Ridge National Laboratory High Flux Research Articles

As-Built Simulation of the High Flux Isotope Reactor

The Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) is an 85 MWt flux trap-type research reactor that supports key research missions, including isotope production, materials irradiation, and neutron scattering. The core consists of an inner and an outer fuel element containing 171 and 369 involute-shaped plates, respectively. The thin fuel plates consist of a U3O8-Al dispersion fuel (highly enriched), an aluminum-based filler, and aluminum cladding. The fuel meat thickness is varied across the width of the involute plate to reduce thermal flux peaks at the radial edges of the fuel elements. Some deviation from the designed fuel meat shaping is allowed during manufacturing. A homogeneity scan of each fuel plate checks for potential anomalies in the fuel distribution by scanning the surface of the plate and comparing the attenuation of the beam to calibration standards. While typical HFIR simulations use homogenized fuel regions, explicit models of the plates were developed under the Low-Enriched Uranium Conversion Program. These explicit models typically include one inner and one outer fuel plate with nominal fuel distributions, and then the plates are duplicated to fill the space of the corresponding fuel element. Therefore, data extracted from these simulations are limited to azimuthally averaged quantities. To determine the reactivity and physics impacts of an as-built outer fuel element and generate azimuthally dependent data in the element, 369 unique fuel plate models were generated and positioned. This model generates the three-dimensional (i.e., radial–axial–azimuthal) plate power profile, where the azimuthal profile is impacted by features within the adjacent control element region and beryllium reflector. For an as-built model of the outer fuel element, plate-specific homogeneity data, 235U loading, enrichment, and channel thickness measurements were translated into the model, yielding a much more varied azimuthal power profile encompassed by uncertainty factors in analyses. These models were run with the ORNL-TN and Shift Monte Carlo tools, and they contained upwards of 500,000 cells and 100,000 unique tallies.

Modeling and simulation of a High Flux Isotope Reactor representative core model for updated performance and safety basis assessments

A high-fidelity neutronics model of the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) with a representative core and experiment loading was developed to serve as the new basis for performance and safety basis assessments. HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world to support high-impact applied and basic neutron science research. The newly developed model better characterizes ongoing and envisioned research activities at HFIR, in comparison to the legacy Cycle 400 (operated in 2004) model. The new reactor model serves as the reference for safety basis, reactor operation, in-core experiment, reactor upgrade, and other research activities. It also serves as the reference for high-enriched uranium to low-enriched uranium conversion studies, enabling consistent performance and safety metrics comparisons.Neutronic performance and safety metrics calculational methods and results including, but not limited to, cycle length estimates, flux and fission distributions, point kinetics data, reactivity coefficients, intra cycle and post shutdown source terms, and control element worths are documented herein. These neutronics results provide essential input to higher accuracy follow-on heat deposition, thermal–hydraulic, thermal-structural, reactor transient, and severe accident analyses and directly support safety analysis upgrades, startup and operations calculations, and fuel storage and transportation evaluations.Furthermore, thermal–hydraulic safety limit calculations for inlet coolant temperature, flux-to-flow, and inlet vessel pressure that utilized the neutronics results obtained with the new reactor model are discussed herein. The use of the new neutronics results led to a significant gain in thermal safety margin and enabled the removal of previous conservatism based on low-fidelity calculations.

The Neutron Macromolecular Crystallography Instruments at Oak Ridge National Laboratory: Advances, Challenges, and Opportunities

The IMAGINE and MaNDi instruments, located at Oak Ridge National Laboratory High Flux Isotope Reactor and Spallation Neutron Source, respectively, are powerful tools for determining the positions of hydrogen atoms in biological macromolecules and their ligands, orienting water molecules, and for differentiating chemical states in macromolecular structures. The possibility to model hydrogen and deuterium atoms in neutron structures arises from the strong interaction of neutrons with the nuclei of these isotopes. Positions can be unambiguously assigned from diffraction studies at the 1.5–2.5 Å resolutions, which are typical for protein crystals. Neutrons have the additional benefit for structural biology of not inducing radiation damage to protein crystals, which can be critical in the study of metalloproteins. Here we review the specifications of the IMAGINE and MaNDi beamlines and illustrate their complementarity. IMAGINE is suitable for crystals with unit cell edges up to 150 Å using a quasi-Laue technique, whereas MaNDi provides neutron crystallography resources for large unit cell samples with unit cell edges up to 300 Å using the time of flight (TOF) Laue technique. The microbial culture and crystal growth facilities which support the IMAGINE and MaNDi user programs are also described.

