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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 793 records · Page 44
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to provide hermetic seal. The replacement of a zirconium alloy using a ferritic material containing chromium and aluminum appears to be the most near term implementation for accident tolerant nuclear fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Accelerating Nuclear-Integrated Data Centers in the USA: SWOT Analysis, Power-Thermal Management Strategies, and Industrial-Scale Demonstration and Potential Deployment

Driven by the growth in digital services, cloud computing, AI, and manufacturing, data centers face rising energy demands that challenge traditional power sources and cooling efficiency. This study explores using nuclear power to meet these demands, focusing on accelerated reactor technology deployment and highlighting needs such as N+1/N+2 power supplies and integrated power-thermal management. A SWOT analysis addresses grid connectivity, reactors, and site selection, particularly DOE sites. Reactor technology demonstration and deployment could be accelerated by leveraging test facilities such as MARVEL, MAGNET, TED, FAS, DOME, LOTUS, ATR, Energy System Proving Grounds, and upcoming Energy Launch Pads, along with modeling and simulation tools such as RELAP5, MOOSE, VERA, RAVEN, and FORCE. The potential power and thermal management options, including various cooling technologies, waste-heat utilization, and an industrial-scale demonstration plan, aim to accelerate the integration of nuclear power and data centers in the USA, while emphasizing community and stakeholder engagement and synergistic efforts.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2020

As-run thermal-hydraulics analysis of the EPRI-2 experiment

The EPRI-2 experiment was designed to irradiate various types of reactor pressure vessel steels at a temperature of 288°C (PLN-3934). The specimens were irradiated in a non-instrumented test train inside a pressurized water loop in the center lobe of the ATR during PALM cycle 153B. Temperature was indicated by post-irradiation examination of melt wires placed within the test train. The purpose of this analysis is to calculate specimen temperature using measured data on reactor power and as-run calculations of heating rates of the test train. The experiment contains several melt wires spanning the temperature range 239°C to 327°C. The accuracy of the model is assessed by comparing the measured and calculated in-pile tube inlet to outlet temperature difference, and comparing the calculated specimen temperature to the temperature range indicated by examination of the melt wires.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

As-run thermal-hydraulic analysis of the EPRI-3 experiment

The EPRI-3 experiment was designed to irradiate various types of reactor internal component steels at a temperature of 288°C. The specimens were irradiated in an instrumented test train inside a pressurized water loop in the center lobe of the ATR during cycles 155B and 158B. Temperature was monitored using thermocouples placed at the top of the test train and melt wires placed within the test train. The purpose of this analysis is to calculate specimen temperature using measured data on reactor power and as-run calculations of heating rates of the test train. The experiment contains four thermocouples and several melt wires spanning the temperature range 239°C to 327°C. The accuracy of the model is assessed by comparing the measured and calculated in-pile tube inlet to outlet temperature difference, comparing the measured and calculated thermocouple temperatures, and comparing the calculated melt wire temperature to the temperature range indicated by examination of the melt wires.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

The Fission Accelerated Steady State Test (FAST) -- A Revised Capsule Design for the Accelerated Testing of Advanced Reactor Fuels

In an effort to improve the development and testing of novel nuclear fuels, a detailed evaluation on reducing the size of fuel specimens for irradiation has been performed. Theoretically speaking, reducing the fuels diameter and incorporating an inner capsule while maintaining the original outside diameter of the experiment significantly reduces the experiment’s sensitivity to fabrication tolerances and drastically reduces the time necessary to reach a desired fuel burnup; both of which improve the cost and schedule associated with the fuel testing. Simulations of this reduced diameter, double encapsulated concept have been performed using Abaqus, MCNP, and ORIGEN. The results of the Abaqus simulations are beyond the scope of this summary and will not be presented herein. For results of these Abaqus simulations, see Beausoleil [2]. The models were tested using various parameters and design deviations which were then compared to the original Advanced Fuels Campaign (AFC) designs. The results of these simulations confirmed that the introduction of a second capsule coupled to a reduction in the fuel pin diameter significantly reduces the experiment’s sensitivity to fabrication tolerances and reduces the time to reach a desired burnup by nearly an order of magnitude. The original AFC design requires approximately 11.7 years to reach 30 Atom % Heavy Metal (at% HM) burnup. A burnup of 30 at% HM can be reached in approximately 2.5 to 3.5 years (irradiation position dependent) with a fuel diameter ½ (~0.097 in. diameter) that of the original AFC design. An added benefit to the fuel pin diameter reduction is that more specimens can be irradiated in any given cycle. Essentially, significantly more specimens can be tested and incur more burnup over a given operating cycle relative to the AFC design. This allows the development of a large test matrix of samples with rapid throughput in ATR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Projection Analysis for ATR Irradiation of the AFC-FAST Experiment

