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The proper calculation of core periphery discontinuity factors is important for accurate modeling when using an advanced nodal diffusion simulator. In cores with hexagonal assemblies, such as in the VVER-1000, most fuel assemblies share two faces with the radial reflector, and some even three faces. For this reason, use of a two-dimensional (2D) reflector model will more accurately capture the neutron physics near the core periphery. This article illustrates key points related to the use of 2D discontinuity factors in the reflector region, first by using an algorithm that applies the methodology proposed by Mittag et al. (2003) after correcting some minor typographical errors in the original publication, and then by employing the SCALE transport module NEWT to compute the appropriate quantities. Large and even negative discontinuity factors are an acceptable fact of this methodology when the diffusion approximation becomes invalid due to the problem’s localized features and the large flux gradients. [ABSTRACT FROM AUTHOR]

The Monte Carlo code KENO contains thermal scattering data for a wide variety of thermal moderators. These data are processed from Evaluated Nuclear Data Files (ENDF) by AMPX and stored as double differential probability distribution functions. The method examined in this paper uses S ( α , β ) probability distribution functions derived from the ENDF data files directly instead of being converted to double differential cross sections. This allows the size of the cross section data on the disk to be reduced substantially amount. KENO has also been updated to allow interpolation in temperature on these data so that problems can be run at any temperature. Results are shown for several simplified problems for a variety of moderators. In addition, benchmark models based on the KRITZ reactor in Sweden were run, and the results are compared with the previous versions of KENO without the direct S ( α , β ) method. Results from the direct S ( α , β ) method compare favorably with the original results obtained using the double differential cross sections. Sampling the data increases the run-time of the Monte Carlo calculation, but memory usage is decreased substantially. [ABSTRACT FROM AUTHOR]

The isotope 232U is being considered for additive to uranium fuel for use as a tracer. The 232U decay chain has high energy gammas that can be used for tracer purposes. However, the presence and intensity of such gammas may increase effective dose rates to workers around such materials This study examines the dose rate from different uranium materials that varying amounts of 232U has been added to. Several materials and their respective storage geometries are modeled for particle transport calculations. For each material and 232U concentration, gamma source terms were generated using SCALE 6.2 ORIGEN. These source terms were then used in MCNP for each material and geometry. Upon analysis, it was determined that the baseline dose from enriched uranium dominates until the 232U concentration reaches about 10–100 parts per trillion (ppt), after which the dose rate increases linearly. • Uranium-232 is considered for use as a tracer in uranium fuels, but could lead to increased radiation dose. • Several stages of the nuclear fuel cycle are modeled with uranium-232 added. • Various concentrations of uranium-232 are examined. • Dose from uranium-232 becomes the primary contributor to dose between 10 and 100 parts per trillion. [ABSTRACT FROM AUTHOR]

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]

Highlights: [•] Modeled the flux trap of the High Flux Isotope Reactor for transcurium production. [•] Identified sources of feedstock losses, primarily due to fission. [•] Identified energies where ratios of fission to (n,γ) absorption are highest. [•] Incorporated resonance absorbers at these energies to reduce fission losses in feedstock. [•] Shielding certain energy bands results in an increase in the efficiency of transcurium production. [Copyright &y& Elsevier]

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]

Abstract: At the University of Cincinnati Nuclear and Radiological Engineering Program, a once moth-balled vintage subcritical reactor facility was reassembled to provide a practical laboratory experience for students in nuclear science. The relative simplicity and accessibility of this facility are some of its most desirable features. Recent work had led to supplementing the experimental aspects of this facility with a three-dimensional, full detail, MCNP model of this reactor. This article emphasizes the latest validations of this simulation model against the most recently available laboratory measurements of neutron flux distributions and subcritical neutron multiplication factors, while illustrating a unique feature of this facility which enables experimentalists to carry out step by step measurements, and the corresponding simulations, while the core is being loaded and unloaded. Furthermore, additional simulations are herein included to evaluate the impact of different moderator and reflector arrangements upon core reactivity, including a few combinations of heavy water and/or beryllium, primarily as a venue for researchers to begin studying potential upgrades or changes to the current design. [Copyright &y& Elsevier]

