Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Nuclear Fusion”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Laser Powder Bed Fusion Additive Manufacture Nb1Zr Development

Next generation fission and fusion nuclear reactors require materials that can withstand operating temperatures greater than 500 °C, neutron irradiation doses of up to 200 displacements per atom (dpa), and potentially corrosive coolants such as the alkali liquid metals sodium, lithium, and NaK (Na33K eutectic alloy). Refractory alloys, such as Nb1Zr (Nb-1wt%Zr) and Molybdenum alloy TZM (Mo-0.5wt%Ti-0.08wt%Zr) have been traditionally considered viable candidates for advanced fission and fusion reactor concepts. However, it is relatively difficult to generate complex geometries of interest from these alloys using traditional manufacturing methods. In addition, there needs to be a concentrated effort to address refractory metal challenges at elevated temperature operation. In order to generate complex geometries of interest, modern manufacturing techniques are considered to increase the technological readiness level (TRL), cost-effectiveness, and schedule savings. This work focused on the continued development of laser powder bed fusion (L-PBF) additive manufacturing (AM) to improve both design flexibility, evaluate microstructure and properties, and ultimately accelerate the TRL and qualification of these processes and alloys for components to potentially be put into service. Niobium alloy Nb1Zr was identified through a down-selection process outlined in previous reports as a candidate to develop in L-PBF AM. Historically, Nb1Zr had been explored for high temperature fast spectrum fission reactors for both terrestrial and space applications. Molybdenum alloy TZM has also been considered for these reactor concepts due to exceptional high-temperature strength, creep resistance, and stability under irradiation. L-PBF AM of TZM has previously been investigated at LANL under the Microreactor program, NASA, ORNL, and in academia. However, due to the crack prone nature of TZM, L-PBF AM of TZM resulted in significant microcracking and additional development is required to pursue viable maturation. Other AM methods have been found to be more successful in printing TZM, and those alternatives approaches are discussed in this effort. The efforts detailed in this report focused on continued development of Nb1Zr through L-PBF and development of TZM via L-PBF and electron powder bed fusion (E-PBF). The objective of this work was to further the development of these AM techniques for the chosen refractory alloys, elucidating and addressing associated challenges through characterization of several demonstration builds. At LANL, Nb1Zr builds were completed using an EOS M290 and M400 machines, and a refractory alloy-dedicated L-PBF system, the Xact Metal XM200G, was installed. The XM200G primary purpose was to do the Nb1Zr parameter development process; however, due to difficulties associated with the machine installation and qualification process, it was decided to pivot development to the larger M400 and M290 machines. Although the supply of Nb1Zr powder was limited, it was sufficient to generate sub-scale metallographic specimens for the purpose of parameter development. This was first accomplished on the EOS M400 then the M290 due to machine schedule availability. Further development of TZM has been initiated at the University of Texas El Paso (UTEP) under contract with LANL to use both a heated build envelope L-PBF machine and E-PBF machine that have been found in the literature to mitigate microcracking. UTEP was provided with TZM powder and build plates to support parallel TZM parameter development across both machines. As part of the contract, UTEP will also be conducting microstructural characterization once optimized process parameters have been identified. The optimized process parameters for each machine will be used to generate a series of metallographic, mechanical, and surface finish specimens for subsequent characterization and testing. In the next section, we provide a detailed discussion of the methodology used for investigating the feasibility of leveraging these alloys for use in advanced reactor applications.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Image fusion for neutron tomography of nuclear fuel

Image fusion, the process of combining different images together, can be useful to create a more complete picture. In this work, image fusion is applied to neutron tomography of nuclear fuel with the goal of enhancing the information obtained about the fuel. Different reconstruction methods, such as Feldkamp, Davis and Kress filtered back projection and Simultaneous Reconstruction Technique, were combined to enhance image quality. This methodology was shown to reduce noise and ring artifacts without sacrificing sharp edges, allowing for a more accurate representation of sample geometry. Technique enhancements and future applications for the neutron imaging community are also discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

