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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Deploying Extended Reality (XR) for Digital Operations and Maintenance (O&M) at the Mechanisms Engineering Test Loop (METL)

This report documents the preliminary efforts to digitize operation and maintenance (O&M) activities for the Mechanisms Engineering Test Loop (METL). METL became operational in September 2018 with the mission to provide an ecosystem for Advanced Reactor Development (ARD). METLs flagship facility’s primary purpose is conducting small to intermediate scale tests for Sodium Fast Reactors (SFR). Its resemblance to commercial SFR’s intermediate heat transport system, prototypic operating conditions, and industrial construction practices/materials provides the overarching benefit of establishing a proving ground for emerging operations and maintenance (O&M) activities such as incorporating Extended Reality (XR) applications throughout the program lifecycle.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

High-Temperature Sodium System Structural Health Monitoring (SHM) Technologies at the Mechanisms Engineering Test Loop (METL)

This report documents the deployment of non-destructive testing (NDT) technologies at the Mechanisms Engineering Test Loop (METL). METL became operational in September 2018 with the mission to provide an ecosystem for Advanced Reactor Development (ARD). METLs flagship facility’s primary purpose is conducting small to intermediate scale tests for Sodium Fast Reactors (SFR). Its resemblance to a commercial SFR’s intermediate heat transport system, prototypic operating conditions, and industrial construction practices/materials provides the overarching benefit of establishing a proving ground for devices of all technology readiness levels as well as commercial off-the-shelf equipment yet to be demonstrated on a sodium system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Molten Salt Reactor Safety Assessment - Three Approaches to A Common Objective

Safety adequacy assessment is central to nuclear power plant (NPP) licensing. Either NPP accident mitigation or prevention can result in adequate safety. Accident prevention, along with the the prevention of accident escalation, can be evaluated using either deterministic or probabilistic methods. Acceptable means to develop principal design criteria (PDCs) via probabilistic methods are provided in NRC Regulatory Guide 1.233 while advanced reactor design criteria are provided in NRC Regulatory Guide 1.232 (RG 1.232). Accident mitigation-based regulatory guidance for non-power reactors is provided in NUREG 1537. While RG 1.232 included class specific guidance for both sodium-cooled fast reactors and modular high-temperature gas-cooled reactors, it did not provide molten salt reactor (MSR) class specific guidance. Class specific design criteria more closely align with reactors in their class so will require less effort to employ to develop design specific PDCs. The American Nuclear Society has recently released a liquid-fueled MSR design safety standard (ANSI/ANS-20.2-2023) that provides class specific design criteria.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Segmentation and Classification of Fission as Pores in Reactor Irradiated Annular U–10Zr Metallic Fuel Using Machine Learning Models

Metallic fuels, particularly U—10Zr, are promising candidates for next-generation sodium-cooled fast reactors. Irradiation of nuclear fuels in reactors can lead to the formation of solid and gas fission product which subsequently forms microstructural pores, deteriorating fuel performance. Due to the massive amount of pores and complex phases formed, a quantitative description of fission gas pores is not yet available, preventing the development of microstructure-informed fuel performance modeling for fuel qualification. This paper applied a pre-trained deep learning model to ~10,260 high magnification scanning electron microscopy images. This method increased the accuracy of fission gas pore segmentation and allows statistical features to be extracted which cannot be achieved manually. A pre-trained decision tree model worked on the segemenation results and further classified the pores into different categories to produce a correlation between the pores, movement of lanthanides, and temperature gradient during irradiation. Finally, this paper emphasizes the potentials of machine learning models to accelerate fuel research, development, and qualification for advanced reactors.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Cooperative Research and Development Agreement (CRADA) NFE-18-07194 with TerraPower LLC (Final Report)

Because of the potential economic and safety benefits of the molten salt reactor (MSR) concept, development of several designs has been initiated around the world over the past decade. New international nuclear safeguards needs and verification challenges are likely to arise because of the commercial interests in MSRs and the number of MSR design variants. As a result, work was undertaken to explore the cross-cutting issues specific to the application of international nuclear safeguards (i.e., safeguards) to liquid-fueled molten salt reactors (LFMSRs). Through a public-private partnership, a report was developed that focuses on the TerraPower Molten Chloride Fast Reactor (MCFR) design that has received a funding award from the US Department of Energy (DOE), Office of Nuclear Energy. The report is intended to provide a preliminary analysis for a safeguards-by-design (SBD) effort to inform designers about how safeguards could be applied to LFMSRs by the International Atomic Energy Agency (IAEA).Although the report specifically focuses on the TerraPower design, the conclusions are applicable to the main design features of LFMSRs and can be used to extrapolate how existing IAEA safeguards measures for other fuel cycle facilities can be appropriately applied or modified. The report evaluates the appropriate safeguards approaches for LFMSRs, presents existing safeguards inspection technologies are still valid for LFMSRs, and identifies new challenges that will require novel measurement instruments to meet verification standards of the IAEA and the international safeguards regime. This document summarizes the results of the work performed under the full report developed as part of the Cooperative Research and Development Agreement (CRADA). This summary does not contain any protected CRADA information and is intended for public release.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

