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At least 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
NASA NTRS1991

Review and statistical analysis of the use of ultrasonic velocity for estimating the porosity fraction in polycrystalline materials

A review and statistical analysis of the ultrasonic velocity method for estimating the porosity fraction in polycrystalline materials is presented. Initially, a semiempirical model is developed showing the origin of the linear relationship between ultrasonic velocity and porosity fraction. Then, from a compilation of data produced by many researchers, scatter plots of velocity versus percent porosity data are shown for Al2O3, MgO, porcelain-based ceramics, PZT, SiC, Si3N4, steel, tungsten, UO2,(U0.30Pu0.70)C, and YBa2Cu3O(7-x). Linear regression analysis produces predicted slope, intercept, correlation coefficient, level of significance, and confidence interval statistics for the data. Velocity values predicted from regression analysis of fully-dense materials are in good agreement with those calculated from elastic properties.

Roth, D. J.↗
NASA NTRS1952

Calculation of Internal Pressures in the Fuel Tube of a Nuclear Reactor

General procedures for computing internal pressures in fuel tubes of nuclear reactors are described and the effects on the pressure of varying neutron flux, fissioning material, and operating temperatures are discussed. A general proof is given that during pile operation each fission product is monotonically increasing and therefore a maximum amount of all elements is present at the time of shit down. The post-shutdown build-up of elements that are held in check during pile operation because of their inordinately high capture cross sections is calculated quantitatively. An account of chemical interactions between the many fission-product elements and the resulting effect on the total pressure completes the discussion. The general methods are illustrated by calculations applied to a system consisting of 90 percent enriched U235 in the form of UO2 packed into a hollow metal cylinder or "pin", operating at a flux of 8 x 10(exp 14) at 2000 F. Calculations of the pressure inside a pin are made with and without a sodium metal heat-transfer additive. The bulk of the pressure is shown to depend on the four elements, xenon, krypton, rubidium, and cesium; the amount of free oxygen, however, was also significant. For a shutdown time of 10(exp 6) seconds, the pressure was about 100 atmospheres.

Rosenbaum, B. M.↗
NASA NTRS2003

High Power MPD Nuclear Electric Propulsion (NEP) for Artificial Gravity HOPE Missions to Callisto

This documents the results of a one-year multi-center NASA study on the prospect of sending humans to Jupiter's moon, Callisto, using an all Nuclear Electric Propulsion (NEP) space transportation system architecture with magnetoplasmadynamic (MPD) thrusters. The fission reactor system utilizes high temperature uranium dioxide (UO2) in tungsten (W) metal matrix cermet fuel and electricity is generated using advanced dynamic Brayton power conversion technology. The mission timeframe assumes on-going human Moon and Mars missions and existing space infrastructure to support launch of cargo and crewed spacecraft to Jupiter in 2041 and 2045, respectively.

McGuire, Melissa L.↗
NASA NTRS1993

Montmorillonite catalysis of RNA oligomer formation in aqueous solution. A model for the prebiotic formation of RNA

Oligomers of adenylic acid of up to the 11-mer in length are formed by the reaction of the phosphorimidazolide of adenosine (ImpA) in pH 8 aqueous solution at room temperature in the presence of Na(+)-montmorillonite. These oligomers are joined by phosphodiester bonds in which the 3',5'-linkage predominates over the 2',5'-linkage by a 2:1 ratio. Reaction of a 9:1 mixture of ImpA, A5'ppA results in the formation of oligomers with a 3:1 ratio of 3',5'- to 2',5'-linked phosphodiester bonds. A high proportion of these oligomers contain the A5'ppA grouping. A5'ppA reacts much more rapidly with ImpA than does 5'-ADP (ppA) or 5'-ATP (pppA). The exchangeable cation associated with the montmorillonite effects the observed catalysis with Li+, Na+, NH4+, and Ca2+ being the more effective while Mg2+ and Al3+ are almost ineffective catalysts. 2',5'-Linked oligomers, up to the tetramer in length, are formed using UO2(2+)-montmorillonite. The structure analysis of individual oligomer fractions was performed by selective enzymatic and KOH hydrolytic studies followed by HPLC analysis of the reaction products. It is concluded from the composition of the oligomers that the rate of addition ImpA to a 3'-terminus containing a 2',5'-linkage is slower than the addition to a nucleoside joined by a 3',5'-linked phosphodiester bond. The potential importance of mineral catalysis of the formation of RNA and other oligomers on primitive Earth is discussed.

