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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 721 records · Page 40
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Behaviour of IPG waste forms bearing BaSO{sub 4} as the dominant sludge constituent generated from the treatment of water used for cooling the stricken power reactors

The great amount of water used for cooling the stricken power reactors at Fukushima Dai-ichi following the earthquake and tsunami of 11 March 2011 has resulted in accumulation of cooling water so-called the remaining water in some buildings. As the cooling water is subsequently contaminated by fission products (FPs) and some other radioactive substances, it is necessary to decontaminate this 'cooling water' to reduce the volume of liquid radioactive waste and to reuse it again for cooling the affected reactors. Some methods are applied to remove the radioactive substances from the cooling water. However, after treatments of water, there arises a secondary radioactive waste, the sludge. Steps are now taken to immobilize this sludge. In this paper, BaSO{sub 4}, as one of main constituents of the sludge, was chosen as an immobilizing target substance. The appropriate manufacturing condition of glass waste form for loading the sludge (BaSO{sub 4}) was studied and the chemical durability was evaluated by measuring the dissolution rate. For this experiment, the iron phosphate glass (IPG), which is known to possess a large loading capacity for a variety of chemical substances, was employed as the glass medium. Based on experimental results, it is evident that BaSO{sub 4} can be loaded into the IPG medium when it possesses the appropriate composition and melting temperature. During loading, BaSO{sub 4} converted into BaO, acting as a network modifier, which leads to enhanced stability of IPG. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Experiences with new neutralization technologies for remediation after ISL mining of uranium in Straz Pod Ralskem

A big affection of the rock environment and groundwaters occurred during the chemical mining of uranium in the years 1966 to 1996 in the neighbourhood of the town Straz pod Ralskem in the Czech Republic. It is necessary to clean the residual technological solutions (RTS) from the underground. The pH of the solutions in some places is still less than 2 and the concentration of sulphates reaches up to 65 g/l. The remedial activities consist of pumping of the RTS from the ground and reprocessing of the RTS in the surface technologies. The implementation of the new neutralization technologies NDS ML and NDS 10 help us with increasing of the efficiency of the remedial process. The NDS ML technology ('Mother liquor reprocessing station') started its operation in 2009 and it processes the concentrated technological solution from the evaporation station after the alum crystallization (mother liquor) with the concentration of total dissolved solids up to 250 g/l. The principle is the neutralization of the acid solutions with the aid of the lime milk. The suspension is then filtrated in the filter press, the filter cake is deposited in the tailings pond and the filtrate is injected back into the underground rock environment. The NDS 10 technology ('Neutralization and Decontamination Station NDS 10') started its operation in 2012 and it works on the same technological principle as the NDS ML station. The difference is that the NDS 10 station can process higher volume (4.4 m{sup 3}/min) of the RTS with lower concentration of total dissolved solids 20 - 25 g/l. This poster describes the experiences of the state enterprise DIAMO with putting of these two neutralization technologies into operation and with using of the lime milk neutralization in such a large scale. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Management of the decommissioning of the Thetis reactor

The Thetis research reactor on the site of the Nuclear Sciences Institute of the Ghent University has been in operation from 1967 until December 2003. This light-water moderated graphite-reflected low-enriched uranium pool-type reactor has been used for various purposes e.g. the production of radioisotopes and activation analyses. During the first years its core power was 15 kW. In the early '70, a core enlargement allowed for operation at typically 150 kW, while the maximum was allowed to be 250 kW In September 2007, Ghent University entrusted to SCK-CEN the management of the back-end of the spent fuel and the decommissioning of the reactor. In 2010, the spent fuel was removed from the reactor and transported to Belgoprocess for cementation in 400 l drums and interim storage awaiting final disposal. This activity allows tackling the decommissioning of the reactor. The objective is to complete its decommissioning by the end of 2014. In the framework of the decommissioning of the Thetis reactor, SCK-CEN set-up the final decommissioning plan and the decommissioning licensing file. These documents include among others a radiological inventory of the reactor. The graphite moderator blocks, the control and the safety pates, the liner of the pool were modeled to assess the activation products (isotopic vector and intensity). At the end of the unloading of the reactor in 2010 a brief mapping of the equipment's and internals of the reactor pool was performed. In 2012, we realized a more detailed mapping. These results confirmed those performed earlier and allowed to confirm the assumptions made in the final decommissioning plan. We set-up the terms of reference for the first decommissioning phase of the reactor namely the dismantling of the reactor i.e. reactor pool, circuits and rabbit system, equipment's and ventilation ducts. The removal of asbestos is also included into this phase. We conducted the selection process and the awarding of this decommissioning job. We gained the decommissioning license in May 2012. We also prepared the software tool allowing managing the decommissioning project by updating the inventory and recording the progress, the characterization measurements and the material and waste production. This software allows also to trace all the material streams and to report to the Authorities. This software is a simplified release of the ones developed by SCK-CEN in the framework of other decommissioning projects like BR3 and Belgonucleaire. The dismantling of the reactor i.e. reactor pool, circuits and rabbit system, will be performed in 2013. In 2014, it is planned to map all the surfaces of the infrastructure to highlight residual contamination of floor, walls and ceiling. The contaminated surfaces will be decontaminated and controlled. The objective is to reach the free release of the reactor building and laboratories by the end of 2014. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Characterization of the cleaning process on a transferred graphene

