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All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Hard disk drive dismantlement for critical material recovery

A system and method for recycling rare earth materials from dissimilar hard disk drives are provided. The system and method generally include scanning each hard disk drive, sorting and aligning each hard disk drive, rapid fastener removal or diversion to a metrology station, and the collection of separated value streams, optionally for formation into new magnetic stock. For each scanned hard disk drive having a match in an inventory database, the method includes the separation of an internal magnet from residual components. For each scanned hard disk drive lacking a match in the inventory database, the method includes generating a metrology data collection record containing the location of each fastener on multiple surfaces of the corresponding hard disk drive. The system and method are commercially scalable with the potential to generate between 600 and 700 metric tons of rare earth elements from a single processing facility annually, including neodymium for example.

McIntyre, Timothy J.↗
DOE OSTI2023

Automated recovery of rare earth permanent magnets from electric machines

A method for the automated recovery of rare earth permanent magnets from electric machines are provided. The method includes identifying electric machines in a mixed product stream for performing a unique robotic disassembly routine. Electric machines that are not identified are diverted to a robot training station, during which time the system and the method include implementing a suitable disassembly routine. A conveyor delivers the remaining electric machines to a rotary platform having multiple stations for the simultaneous disassembly of multiple electric machines. Permanent magnets are removed from the electric machines and are then sorted for recycling operations.

McIntyre, Timothy J.↗
DOE OSTI2023

PV Validation Hub

The Validation Hub will be a clearinghouse for the transfer of novel algorithms and software from the PV research community to industry. Potential algorithms tested in the Hub could include the estimation of various PV loss factors and the detection of various operational issues. The primary function of the Hub will be for developers to submit executable code which will run on hosted data sets. Developers will receive private reports on the accuracy and performance (e.g., run-time) of the submitted algorithms, and public high level summaries will be hosted. These summaries will indicate the organization who submitted the algorithm (e.g., links to GitHub pages, documentation websites, etc.), high-level accuracy metrics, and standardized performance metrics. These results will be stored in a publicly available database, accessible through the Hub, with the ability for users to sort and filter the results. In short, the Hub will be presented to public users as a collection of interactive leaderboards, organized around specific analysis tasks pertinent to the PV data science community. These tasks include things such as the estimation of various PV loss factors and the detection of various operational issues. We will present progress on the development of this hub, including preliminary results of comparative validation of PV data science algorithms and progress towards building the platform itself.

algorithm↗
DOE OSTI2013

A Cask Processing Enclosure for the TRU Waste Processing Center - 13408

This paper will discuss the key elements considered in the design, construction, and use of an enclosure system built for the TRU Waste Processing Center (TWPC). The TWPC system is used for the repackaging and volume reduction of items contaminated with radioactive material, hazardous waste and mixed waste. The modular structural steel frame and stainless steel skin was designed for rapid field erection by the use of interchangeable self-framing panel sections to allow assembly of a sectioned containment building and for ease of field mobility. The structure was installed on a concrete floor inside of an outer containment building. The major sections included an Outer Cask Airlock, Inner Cask Airlock, Cask Process Area, and Personnel Airlocks. Casks in overpacks containing transuranic waste are brought in via an inter-site transporter. The overpack lid is removed and the cask/overpack is transferred into the Outer Cask Airlock. A contamination cover is installed on the overpack body and the Outer Cask Airlock is closed. The cask/overpack is transferred into the Inner Cask Airlock on a cask bogie and the Inner Cask Airlock is closed. The cask lid is removed and the cask is transferred into the Cask Process Area where it is placed on a cask tilting station. Once the Cask Processing Area is closed, the cask tilt station is activated and wastes are removed, size reduced, then sorted and re-packaged into drums and standard waste boxes through bag ports. The modular system was designed and built as a 'Fast Track' project at IP Systems in Broomfield Colorado and then installed and is currently in use at the DOE TWPC located near Oak Ridge, Tennessee. (authors)

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

Unique Construction and Social Experiences in Residential Remediation Sites - 13423

