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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 559 records · Page 31
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Additional Confirmatory LBL Analysis of AGR-5/6/7 Compacts and Over-Coated Particles

Fuel compacts for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) were fabricated by BWX Technologies Nuclear Operations Group (BWXT NOG) located in Lynchburg, Virginia. Two compact packing fractions (PFs) were produced - nominally, 40% PF and 25% PF - in which the tristructural-isotropic (TRISO)-coated particle volume was targeted to be approximately 40% and 25% of the total compact volume, respectively. The TRISO coatings were deposited using a 150 mm diameter fluidized-bed chemical vapor deposition (CVD) furnace on spherical kernels that were nominally 425 μm in diameter. The kernels were from kernel composite Lot J52R-16-69317, which contained low-enriched uranium (15.5% 235 U) in a mixture of uranium carbide and uranium oxide (UCO). Kernels were coated with four concentric CVD layers: a porous carbon buffer layer that was nominally 100 μm thick, an inner pyrolytic carbon (IPyC) layer that was nominally 40 μm thick, a silicon carbide (SiC) layer that was nominally 35 μm thick, and an outer pyrolytic carbon (OPyC) layer that was nominally 40-μm-thick. Coated particle composite J52R-16-98005 was overcoated (OC) with a graphite/resin blend, and these OC TRISO particles were pressed to form cylindrical compacts that were nominally a half inch in diameter and one inch long.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1970

NERVA reactors

NERVA nuclear rocket reactors, considering NRX and XE technology tests

Johnson, M. T.↗
NASA NTRSDate not supplied

Mini-Nozzles for Individual Fuel Elements for NTP Systems NETS

The Testing Reference Design (TRD) reactor and engine for Nuclear Thermal Propulsion (NTP) systems typically use a large single nozzle with regenerative cooling which contributes a significant portion of the engine mass. Furthermore, if a fuel element is damaged or the flow is otherwise compromised, stagnant or backflow is expected which would further damage the reactor and engine. Therefore, the current analysis attempts to solve this potential issue by considering individual, smaller nozzles allocated for each individual fuel element while also neglecting regenerative cooling flow due to lower hoop stresses from the chamber pressure. However, a material that can withstand the chamber temperatures should be considered such as molybdenum or tungsten or a mixture of the two.

Daria Nikitaeva↗
DOE OSTI2018

Drawings for Vendor Reference

This information pertains to process instrumentation and control (P&ID) for systems and subsystems of the Advanced Test Reactor. The systems and subsystems are NOT nuclear safety related in nature, but are support systems for the reactor. This information is being released to support future modifications, replacement or additions to the existing plant systems.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Equipment Procurement and Final Design for the MARVEL High-Grade Heat Extraction System (HGHES)

This summer I worked on the design of a high-grade heat extraction system (HGHES) of the MARVEL Project. The MARVEL project at INL is a first-of-a-kind terrestrial microreactor to help provide a test platform to demonstrate the use of a microreactor to end customers and enable technology developers to test enabling microreactor technologies with a functional nuclear reactor. One of the project's main requirements is the reactors’ ability to deliver high-grade thermal power to end users, both for industrial and commercial use. The project required communication with multiple vendors for custom-made equipment, working with a high-temperature and high-pressure helium coolant system, and research of a Thermal Storage Unit (TSU).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRSDate not supplied

Optimization Considerations for Annular Linear Electromagnetic Pumps for Space Nuclear Power and Propulsion

The Need for High Performance Liquid Metal Pumps Testing Power and Efficiency Considerations Liquid metal cooled reactors is one option for space nuclear power for settlements and electric propulsion vehicles (Snap 10 A, SP100, Soviet Topaz reactors) • High heat capacity • High boiling point (NaK~1200K) • Simplified heat transfer loop • Good neutronics Liquid metal EM pumps are desired over conventional pumps owing to lack of seals, bearings elimination of cavitation issues, lower mass and eliminates mechanical impeller • Conduction pumps (typically DC-high current, low voltage; ~5% efficiency) • Induction pumps (typically AC, high efficiency ~40%) Annular linear induction pump is an EM pump with no moving parts and of cylindrical/tubular geometry which makes it straightforward to integrate in conventional piping architectures • Higher inherent reliability • Invented by Einstein and Szilard

