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At least 1,387 records · Page 77
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Views on Long-Term Safety of Deep Borehole Disposal - 16657

The Swedish Nuclear Fuel and Waste Management Co., SKB, has since the end of the 1980s evaluated Deep Borehole Disposal (DBD), of spent nuclear fuel (SNF). This has been done in parallel with the development of a mined repository concept based on the KBS-3V method. The most recent evaluation involved a broad comparison between the KBS-3V method and DBD. The comparison highlights differences between the two methods in the whole chain of the SNF handling including premises for siting, licensing, repository construction, safety during the operational phase, post-closure safety, physical protection, nuclear safeguards, retrieval of the deposited SNF, time planning, need for development, costs and nontechnical project risks. The ambition has been to make an as objective comparison between the two methods as possible, bearing in mind that there are big differences between the two concepts in terms of technical maturity as well as in terms of both quantity and quality of available data. While the primary safety function of the KBS-3V method is long-term containment in a corrosion resistant copper canister, the safety of DBD relies on stagnant groundwater at depth. It is argued in the aforementioned study that this groundwater stagnancy may be jeopardised because of heat and gas evolution in the disposal boreholes, and that deformation of the boreholes make efficient sealing of the deposition boreholes difficult. SKB has taken the position that the DBD method is associated with too many question marks to make it interesting as an alternative method in the Swedish programme. (authors)

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DOE OSTI2018

Improvement of Discharge Regulation System for Liquid and Gaseous Radioactive Materials in Korea - 18050

In this study, we analyzed radioactive waste management practices and discharge regulations for liquid and gaseous radioactive materials generated from the nuclear power plants. Liquid and gaseous radioactive materials are discharged into the environment in a controlled and monitored state after proper treatment. In order to prevent radiological hazards due to the release of radioactive materials, the standards for effluent control limits and off-site dose constraints at the boundary of exclusion area shall be met. In December 2015, the Nuclear Safety Act was partially amended to add the 'Discharge Plan for Liquid and Gaseous Radioactive Materials' to the application form for the power reactor operating licenses. The Discharge Plan shall include the total amount of liquid and gaseous radioactive materials per site, period (annual, quarterly, etc.), and radionuclide group. Therefore, nuclear power plant operators are required to set in advance the total amount of liquid and gaseous radioactive materials for each radionuclide group per site, and keep them below the control limits during the operation of the facilities. Accordingly, it is expected that public trust regarding regulations on effluent discharge as well as effluent control itself can be further improved. (authors)

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DOE OSTI2018

3-D Gamma-ray Mapping from Unmanned Aerial Systems for Nuclear Decommissioning - 18439

Advanced gamma-ray detection technologies integrated with sophisticated robotic platforms will enable safer and more cost and time-efficient capabilities in support of nuclear decommissioning, waste management, and environmental remediation activities within the United States Department of Energy (U.S. DOE) complex and in the global nuclear power industry. In the Applied Nuclear Physics program at Lawrence Berkeley National Laboratory, we have developed the Nuclear Scene Data Fusion (SDF) capability to map gamma-ray emitting sources in large-scale, 3-D environments in real-time. This powerful capability has been successfully deployed in real-world environments, from a test-bed in Berkeley, CA to the exclusion zone in Fukushima Prefecture, Japan to support environmental remediation and contamination mapping efforts within evacuated communities, and is also currently being further developed in support of decommissioning efforts at the Fukushima Dai-ichi Nuclear Power Plant (FDNPP) in collaboration with the Japan Atomic Energy Agency (JAEA). In this work, we will provide an overview of SDF adapted for gamma-ray mapping onboard an UAS, as well as results from measurements demonstrating point and distributed source mapping. (authors)

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DOE OSTI2018

Mercury Recovery at Former Lithium Production Site at the Miramas in France - 18529

