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At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Performance Assessment of a New Gamma Spectroscopy Toolkit: Peak Doctor and the Spectral Nondestructive Assay Platform (SNAP) - 17397

High resolution gamma spectroscopy (HRGS) is a widely used technique for non-destructive characterization of nuclear materials, and assay of radioactive waste. The ability to accurately evaluate raw spectral data is an important step in producing high-quality results. A spectral analysis application known as Peak Doctor has been developed under a Cooperative Research and Development Agreement (CRADA) between LANL and Pajarito Scientific Corporation (PSC). The software has undergone several years of testing and validation by the Radioactive Waste Characterization Team of the LANL Waste Management Services Group (WM-SVS). PeakDoctor is designed to work with the Spectral Nondestructive Assay Platform (SNAP{sup TM}) to provide a complete HRGS analysis engine. SNAP{sup TM} performs model specific corrections for attenuation and geometry. These applications are the major components of a new toolkit designed primarily for portable far-field non-destructive assay. Peak Doctor performs a very accurate peak fit analysis with the goal of improving upon the capabilities of existing commercially available routines. It is designed to be user-configurable and can be customized to any given type of detector and any given type of spectrum. Peak Doctor was developed to resolve many of the inherent complexities in gamma ray spectra including overlapping peaks, complex multiplet regions, large step peaks, backscatter regions and Compton edges. Unlike most spectral analysis platforms, Peak Doctor does not rely on the energy calibration and detector resolution parameters that are saved by the operator in the spectrum file. Instead, key parameters are pre-defined for individual detectors. The software fits a continuum using a series of splines between a series of junctions or 'knots'. Fitting techniques are used to join the splines together to create a near-continuous multi-element curve. Optimizing the location of the knots is one of Peak Doctor's most powerful features. The knots are initially configured using a pre-defined list and are removed or added automatically during the fitting process. Knots are also created when new features are identified such as Step Peaks and Compton Edges. The user can also manually specify knots in order to improve the continuum fit. The initial fit is improved on by removing any outlier regions (which are converted to peaks) and the continuum fitting process is repeated. At this stage, an experienced user can work interactively with the fit by manually converting peak regions to continuum or vice versa. The final step is to fit a series of peaks above the continuum based on detector parameters. Single and multiplet peaks are then processed by either the tailing (piecewise) or non-tailing (Gaussian) Levenberg Marquardt non-linear peak fitting routine depending on the user defined selection. Only a rough knowledge of the detector characteristics (energy calibration and resolution) is required for the fitting process to proceed with good accuracy. Any peaks that are missed can be manually added by the user and will also be fitted. Discontinuities in the peak / continuum fit can be rapidly identified by reviewing the 'residuals scale' plot. For waste streams that are well characterized, it is often the case that routinely encountered spectra will have similar peaks and continuum features. For these cases a handy 'One Step' method is available that proceeds in an automated fashion without user interaction. Peak Doctor has been configured with a set of default spectrum types including plutonium / transuranic waste (including Weapons Grade Pu, Heat Source Pu, Am-241, Am-243 and Np-237), depleted or enriched uranium, nuclear power plant / intermediate level waste (with fission and activation products such as Cs-137 and Co-60) as well as Naturally Occurring Radioactive Materials (NORM). The user can easily add new spectrum types for specific waste streams or other types of radiological / special nuclear materials. The output from PeakDoctor (including net peak area, centroid energy and a list of minimum detectable activity levels at pre-defined regions) is used by SNAP{sup TM} to complete the data analysis process. Isotope activities are calculated using configurable template-driven geometry and attenuation correction methods. The software accounts for matrix and container wall effects, collimator / angular corrections, detector efficiency (based on calibration curves) and stand-off distance. SNAP{sup TM} uses published algorithms incorporating a series of decision points during its operation to allow the analyst to provide logical input to guide the software during the analysis of each unique spectrum. The Peak Doctor / SNAP{sup TM} toolkit has been extensively tested and validated under a variety of conditions using 'benchmark' reference spectra. Test scenarios reported include HPGe-based measurements of well characterized uranium and plutonium standards and radioactive sources with a variety of different geometries and matrix materials including surrogate drums and boxes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Draft Summary Report on Irradiated Fuel Handling and Management for LOTUS

