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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 1,117 records · Page 62
All sourcesNASA NTRSDOE OSTI
NASA NTRS2001

The Effect of Stochastic Perturbation of Fuel Distribution on the Criticality of a One Speed Reactor and the Development of Multi-Material Multinomial Line Statistics

The effect of random fuel redistribution on the eigenvalue of a one-speed reactor is investigated. An ensemble of such reactors that are identical to a homogeneous reference critical reactor except for the fissile isotope density distribution is constructed such that it meets a set of well-posed redistribution requirements. The average eigenvalue, , is evaluated when the total fissile loading per ensemble element, or realization, is conserved. The perturbation is proven to increase the reactor criticality on average when it is uniformly distributed. The various causes of the change in reactivity, and their relative effects are identified and ranked. From this, a path towards identifying the causes. and relative effects of reactivity fluctuations for the energy dependent problem is pointed to. The perturbation method of using multinomial distributions for representing the perturbed reactor is developed. This method has some advantages that can be of use in other stochastic problems. Finally, some of the features of this perturbation problem are related to other techniques that have been used for addressing similar problems.

Jahshan, S. N.↗
DOE OSTI2022

Impacts of LEU+ and HBU Fuel on Decay Heat and Radiation Source Term

Using low-enriched uranium with 235 U enrichment slightly greater than 5.0 wt % (LEU+) fuel is desirable for light water reactors (LWR) because it can enable longer cycles and/or smaller fresh fuel batches. This report analyzes the impact of LEU+ on back-end considerations, such as its impact on the spent fuel pool, and on the thermal and shielding performance of dry storage systems. For both pressurized water reactor (PWR) and boiling water reactor LEU+ fuel, this report concludes that additional cooling time or regionalized loading plans are required to maintain the same level of thermal and shielding performance of dry storage systems loaded with LEU fuel. For PWR fuel, this report concludes that the impact of LEU+ fuel on the time-to-boil of a spent fuel pool (SFP) is minimal and does not require considerable changes in SFP management.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

SCALE 6.2.4 Validation for Light Water Reactor Decay Heat Analysis

Energy release from the decay of radionuclides in nuclear fuel after its discharge from reactor is a critical parameter for design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. Well-validated computational tools and nuclear data are essential for decay heat prediction. This paper summarizes the validation of the SCALE nuclear analysis code system version 6.2.4, used with ENDF/B-VII.1 libraries, for decay heat analysis of light water reactor used fuel. The experimental data used for validation include full-assembly decay heat measurements that cover assembly burnups of 5 to 51 GWd/tonne U, cooling times after discharge in the 2- to 27-year range, and initial fuel enrichments up to 4 wt% 235U. The comparison between calculated (C) and experimental (E) decay heat showed very good agreement, with an average C/E over all considered measurements of 1.006 (σ = 0.016) for pressurized water reactor and 0.984 (σ = 0.077) for boiling water reactor assembly measurements. The effect of using assembly-average versus axially varying modeling data on the calculated decay heat, important to thermal analyses for used fuel transportation and storage systems, is discussed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Scoping Analysis of Pebble-Bed Reactors for the Destruction of the Transuranic Inventory of LWR Spent Nuclear Fuel

With the forecasted increase in the construction and operation of nuclear reactors, there will be a corresponding increase in the quantity of spent nuclear fuel (SNF) that requires long-term storage. In SNF, transuranic isotopes contribute the most to the long-term radiotoxicity of the fuel and pose a proliferation risk. One option that has been explored to address these issues is the removal of the transuranic isotopes from SNF and the conversion of these isotopes into transuranic fuel (TRU fuel). Here, this work sought to determine how effective a micro-modular Pebble-Bed High-Temperature Gas-Cooled Reactor (PB-HTGR); the 10-MW High Temperature Gas-cooled Test Reactor (HTR-10); and a salt-cooled small-modular pebble-bed reactor (PBR), i.e. the generic Fluoride-cooled High-temperature Reactor (gFHR), are at reducing the inventory of transuranic isotopes while still maintaining the intrinsic safety features of the PBR designs, such as negative temperature coefficients of reactivity. Optimized pebble designs utilizing TRU fuel were found for both reactors through the adjustment for the packing fraction of fuel in each pebble. The Axial Zone Equilibrium Modeling (A-ZEM) method was used in this work to help select the optimized pebble design. Once an optimized pebble design was selected and an equilibrium model was produced, the results from the deep burn (DB) HTR-10 and gFHR designs were compared to the results of two models from the literature. While both the DB gFHR and the DB HTR-10 were able to reduce the weapons-usable transuranic inventory, the performance of these reactors did not match that of the small-modular PB-HTGRs in the literature. Therefore, a need was identified for further refinement of the gFHR design using TRU fuel, as the results for this model were more promising than those of the DB HTR-10, which was strongly limited by the high leakage intrinsic to micro-modular PB-HTGRs.

