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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Understanding Materials Characteristics of Palladium Lanthanide Cermet Wires as a Mimic of Californium Wire Production

Californium-252 (Cf-252) has been synthesized on a large scale (mg) since the 1970s for use as a portable neutron source. With a half-life of 2.645 years and a spontaneous fission fraction of 3.09%, Cf-252 emits 2.31 × 10 12 neutrons/gram second. Today most of the Cf-252 produced in the world is made at Oak Ridge National Laboratory (ORNL) using the High Flux Isotope Reactor (HFIR) and the Radiochemical Engineering Development Center (REDC). Because of its highly radioactive nature, all manipulations of Cf-252 need to be performed in hot cell facilities by highly trained personnel. The Cf-252 produced at ORNL is done at the REDC, which is conveniently located next to the HFIR. Cf-252 is sold as bulk wire, which has been described as palladium (Pd) metal incasing californium oxide (Cf 2 O 3 ). The process was developed by Mosley et al. at Savannah River Laboratory (SRL) where it was modeled after the thorium work done by Fuschillo et al. This process has been used since the 1970s to synthesize this unique wire, but its fundamental chemistry and materials characteristics were never openly reported. Here we aim to better understand the chemistry and the materials characteristics of Cf-252 wire production for process improvements and to better our fundamental understanding of the mid-late actinides. Though there are many challenges in working with radioactive elements such as Cf, the isotope Cf-252 is even more complicated due to its neutron emission. One challenge with neutron emission is the ability of neutrons to penetrate much deeper through materials, causing radiation damage. This makes using instrumentation to characterize the chemistry and materials properties of this process difficult. For this reason, this study will first focus on terbium (Tb) as a surrogate for Cf. Currently Tb is used as a carrier or filler in Cf production when small amounts of Cf are being processed. This, along with their similar size and charge, makes Tb a reasonable surrogate for this study.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Status and Availability of OPTD, the Out-of-Pile Transient Database

The DOE Advanced Reactor Technology program has supported knowledge preservation efforts to recover and preserve fuel data from the US sodium cooled fast reactor (SFR) program. Databases have been established containing experimental data generated during the Integral Fast Reactor program (from in-pile experiments at EBR-II, FFTF, and TREAT) and during related outof-pile examinations of EBR-II-irradiated fuels. The information in these databases is essential to support further development and licensing activities for advanced fast reactor designs. OPTD, the Out-of-Pile Transient Database, is an organized, searchable archive of records describing out-ofpile furnace transient tests conducted on metallic fuels. It includes records of over 150 tests on irradiated fuel pins conducted in furnace apparatuses constructed in the Alpha-Gamma Hot Cell Facility at ANL. This report presents the content, functions, and accessibility of the Out-of-Pile Transient Database (OPTD). It describes the information presently available in the database, the web interface created to provide access to this information, the activities undertaken to build OPTD, and the enabling of online access to the database for DOE users. Plans for FY2021 and beyond are also included.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Fiscal Year 2020 Filtration of Hanford Tank Waste 241-AP-105

Bench-scale filtration testing of 15 liters of 241-AP-105 supernatant was conducted using a Mott inline filter Model 6610 (media grade 5) backpulsed dead-end filter (BDEF) in the hot cells of the Radiochemical Processing Laboratory at Pacific Northwest National Laboratory. The as-received samples were diluted with raw water from the Hanford Site to assess the propensity for solids to form upon dilution and to assess the impact those solids have on filtration performance. The feed was split into two batches. The first batch was settled and decanted as planned in Tank Side Cesium Removal operations. The BDEF was used to filter this AP-105 feed at a targeted flux of 0.065 gpm/ft 2 and exhibited no measurable pressure increase during filtration. The second batch was fed to the BDEF with solids resuspended. This was also operated at the targeted flux of 0.065 gpm/ft 2 . In this case, the target pressure differential of 2 psid (resistance of 6.24 × 10 10 m -1 ) was reached in slightly more than 12 h. A backpulse restored the filter flux; however, the time to the next backpulse was slightly shorter. These tests indicate that after the addition of raw water, settle and decant effectively removed solids from the feed. It is not possible to assess the time required for these solids to settle based on the result from these tests. However, if sufficient time is allowed for the solids to settle, no measurable filter fouling was observed in current testing. In contrast, if insufficient time is allowed for solids to settle, filter fouling will occur relatively quickly and will likely require backpulses more frequently than every 12 hours (1.91 m 3 /m 2 volume filtered). Solids concentrated from backpulse solutions displayed large variability in composition with Al-silicates and mixed chromium-aluminum oxide phases. Amounts of Mg, Fe, and Ca were also found in the presence of Na and O. These solids could be grouped in three categories: (1) soluble Na-salts that formed large block-like particles but were made up of fibrous materials, (2) amorphous Al-phases, and (3) colloidal iron oxides with an average particle diameter of 4.2 µm.

