The Transient Reactor Test Facility (TREAT)

Constructed in the late 1950s, the Transient Reactor Test facility (TREAT) provided numerous transient irradiations until operation was suspended in 1994. It was later refurbished, and resumed operations in 2017 to meet the data needs of a new era of nuclear fuel safety research. TREAT uses uranium oxide dispersed in graphite blocks to yield a core that affords strong negative temperature feedback. Automatically controlled, fast-acting transient control rods enable TREAT to safely perform extreme power maneuvers—ranging from prompt bursts to longer power ramps—to broadly support research on postulated accidents for many reactor types. TREAT’s experiment devices work in concert with the reactor to contain specimens, support in situ diagnostics, and provide desired test environments, thus yielding a uniquely versatile facility. This chapter summarizes TREAT’s design, history, current efforts, and future endeavors in the field of nuclear-heated fuel safety research

AFIP-4 Irradiation Summary Report

The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-4 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a scale prototypic of research reactor fuel plates. The AFIP-4 test further examine the fuel/clad interface and its behavior under extreme conditions. After irradiation, fission gas retention measurements will be performed during post irradiation (PIE)1,2. The following report summarizes the life of the AFIP-4 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results

Monolithic Fuel Fabrication Process Development at the Idaho National Laboratory

Within the Reduced Enrichment for Research and Test Reactors (RERTR) program directed by the US Department of Energy (DOE), UMo fuel-foils are being developed in an effort to realize high density monolithic fuel plates for use in high-flux research and test reactors. Namely, targeted are reactors that are not amenable to Low Enriched Uranium (LEU) fuel conversion via utilization of high density dispersion-based fuels, i.e. 8-9 gU/cc. LEU conversion of reactors having a need for >8-9 gU/cc fuel density will only be possible by way of monolithic fuel forms. The UMo fuel foils under development afford fuel meat density of ~16 gU/cc and thus have the potential to facilitate LEU conversions without any significant reactor-performance penalty. Two primary challenges have been established with respect to UMo monolithic fuel development; namely, fuel element fabrication and in-reactor fuel element performance. Both issues are being addressed concurrently at the Idaho National Laboratory. An overview is provided of the ongoing monolithic UMo fuel development effort at the Idaho National Laboratory (INL); including development of complex/graded fuel foils. Fabrication processes to be discussed include: UMo alloying and casting, foil fabrication via hot rolling, fuel-clad interlayer application via co-rolling and thermal spray processes, clad bonding via Hot Isostatic Pressing (HIP) and Friction Bonding (FB), and fuel plate finishing

Evaluation of Annealing Treatments for Producing Si-Rich Fuel/Matrix Interaction Layers in Low-Enriched U-Mo Dispersion Fuel Plates Rolled at a Low Temperature

During fabrication of U-7Mo dispersion fuels, exposure to relatively high temperatures affects the final microstructure of a fuel plate before it is inserted into a reactor. One impact of this high temperature exposure is a chemical interaction that can occur between dissimilar materials. For U-7Mo dispersion fuels, the U-7Mo particles will interact to some extent with the Al or Al alloy matrix to produce interaction products. It has been observed that the final irradiation behavior of a fuel plate can depend on the amount of interaction that occurs at the U-7Mo/matrix interface during fabrication, along with the type of phases that develop at this interface. For the case where a U-7Mo dispersion fuel has a Si-containing Al alloy matrix and is rolled at around 500°C, a Si-rich interaction product has been observed to form that can potentially have a positive impact on fuel performance during irradiation. This interaction product can exhibit stable irradiation behavior and it can act as a diffusion barrier to additional U-Mo/matrix interaction during irradiation. However, for U-7Mo dispersion fuels with softer claddings that are rolled at lower temperatures (e.g., near 425°C), a significant interaction layer has not been observed to form. As a result, the bulk of any interaction layer that develops in these fuels happens during irradiation, and the layer that forms may not exhibit as stable a behavior as one that is formed during fabrication. Therefore, it may be beneficial to add a heat treatment step during the fabrication of dispersion fuel plates with softer cladding alloys that will result in the formation of a uniform, Si-rich interaction layer that is a few microns thick around the U-Mo fuel particles. This type of layer would have characteristics like the one that has been observed in dispersion fuel plates with AA6061 cladding that are fabricated at 500°C, which may exhibit increased stability during irradiation. This report discusses the result of annealing experiments that were performed using samples from fuel plates that were fabricated at 425°C that had Alloy 5052 cladding. As part of these experiments, samples with Al-Si matrices that had different Si contents were tested. The samples had Al-2Si, Al-4Si, Al-5Si, or Al-6Si as the matrix alloy. The heat treatment temperatures and times that were investigated were 450°C (4 hours), 475°C (4 hours), and 500°C (2 hours) for all the matrix alloy compositions and 525°C (1 hour) for just the Al-4Si and Al-6Si matrix alloy compositions. The results of these experiments showed that the initial interaction layers that form around the U-7Mo particles during fabrication at 425°C continue to grow in thickness and uniformity during each of the heat treatments, though the composition of the layers remains similar to that observed in the as-fabricated samples. The Al-6Si matrix sample annealed at 450°C for 4 hours and the Al-5Si and Al-6Si matrix samples annealed at 475°C for 4 hours formed fuel/matrix interaction layers most similar to those observed in fuel plates with AA6061 cladding that are fabricated at 500°C