A generalized method for characterization of 235U and 239Pu content using short-lived fission product gamma spectroscopy

Rapid nondestructive assay methods for trace fissile material analysis are needed in both nuclear forensics and safeguards communities. To address these needs, research at the Oak Ridge National Laboratory High Flux Isotope Reactor Neutron Activation Analysis facility has developed a generalized nondestructive assay method to characterize materials containing fissile isotopes. This method relies on gamma-ray emissions from short-lived fission products and makes use of differences in fission product yields to identify fissile compositions of trace material samples. Although prior work has explored the use of short-lived fission product gamma-ray measurements, the proposed method is the first to provide a complete characterization of isotopic identification, mass ratios, and absolute mass determination. Successful single fissile isotope mass recoveries of less than 6% recovery bias have been conducted on standards of 235U and 239Pu as low as 12ng in less than 10 minutes. Additionally, mixtures of fissile isotope standards containing 235U and 239Pu have been characterized as low as 198ng of fissile mass with less than 7% recovery bias. The generalizability of this method is illustrated by evaluating different fissile isotopes, mixtures of fissile isotopes, and two different irradiation positions in the reactor. It is anticipated that this method will be expanded to characterize additional fissile nuclides, utilize various irradiation facilities, and account for increasingly complex sample matrices.

Reactor production of Thorium-229

Limited availability of 229Th for clinical applications of 213Bi necessitates investigation of alternative production routes. In reactor production, 229Th is produced from neutron transmutation of 226Ra, 228Ra, 227Ac and 228Th. Irradiations of 226Ra, 228Ra, and 227Ac targets at the Oak Ridge National Laboratory High Flux Isotope Reactor result in yields of 229Th at 26 days of 74.0±7.4MBq/g, 260±10MBq/g, and 1200±50MBq/g, respectively. Intermediate radionuclide yields and cross sections are also studied.

The CG-1D Neutron Imaging Beamline at the Oak Ridge National Laboratory High Flux Isotope Reactor

The Oak Ridge National Laboratory Neutron Sciences Directorate has installed a neutron imaging beamline at the High Flux Isotope Reactor (HFIR) cold guide hall. CG-1D is one of the three instruments that make up the CG1 instrument suite. The beamline optics and detector have recently been upgraded to meet the needs of the neutron imaging community (better “smoothing” of guide system artifacts, higher flux or spatial resolution). These upgrades comprise a new diffuser/aperture system, two new detectors, a He-filled flight tube and silicon (Si) windows. Shielding inside the flight tube, beam scrapers and a beam stop ensure that biological dose is less than 50μSv/hr outside of the radiation boundary.A set of diffusers and apertures (pinhole geometry) has been installed at the exit of the guide system to allow motorized L/D variation. Samples sit on a translation/rotation stage for alignment and tomography purposes. Detectors for the CG-1D beamline are (1) an ANDOR DW936 charge coupled device (CCD) camera with a field of view of approximately 7cm x 7cm and ∼ 80 microns spatial resolution and 1 frame per second time resolution, (2) a new Micro-Channel Plate (MCP) detector with a 2.8cm x 2.8cm field of view and 55 microns spatial resolution, and 5μs timing capability. 6LiF/ZnS scintillators of thickness varying from 50 to 200 microns are being used at this facility. An overview of the beamline upgrade and preliminary data is presented here.

