Advanced Test Reactor Research Articles

Microstructural changes and their effect on hardening in neutron irradiated Fe-Cr alloys

A series of Fe-3 to 18 at.% Cr binary ferritic alloys were neutron irradiated side by side, in the Advanced Test Reactor (ATR) at a temperature of ∼320 °C to a dose of ∼1.8 dpa. Three types of features that form under irradiation are relevant: (i) solute clusters; (ii) α′ precipitates; and, (iii) dislocation loops. The size and number density of the precipitates and loops were measured by transmission electron microscopy (TEM) and compared to previous atom probe tomography (APT) observations. The loop density systematically decreases with increasing Cr, while the number of α′ precipitates increase at 9Cr and above. A major objective of this work is to estimate the dispersed barrier obstacle strength factors (αj) for loops, clusters and α′ precipitates, based on the combined microstructural observations and corresponding irradiation hardening measurements. Standard dispersed barrier-hardening models and computationally derived superposition rules were least square fit to determine the αj. The optimized hardening predictions are in very good agreement with experiment if mixed linear sum and root square sum superposition rules are used. Five increments of 168 h isochronal anneals of the 6Cr alloy between 300 and 400 °C coarsened the loops, while 300 h anneals of the 18Cr at 500 and 600 °C coarsened and dissolved the α′ precipitates, respectively, consistent with the Fe-Cr phase diagram.

Postirradiation examinations of low burnup U3Si2 fuel for light water reactor applications

The United States Department of Energy Advanced Fuels Campaign in collaboration with industry is working to develop enhanced accident tolerant fuels for light water reactors. The initial irradiations of these concepts were performed in the Idaho National Laboratory Advanced Test Reactor using drop-in style irradiations. These irradiations are collectively referred to as Accident Tolerant Fuels 1 (ATF-1). The postirradiation examination of ATF-1 has begun and some concepts have now completed initial examination. This work is focused on postirradiation examination data of U3Si2 fuels at low burnup (i.e., < 20 GWd/tHM) for application in Light Water Reactors (LWRs). The U3Si2 pellets show limited cracking in comparison to the expected behavior of UO2 at same power level. In addition, gamma spectrometry data did not reveal migration of fission products. Minor homogeneous hardening along the pellet radius due to accumulation of fission products and radiation damage was measured by microindentation. The fission gas release and swelling remain very low. Formation of fission gas bubbles resolvable with optical microscopy occurs from the pellets center outward to approximately 60% of the fuel pellet radius. The overall data suggest a good performance of this accident tolerant fuel candidate at low burnup.

Safety margin characterization for high power fueled experiments

The updated safety basis of the Advanced Test Reactor (ATR) has specific criteria for experiments that are cooled by the primary coolant. It requires that during a Condition 2 transient (where flow coastdown is typically the most limiting), the Critical Heat Flux ratio (CHFR) is greater than 2.0 after consideration of several power multipliers. This safety criteria is often overly conservative for fueled experiments and restricts the acceptance of experiments. In the updated safety basis, a provision for a stochastic approach is included. In this study, coupled Dakota-RELAP5 and RAVEN-RELAP5 are used to propagate uncertainties. The objective of this study is to provide a quantitative basis for utilizing the provisions in the ATR experiment safety criteria.Three different fueled experiments, MP-1, FSP-1, and AFIP-6 are considered. RELAP5 models for each experiment are used to simulate the flow coastdown transient. Several thermal-hydraulic correlations are assessed to determine the Onset of Significant Voiding (OSV), Onset of Flow Instability (OFI), and Critical Heat Flux (CHF). The OSV and OFI correlations considered indicate that there is a large safety margin. The CHF correlations also indicate that there is a moderate to large safety margin. The high mass flux of the experiments exceed the Look-up Table (LUT) correlation range. This results in an artificially restricted prediction of CHF value. A high mass flux modifier was implemented to improve LUT applicability. Due to the specifications of the experiments considered, the correlations must be applicable to higher mass flux ranges. It was also found that, often, the conventional CHFR >2.0 criterion is not met. However, the statistical CHFR margin (μ−3σ>1.2) is met. This result suggests that new criteria, while maintaining adequate safety margins, can be used to expand the operating envelope of high power fueled experiments at ATR. The Pearson ranking indicates that the power level and mass flux have the largest impact on the CHFR and OFI predictions. If a reduction of uncertainty in predictions is sought, the factors affecting fuel experiment fission power should be prioritized for scrutiny, followed by the mass flow rate.