Transuranic fuel↗
DOE OSTI2021

Methods of forming fuel rods comprising porous fuel materials

A method of forming a fuel rod for a nuclear reactor comprises disposing a powder comprising particles of a fuel material on a substrate, exposing the powder to energy from an energy source to form a first layer of a nuclear fuel, the first layer comprising inter-granular bonds between the particles of the fuel material, disposing additional powder comprising particles of the fuel material over the first layer of the nuclear fuel, and exposing the additional powder to energy from the energy source to form a second layer of the nuclear fuel and to form the nuclear fuel to have a void fraction greater than about 0.20, the second layer comprising inter-granular bonds between the additional powder and the first layer of the nuclear fuel. Related nuclear fuels comprising a porous structure, fuel rods, nuclear reactors, and methods are disclosed.

O'Brien, Robert C.↗
DOE OSTI2022

Thermal-shock experiments for separate-effects validation of UO 2 fuel fracture models

Because of the important role that fracture plays in the behavior of ceramic UO2 fuel in a nuclear reactor environment, fracture models are a major component of fuel performance codes. As with any aspect of fuel performance, it is crucial to validate these fracture models against experimental data; however, obtaining well-controlled data for conditions representative of a reactor environment is difficult. Quenching is proposed here as a relatively simple approach for using a laboratory environment to achieve conditions that approximate those of a reactor environment. In this paper, an experimental apparatus containing a single instrumented fuel pellet in a sealed section of copper tubing is developed. It is then applied to a series of seven experiments in which the apparatus is first heated to a high temperature (580-680°C) by immersion in a molten salt bath, then quenched in a cold bath (-10-4 °C). Development of these experiments was guided by numerical simulations, and post-test simulations were performed to aid in understanding the experimental behavior and assessing the accuracy of the predictions of fracture initiation and propagation. In addition, similar experiments were performed on solid copper rods to provide temperature-dependent heat transfer coefficients for use in simulations of these experiments. Here, this study demonstrates that quenching is a viable approach for generating thermal gradients representative of those in prototypical light-water reactor conditions at powers of about 5–10 kW/m. Fracture is expected to begin at these power levels, and moderate amounts of radial and axial cracking was observed in these quenching tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Multiphysics simulations of Self-Regulating performance of an optimized molten metal fuel microreactor design

Advanced microreactors are expected to play an indispensable role in reliable energy solutions for civilian/ military applications in off-grid regions and space applications such as power supplies for Lunar and Mars bases. An innovative fast neutron spectrum heat pipe microreactor (HP-MR) concept was recently proposed with two unique features: molten U-Mn fuel and heat pipes as the heat removal mechanism. The microreactor was designed to be self-regulated, solely relying on reactivity feedbacks from fuel temperature. The reactor design has been further optimized to provide sufficient engineering safety margin and effective heat removal. In this study, the self-regulating performance of the optimized molten metal fuel microreactor was investigated by multi-physics simulations (neutronics, heat-transfer and heat pipe modeling) based on the MOOSE MultiApps system. In conclusion, the responses of the reactor to a series of transient scenarios were simulated to demonstrate the exceptional and inherent safety features of the novel microreactor concept.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

FIPD: The SFR metallic fuels irradiation & physics database

The DOE Advanced Reactor Technology (ART) program has supported efforts to recover and preserve metallic fuel data generated throughout the US sodium-cooled fast reactor (SFR) program. Those efforts have been focused on establishing databases of the experimental data that were mainly generated during the Integral Fast Reactor (IFR) program including data generated at Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and Transient Reactor Test Facility (TREAT) reactors, as well as out of pile data. The data is essential for future licensing activities of metallic fuel based advanced fast reactors. This paper describes the development of the SFR Metallic Fuels Irradiation & Physics Database (FIPD) and covers the scientific knowledge available in the database. Furthermore, the architecture of the FIPD is described by showing the available reactor operation data, fabrication, and post-irradiation examination (PIE) data, and other documents. The applications of the fuel database are also discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Development and Test Results of Thermocouples Used in the TRISO-Fuel Irradiation Experiment AGR-5/6/7