Non-NASA Center↗
NASA NTRS2013

Nuclear Thermal Rocket Simulation in NPSS

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic- metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

Belair, Michael L.↗
NASA NTRS2013

Nuclear Thermal Rocket Simulation in NPSS

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic-metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

Belair, Michael L.↗
NASA NTRS2014

Raising Nuclear Thermal Propulsion (NTP) Technology Readiness Above 3

NTP development is currently supported by the NASA program office "Advanced Exploration Systems". The concept is a main propulsion option being considered for human missions to Mars in the 2030's. Major NTP development took place in the 1960's and 1970's under the Rover/NERVA program. The technology had matured to TRL 6 and was preparing to go to TRL 7 with a prototype flight engine before the program was cancelled. Over the last 40 years, a variety of continuations started, but only lasted a few years each. The Rover/NERVA infrastructure is almost all gone. The only remains are a few pieces of hardware, final reports and a few who worked the Rover/NERVA. Two types of nuclear fuel are being investigated to meet the current engine design specific impulse of 900 seconds compared to approximately 850 seconds demonstrated during Rover/NERVA. One is a continuation of composite fuel with new coatings to better control mid-band corrosion. The other type is a CERMET fuel made of Tungsten and UO2. Both fuels are being made from Rover/NERVA lessons learned, but with slightly different recipes to increase fuel endurance at higher operating temperatures. The technology readiness level (TRL) of these current modified reactor fuels is approximately TRL 3. To keep the development cost low and help mature the TRL level past 4 quickly, a few special non-nuclear test facilities have been made to test surrogate fuel, with depleted uranium, as coupons and full length elements. Both facilities utilize inductive heating and are licensed to handle depleted uranium. TRL 5 requires exposing the fuel to a nuclear environment and TRL 6 requires a prototype ground or flight engine system test. Currently, three different NTP ground test facility options are being investigated: exhaust scrubber, bore hole, and total exhaust containment. In parallel, a prototype flight demonstration test is also being studied. The first human mission to Mars in the 2030's is currently 2033. For an advanced propulsion concept to be seriously considered for use, the engine development plans need to show it is feasible and affordable to reach TRL 8 by 2027 and can be qualified for human mission use.

Gerrish, Harold P., Jr.↗
NASA NTRS2015

Affordable Development and Demonstration of a Small NTR Engine and Stage: A Preliminary NASA, DOE, and Industry Assessment

