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CASL VMA FY16 Milestone Report (L3:VMA.VUQ.P13.07) Westinghouse Mixing with COBRA-TF

COBRA-TF (CTF) is a low-resolution code currently maintained as CASL's subchannel analysis tool. CTF operates as a two-phase, compressible code over a mesh comprised of subchannels and axial discretized nodes. In part because CTF is a low-resolution code, simulation run time is not computationally expensive, only on the order of minutes. Hi-resolution codes such as STAR-CCM+ can be used to train lower-fidelity codes such as CTF. Unlike STAR-CCM+, CTF has no turbulence model, only a two-phase turbulent mixing coefficient, β. β can be set to a constant value or calculated in terms of Reynolds number using an empirical correlation. Results from STAR-CCM+ can be used to inform the appropriate value of β. Once β is calibrated, CTF runs can be an inexpensive alternative to costly STAR-CCM+ runs for scoping analyses. Based on the results of CTF runs, STAR-CCM+ can be run for specific parameters of interest. CASL areas of application are CIPS for single phase analysis and DNB-CTF for two-phase analysis.

97 MATHEMATICS AND COMPUTING↗

Validation of Pin Span Superposition to Reconstruct Gridded Bundle Flows

This milestone shows a demonstration of the surface map model to map the surface temperature for single-phase computational fluid dynamics simulation data from STAR-CCM+ to the subchannel code CTF using a two-step process that captures global and local rod surface temperature behavior based on simulation boundary conditions and position. This model can be used to improve results from CTF by transforming surface temperatures obtained from CTF to model data that resembles data from STAR-CCM+. A summary of the current two-step model process and results of the initial investigations with a single subchannel and a 5 x 5 set of fuel rods are shown and discussed within, with suggestions for further improvements.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NRC Multiphysics Analysis Capability Deployment (FY2021 - Part 2)

This report details the progress and activities of Idaho National Laboratory (INL) on the Nuclear Regulatory Commission (NRC) project “Development and Modeling Support for Advanced Non-Light Water Reactors.” Task 4b was completed for this report. INL developed a sample problem showing how to use Serpent 2 to calculate macroscopic cross sections for use in Griffin for a typical sodium fast reactor Unprotected Loss of Flow (ULOF) transient. The complicating factor is the relatively large axial gradient of the coolant (and hence fuel) temperature and its evolution during the transient. A 3D Griffin model of the Advanced Burner Test Reactor (ABTR) is coupled to the System Analysis Module (SAM) to perform the ULOF transient. The model includes various explicit feedback mechanisms for fast reactors, including Doppler, radial expansion from the displacement of the support plate, and axial expansion from the displacement of the fuel pins. The results for the ULOF transient are consistent with published values. We observed small effects in the transient results that arise from the superhomogenization equivalence correction of uniform and nonuniform temperature data sets, but they diminish as more dominant reactivity mechanisms are added to the model. Potential improvements to the neutronics model include: adding the differential control rod and sodium feedback mechanisms, implementing cross section re-homogenization based on the partial volume of the various materials, and enhancing the kinetics parameters. Furthermore, the fluids model only includes four subchannels to represent the entire active core due to performance limitations in SAM when attempting to model one subchannel per assembly.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Ducted Assembly Steady-State Heat Transfer Software (DASSH): Theory Manual

The Ducted Assembly Steady-State Heat transfer software (DASSH) is being developed at Argonne National Laboratory to obtain a steady-state thermal fluids result for the coolant flow and temperature distribution in a hexagonally gridded reactor core with ducted assemblies. In DASSH, each assembly is assumed to contain a hexagonal bundle of wire-wrapped pins. DASSH divides the coolant in the pin bundle into subchannels on which it applies energy balances to determine coolant and duct temperatures. DASSH offers multiple models for inter assembly heat transfer and features the ability to calculate pin temperatures. This document describes the models and equations implemented in DASSH. It includes a description of the geometric system to which the models are applied and a derivation of the subchannel energy balance formulation is presented. It also reviews the available correlations and provides an overview of the process by which power distributions are generated. Finally, it covers the code execution procedure, discussing the main code objects and their interactions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Analysis of Approximations in Modeling of BWR Bundle Void Distributions

