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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 577 records · Page 32
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Review of CTF's Fuel Rod Modeling Using FRAPCON-4.0's Centerline Temperature Predictions

Coolant Boiling in Rod Arrays-Two Fluid (COBRA-TF), or CTF1, is a nuclear thermal hydraulic subchannel code used throughout academia and industry. CTF's fuel rod modeling is originally developed for VIPRE code. Its methodology is based on GAPCON and FRAP fuel performance codes, and material properties are included from MATPRO handbook. This work focuses on review of CTF's fuel rod modeling to address shortcomings in CTF's temperature predictions. CTF is compared to FRAPCON which is U.S. NRC's steady-state fuel performance code for light-water reactor fuel rods. FRAPCON calculates the changes in fuel rod variables and temperatures including the effects of cladding hoop strain, cladding oxidation, hydriding, fuel irradiation swelling, densification, fission gas release and rod internal gas pressure. It uses fuel, clad and gap material properties from MATPRO. Additionally, it has its own models for fission gas release, cladding corrosion and cladding hydrogen pickup. It allows finite difference or finite element approaches for its mechanical model. In this study, FRAPCON-4.0 is used as a reference fuel performance code. In comparison, Halden Reactor Data for IFA432 Rod 1 and Rod 3. CTF simulations are performed in two ways; informing CTF with gap conductance value from FRAPCON, and using CTF's dynamic gap conductance model. First case is chosen to show temperature is predicted correctly with CTF's models for thermal and cladding conductivities once gap conductance is provided. Latter is to review CTF's dynamic gap conductance model. These Halden test cases are selected to be representative of cases with and without any physical contact between fuel-pellet and clad while reviewing functionality of CTF's dynamic gap conductance model. Improving the CTF's dynamic gap conductance model will allow prediction of fuel and cladding thermo-mechanical behavior under irradiation, and better temperature feedbacks from CTF in transient calculations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Measurements of fission gas release from nuclear fuel in support of BISON fuel performance analysis

This project established a fuel irradiation and fission gas release (FGR) loop that has been designed for implementation at the PULSTAR reactor of North Carolina State University (NCSU). The FGR loop is capable of analyzing various types of fuel samples including, e.g., uranium dioxide fuel (UO 2 ) and uranium silicide (U 3 Si 2 ) samples. The fuel samples were produced and characterized at Idaho National Laboratory (INL) and shipped NCSU. The samples are ready for irradiation in the FGR loop to a given total thermal neutron fluence and under controlled temperature conditions. An innovative high resolution gamma-ray spectrometry approach and instrumentation have been established to monitor the loop’s sweep gas and perform high accuracy measurements of the gas release rates relative to predicted birth rates to establish the release-to-birth ratio (R/B) for a given radionuclide. In addition, uncertainty minimization is achieved through the formulation of the measurement observables into relative release-to-birth indicators for selected radionuclides. The experimental findings will be utilized to enhance and validate the fission gas release models that are currently implemented in the BISON fuel performance code. The experimentally deduced physical parameters will be incorporated in BISON’s fission gas release models. Atomistic simulations of fuel behavior have been performed to investigate the mechanisms of fission gas migration in the fuel. At this stage, performance of the fuel irradiation experiments awaits licensing approval by the US NRC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

AGR-5/6/7 Irradiation Experiment Test Plan

This document presents the test plan for the AGR-5/6/7 irradiation experiment, the combined fifth, sixth, and seventh irradiations of the United States Department of Energy Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The combined irradiation tests will consist of fuel qualification tests (AGR-5/6) and a margin test (AGR-7) for the industrially produced uranium oxycarbide (UCO) tristructural isotropic (TRISO) coated particle fuel developed by the AGR Program. The objective of the AGR-5/6/7 irradiation experiment is to establish the performance of UCO TRISO fuel fabricated using engineering-scale equipment in a production environment under and beyond normal operating conditions. Specifically, the goals of the AGR 5/6/7 irradiation experiment are: 1. Irradiate reference design fuel containing low-enriched UCO TRISO fuel particles to support fuel qualification. 2. Establish the operating margins for the fuel beyond normal operating conditions. 3. Provide irradiated fuel performance data and irradiated fuel samples for post-irradiation examination and safety testing. 4. Support the refinement of fuel performance and fission product transport models with online, post-irradiation examination and safety test data. In order to achieve the test objectives, the AGR-5/6/7 experiment will be irradiated in the northeast flux trap position of the Advanced Test Reactor at Idaho National Laboratory. The larger diameter of the northeast flux trap provides greater flexibility for test train design, significantly enhancing the capability for the combined irradiations. The test train contains five separate and independently controlled and monitored capsules. The middle capsule contains the AGR 7 margin test fuel. The two bottom and two top capsules contain the AGR 5/6 fuel. Each capsule contains a variable number of fuel compacts filled with UCO TRISO particles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Report on Assessment of Advanced Reactor Fuel Performance

