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At least 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Statistical Approaches for Pebble Bed Reactor Operations and Safeguards

The design of pebble bed reactors (PBRs) and their method of operation align more closely with statistical approaches used in manufacturing and process control than traditional safeguards statistical approaches. The reason is PBRs will employ a nondestructive assay (NDA) measurement (burnup measurement system [BUMS]) that is part of the fuel handling system supporting discharge decisions in addition to the reactor code which monitor performance. The integration of these two approaches provides the opportunity to monitor reactor performance statistically for both operations and safeguards in ways not achievable using other reactor designs. For light water reactors (LWRs), knowledge about reactor code performance in predicting irradiated special nuclear material (SNM) content historically was only achieved from special measurement campaigns or from fuel reprocessing. Conversely, through statistical comparison of the BUMS with the reactor code-predicted values, PBRs can achieve this in real time. The resulting SNM distribution is an indicator of reactor performance because factors such as transit time and path of the pebble fuel through the reactor determine the plutonium production and uranium depletion. By analyzing the predicted and measured values, opportunities exist to adjust operating parameters, fuel design, and other characteristics to optimize performance and fuel utilization. From a safeguards perspective, this approach also provides the information necessary to validate declared values and evaluate whether the reactor is being operated as expected. This paper outlines statistical approaches for PBRs that can be used to support both operations and safeguards.

Ball, Cory↗
DOE OSTI2023

Neutron Coincidence Measurements of Uranium-233 Oxide

Renewed international interest in thorium-fueled advanced reactors has challenged the safeguards community to address future proliferation concerns. Thorium-based technology presents many benefits but does not eliminate the proliferation risks associated with producing and processing fissile material. A byproduct of thorium-fueled reactors is uranium-233, which is classified as a direct-use material. As a result, the development of new or improved methods to characterize and measure materials containing 233U must mirror the pace of development of reactors and facilities that produce such material. Research is underway to assess, develop, and test approaches for safeguarding nuclear materials within the thorium fuel cycle. Neutron signatures from the nondestructive assay (NDA) of materials containing 233U are being quantified to inform the potential characterization of these materials. Using a traditional neutron coincidence counter and a series of well-documented 233U oxide samples, initial measurements have been made to assess the feasibility of 233U characterization and discrimination from other uranium isotopes, primarily 235U, using a combination of measurement techniques and analysis methods. Data acquisition is performed in list mode, allowing for a variety of analyses to be performed on the raw data that is not available using traditional shift register technology. Measurements were performed in passive and active configurations to quantify the strength of signal and to validate simulations in support of this work. This paper presents and discusses the results of the initial measurements of 233U oxide performed at Oak Ridge National Laboratory.

Lockhart, Madeline↗
DOE OSTI2013

Decommissioning of the Dragon High Temperature Reactor (HTR) Located at the Former United Kingdom Atomic Energy Authority (UKAEA) Research Site at Winfrith - 13180

The Dragon Reactor was constructed at the United Kingdom Atomic Energy Research Establishment at Winfrith in Dorset through the late 1950's and into the early 1960's. It was a High Temperature Gas Cooled Reactor (HTR) with helium gas coolant and graphite moderation. It operated as a fuel testing and demonstration reactor at up to 20 MW (Thermal) from 1964 until 1975, when international funding for this project was terminated. The fuel was removed from the core in 1976 and the reactor was put into Safestore. To meet the UK's Nuclear Decommissioning Authority (NDA) objective to 'drive hazard reduction' [1] it is necessary to decommission and remediate all the Research Sites Restoration Ltd (RSRL) facilities. This includes the Dragon Reactor where the activated core, pressure vessel and control rods and the contaminated primary circuit (including a {sup 90}Sr source) still remain. It is essential to remove these hazards at the appropriate time and return the area occupied by the reactor to a safe condition. (author)

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

Repackaging of High Fissile TRU Waste at the Transuranic Waste Processing Center - 13240

