Proceedings of the Magnetics Workshop
Conference on measurement and reduction of spacecraft magnetic fields - materials and design criteria - magnetic decontamination of components - testing facilities and human tests
CONFERENCE↗
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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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Research on Microbiological Sterilization Problems Quarterly Report, Jul. - Sep. 1966
Microbiological sterilization studies for spacecraft decontamination
SPACECRAFT CONTAMINATION↗
Polymers for spacecraft hardware Monthly technical progress report, 10 Sep. - 9 Oct. 1966
Spacecraft environmental and decontamination tests on polymers
ENVIRONMENTAL TESTING↗
NASA requirements for the sterilization of spacecraft
NASA requirements for decontamination and sterilization of planetary landing capsules
NASA PROGRAM↗
Microbiological burden on the surfaces of the AIMP spacecraft, part 1
Decontamination procedure with isopropyl alcohol for reducing microbial population from surfaces of Anchored Interplanetary Monitoring Platform /AIMP/
DECONTAMINATION↗
Microbiological burden on the surfaces of the AIMP spacecraft, part 3
Decontamination and microbiological monitoring of surfaces of Anchored Interplanetary Monitoring Platform spacecraft
INTERPLANETARY SPACECRAFT↗
Scope of Work for Scientific and Technical Assistance for the Planetary Quarantine Mission Quarterly Report, Period Ending Dec. 31, 1966
Mission planning, and spacecraft sterilization model for microbe decontamination in application to planetary travel studies
MISSION PLANNING↗
Separation of terbium from proton-irradiated gadolinium oxide targets – development of an effective, scalable and automatable process
This work reports an effective and scalable radiochemical separation process for isolating terbium from Gd 2 O 3 . The separation process uses three commercially available extraction chromatography resin columns, has been implemented on a computer-controlled chemistry module, and tested with 100 mg quantities of proton-irradiated nat Gd 2 O 3 . Here, the 4-hour separation procedure isolated radioterbium in 1.3 mL of 0.01 M HCl with 80 ±8% radiochemical yield and a Gd decontamination factor >(1.2 ± 0.3)·10 5 .
Adjacent lanthanide separation↗
Evaluation of a high-throughput method for processing sponge-stick samples to detect viable, non-spore-forming biothreat agents
After a bioterrorism incident, surface sampling is often used to determine the extent of contamination and exposure, guiding decontamination efforts and decisions for re-occupancy of affected sites. The sponge-stick (SS) is a preferred and commonly used device for sample collection to detect both spore-forming and non-spore-forming biothreat agents from non-porous surfaces. Here, in this study, a recently developed high-throughput method (HTM) for processing SS samples to detect viable Bacillus anthracis spores was adapted for detection of non-spore-forming biothreat agents, Yersinia pestis and Francisella tularensis. The scalable HTM was used to process up to 20 SS samples simultaneously, compared to the current stomacher-based method which processes one SS at a time. Comparisons of the HTM and the stomacher-based method were statistically indistinguishable for most experiments (P > 0.05) with HTM recoveries of 37–60 % for Y. pestis inoculated at 102–103 cells/SS and held 48 h at 4 °C to mimic sample transport/storage. The HTM was integrated with Rapid Viability-Polymerase Chain Reaction (RV-PCR) analysis to detect viable Y. pestis in the presence of particulate contamination (Arizona Test Dust, ATD). This approach detected Y. pestis inoculated at 20 cells/SS and ATD did not impact detection (P > 0.05). F. tularensis showed significantly lower recoveries between no-hold time and 48-h hold time (4 °C, P < 0.05) using the HTM, which further testing showed could be due to toxicity of the neutralizing buffer used for SS pre-wetting. With modifications, this method could enhance throughput capacity while maintaining similar recovery efficiencies to current methods for other non-spore-forming bacterial pathogens.
Biological and medical sciences↗
Capture of volatile RuO 4 from oxidized simulated used nuclear fuel solutions
Ruthenium is a challenging fission product in used nuclear fuel (UNF) reprocessing due to its complex redox chemistry, variable speciation in nitric acid, partial extractability, and volatility. This work presents a strategy for Ru removal based on the volatilization of RuO 4 using the strong oxidant sodium bismuthate, followed by RuO 2 deposition onto various substrates. Among the materials tested, polymer-based substrates such as polyolefin wax film (Parafilm®), exhibited superior performance, achieving up to near-quantitative Ru removal from solution. After dissolution of the substrate, 99.6% of the Ru was recovered as RuO 2 . The deposition mechanism onto the polyolefin wax film involves both physisorption and chemical reaction through oxidative cleavage of olefinic bonds within the polymer. In contrast, on inorganic substrates such as Al foil, RuO 4 undergoes incomplete surface adsorption and reduction, as confirmed by FTIR and XPS analyses. This approach remains effective under conditions of simulated oxidation of UNF solutions, selectively removing Ru while leaving other fission products in solution. As a result, capture of RuO 4 offers a practical and efficient strategy for ruthenium decontamination and recovery, offering a route for integration into existing UNF reprocessing flowsheets to enhance overall process safety and performance.
