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DOE OSTI2016

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7 LiF-BeF 2 ) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Corrosion of Structural Materials for Advanced Supercritical Carbon- Dioxide Brayton Cycle

The supercritical carbon-dioxide (referred to as SC-CO 2 hereon) Brayton cycle is being considered for power conversion systems for a number of nuclear reactor concepts, including the sodium fast reactor (SFR), fluoride saltcooled high temperature reactor (FHR), and high temperature gas reactor (HTGR), and several types of small modular reactors (SMR). The SC-CO 2 direct cycle gas fast reactor has also been recently proposed. The SC-CO 2 Brayton cycle (discussed in Chapter 1) provides higher efficiencies compared to the Rankine steam cycle due to less compression work stemming from higher SC-CO 2 densities, and allows for smaller components size, fewer components, and simpler cycle layout. For example, in the case of a SFR using a SC-CO 2 Brayton cycle instead of a steam cycle would also eliminate the possibility of sodium-water interactions. The SC-CO 2 cycle has a higher efficiency than the helium Brayton cycle, with the additional advantage of being able to operate at lower temperatures and higher pressures. In general, the SC-CO 2 Brayton cycle is well-suited for any type of nuclear reactor (including SMR) with core outlet temperature above ~ 500°C in either direct or indirect versions. In all the above applications, materials corrosion in high temperature SC-CO 2 is an important consideration, given their expected lifetimes of 20 years or longer. Our discussions with National Laboratories and private industry early on in this project indicated materials corrosion to be one of the significant gaps in the implementation of SC-CO 2 Brayton cycle. Corrosion can lead to a loss of effective load-bearing wall thickness of a component and can potentially lead to the generation of oxide particulate debris which can lead to three-body wear in turbomachinery components. Another environmental degradation effect that is rather unique to CO 2 environment is the possibility for simultaneous occurrence of carburization during oxidation of the material. Carburization can potentially lead to embrittlement of structural alloys in SC-CO 2 Brayton cycle. An important consideration in regards to corrosion is that the temperatures can vary widely across the various sections of the SC-CO 2 Brayton cycle, from room temperature to 750°C, with even higher temperatures being desirable for higher efficiencies. Thus the extent of corrosion and corrosion mechanisms in various components and SC-CO 2 Brayton cycle will be different, requiring a judicious selection of materials for different sections of the cycle. The goal of this project was to address materials corrosion-related challenges, identify appropriate materials, and advance the body of scientific knowledge in the area of high temperature SC-CO 2 corrosion. The focus was on corrosion of materials in SC-CO 2 environment in the temperature range of 450°C to 750°C at a pressure of 2900 psi for exposure duration for up to 1000 hours. The Table below lists the materials tested in the project. The materials were selected based on their high temperature strength, their code certification status, commercial availabilities, and their prior or current usage in the nuclear reactor industry. Additionally, pure Fe, Fe-12%Cr, and Ni-22%Cr were investigated as simple model materials to more clearly understand corrosion mechanisms. This first phase of the project involved testing in research grade SC-CO 2 (99.999% purity). Specially designed autoclaves with high fidelity temperature, pressure, and flow control capabilities were built or modified for this project.

36 MATERIALS SCIENCE↗

DOE OSTI2018

MELCOR Modeling of Non-LWR Systems Draft Report for the Nuclear Regulatory Commission

This report provides an overview of technical issues and design features relevant to advanced reactors and reviews MELCOR's current readiness for modeling accidents in such reactor types. This report describes advanced reactor physics models currently available or under development, and gauges the level of effort required to develop new models and capabilities applicable to assessing advanced reactor safety issues. Finally, this report reviews the available database that can be used in verification and validation of new models. Four general advanced reactor types are considered in this report: 1) High Temperature Gas-Cooled Reactor (HTGR); 2) Sodium Fast Reactor (SFR); 3) Molten Salt Reactor (MSR); 4) Fluoride Salt-Cooled High Temperature Reactor (FHR)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Initial assessment of metallurgical interaction of clad/base metal systems

