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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 487 records · Page 27
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

FORCE Development Status Update: Vertical Integration and Benchmarking of System Dynamics

Recent efforts to establish effective models for grid energy analysis, especially given the increase in variable renewable energy (VRE) sources and the economic challenges faced by traditional nuclear energy, have generated new technological considerations. One effort to improve the economic viability of nuclear power involves investigating integrated energy systems (IES) which include secondary energy systems that introduce flexibility and secondary market possibilities to existing and perceived future nuclear energy generation technologies. To analyze the technical and economic potential of IES, the Framework for Optimization of Resources and Economics (FORCE) tool suite was developed through a collaboration among national laboratories. Within the FORCE tool suite, the Holistic Energy Resource Optimization Network (HERON) provides algorithms for analyzing the long-term viability of potential IES technologies, while HYBRID provides algorithms and models to achieve high-resolution analysis of coupling physics over a short time period. Continued maturing of the FORCE tool suite requires further interconnections between the various tools in the suite in order to ensure consistent analysis. Analyses performed by applying HYBRID to transient process modeling should be easily harvestable as inputs to HERON analyses. The first item in this status update is a demonstration of an automated data pipeline for loading data from HYBRID into HERON analyses. Application of HYBRID results to HERON, as part of using the FORCE tool suite, relies on robust modeling assumptions for the various models included in HYBRID. The second result of this status update is the benchmarking and validation of cost and operational data, with a particular focus on natural gas energy generators. These generators are benchmarked with a focus on contrasting them with proposed thermal energy storage (TES) technologies.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2017

International Developments on Uncertainties in NPP Decommissioning costs and planned work on Benchmarking at the OECD-NEA - 17074

The ability to decommission NPPs shut down after normal operation has been proven technically. Issues of current concern include the ability to accurately calculate and demonstrate the validity of decommissioning cost estimates, and to control costs during decommissioning. This paper describes recent international developments concerning addressing uncertainties in NPP decommissioning cost estimation, and work currently being planned relating to decommissioning cost benchmarking at the OECD Nuclear Energy Agency. The process of decommissioning cost estimation is evolving, with a general trend towards showing greater levels of detail in the estimate. The International Structure for Decommissioning Costing was published in 2012 and presents a common reporting format for decommissioning costing undertaken on a deterministic basis. However, this does not address probabilistic methods or associated presentation formats for the inclusion of uncertainties in decommissioning estimates. A consistent and comparable treatment of uncertainties in decommissioning cost estimation would further facilitate comparison between different cost estimates and enhance understanding of and confidence in the estimates themselves. There is growing interest in benchmarking decommissioning costs, reflecting a need to better understand the relationships between estimates and actual costs, to give greater confidence that funding will be sufficient, and as a useful tool in project management and enhancing project delivery. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Engaging Canadian Indigenous Communities and Public Stakeholders Through a US Site Benchmarking Trip - 19358

Canadian Nuclear Laboratories (CNL) is proposing a new approach to decommission the WR-1 reactor at its Whiteshell site near Pinawa, Manitoba Canada. To tackle WR-1 CNL is proposing in situ decommissioning, this new approach is subject to a Canadian federal environmental assessment. As part of this CNL is engaging with Indigenous communities and public stakeholders, both regionally and across the country. Nuclear in situ decommissioning is new to Canada but has been used numerous times in the United States. Throughout the environmental assessment process CNL has considered the feedback public and Indigenous groups have provided and has tried to enhance the way information is shared and respond where possible and reasonable. Feedback from some communities has been related to a greater understanding of what the in situ method looks like, how it is monitored, will it perform, and anxiety about having it in the community. CNL sought to respond to this feedback by coordinating a benchmarking trip to an in situ reactor in Hallam, Nebraska. The visit was coordinated around an annual inspection. Several regional stakeholder and Indigenous communities joined CNL on the benchmarking trip. The group witnessed and learned what a decommissioned in situ reactor looked like, how it was performing and the effect on the surrounding communities. The group learned important lessons during the visit, key among them that the entombment is working as designed and has had zero significant detectable effects to the surrounding community. All attendees are now better informed on what CNL's proposal for WR-1 represents and are in a position to better inform their communities and the process moving forward. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Subcritical Benchmark of the BeRP Ball Reflected by Tungsten

