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Modeling of Cr 2 O 3 -doped UO 2 as a near-term accident tolerant fuel for LWRs using the BISON code
The near-term Accident Tolerant Fuel (ATF) concepts consist of minor modifications of the standard UO 2 -Zr fuel rod materials to provide enhanced accident tolerance. Uranium dioxide fuel with chromia dopant, developed by AREVA, is of interest in view of its enhanced fission gas retention. The predictive capability of the BISON fuel performance code for chromia-doped UO 2 fuel is assessed through simulation of the Halden IFA-677.1 fuel rod experiment. A sensitivity analysis is conducted to interpret the deviation of code’s predictions from the experimental data. A power ramp test is also modeled with BISON and calculations are compared to the experimental database. Finally, a large-break LOCA (LBLOCA) case is simulated with BISON, leading to an initial assessment of the enhanced safety associated with chromia-doped fuel compared to the standard UO 2 fuel in LWRs
Sensitivity of UO 2 fuel performance to microstructural evolutions driven by dilute additives
Use of dilute additives to nuclear fuel is being considered to increase the security of commercial fuel management through traceability of fabricated fuel elements. Taggants, as additives are denoted when included for traceability purposes, may also improve fuel performance, as demonstrated in Cr-containing uranium dioxide as described in the literature, and they may also improve fuel safety. In fact, studies have shown that some additives affect fuel material properties such as grain size and density after sintering. Given the possible range of elements that could be used as additives, the impact of such fuel property variations on the fuel’s thermomechanical behavior becomes relevant. These effects can be evaluated through a sensitivity study of standard fuel models to analyze changes in these properties using a fuel performance code. In this work, the BISON code is being used to investigate these effects through a 2D axisymmetric model of smeared UO 2 fuel pellets and ZIRLO® cladding under realistic pressurized water reactor core irradiation conditions. Here, randomly sampled densities and grain sizes within specified ranges are used as input parameters in the simulations, and several fuel model-related outputs are evaluated. The thermomechanical response of the cladding is also addressed in this study. The simultaneous variation of both input parameters offers a more comprehensive path to identify key sensitivities. Outputs explored include temperature, fission gas release, creep, and radial stress. Results show that although most of these outputs are sensitive to grain size to a certain extent, density mainly affects fuel temperature and elastic strain. Furthermore, sensitivities can vary depending on the radial position within the fuel pellet.
Compare Mechanistic Predictions for Doped UO 2 Mechanical Response and Other Properties with Empirical Models and Experimental Measurements
The U.S. Department of Energy’s Nuclear Energy Advanced Modeling and Simulation program develops predictive capabilities using computational methods for the analysis and design of advanced reactor and fuel cycle systems. This program has been supporting the development of BISON, a high-fidelity, high resolution fuel performance tool at the engineering scale. As part of its development, additional modeling capabilities and improvements have been developed for relevant fuel forms. In this work, a fuel creep deformation model for Cr-doped fuel has been implemented into BISON, along with improvements to the empirical UO 2 fuel creep model based on experimental data and improvements to the radial power factor calculation for doped fuels. This work allows for more accurate simulation analyses for both UO 2 and doped-UO 2 fuels.
In Situ Study of Particle Precipitation in Metal-Doped CeO2 during Thermal Treatment and Ion Irradiation for Emulation of Irradiating Fuels
Metallic particles formed in oxide fuels (e.g., UO 2 ) during neutron irradiation have an adverse impact on fuel performance. A fundamental investigation of particle precipitation is needed to predict the fuel performance and potentially improve fuel designs and operations. In this study, we report on the precipitation of Mo-dominant β-phase particles in polycrystalline CeO 2 (surrogate for UO 2 ) films doped with Mo, Pd, Rh, Ru, and Re (surrogate for Tc). In situ heating scanning transmission electron microscopy indicates that particle precipitation starts at ~1073 K with a limited particle growth to ~10 nm. While particle concentration increases with increasing temperature, particle size remains largely unchanged up to 1273 K. There is a dramatic change in the microstructure following vacuum annealing at 1373 K, probably due to phase transition of reduced cerium oxide. At the high temperature, particles grow up to 75 nm or larger with distinctive facets. The particles are predominantly composed of Mo with a body-centered cubic structure (β phase). An oxide layer was observed after storage at ambient conditions. In situ heating X-ray photoelectron spectroscopy reveals an increasing reduction of Ce charge state from 4+ to 3+ in the doped CeO 2 film at temperatures from 673 to 1273 K. In situ ion irradiation transmission electron microscopy with 2 MeV Al 2+ ions up to a dose of ~20 displacements per atom at nominally room temperature does not lead to precipitation of visible particles. However, irradiation with 1.7 MeV Au 3+ ions to ~10 dpa at 973 K produces ~2 nm sized pure Pd particles; Au 3+ irradiation at 1173 K appears to result in precipitates of ~6 nm in size. Some of the defects produced by ion irradiation could be nucleation sites for precipitation, leading to generation of smaller particles with a higher concentration.
