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DOE OSTI2017

A Summary of the Department of Energy's Advanced Demonstration and Test Reactor Options Study

An assessment of advanced reactor technology options was conducted to provide a sound comparative technical context for future decisions by the Department of Energy (DOE) concerning these technologies. Strategic objectives were established that span a wide variety of important missions and advanced reactor technology needs were identified based on recent Department of Energy and international studies. A broad team of stakeholders from industry, academia and government was assembled to develop a comprehensive set of goals, criteria and metrics to evaluate advanced irradiation test and demonstration reactor concepts. Point designs of a select number of concepts were commissioned to provide a deeper technical basis for evaluation. The technology options were compared on the bases of technical readiness and the ability to meet the different strategic objectives. Using the study’s evaluation criteria and metrics, an independent group of exerts from industry, universities and national labs scored each of the point designs. Pathways to deployment for concepts of varying technical maturity were estimated for the different demonstration systems with regard to cost, schedule and possible licensing approaches. Furthermore, this study also presents the tradeoffs that exist among the different irradiation test reactor options in terms of the ability to conduct irradiations in support of advanced reactor research and development (R&D) and to serve potential secondary missions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

A Summary of the Department of Energy’s Advanced Demonstration and Test Reactor Options Study

An assessment of advanced reactor technology options was conducted to provide a sound comparative technical context for future decisions by the U.S. Department of Energy (DOE) concerning these technologies. Strategic objectives were established that span a wide variety of important missions, and advanced reactor technology needs were identified based on recent DOE and international studies. A broad team of stakeholders from industry, academia, and government was assembled to develop a comprehensive set of goals, criteria, and metrics to evaluate advanced irradiation test and demonstration reactor concepts. Point designs of a select number of concepts were commissioned to provide a deeper technical basis for evaluation. The technology options were compared on the bases of technical readiness and the ability to meet the different strategic objectives. Using the study's evaluation criteria and metrics, an independent group of experts from industry, universities, and national laboratories scored each of the point designs. Pathways to deployment for concepts of varying technical maturities were estimated for the different demonstration systems with regard to cost, schedule, and possible licensing approaches. This study also presents the trade-offs that exist among the different irradiation test reactor options in terms of the ability to conduct irradiations in support of advanced reactor research and development and to serve potential secondary missions. The main findings of the study indicate the following: (1) for industrial process heat supply, a high-temperature gas-cooled reactor is the best choice because of the high outlet temperature of the reactor and its strong passive and inherent safety characteristics; (2) for resource utilization and waste management, a sodium-cooled fast reactor (SFR) is best because of the use of a fast flux to destroy actinides; (3) to realize the advantages of a promising but less-mature technology, a fluoride salt-cooled high-temperature reactor and a lead-cooled fast reactor fare about the same; (4) for fulfilling the needs of a materials test reactor, a SFR is considered best because of its ability to produce high fast flux, incorporate test loops, and provide additional large volumes for testing.

Advanced reactors↗

DOE OSTI2023

HydroGEN Seedling: High-Temperature Reactor Catalyst Material Development for Low-Cost and Efficient Solar-Driven Sulfur-Based Processes

The HyS process, driven by solar power, has great potential to reach high-efficiency and low-cost hydrogen production without greenhouse gas emissions. The high-temperature section of the HyS cycle, which operates the catalytic decomposition of sulfuric acid into sulfur dioxide, oxygen, and water, is a fundamental part of the cycle affecting the overall plant efficiency and cost. Therefore, a high-performance catalyst (i.e., low cost, high catalytic activity, and low degradation catalyst) is of critical importance to achieve high efficiency and low hydrogen cost. Research and development has highlighted that a Pt-based monometallic catalyst had unacceptable catalytic activity and performance degradation for a high-efficiency and low-cost hydrogen production process. A high-efficiency solar receiver-reactor system, which incorporates the new catalyst, also needs to be developed to achieve the required plant efficiency and cost. Greenway Energy (GWE) and the University of South Carolina (USC), partnering with HydroGEN node laboratories Idaho National Laboratory (INL), Savannah River National Laboratory (SRNL), and National Renewable Energy Laboratory (NREL), propose the development of a new catalyst formulation, included in a novel solar receiver-reactor concept, to be tested experimentally in the last part of the project.