Radiation-tolerant joining technologies for silicon carbide ceramics and composites

Silicon carbide (SiC) for nuclear structural applications, whether in the monolithic ceramic or composite form, will require a robust joining technology capable of withstanding the harsh nuclear environment. This paper presents significant progress made towards identifying and processing irradiation-tolerant joining methods for nuclear-grade SiC. In doing so, a standardized methodology for carrying out joint testing has been established consistent with the small volume samples mandated by neutron irradiation testing. Candidate joining technologies were limited to those that provide low induced radioactivity and included titanium diffusion bonding, Ti–Si–C MAX-phase joining, calcia–alumina glass–ceramic joining, and transient eutectic-phase SiC joining. Samples of these joints were irradiated in the Oak Ridge National Laboratory High Flux Isotope Reactor at 500 or 800°C, and their microstructure and mechanical properties were compared to pre-irradiation conditions. Within the limitations of statistics, all joining methodologies presented retained their joint mechanical strength to ∼3dpa at 500°C, thus indicating the first results obtained on irradiation-stable SiC joints. Under the more aggressive irradiation conditions (800°C, ∼5dpa), some joint materials exhibited significant irradiation-induced microstructural evolution; however, the effect of irradiation on joint strength appeared rather limited.

Irradiation effects on graphite foam

The solid state reactor is an advanced reactor concept that takes advantage of newly developed materials with enhanced heat transfer characteristics to provide an inherently safe, self-regulated heat source. High conductivity graphite foam, developed and produced at Oak Ridge National Laboratory, is being evaluated as a candidate material for the core of basic heat source modules. Irradiation studies at the Oak Ridge National Laboratory High Flux Isotope Reactor were conducted to obtain preliminary data on the effects of neutron damage on the thermal properties and volume change behavior of the graphite foam as a function of neutron dose up to 2.6 displacements per atom at an irradiation temperature of ∼740 °C. Samples were characterized for dimensional and structural changes, and thermal transport as a function of dose. Following the initial effects of the irradiation, the samples were annealed at 1000 and 1200 °C and the thermal diffusivity measured as a function of temperature. A simple microstructural model was developed for graphite foam and, by coupling this model to the known single crystal and polycrystalline irradiation behavior of graphite; a mechanism by which the irradiation-induced volume and dimensional changes in graphite foam may be explained is postulated.

Use of Autoassociative Neural Networks for Signal Validation

Recently, the use of Autoassociative Neural Networks (AANNs) to perform on-line calibration monitoring of process sensors has been shown to not only be feasible, but practical as well. This paper summarizes the results of applying AANNs to instrument surveillance and calibration monitoring at Florida Power Corporation’s Crystal River #3 Nuclear Power Plant and at the Oak Ridge National Laboratory High Flux Isotope Reactor. In both cases sensor drifts are detectable at a nominal level of 0.5% instrument’s full scale range. This paper will discuss the selection of a five layer neural network architecture, a robust training paradigm, the input selection criteria, and a retuning algorithm.

A novel approach to aluminum determination in biological tissues using a pair of pneumatic tube irradiation facilities

A novel method for the determination of trace aluminum (AI) in the presence of high levels of phosphorus (P) has been developed at the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR). Using successive irradiations in HFIR's two pneumatic tube facilities (PT-1, PT-2) to measure the difference between the two apparent Al concentrations, the true Al concentration in the sample can be calculated without the need for an independent determination of P. Results are presented for brain samples from various regions, some that are strongly affected by Alzheimer's disease (AD) and for NIST SRM 1577b, Bovine Liver.

Flow-induced plastic collapse of stacked fuel plates

Flow-induced plastic collapse of stacked fuel plate assemblies was first noted in experimental nuclear reactors such as the Oak Ridge National Laboratory High Flux Reactor Assembly and the Engineering Test Reactor (ETR). The ETR assembly is a stack of 19 thin flat rectangular fuel plates separated by narrow channels through which a coolant flows to remove the heat generated by the nuclear fission of the fuel within the plates. The uranium alloyed plates have been noted to buckle laterally and plastically collapse at the system design coolant flow rate of 10.7 m/s, thus restricting the coolant flow through adjacent channels. In this paper a methodology and criterion are developed for predicting the plastic collapse of ETR fuel plates. The criterion is compared to some experimental results and the Miller critical velocity theory.