The former Lithium production facility located at Miramas in the South-East of France recently completed a 10-year program of process equipment removal, building demolition and soil remediation to recover mercury (Hg) that had been lost during operations at the site. After over 100 years ago chemical production, all production activities stopped in 2009, and immediately afterwards the clean-up project began. The main project focused on the demolition of all facilities and the decontamination of building materials and soil clean-up, with the goal of remediating the site for future industrial purposes. Prior to starting the decommissioning and clean-up work at the site a detailed site survey was conducted to identify the extent of the contamination of the site. After investigations into available technologies, the decision was made to install on-site treatment processes. This resulted in fewer residual environmental effects (trucking pollution, shipping containers, volume of materials, etc), a reduction in logistic complications, and, importantly, was cost effective compared to sending all untreated material to external processing facilities. This decision resulted in a need to optimize demolition waste management, while ensuring the highest clean-up and safety levels. A variety of treatment processes were investigated and compared and two technologies were finally selected and implemented. The first process was a 'thermal desorption unit (TDU),' which heated excavated soils and captured the vaporized mercury. The second was a washing process that was used to remove surface contamination. After the contaminated areas were excavated, materials were crushed, screened and sorted into different sizes, with each pile being routed to the appropriate process. The smaller-sized particles were treated in the TDU. A batch of around 12 MT was fed into a moisture evaporator and then into a mercury evaporator. The soils were then cooled, sampled to ensure no residual Hg contamination before being mixed with a binding agent, and returned to the excavation site. Larger materials with surface contamination were washed to remove the contamination. The liquid effluents were sent to a tank where the contaminated fines were removed, dried, pressed and sent for processing. If the washed, coarse materials failed to reach acceptable mercury levels, they were crushed again and sent to be treated in the TDU. The washing fluid was recycled for the next batch. (author)

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DOE OSTI2019

Enhanced Techniques for Concrete Decontamination During Decommissioning - 19005

As the primary material of construction for nuclear power plant structures, concrete naturally represents the largest fraction of contaminated material by surface area. As such, decontamination of concrete is a major activity during decommissioning projects. As reported in EPRI 3002005412, Characterization and Remediation of Contaminated Concrete, currently used techniques for decontamination of contaminated concrete are generally slow and labor-intensive to apply. This report notes that little change to the techniques used for concrete decontamination have been made during the past 20 years. There is a clear need for improvement to the techniques available for decontamination of concrete. Needed enhancements include substantial improvements in efficiency, waste management, industrial safety, and worker radiation exposure. Potential areas for development include increased automation of existing techniques, improvement in delivery systems for decontamination equipment, and further development of alternative techniques such as microwave and laser scabbling, and nitrogen blasting. To address this need, EPRI is conducting a project to develop enhanced techniques for concrete decontamination. The project focuses on two primary areas: 1. Adaption of existing techniques to improve efficiency through automation and/or altering the delivery systems for the decontamination equipment (for example, through use of exiting robotic systems); and 2. Field demonstration of technologies currently in the engineering development stage, such as nitrogen blasting and laser scabbling. This paper will report the initial results of this EPRI project, which includes identification of promising candidate technologies that will be demonstrated during the next project phase. (authors)

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DOE OSTI2019

Complex Sources and Impacts on Remediation and Monitoring Approaches - 19058

Remedy selection, implementation, and coordination with tank farm activities for adjacent waste sites need to be facilitated by a technically defensible description of sources and their relative strength and temporal contributions to groundwater contamination. Detailed data describing contaminant source releases are typically not available, so source locations are highly uncertain. Furthermore, data on sources and model parameters are not readily available through direct measurements. Instead, source data are often derived by extrapolation from a few direct and/or indirect measurements, while model parameters are commonly estimated using inverse modeling or calibration methods. Because the conceptual model of sources will influence the mass and timing of the flux of contaminants to groundwater, uncertainty in the source terms should be considered. Even though a best-fit model can often be identified, it may not be the best representation of the sources given the sparse, uncertain nature of the data used for estimating source terms. To support future feasibility studies, a systematic approach is needed for identifying remediation and monitoring activities appropriate under conditions of uncertainty. In this study, an integrated assessment of alternative conceptual model elements has been examined using the B-Complex site in the Hanford Central Plateau to demonstrate how multiple sources in a waste management complex can be described to identify their relative contributions to groundwater. The heterogeneous structure of the system is modeled based on available field information on the lithological data for both the vadose zone and aquifer. The approach includes a simulation assessment and linkages to site data, with an uncertainty analysis applied to evaluate the effects of source term uncertainty on contaminant mass flux to groundwater. An integrated assessment of alternative conceptual model elements was evaluated to identify how parameter uncertainty impacts both temporal and spatial uncertainties associated with vadose zone source terms. Modeling scenarios identify important characteristics of remediation target zones to provide context for how surface barriers, in situ remedies, and water-table treatment zones can be designed to account for source dynamics. Ramifications for remediation and monitoring are discussed within the context of the source term uncertainty. The described approach facilitates systematically addressing alternative conceptual model questions in relation to remediation and monitoring and supports the remedial investigation/ feasibility study process and the remediation strategy development for the Hanford Central Plateau. (authors)

