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DOE OSTI2023

Development of Surveillance Test Articles with Reduced Dimensions and Material Volumes to Support MSR Materials Degradation Management

This report details the efforts toward developing new surveillance test article designs with reduced dimensions and material volumes to support materials surveillance technology development for advanced reactors. Two fabrication methods for the surveillance test articles are described. Welded test articles were fabricated with 316H and A617 materials, and interlocking test articles were fabricated using A709 and titanium-zirconium-molybdenum (TZM). The preliminary results demonstrate the successful design and testing of the flat surveillance test articles. The report also describes ongoing efforts to use an induction heating test setup to increase the heat up and cool down rates in testing the surveillance test articles. A brief description of the planned FY-24 work is provided.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

FY23 status report on the development of new ASME Section III, Division 5 Class B rules

This report summarizes the work done in Fiscal Year 2023 on the development of the new American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Division 5, Class B rules to address the gaps identified for high temperature reactor designs. The summary of the design-by-analysis strain limit evaluation and creep-fatigue damage assessment is presented. The proposed design-by-analysis creep-fatigue damage calculation approach uses a new elastic follow-up-based Isochronous Stress Strain Curve stress relaxation procedure. This approach captures the elastic follow-up generated due to interactions of components with adjacent components, supports, and other connections in the power plant. A set of sample problems are selected to validate the proposed design-by-analysis rules for creep-fatigue damage assessment. The proposed Class B rules are evaluated against the Class A elastic design rules, and the experimental data obtained from a family of Simplified Model Test based key-feature test results. The proposed Class B creep-fatigue damage assessment methodology yields conservative design cycles estimates compared to the experimental results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Open Architecture for Cost Savings in Advanced Nuclear Reactors

Recently, nuclear power plant build projects in the West have run over budget due to high capital costs and schedule overruns. Compared to other sources of energy, nuclear power plants have higher capital costs. Reactors are often different at every site, resulting in a lack of standardization. Nuclear is expected to compete with other low carbon sources of energy which have lower capital costs making it essential for nuclear to develop ways of reducing costs. Strategies such as standardization, learning rates, modularization, and schedule reduction in advanced reactors can reduce nuclear costs by about 40%. Standardization as a way of cutting capital costs has been explored even in large nuclear power plants. Standardization of certain plant components can result in lower component and installation costs and higher learning from experience. Standardization can be achieved by adopting a criterion of key performance indicators and general design principles for a specific system or component such as the balance of plant. Modularization allows the construction of certain components of SMRs in a factory, which saves time, increases productivity, and encourages higher learning rates. Production learning decreases the time and the cost related to an activity. The potential for modularized components of advanced reactors to be manufactured in factories makes it conducive to achieving higher learning rates. Developing large-capacity nuclear programs through sequential builds cultivates a higher learning rate, which in effect may reduce schedule overruns. Open architecture has been identified as a way to drive standardization among advanced reactor designs and result in cost savings. Open architecture (OA) is defined as a design enabling a diverse supply chain by defining and publishing requirements of systems or equipment in functional and/or interface terms, utilizing technical standards in widespread use. Currently, the nuclear industry’s approach is to use closed architecture, making most designs proprietary. However, collaboration between various advanced reactor vendors and suppliers utilizing the concept of open architecture can result in modular and standardized architecture of subsystems or subcomponents of a nuclear power plant. Completely standardizing nuclear power plants may be impossible, however, certain common subsystems amongst the various reactor designs could be standardized and/or access a wider supply chain and leverage existing learning from other sectors. Open architecture will save time and allocate resources to the parts of the plants that have the most unique features. A key advantage of open architecture is its ability to improve production learning across advanced reactors (AR) types in the industry, by providing and utilizing the same kind of component. Sodium fast reactor (SFR), High Temperature Gas Reactor (HTGR) and Molten Salt Reactor (MSR) are the advanced reactors considered for this project. This paper aims to determine the cost savings in advanced reactor programs due to open architecture learning rate. This work is an extension of work done on light water reactor small modular reactors; the cost methodology was utilized to investigate the impact of open architecture on advanced reactors with a particular focus on sodium fast reactors. The cost data on sodium fast reactors used in the model presented the most adequate information required for the analysis.

