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At least 109 records · Page 6
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DOE OSTI2021

Technical analysis of filter testing at the U.S. Department of Energy Filter Test Facility

The U.S. Department of Energy (DOE) technical standard DOE-STD-3020-2015 requires verification testing at the Filter Test Facility (FTF) of high-efficiency particulate air (HEPA) filters to be installed in nuclear facility (Hazard Category 1, 2, 3, and radiological facilities) confinement ventilation systems, or to be installed in habitability systems (e.g., filters for protection of workers required to act for the control and mitigation of emergency situations). Filters in these applications perform a safety function in accident situations, are designated as important to safety, or are necessary for habitability systems. Different FTFs have operated for decades, and results of tests are available to examine the need for and benefits of continued execution of 100 percent QA verification testing. The objective of this study and resulting report was to assess the need for 100 percent QA testing at the current FTF in Maryland, based on an examination of a database of test results from 2013 to August 2020. Filters that require independent testing per DOE-STD-3020-2015 to be installed in nuclear facilities are sent to the FTF before being shipped to the facilities. The FTF examines the documentation (e.g., certificates of compliance, results of particle penetration and airflow resistance testing by the manufacturers, labels affixed to filters and packages, and other requirements specified in purchase orders by the nuclear facility operators), implements visual inspections of the physical state of the filters, and checks the dimensions of frames. Filters that meet the documentation and visual inspection requirements are subjected to particle penetration and airflow resistance testing. In case of rejections, the filters are shipped back to the manufacturers, at the cost of the manufacturers, for replacement. Filters that pass all tests are shipped to the nuclear facilities.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2021

Gear Test Assembly – Report Experimental Testing and Gear Analysis (FY2021 Midyear Report)

The Gear Test Assembly (GTA) has completed three experimental campaigns at the Mechanisms Engineering Test Loop (METL) facility. The most recent campaign tested Inconel 718 spur gears, tapered roller bearings of 52100 bearing steel with no heat treatment, and a lantern ring shaft seal with spring-loaded fasteners. Testing was performed in Test Vessel 1 (TV1) with 250°C sodium that had an oxide concentration of <10ppm. Testing began in February 2021 and was ended in March 2021 when a thrust bearing assembly failed, requiring shutdown and maintenance. The third experimental campaign completed 1,568 simulated fuel assembly maneuvers before the thrust bearing failure. The GTA was removed from TV1 and cleaned in the Carbonation System. The GTA is currently being prepared for the next experimental campaign. Nondestructive examination methods have been developed at Argonne to monitor the health of the GTA spur gears. Eddy current testing (ECT) and ultrasonic testing (UT) were performed prior to sodium testing and after each experimental campaign. The NDE has shown that the primary damage mechanism is mechanical wear on the gear tooth surface. Several larger nicks are observable on the gear faces, with additional large nicks present on the edges of the gear teeth. While this damage is observable, the overall health of the gears is adequate for more in-sodium testing. A total of 12,752 simulated fuel assembly maneuvers have been completed using this set of gears.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Design and Evaluation of a Ring Tension Test Grip for Remote Mechanical Testing of Irradiated Tubular Specimens

Here, the ring tension test (RTT) is a mechanical testing method for determining bulk mechanical behavior in the circumferential or hoop direction for tubular materials. The test is especially useful for testing materials with anisotropic mechanical properties, such as zirconium alloys, which are commonly used as nuclear fuel cladding. Anisotropy requires direction-specific testing to determine the hoop strength. Historically, several RTT methods and grips have been used, each method has its strengths and weaknesses, and, in all cases, the measured strength is subject to uncertainty due to variations of the testing geometry and experimental tolerances. Recent analysis has shown that grips with a hemicylindrical mandrel configuration are recommended as the most robust configuration. The two strictest aspects to be controlled are the ability to determine gage region orientation and closely matching the size of the mandrels to the test specimen. This last requirement is particularly challenging when the dimensions of the specimen vary because of environmental effects such as dimensional changes due to irradiation. This paper presents a new RTT grip designed to incorporate this mandrel shape, hold the gage at the desired orientation, be suitable for remote operation in a hot-cell environment, and be adaptable for different sizes or variations in the specimen size. The general description and the unique design features of the test specimen and grips are given in detail. The performance of the grips in mechanical testing, including in a remote hot-cell environment, is also provided.

