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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 1,441 records · Page 80
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Particle Mass Flow Control for High-Temperature Concentrating Solar Receivers

This report summarizes the results of a two-year project funded by the U.S. Department of Energy's Solar Energy Technologies Office (SuNLaMP 1506) to evaluate the performance of high-temperature (>700 °C) particle receivers for concentrating solar power (see Appendix A for project information). In the first year, novel particle release patterns were designed and tested to increase the effective solar absorptance of the particle curtain. Modeling results showed that increasing the magnitude and frequency of different wave-like patterns increased the effective absorptance and thermal efficiency by several percentage points, depending on the mass flow rate. Tests showed that triangular-wave, square-wave, and parallel-curtain particle release patterns could be implemented and maintained at flow rates of ~10 kg/s/m. The second year of the project focused on the development and testing of particle mass-flow control and measurement methods. An automated slide gate controlled by the outlet temperature of the particles was designed and tested. Testing demonstrated that the resolution accuracy of the slide-gate positioning was less than ~1 mm, and the speed of the slide gate enabled rapid adjustments to accommodate changes in the irradiance to maintain a desired outlet temperature range. Different in-situ particle mass-flow measurement techniques were investigated, and two were tested. The in-situ microwave sensor was found to be unreliable and sensitive to variations in particle flow patterns. However, the in-situ weigh hopper using load cells was found to provide reliable and repeatable measurements of real-time in-situ particle mass flow. On-sun tests were performed to determine the thermal efficiency of the receiver as a function of mass flow rate, particle temperature, and irradiance. Models of the tests were also developed and compared to the tests.

14 SOLAR ENERGY↗
DOE OSTI2018

A Low Order Mimetic Finite Difference Method for Resistive Magnetohydrodynamics in 2D

We developed a low order Mimetic Finite Difference (MFD) discretization for the equations of magnetohydrodynamics (MHD) in two space dimensions. These equations describe the evolution of the electric and magnetic fields in the presence of prescribed velocity field.The method is designed to work on general polygonal meshes and preserves the divergence-free condition on the magnetic field. The electric field is discretized at the vertices/nodes and the magnetic field uses edge-based discretization. The method reconstructs the magnetic field to extract nodal values necessary to approximate some terms present in Ohm’s law. We test the robustness of our numerical scheme on three different types of meshes: with triangular elements, quadrilateral and unstructured polyhedrons obtain from a Voronoi tesselation. Analysis of the convergence for each of the aforementioned mesh types is presented. We finish with a test problem that shows the method is capable of modelling magnetic reconnection.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2018

PROTEUS-NODAL User Manual

The PROTEUS-NODAL code is being developed as a three-dimensional (3D) neutron transport code based on homogenized geometry models of nuclear reactors. The code at present includes two solution options: diffusion theory and a nodal simplified P 3 (SP 3 ). The diffusion solution option is based upon the DIF3D-VARIANT PN transport methodology where the transport options have not been implemented due to the desire to create a SN based solution methodology. The diffusion solver works for Cartesian, hexagonal, and triangular-z geometries, while the SP 3 solution option is currently available for hexagonal-z geometry only. Both diffusion and SP 3 capabilities provide comparable solutions to DIF3D-VARIANT when using similar approximations. This document provides brief descriptions of inputs and outputs of the code.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

Steady-State Thermal-Hydraulic Analysis and Bowing Reactivity Evaluation Methods Based on Neutron and Gamma Transport Calculations