Ruthenium↗
Direct Extraction of Uranium Oxides with N,N -Di(2-ethylhexyl)isobutyramide
The direct extraction of uranium from voloxidized nuclear fuel into an organic solvent offers several potential advantages over conventional hydrometallurgical reprocessing, including reducing the reprocessing plant footprint, providing an initial degree of decontamination from fission products, and minimizing the amount of secondary waste from nitric acid. In this work, the direct extraction of uranium oxides into 1.5 M N,N-di(2-ethylhexyl)isobutyramide (DEHiBA) in n-dodecane is examined. UV–vis spectra and distribution ratios of HNO 2 in 1.5 M DEHiBA as well as the equilibrium organic phase H 2 O concentrations in HNO 3 -loaded 1.5 M DEHiBA are also reported. Hypothesized reaction stoichiometries for the direct extraction of uranium from UO 2 , α-U 3 O 8 , and ε-UO 3 are verified through analysis of organic-phase U, HNO 2 , and HNO 3 concentrations after dissolution. Water generated by the dissolution results in the formation of a separate aqueous phase, which will need to be accounted for in future flowsheet design.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Separation of dilute plutonium(IV) from nitric acid solutions via Reillex HPQ anion exchange resin
Anion exchange is a common method for the separation of plutonium (Pu) in aqueous systems, providing high recovery and decontamination from impurities. Reillex HPQ anion exchange resin has been used around the Department of Energy complex in this application for decades but has typically been used for Pu feed concentrations greater than 1 g/L. It is desirable to recover Pu at more dilute concentrations from a dissolved used fuel or target element, for example. In this work, the performance of Reillex HPQ anion exchange resin was characterized with feed solutions of ca. 0.24 g/L Pu. A significant reduction in the performance of the resin was realized with this dilute concentration relative to a feed solution of 2.35 g/L Pu with similar feed flow rates and the same resin column, likely due to the increased driving force for mass transfer with a higher feed concentration. The maximum capacity of Reillex HPQ resin for Pu can be improved by reducing the flow rate and increasing the feed Pu concentration. Furthermore, the breakthrough point of Pu can be improved in the same manner and by increasing the resin bed height. Diffusion of Pu into the resin was illustrated to be the rate limiting step.
Plutonium separations↗
An Application of Molecular Recognition for the Efficient Removal of Cesium from Hanford Nuclear Waste by Modular Solvent Extraction
In this work, experimental results leading to flowsheet design are presented showing how a calixarene-crown ether based solvent-extraction process can meet the challenge of cesium removal from nuclear tank wastes stored at the US Department of Energy Hanford site. Cleanup of legacy Cold War nuclear waste stored in underground tanks represents one of the greatest environmental challenges facing the US Department of Energy in terms of risk, cost, and effectiveness of applicable science and technology. Planning for the cleanup at the Hanford Site calls for the removal of the radioactive fission product 137Cs from its alkaline salt waste, including the use of modular processes that can be deployed near the tank farms. To meet the resulting need for extremely high selectivity, the Next-Generation Caustic-Side Solvent Extraction (NG-CSSX) process employing a calix[4]arene-crown ether in modified kerosene has been adapted to remove sub-millimolar cesium in competition with molar sodium and potassium in a high-nitrate alkaline matrix. Potassium loading in the solvent was determined in extraction, scrubbing, and stripping, leading to an empirical model closely approximating cesium distribution ratios for a variety of Hanford waste types. Process chemistry has been developed based on this molecular-recognition approach, focusing on the competitive effect of potassium loading and the mitigating process modifications needed, including extending the scrub section. The result is a modular flowsheet design that can achieve cesium decontamination factors well in excess of 15,000 even for the worst-case Hanford waste.
Williams, Neil [ORNL] (ORCID:000000023159226X)↗
Integrated Off-Gas System Tests on the DM1200 Melter with RPP-WTP LAW Sub-Envelope Bl Simulants, VSL-03R3851-1, Rev. 0 (Oct 2003)
The report documents melter and off-gas performance results obtained on the DM1200 melter during processing of LAW Sub-Envelope B1 feed. The principal objectives of the DM1200 melter testing were to demonstrate suitable processing and production rates for LAW B1 feed; characterize the glass product for elemental composition; measure elemental emission rates and decontamination factors (DFs) across the melter and off-gas treatment components; and provide additional data on the performance of the thermal catalytic oxidation and selective catalytic reduction (TCO/SCR) unit efficiency in treating organic emissions. The feed was spiked with organic compounds for part of the testing duration. All data were collected for use in engineering and system design purposes. No sampling and analysis to collect environmental data was performed during these tests.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Investigation of Anion-Exchange Method for Purification of Fission-Produced 99 Mo
The LEU-Modified Cintichem (LMC) process is a small-scale (50- to 100-mL) method for purifying fission-produced 99 Mo. Traditionally, the process utilizes custom-made glassware and requires significant manual operations with hot cell manipulators. As an alternative to the LMC process, there is great interest in developing a 99 Mo purification method that could be automated to minimize remote handling while achieving the same purity specifications for the 99 Mo product. This report explores an ion-exchange approach to purifying fission-produced 99 Mo and discusses the decontamination factors for various fission products that are present after initial recovery of Mo on a titania column from a sulfate-based and uranium-containing solution.