08 HYDROGEN↗

NASA NTRS2003

Thermally Simulated Testing of a Direct-Drive Gas-Cooled Nuclear Reactor

This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet-sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrical thermal simulation of reactor components and concepts.

Godfroy, Thomas↗

DOE OSTI2023

Draft Summary Report on Irradiated Fuel Handling and Management for LOTUS

The National Reactor Innovation Center (NRIC) has conceptualized the design of the Laboratory for Operation and Testing in the United States (LOTUS) test bed to provide the United States Department of Energy (DOE) with the infrastructure necessary to make advanced reactor designs available for commercial developers. LOTUS will provide a test bed to developers with the capabilities of supporting a wide range of experiment design possibilities. Upon completion of the developers’ operations and experiments within the test bed, the irradiated fuel, reactor components, and other experiment materials must be removed. Idaho National Laboratory (INL) possesses significant capabilities for radioactive material handling such as casks, carts, and forklifts. However, given the unique environment presented by the NRIC-LOTUS test bed, located inside the Zero Power Physics Reactor (ZPPR) Perimeter Intrusion Detection and Assessment System (PIDAS) area at the Materials and Fuels Complex (MFC) and the complexity of novel removal activities of recently operated reactor experiments through the new proposed access tunnel. The efficacy was not apparent for existing equipment to provide all the needed capability. To bridge the potential gaps in cask designs, storage, and transportation, NRIC requested the development of trade studies for the transfer, handling, and storage requirements of irradiated fuel salts and other radioactive materials. NRIC directed Boston Government Services, LLC (BGS) to perform the trade studies and develop a report analyzing alternatives. In addition to the BGS reports, the Idaho National Lab’s (INL), provided by the first potential user’s Advanced Reactor Development (ARD) team, prepared a feasibility study for the storage of specific irradiated fuel within the existing ZPPR vault.This summary report is intended to present the trade studies, options, and alternatives that were investigated. The maturity level of LOTUS, the reactor, fuel salt containers, gloveboxes, and reactor testing campaign and concept of operations were not at a level sufficient to base critical decisions on. This report is not intended to present a final recommendation. The final recommendations for fuel storage location, transport, handling equipment, and operations will be made in FY 2024 and will be based on known materials, test campaign requirements, funding, and final analysis of the fuel and equipment to be used.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Conceptual Design of a Pumped Water Capsule for TREAT

This study performed mechanical design, thermal hydraulic, and fuel performance investigations to determine whether an existing large water capsule design could be modified to create forced convection boundary conditions on nuclear fuel specimens in the Transient Reactor Test facility. This concept was found to be viable and offers several advantages compared to a static water environment for a variety of transient power shape categories. The pumped capsule is not foreseen to be superior to a past concept for a larger water loop, but the pumped capsule approach is recommended for further engineering and deployment based on its high cost-to-value ratio primarily as a stop gap capability until a larger water loop can be realized.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Methodology for Irradiation Creep Testing of SiC / SiC Composite Cladding

Irradiation creep compliance is a critical parameter for evaluating the stress evolution of silicon carbide (SiC) components during operation in various reactor concepts. Recent instrumented irradiation creep testing of SiC, performed at low dose in the Halden reactor, showed a creep compliance that is orders of magnitude higher than previous un-instrumented bend relaxation experiments performed in the High Flux Isotope Reactor (HFIR). This discrepancy illustrates the need for additional irradiation creep data for SiC. To this end, a novel irradiation capsule design has been developed to test miniature SiC tensile specimens under loading during irradiation in the HFIR. The experiment design uses the coolant pressure of the HFIR to compress a bellows, which applies the loading on the specimen. Preliminary thermal finite element models were developed to estimate the temperature distribution within the specimen and surrounding capsule components. Similar structural models were developed to evaluate the stress distribution within the specimen. The thermal analyses show that within the gauge section, the specimen temperatures remain within ~10°C of the nominal 300°C design temperature. Similarly, the structural analysis predicts axial stresses within ~10 MPa of the target stress of 100 MPa. These initial analyses show that irradiation creep testing of SiC in the HFIR is feasible and could provide a low-cost, high-throughput vehicle for gathering needed irradiation creep data.

