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DOE OSTI2021

Flowsheet for Conversion of Anion Resin from Nitrate to Sulfate Form for Disposal

Nitrate form anion exchange resins are specifically prohibited for disposal in TRU waste. At some DOE sites, spent resin was mixed with cement to immobilize the resin in a monolith with varying success. At SRS starting in the 1980s, spent anion resin has generally been converted to a more chemically stable form or it was digested with permanganate and transferred as a slurry to the HLW tank farm. In the early years of the site it is reported that over 20 metric tons of organic resin were transferred directly to the tank farm without any attempt to digest it. It appears that in the future conversion to a stable sulfate form could well be the favored disposal path for spent nitrate-form anion resin. This report will document the site history of converting between ionic forms (nitrate to sulfate) and will document process flowsheet conditions for performing that conversion.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Study on the heat removal for fuel debris using heat pipes

Simple method to retrieve fuel debris by meta-stabilizing: Heat removal system is needed. Heat removal using heat pipes: Cover fuel debris with geopolymer. What is a heat pipe: Heat transport medium with cycles of evaporation and condensation. It consists of the container, working fluid, and the wick. Advantage: Internal vacuum structure, small amount of working fluid, Massive heat transfer due to high-speed movement of hydraulic fluid. Cycle of hydraulic fluid with capillary structure. Because the cycle works independently of gravity, we can set heat pipes at free angle. We evaluated the availability of heat pipes within PCV. Investigation of hydrogen and oxygen concentration of internal gas collected by water replacement method: Non-condensable gas was confirmed inside the heat pipe irradiated by gamma rays. The heat removal effect by the heat pipe greatly descends to the extent close to the copper pipe by Gamma ray irradiation. The above reason was presumed to be that non-condensable gas was generated by radiolysis and the active length of the condensation part of the heat pipe decreased. Counter measures to decrease the amount of non-condensable gas is necessary for using heat pipes under radiation environment. We are doing the experiments to check changes in non-condensable gas generation by using catalysts or other kinds of liquids under Gamma ray irradiation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

X-ray tomography of feed-to-glass transition of simulated borosilicate waste glasses

The feed composition of a high level nuclear waste (HLW) glass melter affects the overall melting rate by influencing the chemical, thermophysical, and morphological properties of a relatively insulating cold cap layer over the molten phase where the primary feed vitrification reactions occur. Data from X ray computed tomography imaging of melting pellets comprised of a simulated high-aluminum HLW feed heated at a rate of 10°C/min reveal the distribution and morphology of bubbles, collectively known as primary foam, within this layer for various SiO 2 /(Li 2 CO 3 +H 3 BO 3 +Na 2 CO 3 ) mass fractions at temperatures between 600°C and 1040°C. To track melting dynamics, cross-sections obtained through the central profile of the pellet were digitally segmented into primary foam and a condensed phase. Pellet dimensions were extracted using Photoshop CS6 tools while the DREAM.3D software package was used to calculate pellet profile area, average and maximum bubble areas, and two-dimensional void fraction. The measured linear increase in the pellet area expansion rates – and therefore the increase in batch gas evolution rates – with SiO 2 /(Li 2 CO 3 +H 3 BO 3 +Na 2 CO 3 ) mass fraction despite an exponential increase in viscosity of the final waste glass at 1050°C and a lower total amount of gas-evolving species suggest that the retention of primary foam with large average bubble size at higher temperatures results from faster reaction kinetics rather than increased viscosity. However, viscosity does affect the initial foam collapse temperature by supporting the growth of larger bubbles. Because the maximum bubble size is limited by the pellet dimensions, larger scale studies are needed to understand primary foam morphology at high temperatures. This temperature-dependent morphological data can be used in future investigations to synthetically generate cold cap structures for use in models of heat transfer within a HLW glass melter.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Integrated Waste Treatment Unit Plan for Vessel and Canister Decontamination - 19283

