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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↗

NASA NTRS2020

Higher Power Design Concepts for NASA's Kilopower Reactor

The successful testing of the Kilopower reactor during the KRUSTY (Kilopower Reactor Using Stirling TechnologY) experiment significantly reduced the risk to fly fission power systems by demonstrating stable reactor operation through nominal and severe simulated mission scenarios. The experiment validated the neutronics, heat transfer, and power conversion systems needed for 1 kilowatt of electrical power production from the Kilopower reactor. The need for higher power reactors to support human exploration missions to the moon and Mars has become increasingly important due to the urgency to put boots on the moon by 2024 and have a sustainable presence in the following years. This desire has prompted NASA to continue the development of the Kilopower reactor to extend the power up to 10 kilowatts of electricity in support of a lunar base. These 10 kilowatt units are expected to be used as standalone units or be ganged together to create a modular power grid for propellent production, human habitats, and robotic exploration to name a few. The Kilopower reactor was originally designed to produce electrical power from 1 to 10 kilowatts using the same highly enriched uranium fuel, sodium heat pipes, and Stirling convertors at the proper scale. Consideration has also been given to the use of low enriched uranium fuel for these missions and will be studied along with the other aspects of the reactor. This paper will focus on the design concepts and trades associated with the scale up of the Kilopower power conversion system and heat transfer system to support human exploration of the moon and Mars.

Gibson, Marc↗

DOE OSTI2018

FY 2018 Annual Status Report for the Nuclear Fuels and Materials Library

The Nuclear Fuels and Materials Library (NFML), which is owned by U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) and curated by the Nuclear Science User Facilities (NSUF), is a collection of specialized information and nuclear fuel and material samples from past and ongoing irradiation test campaigns. The NFML also includes real-world components retrieved from decommissioned power reactors and donations from other sources. The NFML can be accessed through a web portal at the NSUF web page NSUF.INL.gov or directly at https://nsuf-infrastructure.inl.gov/. The NFML was conceived shortly after the NSUF was established in 2007. The concept was to offer Advanced Test Reactor (ATR) irradiated materials and NSUF project samples that were stored in Idaho National Laboratory (INL) hot cells to researchers for use in NSUF-awarded research projects. The NFML catalogue began as a series of electronic spreadsheets gathered from principal investigators (PI) and INL programs. In the summer of 2016, the spreadsheets were migrated into an online database within the NSUF website. The NFML inventory has continued to grow with the addition of NSUF-funded project specimens, Experimental Breeder Reactor No. 2 (EBR-II) legacy samples, and donated material from Oak Ridge National Laboratory (ORNL). The online library technology further evolved with upgrades for users and administrators of the database. NFML outreach in FY 2018 included NSUF representatives at conferences and workshops domestically and internationally. This document provides a status report on the NFML for FY 2018.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Horizontal Split Table Conceptual Design for Validation of Nuclear Data used in Advanced Reactors [Slides]

A methodology to create conceptual designs of benchmark critical experiments for advanced reactors nuclear data testing and validation was developed. A first concept was explored, pebble-bed high-temperature gas cooled reactor, based on the HTR-10 reactor. The very high correlation is a proof of concept that the design is similar to the application, and performing such critical experiments would help nuclear data testing and validation. This proof-of-concept was included in the IER-539 CED-1:Preliminary Design of a New Horizontal Split Table report. Other concepts could be explored if needed, such as Molten-salt reactor, Sodium-cooled fast reactor or Heat pipe reactors/Microreactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Applications of Similarity Analysis of Reactivity-initiated Accident Experiments in TREAT

Experimental testing capabilities have been created to perform reactivity-initiated accident experiments in the Transient Reactor Test Facility for fuel safety testing of Accident Tolerant Fuel concepts and to extend the burnup limits for our current light water reactor fuel designs. Completely prototypic test conditions are not possible and compromises naturally have to be made. To date, the representativity of the small scale safety testing to a similar event in a commercial reactor is made by comparing a few specific phenomena of interest. The goal of this work is to apply similarity analysis as another metric to judge the representativity of in-pile safety tests to full-scaled accident scenario.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Postirradiation Examination of the ATF-1 Experiments - 2018 Status

