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At least 757 records · Page 42
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

IER 517: Molybdenum Critical Experiment Design [Slides]

In nuclear fuels, U-Mo metallic fuels are used in new space reactor designs and new research reactor fuels. MITR, MURR, and NBSR reactors plan on converting from different HEU fuels to a U-Mo HALEU fuel. In spend nuclear fuel, 95 Mo is one of the 15 main absorbing fission products in irradiated LWR fuel. This makes it important for criticality safety studies in transportation and reprocessing. In structural materials. Molybdenum is found commonly in alloys that make up the structural materials of nuclear reactors such as type 316 SS. Molybdenum helps to improve high temperature performance and corrosion resistance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Massachusetts Institute of Technology Reactor LEU Fuel Element Flow Test Conceptual Design

As part of the U.S. National Nuclear Security Administration’s (NNSA’s) mission to minimize the civilian use of weapon-grade highly enriched uranium (HEU) fuels, the NNSA Office of Material Management and Minimization (M 3 ) Conversion Program is collaborating with six U.S. High Performance Research Reactors (USHPRR), including one critical facility, to convert from the use of HEU to low-enriched uranium (LEU) fuel. The M 3 conversion objectives for the USHPRR are to develop LEU fuel element designs that will ensure safe reactor operations and maintain the existing experimental performance of each facility. The work is being conducted through many interrelated activities that are being completed by stakeholders across organizations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1972

Conceptual design studies and experiments related to cavity exhaust systems for nuclear light bulb configurations

Investigations of various phases of gaseous nuclear rocket technology have been conducted. The principal research efforts have recently been directed toward the closed-cycle, vortex-stabilized nuclear light bulb engine and toward a small-scale fissioning uranium plasma experiment that could be conducted in the Los Alamos Scientific Laboratory's Nuclear Furnace. The engine concept is based on the transfer of energy by thermal radiation from gaseous fissioning uranium, through a transparent wall, to hydrogen propellant. The reference engine configuration is comprised of seven unit cavities, each having its own fuel transparent wall and propellant duct. The basic design of the engine is described. Subsequent studies performed to supplement and investigate the basic design are reported. Summaries of other nuclear light bulb research programs are included.

Kendall, J. S.↗
DOE OSTI2021

25-Pin metallic fuel performance benchmark case based on the EBR-II X430 experiments series

A metallic fuel benchmark case was developed for the fuel performance code BISON based on 25 uranium-zirconium and uranium-plutonium-zirconium pins of the Experimental Breeder Reactor II X430 experiment series. Results of the benchmarks were compared with measurements and calculations made at the time of the experiment as well as subsequent measurements reported in 2019. The comparisons were used to quantify the accuracy of the BISON predictions and to identify patterns in the BISON differences.BISON predicted burnup, plenum pressure, and fission gas release fractions accurately. BISON temperature predictions were somewhat cooler than the temperatures determined at the time of the experiment but appeared to be reasonably accurate considering uncertainties in the legacy temperature calculations, uncertainties in the legacy linear heat rate calculations, and the high sensitivities of the BISON-predicted temperatures to BISON inputs. Fuel axial elongation predictions had errors correlated to fuel composition; BISON tended to underpredict the elongation of binary fuels and overpredict the elongation of ternary fuels. BISON cladding radial dilation predictions were also significantly lower than legacy PIE measurements.Recommendations were made to improve the BISON fuel gaseous swelling model to account for fuel composition, to add additional capabilities to the coolant channel temperature model to ease benchmark development, and to continue developing benchmark cases based on a wide range of experiments in several reactors. Once a wide array of benchmarks is developed, an attempt can be made to enhance or calibrate BISON models to improve the cladding dilation predictions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRSDate not supplied

Effects of Varying the Heat Transfer Coefficient on Engine Performance

A physics-based Nuclear Thermal Propulsion (NTP) Testing Reference Design (TRD) power balance model was coded in Simulink to investigate engine performance for various design and parameter modifications. Since the primary mode of heat transfer in NTP engines is convective, the convective heat transfer coefficient (HTC) is a key parameter that requires accurate representation. The industry standard Westinghouse correlation has an uncertainty of ±20% which was investigated in this study. The results showed that a 20% decrease in the HTC led to a 4.14% increase in maximum fuel temperature while a 20% decrease in the HTC led to a 1.81% decrease in maximum fuel temperature suggesting that narrowing the uncertainty of this correlation through experimental work would be a critical step in the development of NTP engines. Furthermore, a maximum fuel temperature relationship with specific impulse was developed for the TRD engine which showed potential engine operation between specific impulse values of 715 and 900 seconds with minimal changes to the engine design. This graph could be useful for high level vehicle performance estimations for fuel types with different maximum operating temperatures.