The Maintainable Fusion Pilot Plant

The US fusion community has coalesced around the goal of building an FPP as described by the National Academies of Science, Engineering, and Medicine (NASEM). In addition to demonstrating the viability of the technologies necessary to operate such a plant, including demonstration of net energy and electricity production, NASEM found that a “fusion pilot plant will need to demonstrate the ability to efficiently perform remote maintenance and replacement in support of the design of a power plant, taking into account details of the consequences of the fusion environment, such as material activation and tritium retention in components.” Current designs of fusion demonstration reactors do usually foresee a regular exchange of their first wall modules, including the tritium breeding blankets. In the European Power Plant Conceptual Studies, it is assumed that a fusion reactor will need to change its divertor every 2 years and its first wall blanket module every 5 to 6 years to reach acceptable availability. Underlying this capability are remote-handling technologies to keep the outage for the exchange of these components short. There are many uncertainties in the remote-handling schemes, and most schemes are at a preconceptual level at best. In addition, the exchange of these components would either produce an enormous rad-waste stream or would require an enormous refurbishment activity with huge cost-prohibitive hot-cells. Past Fusion Nuclear Science Facility (FNSF) preconceptual studies have led to hot cell dimensions of an unbelievable size, likely costing tens of billions of dollars. Already at The Way (previously International Thermonuclear Experimental Reactor, ITER), hot-cells have become cost-prohibitive, demanding redesigns of the ITER first wall to reduce the toxic rad-waste/inventory. In this in-situ PFC repair project, a concept for a long-life, maintainable first wall module concept is developed and tested. This first wall concept relies on innovative remote handling to repair the first wall modules in-situ, avoiding costly refurbishments outside of the tokamak vessel. This approach was highlighted in the Fusion Energy Sciences Advisory Committee (FESAC) report on Transformative Enabling Capabilities for Efficient Advance Toward Fusion Energy. In general, the damage of the first wall armor is due to particle and radiation exposures. Load conditions vary from one fusion reactor design to another. In tokamaks, first wall Plasma Facing Components (PFCs) are exposed to far-Scrape-Off-Layer plasma fluxes, electromagnetic radiation, energetic CX neutrals, and potentially runaway electron beams. Protecting the first wall to the worst-case load conditions would require the design of a very thick first wall armor. Transient heat and particle fluxes due to disruptions or edge localized modes can lead to excessive heat loads resulting potentially in melting PFCs down to the cooling channel. Catastrophic events like these need to be avoided by appropriate disruption mitigation systems. However, failure of these systems will still put a first wall at an unacceptable risk. Hence, a first wall design needs to accommodate the occasional transient heat loads by introducing sacrificial limiters, which will absorb these transients.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Data Evaluation of Actinide Cross Sections: 230 Th and 231 Th

A new evaluation of the ENDL cross section set for Thorium (Z=90) is developed using the TALYS statistical model cross section code. The primary goal of this effort is to produce an evaluation that attempts to match as closely as possible fission cross sections developed through surrogate reaction techniques on actinide targets 230 Th and 231 Th. This evaluation effort and the processing needed to render its results into data libraries is a necessary step in making the efforts of nuclear experimentalists useful to the broad community of researchers engaged in simulations of nuclear fusion for basic and applied science. Another aspect, verification and validation against various AGEX experiments, is also presented. The end-product is an updated library that includes the latest measurements of fission cross sections for comparison against those measured via traditional techniques. All the steps in the evaluation, processing, validation and verification, and library release are described in the following sections. For completeness, the appendix contains all the parameters used in the TALYS cross section evaluation for neutrons incident on 230 Th and 231 Th.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Neutron irradiation-enhanced grain growth in tungsten and tungsten alloys

Herein, to understand the microstructural stability of candidate plasma-facing materials under fusion-relevant environments, neutron irradiation of W and W-3%Re alloys with and without K and La dopants was performed in the mixed-spectrum High Flux Isotope Reactor at nominal temperatures of ~850 °C and ~1100 °C to calculated doses between 0.42 and 0.47 dpa. To the best of our knowledge, this study presents the first experimental evidence of radiation-enhanced recrystallization in W and undoped W–Re alloys at ~850 °C, conditions where thermal annealing does not cause any grain growth in a similar timescale. Potassium- or lanthanum-doped tungsten alloys showed more resistance to radiation-enhanced grain growth. We explain the acceleration of grain growth by analyzing the self-diffusion constant under atomic displacement environments. The microstructural observations of the studied W variants suggest that La doping is more effective than K doping for mitigating recrystallization. This study also found that radiation-enhanced recrystallization is an important consideration when designing and applying W to plasma-facing components in future nuclear fusion reactors.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Manufacturing Li 2 TiO 3 -based tritium breeder materials by volume-controlled spark plasma sintering with an optimized microstructure