The Nuclear Thermal Rocket (NTR) represents the next evolutionary step in cryogenic liquid rocket engines. Deriving its energy from fission of uranium-235 atoms contained within fuel elements that comprise the engine's reactor core, the NTR can generate high thrust at a specific impulse of approx. 900 seconds or more - twice that of today's best chemical rockets. In FY'11, as part of the AISP project, NASA proposed a Nuclear Thermal Propulsion (NTP) effort that envisioned two key activities - "Foundational Technology Development" followed by system-level "Technology Demonstrations". Five near-term NTP activities identified for Foundational Technology Development became the basis for the NCPS project started in FY'12 and funded by NASA's AES program. During Phase 1 (FY'12-14), the NCPS project was focused on (1) Recapturing fuel processing techniques and fabricating partial length "heritage" fuel elements for the two candidate fuel forms identified by NASA and the DOE - NERVA graphite "composite" and the uranium dioxide (UO2) in tungsten "cermet". The Phase 1 effort also included: (2) Engine Conceptual Design; (3) Mission Analysis and Requirements Definition; (4) Identification of Affordable Options for Ground Testing; and (5) Formulation of an Affordable and Sustainable NTP Development Strategy. During FY'14, a preliminary plan for DDT&E was outlined by GRC, the DOE and industry for NASA HQ that involved significant system-level demonstration projects that included GTD tests at the NNSS, followed by a FTD mission. To reduce development costs, the GTD and FTD tests use a small, low thrust (approx. 7.5 or 16.5 klbf) engine. Both engines use graphite composite fuel and a "common" fuel element design that is scalable to higher thrust (approx. 25 klbf) engines by increasing the number of elements in a larger diameter core that can produce greater thermal power output. To keep the FTD mission cost down, a simple "1-burn" lunar flyby mission was considered along with maximizing the use of existing and flight proven liquid rocket and stage hardware (e.g., from the RL10-B2 engine and Delta Cryogenic Second Stage) to further ensure affordability. This paper provides a preliminary NASA, DOE and industry assessment of what is required - the key DDT&E activities, development options, and the associated schedule - to affordably build, ground test and fly a small NTR engine and stage within a 10-year timeframe.

Nuclear rocket engines↗
NASA NTRS2016

Fission Surface Power Technology Demonstration Unit Test Results

The Fission Surface Power (FSP) Technology Demonstration Unit (TDU) is a system-level demonstration of fission power technology intended for use on manned missions to Mars. The Baseline FSP systems consists of a 190 kWt UO2 fast-spectrum reactor cooled by a primary pumped liquid metal loop. This liquid metal loop transfers heat to two intermediate liquid metal loops designed to isolate fission products in the primary loop from the balance of plant. The intermediate liquid metal loops transfer heat to four Stirling Power Conversion Units (PCU), each of which produce 12 kWe (48 kW total) and reject waste heat to two pumped water loops, which transfer the waste heat to titanium-water heat pipe radiators. The FSP TDU simulates a single leg of the baseline FSP system using an electrically heater core simulator, a single liquid metal loop, a single PCU, and a pumped water loop which rejects the waste heat to a Facility Cooling System (FCS). When operated at the nominal operating conditions (modified for low liquid metal flow) during TDU testing the PCU produced 8.9 kW of power at an efficiency of 21.7 percent resulting in a net system power of 8.1 kW and a system level efficiency of 17.2 percent. The reduction in PCU power from levels seen during electrically heated testing is the result of insufficient heat transfer from the NaK heater head to the Stirling acceptor, which could not be tested at Sunpower prior to delivery to the NASA Glenn Research Center (GRC). The maximum PCU power of 10.4 kW was achieved at the maximum liquid metal temperature of 875 K, minimum water temperature of 350 K, 1.1 kg/s liquid metal flow, 0.39 kg/s water flow, and 15.0 mm amplitude at an efficiency of 23.3 percent. This resulted in a system net power of 9.7 kW and a system efficiency of 18.7 percent.

Facility Cooling System↗
NASA NTRS2016

Fission Surface Power Technology Demonstration Unit Test Results

The Fission Surface Power (FSP) Technology Demonstration Unit (TDU) is a system-level demonstration of fission power technology intended for use on manned missions to Mars. The Baseline FSP systems consists of a 190 kWt UO2 fast-spectrum reactor cooled by a primary pumped liquid metal loop. This liquid metal loop transfers heat to two intermediate liquid metal loops designed to isolate fission products in the primary loop from the balance of plant. The intermediate liquid metal loops transfer heat to four Stirling Power Conversion Units (PCU), each of which produce 12 kWe (48 kW total) and reject waste heat to two pumped water loops, which transfer the waste heat to titanium-water heat pipe radiators. The FSP TDU simulates a single leg of the baseline FSP system using an electrically heater core simulator, a single liquid metal loop, a single PCU, and a pumped water loop which rejects the waste heat to a Facility Cooling System (FCS). When operated at the nominal operating conditions (modified for low liquid metal flow) during TDU testing the PCU produced 8.9 kW of power at an efficiency of 21.7% resulting in a net system power of 8.1 kW and a system level efficiency of 17.2%. The reduction in PCU power from levels seen during electrically heated testing is the result of insufficient heat transfer from the NaK heater head to the Stirling acceptor, which could not be tested at Sunpower prior to delivery to GRC. The maximum PCU power of 10.4 kW was achieved at the maximum liquid metal temperature of 875 K, minimum water temperature of 350 K, 1.1 kg/s liquid metal flow, 0.39 kg/s water flow, and 15.0 mm amplitude at an efficiency of 23.3%. This resulted in a system net power of 9.7 kW and a system efficiency of 18.7 %.