High-temperature gas reactor irradiation experiments create unique challenges for thermocouple-based temperature measurements. High-temperature industrial thermocouples suffer rapid decalibration due to transmutation of the thermoelements from neutron absorption. For lower-temperature applications, Type K and Type N thermocouples are affected by neutron irradiation only to a limited extent. But until recently, the use of these nickel-based thermocouples was limited when the temperature exceeded 1050°C due to drift related to phenomena other than nuclear irradiation. Certain portions of the AGR-5/6/7 experiment experienced temperatures higher than any of the previous AGR tests, up to 1500°C. Recognizing the limitations of existing thermometry to measure such high temperatures, the sponsor of the AGR-5/6/7 test supported a development and testing program for thermocouples capable of low-drift operation at temperatures above 1100°C. This program included additional development of high-temperature irradiation-resistant thermocouples based on molybdenum/niobium thermoelements, which have been studied at Idaho National Laboratory since circa 2004. A step change in accuracy and long-term stability of this thermocouple type was achieved as part of the AGR-5/6/7 thermometry development program. Additionally, long-term testing (7000+ h) at 1250°C of Type N thermocouples utilizing a customized sheath developed at the University of Cambridge has been completed with excellent low-drift results. The results of this testing as well as testing of the improved high-temperature irradiation-resistant design are reported herein. Both the improved high-temperature irradiation-resistant and the Cambridge Type N thermocouple types were incorporated into the AGR-5/6/7 test, which began irradiation in February 2018 and was completed in July 2020. Finally, a summary of the performance of the thermocouples incorporated into the AGR-5/6/7 test is included herein.

In-pile instrumentation↗
DOE OSTI2022

PNNL's Characterization Summary for MP-2 Experiment

Characterization of as-fabricated fuel was performed at Pacific Northwest National Laboratory (PNNL) in accordance with the characterization plan for the fabrication of U 10Mo plate fuel for the U.S. High Performance Research Reactor conversion program’s Fuel Fabrication Pillar (INL 2021). Similar characterization work is also being performed at Idaho National Laboratory to provide a detailed understanding of the as-fabricated foils that would be irradiated in the Mini-Plate 2 (MP 2) experiment. Under the MP 2 characterization plan, foils are studied that have different fabrication parameters (such as rolling condition, rolling thickness reduction, co-rolling with Zr layers). Similar samples from master foils were sent to both the organizations, so that the testing and analysis can be done independently using similar equipment and standardized measurement and analysis procedures. A final, consolidated report will be prepared based on this work and will summarize all the information obtained from the two laboratories. The MP 2 experiment will provide an opportunity to understand the effects of processing conditions on the final fuel microstructure, to compare results obtained independently, and achieve a two-way validation. In Fiscal Year 2022, PNNL received five MP 2 cast (PD STD2) samples to examine the foils’ chemistry and microstructure. For each cast sample, PNNL received samples from three different locations. PNNL also received and characterized 24 U 10Mo foil samples, by sectioning four pieces/specimens from each foil, in accordance with the MP 2 Characterization Plan (INL 2021). These 24 samples consist of four types of foils from BWX Technologies: 0.047 in. thick hot-rolled and annealed samples with Zr layers; 0.025 in. thick cold-rolled and annealed samples with Zr layers; 0.0105 in. thick cold-rolled and annealed samples with Zr layers. Along with these, PNNL also received four plates with Zr layers that were 0.025 in. and 0.0105 in. thick. This report describes the results of PNNL’s MP 2 foil characterization. Microstructure, Mo homogeneity, carbide fraction and morphology, U 10Mo foil thickness, and Zr thickness were evaluated in both the longitudinal and transverse directions for all the foils of the three different thicknesses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Improved Fuel Cycle Capability of Griffin for Fast Reactor Applications

Griffin is a MOOSE based reactor multiphysics analysis application jointly developed by Idaho National Laboratory and Argonne National Laboratory under the Department of Energy Office of Nuclear Energy Nuclear Energy Advanced Modeling and Simulation Program. This fiscal year, the fuel cycle capability has been significantly extended by improving the assembly shuffling option to allow flexible fuel reloading in the multi-cycle depletion calculation and incorporating decay between cycles. An equilibrium core calculation capability was also implemented to find an equilibrium core. Additionally, an enrichment search capability was added to determine the enrichment condition that allows a core to reach an equilibrium cycle with the end-of-cycle k-effective meeting a user-specified target value. The updated fuel management capability has been extensively tested using the three-dimensional ABTR problem with different batch schemes, exhibiting reasonable solutions in terms of manual shuffling, equilibrium cycle, and enrichment search calculations. The cross section generation workflow capability for fast reactors was further verified to produce microscopic cross sections as well as Griffin core inputs for an ring-heterogeneous configuration. Rigorous verification tests using the ABTR problem demonstrated that the RH core calculations, with cross sections and Griffin inputs generated from the cross section workflow of Griffin, produced accurate solution for fast reactor problems. In addition, an option to convert delay neutron parameter data generated from MC 2 -3 in the DLAYXS format into XML format was added to support transient calculations using MC 2 -3-generated data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Molten Salt Reactor Fundamental Safety Function PIRT