Multifunctional ceramic breeder materials are highly desirable for the deuterium-tritium fusion to achieve high efficiency in breeding tritium through neutron irradiation of lithium-containing blankets. Li 2 TiO 3 displays unique attributes as a potential ceramic breeder material. An optimized microstructure with three-dimensional interconnected pore structure is required for rapid transport of the tritium for effective fuel cycle, which however enviably results in the degradation of the thermal-mechanical properties of the breeding materials. In this work, nanocrystalline porous Li 2 TiO 3 ceramic pellets with controlled porosities of 14% and 20% are manufactured by volume-controlled spark plasma sintering. An optimized 3D interconnected pore structure is achieved consisting of both micro-sized pores and nano-sized pores embedded in nanocrystalline matrix, which could be beneficial to facilitate easy removal of bred T and He. Further, the 3D interconnected porous structure is well maintained upon isothermal annealing of the SPS-fabricated pellets at relevant operation temperature of the solid breeding materials. Single-phasic porous pellets also display enhanced thermal-mechanical properties, superior to current state-of-the-art materials which establish their potential as a promising tritium breeder material for nuclear fusion applications.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Multiphysics Modeling of Impurity Transport for FNSF Startup Scenario with ERO2.0

This paper focuses on performing a multiphysics study using the ERO2.0 and UEDGE codes for two standard double null configurations for the Fusion Nuclear Science Facility: (a) 100% recycling and (b) 99% recycling. Results show that the main contributor to tungsten erosion along the divertor plates is impurities from the midplane waveguides. In addition, the standard high-recycling case (100% recycling) shows a significantly higher buildup of impurities along the divertor tiles during the startup phase, which can lead to a higher increase of energy loss in the plasma during steady-state operation. Last, for high recycling, anomalous diffusion can dominate over parallel field diffusion. The work performed in this study can be iteratively applied to a full operation scenario with additional physics such as those from neutrals, wall shaping, and additional external fields.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Suppression of irradiation hardening in tungsten-coated ferritic steel for fusion reactor blanket applications

W-coated reduced activation ferritic steels have been developed for use as plasma facing components in fusion reactor blankets, offering excellent sputtering resistance and structural strength. Previous high-temperature coating methods, such as diffusion bonding and brazing, caused interfacial deterioration due to thermal stress from mismatched thermal expansion between W and reduced activation ferritic steel. To address this, underwater explosive welding was introduced as a high-velocity cold process that joins dissimilar materials while maintaining a strong, thin interface without the thermal issues associated with traditional methods. In this study, the effects of neutron irradiation on the hardness and microstructure in W-coated F82H reduced activation ferritic steel (W/F82H) joined by underwater explosive welding are investigated. Following neutron irradiation at 290 °C, irradiation hardening is suppressed in W, F82H, and their interface within the W/F82H material. Furthermore, microstructural observations indicate that the recovery of work hardening and relaxation of elastic strain introduced during coating significantly contribute to the suppression of irradiation hardening in W/F82H. In conclusion, W/F82H exhibits significantly suppressed irradiation hardening compared with those in stand-alone materials. This suppression is explained by residual stress from thermal expansion mismatch and the unique microstructure at the interface. These results provide valuable insights for the development of more durable materials in nuclear fusion applications.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Liquid metal walls

Here, the plasma performance of fusion devices depends strongly on the chosen wall materials. Solid plasma-facing components (PFCs) are predominantly used in present devices, and are the most investigated candidates for fusion designs. High-Z materials such as tungsten (W) are the leading solid PFC material candidates. To date, a material choice that scales to steady-state reactor conditions has not been identified. Moreover, if plasma transient events such as edge-localized modes and disruptions cannot be altogether avoided or sufficiently mitigated, the projected peak heat and particle loads far exceed the power exhaust capabilities of solids. Liquid metal (LM) PFCs represent an intrinsic self-healing boundary that are both resilient to surface damage from transients, and that could handle high steady-state heat and particle fluxes. Flowing LM PFCs can be designed to remove “slag,” the buildup of material erosion due to plasma-material interactions, including charge exchange sputtering in the main chamber. Further, LM offer the prospect to manage hydrogenic species otherwise retained in the PFCs, which is important from a safety and inventory standpoint. The two most promising LMs are lithium (Li) and tin (Sn), although Sn–Li eutectics may be considered. While Sn offers a higher temperature window with low vapor pressure and low hydrogen retention, Li offers the prospect of enhanced energy confinement and higher acceptable core contamination limits, and this section focuses on Li PFCs. An LM PFC development research program developed LM PFC concepts for a nuclear fusion device via engineering design calculations, single-effect experiments, and staged prototypical experiments. A self-consistent design window was identified with liquid Li flow speeds ~5–10 m/s; plasma contamination was negligible for predicted Li evolution rates. While these preconceptual designs hold promise, there is substantial R&D needed to advance the technical readiness levels of LM PFCs for application to fusion power plants.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Effects of transient thermal shock on the microstructures and corrosion properties of a reduced activation high-entropy alloy