Analyses of the Advanced Fuels Campaign Fission Accelerated Steady-state Test (AFC-FAST) in the Advanced Test Reactor are presented. A detailed methodology was employed to better account for uncertainties in the planned power and duration of sequential reactor loading cycles. By performing coupled depletion analyses at multiple power levels and durations, the differences in experiment heating outputs can be found. The effects of these uncertainties upon multiple experiment configurations were assessed in an effort to streamline the process of planning for and documenting future irradiations. In conclusion, the data generated from this work have been used to help inform assumptions on subsequent projections to perform only a nominal case depletion.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Challenges and Solutions for Fast Neutron Irradiation of Bulk Material Specimens

Reactor developers continue to recognize opportunities for further enhancing fast spectrum reactor designs with advanced core materials, but all the material test reactors currently available to the United States are thermal spectrum designs. Fortunately, the Advanced Test Reactor and High Flux Isotope Reactor are versatile high flux facilities where spectral modification strategies can be used to reduce undesirable thermal neutron capture transmutation damage and augment fast flux delivered to specimens. New opportunities to leverage high flux regions and specially designed fast flux boosting experiment configurations can be used to achieve meaningful fast fluences on large specimens in ATR. New optimization potentials can be employed to achieve even higher fluences, albeit for smaller specimens, using thermal neutron filters in HFIR test positions. These capabilities, while not true fast reactors, can provide highly relevant environments for researchers needing to study the effects of fast neutron damage in bulk material specimens.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Property analysis and advanced manufacturing technique development for light water reactor annular composite fuel

The University of Florida (UF) has been developing a composite UO{sub 2}-Diamond Fuel design for water-cooled reactors that offers greatly increased thermal conductivity as measured at BOL. Currently, Diamond composite fuel pellets are being irradiated in the ATR to determine whether the thermal conductivity increase is maintained as a function of burnup. UF has further increased the benefits of this increased thermal conductivity of the additive UO{sub 2} fuel by implementing an annular pellet design. The central core material for our annular fuel design must have a high thermal conductivity, a high melting point, a low absorption cross section, a low coefficient of thermal expansion and a neutron thermal moderating capability, which will enhanced the reactivity and performance of the cored pellet. The chosen ZrC core material has a very high melting temperature of 3,700 Celsius degrees, a high thermal conductivity, a good moderating effect, and a low value of thermal expansion. UF has successfully manufactured UO{sub 2} annular pellets. The University of Florida (UF) team believes such an annular pellet can eliminate pellet cracking, reduce fission gas release by over 50% and reduce the maximum pellet temperature by between 300 to 600 Celsius degrees. It is believed that such a pellet design would tolerate a loss of coolant accident (LOCA). (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Ultrasonic Transducer Irradiation Test Results

Ultrasonic technologies offer the potential for high-accuracy and -resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. Other ongoing efforts include an ultrasonic technique to detect morphology changes (such as crack initiation and growth) and acoustic techniques to evaluate fission gas composition and pressure. These efforts are limited by the lack of identified ultrasonic transducer materials capable of long term performance under irradiation test conditions. For this reason, the Pennsylvania State University (PSU) was awarded an ATR NSUF project to evaluate the performance of promising magnetostrictive and piezoelectric transducers in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 10 21 n/cm 2 . The goal of this research is to characterize and demonstrate magnetostrictive and piezoelectric transducer operation during irradiation, enabling the development of novel radiation-tolerant ultrasonic sensors for use in Material Testing Reactors (MTRs). As such, this test is an instrumented lead test and real-time transducer performance data is collected along with temperature and neutron and gamma flux data. The current work bridges the gap between proven out-of-pile ultrasonic techniques and in-pile deployment of ultrasonic sensors by acquiring the data necessary to demonstrate the performance of ultrasonic transducers. To date, one piezoelectric transducer and two magnetostrictive transducers have demonstrated reliable operation under irradiation. The irradiation is ongoing.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