The National Reactor Innovation Center (NRIC) has conceptualized the design of the Laboratory for Operation and Testing in the United States (LOTUS) test bed to provide the United States Department of Energy (DOE) with the infrastructure necessary to make advanced reactor designs available for commercial developers. LOTUS will provide a test bed to developers with the capabilities of supporting a wide range of experiment design possibilities. Upon completion of the developers’ operations and experiments within the test bed, the irradiated fuel, reactor components, and other experiment materials must be removed. Idaho National Laboratory (INL) possesses significant capabilities for radioactive material handling such as casks, carts, and forklifts. However, given the unique environment presented by the NRIC-LOTUS test bed, located inside the Zero Power Physics Reactor (ZPPR) Perimeter Intrusion Detection and Assessment System (PIDAS) area at the Materials and Fuels Complex (MFC) and the complexity of novel removal activities of recently operated reactor experiments through the new proposed access tunnel. The efficacy was not apparent for existing equipment to provide all the needed capability. To bridge the potential gaps in cask designs, storage, and transportation, NRIC requested the development of trade studies for the transfer, handling, and storage requirements of irradiated fuel salts and other radioactive materials. NRIC directed Boston Government Services, LLC (BGS) to perform the trade studies and develop a report analyzing alternatives. In addition to the BGS reports, the Idaho National Lab’s (INL), provided by the first potential user’s Advanced Reactor Development (ARD) team, prepared a feasibility study for the storage of specific irradiated fuel within the existing ZPPR vault.This summary report is intended to present the trade studies, options, and alternatives that were investigated. The maturity level of LOTUS, the reactor, fuel salt containers, gloveboxes, and reactor testing campaign and concept of operations were not at a level sufficient to base critical decisions on. This report is not intended to present a final recommendation. The final recommendations for fuel storage location, transport, handling equipment, and operations will be made in FY 2024 and will be based on known materials, test campaign requirements, funding, and final analysis of the fuel and equipment to be used.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Safety aspects of dry spent fuel storage and spent fuel management

As with the storage of all radioactive materials, the storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Safety aspects like safe enclosure of radioactive materials, safe removal of decay heat, nuclear criticality safety and avoidance of unnecessary radiation exposure must be achieved throughout the storage period. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. After the events of Fukushima, the advantages of passively and inherently safe dry storage systems have become more obvious. TUV and BAM, who work as independent experts for the competent authorities, present the licensing process for sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel, based on their long experience in these fields. All safety relevant issues like safe enclosure, shielding, removal of the decay heat or behavior of cask and building under accident conditions are checked and validated with state-of-the-art methods and computer codes before the license approval. It is shown how dry storage systems can ensure the compliance with the mentioned safety criteria over a long storage period. Exemplarily, the process of licensing, erection and operation of selected German dry storage facilities is presented. (authors)

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

Opportunities for Cost Effective Disposal of Radioactively Contaminated Solid Waste on the Oak Ridge Reservation, Oak Ridge, TN - 13045

Department of Energy (DOE) requirements for the release of non-real property, including solid waste, containing low levels of residual radioactive materials are specified in DOE Order 458.1 and associated guidance. Authorized limits have been approved under the requirements of DOE Order 5400.5, predecessor to DOE Order 458.1, to permit disposal of solid waste containing low levels of residual radioactive materials at solid waste landfills located within the DOE Oak Ridge Reservation (ORR). Specifically, volumetric concentration limits for disposal of solid waste at Industrial Landfill V and at Construction/Demolition Landfill VII were established in 2003 and 2007, respectively, based on the requirements in effect at that time, which included: an evaluation to ensure that radiation doses to the public would not exceed 25 mrem/year and would be as low as reasonably achievable (ALARA), with a goal of a few mrem/year or less (in fact, these authorized limits actually were derived to meet a dose constraint of 1 mrem/year); an evaluation of compliance with groundwater protection requirements; and reasonable assurance that the proposed disposal is not likely to result in a future requirement for remediation of the landfill. Prior to approval as DOE authorized limits, these volumetric concentration limits were coordinated with the Tennessee Department of Environment and Conservation (TDEC) and documented in a Memorandum of Understanding (MOU) between the TDEC Division of Radiological Health and the TDEC Division of Solid Waste Management. These limits apply to the disposal of soil and debris waste generated from construction, maintenance, environmental restoration, and decontamination and decommissioning (D and D) activities on the DOE Oak Ridge Reservation. The approved site-specific authorized limits were incorporated in the URS/CH2M Oak Ridge LLC (UCOR) waste profile system that authorizes disposal of special wastes at either of the RCRA Subtitle D landfills. However, a recent DOE assessment found that implementation of the site-specific authorized limits for volumetrically contaminated waste was potentially limited due in part to confusion regarding the applicability of volumetric concentration limits and/or surface activity limits to specific wastes. This paper describes recent efforts to update the authorized limits for Industrial Landfill V and Construction/Demolition Landfill VII and to improve the procedures for implementation of these criteria. The approved authorized limits have been evaluated and confirmed to meet the current requirements of DOE Order 458.1, which superseded DOE Order 5400.5 in February 2011. In addition, volumetric concentration limits have been developed for additional radionuclides, and site-specific authorized limits for wastes with surface contamination have been developed. Implementing procedures have been revised to clarify the applicability of volumetric concentration limits and surface activity limits, and to allow the use of non-destructive waste characterization methods. These changes have been designed to promote improved utilization of available disposal capacity of the onsite disposal facilities within the DOE Oak Ridge Reservation. In addition, these changes serve to bring the waste acceptance requirements at these DOE onsite landfills into greater consistency with the requirements for commercial/ public landfills under the TDEC Bulk Survey for Release (BSFR) program, including two public RCRA Subtitle D landfills in close proximity to the DOE Oak Ridge Reservation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Remote monitoring of Alpha Gamma hot cell facility by ARG-US RFID system - 15177