Sevenson Environmental Services, Inc., (Sevenson) has performed several radiological remediation projects located in residential urban areas. Over the course of these projects, there has been a wide variety of experiences encountered from construction related issues to unique social situations. Some of the construction related issues included the remediation of interior basements where contaminated material was located under the footers of the structure or was used in the mortar between cinder block or field stone foundations. Other issues included site security, maintaining furnaces or other utilities, underpinning, backfilling and restoration. In addition to the radiological hazards associated with this work there were occupational safety and industrial hygiene issues that had to be addressed to ensure the safety and health of neighboring properties and residents. The unique social situations at these job sites have included arson, theft/stolen property, assault/battery, prostitution, execution of arrest warrants for residents, discovery of drugs and paraphernalia, blood borne pathogens, and unexploded ordnance. Some of these situations have become a sort of comical urban legend throughout the organization. One situation had historical significance, involving the demolition of a house to save a tree older than the Declaration of Independence. All of these projects typically involve the excavation of early 20. century items such as advertisement signs, various old bottles (milk, Listerine, perfume, whisky) and other miscellaneous common trash items. (authors)

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2013

UP2 400 High Activity Oxide Legacy Waste Retrieval Project Scope and Progress-13048

The High Activity Oxide facility (HAO) reprocessed sheared and dissolved 4500 metric tons of light water reactor fuel the fuel of the emerging light water reactor spent fuel between 1976 and 1998. Over the period, approximately 2200 tons of process waste, composed primarily of sheared hulls, was produced and stored in a vast silo in the first place, and in canisters stored in pools in subsequent years. Upon shutdown of the facility, AREVA D and D Division in La Hague launched a thorough investigation and characterization of the silos and pools content, which then served as input data for the definition of a legacy waste retrieval and reconditioning program. Basic design was conducted between 2005 and 2007, and was followed by an optimization phase which lead to the definition of a final scenario and budget, 12% under the initial estimates. The scenario planned for the construction of a retrieval and reconditioning cell to be built on top of the storage silo. The retrieved waste would then be rinsed and sorted, so that hulls could subsequently be sent to La Hague high activity compacting facility, while resins and sludge would be cemented within the retrieval cell. Detailed design was conducted successfully from 2008 until 2011, while a thorough research and development program was conducted in order to qualify each stage of the retrieval and reconditioning process, and assist in the elaboration of the final waste package specification. This R and D program was defined and conducted as a response and mitigation of the major project risks identified during the basic design process. Procurement and site preparatory works were then launched in 2011. By the end of 2012, R and D is nearly completed, the retrieval and reconditioning process have been secured, the final waste package specification is being completed, the first equipment for the retrieval cell is being delivered on site, while preparation works are allowing to free up space above and around the silo, to allow for construction which is scheduled to being during the first semester of 2013. The elaboration of the final waste package is still undergoing and expected to be completed by then end of 2013, following some final elements of R and D required to demonstrate the full compatibility of the package with deep geological repository. The HAO legacy waste retrieval project is so far the largest such project entering operational phase on the site of La Hague. It is on schedule, under budget, and in conformity with the delivery requirements set by the French Safety Authority, as well as other stakeholders. This project paves the way for the successful completion of AREVA La Hague other legacy waste retrieval projects, which are currently being drafted or already in active R and D phase. (authors)

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

Handling Radioactive Waste from the Proton Accelerator Facility at the Paul Scherrer Institut (PSI) - Always Surprising? - 13320

