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179 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Preliminary Conceptual Design of Nuclear Thermal Rocket Reactor Cores Using Ceramic Fuels with Beryllium or Composite Neutron Moderators

Herein several preliminary conceptual designs of nuclear thermal rocket reactor cores are presented that use tin-bonded monolithic ceramic [mononitride (UN), monocarbide (UC), and uranium dioxide (UO 2 )] fuel plates or pins with molybdenum-tungsten alloy clad. Neutron moderation is provided by a block of Be metal or composite materials using metal hydrides such as ZrH 1.6 or YH 1.6 with different matrices (MgO or Be). Mainly high-assay low-enriched uranium is considered, but highly enriched uranium is also assessed for a few configurations. Nominal core thermal power is 300 MW corresponding to about 66 kN (15 klbf) of thrust, and with minimal modifications, 500 MW may be possible (25 klbf of thrust). Depending on the configurations, the amount of 235 U needed for criticality is 30 to 90 kg, and reactor weight is 2.5 to 3.8 tonnes.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Expanded Validation of Uranium Systems with the KENO Monte Carlo Codes and SCALE 6.2.4

The development of accident-tolerant fuel and advanced reactors fueled with high-assay low-enriched uranium has created an increased focus on systems with uranium enrichments between 5 and 20 wt % 235 U. SCALE criticality safety validation reports generated at Oak Ridge National Laboratory contain very few benchmark experiments in this range. The work presented herein is an initial investigation of expanded validation of the entire range of enrichments for uranium-fueled systems, including consideration of over 300 newly developed KENO models. The vast majority of the new models are between 4.4 wt % and 37.5 wt % enrichment and include a variety of thermal, intermediate, and mixed neutron spectra. This work finds no strong evidence of an enrichment-dependent bias, but there are indications of a spectrum-dependent bias. There are, however, only a limited number of independent experiments included at this time to distinguish intermediate enrichment effects from intermediate spectrum effects. Further work in this area is ongoing for intermediate enrichment systems. In the long run, the addition of these models to the Verified, Archived Library and Inputs and Data is planned, as funding and staff availability allow.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Analysis of Radiological Impacts to Terrestrial Biota in Support of Environmental Assessment for Use of DOE-Owned High-Assay Low-Enriched Uranium Stored at Idaho National Laboratory

There are many species of plants and animals (biota) on the INL Site and facilities such as INTEC and MFC have surrounding habitat supporting resident plant and animal populations. Consequently, radionuclide releases from the HALEAU processing facilities could pose a potential radiological impact to nearby flora and fauna populations. Therefore, the RESRAD-BIOTA (Version 1.8) computer code (http://resrad.evs.anl.gov/codes/resrad-biota/) was used to assess the impact of radiation exposures to terrestrial biota resulting from soil contaminated by airborne releases from the HALEAU processing facilities proposed for the INTEC and/or MFC complexes. The contaminated soil subsequently results in contamination in air and in different food sources used by biota. Both external radiation and internal radiation are considered in the dose evaluation.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2022

Expanded Validation of Uranium Systems with the KENO Monte Carlo Codes and SCALE 6.2.4 [Slides]

There is likely a trend in bias for uranium-fueled systems as a function of spectrum, but probably not as a function of enrichment. Some enrichment trends appear, but are potentially the result of outlier experimental results; further investigation is needed where possible. There are good examples in this data of markedly different biases from different systems. Future work includes review and vetting for inclusion in the VALID library and analysis with TSUNAMI.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Critical Experiments in Support of Current Reactors and Advanced Reactor Deployment [Slides]

Limited, specific needs exist to support current reactor fleet and fuel cycle. Primarily, MOX experiments representative of low burnup LWR fuel are needed. LEU+ power distribution measurements may also be useful. Advanced reactors present many challenges and opportunities. New fuel forms and associated processes lack validation. Initial core criticality calculations can be validated with appropriate, highly similar experiments. Transportation of LEU+/HALEU material will be essential, and will rely on new experiments to support validation