The devices would subject transient thermal shocks (TTS) during operation under extremely harsh conditions of nuclear fusion reactors, which inevitably exert significant impact on the microstructure and performance of structural materials. Here, a reduced activation VCrFeTa 0.2 W 0.2 high-entropy alloy (HEA) was developed by vacuum arc melting. The effects of electron beam induced TTS on its microstructure, microhardness, and corrosion properties were investigated. The results indicate that the weight fraction of each phase changes after TTS, showing a significant decrease in the content of BCC1 phase and an increase in the content of BCC2 and Laves phases. The content of BCC1 phase continues to decrease slightly with increasing the power of TTS. Besides, the microhardness of the alloy increases from ~673 HV to ~714 HV after TTS treatments. In the reduced activation HEA, TTS results in a relatively low corrosion current density of ~0.472 μA/cm 2 in 3.5 wt% NaCl solution, around one-third of the current density observed in the as-cast sample. Furthermore, the VCrFeTa 0.2 W 0.2 HEA after TTS exhibits a pitting potential of ~1.165 V SCE , which is much higher than that of the as-cast sample. The refined composite multiscale entropy method is employed to analyze the influence of TTS on current fluctuation behavior during the corrosion process. The reduced activation VCrFeTa 0.2 W 0.2 HEA exhibits excellent properties in harsh environments after TTS, thereby showing advantageous property in the field of nuclear structural materials. Furthermore, TTS is an efficient and controllable strategy for the improvement of the HEAs’ microstructures and performances.

36 MATERIALS SCIENCE↗
DOE OSTI2022

How Bayesian methods can improve R -matrix analyses of data: The example of the d t reaction

The 3 H(d, n) 4 He reaction is of significant interest in nuclear astrophysics and nuclear applications. It is an important, early step in big-bang nucleosynthesis and a key process in nuclear fusion reactors. We use one- and two-level R-matrix approximations to analyze data on the cross section for this reaction at center-of-mass energies below 215 keV. We critically examine the data sets using a Bayesian statistical model that allows for both common-mode and additional point-to-point un- certainties. We use Markov Chain Monte Carlo sampling to evaluate this R-matrix-plus-statistical model and find two-level R-matrix results that are stable with respect to variations in the channel radii. The S factor at 40 keV evaluates to 25.36(19) MeV b (68% credibility interval). We discuss our Bayesian analysis in detail and provide guidance for future applications of Bayesian methods to R-matrix analyses. We also discuss possible paths to further reduction of the S-factor uncertainty.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

A Domestic Program for Liquid Metal PFC Research in Fusion

While high-Z solid plasma-facing components (PFCs) are the leading candidates for reactors, it is unclear that they can survive the intense plasma material interaction (PMI). Liquid metals (LM) PFCs offer potential solutions since they are not susceptible to the same type of damage, and can be “self-healing”. Following the Fusion Energy System Study on Liquid Metal Plasma Facing Components study that recently was completed by Kessel et al. (Fusion Sci Technnol 75:886, 2019) a domestic LM PFC design program has been initiated to develop reactor-relevant LM PFC concepts. This program seeks to evaluate LM PFC concepts for a Fusion Nuclear Science Facility (FNSF) or a Compact Pilot Plant via engineering design calculations, modeling of PMI and PFC components and laboratory experiments. The latter involves experiments in dedicated test stands and confinement devices and seeks to identify and answer open questions in LM PFC design. The new national LM PFC program is first investigating lithium as the plasma facing material for a flowing divertor PFC concept. Several flow speeds will be evaluated, ranging from ~ cm/s to m/s. The surface temperature will initially be held below the strongly evaporative limit in the first design; higher temperatures with strong evaporation will be considered in future concepts. Other topics of interest include: understanding of the hydrogen and helium interaction with the liquid lithium; single effect experiments on wetting, compatibility and embrittlement; and prototypical experiments for control and characterization of flowing LM. A path to plasma and future tokamak exposure of these concepts will be developed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

An atom probe tomography and inventory calculation examination of second phase precipitates in neutron irradiated single crystal tungsten