Work updates for Prof. Lee
I want to update my work which I have done so far in INL, including matlab GUI for neutron radiography and some fuel EBSD works.
AGR-2 TRISO Fuel Post-Irradiation Examination Final Report
An array of post-irradiation exams and experiments were employed to assess the irradiation performance and high-temperature behavior of AGR-2 UCO TRISO fuel produced in an engineering-scale coater. Work was also conducted on UO2 TRISO fuel produced for comparison with the AGR UCO fuel. Emphases we replaced on enumerating the frequency of SiC layer failures (failure of the SiC layer with at least one PyC layer remaining intact) and TRISO failures (failure of all three dense layers) and evaluating the retention of key fission products within the fuel. Examinations of the fuel kernel and coating morphologies were used to understand the features typical of normal intact fuel, degraded or failed fuel, and mechanisms of degradation. Radiochemical analyses were used to determine their retention within the fuel. The results confirm excellent performance of the fuel that exceeds historical HTGR design requirements with significant margin.
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.
Molybdenum and Tungsten isotope compositions of UO 2 fuel pellets: Implications for isotopically enriched taggants
The addition of isotopically enriched taggants to material at the front end of the nuclear fuel cycle could be a powerful tool used to assist law enforcement authorities should material outside of regulatory control be found. Two potential candidates for this purpose are molybdenum (Mo) and tungsten (W) as both elements have five or more stable isotopes and are trace elements contained within nuclear fuel. So there is a concern that Mo and W could undergo isotope fractionation during processes like uranium enrichment and to date, it is unknown if nuclear fuels have natural Mo and W isotope compositions. If Mo and W isotopic variability is present in nuclear fuels, this would hinder the use of these elements as isotopic taggants because it would be difficult to discern the original taggant isotope composition with high confidence. Therefore, a set of 16 low enriched uranium (LEU) fuel pellets from US commercial producers was analyzed using multi collector-inductively coupled plasma mass spectrometry (MC-ICPMS) to determine Mo and W isotope compositions (i.e., 94 Mo/ 92 Mo, 95 Mo/ 92 Mo, 96 Mo/ 92 Mo, 97 Mo/ 92 Mo, 98 Mo/ 92 Mo, 183 W/ 182 W, 184 W/ 182 W, and 186 W/ 182 W). Relative to terrestrial standards, LEU fuel pellets have variable Mo and W isotope compositions, thereby complicating the use of these elements as isotopically enriched taggants. As such, this work demonstrates that the isotope composition of any potential taggant must be well characterized in the base nuclear fuel prior to any taggant addition. Furthermore, these results suggest that Mo and W are not ideal candidates for isotopically enriched taggants.
Comparative chemical and structural analyses of two uranium dioxide fuel pellets
Here, a comparative investigation of the trace element, U and Pb isotopic compositions, and structural attributes of two uranium dioxide fuel pellets was conducted for nuclear forensic applications. One fuel pellet consists of natural UO 2 (0.71% 235 U), while the other contains enriched UO 2 (3.98% 235 U). Numerous impurities, including transition metals, lanthanides, and main group elements were detected in both fuel pellets with total trace element contents that vary between 9.77 and 11.37 μg/g. The relative abundances of the transition metal impurities in the natural UO 2 fuel pellet are linked to the source U ore and provide important insights into its provenance. X-ray fluorescence spectroscopy was utilized to investigate and compare the macro-(cm)-scale distribution of impurities within each of the fuel pellets. In addition, structural properties of the fuel pellets were examined using powder X-ray diffraction (PXRD) and Raman and infrared spectroscopy. Differences in the intensity of the T2 g U-O stretching vibrational mode were observed in the Raman spectra of the fuel pellets. Lastly, Pb isotope ratios were investigated to potentially differentiate between the natural and enriched UO 2 samples and are linked to the process history of these materials.