Advanced Nuclear Reactors↗

DOE OSTI2024

Evaluation of the Simplified Assessment Peak Equivalent Stress Design Limit Probability of Failure

The ASME Boiler and Pressure Vessel Code Section III Division 5 Subsection HH Subpart A (HHA) outlines requirements and qualification methods for graphite core components in high temperature reactor environments. One such structural qualification method is the Simplified Assessment outlined in HHA-3220, which represents a comparison of service level specific equivalent stresses to allowable stress values derived from target probability of failures (POFs). The target POF is based on the structural reliability class (SRC) for the components and service condition of interest, and is used to establish design limits for combined membrane and peak equivalent stresses. Combined membrane stress limits are derived from two parameter Weibull distribution fits of the graphite grade specific tensile strength data and the target POF. The combined membrane stress limits are then scaled based on the ratio of flexural strength to tensile strength (referred hereinafter as Rtf) to establish the corresponding peak equivalent stress limits. It is desired that the target POF is at least maintained for both combined membrane and peak equivalent design limits, especially with the use of Rtf. The assessment presented herein examines tensile and flexural data for medium and fine grade graphites relative to the calculated combined membrane and peak equivalent stress limits per HHA-3220 and illustrates the simplified assessment method results in anticipated, or at least conservative, peak equivalent design stress limits.

36 - MATERIALS SCIENCE↗

DOE OSTI2025

AGC-2 Graphite Preirradiation Data Analysis Report

This report describes the specimen loading order and documents all preirradiation examination material property measurement data for graphite specimens contained within the Second Advanced Graphite Capsule (AGC 2) irradiation capsule. The AGC 2 capsule is the second in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens in order to garner quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear grade graphites. This testing will ascertain the in service behavior of the graphite for pebble bed and prismatic very high temperature reactor designs. Similar to the First Advanced Graphite Capsule (AGC 1) preirradiation examination report, material property tests were conducted on specimens from 18 nuclear grade graphite types. However, AGC 2 tested an increased number of specimens (i.e., 512) prior to loading them into the AGC 2 irradiation assembly. All AGC 2 specimen testing was conducted at Idaho National Laboratory from July 2009 to August 2010. This report also details the specimen loading methodology for graphite specimens inside the AGC 2 irradiation capsule. The AGC 2 capsule design requires “matched pair” creep specimens that have similar dose levels above and below the neutron flux profile mid plane. This provides similar specimens with and without an applied load. Analysis in this document utilizes the neutron flux profile calculated for the AGC 2 capsule design, the capsule dimensions, and the size (i.e., length) of the selected graphite specimens to create a stacking order that produces “matched pairs” of graphite specimens above and below the AGC 2 capsule elevation mid point, thus providing specimens with similar neutron dose levels.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Validation of RANS Models for Predicting Bypass Flow in Prismatic VHTRs

In prismatic Very-High Temperature Reactors (VHTRs), gaps occur between neighboring fuel assemblies referred to as bypass gaps. This work aims to measure the flow distribution between the bypass gaps between neighboring assemblies and the coolant channels within the assemblies. Computational fluid dynamics (CFD) was done using Reynolds-Averaged Navier Stokes (RANS) modeling in STAR-CCM+ to assess the ability of RANS models to accurately capture this phenomena. The RANS results were compared to experimental worked performed by Seoul National University (SNU) and Korea Atomic Energy Research Institute (KAERI) for validation. The CFD results showed good comparison with the experimental data available from SNU and flow split was characterized for multiple configurations, inlet mass flow rates, and bypass gap sizes.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Simulations of Bypass Flow in Prismatic VHTRs

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Coupled SAM/Griffin Model of a Reference Pebble Bed High-Temperature Gas Cooled Reactor for Multi-Physics Simulations