The Integrated Waste Treatment Unit (IWTU) was designed to use a proprietary fluidized-bed steam reforming process to treat 3,400,000 L (898,184 gal) of radioactive sodium bearing waste (SBW). IWTU is currently in commissioning and startup phase prior to future processing of SBW. The current simulant SBW is introduced into the denitration mineralization reformer and transformed into a dry, granular product that is filtered and eventually transferred into a product canister. The intention of the facility is to process the SBW and store the product canisters in vaults to await shipping to a depository at a later date. The IWTU is a part of the Idaho Cleanup Project Core, located at the INL Site and operated by Fluor Idaho, LLC. During early operation and testing of the facility, it was observed that the original process equipment decontamination system did not possess the capability to recover from plant shutdowns that result from upset conditions or maintenance outages; thus, a need for a redesigned decontamination system was identified. This paper discusses current decontamination methods and their capabilities and areas where these methods fall short of decontaminating process equipment and product canisters. This paper discusses the design and purpose of the decontamination methods being tested and developed for implementation into the IWTU. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

System efficiency and power assessment of the all-aqueous copper thermally regenerative ammonia battery

Thermally regenerative ammonia batteries (TRABs) use low-temperature (T < 100 °C) heat to provide stationary energy and power with higher power densities and efficiency relative to other waste heat devices. TRABs are an active area of research in waste heat devices, but currently there is little consensus on what aspect of the system is limiting TRAB performance and what maximum efficiencies are possible. Experiments and numerical models were used here to examine the sensitivity of the battery and distillation column in a TRAB system to key operating variables, thereby establishing practical limits and identifying focus areas for improving performance. Battery power was eight times more sensitive to ohmic losses than kinetic and mass transfer losses, regardless of the operating temperature, and the peak power density was simulated to be 18.8 mW cm –2 at 75 °C. Theoretical energy efficiency limits were defined for a series of ammonia concentrations and operating pressures, ranging from 5 – 12%, which is 2–3 times higher than previous experimental estimations. Atmospheric pressure column operation used a larger amount of waste heat compared to sub-atmospheric pressure. It is estimated that the volume of the battery would take up 9.2m 3 for every 1% of the power output of a natural gas turbine, but with realistic improvements to cell conductivity, the size would reduce to 2.5 m 3 . Finally, the results presented in this work will help streamline future development by focusing on minimizing ohmic losses and provide specific data for full system evaluation of future TRABs.

25 ENERGY STORAGE↗

DOE OSTI2017

Reviewing Spent Nuclear Fuel Dry Cask Storage Programs to Identify Challenges for an Integrated Waste Management System - 17535

Under the Nuclear Waste Policy Act (NWPA), the U.S. Department of Energy (DOE) has responsibility for removing commercial spent nuclear fuel (SNF) from nuclear power plant sites and disposing of the SNF in a deep geologic repository. However, the methods used by commercial nuclear power plants for onsite storage of their SNF define the starting conditions for DOE efforts to develop a waste management system that includes packaging, transportation, storage, and disposal of commercial SNF. The NWPA (as amended) also charges the U.S. Nuclear Waste Technical Review Board (Board) with evaluating the technical and scientific validity of DOE activities related to the management and disposal of SNF and high-level radioactive wastes (HLW). DOE's efforts related to developing an integrated program for storing, transporting, and disposing of SNF and HLW were the focus of a public meeting held by the Board on August 24, 2016, in Washington, D.C. To prepare for the public meeting, the Board's staff compiled information on the dry cask storage systems in service at nuclear power plant sites in order to identify issues that would need to be addressed by DOE in planning to transport SNF away from the sites. This paper is based on a briefing presented to the Board members by the Board's staff before the public meeting. Post-meeting findings and recommendations of the Nuclear Waste Technical Review Board are documented in the Board letter to the Department of Energy dated December 8, 2016. A note about terminology: in this paper, the phrase 'dry cask storage system' is generic and applies to a complete system used for dry storage of commercial SNF.Within these storage systems, commercial SNF is sealed inside a container called either a cask or a canister depending upon the specific type of storage system used. This distinction is discussed in more detail in the body of this paper. When referring to the containers holding SNF in particular systems, this paper uses the terms 'casks' or 'canisters,' as appropriate. When discussing the collective group of all containers, this paper uses the phrase 'casks and canisters'. Dry cask storage of SNF is a widespread practice throughout the U.S. commercial nuclear industry. As of August 1, 2016, more than 2300 storage casks and canisters were loaded at U.S. commercial nuclear power plant sites. This number is expected to increase by more than 200 casks and canisters annually, reaching a total of approximately 4500 by 2025. Multiple vendors have developed their own unique designs of dry cask storage and transportation systems. The majority of storage system designs (75 percent) feature SNF loaded in welded stainless steel canisters that are stored in ventilated concrete overpacks that allow for convection cooling. Some older dry cask storage system designs do not include a welded stainless steel canister; instead, SNF is loaded directly into a heavily shielded metal cask to minimize radiation exposure to workers and prevent atmospheric release of radionuclides. These types of systems are sealed by O-rings compressed under a bolted lid. After loading, dry cask storage systems are located on concrete pads at Independent Spent Fuel Storage Installations (ISFSI) at nuclear power plant sites and will remain there until the time when they will be transported to a consolidated interim storage facility or a deep geologic repository. Eventual transportation of canistered SNF will include removing the stainless steel canisters from the concrete overpacks, possibly repackaging the SNF, and loading the canisters into specially designed transportation casks for shipment by truck, rail, or barge. In the case of the non-canistered casks, some are approved for transportation and can be shipped after impact limiters are installed on the casks. SNF will need to be removed from casks not approved for transportation and transferred to transport-approved casks. This paper highlights five issues and challenges presented by the nuclear industry's use of dry cask storage systems that will need to be addressed by DOE in developing an integrated waste management system. One example, among the five, is that several dry-storage casks and canisters have received Nuclear Regulatory Commission certificates of compliance for storage only and do not have certificates of compliance for transportation. This will pose a challenge when the SNF must be moved to a consolidated interim storage facility or a deep geologic repository. Another example is that, when loaded with SNF, the total weight of new, larger transportation cask designs will exceed 142 metric tonnes (156 tons) which, when combined with the weight of the rail-car, will exceed the gross weight limit for unrestricted operation on some rail lines. This will also affect future transportation planning for shipments of SNF. All five issues and challenges are discussed in detail in the body of the paper. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2012