In September 2012, eight subcritical configurations were measured with the clad a-phase plutonium sphere known as the Beryllium-Reflected Plutonium (BeRP) ball at the National Criticality Experiments Research Center (NCERC). The configurations had varying tungsten thicknesses including: 0 (bare), 0.5, 1.0, 1.5, 2.0, 2.5, 2.75, and 3.0 inches. Measurements were conducted to provide subcritical benchmark data. This work includes measured and simulated results from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP) evaluation. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Benchmark Comparisons of Lattice Physics Calculations for Thorium-Based Fuels in Pressure-Tube Heavy Water Reactors

Thorium is a fertile nuclear fuel resource that is nearly three times as abundant as uranium in the earth's crust and that can be exploited and used to extend uranium resources. The pressure tube heavy water reactor (PTHWR) is an excellent existing technology for utilizing thorium-based fuels, due to its high neutron economy achieved by using heavy water as a moderator and coolant, the ability to perform online refueling, and its flexibility in fuel management. Reactor physics studies are underway at the Canadian Nuclear Laboratories (CNL) to evaluate several types of thorium-based fuels that could be implemented in PT-HWRs in the near-term and intermediate future, building upon past experience. For preliminary, conceptual studies, deterministic lattice physics and core physics codes such as WIMS-AECL and RFSP are used to provide quick and more approximate estimates of performance and safety characteristics. The objective of this study was to perform 2D lattice physics benchmark comparisons between WIMS-AECL and MCNP5 for several types of thorium-based fuel bundle lattices for potential use in PT-HWRs. Benchmark comparisons will help confirm the validity of deterministic calculation results, and provide estimates of the potential magnitude of calculation errors and uncertainties. Areas for potential improvements in the WIMS-AECL calculations may also be identified. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Uncertainty Quantification by Monte Carlo Analysis using CFD Simulations for GEMIX Benchmark Activities

Uncertainty quantification (UQ) methods can be applied to computational fluid dynamics (CFD) simulations for determining the uncertainty bands of the simulations results. In order to determine the applicability of these different methods, simple benchmark activities have been conducted by various group for comparison. The GEMIX (Generic Mixing experiment) test facility was used to develop benchmark data for this purpose by the Paul Scherrer Institute (PSI). The facility involves the mixing of two fluids that are initially separated. One of the two fluids is water with sugar dissolved in it to create a fluid with similar fluid properties to water, except for density. The other fluid was distilled water to reduce the uncertainty in the fluid's properties. During the experiments, velocity fields and concentration measurements were acquired at several planes of interest. The velocity fields were acquired using particle image velocimetry (PIV) and concentrations of each fluid were acquired using laser induced fluorescence (LIF). The inlet velocity profiles were acquired using laser Doppler velocimetry (LDV). The activity provided the opportunity to vary turbulence models, meshing methodologies, and numerical schemes for creation of uncertainty estimates. The present study investigates the usage of a simple Monte Carlo (MC) technique for varying the inlet boundary conditions of each inlet. The velocity profiles and densities of each inlet were specifically varied based on experiment measurement uncertainty. The intent was to isolate the uncertainty due to inlet quantities and not related to meshing, numerical scheme, or turbulence models. The comparisons of experimental and simulation results were found to be reasonably accurate. Though deficiencies were identified in the modeling approach to be corrected in future works by the authors. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Evaluation of Rapid for a UNF cask benchmark problem

