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In this report, building on previous reports issued in FY13 we describe our continued efforts to integrate thermal/hydraulics, neutronics, and structural mechanics modeling codes to perform coupled analysis of a representative fast sodium-cooled reactor core. The focus of the present report is a full core simulation with off-line mesh deformation.
Reactor core simulations require the construction and mesh generation for core models consisting of lattices of fuel and other rods grouped into assemblies, and lattices of assemblies of several types grouped into a core model. A set of tools has been described for generating assembly and core lattice models. Both rectangular and hexagonal lattices are supported. The tools operate in three stages. First, assembly models of various types can be generated by the AssyGen tool, based on input describing the content of unit cells, the arrangement of unit cells in the lattice, and the extent of the lattice and any surrounding material. After generating the assembly model, the model is meshed with the CUBIT mesh generation toolkit, optionally based on a journal file output by AssyGen. After one or more assembly model meshes have been constructed, they are arranged in a core model using the CoreGen tool. The input for CoreGen is similar to that of AssyGen, with assembly models substituted for unit cells. AssyGen and CoreGen also annotate the models with material and volume groupings necessary for specifying materials and boundary conditions required by the analysis. The AssyGen and CoreGen tools are packaged in the open-source MeshKit library for mesh generation; download and build instructions are included in this document.
The new DOE-NE Center of Excellence for Thermal-fluids applications in Nuclear Energy inaugurated in April 2018 considers and researches novel new solution strategies for historically challenging flow issues that still plague the current fleet of deployed Light Water nuclear Reactors (LWRs) as well as predicting various fluid flow and fluid related issues with advanced reactor technologies. Our advanced thermal-fluids research and development approach synergistically combines three natural, though overlapping, length and time scales in a hierarchal multi-scale approach to avoid the temptation and pitfalls of attempting to develop a single solve-all algorithm for physical fluid flow problems that will span nine orders of magnitude in spatial and temporal scales. To demonstrate the multi-scale philosophy of the center we focus on Fluoride Cooled High Temperature Reactors (FHRs), and in particular on the Berkley’s PB-FHR Mark-I design. The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxiliary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid Fluoride salt working fluids - such as Flibe - from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. The PB-FHR reactor relies on a pebble bed approach and pebble bed reactors are in a sense the poster child for multi-scale analysis. The lower length-scale simulator for pebble reactor cores comprises three physics: neutronics (OpenMC), thermal-fluids (Nek5000/NekRS) and fuel performance (BISON). As part of Center of Excellence ongoing research efforts we have developed Cardinal, a new tool platform for lower length-scale simulation. Cardinal tightly couples all three physics and leverages advances in MOOSE such as the MultiApp system and the concept of MOOSE-wrapped Apps. The present report provides an update on the development of Cardinal with an extension of Cardinal to GPUs. We also perform a first-of-a-kind demonstration simulation on Summit representing a 10x capability increase in terms of pebble count for Cardinal.
Here we revisit the diffuse supernova neutrino background in light of recent systematic studies of stellar core collapse that reveal the quantitative impacts of the progenitor conditions on the collapse process. In general, the dependence of the core-collapse neutrino emission on the progenitor is not monotonic in progenitor initial mass, but we show that it can, at first order, be characterized by the core compactness. For the first time, we incorporate the detailed variations in the neutrino emission over the entire mass range 8–100 M ⊙ , based on (i) a long-term simulation of the core collapse of an 8.8 M ⊙ ONeMg core progenitor, (ii) over 100 simulations of iron core collapse to neutron stars, and (iii) half a dozen simulations of core collapse to black holes (the ‘failed channel’). The fraction of massive stars that undergo the failed channel remains uncertain, but in view of recent simulations which reveal high compactness to be conducive to collapse to black holes, we characterize the failed fraction by considering a threshold compactness above which massive stars collapse to black holes and below which the final remnant is a neutron star. We predict that future detections of the diffuse supernova neutrino background may have the power to reveal this threshold compactness, if its value is relatively small as suggested by interpretations of several recent astronomical observations.