Molten Salt Reactors (MSRs), both liquid fuel and sold fuel (e.g., the fluoride salt-cooled high temperature reactors, FHR’s) offer several advantages relative to conventional, water-cooled nuclear power systems. However, currently approved ASME structural materials have not been optimized for use in molten salt environments and demonstrating materials performance in molten salts is challenging. Toward the goal of enabling MSR designs, corrosion resistant claddings offer an attractive solution to help mitigate corrosion concerns and enable the use of approved ASME materials. This report assesses metals and alloys that may be effective as corrosion resistant cladding with select structural alloys. The structural alloys of interest are capable of use at high temperatures ≥600°C for times ≥100,000 hours. Potential cladding alloys were assessed based on their corrosion resistance in both chloride and fluoride based salts, physical properties, mechanical properties, nuclear performance, and resistance to environmental-assisted cracking. Additionally both thermodynamic and kinetic computational modeling were used to assess interactions between the base metal and the cladding. Results to date have identified three candidate alloy systems that help enable MSR designs; nickel, molybdenum, and tungsten based claddings. This report discusses pros and cons of each class of cladding and outlines future work to develop these base metal / clad systems into industrial products suitable for use in commercial nuclear power systems.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Heat Storage Options for Sodium, Salt and Helium Cooled Reactors to Enable Variable Electricity to the Grid and Heat to Industry with Base-Load Reactor Operations

Electricity markets are changing rapidly because of (1) the addition of wind and solar and (2) the goal of a low-carbon electricity grid. These changes result in times of high electricity prices and very low or negative electricity prices. California has seen its first month where more than 20% of the time (mid-day) the wholesale price of electricity was zero or negative. This creates large incentives for coupling heat storage to advanced reactors to enable variable electricity and industrial-heat output (to maximize revenue) while the reactor operates at base load (minimize cost). Recent studies have examined coupling various types of heat storage to Rankine and Brayton power cycles. However, there has been little examination of heat-storage options between (1) the reactor and (2) the power-conversion system or industrial customer. Heat-storage systems can be incorporated into sodium , helium-, and salt-cooled reactors. Salt-cooled reactors include the fluoride-salt-cooled high-temperature reactor (FHR) with its solid fuel and clean coolant and the molten salt reactor (MSR) with its fuel dissolved in the salt. For sodium and salt reactors, it is assumed that a heat-storage system would be in the secondary loop between the reactor and power cycle. For helium-cooled reactors, heat storage can be in the primary or secondary loop. This report is a first look at the rational and the heat storage options for deploying gigawatt-watt hour heat-storage systems with GenIV reactors. Economics and safety are the primary selection criteria. The leading heat-storage candidate for sodium-cooled systems (a low-pressure secondary system with small temperature drop across the reactor core) is sodium coolant stored in large steel tanks, with the coolant transferring heat in channels. The design minimizes sodium volume in the storage and, thus, the risks and costs associated with sodium. For helium systems (high-pressure with large temperature drop across the core), the leading heat storage options are (1) varying the temperature of the reactor core, (2) steel or alumina firebrick in a secondary pressure vessel and (3) nitrate or hot-rock/firebrick at atmospheric pressure. For salt systems (low pressure, high temperatures, and small temperature drop across the reactor core) the leading heat-storage systems are secondary salts. In each case, options are identified and questions to be addressed are identified. In some cases there is a strong coupling between the heat-storage technology and the power cycle. The leading sodium heat-storage technology may imply changes in the power cycle. High-temperature salt systems couple efficiency to Brayton power cycles that may create large incentives for the heat storage to remain within the power cycle rather than in any intermediate heat transfer loop.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Initial Industry Collaborations of the Center of Excellence

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. In particular the NEAMS program is supporting the development of novel thermal-hydraulic codes. The Center of Excellence for Thermal Fluids Application in Nuclear Energy, launched in 2018, has as its key goals to serve as a front door to industry. The Center of Excellence for Thermal Fluids Applications in Nuclear Energy has recently launched a program to start collaborative efforts between the laboratories and industry with the objective of stimulating cooperation and increasing adoption of T/H tools developed under NEAMS by the industry-at-large. In particular, two industry partners agreed to participate in pilot short-term collaborations aimed at demonstrating the value of NEAMS tools to their designs. These are BWXT and Kairos who proposed projects related to HTGR and FHR designs respectively. In this report we present the results of these collaborations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