OTERR User Manual

OTERR (Optimization of TEst Reactor Reloading) is a software tool created to assist in the determination of optimal fuel reloading patterns for test reactors. The intended application is for the Versatile Test Reactor (VTR) program, but it provides functions that could be useful for analysis and optimization of many types of fast reactors. OTERR does not perform neutron/gamma transport, heat transfer, thermal hydraulics, or depletion calculations. Instead, it acts as a wrapper around codes that provide these capabilities, with a native genetic algorithm optimization capability. At this time, wrapping is only implemented for Argonne Reactor Computation (ARC) codes DIF3D, REBUS, and GAMSOR, and SE2-ANL. OTERR also has capabilities to facilitate input creation for DASSH, a thermal hydraulics code similar to SE2- ANL being developed for the VTR program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Special Considerations for the Removal and Disposal of Micro-Reactor Experiments

Idaho National Laboratory (INL) is preparing to host several microreactor experiments through 2030 and beyond. Two new test beds currently under development will operate as microreactor or nuclear system experiment user facilities. Test bed experiments will be performed in series, with each nuclear experiment installed, operated, and removed before installation of the subsequent experiment. The necessarily brief transition period between experiments introduces unique equipment removal and radioactive waste disposal challenges. This paper evaluates equipment removal and radioactive waste management topics associated with tight sequencing of nuclear experiments and presents some of the solutions and approaches currently planned to meet these special considerations for one such experiment, the Molten Chloride Reactor Experiment (MCRE), slated for operation at INL’s Laboratory for Operation and Testing in the United States (LOTUS). The MCRE project is a collaboration between Southern Company Services, TerraPower, and INL, among others, to provide integral nuclear data that will advance molten salt fast reactor technology. MCRE will be the first critical fast-spectrum circulating fuel system ever operated and the first experiment operated in the LOTUS test bed. The experiment will nominally operate at zero power with planned low power excursions as part of operational planning for MCRE has been to minimize at-power operations to limit fission product formation while still achieving experimental objectives. After the experiment is complete, MCRE will be allowed to radioactively decay for a short period (nominally 90 days) prior to system defueling, flushing, removal, and disposal of all MCRE equipment, readying the test bed for the next nuclear experiment. Equipment removal and radioactive waste disposal have been integral to MCRE project planning since the Cooperative Research and Development Agreement was formalized in 2021. Due to requirements for future use of the test bed, the MCRE system, of necessity, must be removed in a much shorter time frame than typical for historic reactor decommissioning projects at INL. This results in minimal time for radioactive decay, resulting in not only elevated radiation dose rates but also the presence of short-to-medium-lived isotopes not typically encountered in the reactor decommissioning and radioactive waste management space. Additional unique constraints placed on the project include lack of intrinsic remote-operations capabilities in the test bed, space constraints in the test bed once MCRE has been installed, and contamination minimization requirements to return the test bed to as-found conditions to enable future use. This paper discusses planned solutions to these challenges. Approaches for implementing remote or semi-remote technologies in a non-hot cell environment with limited space availability are discussed. The paper also summarizes the systems engineering approach for concept development and design of equipment removal systems, which are being implemented concurrent with the MCRE design phase, providing feedback to system designers to incorporate features enabling efficient and safe equipment removal approaches. The timing of this paper at a relatively early phase of the project is meant to highlight the importance of early planning for nuclear system decommissioning while reactor design is ongoing to allow design feedback on componentry driven from decommissioning system needs. As new, innovative nuclear reactor technologies enter the nuclear market sector, reactor experiments are crucial for providing new integral nuclear data that establish safe operational margins for technology advancement. Microreactor experiments at INL will require safe, effective, and timely decommissioning approaches, which in turn require nuclear systems designed to expedite decommissioning. In addition to the advancement of nuclear reactor technologies, these nuclear experiments provide an opportunity to demonstrate, deploy, and test new removal and disposal capabilities to support the next generation of nuclear reactor technology.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