Postirradiation Examination of FUTURIX-FTA metallic alloy experiments

Two different metallic nuclear fuel alloys intended for transmutation applications in fast neutron spectrum nuclear reactors have been irradiated, and their performance has been evaluated after irradiation. These alloys contained elevated levels of plutonium and other minor actinides (Am, Np) compared to historically irradiated fast reactor fuel to ascertain the impact of these isotopes on fuel performance. The irradiation of these alloys was performed at the Phénix fast reactor in France to facilitate comparisons between true fast spectrum irradiations and pseudo-fast spectrum irradiations performed at the thermal neutron spectrum Idaho National Laboratory Advanced Test Reactor. The irradiation tests were designated FUTURIX-FTA DOE1 which contained a low-fertile alloy and FUTURIX-FTA DOE2 which contained a non-fertile alloy. The DOE1 fuel was irradiated to a measured burnup of 9.5% fissions per initial heavy metal atom (FIMA), and the DOE2 fuel was irradiated to a measured burnup of 12.7% fissions per initial heavy metal atom. This work reports baseline Postirradiation Examination (PIE) results for DOE1 and DOE2. Fuel swelling, fission product distribution, cladding strain, fission gas release, fuel microstructure, fuel cladding chemical interaction and minor actinide transmutation were all evaluated. Cladding strain was negligible and fission gas release was 52% for DOE1 and 69% for DOE2. The exams show the 2 pins behaved very similar to EBR-II metallic fuel experience. Minor actinides seem to not affect the performance of this candidate transmutation fuel. Performance data from these irradiations can be used to inform the feasibility of minor actinide transmutation in future reactor systems.

Mechanical properties of irradiated U[sbnd]Mo alloy fuel

This paper presents results of studies characterizing the mechanical properties of irradiated U10Mo fuel in support of low-enriched fuel conversion of several United States high-performance research reactors (USHPRR), specifically the monolithic base fuel qualification. Specimens were tested from ten test fuel plates irradiated at the Advanced Test Reactor (ATR) from the RERTR-12 and AFIP-6 Mk. II irradiation campaigns. Tests were conducted at Idaho National Laboratory (INL) in the Hot Fuel Examination Facility (HFEF) inert atmosphere Main Cell. The monolithic fuel plates consist of a U10Mo fuel foil meat covered with a Zr diffusion barrier layer fabricated by co-rolling, and clad in 6061 Al using a hot isostatic press (HIP) bonding process. Initial unirradiated specimens exhibited nominal (fresh) fuel meat thickness ranging from 0.25 mm to 0.64 mm, and irradiated (as tested) fuel plate average burnup ranged from approximately 0.36 × 1021 to 6.2 × 1021 fissions/cm3. After sectioning individual specimens from the fuel plates, the 6061 Al cladding was removed by dissolution in concentrated NaOH. Pre-dissolution measurement of width and post-dissolution measurement of thickness were conducted on test specimens to facilitate accurate analysis of bend test results. The final thickness values varied from 0.3 to 0.8 mm. Four-point bend testing was conducted on the HFEF Remote Load Frame (RLF) using custom-designed test fixtures and calibrated force and displacement sensors. All specimens exhibited dominantly elastic behavior and failed in a brittle manner. Analysis of test results showed decreasing elastic modulus and bending failure strength with increasing fission density. The trend of modulus decrease intersected with unirradiated modulus at zero fission density. The trend of decreasing bending strength with increased fission density is apparent, but has significantly higher uncertainty levels than the modulus results over the fission density range evaluated. A linear trend fit of the calculated bend strengths versus fission density does not coincide with unirradiated tensile strength, which is significantly higher. Porosities measured on the test specimen material are used with several model equations to predict irradiated modulus values. Over the range of available porosity information, comparison to the experimental modulus values are favorable. Other factors and their potential influence on modulus and strength calculations are discussed.