SAM User's Guide

The System Analysis Module (SAM) is a modern system analysis tool being developed at Argonne National Laboratory for advanced non-LWR safety analysis. It aims to provide fast-running, whole-plant transient analyses capability with improved-fidelity for Sodium-cooled Fast Reactors (SFR), Lead-cooled Fast Reactors (LFR), and Molten Salt Reactors (MSR) or Fluoride-cooled High-temperature Reactors (FHR). SAM takes advantage of advances in physical modeling, numerical methods, and software engineering to enhance its user experience and usability. It utilizes an object-oriented application framework (MOOSE), and its underlying meshing and finite-element library (libMesh) and linear and non-linear solvers (PETSc), to leverage the modern advanced software environments and numerical methods. This document provides a user’s guide, which will help users understand the input description and core capabilities of the SAM code. A brief overview of the code is presented, as well as how to obtain and run it. The input syntax for various parts of the code is provided. Additionally, a number of example problems, starting with simple unit component problems to problems with increasing complexity, are provided. Because the code is still under active development, this SAM User’s Guide will evolve with periodic updates.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2019

Experimental Validation of a Compact Double-walled Twisted-Tube Heat Exchanger Concept

This report summarizes the results of a 3-year NEUP 15-8667 project sponsored by the U.S. Department of Energy to Experimentally Validate the Compact Double-walled Twisted-Tube Heat Exchanger Concept to support the development of the Fluoride-salt-cooled High-temperature Reactor (FHR). The NEUP activities led by the University of New Mexico (UNM) in collaboration with Sandia National Laboratory (SNL) are split into two major parts, Part-I Simulation for optimized heat exchanger design and Part II Experimental evaluation of the HX performance. The executive summary lists the major findings of the simulation and experimental activities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Mechanistic Source Term Considerations for Advanced Non-LWRs

This report is a functional review of the radionuclide containment strategies of fluoride-salt-cooled high temperature reactor (FHR), molten salt reactor (IVISR) and high temperature gas reactor (HTGR) systems. This analysis serves as a starting point for further, more in-depth analyses geared towards identifying phenomenological gaps that still exist, preventing the creation of a mechanistic source term for these reactor types. As background information to this review, an overview of how a mechanistic source term is created and used for consequence assessment necessary for licensing is provided. How mechanistic source term is used within the LMP is also provided. Third, the characteristics of non-LWR mechanistic source terms are examined This report does not assess the viability of any software system for use with advanced reactor designs, but instead covers system function requirements. Future work within the Nuclear Energy Advanced Modeling and Simulations (NEAMS) program will address such gaps.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Development of HPWH System Model, Using All Microchannel Heat Exchangers

In this report, 24-hour energy factor (EF) and first-hour draw rating (FHR) simulations were conducted on two heat pump water heater systems (HPWHs), one using a fin-and-tube evaporator and wrapped-tank condenser coil with D-shape tubes, and the other using a micro-channel evaporator and wrapped-tank condenser coil with microchannel tubes. Alternative refrigerants as drop-in replacements of R-134a were evaluated, including R-1234yf, R-1234ze(E), propane (R-290), R-450A and R-513A. The results are summarized below: 1.Microchannel heat exchangers can decrease the system charge down to 30% relative to fin-and-tube evaporator and wrapped-tank D tubes.2.Having the same contact surface area to the water tank and evaporator frontal flow area, the system using microchannel heat exchangers (MHXs) results in 5% efficiency degradation. Due to the very small cross-sectional flow area of a microchannel tube, MHXs tend to have multiple parallel tubes to restrict the refrigerant pressure drop. The parallel pattern can’t utilize the water temperature stratification as effectively as the D-shape tubes, which are arranged in one or two circuits and counter flow to the water stratification. 3.Refrigerant drop-in system simulations were conducted for R-1234yf, R-1234ze(E), propane, R450A and R513A in comparison to R-134a. Among the alternatives, propane has the largest volumetric capacity, which results in shortest HPWH run time and highest first-hour rating. Propane leads to the smallest energy factor because its largest volumetric capacity overloads the heat exchangers. R-1234ze(E) has the smallest volumetric capacity, highest energy factor, but smallest first-hour rating. R-1234yf, R-450A and R-513A achieve similar performance indices as R-134a. 4.Propane requires the least refrigerant mass (45% relative to R-134a). If a wrapped-tank microchannel condenser and microchannel evaporator are used, the required system charge can be limited below 150 grams, which makes it feasible to use propane for domestic HPWH.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

