R. T. Primm

Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel

A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U{sub 3}O{sub 8} mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties.

Simulation of Mixed-Oxide and Low-Enriched Uranium Fuel Burnup in a Pressurized Water Reactor and Validation Against Destructive Analysis Results

Abstract This work examines the capabilities of simulation codes to predict the concentration of nuclides in spent reactor fuel, in particular mixed-oxide (MOX) fuel, via comparisons with destructive radiochemical analyses performed on irradiated samples. We report on three MOX samples irradiated in a pressurized water reactor (PWR) and two UO2 samples irradiated in a different PWR. Actinide and fission-product concentrations were measured and were compared with concentration values obtained from simulation studies. The actinides include isotopes of uranium, neptunium, plutonium, americium, and curium. The fission products include isotopes of cesium, neodymium, samarium, europium, and gadolinium as well as 90Sr, 95Mo, 99Tc, 101Ru, 106Ru, 103Rh, 109Ag, 125Sb, 129I, and 144Ce. For many of the actinides, the predictions are quite good when compared with the measured values; but concentrations of some tend to be overpredicted. The cesium and neodymium, and some samarium concentrations, are well predicted, but some of the other fission products show variable results. The sensitivity of some of the results to sample-burnup estimates is discussed.

Neutronic Analysis of an Advanced Fuel Design Concept for the High Flux Isotope Reactor

Abstract This study presents the neutronic analysis of an advanced fuel design concept for the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) that could significantly extend the current fuel cycle length under the existing design and safety criteria. A key advantage of the fuel design herein proposed is that it would not require structural changes to the present HFIR core, in other words, maintaining the same rated power and fuel geometry (i.e., fuel plate thickness and coolant channel dimensions). Of particular practical importance, as well, is the fact that the proposed change could be justified within the bounds of the existing nuclear safety basis. The simulations herein reported employed transport theory–based and exposure-dependent eigenvalue characterization to help improve the prediction of key fuel cycle parameters. These parameters were estimated by coupling a benchmarked three-dimensional MCNP5 model of the HFIR core to the depletion code ORIGEN via the MONTEBURNS interface. The design of an advanced HFIR core with an improved fuel loading is an idea that evolved from early studies by R. D. Cheverton, formerly of ORNL. This study contrasts a modified and increased core loading of 12 kg of 235U against the current core loading of 9.4 kg. The simulations performed predict a cycle length of 39 days for the proposed fuel design, which represents a 50% increase in the cycle length in response to a 25% increase in fissile loading, with an average fuel burnup increase of ~23%. The results suggest that the excess reactivity can be controlled with the present design and arrangement of control elements throughout the core’s life. Also, the new power distribution is comparable or even improved relative to the current power distribution, displaying lower peak to average fission rate densities across the inner fuel element’s centerline and bottom cells. In fact, the fission rate density in the outer fuel element also decreased at these key locations for the proposed design. Overall, it is estimated that the advanced core design could increase the availability of the HFIR facility by ~50% and generate ~33% more neutrons annually, which is expected to yield sizeable savings during the remaining life of HFIR, currently expected to operate through 2014. This study emphasizes the neutronics evaluation of a new fuel design. Although a number of other performance parameters of the proposed design check favorably against the current design, and most of the core design features remain identical to the reference, it is acknowledged that additional evaluations would be required to fully justify the thermal-hydraulic and thermal-mechanical performance of a new fuel design, including checks for cladding corrosion performance as well as for industrial and economic feasibility.

Power Distribution Analysis for the ORNL High Flux Isotope Reactor Critical Experiment 3

Abstract The mission of the Reduced Enrichment for Research and Test Reactors 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 Laboratory (ORNL) is currently 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. Power distributions were calculated in the fuel elements of the HFIR, a research reactor at ORNL, via MCNP and were compared to experimentally obtained data. This study was performed to validate Monte Carlo methods for power density calculations and to observe biases. A current three-dimensional MCNP model was modified to replicate the 1965 HFIR Critical Experiment 3 (HFIRCE-3). 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. This analysis 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 reference foil. Power distributions were obtained for clean core (no poison in moderator and symmetrical rod position at 44.536 cm withdrawn with respect to the core axial midplane) and fully poisoned moderator (1.35 grams of boron per liter in moderator and rods fully withdrawn) conditions. The observed deviations between the experimental and calculated values for both conditions were within the reported experimental uncertainties except for some foils located on the top and bottom edges of the fuel plates.