The Advanced Fuels Campaign in collaboration with industry are working to develop enhanced accident tolerant fuels for light water reactors. The initial irradiations of these concepts were performed in the Idaho National Laboratory Advanced Test Reactor using simple drop-in style irradiations. These irradiations are collectively referred to as ATF 1. The postirradiation examination of ATF 1 has begun and some concepts have now completed initial examination. This report describes the results of examinations on ATF 1 irradiations designed to test the feasibility of enhancing UO 2 thermal conductivity with additives, ATF 1 irradiations of uranium silicide, and the current status of ATF 1 irradiations that are in process at the hot cell. Enhanced thermal conductivity uranium dioxide composites containing silicon carbide (UO 2 SiC) and diamond (UO 2 diamond) have been irradiated to low burnup. The conditions of this irradiation test and subsequent postirradiation examinations are discussed. These irradiations evaluate fuel microstructure and potential fuel cladding interaction under representative light water reactor power conditions. Both non destructive and destructive techniques have been used to evaluate fuel integrity, fission gas release, fission product distribution, burnup, fuel swelling and cladding strain. Examination of the UO 2 SiC pellets revealed enhanced cracking when compared to a UO 2 rodlet also irradiated under similar conditions. Additionally, instability of the SiC whiskers in the uranium dioxide matrix was observed in the pellet central region, where the local temperatures exceeded 1300 °C. The microstructure of the diamond added UO 2 was severely disrupted during irradiation, resulting in local migration of cesium along the fuel stack and increased fission gas release when compared with the expected release from the Vitanza curve at corresponding values of burnup and irradiation temperature. The postirradiation examination results cast doubt on the suitability of these additives to improve UO 2 fuel performance in a way that would lead to enhanced accident tolerance. Additionally, postirradiation examination data of U 3 Si 2 fuels at low burnup (i.e., < 20 GWd/tHM) for application in advanced Light Water Reactor (LWR) is presented. The U 3 Si 2 pellets show limited cracking in comparison to the expected behavior of UO 2 at same power level. In addition, gamma scanning data did not reveal migration of fission products. Minor homogeneous hardening along the pellet radius due to accumulation of fission products and radiation damage was measured by microindentation. The fission gas release and swelling remains very low. Formation of fission gas bubbles resolvable with optical microscopy occurs from the pellets center outward to approximately 60% of the fuel pellet radius. The overall data suggest a good performance of this accident tolerant fuel candidate at low burnup. Finally, the current status of postirradiation examination on other ATF 1 irradiations will be discussed. These irradiations include UO 2 clad in an FeCrAl alloy, an alternative uranium silicide compound U 3 Si 5 also clad in a FeCrAl alloy, and an interaction test between UO 2 and different FeCrAl alloys.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

High Temperature Reactor Research and Development Roadmap, Draft for Public Comment

Nuclear power provides clean, reliable energy, contributing about 20 percent of the electricity generated in the United States. It supplies approximately 60 percent of our non-greenhouse-gas emitting power, making it our nation’s single largest contributor of carbon-free electricity. This vital component of the U.S. energy portfolio avoids hundreds of millions of tons of carbon dioxide emissions each year and supports hundreds of thousands of U.S. jobs, and yet is facing unprecedented challenges. Complex market factors, falling alternative generation costs and lower electricity demand forecasts have made operating nuclear power plants uneconomical in some parts of the country. The industry is confronting premature shut downs, a lack of new plants in the pipeline, profound market challenges and intense financing requirements. However, none of these challenges reflect a reduced need for this reliable and clean source of baseload power. Our nation’s nuclear sector urgently needs to adjust to these challenges to ensure continued availability of this vital national energy resource. The Department of Energy and its national laboratories are aggressively working to revive, revitalize, and expand U.S. nuclear energy capacity. We are advancing nuclear energy technologies through targeted early-stage investments to ensure a strong domestic industry now and into the future. By leveraging public-private partnerships and the national laboratory system, we are developing an advanced nuclear infrastructure, encouraging a resilient supply chain, and promoting a strong nuclear pipeline. A key element of this effort is to support the recent and rapid expansion in innovative advanced reactor development already being led by the U.S. nuclear industry. Advanced reactors, particularly non-light water reactor concepts, offer the potential for significant improvements to safety, economics and environmental performance, to help sustain and expand the availability of nuclear power as a clean, reliable and secure power source for our nation. As part of this strategy, the Department of Energy commissioned the development of technology roadmaps for advanced non-light-water reactor concepts. The starting point for the roadmaps is the technical readiness assessment performed as part of an advanced test and demonstration reactor study released in 2016 [ ]. The roadmaps show the Research and Development (R&D) needed to support demonstration of advanced reactor concepts, for either engineering demonstration or commercial demonstration depending on concept maturity, by 2035. The starting point for the roadmaps is the technical readiness assessment performed as part of an advanced test and demonstration reactor study released in 2016. The roadmaps were developed based on a review of technical reports and vendor literature summarizing the technical maturity of each concept and the outstanding R&D needs. Tasks for specific systems were highlighted on the basis of time and resources needed to complete the tasks and the importance of the system to the performance of the reactor concept. The roadmaps are intended to be generic (i.e., not specific to a particular vendor’s design); however, features of the General Atomics Energy Multiplier Module were included to provide specific examples for illuminating needed R&D tasks to be conducted in support of demonstration in the near future.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS2008