Twenty-six drums of high fissile transuranic (TRU) waste from Oak Ridge National Laboratory (ORNL) operations were declared waste in the mid-1980's and placed in storage with the legacy TRU waste inventory for future treatment and disposal at the Waste Isolation Pilot Plant (WIPP). Repackaging and treatment of the waste at the TRU Waste Packaging Center (TWPC) will require the installation of additional equipment and capabilities to address the hazards for handling and repackaging the waste compared to typical Contact Handled (CH) TRU waste that is processed at the TWPC, including potential hydrogen accumulation in legacy 6M/2R packaging configurations, potential presence of reactive plutonium hydrides, and significant low energy gamma radiation dose rates. All of the waste is anticipated to be repackaged at the TWPC and certified for disposal at WIPP. The waste is currently packaged in multiple layers of containers which presents additional challenges for repackaging activities due to the potential for the accumulation of hydrogen gas in the container headspace in quantities than could exceed the Lower Flammability Limit (LFL). The outer container for each waste package is a stainless steel 0.21 m{sup 3} (55-gal) drum which contains either a 0.04 m{sup 3} or 0.06 m{sup 3} (10-gal or 15-gal) 6M drum. The inner 2R container in each 6M drum is ∼12 cm (5 in) outside diameter x 30-36 cm (12-14 in) long and is considered to be a > 4 liter sealed container relative to TRU waste packaging criteria. Inside the 2R containers are multiple configurations of food pack cans, pipe nipples, and welded capsules. The waste contains significant quantities of high burn-up plutonium oxides and metals with a heavy weight percentage of higher atomic mass isotopes and the subsequent in-growth of significant quantities of americium. Significant low energy gamma radiation is expected to be present due to the americium in-growth. Radiation dose rates on inner containers are estimated to be 1-3 mSv/hr (100-300 mrem/hr) with an unshielded dose rate on the waste itself of over 10 mSv/hr (1 rem/hr). Additional equipment to be installed at the TWPC will include a new perma-con enclosure and a shielded/inert glovebox in the process building to repackage and stabilize the waste. All of the waste will be repackaged into Standard Pipe Overpacks. Most of the waste (21 of the 26 drums) is expected to be repackaged at the food-pack can level (i.e. the food-pack cans will not be opened). Five of the incoming waste containers are expected to be repackaged at the primary waste level. Three of the containers exceed the 200 gram Pu-239 Fissile Gram Equivalent (FGE) limit for the Standard Pipe Overpack. These three containers will be repackaged down to the primary waste level and divided into eight Standard Pipe Overpacks for shipment to WIPP. Two containers must be stabilized to eliminate any reactive plutonium hydrides that may be present. These containers will be opened in the inert, shielded glovebox, and the remaining corroded plutonium metal converted to a stable oxide form by using a 600 deg. C tube furnace with controlled oxygen feed in a helium carrier gas. The stabilized waste will then be packaged into two Standard Pipe Overpacks. Design and build out activities for the additional repackaging capabilities at the TWPC are scheduled to begin in Fiscal Year 2013 with repackaging, stabilization, and certification activities scheduled to begin in Fiscal Year 2014. Following repackaging and stabilization activities, the Standard Pipe Overpacks will be certified for disposal at WIPP utilizing Non-Destructive Examination (NDE) to verify the absence of prohibited items and Non-Destructive Assay (NDA) to verify the isotopic content under the TWPC WIPP certification program implemented by the Central Characterization Project (CCP). (authors)

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

Investing in International Information Exchange Activities to Improve the Safety, Cost Effectiveness and Schedule of Cleanup - 13281

With decreasing budgets and increasing pressure on completing cleanup missions as quickly, safely and cost-effectively as possible, there is significant benefit to be gained from collaboration and joint efforts between organizations facing similar issues. With this in mind, the US Department of Energy (DOE) and the UK Nuclear Decommissioning Authority (NDA) have formally agreed to share information on lessons learned on the development and application of new technologies and approaches to improve the safety, cost effectiveness and schedule of the cleanup legacy wastes. To facilitate information exchange a range of tools and methodologies were established. These included tacit knowledge exchange through facilitated meetings, conference calls and Site visits as well as explicit knowledge exchange through document sharing and newsletters. A DOE web-based portal has been established to capture these exchanges and add to them via discussion boards. The information exchange is operating at the Government-to-Government strategic level as well as at the Site Contractor level to address both technical and managerial topic areas. This effort has resulted in opening a dialogue and building working relationships. In some areas joint programs of work have been initiated thus saving resource and enabling the parties to leverage off one another activities. The potential benefits of high quality information exchange are significant, ranging from cost avoidance through identification of an approach to a problem that has been proven elsewhere to cost sharing and joint development of a new technology to address a common problem. The benefits in outcomes significantly outweigh the costs of the process. The applicability of the tools and methods along with the lessons learned regarding some key issues is of use to any organization that wants to improve value for money. In the waste management marketplace, there are a multitude of challenges being addressed by multiple organizations and the effective pooling and exchange of knowledge and experience can only be of benefit to all participants to help complete the cleanup mission more quickly and more cost effectively. This paper examines in detail the tools and processes used to promote information exchange and the progress made to date. It also discusses the challenges and issues involved and proposes recommendations to others who are involved in similar activities. (authors)