ALIP↗
DOE OSTI2016

Microstructural Characterization of In-reactor Corrosion Tested Hastelloy N{sup R} and 316 Stainless Steel

The success of the Molten Salt Reactor Experiment (MSRE) at the Oak Ridge National Laboratory (ORNL) in the period 1950's-70's has rekindled the interested in molten salt-cooled reactors. The most recent form of this type of reactor, the Fluoride salt-cooled High-temperature nuclear Reactors (FHRs) is emerging as a leading reactor concept for the next generation nuclear reactors because it offers, among other benefits, a high degree of passive safety, high thermal efficiency, and low spent fuel. Since 2011, two three-year Integrated Research Projects (IRPs) were awarded by the U. S. Department of Energy (DoE) to the Massachusetts Institute of Technology (MIT) in partnership with the University of California, Berkeley and the University of Wisconsin-Madison to develop a 'path forward' for the development of the FHR. One primary challenge in these projects is to evaluate the corrosion and compatibility of structural alloys in the potential primary coolant of molten Li{sub 2}BeF{sub 4} (FLiBe) salt at 700 deg. C in a neutron irradiation environment. In support of structural materials development for the FHR, the static in-reactor corrosion tests of two promising alloys, nickel-based Hastelloy N{sup R} and 316 stainless steel, in {sup 7}Li-enriched molten FLiBe have been conducted in both metal-lined and graphite crucibles in the MIT nuclear research reactor in 2013. This work focuses on the microstructural characterization of the in-reactor corrosion tested alloys for understanding of their performance in high-temperature molten FLiBe salt. Our preliminary results indicate significant thermal effects, graphite effects, and irradiation effects on the alloys' performance in high-temperature molten FLiBe salt, particularly for the in-reactor corrosion tests. The neutron irradiation induces tritium generation in molten salt which could combine with fluorine ions to form corrosive HF to dramatically attack alloys. Additionally, the neutron irradiation introduces structural defects into alloys which probably alter alloys' corrosion resistivity to high temperature molten FLiBe salt. The effect of graphite on the alloys' corrosion in a nuclear reactor is also possibly different from the out-of-reactor corrosion. All these issues will be better understood by characterizing the microstructure of in-reactor corrosion tested alloy samples in our near future plans. The experimental data in this study provides valuable input for evaluating structural materials performance in the FHR. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2024

Refining Absorber Shroud Geometry to Maximize Power Output and Reduce Power Peaking in ATF Test Train

In the wake of the Fukushima Daiichi nuclear power plant accident in 2011, the accident tolerant fuels (ATF) program was initiated to enhance the safety of light-water reactor fuels, placing significant emphasis on cladding. A crucial step for the broad implementation of ATF in commercial reactors involves irradiation testing of the fuel designs. ATF-2D, the latest experiment in the ATF series, is slated to undergo irradiation in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). ATF-2D is a joint effort of the INL with industry partners General Electric Global Research; Framatome; General Atomics; the Japan Atomic Energy Agency; Hitachi-GE Nuclear Energy, Ltd; Global Nuclear Fuel-Japan Co., Ltd; Nippon Nuclear Fuel Development Co., Ltd; and Mitsubishi Heavy Industries, Ltd. The ATF-2D test train design consists of four tiers, each housing six rodlets. The device will be inserted in Loop 2A within the central flux trap of the ATR, and is anticipated to undergo irradiation throughout three 60-day cycles. Typical pressurized water reactor conditions will be emulated during the irradiation. The objective of the work presented here is to dimension neutron-absorbing hafnium (Hf) components surrounding the fuel rodlets, such that the axial power profile is flattened, while simultaneously ensuring that the total fission power output of the entire test train remains below 200 kW.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Fabrication, Welding, and Inspection Techniques in Preparation for In-Reactor Testing of Accident Tolerant Fuels