Three activities have been performed in FY 2019 to support the Advanced Fuel Cycle Campaign, which include assessment of annular fuel performance in an ultra-high burnup sodium-cooled reactor (SFR), modeling of annular fuel behavior during a transient scenario, and feasibility of a sustainable closed fuel cycle utilizing melt-refining process. Annular fuel performance parameters, including temperature profile, swelling strain, and cumulative damage fraction (CDF), in a 3000 MWt Sustainable Sodium-cooled Fast Reactor (SSFR) were assessed using the BISON fuel performance code. The SSFR is once-through ultra-long-life core based on a 34-batch fuel management scheme with 1.5-year cycle length. The fuel residence time in the SSFR is 51 years and the average discharged burnup is about 30%. Annular fuel has potential to irradiate for ultra-high burnup without major fuel-cladding mechanical interaction. This feature is achieved by using a low smeared density (~55%). In the SSFR design, however, 75% smeared density and a thick fuel pin were preferred to maintain criticality for 51 years. Due to the high smeared density and thick fuel pin configuration, the annular fuel was predicted to suffer premature cladding failure before reaching the discharge burnup of the SSFR. The current fuel creep model predicts a prominent creep strain due to gravity over the half-century irradiation, which may significantly affect the evolution of CDF. Thus, in order to use the annular fuel in an ultra-high burnup core such as the SSFR, a low smeared density and thin annular fuel is strongly recommended.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Highly efficient deep desulfurization of fuels by meso/macroporous H3PW12O40/TiO2 at room temperature

Highlights: • A new kind of meso/macroporous structured HPW/TiO{sub 2} catalyst was prepared by monodispersed polystyrene spheres and quaternary ammonium bromide. • Quaternary ammonium bromide with longer alkyl chains length (STAB) was found to be the best mesostructure-directing agent in the synthesis of meso/macroporous HPW/TiO{sub 2} catalyst. • The influence of dosage of macroporous templates (monodispersed polystyrene spheres) and alkyl chains length of quaternary ammonium bromide on the textural of meso/macroporous HPW/TiO{sub 2} catalyst were investigated. • With proper PS spheres and quaternary ammonium bromide usage, disordered macropores arranged periodicity and led relatively larger pore size and pore volume. • The meso/macroporous HPW/TiO{sub 2} materials exhibited enhanced excellent desulfurization performance of thiophenic sulfur in model fuel than other samples. Dibenzothiophene can be completely removed within 2 h at 30 °C. - Abstract: This paper described the synthesis of meso/macroporous phosphotungstic acid (HPW) /TiO{sub 2} catalyst with quaternary ammonium bromide and monodispersed polystyrene spheres (PS) as dual-template. The synthetic parameters, such as concentration of the templates and the alky chain length of quaternary ammonium bromide was studied carefully and optimized to generate catalyst with regulated porous features. The chemical properties of the catalyst was also investigated, X-ray diffraction (XRD) and Fourier transform infrared (FT-IR) measurements confirmed the Keggin-type HPW were highly dispersed on the TiO{sub 2} framework. The optimal meso/macroporous HPW/TiO{sub 2} catalyst exhibited excellent catalytic activity in the oxidation of refractory sulphur compounds such as benzothiophene (BT), dibenzothiophene (DBT), and 4,6-dimethyldibenzothiophene (4,6-DMDBT). DBT was removed within 2 h at 30 °C under selected reaction conditions and the apparent activation energy for DBT oxidation was calculated to be 45.7 kJ/mol. The high ODS reaction rate of catalyst could be attributed to the combination of unique meso/macroporous architecture allowing for efficient mass transport of reactants and products in pore channel and its high surface area, large mesoporous size, and pore volume enabling sufficient exposure of catalytic active sites. Moreover, the meso/macroporous HPW/TiO{sub 2} catalyst also exhibited excellent reusability with no obviously degradation even after eight cycles.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2026

