International Multi-Modal Spent Nuclear Fuel Transportation Test: The Transportation Test Triathlon.
Abstract not provided.
Saltzstein, Sylvia J.↗
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International Multi-Modal Surrogate Spent Nuclear Fuel Transportation Test: The SNF Transportation Test Triathlon.
Abstract not provided.
Saltzstein, Sylvia J.↗
Functionality Isolation Test for Fuel Cycle Code ORION - MOX Fuel Fabrication
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Bae, Jin Whan↗
A COMPARISON OF NEUTRON COINCIDENCE RATES DETECTED FROM SIMULATED MATERIAL TEST REACTOR FUEL ASSEMBLIES
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Long, Grace↗
Experimental clean combustor program: Diesel no. 2 fuel addendum, phase 3
A CF6-50 engine equipped with an advanced, low emission, double annular combustor was operated 4.8 hours with No. 2 diesel fuel. Fourteen steady-state operating conditions ranging from idle to full power were investigated. Engine/combustor performance and exhaust emissions were obtained and compared to JF-5 fueled test results. With one exception, fuel effects were very small and in agreement with previously obtained combustor test rig results. At high power operating condition, the two fuels produced virtually the same peak metal temperatures and exhaust emission levels. At low power operating conditions, where only the pilot stage was fueled, smoke levels tended to be significantly higher with No. 2 diesel fuel. Additional development of this combustor concept is needed in the areas of exit temperature distribution, engine fuel control, and exhaust emission levels before it can be considered for production engine use.
Gleason, C. C.↗
Pre-Test Feasibility Study Of Reactivity Transient Testing On TRISO Fuel In The TREAT Reactor
Uranium oxycarbide (UCO)-bearing tri-structural isotropic (TRISO) particle fuels are expected to be used in numerous US commercial reactor applications within the next decade. In this work, we reviewed historical particle fuel transient experiments to identify gaps in the TRISO fuel performance transient testing. A BISON--Griffin modeling framework was then developed to conduct preliminary TRISO transient analyses and begin to address these gaps. The framework was demonstrated using limiting-case transient conditions from a prototypic high-temperature gas-cooled reactor. It was then applied to develop a matrix of experiments that could be performed in the Transient Reactor Test Facility (TREAT) to (1) evaluate UCO-fueled particle performance at moderate and high heat rates, (2) assess whether historical testing involving UO$_2$-fueled particles is applicable to modern UCO-fueled particles, (3) deconvolute the impacts of temperature and heat rate on particle transient response, and (4) collect the data needed for fuel performance model validation and/or further development.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Status Report on Advanced Reactor Fuels Hot Cell Furnace Testing
The Advanced Fuels Campaign is developing a series of out-of-pile steady state furnace tests on irradiated nuclear fuel pins in order to better characterize and understand isolated fuel performance phenomena. These tests will also provide valuable datasets for validation and improvement of modelling and simulation tools. The Blister Anneal Furnace (BAF) in the INL Hot Fuel Examination Facility (HFEF) has been identified as the ideal location for these tests, and, in order to perform them, a new fixture for the BAF has been designed and commissioned. This report summarizes the objectives, design, characteristics and status of this fixture. The design was finalized, materials procured and tested, and the assembly fixture is now fully fabricated in the mock-up shop of the Fuel Conditioning Facility (FCF) at MFC.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Development of carbon slurry fuels for transportation (hybrid fuels, phase 2)
Slurry fuels of various forms of solids in diesel fuel are developed and evaluated for their relative potential as fuel for diesel engines. Thirteen test fuels with different solids concentrations are formulated using eight different materials. A variety of properties are examined including ash content, sulfur content, particle size distribution, and rheological properties. Attempts are made to determine the effects of these variations on these fuel properties on injection, atomization, and combustion processes. The slurries are also tested in a single cylinder CLR engine in both direct injection and prechamber configurations. The data includes the normal performance parameters as well as heat release rates and emissions. The slurries perform very much like the baseline fuel. The combustion data indicate that a large fraction (90 percent or more) of the solids are burning in the engine. It appears that the prechamber engine configuration is more tolerant of the slurries than the direct injection configuration.