In boiling water reactors, complex heterogeneous bundle designs, control blades adjacent to the corner of bundles, and the presence of boiling can lead to complex internal void distributions. A few approximations exist to model these void distributions. They could be modeled using a 1D axial solver in which each axial node is assumed to be at an average void, or each pin cell could be modeled with its own void concentration. In the latter case, the void could be discretized in pin-centered or coolant-centered channels. The goal of this project was to quantify the effect of using the different approximations for modeling internal void distributions on neutronics calculations. Using 3D void distributions calculated with CTF, Monte Carlo Neutral Particle (MCNP) transport code models were created for GE-9 and GE-14 lattices. For each model, the internal void distribution from CTF at a given axial node was selected, and a lattice calculation was carried out with MCNP. Comparisons between models using a lattice-averaged void, or using a void distribution in coolant-centered channels, showed large differences in reactivity which in some cases were well above 1,000 pcm, and it also showed differences in normalized fission rates greater than 20%. It was also found that using a lattice average void can lead to a significant difference in the estimation of the worth of a control blade. The differences found when comparing results from models using pin-centered and coolant-centered channels were up to 200 pcm in reactivity and up to 1.4% in the normalized fission rates. In addition to these two sets of comparisons, MCNP models were set up so that each subchannel had a saturated liquid component around the fuel pins and a saturated vapor component in the center to approximate annular flow. In comparison to the models using coolant-centered subchannels, up to 1–3% differences in normalized fission rates could be found.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Summary of DASSH-CFD Inter-Assembly Heat Transfer Comparison

Fast reactor core design requires integration of detailed power distributions obtained from neutronics models and thermal fluids calculations to determine temperature distributions and flow conditions. During the design phase, calculations may be repeated as specifications change and computational expense must be balanced with solution accuracy, creating an opportunity for the use of intermediate-fidelity codes. To that end, the Ducted Assembly Steady-State Heat Transfer Software (DASSH) has been developed at Argonne National Laboratory to perform full-core, steady state coolant and fuel pin temperature calculations in reactors consisting of hexagonal, ducted assemblies. DASSH is a first-order, finite volume subchannel code based on methodologies initially introduced in the legacy thermal hydraulics codes SUPERENERGY-2 and SE2-ANL. It can work coupled with the Argonne Reactor Computational (ARC) code suite to obtain neutron and gamma power distributions based on neutron transport solutions and features many improvements and new capabilities compared to its predecessors. DASSH calculates temperatures by balancing energy between subchannels and relies on correlations to account for flow distribution and coolant mixing.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

MOSCATO Development and Integration in Fiscal Year 2025: Implementation of Multiphase, Multiphysics Modeling Capabilities for Molten Salt Systems