Intrinsic graphene possesses many outstanding physical and chemical properties, but their full explorations are often hindered by the effects of substrate and/or contamination. The authors employ the ultrahigh vacuum transmission electron microscopy equipped with a residual gas analyzer to in-situ characterize an effective decontamination process on a suspended graphene. Raman spectroscopic spectra further verify the cleanness of the resultant graphene membrane. The authors also present two contrasting growth morphologies of copper nanoparticles obtained on both clean and unclean graphene surfaces and show that the intrinsic growth dynamics can only manifest on the surface without contaminations.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2014

Ecological succession and viability of human-associated microbiota on restroom surfaces

Human-associated bacteria dominate the built environment (BE). Following decontamination of floors, toilet seats, and soap dispensers in four public restrooms, in situ bacterial communities were characterized hourly, daily, and weekly to determine their successional ecology. The viability of cultivable bacteria, following the removal of dispersal agents (humans), was also assessed hourly. A late-successional community developed within 5 to 8 h on restroom floors and showed remarkable stability over weeks to months. Despite late-successional dominance by skin- and outdoor-associated bacteria, the most ubiquitous organisms were predominantly gut-associated taxa, which persisted following exclusion of humans. Staphylococcus represented the majority of the cultivable community, even after several hours of human exclusion. Methicillin-resistant Staphylococcus aureus (MRSA)-associated virulence genes were found on floors but were not present in assembled Staphylococcus pan-genomes. Viral abundances, which were predominantly enterophages, human papilloma virus, and herpesviruses, were significantly correlated with bacterial abundances and showed an unexpectedly low virus-to-bacterium ratio in surface-associated samples, suggesting that bacterial hosts are mostly dormant on BE surfaces.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2021

Point Source Likelihood Framework (psl) v1.0

PSL is a software framework for localizing radiation point sources in a 3D environment using a freely moving radiation detector system. The framework assumes a single point source of radiation is in the environment and uses known detector positions and orientations along with their radiation measurements to find the most likely location of the point source. This approach to the problem enables analyses that improve upon the capability of a standard Maximum Likelihood Expectation Maximization (MLEM) approach, such as setting a source detection threshold, finding a likelihood interval for the position of the source, finding a likelihood interval for the activity (intensity) of the source, determining whether more than one point source may be present, and generating a map of Minimum Detectable Activity (MDA) for a given source type. Applications include radiological and nuclear security, consequence management, decontamination and decommissioning of nuclear facilities, and radiological emergency response.