Argonne National Laboratory (Argonne) has developed a monitoring and tracking system for nuclear materials based on radio frequency identification (RFID) technology. The system, called ARG-US ('Watchful Guardian'), has been in fielding testing and applications at selected U.S. Department of Energy (DOE) sites since 2010. ARG-US utilizes battery-powered tags to remotely and continuously monitor the environmental parameters, e.g., radiation levels, of the tagged items such as containers of radioactive materials. To demonstrate that the system can monitor nuclear and radiological facilities as well, a compact system was installed in the Alpha Gamma Hot Cell Facility (AGHCF) at Argonne in 2013. AGHCF is a Hazard Category 2 non-reactor nuclear facility that is being decommissioned. It is also being used to support de-inventory of other facilities on the Argonne site. As such, large quantities of radioactive materials and wastes are being discharged from the hot cell, loaded into transport containers, and shipped away. The deployed ARG-US system monitors worker areas, material transfer areas, cell exhaust HEPA filters, and?on a rover/robot?in-cell locations of AGHCF. Thus far, the system is performing as designed and yielding results that can be corroborated by existing surveillance means. The in-cell unit is particularly valuable, as it yields information that would otherwise be hard to obtain. In time, the benefits of automated monitoring 24/7 by ARG-US should become apparent and may lead to a fuller system with more distributed tags in the hot cell areas until the facility is fully decommissioned. (authors)

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

On the Alpha Radiation Damage to the O-Ring Seal of the SAVY-4000 Container

In this memo, the dose to an O-ring from the alpha decay of 238 PuO 2 is calculated. These calculations were made in order to assess possible radiation damage to the O-ring, which could lead to failure and loss of integrity of the seal created by the O-ring. The O-ring being discussed here provides the seal between the lid and bottom of a nuclear materials container. The container will be used for the storage and transport of a number of different radioactive materials. The analysis is made using 238 PuO 2 as it would create the greatest damage to the O-ring when considering the materials that may be stored in it. Contamination of the O-ring is not expected to occur. But in the event radioactive material comes in contact with the O-ring, these calculations provide an estimate of the radiation damage.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Nanocomposite Materials for Radionuclide Sequestration from Groundwater Environments

The half-lives of radionuclides range from fractions of a second to billions of years. Since no practical method of altering radioactive decay exists, and since exposure to either the energy emitted from radioactive decay or chemical properties of radionuclides poses dire health risks, radioactive materials must be segregated and controlled. The capture, treatment, and disposition of radioactive materials remain an extraordinary challenge. In here, we focus our attention on the synthesis and characterization of a unique class of nanocomposite materials that have potential for removal of radionuclide contamination. Specifically, we report a simple approach to decorate the surface of iron-based (Fe/FexOy) material with various nano-catalysts. Specifically, copper (Cu), tin (Sn), and silver (Ag) nanoparticles were prepared through two different reduction approaches, namely, citrate and cetyltrimethylammonium bromide (CTAB) methods, on the iron-based material surface. All samples were characterized by a variety of analytical tools, which included scanning electron microscopy (SEM), electron-dispersive X-ray microanalysis (EDS), and EDS mapping to elucidate materials’ morphology as well as nano-catalysts’ loading and location on the iron-based structures.

