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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 811 records · Page 45
All sourcesNASA NTRSDOE OSTI
DOE OSTI2010

Clean Energy Policy Analyses: Analysis of the Status and Impact of Clean Energy Policies at the Local Level

This report takes a broad look at the status of local clean energy policies in the United States to develop a better understanding of local clean energy policy development and the interaction between state and local policies. To date, the majority of clean energy policy research focuses on the state and federal levels. While there has been a substantial amount of research on local level climate change initiatives, this is one of the first analyses of clean energy policies separate from climate change initiatives. This report is one in a suite of reports analyzing clean energy and climate policy development at the local, state, and regional levels.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2015

State of Modeling Symmetry in Hohlraums

Modeling radiation drive asymmetry is challenging problem whose agreement with data depends on the hohlraum gas fill density. Modeling to date uses the HYDRA code with crossbeam energy transfer (CBET) calculated separately, and backscattered light removed from the input laser. For high fill hohlraums (~>1 mg/cc), matching symmetry requires ad hoc adjustments to CBET during picket and peak of drive. For near-vacuum hohlraums, there is little CBET or backscatter, and drive is more waist-high than predicted. For intermediate fill densities (~0.6 mg/cc) there appears to be a region of small CBET and backscatter where symmetry is reasonably well modeled. A new technique where backscatter and CBET are done “inline” appears it could bring high fill simulations closer to data.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Sensitivity-Uncertainty Based Nuclear Criticality Safety Validation

These are slides from a seminar given to the University of Mexico Nuclear Engineering Department. Whisper is a statistical analysis package developed to support nuclear criticality safety validation. It uses the sensitivity profile data for an application as computed by MCNP6 along with covariance files for the nuclear data to determine a baseline upper-subcritical-limit for the application. Whisper and its associated benchmark files are developed and maintained as part of MCNP6, and will be distributed with all future releases of MCNP6. Although sensitivity-uncertainty methods for NCS validation have been under development for 20 years, continuous-energy Monte Carlo codes such as MCNP could not determine the required adjoint-weighted tallies for sensitivity profiles. The recent introduction of the iterated fission probability method into MCNP led to the rapid development of sensitivity analysis capabilities for MCNP6 and the development of Whisper. Sensitivity-uncertainty based methods represent the future for NCS validation – making full use of today’s computer power to codify past approaches based largely on expert judgment. Validation results are defensible, auditable, and repeatable as needed with different assumptions and process models. The new methods can supplement, support, and extend traditional validation approaches.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Procedure for Creating, Reviewing, and Submitting Input Files into the Benchmark Library

Los Alamos National Laboratory (LANL) has a number of criticality benchmark input files, modeled for use with MCNP, developed and maintained by various organizations including the Nuclear Criticality Safety (NCS) Division, XCP-3, and XCP-5. The purpose of this procedure is to standardize input file creation, review, and maintenance. This procedure incorporates existing practices used by IRSN as documented in the original report and the translation from French in addition to practices used by the LANL NCS Division as documented in. The report relates to input files used for the validation of the continuous energy Monte Carlo code MCNP6 but is general enough to be extended to other codes. This report describes the procedure on how to prepare and use benchmarks (taken mainly from the ICSBEP Handbook) for the validation of calculation routes. In particular, this report describes the procedure to follow in the creation of the input files and their integration in the experimental validation databases.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Internal Material Melting Development Progress [Slides]

Slides for an AFC workshop giving a brief overview of the current status of Bison code development for correct energy balance of melting material. The slides provide background information of why the development is needed. The planned approach of casting the energy balance as a change in enthalpy is described. Status of current code additions is outlined. Extra slides are provided with more mathematical details of the approach derivation and verification.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Development of Mechanistic Fission Gas Release and Swelling Models for UN Fuels in BISON