Closure Report for Corrective Action Unit 547: Miscellaneous Contaminated Waste Sites, Nevada National Security Site, Nevada

This Closure Report (CR) presents information supporting closure of Corrective Action Unit (CAU) 547, Miscellaneous Contaminated Waste Sites, and provides documentation supporting the completed corrective actions and confirmation that closure objectives for CAU 547 were met. This CR complies with the requirements of the Federal Facility Agreement and Consent Order (FFACO) that was agreed to by the State of Nevada; the U.S. Department of Energy (DOE), Environmental Management; the U.S. Department of Defense; and DOE, Legacy Management (FFACO, 1996 as amended). CAU 547 consists of the following three Corrective Action Sites (CASs), located in Areas 2, 3, and 9 of the Nevada National Security Site: (1) CAS 02-37-02, Gas Sampling Assembly; (2) CAS 03-99-19, Gas Sampling Assembly; AND (3) CAS 09-99-06, Gas Sampling Assembly Closure activities began in August 2011 and were completed in June 2012. Activities were conducted according to the Corrective Action Decision Document/Corrective Action Plan (CADD/CAP) for CAU 547 (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office [NNSA/NSO], 2011). The recommended corrective action for the three CASs in CAU 547 was closure in place with administrative controls. The following closure activities were performed: (1) Open holes were filled with concrete; (2) Steel casings were placed over vertical expansion joints and filled with cement; (3) Engineered soil covers were constructed over piping and exposed sections of the gas sampling system components; (4) Fencing, monuments, Jersey barriers, radiological postings, and use restriction (UR) warning signs were installed around the perimeters of the sites; (5) Housekeeping debris was picked up from around the sites and disposed; and (6) Radiological surveys were performed to confirm final radiological postings. UR documentation is included in Appendix D. The post-closure plan was presented in detail in the CADD/CAP for CAU 547 and is included as Appendix F of this report. The requirements are summarized in Section 5.2 of this report. The proposed post-closure requirements consist of visual inspections to determine the condition of postings and radiological surveys to verify contamination has not migrated. NNSA/NSO requests the following: (1) A Notice of Completion from the Nevada Division of Environmental Protection to NNSA/NSO for closure of CAU 547; and (2) The transfer of CAU 547 from Appendix III to Appendix IV, Closed Corrective Action Units, of the FFACO.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Removal of Chalk River unidentified deposit (CRUD) radioactive waste by enhanced electrokinetic process

Decontamination techniques proposed and used to remove Chalk River unidentified deposit (CRUD) in radioactive waste management. In cases of huge volumes of metal or radionuclides contaminated by CRUD, removal of CRUD by mechanical or chemical decontamination is difficult. An advanced electrokinetic process combined with chemical decontamination was applied to remove CRUD and experimentally evaluated. We used oxalic acid for CRUD removal, and cobalt (Co) released from the CRUD was transferred to the cathode in an electrokinetic reactor. Our results indicate that the combined system is efficient for CRUD removal with enhanced, efficiency by use of the cation exchange membrane and zeolite.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2013

