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Design Study for a Low-enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2007

Description: This report documents progress made during fiscal year 2007 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low enriched uranium fuel (LEU). Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. A high volume fraction U/Mo-in-Al fuel could attain the same neutron flux performance as with the current, HEU fuel but materials considerations appear to preclude production and irradiation of such a fuel. A diffusion barrier would be required if Al is to be retained as the interstitial medium and the additional volume required for this barrier would degrade performance. Attaining the high volume fraction (55 wt. %) of U/Mo assumed in the computational study while maintaining the current fuel plate acceptance level at the fuel manufacturer is unlikely, i.e. no increase in the percentage of plates rejected for non-compliance with the fuel specification. Substitution of a zirconium alloy for Al would significantly increase the weight of the fuel element, the cost of the fuel element, and introduce an as-yet untried manufacturing process. A monolithic U-10Mo foil is the choice of LEU fuel for HFIR. Preliminary calculations indicate that with a modest increase in reactor power, the flux performance of the reactor can be maintained at the current level. A linearly-graded, radial fuel thickness profile is preferred to the arched profile currently used in HEU fuel because the LEU fuel media is a metal alloy foil rather than a powder. Developments in analysis capability and nuclear data processing techniques are underway with the goal of verifying the preliminary calculations of LEU flux performance. A conceptual study of the operational cost of an LEU fuel fabrication facility yielded the conclusion that the annual fuel cost to the HFIR would increase significantly from ...
Date: November 1, 2007
Creator: Primm, Trent; Ellis, Ronald James; Gehin, Jess C; Ilas, Germina; Miller, James Henry & Sease, John D
Partner: UNT Libraries Government Documents Department

Validating MCNP for LEU Fuel Design via Power Distribution Comparisons

Description: The mission of the Reduced Enrichment for Research and Test Reactors (RERTR) Program is to minimize and, to the extent possible, eliminate the use of highly enriched uranium (HEU) in civilian nuclear applications by working to convert research and test reactors, as well as radioisotope production processes, to low enriched uranium (LEU) fuel and targets. Oak Ridge National Lab (ORNL) is reviewing the design bases and key operating criteria including fuel operating parameters, enrichment-related safety analyses, fuel performance, and fuel fabrication in regard to converting the fuel of the High Flux Isotope Reactor (HFIR) from HEU to LEU. The purpose of this study is to validate Monte Carlo methods currently in use for conversion analyses. The methods have been validated for the prediction of flux values in the reactor target, reflector, and beam tubes, but this study focuses on the prediction of the power density profile in the core. A current 3-D Monte Carlo N-Particle (MCNP) model was modified to replicate the HFIR Critical Experiment 3 (HFIRCE-3) core of 1965. In this experiment, the power profile was determined by counting the gamma activity at selected locations in the core. Foils (chunks of fuel meat and clad) were punched out of the fuel elements in HFIRCE-3 following irradiation and experimental relative power densities were obtained by measuring the activity of these foils and comparing each foil s activity to the activity of a normalizing foil. The current work consisted of calculating corresponding activities by inserting volume tallies into the modified MCNP model to represent the punchings. The average fission density was calculated for each foil location and then normalized to the normalizing foil. Power distributions were obtained for the clean core (no poison in moderator and symmetrical rod position at 17.5 inches) and fully poisoned-moderator (1.35 g B/liter in moderator and ...
Date: November 1, 2008
Creator: Primm, Trent; Maldonado, G Ivan & Chandler, David
Partner: UNT Libraries Government Documents Department