Heat Transfer Coefficient↗
DOE OSTI2021

University of Missouri Research Reactor LEU Fuel Element Flow Test Conceptual Design

As part of the U.S. National Nuclear Security Administration’s (NNSA’s) mission to eliminate or minimize the civilian use of weapon-grade highly enriched uranium (HEU) fuels, the NNSA Office of Material Management and Minimization (M 3 ) Conversion Program is collaborating with six U.S. High Performance Research Reactors (USHPRR), including one critical facility, to convert from the use of HEU to low-enriched uranium (LEU) fuel. The M 3 conversion objectives for the USHPRR are to develop LEU fuel-element designs that will ensure safe reactor operations, as well as maintain the existing experimental performance of each facility. The work is being conducted through many interrelated activities that are being completed by stakeholders across organizations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Summary of BISON Milestones: NEAMS FY2020 Report

This summary report contains an overview of work performed under the work package entitled "FY2020 NEAMS Advanced Fuels Performance'', which is focused on the development and support of the fuel performance code BISON. The second chapter lists FY20 milestones titles, completion schedule, and milestone level. Subsequent chapters summarize and demonstrate completion of the milestones. The last chapter outlines FY21 proposed future work. In FY20, the NEAMS program emphasized development of BISON for its application to advanced reactors. While there are a variety of advanced fuel concepts, based on interaction with industry and the Nuclear Regulatory Commission, the fuel types we chose to develop were metallic fast reactor, UN/UC and particle fuels. The last chapter of this report documents proposed work for FY21. We plan to continue work on metallic and particle fuel in terms of developing/calibrating models and to begin rigorous validation/assessment for both fuel types. Due to the merger of the NEAMS and CASL programs, FY21 will see a return to light water reactor model development and simulation; this time focused on advanced technology fuels. Additionally, we seek to improve BISON, fundamentally. As such, we plan improvements to BISON and MOOSE in terms of algorithmic robustness, performance, ease-of-use, and quality assurance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Advancements in modeling fuel pulverization and cladding behavior during a LOCA

During a loss of coolant accident (LOCA), there is the possibility of nuclear fuel rods to undergo a three step process known as fuel fragmentation, relocation, and dispersal (FFRD). The chance of FFRD occurring increases as the fuel burnup increases. To support the nuclear industry's desire to increase the discharge burnup of nuclear fuels in light water reactors (LWRs) it is imperative to understand the mechanisms driving the evolution of FFRD. In this work, a multiscale modeling modeling approach is used to garner insight into underlying mechanisms leading to the ne fragmentation (also known as pulverization) of nuclear fuel during a LOCA. This report includes a summary of the atomistic and phase-field studies to develop a new pulverization criterion for use in the engineering scale Bison fuel performance code. Details are also provided on cladding modeling improvements related to hydrogen/hydride embrittlement and damage, and anisotropic thermal creep. The new models are used on the existing integral and separate effects LOCA validation cases available in Bison.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Values of Recovered Uranium from HALEU Used Nuclear Fuels (Rev. 1)

The value of the recovered uranium (RU) from high assay low-enriched uranium (HALEU) used nuclear fuels was evaluated. Three utilizations of the recovered uranium were considered in this study, which include the cases that RU is used as a fissile material of nuclear fuel, RU is reused in the original advanced reactor after reenrichment, and RU is reused in conventional light water reactors after down-blending. In this study, the RU values were identified by comparing the cost of making a unit mass of fuel with RU versus the fuel cost with the equivalent fresh enriched uranium (EU). A series of bounding analyses for calculating the fuel costs were conducted using several selected reactor types, which include microreactors, advanced thermal reactors, and fast reactors having a burnup of 2 – 165 GWd/t (with residual U-235 content in discharged fuels of 0.8 - 19.6%). This study concludes that RU having a residual U-235 content higher than ~7% would cost less than the fresh EU. The affordability increases as the residual U-235 content in RU increases. For instance, the fuel cost with RU having the residual U-235 content of 19.6% is about 85% cheaper than the fuel cost with the equivalent fresh EU. This study observed that reusing RU after reenrichment in the original microreactor is impractical because the U-235 content in the re-enriched RU fuel would need to be higher than the limit for low-enriched uranium (<20%) to provide the same burnup performance due to parasitic absorption from U-236. It is noted that this study focused on the recovery of uranium only, and the value of other fissile materials (such as Pu) in the used nuclear fuel was not considered even though those are bred significantly in fast reactors. In addition, the impacts of uncertainties in the cost data and the value of RU of TRISO fuels were not evaluated in this study due to the limited information on the cost data uncertainties and the separation cost from TRISO fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Accelerating nuclear fuel development and qualification: Modeling and simulation integrated with separate-effects testing