The Paul Scherrer Institut (PSI) is the largest national research centre in Switzerland. Its multidisciplinary research is dedicated to a wide field in natural science and technology as well as particle physics. In this context, PSI is operating, amongst others, a large proton accelerator facility since more than 30 years. In two cyclotrons, protons are accelerated to high speeds and then guided along roughly 100 m of beam line to three different target stations to produce secondary particles like mesons and neutrons for experiments and a separately beam line for UCN. The protons induce spallation processes in the target materials, and also at other beam loss points along the way, with emission of protons, neutrons, hydrogen, tritium, helium, heavier fragments and fission processes. In particular the produced neutrons, due to their large penetration depth, will then interact also with the surrounding materials. These interactions of radiation with matter lead to activation and partly to contamination of machine components and the surrounding infrastructures. Maintenance, operation and decommissioning of installations generate inevitably substantial amounts of radioactive operational and dismantling waste like targets, magnets, collimators, shielding (concrete, steel) and of course secondary waste. To achieve an optimal waste management strategy for interim storage or final disposal, radioactive waste has to be characterized, sorted and treated. This strategy is based on radiation protection demands, raw waste properties (size, material, etc.), and requirements to reduce the volume of waste, mainly for legal and economical reasons. In addition, the radiological limitations for transportation of the waste packages to a future disposal site have to be taken into account, as well as special regulatory demands. The characterization is a task of the waste producer. The conditioning processes and quality checks for radioactive waste packages are part of an accredited waste management process of PSI, especially of the Section Dismantling and Waste Management. Strictly proven and accepted methods needed to be developed and enhanced for safe treatment, transport, conditioning and storage. But in the field of waste from research activities, individual and new solutions have to be found in an increasingly growing administrative environment. Furthermore, a wide variety of components, with a really large inventory of radioactive nuclides, has to be handled. And there are always surprising challenges concerning the unusual materials or the nuclide inventory. In case of the operational and dismantling radioactive accelerator waste, the existing conditioning methods are in the process of a continuous enhancement - technically and administratively. The existing authorized specifications of conditioning processes have to be extended to optimize and fully describe the treatment of the inevitably occurring radioactive waste from the accelerator facility. Additional challenges are the changes with time concerning the legal and regulatory requirements - or do we have to consider it as business as usual? This paper gives an overview of the current practices in radioactive waste management and decommissioning of the existing operational accelerator waste. (authors)

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DOE OSTI2013

Estimation of Performance of an Active Well Coincidence Counter Equipped with Boron-Coated Straw Neutron Detectors - 13401

He-3, a very rare isotope of natural helium gas, has ideal properties for the detection of thermal neutrons. As such it has become the standard material for neutron detectors and sees ubiquitous use within many radiometric applications that require neutron sensitivity. Until recently, there has been a fairly abundant supply of He-3. However, with the reduction in nuclear weapons, production of tritium ceased decades ago and the stockpile has largely decayed away, reducing the available He-3 supply to a small fraction of that needed for neutron detection. A suitable and rapidly-deployable replacement technology for neutron detectors must be found. Many potential replacement technologies are under active investigation and development. One broad class of technologies utilizes B-10 as a neutron capture medium in coatings on the internal surfaces of proportional detectors. A particular implementation of this sort of technology is the boron-coated 'straw' (BCS) detectors under development by Proportional Technologies, Inc. (PTi). This technology employs a coating of B-10 enriched boron carbide (B{sub 4}C) on the inside of narrow tubes, roughly 4 mm in diameter. A neutron counter (e.g. a slab, a well counter, or a large assay counter designed to accommodate 200 liter drums) could be constructed by distributing these narrow tubes throughout the polyethylene body of the counter. One type of neutron counter that is of particular importance to safeguards applications is the Active Well Coincidence Counter (AWCC), which is a Los Alamos design that traditionally employs 42 He-3 detectors. This is a very flexible design which can accurately assay small samples of uranium- and plutonium-bearing materials. Utilizing the MCNPX code and benchmarking against measurements where possible, the standard AWCC has been redesigned to utilize the BCS technology. Particular aspects of the counter performance include the single-neutron ('singles') detection efficiency and the time constant for the decrease in neutron population in the counter following a fission event (a.k.a. the die-away time). Results of the modeling and optimization are presented. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2012

Spent Sealed Sources Management in Switzerland - 12011

Information is provided about the international recommendations for the safe management of disused and spent sealed radioactive sources wherein the return to the supplier or manufacturer is encouraged for large radioactive sources. The legal situation in Switzerland is described mentioning the demand of minimization of radioactive waste as well as the situation with respect to the interim storage facility at the Paul Scherrer Institute (PSI). Based on this information and on the market situation with a shortage of some medical radionuclides the management of spent sealed sources is provided. The sources are sorted according to their activity in relation to the nuclide-specific A2-value and either recycled as in the case of high active sources or conditioned as in the case for sources with lower activity. The results are presented as comparison between recycled and conditioned activity for three selected nuclides, i.e. Cs-137, Co-60 and Am-241. (author)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2012