07 ISOTOPE AND RADIATION SOURCES↗
Technical Assessment of Off-Gas System Technologies for Potential Use in Molten Salt Reactors
Molten salt reactors (MSRs) are a class of Generation IV advanced reactor technologies aimed to enhance and improve the safety, fuel utilization and cost-effectiveness of nuclear power generation. MSR concepts are based on using a molten salt mixture as a primary nuclear reactor coolant, while the fuel can be either directly dissolved in the coolant (i.e., fluid-fueled) or can be in a separate solid form (i.e., salt-cooled solid-fueled). MSRs can be operated in either thermal or fast neutron spectra, as simple fissile convertors, or breeder reactors (e.g., utilizing both fissile and fertile fuels), as well as accelerator-driven sub-critical reactors or transuranic element burners. Operation of MSRs will necessitate management and treatment of the off-gas streams released to the headspace of the reactor where, in some designs, a cover gas can be circulated to remove certain fission products and maintain an inert atmosphere. An inert MSR cover gas swept over or sparged into the salt will confine radionuclides emerging from the free surface, including radioactive noble and non-noble gases, aerosols, and volatile species. The elements and compounds of these species will have a wide range of thermochemical and physical properties, some of which will decay to different products in the off-gas stream. Therefore, system components used to manage and treat MSR off-gas streams will need to be robust under dynamic reactor and chemistry conditions to support adequate reactor performance by managing the composition and species in the salt coolant and cover gas space. Off-gas system components would perform as a radionuclide boundary to the reactor vessel and a primary means for preventing radionuclide release; therefore, they are of high safety significance for MSR operations. This report reviews various technologies and materials for potential implementation in the design of an MSR off-gas system per a general conceptual framework. The reviewed technologies operate based on scrubbing, capture, delay, or separation of species in the off-gas stream. These include molten hydroxide scrubbers, solid sorbents, delay off-gas systems, particle traps, and cryogenic distillation. Since aspects of an off-gas system will be design-specific, the approach is to provide general discussions on attributes of these technologies and materials, including target off-gas stream species, operational parameters and conditions of importance to decontamination factors, implementation maturity, system component monitoring, potential off-normal conditions and considerations for waste products. The information aims to support the U.S. Nuclear Regulatory Commission in their safety evaluations for MSR designs.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Engineering Assessment And Certification of Integrity Of the DWTF Waste Treatment Tanks At Lawrence Livermore National Laboratory
This Engineering Assessment and Certification of Integrity of the Building 695 (B-695) Decontamination and Waste Treatment Facility (DWTF) Waste Treatment Tanks has been prepared using the guidelines of CCR §66270.16(a) “Specific Part B Information Requirements for Tank Systems”, and the CCR §66264.192(b) requirements for the written assessment of tank systems that store hazardous waste and have secondary containment. The regulations require that this assessment be completed and certified by an independent, qualified, California-registered Professional Engineer. This document will be kept on file by the Lawrence Livermore National Laboratory (LLNL) Environmental Functional Area and is applicable for a maximum of five (5) years from the date of issue.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
U.S. Efforts in Support of Examinations at Fukushima Daiichi - September 2024 Meeting Notes
Information obtained from Fukushima Daiichi Nuclear Power Station (Daiichi) is required to inform future Decontamination and Decommissioning (D&D) activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This information also has important implications for the safety and operation of U.S. Commercial nuclear power plants. A collaborative U.S. and Japanese effort was initiated in 2014 by the Department of Energy Office of Nuclear Energy to identify Daiichi examination needs and evaluate recent Daiichi examination data to address these needs. This document summarizes information presented at and findings, action items, and recommendations by U.S. and Japanese experts in reactor safety and plant operations during the September 2024 Forensics Effort meeting. Significant safety insights were obtained in several areas: system and component performance, radionuclide surveys and sampling, debris end-state location, combustible gas effects, and plant operations and maintenance. In addition to reducing uncertainties and knowledge gaps in severe accident modeling progression, these insights continue to be used to assess whether additional updates are needed in guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, Daiichi-related activities, such as code modeling improvements and analysis, testing, and new technology deployment efforts, have the potential to offer additional safety and economic benefits to the operating fleet and new light water reactor (LWR) and non-LWR designs.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