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DOE OSTI2019

Design of an Evapotranspiration Surface Barrier for the Subsurface Disposal Area at INL - 19220

At the INL Site, the US DOE is planning to construct one of its largest CERCLA landfill covers. An evapotranspiration surface barrier that will be approximately 65 ha (160 acres) will cover the 39-ha (97-acre) Subsurface Disposal Area at the INL to isolate legacy mixed waste disposed of prior to 1988, low-level radioactive waste disposed of after 1988, and debris from demolition of structures being used to process some of the legacy mixed waste. The evapotranspiration surface barrier will limit percolation to less than 1 cm (0.39 in.) per year under typical precipitation conditions; provide a minimum 3-m (10-ft)-thick cover over all waste to minimize the potential for plant, animal, and inadvertent human intrusion; support growth of native vegetation; and visually blend in with the surrounding desert landscape. A robust design will allow the evapotranspiration surface barrier to resist erosion and perform for a minimum of 1,000 years. The 1-cm (0.39-in.)-per-year and 1,000-year performance objectives were established by a CERCLA record of decision and its implementing remedial design work plan, and these performance objectives are consistent with a performance assessment conducted under US DOE Order 435.1, 'Radioactive Waste Management.' (authors)

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DOE OSTI2019

Frisker Response to Soil Concentrations (Bq/g) Compared to NaI(Tl) - 19338

The Chalk River Laboratories (CRL) began nuclear operations in 1947. Portions of the site land areas have become radiologically contaminated through the 1952 National Reactor Experimental (NRX) accident and/or waste management activities over the past 70 years of site operations. The Environmental Remediation Projects Group (ERP) characterizes, and performs risk assessment, and remedial actions, as required over these contaminated land areas. ERP work activities are subject to controls established by CRL's Radiation Protection (RP) Group such that RP establishes the radiological controls and monitoring requirements for ERP field work activities. The different methods of monitoring soil contamination between ERP and the RP program present some challenges in the interpretation and comparison of the two sets of results and the soil disposition. ERP utilizes NaI(Tl) scintillation detectors and MARSSIM land scanning techniques to measure soil concentrations between sample locations and in excavated materials. Whereas, RP utilizes surface contamination monitoring equipment in the form of a 15.5 cm{sup 2} open area pancake Geiger-Mueller (GM) tube detector to measure soil contamination levels reported in counts per minute (cpm). This paper presents a method to evaluate the theoretical best responses that could be obtained from surface contamination monitoring instrumentation for a known concentration (1 Bq/g) of various radionuclides. A static efficiency of cpm/Bq{sup -1} g{sup -1} is obtained from these calculations. These contamination monitoring sensitivities are then used in a MARSSIM scanning sensitivity equation and compared to the scanning sensitivity of an NaI(Tl) scintillation detector. The results demonstrate that the sensitivity of the frisker probes to particle radiations (betas for pancake) tend to fall in a range of only a few cpm/Bq{sup -1} g{sup -1}. Rather, the primary static response of the pancake probe to soil contamination is from gamma radiation where a value of 36 cpm/Bq{sup -1} g{sup -1} for Cs-137 was obtained at 2.54 cm distance without the application of scanning statistics. In comparison, a 7.6 cm x 7.6 cm NaI(Tl) detector has a Cs-137 static response factor of 10,000 cpm/Bq{sup -1} g{sup -1} at 15 cm distance without applying scanning statistics. (authors)