Treatment of Uranium Slugs at the CEA Marcoule site - 12026

The decladding units on the Marcoule nuclear site in southeast France were commissioned in the early 1960's to receive spent fuel cartridge for interim storage in pools and to prepare them for dissolution. After decladding, the uranium slugs were sent to the reprocessing plant for dissolution. The units were shut down in 1997. All the operational and process equipment was dismantled and transferred to the ANDRA waste repository between 1998 and 2001. During cleanup of the G2-G3 interim storage pool all the internal components (baskets and machinery) mounted on the civil engineering structures were cut up. In November 2007, a UNGG fuel slug in its aluminum cartridge was found at the bottom of the pool. Similarly, while iron slugs (used to control plutonium-producing reactors) were being sorted, five de-cladded uranium slugs were found between 2006 and 2008. As the processing facilities had already been dismantled, it was decided to send these fuel cartridges to a laboratory for dissolution. The project was to recover and condition the cartridge and slugs and transport them safely to the laboratory. This required the design of dedicated equipment due in particular to the high dose rate and pyro-phoricity hazard. The procedure also required safety clearance by the regulatory authorities for the operations within the facility and for the transport of radioactive materials. The cartridge was dissolved in May 2009, 18 months after it was discovered. Based on the experience acquired, the retrieval of the uranium slugs was completed in March 2011 after a year of preparation. (authors)

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DOE OSTI2012

Treatment of G1 Baskets at the CEA Marcoule Site - 12027

In the dismantling program for the first-generation French reactors in accordance with the nonproliferation treaty, the CEA is in charge of cleanup and dismantling operations for the facilities at Marcoule, including the decladding units. The G1 decladding was built between 1955 and 1957 in order to de-clad spent fuel elements from the G1 plutonium-producing reactor and prepare them for dissolution. The facility was also used for interim storage of G1, G2 and G3 fuel dissolution baskets, which had been used during plant operation for transfer (from the decladding facility to the UP1 plant) and/or dissolution of spent fuel elements. One of the cleanup projects involves recovery of the baskets, which will be cut up, sorted, and conditioned in metal bins. The bins will be immobilized with cement grout, then transferred to the onsite solid waste conditioning facility (CDS) and to the repository operated by the French National Radioactive Waste Management Agency (ANDRA). The project is now in progress, after special safety permits were issued and measurement stations and dedicated tools were developed to handle all types of baskets (which differed according to their origin and use). The disposal of all the baskets is scheduled to last 2 years and will produce 55 metal waste bins. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Dismantling Structures and Equipment of the MR Reactor and its Loop Facilities at the National Research Center 'Kurchatov Institute' - 12051

In 2008 a design of decommissioning of research reactors MR and RFT has been developed in the National research Center 'Kurchatov institute'. The design has been approved by Russian State Authority in July 2009 year and has received the positive conclusion of ecological expertise. In 2009-2010 a preparation for decommissioning of reactors MR and RFT was spent. Within the frames of a preparation a characterization, sorting and removal of radioactive objects, including the irradiated fuel, from reactor storage facilities and pool have been executed. During carrying out of a preparation on removal of radioactive objects from reactor sluice pool water treating has been spent. For these purposes modular installation for clearing and processing of a liquid radioactive waste 'Aqua - Express' was used. As a result of works it was possible to lower volume activity of water on three orders in magnitude that has allowed improving essentially of radiating conditions in a reactor hall. Auxiliary systems of ventilation, energy and heat supplies, monitoring systems of radiating conditions of premises of the reactor and its loop-back installations are reconstructed. In 2011 the license for a decommissioning of the specified reactors has been received and there are begun dismantling works. Within the frames of works under the design the armature and pipelines are dismantled in a under floor space of a reactor hall where a moving and taking away pipelines of loop facilities and the first contour of the MR reactor were replaced. A dismantle of the main equipment of loop facility with the gas coolant has been spent. Technologies which were used on dismantle of the radioactive contaminated equipment are presented, the basic works on reconstruction of systems of maintenance of on the decommissioning works are described, the sequence of works on the decommissioning of reactors MR and RFT is shown. Dismantling works were carried out with application of means of a dust suppression that, in aggregate with standard means at such works of individual protection of the personnel and devices of radiating control, has allowed to lower risk of action of radiation on the personnel, the population and environment at the expense of reduction of volume activity of radioactive aerosols in air. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Regulator Interface Strategies Implemented at the Y-12 National Security Complex Old Salvage Yard Soils Remediation Project, Oak Ridge, TN - 12162

