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DOE OSTI2023

Fuel Flow Modeling under Thermal Fluidics Considerations within Operational Domains of Molten Salt Reactors

Molten Salt Reactors (MSRs) offer significant versatility supported by their design flexibility and safety advantages due to high temperatures, design customization capabilities, strong negative reactivity feedback effects, and salt chemistry characteristics. Different control schedules and different mass flow rates in the MSR primary systems impact neutronics and thermal hydraulics coupling. The result is two-fold. First, it leads to adaptability in various applications when the customization of a MSR unit is desired. Second, it calls for the quantification of dynamics and safety characteristics to assess the operational domain and make sure reactor stability and inherent safety characteristics are maintained. This paper provides an in-depth transient analysis of the operational implications of MSRs having heat-generating fuel and heat-transporting salt mixed forming liquid fuel salts contained within primary systems. A coupling model was developed to simulate fuel flow in molten salt reactors. A zero-dimensional reactor kinetics model was used with a one-dimensional heat transfer model in the reactor core. The temperature reactivity feedbacks of the fuel salt resulting from the thermal-hydraulics model were used in the reactor kinetics model to complete the coupled code. The model was applied to simulate the Molten Salt Breeder Reactor (MSBR) response at steady state and transients. The developed model was shown to be a suitable tool for the dynamic analysis of molten salt reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Alloy 709 Advanced Austenitic Stainless Steel

Provide material solutions to enabling design, construction, licensing and operation of advanced reactors Including Fast Reactors, Gas-cooled Reactors and Molten Salt Reactors (solid or liquid fuel) Could be of modular or micro designs, and from hundreds of MWe to kWe Provide technical bases needed for design and licensing of advanced reactor components Develop & validate improved high temperature design methodology Provide qualification data (to NQA-1 or equivalent) on structural materials Utilize consensus standards when appropriate (e.g., ASME, ASTM, etc.) Target resolution of issues needed for near to mid-term deployment of advanced reactors

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Characterization of Heat Transfer in a Trapped-Vortex Combustor Designed for High-Temperature Material Testing

Next generation gas turbines will reduce carbon emissions by improving their cycle efficiencies and implementing hydrogen as a low-carbon fuel. The combined-cycle efficiency can be improved by increasing the combustor outlet temperature and decreasing the cooling air required for hot-section parts. Ultra-high temperature ceramic matrix composites (CMC) can survive increasingly extreme environments due to their strong mechanical properties at expected conditions. The introduction of CMCs in high-hydrogen flame environments requires testing at realistic conditions that mimic the thermochemical and fluid mechanic states in gas-turbine combustors. In this work, we describe a new experiment that has been designed to test high-temperature materials in a combustor-relevant environment. Its trapped-vortex combustor chamber design allows for high levels of fuel flexibility and wide flame stability limits. Testing of monolithic silicon carbide (SiC) samples in this environment was done to characterize the heat transfer to material samples over a wide range of operating conditions, including variations in fuel composition and thermal power. Heat flux measurements were verified in non-reacting and reacting environments after which they were benchmarked with previous literature to verify the operation of the facility with heated air. Testing in combustor-relevant conditions shows high levels of heat flux to the material samples in the combustor, particularly downstream of the anchored flame.

Richins, Porter [Pennsylvania State University , ,↗

DOE OSTI2021

Advanced Materials Program Summary

Advanced Reactor Technologies (ART) Advanced Materials Program: Provide material solutions to enable design, construction, and operation of licensable advanced reactors Including Gas-cooled Reactors, Fast Reactors, Molten Salt Reactors (solid or liquid fuel) Could be of modular design, and from 350 to 1 Mwe Conduct developmental R&D on structural materials that are best addressed by the national program Integrate program-directed work at national labs & universities and collaboration with international partners to address advanced reactor developer needs Provide qualification data (to NQA-1 or equivalent) on structural materials and develop & validate improved high temperature design methodology Utilize consensus standards organizations when appropriate (e.g., ASME, ASTM, etc.) Target resolution of issues needed for near to mid-term deployment of advanced reactors

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Training material models using gradient descent algorithms

High temperature design requires accurate constitutive models to describe material inelastic deformation and failure behavior. Oftentimes, calibrating accurate models devolves into the problem of fitting the model parameters against experimental test data. Here, we present the pyopmat package, an open source framework for calibrating constitutive models against experiment data subjected to various loading conditions using machine learning techniques. The package calculates the exact gradient of the model response with respect to the parameters using a combination of automatic differentiation and the adjoint method. Given this exact gradient, we compare the performance of several gradient-based optimization techniques in fitting realistic constitutive models against data. Here, we demonstrate the efficiency and accuracy of our package through example problems using both synthetic data, generated using known parameter sets, under monotonic and cyclic loading conditions and also with an example applying the techniques developed here to actual high temperature creep-fatigue test data.

