PBMR 400 Coupled Code Benchmark: Challenges and Successes with NEM-THERMIX

Presented are preliminary results of a sensitivity study on mesh sizing and diffusion coefficient on a the core multiplication factor in a pebble bed reactor core simulation. Nonphysical behavior is observed for ranges of mesh size and diffusion coefficients. A guideline was established for avoiding this problem in subsequent models

SHARP/PRONGHORN Interoperability: Mesh Generation

Progress toward collaboration between the SHARP and MOOSE computational frameworks has been demonstrated through sharing of mesh generation and ensuring mesh compatibility of both tools with MeshKit. MeshKit was used to build a three-dimensional, full-core very high temperature reactor (VHTR) reactor geometry with 120-degree symmetry, which was used to solve a neutron diffusion critical eigenvalue problem in PRONGHORN. PRONGHORN is an application of MOOSE that is capable of solving coupled neutron diffusion, heat conduction, and homogenized flow problems. The results were compared to a solution found on a 120-degree, reflected, three-dimensional VHTR mesh geometry generated by PRONGHORN. The ability to exchange compatible mesh geometries between the two codes is instrumental for future collaboration and interoperability. The results were found to be in good agreement between the two meshes, thus demonstrating the compatibility of the SHARP and MOOSE frameworks. This outcome makes future collaboration possible

Improved Prediction of the Temperature Feedback in TRISO-Fueled Reactors

The Doppler feedback mechanism is a major contributor to the passive safety of gas-cooled, graphite-moderated high temperature reactors that use fuel based on Tristructural-Isotropic coated particles. It follows that the correct prediction of the magnitude and time-dependence of this feedback effect is essential to the conduct of safety analyses for these reactors. We present a fuel conduction model for obtaining better estimates of the temperature feedback during moderate and fast transients. The fuel model has been incorporated in the CYNOD-THERMIX-KONVEK suite of coupled codes as a single TRISO particle within each calculation cell. The heat generation rate is scaled down from the neutronic solution and a Dirichlet boundary condition is imposed as the bulk graphite temperature from the thermal-hydraulic solution. This simplified approach yields similar results to those obtained with more complex methods, requiring multi-TRISO calculations within one control volume, but with much less computational effort. An analysis of the hypothetical total control ejection in the PBMR-400 design verifies the performance of the code during fast transients. In addition, the analysis of the earthquake-initiated event in the PBMR-400 design verifies the performance of the code during slow transients. These events clearly depict the improvement in the predictions of the fuel temperature, and consequently, of the power escalations. In addition, a brief study of the potential effects of particle layer de-bonding on the transient behavior of high temperature reactors is included. Although the formation of a gap occurs under special conditions its consequences on the dynamic behavior of the reactor should be analyzed. The presence of a gap in the fuel can cause some unusual reactor behavior during fast transients, but still the reactor shuts down due to the strong feedback effects

A Newton solution for the superhomogenization method: The PJFNK-SPH

RÉSUMÉ: This work presents two novel topics regarding the Superhomogenization method: 1) the formalism for the implementation of the method with the linear Boltzmann Transport Equation, and 2) a Newton algorithm for the solution of the nonlinear problem that arises from the method. These new ideas have been implemented in a continuous finite element discretization in the MAMMOTH reactor physics application. The traditional solution strategy for this nonlinear problem uses a Picard, fixed-point iterative process whereas the new implementation relies on MOOSE's Preconditioned Jacobian-Free Newton Krylov method to allow for a direct solution. The PJFNK-SPH can converge problems that were either intractable or very difficult to converge with the traditional iterative approach, including geometries with reflectors and vacuum boundary conditions. This is partly due to the underlying Scalable Nonlinear Equations Solvers in PETSc, which are integral to MOOSE and offer Newton damping, line search and trust region methods. The PJFNK-SPH has been implemented and tested for various discretizations of the transport equation included in the Rattlesnake transport solver. Speedups of five times for diffusion and ten to fifteen times for transport were obtained when compared to the traditional Picard approach. The three test problems cover a wide range of applications including a standard Pressurized Water Reactor lattice with control rods, a Transient Reactor Test facility control rod supercell and a prototype fast-thermal reactor. The reference solutions and initial cross sections were obtained from the Serpent 2 Monte Carlo code. The SPH-corrected cross sections yield eigenvalues that are near exact, relative to reference solutions, for reflected geometries, even with reflector regions. In geometries with vacuum boundary conditions the accuracy is problem dependent and solutions can be within a few to a few hundred pcm. The root mean-square error in the power distribution is below 0.8% of the reference Monte Carlo. There is little benefit from SPH-corrected transport in typical scoping calculations, but for more detailed analyses it can yield superior convergence of the solution in some of the test problems. This PJFNK-SPH approach is currently being used in the modeling of the Transient Test Reactor at Idaho National Laboratory, where full reactor core SPH-corrected cross sections are employed to reduce the homogenization errors in transient or multi-physics calculations. This base implementation of the PJFNK-SPH provides an extremely robust solver and a springboard to further improve the Superhomogenization method in order to better preserve neutron currents, one of the primary deficiencies "of the method. (C) 2017 The Authors. Published by Elsevier Ltd

Preservation of kinetics parameters generated by Monte Carlo calculations in two-step deterministic calculations