Carbon-14 content in tree and soil samples at the Idaho National Laboratory nuclear site

Idaho National Laboratory (INL) is a nuclear research facility located in southeastern Idaho, USA; over the course of its operational history, INL has operated 52 reactors and 1 reactor fuel reprocessing facility. To determine the extent to which previous nuclear operations at INL have impacted local environmental carbon-14 (14C) concentrations, tree and soil samples from the INL desert and surrounding areas were collected, combusted, and analyzed by Accelerator Mass Spectrometry (AMS). Transport models of the plumes from the Advanced Test Reactor Facility (ATR) and the Idaho Chemical Processing Plant (ICPP) suggest the historic annual integrated plume distributions from each of these source terms was most likely in the northeast-southwest direction, with very little ground contact from stack emissions in the immediate vicinity of the facilities and maximum estimated ground contact arising ∼400–1000 m to the northeast/southwest from each facility. 14C data from annual growth ring data from trees located immediately adjacent to the ATR, INL’s Central Facilities Area (CFA), and Mud Lake, Idaho (ML) are in statistical agreement with regional nuclear weapons testing fallout backgrounds. Surface soil samples taken near a low level radioactive disposal facility and downwind from the ICPP show percent modern carbon (pMC) values ranging from 28 ± 2 to 92 ± 4, suggesting a mixture of aged and modern background carbon containing materials. Taken together, these data suggest that 14C in the INL region is predominantly derived from a mixture of aged and modern background sources (e.g. nuclear weapons testing fallout), with insignificant contributions from INL source terms.

Safety analysis of the FASt TEst reactor (FASTER) preconceptual design

The FASt TEst Reactor (FASTER) plant is a sodium-cooled fast spectrum test reactor design that was developed as part of an advanced demonstration and test reactor study, and provides high levels of fast and thermal neutron flux for scientific research and development. The 120 MWe FASTER plant has been designed with extended testing capabilities in mind, while keeping the design as simple as possible in order to make it efficient and cost attractive. The main function of the reactor is to provide neutrons for irradiation testing utilizing existing proven technologies to ensure a high technology readiness level. A preliminary safety analysis was performed of the FASTER design using the fast reactor safety analysis code SAS4A/SASSYS-1 to assess the reactor's safety performance during beyond design basis accidents. Three types of transient scenarios were analyzed for the FASTER design: unprotected transient overpower, unprotected loss of heat sink, and unprotected station blackout. These transients were simulated at beginning of equilibrium cycle and end of equilibrium cycle conditions. For each transient, failures of both the primary and secondary reactor scram systems were assumed. Because of the favorable reactivity feedbacks, relatively low peak fuel, cladding, and coolant temperatures were obtained during the transients and the reactor vessel temperature remained well below the limit for ensuring structural stability.

Postirradiation examination on metallic fuel in the AFC-2 irradiation test series

Irradiation experiments on metallic fuel alloys containing rare earth additions to simulate potential carry-over from pyro-metallurgical reprocessing and minor actinide additions were performed in the Idaho National Laboratory Advanced Test Reactor. These tests were designated AFC-2A and AFC-2B. Additionally irradiation experiments were also performed on metallic fuel alloys containing minor actinide additions that were fabricated by different techniques. These tests were designated AFC-2E. These pins were removed from the reactor after achieving burnups between 6 and 11% fission per initial heave metal atom. Examination of these pins revealed that they were unintentionally operated beyond their desired power and temperature conditions. Highlights from the postirradiation examination of these experiments is discussed. A variety of issue contributed to the poor performance of these irradiation tests including capsule design, reactor power variations and fabrication issues. While all the issues contributed, the major issue was determined to be unanticipated variations in reactor power. Through comparisons between these experiments and historical off-normal experiments, it was possible to conclude bound the operating temperatures of these experiments. There likely was some in-pile melting in the rare earth addition experiments (AFC-2A, AFC-2B), but it is not likely that there was in-pile melting in the fabrication technique experiments (AFC-2E).

Which Parameters Are Important? Differential Importance Under Uncertainty.