ECAR-3386 AS-RUN THERMAL ANALYSIS OF THE AGC-3 EXPERIMENT

The third Advanced Graphite Creep (AGC-3) experiment was designed to irradiate various types of graphite specimens at a temperature of 900°C. The specimens were irradiated in an instrumented leadout capsule experiment in the east flux trap of the ATR during cycles 152B, 154B, 155A, and 155B. Temperature was monitored using twelve thermocouples located at various elevations in the reactor core, and a helium-argon gas mixture was used for gas gap temperature control of the specimens. The purpose of this analysis is to calculate specimen temperature using measured data on reactor power and helium-argon gas flows, and as-run calculations of heating rates and displacement per atom (DPA) in graphite. The accuracy of the model is assessed by comparing measured and calculated thermocouple temperatures. Uncertainty in gas gaps may preclude an accurate temperature calculation. In these cases, adjustments are made to the thermal model in order to reconcile the measured and calculated thermocouple temperature and to ensure the accuracy of the calculated specimen temperature.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

As-run thermal hydraulics analysis of the EPRI-1 Experiment

The EPRI-1 experiment was designed to irradiate various types of reactor pressure vessel steels at a temperature of 288°C (PLN-3934). The specimens were irradiated in an instrumented test train inside a pressurized water loop in the center lobe of the ATR during cycle 157C. Temperature was monitored using thermocouples placed at the top of the test train. Additional temperature indications were obtained by post-irradiation examination of melt wires placed within the test train and spanning the temperature range 239°C to 327°C. The purpose of this analysis is to calculate specimen temperature using measured data on reactor power and as-run calculations of heating rates of the test train. The accuracy of the model is assessed by comparing the measured and calculated in-pile tube inlet to outlet temperature difference, comparing the measured and calculated thermocouple temperatures, and comparing the calculated specimen temperature to the temperature range indicated by the melt wires.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

ECAR-2322 As-Run Thermal Analysis Of The AGC-2 Experiment

The second Advanced Graphite Creep (AGC-2) experiment was designed to irradiate various types of graphite specimens at a temperature of 600°C. The specimens were irradiated in an instrumented leadout capsule experiment in the south flux trap of the ATR during cycles 149A, 1498, 1508, 151A, and 1518. Temperature was monitored using twelve thermocouples located at various elevations in the reactor core, and a helium-argon gas mixture was used for gas gap temperature control of the specimens. The purpose of this analysis is to calculate specimen temperature using measured data on reactor power and helium-argon gas flows, and as-run calculations of heating rates and displacement per atom (DPA) in graphite. The accuracy of the model is assessed by comparing measured and calculated thermocouple temperatures. Uncertainty in gas gaps may preclude an accurate temperature calculation. In these cases, adjustments are made to the thermal model in order to reconcile the measured and calculated thermocouple temperature and to ensure the accuracy of the calculated specimen temperature.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

Summary of Post Irradiation Examination Results of the AFIP-6 Failure

The AFIP-6 test assembly was irradiated for one cycle in the Advanced Test Reactor at Idaho National Laboratory. The experiment was designed to test two monolithic fuel plates at power and burn-ups which bounded the operating conditions of both ATR and HFIR driver fuel. Both plates contain a solid U-Mo fuel foil with a zirconium diffusion barrier between 6061-aluminum cladding plates bonded by hot isostatic pressing. The experiment was designed with an orifice to restrict the coolant flow in order to obtain prototypic coolant temperature conditions. While these coolant temperatures were obtained, flow restriction resulted in low heat transfer coefficients and the failure of the fuel plates. The results from the post irradiation examinations and some observations of the failure mechanisms are outlined herein.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

