Space Nuclear Propulsion Fuel and Moderator Development Plan Conceptual Testing Reference Design
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Preliminary Evaluation of Driver Fuel Plates and Reduced Sodium Temperature to Increase Fast Test Reactor Experimental Neutron Irradiation Capabilities
All sodium-cooled fast test reactors operated so far have used essentially the same fuel geometrical ar-rangement, namely cylindrical fuel pins assembled into a triangular array in hexagonal fuel assemblies. Fuel compositions and dimensions may vary but the overall geometrical pattern remains identical. Fur-thermore, most sodium-cooled reactors have operated, and still operate, with core-average inlet and outlet sodium temperatures of, respectively, 350-400°C and 500-550°C. The present preliminary study revisits the standard fuel assembly geometry as well as operating tempera-tures to evaluate potential impact on fast test reactor neutron irradiation capabilities. The reference fuel is U-20Pu-10Zr. Pin (reference) and plate geometries are considered together with two sets of core-average sodium inlet/outlet temperatures: 350/500°C (reference) and 200/400°C. More prototypic environments could be obtained inside ad-hoc irradiation testing vehicles that are independent from the reactor primary coolant. Everything else being the same, fuel temperature is significantly lower in plates than in pins. Therefore, plates could in principle accommodate higher plutonium content, thus, reducing the need for uranium enrichment or allowing the use of lower quality plutonium (U-xPu-10Zr thermal conductivity and solidus temperature decrease as x increases, hence, necessitating additional thermal margins). For the reference pin configuration, lowering the inlet/outlet sodium temperatures from the reference 350/500°C down to 200/400°C provides additional thermal margins that can be used to increase the peak fast flux from about 4.5 × 10 15 n/cm 2 -s to 6 ×10 15 n/cm 2 -s for the same core power of 300 MW. Assuming 300 Equivalent Fuel Power Days (EFPD) of operation per calendar year, a (steel) test article could accu-mulate up to 75 dpa/year. The use of fuel plates provides even more thermal margins which may allow the peak fast flux to reach values as high as 8 × 10 15 n/cm 2 -s for both sets of inlet/outlet temperatures. In this environment, a test article could accumulate up to 95 dpa/year assuming 300 EFPD/year, hence greatly accelerating irradiation testing. Allowing a fast test reactor to operate over a wide range of inlet/outlet temperatures could add significant flexibility to its neutron irradiation capabilities. Steel present in the driver fuel assemblies is not expected to accumulate more than 90 displacements per atom (dpa) over its lifetime. Hence, steel alloys that are not suited for very high dpa but can operate over a wide range of temperatures (from low to high) would be appropriate for this application. Finally, because pin and plate assemblies fit on the same grid plate, a fast test reactor could in principle start-up with standard fuel pin assemblies and, later on, move to a plate-type Mark-II fuel to further in-crease its neutron irradiation capabilities while potentially reducing the level of uranium enrichment needed.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Smoking characteristics of various fuels as determined by open-cup and laboratory-burner smoke tests
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FUELS-OLEFINS↗
AGR-1, AGR-2 and AGR-3/4 Dimensional Change Data Analysis
A series of Advanced Gas Reactor (AGR) experiments have been completed in the Advanced Test Reactor at Idaho National Laboratory in support of qualification and development of tristructural isotropic fuel. Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder. These capsules are instrumented with thermocouples embedded in the graphite, enabling temperature control. The fuel compacts are composed of fuel particles surrounded by a graphitic A3 matrix material. Dimensional change in AGR fuel compacts is vital because the swelling or shrinkage affects the size of the gas gaps that are used to control temperatures. Analysis of dimensional change in the AGR fuel compacts is needed to establish the variables directly relating to compact shrinkage. The variables initially identified for consideration were matrix density, compact density, fuel packing fraction, uranium loading, fuel particle diameter, cumulative fast neutron fluence, and volume average time average fuel temperature. In addition to the data available from the AGR experiments, the analysis included specimens formed from the same A3 matrix material used in Advanced Graphite Creep (AGC) experiments, which provide graphite creep data during irradiation for design and licensing purposes. The primary purpose of including the AGC specimens was to encompass dimensional behavior at zero packing fraction, zero uranium loading, and zero particle diameter. All possible combinations of first-order variable regressions were considered in the analysis. The study focused on identifying the best regression models for percent change in diameter, length, and volume. Bootstrap analysis was used to ensure the resulting regression models were robust and well-performing. The variables identified as very significant in predicting change in one or more dimensions (diameter, length, and volume) are volume average time average temperature, fast fluence, compact density, packing fraction and fuel particle diameter. Due to the presence of confounding effects between several variables, interpretation of these results is equivocal; the use of multiple statistical tests provides additional confidence in the conclusion.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Final CRADA Report: Accelerated Burn-up Accumulation Test of Clean Core Thorium Energy Designated ANEEL Fuel