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Fuel Fabrication Capability Assessment in Support of Advanced Reactor Deployments

More than 30 U.S. companies are designing a variety of advanced reactor concepts, and several companies are planning to demonstrate their reactor designs in the mid-2020s to late 2030s time frame. In 2020, the U.S. Department of Energy (DOE) announced a series of awards under the Advanced Reactor Demonstration Program (ARDP) to accelerate the successful deployment of 10 of these reactors under three pathways. TerraPower and X-energy were awarded grants under the Advanced Reactor Demonstration Program to deploy their respective Natrium reactor and Xe-100 reactor designs in the next 7–10 years. These demonstrations are in addition to several parallel programs, including the U.S. Department of Defense’s (DoD’s) interest in the development of microreactors, and interest of the National Aeronautics and Space Administration in space nuclear power and propulsion. The National Reactor Innovation Center’s (NRIC’s) mission is to accelerate the demonstration and deployment of advanced reactors; NRIC is partnering with several reactor developers and harnessing the world-class capabilities of the U.S. National Laboratory system to deliver on its mission. Several of these reactor designs will require advanced fuel forms that are not commercially available today, including metal fuel, molten salt fuel, TRi-structural ISOtropic (TRISO) particle fuel, and uranium nitride fuel. Recognizing that there may be potential gaps in the laboratory-scale process development and pilot-scale first-of-a-kind (FOAK) production of these fuel forms leading to delivery of the FOAK cores, NRIC commissioned this study to look at the challenges that need to be overcome for successful deliveries, including the evaluation of existing facilities and the potential need for a new fuel fabrication facility.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

SPARC - Plans for a New Critical Experiment Facility with a Horizontal Split Table

Several critical experiment facilities, sometimes referred to as zero power reactor facilities, have provided crucial data to aid understanding and validate nuclear-physics models since the beginning of nuclear technology. Indeed, the first man-made reactor, Chicago Pile-1, was essentially this type of reactor. However, there was a downturn in nuclear technology development toward the turn of the millennium, and the need for these specialized research facilities waned. Now there are few of these experimental facilities operational in the world and those that remain have relatively small critical assembly machines. The need for criticality safety benchmark experiments at intermediate neutron energy levels and the modern resurgence of interest in advanced reactors designs, many of which do not have historical precedents in terms of nuclear fuel composition, moderator, and coolant combinations, all combine to create a substantial need for a critical experiment facility with a large horizontal split-table (HST) machine. A HST machine is used to arrange two separate and subcritical parts of a core assembly, bring them together in a precise manner to achieve criticality using remote controls, and separate them to achieve a subcritical configuration again. A new effort was recently performed to develop user needs for a HST, assess candidate locations at the Idaho National Laboratory (INL), and develop a plan for deployment. This project is referred to as the System Physics Advanced Reactor Critical facility (SPARC). A few months after this assessment began, and shortly after as a viable pathway was emerging, a series of important presidential executive orders were issued to revitalize nuclear energy in the United States (U.S.). The relevance of SPARC to these executive orders was immediately apparent. The far-reaching potential of SPARC to these executive orders will reside in its ability to produce data which facilitates licensing of advanced nuclear reactor designs while reducing uncertainties to help increase energy production alongside new criticality safety data to enable more efficient nuclear fuel manufacture, transport, and storage.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Core Physics Characteristics of Extended Enrichment and High Burnup Boiling Water Reactor Fuel

This paper presents the highlights of boiling water reactor (BWR) core physics studies performed at Oak Ridge National Laboratory as part of a series of studies conducted to compare low-enriched uranium (LEU) with LEU+ fuel. The studies analyzed isotopic fuel content, lattice parameters (Phase 1), and core physics (Phase 2) to identify challenges in operation, storage, and transportation for BWRs and pressurized water reactors (PWRs). Because of a lack of publicly available lattice and core designs for modern BWR fuel assemblies and reactor cores, several optimized lattice designs were generated, and different core loading strategies were investigated. Twelve optimized lattice designs with 235 U enrichments ranging from 1.6% to 9% and gadolinia loadings ranging from 3 to 8 wt% were used to model axial enrichment and geometry variations in fuel assemblies for core designs. Each core shares a common set of approximations in design and analysis to allow for consistent comparisons between LEU and LEU+ fuel. The objective is to highlight anticipated changes in core behavior with respect to the reference LEU core. The results of this study show that the differences in LEU and LEU+ core reactor physics characteristics are less significant than the differences in lattice physics characteristics reported in the Phase 1 studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Expression of Interest for off-gas system design build