Hunyadi Murph, Simona E.↗
DOE OSTI2020

Accelerating Nuclear Facility Mission Completion Through an Improved Chemical Hazard Control Strategy - 20460

Code of Federal Regulation (CFR) 10 CFR 830, Nuclear Safety Management [1], explicitly identifies hazardous materials in addition to radioactive materials when establishing the scope of materials that must be evaluated in Hazard Category (HC) 1, 2, and 3 nuclear facilities. This inclusion has led to a near decade-long progression of more and more rigorous application of chemical hazard controls in these facilities. The progressive addition of more rigorous chemical hazard controls has occurred across the Department of Energy (DOE) Complex, having been driven by contractors and regulators alike. However, chemical hazards in other DOE facilities have been, and continue to be, controlled using traditional industrial safety centered strategies. Thus, identical chemical hazards are controlled using dramatically different approaches based solely on the facility type. Not only are the controls different, the analysis and derivation rigor and requirements are also significantly different. In particular, chemical control derivation and application in non-nuclear facilities are much less onerous than the comprehensive hazard analysis process used in nuclear facilities, even in circumstances in which chemicals are not initiators to, or do not exacerbate a nuclear event. While the derivation and control approaches used in these two types of facilities are disparate, the safety results are similarly effective based on a review of reported chemical events from 2014 through 2018. While both approaches achieve comparable safety outcomes, the approach used in nuclear facilities results in significantly greater preparation costs and implementation costs. Furthermore, the problem with the differing approach is not relegated to the control preparation and implementation burden. The operational impact of potentially unnecessary controls slows mission completion without a corresponding increase in safety, not to mention the dilution of the importance of Technical Safety Requirements. The proposed solution must first include a critical and thorough examination of DOE Complex and top-tier chemical industry chemical hazard control practices, coupled with a similar examination of the driving facility regulatory requirements. Results from the examination can then be used to fuel a bold departure from current chemical control practices in Hazard Category 1, 2, and 3 nuclear facilities. Development of an improved and consistent chemical hazard control process would follow, one that continues to ensure safety while accelerating mission completion and significantly reducing the burden associated with chemical hazard control development and implementation. The improved process must establish a hierarchy where radiological hazards are addressed using safety class and safety significant controls and chemical hazards are addressed using overarching safety management programs (with limited unique exceptions). The improved Hazard Category 1,2, and 3 nuclear facility chemical hazard control strategy, while compliant with 10 CFR 830 [1], would be founded on the requirements in 10 CFR 851, Worker Safety and Health Program [2], and the complementary DOE O 151.1D, Comprehensive Emergency Management System [3]. In this way, the Hazard Category 1, 2, and 3 nuclear facility chemical hazard control strategy would, wherever appropriate, be the same process used at other DOE Complex facilities. Ultimately, the improved process will maintain safety and significantly reduce costs and life cycle risk through faster mission completion. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Development of a Course in Gamma Ray Spectrometry

The majority of undergraduate and graduate students in nuclear science and engineering take a general course in nuclear instrumentation where one of the topics is typically in gamma ray spectrometry. Rarely are courses offered solely on this subject matter other than intensified workshops conducted by the International Atomic Energy Agency (IAEA) or private companies which can be expensive especially for large groups of students. Currently there are many world-wide projects in routine monitoring and advanced research that employ gamma rays. Some but not an exhaustive list of these topics include neutron activation analysis, naturally occurring radioactivity materials (NORM), nuclear forensics, nuclear security, in-situ gamma ray detection, galactic gamma ray detection, health physics and advanced instrumentation. At the Nuclear Engineering Teaching Laboratory at the University of Texas at Austin, many of our students are involved in an array of research topics that involve gamma ray spectrometry in their Masters of Science theses, PhD dissertations and special topics courses. It was decided to invest the time and effort to develop a course with a strong technical component to complement their research activities. A detailed stand-alone course in gamma ray spectrometry that includes a series lectures in modular form has been developed. As well we have incorporated fourteen advanced animations to better explain some of the fundamental physical processes and experimental set-ups. Six laboratory sessions and multiple homework assignments have also been established. This course is applicable to senior undergraduate students, graduate students, and novice researchers the above mentioned fields in a traditional class/laboratory setting or for intense one to two week workshops. A great deal of effort was undertaken to prepare a stand-alone course in gamma ray spectrometry which includes many of the fundamental aspects as well as the more advanced instrumentation. Animations of gamma ray interactions and experimental set ups were also developed. Six laboratories and six assignments were also conceived. Every topic has its own module thus a 'pick and choose' subset of lectures can be given to different levels of understanding. This course is applicable to senior undergraduate students, graduate students, novice researchers in neutron activation analysis and naturally occurring radioactive material, and for intense one to two week workshops. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Assessing potential induced radioactivity in materials processed with X-ray energy above 5 MeV: Assessment protocols and practical experience