An effort has been dedicated to developing a reference model for multi-physics coupled simulation of the pebble bed high-temperature gas-cooled reactor (PB-HTGR) with SAM and Griffin computer codes for safety analysis and licensing purpose. The reference problem of a prototypical reactor design serves as the foundation for the U.S. NRC (Nuclear Regulatory Commission) to verify the adequacy of computer codes and evaluation models for specific reactor types. In this work, a SAM model of the HTR-PM reactor has been developed based on publicly available design information and the multi-dimensional Pronghorn model developed by Idaho National Laboratory. The SAM HTR-PM model includes a multi-dimensional core region and 0- D/1-D fluid components. Additionally, a simplified air RCCS loop is modeled for decay heat removal. The Griffin model is based on a recent work by Idaho National Laboratory. The coupling between the models is achieved through the MOOSE MultiApp system. Both steady-state and transient scenarios were simulated to demonstrate the coupled model’s capability for multi-physics simulations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

High Fidelity Simulations of Air-Cooled Reactor Cavity Cooling System

High Temperature Gas Reactor (HTGR) designs incorporate passive safety systems (e.g., the Reactor Cavity Cooling System [RCCS]) that utilize natural principles to manage heat dissipation from the reactor pressure vessel (RPV) during accidents or routine shutdowns. The industry community is experiencing a pressing need for advanced simulation tools that can accurately assess the performance of these types of systems. In the literature, a knowledge gap exists concerning high-fidelity data for the RCCS, and this gap is one of the areas of focus of the present study. This research focuses on a specific RCCS designed for General Atomics' Modular High-Temperature Gas Reactor (GA-MHTGR). Experimental studies on a scaled version (can be seen in Figure 1) of the air-cooled RCCS used in GA-MHTGR were conducted by the University of Wisconsin-Madison (UW-Madison). This work contributes to a broader initiative aimed at establishing a numerical benchmark based on the UW-Madison experiments.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Release of a High Temperature Engineering Test Reactor (HTTR) Steady-State Multiphysics Model to the Virtual Test Bed

The National Reactor Innovation Center (NRIC) accelerates the demonstration and deployment of advanced nuclear energy through inspiring stakeholders and the public, empowering innovators, and delivering successful outcomes. The Virtual Test Bed (VTB) supports this mission by providing an open-source repository of advanced test reactor models for industry, academia, and the public to utilize. During my time at Idaho National Laboratory (INL), I: • Verified the input files of the High Temperature Engineering Test Reactor • Generated computational results • Created user documentation for the HTTR • Uploaded a steady-state HTTR model to the Virtual Test Bed • Authored a conference publication on the model

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2026

Monitoring of Liquid Metal Reactor Heater Zones with Recurrent Neural Network Learning of Temperature Time Series

Advanced high-temperature fluid reactors (ARs), such as sodium fast reactors (SFRs) and molten salt cooled reactors (MSCRs) utilize high-temperature fluids at ambient pressure. To melt the fluid during reactor startup and prevent fluid freezing during cooldown, the thermal–hydraulic systems of such ARs include heater zones consisting of specific heaters with controllers, temperature sensors, and thermal insulation. The failure of heater zones due to insulation material degradation or improper installation, resulting in parasitic heat losses, can lead to fluid freezing. The detection of faults using a heat-transfer model is difficult because of a lack of knowledge of the experimental details. Data-driven machine learning of heater zone temperature time series offers a viable alternative. In this study, we benchmarked the performance of recurrent neural networks (RNNs) in an analysis of heat-up transient temperature time series of heater zones installed on a liquid sodium vessel. The RNN models include long short-term memory (LSTM) and gated recurrent unit (GRU) networks, as well as their bi-directional variants, BiLSTM and BiGRU. Anomalous temperature points were designated using a percentile-based threshold applied to residual fluctuations in the detrended temperature time series. Additionally, the impact of the exponentially weighted moving average (EWMA) method on detection accuracy was examined. The RNN models’ performance was assessed using precision, recall, and F 1 score metrics. Results demonstrated that RNN models effectively detect anomalies in temperature time series with the best models for each heater zone achieving F 1 scores of over 93%. To explain the variations in RNN model performance across different heater zones, we used Kullback–Leibler (KL) divergence to quantify the relative entropy between training and testing data, and the Detrended Fluctuation Analysis (DFA) to assess long-range temporal correlations. For datasets with strong long-range correlations and minimal relative entropy between training and testing data, GRU is the best-performing model. When the data exhibits weaker long-term correlations and a significant relative entropy between training and testing distributions, BiGRU shows the best performance. For the data sets with intermediate values of both KL divergence and DFA, the best performance is obtained with LSTM and BiLSTM, respectively.