Impact of HFIR LEU Conversion on Beryllium Reflector Degradation Factors

Description: An assessment of the impact of low enriched uranium (LEU) conversion on the factors that may cause the degradation of the beryllium reflector is performed for the High Flux Isotope Reactor (HFIR). The computational methods, models, and tools, comparisons with previous work, along with the results obtained are documented and discussed in this report. The report documents the results for the gas and neutronic poison production, and the heating in the beryllium reflector for both the highly enriched uranium (HEU) and LEU HFIR configurations, and discusses the impact that the conversion to LEU may have on these quantities. A time-averaging procedure was developed to calculate the isotopic (gas and poisons) production in reflector. The sensitivity of this approach to different approximations is gauged and documented. The results show that the gas is produced in the beryllium reflector at a total rate of 0.304 g/cycle for the HEU configuration; this rate increases by ~12% for the LEU case. The total tritium production rate in reflector is 0.098 g/cycle for the HEU core and approximately 11% higher for the LEU core. A significant increase (up to ~25%) in the neutronic poisons production in the reflector during the operation cycles is observed for the LEU core, compared to the HEU case, for regions close to the core s horizontal midplane. The poisoning level of the reflector may increase by more than two orders of magnitude during long periods of downtime. The heating rate in the reflector is estimated to be approximately 20% lower for the LEU core than for the HEU core. The decrease is due to a significantly lower contribution of the heating produced by the gamma radiation for the LEU core. Both the isotopic (gas and neutronic poisons) production and the heating rates are spatially non-uniform throughout the beryllium reflector volume. ...
Date: October 1, 2013
Creator: Ilas, Dan
Partner: UNT Libraries Government Documents Department

Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011

Description: This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.
Date: March 1, 2012
Creator: Renfro, David G; Cook, David Howard; Freels, James D; Griffin, Frederick P; Ilas, Germina; Sease, John D et al.
Partner: UNT Libraries Government Documents Department

Low-Enriched Uranium Fuel Design with Two-Dimensional Grading for the High Flux Isotope Reactor

Description: An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.
Date: May 1, 2011
Creator: Ilas, Germina & Primm, Trent
Partner: UNT Libraries Government Documents Department

Preliminary Multiphysics Analyses of HFIR LEU Fuel Conversion using COMSOL

Description: The research documented herein was performed by several individuals across multiple organizations. We have previously acknowledged our funding for the project, but another common thread among the authors of this document, and hence the research performed, is the analysis tool COMSOL. The research has been divided into categories to allow the COMSOL analysis to be performed independently to the extent possible. As will be seen herein, the research has progressed to the point where it is expected that next year (2011) a large fraction of the research will require collaboration of our efforts as we progress almost exclusively into three-dimensional (3D) analysis. To the extent possible, we have tried to segregate the development effort into two-dimensional (2D) analysis in order to arrive at techniques and methodology that can be extended to 3D models in a timely manner. The Research Reactors Division (RRD) of ORNL has contracted with the University of Tennessee, Knoxville (UTK) Mechanical, Aerospace and Biomedical Engineering Department (MABE) to perform a significant fraction of this research. This group has been chosen due to their expertise and long-term commitment in using COMSOL and also because the participating students are able to work onsite on a part-time basis due to the close proximity of UTK with the ORNL campus. The UTK research has been governed by a statement of work (SOW) which clearly defines the specific tasks reported herein on the perspective areas of research. Ph.D. student Isaac T. Bodey has focused on heat transfer, fluid flow, modeling, and meshing issues and has been aided by his major professor Dr. Rao V. Arimilli and is the primary contributor to Section 2 of this report. Ph.D student Franklin G. Curtis has been focusing exclusively on fluid-structure interaction (FSI) due to the mechanical forces acting on the plate caused by the flow ...
Date: June 1, 2011
Creator: Freels, James D; Bodey, Isaac T; Arimilli, Rao V; Curtis, Franklin G; Ekici, Kivanc & Jain, Prashant K
Partner: UNT Libraries Government Documents Department

Validation of a Monte Carlo Based Depletion Methodology Using HFIR Post-Irradiation Measurements

Description: Post-irradiation uranium isotopic atomic densities within the core of the High Flux Isotope Reactor (HFIR) were calculated and compared to uranium mass spectrographic data measured in the late 1960s and early 70s [1]. This study was performed in order to validate a Monte Carlo based depletion methodology for calculating the burn-up dependent nuclide inventory, specifically the post-irradiation uranium
Date: November 1, 2009
Creator: Chandler, David; Maldonado, G Ivan & Primm, Trent
Partner: UNT Libraries Government Documents Department

3D COMSOL Simulations for Thermal Deflection of HFIR Fuel Plate in the "Cheverton-Kelley" Experiments