Idaho National Laboratory (INL) and Pacific Northwest National Laboratory (PNNL) are studying a used nuclear fuel reprocessing system that could be used to recycle used nuclear fuel by separating waste fission products from re-useable, valuable uranium that can be used in making new nuclear fuel. This process includes dissolving the mixture of uranium and fission products in a nitric and hydrofluoric acids, then using an organic extraction process to remove and recover the uranium. The waste fission products remain in the acidic aqueous “raffinate” stream.” The fission products in this raffinate waste stream are expected to require stabilization in a vitrified glass waste form for safe disposal. INL and PNNL plan to construct and operate a facility in which to test the performance of the electric melter process that would receive a non-radioactive but chemically similar simulant of this raffinate. This test system will be located at INL in or near Idaho Falls, ID. In the melter, which operates at a temperature of about 1,050 to 1,150oC, the water, nitric acid, and other volatile and semi-volatile materials evolve into an off-gas stream, while the fission products (in this case, non-radioactive simulants of fission products) are melted with glass forming chemicals into a molten glass. Upon cooling, the molten glass provides a durable waste form for long-term disposal. Expressions of interest are requested from companies who would like to receive a request for proposal to design and build this off-gas system. We expect to request proposals and desire to award a contract prior to the end of September 2019.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Characteristics of TRU Fueled SFR Core during the Transition with Partially Loaded LEU Assemblies

The fast reactor (FR) fuels typically require higher fissile content than the thermal reactor fuels because the cross sections of the fissile isotopes are smaller relative to the fertile isotopes and higher neutron leakage from the reactor core. The medium-sized high-burnup sodium-cooled fast reactors (SFRs) in the framework of uranium-plutonium/transuranic (U/Pu or U/TRU) fuel-cycle can be optimized to operate with less than 17 % of Pu/TRU content in the driver fuel while having higher TRU conversion ratios (CR~1.0) [1]. However, possible shortage or inaccessibility to TRU based fuel could be accommodated by the use of uranium-based fuels on demand such as low enriched uranium (LEU) having significantly higher 235U content than used in the commercial light water reactors (LWRs), which is defined as high-assay LEU. The reactor systems having low TRU conversion ratio are commonly categorized as burner reactors, which include LWRs and even can be achieved by FRs with designed to utilize higher TRU fractions (e.g., increased neutron leakage and lower fuel volume fractions). Such reactor systems need to be initially loaded with high excess reactivity, which decreases toward the end of cycle, in order to achieve target burnup conditions. Therefore, in order to offset the excess reactivity, neutron absorbing materials need to be inserted into the core to compensate (e.g., control rods). By reducing the neutron leakage (e.g., increasing the fuel volume fraction or thicker blankets), the TRU conversion ratio in the FR can be increased. If the TRU conversion ratio becomes 1.0 (technically slightly greater than 1.0 to account for process losses during recycle), the reactor is in the condition known as “breakeven”, in which fuel contains the same amount of fissile material after recycling that was used to fuel the reactor. In such a condition, combined with the fact that fission products do not have strong poisoning effects in FRs, the reactivity swing becomes smaller than that of burner type reactors. For breeders, the reactivity can increase toward the end of the cycle because more fissile material is produced than consumed. In this regard, the breakeven FR core is implicitly optimized in a sense by eliminating the large insertion of control elements to offset the reactivity during the normal operation. However, the breakeven condition here is assumed to be achieved only by the TRU based fuel. Possible lack of TRU based fuel needs to be compensated by available LEU resources containing equivalent amount of fissile (i.e., 235U). The use of LEU will have some impact on important safety parameters that need to be evaluated. U-235 has a higher effective delayed neutron fraction than Pu-239. Also, the number of neutrons produced by fissions per neutron absorption (i.e. ?) of 235U in the fast energy range (>1e4 eV) is lower than that of 239Pu, which may lead to the poor fissile conversion when LEU is used as a high-conversion FR fuel. The full-paper presents the study of the impact of partially loaded LEU fuels within TRU fueled breakeven SFR on the transition of nuclide evolutions and core characteristics particularly focusing on the burnup reactivity swing. The full paper analyzes how the large reactivity swing occurs during the LEU loading and addresses possible options to reduce it to the level equivalent to that of the TRU fueled SFR core.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Transport Modeling of As-Run ATR Cycles to Support U-10Mo Research Reactor Fuel Qualification Experiment

