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At least 883 records · Page 49

SPC-70646 Specification for Reactor Supplemental Shielding for Use in DOME

NRIC is developing the Demonstration of Microreactor Experiments (DOME) test bed at the Idaho National Laboratory (INL) Materials and Fuels Complex (MFC): to allow for testing of advanced reactors in support of future licensing and commercial operations. One such need is a neutron and photon radiation shielding system to protect personnel and equipment from harsh neutron and photon fluxes during reactor operations and post shut down operations, disassembly, and decommissioning.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

An Integrated Framework for Risk Assessment of Safety-related Digital Instrumentation and Control Systems in Nuclear Power Plants: Methodology Refinement and Exploration

This report documents activities performed by Idaho National Laboratory (INL) during Fiscal Year (FY) 2023 for the U.S. Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program, Risk Informed Systems Analysis (RISA) Pathway, digital instrumentation and control (DI&C) risk assessment project. In FY 2019, the RISA Pathway initiated a project to develop a risk assessment strategy for delivering a technical basis to support effective, and secure DI&C technologies for digital upgrades/designs. A risk assessment-informed framework was proposed for this strategy, which aims to (1) provide a best-estimate, risk informed capability to quantitatively estimate the safety margin obtained from plant modernization, especially for safety-related DI&C systems, (2) support and supplement existing risk informed DI&C design guides by providing quantitative risk information and evidence, (3) offer a capability of design architecture evaluation of various DI&C systems, (4) assure the long-term safety and reliability of safety-related DI&C systems, and (5) reduce uncertainty in costs and support integration of DI&C systems in the plant. To achieve these technical goals, the LWRS-developed framework provides a means to address relevant technical issues by: (1) defining a risk informed analysis process for DI&C upgrade that integrates hazard analysis, reliability analysis, and consequence analysis, (2) applying risk informed tools to address common cause failures (CCFs) and quantify corresponding failure probabilities for DI&C technologies, particularly software CCFs, (3) evaluating the impact of digital failures at the component level, system level, and plant level, and (4) providing insights and suggestions on designs to manage the risks, thus to support the development and deployment of advanced DI&C technologies in nuclear power plants (NPPs). Adding diversity within a system or components is the primary means to eliminate and mitigate CCFs, but diversity also increases system complexity and may not address all sources of systematic failures. Optimization of diversity and redundancy applications for the safety-critical DI&C systems remains a challenge. To deal with the technical issues in addressing potential software CCFs in safety-related DI&C systems of NPPs and supporting relevant design optimization, the proposed framework provides: (a) A best-estimate, risk informed capability to address new technical digital issues quantitatively, focusing on software CCFs in safety-related DI&C systems of NPPs; (b) A common and a modularized platform for DI&C designers, software developers, cybersecurity analysts, and plant engineers to predict and prevent risk in the early design stage of DI&C systems; (c) Technical bases and risk informed insights to assist users address the risk informed alternatives for evaluation of CCFs in safety-related DI&C systems of NPPs; and (d) A risk informed tool that offers a capability of design architecture evaluation of various DI&C systems to support system design decisions in diversity and redundancy applications. The research and development efforts of this project in FY 2023 are focused on refining current methods on software CCF modeling and estimation and exploring additional innovative approaches to risk assessment of DI&C systems to enable a more comprehensive and complete assessment of various safety-related DI&C design architectures. The primary audience of this report are DI&C designers, engineers, and probabilistic risk assessment (PRA) practitioners. This includes stakeholders, such as the nuclear utilities and regulators who consider the deployment and upgrade of DI&C systems, DI&C software developers and reviewers, and cybersecurity specialists. It should be noted that all the analyses are performed for the demonstration of the methodology, not for the evaluation of an actual digital control system. Results are obtained based on limited design information and testing data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Summary of Technical Peer Review on the Risk Assessment Framework proposed in Report INL/RPT-22-68656 for Digital Instrumentation and Control Systems