Development and Testing of Space Fission Technology at NASA-MSFC

The Early Flight Fission Test Facility (EFF-TF) at NASA-Marshall Space Flight Center (MSFC) provides a capability to perform hardware-directed activities to support multiple inspace nuclear reactor concepts by using a non-nuclear test methodology. This includes fabrication and testing at both the module/component level and near prototypic reactor configurations allowing for realistic thermal-hydraulic evaluations of systems. The EFF-TF is currently performing non-nuclear testing of hardware to support a technology development effort related to an affordable fission surface power (AFSP) system that could be deployed on the Lunar surface. The AFSP system is presently based on a pumped liquid metal-cooled reactor design, which builds on US and Russian space reactor technology as well as extensive US and international terrestrial liquid metal reactor experience. An important aspect of the current hardware development effort is the information and insight that can be gained from experiments performed in a relevant environment using realistic materials. This testing can often deliver valuable data and insights with a confidence that is not otherwise available or attainable. While the project is currently focused on potential fission surface power for the lunar surface, many of the present advances, testing capabilities, and lessons learned can be applied to the future development of a low-cost in-space fission power system. The potential development of such systems would be useful in fulfilling the power requirements for certain electric propulsion systems (magnetoplasmadynamic thruster, high-power Hall and ion thrusters). In addition, inspace fission power could be applied towards meeting spacecraft and propulsion needs on missions further from the Sun, where the usefulness of solar power is diminished. The affordable nature of the fission surface power system that NASA may decide to develop in the future might make derived systems generally attractive for powering spacecraft and propulsion systems in space. This presentation will discuss work on space nuclear systems that has been performed at MSFC's EFF-TF over the past 10 years. Emphasis will be place on both ongoing work related to FSP and historical work related to in-space systems potentially useful for powering electric propulsion systems.

Polzin, Kurt↗

NASA NTRS2010

Testing of Liquid Metal Components for Nuclear Surface Power Systems

The Early Flight Fission Test Facility (EFF-TF) was established by the 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 non-nuclear test methodology. This includes fabrication and testing at both the module/component level and near prototypic reactor configurations. The EFF-TF is currently supporting an effort to develop an affordable fission surface power (AFSP) system that could be deployed on the Lunar surface. The AFSP system is presently based on a pumped liquid metal-cooled (Sodium-Potassium eutectic, NaK-78) reactor design. This design was derived from the only fission system that the United States has deployed for space operation, the Systems for Nuclear Auxiliary Power (SNAP) 10A reactor, which was launched in 1965. Two prototypical components recently tested at MSFC were a pair of Stirling power conversion units that would be used in a reactor system to convert heat to electricity, and an annular linear induction pump (ALIP) that uses travelling electromagnetic fields to pump the liquid metal coolant through the reactor loop. First ever tests were conducted at MSFC to determine baseline performance of a pair of 1 kW Stirling convertors using NaK as the hot side working fluid. A special test rig was designed and constructed and testing was conducted inside a vacuum chamber at MSFC. This test rig delivered pumped NaK for the hot end temperature to the Stirlings and water as the working fluid on the cold end temperature. These test were conducted through a hot end temperature range between 400 to 550C in increments of 50 C and a cold end temperature range from 30 to 70 C in 20 C increments. Piston amplitudes were varied from 6 to 1 1mm in .5 mm increments. A maximum of 2240 Watts electric was produced at the design point of 550 hot end, 40 C cold end with a piston amplitude of 10.5mm. This power level was reached at a gross thermal efficiency of 28%. A baseline performance map was established for the pair of 1kW Stirling convertors. The performance data will then be used for design modification to the Stirling convertors. The ALIP tested at MSFC has no moving parts and no direct electrical connections to the liquid metal containing components. Pressure is developed by the interaction of the magnetic field produced by the stator and the current which flows as a result of the voltage induced in the liquid metal contained in the pump duct. Flow is controlled by variation of the voltage supplied to the pump windings. Under steady-state conditions, pump performance is measured for flow rates from 0.5-4.3 kg/s. The pressure rise developed by the pump to support these flow rates is roughly 5-65 kPa. The RMS input voltage (phase-to-phase voltage) ranges from 5-120 V, while the frequency can be varied arbitrarily up to 60 Hz. Performance is quantified at different loop temperature levels from 50 C up to 650 C, which is the peak operating temperature of the proposed AFSP reactor. The transient response of the pump is also evaluated to determine its behavior during startup and shut-down procedures.