Ryan, T. W., III↗
BWR Spent Nuclear Fuel Acquisition and Testing to Support DOE-NE High Burnup Spent Fuel Data Project
The Office of Spent Fuel and Waste Disposition (SFWD) within the US Department of Energy (DOE) Office of Nuclear Energy (NE) established the Spent Fuel and Waste Science and Technology (SFWST) campaign to conduct research and development (R&D) activities related to the storage, transportation, and disposal of spent nuclear fuel (SNF) and high-level radioactive waste. The SFWST program was created within SFWD to address issues of extended or long-term SNF storage and transportation. Some near-term objectives of SFWST are to use a science-based, engineering-driven approach to: Support the enhancement of the technical bases to support the continued safe and secure dry storage of SNF for extended periods; Support the enhancement of the technical bases for retrieving SNF after extended dry storage; Support the enhancement of the technical bases for transporting high burnup (HBU) fuel and transporting low burnup fuel and HBU fuel after dry storage DOE-NE, in partnership with the Electric Power Research Institute, developed the High Burnup Spent Fuel Data Project to perform a large-scale demonstration and laboratory-scale testing of HBU pressurized water reactor (PWR) fuels (exceeding 45 gigawatt-days per metric ton of uranium [GWd/MTU]). Under this project, 25 sister rods—which are rods that have the same design, power histories, and other characteristics—were removed from assemblies at the North Anna Nuclear Power Station and sent to Oak Ridge National Laboratory (ORNL) in January 2016. ORNL performed detailed nondestructive examination (NDE) on all 25 rods. The NDE consisted of visual examinations, gamma and neutron scanning, profilometry and rod length measurements, and eddy current examinations. After completing the NDE, 10 of the sister rods were delivered to Pacific Northwest National Laboratory (PNNL) in a NAC International, Inc. legal-weight truck cask in September 2018 for destructive examination (DE). To date, SFWD work has focused on the PWR fuel that is part of the Sister Rod Test program. No boiling water reactor (BWR) fuel has been tested in the program, and the data needs that were identified for the PWR fuel have not been collected for BWR fuel. The goal to obtain six to nine BWR rods and test them at ORNL will support closing this important data gap. BWR fuel comprises approximately 56% of the total fuel assemblies currently in storage at nuclear power plants in the United States. BWR nuclear fuel and cladding designs and manufacturing are significantly different from PWRs. Differences include the following: BWR fuel pellets are larger than PWR pellets; Variations of Zircaloy-2 (including liners) are used instead of the Zircaloy-4 cladding materials used in PWRs; Clad manufacturing and stress-relief processes are different between PWRs and BWRs; The fuel rod dimensions are different because larger rod diameters and thicker cladding are used in BWRs; BWR fuel typically has lower internal rod pressures and sees vastly different operating conditions than PWR fuel (i.e., two-phase flow); BWR assemblies are “canned,” meaning each assembly is surrounded by a metal fuel channel; BWR cladding is often composed of an inner pure Zr liner that has widely different mechanical properties than the Zircaloy-2 alloy and exhibits a stronger affinity for hydrogen; The BWR SNF generally has more total hydrogen in the cladding/liner than typical PWR fuel; The construction of the PWR and BWR assemblies is vastly different; BWR rods are solidly attached to the assembly nozzles and experience a much different vibration and shock load than PWR rods, which are “floating” within a grid system attached to guide tubes, and the rods sit loosely on the bottom end plates. These numerous differences will affect the way the BWR SNF responds under dry storage preparation processes (e.g., vacuum drying) and during transportation. The results collected in the PWR experimental program must be compared with a subset of similar data collected on BWR SNF to establish a technical basis for whether the larger PWR database is sufficient to bound the BWR SNF end-of-life conditions as is currently assumed for several fuel/clad properties. Changes that occur in both fuel types at HBU could exacerbate any mechanical property differences. As the fuel burnup increases, several changes occur that might affect the performance of the fuel, cladding, and assembly hardware in storage and transportation. These changes include increased cladding corrosion layer thickness, increased cladding hydrogen content, increased cladding creep strains, increased fission gas release, and the formation of the HBU structure at the surface of the fuel pellets. The Nuclear Regulatory Commission (NRC) limits the current maximum rod-averaged burnup to 62 GWd/MTU due to these changes and the lack of data at higher burnups.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
High Fidelity Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Results
Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, the integration of thermal hydraulic hardware tests with simulated neutronie response provides a bridge between electrically heated testing and fueled nuclear testing, providing a better assessment of system integration issues, characterization of integrated system response times and response characteristics, and assessment of potential design improvements' at a relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design can developed. Test results presented in this paper correspond to a "first cut" simulator design for a potential liquid metal (NaK) cooled reactor design that could be applied for Lunar surface power. Proposed refinements to this simulator design are also presented.