This report summarizes the peer review activities initiated by Idaho National Laboratory (INL) during fiscal year (FY) 2023 for the evaluation and improvement of the methodology developed under the U.S. Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program, Risk Informed Systems Analysis (RISA) Pathway, digital instrumentation and control (DI&C) risk assessment project. In FY 2019, the RISA Pathway initiated a project to develop a risk assessment strategy for delivering a technical basis to support effective and secure DI&C technologies for digital upgrades/designs. A framework was proposed for this strategy, which aims to (1) provide a best-estimate, risk-informed capability to quantitatively and accurately estimate the risk impact of plant modernization, considering the introduction of high safety-significant safety-related (HSSSR) DI&C systems, (2) support and supplement existing risk-informed DI&C design guides by providing quantitative risk information and evidence, (3) offer a capability of design architecture evaluation of various DI&C systems, (4) assure the long-term safety and reliability of HSSSR DI&C systems, and (5) reduce uncertainty in costs and support integration of DI&C systems in the plant. The R&D efforts of this project from FY 2019 through FY 2022 were focused on methodology improvement and demonstration of the proposed framework for the risk assessment and design optimization of safety-critical DI&C systems. Collaborations with the nuclear industry have been initiated to support the reliability and risk assessment of their DI&C systems by using the proposed framework. In FY 2023, the framework has reached to a point for a technical peer review and obtain stakeholder feedback. This peer review activity includes coordination of the reviews performed by a group of industry stakeholders, documentation of the peer review suggestions, providing resolutions and responses to the peer review comments. The objective of this technical peer review is to obtain representative feedback on the proposed framework to improve the technical qualities of its methodology and readiness for deployment to the industry. Feedback may identify potential areas for improvement and further development. The Subject Matter experts were invited to review the latest project report documenting the methodology developed in the project and provide evaluations of the technical qualities of the proposed framework and relevant methods. The reviewed project report is “An Integrated Framework for Risk Assessment of High Safety-significant Safety-related Digital Instrumentation and Control Systems in Nuclear Power Plants: Methodology and Demonstration” INL/RPT-22-68656 (short as “INL/RPT-22-68656” in this report). This peer review report documents the technical questions provided for technical peer review and introduces the technical peer reviewers from the stakeholders including nuclear utilities, regulators, and universities. Comments from technical peer reviewers and the resolutions and responses to these comments are outlined. Insights and lessons learned from the technical peer review are summarized in conclusions and future work. The primary audience of this report are DI&C designers, engineers, and probabilistic risk assessment (PRA) practitioners. This includes stakeholders, such as the nuclear utilities and regulators who consider the deployment and upgrade of DI&C systems, DI&C software developers and reviewers, and cybersecurity specialists.

99 GENERAL AND MISCELLANEOUS↗

First-Principles Cost Estimation of a Sodium Fast Reactor Nuclear Plant

A multi-tiered cost analysis is performed to estimate full costs of a nuclear power plant (NPP) based on sodium-cooled fast reactor (SFR) technology. To address the lack of fully transparent cost estimations from past undertakings for NPPs, we have developed a detailed and first-principles-based cost estimate for a generalized SFR NPP. Our intent is to achieve a high degree of transparency with our cost assumptions and develop a cost model that is flexible and easily extendable to variations in NPP design and other nuclear reactor types. Furthermore, we strive to achieve a clear organization of costs and complete identification of key cost drivers based on first principles. To this end, the cost results of our analysis as given in Table 24 and Table 25 are organized and categorized into a code of accounts (COA) under development at Idaho National Laboratory (INL). Varying degrees of first-principles methods are employed, such as design for manufacture and assembly® (DFMA® ), to elucidate costs in all process levels of the plant equipment, buildings and site structures, personnel, and other miscellaneous but significant cost elements. These approaches have been successfully applied in past cost analysis projects and are designed to enable rapid and flexible cost estimation. Application of these techniques for evaluating NPP costs is similar in concept to the full, detailed estimation of construction and fabrication costs determined in a later stage of NPP development. Note that our approach tries to avoid use of other past analysis results and data such as those from the legacy Energy Economic Data Base (EEDB) Program, as these resources are based on historical NPP costs and thus may not be indicative of new reactor technologies or construction and fabrication/manufacturing techniques. However, we provide a comparison of our SFR NPP cost results in Table 90 against those included in the EEDB for a representative pressurized water reactor (PWR).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Resilience Development for Electric Energy Delivery Sytems (ResDEEDS): A Tool for Power System Resilience Planning (Rev. 1)

The INL Resilience Framework is built to be customized to a particular system’s characteristics, resilience goals, and hazards that it is likely to face. The challenge with a framework that is built to be customized is that it requires more effort from the user to understand the framework and apply it correctly to a particular system. The goal of the web application described here is to automate the framework application as much as possible. Inputs from users are standardized into a common format, but the modeling platform used is a modular framework built to adapt to any system. Baseline effects from common hazards are programmed with stochastic variables to allow for quick exploration of hazard impacts, but can be customized down to component-level impact if a user desires a high level of detail and specificity.