In accordance with ISO11137-1 section 5.1.2, ‘the potential for induced radioactivity in product shall be assessed’. This article describes how compliance to this requirement may be achieved using qualified test methods. Materials of consideration are conceptually discussed. Results of testing conducted on products processed with a 7.5 MeV X-ray irradiation process are provided. As X-ray becomes more widely used in healthcare sterilization, having standard assessment protocols for activation coupled with a shared database of material test results will benefit all healthcare product manufacturers seeking to avail of this innovative technology.

Kroc, Thomas K.↗
DOE OSTI2022

Traditional Forensic Examinations on Bulk Special Nuclear Material

The Federal Bureau of Investigation (FBI) Laboratory at Quantico, Virginia is responsible for the forensic examination of radiological evidence and evidence contaminated with radioactive materials. Due to these unique hazards and the necessary specialized safety protocols, the FBI has developed a network of Partner Laboratories (PLs) across the United States of America to support conventional forensic examinations of radiological evidence and evidence contaminated with radioactive materials. In support of this, the FBI Laboratory has established the Hazardous Evidence Analysis Team (HEAT), a group of qualified forensic examiners, scientists, technicians, and photographers who can deploy to laboratories outside of the FBI Laboratory at Quantico, VA to perform conventional forensics on evidence containing/contaminated with hazardous materials. One of the FBI’s PLs, the United States Department of Energy’s Los Alamos National Laboratory (LANL), in Los Alamos, New Mexico, has unique facilities, personnel, and procedures to secure, safely handle, and process significant quantities of special nuclear material (SNM). While LANL’s procedures for working with SNM are well established, working with SNM under evidentiary controls is not. Close cooperation between LANL and FBI HEAT is required due to FBI policies on evidence handling, as well as the challenges associated with the exploitation and preservation of conventional forensics (fingerprints, trace evidence, and photography) on SNM evidence.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2019

Radioactive Waste Minimization - Clearance for Reuse, Recycling and Disposal - 19598

Virtually all activities related to radioactive materials and to the utilization of nuclear energy generate radioactive waste. Due to limited storage facilities and waning public acceptance for long-term disposal of radioactive waste as well as the associated cost, implementation of strategies and methods for its effective reduction are mandatory. Clearance of materials is one possible concept to reach this goal. The International Atomic Energy Agency (IAEA) defines the process of clearance as 'the removal of radioactive materials or radioactive objects within authorized practices from any further regulatory control by the regulatory body.' Thus, these materials are no further treated as radioactive waste and may be recycled or disposed of conventionally. In quantitative terms, clearance is generally related to the '10 μSv-Concept' This means that the effective dose expected to be incurred by any member of the public due to cleared materials is of the order of 10 μSv or less in a year. A combination of appropriate clearance concepts, methods and measurement techniques, following an integrated clearance strategy for a project, allows the return of up to 97 % of the mass of a decommissioned NPP to the conventional waste cycle. This includes the concept of clearance for a specific purpose, often referred to as 'conditional clearance', i.e. the case where a certain condition has to be fulfilled before the act of clearance has been completed. As a technical expert organization, TUV NORD EnSys accompanied the development, implementation and execution of clearance procedures in more than 13 nuclear power-plants and installations, either operational or in various states of decommissioning. Thus, TUV Nord EnSys has abundant experience in and a comprehensive overview on the implementation of different clearance strategies and procedures. Considering the increased waste-volumes and additional waste-streams to be handled during decommissioning, the transition to a decommission-focused clearance system is of vital importance for successful waste-minimization. This paper provides a view on the perspective of a technical support organisation (TSO) towards implementation of a holistic clearance strategy into the decommissioning concept for radioactive waste minimization. (authors)

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

Radiological issues associated with the recent boom in oil and gas hydraulic fracturing