After the Fukushima events in 2011, DOE-NE in collaboration with nuclear industry shifted R&D emphasis to accident performance of LWR fuels under extended loss of active cooling and steam exposure. DOE-NE has created a roadmap for the “Development of Light Water Reactor Fuels with Enhanced Accident Tolerance.” The mission of the Accident Tolerant Fuel (ATF) Roadmap is to develop the next generation of LWR fuels with improved performance, reliability, and safety characteristics during normal operations and accident conditions and with reduced waste generation. The ultimate goal of the ATF roadmap is to support the insertion of lead fuel rods (LFRs) or lead fuel assemblies (LFAs) of an Accident Tolerant Fuel into a commercial LWR within 10 years (i.e., by the end of FY-2022). As a step toward this goal, an irradiation test series has been developed to assess the performance of proposed ATF concepts under normal LWR operating conditions. Data generated by this test program will be used to establish the feasibility of certain aspects of proposed ATF concepts, as well as provide information to support screening among concepts; as such, it is an integral part of Phase I: Feasibility Assessment and Down-Selection outlined in the ATF Roadmap. This irradiation test series is planned to be performed as a series of drop-in capsule tests to be irradiated in the Advanced Test Reactor (ATR) operated by the Idaho National Laboratory (INL), and it has been designated as the ATF-1 test series. Current fission reactors use zirconium-based fuel cladding because of its extremely low macroscopic thermal neutron absorption cross-section, good high temperature strength, and decent corrosion resistance. However, advanced, innovative materials may provide these same benefits while increasing reactor safety margin, core power density, and fuel utilization. These advanced fuel cladding systems will allow revolutionary cladding performance and enhanced fuel mechanical designs, however, challenges exist in design, analysis and fabrication of innovative, never before tested, fuel cladding systems for in-reactor testing. This paper highlights the challenges associated with design, fabrication and welding, and inspection of innovative materials and actions taken to address those challenges in preparation for the Phase I ATR irradiation testing. The lessons learned from Phase I of this experiment can be used to guide researchers for design and analysis of future in-reactor testing of advanced fuel cladding systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Refining Absorber Shroud Geometry to Maximize Power Output and Reduce Power Peaking in ATF Test Train

In the wake of the Fukushima Daiichi nuclear power plant accident in 2011, the accident tolerant fuels (ATF) program was initiated to enhance the safety of light-water reactor fuels, placing significant emphasis on cladding. A crucial step for the broad implementation of ATF in commercial reactors involves irradiation testing of the fuel designs. ATF-2D, the latest experiment in the ATF series, is slated to undergo irradiation in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). ATF-2D is a joint effort of the INL with industry partners General Electric Global Research; Framatome; General Atomics; the Japan Atomic Energy Agency; Hitachi-GE Nuclear Energy, Ltd; Global Nuclear Fuel-Japan Co., Ltd; Nippon Nuclear Fuel Development Co., Ltd; and Mitsubishi Heavy Industries, Ltd. The ATF-2D test train design consists of four tiers, each housing six rodlets. The device will be inserted in Loop 2A within the central flux trap of the ATR, and is anticipated to undergo irradiation throughout three 60-day cycles. Typical pressurized water reactor conditions will be emulated during the irradiation. The objective of the work presented here is to dimension neutron-absorbing hafnium (Hf) components surrounding the fuel rodlets, such that the axial power profile is flattened, while simultaneously ensuring that the total fission power output of the entire test train remains below 200 kW.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Magnox Reprocessing TDN Reactors: Utilising 3D Printing and X-Ray Imaging to Re-Design and Test Fluidising Air Nozzles - 16402