With the advancement of computer technology, the variational nodal transport code VARIANT is now routinely used for sodium-cooled fast reactor (SFR) design analysis. In particular, VARIANT transport calculations are usually performed for whole core neutronics analyses and for fuel cycle analyses with REBUS-3. However, the steady-state thermal hydraulics analysis with the SE2-ANL coed and the assembly bowing reactivity calculation with the structural analysis with the NUBOW-3D code are still performed with the DIF3D finite difference diffusion theory code. Furthermore, the DIF3D diffusion calculation for these applications is performed with only six triangular meshes per hexagonal assembly, and the SE2-ANL code determines the pin power distributions approximately by assuming that the power distribution within an assembly is separable in the radial and axial directions. In order to overcome the limitations and to improve the accuracy of the existing methods, it was proposed to develop the relevant computational methods based on the transport calculations with the VARIANT and PROTEUS-SN codes. The objectives of this study are to develop a new heating calculation method based on the coupled neutron and gamma transport calculations with VARIANT, to implement the new heating calculation method into the SE2-ANL steady-state thermal-hydraulics analysis code with improved numerical algorithms and additional capabilities for automatic flow allocation calculations, and to develop a method to calculate bowing reactivity coefficients based on the VARIANT core solutions and the PROTEUS-SN assembly solutions, which can be used in NUBOW-3D calculations. A new, coupled neutron and gamma heating calculation procedure has been developed based on VARIANT transport calculations. In the new heating calculation procedure, the neutron flux distribution is first determined by solving an eigenvalue transport problem using the VARIANT transport code. Using the calculated neutron flux, the intra-nodal gamma source distributions are calculated to be consistent with the neutron flux distribution with the GAMSOR code. For this, the GAMSOR code has been updated to generate the intra-nodal gamma source distributions in the form of the VARSRC dataset of VARIANT. With this gamma source distribution, the gamma flux distribution is determined by solving a fixed source gamma transport problem using VARIANT again. With the neutron and gamma flux distributions, the heat generation rates are evaluated at individual fuel pins and assembly duct walls, without assuming separable axial and radial profiles for the intra-assembly power distribution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Verification of the DIF3D Software to Support Fast Reactor Analysis

Ongoing design activities at Argonne National Laboratory are requiring a thorough verification of the Argonne Reactor Computation codes be performed. DIF3D is central to this system. The driver for this effort requires the 3D Cartesian, triangular-Z, and hexagonal-Z core geometry options of DIF3D be verified. Previous work identified the DIF3D features required to be verified to support current design activities, features of which are generally applicable to hexagonal-Z fast reactor designs. The scope of this verification effort includes verifying DIF3D’s ability to correctly translate the user’s model in to DIF3D’s preferred format, verifying that options planned for use have the desired effect, and verifying the correctness of the eigenvalue, fixed-source, forward, and adjoint solvers in DIF3D-FD and DIF3D-VARIANT. This manuscript provides the verification tasks and their results with respect to the features needed for current design activities. Since analytic solutions of the neutron diffusion and transport equations are either limited in scope or not possible, multiple tiers of problems unique to each solver and geometry type were implemented. Each of these tiers tests features independent and complementary arguments for why the separate testing of functionalities is acceptable. Finally, this separate testing was also supplemented with a high-level integral check of each the diffusion and transport capabilities and applicable geometries. To accommodate cases which an analytic solution is not feasible, MCNP6.2 was relied upon to provide a higher-order reference solution. This therefore required that the capabilities within MCNP6.2 which were relied upon for this work are also verified in this work. No MCNP discrepancies were noted in this effort. Note that the MCNP6.2 verification included in this work does not stand as a full verification of MCNP6.2, but merely verifies the features used in verifying DIF3D. The verification effort identified no issues that are debilitating or otherwise impactful to design usage of DIF3D, and thus DIF3D version 11.0, release 3012 is considered verified. The types of issues that were identified were predominantly in the areas of: unclear documentation, software bugs which were inconsequential to final results, editing options which were ignored in favor of printing more information than requested, bugs in the outputs of intermediate results, or secondary output binary file information which was not present. While not a bug, this verification report also identified that the algorithm used to evaluate the peak fast flux in a nodal transport solution can be quite unreliable due to the polynomial order used and the location of the peak within the mesh. The authors of the report therefore recommend the usage of the EvaluateFlux software (distributed with ARC) as a more robust alternative.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

NEAMS Thermal-Fluids Test Stand for Fluoride-Slat Cooled, High Temperature Reactor Development (Final CRADA Report)

The work performed in this CRADA resulted in significant progress toward the development of a Thermal-fluid test-stand based on NEAMS tools. Significant work has been achieved in this area with simulations performed in Nek5000 of the heat and fluid flow in twisted tube heat exchangers and various software and modeling improvements in SAM geared toward heat exchanger simulations. In particular a set of simulations of flow and heat transfer in twisted elliptical tube geometries has been performed with Nek5000. The method was validated by simulating a case at a moderate Reynolds number and relatively high Froude number. This case was demonstrated to predict a Nusselt number in excellent agreement with the applicable correlation found in literature. It was further shown that at lower Froude numbers, this agreement becomes much worse. Additional cases were then simulated with low Froude numbers for both a square and a triangular unit cell to test the effect of pitch-to-maximum diameter ratio for the purposes of future CFD modeling. The results have proven invaluable to understand the flow physics in twisted tubes and provide reference data for design. Results have been summarized in a conference article. Moreover, work in SAM has led to significant improvement of software quality, the creation of a set of benchmarks for natural circulation and other significant software improvement.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Preliminary Evaluation of Driver Fuel Plates and Reduced Sodium Temperature to Increase Fast Test Reactor Experimental Neutron Irradiation Capabilities