36 MATERIALS SCIENCE↗

NASA NTRS1973

Compatibility tests of materials for a lithium-cooled space power reactor concept

Materials for a lithium-cooled space power reactor concept must be chemically compatible for up to 50,000 hr at high temperature. Capsule tests at 1040 C (1900 F) were made of material combinations of prime interest: T-111 in direct contact with uranium mononitride (UN), Un in vacuum separated from T-111 by tungsten wire, UN with various oxygen impurity levels enclosed in tungsten wire lithium-filled T-111 capsules, and TZM and lithium together in T-111 capsules. All combinations were compatible for over 2800 hr except for T-111 in direct contact with UN.

Sinclair, J. H.↗

DOE OSTI2018

The International Experimental Thermal Hydraulic Systems database – TIETHYS: A new NEA validation tool

Nuclear reactor codes require validation with appropriate data representing the plant for specific scenarios. The thermal-hydraulic data is scattered in different locations and in different formats. Some of the data is in danger of being lost. A relational database is being developed to organize the international thermal hydraulic test data for various reactor concepts and different scenarios. At the reactor system level, that data is organized to include separate effect tests and integral effect tests for specific scenarios and corresponding phenomena. The database relies on the phenomena identification sections of expert developed PIRTs. The database will provide a summary of appropriate data, review of facility information, test description, instrumentation, references for the experimental data and some examples of application of the data for validation. The current database platform includes scenarios for PWR, BWR, VVER, and specific benchmarks for CFD modelling data and is to be expanded to include references for molten salt reactors. There are place holders for high temperature gas cooled reactors, CANDU and liquid metal reactors. This relational database is called The International Experimental Thermal Hydraulic Systems (TIETHYS) database and currently resides at Nuclear Energy Agency (NEA) of the OECD and is freely open to public access. Going forward the database will be extended to include additional links and data as they become available. https://www.oecd-nea.org/tiethysweb/

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS2011

Performance Testing of a Liquid Metal Pump for In-Space Power Systems

Fission surface power (FSP) systems could be used to provide power on the surface of the moon, Mars, or other planets and moons of our solar system. Fission power systems could provide excellent performance at any location, including those near the poles or other permanently shaded regions, and offer the capability to provide on demand power at any time, even at large distances from the sun. Fission-based systems also offer the potential for outposts, crew and science instruments to operate in a power-rich environment. NASA has been exploring technologies with the goal of reducing the cost and technical risk of employing FSP systems. A reference 40 kWe option has been devised that is cost-competitive with alternatives while providing more power for less mass anywhere on the lunar surface. The reference FSP system is also readily extensible for use on Mars, where it would be capable of operating through global dust storms and providing year-round power at any Martian latitude. Detailed development of the FSP concept and the reference mission are documented in various other reports. The development discussed in this paper prepares the way for testing of the Technology Demonstration Unit (TDU), which is a 10 kWe end-to-end test of FSP technologies intended to raise the entire FSP system to technology readiness level (TRL) 6. The Early Flight Fission Test Facility (EFF-TF) was established by NASA s Marshall Space Flight Center (MSFC) to provide a capability for performing hardware-directed activities to support multiple in-space nuclear reactor concepts by using a nonnuclear test methodology. This includes fabrication and testing at both the module/component level and at near prototypic reactor components and configurations allowing for realistic thermal-hydraulic evaluations of systems. The liquid-metal pump associated with the FSP system must be compatible with the liquid NaK coolant and have adequate performance to enable a viable flight system. Idaho National Laboratory (INL) was tasked with the modeling, design, and fabrication of an ALIP suitable for the FSP reference mission. A prototypic ALIP was fabricated under the direction of INL and shipped to MSFC for inclusion in the Technology Demonstration Unit (TDU), a quarter-scale end-to-end reactor simulator system that is scheduled for testing at NASA-GRC. Before inclusion in the TDU, the ALIP was tested in the ALIP test circuit (ATC), which is a rig developed and operated at MSFC for the specific purpose of providing accurate quantification of liquid metal pump performance. Data showing the pump performance curves (pressure, flowrate, and pump efficiency) are presented for various operating power levels, demonstrating the full performance envelope of the pump.