Joshi, Tenzing↗
DOE OSTI2015

Light water detritiation

Hundreds of thousands of tons of tritiated light water have been accumulating from the enterprises of nuclear fuel cycles around the world. The Dual-Temperature Water-Hydrogen (DTWH) process looks like the only practical alternative to Combined Electrolysis and Catalytic Exchange (CECE). In DTWH power-consuming lower reflux device (electrolytic cell) is replaced by a so-called 'hot tower' (LPCE column operating at conditions which ensure relatively small value of elementary separation factor α(hot)). In the upper, cold tower, the tritium transfers from hydrogen to water while in the lower, hot tower - in the opposite direction - from water to hydrogen. The DTWH process is much more complicated compared to CECE; it must be thoroughly computed and strictly controlled by an automatic control system. The use of a simulation code for DTWH is absolutely important. The simulation code EVIO-5 deals with 3 flows inside a column (hydrogen gas, water vapour and liquid water) and 2 simultaneous isotope exchange sub-processes (counter-current phase exchange and co-current catalytic exchange). EVIO-5 takes into account the strong dependence of process performance on given conditions (temperature and pressure). It calculates steady-state isotope concentration profiles considering a full set of reversible exchange reactions between different isotope modifications of water and hydrogen (12 molecular species). So the code can be used for simulation of LPCE column operation for detritiation of hydrogen and water feed, which contains H and D not only at low concentrations but above 10 at.% also. EVIO-5 code is used to model a Tritium Removal Facility with a throughput capacity of about 400 m{sup 3}/day. Simulation results show that a huge amount of wet-proofed catalyst is required (about 6000 m{sup 3}), mainly (90%) in the first stage. One reason for these large expenses (apart from a big scale of the problem itself) is the relatively high tritium separation factor in the hot tower. The introduction of some quantity of deuterium into the gaseous flow before the hot tower lowers the tritium separation factor in that column. One possible variant of deuterium introduction to the hot tower of the first stage was modelled. The decontamination capacity increases by a 2.5 factor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Detection of uranyl fluoride and sand surface contamination on metal substrates by hand-held laser-induced breakdown spectroscopy

Here, a hand-held device for laser-induced breakdown spectroscopy has been investigated for the determination of uranyl fluoride surface contamination. This research demonstrates the ability to successfully detect uranium on surfaces when using a low resolving power (λ/Δλ=4000) spectrograph, with a 5 mJ energy per 1 ns pulsed laser radiation, available as a commercially packaged hand-held system. Sand/uranyl fluoride mixtures are prepared to simulate residue likely encountered during decontamination efforts at facilities that handle uranium hexafluoride. Detection limits are described for four uranium lines with one revealing the capability to detect uranium at a level of 250 parts per million. Lastly, advantages of the studied compact device include that location specific information can be obtained on-site to augment contamination identification.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Ultraviolet germicidal irradiation and its effects on elemental distributions in mouse embryonic fibroblast cells in x-ray fluorescence microanalysis

Rapidly-frozen hydrated (cryopreserved) specimens combined with cryo-scanning x-ray fluorescence microscopy provide an ideal approach for investigating elemental distributions in biological cells and tissues. However, because cryopreservation does not deactivate potentially infectious agents associated with Risk Group 2 biological materials, one must be concerned with contamination of expensive and complicated cryogenic x-ray microscopes when working with such materials. We employed ultraviolet germicidal irradiation to decontaminate previously cryopreserved cells under liquid nitrogen, and then investigated its effects on elemental distributions under both frozen hydrated and freeze dried states with x-ray fluorescence microscopy. We show that the contents and distributions of most biologically important elements remain nearly unchanged when compared with non-ultraviolet-irradiated counterparts, even after multiple cycles of ultraviolet germicidal irradiation and cryogenic x-ray imaging. This provides a potential pathway for rendering Risk Group 2 biological materials safe for handling in multiuser cryogenic x-ray microscopes without affecting the fidelity of the results.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Isolation of 163 Ho from dysprosium target material by HPLC for neutrino mass measurements

The rare earth isotope 163 Ho is of interest for neutrino mass measurements. This report describes the isolation of 163 Ho from a proton-irradiated dysprosium target and its purification. A Dy metal target was irradiated with 16 MeV protons for 10 h. After target dissolution, 163 Ho was separated from the bulk Dy via cation-exchange high performance liquid chromatography using 70 mmol dm –3 α-hydroxyisobutyric acid as the mobile phase. Subsequent purification of the collected Ho fraction was performed to remove the α-hydroxyisobutyrate chelating agent and to concentrate the Ho in a low ionic strength aqueous matrix. The final solution was characterized by MC-ICP-MS to determine the 163 Ho/ 165 Ho ratio, 163 Ho and the residual Dy content. The HPLC purification process resulted in a decontamination factor 1.4E5 for Dy. As a result, the isolated Ho fraction contained 24.8 ±1.3 ng of 163 Ho corresponding to holmium recovery of 72 ± 3%.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2013

Interim salt disposition program macrobatch 6 tank 21H qualification monosodium titanate and cesium mass transfer tests