The influence of iron incorporation on the microstructure and crystal chemistry of UO 2 fuel
The incorporation of dopants into UO 2 is of interest for both doped advanced fuel and nuclear forensics research. Suitable forensics dopants, or taggants, for nuclear fuels are ones that do not adversely affect the performance of the parent material while still being traceable during key stages of the fuel’s life cycle. In this paper, we present our efforts to understand the effects of doping UO 2 with Fe 2 O 3 under multiple parameters, such as Fe concentration (0–3000 wppm), sintering temperature (1673 K, 1773 K, and 1973 K), sintering time (2–12 h), and atmosphere (reducing and inert). Here, we investigated densification behavior, microstructure evolution, and lattice defects based on the geometric densities, scanning electron microscopy (SEM), X-ray diffractograms, and Raman spectra. The effect of Fe on the microstructure and crystal chemistry of UO 2 is very complex, especially at the intermediate temperature (1773 K), where we found that densification kinetics are enhanced and that lattice contraction occurs for doping concentrations greater than 1000 wppm Fe.
An Experimentally Validated Mesoscale Model of Thermal Conductivity of a UO2 and BeO Composite Nuclear Fuel
Not provided.
In-Situ TEM study of microstructural evolution in proton irradiated single crystal UO2 under high-temperature annealing
Not Available
Post-irradiation examinations of UO 2 composites as part of the Accident Tolerant Fuels Campaign
Enhanced thermal conductivity uranium dioxide composites containing silicon carbide (UO 2 -SiC) and diamond (UO 2 -diamond) have been irradiated to low burnup. The conditions of this irradiation test and subsequent postirradiation examinations are discussed. These irradiations evaluate fuel microstructure and potential fuel cladding interaction under representative light water reactor power conditions. Both non-destructive and destructive techniques have been used to evaluate fuel integrity, fission gas release, fission product distribution, burnup, fuel swelling and cladding strain. Examination of the UO 2 -SiC pellets revealed enhanced cracking when compared to a UO 2 rodlet also irradiated under similar conditions. Additionally instability of the SiC whiskers in the uranium dioxide matrix was observed in the pellet central region, where the local temperatures exceeded 1300°C. The microstructure of the diamond-added UO 2 was severely disrupted during irradiation, resulting in local migration of cesium along the fuel stack and increased fission gas release when compared with the expected release from the Vitanza curve at corresponding values of burnup and irradiation temperature. The postirradiation examination results cast doubt on the suitability of these additives to improve UO 2 fuel performance in a way that would lead to enhanced accident tolerance.
In-situ ion irradiation induced nanograin growth in a spent UO 2 fuel
This study investigates the irradiation-driven evolution of nanograins in the early-stage restructured rim region of medium burnup spent uranium dioxide (UO 2 ) fuel. Transmission electron microscopy (TEM) lamellas prepared from Belgium Reactor 3 (BR-3) fuel were subjected to in-situ 300 keV Xe ion irradiations under varying doses, fluxes, and temperatures to evaluate their effect on the evolution of the nanograins. Our results reveal that the nanograins grow during ion irradiation. Additionally, the growth is most pronounced at elevated temperatures (300 °C), moderate at room temperature, and negligible at cryogenic temperature (−223 °C). This behavior indicates that thermal activation, alongside irradiation effects, is essential to overcome grain boundary pinning by fission gas bubbles, metallic precipitates, and porosity. Furthermore, while nanograins (<200 nm) consistently coarsened under irradiation, larger grains did not undergo further restructuring, which can be attributed to the strong defect annihilation at TEM lamella surfaces combined with the limited electronic stopping power of low energy Xe ions used in this work. These findings highlight the roles of thermal spike effects, defect mobility, and impurity pinning in governing grain evolution in the rim region of spent UO 2 fuel during ion irradiation, providing key insights for predictive models of restructuring and performance of the high burnup nuclear fuel.
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.