All sodium-cooled fast test reactors operated so far have used essentially the same fuel geometrical ar-rangement, namely cylindrical fuel pins assembled into a triangular array in hexagonal fuel assemblies. Fuel compositions and dimensions may vary but the overall geometrical pattern remains identical. Fur-thermore, most sodium-cooled reactors have operated, and still operate, with core-average inlet and outlet sodium temperatures of, respectively, 350-400°C and 500-550°C. The present preliminary study revisits the standard fuel assembly geometry as well as operating tempera-tures to evaluate potential impact on fast test reactor neutron irradiation capabilities. The reference fuel is U-20Pu-10Zr. Pin (reference) and plate geometries are considered together with two sets of core-average sodium inlet/outlet temperatures: 350/500°C (reference) and 200/400°C. More prototypic environments could be obtained inside ad-hoc irradiation testing vehicles that are independent from the reactor primary coolant. Everything else being the same, fuel temperature is significantly lower in plates than in pins. Therefore, plates could in principle accommodate higher plutonium content, thus, reducing the need for uranium enrichment or allowing the use of lower quality plutonium (U-xPu-10Zr thermal conductivity and solidus temperature decrease as x increases, hence, necessitating additional thermal margins). For the reference pin configuration, lowering the inlet/outlet sodium temperatures from the reference 350/500°C down to 200/400°C provides additional thermal margins that can be used to increase the peak fast flux from about 4.5 × 10 15 n/cm 2 -s to 6 ×10 15 n/cm 2 -s for the same core power of 300 MW. Assuming 300 Equivalent Fuel Power Days (EFPD) of operation per calendar year, a (steel) test article could accu-mulate up to 75 dpa/year. The use of fuel plates provides even more thermal margins which may allow the peak fast flux to reach values as high as 8 × 10 15 n/cm 2 -s for both sets of inlet/outlet temperatures. In this environment, a test article could accumulate up to 95 dpa/year assuming 300 EFPD/year, hence greatly accelerating irradiation testing. Allowing a fast test reactor to operate over a wide range of inlet/outlet temperatures could add significant flexibility to its neutron irradiation capabilities. Steel present in the driver fuel assemblies is not expected to accumulate more than 90 displacements per atom (dpa) over its lifetime. Hence, steel alloys that are not suited for very high dpa but can operate over a wide range of temperatures (from low to high) would be appropriate for this application. Finally, because pin and plate assemblies fit on the same grid plate, a fast test reactor could in principle start-up with standard fuel pin assemblies and, later on, move to a plate-type Mark-II fuel to further in-crease its neutron irradiation capabilities while potentially reducing the level of uranium enrichment needed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

A virtual element generalization on polygonal meshes of the Scott-Vogelius finite element method for the 2-D Stokes problem

The virtual element method (VEM) is a Galerkin approximation method that extends the finite element method to polytopal meshes. In this paper, we present a conforming formulation that generalizes the Scott-Vogelius finite element method (FEM) for the numerical approximation of the Stokes problem to polygonal meshes in the framework of the virtual element method. In particular, we consider a straightforward application of the virtual element approximation space for scalar elliptic problems to the vector case and approximate the pressure variable through discontinuous polynomials. We assess the effectiveness of the numerical approximation by investigating the convergence on a manufactured solution problem and a set of representative polygonal meshes. We numerically show that this formulation is convergent with optimal convergence rates except for the lowest-order case on triangular and square meshes where the method coincides with the P 1 - P 0 Scott-Vogelius scheme, which is well-known to be unstable.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2021