The assessment of economic viability of a new reactor concept is crucial particularly during the early stages of its concept development. The G4-ECONS methodology provides a standardized top-down estimate of electricity cost and parametric sensitivities, not specifically targeted toward an accurate prediction of the final cost when deployed, but rather seeking an approximation of cost variations relative to other systems. This study presents an analysis of the Integral Molten Salt Reactor (IMSR) concept in comparison with a consistent analysis of an advanced PWR reactor (represented by AP1000). Estimation of levelized unit electricity costs, as well as sensitivity analyses to the discount rate and uranium or SWU prices, are presented using this methodology. [ABSTRACT FROM AUTHOR]

This paper introduces a quick method for improving the accuracy of Monte Carlo simulations by generating one- and two-dimensional cross sections at a user-defined temperature before performing transport calculations. A finite difference method is used to Doppler-broaden cross sections to the desired temperature, and unit-base interpolation is done to generate the probability distributions for double differential two-dimensional thermal moderator cross sections at any arbitrarily user-defined temperature. The accuracy of these methods is tested using a variety of contrived problems. In addition, various benchmarks at elevated temperatures are modeled, and results are compared with benchmark results. The problem-dependent cross sections are observed to produce eigenvalue estimates that are closer to the benchmark results than those without the problem-dependent cross sections. [ABSTRACT FROM AUTHOR]

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]

Abstract: Various arrangements of beryllium loadings to create an internal neutron reflector in the flux trap region of the Oak Ridge National Laboratory’s High Flux Isotope Reactor (HFIR) have been investigated. In particular, the impact upon fuel cycle length has been studied by performing calculations using the HFIR MCNP-based model HFV4.0. This study included examining perturbations in reactivity, flux, and power distribution caused by the various beryllium loadings. The HFIR Cycle 400 core configuration was used as a reference to calculate the impact of beryllium loadings upon cycle length. Three different configurations of beryllium loadings were investigated and compared against the Cycle 400 benchmark calculations; Cases 1 through 3 modeled combinations of 12 and 18 beryllium rods loaded into unused experimental sites. Calculated eigenvalues have shown that potential increases in reactivity between 0.56 and 0.79 dollars are attainable, depending on the various beryllium configurations. These results correspond to possible increases in fuel cycle length ranging between 2.3% and 3.3%. On the basis of their practicality, cost versus benefit, and greater potential for implementation, Cases 2 and 3 (both with 18 beryllium rods) were studied further and are herein reported in greater detail. Neutron flux distributions for Cases 2 and 3 were calculated at the horizontal mid-plane of the flux trap region, which showed no significant changes in the thermal flux magnitude and radial profile in comparison to Cycle 400. Likewise, safety analysis related parameters were contrasted, revealing power increments of up to 2% near the inside edge of the inner fuel element, well below the maximum acceptable value of 9%, a standing guideline employed for experiments at the HFIR. Additionally, the average neutron heat generation rate in beryllium rods and the maximum heat generation rate were evaluated to confirm that the design provides adequate coolant flow inside the rod and around the beryllium targets to carry away any excess heat. Case 2 with 18 beryllium rods was recommended for implementation at the HFIR. In fact, funding has been allocated for this project and beryllium rod fabrication is scheduled. The proposed beryllium loading does not involve permanent design changes and is predicted to yield a meaningful increase in fuel cycle length that could translate into potential annual savings in direct fuel costs of approximately $235,000, or of more than five million dollars over the projected life of this reactor. [Copyright &y& Elsevier]

An analysis of the normal and shear stress distribution on the rake face of an orthogonal cutting tool is presented. A research-licensed version of DYNA3D finite element package was used to model the cutting process based upon mechanistic experiments that provided the empirical parameters of the Johnson–Cook constitutive model. The Johnson–Cook model uses strain, strain rate, and temperature to predict the material’s stress response. The computer model gave predictions for the normal and shear stresses on the rake lace that qualitatively agree with the experimentally determined stress distributions determined by other researchers. [Copyright &y& Elsevier]