Clean Core Thorium Energy (CCTE), LLC, located in Oak Brook, Illinois, is committed to the development of alternative nuclear fuels. CCTE is focused on leveraging the inherent benefits of thorium to create a novel nuclear fuel solution for our worlds growing power demand. Unlike some other fuel cycles proposing to use thorium in advanced reactors, CCTE is focused on deploying solid fuel designs (ceramic pellets in metallic cladding) in existing pressurized heavy-water reactors (PHWRs) and Canada deuterium uranium (CANDU) reactors. CCTE’s fuel design, referred to as Advanced Nuclear Energy for Enriched Life (ANEEL), uses mixed thorium-uranium oxide ((Th,U)O2) to enhance reactor operational strategies while producing significantly low attractiveness material in terms of nuclear proliferation. The overall project objective is to investigate the performance of high burnup ANEEL fuel via an irradiation experiment to be performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). The scope of work for this CRADA was specifically aimed at performing conceptual design and analysis to support development of an irradiation test rig, development of a fuel fabrication process for the experiment pellets, fabrication and qualification of the pellets, and shipment of the pellets to INL so they can be used to construct experiment rodlets.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Cold flow testing of the Space Shuttle Main Engine alternate turbopump development high pressure fuel turbine model
NASA's Marshall Space Flight Center has established a cold airflow turbine test program to experimentally determine the performance of liquid rocket engine turbopump drive turbines. Testing of the SSME alternate turbopump development (ATD) fuel turbine was conducted for back-to-back comparisons with the baseline SSME fuel turbine results obtained in the first quarter of 1991. Turbine performance, Reynolds number effects, and turbine diagnostics, such as stage reactions and exit swirl angles, were investigated at the turbine design point and at off-design conditions. The test data showed that the ATD fuel turbine test article was approximately 1.4 percent higher in efficiency and flowed 5.3 percent more than the baseline fuel turbine test article. This paper describes the method and results used to validate the ATD fuel turbine aerodynamic design. The results are being used to determine the ATD high pressure fuel turbopump (HPFTP) turbine performance over its operating range, anchor the SSME ATD steady-state performance model, and validate various prediction and design analyses.
Gaddis, Stephen W.↗
Accident Tolerant Fuels Series 1 (ATF-1) Irradiation Testing FY 2018 Status Report
This report contains a summary of irradiation testing in the ATR of NTRD ATF 1 experiments performed at the INL in FY 2018. Additionally, four capsules met the sponsor-specified burnup targets, two of which were shipped from the ATR to the HFEF at INL.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
An Overview in the Development of a Multi-Sensor Data Science System for Monitoring a Solvent Extraction Process
To enhance nuclear nonproliferation stewardship, Idaho National Laboratory is building the Beartooth nuclear fuel cycle processing test bed. The Beartooth test bed will allow researchers the opportunity to study nuclear fuel processing operations including the use of centrifugal contactors in solvent extraction processes. The test bed is designed to support data collection and machine learning to monitor process operations. As part of this project, researchers will study data collected from a host of sensors that have not been typically used to monitor solvent extraction processes such as vibration, acoustic, colormetric, and thermal measurement data. The goal of this research is to employ machine learning and data analytics to study the confluence of signals collected from both traditionally and non-traditionally used sensors to provide operator process awareness and discover process anomalies. An overview of planned sensors and experiments that will focus on signal discovery will be presented.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
An Overview in the Development of a Multi-Sensor Data Science System for Monitoring a Solvent Extraction Process
To enhance nuclear nonproliferation stewardship, Idaho National Laboratory is building the Beartooth nuclear fuel cycle processing test bed. The Beartooth test bed will allow researchers the opportunity to study nuclear fuel processing operations including the use of centrifugal contactors in solvent extraction processes. The test bed is designed to support data collection and machine learning to monitor process operations. As part of this project, researchers will study data collected from a host of sensors that have not been typically used to monitor solvent extraction processes such as vibration, acoustic, colormetric, and thermal measurement data. The goal of this research is to employ machine learning and data analytics to study the confluence of signals collected from both traditionally and non-traditionally used sensors to provide operator process awareness and discover process anomalies. An overview of planned sensors and experiments that will focus on signal discovery will be presented.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
CFD-Chemical Model of One-Third Scale Demonstration of DOE Sealed Canister with ATR Fuel