In probabilistic risk assessment, attention is often focused on the expected value of a risk metric. The sensitivity of this expectation to changes in the parameters of the distribution characterizing uncertainty in the inputs becomes of interest. Approaches based on differentiation encounter limitations when (i) distributional parameters are expressed in different units or (ii) the analyst wishes to transfer sensitivity insights from individual parameters to parameter groups, when alternating between different levels of a probabilistic safety assessment model. Moreover, the analyst may also wish to examine the effect of assuming independence among inputs. This work proposes an approach based on the differential importance measure, which solves these issues. Estimation aspects are discussed in detail, in particular the problem of obtaining all sensitivity measures from a single Monte Carlo sample, thus avoiding potentially costly model runs. The approach is illustrated through an analytical example, highlighting how it can be used to assess the impact of removing the independence assumption. An application to the probabilistic risk assessment model of the Advanced Test Reactor large loss of coolant accident sequence concludes the work.

On the flow induced vibration of an externally excited nuclear reactor experiment

An in-pile, drop-in experiment design is presently being designed and studied for the near-term deployment within the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL); this experiment is termed the Miniplate-1 Large-B (MP-1 LB) Experiment. A number of explicit studies are performed during the design- and safety-related stage. Traditionally, a clear and logical methodology has been developed and utilized for analyses such as hydraulics, thermal-loads, mechanical loads, and others for these experiments. Recently a small component from a different experiment assembly mechanically separated while in the reactor’s core. While this experiment didn’t compromise the safety of the reactor, it led to a higher-level question which centered on whether the appropriate level of consideration was being made toward the fluid–structure-interactions of these experiments. The outcome yielded separate flow test experiments of like-for-like geometry in an experimental loop located at Oregon State University which produces experimental data compliant with applicable parts and requirements to ASME’s NQA-1 2008, 2009a standard – suitable for benchmark evaluation. The objectives of this study are to (1) develop a process for handling and interpreting the mechanical response of the test elements during hydraulic testing, to (2) characterize the motion of a specific test element during a flow test which imposes a wide range of hydraulic conditions, and (3) provide objective observations toward the potential safety related implications that are tied to the synthesized data. The outcome of this study has led to a confident process in inspecting the experimental data, synthesized it for interpretation, identified several unique hydraulic characteristics of the experiment design which were previously unknown, and demonstrated that the likelihood for mechanical failure resulting from fluid-structure-interactions in the reactor is far below any criterion for concern of the element’s safety.

Upgrading Limiting Peak-Power Analysis Techniques with Modern Validation and Uncertainty Quantification for the Advanced Test Reactor

This work supports the acceptability of the two-dimensional deterministic transport code HELIOS to replace the legacy diffusion code PDQ for computing the peak-power performance parameters of the Advanced Test Reactor (ATR). The 95% Confidence Rule, commonly used in the commercial reactor sector, is explored to develop the so-called reliability factors that provide statistical confidence that the peak-power limits within the hottest location along a fuel plate, referred to as the hot stripe, will not be exceeded. Additionally, an alternative “legacy” methodology was explored that attempts to mimic the exact PDQ analysis process used for defining the peak-power limits. The legacy methodology involves interpolating power between regions at azimuthal boundaries subtending the regions of interest.In order to apply the 95% Confidence Rule, a statistically significant calculation-to-measurement bias must first be established. Unlike the commercial world, where thousands of power observations can be collected every cycle using online flux-mapping instrumentation, the ATR power distribution must be measured during “depressurized” zero-power measurements using fission wires in polyethylene wands. In 2012, fission wire activation data were collected during a flux run in the Advanced Test Reactor Critical Facility. Also to improve statistical validity, archival data from ATR zero-power flux runs from 1977, 1986, and 1994 were digitized from scanned reports and used to create new benchmark models. Borrowing from least-squares adjustment methods commonly used for neutron activation spectroscopy, adjusted fission wire powers were calculated for all four data sets. The mean and standard deviation of the bias between a priori calculated and adjusted wire powers were then taken as the bias and uncertainty used in the 95% Confidence Rule.