As the worldwide hydraulic fracturing 'fracking' market continued to grow to an estimated $37 Billion in 2012, the need to understand and manage radiological issues associated with fracking is becoming imperative. Fracking is a technique that injects pressurized fluid into rock layer to propagate fractures that allows natural gas and other petroleum products to be more easily extracted. Radioactivity is associated with fracking in two ways. Radioactive tracers are frequently a component of the injection fluid used to determine the injection profile and locations of fractures. Second, because there are naturally-occurring radioactive materials (NORM) in the media surrounding and containing oil and gas deposits, the process of fracking can dislodge radioactive materials and transport them to the surface in the wastewater and gases. Treatment of the wastewater to remove heavy metals and other contaminates can concentrate the NORM into technologically-enhanced NORM (TENORM). Regulations to classify, transport, and dispose of the TENORM and other radioactive waste can be complicated and cumbersome and vary widely in the international community and even between states/provinces. In many cases, regulations on NORM and TENORM do not even exist. Public scrutiny and regulator pressure will only continue to increase as the world demands on oil and gas continue to rise and greater quantities of TENORM materials are produced. Industry experts, health physicists, regulators, and public communities must work together to understand and manage radiological issues to ensure reasonable and effective regulations protective of the public, environment, and worker safety and health are implemented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Data Fusion for a Vision-Radiological System for Source Tracking and Discovery

A multidisciplinary approach to allow the tracking of the movement of radioactive sources by fusing data from multiple radiological and visual sensors is under development. The goal is to improve the ability to detect, locate, track and identify nuclear/radiological threats. The key concept is that such widely available visual and depth sensors can impact radiological detection, since the intensity fall-off in the count rate can be correlated to movement in three dimensions. To enable this, we pose an important question; what is the right combination of sensing modalities and vision algorithms that can best compliment a radiological sensor, for the purpose of detection and tracking of radioactive material? Similarly what is the best radiation detection methods and unfolding algorithms suited for data fusion with tracking data? Data fusion of multi-sensor data for radiation detection have seen some interesting developments lately. Significant examples include intelligent radiation sensor systems (IRSS), which are based on larger numbers of distributed similar or identical radiation sensors coupled with position data for network capable to detect and locate radiation source. Other developments are gamma-ray imaging systems based on Compton scatter in segmented detector arrays. Similar developments using coded apertures or scatter cameras for neutrons have recently occurred. The main limitation of such systems is not so much in their capability but rather in their complexity and cost which is prohibitive for large scale deployment. Presented here is a fusion system based on simple, low-cost computer vision and radiological sensors for tracking of multiple objects and identifying potential radiological materials being transported or shipped. The main focus of this work is the development on two separate calibration algorithms for characterizing the fused sensor system. The deviation from a simple inverse square-root fall-off of radiation intensity is explored and accounted for. In particular, the computer vision system enables a map of distance-dependence of the sources being tracked. Infrared, laser or stereoscopic vision sensors are all options for computer-vision implementation depending on interior vs exterior deployment, resolution desired and other factors. Similarly the radiation sensors will be focused on gamma-ray or neutron detection due to the long travel length and ability to penetrate even moderate shielding. There is a significant difference between the vision sensors and radiation sensors in the way the 'source' or signals are generated. A vision sensor needs an external light-source to illuminate the object and then detects the re-emitted illumination (or lack thereof). However, for a radiation detector, the radioactive material is the source itself. The only exception to this is the field of active interrogations where radiation is beamed into a material to entice new/additional radiation emission beyond what the material would emit spontaneously. The aspect of the nuclear material being the source itself means that all other objects in the environment are 'illuminated' or irradiated by the source. Most radiation will readily penetrate regular material, scatter in new directions or be absorbed. Thus if a radiation source is located near a larger object that object will in turn scatter some radiation that was initially emitted in a direction other than the direction of the radiation detector, this can add to the count rate that is observed. The effect of these scatter is a deviation from the traditional distance dependence of the radiation signal and is a key challenge that needs a combined system calibration solution and algorithms. Thus both an algebraic approach as well as a statistical approach have been developed and independently evaluated to investigate the sensitivity to this deviation from the simplified radiation fall-off as a function of distance. The resulting calibrated system algorithms are used and demonstrated in various laboratory scenarios, and later in realistic tracking scenarios. The selection and testing of radiological and computer-vision sensors for the additional specific scenarios will be the subject of ongoing and future work. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Long-Term Stewardship Optimization for a Decommissioned Nuclear Reactor - 19579