42 ENGINEERING↗
DOE OSTI2021

An Artificial-Intelligence and Machine-Learning-Based Methodology to Conduct Seemingly Strain-Controlled Fatigue Test in a Pressurized-Water-Reactor-Test-Loop-Autoclave, While Not Controlling the Strain

In general, low cycle fatigue analysis of pressurized water reactor (PWR) components, requires strain-controlled fatigue test data such as using strain versus life (ε–N) curves. Conducting strain-controlled fatigue tests under in-air conditions is not an issue. However, controlling strain in a PWR-test-loop-autoclave is a challenge, since an extensometer cannot be placed in a narrow autoclave (typically used in a high-temperature-pressure PWR-test-loop). This is due to lack of space inside an autoclave that houses the test specimen. In addition, installing a contact-type extensometer in the path of a high-pressure flow can be a challenge. These difficulties of using an extensometer inside an autoclave led us to use an outside-autoclave displacement sensor which measures the displacement of pull-rod-specimen assembly. However, in our study (based on in-air fatigue test data), we found that a pull-rod-controlled based fatigue test can lead to substantial cyclic hardening/softening resulting in substantially different cyclic strain amplitudes and their rates compared to the desired cyclic strain amplitudes and its rates. In this paper, we propose an Artificial-Intelligence and Machine-Learning based technique such as using k-means clustering technique to improve the pull-rod-control based fatigue test method, such that the gage-area strain amplitude and rates can reasonably be achieved. In support of this, we present the fatigue test results for both 316 SS base and 81/182 dissimilar-metal-weld specimens.

42 ENGINEERING↗
DOE OSTI2023

Conceptual Neutronics Scoping for Ramp Testing in the Advanced Test Reactor

With the closer of the Halden Boiling Water Reactor in 2018, the global capacity for in-reactor power ramp testing has been lost. As such tests provide valuable data for understanding pellet-cladding interaction (PCI) phenomena, finding new facilities for ramp testing is of interest to the United States Accident-Tolerant Fuel (ATF) Program. Several options at Idaho National Laboratory (INL) are being evaluated for adding ramp test capabilities. In the Transient Reactor Test Facility (TREAT), one option is testing inside a Transient Water Irradiation System in TREAT (TWIST) capsule. In the Advanced Test Reactor (ATR), one option under consideration is testing in an I-Loop toward the outer edge of the core. Another is testing in Loop-2A, located in the Center Flux Trap (CFT) of the ATR. In this paper, we examine the conceptual scoping for the Loop-2A experiment known as ATF-2Ramp.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Conceptual Neutronics Scoping for Ramp Testing in the Advanced Test Reactor

With the closure of the Halden Boiling Water Reactor in 2018, the global capacity for in-reactor power ramp testing has been lost. As such tests provide valuable data for understanding pellet-cladding interaction (PCI) phenomena, finding new facilities for ramp testing is of interest to the United States Accident-Tolerant Fuel (ATF) Program. Several options at Idaho National Laboratory (INL) are being evaluated for adding ramp test capabilities. In the Transient Reactor Test Facility (TREAT), one option is testing inside a Transient Water Irradiation System in TREAT (TWIST) capsule. In the Advanced Test Reactor (ATR), one option under consideration is testing in an I-Loop toward the outer edge of the core. Another is testing in Loop-2A, located in the Center Flux Trap (CFT) of the ATR. In this paper, we examine the conceptual scoping for the Loop-2A experiment known as ATF-2Ramp.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Development of Irradiation Test Devices for Transient Testing