Reducing Uncertainty in Hydrodynamic Modeling of ATR Experiments Via Flow Testing, Validation, and Optimization

The development, characterization, and qualification testing of nuclear fuel at Idaho National Laboratory’s Advanced Test Reactor (ATR) requires extensive design and analysis activities prior to the insertion of an irradiation experiment in-pile. Significant effort is made in the design and development phase of all in-pile experiments to ensure that the maximum feasible impacts of all necessary experimental requirements are satisfied. The advancement of fuel, cladding, and in-reactor materials technology in recent years has introduced complexities associated with the design and construct of in-pile experiments necessitating deeper understanding of boundary conditions and increasingly comprehensive observations resulting from the experiment. Each unique experiment must be assessed for neutronics response, thermal/hydraulic/hydrodynamic performance, and structural integrity. This is accomplished either analytically, computationally, or experimentally, or some combination thereof, prior to insertion into the ATR. The various effects are interrelated to various degrees, such as the case with the experiment temperature affecting the thermal cross section of the fuel or the increased temperature of the experiment’s materials reducing the mechanical strength of the assemblies. Additionally, the feedback between the experiment’s response to a reactor transient could alter the neutron flux profile of the reactor during the transient. Each experiment must therefore undergo a barrage of analyses to assure the ATR operational safety review committee that the insertion and irradiation of the experiment will not detrimentally affect the safe operational envelope of the reactor. In many cases, the nuclear fuel being tested can be double-encapsulated to ensure safety margins are adequately addressed, whereas failed fuel would be encased in a protective capsule. In other cases, the experiments can be inserted in a self-contained loop that passes through the reactor core, remaining isolated from the primary coolant. In the case of research reactor fuel, however, the fuel plates must be tested in direct contact with the reactor coolant, and being fuel designed for high neutron fluxes, they are inherently power-dense plates. The combination of plate geometry, high-power density, and direct contact with primary coolant creates a scenario where the neutronic/thermomechanic/hydrodynamic characteristics of the fuel plates are tightly coupled, necessitating as complete characterization as possible to support the safety and programmatic assessments, thus enabling a successful experiment. This paper explores the efforts of the U.S. High-Performance Research Reactor program to thermomechanically/hydromechanically characterize the program’s wide variety of experiments, which range from stacks of miniplate capsules to full-sized, geometrically representative curved plates. Special attention is given to instances where the combination of experimental characterization and analytical assessment has reduced uncertainties of the safety margins, allowing experiments to be irradiated that would otherwise not have passed the rigorous qualification process for irradiation in the ATR. In some cases, the combined processes have exposed flow and heat transfer characteristics that would have been missed using historical methods, which allows for more accurate and representative postirradiation assessments.

Testing fast reactor fuels in a thermal reactor

The present lack of a domestic fast neutron flux irradiation capability combined with continued development of fast reactor fuels in the U.S. motivated an innovative engineering solution to utilize a unique neutron flux tailoring capability in the Advanced Test Reactor at the Idaho National Laboratory. To achieve the objectives of the fast reactor fuel irradiation tests, the incident neutron flux was hardened substantially by placing fueled irradiation capsules inside specially designed cadmium shrouds. Use of cadmium prevents thermal neutrons from reaching the fuels being tested and alleviates the plutonium self-shielding that would normally arise during irradiations of high density, highly enriched fuels in a thermal neutron spectrum.The present paper illustrates the profound effect this engineered solution has on the efficacy of the experiments. Based on the comparison of post-irradiation measurements of the columnar grain region in fast reactor mixed-oxide fuels with fuel performance calculations, it is demonstrated that thermal conditions achieved in these cadmium-shrouded fuel experiments are substantially prototypic of a sodium fast reactor and are suitable for concept-screening tests supporting development of new fast reactor fuels. It is also shown that if the experiments were conducted in an unmodified ATR neutron spectrum, gross plutonium self-shielding would cause a strong depression of the fission power at the fuel centerline preventing fuel restructuring, a hallmark feature of mixed oxide fuel behavior under fast reactor conditions.Recognizing the need for testing metallic fuels for fast reactors, the impact of neutron flux energy spectrum on the radial temperature distribution in metallic fuel is investigated. It is shown that the use of Cd shrouds allows to attain radial temperature distributions nearly identical to those that exist in an SFR.