DOE OSTI2019

Experimental and Modeling Investigation of Overcooling Transients that Include Freezing, in Fluoride-Salt Cooled High-Temperature Reactors (FHRs). Final Report

The goal of this project was an experimental and modeling investigation of overcooling transients that include freezing in Fluoride-Salt Cooled High-Temperature Reactors (FHRs). The most impactful product of this research collaboration between Professor Scarlat’s group at UW Madison and subsequently at UC Berkeley, and Dr. Rui Hu’s group at Argonne National Laboratory was the initial demonstration of SAM application to FHR transients, and the generation of a freezing module for the SAM code. The national laboratory and industry graduate student internships enabled by this project were valuable for workforce development, and also for knowledge transfer between university, national lab, and industry.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Survey and Assessment of Computational Capabilities for Advanced (Non-LWR) Reactor Mechanistic Source Term Analysis.

A vital part of the licensing process for advanced (non-LWR) nuclear reactor developers in the United States is the assessment of the reactor’s source term, i.e., the potential release of radionuclides from the reactor system to the environment during normal operations and accident sequences. In comparison to source term assessments which follow a bounding approach with conservative assumptions, a mechanistic approach to modeling radionuclide transport, which realistically accounts for transport and retention phenomena, is expected to be used for advanced reactor systems. As the designs of advanced reactors increase in maturity and progress towards licensing, there is a need to advance modeling and simulation capabilities in analyzing the mechanistic source term (MST) of a prospective reactor concept. In the present work, a survey is provided of existing computational capabilities for the modeling of advanced reactors MSTs. The following reactors are considered: high temperature gas reactors (HTGR); molten salt reactors (MSR) which include salt-fueled reactors and fluoride salt-cooled high temperature reactors (FHR); and sodium- and lead-cooled fast reactors (SFR, LFR). A review of relevant codes which may be useful in providing information to MST analyses is also completed, including codes that have been used for source term analyses of LWRs, as well as those being developed for other aspects of advanced reactor system modeling such as reactor physics, thermal hydraulics, and chemistry. A discussion of MST modeling capabilities for each reactor type is provided with additional focus on important phenomena and functional requirements. Additionally, a comprehensive survey is provided of tools for consequence modeling such as atmospheric transport and dispersion (ATD).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

SAM Theory Manual

The System Analysis Module (SAM) is an advanced and modern system analysis tool under development at Argonne National Laboratory for advanced non-LWR reactor safety analysis. It aims to provide fast-running, modest-fidelity, whole-plant transient analyses capabilities, which are essential for fast turnaround design scoping and engineering analyses of advanced reactor concepts. While SAM is being developed as a system-level modeling and simulation tool, advanced modeling techniques being implemented include a reduced-order three-dimensional module, pseudo 3-D conjugate heat transfer modeling in reactor core, flexible and multi-scale modeling of heat transfer between fluid and structures, in addition to the advances in software environments and design, and numerical methods. SAM aims to be a generic system-level safety analysis tool for advanced non-LWRs, including Liquid-Metal-cooled fast Reactors (LMR), Molten Salt Reactors (MSR), Fluoride-salt-cooled High- temperature Reactors (FHR), and High-Temperature Gas-cooled Reactors (HTGR). SAM takes ad- vantage of advances in physical modeling, numerical methods, and software engineering to enhance its user experience and usability. It utilizes an object-oriented computational framework (MOOSE), and its underlying meshing and finite-element library and linear and non-linear solvers, to leverage the modern advanced software environments and numerical methods. This document provides the theoretical and technical basis of the code to help users understand the underlying physical models (such as governing equations, closure models, and component models), system modeling approaches, numerical discretization and solution methods, and the overall capabilities in SAM. As new code capabilities and features are added, the SAM Theory Manual will be updated periodically to keep it consistent with the state of the development.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