The generation of accurate kinetic parameters such as mean generation time Λ and effective delayed neutron fraction βeff via Monte Carlo codes is established. Employing these in downstream deterministic codes warrants another step to ensure no additional error is introduced by the low-order transport operator when computing forward and adjoint fluxes for bilinear weighting of these parameters. Another complexity stems from applying superhomogenization (SPH) equivalence in non-fundamental mode approximations, where reference and low-order calculations rely on a 3D full core model. In these cases, SPH factors can optionally be computed for only part of the geometry while preserving reaction rates and K-effective, but the impact of such approximations on kinetics parameters has not been thoroughly studied. This paper aims at studying the preservation of bilinearly-weighted quantities in the Serpent–Griffin calculation procedure. Diffusion and transport evaluations of IPEN/MB-01, Godiva, and Flattop were carried out with the Griffin reactor physics code, testing available modeling options using Serpent-generated multigroup cross sections and equivalence data. Verifying Griffin against Serpent indicates sensitivities to multigroup energy grid selection and regional application of SPH equivalence, introducing significant errors; these were demonstrated to be reduced through the use of a transport method together with a finer energy grid

Physics-based multiscale coupling for full core nuclear reactor simulation

Numerical simulation of nuclear reactors is a key technology in the quest for improvements in efficiency, safety, and reliability of both existing and future reactor designs. Historically, simulation of an entire reactor was accomplished by linking together multiple existing codes that each simulated a subset of the relevant multiphysics phenomena. Recent advances in the MOOSE (Multiphysics Object Oriented Simulation Environment) framework have enabled a new approach: multiple domain-specific applications, all built on the same software framework, are efficiently linked to create a cohesive application. This is accomplished with a flexible coupling capability that allows for a variety of different data exchanges to occur simultaneously on high performance parallel computational hardware. Examples based on the KAIST-3A benchmark core, as well as a simplified Westinghouse AP-1000 configuration, demonstrate the power of this new framework for tackling—in a coupled, multiscale manner—crucial reactor phenomena such as CRUD-induced power shift and fuel shuffle.Massachusetts Institute of Technology. Department of Nuclear Science and EngineeringIdaho National Laboratory (Contract DE-AC07-05ID14517

SHARP/PRONGHORN Interoperability: Mesh Generation

Progress toward collaboration between the SHARP and MOOSE computational frameworks has been demonstrated through sharing of mesh generation and ensuring mesh compatibility of both tools with MeshKit. MeshKit was used to build a three-dimensional, full-core very high temperature reactor (VHTR) reactor geometry with 120-degree symmetry, which was used to solve a neutron diffusion critical eigenvalue problem in PRONGHORN. PRONGHORN is an application of MOOSE that is capable of solving coupled neutron diffusion, heat conduction, and homogenized flow problems. The results were compared to a solution found on a 120-degree, reflected, three-dimensional VHTR mesh geometry generated by PRONGHORN. The ability to exchange compatible mesh geometries between the two codes is instrumental for future collaboration and interoperability. The results were found to be in good agreement between the two meshes, thus demonstrating the compatibility of the SHARP and MOOSE frameworks. This outcome makes future collaboration possible

Status Report on the Modeling of TRISO Energy Deposition, Time-Dependent Temperature Field and Doppler Feedback

The Doppler feedback mechanism is a major contributor to the passive safety of gas-cooled, graphite-moderated high temperature reactors that use fuel based on Tristructural-Isotropic coated particles. It follows that the correct prediction of the magnitude and time-dependence of this feedback effect is essential to the conduct of safety analyses for these reactors. We present a fuel conduction model for obtaining better estimates of the temperature feedback during moderate and fast transients. The fuel model has been incorporated in the CYNOD-THERMIX-KONVEK suite of coupled codes as a single TRISO particle within each calculation cell. The heat generation rate is scaled down from the neutronic solution and a Dirichlet boundary condition is imposed as the bulk graphite temperature from the thermal-hydraulic solution. This simplified approach yields similar results to those obtained with more complex methods, requiring multi-TRISO calculations within one control volume, but with much less computational effort. We provide an analysis of the hypothetical total control ejection event in the PBMR-400 design that clearly depicts the improvement in the predictions of the fuel temperature

Investigation of Supercells for Preparation of Homogenized Cross Sections for Prismatic Deep Burn VHTR Calculations

In Very High Temperature Reactors (VHTRs), the long mean-free-path and large migration area of neutrons leads to spectral influences between fuel and reflector zones over long distances. This presents significant challenges to the validity of the classic two-step approach of cross section preparation wherein infinite lattice transport calculations are performed on relatively small physical domains (e.g. single assembly) in order to compute homogenized few-group cross sections for whole core analysis. Effects of the inner and outer reflectors render infinite lattice calculations on a single peripheral fuel assembly quite inaccurate, while burnable poison locations affect neighboring assemblies as well. Use of transuranics-only (TRU) Deep Burn fuel in a prismatic VHTR (DB-VHTR) presents the additional challenge of producing vastly different neutron spectra between fresh and burned fuel. ?This paper presents the progress in seeking a systematic method for generation of diffusion theory data in optically thin, multiply-heterogeneous reactors in a production context. A companion paper presents the underlying theory and systematic development of the methodology. In the context of this work, a supercell refers to an extended domain surrounding a region of interest. The extended domain is used to decouple the solution in this region of interest from the boundary conditions of the problem. This is an extension of the concept of color set, which was demonstrated to work very well for light water reactors (LWR). However, a half-assembly in an LWR presents a greater neutronic depth (in mean free paths) than in a VHTR. ??In order to make the supercell calculations more computationally manageable, an initial calculation is performed on a small domain and individual cells (individual compacts or coolant channels with graphite surrounding) are homogenized then used in the supercell calculations. This allows faster computation on the larger domain while retaining the overall hexagonal geometry of the fuel blocks. An application of this supercell concept using the DRAGON transport code is evaluated in this work for its effectiveness and practicality as part of an overall cross section preparation scheme for prismatic DB-VHTR reactors. The sizes of supercells for a peripheral fuel block are evaluated using independence from boundary conditions as an indicator