Search

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]

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]

The High Flux Isotope Reactor (HFIR) located at the Oak Ridge National Laboratory utilizes a large annular beryllium reflector that is subdivided into three concentric regions and encompasses the compact reactor core. Nuclear transmutations caused by neutron activation occur in the beryllium reflector regions, which leads to unwanted neutron-absorbing and radiation-emitting isotopes. During the past year, two topics related to the HFIR beryllium reflector were reviewed. The first topic included studying the neutron poison (³He and 6Li) buildup in the reflector regions and its effect on beginning-of-cycle reactivity. A new methodology was developed to predict the reactivity impact and estimated symmetrical critical control element positions as a function of outage time between cycles due to ³He buildup and was shown to be in better agreement with actual symmetrical critical control element position data than the current methodology. The second topic included studying the composition of the beryllium reflector regions at discharge and during postdischarge decay to assess the viability of transporting, storing, and ultimately disposing of the reflector regions currently stored in the spent-fuel pool. The post-irradiation curie inventories were used to determine whether, for disposal purposes, the reflector regions are discharged as transuranic (TRU) waste or become TRU waste during the decay period and to determine the nuclear hazard category, which may affect the controls invoked for transportation and temporary storage. Two of the reflector regions were determined to be TRU waste at discharge, and the other region was determined to become TRU waste <2 yr after being discharged due to irradiation of the initial uranium impurity content (0.0044 wt% uranium). It was also concluded that all three of the reflector regions could be classified as nuclear hazard category 3 (potential for localized consequences only). [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]

Approximately one year before a single UO2 fuel pellet is pressed and fresh fuel assemblies manufactured, design calculations are performed to effectively guarantee the performance of a light water reactor (LWR) core for an upcoming fuel cycle. Energy requirements must be fulfilled in conjunction with all other reactivity, thermal, and operational limits. Furthermore, several months prior to a reactor startup, the proposed core must be licensed by the regulating authority. Therefore, all built-in design conservatisms (i.e., target margins) established a priori must satisfy all safety, operational, and regulatory constraints, a posteriori. The magnitude of target margins directly impacts cycle energy efficiency, which is why this design cushion is often referred to as the "cost of margin" because it ultimately affects the cost per generated kilowatt-hour by a LWR. This article illustrates the modern role of nuclear fuel management optimization in the LWR core reload design process, highlighting some of the history and recent advancements in the field, particularly, in the area of pin-by-pin optimization. Also, important limitations are highlighted to help define the new level of sophistication which the field must conquer for designers to ultimately be able to optimize LWRs "one fuel pin at a time." [ABSTRACT FROM AUTHOR]

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]

Abstract: The purpose of this study is to validate a Monte Carlo based depletion methodology by comparing calculated post-irradiation uranium isotopic compositions in the fuel elements of the High Flux Isotope Reactor (HFIR) core to values measured using uranium mass-spectrographic analysis. Three fuel plates were analyzed: two from the outer fuel element (OFE) and one from the inner fuel element (IFE). Fuel plates O-111-8, O-350-I, and I-417-24 from outer fuel elements 5-O and 21-O and inner fuel element 49-I, respectively, were selected for examination. Fuel elements 5-O, 21-O, and 49-I were loaded into HFIR during cycles 4, 16, and 35, respectively (mid to late 1960s). Approximately one year after each of these elements were irradiated, they were transferred to the High Radiation Level Examination Laboratory (HRLEL) where samples from these fuel plates were sectioned and examined via uranium mass-spectrographic analysis. The isotopic composition of each of the samples was used to determine the atomic percent of the uranium isotopes. A Monte Carlo based depletion computer program, ALEPH, which couples the MCNP and ORIGEN codes, was utilized to calculate the nuclide inventory at the end-of-cycle (EOC). A current ALEPH/MCNP input for HFIR fuel cycle 400 was modified to replicate cycles 4, 16, and 35. The control element withdrawal curves and flux trap loadings were revised, as well as the radial zone boundaries and nuclide concentrations in the MCNP model. The calculated EOC uranium isotopic compositions for the analyzed plates were found to be in good agreement with measurements, which reveals that ALEPH/MCNP can accurately calculate burn-up dependent uranium isotopic concentrations for the HFIR core. The spatial power distribution in HFIR changes significantly as irradiation time increases due to control element movement. Accurate calculation of the end-of-life uranium isotopic inventory is a good indicator that the power distribution variation as a function of space and time is accurately calculated, i.e. an integral check. Hence, the time dependent heat generation source terms needed for reactor core thermal hydraulic analysis, if derived from this methodology, have been shown to be accurate for highly enriched uranium (HEU) fuel. [Copyright &y& Elsevier]