In this work, an approach to transform and accelerate nuclear fuel development and qualification is outlined. The approach exploits advanced modeling and simulation at the outset to inform constituent and system selection and to enable integral fuel performance analyses. Analyses using these tools identify and prioritize the most important fuel performance parameters and phenomena for subsequent targeted characterization with separate-effects tests. Separate-effects testing spans out-of-pile and in-pile tests and is meant to iterate with and inform engineering-scale integral fuel performance analyses throughout the development process. Exercising this cycle in an agile fashion will increase confidence in the integral fuel performance predictions while reducing uncertainties. This process sets the stage for executing a much more limited set of well-defined integral irradiation tests designed to validate engineering-scale fuel performance codes and to confirm the performance and safety of the fuel system under prototypic conditions. This approach will reduce the time for development and qualification of a new fuel system, and it will also reduce associated costs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Enabling Load Following Capability in the Transatomic Power MSR

This project is dedicated towards designing a fuel processing system that enables liquid-fueled molten salt reactors (MSR) to load follow by removing the dissolved xenon in the fuel salt. As one of the Gen-IV nuclear reactor concepts, the molten salt reactor receives increasing development interests in the recent years. One distinguishing feature of the liquid-fueled molten salt reactor is its improved ability to operate in a load-following mode by including the unique online fission product removal system. Load-following means that the reactor changes its power output based on the demand on the grid. Most of the current operating nuclear reactors have limited load-following ability and operate as the base load on the grid. Due to the rapid increase of solar energy, the requirement on load-following capacity is significantly increased because of the varying power output of the solar panels, yet the traditional load-following capacity is expected to decrease as the decarbonization of the grid continues. Therefore, the ability to perform load-following operation for the nuclear reactors will greatly enhance the resilience of the grid and make nuclear energy more economically competitive. This load-following feature is included in many commercial molten salt reactor designs, such as the designs by Transatomic Power, Terrestrial Energy, and Flibe Energy. Unfortunately, detailed analysis of the fuel processing system for commercial scale MSRs is still lacking, as well as how the fuel processing quantitively impacts the load-following operation. Moreover, experimental data for many of the underlying physics of fuel processing is limited. This project aims to pave the way for the fuel processing technology to advance to the commercial stage by performing combined experimental and simulation research. During the project period, four interconnected aspects of the development of the fuel processing system in liquid-fueled molten salt reactors are investigated. These aspects are the simulation and analysis of the fission product removal system, the fuel cycle simulation, the coupled reactor neutronics and thermal hydraulics transient simulation, and the gaseous fission product removal experiment. Multiphase CFD simulations are performed for components of the processing systems, and simplified air-water experiments are carried out to provide validation data. It is concluded that the CFD simulation can satisfactorily predict the system level performance of the components, and engineering models are constructed based on this success. Fuel cycle analysis is performed for two representative MSR design, the MSBR and the Transatomic Power MSR. Open-source code SaltProc is developed to incorporate the unique fuel processing system of the MSRs. It is concluded that the removal of xenon is essential for load-following operation in thermal spectrum MSR and Molten Salt Breeder Reactor. For the Transatomic Power MSR, the xenon poisoning effect is negligible due to its relatively fast neutron spectrum, though the overall fuel cycle economics still benefits from the removal of xenon. Coupled reactor neutronics and thermal hydraulics transient simulation is performed specifically for the Transatomic Power MSR. It is concluded that the reactor core design could perform power ramping fast enough to satisfy load-following operation. Combining the findings from each aspect, it is concluded that the load-following operation of a thermal neutron MSR is dependent upon the removal of xenon, which could be achieved for a commercial sized reactor using continuous inert gas sparging in a separate system with reasonable dimensions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

SNF Canister Coatings for Corrosion Prevention and Mitigation (FY21 Status Report)

