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DOE OSTI2017

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↗
DOE OSTI2017

Uncertainty Quantification and Sensitivity Analysis for Coupled VERA-CS and FRAPCON

The current nuclear power plant fleet in the U.S. is facing numerous challenges economically ranging from low natural gas price to the rapidly growing renewable energy sources such as wind and solar power. In order for the nuclear industry to keep the viability of its operating fleet, new innovations have to be introduced to lower the costs to operate the fleet. One area that could contribute to the cost reduction is to lower the fuel cycle cost and improve the plant safety through initiatives such as higher burnup fuel and accident tolerant fuel. At the same time, the nuclear industry has to address future challenges on the horizon such as load following and other flexible operating strategies as well as those associated with new regulations such as 10 CFR 50.69 and 10 CFR50.46c etc. All these require coupled analysis of core neutronics, thermal hydraulics and fuel performance, which necessitates the development of a multi-physics toolkit. Since different models, assumptions, and fidelities have gone into various computer codes, how to consistently propagate uncertainties in a multi-physics environment warrants careful evaluation. The U.S. DOE's Light Water Reactor Sustainability (LWRS) Program initiated a project to develop analytical capabilities to support the industry to address these challenges. The general idea behind the initiative was the development of an integrated multiphysics evaluation model. This integrated evaluation model is called LOTUS which stands for LOCA Toolkit for the U.S. Light Water Reactors. The focus of LOTUS is to establish the automation of the interface between the five disciplines. LOTUS framework ensures that uncertainties can be easily propagated to solve the problems forward and backward (inverse) with large numbers of simulations. The objective of this work is to perform uncertainty quantification (UQ) and sensitivity analysis (SA) on a coupling of CASL's Virtual Environment for Reactor Applications-Core Simulator (VERA-CS) and NRC's fuel performance code FRAPCON. The work presented herein is a subset of the overall LOTUS framework. The figures of merit (FOM) selected for analysis are minimum departure from nucleate boiling ratio (MDNBR), maximum fuel centerline temperature (MFCT), and the gap conductance at peak power (GCPP). All FOM are obtained at the end of the of a depletion case for a half cycle lasting 263 days. The minimum is taken over all spatial location for each pin at the end of the half cycle. This value is calculated in VERACS alone and has been analyzed in previous work by Brown and Zhang. The MFCT is available in both VERA-CS and FRAPCON, but is calculated via different methods (see Model Inconsistencies section for more details). For this work, only the FRAPCON values are reported due to the code's maturity from extensive validation in fuel performance. Previous work includes SA on a standalone nuclear fuel performance code VTT-modified FRAPCON- 3.4 by Ikonen. The GCPP is the gap conductance at the location corresponding to maximum power in the fuel rod. This value is available in FRAPCON alone and highlights the capability of LOTUS to obtain unique fuel performance metrics for each pin in a core simulation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Diffusion Synthetic Acceleration for High Order S{sub N} Transport on Meshes with Curved Surfaces

Thermal radiation transport is integral to multiphysics applications such as inertial confinement fusion and astrophysics. Materials in these problems can be exceptionally optically thick and diffusive. A transport solver must be capable of accurate and efficient solutions to such problems. Previously, Woods et al. established a proof-of-concept for high order (HO) discontinuous Galerkin finite element (DGFEM) transport on meshes with curved surfaces. They numerically demonstrated the spatial convergence is order p + 1. Remarkably, these convergence rates are relatively invariant to the amount of curvature of the spatial mesh. We present here an extension of this work that demonstrates a rapidly-convergent iterative solver for the FEM transport equations on meshes with curved sides. We use the modified interior penalty (MIP) diffusion synthetic acceleration (DSA) established by Wang and Ragusa. The partially consistent MIP equations stem from the symmetric interior penalty (IP) derivation of the diffusion equation. The IP method is not stable for optically thick cells so it was combined with the diffusion conforming form (DCF), a discretized diffusion equation derived from the discretized S N equations that is not stable for intermediate optical thicknesses. The MIP equations adapt a 'switch' to the IP method in optically thinner regions and the DCF method in optically thicker regions. The literature does not reveal any Fourier analyses performed for iterative transport on meshes with curved surfaces. Wang and Ragusa performed a Fourier analysis for orthogonal meshes and our results appear to be as expected. Previous literature has reported on some of the sensitivities that affect the spectral radius of DSA schemes: polynomial order, and scattering ratios. Presently, we focus on the sensitivity of the spectral radius to polynomial order, scattering ratio, the constant C, and mesh curvature for varying cell thicknesses. We also numerically study the behavior of the spectral radius to verify our DSA implementation. A sequence of problems is designed to exercise the acceleration. As the problems become more optically thick and diffusive, the accelerated solution continues to converge toward the analytic solution. We employ the Modular Finite Element Methods (MFEM) package, developed at Lawrence Livermore National Laboratory, to generate the linear algebraic system of equations for each angle in the discrete ordinates quadrature set, and use UMFPack to solve this system of equations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Flexible Spatial Partitions in MPACT Through Module-Based Data Passing