gated recurrent unit↗

DOE OSTI2023

Release of a High Temperature Engineering Test Reactor (HTTR) Steady State Multiphysics Model to the Virtual Test Bed

In response to climate change, global governments and private industry have established a common goal of achieving net-zero emissions by 2050 \cite{osti_1865910}. This goal requires a reassessment of current energy demands and production methods. Reducing emissions at an affordable cost while maintaining grid reliability requires a nationwide collaborative effort among government and industry in the United States. Nuclear power is the leading low-carbon electricity generation method. In the past 50 years, the use of nuclear power has reduced carbon dioxide emissions by over 60 gigatons and has played a crucial role in the security of energy supply~\cite{IEA}. In the U.S., nuclear power accounts for 20\% of the electrical supply and provides energy reliably. Advanced reactors will operate at higher temperatures, operate more efficiently, utilize more energy stored within fuel, and reduce the amount of waste produced \cite{osti_1616270}. To face these challenges and goals, the U.S. Department of Energy has created an initiative to focus on the modeling and simulation tools to support future nuclear power plant design, licensing, and operations. The Virtual Test Bed (VTB)~\cite{vtb2023} was launched by the National Reactor Innovation Center (NRIC) in collaboration with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program to support the advanced nuclear reactor community. The VTB involves teams from both Idaho National Laboratory and Argonne National Laboratory and aims to provide example models for a broad range of both current and future advanced reactor designs. A feature of the VTB is the automatic testing of these models to ensure continued functionality as simulation tools are further developed. The VTB and the advanced reactor models documented there are important resources for this initiative. This work describes the inclusion of a new model on the VTB---a High Temperature Engineering Test Reactor (HTTR) steady-state model \cite{LABOURE2023109838}.

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 OSTI2025

A MOOSE-Based Model for Fission Product Transport and Source Term Estimation for High-Temperature Gas-Cooled Reactors

Thanks to fuel elements containing tristructural isotropic (TRISO) particles combined with a low core power density and passive feedback mechanisms leading to modest temperature rises in the event of accidental events, high-temperature gas-cooled reactors (HTGRs) offer a high degree of reliability in terms of fission product retention. While the anticipated source term for HTGRs is expected to be very low, it is important to provide a quantitative estimate of radiological releases during nominal and accidental conditions. Here, we propose a computationally efficient mechanistic source term methodology relying on the Multiphysics Object Oriented Simulation Environment (MOOSE) for tracking fission product transport from TRISO particles up to the coolant pressure boundary, as well as modeling the transport and potential deposition of these nuclides inside the reactor coolant loop. The proposed computational scheme is applied to estimate source term inventories for a representative 10-MW(thermal) prismatic high-temperature microreactor and is qualitatively compared against known release fractions. In addition to providing an alternate analysis tool, this MOOSE model can help reactor designers quantify the influence of key design parameters relevant for studies of radiological dose consequences.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Self-welding of Inconel 617 under high-pressure-high-temperature conditions for nuclear reactors