This report describes the work in NEAMS (Nuclear Energy Advanced Modeling and Simulation) to develop a mechanistic fission gas model for uranium nitride fuels in BISON. The existing Sifgrs (Simple integrated fission gas release and swelling) model tracks the average properties of two bubble populations in the bulk and at the grain boundaries. It was recognized that dislocations play a crucial role in the fission gas swelling of UN, and an irradiation-induced dislocation density model was needed, as well as a model describing how fission gas interacts with dislocations creating a third population of bubbles along dislocations. A mechanistic model that tracks dislocation bubbles was implemented in Sifgrs. This model was used to simulate fission gas swelling and release in UN. Lower-length-scale calculations and experimental observations from carbide fuel were leveraged to help populate the model with essential parameters. As a placeholder, an empirical function was formulated for the evolution of the dislocation network. Because of the difference in evolution of this network at different temperatures, the dislocation bubbles are able to capture behavior that the bulk intragranular bubbles cannot. It was found that the bulk bubbles dominate microscopic swelling at low temperatures, and the dislocation bubbles dominate at higher temperatures. Based on this model, the transition between the two bubble types is the main factor behind the breakaway swelling phenomenon in UN. In order to test the model, a couple of assessments were run. For the lower temperature JOYO pins, the model produces reasonable fission gas release and swelling values. For the higher temperature SP1 pin, the model dramatically underestimates the fission gas release. This issue is attributed to the gas being trapped inside dislocation bubbles, unable to escape to grain boundaries to cause fission gas release, and could be remedied by a mechanistic dislocation model that allows the dislocation density to decrease at very high temperatures. A preliminary mechanistic dislocation model was developed supported by first-principle calculations. These calculations provided valuable insight into how interstitial defects cluster in UN in the {110} orientation, and may eventually form dislocation loops, leading to the conclusion that dislocation loops may nucleate from these clusters. An estimate of the dislocation line energy in UN was obtained and will be improved in future work. The free-energy cluster dynamics code Centipede was used to track interstitial and vacancy absorption at dislocation loops and calculate their growth. In addition, improvements to Centipede were made including new convergence criteria for transient simulations and sink driving force updates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Neutronic Characteristics of ENDF/B-VIII.0 Compared to ENDF/B-VII.1 for Light-Water Reactor Analysis

The Evaluated Nuclear Data File (ENDF)/B-VIII.0 data library was released in 2018. To assess the new data during development and shortly after release, many validation calculations were performed with static, room-temperature benchmarks. Recently, when performing validation of ENDF/B-VIII.0 for pressurized water reactor depletion calculations, a regression in performance compared to ENDF/B-VII.1 was observed. This paper documents extensive benchmark calculations for light-water reactors performed using continuous energy Monte Carlo code with ENDF/B-VII.1 and -VIII.0 and neutronic characteristics of ENDF/B-VIII.0 are discussed and compared to those of ENDF/B-VII.1. It is our recommendation that ENDF/B data library assessment should include reactor-specific benchmark assessments, including depletion cases, such that these types of regressions may be caught earlier in the data development cycle.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Pulsed-Neutron Experiments at the Inherently Safe Subcritical Assembly

The pulsed neutron technique is a powerful, dynamic method to assay the reactivity of a multiplying system. This work presents the novel application of the pulsed neutron technique to the Inherently Safe Subcritical Assembly, an experimental configuration accepted by the International Criticality Safety Benchmark Evaluation Project Handbook. The experiments were replicated with COG11.3, a continuous-energy Monte Carlo code. The pulsed neutron data were analyzed using the Sjöstrand and Gozani area-ratio methods and by extracting the prompt neutron decay constant. Subsequent static k-eigenvalue and 𝛼-eigenvalue simulations were also performed for the same configurations. Neutron detector dead-time effects from the experiments were corrected using the Backwards Extrapolation Method and shown to have a negligible impact on the estimated reactivities. The results highlight that capturing time-dependent effects like delayed neutron precursor buildup are essential to accurately reproduce experimental results. They also highlight the importance of shielding the detectors from generator source neutrons in deeply subcritical configurations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Time stretching of the GeV emission of GRBs: Fermi-LAT data versus geometrical model