17 WIND ENERGY↗

Application of NEAMS Multiphysics Framework for Species Tracking in Molten Salt Reactors

This report from Idaho National Laboratory (INL) summarizes the key modeling and simulation activities conducted under the Department of Energy (DOE) Molten Salt Reactor (MSR) Campaign during the Fiscal Year 2023 (FY23). The focus of the work was to leverage state-of-the-art modeling capabilities from the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) codes to enable novel multiphysics and multiscale modeling and simulation of MSRs. Through collaboration with NEAMS code developers, advanced multiphysics analysis capabilities for MSR systems were demonstrated by coupling depletion, thermal-hydraulics, and thermochemistry into an innovative framework for chemical species transport in MSRs. As a result, the framework can track nuclides throughout their lifetimes in the core, from production (depletion) to advection throughout the salt volume (thermal-hydraulics) and off-gassing or precipitation outside of the salt (thermochemistry). This work supports the near-term deployment of MSRs by integrating the synergistic efforts between the DOE’s MSR Campaign and NEAMS program. The resulting framework will help better connect system design modelers with experimentalists to better understand and predict complex physical behaviors in MSRs. Researchers and MSR developers alike can now leverage these new modeling and simulation capabilities to perform novel analyses with applications including: • MSR dynamics during normal operational transients and accident scenarios • Off-gas system design and performance for fuel cycle and depletion analysis • Corrosion and active chemistry control for reactor component health and lifetime determination • Source term, decay heat and activity determination in accident scenarios • Special nuclear material accountancy and chemical forensic analysis for safeguards • Digital twin development of experiments and experimental reactor demonstrations • Measurement requirements for instrumentation and control design • Uncertainty and sensitivity analysis of missing data to inform future experimental data collection.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Modeling of Prismatic High Temperature Reactors in Pronghorn

Pronghorn is a Multiphysics Object-Oriented Simulation Environment (MOOSE) based thermal-hydraulics code developed at Idaho National Laboratory (INL) for advanced reactor analysis. It has been previously applied to model Pebble-Bed High Temperature Reactors (High Temperature Reactor (HTR)s), Liquid-Metal Cooled Reactors, and Molten Salt Reactors, among others. This work applies the coarse-mesh thermal hydraulics capabilities in Pronghorn to model Prismatic-Core HTRs. In particular, the Oregon State University (OSU)’s High Temperature Test Facility (HTTF) is modeled with Pronghorn. The HTTF is a 1:4 height scaled-down facility of General Atomics’ Modular High Temperature Gas-cooled Reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Final Design for Thermal/Epithermal eXperiments (TEX) with Chloride Absorbers to Provide Validation Benchmarks for Y-12 Electrorefining Facility