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

Optimisation of the Management of Higher Activity Waste in the UK - 13537

The Upstream Optioneering project was created in the Nuclear Decommissioning Authority (UK) to support the development and implementation of significant opportunities to optimise activities across all the phases of the Higher Activity Waste management life cycle (i.e. retrieval, characterisation, conditioning, packaging, storage, transport and disposal). The objective of the Upstream Optioneering project is to work in conjunction with other functions within NDA and the waste producers to identify and deliver solutions to optimise the management of higher activity waste. Historically, optimisation may have occurred on aspects of the waste life cycle (considered here to include retrieval, conditioning, treatment, packaging, interim storage, transport to final end state, which may be geological disposal). By considering the waste life cycle as a whole, critical analysis of assumed constraints may lead to cost savings for the UK Tax Payer. For example, it may be possible to challenge the requirements for packaging wastes for disposal to deliver an optimised waste life cycle. It is likely that the challenges faced in the UK are shared in other countries. It is therefore likely that the opportunities identified may also apply elsewhere, with the potential for sharing information to enable value to be shared. (authors)

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

High Resolution Scanning Configurations for the Canberra Tomographic Gamma Scanner - 17323

The Canberra Tomographic Gamma Scanner (TGS) is a Non-Destructive Assay (NDA) system that combines high resolution gamma ray spectroscopy (HRGS) with low-spatial-resolution imaging for the measurement of heterogeneous drummed waste with matrix densities in the range of 0 to 1 g/cc. The TGS methodology uses a two-pass scanning protocol; a 10-15 μCi {sup 152}Eu transmission source is used to interrogate density and composition of the matrix in the first pass, and passive gamma ray spectroscopy is used to collect emission spectra in the second, while the drum is rotated and translated to vary the view path. In standard implementations, the transmission and emission data are used to reconstruct three dimensional images of the matrix and activity distribution inside the drum on a 10x10x16 grid of discrete volume elements or 'voxels'. These images allow attenuation corrections to be calculated and applied on a view by view, and voxel by voxel basis, reducing the over- and under-correction biases obtained by assay techniques which assume homogeneous distributions of material. The most recent TGS system deployed by Canberra featured, among other new developments, the ability to perform higher resolution image re-construction, which may prove beneficial for the assay of more highly mixed matrices or oddly shaped assay items within a container. These additional scanning protocols reconstruct the image data into 14x14 and 20x20 voxel grids, using a smaller collimator aperture size. The hardware and software implementations of these higher resolution scans are discussed, and the calibration and initial evaluations of performance are presented. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Computational Safeguards Analysis of PWR Spent Nuclear Fuel in Dry Storage - 17435

The master plan for spent nuclear fuel (SNF) management in Korea recently adopted by Korean atomic energy promotion council. According to the master plan, the detailed practical and R and D strategies for SNF transferring, storage and disposal should be determined in the near future. Besides, it is also necessary to make plan how to secure SNF at each step in terms of nuclear safeguards, safety, security and nonproliferation. Several nondestructive analysis (NDA) techniques have been studied for assuring the safety of SNF in PWR methods. Obviously, gamma-ray is most dominant radiation emitted by SNF so several gamma-ray measurement methods have been developed and applied for the purpose of SNF monitoring. Those methods, however, suffer from a lack of nuclear data and interfering background radiation. To date, several neutron measurement methods have been developed as a solution to the problem of gamma-ray methods. Korea Institute of Nuclear Nonproliferation and Control (KINAC) currently develops a new novel technique of fast neutron detection for spent fuel monitoring at dry spent fuel storage facilities. As the first step, a computational program has been developed in order to evaluate neutron flux on the surface of PWR dry storage cask and study for some potential sabotage events. Typically, neutron flux on the surface of dry storage cask can be evaluated via two steps. The source term, based on burn-up rate of SNF, its cooling time and others related to operation condition at nuclear power plants, can be firstly defined by ORIGEN code. The surface flux can be then evaluated via MCNP code. These two separated steps are complicated and time consuming so KINAC developed a new algorithm and user-interface, coupled ORIGEN and MCNP. (authors)