The Oak Ridge Y-12 National Security Complex housed an area known as the Old Salvage Yard (OSY) that was approximately 7 acres. The OSY was used as an area for the accumulation, processing and storage of scrap metal and equipment from Y-12 operations extending from 1968 until 2009. Areas in the northern sections of OSY also have been used for the storage of used oils containing solvents and the accumulation and recycling or de-heading and crushing of 55-gal metal drums. Scrap metal operations historically involved the accumulation, sorting, storage, public sale or disposal of scrap metal and equipment. Non-containerised storage of scrap metal was routine until 1995 when scrap metal received at OSY was placed in B-24 and B-25 boxes. Under the American Recovery and Reinvestment Act (ARRA), approximately 26,759 cubic meters of scrap metal and debris were removed and disposed at both on and off-site disposal facilities including the on-site, Oak Ridge Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) landfill in 2010 and 2011. This removal action was performed in accordance with a CERCLA Record of Decision (ROD) and a close working relationship with both the U.S. Environmental Protection Agency (EPA) Region IV and Tennessee Department of Environment and Conservation (TDEC). Due to efficiencies and the excellent cooperative relationship forged with EPA Region IV and TDEC for Y-12 ARRA Cleanup Projects, a surplus of funding was available for additional remediation work that was completed in fiscal year (FY) 2011. The underlying OSY soils were targeted for characterization and potential remediation. To expedite these important activities, the U.S. Department of Energy Oak Ridge Environmental Management partnered with the regulators during detailed planning sessions through a variety of means to quickly and efficiently characterize and pinpoint areas requiring remediation according to previous ROD commitments. Data Quality Objectives (DQOs), data-sharing, real-time characterization reporting, surface and groundwater modeling and other interface planning activities were utilized to help facilitate and complete characterization and remediation activities. As a result of these strategies, the surgical extraction of one contiguous area of soil approximately 354 cubic meters is planned for FY12. The strategies discussed resulted in a major reduction of footprint remediation (i.e., 2.8% of the original estimate) which was originally estimated at over 26,759 cubic meters. The original estimate was developed using historical data collected at various times over the period of 20 years. By leveraging a hybrid sampling approach that involved both statistically-based and biased sampling locations, the area of contamination was significantly reduced resulting in both a compliant remedial design that is cost effective while mitigating a principle threat sources to surface and groundwater at the Y-12 plant. One remedial action boundary of 354 cubic meters was verified in the northern section of the Western OSY area known as the old drum de-header station for VOCs. The original estimate for disposal was in excess of 26,759 cubic meters. This area is scheduled for waste characterization and profile development in the first half of fiscal year 2012. The anticipated disposal facility is an on-site Oak Ridge CERCLA disposal landfill known as the Environmental Management Waste Management Facility (EMWMF). By utilizing the careful strategic planning, field-based screening and close cooperation of regulatory stakeholders as detailed in this paper, the total area of soil requiring remedial action within the Y-12 OSY footprint was 354 cubic meters or 2.8% of the original planned estimate. A potential waste reduction of 97.2% was realized over the original planned estimate for OSY Soils. Significant cost savings were achieved by - Minimizing the footprint of the remedial action; - Confirmatory analysis of soils instead of use of historical sampling results for waste profile development; - Targeting traditional laboratory analysis for soils that failed field screen protocols for radionuclides, heavy metals and VOCs; - Providing accurate and precise definition of RABs; and - Reduction of target contaminants for focused laboratory analysis instead of broad-range contaminants. (authors)