It is known that the high-energy (>100 MeV) emission of gamma-ray bursts is delayed with respect to the low-energy emission. However, the dependence of light curves on energy has not been studied for the high-energy bands. In this paper, we consider the bursts observed by Fermi LAT from 2008 August 4 to 2011 August 1, for which at least 10 photons were observed with energy greater than 1 GeV. These include four bursts: GRB 080916C, GRB 090510, GRB 090902B, and GRB 090926A. We use the Kolmogorov–Smirnov test to compare the light curves in the two bands 100 MeV 1 GeV. For GRB 080916C and GRB 090510 the light curves in the two bands are statistically compatible. However, for GRB 090926A, the higher-energy light curve is stretched compared to the lower-energy one with a statistical significance of 3.3σ and, for GRB 090902B, on the contrary, the lower-energy curve is stretched with 2.3σ significance. We argue that the observed diversity of stretching factors may be explained using a simple geometrical model. The model assumes that the jet opening angle depends on the emission energy in a way that the most energetic photons are radiated near the axis of the jet. All of the bursts are considered equivalent in their rest frames, and the observed light curves differ only due to different redshifts and view directions. The model conforms to the total burst energy constraint and matches the Fermi-LAT observations of the fraction of GRBs visible in the 100 MeV<E<1 GeV band, which may be observed at higher energies. The model predicts the distribution of observable stretching factors, which may be tested in future data. Finally, we propose a method to estimate the observer's off-axis angle based on the stretching factor and the fraction of the high-energy photons. The code for modeling is open source and is publicly available on GitHub (https://github.com/maxitg/GammaRays).

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2011

Nuclear Data Performance Testing Using Sensitive, but Less Frequently Used ICSBEP Benchmarks

The International Criticality Safety Benchmark Evaluation Project (ICSBEP) has published the International Handbook of Evaluated Criticality Safety Benchmark Experiments annually since 1995. The Handbook now spans over 51,000 pages with benchmark specifications for 4,283 critical, near critical, or subcritical configurations; 24 criticality alarm placement/shielding configurations with multiple dose points for each; and 200 configurations that have been categorized as fundamental physics measurements relevant to criticality safety applications. Benchmark data in the ICSBEP Handbook were originally intended for validation of criticality safety methods and data; however, the benchmark specifications are now used extensively for nuclear data testing. There are several, less frequently used benchmarks within the Handbook that are very sensitive to thorium and certain key structural and moderating materials. Calculated results for many of those benchmarks using modern nuclear data libraries suggest there is still room for improvement. These and other highly sensitive, but rarely quoted benchmarks are highlighted and data testing results provided using the Monte Carlo N-Particle Version 5 (MCNP5) code and continuous energy ENDF/B-V, VI.8, and VII.0, JEFF-3.1, and JENDL-3.3 nuclear data libraries.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Updating the Building Science Advisor (BSA): A Tool to Assist in the Design of Durable Building Envelopes

Predicting the moisture durability of building envelope components remains challenging due to multiple influencing factors, including material selection, assembly positioning, local climate conditions, air tightness, interior environment, and construction quality. Building codes increasingly emphasize energy efficiency through enhanced insulation and tighter envelopes but offer limited guidance on moisture durability considerations. Consequently, builders face uncertainty, particularly as new materials and assemblies enter the market.The Building Science Advisor (BSA) is a free, web-based expert system developed to address these challenges by providing actionable insights into the moisture durability and energy efficiency of both new and retrofit wall designs. Recently updated, we are now providing version 3.0 of the tool. BSA features significant user interface improvements, enhancing navigation and user interaction through a refreshed, intuitive design. Additionally, the tool incorporates a newly developed database containing pre-simulated wall assembly cases, significantly reducing response times and improving the accuracy of moisture durability assessments. Furthermore, the updated BSA includes moisture content as a performance criterion, providing users with a more comprehensive understanding of moisture-related durability risks. These enhancements enable rapid, reliable assessments tailored to specific climate zones and local building practices. BSA continues to offer targeted guidance on wall retrofit scenarios and delivers access to an expanded library of location-specific building science resources.This paper describes these key updates, highlighting the enhanced features, expanded capabilities, and overall improvements to user experience and educational content. The paper includes a demonstration that illustrates how the revised BSA effectively supports practitioners in designing durable, energy-efficient building envelope assemblies.