The Department of Energy’s (DOE) Office of Materials Management and Minimization has been tasked with converting the five remaining United States High Performance Research Reactors (U.S. HPRR) from highly enriched uranium to low-enriched uranium. The Nuclear Regulatory Commission (NRC) regulates Massachusetts Institute of Technology Reactor (MITR), Missouri University Research Reactor (MURR), and National Bureau of Standards Reactor (NBSR); and DOE regulates the High Flux Isotope Reactor and Advanced Test Reactor (ATR). To meet the high demands of these reactors, the U.S. HPRR program has chosen to use 90% uranium - 10% molybdenum (U-10Mo) monolithic fuel. This plate-type fuel will undergo multiple irradiation experiment campaigns in ATR, from mini-plates to full element tests, over a large range of operating conditions. This will provide data in support of the fuel qualification of each reactor. This summary focuses on the as-run neutronic analysis of the first series of mini-plate (MP-1) experiments, which have been irradiated in the ATR. MP-1 experiment’s main goal is to demonstrate the fabrication process and meet the fuel irradiation performance requirements with primary focus on the fuel plates associated with the three NRC reactors, with a small focus on the low power requirements of the ATR fuel. The MP-1 experiments are planned to be the basis of the monolithic U-10Mo fuel for qualification through the NRC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Transport Modeling of As-Run ATR Cycles for U-10Mo Fuel Qualification Experiment

The Department of Energy’s (DOE) Office of Materials Management and Minimization has been tasked with converting the five remaining United States High Performance Research Reactors (U.S. HPRR) from highly enriched uranium to low-enriched uranium. The Nuclear Regulatory Commission (NRC) regulates Massachusetts Institute of Technology Reactor (MITR), Missouri University Research Reactor (MURR), and National Bureau of Standards Reactor (NBSR); and DOE regulates the High Flux Isotope Reactor and Advanced Test Reactor (ATR). To meet the high demands of these reactors, the U.S. HPRR program has chosen to use 90% uranium - 10% molybdenum (U-10Mo) monolithic fuel. This plate-type fuel will undergo multiple irradiation experiment campaigns in ATR, from mini-plates to full element tests, over a large range of operating conditions. This will provide data in support of the fuel qualification of each reactor. This summary focuses on the as-run neutronic analysis of the first series of mini-plate (MP-1) experiments, which have been irradiated in the ATR. MP-1 experiment’s main goal is to demonstrate the fabrication process and meet the fuel irradiation performance requirements with primary focus on the fuel plates associated with the three NRC reactors, with a small focus on the low power requirements of the ATR fuel. The MP-1 experiments are planned to be the basis of the monolithic U-10Mo fuel for qualification through the NRC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

AFC Metal Fuel Workshop Presentations

Compilation of PowerPoint slide decks from presentations given at the 3rd Annual AFC Metal Fuels Workshop. The purpose of this workshop is to bring the national metallic fuel stakeholders together to discuss status and needs of metallic fuel and review the DOE AFC Metal Fuel R&D. Desired Outcomes - Advance Metallic Fuel Technology: 1. Stakeholder Community Building 2. National Program R&D Accountability, Planning Support, and Increased Stakeholder Advocation.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