The Transient Reactor Test facility (TREAT) resumed operations in 2017 in order to reclaim its crucial role in nuclear-heated fuel safety research. TREAT’s historic era of operation (1959-1994) was best known for integral-scale testing of large fuel specimens/bundles under postulated reactor plant accident conditions, but TREAT also supported smaller-scale phenomena identification tests that elucidated fundamental behaviors and paved the way for these integral-scale tests. Advances in modern computational capabilities and a resurgence of interest in novel reactor technologies have created an opportunity for emphasizing modernized science-based and separate effects tests once again at TREAT. An innovative approach to this type of testing has been developed to leverage minor radioactivity built-in during brief TREAT irradiations by arranging smaller fuel specimens in low activation hardware so that they can be easily extracted and shipped for post irradiation examination within weeks. This recently-established capability, termed the Minimal Activation Retrievable Capsule Holder (MARCH) irradiation vehicle system, includes capabilities for cost-effective simplified-environment testing of centimeter-scale fuel samples of various geometries, temperature-controlled irradiations of millimeter-size samples for lower-length-scale model development, liquid-metal bonded heat sink capsules for controlling transient temperature response in fuel rodlets, and an innovative approach to high throughput irradiation of transient sensors and instrumentation. The MARCH system’s capabilities will also set the foundation for fuel safety research performed in larger integral-scale test devices with coolant environments representing reactor plants. Based upon historic approaches, but modernized to meet current nuclear technology needs, these larger irradiation devices include flowing pressurized water (including the ability to depressurize to steam) as well liquid metal cooling loops for various fuel rod and small bundle specimens. Here, this paper describes the recently-established MARCH system and current trajectory to enabling advanced transient science with a suite of irradiation test devices.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Equivalent Test Bed in PSCAD and PSLF for Studying Advanced Power Systems Controller Performance

With increasing popularity of sustainable energy resources, research on power electronic-based renewable generation modeling is gaining impetus to enhance reliability of electric grids with high penetrations of renewable energy. A set of generic and public models for renewable energy are available in commercial positive sequence power system simulation software. However, advanced controller models applicable for renewable resources are typically not included in these tools. To ensure the integrity of the newly developed controller models for renewables in positive sequence simulators for large power systems, it is necessary to validate the dynamic performance of the proposed model with detailed electromagnetic models. This paper creates an equivalent test bed in PSCAD and PSLF for validating the performance of advanced controllers associated with renewable energy sources. To demonstrate the usefulness of the test-bed, a user-defined synthetic inertia (SI) controller model was developed in PSLF and validated through the proposed test bed.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2019

Estimation of Radiation Doses to U.S. Military Test Participants from Nuclear Testing: A Comparison of Historical Film-Badge Measurements, Dose Reconstruction and Retrospective Biodosimetry

Retrospective radiation dose estimations, whether based on physical or biological measurements, or on theoretical dose reconstruction, are limited in their precision and reliability, particularly for exposures that occurred many decades ago. Here, we studied living U.S. military test participants, believed to have received high-dose radiation exposures during nuclear testing-related activities approximately six decades ago, with two primary goals in mind. The first was to compare three different approaches of assessing past radiation exposures: 1. Historical personnel monitoring data alone; 2. Dose reconstruction based on varying levels of completeness of individual information, which can include film badge data; and 3. Retrospective biodosimetry using chromosome aberrations in peripheral blood lymphocytes. The second goal was to use the collected data to make the best possible estimates of bone marrow dose received by a group with the highest military recorded radiation doses of any currently living military test participants. Six nuclear test participants studied had been on Rongerik Atoll during the 1954 CASTLE Bravo nuclear test. Another six were present at the Nevada Test Site (NTS) and/or Pacific Proving Ground (PPG) and were believed to have received relatively high-dose exposures at those locations. All were interviewed, and all provided a blood sample for cytogenetic analysis. Military dose records for each test participant, as recorded in the Defense Threat Reduction Agency's Nuclear Test Review and Information System, were used as the basis for historical film badge records and provided exposure scenario information to estimate dose via dose reconstruction. Dose to bone marrow was also estimated utilizing directional genomic hybridization (dGH) for high-resolution detection of radiation-induced chromosomal translocations and inversions, the latter being demonstrated for the first time for the purpose of retrospective biodosimetry. As the true dose for each test participant is not known these many decades after exposure, this study gauged the congruence of different methods by assessing the degree of correlation and degree of systematic differences. Overall, the best agreement between methods, defined by statistically significant correlations and small systematic differences, was between doses estimated by a dose reconstruction methodology that exploited all the available individual detail and the biodosimetry methodology derived from a weighted average dose determined from chromosomal translocation and inversion rates. Employing such a strategy, we found that the Rongerik veterans who participated in this study appear to have received, on average, bone marrow equivalent doses on the order of 300–400 mSv, while the NTS/ PPG participants appear to have received approximately 250–300 mSv. The results show that even for nuclear events that occurred six decades in the past, biological signatures of exposure are still present, and when taken together, chromosomal translocations and inversions can serve as reliable retrospective biodosimeters, particularly on a group-average basis, when doses received are greater than statistically-determined detection limits for the biological assays used.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2014

Final Report: Part 1. In-Place Filter Testing Instrument for Nuclear Material Containers. Part 2. Canister Filter Test Standards for Aerosol Capture Rates.