Road-ready and final disposition packaging configurations for the advanced test reactor (ATR) fuel currently specifies storage within helium backfilled DOE sealed standard canisters. The aluminum cladding of the ATR fuel contains an oxyhydroxide layer of boehmite/bayerite that generates hydrogen when subjected to irradiation. Understanding the effect of this hydrogen buildup over time to important for long term storage considerations. Previous modeling efforts have built a coupled CFD-chemical model to simulate the temperature gas phase concentrations within the DOE sealed standard canisters. A demonstration case for these DOE sealed canisters will be eventually performed with a one-third scale mockup that has been instrumented with thermocouples and gas concentration probes. This study seeks to model the planned canister for validation of the previously developed model, such that confidence in its long-term prediction can be increased. The deployment of the instrumented lid for online monitoring in intended for a period of 10+ years based on previous monitoring of commercial fuel storage; however, the model is still run for the previously used 50-year storage periods. This case is modeled with a G-value using a bi-linear function such that it decreases at higher dose rates to be consistent with experiments. Validation efforts of the model would likely revolve around the results for the 1st year, so the results of this timeframe are also highlighted. For dried fuel of the nominal decay heat (18W), the predicted hydrogen concentration is 0.19% after 1 year, 0.91% after 10 years, and 2.8% after 50 years, with a maximum pressure of 1.26 atm. Consistent with previous modeling, the decay heat of the fuel is the main factor that influences the results. For undried fuel in pure helium, the hydrogen concentration ranges from 0.23 to 1.2% after 1 year, 1.25-6.2% after 10 years and 8.8 to 18.84% after 50 years. For dried fuel in pure helium the hydrogen concentration ranges from 0.06% to 0.34% after 1 year, 0.35-2.11% after 10 years and 0.95 to 6.6% after 50 years. While long-term results in the presence of residual air are mostly the same, early reactions with O 2 can delay significant production of H 2 until it is consumed to form more water vapor, lowering the range to 0.14 to 1.0% after 1 year. The maximum absolute pressure that is reached across any scenario is 2.06 atm. In the event of residual air, the presence of nitric acid is possible in the range of 18-131 ppm after 1 year, 174-1180 ppm after 10 years, and 586-3500 ppm after 50 years. As long as the fuel is sufficiently dried, or of nominal decay heat, a 4% lower flammability limit of hydrogen will not be reached within a 10-year monitoring period.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Evaluation of converters fueled with uranium nitride
Performance tests of two thermionic energy converters fueled with uranium nitride
Cassell, P. L.↗
Handbook for Performing Hydrocarbon Fuel Fires in XTF For Fast-Heat System Level Tests
This document serves to provide information on all aspects of a system level thermal qualification test including a description of the setup and conduct of a full system fuel fire test in the Thermal Test Complex (TTC) Crosswind Test Facility (XTF). This might be referred to as a “handbook” for future tests. This information is intended to assist technologists and test directors in performing these types of tests in the future.
42 ENGINEERING↗
A cermet fuel reactor for nuclear thermal propulsion
Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that the work done in the 1960's has demonstrated that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk.
Kruger, Gordon↗
Leaner Lifted-Flame Combustion Enabled by the Use of an Oxygenated Fuel in an Optical CI Engine
Leaner lifted-flame combustion (LLFC) is a mixing-controlled combustion strategy for compression-ignition engines that does not produce soot because the equivalence ratio at the lift-off length, Φ(H), is less than or equal to approximately two. In addition to preventing soot formation, LLFC can simultaneously control emissions of nitrogen oxides because it is tolerant to the use of exhaust-gas recirculation for lowering in-cylinder temperatures. LLFC can be achieved through the use of oxygenated fuels and enhanced fuel/charge-gas mixing upstream of the lift-off length. Enhanced mixing can be obtained via higher injection pressures, smaller injector orifice diameters, lower intake-manifold and coolant temperatures, and/or more retarded injection timings. This study focuses on the effects of an oxygenated fuel blend (T50) made up of 50% by volume tri-propylene glycol mono-methyl ether with a #2 ULSD emissions-certification fuel (CFA), compared against baseline testing of the CFA fuel without the oxygenate. Experimental measurements include crank-angle resolved natural luminosity (NL) and chemiluminescence (CL) imaging diagnostics. EGR is simulated by adding nitrogen and carbon dioxide to the intake charge to produce a 16% intake-oxygen mole fraction (XO2), and results are compared against cases with no charge dilution (i.e., 21% XO2). Initial experiments with a two-hole tip achieved soot-free LLFC at low loads with the T50 fuel, 240 MPa injection pressure, 50 °C intake-manifold temperature (IMT), 95 °C coolant temperature, and -5 crank-angle degree (CAD) after top-dead-center (ATDC) start of combustion (SOC) timing. The strategy was extended to more moderate loads by employing a 6-hole injector tip, where lowering the IMT to 30 °C, reducing the coolant temperature to 85 °C, and retarding the SOC timing to +5 CAD ATDC resulted in sustained LLFC at both 21% and 16% XO2 at up to 6.2 bar gross indicated mean effective pressure (gIMEP) with T50. The achievement of LLFC was confirmed by independent soot measurements using a smoke meter, spatially integrated natural luminosity from the NL diagnostics, and laser-induced incandescence for measuring soot volume fraction in the exhaust stream. In contrast to the results with T50, LLFC was not achieved under any of the test conditions with CFA. Combustion was stable under LLFC conditions, with a coefficient of variation of gIMEP less than 1.5%. Nitrogen oxides and hydrocarbon emissions were also lowered by up to 25% each for LLFC with T50 relative to using the same operating conditions with CFA. Combustion noise was similarly reduced by 4-6 dBA, and ringing intensity by 60-80%, for LLFC with T50.