fission power technology↗
DOE OSTI2019

Initial Application of TSUNAMI for Validation of Advanced Fuel Systems

A number of advanced reactor concepts and accident-tolerant fuels (ATFs) are under development or in the initial stages of deployment and testing in the nuclear power industry. Many of these concepts incorporate chemical or material forms that have not traditionally been used in the nuclear industry and may present challenges for nuclear criticality safety validation. Sensitivity/uncertainty methods are ideally suited for assessing the applicability of existing critical experiments to these potentially challenging applications. This paper presents two case studies recently completed at Oak Ridge National Laboratory (ORNL) which investigated the use of the TSUNAMI tools within the SCALE code system to perform such assessments.Many advanced reactor concepts include uranium enrichments above 5 wt% 235U but below the 20 wt% limit of low-enriched uranium (LEU). This material, referred to as high assay LEU (HA-LEU), poses difficulties across the entire infrastructure of nuclear power, including enrichment, transportation, fabrication, and storage. A potentially representative transportation package was selected to perform a criticality safety validation assessment for UF6 containing HA-LEU. Sensitivity data were generated using the TSUNAMI-3D sequence, and an assessment of the applicability of a range of critical experiments was performed using the TSUNAMI-IP tool. The critical experiments were drawn from the set of available sensitivity data files distributed with the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. The results indicate that a sufficient number of available experiments may be available to perform validations for these systems.Many different cladding materials and fuel dopants are included in ATFs to mitigate the consequences of a loss of coolant accident. Some of these systems have already been fabricated and introduced as lead test assemblies in commercial reactors, while others are still being developed. Several concepts were considered for pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) in comparison to the current Zircaloy and UO2 fuel system. In most of these cases, the ATF materials did not significantly impact the number of applicable experiments for validation. Some fuel and cladding materials were identified, however, which may reduce the number of available experiments and which will increase the nuclear data–induced uncertainty in neutron multiplication.

Marshall, William (B.J.)↗
DOE OSTI2019

Reduction-oxidation of actinides extraction process (ROANEX) for used nuclear fuel recycling

The invention relates to the ROANEX method, which extracts actinides from used nuclear fuel in a single purification cycle. The used nuclear fuel contains actinides, U, Am, Pu, Np. and Cm, and fission products, Cs, Sr and Tc. The fission products are separated first from the used nuclear fuel. The actinides are reduced to their lowest oxidation states and then oxidized to their highest oxidations states. Uranium, Pu and Np move to an organic phase solution and Am and Cm move to a nitrate solution. Uranium, Pu, and Np are stripped from the organic phase solution, and then treated with an oxalic acid to form a precipitate. Americium and Cm are treated with a potassium carbonate solution and Am precipitates. Actinides Am, U, Pu, and Np precipitates are heated in an oven and then blended together to form a mixed oxide fuel of UO2, PuO2, NpO2 and AmO2.