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DOE OSTI2019

Panel Session 4B: Best Practices in Project Communication Involving Stakeholders

This panel focused on a sustainable, holistic and integrated approach to structure nuclear waste management, disposal and remediation decisions. This panel had participation from internal and external stakeholders and discuss various approaches, tools, and techniques to effectively communicate factors that inform decisions. The panelists presented a summary of their experiences and share key points that were helpful in gaining common understanding with stakeholders. Presentations proposed include: 1 - The Evolution of U.S. Department of Energy and Stakeholder Relationships at the Fernald Preserve, Ohio, Site (Penny Borgman, Navarro Research and Engineering, Inc. (USA)); 2 - The Site Specific Advisory Board Experience - A Members Perspective - 19069 (Belinda Price, Oak Ridge Site Specific Advisory Board (USA)); 3 - Preserving the Mound Laboratory Legacy-The Mound Cold War Discovery Center - 19550 (Susan Smiley, US DOE Office of Legacy Management (USA); Melissa Lutz, Navarro (USA))

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DOE OSTI2019

Panel Session 36: Panel: Young Professionals in Nuclear Science and Engineering an International Perspective

This panel focuses on the young professional views from around the world on radioactive waste management. An informal panel will enable these future leaders of our industry to share ideas and experiences and develop contacts within peer groups on a genuine international level. The informal and interactive panel format will stimulate valuable discussion among the participants and the audience. The panel will be led by young professionals who will actively encourage participation from all those attending. Mentoring will be discussed. This panel aims to encourage fresh thinking and provide an opportunity for an open and frank discussion of issues. Presentations proposed include: 1 - International Youth Nuclear Congress (Alexander Borovskis, AusYGN (Australia)); 2 - AECOM Early Career Professionals Technology Exchange (Natalie Young, Washington River Protection Solutions (USA)); 3 - Young Professionals in Nuclear Science and Engineering - An International Perspective (Shona Thompson, Canadian Nuclear Safety Commission (Canada))

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DOE OSTI2019

Panel Session 62: Progress on Deep Repository Programs Around the World

This panel session focused on the progress of deep repository programs, worldwide. While efforts on a deep geological repository for SNF/UNF and HLW in the USA were terminated in 2010, several other repository programs continue to make considerable progress and deserve to be recognized. Panelists were invited from national, multi-national, and international repository programs and organizations to report upon the status and future as well as the perceived reasons underlying the success and the failure of components of their respective repository program. This panel provided a forum for interested parties to present, exchange, and take advantage of, as appropriate, repository progress, challenges, concepts/designs and other lessons learned around the world, including site characterization, site selection, research and development, licensing, construction, operation, and public acceptance and participation. Presentations proposed include: 1 - Cigeo (French Deep Geological Repository Project) update 2018 (Daniel Delort, Andra (France)); 2 - Progress on Deep Geological Repository Programs Around the World (Bruce McKirdy, Radioactive Waste Management (United Kingdom)); 3 - Geological disposal facility in Belgium (Didier Leonard, ONDRAF/NIRAS (Belgium))

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DOE OSTI2019

Waste Facility Groundwater Monitoring Design Under Uncertainty - 19462

A systematic approach was developed to evaluate current and future monitoring networks designed to detect potential contaminant releases at a number of Dangerous Waste Management Units within the 200 West Area of the Central Plateau at the U.S. Department of Energy's Hanford Site near Richland, Washington. The analyses account for uncertain future flow directions arising from the operation of a high-capacity groundwater pump-and-treat remedy that will operate for the foreseeable future. The developed approach and tools provide a novel means to evaluate and compare the relative efficacy of current or potential monitoring locations in complex and uncertain flow fields without recourse to computationally prohibitive contaminant transport modeling. (authors)