Polzin, Kurt A.↗

DOE OSTI2019

Postirradiation examinations of low burnup U 3 Si 2 fuel for light water reactor applications

The United States Department of Energy Advanced Fuels Campaign in collaboration with industry is working to develop enhanced accident tolerant fuels for light water reactors. The initial irradiations of these concepts were performed in the Idaho National Laboratory Advanced Test Reactor using drop-in style irradiations. These irradiations are collectively referred to as Accident Tolerant Fuels 1 (ATF-1). The postirradiation examination of ATF-1 has begun and some concepts have now completed initial examination. This work is focused on postirradiation examination data of U 3 Si 2 fuels at low burnup (i.e., < 20 GWd/tHM) for application in Light Water Reactors (LWRs). The U 3 Si 2 pellets show limited cracking in comparison to the expected behavior of UO 2 at same power level. Moreover, gamma spectrometry data did not reveal migration of fission products. Minor homogeneous hardening along the pellet radius due to accumulation of fission products and radiation damage was measured by microindentation. The fission gas release and swelling remain very low. Formation of fission gas bubbles resolvable with optical microscopy occurs from the pellets center outward to approximately 60% of the fuel pellet radius. With this being said, the overall data suggest a good performance of this accident tolerant fuel candidate at low burnup.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Commercial-scale, fluoride-salt-cooled, high-temperature reactors with novel refueling and decay heat removal capabilities

The fluoride-salt-cooled high-temperature reactor (FHR) is a new solid-fuel, graphite-moderated reactor concept. Recent work has developed test and demonstration-scale FHR designs. From this work, a new 'fuel-inside-radial-moderator' (FIRM) assembly concept has emerged that offers several important advantages, from neutron economy to refueling. In this work we discuss the potential benefits and challenges associated with designing a commercial-scale, FIRM-based FHR. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Functional and Operating Requirements for the Microreactor Agile Non-Nuclear Experimental Test Bed (MAGNET)

Microreactors, or small, transportable reactors with a capacity of < 20 MWTH, are sought to provide heat and power for myriad applications in remote areas, military installations, emergency operations, humanitarian missions and disaster relief zones as illustrated in Figure 1. These small, transportable reactor designs, while offering many advantages, are largely untested and unproven. System and component testing is needed to demonstrate to regulators that these designs are safe and to convince customers that the systems are robust, reliable, and efficient. The microreactor agile non-nuclear experimental test bed (MAGNET) is being constructed at Idaho National Laboratory to assist with the development, demonstration, and validation of microreactor components and systems. MAGNET will support technology maturation to reduce uncertainty and risk relative to the operation and deployment of this unique class of systems. Stakeholders for this test bed include microreactor developers, energy users, and regulators. Regulators will be engaged early in the design and testing to expedite regulatory approval and licensing. Within MAGNET, systems and components can be safely tested, providing valuable information on operating regimes, failure modes, and thresholds. Since there are various types of microreactors being proposed, which can be classified according to their core cooling method, heat-pipes, gas-cooled (pebble bed or prismatic), molten salt, light water, or light water, the goal is to provide a test bed that is broadly applicable to multiple microreactor concepts. Each reactor type poses a different set of design and operational challenges and performance claims stated by commercial vendors have not been independently verified through rigorous testing. The initial set of tests to be performed in MAGNET are targeted towards demonstrating the feasibility and performance of heat-pipe cooled reactors, since this concept is unique to very small nuclear reactors. However, the testbed will be constructed to accommodate other designs in addition to heat-pipe cooled reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Extension of Virtual Test Bed Advanced Modeling and Simulation Capabilities for Fusion Energy

