate the reactivity feedback due to structural deformation which relies upon perturbation theory and coupling legacy physics codes with limited geometry capabilities. With advancements in modeling and simulation, radial core expansion phenomena can now be modeled directly, providing an assessment of the accuracy of the reactivity feedback coefficients generated by indirect legacy methods. Recently a new capability was added to the PROTEUS-SN unstructured geometry neutron transport solver to analyze deformed meshes quickly and directly. By supplying the deformed mesh in addition to the base configuration input files, PROTEUS-SN automatically processes material adjustments including calculation of region densities to conserve mass, calculation of isotopic densities according to material models (for example, sodium density as a function of temperature), and subsequent re-homogenization of materials. To verify the new capability of directly simulating deformed meshes, PROTEUS-SN was used to compute reactivity feedback for a series of contrived yet representative deformed configurations for the Advanced Burner Test Reactor design. The indirect legacy procedure was also performed to generate reactivity feedback coefficients for the same deformed configurations. Interestingly, the legacy procedure consistently overestimated reactivity feedbacks by 35% compared to direct simulations by PROTEUS-SN. This overestimation indicates that the legacy procedures are in fact not conservative and could be overestimating reactivity feedback effects that are closely tied to reactor safety. We conclude that there is indeed value in performing direct simulation of deformed meshes despite the increased computational expense. PROTEUS-SN is already part of the SHARP multi-physics toolkit where both thermal hydraulics and structural mechanical feedback modeling can be applied but this is the first comparison of direct simulation to legacy techniques for radial core expansion.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Effects of Boron and Graphite Uncertainty in Fuel for TREAT Simulations

As a result of the accident at Fukushima, the United States has made a concerted effort to develop fuel with enhanced accident tolerance for the current fleet of reactors. In 2012 Congress laid out a schedule for the development of accident tolerant fuel (ATF), the goal of which is to test in a commercial reactor by 2022. Idaho National Lab's (INL) Transient Reactor Test Facility (TREAT) is one of the locations that will be performing ATF testing and is scheduled for restart in 2018. TREAT was built to subject fuel to conditions ranging from mild transients to severe reactor accidents. However, typically a number of pre-test calibrations are required before evaluation of fuel can begin, which is costly and time consuming. Advanced modeling techniques and current computational capacity make full core TREAT simulations possible, with the goal of such simulations to understand the pre-test core and minimize the number of required calibrations. But, in order to simulate TREAT with a high degree of precision the reactor materials and geometry must also be modeled with a high degree of precision. This paper examines how uncertainty in the reported values of boron and graphite have an effect on simulations of TREAT.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Dataset of Simulated Intracardiac Transmembrane Voltage Recordings and ECG Signals

The dataset consists of pairs of computationally simulated intracardiac transmembrane voltage recordings and ECG signals. In total, 16140 organ-level simulations were conducted to create this dataset, using a range of cardiac geometries and physiological parameters. Simulations were performed at LLNL's Lassen supercomputer, concurrently utilizing 4 GPUs and 40 CPU cores. Each simulation produced pairs of 500ms-by-10 ECG signals and 500ms-by-75 transmembrane voltage signals. For convenience, the signals are concatenated and saved as matrices. Each of these matrices is then stored as a numpy array. See the documentation for further details.

Landajuela, Mikel↗
DOE OSTI2025

Deterministic High-Fidelity Neutronics Simulation of Pebble Bed Reactors Using Pebble Tracking Transport

The pebble tracking transport (PTT) algorithm offers a high-fidelity deterministic approach for neutron transport for pebble bed reactors (PBRs). This approach requires the mesh for the active-core region to consist exclusively of tetrahedral elements, where each node in the pebble-packing region represents a pebble centroid. This paper investigates the application of PTT for full-scale PBRs, considering both the isothermal and the temperature-dependent core conditions. Macroscopic cross sections are generated using Serpent 2 full-core eigenvalue simulations where pebbles are grouped into disjoint subsets using machine learning. To minimize the need for individual cross-section sets for each pebble in the core, K-means clustering is used to group pebbles by temperature and neutronic environment parameters. Here, we compare the multiplication factor and power rate distributions between PTT simulations using the Griffin reactor physics software and reference solutions from Serpent 2. Our analysis shows that a full-core, high-fidelity PTT calculation produces accurate results with minimal local (pebblewise) errors. Additionally, timing results indicate that PTT simulations converge rapidly on modern supercomputing platforms.

Griffin↗
DOE OSTI2021

Assessment of MOOSE-Based Tools for Calculating Radial Core Expansion

Radial core expansion in liquid-metal cooled fast reactor systems is a well-known phenomenon that produces strong reactivity feedback effects. An inherently safe reactor design takes advantage of negative reactivity feedback in accident scenarios by utilizing a core restraint system which produces a bowed shape that allows for radial expansion of the fuel regions. Detailed modeling and simulation of core radial expansion itself as well as subsequent reactivity feedback is a challenging task involving contact of many fuel assembly elements and physics feedback from neutronics, thermal hydraulics, and thermal mechanical response. A variety of physics codes have been developed to model aspects of radial core expansion but in general invoke geometrical or physics approximations. No code system currently exists which tightly and robustly couples these physics with enough detail to fully resolve the complex core radial expansion reactivity feedback effects. The future availability of such a code system is of vital importance to fully understanding the reactivity feedback effects that occur due to radial expansion, and consequently to optimizing the design of the core restraint system. A high-fidelity code will also be used to benchmark existing lower fidelity, faster running models to understand their benefits, limitations, and range of applications. A code development path using MOOSE-based tools is proposed in order to leverage the detailed geometry capabilities and natural tight coupling and robustness of MOOSE-based applications for modeling this complex phenomena. While simulation of the full phenomenon involves several physics, an assessment has been initiated on the capabilities and readiness of the currently available Tensor Mechanics module within MOOSE for calculation of the structural mechanical responses which occur within the reactor core. This report focuses on modeling the force-deformation response which mimics the physics of a duct contact deformation, as well as differential thermal expansion which produces a thermal bowed shaped for the fuel assemblies in a core. Simple examples were initially performed such as simple supported beam bending under load. The complexity of examples was progressively increased to better mimic the duct behavior by including a differential thermal example and inclusion of hexagonal cross-sections in the geometry. Further assessment of the structural mechanical response simulation capability is still required for modeling duct contact interactions and irradiation creep and swelling. Companion thermal hydraulic and neutronics assessments will also be required; these activities are planned for future years. Finally, integration of the multiple physics components through MOOSE is required to predict the core radial expansion and subsequent feedback effects.

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↗
NASA NTRS1985

Particle distributions in approximately 10(13) - 10(16) eV air shower cores at mountain altitude and comparison with Monte Carlo simulations

Photographs of 521 shower cores in an array of current-limited spark (discharge) chambers at Sacramento Peak (2900m above sea level, 730 g /sq cm.), New Mexico, U.S.A., have been analyzed and the results compared with similar data from Leeds (80m above sea level, 1020 g sq cm.). It was found that the central density differential spectrum is consistent with a power law index of -2 up to approx. 1500/sq m where it steepens, and that shower cores become flatter on average with increasing size. Scaling model predictions for proton primaries with a approx E sup -2.71 energy spectrum account well for the altitude dependence of the data at lower densities. However, deviations at higher densities indicate a change in hadron interaction characteristics between approx few x 10 to the 14th power and 10 to the 15th power eV primary energy causing particles close to the shower axis to be spread further out.

Ash, A. G.↗