History and status of spent fuel treatment at the INL Fuel Conditioning Facility

Experimental Breeder Reactor II (EBR-II) had an illustrious operational history from 1964 to 1994 as one of the first sodium-cooled fast reactors to use sodium-bonded metallic driver fuels and blankets. EBR-II was a testbed for fuel development and reactor design principles. Following shutdown, it was decided to use electrometallurgical methods to treat the spent fuel and blanket materials for final disposition. Here, this paper provides a brief history of U.S. liquid metal cooled reactor development with emphasis on EBR-II, and a technical description of the electrometallurgical methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

NEAMS Advanced Reactor Model Contributions to the NRIC Virtual Test Bed

The U.S. Department of Energy (DOE) Office of Nuclear Energy’s Advanced Modeling and Simulation (NEAMS) Program develops models of advanced reactor phenomena to demonstrate code applicability to challenging physics problems, drive code development through user assessment, and perform code verification and validation. Meanwhile, the U.S. DOE’s National Reactor Innovation Center (NRIC) hosts an open-source website and associated GitHub repository called the Virtual Test Bed (VTB) on which computational models for advanced reactors are documented and shared with the reactor community. Several NEAMS advanced reactor models (including input files, documentation, and discussion of results) have recently been contributed to the NRIC VTB. The open sharing of these models benefits the reactor community by providing “best practice” examples using NEAMS tools for advanced reactor physics problems. This report summarizes and provides links to these new models, which include modeling phenomena important to liquid metal cooled fast reactors, molten salt reactors, high temperature gas-cooled reactors, and microreactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

NSUF BOILER Pre-Irradiation Characterization and High Flux Isotope Reactor Experiment Design

Alumina-forming austenitic (AFA) stainless steels have emerged as a candidate alloy because of their high-temperature strength, formability, cost, and compatibility with primary coolants for lead-cooled fast reactors (LFRs). This class of steels has exceptional high-temperature oxidation performance; however, a high concentration of Ni is required to stabilize the austenite phase and to provide sufficient creep strength. AFA stainless steels are also susceptible to liquid metal embrittlement (LME). Additionally, under neutron irradiation, Ni will enrich at grain boundaries due to radiation-induced segregation (RIS). Nickel RIS can increase the LME under these coupled effects. Oak Ridge National Laboratory (ORNL) and the NSUF program have leveraged its High Flux Isotope Reactor (HFIR) and experience with complex irradiation experiments to design experiment capsules that test the aforementioned coupled effects. These capsules are designed for insertion in the central flux trap, the highest flux region, of HFIR. The experiment capsules will be filled with Pb, designed to passively melt from the gamma heating in HFIR. The specimens were fabricated into miniature tensile specimens from two different alloys, GA05-25Ni and GA05-20Ni, varying Ni concentrations. The experiment capsules are designed to achieve target temperatures of 400 °C and 650 °C with accumulated dosage of 3 dpa. This report documents the specimen alloy characterization, experimental design, and expected performance of the capsules.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

High bias machine learning for antineutrino-based safeguards for small reactors

The statistical methods used for antineutrino detection will need to be improved to effectively monitor the inventory of next-generation nuclear reactors. In this sensitivity study, we evaluate machine learning models compared to previously used statistical approaches to identify diversion scenarios in a simulated Advanced Fast Reactor (AFR)-100. A chi-square goodness-of-fit technique, which individually compares the simulated antineutrino yields to the expected antineutrino yield, resulted in precise but low diversion detection probability. Various support vector machine (SVM) models were applied with diverse training datasets to evaluate the robustness of the method towards unexpected or “unseen” diversion scenarios. Furthermore, our results indicate that while the SVM models significantly improved the detection probability of near-field antineutrino-based safeguards, up to a probability of ~0.04, for the simulated small reactor, the detection system still needs improvements to reach the 0.2 detection limit established by the International Atomic Energy Agency.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2020

Status of GTA Gear Inspection after Testing in Sodium

The Mechanisms Engineering Test Loop (METL) is an intermediate-scale liquid metal experimental facility that provides purified reactor-grade (R-grade) sodium to various experimental test vessels. In these test vessels, components that are required to operate in an advanced fast reactor can be tested in a prototypical sodium environment. Experiments conducted in METL significantly assist in the development and maturation of systems and components for advanced reactors. The METL facility consists of multiple test segments including: a purification and diagnostic loop (cold trap was designed after EBR-II), two 18” diameter test vessels with 150 L capacities capable of operating in static or dynamic flow at 538°C, two 28” diameter test vessels with 644 L capacities capable of operating in static or dynamic flow at 650°C, a 3,180 L dump tank with 21 instrumentation ports, and the infrastructure to expand capacity in the future.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