36 MATERIALS SCIENCE↗

DOE OSTI2023

Counterintuitive Catalytic Reactivity of the Aluminum Oxide “Passivation” Shell of Aluminum Nanoparticles Facilitating the Thermal Decomposition of exo -Tetrahydrodicyclopentadiene (JP-10)

High energy density aluminum nanoparticles (AlNP) have been at the center of attention as additives to hydrocarbon jet fuels like exo-tetrahydrodicyclopentadiene (JP-10, C 10 H 16 ) aiming at a superior performance of volume-limited air-breathing propulsion systems. However, a fundamental understanding of the ignition and combustion chemistry of JP-10 in presence of AlNPs has been elusive. Exploiting an isomer-selective comprehensive identification of the decomposition products in a newly designed high temperature chemical microreactor coupled to vacuum ultraviolet photoionization, we reveal an active low temperature heterogeneous surface chemistry commencing at 650 K involving the alumina (Al 2 O 3 ) shell. Contrary to textbook knowledge of an “inactive alumina surface”, this unconventional reactivity, where oxygen is transferred from alumina to JP-10, leads to generating cyclic, oxygenated organics like phenol (C 6 H 5 OH) and 2,4-cyclopentadiene-1-one (C 5 H 4 O) - key tracers of an alumina-mediated interfacial chemistry. Furthermore, this counterintuitive reactivity transforms our knowledge on the (catalytic) processes of alumina-coated AlNPs on the molecular level.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2021

A Minimum Creep Rate for 2-1/4Cr-1Mo Steel Consistent With the ASME Section III, Division 5 Rules

This technical note describes a minimum creep rate model for 2-1/4Cr-1Mo steel that is consistent with the current creep strain equation embedded in the ASME Boiler & Pressure Vessel Code Section III, Division 5, Subsection HB, Subpart B isochronous stress–strain curves. Minimum creep rate models for all the Section III, Division 5 Class A materials are required for the development of improved high temperature design methods. Of all the Class A materials, only 2-1/4Cr-1Mo does not have a readily identifying minimum creep rate term in the current isochronous stress–strain curve model.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Recommendation for Limiting Conditions for ASME BPVC Section III Division 5 Allowable Stress Criteria

Section III, Division 5 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) specifies rules for elevated temperature nuclear reactors. Time dependent allowable stresses are an important aspect of the high temperature design models and are related to the time to 1% strain, time to rupture, and time to onset of tertiary creep. The time to onset of tertiary creep criterion has an overly strong negative influence on the allowable stresses for some structural materials, likely due to very limited datasets and misidentification of microstructural changes rather than onset of significant creep damage, as tertiary creep is meant to represent. It is recommended that the onset of tertiary creep be determined through an alloy specific ratio between onset of tertiary creep and creep rupture time, which would give access to the significantly larger creep rupture time datasets.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Class B Materials Code Case

Class B Materials Code Case to include: FY25 - High temperature design methodology, overview of ASME Section III (Division 5), Design for elevated temperature service, Goal to address current gaps, evaluation of proposed rules, allowable stress development, Class A materials (existing ASME approach), variable confidence index, data extrapolation, bounding value for SEE, and next steps.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Design of a Combined Proportional Integral Derivative Controller to Regulate the Temperature Inside a High-Temperature Tubular Solar Reactor

Solar fuels are proven to be promising candidates for thermochemical energy storage. However, the transient nature of solar radiation is an obstacle to maintaining a stable operational temperature inside a solar reactor. To overcome this challenge, the temperature of a solar reactor can be regulated by controlling the incoming solar radiation or the feedstock flowrate inside the reactor. In this work, a combined proportional integral derivative (PID) controller is implemented to regulate the temperature inside a high-temperature tubular solar reactor with counter-current flowing gas/particles. The control model incorporates two control systems to regulate incoming solar radiation and gas flow simultaneously. The design of the controller is based on a reduced-order numerical model of a high-temperature tubular solar reactor that is vertically oriented with an upward gas flow and downward particle flow. The reactor receives heat circumferentially through its wall over a finite segment of its length. Formulation of the heat transfer model is presented by applying the energy balance for the reactor tube and considering heat and mass transfer inside. A set of governing differential equations are solved numerically by using the finite volume method to obtain reactor wall, particles, and gas temperatures along the reactor length with various boundary conditions. Simulation results are used to tune the PID controller parameters by utilizing the Ziegler–Nichols tuning method. Both the simulation results and the controller performance are visualized on the LABVIEW platform. The controller is challenged to track different temperature setpoints with different scenarios of transient solar radiation. Here, the performance of the PID controller was compared to experimental results obtained from an industrial PID controller embedded in a 7 kW electric furnace. Results show that the combined PID controller is successful in maintaining a stable temperature inside the reactor by regulating the incoming solar radiation and the flowrate via small steady-state error and reasonable settling time and overshoot.