Riddle, Catherine Lynn↗
DOE OSTI2018

Summary Technical Report for Conceptual Design of MARCH-SERTTA

The Transient Reactor Test facility (TREAT) was constructed in the late 1950’s, provided thousands of transient irradiations before being placed in standby in 1994, and resumed reactor operations in 2017 in order to reclaim its crucial role in nuclear-heated safety research [1]. Starting in September of 2018, TREAT performed its first series of new fueled transient experiments using a novel modular irradiation vehicle, termed the Minimal Activation Retrievable Capsule Holder (MARCH) system. The MARCH system was developed to support high throughput testing and post transient examination by using fresh fuel samples arranged in low-activation hardware materials. Deployment of this irradiation vehicle system was supported by the Accident Tolerant Fuels (ATF) program. These initial experiments used inert gas atmosphere capsules (referred to as the Separate Effects Test Holder, SETH) containing small rodlets composed of UO2 pellets in zirconium alloy cladding with temperature measurement instruments. The initial SETH campaign was conducted in order to commission irradiation capabilities and deepen the understanding of specimen to core energy coupling. Building upon the successes of the initial SETH campaign, an evolution of the capsule design has been conceptualized in order to support the same kind of high throughput, cost effective, rodlet scale testing, but with the notable improvement of being able to immerse specimens in liquid water for testing conditions better representing Light Water Reactors (LWR). This capsule is referred to as the MARCH system Static Environment Rodlet Transient Test Apparatus (MARCH-SERTTA).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Post-irradiation Examinations of Accident Tolerant Fuel Candidates with Iron-alloy Cladding

A collaborative effort of the Advanced Fuels Campaign together with industry consortia is focused on the development of enhanced Accident Tolerant Fuels (ATF) for Light Water Reactors (LWRs). The scoping studies, referred to as ATF 1 irradiations, are being performed in the Advanced Test Reactor of Idaho National Laboratory using drop-in style irradiations. The Post Irradiation Examinations (PIEs) of ATF-1 began in 2017 and continued in 2019. This report describes the results of examinations on ATF-1 irradiations designed to test the performance of iron-based cladding. Two rodlets with commercial iron-based cladding and UO2 pellets have been irradiated to low burnup (i.e., < 20 GWd/tHM). The first rodlet had an Alloy33® cladding, while the other had a Kanthal APMT® cladding. The performed PIEs were focused on the determination of fuel microstructure, evaluation of fuel cladding interaction and irradiation-induced variations of cladding mechanical properties. For both rodlets, no significant changes in the cladding hardness were measured, and the cladding hoop strain remained limited. Gamma tomography revealed no Cs axial migration as expected for LWR fuel, but significant cesium radial migration, which was induced by thermal migration consequent to high irradiation temperatures compared to standard LWR irradiation temperatures. Microstructure analyses revealed formation of secondary phases along the pellet rim of samples from both rodlets. No wastage or extensive chemical interaction between the fuel and the internal wall of the cladding was observed, suggesting that the secondary phases accumulated at the pellet rim might be due to formation of Cs-U-O compounds. However, further advanced PIEs are necessary to confirm this hypothesis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

EFFICIENT SURROGATE FOR LOW ENERGY DAMAGE EFFECTS IN POLYATOMIC MATERIALS

We present an efficient surrogate for binary-collision Monte-Carlo (BCMC) calculations in polyatomic materials for low ion energies based on Parkin and Coulter’s net displacement function. Most of BCMC’s execution time is spent on the low energy tail of cascades when recoils’ residual travel distance is small compared to the desired resolution of the damage calculation. We propose to curtail an ions’ random walk if its energy reduces below a threshold Et and locally deposit an estimate of the expected number of defects it creates; the estimate of the number of defects is obtained from Parkin and Coulter’s net displacement functions. We find that results obtained with the BCMC code MyTRIM with and without truncation of the random walks compare well for large energy thresholds Et > 5 keV. For Et = 5 keV larger discrepancies are observed because of large differences between the net displacement function and MyTRIM for very low energy recoils. Future work will investigate and resolve this discrepancy. The energy threshold should be chosen so that ions’ average distance to next nuclear scattering is small compared to the desired resolution. A reduction of execution time of damage cascades by over 90 % is observed for an energy threshold of Et = 5 keV and a 1 MeV Xe-135 primary impinging on UO2. The desired application of this surrogate model are coupled FEM-BCMC calculations with varying compositions where net displacement functions are tabulated by composition. For reducing the number of required net displacement functions, derivatives of the net displacement function with respect to number fraction are computed.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2019