Structure-preserving numerical discretizations for domains with boundaries

This SAND report documents Exploratory Express LDRD Project 223790, "Structure-preserving numerical discretizations for domains with boundaries", which developed a method to incorporate consistent treatment of domain boundaries and arbitrary boundary conditions in discrete exterior calculus (DEC) for arbitrary polygonal (2D) and tensor-product structure prism (3D) grids. The new DEC required the development of novel discrete exterior derivatives, boundary operators, wedge products and Hodge stars. This was accomplished through the use of boundary extension and the blending of known 2D operators on the interior with 1D operators on the boundary. The Hodge star was based on the Voronoi Hodge star, and retained the limitation of a triangular circumcentric primal or dual grid along with low-order accuracy. In addition to the new DEC, two related software packages were written: one for the study of DEC operators on arbitrary polygonal and polyhedral grids using both symbolic and numerical approaches and one for a (thermal) shallow water testbed using TRiSK-type numerics. Immediately relevant (already funded, through CANGA) followup work is the development of a high-order, geometrically flexible Hodge star and structure-preserving, high-order, oscillation-limiting transport operators (using WENO) for n-forms on arbitrary 2D and 3D grids. This will provide all of the machinery required for a high-order version of TRiSK with boundaries on arbitrary 2D and tensor-product 3D grids, which is applicable to both the atmospheric (CRM in E3SM-MMF) and oceanic (MPAS-O) components of E3SM.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2021

Cross Section Measurements of Photonuclear Reaction Pathways Towards Promising Medical Radioisotopes

Project Objectives: The goal of this project was to generate data relevant to radioisotope production while developing innovative technologies that foster and enhance novel production of radioisotopes, and, also, while providing opportunities for cultivating and training future generations of scientists. This work has provided the foundation for methodologies for determination of photonuclear cross sections over multiple energies in a single irradiation. Simultaneously, the feasibility of electron LINAC production of several in-demand radioisotopes such as 47 Sc, 67 Cu, 77 As, and 186 Re has been demonstrated. To accomplish these objectives a collaboration was formed between two complimentary facilities, the Low Energy Accelerator Facility (LEAF) at Argonne National Laboratory and the High Intensity Gamma-ray Source (HIGS) at Triangular Universities Nuclear Laboratory (TUNL). The involvement of the research group from North Carolina Central University gave students at this Historically Black University experience in forefront nuclear-physics research relevant to addressing a high-priority interdisciplinary issue. Project Description: HIGS provides a nearly monoenergetic gamma-ray beam by intra-cavity Compton backscattering of free-electron photons from electrons circulating in a storage ring. This beam can be collimated to produce a very precise energy beam. If the beam is un-collimated a calculated and precise energy spread of the beam occurs radially. The γ-flux can be evenly distributed over the radial distribution of energy and used to perform activation experiments on concentric ring targets. Thus providing multiple energy ranges in a single irradiation. Each concentric ring target can be counted separately in order to determine activation at the given energy and successively be correlated to the activation cross section. Targets were activated to determine production feasibility using electron beams at LEAF. Potential Impact: The Nuclear Science Advisory Committee recently named production of radioisotopes with electron LINACs as one of the most compelling and largest-impact opportunities for the production of high specific activity radioisotopes. Improving the photonuclear cross sectional data base with experimentally verified results will greatly enhance a researcher’s ability to rationalize electron LINAC production routes towards desired radioisotopes. This work will provide the foundation for methodologies for determination of photonuclear cross sections over multiple energies and multiple targets in a single irradiation. The techniques developed in this project will enable future studies to continue verifying theoretically predicted photonuclear cross section with experimental results. These data will also enable adaptation of models and support more precise theoretical calculation of photonuclear cross sections. This research will involve undergraduates, graduate students, and post-docs to give them a valuable research experience leading towards the next generation of scientists in the field of medical isotopes.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Computational and Experimental Validation of Fractal-Fin, Dimpled Solar Heat Collector

A 3 foot x 3 foot x 3 foot aluminum solar collector was manufactured using computer numerical control. The interior of the device included six triangular dimpled fins for enhanced heat transfer. The interior vertical wall on the south side was also dimpled. The solar collector working fluid was based on water, and the collector consisted solely of passive heat transfer mechanisms (no moving parts), making it ideal for off-the-grid and rural applications. Two types of heat transfer experiments were conducted. One experiment had external flat heaters attached on the top and the front side, while the other four sides were insulated. Except for the bottom surface, the second experiment had all its exterior surfaces sprayed with black solar paint to collect as much solar heat as possible. Temperature data as a function of time was collected using 14 thermocouples spread strategically throughout the solar collector. In addition, computational fluid dynamics (CFD) simulations were conducted using the dynamic Smagorinsky large eddy simulation turbulence model. The first simulation considered that both the top and front surfaces were exposed to a fixed temperature of 313.7 K (105 °F), while the remaining four surfaces were insulated. For the second simulation, all conditions were the same, except that the temperature for both heated surfaces was raised to 350 K (170.3 °F). The two temperatures are expected to bound the solar collector operational temperature during the late- Spring, Summer, and early-Fall months. The solar collector design, experimental data, CFD output, and a discussion of five manufacturing approaches and costs are documented in this report.

