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At least 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

SCCTE: An LEU NTP Concept with Tungsten Cermet Fuel

Presented in this paper is the Space Capable Cryogenic Thermal Engine (SCCTE), a Low Enriched Uranium (LEU) W-UO2 cermet fuel, ZrH1.8 moderated Nuclear Thermal Propulsion (NTP) Concept. The SCCTE project was undertaken by a collaboration with NASA Marshall Space Flight Center (MSFC) and Center for Space Nuclear Research (CSNR) to investigate the applicability of LEU cermet NTP technology for a human Mars mission in the 2030's time frame. This paper focuses on the nuclear design aspects of SCCTE.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Performance of U3Si2 Fuel in a Reactivity Insertion Accident

In this study we examined the performance of the U3Si2 fuel cladded with Zircaloy (Zr) in a reactivity insertion accident (RIA) in a PWR core. The power excursion as a result of a $1 reactivity insertion was calculated by a TRACE PWR plant model using point-kinetics, for alternative cores with UO2 and U3Si2 fuel assemblies. The point-kinetics parameters (feedback coefficients, prompt-neutron lifetime and group constants for six delayed-neutron groups) were obtained from beginning-of-cycle equilibrium full core calculations with PARCS. In the PARCS core calculations, the few-group parameters were developed utilizing the TRITON/NEWT tools in the SCALE package. In order to assess the fuel response in finer detail (e.g. the maximum fuel temperature) the power shape and thermal boundary conditions from the TRACE/PARCS calculations were used to drive a BISON model of a fuel pin with U3Si2 and UO2 respectively. For a $1 reactivity transient both TRACE and BISON predicted a higher maximum fuel temperature for the UO2 fuel than the U3Si2 fuel. Furthermore, BISON is noted to calculate a narrower gap and a higher gap heat transfer coefficient than TRACE. This resulted in BISON predicting consistently lower fuel temperatures than TRACE. This study also provides a systematic comparison between TRACE and BISON using consistent transient boundary conditions. The TRACE analysis of the RIA only reflects the core-wide response in power. A refinement to the analysis would be to predict the local peaking in a three-dimensional core as a result of control rod ejection.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

BISON Modeling of Reactivity-Initiated Accident Experiments in a Static Environment

In conjunction with the restart of the TREAT reactor and the design of test vehicles, modeling and simulation efforts are being used to model the response of Accident Tolerant Fuel (ATF) concepts under reactivity insertion accident (RIA) conditions. The purpose of this work is to model a baseline case of a 10 cm long UO2-Zircaloy fuel rodlet using BISON and RELAP5 over a range of energy depositions and with varying reactor power pulse widths. The results show the effect of varying the pulse width and energy deposition on both thermal and mechanical parameters that are important for predicting failure of the fuel rodlet. The combined BISON/RELAP5 model captures coupled thermal and mechanical effects on the fuel-to-cladding gap conductance, cladding-to-coolant heat transfer coefficient and water temperature and pressure that would not be capable in each code individually. These combined effects allow for a more accurate modeling of the thermal and mechanical response in the fuel rodlet and thermal-hydraulics of the test vehicle.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Thermomechanical simulation of the DIAMINO irradiation experiment using the LICOS fuel design code

Two separate-effect experiments in the HFR and OSIRIS Material Test Reactors (MTRs) are currently under Post- Irradiation Examinations (MARIOS) and under preparation (DIAMINO) respectively. The main goal of these experiments is to investigate gaseous release and swelling of Am-bearing UO2-x fuels as a function of temperature, fuel microstructure and gas production rate. First, a brief description of the MARIOS and DIAMINO irradiations is provided. Then, the innovative experimental in-pile device specifically developed for the DIAMINO experiment is described. Eventually, the thermo-mechanical computations performed using the LICOS code are presented. These simulations support the DIAMINO experimental design and highlight some of the capabilities of the code. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

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↗
DOE OSTI2023

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.

58 GEOSCIENCES↗
DOE OSTI2019

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.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Radioactive Semivolatiles in Nuclear Fuel Reprocessing

In nuclear fuel reprocessing, various radioactive elements enter the gas phase from the unit operations found in the reprocessing facility. In previous reports, the pathways and required removal were discussed for four radionuclides known to be volatile, 14 C, 3 H, 129 I, and 85 Kr. Other, less volatile isotopes can also report to the off-gas streams in a reprocessing facility. These were reported to be isotopes of Cs, Cd, Ru, Sb, Tc, and Te. In this report, an effort is made to determine which, if any, of 24 semivolatile radionuclides could be released from a reprocessing plant and, if so, what would be the likely quantities released. As part of this study of semivolatile elements, the amount of each generated during fission is included as part of the assessment for the need to control their emission. Also included in this study is the assessment of the cooling time (time out of reactor) before the fuel is processed. This aspect is important for the short-lived isotopes shown in the list, especially for cooling times approaching 10 y. The approach taken in this study was to determine if semivolatile radionuclides need to be included in a list of gas-phase radionuclides that might need to be removed to meet Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) regulations. A list of possible elements was developed through a literature search and through knowledge and literature on the chemical processes in typical aqueous processing of nuclear fuels. A long list of possible radionuclides present in irradiated fuel was generated and then trimmed by considering isotope half-life and calculating the dose from each to a maximum exposed individual with the US EPA airborne radiological dispersion and risk assessment code CAP88 (Rosnick 1992) to yield a short list of elements that actually need to be considered for control because they require high decontamination factors to meet a reasonable fraction of the regulated release. Each of these elements is then discussed with respect to what is known in the literature about their behavior in a reprocessing facility. The context for the evaluation in this document is a UO2-based fuel processed through an aqueous-based reprocessing system with a TBP-based solvent extraction chemistry. None of these elements form sufficiently volatile compounds in the context of the reprocessing facility to be of regulatory concern.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

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.

Crystallography↗
DOE OSTI2019

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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

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.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2020

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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