Bragg-Sitton, Shannon M.↗
Visualization of the High-Burnup Spent Fuel Rod Phase 1 Test Plan: Technical Memo - Spent Fuel and Waste Disposition (Technical Memo)
Twenty-five high-burnup fuel rods were extracted from seven different fuel assemblies used for power production at the North Anna nuclear power plant and shipped to Oak Ridge National Laboratory (ORNL) in 2016 for detailed non-destructive examination (NDE) and destructive examination (DE). The spent fuel rods were from 17×17 lattices and consist of four cladding types—Zirlo ® , M5 ® , Zircaloy-4, and low tin Zircaloy-4 (Zirc-4). These spent fuel rods are being tested to provide: (a) baseline characterization and mechanical property data that can be used as a comparison to fuel that was loaded into a modified TN-32B cask in November 2017, as part of the high-burnup confirmatory data project and (b) data applicable to high-burnup fuel rods (>45 GWd/MTU) currently stored and to be stored in the dry-cask fleet. The TN-32B cask is referred to as the “Demo” cask and is currently expected to be transported to a separate location and the internal contents inspected in approximately ten years. ORNL has completed the NDE of the twenty-five fuel rods. The purpose of this technical memorandum is to present a simplified summary of the first phase of destructive examinations and test conditions that will be used for communicating with various stakeholders. The destructive examinations will leverage the expertise and capabilities from multiple national laboratories for performing independent measurements of relevant data. Close coordination is required to ensure that all examinations follow well documented procedures and are performed so that measured data and characteristics can be readily compared. Pacific Northwest National Laboratory (PNNL) has published a detailed overview of the test program. ORNL and PNNL developed detailed draft test plans for testing to be performed at their facilities. ORNL and PNNL are in the process of refining these test plans to apply specifically to the testing described in this memorandum. Argonne National Laboratory (ANL) contributed to the ORNL test plan by describing tests to be conducted at ANL. Testing will be based on continuous learning. If a test produces results that are inconsistent with expectations or current trends, further testing will be paused until a path forward is established to understand the results and to identify follow-on testing.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Testing Accident Tolerant Fuels in Reactivity Initiated Accident Conditions
The Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) has been successfully restarted. A principal goal of restarting the reactor was to perform transient testing of Accident Tolerant Fuels (ATF), supporting their safe adoption by the commercial Light Water Reactor (LWR) fleet. This paper will describe the efforts underway at INL to conduct testing of ATF under Reactivity Initiated Accident (RIA) conditions in an inert helium atmosphere. Tests in an inert helium atmosphere capsule provide valuable information about initial deposition and distribution of energy in the nuclear fuel. Tests in an inert helium atmosphere are especially suited for ATF concepts where the thermophysical properties of the new fuel material differ from UO2 such as the U3Si2 fuel concept and other non-oxide materials. These tests are also useful for exploring the thermal mechanical interaction of pellets with cladding where the cladding mechanical properties do not change much with temperature such as Silicon Carbide Cladding. Tests with nominal dimensions will be used to determine how fresh fuel failure and core coolability limits of ATF pellet and cladding materials compare with that of standard LWR fuel made up of UO2 pellets and Zircaloy cladding. Tests with a reduced pellet cladding gap will be used as a preliminary indicator of any pellet cladding mechanical interaction (PCMI) vulnerabilities of the new fuel concepts. The materials, methods, and schedule for the ATF RIA tests will be discussed in the paper.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Use of fluidic oscillator to measure fuel-air ratios of combustion gases
A fluidic oscillator was investigated for use in measuring fuel-air ratios in hydrocarbon combustion processes. The oscillator was operated with dry exhaust gas from an experimental combustor burning ASTM A-1 fuel. Tests were conducted with fuel-air ratios between 0.015 and 0.031. Fuel-air ratios determined by oscillator frequency were within 0.001 of the values computed from separate flow measurements of the air and fuel.