As part of the Consortium for Advanced Simulation of Light Water Reactors (CASL), the Virtual Environment for Reactor Applications (VERA) is being developed to provide high-fidelity multiphysics simulations of nuclear reactor cores. The MPACT code being developed collaboratively by Oak Ridge National Laboratory and the University of Michigan is the primary deterministic neutron transport solver available within VERA. MPACT employs the 2D/1D method to solve 3D problems using 2D method of characteristics (MOC) radial transport solvers with 1D nodal expansion method-based P3 (NEM-P3) axial transport solvers, which are coupled using axial and radial transverse leakages. A 3D coarse mesh finite difference (CMFD) solver is used to provide accelerated convergence and stability. Quarter-core problems in MPACT are typically run on moderately sized clusters ({approx}500-10,000 cores). As such, it relies heavily on spatial decomposition, both radially and axially, using MPI data communication along spatial domain boundaries to pass angular flux boundary condition and neighboring data. The initial spatial decomposition strategy implemented into MPACT overlaid a Cartesian grid over the core to form the domain boundaries. With this approach, a single buffer spanning across the entire boundary between two spatial domains is passed to a domain's neighbor. Additionally, only one neighboring domain index was allowed in each direction. For example, the entire eastern boundary of a domain must be the entire western boundary of the corresponding neighbour's domain. This implementation helped to execute cases on several thousand cores, where substantial radial decomposition was used. The default for several years was to use assembly decomposition, which yielded 73 radial partitions in the Watts Bar Unit 1 (WBN1) models being analyzed. However, as CASL formed stronger connections with industry partners, there was a strong push to allow MPACT to run on fewer cores, preferably in the 500-1,000 core range. Such cases require much less radial decomposition, typically with 8-16 radial domains on each plane. Unfortunately, the previous decomposition strategy is very limited in its ability to yield well-balanced domains, and it does not permit partitions with an arbitrary number of domains, particularly with small numbers of radial domains (3, 5, 7, 11, 13, etc.). With these limitations, a more flexible approach that provided more even-load balancing was needed. The details of this new strategy are outlined here, along with insights to its advantages. The examples provided are for the WBN1 core, which has been the focus of several initial CASL analyses. It is worth noting that spatial decomposition strategies have been employed in many codes with focus on a number of different transport methodologies. The references cited are not exhaustive as this has been very important research topic across many areas.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

A New Look at Historic CRAC 'Divergence' Experiments - Implications for a CAAS Minimum Accident of Concern