Savannah River National Laboratory (SRNL) performed experiments on qualification material for use in the Interim Salt Disposition Program (ISDP) Batch 6 processing. This qualification material was a set of six samples from Tank 21H in October 2012. This sample was used as a real waste demonstration of the Actinide Removal Process (ARP) and the Extraction-Scrub-Strip (ESS) tests process. The Tank 21H sample was contacted with a reduced amount (0.2 g/L) of MST and characterized for strontium and actinide removal at 0 and 8 hour time intervals in this salt batch. 237 Np and 243 Am were both observed to be below detection limits in the source material, and so these results are not reported in this report. The plutonium and uranium samples had decontamination factor (DF) values that were on par or slightly better than we expected from Batch 5. The strontium DF values are slightly lower than expected but still in an acceptable range. The Extraction, Scrub, and Strip (ESS) testing demonstrated cesium removal, stripping and scrubbing within the acceptable range. Overall, the testing indicated that cesium removal is comparable to prior batches at MCU.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Miscibility Evaluation Of The Next Generation Solvent With Polymers Currently Used At DWPF, MCU, And Saltstone

The Office of Waste Processing, within the Office of Technology Innovation and Development, funded the development of an enhanced Caustic-Side Solvent Extraction (CSSX) solvent for deployment at the Savannah River Site for removal of cesium from High Level Waste. This effort lead to the development of the Next Generation Solvent (NGS) with Tris (3,7-dimethyl octyl) guanidine (TiDG). The first deployment target for the NGS solvent is within the Modular CSSX Unit (MCU). Deployment of a new chemical within an existing facility requires verification that the new chemical components are compatible with the installed equipment. In the instance of a new organic solvent, the primary focus is on compatibility of the solvent with organic polymers used in the affected facility. This report provides the calculated data from exposing these polymers to the Next Generation Solvent. An assessment of the dimensional stability of polymers known to be used or present in the MCU, Defense Waste Processing Facility (DWPF), and Saltstone facilities that will be exposed to the NGS showed that TiDG could selectively affect the elastomers and some thermoplastics to varying extents, but the typical use of these polymers in a confined geometry will likely prevent the NGS from impacting component performance. The polymers identified as of primary concern include Grafoil ® (flexible graphite), Tefzel ® , Isolast ® , ethylene-propylene-diene monomer (EPDM) rubber, nitrile-butadiene rubber (NBR), styrene-butadiene rubber (SBR), ultra high molecular weight polyethylene (UHMWPE), and fluorocarbon rubber (FKM). Certain polymers like NBR and EPDM were found to interact mildly with NGS but their calculated swelling and the confined geometry will impede interaction with NGS. In addition, it was found that Vellumoid (cellulose fibers-reinforced glycerin and protein) may leach protein and Polyvinyl Chloride (PVC) may leach plasticizer (such as Bis-Ethylhexyl-Phthalates) into the NGS solvent. Either case will not impact decontamination or immobilization operations at Savannah River Site (SRS). Some applications have zero tolerance for dimensional changes such as the operation of valves while other applications a finite dimensional change improves the function of the application such as seals and gaskets. Additional considerations are required before using the conclusions from this work to judge outcomes in field applications. Decane, a component of Isopar L that is most likely to interact with the polymers, mildly interacted with the elastomers and the propylene based polymers but their degree of swelling is at most 10% and the confined geometry that they are typically placed in indicate this is not significant. In addition, it was found that Vellumoid may leach protein into the NGS solvent. Since Vellumoid is used at the mixer in Saltstone where it sees minimum quantities of solvent, this leaching has no effect on the extraction process at MCU or the immobilization process at saltstone. No significant interaction is expected between MaxCalix and the polymers and elastomers used at MCU, DWPF, and Saltstone. Overall, minimal and insignificant interactions are expected on extraction and immobilization operations when MCU switches from CSSX to NGS solvent. It is expected that contacting NGS will not accelerate the aging rate of polymers and elastomers under radiation and heat. This is due to the minimal interaction between NGS and the polymers and the confined geometries for these polymers. SRNL recommends the use of the HSP method (for screening) and some testing to evaluate the impact of other organic such as alcohols, glycolate, and their byproducts on the polymers used throughout the site.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Assessment of the Impact of a New Guanidine Suppressor In NGS on F/H Laboratory Analyses For DWPF and Saltstone MCU Transfers