The National Reactor Innovation Center (NRIC) was established to accelerate the deployment of novel reactor concepts. This is achieved by providing physical and virtual spaces for building and testing reactor experiments. The Virtual Test Bed (VTB) represents the virtual counterpart to the physical test beds. It is a collaboration with the Department of Energy’s (DOE) Nuclear Energy Advanced Modeling and Simulation (NEAMS) program with the mission to accelerate the deployment of advanced reactors by facilitating the adoption of advanced modeling and simulation (M&S) tools developed by the DOE. This mission has been carried out by the VTB since 2020 by hosting and featuring dozens of advanced fission nuclear reactor models developed by national laboratories and academia. The charter of the NRIC’s definition of advanced reactors also includes fusion nuclear reactors. As the tools developed by the NEAMS program are increasingly used for modeling fusion energy devices, there is an increasing need to host fusion reactor models on the VTB repository. The VTB will be extended in 2024 to support fusion energy modeling and simulation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2025

Lithium Divertor Targets and Walls for the ASTER Liquid Stellarator Reactor, Distributed Divertor

Stellarators may have advantages for certain liquid metal options as Plasma Facing Components (PFC) for divertor targets and first walls due to the wide range of possible magnetic configurations, which additionally are free of disruptions and fast field variations. In a previous work (V. Queral et al., IEEE Trans. Plasma Sci. 52, 2024), a concept of stellarator reactor (ASTER-CP) based on swirling Li-molten salts and liquid lithium floating on the molten salt as PFC was presented. The divertor matters were not studied then and, thus, they are being studied and experimentally tested now. The ASTER-CP reactor concept, the initial liquid metal experiments and potential concepts for the ASTER-CP divertor and first wall are reported. Concerning the experiments, several small scale experiments of galinstan in a small rotating cylinder under magnetic field have been produced, including one experiment with high viscosity galinstan-mixture for increased thickness of layer. An experiment of floating lithium on the molten salt LiCl-PbCl2 gave fast volatilization/decomposition of the molten salt. Particularly for divertors, the traditional free-flow, Capillary Porous Systems and ‘divertorlets’ have been studied for application to ASTER-CP. Surface waves (hot spots), lack of enough surface fluid turbulence and excessive fluid speed are the main issues found in fast free-flow. The perhaps original concept of Distributed Divertor and Equi-power Surface is tentatively proposed and studied, taking advantage of stellarator fields and low recycling regime.

distributed divertor↗

DOE OSTI2024

A Workflow to Optimize Fast Neutron Irradiation in A Thermal Neutron Spectrum Test Reactor Leveraging Open-Source Tools

The Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL) is one of the key nuclear engineering research and testing facilities within the US Department of Energy (DOE). The ATR is one of few high-power research reactors in the world with different application including accelerated testing of nuclear fuel, materials irradiation in a very high neutron flux environment, and medical radioisotope production [1]. Also, the ATR offers opportunities for testing fast spectrum fission and fusion reactor materials. The key challenges in this area are in further detailing and optimizing a fast spectrum environment within a thermal test reactor. This challenge involves researching, developing, and testing novel concepts for the multiplying of neutron populations into ever higher energy spectra in high flux test reactors like ATR. The main objective of this work is to investigate candidate materials for establishing a fast neutron experiment irradiation in thermal neutron spectrum test reactors which can be accomplished by filtering thermal and epithermal neutrons and boosting fast neutrons at designated irradiation positions. However, adding these filters will render the neutron spectrum and the criticality of the system. The selection of the thickness and material layers should be accomplished by developing an optimization design algorithm that is applicable for ATR to enhance the fast neutron spectrum irradiation utilizing high-fidelity Monte Carlo methods along with advanced machine learning capabilities. This paper presents workflow for design optimization to enhance fast neutron irradiation in the ATR. The workflow leverages open-source tools to develop an algorithm that is viable to ATR and can be leveraged in other reactors. The following sections discuss the development of the experiment design optimization workflow and its application to ATR irradiation positions.