Uranium electrorefining operations at Y-12 require validation for chlorine absorption and reflection. Plutonium chloride solution operations at Los Alamos National Laboratory require chlorine absorption validation. The growing need for chlorine validation is evident to the wider criticality safety community, with multiple attendees at the recent TEX 2.0 meeting at Lawrence Livermore National Laboratory (May 2023) requesting validation for chlorine (Idaho National Laboratory (INL), Institute de radioprotection et de surete nucleaire (IRSN), Savannah River Nuclear Site (SRNS), LANL, and Y-12). Los Alamos National Laboratory has recently performed, and benchmarked, an experiment titled Chlorine Worth Study (CWS) for the internal operations at the lab, but due to the difficulties in precisely characterizing the material compositions of the chlorine absorbers it is advantageous to perform a complimentary study with a different chlorine-based absorber material. Furthermore, having a uranium-based vs plutonium-based experiment provides a separate and important validation basis for criticality safety and nuclear data evaluation. The original final design report for the study of chlorine absorption using the TEX-HEU experimental base was presented in 2022, but used the same chlorine-bearing materials that were found to be difficult to characterize in the LANL benchmark. A complete redesign of the experiment has been performed looking at alternative absorber materials in various forms to produce an experiment that is fully characterizable. This report presents five novel chlorine experiments using the TEX-HEU test bed which provides direct comparison to the Y-12 and INL/Terrapower application needs, utilizing sodium chloride (NaCl) absorber plates. The absorber plates will consist of granulated NaCl (≥99.5% pure), which will be fully encapsulated in aluminum tins, providing a simple but effective chlorine-based absorber material that can be completely characterized. Of the five configurations presented in this report, it is expected that two or three configurations will be down selected for the actual experiment, with the other configurations being alternates. Three of the configurations are in the standard configuration, where the absorber is placed directly on the HEU fuel plates, and two in the sandwich configuration, where the absorber is surrounded by polyethylene moderators to force additional neutron thermalization prior to reaching the absorber. There are two thicknesses of NaCl absorber plates: 3/16” and 1/4” active thicknesses (i.e. not including the encapsulation). Both variations of the absorbers have an active absorber radius of 6” and a total radius of 7.5” to match the diameter of the HEU plates, with the outer 1.5” being aluminum encapsulation. The high-density polyethylene (HDPE) moderators are of the thicknesses: 27/16”, 7/4”, 1/8”, 11/16”, and 3/4”. The final configurations have six (one sandwich and one standard configuration), eight (one sandwich and one standard configuration), and 18 total fuel layers (standard configuration). The standard and sandwich configurations were designed such that the differences in moderator thicknesses are 1” HDPE, which were already procured for the original CED-2. The proposed configurations were precisely tuned to closely match the sensitivity profiles and neutron spectra of the Y-12 upset cases and were also compared to the INL/Terrapower upset cases. The assessment of experimental uncertainties of the non-absorber components was predicted to be 0.00114 Δk eff . The assessment of uncertainties resulting from the absorbers was predicted to be 0.00029 Δk eff . This results in a total uncertainty of 0.00118 Δk eff . Many of the largest uncertainties, namely the moderator densities, may be reduced with precision dimensional inspection of the components. The 1” HDPE moderators as well as the HDPE reflectors from the original CED-2 were incorporated in the final designs presented here. Additional HDPE moderator plates must be fabricated to complete the configurations. NaCl absorber plates will by fabricated at LLNL. The total additional cost is expected to be $\$$54,250 for the remaining components. Precise inspection, including dimensional, mass, density, and impurity, is recommended for all components. LLNL estimates that these costs are around $\$$12,000. It is expected, based on previous TEX-HEU experiments, that three weeks of experimental facility time is needed to complete the experiments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Impact to Groundwater All-Pathways Dose Estimates for the Remote-Handled Low-Level Waste Disposal Facility Performance Assessment Using Updated Dose Coefficients from DOE-STD-1196-2022

The Performance Assessment (PA) for the Remote-Handled Low-Level Waste (RHLLW) Disposal Facility at Idaho National Laboratory (INL) was completed in 2018 (DOE-ID 2018) using dose coefficients from U.S. Department of Energy (DOE) Standard DOE-STD-1196-2011 (DOE 2011). Internal and external dosimetry was updated in 2021 and a new technical standard was published in 2022 (DOE-STD-1196-2022) (DOE 2022). This technical memorandum provides a comparison of the ingestion dose coefficients between those published in DOE (2011) and those published in DOE (2022). The dose coefficients in DOE (2022) were then used to calculate the all-pathways dose for the groundwater pathway and the results between the doses published in the 2018 PA and those calculated using the updated dose coefficients in DOE (2022) were compared. Several other issues in the 2018 RHLLW Disposal Facility PA were also addressed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Modeling Enhancements and Demonstration of Shift Capabilities for PBRs and MSRs

This technical report documents the modeling enhancements and demonstrations with the Shift Monte Carlo (MC) code targeted at pebble-bed reactors (PBRs) and molten salt reactors (MSRs) under the US Department of Energy (DOE) Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in fiscal year (FY) 2023. The work performed included several enhancements, such as improvements for multigroup cross section generation, a new eigenvalue mode considering only prompt fission neutrons, and enhancements to the Titan frontend for Shift to allow for new geometry types and tally functionality. Additionally, new PBR equilibrium core search reference calculations were generated with Shift and compared to Serpent calculations provided by Idaho National Laboratory (INL). These enhancements provide a robust foundation for applying Shift for both reference and two-step neutronics analysis for advanced reactor simulation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

National Reactor Innovation Center (NRIC) Grid Connection Roadmap (Rev. 1)

In an effort to define the roadmap for establishing the necessary electrical infrastructure at the Idaho National Laboratory to fully support connection of NRIC reactors, this roadmap serves to identify the current forecasted reactors planned for installation at INL and provide a consolidated list of the identified infrastructure improvements that will be necessary to support testing of NRIC reactors. Additionally, the roadmap will leverage this information to establish an integrated timeline to support detailed planning and execution of system improvements. To drive this planning, requirements, as derived from authoritative sources, will be identified. A clear definition of impacted stakeholders will be established, as well as a list of risks that will need to be managed as the program moves forward.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