14 SOLAR ENERGY↗
DOE OSTI2022

Verification of the DIF3D Software to Support Fast Reactor Analysis

Ongoing design activities at Argonne National Laboratory are requiring a thorough verification of the Argonne Reactor Computation codes be performed. DIF3D is central to this system. The driver for this effort requires the 3D Cartesian, triangular-Z, and hexagonal-Z core geometry options of DIF3D be verified. Previous work identified the DIF3D features required to be verified to support current design activities, features of which are generally applicable to hexagonal-Z fast reactor designs. The scope of this verification effort includes verifying DIF3D’s ability to correctly translate the user’s model in to DIF3D’s preferred format, verifying that options planned for use have the desired effect, and verifying the correctness of the eigenvalue, fixed-source, forward, and adjoint solvers in DIF3D-FD and DIF3D-VARIANT. This manuscript provides the verification tasks and their results with respect to the features needed for current design activities. Since analytic solutions of the neutron diffusion and transport equations are either limited in scope or not possible, multiple tiers of problems unique to each solver and geometry type were implemented. Each of these tiers tests features independent and complementary arguments for why the separate testing of functionalities is acceptable. Finally, this separate testing was also supplemented with a high-level integral check of each the diffusion and transport capabilities and applicable geometries. To accommodate cases which an analytic solution is not feasible, MCNP6.2 was relied upon to provide a higher-order reference solution. This therefore required that the capabilities within MCNP6.2 which were relied upon for this work are also verified in this work. No MCNP discrepancies were noted in this effort. Note that the MCNP6.2 verification included in this work does not stand as a full verification of MCNP6.2, but merely verifies the features used in verifying DIF3D. The verification effort identified no issues that are debilitating or otherwise impactful to design usage of DIF3D, and thus DIF3D version 11.0, release 3012 is considered verified. As some additional changes have been made to the ARC software since this point all versions between release 3012 and 3266 can be considered verified as version 3253 was used for all updates in this revision. The types of issues that were identified were predominantly in the areas of: unclear documentation, software bugs which were inconsequential to final results, editing options which were ignored in favor of printing more information than requested, bugs in the outputs of intermediate results, or secondary output binary file information which was not present. While not a bug, this verification report also identified that the algorithm used to evaluate the peak fast flux in a nodal transport solution can be quite unreliable due to the methodology used and the location of the peak within the mesh. The authors of the report therefore recommend the usage of the EvaluateFlux software (distributed with ARC) as a more robust alternative noting that DIF3D will properly notify the user when the peaking values it is providing are potentially incorrect.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2022

Fractal-Fin, Dimpled Solar Heat Collector with Solar Glaze

Exterior solar glaze was added to a 3 foot x 3 foot x 3 foot aluminum solar collector that had six triangular dimpled fins for enhanced heat transfer. The interior vertical wall on the south side was also dimpled. The solar glaze was added to compare its solar collection performance with unglazed solar collector experiments conducted at Sandia in 2021. The east, west, front, and top sides of the solar collector were encased with solar glaze glass. Because the solar incident heat on the north and bottom sides was minimal, they were insulated to retain the heat that was collected by the other four sides. The advantages of the solar glaze include the entrapment of more solar heat, as well as insulation from the wind. The disadvantages are that it increases the cost of the solar collector and has fragile structural properties when compared to the aluminum walls. Nevertheless, prior to conducting experiments with the glazed solar collector, it was not clear if the benefits outweighed the disadvantages. These issues are addressed herein, with the conclusion that the additional amount of heat collected by the glaze justifies the additional cost. The solar collector glaze design, experimental data, and costs and benefits are documented in this report.

14 SOLAR ENERGY↗
DOE OSTI2022

Integral Experiment Request 305 CED-3a Summary Report

Under IER-305, critical experiments will be done with and without molybdenum sleeves on 7uPCX fuel rods. New critical assembly hardware has been designed and procured to accomplish the experiments with the fuel supported by in a 1.55 cm triangular-pitched array.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Development and Analysis of Optimal Multilevel Solvers on Advanced Computers. Final Report

Constrained optimization in the context of time dependent, partial differential equations (PDE) leads to a symmetric, block-tridiagonal system of nonlinear equations that must be solved repeatedly in an iterative solution strategy. The blocks represent spatial discretization, while the connection between the blocks represents a forward and backward integration in time. The focus of this project is to apply a parallel-in-time (PiT) solution technique to the large block-triangular system.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2022