This paper examines power history parameters and results based on historic fissile solution criticality 'divergence' experiments (referenced here as Divergence Experiments) from the 1960's-70's' Consequences Radiologiques d'un Accident de Criticite' (CRAC) campaign. The results from specific Divergence Experiments have implications for an ANSI/ANS-8.3 (1997) Criticality Accident Alarm System (CAAS) Minimum Accident of Concern (MAC). Specifically, this paper examines select CRAC Divergence Experiments in conjunction with 'qualifying reactivity additions' for a MAC defined by an absolute time constraint. A previous paper presented at the 2011 International Conference on Nuclear Criticality (ICNC 2011) illustrated the differences in 'qualifying reactivity additions' above delayed critical (and below prompt) for a variety of MAC constructs. That paper was the first to examine delayed critical kinetics in greater detail over a range of process and MAC considerations. The ICNC2011 work did not consider an absolute MAC time constraint from the excursion onset and assumed an initial excursion power. That study evaluated MAC constructs based on 'relative time differences' between reaching a specific fission rate and total fissions. The work presented here thus picks up where the ICNC2011 study left off. The primary focus of this work provides simulated power histories for specific Divergence Experiments that served as a basis for justifying a MAC noted in ANS-8.3 Appendix A. This paper documents computational modeling and dynamic multiphysics simulation with known Divergence Experiment parameters for reactivity additions at, below, and above the reported terminal reactivity. The initial step considers replicating what could be a reasonable power history for the terminal reactivity addition to the extent the same peak power and total fissions are obtained over the total time considered. Power histories for all other reactivity additions are simulated by varying the input reactivity and solution heights for the same model. A relationship is obtained between reactivity addition and the time to reach specific MAC power and fission levels within a given time constraint (e.g. 60 s). The 'qualifying reactivity addition' is determined for the total MAC fissions within that time. This value is then compared to the historic terminal value (e.g. $0.33) or any other values (e.g. 'excess reactivity of only a few cents') noted in ANS-8.3 Appendix A. In general, calculated MAC power levels have ranged from 1.8 kW to 173 kW depending on material type and form, H/U, enrichment, critical volume, and dose response function. A minimum accident consideration noted a SHEBA experiment with an excess reactivity 7 cents above delayed critical (TD ∼110 s). This excursion provided an additional 10{sup 15} fissions after an undetermined buildup time resulting in a 'few times 10{sup 15} fissions'. ICNC 2011 LEU solutions calculations show 0.2 Gy in air corresponds to 1.4 x10{sup 16} fissions. Thus, while historically, 0.2 Gy has been loosely associated with 'several times 10{sup 15} fissions', in reality, the actual total fissions could be an order of magnitude larger, and the power associated with 0.2 Gy/min (7.5 kW) larger as well. The reactivity addition required to obtain the actual 0.2 Gy total fissions over 60 s could be much larger than considered. Another reason the qualifying reactivity level could be larger than previously noted, involves temperature reactivity feedback. Compensated reactivity results in a departure from a pure exponential power rise, increasing the doubling time (TD) as the excursion approaches peak power. Uncompensated reactivity additions and constant TD have been traditionally assumed for a pure exponential power rise through the kW range to determine either total MAC fissions, power, or arrival time to these values.

42 ENGINEERING↗
DOE OSTI2017

Simulation of the RPI Reactor Critical Facility Core Using High-Fidelity Neutronics Code PROTEUS

The PROTEUS code is a high-fidelity three-dimensional (3D) deterministic neutron transport code developed by ANL under the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. The code contains the SN2ND (second order discrete ordinates method) and two MOC transport solvers, MOCFE and MOCEX (first order methods of characteristics), based on unstructured finite element meshes which allow users to model complex or unconventional geometry reactor problems. Among the two MOC solvers which normally perform better than SN for heterogeneous geometry problems, MOCEX is based on a rigorous formulation using the two-dimensional (2D) MOC radially and the discontinuous Galerkin finite element method axially to overcome the drawbacks of MOCFE with the full 3D MOC representation which requires enormous memory and computational efforts. The Reactor Critical Facility (RCF), an open tank research reactor at the Rensselaer Polytechnic Institute (RPI), became operational in 1956 under the American Locomotive company and has been operating using low-enriched uranium fuel pins, which makes it the only university research facility to use fuel rods similar to operating commercial light water reactors (LWRs). This paper presents modeling and simulation results of the RCF using PROTEUS for code verification and validation which also supports an ongoing NEUP project in which a set of benchmark validation experiments for multiphysics coupling is being developed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Surrogate Models for TREAT Transient Calculations

The Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is currently preparing for resumption of operation after more than 20 years in a long-term standby state. In parallel with preparation for operations and a new transient testing program, the DOENE Advanced Modeling and Simulation program (NEAMS) has been supporting the development of three-dimensional transient multi-physics modeling capabilities within the reactor physics application MAMMOTH. Recently, the MAMMOTH reactor physics team has begun development of uncertainty and sensitivity analysis for the TREAT core. This research will be used to support the analysis of reactor operation and experiment planning. Traditional approaches, i.e., Monte Carlo simulations, for such tasks require a large number of model evaluations under both steady state and transient conditions. However, exhaustive sets of sampling runs are simply not practical for TREAT since evaluations of each model can take a substantial amount of time. A solution to this problem is to build a surrogate model employing a smaller number of model evaluations. The surrogate model is designed to provide an accurate and fast approximation to the detailed TREAT model. In this manner, the surrogate model can be used in place of the TREAT transient model for uncertainty and sensitivity analysis at a fraction of the computational cost. The high dimensional model representation (HDMR) method is employed here for approximating the responses of interest in terms of functions of lower dimensions. This method has been implemented in RAVEN and all TREAT transient calculations are performed using the multiphysics code MAMMOTH. In addition, a code interface has also been developed in order to couple RAVEN with MAMMOTH. The surrogate model is then tested and compared for TREAT startup testing program Transient Test no. 15 core, and its accuracy is measured by its ability to estimate the mean, coefficient of variance, skewness, excess kurtosis and probability density function of the outputs of Transient Test no. 15 model, as well as root mean square error metric.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Demonstration of Load-Follow Operations with Standalone BISON through VERA

As part of the Consortium for Advanced Simulation of Light Water Reactors (CASL), the Virtual Environment for Reactor Applications (VERA) is being developed to provide high-fidelity multiphysics simulations of nuclear reactor cores. VERA contains a number of single physics packages that are coupled together. MPACT is the primary deterministic neutron transport solver, CTF provides thermal hydraulics solutions, and BISON - which is built off of the MOOSE framework - is used for fuel performance simulations. Over the past year, a standalone BISON capability has been developed using a one-way coupling between MPACT/CTF coupling calculations and BISON. This has allowed CASL contributors to gain insight into the fuel performance characteristics while a more tightly coupled methodology between MPACT, CTF, and BISON has been under development. The standalone capability has been used to simulate Watts Bar Unit 1 (WBN1) Cycles 1-3 and to screen for quantities of interest to pellet-clad interaction in Cycles 6-7. The simulation and screening results are presented here. CASL recently partnered with Exelon and the University of Illinois at Urbana Champaign (UIUC) to begin simulating some of Exelon's plants, with particular focus on load-follow operations. In such operations, the power output from the plant will vary based on the demand or anticipated demand at the time. For example, demand is much lower late at night than during the day when people are awake, so it operating the plant at lower power in the evenings could be advantageous. Historically, nuclear power generation has been among the cheapest sources of energy, so most plants in the United States have been operated at full power except for outages and various small scale events. However, in the current economic climate, priorities have changed. A number of questions arise when load-follow operations are considered, particularly with respect to VERA's tools and capabilities. Can the coupled neutronics and thermal hydraulics simulations with MPACT and CTF perform well without substantial convergence issues? Can BISON handle the somewhat rapid power changes present with load-follow? Will clad hoop stresses be alarmingly high as a result of these power changes? This work addresses these questions and demonstrates operations to assess feasibility and robustness before Exelon and UIUC proceed with their simulations using VERA. To demonstrate load-follow capability, two test problems were developed. The first is a single rod that experiences nominal operations during Cycle 1, and then it simulates 18 months of representative load-follow operations in Cycle 2. This single-rod case was used to test the robustness of MPACT, CTF, and BISON on a typical rod. The second problem is a quarter-core model based on WBN1, Cycle 3, with a single month of load-follow operations inserted into the middle of the cycle. This provides better understanding of the stresses to the clad during the numerous power changes. Sections of this article are adapted from the original, full technical report on this capability.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Scientific Machine Learning using MOOSE

A machine learning interface has been recently developed for the Multiphysics Object-Oriented Simulation Environment (MOOSE). This presentation summarizes the motivation behind using machine learning in scientific computing together with several examples for applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Applying a predictive modeling methodology to a radiation transport inverse problem - 25588