The phenomenon identification and ranking table (PIRT) process was employed to evaluate the current capability to assess the ability of liquid-salt–fueled molten salt reactors to achieve their fundamental safety functions (FSFs). The PIRT process provides a structured mechanism to elicit and document expert opinions on the most important phenomena and the corresponding level of knowledge with regard to achieving the FSFs. The PIRT panel included reactor developers, accident progression evaluation tool developers, US Department of Energy (DOE) national laboratory technical staff, university researchers, and US Nuclear Regulatory Commission (NRC) staff. The information-gathering process concentrated on phenomena related to accidents in which the fuel salt—including any in the cover gas—has been released from the first barrier layer. This information-gathering process included identifying all potential accidents that could result in the release of a substantial amount of radionuclides that could breach or bypass the first barrier. The PIRT did not identify any previously unrecognized systemic vulnerabilities. However, the elicitation process identified several areas with low levels of knowledge and significant potential impact on accident progression. These areas either require additional research to improve the state of knowledge or additional design conservatism to accommodate the remaining unknowns.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1978

Correlations of catalytic combustor performance parameters

Correlations for combustion efficiency percentage drop and the minimum required adiabatic reaction temperature necessary to meet emissions goals of 13.6 g CO/kg fuel and 1.64 g HC/kg fuel are presented. Combustion efficiency was found to be a function of the cell density, cell circumference, reactor length, reference velocity, and adiabatic reaction temperature. The percentage pressure drop at an adiabatic reaction temperature of 1450 K was found to be proportional to the reference velocity to the 1.5 power and to the reactor length. It is inversely proportional to the pressure, cell hydraulic diameter, and fractional open area. The minimum required adiabatic reaction temperature was found to increase with reference velocity and decrease with cell circumference, cell density and reactor length. A catalyst factor was introduced into the correlations to account for differences between catalysts. Combustion efficiency, the percentage pressure drop, and the minimum required adiabatic reaction temperature were found to be a function of the catalyst factor. The data was from a 12 cm-diameter test rig with noble metal reactors using propane fuel at an inlet temperature of 800 K.

Bulzan, D. L.↗
DOE OSTI2022

Estimation of China Fusion Engineering Test Reactor performance and burning fraction in different pellet fueling scenarios by a multi-species radial transport model

Tritium self-sufficiency in future deuterium–tritium fusion reactors is a crucial challenge. As an engineering test reactor, the China Fusion Engineering Test Reactor requires a burning fraction of 3% for the goal to test the accessibility to the future fusion plant. To self-consistently simulate burning plasmas with profile changes in pellet injection scenarios and to estimate the corresponding burning fraction, a one-dimensional multi-species radial transport model is developed in the BOUT++ framework. Here several pellet-fueling scenarios are then tested in the model. The results show that the increased fueling depth improves the burning fraction by particle confinement improvement and fusion power increase. Nevertheless, by increasing the depth, the pellet cooling-down may significantly lower the temperature in the core region. Taking the density perturbation into consideration, the reasonable parameters of the fueling scenario in these simulations are estimated as pellet radius ${r}_{{\rm{p}}}=3\,{\rm{mm}},$ injection rate $=\,4\,\mathrm{Hz},$ and pellet injection velocity $=\,1000\mbox{-}2000\,{\rm{m}}\,{{\rm{s}}}^{-1}$ without drift or $450\,{\rm{m}}\,{{\rm{s}}}^{-1}$ with high-field-side drift.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

A Robust Processing Approach for Producing Highly Loaded Dispersion Fuels

A robust fabrication method, resulting in higher yields, to produce highly loaded U3Si2-Al dispersion fuels for converting research reactors from a high enriched to a low enriched uranium fuel needs to be developed. To reliably produce a highly loaded dispersion fuel, process changes need to be implemented where the traditional approach has experienced challenges and poor yields. The present work describes the key changes needed. Parts of the work were done with uranium silicide and parts were completed with a representative silicide surrogate. The major deviations from traditional fabrication methods are associated with a refinement in particle size distribution, method for compacting to achieve complex shapes, and welding of the aluminum picture frame used to encapsulate the fuel compact. Methods for rolling and arc melting are also discussed. By using the methods described within, fabricating a highly loaded dispersion fuel that can meet stringent fuel homogeneity and geometry requirements at higher yields and lower costs may be possible.