Beyond interpolation: Physics-inspired gating transformers for extrapolating irradiation conditions to novel nuclear fuels

The qualification of advanced nuclear fuels relies on irradiation experiments in test reactors that emulate commercial conditions. Designing these tests requires accurate prediction of key irradiation quantities, particularly heat generation rate and burnup, yet obtaining them typically involves computationally expensive multi-step simulation workflows. We propose a physics-inspired gating transformer (PIGT) that integrates an inverse-square, distance-based attenuation into the encoder representation to bias attention toward physically relevant spatial relationships while retaining data-driven flexibility. Using MiniFuel irradiation data from the High Flux Isotope Reactor at Oak Ridge National Laboratory, we benchmark against ensemble methods, feedforward and recurrent networks, convolutional models, and standard transformers. While baseline models perform well under interpolation, they exhibit a pronounced generalization gap when evaluated on fuels not included in the training set. The proposed model consistently improves extrapolative accuracy and stability, yielding the strongest performance on unseen fuel configurations. These results indicate that a lightweight physics structure embedded within attention mechanisms can substantially improve robustness, enabling more reliable surrogate predictions to accelerate the design of nuclear fuel irradiation experiments.

Fuel qualification↗
DOE OSTI2023

AGR-1, AGR-2, AGR-3/4, and AGR-5/6/7 DimensionalChange Analysis

A series of fuel irradiation experiments have been planned in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to support the licensing and operation of the Advanced Reactor Technologies high temperature gas-cooled reactor. The advanced gas reactor (AGR) experiments are comprised of multiple independent capsules containing multiple cylindrical fuel compacts, placed inside of a graphite cylinder in ATR. The purpose of the AGR experiments is to provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for accident testing, and support the development of fuel performance and fission product transport models. The advanced graphite creep (AGC) experiments provide irradiation creep data for design and licensing. To date, six irradiation campaigns have been completed: AGR-1 (December, 2006 – November, 2009); AGR-2 (June, 2010 – October, 2013); AGR-3/4 (December, 2011 – April, 2014); AGC-1 (September, 2009 – January, 2011); AGC-2 (April, 2011 – May, 2012); and AGC-3 (November, 2012 – April, 2014).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Impact of recent ENDF nuclear data update, high initial enrichment and high burnup fuel on critical experiments applicability determination via the integral index c k for burnup credit validation

In 2012, NUREG/CR-7109 reported on the validation of burnup credit calculations involving major and minor actinides and major fission products which was investigated for pressurized and boiling water reactor (PWR and BWR) fuel enrichments up to 5 wt% 235 U and assembly-average burnups up to 60 GWd/MTU. Recently, there has been interest in increasing the maximum enrichment used in PWR fuel as high as 8 wt% 235 U and correspondingly increasing the maximum assembly-average burnups to approximately 75 GWd/MTU. These proposed increases in enrichment and burnup necessitate reinvestigation of the validation basis for k eff calculations for this expanded application space. Additionally, the 2012 study was performed by using the Evaluated Nuclear Data File (ENDF)/B-VII.0 nuclear data with the SCALE 6 covariance library, and the effects of using the newly released ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data and covariance libraries should be evaluated. In this work, published in NUREG/CR-7309 in 2025, the validation assessment was performed consistently with NUREG/CR-7109: modeling irradiated fuel assemblies in the Generic Burnup Credit (GBC)-32 cask defined in NUREG/CR-6747. The TSUNAMI-3D sequence was used to generate sensitivity data for the application model, and the data were compared with sensitivity data from select benchmark models. The integral parameter c k is the metric of similarity used in this study and is consistent with NUREG/CR-7109, where a c k value in excess of 0.8 indicates sufficient similarity for use in validation. A new set of benchmark experiments with sensitivity data has been assembled for this effort. The number of experiments with available sensitivity data is now 2,104, compared to 474 in NUREG/CR-7109. This increase was facilitated by the efforts of the Nuclear Energy Agency to generate sensitivity data for a majority of the experiments in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook to supplement the data available in the Oak Ridge National Laboratory (ORNL) Verified, Archived Library of Inputs and Data (VALID). The complete set of benchmarks considered here includes experiments for low-enriched uranium (LEU), intermediate enriched uranium (IEU), and a mixture of uranium and plutonium (MIX) from the ICSBEP Handbook and VALID, as well as ORNL models of the Haut Taux de Combustion (HTC) experiments and other potentially relevant models not included in VALID. The updated similarity study shows that none of the extended burnup and higher enrichment combinations considered show a significant decrease in the number of potentially applicable experiments, meaning sufficient critical experiments exist for the validation of BUC criticality safety calculations, with initial enrichments up to 8 wt% 235 U and burnups up to 80 GWd/MTU. Additionally, both the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries can be used for validation since the number of critical experiments applicable for validation increases for most cases with the most recent nuclear data compared to the previous one. As in previous BUC validation studies, the French HTC experiments are the most similar in a majority of the application cases studied, especially from representative discharge burnups ranging from 40 to 80 GWd/MTU. In conclusion, these results match the conclusions presented in NUREG/CR-7109 regarding validation of the primary actinides in BUC analyses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Hybrid Solid Oxide Fuel Cell/Gas Turbine Model Development for Electric Aviation