Implementation of the Next Generation Solvent (NGS) in the Modular Caustic-Side Solvent Extraction Unit (MCU) will now proceed with a new suppressor compound, 1,2,3-tris(3,7-dimethyloctyl)guanidine (TiDG), replacing the originally planned suppressor for NGS, 1,3-dicyclohexyl-2-(11-methyldodecyl) guanidine (DCiTG). The Savannah River National Laboratory (SRNL) was tasked with evaluating the potential impact to F/H Laboratory analyses supporting the Defense Waste Processing Facility (DWPF) Waste Acceptance Criteria (WAC) used to qualify transfers of MCU Strip Effluent (SE) into the facility and the Saltstone WAC used to qualify transfers of Tank 50 containing Decontaminated Salt Solution (DSS) from MCU into Saltstone. This assigned scope is covered by a Task Technical and Quality Assurance Plan (TTQAP). Previous impact evaluations were conducted when the DCiTG suppressor was planned for NGS and concluded that there was no impact to either the determination of MCU SE pH nor the analysis of Isopar ® L carryover in the MCU SE and DSS streams. SRNL reported on this series of cross-check studies between the SRNL and F/H Laboratories. The change in suppressor from DCiTG to TiDG in the NGS should not impact the measurement of Isopar ® L or pH in SE or DSS necessary to satisfy DWPF and Saltstone WAC (Tank 50) criteria, respectively. A statistical study of the low bias observed in Isopar ® L measurements in both SRNL and F/H Laboratories may be necessary now that the final NGS composition is fixed in order to quantify the low bias so that a proper correction can be applied to measurements critical to the DWPF and Saltstone WACs. Depending upon the final DWPF WAC requirement put in place for SE pH, it could become necessary to implement an alternative ICP-AES measurement of boron. The current blended solvent system testing in SRNL should address any impacts to Isopar ® L carryover into either the DSS or the SE. It is recommended that SRNL monitor the current blended solvent work underway with simulants in SRNL as well as any DWPF CPC testing done with the new SE stream to ascertain whether any need develops that could result in modification of any currently planned F/H Laboratory testing protocols.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Investigation of Polymetric Biofilms Formed by Dangerous Pathogens, and the Prevention of Their Formation by Disinfectants and Bacteriophages (CRADA Final Report)

This was a collaborative effort between Lawrence Livermore National Security, LLC as manager and operator of Lawrence Livermore National Laboratory (LLNL) and Intralytix Inc., to study biofilms formed on the surfaces of polymers by especially dangerous pathogens and nosocomial infections, to explore the possibility of decontamination of surfaces using different classes of disinfectants and specific bacteriophages. The research was conducted under the Global Initiative for Proliferation Prevention (GIPP) Program of DOE, the United States Industry Coalition, Inc. (USIC), and cooperating New Independent States (NIS) of the Former Soviet Union (FSU). The Russian institutes involved with this project were from Ukrainian I.I.Mechnikov Anti-Plague Science Research Institute (UAPRI), Odessa, Ukraine; and G. Eliava Institute of Bacteriophage, Microbiology and Virology of the Georgian Academy of Sciences (Eliava Institute), Tbilisi, Georgia. The technical objectives of this CRADA project were to develop novel bacteriophage preparations with polyvalent characteristics for use in variety of applications and useful for the treatment of antibiotic resistant pathogens and those associated with biofilm matrices. This would be achieved through the identification, testing and improvement of bacteriophages capable of lysing specific, targeted nosocomial and other pathogens with the above described properties.

Garcia, Emilio↗
DOE OSTI2013

Quick Start User’s Guide for the PATH/AWARE Decision Support System

The Prioritization Analysis Tool for All-Hazards/Analyzer for Wide Area Restoration Effectiveness (PATH/AWARE) software system, developed by Sandia National Laboratories, is a comprehensive decision support tool designed to analyze situational awareness, as well as response and recovery actions, following a wide-area release of chemical, biological or radiological materials. The system provides capability to prioritize critical infrastructure assets and services for restoration. It also provides a capability to assess resource needs (e.g., number of sampling teams, laboratory capacity, decontamination units, etc.), timelines for consequence management activities, and costs. PATH/AWARE is a very comprehensive tool set with a considerable amount of database information managed through a Microsoft SQL (Structured Query Language) database engine, a Geographical Information System (GIS) engine that provides comprehensive mapping capabilities, as well as comprehensive decision logic to carry out the functional aspects of the tool set. This document covers the basic installation and operation of the PATH/AWARE tool in order to give the user enough information to start using the tool. A companion user's manual is under development with greater specificity of the PATH/AWARE functionality.