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

A nano-structured ZnO film as diagnostic X-ray sensor

Currently some international organizations such as WHO and IAEA have shown concerns about the quality of diagnostic services in clinics and hospitals that use ionizing radiation. In fact, the IAEA recommend that the characteristics of the X-ray beam must be adjusted to obtain the highest quality of the radiographic image with the minimum exposure to the patient. Several types of detectors may be used for monitoring X-ray beams, such as: ionization chamber, photodiode, phototransistor, among others. Recently nano-structured films made of various types of metal oxide materials have been used for various technological applications. Accordingly, the purpose of this paper is to present a sort of device based on a nano-structured zinc oxide (ZnO) to operate as a diagnostic X-ray sensor. By depositing a thin film on the glass substrate some ZnO semiconductor samples were built by sputtering techniques and then mounted in a BNC type connector to perform the electrical characterization. To test the device, we choose a standard X-Ray beam, the RQR9 radiation quality, which is normally used as the tool and condition for calibrating diagnostic X-Ray instruments in the energy range of computed tomography, in accordance with the stated requirements of IEC 61267. A 6430 sub-femto-ammeter, Keithley, was used as electrometer to perform the output readings and simultaneously bias the ZnO sensor. Analysis of the angular dependence and the dose rate were performed to evaluate how the device responds under the RQR9 radiation spectra. Although the results have shown that the ZnO film presents a certain angular dependence, if an angle of incidence of photons is selected, the device displays reproducibility as X-ray sensor and has the feature of radiation hardness unlike other types of semiconductor electronic devices typically used as an X-ray detector. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Radiation Measurement Systems and Experiences in Japan after the Fukushima Accident

In the approximately 4 years since the March 2011 earthquake and subsequent Tsunami damaged the three Dai-Ichi nuclear power plants, the AREVA and Canberra team have deployed a variety of different radiation measurement machines in Japan. These measurement projects or instruments include: - Dose-rate mapping of the close-in areas on the damaged reactor site using the stored output of dose rate meters carried by workers planning for the installing of a water processing unit; - Spectroscopic on-line measurements of the first water processing unit on the site, where these nuclide-specific measurements at multiple points allowed remote users to determine the optimum processing parameters - General purpose food measurement units, both high precision HPGe systems and low cost NaI systems, which were used throughout the country for food quality measurements; - In-vivo measurement systems, both fixed and mobile whole body counting units, that were used on at least 500,000 people, demonstrating to them that internal doses were very low; - In-vivo measurements with special high-sensitivity whole body counters on babies and children, demonstrating to the parents of these children that internal doses are very low; - Automatic systems for high throughput assay of bags of rice that are part of the consumer confidence plan to prove to customers that Fukushima Prefecture rice is safe; - High sensitivity automatic system to assay boxes of special semi-dry persimmons and to report the activity of each individual item within the box meets the regulatory limit; - System to assay soil on a conveyor belt and sort the output according to level of radioactivity, which would then reduce the volume of material that must be treated as radioactive; - System to assay a truck loaded with 1 cubic meter sacks of soil and vegetation, and report the results of each individual sack. - On-line water measurement system for SrY90 at levels that are suitable for release to the environment; - Mobile system to prove that rolls of grass harvested for animal feed meet the regulatory requirements. This document briefly describes the purpose of each measurement project, describes the instruments and assay method that used, and discusses some of the operational experiences from each of these projects. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Design and Validation Testing of TruckScan to Assay Large Sacks of Fukushima Radioactive Debris on a Truck

As a result of the March 2011 earthquake and resulting tsunami in Japan, there was a serious accident at the Fukushima Dai-ichi Nuclear Power Plant. This accident has contaminated soil and vegetation in a wide area around the plant. Decontamination projects over the last 4 years have resulted in large numbers of 1 cubic meter canvas bags of debris, commonly called Super Sacks [SS]. These are currently stored nearby where they were generated, but starting in 2015, they will be moved to various Interim Storage Facilities [ISF]. Trucks will typically carry 8-20 of these SSs. When the trucks arrive at the ISF they need to be rapidly sorted into groups according to radioactivity level, for efficient subsequent processing. Canberra Industries, Inc. [CI] has designed a new truck monitoring system 'TruckScan' for use at these ISFs. The TruckScan system must measure the entire truck loaded with multiple closely packed SSs, and generate a nuclide specific assay report showing the radioactivity in each individual SS. The Canberra-Japan office, along with Obayashi Corporation have performed validation testing to demonstrate to the regulatory authorities that the proposed technique was sufficiently accurate. These validation tests were conducted at a temporary storage area in Fukushima prefecture. Decontaminated waste of various representative types and of various levels of radioactivity was gathered and mixed to create homogeneous volumes. These volumes were sampled multiple times and assayed with laboratory HPGe detectors to determine the reference concentration of each pile. Multiple SSs were loaded from each pile. Some of the SSs were filled 50% full, others 75% full, and others 100% full, to represent the typical loading configuration of the existing SSs in the field. The content of the SSs are either sand, soil, or vegetation with densities ranging from 0.3 g/cc - 1.6 g/cc. These SSs with known concentrations of Cs-134 and Cs-137 were then loaded onto trucks in a variety of configurations, typical of how they might be on the real trucks. A partial system was installed at the site and used to assay these trucks with the various loading configurations. Whereas the full system will have 8 collimated 3 x 3'' NaI detectors, the test system only had two detectors; therefore the truck was moved and counted 4 times. The data were acquired and analyzed with the Canberra Genie software to determine the peak counts for both Cesium nuclides. That data was then analyzed with a prototype version of a Maximum Entropy algorithm, to determine the individual SS activity. The goal of the validation tests was to demonstrate that the system could detect 1000 Bq/kg in 15 seconds, and to determine how accurately it could quantify individual SSs. The validation tests demonstrated that the product would perform as predicted. The TruckScan results were consistent with the sample assay results [y = 1.0029 x, R{sup 2} = 0.914]. The Total Propagated Uncertainty, including both uncertainties from these tests and others that were estimated but not tested was 16.6% percent. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Remote Decommissioning of the Plutonium Experimental Extraction Facility, Sellafield - 17031