To ensure criticality safety and maintain a high level of security of Used Nuclear Fuel (UNF), it is necessary to perform detailed Particle Transport simulations. The commonly used technique is the 3-D, continuous energy Monte Carlo method that is time-consuming and, for eigenvalue calculations, may have difficulties with the source convergence. Wenner and Haghighat and Dufek demonstrated that in criticality problems with high dominance ratio and/or under-sampling possibilities, the fission matrix (FM) methodology should be used in order to be able to obtain unbiased solution. The RAPID (Real-time Analysis for Particle transport and In-situ Detection) code system is based on the Multi-stage Response-function Transport (MRT) methodology, and uses the FM formulation to determine eigenvalue, subcritical multiplication and fission density distribution of nuclear systems. This paper uses a benchmark problem similar to the GBC-32 Cask to evaluate the accuracy and performance of RAPID against the results of a reference MCNP solution. We performed a detailed sensitivity analysis on a single-assembly model in order to determine the MCNP eigenvalue parameters necessary to obtain a reliable reference solution. These results provided a guide for the selection of an appropriate set of parameters for the eigenvalue calculation for the whole cask. The paper is organized as follows: initially, the RAPID code system is briefly discussed; then, the benchmark problem is introduced. The methodology used to analyze results is proposed and some of the results of different cases are analyzed to identify the best set of eigenvalue parameters for a reference MCNP calculation. The selected case is then compared with the RAPID calculation for both a single assembly and a full cask model. In the end some conclusions and future work perspectives are drawn.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

RELAP5-3D Solutions to Exercise 1 of the OECD-NEA HTTF Benchmark

The high-temperature gas-cooled reactor (HTGR) is an advanced reactor concept that has received considerable attention over the last 60 years. HTGRs boast high coolant outlet temperatures, passive safety, and large margins to fuel failure. Recent years have seen increasing interest in HTGRs, whether large-scale HTGRs for electricity or process heat applications or as microreactors serving remote communities. HTGRs have considerable potential to fulfill these applications, but modeling and simulation tools typically used for reactor safety analysis have not been validated for HTGR modeling. The High-Temperature Test Facility (HTTF) was constructed at Oregon State University (OSU) to provide validation data for HTGR modeling. Recently, Idaho National Laboratory (INL), OSU, Argonne National Laboratory, Canadian Nuclear Laboratories, and the University of Tennessee Knoxville have collaborated to develop an HTGR thermal hydraulics benchmark based on HTTF experiments. This summary presents results from a portion of that benchmark using RELAP5-3D

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Whole core analysis of OECD/NEA TVA-WB1 benchmark exercise 1 using continuous-energy Monte Carlo code MVP with JENDL-4.0u and ENDF/B-VII.0

The TVA WB1 benchmark problem was precisely modeled in accordance with the geometry and material specifications, and benchmark exercise 1 was solved with the MVP code and the nuclear data libraries based on JENDL-4.0u (JENDL-4.0 updated file) and ENDF/B-VII.0. The criticality, the control rod bank worth, the isothermal temperature coefficient (ITC), the integral rod worth of bank D, and differential soluble boron worth (DBW) were obtained. The criticality and the control rod bank worth were compared with the measured data and published calculation results with other codes, Serpent and KENO. The criticality and the control rod bank worth calculated by MVP and the libraries agree with the measured and the published calculation results within the criteria specified in the ANSI/ANS standard. From these results, the calculation model for the MVP and the libraries was verified. The other parameters such as ITC, the integral rod worth, and DBW were also compared between JENDL-4.0u and ENDF/B-VII.0. The comparison of the results shows JENDL-4.0u and ENDF/B-VII.0 give comparable results within the standard deviation, the difference between libraries is small. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

OECD/NEA MPCMIV Benchmark - Preliminary fuel performance results

The on-going OECD/NEA Multi-physics Pellet Cladding Mechanical Interaction Validation (MPCMIV) benchmark aims to provide guidance on multi-physics validation through the modelling of two cold ramps. In this paper, the first results for the base irradiation of the father rod and fuel rodlet (refabricated from the father rod), and the first cold ramp are presented. The base irradiation consists of 3 years of irradiation in the Forsmark-2 reactor. The cold ramp encompasses a steady-state pre-ramp period of less than one hour at a low constant linear heat rate (LHR) followed by a ramp test (< 1 min) with a much higher maximal LHR. The base irradiation is modelled using the fuel performance codes FRAPCON and FAST, while the cold ramp modeling is using the fuel performance code FRAPTRAN. Several missing parameters for FRAPCON/FAST base irradiation models have been selected using multiple references such as the OECD/NEA light water reactor Uncertainty Analysis in Modelling (UAM) benchmark specifications and the FRAPCON Integral Assessment report. The obtained results for the base irradiation such as the cladding outer diameter show reasonable agreement with the experimental measurements. The results for the cold ramp show larger differences with the experimental measurements (e.g., cladding axial elongation). However, such differences have been observed as well in other studies involving pellet cladding mechanical interaction analyses and are attributed to the fuel performance modelling assumptions and to the tuned modelling parameters that were not covered by the specifications. Uncertainty and sensitivity analysis might allow a better quantification of these missing parameters. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