In response to continuously declining availability of fossil fuels, nuclear power has become an important alternative source of power generation. Growing application of nuclear power has relieved the shortage of electricity. However, it is also accompanied with new drawbacks, such as radioactive waste, and potential catastrophic consequences in case of major accidents. For nuclear power plants, severe accidents mostly originate from the reactor, making the reliability of the reactor of paramount importance to the overall safety of a nuclear power plant. Massive heat generation through nuclear fission/fusion and heat exchange render the whole reactor operating at extremely high temperatures (i.e., 900–1000 °C), which, in conjunction with large load from subsystems/ components made of special heavy alloys, renders the reactor under the risk of self-welding, a phenomenon that could lead to strong bonding between components and paralyze the operability of the nuclear reactor. This paper investigates the self-welding behavior of Inconel 617, a primary candidate nickel alloy for key components of the next generation gas cooled nuclear plant, inside a controlled atmospheric furnace with preset static load, simulating the high-pressure high-temperature conditions inside nuclear reactors. Therefore, test results show that Inconel 617 experiences self-welding through common oxide zone at the interface, and it has remarkable bonding strength, requiring 4114 N (57.15 MPa) and 4338 N (60.25 MPa) to break apart a mated sample pair, under an apparent contact pressure of 0.341 MPa after 50-h aging, in helium and air atmospheres, respectively. The higher oxygen presence in air results in slightly stronger oxide bond at the interface, compared to helium, and the bond strength increases with dwell time.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2023

Enhancing the Chemical Energy Flux in a High-Temperature Tubular Counterflow Solid Fuel Synthesis Reactor Using a Bypass

Redox reactions of metal oxides offer a path towards using intermittent renewable resources for high-density thermochemical energy storage. Thermochemical energy storage often involves the flow of a particulate media. We describe a novel method to increase the throughput in a gravity-driven high-temperature thermochemical storage reactor flowing pelletized MgMnO. The moving bed reactor operates under counter-flow conditions and encounters particle flowability problems at temperatures of 1500 °C leading to sintering of the bed. Inertial forces of a counter-flowing gas can overcome the gravitational forces on the particles and limit the chemical energy storage rate of the reactor. We found that the insertion of a gas bypass (a slotted tube) into the reactor results in a 100% increase of the flow rates and achieved a 50% higher chemical energy storage flux compared to the operation without a bypass tube while mitigating the effects of sintering on the particles. As a result, the higher solid flow rates require a longer heated zone to reach a comparable residence time and extent of reduction compared to the lower flow rates.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2024

A Framework to Assess Advanced Reactor Spent Fuel Management Facility Deployment

Advanced nuclear reactors offer various operational advantages over existing light water reactors but could produce types of spent nuclear fuel (SNF) with a wide variety of forms and characteristics depending on how many different concepts are deployed. Each advanced reactor SNF type potentially poses unique management challenges. New planning efforts will be necessary to anticipate how the management requirements of advanced reactor SNF will affect the deployment of an integrated waste management system. This paper applies a framework of high-level facility deployment milestones to a generic SNF management system, reviewing them together with the advanced reactor SNF characteristics and management requirements. This allows for the investigation of factors that influence facility and system deployment, and ultimately, the identification of challenges facing the deployment of different kinds of SNF management facilities. Here, the back end of the once-through fuel cycle is examined for four advanced reactor system technology types: sodium-cooled fast reactors, high-temperature gas-cooled reactors, liquid-fuel molten salt reactors, and lead-cooled fast reactors. It is observed that milestones earlier in the facility deployment process (e.g., siting and facility design) are more impacted by the uniqueness of advanced reactor SNF characteristics than others (e.g., construction and testing). Ultimately, none of the differences are seen as fundamentally disqualifying in a technical sense; however, they should be considered early, potentially as part of reactor design, to avoid issues in the future.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Editorial: Benchmark experiments, development and needs in support of advanced reactor design

Advanced nuclear reactor designs will for the most part be a departure from low enrichment light water reactor (LWR) designs currently operated around the world. Such advanced designs include but are not limited to new TRISO-fueled high temperature gas reactors, heat-pipe cooled micro-reactors, fluoride salt cooled high-temperature reactors, molten salt reactors, lead cooled fast reactors, nuclear thermal propulsion concepts, and include LWR designs with advanced fuel and clad types. Modeling and simulation methods for advanced reactors is necessary for regulators to approve license requests. However, regulators also require that modeling approaches be validated against experimental measurements. Hence, there is a crucial need for data for advanced reactor systems that will support validation of analysis methods. To this end, this Research Topic includes eleven papers organized into topical seven categories relevant for advanced reactor design.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