Estimation of Sorption Behavior of Europium(III) Using Biotite Flakes - 13272

The interaction of biotite and Eu(III) (europium (III)) was examined by using secondary ion-microprobe mass spectrometer (SIMS), fluorescence emission spectrum and decay behavior of fluorescence emission spectrum in addition to the time-changes of Eu(III) and potassium ions concentrations in a solution, using the flake form samples. The results of SIMS showed that the intensity of Eu was gradually decreasing with depth, while the intensity of Eu in the case shaken for 30 days exceeded that in the case for 1 day. Furthermore, the spatial distribution of Eu(III) and potassium ions in the flake of biotite suggested that Eu ions diffuse mainly from the edges of biotite flake, while Eu ions can slightly diffuse through some small cracks existing on the flake surface far from the edges. Besides, the elution amount of potassium from the biotite flakes into a solution was proportional to the sorption amount of Eu(III). The changes nearly revealed ion exchange between these ions, while muscovite flake sample did not show such ion exchange reaction. In addition, from the time-change of Eu(III) concentration, an apparent diffusion coefficient was estimated to be 8.0x10{sup -12} m{sup 2}/s, by using two-dimensional diffusion model coupled with a film between the solid phase and the liquid phase. Furthermore, the fluorescent intensity decreased with the shaking (contacting) time. This means that Eu(III) gradually diffuses into the inside of biotite edges of the biotite flakes, after the sorption of Eu(III) in the edges. This tendency was observed also in the powder samples. The observed fluorescence decay (at 592 nm in wave length) showed almost similar curve in any samples, indicating a certain sorption form of Eu(III) onto the edges of the biotite flakes. These results mentioned above suggest that the diffusion processes through internal layer in biotite mainly control the sorption behavior of multivalent ions. Such diffusion processes affect the retardation-effects on fracture surfaces in the rock matrix, depending on the fluid flow velocity of groundwater. That is, a more reliable model considering the mass transfer in the internal layer of biotite may be required to estimate the sorption behavior of RNs with biotite which controls the whole sorption behavior of granite. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Unitary Thermal Energy Management for Propulsion Range Augmentation (UTEMPRA)

Battery Electric Vehicles (BEVs) encounter weather-dependent driving range. A major drain on the vehicle battery is the HVAC system which, when typically using resistive heating only, can result in a reduction of up to 50% of the driving range in winter weather. In this project, a BEV thermal management system is designed and demonstrated on a Fiat 500e BEV. The system comprises a simple, compact, semi-hermetic refrigerant sub-system and an accompanying a coolant circulatory system which, through an innovative valve arrangement, is used to transfer heat between the several heat sources and sinks in the BEV. The system is capable of A/C and heat pump operation, thermal management of the traction battery and powertrain electronics and also recycles waste heat of the vehicle powertrain to heat the cabin to further converse energy. An automated control system is implemented and tested that selects the relevant mode of operation and optimizes energy usage while meeting vehicle thermal requirements. In wind tunnel tests, the vehicle and UTEMPRA system demonstrated 13% (the target of the project was set at 15%) range increase at -10°C driving at a steady 40 kph speed. The vehicle also demonstrated full capability to automatically control thermal conditions at a wide range of ambient temperatures thus demonstrating the viability of such a system in practice.

33 ADVANCED PROPULSION SYSTEMS↗

NASA NTRS1988

Mars sample return power supply

A power supply is designed for a vehicle able to operate on the surface of Mars for a period of 5 to 10 years. This vehicle will be used for sample and data collection. The design is based on the assumption that the vehicle will be unmanned. Also, there will be no means by which components could be repaired or replaced while on the Martian surface. A consequence of this is that all equipment must meet high standards of reliability and, if possible, redundancy. Power will be supplied to the vehicle by means of a General Purpose Heat Source capable of producing a minimum of 7 kW of thermal power. The heat generated from the General Purpose Heat Source will be transferred to a Stirling engine via hot side heat pipes. The Stirling engine will then convert this heat into 2 kW of electrical power. Cold side heat pipes will be used to carry away waste heat, which will be released to the Martian environment via radiators connected to the end of the cold side heat pipes.