In this paper, the predictive modeling of coupled multiphysics systems (PM-CMPS) procedure, developed by Cacuci and embodied in the software package MULTIPRED, has been applied to determine unknown parameters of a radioactive system from its gamma-ray and neutron signature. PM-CMPS has been integrated in two different ways with the inverse transport code INVERSE to determine the parameters. The first uses PM-CMPS in the inverse mode to directly estimate the parameters while the second mode uses a traditional nonlinear optimization routine and has PMCMPS complete a final iteration to estimate the uncertainty in the parameter best estimates. A test problem has been created and used to test both methods for varying numbers and types of responses and parameters. The first method was capable of accurately determining parameters but had a number of drawbacks. The second method functioned well at determining the correct parameter values in cases with limited numbers of parameters. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

Anderson acceleration stability in NDA-accelerated k-eigenvalue problems

Anderson acceleration (AA) has been used to improve the stability and convergence rate of multiphysics iterative methods for reactor analysis. Most applications studied assume a tightly converged solution for the different physics problems, and AA is usually applied to state variables like temperature, density, and heat generation rate. In this paper, we study the theoretical performance of AA in NDA-accelerated k-eigenvalue problems. The problems and algorithms studied are simplified from the coupled iteration scheme adopted by MPACT and many other high-fidelity whole-core reactor codes. Compared to previous analyses of AA for these iteration schemes, we study the case with a partially converged neutronics solution and possibly partially converged nonlinear diffusion acceleration (NDA)/coarse mesh finite difference (CMFD) solutions. We observe that the performance of the iteration scheme with AA is very sensitive to the initial guess and is affected by the partially converged CMFD solutions. When the NDA solution is fully converged, using AA cannot achieve the optimal convergence rate in large-sized problems. Conversely, if the NDA solution is partially converged, the iteration scheme with AA can diverge or converge extremely slowly. It is found that the loss of robustness for AA is due to the fact that it is applied to the iterative subspace of state variables rather than the fundamental unknowns of the governing equations. To improve the robustness, the scalar flux should also be considered in the implementation of AA. After considering the residuals of flux, we observe that the stability is regardless of the partial convergence of NDA solutions. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Multigroup cross section generation capability in GRIFFIN

GRIFFIN is an advanced reactor multiphysics application built on the object-oriented simulation environment (MOOSE) and is jointly developed by Idaho National Laboratory and Argonne National Laboratory. The cross section application programming interface, originally developed for the PROTEUS code, has been integrated into GRIFFIN to prepare cross sections for thermal reactor applications with heterogeneous geometries. Additional improvements have been made by implementing an on-the-fly slowing down method, a double heterogeneity treatment capability, and updating the procedure to generate the fine multigroup library. The cross section preparation capability in GRIFFIN was verified for graphite-moderated TRISO fuel-based reactor benchmark problems: unit-cell problems of VHTR and EMPIRE micro reactor and HTTR assembly problems. Eigenvalues and multigroup cross sections of GRIFFIN agreed very well with those of the continuous-energy Monte Carlo code Serpent2 within 200 pcm in eigenvalue and 2% in cross sections. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

MOOSE framework enhancements for meshing reactor geometries

MOOSE is an open-source, parallel finite element framework designed to permit rapid development of robust multi-physics modeling capabilities. Under the DOE-NEAMS program, numerous solvers have been developed utilizing the open-source MOOSE framework for multiphysics reactor analysis. These solvers require input finite element meshes representing the discretized geometry. Typically, reactor analysts turn to licensed external tools for creation of reactor geometry meshes. Recently, enhancements have been added to the MOOSE framework to mesh common reactor geometries and improve MOOSE-based application user workflows. Support for hexagonal pins, assemblies, and cores has been added, and Cartesian support has been extended. Options for modeling static and rotating control drums within a hexagonal assembly are now available. Pin, assembly, and plane regions can be identified through automatically applied tags on the mesh called 'reporting IDs' for easier post- processing of physics results. An external open-source triangle routine has been leveraged within MOOSE to mesh core periphery zones. A set of reactor geometry builder routines further streamline the construction of hexagonal and Cartesian cores and include the ability to assign materials to regions during mesh generation. The new meshing routines are available through the MOOSE framework in the open-source 'Reactor' module, and the resulting directly within MOOSE-based applications or exported as Exodus II files for use in other finite element solvers. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Release on the Virtual Test Bed of an MSRE thermal hydraulics model