The decommissioned Hallam Nuclear Power Facility (Hallam), located in Lancaster County, Nebraska, approximately 19 miles south of Lincoln, was developed as a prototype to investigate the technical and economic feasibility of a sodium-cooled, graphite-moderated nuclear reactor. As part of its Power Demonstration Program, the US Atomic Energy Commission (US AEC), a predecessor agency of the US Department of Energy (US DOE), operated the Hallam reactor from 1962 to 1964. The radiological surveillance of the Hallam site was established when the operating agreement between AEC and the Nebraska Public Power District was terminated. The US DOE Chicago Operations Office and the Nebraska Department of Health (NDH) agreed to install a shallow groundwater monitoring system as part of the environmental surveillance program. Surveillance at the site was scheduled to end in fiscal year 2005, with no further activities planned after that. At that time, the plan was to transition the site to the Nebraska Public Power District with future use of the facility to remain restricted. US DOE has title to and responsibility for the entombed radioactive materials. A surveillance and monitoring program was initiated in 1970 by NDH and was funded by US AEC. It included the analysis of samples from deep production wells (groundwater from the regional aquifer at depths greater than 55 meters [180 feet]) at the Sheldon Station. NDH was concerned in 1990 about the possibility of shallow groundwater coming in contact with the buried radiological materials along the buried walls of the reactor. Subsequently, DOE and NDH agreed to further characterize hydrologic conditions and establish a monitoring program in the shallow, perched groundwater zones. The current monitoring program focuses on the shallow, perched groundwater. Today, decades since facility decommissioning, site inspections and groundwater monitoring continue at this Cold War-era legacy site. The US DOE Office of Legacy Management (LM) monitors groundwater as a best management practice in response to a request from the Nebraska Department of Health and Human Services (DHHS). In 2006, LM recommended discontinuing groundwater monitoring because analytical results since 1970 demonstrated there had been no impact to shallow, perched groundwater and no current or anticipated unacceptable risk to human health and the environment. The State of Nebraska did not concur with the recommendation to stop monitoring but did agree to a reduction in sampling and analysis from once a year to once every 2 years--a schedule that began in 2008. An updated assessment of the groundwater monitoring effort was issued in 2016. The entire body of water-quality data continues to demonstrate that there have been no negative impacts to the shallow, perched groundwater from radioactive materials entombed at the site 46 years ago. LM, with concurrence from Nebraska DHHS, changed the sampling frequency, effective in 2018, as follows: once every 5 years from 2021 to 2041 and once every 10 years from 2041 to 2071. In summary, LM remains committed to its primary mission to protect human health and the environment at each of its 92 legacy sites. With results from groundwater sampling events demonstrating compliance for 48 years, the Hallam site was able to set an LM benchmark to optimize long-term stewardship and define an end-state strategy for other decommissioned reactor facilities across the DOE complex. Continued collaboration with local communities, stakeholder, and regulators often leads to outcomes of sound protectiveness without compromise to safety or compliance, as demonstrated by the optimization of the long-term stewardship at the Hallam site. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Radioactivity-in-materials lead for nEXO (Final Scientific Report)

Abstract, introduction, and summary of work completed and products produced by DOE award SC002466. Primary products were six peer-reviewed publications and two PhD theses. This award also supported the operation of an underground HPGe detector and several radioassay measurements: one in a stand alone publication in Phys. Rev. C and the others in an upcoming nEXO radioassay summary paper. The nEXO collaboration aims to demonstrate the Majorana nature of the neutrino by observing the neutrinoless double-beta decay of 136 Xe with a next-generation experiment. The University of Kentucky nEXO group (UKY) operates a world-leading ultra-low-background γ-ray spectrometer as part of the nEXO radioactive background control R&D effort. The nEXO project is a proposed ton-scale neutrinoless double-beta decay experiment. The UKY PI is leading the radioactivity content assessment for all materials required by the nEXO project. As a subset of this task, UKY is also responsible for project-wide management of all low-background γ-ray spectrometry measurements; material assays with these types of instruments, together with ICP-MS, neutron-activation-analysis, and α-spectrometry, are critical to ensuring that the nEXO project can achieve its design sensitivity

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Detection with Enhanced Energy Windowing Phase I Report