Potential Annealing Treatments For Tailoring The Starting Microstructure Of Low-Enriched U-Mo Dispersion Fuels To Optimize Performance During Irradiation

Low-enriched uranium-molybdenum (U-Mo) alloy particles dispersed in aluminum alloy (e.g., dispersion fuels) are being developed for application in research and test reactors. To achieve the best performance of these fuels during irradiation, optimization of the starting microstructure may be required by utilizing a heat treatment that results in the formation of uniform, Si-rich interaction layers between the U-Mo particles and Al-Si matrix. These layers behave in a stable manner under certain irradiation conditions. To identify the optimum heat treatment for producing these kinds of layers in a dispersion fuel plate, a systematic annealing study has been performed using actual dispersion fuel samples, which were fabricated at relatively low temperatures to limit the growth of any interaction layers in the samples prior to controlled heat treatment. These samples had different Al matrices with varying Si contents and were annealed between 450 and 525°C for up to 4 h. The samples were then characterized using scanning electron microscopy (SEM) to examine the thickness, composition, and uniformity of the interaction layers. Image analysis was performed to quantify various attributes of the dispersion fuel microstructures that related to the development of the interaction layers. The most uniform layers were observed to form in fuel samples that had an Al matrix with at least 4 wt.% Si and a heat treatment temperature of at least 475°C. © 2011 Elsevier B.V. All rights reserved

Correction To: A Source Biasing and Variance Reduction Technique for Monte Carlo Radiation Transport Modeling of Emission Tomography Problems (Journal of Radioanalytical and Nuclear Chemistry, (2019), 320, 1, (37-45), 10.1007/S10967-019-06457-1)

In the original article, Equation 4 was incorrectly published. The correct equation is provided in this correction

Accelerated Radiation Transport Modeling Techniques for Pencil Beam Computed Tomography using Gamma Rays

Monte Carlo radiation transport modeling studies were performed for a compact, and high-resolution gamma-ray computed tomography system designed for imaging irradiated nuclear fuel. The system comprises a 60Co source – chosen for its highly penetrating 1173 keV and 1332 keV gamma rays – a pair of high-aspect-ratio pencil beam collimators, and an inorganic scintillator detector. Two acceleration methods are proposed to rapidly model a transmission type gamma-ray tomography system. The first, a variance reduction technique, is based on performing Monte Carlo simulations with a monodirectionally-biased source, sampled from a characteristic sub-volume of the full source volume. The second acceleration method is based on the deterministic calculations using the Beer–Lambert law and detector response characteristics. Comparison of simulations using acceleration approaches with analog simulations of the fully isotropic, full-volume equivalent, show that the Monte Carlo variance reduction technique gives quantitatively accurate predictions for large collimator aspect ratios while the deterministic calculations are semi-quantitative but converge close to the correct result as the collimator aspect ratio increases. As such, these techniques can be used to reduce the computational cost in generating simulated radiographs and tomographs by several orders of magnitude. Experimental validation efforts are currently underway and will be demonstrated in future work

AFIP-4 Irradiation Summary Report

The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-4 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a scale prototypic of research reactor fuel plates. The AFIP-4 test further examine the fuel/clad interface and its behavior under extreme conditions. After irradiation, fission gas retention measurements will be performed during post irradiation (PIE)1,2. The following report summarizes the life of the AFIP-4 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results

AFIP-4 Irradiation Summary Report

The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-4 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a scale prototypic of research reactor fuel plates. The AFIP-4 test further examine the fuel/clad interface and its behavior under extreme conditions. After irradiation, fission gas retention measurements will be performed during post irradiation (PIE). The following report summarizes the life of the AFIP-4 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results