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AbstractThe goal of our research is to build upon the capability of RELAP5-3D to model molten lead systems. Molten lead has several potential uses in future advanced reactors, like the lead fast reactor or fusion reactors that utilize dual-coolant lead lithium blankets. This potential for use in future generations of reactors highlights the necessity of developing molten lead models to ensure that they can accurately predict thermohydraulic behavior. We have developed a RELAP5-3D model of the Lobo Lead Loop facility located at the University of New Mexico to verify the accuracy of RELAP5-3D via comparison to existing computational fluid dynamics results and analytical calculations.It was found that RELAP5-3D accurately calculated radiative heat transfer (within <1%) when compared to theoretical calculations. In addition, pressure drop calculations done in RELAP5-3D demonstrated reasonable agreement within 20 kPa, mostly within ~7% to 15%, when compared to the computational fluid dynamics model of the facility developed by the University of New Mexico, and captured the dependence of pressure drop on flow velocity accurately.Finally, a hypothetical loss-of-flow transient was imposed on the RELAP5-3D model to determine the feasibility of performing a similar experiment with the Lobo Lead Loop. It was found that such an experiment could be possible, as the RELAP5-3D model indicated that the temperatures of the fluid would not exceed the limiting temperatures of the structure (1658 K) nor the maximum temperature of the electromagnetic pump inlet (823 K). Although there are no experimental data to begin validation, the model will be readily available for future validation studies when the experimental data are generated, especially as the model continues to evolve over time. The results so far demonstrate a promising first step in the verification/validation of the RELAP5-3D model of the Lobo Lead Loop.The highlights from our research are as follows:1. The Lobo Lead Loop facility at the University of New Mexico is a good candidate for molten lead system code validation.2. The Lobo Lead Loop currently has extensive pressure drop results from a high-fidelity computational fluid dynamics model, which offers the opportunity for code-to-code verification of pressure drop in RELAP5-3D.3. A RELAP5-3D model of the Lobo Lead Loop has been developed to begin verification studies and to prepare for potential validation studies.4. Development of the model will continue throughout the future to prepare for potential validation studies. [ABSTRACT FROM AUTHOR]

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]

A new computer code for performing BWR in-core and out-of-core fuel cycle optimization for multiple cycles simultaneously has been developed. Parallel simulated annealing (PSA) is used to optimize the new fuel inventory and placement of new and reload fuel for each cycle considered. Several algorithm improvements were implemented and evaluated. The most significant of these are variable sampling probabilities and sampling new fuel types from an ordered array. A heuristic control rod pattern (CRP) search algorithm was also implemented, which is useful for single CRP determinations, however, this feature requires significant computational resources and is currently not practical for use in a full multi-cycle optimization. The PSA algorithm was demonstrated to be capable of significant objective function reduction and finding candidate loading patterns without constraint violations. The use of variable sampling probabilities was shown to reduce runtime while producing better results compared to using constant sampling probabilities. Sampling new fuel types from an ordered array was shown to have a mixed effect compared to random new fuel type sampling, whereby using both random and ordered sampling produced better results but required longer runtimes. [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]