Local atomic structure of Pd and Ag in the SiC containment layer of TRISO fuel particles fissioned to 20% burn-up

The structure and speciation of fission products within the SiC barrier layer of tristructural-isotropic (TRISO) fuel particles irradiated to 19.6% fissions per initial metal atom (FIMA) burnup in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) was investigated. As-irradiated fuel particles, as well as those subjected to simulated accident scenarios, were examined. The TRISO particles were characterized using synchrotron X-ray absorption fine-structure spectroscopy (XAFS) at the Materials Research Collaborative Access Team (MRCAT) beamline at the Advanced Photon Source. The TRISO particles were produced at Oak Ridge National Laboratory under the Advanced Gas Reactor Fuel Development and Qualification Program and sent to the ATR for irradiation. XAFS measurements on the palladium and silver K-edges were collected using the MRCAT undulator beamline. Analysis of the Pd edge indicated the formation of palladium silicides of the form PdxSi (2 ≤ x ≤ 3). In contrast, Ag was found to be metallic within the SiC shell safety tested to 1700 °C. To the best of our knowledge, this is the first result demonstrating metallic bonding of silver from fissioned samples. Knowledge of these reaction pathways will allow for better simulations of radionuclide transport in the various coating layers of TRISO fuels for next generation nuclear reactors. They may also suggest different ways to modify TRISO particles to improve their fuel performance and to mitigate potential fission product release under both normal operation and accident conditions.

The AGR-3/4 fission product transport irradiation experiment

AGR-3/4 was the combined third and fourth planned irradiations for the U.S. Department of Energy (DOE) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The primary purpose of the AGR program is to support the development and qualification of tristructural isotropic (TRISO)-coated particle fuel for use in High Temperature Gas-cooled Reactors. AGR-3/4 was designed as a fission product transport irradiation experiment whose specific objectives were to: (1) irradiate fuel containing UCO (uranium oxycarbide) designed-to-fail (DTF) fuel particles that provide a fixed source of fission products for subsequent transport through compact matrix and structural graphite materials; (2) assess the effects of sweep gas impurities on fuel performance and fission product transport; (3) provide irradiated fuel and material samples for post-irradiation examination (PIE) and post-irradiation heating; and (4) support the refinement of fuel performance and fission product transport models. The AGR-3/4 test train was irradiated in the northeast flux trap of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for 369.1 effective full power days from December 2011 to April 2014. The experiment was successful in achieving its specification goals in terms of burnup and fast fluence levels reached at the end of irradiation and fuel temperature levels maintained throughout irradiation: peak compact burnup reached 15.27% fissions per initial heavy-metal atom and peak compact fast fluence reached 5.32 × 1025 n/m2 (E > 0.18 MeV), while the time-average volume-average temperatures of the compacts ranged from 854 to 1345 °C. Fission product release–to–birth ratios reached values in the 10−4–10−3 range early during irradiation as the DTF particles started to fail. Subsequent post-irradiation examination will provide information on fission product distributions in matrix and core graphite materials, enabling refinement of fission product transport models.

Ceramography of irradiated TRISO fuel from the AGR-2 experiment

Ceramography was performed on cross sections from four tristructural isotropic (TRISO) coated particle fuel compacts selected from the AGR-2 experiment, which was irradiated between June 2010 and October 2013 in the Advanced Test Reactor (ATR). The individual fuel specimens examined in this study contained coated particles with either uranium oxide (UO2) kernels or uranium oxide/uranium carbide (UCO) kernels that were irradiated to final burnup values between 9.0 and 11.5% FIMA. These examinations were intended to explore kernel and coating morphology evolution during irradiation. This included observations of kernel porosity, kernel swelling, and irradiation-induced TRISO coating layer fracture and separation. Variations in behavior within a specific cross section, which could be related to temperature or burnup gradients within the fuel compact, were also explored. Results were compared with similar investigations performed as part of the earlier AGR-1 irradiation experiment. The criteria for categorizing AGR-1 particle post-irradiation morphologies during ceramographic exams were applied to the AGR-2 compact particles examined. This paper presents the results of the AGR-2 examinations and discusses the key implications for fuel irradiation performance.