Hoang, Don↗

DOE OSTI2017

The Road to 2020: Comprehensive Environmental Response, Compensation and Liability Act of 1980 (CERCLA): Cleanup of the East Tennessee Technology Park - 17425

Uranium enrichment activities at the 890-ha (2,200-acre) East Tennessee Technology Park (ETTP), formerly the Oak Ridge Gaseous Diffusion Plant, date back to the World War II Manhattan Project. The site produced nuclear fuel for weapons and reactors for four decades. Because of these operations, ETTP has a legacy of radiologically and chemically contaminated buildings, soil, sediment, and groundwater that required remedial action to reduce the risk to human health and the environment. Since the early 1990's, the DOE Oak Ridge Office of Environmental Management (OREM) has directed the environmental cleanup of this contaminated site through the Comprehensive Environmental Response, Compensation and Liability Act of 1980 (CERCLA) process. Cleanup activities have included the demolition of contaminated facilities and structures, including five gaseous diffusion buildings; remediation of contaminated soils; disposal of associated waste; and treatment of contaminated groundwater. In August 2016, DOE realized its Vision 2016, demolition of all five gaseous diffusion buildings at ETTP. This milestone marked the first time that an entire uranium enrichment complex has been demolished. The landmark cleanup included deactivation and demolition of 418,064 m{sup 2} (4.5 million ft{sup 2}) of radiologically and chemically contaminated facilities, and disposal of more than 55,000 truckloads of debris. Fulfillment of Vision 2016 was a major step toward completing ETTP cleanup by 2020. To attain 2020 cleanup, DOE and its cleanup contractor, URS - CH2M Oak Ridge LLC (UCOR), must complete demolition of the following areas: - Poplar Creek Facilities; - K-731 Switch House; - Building K-1037; - Central Neutralization Facility; - K-1200 Complex; - Toxic Substances Control Act Incinerator facilities; - Balance of Site Facilities. In addition, DOE will complete all remedial actions in accordance with current Records of Decision (RODs); as well as cleanup activities involving surface water, sediment, and groundwater that will be addressed in a final site-wide ROD. All waste, including soils, debris, and rubble from demolition and remediation activities, will be dispositioned. Final cleanup of ETTP further advances DOE's goal for re-industrialization of the site as a private sector industrial park. With risks reduced, hazards removed, and the environment restored, DOE can transfer usable ETTP building and land assets to the community for redevelopment and reuse. A history center, equipment building, viewing tower, and wayside exhibits will tell the story of the site, honoring its contributions to the nation's defense and energy needs. Approximately 1,214 ha (3,000 acres) of adjoining lands, now the Black Oak Ridge Conservation Easement (BORCE), will be used for conservation and recreational purposes, including green-ways and trail systems. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

A Novel Mobile Examination and Remediation Facility for On-Site Remediation of Dry Storage Systems - 18389

Most of the spent nuclear fuel (SNF) in dry storage in the United States is safely and securely stored in dual-purpose (i.e., storage and transportation) canisters [1]. However, some of the SNF is stored in storage-only (i.e., non-transportable) packages, or packages that may become nonconforming for continued storage or transportation for a variety of reasons. Currently, the capability to remediate dry storage systems (DSSs) for SNF that are not transportable or that need to be repaired to maintain confinement boundary integrity during storage service does not exist without on-site access to a spent fuel pool. Hence, there is a potential future need for a mobile facility with remediation capabilities that can be used at multiple storage installations. These capabilities would permit needed repairs to an existing DSS to support continued storage or for the dry transfer of fuel to a new conveyance suitable for transport, if necessary. This is a technical paper that does not take into account the contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in canisters to be an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail. A novel design concept for a Mobile Examination and Remediation Facility (MERF) is proposed with functions and capabilities to remediate a DSS. The proposed MERF would be a multi-purpose facility, remotely operated and computer-controlled with semi-automated equipment to allow trained operators to safely perform remediation activities while keeping worker radiation dose as low as reasonably achievable. It would include a shielded rotating platform with support towers and would use commercial off- the-shelf examination and repair equipment. A slip ring arrangement underneath the rotating platform would allow for the power supply without impeding the continuous rotation and would simplify cable management. The support towers would be mounted on the rotating platform, and each tower would have vertical and horizontal positioning capability for axial and radial tool location. The rotating platform would be attached to the main MERF structure. Capabilities would include visual, surface, and volumetric examinations to fully characterize physical conditions on canister surface before and after a repair; repair techniques, e.g., laser beam welding, friction stir welding, or cold spray, to perform needed remediation activities; and associated equipment for surface preparation such as cutting and grinding equipment. The proposed design would be simple, modular, robust, economically viable, and easily transportable and would have a small footprint. The modular design would present several advantages including compatibility with a variety of different DSSs, on-site assembly, and transportability; it also would be built in interchangeable sections to facilitate decontamination. Further, shielding would be supplemented through interface with the existing transfer cask and storage overpack. Robustness and canister stability would be achieved using pneumatic holders and steel support structures designed with vibration absorbers to reduce impacts and/or a 'pendulum' effect during a seismic event. Thus, the proposed MERF would be designed to: (i) minimize handling of an affected DSS to the extent practicable, (ii) provide radiation shielding and minimize dose to operators and the public during remediation activities, and (iii) enable examination, repair, and repackaging. This paper describes the proposed MERF design concept and capabilities, main components, equipment, different modes of operation, and operating procedures. Structural and safety assessments would need to be performed before final design, licensing, and deployment. The MERF is evaluated with respect to regulatory impact (first of a kind, prior experience), operational impact (safety, security, staffing, ongoing operations, constraints), worker dose (low, medium, high), cost (low, medium, high), schedule (duration, time sequence), and technology readiness level. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2012