This report summarizes the current actives in FY21 related to the effort by Sandia National Laboratories to identify and test coating materials for the prevention, mitigation, and repair of spent nuclear fuel dry storage canisters against potential chloride-induced stress corrosion cracking. This work follows up on the details provided in Sandia National Laboratories FY20 report on the same topic, which provided a detailed description of the specific coating properties desired for application and implementation on spent nuclear fuel canisters, as well as provided detail into several different coatings and their applicability to coat spent nuclear fuel canisters. In FY21, Sandia National Laboratories has engaged with private industry to create a Memorandum of Understanding and established a collaborative R&D program building off the analytical and laboratory capabilities at Sandia National Laboratories and the material design and synthesis capabilities of private industry. The resulting Memorandum of Understanding included four companies to date (Oxford Performance Materials, White Horse R&D, Luna Innovations, and Flora Coating) proposing six different coating technologies (polyetherketoneketone, modified polyimide/polyurea, modified phenolic resin, silane-based polyurethane hybrid with and without a Znrich primer, and a quasi-ceramic sol-gel polyurethane hybrid) to be tested, evaluated, and optimized for their potential use for this application. This report provides a detailed description of each of the coating systems proposed by the participating industry partners. It also provides a description of the planned experimental actives to be performed by Sandia National Laboratories including physical tests, electrochemical tests, and characterization methods. These analyses will be used to identify specific ways to further improve coating technologies toward their application and implementation on spent nuclear fuel canisters. In FY21, Sandia National Laboratories began baseline testing of the base metal material in according with activities of the Memorandum of Understanding. In FY22, Sandia National Laboratories will receive coated coupons from each of the participating industry partners and begin characterization, physical, and electrochemical testing following the test plan described herein.

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

Corroborating Channel Gap Probe Measurements with an Optical Profiler

As nuclear fuel plates get irradiated, the fuel plate thickness changes. How the fuel plate changes with irradiation is a significant parameter that is important to predict fuel performance in new fuel systems. The channel gap probe is used to indirectly infer changes in fuel plate thicknesses by measuring the gaps between the fuel plates contained within capsules after each irradiation cycle. Independent measures should be used to corroborate salient measurements such as changes in fuel thickness which help ascertain fuel performance. The use of traditional metrological tools to validate channel gap probe measurements are not practical due to the capsule geometry. The accuracy of the channel gap probe inserted within a capsule has not been verified. Since there is no “gold” standard measurement to verify CGP measurements, an optical profilometer has been used to corroborate the CGP measurements. The optical profilometer has been able to corroborate the capsules channel gaps to within 0.04 mm for both ends of the capsule. The optical profiler is also able to measure plate thickness that matched micrometer measurements to within 0.025 mm. In conclusion, the profilometer corroborated the CGP measurements but a discrepancy was found at one end of the capsule for the Inside Capsule Width from the profilometer when compared to a caliper measurement.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2026

Pulse radiolysis and transient absorption spectra of aqueous solutions of sodium sulfamate

Chemical kinetics for the reactions of sulfamate ions (NH 2 SO 3 − ) with the primary products of water and nitric acid radiolysis were measured in aqueous solutions at ambient temperature. Using time-resolved electron pulse radiolysis techniques with a custom multichannel detection system, we examined the reactivity of NH 2 SO 3 − with the hydroxyl radical ( • OH), hydrogen atom (H • ), and nitrate radical (NO 3 • ). The sulfamate ion was found to react with • OH and H • with second-order rate coefficients of k • OH = (5.60 ± 0.04) × 10 6 M −1 s −1 and k H • = (7.96 ± 0.10) × 10 6 M −1 s −1 , respectively, and with NO 3 • with a rate coefficient of k NO 3 • = (1.67 ± 0.06) × 10 7 M −1 s −1 . The reactions of NH 2 SO 3 − with • OH and H • resulted in the formation of two transient radical species, one with maximum absorbance at 300 nm and a second with maxima at both 300 nm and 600 nm. These spectra are tentatively assigned to • NH 2 SO 3 and • NHSO 3 − , respectively. By measuring the absorbance of these radicals as a function of pH, the radical pK a was determined to be 9.5 ± 0.1. Overall, this work has implications for the longevity and performance of ferrous sulfamate, Fe(NH 2 SO 3 ) 2 , as a plutonium reductant in the reprocessing of used nuclear fuel.