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

MINOS Infrasound Analysis Synopsis

This report was written as a guide to working with infrasound data collected as part of the Multi-Informatics for Nuclear Operations Scenarios (MINOS)project, an NA-22 funded venture. The main purpose of overall MINOS project is the combination of multiple, disparate data modalities to characterize the operations at a nuclear facility, specifically instrumenting and studying the High-Flux Isotope Reactor (HFIR) and Radiochemical Engineering Development Center (REDC) locate at Oak Ridge National Laboratory in Oak Ridge, TN. HFIR is an 85 MW research reactor and is used primarily for production of medical radioisotopes, material irradiation experiments, neutron activation, and neutron scattering. Targets for the reactor are constructed, processed, and dissolved at REDC. REDC also hosts other glove-box and hot-cell type activities.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

ORNL Post-irradiation Examination (LAMDA) Final Report on Characterization of Irradiated Grade 92 Ferritic-Martensitic Steels

Irradiation effects on ferritic-martensitic steel Grade 92 were investigated in this project. Samples from two heats of Grade 92, i.e., G92-2b and NF616, were selected, which were irradiated in the High Flux Isotope Reactor (HFIR) and the Advanced Test Reactor (ATR), respectively. Samples from two heats of ferriticmartensitic steel T91, irradiated in the ATR, were also selected as references for the Grade 92 samples. The selected G92-2b samples were irradiated to 0.5–14.7 displacements per atom (dpa) at two temperature regimes: 400–497 °C and 683–720°C. The selection of the high irradiation temperatures up to ~720°C is to test the performance at such upper bound temperatures, which would provide insights for the upper bound applicable temperature for Grade 92. The selected NF616 and T91 samples were irradiated to 3–8.2 dpa at 292–448°C. Except for T91 heat2 samples in 3-mm diameter discs, all the other selected samples are of type SS-J2 miniature tensile specimens. Two Nuclear Science User Facilities (NSUF), i.e., the Irradiated Materials Examination and Testing (IMET) hot cell facility and the Low Activation Materials Design and Analysis (LAMDA) facility, were used in the Post-Irradiation Examination (PIE) activities for the selected samples, which include: Tensile tests at room temperature (IMET); Fractography of the tensile fracture surfaces (LAMDA: scanning electron microscopy and energy; dispersive spectroscopy); Vickers hardness measurements (LAMDA); and Microstructural characterization (LAMDA: electron backscatter diffraction, focused-ion beam, and transmission electron microscopy).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Demonstration of Integrated PostIrradiation Process Steps forMultikilogram Amounts of Low- and High-Burnup Used Nuclear Fuels

sponsored research and development (R&D) on the advanced, closed nuclear fuel cycle, including post-irradiation processing of used fuels, recovery of separated components, and preparation of recycle nuclear fuels. During the period 2007–2010, kilogram-scale demonstrations of the integrated processing steps were performed for actual light water reactor (LWR) used nuclear fuels (UNF). These integrated processing demonstrations were called “the Coupled-End-to-End” (CETE) campaigns and were associated with the Global Nuclear Energy Partnership program, which ran concurrently. The demonstrations were carried out in the shielded hot cells and associated gloveboxes located in the Irradiated Fuels Examination Laboratory (IFEL) and the Radiochemical Engineering Development Center (REDC) at the Oak Ridge National Laboratory (ORNL). Limited circulation program reports and a few technical papers were issued. The subject document provides a comprehensive summary of results, technical comparisons, and conclusions obtained from the demonstration campaigns and subsequent associated R&D.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Northstar Accelerator Based Mo99 Production Facility Design Support

Facility design support in FY20 centered on transfer of the target insertion and design details to MBraun and Springs, the companies under contract via Northstar to do the hot cell, target insertion and the local target shielding. A significant number of changes have been introduced by these vendors which have been reviewed and critiqued. In addition, the company contracted to produce the Mo100 disks announced that the tolerance on the thickness of 0.5 mm would be ±25%. An alternative design of the target holder was proposed. Bulk shielding and local target shielding design work has continued also and will be reported. Ongoing design and cooling issues continue to arise as the facility takes shape and moves toward completion. LANL is part of the team identifying, analyzing and designing solutions to these issues. Many of these are covered in this report.