A thermodynamic model was developed and validated to analyze a high-performance solid oxide fuel cell and gas turbine (SOFC-GT) hybrid power system for electric aviation. This study used a process simulation software package (ProMax) to study the role of SOFC design and operation on the feasibility and performance of the hybrid system. Standard modules, including compressor, turbine, heat exchanger, reforming reactor, and combustor were used from the ProMax tool suite while a custom module was created to simulate the SOFC stack. The model used an SOFC test data set as an input. Additional SOFC stack performance effects, such as pressure, temperature, and utilization of air and fuel, were added from open source data. System performance predictors were SOFC specific power, fuel-to-electricity conversion efficiency, and hybrid system efficiency. Using these input data and predictors, a static thermodynamic performance model was created that can be modified for different system configurations and operating conditions. Prior to creating the final aircraft performance model, initial demonstration models were developed to validate output results. We used the NASA SOFC model as a benchmark, which was created with their Numerical Propulsion System Simulator (NPSS) software framework. Our output results matched within 1% of both the NASA model and open source SOFC performance data. With confidence gained in the accuracy of this model, a 1-MW SOFC-GT hybrid power system was constructed for an aircraft propulsion concept. Overall hybrid system efficiencies of > 75% FTE were observed during standard 36,000 feet cruise flight conditions.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2021

Metallic fuel cladding degradation model development and evaluation for BISON

Fuel cladding chemical interaction (FCCI) and coolant cladding chemical interaction (CCCI), both consume stainless-steel cladding in sodium-cooled fast reactors (SFRs), and effectively degrade cladding mechanical performance through thickness reduction. Therefore, FCCI/CCCI and corresponding cladding degradation models are essential for advanced fuel performance codes, such as BISON, to reliably predict cladding behavior in SFRs. Here, we report the development and implementation efforts for BISON's FCCI/CCCI correlations and cladding degradation models based on U.S. legacy metallic fuel data. The models were evaluated using Integral Fast Reactor (IFR) program X447/A experiment data supported by the ongoing integration project enabling coordinated application of BISON and EBR-II fuel irradiation and physics database (FIPD). Furthermore, the implemented models were demonstrated to improve BISON's capabilities of predicting cladding damage and degradation behavior that is more consistent with post-irradiation examination observations. Additionally, some limitations of current BISON modules are identified, which are to be overcome through the BISON-FIPD integration efforts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Impact of uranium oxide (UO 2 ) fuel with molybdenum (Mo) inserts on pressurized water reactor performance and safety