MOSCATO (Molten Salt Chemistry and Transport) is a multiphysics code that provides high-fidelity, coupled simulations of fluid flow, heat transfer, mass transfer, chemistry, electrochemical phenomena, and alloy corrosion for molten salt systems. In FY25, significant developments were made to the code package, enhancing its capabilities for modeling all relevant phenomena within operating moltens salt reactors (MSRs). The developments and activities in FY25 included: 1. Implementation of Level-Set methods to enable modeling of single-bubble behavior in molten salts. In FY25, the Level-Set two-phase flow modeling implementation was improved to simulate single bubble behavior with molten salt media. The large density and viscosity ratios between typical gases and molten salt liquids present challenges for these types of numerical solvers. With enhancements to the pressure projection method, MOSCATO’s Level-Set solver was able to be successfully validated to experiments related to helium bubble rise in stagnant molten salt. The simulated bubble rising velocity showed reasonable good agreement with experimental measurements. The bubble shape and dynamics were also visually compared with experimental snapshots, demonstrating a good qualitative match. 2. Generation of mass transfer correlations for multiphase flow systems. To enable calculations of the tritium transport across the interface between gas bubbles and salt, we modeled high- Schmidt-number mass transfer around a sphere across a broad range of Reynolds numbers. The mesh near the sphere surface was highly refined to resolve steep concentration gradients caused by the low diffusion coefficient. Literature-based mass transfer correlations were compared with the numerical results, and modifications were proposed to improve agreement, particularly at higher Schmidt numbers. These mass transfer correlations were subsequently provided to other national laboratories to help enable high quality mass transfer simulations using lower-order solvers under development within the NEAMS program. 3. Preliminary implementation of a bubbly flow solver. To model bubbly flow in molten salt, we implemented a bubbly flow solver for void fractions less than 5%. To do so, an algebraic relative velocity model that assumes small bubbles with rapid momentum equilibration was added to MOSCATO to compute bubble velocities. Preliminary comparisons with experimental data showed reasonable agreement, and further improvements are underway. 4. Generation of mass transfer correlations for MSRE subchannel The Molten-Salt Reactor Experiment (MSRE) was a landmark historical project that demonstrated the feasibility of molten-salt reactor technology. The MSRE campaign also generated a significant body of experimental data and reports that continue to support molten-salt–related research. In this report, the MSRE core subchannel was used as the reference geometry for a mass transfer study performed with MOSCATO. The geometry and computational mesh were adapted from a previous study, providing adequate resolution for the relatively low Reynolds number in this case. Additional mesh refinement was applied to reach higher Schmidt numbers, enabling the derivation of a reliable mass-transfer correlation for the present scenario. 5. Simulations of oxygen ingressions into molten salt. In the previous fiscal year, we initiated a study to simulate oxygen ingression in stagnant salt. As oxygen enters the salt through its surface, it reacts with Ce 3+ to form solid CeO 2 and other reaction products. To more fully capture the complex diffusion-convection-reaction mechanisms, capabilities for modeling natural convection in the salt vessel were added. These were needed as the flow of the ingressed gas induced flow in the salt caused by surface shear and non-isothermal effects. With these updated physics in place, we were able to successfully reproduce the experimental results for the rate of change of CeCl 3 concentrations versus time. 6. Flow corrosion model validation. In FY24, MOSCATO’s corrosion model was validated against static corrosion experiments. In FY25, this work was extended to a flow corrosion experiment, where FLiNaK salt was driven by natural convection, with initial salt impurities to initiate corrosion. Despite uncertainties in parameters such as elemental diffusion coefficients in the alloy and unknown H + concentrations, the simulations achieved good agreement with experimental results, especially in predicting sample mass losses.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Development and Assessment of CTF for Pin-resolved BWR Modeling

CTF is the modernized and improved version of the subchannel code, COBRA-TF. It has been adopted by the Consortium for Advanced Simulation for Light Water Reactors (CASL) for subchannel analysis applications and thermal hydraulic feedback calculations in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS). CTF is now jointly developed by Oak Ridge National Laboratory and North Carolina State University. Until now, CTF has been used for pressurized water reactor modeling and simulation in CASL, but in the future it will be extended to boiling water reactor designs. This required development activities to integrate the code into the VERA-CS workflow and to make it more ecient for full-core, pin resolved simulations. Additionally, there is a significant emphasis on producing high quality tools that follow a regimented software quality assurance plan in CASL. Part of this plan involves performing validation and verification assessments on the code that are easily repeatable and tied to specific code versions. This work has resulted in the CTF validation and verification matrix being expanded to include several two-phase flow experiments, including the General Electric 3 3 facility and the BWR Full-Size Fine Mesh Bundle Tests (BFBT). Comparisons with both experimental databases is reasonable, but the BFBT analysis reveals a tendency of CTF to overpredict void, especially in the slug flow regime. The execution of these tests is fully automated, analysis is documented in the CTF Validation and Verification manual, and the tests have become part of CASL continuous regression testing system. This paper will summarize these recent developments and some of the two-phase assessments that have been performed on CTF.

Salko, Robert K↗

MC21/CTF and VERA-CS Mutiphysics Solutions to VERA Core Physics Benchmark Problems 6 and 7

The continuous energy Monte Carlo neutron transport code, MC21, was coupled to the CTF subchannel thermal-hydraulics code using a combination of CASL tools and in-house Python scripts. An MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 6 demonstrated excellent agreement with MC21/COBRA-IE and VERA solutions. The MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 7, Watts Bar Unit 1 at beginning of cycle (BOC) hot full power (HFP) equilibrium xenon conditions, is the first published coupled Monte Carlo neutronics / subchannel T-H solution for this problem. MC21/CTF predicted a critical boron concentration of 854.5 ppm, yielding a critical eigenvalue of 0.99994 ± 6.8E-6 (95% CI). Excellent agreement with a VERA solution of Problem 7 was also demonstrated for integral and local power and temperature parameters.