FORCE Integration with DRAFT and IDAES

Integrated energy systems (IES) combine, in mutually beneficial ways, power from variable renewable energy sources and nuclear power plants (NPP) to produce multiple commodities and improve economic viability under uncertain market conditions. Technical and economic analysis of IES requires modeling of complex processes with software models having enough fidelity to capture real-world dynamics while still being capable of running using reasonable computing resources. The open-source Framework for Optimization of Resources and Economics (FORCE) tool suite, developed at Idaho National Laboratory (INL), has enabled comprehensive modeling and simulation of IES. The capabilities within FORCE include grid portfolio optimization through the Holistic Energy Resource Optimization Network (HERON) and the transient process model analysis library HYBRID, among others. Recent developments for the FORCE toolset have centered on strengthening the flexibility and modularity to link with external models and other available software to enhance IES simulations. Adding versatility to the FORCE toolset improves capability resulting in better techno-economic simulation of nuclear and IES components (both in potential higher fidelity and accuracy to expected performance after deployment). It also helps leverage existing work in the IES field and improve efficiency in national code development. This report focuses on two such endeavors: integration of the Dynamic Reliability Analysis Framework Tool (DRAFT) and the Design Integration and Synthesis Platform to Advance Tightly Coupled Hybrid Energy Systems (DISPATCHES) software packages into FORCE.

97 MATHEMATICS AND COMPUTING↗

Integrated Risk-Informed Condition Based Maintenance Capability and Automated Platform: Technical Report 1

Due to continuing global energy market trends, driven heavily by the abundant preserves of natural gas, there is an immediate need to reduce costs associated with operation and maintenance (O&M) for the current domestic nuclear power industry and for future reactor developments. This is to ensure that nuclear power generation remains an economically competitive and viable option in the energy market. O&M costs include labor-intensive preventive maintenance (PM) programs, which involve manually-performed inspection, calibration, testing, and maintenance of plant assets at periodic frequency and time-based replacement of assets, irrespective of their condition. This has resulted in an expensive, labor-centric business model to achieve high capacity factors. Fortunately, there are technologies (advanced sensors, data analytics, and risk assessment methodologies) that can enable the transition from a labor-centric business model to a technology-centric business model. The technology-centric business model will result in a significant reduction of PM activities, laying the foundation for real-time condition assessment of plant assets, reducing overall labor and part costs. To enable this transition, PKMJ Technical Services LLC is partnering with the U.S. Department of Energy’s Idaho National Laboratory (operated by the Battelle Energy Alliance, LLC) and the Public Services Enterprise Group (PSEG) Nuclear, LLC in the Integrated Risk-Informed Condition-Based Maintenance Capability and Automated Platform Project. In this report, the configuration of a digital cloud platform using Microsoft Azure is discussed, data from the PSEG Salem Nuclear Generating Station Units 1 & 2 are imported into a digital cloud platform, and the data is used for an evaluation of several key areas: cost impact analysis, risk-informed model development, and preventive maintenance strategy optimization. First, the cost impact analysis reviews which plant assets are potential good candidates for condition-based monitoring. Next, INL utilized the data in their local environment to develop the risk-informed model; which provides estimates of failure rates and probability of failures of assets based upon their past performance. The developed model is performed on assets selected from the cost impact analysis. Lastly, engineers assess the preventive maintenance strategy for the selected assets at PSEG against maintenance strategies in the nuclear industry for similar assets to potentially identify acceptable justification for the extension of current maintenance frequencies.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

A User Guide to PARET/ANL

PARET was originally created in 1969 at what is now Idaho National Laboratory (INL), to analyze reactivity insertion events in research and test reactor cores cooled by light or heavy water, with fuel composed of either plates or pins. The use of PARET is also appropriate for fuel assemblies with curved fuel plates when their radii of curvatures are large with respect to the fuel plate thickness. The PARET/ANL version of the code has been developed at Argonne National Laboratory (ANL) under the sponsorship of the U.S. Department of Energy/NNSA since the inception of the Reactor Conversion Program. Since Reduced Enrichment for Research and Test Reactors (RERTR) began in 1978, PARET/ANL has been benchmarked to experimental data including SPERT testing, and used to determine the expected transient behavior of many reactors both inside and outside the Reactor Conversion Program. This document provides the pertinent information for the use of PARET/ANL Version 7.6.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Demonstrate FARM supervisory capabilities for a thermal energy storage problem for the DETAIL facility: IES Simulation Ecosystem Control System Development