Description: Three dimensional simulation capabilities are currently being developed at Oak Ridge National Laboratory using COMSOL Multiphysics, a finite element modeling software, to investigate thermal expansion of High Flux Isotope Reactor (HFIR) s low enriched uranium fuel plates. To validate simulations, 3D models have also been developed for the experimental setup used by Cheverton and Kelley in 1968 to investigate the buckling and thermal deflections of HFIR s highly enriched uranium fuel plates. Results for several simulations are presented in this report, and comparisons with the experimental data are provided when data are available. A close agreement between the simulation results and experimental findings demonstrates that the COMSOL simulations are able to capture the thermal expansion physics accurately and that COMSOL could be deployed as a predictive tool for more advanced computations at realistic HFIR conditions to study temperature-induced fuel plate deflection behavior.
Date: August 1, 2012
Creator: Jain, Prashant K; Freels, James D & Cook, David Howard
Partner: UNT Libraries Government Documents Department

DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

Description: This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.
Date: February 2011
Creator: Cook, David Howard; Freels, James D.; Ilas, Germina; Jolly, Brian C.; Miller, James Henry; Primm, R. Trent, III et al.
Partner: UNT Libraries Government Documents Department

Neutron Diffraction Residual Strain Tensor Measurements Within The Phase IA Weld Mock-up Plate P-5

Description: Oak Ridge National Laboratory (ORNL) has worked with NRC and EPRI to apply neutron and X-ray diffraction methods to characterize the residual stresses in a number of dissimilar metal weld mockups and samples. The design of the Phase IA specimens aimed to enable stress measurements by several methods and computational modeling of the weld residual stresses. The partial groove in the 304L stainless steel plate was filled with weld beads of Alloy 82. A summary of the weld conditions for each plate is provided in Table 1. The plates were constrained along the long edges during and after welding by bolts with spring-loaded washers attached to the 1-inch thick Al backing plate. The purpose was to avoid stress relief due to bending of the welded stainless steel plate. The neutron diffraction method was one of the methods selected by EPRI for non-destructive through thickness strain and stress measurement. Four different plates (P-3 to P-6) were studied by neutron diffraction strain mapping, representing four different welding conditions. Through thickness neutron diffraction strain mappings at NRSF2 for the four plates and associated strain-free d-zero specimens involved measurement along seven lines across the weld and at six to seven depths. The mountings of each plate for neutron diffraction measurements were such that the diffraction vector was parallel to each of the three primary orthogonal directions of the plate: two in-plane directions, longitudinal and transverse, and the direction normal to the plate (shown in left figure within Table 1). From the three orthogonal strains for each location, the residual stresses along the three plate directions were calculated. The principal axes of the strain and stress tensors, however, need not necessarily align with the plate coordinate system. To explore this, plate P-5 was selected for examination of the possibility that the principal axes of strain ...
Date: September 1, 2011
Creator: Hubbard, Camden R
Partner: UNT Libraries Government Documents Department

Density of Gadolinium Nitrate Solutions for the High Flux Isotope Reactor

Description: In late 1992, the High Flux Isotope Reactor (HFIR) was planning to switch the solution contained in the poison injection tank from cadmium nitrate to gadolinium nitrate. The poison injection system is an emergency system used to shut down the reactor by adding a neutron poison to the cooling water. This system must be able to supply a minimum of 69 pounds of gadolinium to the reactor coolant system in order to guarantee that the reactor would become subcritical. A graph of the density of gadolinium nitrate solutions over a concentration range of 5 to 30 wt% and a temperature range of 15 to 40{sup o}C was prepared. Routine density measurements of the solution in the poison injection tank are made by HFIR personnel, and an adaptation of the original graph is used to determine the gadolinium nitrate concentration. In late 2008, HFIR personnel decided that the heat tracing that was present on the piping for the poison injection system could be removed without any danger of freezing the solution; however, the gadolinium nitrate solution might get as cold as 5{sup o}C. This was outside the range of the current density-concentration correlation, so the range needed to be expanded. This report supplies a new density-concentration correlation that covers the extended temperature range. The correlation is given in new units, which greatly simplifies the calculation that is required to determine the pounds of gadolinium in the tank solution. The procedure for calculating the amount of gadolinium in the HFIR poison injection system is as follows: (1) Calculate the usable volume in the system; (2) Measure the density of the solution; (3) Calculate the gadolinium concentration using the following equation: Gd(lb/ft{sup 3}) = measured density (g/mL) x 34.681 - 34.785; (4) Calculate the amount of gadolinium in the system using the following ...
Date: May 1, 2009
Creator: Taylor, Paul Allen & Lee, Denise L
Partner: UNT Libraries Government Documents Department