High Temperature Gas-Cooled Reactor Thermal Hydraulics Benchmark Problem 2 Description: Depressurized Conduction Cooldown

Problem 2 of the HTGR T/H Benchmark represents the depressurized conduction cooldown transient and is based on HTTF experiment PG-29. The problem includes an exercise for code-to-code comparison, an exercise for code-to-data comparison, and an exercise for error scaling. This presentation provides an overview of the problem description and provides benchmark participants an opportunity to ask questions about the problem definition.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Code Benchmark of Depressurized Conduction Cooldown Transient in the High Temperature Test Facility

This paper presents results from modeling of a depressurized conduction cooldown (DCC) transient at the High Temperature Test Facility (HTTF) as part of the OECD-NEA Thermal Hydraulics Code Validation Benchmark for High-Temperature Gas-Cooled Reactors using HTTF Data . This paper briefly describes the benchmark and the models being used. It then presents a comparison of steady state and transient results based on the Problem 2 Exercise 1A and 1B definitions. We compare block and helium temperature distributions, mass flow distribution, and energy balance in steady state. All models show comparable mass flow distributions and energy balances. The temperatures within the core and outer regions are comparable in all models too, but inner reflector temperatures can vary significantly. Despite that, we find that the models are in good agreement for the full-power steady state. In the DCC, we look at block temperature at the core midplane and RCCS water exit temperature. The INL and ANL models are found to be in excellent agreement with one another on block temperature over time, while the agreement when the KAERI and NRG models are added into consideration is good. Differences in the transient heat removal from the RCCS cause the differences in block temperature over time in these models. The CNL models show similar trends to the INL, ANL, KAERI, and NRG models, but the temperatures are high because the volumes used in calculating the average temperature include the heater rods in the CNL models only. The HUN-REN model shows results that suggest significantly lower heat removal in the RCCS which merit further investigation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Progress in the HTTF Benchmark and RELAP5-3D Gas-Cooled Reactor Validation

This slide set presents results of analysis done in the first year after the kickoff of the HTTF benchmark. This work includes the development of a new RELAP5-3D model of the facility and collection and presentation of results from one of the exercises in the benchmark. Highlights include good agreement between the new and legacy models for full-power steady state and similar predictions of maximum block temperature during a pressurized conduction cooldown. Additionally, code-to-code comparisons of a full-power steady state and depressurized conduction cooldown (DCC) from full-power steady state as part of Problem 2 Exercises 1A and 1B show good agreement on temperatures predicted in the core but considerable spread in central reflector temperatures. During the DCC, temperature behavior is similar across all models, but cooldown rates are dictated by the performance of the reactor cavity cooling system

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

International collaboration framework for the calculation of performance loss rates: Data quality, benchmarks, and trends (towards a uniform methodology)

Abstract The IEA PVPS Task 13 group, experts who focus on photovoltaic performance, operation, and reliability from several leading R&D centers, universities, and industrial companies, is developing a framework for the calculation of performance loss rates of a large number of commercial and research photovoltaic (PV) power plants and their related weather data coming across various climatic zones. The general steps to calculate the performance loss rate are (i) input data cleaning and grading; (ii) data filtering; (iii) performance metric selection, corrections, and aggregation; and finally, (iv) application of a statistical modeling method to determine the performance loss rate value. In this study, several high‐quality power and irradiance datasets have been shared, and the participants of the study were asked to calculate the performance loss rate of each individual system using their preferred methodologies. The data are used for benchmarking activities and to define capabilities and uncertainties of all the various methods. The combination of data filtering, metrics (performance ratio or power based), and statistical modeling methods are benchmarked in terms of (i) their deviation from the average value and (ii) their uncertainty, standard error, and confidence intervals. It was observed that careful data filtering is an essential foundation for reliable performance loss rate calculations. Furthermore, the selection of the calculation steps filter/metric/statistical method is highly dependent on one another, and the steps should not be assessed individually.