OSRP Source Repatriations-Case Studies: Brazil, Ecuador, Uruguay

The Global Threat Reduction Initiative's (GTRI) Offsite Source Recovery Project (OSRP) began recovering excess and unwanted radioactive sealed sources (sources) in 1999. As of February 2012, the project had recovered over 30,000 sources totaling over 820,000 Ci. OSRP grew out of early efforts at Los Alamos National Laboratory (LANL) to recover disused excess Plutonium- 239 (Pu-239) sources that were distributed in the 1960's and 1970's under the Atoms for Peace Program. Source recovery was initially considered a waste management activity. However, after the 9/11 terrorist attacks, the interagency community began to recognize that excess and unwanted radioactive sealed sources pose a national security threat, particularly those that lack a disposition path. After OSRP's transfer to the U.S. National Nuclear Security Administration (NNSA) to be part of GTRI, its mission was expanded to include all disused sealed sources that might require national security consideration. Recognizing the transnational threat posed by porous borders and the ubiquitous nature of sources, GTRI/OSRP repatriates U.S. origin sources based on threat reduction prioritization criteria. For example, several recent challenging source repatriation missions have been conducted by GTRI/OSRP in South America. These include the repatriation of a significant amount of Cs-137 and other isotopes from Brazil; re-packaging of conditioned Ra-226 sources in Ecuador for future repatriation; and, multilateral cooperation in the consolidation and export of Canadian, US, and Indian Co-60/Cs-137 sources from Uruguay. In addition, cooperation with regulators and private source owners in other countries presents opportunities for GTRI/OSRP to exchange best practices for managing disused sources. These positive experiences often result in long-term cooperation and information sharing with key foreign counterparts. International source recovery operations are essential to the preservation of U.S. national security interests. They are also mutually beneficial for fostering positive relationships with other governments and private industry, and demonstrate that responsible end-of-life options are given to legacy U.S.-origin sources in other countries. GTRI/OSRP does not take back sources that have a viable path for commercial disposal. Most US origin sources were sold commercially and were not provided by the US government. Below is a synopsis of cooperative efforts with Brazil, Ecuador, and Uruguay. Bilateral and multilateral efforts have been successful in removing hundreds of U.S.origin sealed radioactive sources from Latin American countries to the U.S. As many disused sources remain in the region, and since repatriation is not always an option, GTRI will continue to work with those countries to ensure that these sources are stored securely for the long-term. Successful Latin America operations should serve as a model for other regional cooperation in the repatriation of sealed sources, encouraging other source exporting countries to implement similar programs. Securing and removing sources, both domestically and internationally, is crucial to strengthening the life-cycle management of radioactive sources worldwide. Such efforts not only prevent these materials from being used maliciously, but also address public health and safety concerns, and under-gird the IAEA Code of Conduct on the Safety and Security of Radioactive Sources. (authors)

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2022

Exfoliating silica bilayers via intercalation at the silica/transition metal interface