Tungsten is the prime candidate material for the plasma facing divertor system and first wall armor of future nuclear fusion reactors. However, the understanding of the microstructural and chemical evolution of pure tungsten under neutron irradiation is relatively unknown, in part due to a lack of experimental data on this topic. Here, single crystal tungsten has been irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory at a nominal temperature of 700–800 °C to damage levels of 0.1 and 1.8 displacements per atom (dpa). Inventory calculations of the neutron irradiation experiments have been used to track isotope generation and decay to inform atom probe tomography (APT) results in the determination of the transmutation-induced precipitate compositions. Furthermore, APT crystallography has been used to show the relationship between precipitates and matrix. The composition of the precipitates is shown to progress towards that of the σ-phase at the highest dose studied, with those precipitates lying along crystallographic planes similar to those of displacement-induced dislocations. This work also sets the framework for APT studies of materials that contain isotopic ratios far from those observed in the natural state.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Radiation effects on stress evolution and dimensional stability of large fusion energy structures

Here we assess the effects of neutron irradiation on the deformation and stress evolution of large-scale fusion energy structures. This is accomplished through non-linear finite element structural analysis of the coupled thermal and mechanical fields at the Beginning-Of-Life (BOL), at 45 dpa, and at 90 dpa. Radiation effects include volumetric swelling and the influence of radiation on the mechanical properties. The system studied here is a large section of a full inboard module of an integrated structure comprising the First Wall and Blanket (FW/B) of a Dual Cooled Lithium-Lead (DCLL) energy conversion unit in the Fusion Nuclear Science Facility (FNSF). The structural material is the ferritic/martensitic steel F82H. We analyze several radiation effects phenomena that can lead to significant impact on the mechanical design and lifetime of the structure. These include volumetric swelling effects on the assembly and disassembly of modules and the impact of spatial gradients in neutron damage and helium generation on deformation, stress evolution, and plastic strain accumulation. We show that the stress state is strongly influenced by the accumulation of swelling strains, radiation hardening and softening, and by spatial gradients in displacement damage and helium gas generation. Several key regions in the FW/B structure are identified where severe plastic strains accumulate and may be potential sites for failure, especially during cyclic reactor operations. We finally show that minimum gaps between Inboard (IB) blanket sectors must be present during assembly so as to avoid excessive stress and strain in FW/B modules critical regions and raising chances of failure. A range of structural displacements are shown, from as low as 4 mm at BOL for free side walls to as high as 46 mm at 90 dpa for fully constrained side walls.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Assessment of thermal and radiation induced creep in the dual cooled lead lithium blanket

The creep in recently designed blanket for the fusion nuclear science facility is described by decomposing it into thermal, irradiation and cavity swelling creep. Initial estimates of thermal creep using elastic high temperature rules showed that the recent heat transfer improvements at the first wall (FW) is feasible, and the FW can withstand heat flux of 0.35 MW/m 2 at 3000 hrs without rupture. A viscoplastic model that is based on Norton's power law and relies on Multiphysics coupling of solid mechanics and heat transfer modules is used to capture the inelastic deformation in the blanket. Results showed that the design peak heat flux of 0.25 MW/m 2 produced maximum thermal creep strain of 0.45 % and the relaxation of the thermal stress at the FW. Irradiation creep is prescribed to be proportional to the displacement damage dose and applied stress. The displacements from irradiation creep radially decrease from the FW to the back wall, but the maximum deformation is found at the back wall that is connected to the Helium manifold due to the high stresses at the region. In conclusion, cavity swelling creep is dominant at steady state and combined with radiation creep to produce displacement of about 8 mm at the FW.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Morphological and nanomechanical changes in tungsten in high heat flux conditions

Abstract Morphological and nanomechanical alteration of tungsten in extreme environments, like those in edge localized modes in nuclear fusion environments, up to 46.3 GWm −2 heat fluxes were experimentally simulated using electrothermal plasma. Surface and subsurface damage to the tungsten is seen mainly in the form of pore formation, cracks, and resolidified melt instabilities. Mirco voids, rosette-type microfeatures, core-shell structure, particle enrichment, and submicron channels all manifest in the damaged subsurface. The formation of voids in the subsurface was determined to originate from the ductile fracture of hot tungsten by plastic flow but not developed to cracking. The voids were preferentially settled in grain boundaries, interfaces. The directionality of elongated voids and grains is biased to the heat flow vector or plasma pathway, which is the likely consequence of the thermally driven grain growth and sliding in the high-temperature conditions. The presence of a border between the transient layer and heat-affected zone is observed and attributed to plasma shock and thermal spallation of fractural tungsten at high temperature. Plasma peening-like hardening effects in tungsten were observed in the range of 22.7–46.3 GWm −2 but least in the case of the lowest heat flux, 12.5 GWm −2 .

36 MATERIALS SCIENCE↗
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↗