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DOE OSTI2019

Corrosion of Carbon Steel Inside Used Nuclear Fuel Containers - 19492

This paper presents an overview of the carbon steel (CS) corrosion research being conducted by the Radiation Assisted Materials Performance Science (RAMPS) group at Western University, in collaboration with the Canadian Nuclear Waste Management Organization (NWMO) and Canada Natural Science and Engineering Research Council (NSERC). The objective of our studies is to develop a high-fidelity CS corrosion dynamic model that can be used to assess the integrity of the current Canadian used fuel container (UFC) design with confidence. Our key findings, the implications of the findings on corrosion science in general, and the potential consequences of these results on the predictions for long-term UFC integrity are described. Results are presented on the following five themes: chemical kinetics of γ-radiolysis of liquid water and humid air, aqueous corrosion of CS under γ-radiation, corrosion of CS under small water volumes and water filled gaps, Liesegang ring formation in a gel medium, and radiation-induced iron oxide nanoparticle formation from dissolved metal ions. These studies together represent a detailed and integrated examination of the whole CS corrosion system, and in particular, its behaviour under γ-radiation. The key finding of this research is that under certain conditions, the CS corrosion system can exhibit non-linear dynamic behaviours - which can be seen in periodic precipitation patterns and oscillations in measurable system properties such as corrosion potential. This throws into question existing corrosion models that do not account for these types of behaviours. We describe the preliminary steps we have taken in integrating these new understandings into a CS corrosion dynamic model that will be able to make long-term predictions of CS corrosion in the UFC environment. (authors)

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DOE OSTI2019

Radiation Induced Corrosion of Copper in Deep Geological Repositories - 19493

This paper presents an overview of the Cu corrosion research being conducted by the Radiation Assisted Materials Performance Science (RAMPS) group at Western University, in collaboration with the Canadian Nuclear Waste Management Organization (NWMO) and Canada Natural Science and Engineering Research Council (NSERC). This paper presents an overview of our research in three areas to develop a high-fidelity copper corrosion model. Copper is an important candidate material for corrosion protection of used nuclear fuel containers (UFC). The UFC and its environment will be exposed to a continuous flux of γ-radiation. The three areas are chemical kinetics of γ-radiolysis of humid air and groundwater, the effects of solution parameters on Cu corrosion dynamics without γ-radiation and the effects of γ-radiation on Cu corrosion dynamics. These studies show that because of the strong feedback loops between processes occurring during copper corrosion, the overall corrosion mechanism may not follow linear dynamics. Presented here are our key findings, their implications for corrosion science in general, and the potential implications of these results for predictions of the adequacy of a UFC copper corrosion allowance. (authors)

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DOE OSTI2019

Remote Technology to Retrieve Waste Below the 324 Building at the Hanford Site - 19622