Conrad, Jacy K. [Idaho National Laboratory (INL), ↗
DOE OSTI2024

Iodine Capture Studies of Silver Mordenite and Novel Alternative Metal Sorbents

Radioiodine is one of the radionuclides of concern when considering reprocessing of used nuclear fuel (UNF). It is expected to be released primarily in the dissolver off-gas (DOG) stream. Sorbents that target I 2 removal from a UNF processing off gas must perform efficiently under elevated temperatures in the presence of water vapor and nitrogen oxide gasses (NOx). Many studies have been conducted that examine the adsorption of iodine species in the presence of NOx gases through reduced silver-exchanged mordenite (Ag 0 Z), a zeolite mineral. This study characterizes the effects of iodine, water vapor, and nitrogen dioxide on the adsorption of I 2 by Ag 0 Z at lower temperatures suitable for comparison with novel Cu- and Bi- bearing sorbents which offer a non-RCRA alternative. When exposed to ‘harsh’ conditions such as I 2 , H 2 O, and NO 2 vapors carried by air, Cu 2 S-polyacrylonitrile (PAN) is more efficient than Ag 0 Z and all other novel sorbents, providing with 11% I capture efficiency. However, under ‘ideal’ conditions in which I 2 is carried by dry air), Ag 0 Z is more efficient than Cu 2 S-PAN and other novel sorbents, providing 74% I capture efficiency. To investigate this result and the overall performance of all sorbents tested, several characterization techniques were employed, including thermal gravimetric analysis, scanning electron microscopy – electron dispersion X-Ray spectroscopy, powder X-ray diffraction, and X-ray photoelectron spectroscopy. These techniques are discussed within this report. This document also includes a design and test plan update from Idaho National Laboratory as the next step in testing the high performing sorbents from thin-bed tests carried out by Oak Ridge National Laboratory.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Mechanical Properties of Irradiated U-10wt%Mo Alloy Degraded by Porosity Development

A plate-type nuclear fuel consisting of a solid monolithic foil of U-10wt%Mo is under development for use in the United States’ high performance research reactors. In support of developing this fuel, the fuel has been fabricated for the first time by a commercial fuel vendor and subsequently irradiated in a test reactor. This provides an opportunity to evaluate post-irradiation mechanical properties of commercially fabricated fuel. Four-point bend testing was conducted on the irradiated U-10Mo fuel and the data produced includes bending strength and Young’s modulus. Although the material behaves in a brittle manner due to the developed porosity, a general trend of strength and modulus reduction are found as fission density increases. The data produced is evaluated using both Weibull statistics and a modulus degradation model with recommendations provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Advancements in modeling fuel pulverization and cladding behavior during a LOCA

During a loss-of-coolant accident (LOCA), it is possible for nuclear fuel rods to undergo a three-step process known as fuel fragmentation, relocation, and dispersal (FFRD). The chance of FFRD occurring increases as the fuel burnup increases. To support the nuclear industry's desire to increase the discharge burnup of nuclear fuels in light-water reactors (LWRs), it is imperative to understand the mechanisms driving the evolution of FFRD. In this work, a multiscale modeling approach is used to garner insight into underlying mechanisms leading to the fine fragmentation (also known as pulverization) of nuclear fuel during a LOCA. This report includes a summary of the atomistic and phase-field studies to develop a new pulverization criterion for use in the engineering-scale Bison fuel performance code. Details are also provided on cladding model improvements related to hydrogen/hydride embrittlement and damage and anisotropic thermal creep. The new models are used on the existing integral and separate effects LOCA validation cases available in Bison.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

A realistic TRACE PWR LOCA model with application to high-burnup analysis

This paper presents a TRACE model of a postulated large-break loss-of-coolant accident (LBLOCA) event in a four-loop pressurized water reactor (PWR). The input files have been made publicly available and serve as a starting point for researchers to improve upon or apply to their purposes. The model, based on a legacy PWR model, contains numerous improvements and modifications consistent with US Nuclear Regulatory Commission LBLOCA analysis guidelines. The model is applied to analyze high-burnup (>62 GWD/MTU) fuel behavior during LBLOCA, improving on previous work to provide realistic thermal hydraulic predictions for future fuel performance analyses and experiments related to fission fragment, relocation, and dispersal (FFRD). Sensitivity studies are presented to quantify the impact of the vessel-modeling approach, core radial discretization, and other phenomena. In conclusion, the model is intended for research purposes and contains sufficient plant geometric, operational, and safety system details to represent the main physical phenomena pertaining to LBLOCA.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