A portable instrument has been developed to assess the functionality of filter sand o-rings on nuclear material storage canisters, without requiring removal of the canister lid. Additionally, a set of fifteen filter standards were procured for verifying aerosol leakage and pressure drop measurements in the Los Alamos Filter Test System. The US Department of Energy uses several thousand canisters for storing nuclear material in different chemical and physical forms. Specialized filters are installed into canister lids to allow gases to escape, and to maintain an internal ambient pressure while containing radioactive contaminants. Diagnosing the condition of container filters and canister integrity is important to ensure worker and public safety and for determining the handling requirements of legacy apparatus. This report describes the In-Place-Filter-Tester, the Instrument Development Plan and the Instrument Operating Method that were developed at the Los Alamos National Laboratory to determine the “as found” condition of unopened storage canisters. The Instrument Operating Method provides instructions for future evaluations of as-found canisters packaged with nuclear material. Customized stainless steel canister interfaces were developed for pressure-port access and to apply a suction clamping force for the interface. These are compatible with selected Hagan-style and SAVY-4000 storage canisters that were purchased from NFT (Nuclear Filter Technology, Golden, CO). Two instruments were developed for this effort: an initial Los Alamos POC (Proof-of-Concept) unit and the final Los Alamos IPFT system. The Los Alamos POC was used to create the Instrument Development Plan: (1) to determine the air flow and pressure characteristics associated with canister filter clogging, and (2) to test simulated configurations that mimicked canister leakage paths. The canister leakage scenarios included quantifying: (A) air leakage due to foreign material (i.e. dust and hair) fouling of o-rings, (B) leakage through simulated cracks in o-rings, and (C) air leakage due to inadequately tightened canister lids. The Los Alamos POC instrument determined pertinent air flow and pressure quantities, and this knowledge was used to specify a customized Isaac® (Z axis, Salt Lake City, UT) leak test module. The final Los Alamos IPFT (incorporating the Isaac® leak test module) was used to repeat the tests in the Instrument Development Plan (with simulated filter clogging tests and canister leak pathway tests). The Los Alamos IPFT instrument is capable of determining filter clogging and leak rate conditions, without requiring removal of the container lid. The IPFT measures pressure decay rate from 1.7E-03 in WC/sec to 1.7E-01 in WC/sec. On the same unit scale, helium leak testing of canisters has a range from 5.7E-07 in WC/sec to 1.9E-03 in WC/sec. For a 5-quart storage canister, the IPFT measures equivalent leak flow rates from 0.03 to 3.0 cc/sec. The IPFT does not provide the same sensitivity as helium leak testing, but is able to gauge the assembled condition of as-found and in-situ canisters.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

Choreographer Pre-Testing Code Analysis and Operational Testing.