The Waste Technology Engineering Laboratory (324 Building) is a Category 2 Nuclear Facility at Hanford. The 324 Building Disposition Project is responsible for the removal and remedial actions under CERCLA. Additionally, there are RCRA Dangerous Waste Management Units which the CERCLA actions are presumed to satisfy. The Radiochemical Engineering Complex (REC) is a group of four hot cells (A, B, C and D Cells) with an interconnecting airlock inside of the 324 Building. Underneath the REC B-Cell are contaminated soils that have an estimated 12,793 R/Hr dose rate and remediation is under CERCLA 300-FF-2 Operable Unit. The current action consists of debris removal from the floor of B-Cell, B-Cell floor removal, and excavation of the soils beneath the floor to a depth of up to twelve feet with four Remote Excavator Arms (REA). This paper provides insight into the preparation and performance of high tolerance core drilling into a hot cell complex. Installation of the REA Through Support (REA TS) mechanism through the 4'-6' concrete shield walls is a complex series of tasks with demanding tolerances at each of four locations around the perimeter of B-Cell. The use of training mock ups was included in the completion of this work. The REA TS is two 11 inch diameter high strength steel sleeves that have expandable collars at each end allowing the interstitial space between the outside of the sleeve and inside of the core drill to be grouted to form a substantial structural connection, and a 10 inch pipe with hydraulic hoses to power the REA. There are three cores required for each REA location. The REAs themselves are customized excavator arms, mounted on support columns to be used in the excavation and transfer of the contaminated soil into storage facilities. The tasks for the installation of the REA TSs are: 1) Characterization of B-Cell Cubicles. 2) Embedded Pipe Fill. 3) Interference Removal. a. Internal Walls, b. External Walls. 4) Core Drilling. 5) Go/No-Go gauge, and video inspection. 6) Installation of REA Through Support Assembly. The 324 Building REC Hot Cells have internal, external, and embedded utilities throughout the shield walls in conflict with the REA TS locations. Some of the embedded utilities have connection points inside of the B-Cell Cubicles. There were two separate issues at the conflicted locations. The first issue was the core drill cooling water would flow along the embedded utility potentially spreading legacy chemical and radiological contamination to areas of the facility that was unwanted. The second issue was that the REA TS required a substantial structural connection to the B-Cell wall, the cored diameter is larger than the REA TS and these interstitial areas required a full grouted connection. The embedded utilities posed a hazard that if unfilled, the interstitial grout would flow into the embedded utility and destabilize the required grouted structural connection. The B-Cell Cubicles had not been opened in over 10 years. A mockup was completed to install a glovebag over the cubicle opening and then to verify functionality and port requirements. The embedded utility pipe connections in the cubicles were then opened inside a glovebag to verify the correct connection points, connection method for future embedded pipe fill material delivery system, radiological characterization, and air flow to show that the embedded utility pipes were not blocked and would not pressurize. The embedded pipe fill material and grout system was mocked up and run several times to verify methodology and functionality. The B-Cell Cubicles were then reentered and the embedded utility pipe fill was completed using a hand pump. External interferences at the core drill locations were removed using standard interference removal techniques. Internal interferences at the core drill locations were removed using a remote cutting tool that was controlled through the use of the B-Cell crane, remote cameras and remote lighting. The core drilling was mocked up two separate times, the project's purpose built 'Mock Up Facility' was core drilled to verify tolerances and methodology prior to subcontract award, the core drill subcontractor built a 'dummy wall' to train and practice on achieving the demanding tolerances, coordinated alignments between the cores, short term and long term dose shielding techniques, and contamination control techniques. A core drill template, and additional anchoring techniques were used to maximize accuracy and precision of the REA TS core drills. A 'Go/No-Go' gauge was built and installed at each of the four REA TS locations to verify that the core drill met tolerances. The core drill locations were visually inspected using a boroscope. The REA TS installation was mocked up at the projects 'Mock Up Facility' beginning with basic hazard controls, then with intermediate hazard controls, and finally with full occupational safety, industrial hygiene, and radiological controls. The REA TSs were installed and grouted into place. (authors)

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DOE OSTI2020

2019 Public Debate on the French Waste Plan - 20181

For the first time, a public debate has preceded the preparation of the National Radioactive Materials and Waste Management Plan (PNGMDR). This debate, which took place over 5 months from 17 April to 25 September, was held by decision of the French National Public Debate Commission (CNDP), an independent administrative authority responsible for public participation in major projects and plans/programmes with environmental impacts. To organise and lead the debate, the CNDP appointed a Special Commission. (authors)