An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

Description: The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.
Date: August 1, 2009
Creator: Rosenthal, Murray Wilford
Partner: UNT Libraries Government Documents Department

MEASURED AND CALCULATED HEATING AND DOSE RATES FOR THE HFIR HB4 BEAM TUBE AND COLD SOURCE

Description: The High Flux Isotope Reactor at the Oak Ridge National Laboratory was upgraded to install a cold source in horizontal beam tube number 4. Calculations were performed and measurements were made to determine heating within the cold source and dose rates within and outside a shield tunnel surrounding the beam tube. This report briefly describes the calculations and presents comparisons of the measured and calculated results. Some calculated dose rates are in fair to good agreement with the measured results while others, particularly those at the shield interfaces, differ greatly from the measured results. Calculated neutron exposure to the Teflon seals in the hydrogen transfer line is about one fourth of the measured value, underpredicting the lifetime by a factor of four. The calculated cold source heating is in good agreement with the measured heating.
Date: March 1, 2009
Creator: Slater, Charles O; Primm, Trent; Pinkston, Daniel; Cook, David Howard; Selby, Douglas L; Ferguson, Phillip D et al.
Partner: UNT Libraries Government Documents Department

Analysis of HFIR Dosimetry Experiments Performed in Cycles 400 and 401

Description: The High Flux Isotope Reactor (HFIR) has been in operation at Oak Ridge National Laboratory since 1966. To upgrade and enhance capabilities for neutron science research at the reactor, a larger HB-2 beam tube was installed in April of 2002. To assess, experimentally, the impact of this larger beam tube on radiation damage rates [i.e., displacement-per-atom (dpa) rates] used in vessel life extension studies, dosimetry experiments were performed from April to August 2004 during fuel cycles 400 and 401. This report documents the analysis of the dosimetry experiments and the determination of best-estimate dpa rates. These dpa rates are obtained by performing a least-squares adjustment of calculated neutron and gamma-ray fluxes and the measured responses of radiometric monitors and beryllium helium accumulation fluence monitors. The best-estimate dpa rates provided here will be used to update HFIR pressure vessel life extension studies, which determine the pressure/temperature limits for reactor operation and the HFIR pressure vessel's remaining life. All irradiation parameters given in this report correspond to a reactor power of 85 MW.
Date: September 1, 2008
Creator: Remec, Igor & Baldwin, Charles A
Partner: UNT Libraries Government Documents Department

Reactivity Accountability Attributed to Reflector Poisons in the High Flux Isotope Reactor

Description: The objective of this study is to develop a methodology to predict the reactivity impact as a function of outage time between cycles of 3He, 6Li, and other poisons in the High Flux Isotope Reactor s (HFIR) beryllium reflector. The reactivity worth at startup of the HFIR has been incorrectly predicted in the past after the reactor has been shut-down for long periods of time. The incorrect prediction was postulated to be due to the erroneous calculation of 3He buildup in the beryllium reflector. It is necessary to develop a better estimate of the start-of-cycle symmetric critical control element positions since if the estimated and actual symmetrical critical control element positions differ by more than $1.55 in reactivity (approximately one-half inch in control element startup position), HFIR is to be shutdown and a technical evaluation is performed to resolve the discrepancy prior to restart. 3He is generated and depleted during operation, but during an outage, the depletion of 3He ceases because it is a stable isotope. 3He is born from the radioactive decay of tritium, and thus the concentration of 3He increases during shutdown. SCALE, specifically the TRITON and CSAS5 control modules including the KENO V.A, COUPLE, and ORIGEN functional modules were utilized in this study. An equation relating the down time (td) to the change in symmetric control element position was generated and validated against measurements for approximately 40 HFIR operating cycles. The newly-derived correlation was shown to improve accuracy of predictions for long periods of down time.
Date: December 1, 2009
Creator: Chandler, David; Maldonado, G Ivan & Primm, Trent
Partner: UNT Libraries Government Documents Department

Partial Safety Analysis for a Reduced Uranium Enrichment Core for the High Flux Isotope Reactor