Useability and Optimization Improvements in MOOSE

The Multiphysics Object-Oriented Simulation Environment (MOOSE) framework is a foundational capability used by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program to create over 15 different simulation tools for advanced nuclear reactors. Due to this broad use, improvements to the framework in support of modeling and simulation goals are critical to the program. These improvements can take many forms, including optimization, improved user experience, streamlined application programming interfaces (APIs), parallelism, and new capabilities. The work transcribed in this report was conducted in direct support of the simulation tools and has already been deployed or will be deployed in the coming months. The capabilities were implemented in the same order as they are covered in this report: increased support of face variables, arbitrary spatial and temporal evaluation of material properties, and the addition of a triangular meshing library in libMesh.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

Verification of the DIF3D Software to Support Fast Reactor Analysis (Rev. 3)

Ongoing design activities at Argonne National Laboratory are requiring a thorough verification of the Argonne Reactor Computation codes be performed. DIF3D is central to this system. The driver for this effort requires the 3D Cartesian, triangular-Z, and hexagonal-Z core geometry options of DIF3D be verified. Previous work identified the DIF3D features required to be verified to support current design activities, features of which are generally applicable to hexagonal-Z fast reactor designs. The scope of this verification effort includes verifying DIF3D’s ability to correctly translate the user’s model in to DIF3D’s preferred format, verifying that options planned for use have the desired effect, and verifying the correctness of the eigenvalue, fixed-source, forward, and adjoint solvers in DIF3D-FD and DIF3D-VARIANT. This manuscript provides the verification tasks and their results with respect to the features needed for current design activities. Since analytic solutions of the neutron diffusion and transport equations are either limited in scope or not possible, multiple tiers of problems unique to each solver and geometry type were implemented. Each of these tiers tests features independent and complementary arguments for why the separate testing of functionalities is acceptable. Finally, this separate testing was also supplemented with a high-level integral check of each the diffusion and transport capabilities and applicable geometries. To accommodate cases which an analytic solution is not feasible, MCNP6.2 was relied upon to provide a higher-order reference solution. This therefore required that the capabilities within MCNP6.2 which were relied upon for this work are also verified in this work. No MCNP discrepancies were noted in this effort. Note that the MCNP6.2 verification included in this work does not stand as a full verification of MCNP6.2, but merely verifies the features used in verifying DIF3D. The verification effort identified no issues that are debilitating or otherwise impactful to design usage of DIF3D, and thus DIF3D version 11.0, release 3012 is considered verified. As some additional changes have been made to the ARC software since this point all versions between release 3012 and 3266 can be considered verified as version 3253 was used for all updates in this revision. The types of issues that were identified were predominantly in the areas of: unclear documentation, software bugs which were inconsequential to final results, editing options which were ignored in favor of printing more information than requested, bugs in the outputs of intermediate results, or secondary output binary file information which was not present. While not a bug, this verification report also identified that the algorithm used to evaluate the peak fast flux in a nodal transport solution can be quite unreliable due to the methodology used and the location of the peak within the mesh. The authors of the report therefore recommend the usage of the EvaluateFlux software (distributed with ARC) as a more robust alternative noting that DIF3D will properly notify the user when the peaking values it is providing are potentially incorrect.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

User-oriented Improvements in the MOOSE Framework in Support of Multiphysics Simulation

The Multiphysics Object-Oriented Simulation Environment (MOOSE) framework is a foundational capability used by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program to create over 15 different simulation tools for advanced nuclear reactors. Due to this ubiquity, improvements to the framework in support of modeling and simulation goals are critical to the program. These improvements can take many forms, including optimization, improved user experience, streamlined application programming interfaces (APIs), parallelism, and new capabilities. The work transcribed in this report was conducted in direct support of the simulation tools and has already been deployed or will be deployed in the coming months. The capabilities were implemented in the same order as they are covered in this report: chainable execution objects or executors, support for transfers between same-level applications in a coupling scheme, support for boundary/subdomain restricted transfers, support for transfers between applications with different coordinate or unit systems, support for MOOSE applications in the NEAMS Workbench, deployment of the MOOSE application of the Idaho National Laboratory (INL) high-performance computing (HPC) OnDemand platform, the addition of a triangular meshing library in libMesh, and increased support of face variables.

97 MATHEMATICS AND COMPUTING↗