43 PARTICLE ACCELERATORS↗
DOE OSTI2021

Monitoring Radiochemical Processing Streams for the 238 Pu Supply Program with Process Pulse II

Oak Ridge National Laboratory (ORNL) is developing advanced spectroscopic and real-time monitoring capabilities to improve the timeliness of analytical measurements and process decisions for the 238Pu Supply Program. Reducing the time, resources, and costs associated with each production campaign is critical because overlapping campaigns will be required to meet the production goals of the National Aeronautics and Space Administration. Real-time, in situ analytical measurements in the heavily shielded hot cells at the Radiochemical Engineering Development Center (REDC) will allow for rapid process information feedback and operational benefits that help the 238Pu supply program scale-up production efforts. Noteworthy steps were taken during Campaign 5 to establish the ability to monitor processing streams in real time with spectrophotometry and a commercially available online monitoring software called The Unscrambler X Process Pulse II (PP) multivariate statistical process monitoring system by Camo Analytics (version 5.60). PP automates univariate-type calculations within the software itself and executes multivariate models built using The Unscrambler X (version 10.4 or newer). The Unscrambler is a commercially available data analysis software made by the same company. PP is composed of easy-to-use-tools for all personnel, including data scientists and technicians. The software can be used to plot analyte concentration profiles, spectral data, and other process variables in real time. All process data are represented in a single view with interactive charts useful for viewing how a process evolves over time.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Out-of-Pile Furnace Tests on Fast Reactor Metallic Fuels Conducted at the AGHCF

An extensive out-of-pile furnace test program was conducted at Argonne’s Alpha-Gamma Hot Cell Facility (AGHCF) from 1987-1994 to evaluate the fuel/clad compatibility and performance of fast reactor metallic fuels. This test program included over 150 tests on irradiated fuels conducted in two furnace apparatuses. The available records of these tests have been preserved with the support of the Advanced Reactor Technology program and organized in the OPTD (Out-of-Pile Transient Database). This report provides at-a-glance summary information for each of the out-of-pile furnace tests, including information about the tested fuel samples, test conditions, purpose of the tests, and key results. It is intended for open and unlimited distribution to allow all interested persons to view key information about the out-of-pile tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Tritiated Sulfur Hexafluoride Disposition Strategies

Tritium (T 2 ) is a radioactive isotope of hydrogen that is produced in nuclear fission reactions and is often used in nuclear fusion reactions and accelerator-based applications for medical isotope production. As a hydrogen isotope, tritium can readily bind to hydroxyl radicals (OH), forming tritiated water (HTO or T 2 0), and to carbon atoms. Tritium decays to helium-3 ( 3 He) via beta-decay with max decay energy of 18.6 keV. While it is not an external radiation hazard, it can be an internal radiation hazard if tritium is inhaled, ingested, or absorbed through the skin. In applications where tritium is handled, tritium confinement is performed using different barriers to minimize releases to the environment. For gaseous (elemental), liquid (oxide), and metal (hydride) tritium, process piping and components provide the primary confinement function. Secondary tritium confinement is typically provided by inert (i.e. non-flammable gases such as nitrogen, argon, or helium) gloveboxes which are connected to a tritium stripper system. Primary tritium confinement barriers typically produce low volumes of high tritium concentrations of tritiated methane from carbon in steels or organic materials along with tritiated water/tritium oxide (e.g. HTO) and tritiated ammonia from reactions with oxygen and nitrogen. Tritium escaping primary confinement into secondary confinement atmospheres (e.g. gloveboxes) produce higher volumes of lower activity contamination than found in process piping. Tritium contamination also occurs by leaks or tritium permeation/diffusion through confinement materials. Tritium from inside primary confinement barriers will diffuse or leak out of the primary confinement barrier and usually into the air, if the system is inside an air hood or ventilated hot cell, or into the secondary confinement (e.g. glovebox) atmosphere which is either exhausted or stripped based on the function of the secondary confinement (glovebox) system. Accelerator based processes for medical isotope production represent an atypical tritium contamination challenge. In medical isotope production, deuterium supply gas is ionized and accelerated to a tritium gas target to produce neutrons that are then used to produce the medical isotopes through additional nuclear fission reactions. To create large voltage differentials for accelerator operations, an electrical insulation medium is needed to prevent or rapidly quench electric discharges. A common electrical insulation medium utilized in accelerator applications is sulfur hexafluoride (SF 6 ) gas. SF 6 has a high dielectric strength and allows for the construction of smaller accelerator systems compared to other electrical insulation mediums such as air or dry nitrogen. Due to tritium permeation/diffusion through accelerator process and confinement materials, there is the possibility that tritium can contaminate the electrical insulation medium of the accelerator. Tritium contaminated SF 6 creates a material without any obvious processes for managing the contamination, reuse, or disposal of the used SF 6 . This document will discuss possible management strategies for tritium contaminated SF 6 for accelerator-based processes for Molybdenum-99 (Mo-99) production.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Radiological Handling and Containment Considerations in Support of an American Medical Isotope Producer