• FeCrAl, cladding/channel box halved, required ∼0.6% more 235U average enrichment. • Geometric/enrichment-altered full core models reached 527 effective full power days. • Hybrid design—FeCrAl cladding and SiC/Ni/Cr channel box—matched cycle length. • Thermomechanical performance of a Zircaloy-cladded fuel rod was compared to FeCrAl. The impact of replacing Zircaloy with FeCrAl, a candidate enhanced accident-tolerant fuel cladding material, was evaluated for 10 × 10 boiling water reactor fuel bundles. Results from a series of full-core parametric studies estimated that replacing UO 2 /Zircaloy with UO 2 /FeCrAl would require an average enrichment increase of 0.6% 235U throughout the fuel lattice with the cladding and channel box thicknesses halved and fuel pellet diameter increased. Full-core results indicated that UO 2 /FeCrAl models with these geometric/enrichment specifications matched the base UO 2 /Zircaloy cycle length of 527 effective full power days. Optimization studies of the full-core design established loading and control blade patterns for both Zircaloy and FeCrAl models. A side study was conducted modeling a hybrid fuel bundle consisting of FeCrAl cladding and a SiC/Ni/Cr channel box. By halving the cladding thickness, the enrichment level required was less than that of the Zircaloy base case design after performing loading pattern optimization of the hybrid bundle core. Lastly, the thermomechanical performance of a Zircaloy-cladded fuel rod was compared to that of a FeCrAl system. Results from this analysis show that, if starting from the same fuel-cladding gap thickness, a FeCrAl-clad fuel rod operates with a greater average fuel centerline temperature, comparable axial elongation and radial displacement, and longer time to gap closure compared to a Zircaloy-clad fuel rod. This fuel performance analysis was primarily based on the commercial Kanthal APMT FeCrAl alloy but also used available data for the C35M FeCrAl alloy developed at Oak Ridge National Laboratory. [ABSTRACT FROM AUTHOR]

This paper presents results obtained by reproducing a VVER-1000 lattice physics benchmark using the Serpent2 code. The benchmark data are related to the initial cycles of the Khmelnitskiy nuclear power plant. This study compares results obtained with the Serpent2 code against those calculations published in the benchmark document. Detailed descriptions are provided for two fuel assemblies; one without gadolinium in the fuel pins, and the other with gadolinium. A sensitivity study is performed utilizing different nuclear data libraries, different depletion sequences (including pin by pin burnup) and different number of radial regions within the fuel pins containing gadolinium. The infinite multiplication factor behavior and isotopic concentrations of major isotopes are compared and discussed. It is concluded that Serpent2 provided reasonable results in comparison with the other codes used in the benchmark. [ABSTRACT FROM AUTHOR]

• Gas sparging in molten salt reactors. • Majority of the xenon is found in the circulating gas bubbles. • Species transport implemented for molten salt reactor analysis. Molten salt reactors (MSRs) are a class of next-generation nuclear reactors that have received recent industrial and research interest. A generalized species transport solver was implemented in the Virtual Environment for Reactor Applications (VERA) computing suite to extend this tool to analyze liquid-fueled MSRs. This core simulator has been extended to model the transport of fission product gases into a collection of circulating gas bubbles with the purpose of removing the gases. This paper presents the governing species transport equation, along with various nuclear source terms. Development of the source term for phase migration is discussed, along with a simplified interfacial area tracking method. Finally, a case study on a simplified MSR loop is presented in which modeling parameters were varied to assess their impact on gas removal. The steady state results show that parameters such as bubble diameter, gas injection rate and mass transfer coefficient have a low to moderate effect on the fraction of xenon in the core region. Removal efficiency has the greatest effect on the fraction in the core region. After the pump bowl, bubble diameter has a minor effect on the fraction of xenon in the gas void. These results point out that increasing parameters such as mass transfer coefficient, gas injection rate, and removal efficiency drives the xenon into the circulating gas void, while decreasing bubble diameter also drives xenon into the gas void by increasing interfacial area. [ABSTRACT FROM AUTHOR]