Aviles, Brian↗

Hydraulic performance evaluation of rod bundle with a concept spacer grid design

In a pressurized water reactor the spacer grid has a great influence on the thermal-hydraulic and mechanical properties of the fuel assembly. In order to improve the overall performance of the fuel assembly, a new concept spacer grid design names CATF with 'non-bending' guide vane is proposed. The hydraulic performance of the CATF grid design is evaluated by using a commercially computational fluid dynamics (CFD) code, ANSYS CFX12.1, and the evaluation result is compared with ordinary spacer grid including bending guide vane. Three dimensional rod bundles including CATF spacer grid and ordinary spacer grid have been simulated. The analysis results show that local resistance coefficient of CATF spacer grid is smaller than that of ordinary spacer grid by about 29.7% because of the non-bending guide vanes. Furthermore, the cross flow strength between subchannels near the outer strap area, the swirl strength and cross flow strength in subchannels near the outer strap area in 1-10 times hydraulic diameter downstream the spacer grid area of the ordinary grid, are larger than that of CATF spacer grid. The two types of spacer grids appear to have the same mixing characteristics in inner-grid area. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Multiple RANS Scheme for the CFD Simulation of PWR

Though best-estimate system codes have been well established, the assumptions for the simplification of the conservation equations will be invalid when 3D effects play an important role in the object. The inner domain of PWR can create high 3D effects because of the disturbance of mixing vanes and the complex distribution of flow resistance. With the finest mesh to study the flow and heat transfer process of a nuclear reactor, a CFD simulation can give a perfect understanding of the physical processes in a whole reactor. However, more than 10 thousands of subchannels exist in a typical pressurized water reactor (PWR). These subchannels are narrow, comprising complex spacer grids and mixing vanes. The need of both the computational resources and the simulation time is great, which is not good for the simulation efficiency. However, the efficiency is critical for the CFD simulation of a PWR. It is meaningless if the simulation does not end. To improve the efficiency, one approach is to well consider the turbulence model used in a CFD simulation, because some models are complex and require great computing resources and time. On the contrary, some models are simple, but are also suitable to reflect the physical status in a PWR. Furthermore, the effects of the typical structures of PWR should also be clarified. They are closely related to the flow mechanism. The CFD simulation may be simplified by such effects. The objective of this research is to study the flow characteristics affected by the turbulence models and typical structures. The authors then design a scheme based on the regularities of the flow characteristics to improve the efficiency of the CFD simulation for PWR. By this research, we found: 1) The mixing vanes, a typical structure of a PWR, make the anisotropic eddy viscosity cannot be ignored for the region with short distance downstream of this structure; 2) Outside of this region, the RANS models, regardless of whether they are isotropic or anisotropic, can obtain similar results. To improve the simulation efficiency, a CFD scheme is designed by the hybrid application of multiple RANS models (MRANS). In this scheme, the major domain with low anisotropic viscosity uses the turbulence model which requires the least time. The difference is just little in the comparison with a simulation using the anisotropic RANS model for the whole computational domain. The MRANS scheme thus will be significant to improve the efficiency of the CFD simulation for a whole PWR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Integration and coupling of ATHLET, CTF and DYN3D on the Salome Platform. Application to A MSLB Transient

In the framework of the EU founded NURESAFE project, a fully functional multi-physics multi-scale simulation tool was developed. The TH system code ATHLET, the subchannel code CTF and the neutronics code DYN3D were integrated and coupled on the Salome platform. The developments achieved during this three-year project include assembly level and pin-by-pin multi-physic TH) / NK coupling, as well as a multi-scale TH/TH coupling between ATHLET system code and CTF subchannel code. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Characteristics of Axial Power Shape and Flow Field at Low Flow Condition

The representative event which experiences the low flow condition is SLB (Steam Line Break) event. During the SLB event with LOOP (Loss of Offsite Power), the reactor core is tripped and all RCPs (Reactor Coolant Pump) are stopped. The biggest worth control rod also stuck out. As a result, the flow rate and pressure in reactor core decrease and the peaking factor is relatively high compared to a nominal operating condition. In this paper, we study the tendency of the core axial power distribution and flow field with respect to the core average power at the SLB accident condition. Generally, the condition has been analyzed using the neutronics code with one dimensional flow result of the assemblies. However, if the core average power increases, the coolant density in the hot channel decreases and mass flux increases. The assumption that the mass flux is constant at the one dimensional analysis is incorrect. Hence, it is needed to couple the neutronics code and the subchannel analysis code in order to give feedback to neutronics code with the 3-D flow information. In this study, we coupled the nuclear design code and the subchannel analysis code and showed the characteristics of the axial power distribution and the flow field at the low flow condition. Since there is no experimental data at the condition, the characteristics of the power distribution and flow field were analyzed to understand the physical phenomena. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Critical Heat Flux Model Improvement in CTF for Natural Circulation Type Reactors