The growth of the silica (SiO 2 ) bilayer (BL) films on transition metal (TM) surfaces creates a new class of two-dimensional (2D) crystalline, self-contained materials that interact weakly with the TM substrate. The BL-silica/TM heterojunction has shown unique physical and chemical properties that can lead to new chemical reaction mechanisms under the sub-nm confinement and broad potential applications ranging from surface protection, nano transistors, molecular sieves to nuclear waste removal. Novel applications of BL-silica can be further explored as a constituent of van der Waals assembly of 2D materials. Key to these applications is an unmet technical challenge to exfoliate and transfer BL-silica films in a large area from one substrate to another without material damage. In this study, we propose a new exfoliation mechanism based on gas molecule intercalation from density functional theory studies of the BL-silica/TM heterojunction. Here, we found that the intercalation of O atoms and CO molecules at the BL-silica/TM interface weakens the BL-silica—TM hybridization, which results in an exponential decrease of the exfoliation energy against the interface distance as the coverage of interfacial species increases. This new intercalation mechanism opens up the opportunity for non-damaging exfoliation and transfer of large area silica bilayers.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Integrating the Safety Evaluation for a Molten Salt Reactor Operation and Fuel Cycle Facility Application

Molten salt reactor (MSR) sites may include additional elements of the nuclear fuel cycle beyond those of the existing fleet. In the existing fleet, the individual elements of the fuel cycle typically have been located on different sites and licensed separately. Providing robust separation between hazards remains a useful safety practice for MSRs. Although the different elements of the fuel cycle at a nuclear site that includes MSRs may transfer material between processes more frequently than prior practices, providing adequate separation between distinct facilities avoids the potential for adverse interactions. Additionally, some elements of the MSR fuel cycle, such as fuel salt preparation or waste stabilization, may be more efficient to share among multiple nearby reactors, and nuclear sites that include MSRs may also include other reactor classes. Hence, discrete MSR fuel cycle facilities located at a common site could be physically separated with robust barriers—albeit potentially connected by piping—and licensed individually. This report describes the hazards of individual elements of representative MSR fuel cycle facilities, including their relationship to overall site level hazards. The report maps the regulatory compliance aspects of the individual MSR fuel cycle elements (e.g., fuel salt preparation, reactor, waste stabilization) to existing and developing regulations, as well as describes current and developing site-level regulations from an MSR perspective.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1980

Liquid droplet radiators for heat rejection in space

A radiator for heat rejection in space is described which utilizes a stream of liquid droplets to radiate waste heat. The large surface area per mass makes the liquid droplet radiator at least an order of magnitude lighter than tube and fin radiators. Generation and collection of the droplets, as well as heat transfer to the liquid, can be achieved with modest extensions of conventional technology. Low vapor pressure liquids are available which cover a radiating temperature range 250-1000 K with negligible evaporation losses. The droplet radiator may be employed for a wide range of heat rejection applications in space. Three applications - heat rejection for a high temperature Rankine cycle, cooling of photovoltaic cells, and low temperature heat rejection for refrigeration in space illustrate the versatility of the radiator.

Mattick, A. T.↗

NASA NTRS1986

A lunar transportation system

Due to large amounts of oxygen required for space travel, a method of mining, transporting, and storing this oxygen in space would facilitate further space exploration. The following project deals specifically with the methods for transporting liquid oxygen from the lunar surface to the Lunar Orbit (LO) space station, and then to the Lower Earth Orbit (LEO) space station. Two vehicles were designed for operation between the LEO and LO space stations. The first of these vehicles is an aerobraked design vehicle. The Aerobrake Orbital Transfer Vehicle (OTV) is capable of transporting 5000 lbm of payload to LO while returning to LEO with 60,000 lbm of liquid oxygen, and thus meet mission requirements. The second vehicle can deliver 18,000 lbm of payload to LO and is capable of bringing 60,000 lbm of liquid oxygen back to LEO. A lunar landing vehicle was also designed for operation between LO and the established moon base. The use of an electromagnetic railgun as a method for launching the lunar lander was also investigated. The feasibility of the railgun is doubtful at this time. A system of spheres was also designed for proper storing and transporting of the liquid oxygen. The system assumes a safe means for transferring the liquid oxygen from tank to tank is operational. A sophisticated life support system was developed for both the OTV and the lunar lander. This system focuses on such factors as the vehicle environment, waste management, water requirements, food requirements, and oxygen requirements.

Source record↗