This work will discuss evaluation and design of new and modified facility level radiological containment, transfer, and handling systems in support of the United States’ American Medical Isotope Production Act. In the United States alone, Molybdenum-99 (Mo-99) is used as a precursor to locally generate its decay product, technetium-99m (Tc-99m), for use in approximately 40,000 medical procedures every day to detect cancer and heart disease. Currently, however, Mo-99 is primarily produced overseas and typically requires highly enriched uranium (HEU) which is classified as weapon -usable. This poses a serious security risk to the United States and, in 2012, Congress passed the American Medical Isotope ACT which aimed to cease all Mo-99 production that requires HEU and support American based companies who have found safer production methods. Through this effort, a system was designed and developed to access irradiated material in processing hot cells for ease of material introduction and removal in the confined area. The proposed solution uses a double-door design that allows for easy access to material while also providing shielding through a mechanically manipulated carousel. A system was also designed to transport irradiated materials from target reactors to a target receipt box using a rail guide system, pneumatic motor, and lead screw drive.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Physical Sciences Vistas Issue 3 2021

In this issue, highlights of our contributions to mission operations include descriptions of the following. • How exact and scrupulous planning in concert with sophisticated science enabled a first-ever direct measurement of a radionuclide with a half-life as short as six days. The work, which included operations at the Los Alamos Neutron Science Centers (LANSCE) Weapons Neutron Research Facility and the Isotope Production Facility’s hot cell facility, is a boon for both astrophysics and weapons science. • An introduction to Christie Davis and her role in ensuring the Lab’s execution of simultaneous excellence. • A look at progress by AOT’s Target and Experimental Support Team in reclaiming long-dormant space in a radiological controlled area for new work improving target systems for the Lujan Center. • A story showcasing a cross-organizational effort to minimize the directorate’s legacy and environmental footprint. Staff from across the Lab joined together to ensure the safe and efficient disposal of Rocky Flats legacy waste stored in a transportainer at the Target Fabrication Facility. • How the Lab’s Utilities and Infrastructure experts collaborated with stakeholders in our directorate to safety execute preventive maintenance on the LANSCE mesa’s large electrical substation. The Safe Conduct of Research principles provided a common framework for the planning and process. • A description of the resources called upon and the procedures undertaken by MPA to inventory the division’s time-sensitive chemicals, including the careful and conscientious response of an alert team member when something seemed amiss.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Conduct weld campaign (FY-21-1) on irradiated materials provided by the Canadian Nuclear Laboratory (CNL), including baseline post-weld evaluation and testing

A collaborative research on developing advanced welding technologies for irradiated stainless steel 304 materials between CNL and ORNL has been established to support both U.S. and Canadian interests in evaluation of weld repair techniques on irradiated materials to support continuous operation of commercial nuclear power. The work utilizes unique material from the National Research Universal (NRU) reactor and the specialized welding hot cell facility at ORNL. The objective is to explore suitable welding technique and parameters and to determine the helium concentration limitation in terms of irradiated stainless-steel weldability. An additional objective is to develop the knowledge and understanding of cracking mechanisms induced by high concentration of helium in the process of weld repair of the irradiated structural alloys. This report summarizes the experimental welding evaluation on irradiated material from the NRU reactor containing helium concentrations ranging from 12 to 45 atomic parts per million (appm).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Measuring Neptunium Concentration using Optical Spectrometry for the Pu-238 Supply Program