MOOSE Framework Meshing Enhancements to Support Reactor Analysis

MOOSE-based physics codes require an input finite element mesh on which the physics solution is calculated, reported, and transferred to other physics codes. The use of difficult-touse, external licensed software is often required to generate high quality meshes for reactor geometries. High-fidelity geometry modeling also requires elaborate tracking of groups of elements for material property assignment and output reporting which can be considerably complex for the user to identify and maintain. Under the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, several meshingrelated enhancements have been developed for the MOOSE framework to address user challenges in creating finite element meshes for advanced reactor geometries. MOOSE mesh generators have been developed to mesh hexagonal geometries (pins, ducted assemblies, and cores) commonly found in liquid-metal cooled fast reactor concepts. The mesh generator used for hexagonal pin cells is generic for regular polygons and therefore may also be used for Cartesian pin cells. Hexagonal pin cells can be stitched into ducted assemblies, and assemblies can be stitched together into a core. The user may specify region ids, region names, and other preferences on the mesh. This control is useful for later material mapping in the MOOSE-based physics codes input. A capability was also developed for meshing rotating control drums including determination of material volume fractions in each mesh element as a function of time. Control drum meshes may be stitched to other hexagonal assemblies to create a core configuration. Additional mesh generators were developed that wrap around the hexagonal meshing capabilities and utilize “extra element integer” ID values on each element. In regular Cartesian or hexagonal assemblies or cores, the bookkeeping of element groups for both material assignment and output reporting can now be automated through assignment of pin, assembly, core, axial and depletion id values stored as extra element integers. The extra element tags on the mesh greatly speed the reactor analyst’s efforts to map materials to meshes, track depletion zones, and parse output such as axial pin power distributions. At the highest level, pin, assembly, and core mesh generators (with this reactor terminology) have also been developed to easily generate regular Cartesian and hexagonal cores, including axial extrusion. These reactor geometry builders call upon the previously mentioned capabilities to produce analysis-ready 3D meshes including material assignments. Open source mesh triangulation capabilities were also investigated for integration into the MOOSE framework to address the need for meshing the core periphery region which extends from the irregular outer assembly border to a cylindrical boundary. Options are limited due to licensing constraints, and the recommendation is pursue building a native MOOSE Delaunay triangulator routine with full functionality. Finally, a series of verification problems were performed with NEAMS physics tools. All developed capabilities will be available in the new open-source “Reactor” module of the MOOSE framework, which is accessible to any MOOSE-based NEAMS physics tool.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Thermal Hydraulic Experimental Test Article - Fiscal Year 2023 (Final Report)

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that is used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility simulates nominal conditions as well as protected/unprotected loss of flow accidents in a sodium-cooled fast reactor (SFR). High fidelity distributed temperature profiles of the developed flow field may be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility was designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. THETA is comprised of a traditional primary coolant and secondary coolant system. The primary system is submerged in the pool of sodium and consists of a pump, electrically heated core, intermediate heat exchanger, and connected piping and thermal barriers (redan). The secondary system, located outside of the sodium pool, consists of a pump, sodium to air heat exchanger, and connected piping and valves. To date a test matrix has been completed utilizing the primary system of THETA. These tests, along with computational fluid dynamics and systems code models, determined the heat transfer across the core barrel and intermediate heat exchanger outlet was too great to effectively represent scaled thermal hydraulic phenomena of a liquid metal cooled reactor. Therefore, a significant effort was made to remove the primary system from the METL 28” test vessel #4 and clean the residual sodium from the primary system to facilitate upgrades. Thermal insulation was then incorporated in the core barrel and intermediate heat exchanger outlets. The primary system was then replaced, and a series of tests were performed to assess the performance of the thermal insulation. With primary system testing and upgrades complete, the secondary system could then be brought online. The tube side of the shell-and-tube intermediate heat exchanger was installed onto the primary system flange to begin installing the secondary system. The support structure for the secondary system was then erected on the METL mezzanine alongside the THETA primary system to facilitate installation of the secondary system components (sodium-to-air heat exchanger, flowmeter and pump). The piping and expansion tank were welded into the secondary system. Non-destructive examination of the secondary system welds was completed in order to satisfy ASME B31.3 pipe code for class M process fluids. The heating system and insulation were then added to prepare the system to be filled with sodium. The ancillary electrical equipment was installed which included the pump control box, blower VFD, pipe heater control system, etc. The secondary system will be filled, and a test matrix will be completed in early FY2024.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Cladding Profilometry Analysis of Experimental Breeder Reactor-II Metallic Fuel Pins with HT9, D9, and SS316 Cladding