U3Si2, HFIR, dispersion fuel, Uranium, Research Re↗
DOE OSTI2022

Irradiation of MiniFuel Targets Bearing TRISO Fuel Compacts (Status Report)

Irradiation testing of MiniFuel compacts bearing tristructural isotropic (TRISO) fuel particles was performed at Oak Ridge National Laboratory (ORNL) to support the development of Kairos Power’s (KP’s) fluoride salt–cooled high-temperature reactor concept. The fuel compacts were fabricated with TRISO fuel particles of different types—including low-enriched uranium oxide, uranium carbide (LEUCO), natural uranium oxide, uranium carbide (NUCO), and low-enriched uranium dioxide (LEUO 2 )—and inserted into MiniFuel irradiation targets. Five targets were assembled and inserted in the High Flux Isotope Reactor (HFIR) for four cycles. The data collected post-irradiation will provide experimental input to validate TRISO fuel performance models for high particle power operations. This report summarizes the completion of the HFIR irradiation, the as-irradiated numerical analysis, and the post-irradiation work performed to date. This work was performed under the Nuclear Science User Facility program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI1992

Nuclear Safety [Vol. 33, No. 1, January-March 1992]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 Technical Note: A New Approach to Fission Reactor Safety, Yu. V. Petrov; 5 Erratum; ACCIDENT ANALYSIS: 6 Analysis and Modeling of Fission Product Release from Various Uranium-Aluminum Plate-Type Reactor Fuels, R. P. Taleyarkhan; CONTROL AND INSTRUMENTATION: 23 Applications of a Surveillance and Diagnostics Methodology Using Neutron Noise From a Pressurized-Water Reactor, R. T. Wood, L. F. Miller, and R. B. Perez; DESIGN FEATURES: 36 Westinghouse Advanced Passive 600 Plant, B. A. McIntyre and R. K. Beck; 47 System 80+™ PWR Safety Design, C. W. Bagnal, R. A. Matzie, and R. S. Turk; ENVIRONMENTAL EFFECTS: 58 The MATS Experiments—Mesoscale Atmospheric Transport Studies at the Savannah River Site, A. H. Weber, S. Berman, R. J. Kurzeja, and R. P. Addis; 75 Book Review: Environmental Radioactivity in the European Community 1984-1985-1986, C. A. Little; WASTE AND SPENT FUEL MANAGEMENT: 76 Activities Related to Waste and Spent Fuel Management, M. D. Muhlheim and E. G. Silver; OPERATING EXPERIENCES: 87 Aging Assessment of BWR Control Rod Drive Systems, R. H. Greene; 100 Reactor Shutdown Experience, Compiled by J. W. Cletcher; 103 Selected Safety-Related Events, Compiled by G. A. Murphy; 110 Operating U.S. Power Reactors, Compiled by M. D. Muhlheim and E. G. Silver; RECENT DEVELOPMENTS: 129 General Administrative Activities, Compiled by M. D. Muhlheim and E. G. Silver; 139 Reports, Standards, and Safety Guides, D. S. Queener; 143 Proposed Rule Changes as of Sept. 30 1991; ANNOUNCEMENTS: 128 Harvard School of Public Health Announces Short Courses; 128 16th Biennial ANS Topical Meeting on Reactor Operating Experience: Present and Future Technologies—Applying Lessons Learned (Call for Papers); 147 The Authors; 150 Indexes to Nuclear Safety, Volume 32; 154 Errata.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Comparative Structural Rigidity Analysis of the NBSR DDE and the NBSR LEU Fuel Element

The National Bureau of Standards Reactor (NBSR) is a 20-MW heavy-water moderated and cooled, enriched-fuel, tank-type reactor located in the NIST Center for Neutron Research laboratory complex. NBSR is one of six U.S. high performance research reactors (USHPRR), including one critical facility, that actively collaborates with the National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program to convert from the use of highly enriched uranium fuel to the use of low-enriched uranium (LEU, < 20 wt% U-235) fuel. On the LEU U-10Mo fuel qualification side, mini-plate and large-plate irradiations have been successfully performed, and more experiments are either ongoing or planned. As an additional experimental campaign to support licensing of the LEU fuel for the use in the NBSR, the NBSR Design Demonstration Element (DDE) will be a full-sized test element similar to the NBSR LEU fuel element and irradiated in a test reactor under conditions as prototypic of the NBSR LEU fuel element operating conditions as possible.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