The goal of the power dispatch problem for an Integrated Energy System (IES) is to adjust the power output and the heat flow of each component to maximize the profitability of the whole unit. Facilities that can integrate real-time digital signals, mock nuclear power, thermal energy storage and industrial heat use via high-temperature electrolysis were constructed at INL to support the research activities. The Dynamic Energy Technology and Integration Laboratory (DETAIL) houses the Microreactor Agile Non-nuclear Experimental Test Bed (MAGNET) and the Thermal Energy Distribution System (TEDS). In this report, the hierarchical control system architecture proposed in June 2023 milestone for the flexible operation of DETAIL facility is finalized and demonstrated. A brief description of the components and the corresponding Dymola models from the HYRBID repository is first provided. Then, the current control strategy is presented. In particular, the approach for generating the set-point trajectories to be fed to the PI controllers is analyzed, and its limits were identified. To preserve safe operation over both long-time and real-time horizons, the integration of a Supervisory Control layer embedding a modified version of FARM (Feasible Actuator Range Modifier) module is proposed. FARM is a component of the RAVEN-based FORCE framework designed to support HERON module at optimizing the operation of IES units. The proposed control system for DETAIL foresees FARM to be applied twice, i.e., the original version (“FARM-Validator”) aiding the solution of the power dispatch problem, and a modified version (“FARM-Supervisory”) coordinating the PID controllers. Despite the kernel of the two modules is the same, their tasks are quite different. The former intervenes at the beginning of each hour to prevent constraint violations over long time periods, the latter addresses real-time control tasks and monitors the response of constrained variables at a much finer time resolution. A tentative procedure for training the embedded Digital Twins with the experimental data is also proposed. Finally, the capabilities of the designed architecture and the impact of the added Supervisory Control layer are demonstrated by simulating a representative power dispatch scenario.

25 ENERGY STORAGE↗

Advanced Reactors Integrated Energy System: Thermal Energy Storage Island Design

The main topic of this research is integrated energy systems (IES) designed for pairing industrial thermal energy loads with advanced reactors (ARs). The Idaho National Laboratory (INL) Crosscutting Technology Development IES program and the National Reactor Innovation Center (NRIC) are seeking to develop, design, and construct an AR-IES demonstration platform that couples the thermal output from an AR operating at the INL/NRIC Demonstration of Microreactor Experiments (DOME) test bed in the Experimental Breeder II dome to a variable capacity load emulator (i.e., air-cooled radiator) and sensible thermal energy storage (TES) via a molten salt thermal energy transfer fluid. In the rapidly evolving landscape of energy supply and distribution, flexibility has emerged as a prized attribute, surpassing the traditional notions of stability and baseload generation capability. This shift in priorities is particularly evident in the context of nuclear power plants (NPPs), where adaptability over constant output is becoming more important. As our energy infrastructure and resources embraces the rise of distributed energy generation, the inherent variability in net demand continues to grow. Moreover, the use of nuclear energy as a source of heat for decarbonizing the industrial sector is becoming a very pressing topic. In such environment, advanced NPPs are poised to enter a more competitive energy market, delivering both, flexible electricity and heat. This shift motivates the exploration of TES systems, designed to empower NPPs with nimble responsiveness to market fluctuations, flexible heat delivery capabilities, and redefine their role in the energy field. TES systems offer the unique advantage of storing nuclear energy in its original form as heat, thereby affording unparalleled flexibility in its subsequent utilization.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Coronado Generating Station Repowering Evaluation - Siting Evaluation

The purpose of this initial siting evaluation is to consider the suitability of the Salt River Project (SRP) owned land at/near Coronado Generating Station (CGS) for a nuclear generating station (including advanced light water reactors, light water small modular reactors, and non-light-water designs1) and to assess if there are any exclusionary or avoidance factors associated with coal to nuclear repowering at CGS. This initial siting evaluation considers several characteristics (e.g., environmental conditions, seismic concerns, site footprint, water use, etc.) and highlights favorable/preferred characteristics as well as potential risks to feasibility. The results are intended to inform SRP on the strengths and weaknesses associated with the CGS site and inform the selection of candidate nuclear technologies for the CGS site that align with SRP’s mission and business objectives. This report relies on industry-recognized siting guidance, including Electric Power Research Institute’s (EPRI’s) Siting Guide (Reference 1) 2 and Nuclear Regulatory Commission’s (NRC’s) Regulatory Guide (RG) 4.7 (Reference 3), as well as nuclear domain expertise within the Gateway for Accelerated Innovation in Nuclear (GAIN), MPR Associates, Inc. (MPR), and the Idaho National Laboratory (INL).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