A redundant suite of gloveboxes formerly used for an experimental plutonium extraction process required decommissioning to avoid further degradation and a risk to other users of the Analytical Services building on Sellafield site. Following a full option selection process, remote techniques were chosen as the most appropriate (i.e. ALARP) decommissioning method. A nearby ventilation system from a previous project was re-engineered to ventilate the laboratory. Gloveboxes in neighbouring labs were cleared to create space to install a personnel access facility. The containment was reinforced where necessary to provide adequate protection from the remote machinery. Following a campaign of manual waste removal, including approximately 21 tonnes of lead chevron bricks released as exempt waste, a Brokk{sup R} 90 remotely-operated vehicle (ROV) was deployed to perform remote size reduction of the glovebox suite. The resulting plutonium-contaminated material (PCM) was delivered to a Kraft Raptor manipulator for sorting and consignment to an on-site store in 200-litre drums. Despite indications from pre-decommissioning surveys of the presence of ILW in the shielded section, none was found. Following the successful removal of the glovebox suite, and the emergence during the project of additional redundant gloveboxes within the Analytical Services building, the Remediation team has taken the opportunity to retain the decommissioning infrastructure as a remote glovebox breakdown facility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Feedback of the Dutch Historical Waste Management Program - 17085

This paper provides a feedback on the Dutch historical waste management program. Over the past decades, the High Flux Reactor located in Petten in the province of North Holland has tested numerous types of materials and fuels irradiation, for both medical and nuclear energy applications. A number of waste containers was accumulated on site, where the storage facility was initially built. Following Dutch political decisions in the 1980's, the contents of the waste storage facility in Petten must be transferred to another location. The consequences of this decision included the transportation of waste containers from Petten to the new location, which is located in the South-West of the Netherlands. This project was initiated and is now running. It includes a complex structure and can be described as follows: each historical-waste container will be opened, its contents segregated in three categories depending on the dose rate emission, characterized and repacked in new containers suitable for transportation and storage at the new site. The approach chosen is based on the final dose rate emitted by the repacked containers, which are sub-divided into 3 categories.. The lowest dose-emitting containers are transported directly to the new storage facility. The 2 other categories are first transported to a third party to be super compacted and subsequently transported to the new storage facility. So far, the segregation method is based on Co-60 gamma emission and nuclide vectors associated to the different materials. A pre-sorting has been established to fit as much as possible IAEA recommendations. Historical waste has then to be classified in families sharing common history. Thus, a set of canisters containing the pieces of the old reactor vessel, which was replaced in 1984 was the first challenge, followed by a set of canisters containing residues of instruments only and pieces coming from the top of the control rods of the reactor which were not directly irradiated in the reactor. So far, the two first sets of canisters are being treated and characterized as well as possible. Reporting and official documents have been filled up to comply to Dutch transport regulations and to define waste acceptance criteria for the storage facility. The first transports of containers with Low Level Waste occurred in October 2015, and another transport took place at the end of 2016. (authors)

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