Choreographer is a "moving target defense system", designed to protect against attacks aimed at IP addresses without corresponding domain name system (DNS) lookups. It coordinates actions between a DNS server and a Network Address Translation (NAT) device to regularly change which publicly available IP addresses' traffic will be routed to the protected device versus routed to a honeypot. More details about how Choreographer operates can be found in Section 2: Introducing Choreographer. Operational considerations for the successful deployment of Choreographer can be found in Section 3. The Testing & Evaluation (T&E) for Choreographer involved 3 phases: Pre-testing, Code Analysis, and Operational Testing. Pre-testing, described in Section 4, involved installing and configuring an instance of Choreographer and verifying it would operate as expected for a simple use case. Our findings were that it was simple and straightforward to prepare a system for a Choreographer installation as well as configure Choreographer to work in a representative environment. Code Analysis, described in Section 5, consisted of running a static code analyzer (HP Fortify) and conducting dynamic analysis tests using the Valgrind instrumentation framework. Choreographer performed well, such that only a few errors that might possibly be problematic in a given operating situation were identified. Operational Testing, described in Section 6, involved operating Choreographer in a representative environment created through Emulytics TM . Depending upon the amount of server resources dedicated to Choreographer vis-á-vis the amount of client traffic handled, Choreographer had varying degrees of operational success. In an environment with a poorly resourced Choreographer server and as few as 50-100 clients, Choreographer failed to properly route traffic over half the time. Yet, with a well-resourced server, Choreographer handled over 1000 clients without missrouting. Choreographer demonstrated sensitivity to low-latency connections as well as high volumes of traffic. In addition, depending upon the frequency of new connection requests and the size of the address range that Choreographer has to work with, it is possible for all benefits of Choreographer to be ameliorated by its need to allow DNS servers rather than the end client to make DNS requests. Conclusions and Recommendations, listed in Section 7, address the need to understand the specific use case where Choreographer would be deployed to assess whether there would be problems resulting from the operational considerations described in Section 3 or performance concerns from the results of Operational Testing in Section 6. Deployed in an appropriate architecture with sufficiently light traffic volumes and a well-provisioned server, it is quite likely that Choreographer would perform satisfactorily. Thus, we recommend further detailed testing, to potentially include Red Team testing, at such time a specific use case is identified

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Comparison of Light-Weight Radioisotope Heater Unit (LWRHU) Historical Testing to Special Form Testing Requirements

This paper compares the Light-Weight Radioisotope Heater Unit (LWRHU) historical testing to Department of Transportation (DOT) Special Form criteria per 49 CFR Part 173, Shippers-General Requirements for Shipments and Packages, Section 469, Tests for special form Class 7 (radioactive) materials as specified in DOE-STD-1027, Hazard Categorization of DOE Nuclear Facilities. Compliance with the aforementioned criteria allows for the exclusion of radioactive sources from the facility’s nuclear material inventory. As part of a new program at Los Alamos National Laboratory (LANL), LWRHUs are planned to be moved into TA-55, PF-5 Power Source Assembly Area (PSAA) for assembly and testing activities. The PSAA Facility is categorized as a low hazard, non-nuclear facility per DOE STD-1027 and is described in TA55-PF-5-FHC-11-001, Facility Hazard Categorization of the PSAA at PF-5, TA-55. The Facility Hazard Categorization document states, there must be documentation that radioactive sources are engineered to pass the special form testing specified by DOT in 49 CFR 173.469, Tests for special form Class 7 (radioactive) materials or ANSI N43.6, Sealed Radioactive Sources - Classification. This paper documents the extensive historical environmental safety testing of the LWRHU and demonstrates its ability to survive potential launch safety accident scenarios. The testing results are used herein to demonstrate through a comparative analysis that the robust design of the LWRHU meets or exceeds the special form testing specifications in 49 CFR 173.469, enabling the LWRHU to be transported and handled in PF-5 in support of testing and assembly program.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Mechanisms Engineering Test Loop (METL) Operations and Testing Report (FY2022)

This report documents the operations, testing, maintenance, and improvements that were performed at the Mechanisms Engineering Test Loop (METL) during FY2022. The METL facility had a very successful fourth year of operations and went through some new experiences such as taking the facility from its frozen state in October 2021 and thawing the facility and also recovering from an oxide plug in one of the cold trap lines. After the facility was fully thawed, the METL staff maintained the facility in a continuous molten state – either flowing sodium or a static flow condition. METL facility continued supporting the Gear Test Assembly (GTA) testing and hosted its first 28” test vessel experiment, called the Thermal Hydraulic Experimental Test Article (THETA) which was inserted into Test Vessel 4. Work to accommodate two additional experiments, a gripper test article (GrTA) and a flow sensor test article (F-STAr) continued as they are expected to make their debut in FY2023. In addition, the steam piping for the B308 scrubber unit was replaced with stainless steel piping and controls and new parts for the scrubber were purchased – such as a new pump, new storage tank, and new blower and motor. These components will ultimately replace the existing scrubber components when those 1970’s era require replacement.