42 - ENGINEERING↗

DOE OSTI2015

Development of ASTM Standard for SiC-SiC Joint Testing Final Scientific/Technical Report

As the nuclear industry moves to advanced ceramic based materials for cladding and core structural materials for a variety of advanced reactors, new standards and test methods are required for material development and licensing purposes. For example, General Atomics (GA) is actively developing silicon carbide (SiC) based composite cladding (SiC-SiC) for its Energy Multiplier Module (EM2), a high efficiency gas cooled fast reactor. Through DOE funding via the advanced reactor concept program, GA developed a new test method for the nominal joint strength of an endplug sealed to advanced ceramic tubes, Fig. 1-1, at ambient and elevated temperatures called the endplug pushout (EPPO) test. This test utilizes widely available universal mechanical testers coupled with clam shell heaters, and specimen size is relatively small, making it a viable post irradiation test method. The culmination of this effort was a draft of an ASTM test standard that will be submitted for approval to the ASTM C28 ceramic committee. Once the standard has been vetted by the ceramics test community, an industry wide standard methodology to test joined tubular ceramic components will be available for the entire nuclear materials community.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Design Report for the ½ Scale Air-Cooled RCCS Tests in the Natural convection Shutdown heat removal Test Facility (NSTF)

The Natural convection Shutdown heat removal Test Facility (NSTF) is a large scale thermal hydraulics test facility that has been built at Argonne National Laboratory (ANL). The facility was constructed in order to carry out highly instrumented experiments that can be used to validate the performance of passive safety systems for advanced reactor designs. The facility has principally been designed for testing of Reactor Cavity Cooling System (RCCS) concepts that rely on natural convection cooling for either air or water-based systems. Standing 25-m in height, the facility is able to supply up to 220 kW at 21 kW/m 2 to accurately simulate the heat fluxes at the walls of a reactor pressure vessel. A suite of nearly 400 data acquisition channels, including a sophisticated fiber optic system for high density temperature measurements, guides test operations and provides data to support scaling analysis and modeling efforts. Measurements of system mass flow rate, air and surface temperatures, heat flux, humidity, and pressure differentials, among others; are part of this total generated data set. The following report provides an introduction to the top level-objectives of the program related to passively safe decay heat removal, a detailed description of the engineering specifications, design features, and dimensions of the test facility at Argonne. Specifications of the sensors and their placement on the test facility will be provided, along with a complete channel listing of the data acquisition system.

42 ENGINEERING↗

DOE OSTI2024

New Virtual Test Bed Capabilities: Virtual DOME Model and New Updates to Repository

The Department of Energy (DOE) Office of Nuclear Energy National Reactor Innovation Center accelerates the deployment of novel reactor concepts by establishing both physical and virtual spaces for building and testing various components, systems, and complete pilot plants. The Virtual Test Bed represents the virtual arm of the National Reactor Innovation Center and is a joint effort with the DOE Nuclear Energy Advanced Modeling and Simulation Program. The Virtual Test Bed mission is to accelerate the deployment of advanced reactors by facilitating the adoption of cutting-edge DOE advanced modeling and simulation tools to design, evaluate, and license reactors. This is primarily achieved by storing example challenge problems in an externally available repository and by developing models to fill the M&S gaps needed for potential demonstrators. Activities conducted this fiscal year focused on developing of a Demonstration of Microreactor Experiments shield model to help accelerate the confirmatory analysis required for the reactor demonstration. This model and workflow will allow developers to leverage advanced modeling and simulation tools to ensure their reactor demonstration concept will meet dose requirements and that the surrounding shield will stay within concrete temperature limits during steady-state and transient operation conditions. An initial model has been developed to evaluate the temperature distribution in the concrete shield during steady-state operation, including neutron and gamma heating effects. Various modeling strategies have been examined to understand their applicability and limitations with different reactor designs to make the workflow as reactor-agnostic as possible and computationally effective to maximize its usability. In addition to describing the Demonstration of Microreactor Experiments shield model and associated results, this report summarizes other accomplishments regarding repository maintenance and improvement and new external models hosted on the repository.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