This document reviews the progress of Phase I of the Detection with Enhanced Energy Windowing (DEEW) project. The DEEW project is the implementation of software incorporating an algorithm which reviews data generated by radiation portal monitors and utilizes advanced and novel techniques for detecting radiological and fissile material while not alarming on Naturally Occurring Radioactive Material. Independent testing indicated that the Enhanced Energy Windowing algorithm showed promise at reducing the probability of alarm in the stream of commerce compared to existing algorithms and other developmental algorithms, while still maintaining adequate sensitivity to threats. This document contains a brief description of the project, instructions for setting up and running the applications, and guidance to help make reviewing the output files and source code easier.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

Assessment of Potential Dose and Environmental Impacts from Proposed Testing at the INL Radiological Response Training Range

This assessment uses screening-level models to calculate potential environmental impacts from proposed tests at the Idaho National Laboratory (INL) Radiological Response Training Range (RRTR) site. Proposed tests could be conducted using 11 different radioactive material types that include K2O, LaBr3, KBr, Cu, Zr, F, Ga, Ga2O3, NaNO2, Ga-68, and Tc-99m. The tests could potentially release radioactive material to the atmosphere and radionuclides and other contaminants to the soil, which could leach into the unsaturated zone and migrate to the aquifer. Atmospheric transport of radionuclides to potential human receptors and time-integrated air concentrations were calculated with a Gaussian plume model and three years of hourly meteorological data. Potential surface-soil impacts were calculated with the computer program mixing-cell model (MCM). Groundwater impacts were calculated with the computer programs MCM and GWSCREEN. Radiological doses from potential atmospheric releases were calculated for public receptors off the INL Site and for workers at nearby INL facilities. Results were compared to regulatory dose limits. Maximum potential groundwater concentrations were estimated in the aquifer below the NSTR site and compared to drinking water standards or risk-based screening levels for resident tap water. Soil concentrations were calculated and compared to risk-based screening levels for workers and potential future residents. All impacts were estimated assuming 12 tests are conducted annually using all 11 material types for a period of 15 years. This document provides the resources to enable a subject matter expert in the field of environmental assessments to replicate the modeling and calculations. The methodology and parameters are presented in the text. All electronic files, including computer-code input, output, executable files, batch files, scripts, and spreadsheet files are contained in a zip file that can be accessed by selecting “Additional Information” (select Native File) in the INL Electronic Document Management System (EDMS). It is highly unlikely the test scenarios evaluated in this ECAR will adversely impact human health based on comparisons of calculated dose and concentration against regulatory standards and risk-based screening levels. Conservative estimates of dose to workers and the public from atmospheric transport of possible radionuclide releases are far below federal radiation protection standards. Conservative estimates of potential contaminant concentrations in groundwater are less than federal drinking water standards or screening levels. Predicted radionuclide concentrations in surface soils are below risk-based screening levels, except for Ge-68 (material Ga-68) for the worker. The Ge-68 soil concentration can be made less than the worker PRG, if the number of annual tests using Ga-68 is reduced from 12 to 6. However, the sum of ratios still exceeds one because of the high K-40 ratio. If the EF of the worker (number of days the worker is in the contaminated testing area) is reduced from 225 days/yr (default value for full time worker) to 112 days/yr, the Ge-68 ratio is less than one and the sum of ratios is less than one. Actual radiation doses and groundwater and surface-soil concentrations are likely to be much less than those calculated because of the conservative assumptions and parameters employed in the modeling. For example, atmospheric-transport calculations assume the entire inventory of each material type is readily released to the atmosphere and no plume deposition, depletion, or radioactive decay occurs during transport. The calculations also assume the same meteorological conditions (e.g., wind velocity, wind direction, stability class) that produce the maximum 95th percentile concentration (i.e., concentration representing the 95th percentile of a distribution of concentrations derived from 3 years of hourly meteorological data) at each receptor location are the same for all 12 tests during the year, and each receptor is assumed to be present during all 12 tests. The surface-soil assessment assumes the entire inventory of each test is deposited in the top 5 cm of soil. No atmospheric dispersal is assumed, and the radionuclides are subject only to leaching and radioactive decay. The groundwater-pathway modeling is conservative in that it is one-dimensional in the unsaturated zone (no lateral spreading/dilution) and assumes the entire inventory of contaminants infiltrates into the ground at the same location for every test. This is especially conservative for particulate radionuclides because they would have to dissolve or corrode first and some would be dispersed into the atmosphere. The groundwater receptor is also assumed to consume water directly from a hypothetical well positioned in the location of maximum concentration. In addition, conservative degradation rates were used, and volatilization was not considered for the nonradioactive chemical

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