EFFICIENT SURROGATE FOR LOW ENERGY DAMAGE EFFECTS IN POLYATOMIC MATERIALS

We present an efficient surrogate for binary-collision Monte-Carlo (BCMC) calculations in polyatomic materials for low ion energies based on Parkin and Coulter’s net displace- ment function. Most of BCMC’s execution time is spent on the low energy tail of cascades when recoils’ residual travel distance is small compared to the desired resolution of the damage calculation. We propose to curtail an ions’ random walk if its energy reduces below a threshold Et and locally deposit an estimate of the expected number of defects it creates; the estimate of the number of defects is obtained from Parkin and Coulter’s net displacement functions. We find that results obtained with the BCMC code MyTRIM with and without truncation of the random walks compare well for large energy thresholds Et > 5 keV. For Et = 5 keV larger discrepancies are observed because of large differ- ences between the net displacement function and MyTRIM for very low energy recoils. Future work will investigate and resolve this discrepancy. The energy threshold should be chosen so that ions’ average distance to next nuclear scattering is small compared to the desired resolution. A reduction of execution time of damage cascades by over 90 % is observed for an energy threshold of Et = 5 keV and a 1 MeV Xe-135 primary impinging on UO2. The desired application of this surrogate model are coupled FEM-BCMC cal- culations with varying compositions where net displacement functions are tabulated by composition. For reducing the number of required net displacement functions, derivatives of the net displacement function with respect to number fraction are computed.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Dynamic System Scaling Applied to Zr-4 Cladding under RIA Conditions

The research and development lifecycle for new fuel designs currently has an ideal 20-25 year timeline. Efforts are currrently underway to accelerate fuel design and development using high performance fuel modeling codes and reduced scale experiments with new instrumentation techniques. However current tools available require a full scale prototype to be designed and tested for fuel qualification. The thermal-hydraulics community has leveraged scaling methodologies to use reduced scale integral effect test facilities (Westinghouse AP600/1000, NuScale Power Module) in lieu of building full scale prototypes for reactor design certification. Dynamic System Scaling (DSS) is a recent development to scaling methodologies that enable the tracking and quantification of dynamic scaling distortions experienced throughout a transient. Recent TREAT experiments performed reactivity initiated accident (RIA) pulses on UO2 fuel rodlets in Zr-4 cladding and were accompanied by BISON predictions for the cladding temperature. DSS is applied to pyrometer data of outer cladding temperatures and two BISON predictions after peak cladding temperature has been reached. One BISON case uses a constant gap conductance value between the fuel and cladding and the other using a conventional gap conductance model. Results from the DSS analysis show that there is good agreement between both BISON predictions and the experimental data and that there is not a significant distortion seen between using either gap conductance model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Advanced Microscopy Report on UCO Fuel Kernels from Selected AGR-1 and AGR-2 Experiments

A variety of neutron irradiation experiments were completed at Idaho National Laboratory (INL) on tristructural isotropic (TRISO) coated particles contained in a graphitic matrix as part of the Advanced Gas Reactor (AGR) fuel development and qualification program. Although the initial advanced microscopy and microanalysis studies on AGR-1 particles focused predominantly on the SiC layer?s role as the main fission-product containment, studies were further expanded to evaluate UCO kernels of selected AGR-1 and AGR-2 coated particles. Fuel kernels tested in the AGR-1 and AGR-2 experiments is composed of a mixture of UO2 and UC, with a nominal as-fabricated stoichiometry of UO1.37 C0.35 and UO1.43 C0.39, respectively. These as-fabricated kernel-stoichiometry values are determined at the completion of kernel fabrication and do not account for any changes that will occur during the subsequent coating and compacting processes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