Salonvaara, Mikael [ORNL] (ORCID:0000000318991554)↗
DOE OSTI2016

SINGLE PHASE ANALYTICAL MODELS FOR TERRY TURBINE NOZZLE

All BWR RCIC (Reactor Core Isolation Cooling) systems and PWR AFW (Auxiliary Feed Water) systems use Terry turbine, which is composed of the wheel with turbine buckets and several groups of fixed nozzles and reversing chambers inside the turbine casing. The inlet steam is accelerated through the turbine nozzle and impacts on the wheel buckets, generating work to drive the RCIC pump. As part of the efforts to understand the unexpected “self-regulating” mode of the RCIC systems in Fukushima accidents and extend BWR RCIC and PWR AFW operational range and flexibility, mechanistic models for the Terry turbine, based on Sandia National Laboratories’ original work, has been developed and implemented in the RELAP-7 code to simulate the RCIC system. RELAP-7 is a new reactor system code currently under development with the funding support from U.S. Department of Energy. The RELAP-7 code is a fully implicit code and the preconditioned Jacobian-free Newton-Krylov (JFNK) method is used to solve the discretized nonlinear system. This paper presents a set of analytical models for simulating the flow through the Terry turbine nozzles when inlet fluid is pure steam. The implementation of the models into RELAP-7 will be briefly discussed. In the Sandia model, the turbine bucket inlet velocity is provided according to a reduced-order model, which was obtained from a large number of CFD simulations. In this work, we propose an alternative method, using an under-expanded jet model to obtain the velocity and thermodynamic conditions for the turbine bucket inlet. The models include both adiabatic expansion process inside the nozzle and free expansion process out of the nozzle to reach the ambient pressure. The combined models are able to predict the steam mass flow rate and supersonic velocity to the Terry turbine bucket entrance, which are the necessary input conditions for the Terry Turbine rotor model. The nozzle analytical models were validated with experimental data and benchmarked with CFD simulations. The analytical models generally agree well with the experimental data and CFD simulations. The analytical models are suitable for implementation into a reactor system analysis code or severe accident code as part of mechanistic and dynamical models to understand the RCIC behaviors. The cases with two-phase flow at the turbine inlet will be pursued in future work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Optimization of Monte Carlo transport simulations in stochastic media

This paper presents an accurate and efficient approach to optimize radiation transport simulations in a stochastic medium of high heterogeneity, like the Very High Temperature Gas-cooled Reactor (VHTR) configurations packed with TRISO fuel particles. Based on a fast nearest neighbor search algorithm, a modified fast Random Sequential Addition (RSA) method is first developed to speed up the generation of the stochastic media systems packed with both mono-sized and poly-sized spheres. A fast neutron tracking method is then developed to optimize the next sphere boundary search in the radiation transport procedure. In order to investigate their accuracy and efficiency, the developed sphere packing and neutron tracking methods are implemented into an in-house continuous energy Monte Carlo code to solve an eigenvalue problem in VHTR unit cells. Comparison with the MCNP benchmark calculations for the same problem indicates that the new methods show considerably higher computational efficiency. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Evaluation for 4S core nuclear design method through integration of benchmark data

The 4S is a sodium-cooled small fast reactor which is reflector-controlled for operation through core lifetime about 30 years. The nuclear design method has been selected to treat neutron leakage with high accuracy. It consists of a continuous-energy Monte Carlo code, discrete ordinate transport codes and JENDL-3.3. These two types of neutronic analysis codes are used for the design in a complementary manner. The accuracy of the codes has been evaluated by analysis of benchmark critical experiments and the experimental reactor data. The measured data used for the evaluation is critical experimental data of the FCA XXIII as a physics mockup assembly of the 4S core, FCA XVI, FCA XIX, ZPR, and data of experimental reactor JOYO MK-1. Evaluated characteristics are criticality, reflector reactivity worth, power distribution, absorber reactivity worth, and sodium void worth. A multi-component bias method was applied, especially to improve the accuracy of sodium void reactivity worth. As the result, it has been confirmed that the 4S core nuclear design method provides good accuracy, and typical bias factors and their uncertainties are determined. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Benchmarking ENDF/B-VII.1, JENDL-4.0 and JEFF-3.1

Three nuclear data libraries have been tested extensively using criticality safety benchmark calculations. The three libraries are the new release of the US library ENDF/B-VII.1 (2011), the new release of the Japanese library JENDL-4.0 (2011), and the OECD/NEA library JEFF-3.1 (2006). All calculations were performed with the continuous-energy Monte Carlo code MCNP (version 4C3, as well as version 6-beta1). Around 2000 benchmark cases from the International Handbook of Criticality Safety Benchmark Experiments (ICSBEP) were used. The results were analyzed per ICSBEP category, and per element. Overall, the three libraries show similar performance on most criticality safety benchmarks. The largest differences are probably caused by elements such as Be, C, Fe, Zr, W. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Calculation of neutron and gamma fluxes in support to the interpretation of measuring devices irradiated in the core periphery of the OSIRIS Material Testing Reactor

