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This study examines the production of 232U via neutron irradiation. Uranium-232 is considered for use as a tracer in nuclear fuel. However, a source of 232U is needed. This study examines the production of 232U via neutron irradiation of targets constructed out of either 231Pa or 230Th in the High Flux Isotope Reactor. HFIR targets were modeled in MCNP to determine 232U yield in protactinium and thorium targets. Flux tallies were used inside the targets to determine the neutron flux inside each target. This flux was then used in SCALE 6.2 ORIGEN to determine the 232U yield, as well as the buildup of the byproduct 233U. Several 230Th enrichments were examined to determine how 232U yield is affected by 230Th enrichment, as well as the effects of the presence of 232Th. The buildup of 228Th and 229Th in thorium targets was also examined, as these isotopes may impact the feasibility of recycling of thorium target materials. • Uranium-232 is considered for use as a tracer in uranium fuels. • Reactor irradiation of protactinium-231 and thorium-230 are considered to be the most viable means of production. • Protactinium-231 is considered the preferred target material, requiring only one neutron capture. • The High Flux Isotope Reactor was considered for production, with the flux trap being the preferred irradiation facility. • Target modeling determines that uranium-232 yield from thorium-230 targets is significantly impacted by resonance absorption. [ABSTRACT FROM AUTHOR]

Abstract: The High Flux Isotope Reactor located at the Oak Ridge National Laboratory is a versatile 85MWth research reactor with cold and thermal neutron scattering, materials irradiation, isotope production, and neutron activation analysis capabilities. HFIR staff members are currently in the process of updating the thermal hydraulic and reactor transient modeling methodologies. COMSOL Multiphysics has been adopted for the thermal hydraulic analyses and has proven to be a powerful finite-element-based simulation tool for solving multiple physics-based systems of partial and ordinary differential equations. Modeling reactor transients is a challenging task because of the coupling of neutronics, heat transfer, and hydrodynamics. This paper presents a preliminary COMSOL-based neutronics study performed by creating a two-dimensional, two-group, diffusion neutronics model of HFIR to study the spatially-dependent, beginning-of-cycle fast and thermal neutron fluxes. The 238-group ENDF/B-VII neutron cross section library and NEWT, a two-dimensional, discrete-ordinates neutron transport code within the SCALE 6 code package, were used to calculate the two-group neutron cross sections required to solve the diffusion equations. The two-group diffusion equations were implemented in the COMSOL coefficient form PDE application mode and were solved via eigenvalue analysis using a direct (PARDISO) linear system solver. A COMSOL-provided adaptive mesh refinement algorithm was used to increase the number of elements in areas of largest numerical error to increase the accuracy of the solution. The flux distributions calculated by means of COMSOL/SCALE compare well with those calculated with benchmarked three-dimensional MCNP and KENO models, a necessary first step along the path to implementing two- and three-dimensional models of HFIR in COMSOL for the purpose of studying the spatial dependence of transient-induced behavior in the reactor core. [Copyright &y& Elsevier]

This study is focused on a comparison of experimental neutron flux data from the VR-1 reactor with results from Serpent calculation. Serpent data were calculated by the dynamic part of the code and the steady-state criticality calculation. The Serpent calculation was performed with the JEFF 3.3 and ENDF/B-VIII.0 nuclear data library. This paper expands previous work and focuses on the integral control rod worth. Comparison of the ENDF/B-VIII.0 and JEFF 3.3 nuclear data libraries with experimental data is especially valuable. The comparison shows better performance of JEFF 3.3 nuclear data instead of ENDF/B-VIII.0 for the transient calculation of the cases presented in this article. In addition, the comparison of static and dynamic reactivity provided an interesting difference, especially for small rod insertion. Moreover, the study confirms that the Serpent dynamic external source simulation mode provides an accurate description of reactor behavior, and that the shape of the integral control rod worth is close to one experimentally measured. • Control rod drop experiment was performer at VR-1 reactor. • Results from experiment were compared with Serpent dynamic calculation. • Diff. nuclear data libraries were used to study difference in delayed neutron data. • JEFF 3.3 nuclear data library provides better description compared to ENDF/B-VIII.0. [ABSTRACT FROM AUTHOR]