Targets containing 237 Np are irradiated at the High Flux Isotope Reactor at Oak Ridge National Laboratory to produce 238 Pu, as plutonium dioxide (PuO 2 ), a heat source material for use in future NASA space missions. Kilogram quantities of Np are processed yearly and developing methods to efficiently characterize and monitor the Np processing streams with optical spectroscopy and real-time monitoring software is a critical aspect of this effort. Here, we evaluate how ultraviolet/visible/near-infrared (UV-Vis-NIR) absorbance and Raman spectroscopic techniques are applicable to characterizing concentrated Np solutions and provide proof-of-principle and proof-of-concept results for each technique. Both techniques were tested in lab-scale studies and should improve the timeliness of production efforts. However, spectrophotometry is the more mature option and has been successfully demonstrated in a hot cell environment. An overview of the initial findings, a brief demonstration, and areas for future work will be highlighted in this report.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Isotope Production at the Jefferson Lab LERF Facility (Final Report)

The project initially proceeded well, meeting all of the Milestones. However, the need to create a new irradiation area in the LERF (the initial proposal was to use the CEBAF Injector) used up a larger fraction of the budget than foreseen. A further problem arose when the hot-cell became unavailable at VCU, followed by VCU dropping out of the collaboration. This meant that we were unable to separate the 67 Cu from the gallium. We replaced this by carrying out detailed gamma spectroscopy to determine the quantity of 67 Cu produced.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Commissioning of a Cask Enabling Characterization of Irradiated Nuclear Fuels with Pulsed Neutrons

Irradiation tests are a key component of nuclear fuel development and identifying typical and atypical regions in the irradiated fuel volume relies on very few characterization techniques. The goal of the effort reported here is to provide complementary measurements adding to the available parameter space for post irradiation examination as well as to inform subsequent hot cell PIE examinations by identifying typical and atypical regions with respect to microstructure, tomographic data, or isotope densities. Pulsed neutrons, enabling diffraction as well as energy-resolved neutron imaging and neutron absorption resonance spectroscopy, offer unique capabilities for this purpose. Time-of-flight neutron diffraction has the potential to offer efficient, non-destructive and non-contact microstructural characterization of irradiated fuel pins with spatial resolution of 1 mm 3 to 1 cm 3 while for energy-resolved neutron imaging (and by extension tomography) a resolution of 100 μm was demonstrated. The potential results include crystallinity vs. amorphous volumes and microstructural information such as phase compositions, lattice strains (indicative of residual stresses or chemistry variations) and textures from the diffraction data as well as distances (e.g. pellet to cladding), cracks, and isotopic distributions of minor actinides, fission products as well as fission gas partial pressures e.g. in the plenum from energy resolved neutron imaging.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Fiscal Year 2021 Filtration of Hanford Tank 241-AP-107 Supernatant Samples Obtained at Prototypic Tank Level and Filtered at 16 °C