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]

• Exact 3D full core burnup calculation via stochastic codes is complicated and challenging problem. • Two step approach is currently widely used for full core neutronic calculation. • Microdepletion method can be implemented in the two step approach and it can evaluate burnup transmutation of the fuel in the core with limited calculation requirements. • Analysis of Plutonium quality after first year of operation of VVER-1000 nuclear reactor was conducted • Spent nuclear fuel contains around 500 kg of Plutonium after one year of operation. • The Plutonium amount with Pu-239 fraction higher than 80% is less than 8 Kg after irradiation process. Detailed three-dimensional (3D) spent nuclear fuel (SNF) isotopic composition analysis using a full-core 3D nodal simulator is not a widespread practice within production-level nuclear analysis fuel management codes. A 3D full-core pin-wise detailed isotopic analysis can be very time consuming, so more often than not, reduced geometries are used, either based on two-dimensional (2D) lattice-physics, single fuel pins, or point-reactor (0D) estimations. In fact, the typical macroscopic cross section based codes do not usually have the ability to evaluate the full detailed composition of SNF, but instead may track only a few isotopes in spatially homogenized regions. Accordingly, a microscopic cross section based depletion sequence was revisited within the otherwise macroscopic-based NESTLE code. To illustrate this capability, the plutonium content of SNF within a VVER-1000 benchmark was calculated and its non-proliferation risk was estimated. The total plutonium masses and 239Pu fractions were calculated on a 3D node-by-node basis using an established VVER-1000 NESTLE benchmark model. The total plutonium mass after the first cycle is estimated at 502.6 kg, and it increased to 523.4 kg after 90 days of decay. The mass of plutonium with a 239Pu content greater than 80% at the end of cycle is about 7.8 kg, while all remaining content shows to have a 239Pu fraction below 87%. The plutonium content with a 239Pu fraction greater than 80% after 20 days of the end of cycle is less than 7.9 kg. This article highlights a feature of the NESTLE code that can help make isotopic assessments radially and axially for fuel assemblies located in different regions of the core, thus accounting for the varying 3D conditions and flux distributions as a function of exposure. [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]

• Benchmark study of a VVER 1000 core simulation of hot zero power (HZP) tests. • The Serpent-NESTLE combination is capable of providing high quality simulations. • The Serpent-NESTLE combination and MCNP, both, provided reasonable results. This article presents a benchmark study of a VVER 1000 core simulation of hot zero power (HZP) tests. Measured experimental data from the Temelin Unit 2 nuclear power plant's operation test during its first reactor startup were compared using MCNP as well as the combination of Serpent + NESTLE lattice-to-diffusion nodal simulation. Results from full-core 3D models included multiplication factor and moderator temperature coefficients (MTC) that were compared to experimental data, which revealed a very good agreement with criticality states during HZP tests, and a reasonable agreement for MTC. Therefore, it is concluded that the Serpent-NESTLE combination and MCNP, both, provided reasonable results for this benchmark. The Serpent-NESTLE code combination was used for the first time in this study and has shown to be capable of providing high quality simulations in an efficient fashion. To encourage future benchmarks by other researchers, this article makes a special effort to document the core design information and details that would facilitate such efforts. [ABSTRACT FROM AUTHOR]