Polzin, Kurt↗

DOE OSTI2025

NRIC Annual Report FY 2025

The National Reactor Innovation Center (NRIC), established in August 2019, is a national United States (U.S.) Department of Energy (DOE) program. NRIC’s mission is to partner with industry and national laboratories to bridge the gap between concept, demonstration, and commercialization of advanced nuclear technology. NRIC accomplishes this through building or enhancing existing DOE infrastructure to support testing of components and systems that are key to successfully deploying advanced nuclear technology. NRIC’s vision is that by 2028, NRIC will be partnered with industry and accelerating the demonstration and deployment of advanced nuclear technology using DOE national laboratory infrastructure and expertise. NRIC will establish four new experimental facilities and two large reactor test beds for integrated technology demonstrations and experimentation by 2028 and complete two advanced nuclear technology tests by 2030. Achieving this vision will enable urgently needed abundant and affordable clean energy both domestically and internationally. NRIC’s success will inspire our nation and the global community to embrace the promising contribution of innovative nuclear reactor technologies to the clean energy economy and re-establish the U.S. as the global leader in advanced nuclear energy. NRIC is tasked with expediting the development of advanced nuclear energy technologies by bringing together private-sector technology developers and the world-class capabilities of the DOE national laboratory system. Through this program, the U.S. private sector is given access to the physical infrastructure available at DOE national laboratories to test and demonstrate their reactor concepts. NRIC works closely with the DOE-Nuclear Energy (NE) program that grants access to technical, regulatory, and financial support for commercializing nuclear energy. NRIC builds upon these new reactor concepts and technology successes to effectively strengthen U.S. nuclear leadership.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

National Reactor Innovation Center Annual Report

The National Reactor Innovation Center (NRIC), established in August 2019, is a national United States (U.S.) Department of Energy (DOE) program. NRIC’s mission is to partner with industry and national laboratories to bridge the gap between the concept, demonstration, and commercialization of advanced nuclear technology. NRIC accomplishes this through building or enhancing existing DOE infrastructure to support the testing of components and systems that are key to successfully deploying advanced nuclear technology. NRIC works to inspire stakeholders and the public, empower innovators, and deliver successful outcomes through efficient collaboration and coordination with partners. NRIC’s vision is that by 2028, NRIC will be partnered with industry and accelerating the demonstration and deployment of advanced nuclear technology using DOE national laboratory infrastructure and expertise. NRIC will establish four new experimental facilities and two large reactor test beds for integrated technology demonstrations and experimentation by 2028 and complete two advanced nuclear technology tests by 2030. Achieving this vision will enable urgently needed abundant and affordable clean energy both domestically and internationally. NRIC’s success will inspire our nation and the global community to embrace the promising contribution of innovative nuclear reactor technologies to the clean energy economy and re-establish the U.S. as the global leader in advanced nuclear energy. NRIC is tasked with expediting the development of advanced nuclear energy technologies by bringing together private-sector technology developers and the world-class capabilities of the DOE national laboratory system. Through this program, the U.S. private sector is given access to the physical infrastructure available at DOE national laboratories to test and demonstrate their reactor concepts. NRIC works closely with the Gateway for Accelerated Innovation in Nuclear (GAIN),; which is the DOE-Nuclear Energy (NE) program that grantings access to technical, regulatory, and financial support for commercializing nuclear energy. As observed in Figure 1, NRIC builds upon these new reactor concepts and technology successes to effectively strengthen U.S. nuclear leadership.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2016