The objective of this study was to enhance the departure from nucleate boiling (DNB) modeling capabilities of the subchannel analysis code CTF. The intent was to include consideration for PWR-type reactors which are characterized by relatively low subchannel mass flux and high inlet subcooling during normal operation. Identifying a CHF model appropriate for these type of conditions is required for an accurate thermal-hydraulics performance analysis. Several CHF models that are available in open literature, including those already implemented in CTF, were studied for their applicability. This includes the CHF models by Westinghouse (W-3), Electric Power Research Institute (EPRI-1), Biasi, Bowring, and Groeneveld. Since the models were developed within a certain validity range of operating conditions (mass flux, equilibrium quality, pressure, hydraulic diameter, and others), these ranges were used as a measure of acceptability. It was found in these considerations that the Groeneveld and Bowring models were the most appropriate. Therefore, the models by Bowring and Groeneveld were selected for implementation in CTF. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

VUSAT - A VERA-CS Uncertainty and Sensitivity Analysis Toolkit

VERA-CS (the Virtual Environment for Reactor Applications, Core Simulator) is a high-fidelity multiphysics core simulator under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS includes three main components: MPACT for reactor physics and neutron transport, COBRA-TF for thermal-hydraulics, and ORIGEN for isotopic depletion. VERA-CS has been applied to an array of problems including core physics analysis, and the modeling for all of the fuel cycles of Watts Bar Nuclear Unit 1 (WBN1). MPACT is a 3D pin-resolved reactor transport code developed by Oak Ridge National Laboratory (ORNL) and the University of Michigan. COBRA-TF (CTF) is a three-field two-phase flow subchannel analysis code developed by ORNL and North Carolina State University. CTF solves the mass, momentum, and energy equations for the liquid, vapor, and droplet phases for each subchannel in the core and captures the axial flow within each channel as well as cross-flow between channels. Direct coupling between neutronics (MPACT) and thermal-hydraulics (CTF) in VERA-CS provides temperature and density feedback. CTF includes a number of models important for LWR safety and design analysis including flow regime dependent two-phase wall heat transfer, inter-phase heat transfer and drag, droplet breakup, and quench-front tracking. ORIGEN is developed by ORNL and released through the SCALE package. ORIGEN is directly integrated in MPACT through its API to calculate the changes in nuclides for every depletable cross-section region. Uncertainties exist in safety calculations with computer codes for nuclear reactor modeling and simulation, and they come from code limitations, scaling inaccuracies embedded in the experimental data, and uncertainties associated with the state of the reactor at the initiation of the transient. Consequently, for applications of high fidelity calculations using VERA-CS, proper uncertainty quantification (UQ) and sensitivity analysis (SA) of the computing models of reactor neutronics and thermal-hydraulics are important to comprehensively understand the results. New methods in UQ and SA have to be developed in multi-physics calculations using VERA-CS. Toward this end, a software package called VUSAT which stands for VERA-CS Uncertainty and Sensitivity Analysis Toolkit has been developed to facilitate the applications of VERA-CS within the context of best estimate plus uncertainty (BEPU) analysis framework. In this work, the VUSAT application to analyze departure from nucleate boiling (DNB) is presented.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

High Resolution Stereoscopic PIV Measurements in a 5 x 5 Rod Bundle with Mixing Vane