BISON finite element method fuel performance simulations were conducted using an existing automated process that couples the Fuels Irradiation & Physics Database (FIPD) and the Integral Fast Reactor Materials Information System database by writing input files and comparing the BISON output to post-irradiation fuel pin profilometry measurements contained within the databases. The importance of this work is to demonstrate the ability to benchmark fuel performance metallic fuel models within BISON using Experimental Breeder Reactor-II fuel pin data for a number of similar pins, while building off previous modeling efforts. Changes to the generic BISON input file include implementing pin specific axial power and flux profiles, pin specific fluences, frictional contact, and irradiation-induced volumetric swelling models for cladding. A statistical analysis of irradiation-induced volumetric swelling models for HT9, D9, and SS316 was performed for experiments X421/X421A, X441/X441A, and X486. Between these three experiments, there were 174 post-irradiation examination (PIE) profilometries used for validating the swelling models presented using a standard error of the estimate (SEE) method. Implementation of the volumetric swelling models for D9 and SS316 claddings was found to have a significant impact on the BISON profilometry simulated, where HT9 clad pins had an insignificant change due to low fluence values. BISON profilometry simulated for HT9, D9, and SS316 fuel pins agreed with PIE profilometry measurements, with assembly SEE values being 4.4 × 10−3 for X421A, 2.0 × 10−3 for X441A, and 2.8 × 10−3 for X486. D9 clad pins in X421/X421A had the highest SEE values, which is due to the BISON simulated profilometry being shifted axially. While this work accomplished its purpose to demonstrate the modeling of multiple fuel pins from the databases to help validate models, the results suggest that the continued development of metallic fuel models is necessary for qualifying new metallic fuel systems to better capture some physical performance phenomena, such as the hot pressing of U-Pu-Zr and the fuel cladding chemical interaction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Thermal Hydraulic Experimental Test Article (FY2020 Status Report)

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that will be used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility will simulate nominal conditions as well as protected/unprotected loss of flow accidents in a sodium-cooled fast reactor (SFR). High fidelity distributed temperature profiles of the developed flow field will be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility is being designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. THETA is comprised of a traditional primary and secondary system. The primary system is submerged in the pool of sodium and consists of a pump, electrically heated core, intermediate heat exchanger, and connected piping and thermal barriers (redan). The secondary system, located outside of the sodium pool, consists of a pump, sodium to air heat exchanger, and connected piping and valves. Figure 1 illustrates the main components of the primary system. THETA will be installed in the Mechanisms Engineering Test Loop (METL) with the primary system in the 28 inch Test Vessel #4, Figure 2. The design work for the facility has been completed. Since the FY19 THETA status report, the THETA primary system components have been assembled and tested using de-ionized water. Some results of this testing at select Richardson and Reynolds numbers are included herein. The custom THETA sodium flowmeter (permanent magnet submersible sodium flowmeter) has been built and calibrated using sodium in an auxiliary facility against a NIST traceable vortex shedder type flowmeter. The permanent magnet flowmeter showed excellent performance, following the theoretical equation for measured voltage quite well, with an error < 3.5% at flow rates ranging from 3-24 GPM at temperatures up to 400 °C. The secondary sodium system for heat rejection from intermediate heat exchanger has been analyzed according to ASME code. All secondary side piping, fittings and a U-stamped air-to-sodium heat exchanger have been procured.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Thermal Hydraulic Experimental Test Article - Report of THETA Commissioning for METL Testing

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that will be used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility will simulate nominal conditions as well as protected/unprotected loss of flow accidents in a sodium-cooled fast reactor (SFR). High fidelity distributed temperature profiles of the developed flow field will be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility is being designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. THETA is comprised of a traditional primary and secondary system. The primary system is submerged in the pool of sodium and consists of a pump, electrically heated core, intermediate heat exchanger, and connected piping and thermal barriers (redan). The secondary system, located outside of the sodium pool, consists of a pump, sodium to air heat exchanger, and connected piping and valves. Figure 1 illustrates the main components of the primary system. THETA has been installed in the Mechanisms Engineering Test Loop (METL) with the primary system in the 28 inch Test Vessel #4, Figure 2 and Figure 3. Since the FY20 THETA status report, [1], further water shakedown testing was performed, some final welding modifications were performed, all system components were sanitized, dry assembled, and commissioned before installing into METL test vessel #4. The THETA submersible electromagnetic flowmeter was welded closed to hermetically seal internal components from sodium followed by a helium leak check for the final time

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