Knowlton, Robert G.↗
DOE OSTI2013

Sample Results From The Interim Salt Disposition Program Macrobatch 7 Tank 21H Qualification Samples

Savannah River National Laboratory (SRNL) analyzed samples from Tank 21H in support of qualification of Macrobatch (Salt Batch) 7 for the Interim Salt Disposition Program (ISDP). An ARP and several ESS tests were also performed. This document reports characterization data on the samples of Tank 21H as well as simulated performance of ARP/MCU. No issues with the projected Salt Batch 7 strategy are identified, other than the presence of visible quantities of dark colored solids. A demonstration of the monosodium titanate (0.2 g/L) removal of strontium and actinides provided acceptable 4 hour average decontamination factors for Pu and Sr of 3.22 and 18.4, respectively. The Four ESS tests also showed acceptable behavior with distribution ratios (D(Cs)) values of 15.96, 57.1, 58.6, and 65.6 for the MCU, cold blend, hot blend, and Next Generation Solvent (NGS), respectively. The predicted value for the MCU solvent was 13.2. Currently, there are no models that would allow a prediction of extraction behavior for the other three solvents. SRNL recommends that a model for predicting extraction behavior for cesium removal for the blended solvent and NGS be developed. While no outstanding issues were noted, the presence of solids in the samples should be investigated in future work. It is possible that the solids may represent a potential reservoir of material (such as potassium) that could have an impact on MCU performance if they were to dissolve back into the feed solution. This salt batch is intended to be the first batch to be processed through MCU entirely using the new NGS-MCU solvent.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Single-faced GRAYQB™: a radiation mapping device

GrayQb{trademark} is a novel technology that has the potential to characterize radioactively contaminated areas such as hot cells, gloveboxes, small and large rooms, hallways, and waste tanks. The goal of GrayQb{trademark} is to speed the process of decontaminating these areas, which reduces worker exposures and promotes ALARA considerations. The device employs Phosphorous Storage Plate (PSP) technology as its primary detector material. PSPs, commonly used for medical applications and non-destructive testing, can be read using a commercially available scanner. The goal of GrayQb{trademark} technology is to locate, quantify, and identify the sources of contamination. The purpose of the work documented in this report was to better characterize the performance of GrayQb{trademark} in its ability to present overlay images of the PSP image and the associated visual image of the location being surveyed. The results presented in this report are overlay images identifying the location of hot spots in both controlled and field environments. The GrayQb{trademark} technology has been mainly tested in a controlled environment with known distances and source characteristics such as specific known radionuclides, dose rates, and strength. The original concept for the GrayQb{trademark} device involved utilizing the six faces of a cube configuration and was designed to be positioned in the center of a contaminated area for 3D mapping. A smaller single-faced GrayQb{trademark}, dubbed GrayQb SF, was designed for the purpose of conducting the characterization testing documented in this report. This lighter 2D version is ideal for applications where entry ports are too small for a deployment of the original GrayQb™ version or where only a single surface is of interest. The shape, size, and weight of these two designs have been carefully modeled to account for most limitations encountered in hot cells, gloveboxes, and contaminated areas. GrayQb{trademark} and GrayQb{trademark} SF share the same fundamental detection system design (e.g., pinhole and PSPs). Therefore, performance tests completed on the single face GrayQB in this report is also applicable to the six- faced GrayQB (e.g., ambient light sensitivity and PSP response). This report details the characterization of the GrayQb{trademark} SF in both an uncontrolled environment; specifically, the Savannah River Site (SRS) Plutonium Fuel Form Facility in Building 235-F (Metallurgical Building) and controlled testing at SRS’s Health Physics Instrument Calibration Facility and SRS’s R&D Engineering Imaging and Radiation Systems Building. In this report, the resulting images from the Calibration Facility were obtained by overlaying the PSP and visual images manually using ImageJ. The resulting images from the Building 235-F tests presented in this report were produced using ImageJ and applying response trends developed from controlled testing results. The GrayQb{trademark} technology has been developed in two main stages at Savannah River National Laboratory (SRNL): 1) the GrayQb{trademark} development was supported by SRNL’s Laboratory Directed Research and Development Program and 2) the GrayQb{trademark} SF development and its testing in Building 235-F were supported by the Office of Deactivation and Decommissioning and Facility Engineering (EM-13), U.S. Department of Energy – Office of Environmental Management.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