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2014

Evaluation Metrics Applied to Accident Tolerant Fuels

The safe, reliable, and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and have yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. One of the current missions of the U.S. Department of Energy’s (DOE) Office of Nuclear Energy (NE) is to develop nuclear fuels and claddings with enhanced accident tolerance for use in the current fleet of commercial LWRs or in reactor concepts with design certifications (GEN-III+). Accident tolerance became a focus within advanced LWR research upon direction from Congress following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal of ATF development is to identify alternative fuel system technologies to further enhance the safety, competitiveness and economics of commercial nuclear power. Enhanced accident tolerant fuels would endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system while maintaining or improving performance during normal operations. The U.S. DOE is supporting multiple teams to investigate a number of technologies that may improve fuel system response and behavior in accident conditions, with team leadership provided by DOE national laboratories, universities, and the nuclear industry. Concepts under consideration offer both evolutionary and revolutionary changes to the current nuclear fuel system. Mature concepts will be tested in the Advanced Test Reactor at Idaho National Laboratory beginning in Summer 2014 with additional concepts being readied for insertion in fiscal year 2015. This paper provides a brief summary of the proposed evaluation process that would be used to evaluate and prioritize the candidate accident tolerant fuel concepts currently under development.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

High-Temperature Gas-Cooled Test Reactor Point Design

A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Dynamics of MHD instabilities near a ferromagnetic wall

Prospective fusion component testing and DEMO power reactor concepts are expected to employ low-activation ferritic steels because of their ability to withstand the high neutron flux of the reactor environment. However, theory suggests that ferromagnetic material may amplify certain external MHD instabilities. Using its ferromagnetic-resistive wall mode (FRWM) upgrade, the High Beta Tokamak-Extended Pulse (HBT-EP) experiment has observed approximately doubled growth rates when operating with a close-fitting ferromagnetic first wall, compared to operation with a stainless steel first wall. The presence of a ferromagnetic wall correlates with earlier disruptions, and FRWM growth rates increase with decreasing mode rotation, as expected due to the increased skin depth allowing greater mode interaction with the bulk ferromagnetic material. It is also seen that introducing low-n asymmetries into the toroidal distribution of ferromagnetic material, similar to the ITER test blanket module toroidal asymmetry, changes the phase preference of rotating modes; meanwhile, a similar change in purely conducting material does not significantly change the mode's phase preference.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2019

SAM Code Validation using the Compact Integral Effects Test (CIET) Experimental Data

The primary objective of this study is to validate the system analysis code, SAM, using experimental data from the Compact Integral Effects Test (CIET) facility. The System Analysis Module (SAM) is an advanced and modern system analysis tool being developed at Argonne National Laboratory under the U.S. DOE Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. Originally aiming at supporting sodium fast reactor (SFR) design, it is now extended to cover other non-light water reactor (non-LWR) concepts, such as high-temperature gas-cooled reactor (HTGR) designs, molten salt-fueled, and molten salt-cooled reactor designs. To support SAM code development for the wide range of non-LWR applications, it is of paramount importance to validate the code against experiments relevant to these reactor concepts. The Compact Integral Effects Test (CIET) facility, which was designed to reproduce the thermal-hydraulics response of fluoride salt-cooled high-temperature reactors (FHR) under both forced- and natural-circulation conditions, has been identified and selected as one of benchmark test facilities for SAM code validation. Among available CIET experiments, three sets of them with distinctive characteristics were selected for SAM code validation purpose. The three sets experiments are: 1) Power step change transient tests; 2) DHX-DRACS natural circulation tests; and 3) heater frequency response tests. For all tests, steady-state or transient, SAM-predicted results show very good agreement with experimental data. The successful validation of SAM against these selected CIET experiments demonstrates that the computer code is well suited for thermal-hydraulics analysis of FHR designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