14 SOLAR ENERGY↗
DOE OSTI2020

Verification benchmarks for single-phase flow in three-dimensional fractured porous media

Flow in fractured porous media occurs in the earth’s subsurface, in biological tissues, and in man-made materials. Fractures have a dominating influence on flow processes, and the last decade has seen an extensive development of models and numerical methods that explicitly account for their presence. To support these developments, four benchmark cases for single-phase flow in three-dimensional fractured porous media are presented. Furthermore, the cases are specifically designed to test the methods’ capabilities in handling various complexities common to the geometrical structures of fracture networks. Based on an open call for participation, results obtained with 17 numerical methods were collected. This paper presents the underlying mathematical model, an overview of the features of the participating numerical methods, and their performance in solving the benchmark cases.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

Development of a Benchmark Eddy Flux Evapotranspiration Dataset for Evaluation of Satellite-Driven Evapotranspiration Models Over the CONUS

A large sample of ground-based evapotranspiration (ET) measurements made in the United States, primarily from eddy covariance systems, were post-processed to produce a benchmark ET dataset. The dataset was produced primarily to support the intercomparison and evaluation of the OpenET satellite-based remote sensing ET (RSET) models and could also be used to evaluate ET data from other models and approaches. OpenET is a web-based service that makes field-delineated and pixel-level ET estimates from well-established RSET models readily available to water managers, agricultural producers, and the public. The benchmark dataset is composed of flux and meteorological data from a variety of providers covering native vegetation and agricultural settings. Flux footprint predictions were developed for each station and included static flux footprints developed based on average wind direction and speed, as well as dynamic hourly footprints that were generated with a physically based model of upwind source area. The two footprint prediction methods were rigorously compared to evaluate their relative spatial coverage. Data from all sources were post-processed in a consistent and reproducible manner including data handling, gap-filling, temporal aggregation, and energy balance closure correction. The resulting dataset included 243,048 daily and 5,284 monthly ET values from 194 stations, with all data falling between 1995 and 2021. We assessed average daily energy imbalance using 172 flux sites with a total of 193,021 days of data, finding that overall turbulent fluxes were understated by about 12% on average relative to available energy. Multiple linear regression analyses indicated that daily average latent energy flux may be typically understated slightly more than sensible heat flux. This dataset was developed to provide a consistent reference to support evaluation of RSET data being developed for a wide range of applications related to water accounting and water resources management at field to watershed scales.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

A benchmarking study of Geant4 for Auger electrons emitted by medical radioisotopes

Auger emitting radioisotopes are of great interest in targeted radiotherapy because, once internalised in the tumour cells, they can deliver dose locally to the radiation sensitive targets, while not affecting surrounding cells. Geant4 is a Monte Carlo code widely used to characterise the physics mechanism at the basis of targeted radiotherapy. Here, we benchmarked the modelling of the emission of Auger electrons in Geant4 deriving from the decay of 123 I, 124 I, 125 I radionuclides against existing theoretical approaches. We also compared Geant4 against reference data in the case of 131 Cs, which is of interest for brachytherapy. In the case of 125 I and 131 Cs, the simulation results are compared to experimental measurements as well. Good agreement was found between Geant4 and the reference data. As far as we know, this is the first study aimed to benchmark against experimental measurements the emission of Auger electrons in Geant4 for radiotherapy applications.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

A benchmark of Gutzwiller conjugate gradient minimization method in ground state energy calculations of dimers

Herein we present numerical results of ground-state energies of 9 molecules in the well-established G2 molecule set given by the Gutzwiller conjugate gradient minimization (GCGM) method. The method, beyond the commonly used Gutzwiller approximation, was recently developed based on Gutzwiller variational wave functions. We find that compared to benchmark data given by full configuration interaction, GCGM total energies are reasonably well reproduced with the minimum basis set. To include the dynamical correlation beyond the minimal basis calculations, we adopt the local density approximation for the dynamical correlation energy $E_c$. By comparing the results with benchmark data given by experiments and large-basis configuration interaction, the GCGM total energies with $E_c$ are in general better reproduced, but discrepancies are still observed for some dimers.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