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DOE OSTI2020

Proposed Analytical Methods for Determining Filter Media Properties

High Efficiency Particulate Air (HEPA) filters, commonly used in nuclear filtration applications, are an integral part of the waste management processes in nuclear plants. HEPA filters are 99.97% efficient filtration devices characterized by their high resistance to air flow, or pressure drop. From theoretical models, the initial pressure drop across a clean filter is proven to be a function of the filter media fiber diameters and porosity of the media. These filter properties are relatively difficult to obtain with traditional manual methods; therefore, there is a need to develop analytical methods to find the fiber diameter and porosity to determine the pressure drop. Typically, these values can be found by substituting a calculated representative value based upon measured values, such as equivalent fiber diameter based upon a clean pressure drop. However, these values may also be directly recorded and measured by incorporating Scanning Electron Microscope (SEM) image analysis and other measurement methods to analyze the filter media and determine its physical properties without the need for media testing. DiameterJ, an open source Java plug-in, used with ImageJ, can process images of filter media taken by an SEM to find statistical data such as mean fiber diameter and porosity. To produce the raw data, SEM images of two filter media type samples are taken and segmented in DiameterJ using the traditional and statistical region merging segmentation algorithms. Manual segmentation is necessary after the initial segmentation by the algorithms as the images tend to be too complex for the algorithms to output with the necessary accuracy. However, complications exist in the manual segmentation process as these methods can be time intensive and prone to the individual bias of the user. This in turn can skew the final mean fiber diameter result, and lead to either an over or under prediction of the pressure drop. It was also discovered that the porosity data produced by DiameterJ is inaccurate, as the SEM analyzes the three-dimensional filter media by projecting its geometry onto a plane and analyzing it as a two-dimensional binary image. Thus, the porosity is artificially inflated through the segmentation process, rendering this result incorrect. Alternatively, density determination, gravimetric analysis, and thickness testing of the filter media is collectively used to determine the filter fiber porosity. Together, the SEM image analysis and analytical lab methods produce results through direct measurements which allow for the prediction of the initial pressure drop from clean filter media without the need for prior media testing to collect pressure data. By improving upon this proposed analytical method in the future, there is potential to streamline the process of finding these filter properties into a more direct methodology for determining the pressure drop across HEPA filters.

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DOE OSTI2020

Remote Radiation Sensing Using Aerial and Ground Platforms

Remote sensing of ionizing radiation has a significant role in waste management, nuclear material management and nonproliferation, and radiation safety. Robotic platforms can surpass the number of tasks that are achieved by humans. With this technique, the operator's radiation exposure can be decreased. Remote sensing allows for the evaluation and monitoring of radiological contamination. Gamma-ray and neutron sensors were integrated onto the robotic platforms. This approach allows for the radiation sensor data to be dynamically tracked and mapped thus enabling further analysis of the radiation flux in temporal and spatial domains. The goal is to complete scheduled tasks while the robot is being irradiated. To achieve this, electronic components must be shielded and radiation hardened. CZT Detector: Cadmium Zinc Telluride (CZT) detector technology has been a promising solution for gamma-ray and x-ray measurements. Detector data is transferred to the Odroid minicomputer that controls and powers the module via the USB. Robot Operating System (ROS) was utilized for data acquisition and data fusion. The Mariscotti method was employed for the spectrum analysis. A function was programmed in ROS for the automatic identification of photopeaks. CLYC Detector: A Cs{sub 2}LiYCl{sub 6}:Ce{sup 3+} (CLYC) detector was used for simultaneous medium-resolution gamma-ray measurements and neutron counting. A 2.54 cm diameter photomultiplier tube (PMT) was equipped with a high voltage supply and a miniature digitizer. Gamma-ray excitation: fast core-to-valence luminescence (CVL) with 1 ns decay constant, and prompt Ce{sup 3+} emission with 50 ns decay constant. Neutron excitation: slow cerium self-trapped excitation (Ce{sup 3+} STE), 1000 ns decay constant. Radiation Source Localization: Maximum Likelihood Estimation (MLE) and gradient-based methods were used to locate the position of a radiation source based on measured radiation intensities. Multi-Particle Transport Code FLUKA: Estimation of radiation damage of the electronic components is important in order to optimize the robot's operational time while it is irradiated. Displacement per atom (DPA) represents the radiation damage in materials exposed to the ionizing radiation. Various shielding layers of different thickness t were analyzed (< 5% statistical error). The model of the controller of the UAS was designed in FLUKA. Conclusion: CZT and CLYC detectors were integrated onto the robotic platforms. Radiation source localization and contour mapping using robotic platforms were studied. Functions for data analysis and fusion were developed in ROS. FLUKA code was utilized to analyze DPA values. Layers of low-density and high-density materials were used to shield the UAS electronics.

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