BISON capability and validation for U3Si2, Cr2O3-doped UO2, FeCrAl, and Cr-coated Zircaloy ATF concepts
Development, implementation, and validation of material and behavior models for accident tolerant fuel (ATF) concepts in the BISON fuel performance code began in 2014 in response to the events that occurred at the Fukushima Daichii nuclear power plant in March 2011. Early on, the focus was on U3Si2 fuel and FeCrAl cladding as part of a high-impact problem through the Nuclear Energy and Advanced Model- ing Simulation program [1]. Then, developments for Cr2O3- doped UO2 fuel, as well as SiC-SiC and Cr-coated Zircaloy cladding began based upon industry interests. Near the end of fiscal year 2018 the Consortium for Advanced Simulation of Light Water Reactors took over further Accident Tolerant Fuel (ATF) work in BISON in support of Nuclear Regulatory Commission (NRC) engagement. Discussions with the NRC clarified their list of priority fuel and cladding concepts, which included Cr2O3-doped UO2 and U3Si2 fuels, as well as Cr-coated Zircaloy and FeCrAl claddings. Further, reports similar in form to NUREG/CR-7024 [2] that was developed for traditional light water reactor (LWR) materials, UO2 fuel with Zircaloy claddings were created for the priority fuel and cladding concepts. In this paper select details from the reports entitled "ATF material model development and validation for priority fuel concepts" [3] and "ATF material model development and validation for priority cladding concepts" [4] are presented. First, a brief overview of the available models and their associated uncertainty are provided for the priority fuel and cladding concepts. Then, for the concepts with experimental data, comparisons between simulation and experiment are shown taking into account forward uncertainty quantification (UQ) techniques. For the concepts without publicly available experimental data, parametric studies incorporating forward UQ are completed to verify claims made about the materials’ behavior.
Microstructural and Micro-Chemical Evolutions in the Irradiated UCO Fuel Kernels of AGR-1 and AGR-2 TRISO Fuel Particles
AGR-1 and AGR-2 TRISO fuel particles were fabricated with slightly different fuel kernel chemical compositions, modified fabrication processes, different fuel kernel diameters, and changed 235U enrichments. To correlate those differences with the fuel kernel responses to neutron irradiations in terms of irradiated fuel microstructure, fission products chemical and physical states, and fission gas bubble evolutions, extensive microstructural and analytical characterizations were conducted. The studies used a state of art transmission electron microscopy (TEM) equipped with Energy-dispersive X-ray spectroscopy (EDS) of four silicon solid-state detectors which have super sensitivity and fast speed. The TEM specimens were prepared from selected AGR-1 and AGR-2 irradiated fuel kernels exposed to safety testing after irradiation. The particles were chosen to represent a representative irradiation conditions with a fuel burnup within the range from 10.8 to 18.6% FIMA, and the time-average volume-average temperatures vary from 1070 to 1287°C. The 235U enrichment was 19.74 wt.% for the AGR1 fuel kernels and 14.03 wt.% for the AGR-2 fuel kernels. The TEM results show that there were significant microstructural reconstructions in the irradiated fuel kernels for both the AGR-1 and AGR-2 fuels. There are four major phases including fuel matrix of UO2 and UC, U2RuC2, and UMoC2 in the irradiated AGR2 fuel kernel. Zr and Nb form solid solution in the UC phase. UMoC2 phase often shows a detectable concentration of Tc. Pd was found to mainly locate in the buffer layer or to be associated with fission gas bubble within the UMoC2 phase. The EDS maps qualitatively show that the rare-earth fission products (Nb, et al.) preferentially reside in the UO2 phase. In contrast, in the irradiated AGR1 fuel kernel, no U2RuC2 or UMoC2 precipitates were positively identified. Instead, there is a high number of rod-shape precipitates enriched with Ru, Tc, Rh, and Pd observed in the fuel kernel center and edge zone. The difference of microstructural and micro-chemical evolutions in irradiated fuel kernels between the AGR-1 and AGR-2 TRISO fuel particle may result from a combined factor of irradiation temperature, fuel geometry and chemical composition. However, the irradiation temperature probably play a more deterministic role. Limited electron energy loss spectroscopy (EELS) characterizations on the AGR2 fuel kernel show that there is nearly no carbon in the UO2 phase while a small fraction of oxygen was detected in the UC/UMoC2 phase.