Description: A computational model of the reactor core of the High Flux Isotope Rector (HFIR) was developed in order to analyze non-destructive accidents caused by transients during reactor operation. The reactor model was built for the latest version of the nuclear analysis software package called Program for the Analysis of Reactor Transients (PARET). Analyses performed with the model constructed were compared with previous data obtained with other tools in order to benchmark the code. Finally, the model was used to analyze the behavior of the reactor under transients using a different nuclear fuel with lower enrichment of uranium (LEU) than the fuel currently used, which has a high enrichment of uranium (HEU). The study shows that the presence of fertile isotopes in LEU fuel, which increases the neutron resonance absorption, reduces the impact of transients on the fuel and enhances the negative reactivity feedback, thus, within the limitations of this study, making LEU fuel appear to be a safe alternative fuel for the reactor core.
Date: April 1, 2009
Creator: Primm, Trent & Gehin, Jess C
Partner: UNT Libraries Government Documents Department

Upgraded HFIR Fuel Element Welding System

Description: The welding of aluminum-clad fuel plates into aluminum alloy 6061 side plate tubing is a unique design feature of the High Flux Isotope Reactor (HFIR) fuel assemblies as 101 full-penetration circumferential gas metal arc welds (GMAW) are required in the fabrication of each assembly. In a HFIR fuel assembly, 540 aluminum-clad fuel plates are assembled into two nested annular fuel elements 610 mm (24-inches) long. The welding process for the HFIR fuel elements was developed in the early 1960 s and about 450 HFIR fuel assemblies have been successfully welded using the GMAW process qualified in the 1960 s. In recent years because of the degradation of the electronic and mechanical components in the old HFIR welding system, reportable defects in plate attachment or adapter welds have been present in almost all completed fuel assemblies. In October 2008, a contract was awarded to AMET, Inc., of Rexburg, Idaho, to replace the old welding equipment with standard commercially available welding components to the maximum extent possible while maintaining the qualified HFIR welding process. The upgraded HFIR welding system represents a major improvement in the welding system used in welding HFIR fuel elements for the previous 40 years. In this upgrade, the new inner GMAW torch is a significant advancement over the original inner GMAW torch previously used. The innovative breakthrough in the new inner welding torch design is the way the direction of the cast in the 0.762 mm (0.030-inch) diameter aluminum weld wire is changed so that the weld wire emerging from the contact tip is straight in the plane perpendicular to the welding direction without creating any significant drag resistance in the feeding of the weld wire.
Date: February 1, 2010
Creator: Sease, John D.
Partner: UNT Libraries Government Documents Department

Establishing a Cost Basis for Converting the High Flux Isotope Reactor from High Enriched to Low Enriched Uranium Fuel

Description: Under the auspices of the Global Threat Reduction Initiative Reduced Enrichment for Research and Test Reactors Program, the National Nuclear Security Administration /Department of Energy (NNSA/DOE) has, as a goal, to convert research reactors worldwide from weapons grade to non-weapons grade uranium. The High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab (ORNL) is one of the candidates for conversion of fuel from high enriched uranium (HEU) to low enriched uranium (LEU). A well documented business model, including tasks, costs, and schedules was developed to plan the conversion of HFIR. Using Microsoft Project, a detailed outline of the conversion program was established and consists of LEU fuel design activities, a fresh fuel shipping cask, improvements to the HFIR reactor building, and spent fuel operations. Current-value costs total $76 million dollars, include over 100 subtasks, and will take over 10 years to complete. The model and schedule follows the path of the fuel from receipt from fuel fabricator to delivery to spent fuel storage and illustrates the duration, start, and completion dates of each subtask to be completed. Assumptions that form the basis of the cost estimate have significant impact on cost and schedule.
Date: February 1, 2010
Creator: Primm, Trent & Guida, Tracey
Partner: UNT Libraries Government Documents Department

Evaluation of HFIR LEU Fuel Using the COMSOL Multiphysics Platform

Description: A finite element computational approach to simulation of the High Flux Isotope Reactor (HFIR) Core Thermal-Fluid behavior is developed. These models were developed to facilitate design of a low enriched core for the HFIR, which will have different axial and radial flux profiles from the current HEU core and thus will require fuel and poison load optimization. This report outlines a stepwise implementation of this modeling approach using the commercial finite element code, COMSOL, with initial assessment of fuel, poison and clad conduction modeling capability, followed by assessment of mating of the fuel conduction models to a one dimensional fluid model typical of legacy simulation techniques for the HFIR core. The model is then extended to fully couple 2-dimensional conduction in the fuel to a 2-dimensional thermo-fluid model of the coolant for a HFIR core cooling sub-channel with additional assessment of simulation outcomes. Finally, 3-dimensional simulations of a fuel plate and cooling channel are presented.
Date: March 1, 2009
Creator: Primm, Trent; Ruggles, Arthur & Freels, James D
Partner: UNT Libraries Government Documents Department