TEMPERATURE DEPENDANT BEHAVIOUR OBSERVED IN THE AFIP-6 IRRADIATION TEST

The AFIP-6 test assembly was irradiated for one cycle in the Advanced Test Reactor at Idaho National Laboratory. The experiment was designed to test two monolithic fuel plates at power and burn-ups which bounded the operating conditions of both ATR and HFIR driver fuel. Both plates contained a solid U-Mo fuel foil with a zirconium diffusion barrier between 6061-aluminum cladding plates bonded by hot isostatic pressing. The experiment was designed with an orifice to restrict the coolant flow in order to obtain prototypic coolant temperature conditions. While these coolant temperatures were obtained, the reduced flow resulted in a sufficiently low heat transfer coefficient that failure of the fuel plates occurred. The increased fuel temperature led to significant variations in the fission gas retention behaviour of the U-Mo fuel. These variations in performance are outlined herein.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1983

Autothermal reforming of aliphatic and aromatic hydrocarbon liquids

Results are presented from a study of the autothermal reforming of paraffins and aromatics over nickel catalysts. The trials were performed to examine the carbon products that appear when steam is passed over hydrocarbon liquids to form H2-rich gases, i.e., the autothermal process (ATR). Attention was given to n-hexane, n-tetradecane, benzene, and benzene solutions of naphthalene with reactant preheat to 1000-1150 F. The carbon-formation limit was sought as a function of the steam-to-carbon and oxygen to carbon molar ratios at constant pressure and the preheat temperatures. The catalyst bed was examined after each trial to identify the locations and types of carbon formed using SEM, thermal gravimetric analysis, and X ray diffraction techniques. The hydrocarbon fuels each had a separate temperature and reaction profile, as well as carbon formation characteristics. No carbon formation was observed in the upper layer of the reactor bed, while both gas phase and surface-grown deposits were present in the lower part. The results are concluded of use in the study of No. 2 fuel oil for ATR feedstock.

Flytzani-Stephanopoulos, M.↗
DOE OSTI2023

Development of the generation III (ATR generation I) Plutonium-238 production target design

A domestic supply chain to produce 238 Pu fuel for radioisotope thermoelectric generators is critical for enabling future space exploration. A multi-laboratory effort has worked to establish a common target design to efficiently produce 238 Pu in two research reactors. This approach ensures that the annual production goals set forth by NASA are met, while also establishing redundant production capabilities. Here, this paper describes the effort to develop the common target design as well as considerations for future applications for the irradiation platform.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

A Variable Load LVDT-Based Creep Test Rig for Use in ATR Loop 2A

Understanding the creep behavior of materials in a radiation environment is essential in evaluating safety issues associated with nuclear power plant operation, especially for plants applying for life extension. In US materials testing reactors (MTRs), this understanding has been achieved through a “cook and look” approach (where specimens are irradiated under load for some period of time and then removed from an MTR for post-irradiation examination). A number of “cook and look” cycles can be required for each evaluated material, which makes this approach expensive. In addition, examinations are limited to end-state conditions, which may or may not be sufficient to reveal all phenomena of interest (because examinations are not performed at prototypic conditions relative to temperature, pressure, stress, etc.).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Preliminary Conceptual Neutronic, Thermal, and Mechanical Design for {sup 238}Pu ATR Targets

Pu-238 is a valuable heat source material used to manufacture radioisotope power systems (RPSs), which are needed to provide electrical power for space exploration and other missions. Before 1989, the United States Department of Energy (DOE) produced Pu-238 at the Savannah River Site in South Carolina in nuclear reactors that are now decommissioned. From 1989 through 2009, existing DOE inventories of Pu- 238 were supplemented by quantities of Pu-238 purchased from Russia; however, supplies available for purchase from Russia are limited. DOE has restored domestic production to meet future demand; otherwise, NASA will be in danger of not being able to use RPSs in future missions. The Pu-238 Supply Program has the objective of reestablishing the domestic capability for producing Pu-238. In order to meet the quantities of Pu-238 needed to support NASA space mission and other federal users the ATR is being considered as an important asset to achieve the desired production of the radioisotope. In order to determine the constraints and parameters needed to maximize the production of Pu-238 at ATR a preliminary conceptual design target design was performed. The preliminary conceptual design had two principal goals: 1. Perform a trade study to determine an ATR target length that maximizes production as well as minimize the need to modify ORNL's target. 2. Evaluate the selected preliminary design from the standpoint of the ATR nuclear safety basis. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