One of the key final steps in reprocessing spent nuclear fuel is the conversion of highly concentrated uranyl nitrate hexahydrate liquor [> 1000 gU/litre] into uranium trioxide (UO{sub 3}) powder by thermal decomposition using Thermal Denitration (TDN) reactors. The TDN reactors in the Magnesium non-oxidising (Magnox) fuel reprocessing plant are high aspect ratio fluidising beds, which suffer from poor fluidization and UO{sub 3} powder blockages. It is widely understood that these issues arise from the inadequate design of the fluidising air distributors (nozzles), which are located around the base of the TDN reactors. A non-active test rig was constructed at the National Nuclear Laboratory (NNL) experimental facility in Workington, Cumbria (UK) which used compressed air at ambient temperature to assess the performance of the current fluidising air nozzle designs (datum) against a range of new nozzle designs. Full-scale nozzle designs were manufactured using a 3-Dimensional (3D) printer and tested to obtain data of both quantitative and qualitative nature. NNL then collaborated with the Chemical Engineering Department at University College London (UCL) to further assess the fluidising potential of the new nozzle designs. Images of inside the fluidized bed were taken utilising a unique X-Ray facility. The results from this study will allow new nozzle designs to be recommended, manufactured and installed, which will ensure improved fluidization of the TDN reactor bed and prevent blockages with UO{sub 3} powder. (authors)

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

Research to Develop Flood Barrier Testing Strategies for Nuclear Power Plants

The U.S. Nuclear Regulatory Commission has developed regulations regarding the siting and design of nuclear power plants (NPPs) that are aimed at addressing various natural hazards, including flooding. Flood barriers are designed to prevent water from entering NPP areas containing structures, systems, and components (SSCs) important to safety. The barriers are used at NPPs along with drains, sumps, pumps, valves, plugs, and site grading as part of the plant flood protection features that protect SSCs from experiencing external or internal flooding and mitigate the effects of flooding on NPP operations. The performance of flood protection features, including flood barriers at NPPs, has been an ongoing concern. Domestic and international operational experience provides clear indications that flood barrier performance has significant safety implications, especially for aging NPPs. The observed deficiencies show that flood barriers should be designed and installed properly, then adequately tested, inspected, and maintained in order to ensure that they perform their intended functions during flooding events. Here, this paper reviews available information related to flood barriers employed at U.S. NPPs and provides an overview and categorization of NPP flood barriers. It identifies potential domestic and international flood barrier testing facilities, including operating and decommissioned U.S. NPPs. Finally, this paper presents the technical and logistical considerations that should be made when developing specific testing strategies and protocols for flood barriers, such as the selection of flood barriers, test locations, testing approach, performance criteria, and testing parameters.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

TREAT LEU Conversion Analyses Performed 2016-2018

TREAT is an experimental nuclear reactor designed to produce high neutron flux transients for the testing of reactor fuels and other materials. After more than 20 years of non-operational standby, TREAT has been restarted under the U.S. Department of Energy’s Resumption of Transient Testing Program. The conversion of TREAT from its current highly enriched uranium fuel to a low enriched uranium fuel is being pursued in keeping with the mission of the Department of Energy National Nuclear Security Administration’s Material Management and Minimization Reactor Conversion Program. This report documents additional analyses performed over the timeframe of 2016 to early 2018 to evaluate the LEU Conceptual Design for the TREAT reactor, following the initial design studies documented in the 2016 Conceptual Design Report. The analyses documented in this report include neutronic and thermal hydraulic sensitivity studies to evaluate the effect of key design parameters on predicted core behaviors, as well as a simple investigation into the effect of fuel thermal expansion.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Preliminary Options Assessment of Versatile Irradiation Test Reactor

The objective of this report is to summarize the work undertaken at INL from April 2016 to January 2017 and aimed at analyzing some options for designing and building a versatile test reactor; the scope of work was agreed upon with DOE-NE. Section 2 presents some results related to KNK II and PRISM Mod A. Section 3 presents some alternatives to the VCTR presented in [ ] as well as a neutronic parametric study to assess the minimum power requirement needed for a 235U metal fueled fast test reactor capable to generate a fast (>100 keV) flux of 4.0 x 1015 n /cm2-s at the test location. Section 4 presents some results regarding a fundamental characteristic of test reactors, namely displacement per atom (dpa) in test samples. Section 5 presents the INL assessment of the ANL fast test reactor design FASTER. Section 6 presents a summary.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