Fuel Grading Study on a Low-Enriched Uranium Fuel Design for the High Flux Isotope Reactor

Description: An engineering design study that would enable the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium to low-enriched uranium fuel is ongoing at Oak Ridge National Laboratory. The computational models used to search for a low-enriched uranium (LEU) fuel design that would meet the requirements for the conversion study, and the recent results obtained with these models during FY 2009, are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating high-enriched uranium fuel core. These studies indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations.
Date: November 1, 2009
Creator: Ilas, Germina & Primm, Trent
Partner: UNT Libraries Government Documents Department

Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2008

Description: This report documents progress made during FY 2008 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Scoping experiments with various manufacturing methods for forming the LEU alloy profile are presented.
Date: March 1, 2009
Creator: Primm, Trent; Chandler, David; Ilas, Germina; Miller, James Henry; Sease, John D & Jolly, Brian C
Partner: UNT Libraries Government Documents Department

Development of a Scale Model for High Flux Isotope Reactor Cycle 400

Description: The development of a comprehensive SCALE computational model for the High Flux Isotope Reactor (HFIR) is documented and discussed in this report. The SCALE model has equivalent features and functionality as the reference MCNP model for Cycle 400 that has been used extensively for HFIR safety analyses and for HFIR experiment design and analyses. Numerical comparisons of the SCALE and MCNP models for the multiplication constant, power density distribution in the fuel, and neutron fluxes at several locations in HFIR indicate excellent agreement between the results predicted with the two models. The SCALE HFIR model is presented in sufficient detail to provide the users of the model with a tool that can be easily customized for various safety analysis or experiment design requirements.
Date: March 1, 2012
Creator: Ilas, Dan
Partner: UNT Libraries Government Documents Department

Establishing Specifications for Low Enriched Uranium Fuel Operations Conducted Outside the High Flux Isotope Reactor Site

Description: The National Nuclear Security Administration (NNSA) has funded staff at Oak Ridge National Laboratory (ORNL) to study the conversion of the High Flux Isotope Reactor (HFIR) from the current, high enriched uranium fuel to low enriched uranium fuel. The LEU fuel form is a metal alloy that has never been used in HFIR or any HFIR-like reactor. This report provides documentation of a process for the creation of a fuel specification that will meet all applicable regulations and guidelines to which UT-Battelle, LLC (UTB) the operating contractor for ORNL - must adhere. This process will allow UTB to purchase LEU fuel for HFIR and be assured of the quality of the fuel being procured.
Date: October 1, 2010
Creator: Pinkston, Daniel; Primm, Trent; Renfro, David G & Sease, John D
Partner: UNT Libraries Government Documents Department

2D Thermal Hydraulic Analysis and Benchmark in Support of HFIR LEU Conversion using COMSOL

Description: The research documented herein was funded by a research contract between the Research Reactors Division (RRD) of Oak Ridge National Laboratory (ORNL) and the University of Tennessee, Knoxville (UTK) Mechanical, Aerospace and Biomedical Engineering Department (MABE). The research was governed by a statement of work (SOW) which clearly defines nine specific tasks. This report is outlined to follow and document the results of each of these nine specific tasks. The primary goal of this phase of the research is to demonstrate, through verification and validation methods, that COMSOL is a viable simulation tool for thermal-hydraulic modeling of the High Flux Isotope Reactor (HFIR) core. A secondary goal of this two-dimensional phase of the research is to establish methodology and data base libraries that are also needed in the full three-dimensional COMSOL simulation to follow. COMSOL version 3.5a was used for all of the models presented throughout this report.
Date: September 1, 2010
Creator: Freels, James D; Bodey, Isaac T; Lowe, Kirk T & Arimilli, Rao V
Partner: UNT Libraries Government Documents Department