Image fusion of Secondary Ion Mass Spectrometry and Energy-dispersive X-Ray Spectroscopy data for the characterization of uranium-molybdenum fuel foils

Uranium-molybdenum (U-Mo) monolithic fuels represent one option for converting civilian research and test reactors operating with high enriched uranium (HEU) to low enriched uranium (LEU), effectively reducing the threat of nuclear proliferation world-wide. However, processes associated with fabrication of U-Mo monolithic fuels result in regions of elemental heterogeneity, observed as bands traversing the cross-section of representative samples. Isotopic variations (e.g., 235U and 238U) could also be introduced because of associated processing steps, particularly since HEU feedstock is melted with natural or depleted uranium diluent to produce LEU. This study demonstrates the utility of correlative analysis of Energy-Dispersive X-ray Spectroscopy (EDS) and Secondary Ion Mass Spectrometry (SIMS) with their image data streams using image fusion, resulting in a comprehensive microanalytical characterization toolbox. Elemental and isotopic measurements were made on a sample from the Advanced Test Reactor (ATR) Full-sized plate In-center flux trap Position (AFIP)-7 experiment and compared to previous optical and electron microscopy results. The image fusion results are characteristic of SIMS isotopic maps, but with the spatial resolution of EDS images and, therefore, can be used to increase the effective spatial resolution of the SIMS imaging results to better understand homogeneity or heterogeneity that persists because of processing selections. Visual inspection using the image fusion methodology indicated slight variations in the 235U/238U ratio and quantitative analysis using the image intensities across several FoVs revealed an average 235U atom percent value of 17.9 ± 2.4%, which was indicative of a non-uniform U isotopic distribution in the area sampled. Further development of this capability is useful for understanding the connections between the properties of LEU fuel alternatives and the ability to predict performance under irradiation.

Fuel element design and analysis for potential LEU conversion of the Advanced Test Reactor

This paper provides a summary of the status of the design process for a low-enrichment replacement for the high-enrichment fuel that has historically been used in the Advanced Test Reactor (ATR) at Idaho National Laboratory. Under current long-term Department of Energy policy, the ATR will be converted to operate using low-enrichment fuel by 2030. To this end, an engineering evaluation and design process has been underway for more than ten years. Initially performed as a series of scoping studies for different design options, the last five years have seen a transition to a formal Safety-in-Design approach to the design and safety evaluation process. This paper focuses on such efforts in progress since 2012 to identify pre-conceptual and conceptual design candidates for ATR conversion. The Enhanced LEU Fuel (ELF) design is selected as the most promising conceptual design. Analyses and trade-off studies performed that led to that selection are described. Finally, analysis of certain aspects of core performance for an ELF-fueled core relative to the current high-enrichment fuel show that only small and manageable changes to core operations will be necessary to support conversion of ATR to low-enrichment fuel.

Key results from irradiation and post-irradiation examination of AGR-1 UCO TRISO fuel

The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3×105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time-average, volume-average irradiation temperatures of the individual compacts ranged from 955 to 1136°C. This paper focuses on key results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior. The fuel exhibited zero TRISO coating failures (failure of all three dense coating layers) during irradiation and post-irradiation safety testing at temperatures up to 1700°C. Advanced PIE methods have allowed particles with SiC coating failure that were discovered to be present in a very-low population to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of lessons learned from AGR-1 to fuel fabrication and post-irradiation examination for subsequent fuel irradiation experiments as part of the US fuel program are also discussed.

Evaluation of the Enhanced LEU Fuel (ELF) Design for Conversion of the Advanced Test Reactor to a Low-Enrichment Fuel Cycle

A conceptual low-enrichment uranium (LEU) fuel design has been developed for the Advanced Test Reactor (ATR) at Idaho National Laboratory. The ATR is currently fueled with a high-enrichment fuel but is slated to be converted to LEU under programs led by the National Nuclear Security Administration of the U.S. Department of Energy. A conceptual LEU fuel design, the Enhanced LEU Fuel (ELF), has been developed assuming power peaking control through the use of variable fuel meat thicknesses and no use of burnable poison. In initial work, this design was shown to satisfy performance requirements for ATR operation. Following these design calculations, a safety analysis process was initiated to demonstrate that the ELF design would successfully meet safety limits for postulated accident conditions. Those calculations, performed using RELAP5 and ATR-SINDA, require physics analysis to provide spatial power distributions and kinetics parameters for various core operations configurations. This article describes the findings of the physics analysis and provides predictions for the behavior of a LEU-fueled version of ATR, and compares these to calculations of the performance of the current high-enrichment uranium fuel.