42 ENGINEERING↗
DOE OSTI2023

A simulation study of test fixture attachment points for vibration test optimization (MEEN 5440 Final Project)

The aerospace industry uses vibration shaker tables to perform component durability testing. These dynamic environments tests often have a hard time replicating the service environment of a component due to a mismatch in the applied force or the boundary condition. Figure 1 shows the desired outcome of a dynamic environments test. The measured responses in a lab test, ideally, should match the measured response of the service environment field test. One way to address the mismatch between field environment and lab environment is with test fixture design. The test fixture attaches the component under test to the shaker table. Single-axis shakers are the industry standard type of shaker table, and, therefore, test fixtures are often rigid in order to minimize cross-axis responses.

42 ENGINEERING↗
DOE OSTI2023

Scoping Study for Fast Flux Testing in the Advanced Test Reactor

The value of fast spectrum reactors remains prominent in the nuclear technology portfolio. The performance of these reactors can be maximized with advancements in nuclear fuel technologies, but development of these technologies is currently held back by lack of fast spectrum test reactors available to the United States. Spectral modification of experiment positions in the thermal spectrum Advanced Test Reactor (ATR) has long been used to support fast reactor fuel development, but these methods have not been progressed to their full potential. This study investigated the use of concentric rings of aluminum-clad fuel plates in ATR flux traps and thermal neutron absorbing filters to increase fast neutron flux on test specimens. This concept was termed the Boosted Energy Advanced Spectrum Test (BEAST). This approach will enable irradiation of advanced fuel designs in prototypic-length fuel pins and representative flux environment to support post irradiation exams, enable transient testing, and produce the type of data that will permit lead test assembly irradiations in true Sodium Fast Reactors (SFRs) when they become available. Neutronic predictions were performed to investigate BEAST design options and thermal hydraulic models were produced to ensure feasibility of BEAST. Two versions were considered based on the geometric limitations of ATR’s small and large flux traps. The small version was found to be preferable due to slightly higher fast flux and fast-to-thermal neutron ratio. Perhaps more influentially, the small flux trap option was also preferred to avoid conflict with ongoing very high temperature reactor fuel irradiation programs in ATR’s large northeast flux trap. The small flux trap option provided less than half the test volume of the large version, but still had adequate volume for seven SFR pins in cross section which could be stacked two-high in ATR’s 1.2m long core to accommodate up to 14 EBR-II size pins. The preference for the small flux trap configuration should be revisited if additional collaborative test programs emerge with the need to irradiate a significant volume of additional specimens. Calculations were performed regarding a lithium deuteride ring to convert thermal neutrons into 14 MeV fusion neutrons. At the time this report was written these calculations were partially complete and it remains to be seen whether the concept would be worth including in BEAST. Given the preference for the small flux trap option, which does not afford enough room for the 14 MeV ring, it was concluded to defer future work on the lithium deuteride ring. This decision could be revisited if fusion material research programs emerge for collaborative testing in BEAST. A cadmium-lined specimen holder design was found to be adequate in filtering thermal neutrons and preferred over other neutron absorbers based on past experience with cadmium baskets. It was acknowledged that cadmium-bearing hardware would become depleted and need to be replaced occasionally, which appeared feasible from a mechanical design perspective. Neutronic studies investigated different enrichment levels in the booster fuel using uranium-molybdenum alloy dispersion fuel which has performed well in past ATR irradiations. Both options were able to drive fuel pins to SFR-like fission heating rates. The high enriched booster fuel option outperformed the low enriched option by ~20% on key metrics including fast flux and fast-to-thermal ratio, but the low enriched option was favored in order to broaden options for potential fuel suppliers. The preferred BEAST design options including cadmium filter with low enriched booster fuel in the small flux trap configuration was predicted to achieve 6.2E14 n/cm2sec fast flux (>0.1 MeV) with a fast-to-thermal ratio of 44.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Gear Test Assembly - Experimental Testing and Analysis of Gears and Bearings - FY2026