Bench-scale filtration testing of 8.5 liters of supernatant from Hanford waste tank 241-AP-107, chilled to 16 °C, was conducted using a backpulse dead-end filter (BDEF) filtration system equipped with a feed vessel and a Mott inline filter Model 6610 (Media Grade 5) in the hot cells of the Radiochemical Processing Laboratory at Pacific Northwest National Laboratory. This was done to assess the impact of a lower sampling location within the tank as well the lowered filtration temperature on supernatant stability and fouling. The as-received samples were transferred to 1.5-liter poly bottles and held at 16 °C for approximately 1 week prior to filtration. The feed was filtered through the BDEF system at a targeted flux of 0.065 gpm/ft 2 to match the prototypic operation of the TSCR system. During the initial period of filtration, the differential pressure required to effect filtration at 0.065 gpm/ft 2 increased until it reached 2 psid [the Tank Side Cesium Removal (TSCR) action limit] at 26 hours. After this, the filter was backpulsed to dislodge accumulated solids and reduce this pressure differential. An additional eight backpulses were conducted during the initial filtration period; each time, the target pressure was reached sooner than during the previous interval. Volume filtered decreased from 0.9 to 0.03 m 3 /m 2 over the course of 14 processing hours. After the ninth backpulse, the backpulse frequency had become unsustainable, and it was decided to perform an extensive filter cleaning. The filter was cleaned by draining the chilled AP-107 feed into chilled holding containers, introducing 0.1 M NaOH into the feed vessel, and recirculating the NaOH through the system for 20 minutes before allowing the system to soak for 2 hours without temperature control. A measurable decrease in filter resistance during this recirculation indicated that the 0.1 M NaOH was likely dissolving some of the solids that had deposited on the filter. Post cleaning, the filter resistance was effectively restored to initial conditions as the initial transmembrane pressure was restored to original levels. However, resumed processing of the AP-107 feed at 16 °C continued to result in an increased rate of filter resistance. An additional five backpulses were conducted (four during feed processing, one during subsequent cleaning) before the conclusion of the test. Solids concentrated from the backpulse solutions displayed sodium oxalate-type phases, Al-oxides and sodium Al-oxides/carbonates, iron oxides, and Ca-bearing phases (calcite). The scanning electron microscopy analyses also revealed a large distribution of particles, with some particles, notably sodium phosphate dodecahydrate, having diameters close to 0.5 mm.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Upgrade of Gamma Spectrometry Systems for ORNL TRISO Fuel PIE

Gamma spectrometry is a key element in much of the post-irradiation examination (PIE) work performed under the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program (Demkowicz et al. 2015; Stempien et al. 2021). Gamma spectrometers are integrated into three major capabilities used at the Oak Ridge National Laboratory (ORNL) Irradiated Fuels Examination Laboratory (IFEL) for PIE of tristructural-isotropic (TRISO) coated particles and fuel compacts: the Core Conduction Cooldown Test Facility (CCCTF), the Vertical Counting System (VCS), and the Irradiated Microsphere Gamma Analyzer (IMGA). The CCCTF includes liquid-nitrogen-cooled traps to extract 85 Kr out of the He sweep gas that passes through the furnace in which the fuel compacts are heated during safety testing. Analysis of the 85 Kr activity in the traps is the primary indicator for TRISO failure during safety testing. The VCS is a system used to accurately measure gamma emission from components placed in a lead-shielded chamber. It is used to count the CCCTF deposition cups after removal from furnace. Each cup resides in the CCCTF furnace for typically 12–24 h and is periodically replaced with a fresh cup throughout the safety test. Metallic fission products collect on the water-cooled cups and several gamma-emitting isotopes ( 110 mAg, 134 Cs, 137 Cs, 154 Eu, and 155 Eu) are often measured and provide indication of the retention performance of the TRISO coatings. The VCS is also used to measure the presence of these isotopes on the CCCTF tantalum liner and sweep gas inlet tube for the determination of cup collection efficiency, as well as support other gamma spectrometry needs related to calibration of the 85 Kr fission gas traps and various other special PIE tasks. The IMGA uses gamma spectrometry to measure the inventory of gamma-emitting isotopes in individual TRISO particles. An automated particle handling system within the IMGA hot cell removes each particle from a source vial and positions it in front of a gamma detector, and output from the gamma spectrometer is used by the IMGA software to determine a destination vial such that particles are sorted according to their inventory and retention characteristics. At the conclusion of the AGR-1 and AGR-2 PIE campaigns, the gamma spectrometer systems used at ORNL to support that PIE had reached the end of its life cycle due to gradual obsolescence of the hardware and software. Upgrade of the Canberra Genie 2000 software used by these systems to a Windows 10 version was not a viable option, because the newest Windows 10 version offered by Mirion (the new owner of the Canberra technology) did not include the dynamic-link libraries (DLLs) needed for integration with the custom PIE software used with the CCCTF and IMGA, and Mirion had no current plans for development and release of Windows 10 versions of these DLLs with the Model S560 Genie 2000 Programming Library. Ultimately a switch was made to ORTEC gamma spectrometry systems, which appeared to be a more sustainable solution due to more proactive vendor support. The ORTEC conversion involved replacing the aging detector preamplifier and multichannel analyzer (MCA) hardware, upgrading the obsolete Windows 7 computers to Windows 10 compatible models, adopting ORTEC GammaVision software, and extensive modification of the ORNL-developed Visual Basic .NET (VB.NET) programs that provide the CCCTF and IMGA user interfaces.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