33 ADVANCED PROPULSION SYSTEMS↗
Auto-Thermal Reforming of Jet-A Fuel over Commercial Monolith Catalysts: MicroReactor Evaluation and Screening Test Results
This paper describes the results of a series of catalyst screening tests conducted with Jet-A fuel under auto-thermal reforming (ATR) process conditions at the research laboratories of SOFCo-EFS Holdings LLC under Glenn Research Center Contract. The primary objective is to identify best available catalysts for future testing at the NASA GRC 10-kW(sub e) reformer test facility. The new GRC reformer-injector test rig construction is due to complete by March 2004. Six commercially available monolithic catalyst materials were initially selected by the NASA/SOFCo team for evaluation and bench scale screening in an existing 0.05 kW(sub e) microreactor test apparatus. The catalyst screening tests performed lasted 70 to 100 hours in duration in order to allow comparison between the different samples over a defined range of ATR process conditions. Aging tests were subsequently performed with the top two ranked catalysts as a more representative evaluation of performance in a commercial aerospace application. The two catalyst aging tests conducted lasting for approximately 600 hours and 1000 hours, respectively.
Yen, Judy C. H.↗
Steps in the Development of Nuclear Thermal Propulsion Fuels
Due to the recent increased interest in Mars and deep-space missions it has been determined to be feasible to revisit the ability for the United States to build a Nuclear Thermal Propulsion (NTP) system. The need to make and qualify a NTP fuel that meets mission performance requirements is an essential first step. Graphite fuels and cermet fuels are of particular interest since these fuels have shown significant advantages over other fuel types. This paper will address the history of NTP fuel fabrication technology as related to the NERVA, GE 710, and Argonne National Laboratory’s Nuclear Fuel Programs as well as recent efforts in recapturing heritage fuels and develop new NTP fuels. Substantial experimental databases and supporting documentation exists for the composite fuel. Some irradiation and high temperature test data is available for cermet fuels, but cermet fuels were never tested at prototypic NTP conditions. A first step in the development effort will be a fuel fabrication recapture effort to provide samples to show that the technology works and performance of the fuel is acceptable. Advances in fuel fabrication, materials processing, and coating technology are expected to improve and/or enhance future fuel development, maturation and certification efforts. The current plan is to perform, non-nuclear separate-effects and integrated tests to provide additional test data and insight into the capabilities of each fuel and to use that data to evaluate potential performance within a NTP engine stage.
Qualls, A↗
Accident Tolerant Fuels Series 1 (ATF-1) Irradiation Testing FY 2019 Status Report
This report contains a summary of irradiation testing in the ATR of NTRD ATF 1 experiments performed at the INL in FY 2019. Additionally, three capsules met the sponsor-specified burnup targets, and five capsules were shipped from the ATR to the HFEF at INL.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Expansion of the Fast Neutron Hodoscope at TREAT to Support Fuel Safety Experiments
In 2024 the Fuel Motion Monitoring System (FMMS), or Hodoscope, at the Transient Reactor Test Facility, is being expanded from 96 viewing channels to 192 viewing channels, effectively doubling the FMMS field of view. Here, this increase in capability will allow the FMMS to support larger scale fuel tests, encompassing height-of-core test devices, test devices with multi-pin fuel assemblies, and test devices with recirculating coolant flow. Work supporting the expansion included refurbishing 96 additional proton recoil scintillator (PRS) detectors, doubling the data acquisition system (DAS) installed architecture, improving time synchronization in the DAS, and new research to measure the PRS detector energy-dependent, fast-neutron detection efficiency. In addition, laboratory activities have produced an improved benchtop testing capability for assessing the DAS, time synchronization, and external start triggering, along with an updated capability to scan PRS detectors to develop a preliminary flat-field normalization prior to deployment to TREAT.
46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