We have conducted a series of numerical experiments using spherically symmetric, general relativistic, neutrino radiation hydrodynamics with the code Agile-BOLTZTRAN to examine the effects of modern neutrino opacities on the development of supernova simulations. We test the effects of opacities by removing opacities or by undoing opacity improvements for individual opacities and groups of opacities. We find that improvements to electron capture (EC) on nuclei, namely EC on an ensemble of nuclei using modern nuclear structure models rather than the simpler independent-particle approximation (IPA) for EC on a mean nucleus, plays the most important role during core collapse of all tested neutrino opacities. Low-energy neutrinos emitted by modern nuclear EC preferentially escape during collapse without the energy downscattering on electrons required to enhance neutrino escape and deleptonization for the models with IPA nuclear EC. During shock breakout the primary influence on the emergent neutrinos arises from non-isoenergetic scattering (NIS) on electrons. For the accretion phase, NIS on free nucleons and pair emission by e {sup +} e {sup -} annihilation have the largest impact on the neutrino emission and shock evolution. Other opacities evaluated, including nucleon-nucleon bremsstrahlung and especially neutrino-positron scattering, have little measurable impact on neutrino emission or shock dynamics. Modern treatments of nuclear EC, e {sup +} e {sup -}-annihilation pair emission, and NIS on electrons and free nucleons are critical elements of core-collapse simulations of all dimensionality.
Monte Carlo simulations are the gold standard in particle transport simulations, for shielding and nuclear reactor simulations Full core simulations can require terabytes of memory for tracking pin-resolved (up to 50k pins), nuclide (200 tracked), axial (20-200 zones) reaction rates Domain decomposition, splitting the calculation and the memory storage, among several nodes is challenging as it involves communication of particles across domain boundaries We created a new MOOSE application, MaCaw, to study this problem Algorithms developed for ray tracing in MOOSE were adapted for simulating neutral particle transport Physics and tally routines in OpenMC, dynamically linked, are called from the application
Here, 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.
As the Consortium for Advanced Simulation of Light Water Reactors (CASL) moves forward with more complex multiphysics simulations, there is increased focus on incorporating fuel performance analysis methods. The coupled neutronics/thermal-hydraulics capabilities within the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) have become relatively stable and major advances have been made in analysis efforts, including the simulations of twelve cycles of Watts Bar Nuclear Unit 1 (WBN1) operation. However, VERA-CS approaches for treating fuel pin heat transfer have well-known limitations that could be eliminated through better integration with the BISON-CASL fuel performance code. Several approaches are being taken to improve or replace the VERA-CS fuel models and improve integration with BISON-CASL. Tiamat is being developed to replace the fuel models with a direct coupling of BISON-CASL and the neutronics and coolant heat transfer of VERA-CS. Tiamat will execute in an inline mode that allows for in-memory transfer of VERA-CS data to BISON-CASL. This will provide fuel performance results during core simulation. However, fuel performance typically undergoes an independent analysis using a stand-alone fuel performance code with manually specified input defined from an independent core simulator solution or set of assumptions. This milestone covers initial efforts to facilitate using VERA for core simulation and fuel performance to operate in this mode of execution. This approach will be used for several CASL fuel performance challenges, such as RIAs and DNB. These issues require time-dependent full core results to establish the boundary conditions for the high-fidelity BISON-CASL simulations. The approach will also be used to improve fuel temperature models in VERA-CS, guide the development of Tiamat, and simplify the use of BISON-CASL within VERA for industry users.
The Method of Characteristics (MOC) has seen wide interest in reactor physics because of its accuracy and efficiency in computing lattice physics problems. While most of its use has been in solving 2D problems, there has been recent interest in extending MOC to 3D in order to more accurately calculate 3D power distributions in LWRs. While the method is naturally extensible to 3D, it presents significant computational difficulties. OpenMOC, an open source MOC simulator, has been extended from 2D to 3D capabilities. Simulation results are presented in this study for the full core 3D simulation of the BEAVRS benchmark, adequately resolved in space and angle.