Our research aims to build a multiphysics framework for transient analysis of the Dual-Coolant Lead-Lithium (DCLL) blanket design of the proposed Fusion Nuclear Science Facility (FNSF) for all Anticipated Operational Occurrences (AOOs) using Reduced Order Models (ROMs). This paper presents an assessment of using RELAP5-3D for transient thermal-hydraulic blanket analysis of the DCLL design. Preliminary validation studies for the RELAP5-3D properties were conducted using a representative vertical flow loop model and a comparison study using experimental data from thermal convection corrosion loops at ORNL. These studies inspected the thermal hydraulic response of Lead-Lithium Eutectic (PbLi) systems within the RELAP5-3D code; mainly pressure drop and heat transfer, as compared to analytical calculations and experimental data. The Magnetohydrodynamic (MHD) pressure drop effect for liquid metal under the effects of a uniform magnetic field was implemented into our models using an equivalent forms loss coefficient based on the phasic momentum equations. We verified our implementation of MHD pressure drop within RELAP5-3D using a systematic approach consisting of two studies. Models were developed to perform comparison studies using experimental data from the Argonne Liquid metal EXperiment (ALEX) facility and the Magnetohydrodynamic PbLi Experiment (MaPLE) facility. Using the developed validation basis, we built a simplified DCLL channel model within RELAP5-3D, inspected MHD pressure drop and performed transient temperature analysis. The DCLL channel model used radial heating data from the International Thermonuclear Experimental Reactor (ITER) which was normalized to the FNSF design. A representative startup transient was performed using a relative power curve based on current PWR power ramp limits. This study demonstrated that RELAP5-3D ROMs are capable of accurate transient analysis of the DCLL blanket. We have since utilized MCNP to develop heating profiles based on the FNSF design and are working toward implementing them into our DCLL model. [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]

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]

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]

A numerical framework for modeling depletion and mass transport in liquid-fueled molten salt reactions is presented based on exponential time differencing. The solution method involves using the finite volume method to transform the system of partial differential equations (PDEs) into a much larger system of ordinary differential equations. The key part of this method involves solving for the exponential of a matrix. We explore six different algorithms to compute the exponential in a series of progression problems that explore physical transport phenomena in molten salt reactors. This framework shows good results for solving linear parabolic PDEs with each of the six matrix exponential algorithms. For large problems, the series solvers such as Padé and Taylor have large run times, which can be mitigated by using the Krylov subspace. [ABSTRACT FROM AUTHOR]

This paper presents the development of a lattice physics-to-core simulator two-step procedure for the rapid analysis of the Advanced High Temperature Reactor (AHTR). Lattice physics, reflector, and control blade models were developed from which cross-section libraries could be generated for a nodal core simulator. Few-group structures for the core simulator were also generated to account for the neutronic characteristics of AHTR. After developing the AHTR two-step procedure, cross-section libraries were generated using the SERPENT continuous-energy Monte Carlo code. These libraries were then used in the core simulator NESTLE to perform full-core calculations, which were in turn benchmarked against reference SERPENT full-core models. Benchmarking results showed reasonable accuracy of the developed two-step procedure but revealed an inherent inadequacy in the one-dimensional radial reflector model and showed a likely need for a greater number of energy groups than were used in this study. [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]

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: 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]

• We assessed the hydrogen migration and redistribution model in the BISON code for component-level tritium transport. • Overall, the capabilities of the BISON code for component level modeling of tritium transport are promising. • A calibration method resulted in improved BISON predictions. We assessed the hydrogen migration and redistribution model implemented in the BISON fuel performance code for component-level tritium transport applications in fusion and fission example scenarios. We developed BISON models of an ITER heat exchanger, light water reactor fuel cladding, and a fluoride salt cooled high temperature reactor heat exchanger. Reasonable agreement between reported values in reference studies and BISON predictions demonstrated the ability of the BISON models to predict tritium transport behavior through steel materials. Next, sensitivity and uncertainty analysis was used to understand key parameter sensitivities within the BISON model for evaluation of tritium migration and redistribution.. The sensitivity analysis showed that the diffusivity activation energy was the most important parameter and that the heat of transport was the least important parameter for tritium migration in steels. Overall, the capabilities of the BISON code for component level modeling of tritium transport are promising. [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]

• 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]