This work investigates nuclear reactor performance and safety characteristics of UO 2 with high thermal conductivity Mo insert structures by using multiphysics modeling techniques. Additionally, the purpose of this study is to use scoping analyses to quantify the impact of using Mo inserts from neutronic and heat transfer standpoints. Attention is given to reactor performance parameters, such as cycle length, maximum fuel temperature, temperature gradients in the fuel, and stored energy in the fuel. The finite-element code BISON and the Monte Carlo particle transport code Serpent were used to perform sensitivity analyses on the Mo insert geometry to optimize the insert design and inform larger scale modeling that required the homogenization of the UO 2 and Mo. Although BISON is often used as a fuel performance analysis tool, it is used in this context for heat transfer analysis only. Fuel performance optimization is outside the scope of the current study, but would be important for future work focused on this concept. The results showed that the insert had little impact on neutronic performance and that homogenizing the UO 2 and Mo was acceptable for reactor physics calculations. Reactivity temperature coefficients calculated using homogeneous UO 2 -Mo were shown to be relatively similar to UO 2 , but higher Mo content and 235 U enrichment can reduce the worth of soluble boron and control rods. The effect of insert geometry on heat transfer was much greater, and an approximately 15–20% difference in maximum fuel temperature was predicted between the best and worst performing heat transfer geometries. Furthermore, thermal conductivity calibration based on the finite element analysis results was performed to improve the accuracy of temperature predictions in reactor analysis models that homogenized the UO 2 -Mo fuel. Compared with UO 2 in a pressurized water reactor (PWR), the optimized UO 2 -Mo design increased the margin to fuel melt by 13–32% across the fuel cycle, but it requires the 235U enrichment to exceed 5% to match the cycle length of conventional UO 2 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Developing the test conditions for the first fueled transients in the TREAT reactor

The restart of the Transient Reactor Test (TREAT) facility at Idaho National Laboratory, re-establishes a national in-pile testing capability for nuclear fuels in accident scenarios. The purpose of the first tests in TREAT are to study the response of nuclear fuel durring a reactivity initiated accident (RIA). The initial objective is to demonstrate the ability to create a desired enthalpy increase in a representative fuel test specimen over a desired time step and to better understand the uncertainty. The five test rods initially available for these irradiations are UO2-Zircaloy with standard Pressurized Water Reactor (PWR) geometry. The rods are 15 cm long with 9 cm of active fuel. The initial testing environment is dry helium. BISON fuel performance calculations were run to model the heat transfer during postulated transients. Based on those results energy deposition targets were selected around the range of observable phenomena in the testing environment. Reactor / experiment power coupling factors, step insertion size and transient clipping response times were then calculated using results from a RELAP point kinetics model. Calculated energy deposition targets are 150, 350, 650, 960, and 1300 J/gUO2. The energy coupling factor for the experiments was calculated to be ~1.5 J/gUO2-MJ. Two step insertion sizes were selected: a 0.6% ?k/k step for the 150 J/g transient and a 1.8% ?k/k step for the other transients. Clipping response times ranged from 1.3 seconds to unclipped transients. The experiments will be irradiated in the TREAT facility in the fall of 2018. The first transient will be the 150 J/g pulse with a 0.6% ?k/k step that is unclipped. Follow on tests may modify the reactor operating parameters based on results from previous transients. Any preliminary data available on the results of the transients at the time of the NUMAT Conference will be presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

CRADA NFE-07-00912: Development of Model Fuels Experimental Engine Data Base & Kinetic Modeling Parameter Sets (Final Report)

The automotive and engine industries are in a period of very rapid change being driven by new emission standards, new types of after treatment, new combustion strategies, the introduction of new fuels, and drive for increased fuel economy and efficiency. The rapid pace of these changes has put more pressure on the need for modeling of engine combustion and performance, in order to shorten product design and introduction cycles. New combustion strategies include homogeneous charge compression ignition (HCCI), partial-premixed combustion compression ignition (PCCI), and dilute low temperature combustion which are being developed for lower emissions and improved fuel economy. New fuels include bio-fuels such as ethanol or bio-diesel, drop-in bio-derived fuels and those derived from new crude oil sources such as gas-to-liquids, coal-to-liquids, oil sands, oil shale, and wet natural gas. Kinetic modeling of the combustion process for these new combustion regimes and fuels is necessary in order to allow modeling and performance assessment for engine design purposes. In this research covered by this CRADA, ORNL developed and supplied experimental data related to engine performance with new fuels and new combustion strategies along with interpretation and analysis of such data and consulting to Reaction Design, Inc. (RD). RD performed additional analysis of this data in order to extract important parameters and to confirm engine and kinetic models. The data generated was generally published to make it available to the engine and automotive design communities and also to the Reaction Design Model Fuels Consortium (MFC).

computational fluid dynamics↗