The Gear Test Assembly (GTA) is an experimental test apparatus built and installed in the Mechanisms and Engineering Test Loop (METL) at Argonne National Laboratory (ANL). While designed to accommodate a variety of components intended for use in advanced compact in-vessel transfer machine, the focus of GTA been the testing of large radial spur gears. The performance of the components used in GTA motivates the choices made in the design, components, and material choices for the forthcoming METL test articles; the Gripper Test Assembly (GrTA), and the Bearing Test Assembly (BTA). To date, GTA has completed nine experimental testing campaigns and achieved over 21 million revolutions under prototypic sodium fast reactor loads and operating conditions. This report provides an overview and results for the two most recent experiments Campaign #8 and Campaign #9. `Campaign #8 was the first campaign to make use of a set of nickel alloy radial spur gears, made from Hastelloy C-22HS, and combined with heat-treated tapered roller bearings. The bearings and torque profile used allowed Campaign #8 to be directly compared with Campaign #1 where Inconel 718 gears were used instead. Campaign #8 attained 1,058,880 shaft revolutions within 23% of the 1,314,855 revolutions achieved in Campaign #1. These results demonstrate that the heat-treated tapered roller bearings show consistent increased lifetimes, approximately three times longer than for non heat-treated taper rollers, and that the particular material of the gear does not play a substantial role in performance. Similar to Campaign #7 where bearing failure resulted in fragmentation and transport of bearing material into the gear teeth led to tooth damage in the Inconel gears, a similar process led to the damage of the Hastelloy gears. Subsequent NDE analysis showed that the damage was only limited to the tooth surface and does not extend into the gear interiors. Based on the NDE reports, the Hastelloy gears can be reconditioned and put back into service for future campaigns as was done for the Inconel gears. Reconditioned Inconel 718 gears damaged in Campaign #7 were returned to service in Campaign #9 and paired with non heat-treated tapered roller bearings. In Campaign #9 a novel set of operating conditions were implemented whereby the gears and bearings were cycled between hot, submerged refueling conditions where fuel handling operations occurred and cold, dry, and inert conditions where the sodium was drained and the vessel was allowed to cool simulating long, ex-vessel storage of the fuel handling machine. A total of eight of these operating cycles were achieved before a bearing failure occurred after the accumulation of 475,692 shaft revolutions were achieved. During Campaign #9, one of the drive motors was damaged and over the course of a 240 day standby period, the entire GTA drive system was upgraded from a 480VAC system to a 240VAC system that provided the opportunity to consolidate wall-space in the METL facility and provide more space for additional test articles being installed in the coming year. The over 475,000 shaft revolutions achieved in Campaign #9 eclipses the previous record for lifetime of non heat-treated tapered roller bearings held by Campaign #3 of 392,000 shaft revolutions. Further analysis and future campaigns hope to shed light on the cause of this approximately 17% increase in lifetime.

42 ENGINEERING↗
DOE OSTI2024

Numerically Testing Conceptual Models of the Utah FORGE Reservoir Using July 2023 Circulation Test Data

Over the past several years, many new data sets have become available regarding the characterization of the Utah FORGE reservoir. These include, but are not limited to, the stimulation of Well 16A, the drilling and completion of Well 16B, and interwell circulation confirmatory testing. As part of the characterization efforts, conceptual models of the reservoir are re-examined as new data become available. As part of the planning for FORGE activities, numerical models are often used to predict the reservoir response to the planned testing. Stochastic methods are often employed to bound uncertainty and allow for evaluation of comprehensive ranges of key reservoir parameters. For the most recent interwell circulation confirmatory testing (July 2023), a priori numerical model predictions did bound the observed behavior (Xinj et al., 2023), but key deviations from expected behavior prompted the FORGE team to reevaluate our conceptual model of the reservoir. In early October 2023, key members of the development, testing, and monitoring teams met for 2 days to review newly collected data and discuss ‘interesting’ or ‘key’ observations. From these discussions, 15 Key Observations were documented, with several significant ones being that the discrete fracture network developed from the 16A stimulation data may not be appropriate and that the early time pressure data obtained during the summer 2023 reservoir testing were best described using radial solutions. In July 2023, two campaigns of interwell confirmatory testing were conducted, the first set of tests occurred on July 4-5 and the second set on July 18-19. The second set of circulation tests conducted at the Utah FORGE site between the injection well 16A(78)-32 and production well 16B(78)-32 on July 18 and 19, 2023 are used to calibrate material properties in a thermal-hydraulic-mechanical (THM) simulation of the discrete fracture network connecting the wells. The spatially and temporally varying reservoir properties are calibrated to match the time dependent pressure and production profiles from the circulation tests. In future work, this calibrated model will be coupled to the native state THM model of the FORGE reservoir to predict surface deformation and strains resulting from pumping schedules.

58 GEOSCIENCES↗