Riddlebaugh, S. M.↗
A numerical study on the sooting tendencies of bio-derived fuels for spark-ignition engines
While more efficient vehicles have been developed over the decades, their performance is limited by the properties of existing fuels. In response, the Co-Optimization of Fuels and Engines Initiative (Co-Optima) under the U.S. Department of Energy (DOE) has developed a rigorous screening approach to identify the most promising biomass-derived blendstocks that are suitable for advanced spark-ignition (SI) engines. A detailed kinetic model has also been developed to predict the combustion properties of the selected blendstocks. This kinetic model has been designed mainly targeting ignition and flame propagation properties, and it has not previously been validated for soot formation. In this work, we numerically predicted the sooting tendency of 20 Co-Optima SI blendstocks using the Co-Optima kinetic model. The sooting tendencies are determined quantitatively using the Yield Sooting Index (YSI) methodology. As shown in Fig. 1, the predicted YSIs show good agreement with measurements, except for 2,5-dimethylfuran. We also quantify the sensitivity of the YSI predictions to aromatic growth reactions and fuel decomposition reactions of the test fuels. It is found that PAH growth reactions have little impact on YSIs for all 20 fuels under investigation. On the other hand, fuel decomposition reactions have a significant influence on the YSI of 2,5-dimethylfuran. Perturbing their reaction rates by a factor of two is shown to lead to 11 % YSI prediction uncertainty.
09 BIOMASS FUELS↗
AGR-2 Compact 6-4-1 Post-Irradiation Examination Results
The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform research and development on tristructural isotropic (TRISO)-coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel Program. The overarching goal of the program is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety/heating testing, fuel performance modeling, and fission product transport (INL 2017). Several fuel irradiation experiments have been performed at the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), and additional irradiations are planned. These experiments are intended to provide data on fuel performance under irradiation, support fuel fabrication process development, qualify fuel for operating and accident conditions, provide irradiated fuel for accident testing, and support development of fuel performance and fission product transport models.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Flame radiation
Total radiation and heat flux data was obtained on a combustor liner by advanced instrumentation. If the results obtained by the special instrumentation are considered to be representative of the total radiation and heat flux, then the effect of variation of engine operating parameters and of fuel type can be more easily obtained. The special instrumentation used for these investigations consisted of five total radiometers and two total heat flux gages. The radiometers were arranged axially and circumferentially through sliding air seals in the outer liner. The two heat flux gages were welded in the outer liner between two circumferential radiometers. Static pressures were obtained on both the cold and the hot side of the outer liner in the area of the heat flux gages. Liner metal temperatures were also obtained. The combustor inlet pressure was varied over a nominal range of 0.5 to 2.07 MPa, inlet air temperature from 550 to 670 K, and fuel-air ratio from about 0.015 to 0.040. The two fuels tested were ASTM Jet A and a fuel designated as ERBS V. Results of the tests are discussed.
Wear, J. D.↗