Spacer grids are used in the pressurized water reactor (PWR) to fix, clamp the fuel rods, and enhance the turbulence mixing within the spaces of the rods and overall heat transfer performance. Proper understanding on the thermal-hydraulic characteristics of the turbulent mixing and heat transfer created by the space grids is important. However, spacer grids having complex design structure with dimple, mixing vane and spring have been challenges for computational fluid dynamics (CFD) turbulent models. High fidelity experimental data with spatial-temporal resolutions is essential to validate and assess the performances of CFD models. Baglietto and Ninokata applied an improved realizable anisotropic turbulence models via direct numerical simulation (DNS)-based near-wall formulation data to study the wall shear stresses and flow structures in tight bundle subchannels. These authors have shown that their models are able to capture the stress distribution and velocity field at an acceptable accuracy. The authors numerically studied the flow field characteristics at the region downstream of the spacer grids, and validated their results to particle imagery velocimetry (PIV) measurements. Dominguez and Hassan performed two-dimensional time resolved PIV (TR-PIV) measurements within a 5 x 5 PWR rod bundle with spacer-grids that utilized a matching index of refraction (MIR) approach. Their averaged data showed a rather uniform direction velocity profile after the spacer grid while the downstream illustrated a no slip wall profile. In this paper, we presents flow measurements in the MIR 5 x 5 fuel rod bundle with a spacer grid. Time resolved stereoscopic PIV (TR-SPIV) measurements were performed to obtain high spatial-temporal resolution two-dimensional three-component (2D3C) velocity vector fields in the subchannels between the rods. Flow statistics such as mean velocity, and Reynolds stresses were computed. The obtained experimental data can be used to improve the computational modeling and validate the numerical models that are currently applied to light and pressurized water reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

The MOOSE fluid properties module

The fluid properties module in MOOSE~\cite{lindsay2022moose} serves a variety of fluid simulation applications based on MOOSE, including the MOOSE Navier Stokes module~\cite{moose_ns}, Pronghorn~\cite{pgh}, the MOOSE thermal hydraulics module, SAM~\cite{sam}, RELAP-7~\cite{relap7}, Sockeye~\cite{sockeye}, Pronghorn-subchannel~\cite{subchannel} and the MOOSE porous flow module~\cite{porous}. These applications are used to solve coarse mesh multi-dimensional thermal-hydraulics~\cite{pgh}, 1D systems analysis~\cite{sam,relap7} in nuclear reactor analysis, heat pipe modeling~\cite{sockeye} and porous flow simulations~\cite{porous} for underground gas storage and water seepage. The use of consistent fluid properties across fluid flow applications facilitates coupled flow simulations~\cite{anl_sam_pgh,osti_1889653}. Each application has historically driven the implementation of several fluid properties, which were later extended to be compatible with other applications. The unique diversity of applications of the module, due to its presence in MOOSE, has driven its expansion to new fluids, such as advanced nuclear reactor coolants and, more recently, arbitrary functions or tables-based property definitions, as detailed in section~\ref{content}, as well as numerous thermophysical properties and variable sets, as detailed in subsection~\ref{sec:prop}. The need for different discretizations of flow equations based on the compressibility of the fluid has motivated support for both primitive (pressure- and temperature-based) and conservative (internal energy- and specific volume-based) flow variables in the module. Thermodynamic relations are used to convert between these two formulations, as needed. The module serves a dual purpose of providing fluid properties for direct use in flow simulations and facilitating the implementation of user-specific fluid properties. Contributions of new properties for existing fluids or new fluids are strongly encouraged.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Latest developments in the MOOSE fluid properties module

The fluid properties module in MOOSE serves a variety of fluid simulation applications based on MOOSE, including the MOOSE Navier Stokes module~\cite{moose_ns}, Pronghorn~\cite{pgh}, SAM~\cite{sam}, the MOOSE thermal hydraulics module, RELAP-7~\cite{relap7} and subchannel~\cite{subchannel}. It is used for coarse mesh multi-dimensional thermal-hydraulics~\cite{pgh}, 1D systems analysis~\cite{sam,relap7} in nuclear reactor analysis, and porous flow simulations~\cite{porous} for underground gas storage and water seepage. The use of consistent fluid properties across fluid flow applications facilitates coupled simulations~\cite{anl_sam_pgh}. The module offers a consistent set of interfaces to implement to create a new fluid property. There are numerous fluid properties of interest in the entirety of all fields of fluid flow simulations, and this is exacerbated by the use of different variable sets depending on the compressibility of the fluid. For single-phase fluids, the following variable sets may be used to compute fluid properties: (pressure, temperature) and (specific volume, specific internal energy). Some properties may also be computed using the (pressure, density) or the (specific volume, specific enthalpy) variable sets. In order to reduce the challenge of adding a new fluid property, properties may be implemented partially.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