14 SOLAR ENERGY↗

DOE OSTI2021

Development of High Temperature (>700°C) molten Salt Pump Technology for Gen3 Solar Power Tower Systems

Generation 3 Concentrating Solar Power (CSP) Systems require thermal energy storage that can efficiently receive solar heat and deliver it to a working fluid at temperatures greater than 700°C. For liquid systems, the thermal transfer fluid must be phase stable over large temperature ranges. High temperature molten salts meet this criteria, but are difficult liquids to handle. The high heat and salt content can be very corrosive to several different materials that are wetted by the liquid salt. For power tower collectors, the molten salt must be pumped under high pressure to the top of the collector tower. These pumps have impellers and shafts as part of their rotating elements, and under load, require bearings to control shaft movement while rotating. These high temperature pumps require submerged bearings, and thus bearing materials that can withstand the temperature and molten salt contact over long periods of time. This is essential for keeping pump maintenance costs low, and downtime to a minimum. Long running bearing materials is an enabling technology for integrated liquid systems in Generation 3 CSP Systems. Keeping costs low and systems reliable are critical for the broad adoption of Gen 3 CSP Systems. Hayward Tyler, Inc (HTI) proposed development of journal bearing materials for use in vertical pumps designed for pumping high temperature molten salt on both the hot and cold sides of the solar power tower. To complete this objective, HTI utilized the testing capabilities of Oak Ridge National Laboratory (ORNL) and High Temperature System Design (HTSD). Materials would first be selected for static corrosion testing, then downselected for tribological testing. Successful tribologically tested materials would then be downselected for use in additional test regimes. Ten samples were selected for static corrosion testing. Three passed the static testing regime. Two of these materials were downselected for ball material in tribological testing against three metallic alloys: Sintered Y2O3 Partially Stabilized Zirconia YTZP and Silicon Nitride, Grade 147-31N. HTI also proposed collecting data from a thorough bearing test regime using a test rig designed and built as part of this project, and also a conceptual design for a high temperature molten salt vertical pump. HTI withdrew from the project prior to completion of this work and this report reflects only the work completed prior to withdrawal. The completed work produced two papers, one manuscript submitted for publication, and one paper published in the journal Solar Energy Materials and Solar Cells. The papers are: “Tribological behavior of ceramic-alloy bearing contacts in molten salt lubrication for concentrating solar power” and “Material Selection and Corrosion Studies Of Candidate Bearing Materials For Use In Molten Chloride Salt."

14 SOLAR ENERGY↗

DOE OSTI2024

An Innovative Design of High-Temperature, Sensible Molten Salt Thermal Energy Storage Systems With Geopolymer Insulation

This project aims to mitigate the risks of the current molten salt TES tank designs. The goal is to create a more reliable and cost-effective TES by (1) developing a new material solution for internal thermal insulation and (2) by designing a concrete-based tank structure to eliminate metal-related issues (e.g., corrosion, thermomechanical failure, etc.). Firstly, the project aims to engineer a composite refractory insulation material that possesses sufficient thermal insulation properties without significantly sacrificing the resistance to salt permeation and mechanical strength, which is the major trade-off in traditional refractory insulation design. The primary objective is to perform materials design and optimization of the composite refractory to achieve the required thermal and mechanical properties as suitable internal insulation for molten salt TES. Secondly, the project aims to design a concrete-based TES system with conventional low- cost Portland cement serving as the primary tank structure and the aforementioned composite refractory as the internal thermal insulation and salt containment. The objective is to investigate the technical and economical feasibilities of concrete tanks for high- temperature molten salt storage applications.