The nuclear industry is taking leaps in innovations with companies seeking a sustainable energy future through advanced nuclear reactors. The \gls{DOE}’s \gls{neams} program seeks to substantiate and bolster the deployment of advanced reactors through flexible multifidelity, multiphysics simulations of advanced nuclear reactors. Applications like SAM for one-dimensional systems thermal-hydraulics, and Pronghorn for multidimensional coarse mesh thermal-hydraulics, are geared to support innovations in industry by facilitating design, optimization, and licensing of advanced nuclear reactors. Coupling systems thermal-hydraulics and computational fluid dynamics codes can be difficult as the pressure coupling converges slowly; however, it is important to obtain the desired accuracy in each part of the primary loop. The authors of this model created an \gls{odc}~\cite{Mau23} approach to coupling SAM and Pronghorn. Leveraging this coupling technique, a \gls{msre} model was developed and released to the \gls{neams}/\gls{nric} \gls{vtb}. The \gls{msre} was chosen to be modeled because of the wealth of experimental data available and because of the strong physics coupling between the core and primary circuit \cite{doi:10.13182/NT8-2-118}. This document contextualizes the history of the \gls{msre}, describes the thermal hydraulics models used, and detail the implementation of multidimensional thermal-hydraulics and system codes based on the \gls{odc} method~\cite{Penn} for the \gls{msre} model. Finally, this document presents how other modelers could apply the SAM-Pronghorn \gls{odc} for other advanced reactor models. Current item is the set of slides for ANS Winter 23. The release of the model and the ANS summary have already been approved

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Reactor physics benchmark experiments at the JSI TRIGA MARK II reactor - Current status and future outlook

Full text of publication follows. With the development of new high-fidelity computational methods, improvement of nuclear data, and multiphysics modelling, there is an increased need for benchmark experiments to experimentally validate the models, methods and input data. Many of the nuclear facilities designed to perform reactor physics benchmark experiments have been shut down. Therefore, research reactors offer a great opportunity for benchmark experiments, if they are well designed and performed with great care and accuracy. In this presentation we provide an overview of the past and ongoing activities related to benchmark experiments at the Jozef Stefan Institute TRIGA Mark II research reactor. The following experiments have been performed: criticality with fresh fuel, {sup 197}Au(n,γ) and {sup 27}Al(n,α) reaction rates in irradiation channels, absolute and relative {sup 197}Au(n,γ), {sup 235}U(n,f) and {sup 238}U(n,f) reaction rates in the core, burnup, kinetic parameters, control rod worth, isothermal reactivity coefficient, self-shielding, slow and fast (pulse) transients, nuclear heating, delayed and prompt gamma ray production, temperature profiles for multi-physics. Since the existing fleet of research reactors is ageing very rapidly and new experiments are needed, new research reactors should be designed and built to meet the needs of future advanced reactors, education and training, and other technologies in the coming years. We will review planned activities at the JSI TRIGA reactors and plans for the new research reactor in Slovenia. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Transient convective delayed neutron precursor of {sup 235}U for the molten salt reactor experiment

This work investigates the delayed neutron precursors from the fission of {sup 235}U in molten salt reactors. The six delayed neutron groups predict the spatial distribution of the advective transport for radioactive mass transfer in the molten-salt reactor experiment (MSRE). The Mole code, using the framework of the Multiphysics Object-Oriented Simulation Environment (MOOSE), was used to compute the delayed neutron concentration of the whole system. This approach can be used for analysis of reactivity in circulating conditions. The effect of delayed neutron concentrations on the static or dynamic behavior of the system in MSRE was analyzed. Using this approach, the pattern and validation of delayed neutron for a range of parameters of interest are reported and discussed herein. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Fluid Properties for MOOSE

The Multiphysics Object-Oriented Simulation Environment (MOOSE) enables a wide range of advanced nuclear reactor simulations. Under the guidance of MOOSE's Finite Volume Team, we worked on the fluid properties module. Significant contributions include enabling Tabulated Fluid Properties (TFP) for systems thermal hydraulics analysis, Temperature and Pressure functionalized Fluid Properties, and Lead & Lead-Bismuth properties. Along with improving the capabilities of the fluid properties module, we also improved the documentation to allow for future users and developers to understand how the module works.

97 MATHEMATICS AND COMPUTING↗