Technological irradiations carried out in material testing reactors (MTRs) are used to study the behavior of materials under irradiation conditions required by different types of nuclear power plants (NPPs). For MTRs, specific instrumentation is required for the experiment monitoring and for the characterization of irradiation conditions, in particular the flux of neutrons and photons. To measure neutron and photon flux in experimental locations, different sensors can be used, such as SPNDs (self-powered neutron detectors), SPGDs (self-powered gamma detectors) and ionization chambers. These sensors involve interactions producing ultimately a measurable electric current. Various sensors have been recently tested in the core periphery of the OSIRIS reactor (located at the CEA-Saclay center) in order to qualify their responses to the neutron and the photon flux. One of the key input data for this qualification is to have a relevant evaluation of neutron and gamma fluxes at the irradiation location. The objective of this work is to evaluate the neutron and the gamma flux in the core periphery of the OSIRIS reactor. With this intention, specific neutron-photonic three-dimensional calculations have been performed and are mainly based on the TRIPOLI-4{sup R} three-dimensional continuous-energy Monte Carlo code, developed by CEA (Saclay Center) and extensively validated against reactor dosimetry benchmarks. In the case of the OSIRIS reactor, TRIPOLI-4{sup R} code has been validated against experimental results based on neutron flux and nuclear heating measurements performed in ex-core and in-core experiments. In this work, simultaneous contribution of neutrons and gamma photons in the core periphery is considered using neutron-photon coupled transport calculations. Contributions of prompt and decay photons have been taken into account for the gamma flux calculation. Specific depletion codes are used upstream to provide the decay-gamma sources required by TRIPOLI-4 calculations. In this paper, the neutron-photonic calculation scheme used for the neutron-gamma flux evaluation in the OSIRIS core periphery will be first detailed. Then, a comparison of calculated and measured values of neutron and gamma fluxes will be introduced. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

A new moment-based general-relativistic neutrino-radiation transport code: Methods and first applications to neutron star mergers

ABSTRACT We present a new moment-based energy-integrated neutrino transport code for neutron star merger simulations in general relativity. In the merger context, ours is the first code to include Doppler effects at all orders in υ/c, retaining all non-linear neutrino–matter coupling terms. The code is validated with a stringent series of tests. We show that the inclusion of full neutrino–matter coupling terms is necessary to correctly capture the trapping of neutrinos in relativistically moving media, such as in differentially rotating merger remnants. We perform preliminary simulations proving the robustness of the scheme in simulating ab-initio mergers to black hole collapse and long-term neutron star remnants up to ${\sim }70\,$ ms. The latter is the longest dynamical space-time, 3D, general relativistic simulations with full neutrino transport to date. We compare results obtained at different resolutions and using two different closures for the moment scheme. We do not find evidences of significant out-of-thermodynamic equilibrium effects, such as bulk viscosity, on the post-merger dynamics or gravitational wave emission. Neutrino luminosities and average energies are in good agreement with theory expectations and previous simulations by other groups using similar schemes. We compare dynamical and early wind ejecta properties obtained with M1 and with our older neutrino treatment. We find that the M1 results have systematically larger proton fractions. However, the differences in the nucleosynthesis yields are modest. This work sets the basis for future detailed studies spanning a wider set of neutrino reactions, binaries, and equations of state.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2015

An Update on NiCE Support for BISON

The Nuclear Energy Advanced Modeling and Simulation program (NEAMS) from the Department of Energy s Office of Nuclear Energy has funded the development of a modeling and simulation workflow environment to support the various codes in its nuclear energy scientific computing toolkit. This NEAMS Integrated Computational Environment (NiCE) provides extensible tools and services that enable efficient code execution, input generation, pre-processing visualizations, and post-simulation data analysis and visualization for a large portion of the NEAMS Toolkit. A strong focus for the NiCE development team throughout FY 2015 has been support for the Multiphysics Object Oriented Simulation Environment (MOOSE) and the NEAMS nuclear fuel performance modeling application built on that environment, BISON. There is a strong desire in the program to enable and facilitate the use of BISON throughout nuclear energy research and industry. A primary result of this desire is the need for strong support for BISON in NiCE. This report will detail improvements to NiCE support for BISON. We will present a new and improved interface for interacting with BISON simulations in a variety of ways: (1) improved input model generation, (2) embedded mesh and solution data visualizations, and (3) local and remote BISON simulation launch. We will also show how NiCE has been extended to provide support for BISON code development.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