• CASL VERA represents a new tool for fuel loading pattern optimization for SMR simulations. • The single fuel assembly calculations provided by the CASL VERA are in a good agreement against the Polaris or Serpent calculation. • The 3D full core calculations with water reflector are in a good agreement against Serpent calculation code. • The 3D full core calculations with stainless steel reflector are in a worse agreement with Serpent calculation code. The optimal stainless steel fraction to het reasonably good results against the Serpent code, was set up to 90%. • Stainless steel reflector extends core operation, moreover it balances the power generation in the SMR NuScale core. Three-dimensional (3D) full-core calculations are an integral part of fuel reload design for today's light water reactors (LWRs). The current approaches are typically based on the nodal diffusion codes that calculate criticality state points and power distributions as a function of burnup. The CASL VERA (Consortium for Advanced Simulation of Light Water Reactors, Virtual Environment for Reactor Applications) code represents one of the latest advancements in 3D full-core calculation analysis based on transport theory methods employing the Method of Characteristics (MOC) and coupled multi-physics. In this article, a publicly released version of the NuScale reactor core is analysed with the VERA code (version 3.9 and 4.1) and contrasted against some static Serpent and Polaris based simulations. The analysis shows an excellent agreement for the lattice-level calculations as well as with some of the 3D full-core models. However, larger deviations were found in cases with heavy reflector models, whereby the reflector composition was found to impact the differences between the VERA and Serpent results. In this analysis, it was determined that greater than 90% stainless steel content in the heavy reflector leads to higher deviations between the VERA results and the Serpent results. The burnup calculations showed that the presence of the heavy reflector extends the cycle length and also leads to a flatter power distribution in the core, which can generally be interpreted as more efficient. [ABSTRACT FROM AUTHOR]

• 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]

• State-of-the-art literature review of multiphysics modeling current status and needs. • Separate- and integral-effects experiments for the validation of the multiphysics analysis are needed for particular components of interest for the DCLL blanket. • Robust sensitivity and uncertainty (S/U) analysis must be integrated with the facility design to identify the significance of influencing parameters involved in tritium transport, along with quantification of their uncertainties towards output parameters. The development and demonstration of fusion power plant facilities require advancement of multiphysics modeling capabilities. These capabilities provide an integrated platform for design and analysis of a fusion engineering demonstration power plant. We present a state-of-the-art literature review of the status and needs of multiphysics modeling, including neutronics analysis, and thermal hydraulic/mechanical simulation tools in support of fusion powerplant system design and integration. In this paper we use the Fusion Nuclear Science Facility (FNSF) concept as an example system to assess future needs for advancement of multiphysics modeling capabilities. Specifically, we discuss the latest progress on R&D missions in the literature which are specific to the neutronics and thermal hydraulics design of the breeding blanket concept chosen for the FNSF, the Dual Coolant Lead Lithium (DCLL) blanket. We identify future R&D needs to be addressed through the development of multiphysics modeling capabilities that tie fusion neutronics to tritium breeding ratio (TBR), neutron wall loading (NWL), material damage and transmutation, shutdown dose, and safety assessments. The capabilities will support the design and licensing of a fusion engineering demonstration facility as well as a future engineering demonstration plant. We conclude that more integrated and prototypic tritium experiments must be developed to establish the database needed for the verification and validation of tritium transport simulations. Robust sensitivity and uncertainty (S/U) analysis must be integrated with the facility design to identify the significance of influencing parameters involved in tritium transport, along with quantification of their uncertainties towards output parameters. This enables design optimization of the DCLL blanket during the R&D phase for a fusion engineering demonstration facility. Additional separate- and integral-effects experiments for the validation of the multiphysics analysis is needed for particular components of interest for the DCLL blanket, such as the flow channel inserts (FCIs), along with multi-fidelity simulations which incorporate the magnetohydrodynamic (MHD) effects, and coupled analysis techniques that can capture the interactions between the thermal hydraulic effects and the tritium transport behavior. Simultaneously, the systematic S/U analysis for the multiphysics modeling framework must prove the reliability of the model to help define the technical gaps that need to be addressed by subsequent R&D activities for a fusion engineering demonstration facility. Lastly, S/U analysis of the uncertainties in design parameters for the evaluation of steady and transient safety analysis would be the final requirement that will elucidate the importance of these design parameters, as well as the quantification of their uncertainties on several identified key Figures of Merit (FoM) applicable to the safety analysis to provide guidance for the final design optimization of the DCLL. [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]