14 SOLAR ENERGY↗

DOE OSTI2020

HDG-1 Graphite Preirradiation Data Package Report

This report documents all pre-irradiation examination material-property measurement data for graphite specimens that are going to be used within the first high dose graphite (HDG) -1 irradiation capsule. The two new HDG capsules signify a major change to the AGC Experiment. HDG-1 and HDG-2 will replace the last two Advanced Graphite Creep (AGC) capsules (AGC-5 and AGC-6) which were designed to irradiate graphite at the extreme upper operational temperatures for a very-high-temperature reactor (VHTR) design, 1100°C. These very high temperature AGC-5 and AGC-6 capsules have been repurposed to re-irradiated specimens (from AGC-2, AGC-3, and AGC-4) at the lower temperatures of 600°C and 800°C. HDG-1 will be irradiated at 600°C and HDG-2 will be irradiated at 800°C. By re-irradiating the previous AGC specimens a total maximum neutron dose of around 15 dpa (displacements per atom) can be achieved for all major graphite grades at irradiation temperatures of 600°C and 800°C. Specimens in the HDG-1 capsule are made up of previously irradiated specimens from the AGC-2 capsule and unirradiated specimens prepared for the now discontinued AGC-5 capsule. Utilizing the irradiated specimens, a maximum neutron dose of around 15 dpa is anticipated. These new maximum dose levels will provide irradiated material property data over a total neutron dose range of 1-15 dpa at a temperature of 600°C when combined with the previous AGC-1 and AGC-2 irradiation data. This will provide quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades for use within high temperature reactor designs. Similar to previous AGC test trains, HDG-1 includes the major graphite grades (IG-110, NBG-17, NBG-18, PCEA, and 2114) as well as adding the very fine-grain grade IG-430 which is of interest to the Molten Salt Reactor (MSR) designs. Also new to the HDG-1 capsule are 90 smaller geometry specimens designated as pencil specimens. These specimens take up only one third the space of a standard creep size specimen. This increased number of specimens will enhance property measurement statistics because they will provide 3 times the control specimen data at a position that would otherwise only have a single measurement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

HDG-1 Graphite Preirradiation Data Package Report

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Design of a high-temperature cell for cobalt-60 irradiations of aqueous solutions with in situ UV–visible spectroscopy

To understand the speciation of solutes in aqueous solutions in high temperature radiation environments, we report the design and fabrication of a custom-built, high temperature (≤300 °C) titanium irradiation cell with in situ optical spectroscopy capabilities, as afforded by coupled fiber optic cables. The wetted surfaces of the 8-inch tall cylindrical cell with 3.5 in. diameter are entirely made of titanium, sapphire, and gold, which are chemically and radiolytically inert. The initial benchmarking results are reported, including the baseline spectrum of deionized water as a function of temperature, the stability of a spectrum over 4 h at 100 °C, and an irradiated Fricke dosimetry solution under ambient irradiator temperature conditions (27.0 ± 0.5 °C). Further, the average gamma radiation dose rate in the cell in its current configuration is 26.1 ± 1.3 Gy min -1 . This cell has application in studying several processes throughout the nuclear fuel cycle, including the reactor coolant behavior.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2022

Preliminary Design Needs for High-Temperature Heat Pipe Imaging System

This report provides preliminary design needs for radiological imaging heat pipes to perform in operando assessment for high-temperature alkaline metal charged heat pipes (HPs). The measurement parameters of interest have been summarized for characterizing the operando HP behavior. Based on the current SPHERE experimental facility, the needs and modifications required for the experimental setup have also been identified to enable the X-ray-assisted visualizations for the single HP during its operating conditions.

42 ENGINEERING↗

DOE OSTI2025

Graphite waste classification and disposal cost estimation for high temperature gas and salt reactors

As high-temperature reactor designs progress to demonstration, managing the radioactive wastes from these systems presents unique challenges. This work explores the irradiated graphite source term produced by three reactor designs: The Modular High Temperature Gas reactor (MHTGR), a pebble-bed High Temperature Gas Reactor (pb-HTGR), and a Fluoride-cooled High-temperature Reactor (FHR). We predicted a C-14 concentration of 4.3 Ci/m 3 for the MHTGR, 1.2 Ci/m 3 for the pebble bed HTGR, and 2.5 Ci/m 3 for the gFHR after 20 years of operation. The final C-14 concentration highly depended on the graphite nitrogen impurity, a major precursor for C-14. The C-14 concentration in all reactor types exceeded the 0.8 Ci/m3 threshold, resulting in a Class C waste classification. The costs associated with accepting the graphite after 20 years in a low-level waste disposal facility were projected to be 255 dollars per kWe for the MHTGR, 248 dollars per kWe for the pb-HTGR, and 56.8 dollars per kWe for the FHR.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Designing Cladded Components for High Temperature Nuclear Service: Part I—Analysis Methods

Use of corrosion-resistant cladding can greatly extend the design life of structural components in many advanced reactor systems. However, there are currently no ASME design rules for cladded components to guard against creep-fatigue failure and ratcheting strain accumulation in elevated temperature nuclear service. This paper, presented in two parts, addresses this gap by proposing a design strategy for cladded components that do not require long-term testing of clad materials. The proposed approach relies on approximate design analysis methods for two types of clad materials—a soft clad that creeps much faster than and has lower yield stress than the class A base material and a hard clad that creeps much slower than and has higher yield stress than the class A base material. Part I discusses the approximate analysis strategies for the clad materials—treat a soft clad as perfectly compliant and a hard clad as linear elastic—and Part II develops a complete set of design rules for each of the two types of cladded components. In conclusion, finite element analyses of representative high temperature reactor components show that the proposed design analysis methods can bound the design quantities in soft cladded components and approximate the design quantities in hard cladded components.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