SAM User’s Guide

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Overview of Advanced Reactor Simulation Capabilities to Support Demonstrations

The Virtual Test Bed (VTB) was established to support the deployment of advanced reactors by facilitating the use of state-of-the-art modeling and simulation capabilities. These tools are able to solve previously prohibitive challenge problems in greater levels of fidelity and in a multi-physics framework. A VTB repository was recently set up to host models for a wide variety of reactor types and using a range of NEAMS-developed software. Table I summarizes the different capabilities already available in the repository and up-and-coming near/long term ones. The repository currently hosts models for a wide variety of reactor types, including High-Temperature Gas Cooled Reactors (HTGR), Sodium Fast Reactors (SFR), Fluoride salt cooled High-temperature Reactors (FHR), Molten Salt Reactors (MSR), and Heat-Pipe Microreactors (HP-MR). It should be noted that work is ongoing for all of the models and each example problem is at a different state of maturity.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2021

Improvements of SAM Heat Transfer Models for Molten Salt-Cooled Pebble Bed Reactors

To support the development and utilization of the SAM code for molten salt-cooled pebble bed reactor safety analyses and licensing, selected improvements were made to the one-dimensional system-level heat transfer modeling in SAM with the goal of better capturing heat transfer phenomena specific to these designs. In consideration of molten salt as a coolant, new models were introduced to capture the enhanced radiative heat transfer that must be considered at the higher working temperatures. Radiative heat transfer is handled not only between two structural surfaces, but also in combination between the surfaces and the molten salt fluid, which has a higher opacity and absorption of radiative heat as compared to water. Several examples of analytical verification of the new radiative heat transfer methods are included in this report. Additionally, convective heat transfer correlations were selected from literature and included in SAM to account for the enhanced heat transfer expected in pebble bed geometry flows. These correlations are verified and demonstrated in the present work with a reference FHR example.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Mechanistic Source Term Considerations for Advanced Non-LWRs (Revision 1)

This report is a functional review of the radionuclide containment strategies of fluoride-salt-cooled high temperature reactor (FHR), molten salt reactor (MSR) and high temperature gas reactor (HTGR) systems. This analysis serves as a starting point for further, more in-depth analyses geared towards identifying phenomenological gaps that still exist, hindering the creation of a mechanistic source term for these reactor types. As background information to this review, an overview of how a mechanistic source term is created and used for consequence assessment necessary for licensing is provided. How a mechanistic source term is used within the Licensing Modernization Project (LMP) is also provided. Lastly, the characteristics of non-LWR mechanistic source terms are examined. This report does not assess the viability of any software system for use with advanced reactor designs, but instead covers system function requirements. Future work within the Nuclear Energy Advanced Modeling and Simulations (NEAMS) program will address such gaps. This document is an update of SAND 2020-6730. An additional chapter is included as well as edits to original content.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Fission Product Transport in TRISO Particles and Pebbles

This document demonstrates completion of the goals described in the technical narrative of the FOA project titled: ”Modeling and Simulation Development Pathways to Accelerating KP-FHR Licensing” regarding fission product transport in the Kairos-proposed fuel pebble by INL and Kairos Power. Showcased in this report are code developments and simulations in BISON that extend the state of the art in computation and understanding of fission product transport in a TRISO fuel particle and pebble. These enhancements lay the foundation for making predictions of fission product transport that can be used as input in the fuel licensing process. This was achieved by installing existing fuel material models originally used in PARFUME, developing a new failure probability method that is efficient and multi-dimensional, employing material homogenization, and expanding verification and validation simulations to demonstrate the efficacy of the work. All this work is leveraged to spotlight the main deliverable; a three-dimensional model and corresponding demonstration simulation of a pebble, which will serve as the starting point for models used to predict fission product release.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