The VERA core simulator, VERA-CS, provides the capability for steady-state reactor core simulation for operating nuclear power plant (PWR) conditions across multiple fuel cycles. This capability parallels that of industry core simulators which are typically licensed by the NRC for core design, core surveillance/monitoring activities, and safety-related calculations that support accident analyses. The licensing basis for these codes for these applications is a rigorous process of validating the software against measured data and quantifying the resulting biases and uncertainties. This document provides an initial plan for performing similar activities with VERA-CS, in a manner consistent with benchmarking activities performed in the nuclear industry.
The boiling water reactor (BWR) loading pattern (LP) optimization system FINELOAD-3 (FL3), has been applied for actual BWR reload core design for over fifteen years in TEPCO Systems Corporation (TEPSYS) and for over ten years in several European utilities. FL3 works in combination with the advanced nodal core simulator SIMULATE-3 (S3) or other core simulators. Recently, the channel bow constraint option has been added in FL3. In the BWR reload core design work, the channel bow management is very important in order to avoid the interference between a channel and a control blade and/or to minimize the effect on critical power ratio during the operation cycle. The channel bow constraint option is applied to eliminate channel bow violation against design target. (authors)
The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time step of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.
The natural circulation of a single phase fluid in a scale model of a pressurized water reactor system during a postulated grade core accident is analyzed. The fluids utilized were water and SF6. The design of the reactor model and the similitude requirements are described. Four LDA tests were conducted: water with 28 kW of heat in the simulated core, with and without the participation of simulated steam generators; water with 28 kW of heat in the simulated core, with the participation of simulated steam generators and with cold upflow of 12 lbm/min from the lower plenum; and SF6 with 0.9 kW of heat in the simulated core and without the participation of the simulated steam generators. For the water tests, the velocity of the water in the center of the core increases with vertical height and continues to increase in the upper plenum. For SF6, it is observed that the velocities are an order of magnitude higher than those of water; however, the velocity patterns are similar.
The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3-D whole core transport code being developed for the CASL toolset, Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include (1) a subgroup method for resonance selfshielding and (2) a whole-core transport solver with a 2-D/1-D synthesis method. The MPACT code requires a cross section library to support all the MPACT core simulation capabilities which would be the most influencing component for simulation accuracy.
The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3-D whole core transport code being developed for the Virtual Environment for Reactor Analysis (VERA) CASL toolset. Key characteristics of the MPACT code include (1) a subgroup method for resonance selfshielding and (2) a whole-core transport solver with a 2-D/1-D synthesis method. The MPACT code requires a cross section library to support all of its core simulation capabilities; a cross section library would be MPACT’s most influencing component for simulation accuracy. Due to the limitation of computing capacity even in high performance computing, the cross section library must be developed to enhance computational efficiency in memory and computing time without losing accuracy and generality. This requires a very coarse energy group structure with about 50 groups. Multigroup (MG) neutron cross section libraries for the CASL VERA-CS neutronics simulator MPACT have been created by using the AMPX/SCALE code package developed at Oak Ridge National Laboratory (ORNL). A new 51-group structure was developed for efficient simulation to be applicable for both pressurized water reactor (PWR) and boiling water reactor (BWR) simulations. New v4.3m1 MPACT 51-group library with the ENDF/B- VII.1 nuclear data were developed for MPACT. Since SCALE KENO has some issues in the probability table and has a low cutoff energy for the free gas model, this issue has been corrected. This study focused on the development of the ENDF/B-VII.1 v4.3m1 MPACT MG library, including verification and validation. This document includes the library generation methodology and procedure, the verification procedure, and benchmark results compared to continuous energy Monte Carlo results using newly developed cross section libraries.
We construct a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is based on the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by using relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to {approx} 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect to the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. It is also interesting that the root mean square of the mass number is not very different from the average mass number, since the former is important for the evaluation of coherent scattering rates on nuclei but has been unavailable so far.
CASL (the Consortium for Advanced Simulation of LWRs) is investing in the development of the multi-dimensional Bison fuel performance code for both transient and nominal performance analysis. The MPACT whole-core neutronics capability has been integrated with the Cobra-TF sub-channel solver to provide a neutron-transport based core-simulator with substantially improved fidelity over traditional diffusion-based core simulators. The Tiamat code is being developed to provide an integrated tool for predicting pellet-clad interaction and improving the high-fidelity core simulator.