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

Demolition of the K-27 Building at the East Tennessee Technology Park (ETTP) - 17423

On February 8, 2016, URS - CH2M Oak Ridge LLC (UCOR) initiated the demolition of the East Tennessee Technology Park's (ETTP's) final gaseous diffusion building in DOE's Oak Ridge Environmental Management program. Building K-27 was a 35,582 m{sup 2} (383,000 ft{sup 2}) gaseous diffusion building that contained approximately 92,903 m{sup 2} (1 million ft{sup 2}) of floor space. It was one of five gaseous diffusion buildings at the site, formerly known as the Oak Ridge Gaseous Diffusion Plant. The actual demolition of this facility took approximately seven months to complete; however, planning and preparation took more than three years. The deactivation process was a critical element in preparing the building for demolition. UCOR's K-27 deactivation activities began in April 2012 with the installation of safe walking paths and fall protection systems to allow safe work access. Next, all mechanical and electrical systems associated with the facility were shut down and isolated, a process known as 'cold and dark.' More than 19.3 km (12 mi) of process piping and components were marked and measured. Additionally, workers performed nondestructive assay (NDA) and vent-purge-and-drain activities and applied foam to piping and components as needed [1]. Based on characterization results, approximately 4.8 km (3 mi) of process piping, and 372 converters and compressors were removed and shipped offsite prior to demolition. Lower-hazard materials and equipment were demolished with the building structure and transported to onsite disposal facilities. Upfront waste characterization and ongoing monitoring were key to the successful disposal of demolition debris at onsite facilities. Due to process improvements in waste loading recommended by work crews, the project achieved a 50 percent improvement in the daily hauling rate during the 9,000-truckload shipping campaign. Crews worked in high-risk, high-hazard conditions and were exposed to radiological and chemical contaminants, extreme temperatures, and deteriorating infrastructure. However, the hazard controls that were put in place, based on the Integrated Safety Management System (ISMS) walk-downs and hazard determination, made the effort manageable. The overall safety performance during demolition of the facility was excellent, with no significant injuries, and exceeded industry standards and DOE target levels. Outstanding safety performance was a testament to the dedication of the workforce, given the significant hazards presented by the demolition of the Hazard Category 2 nuclear facility. Another key element to the success of the K-27 demolition was the support and approval of all stakeholders involved in this effort, including DOE, the surrounding communities, the State of Tennessee, the UCOR management team, and, most importantly, the workforce that performed this highly hazardous work. This $52 million demolition project was completed ahead of schedule with no significant injuries and in compliance with all DOE orders, site procedures, and environmental controls. (authors)

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

Feasibility studies for the design of disposal vaults for multi-purpose containers (MPCs): construction, operation, ventilation and management of heat - 15175

The Nuclear Decommissioning Authority (NDA) subsidiary, Radioactive Waste Management Ltd (RWM) is responsible for the safe implementation of a Geological Disposal Facility (GDF) for radioactive waste in the UK. As part of their generic programme to underpin the planned designs for a GDF, work is ongoing to explore a range of possible geological disposal concepts. This work developed GDF concept designs for Multi-Purpose Containers (MPC) disposal in the generic host rock types considered: higher strength (crystalline), lower strength sedimentary (sedimentary / clay formations) and evaporite rocks. This paper is specific to concepts for the disposal of MPC for Pressurised Water Reactor (PWR) spent nuclear fuel (SF). The key outcomes of the feasibility studies to improve the understanding of the MPC disposal concept and the specific challenges to construct, operate, ventilate, manage heat and backfill associated with this concept are summarised in this paper. The management of the high loading of radiogenic heat (from 12 PWR SF assemblies in MPCs compared to 4 in 'standard' disposal containers) during an extended open phase (delayed backfilling of the disposal area) of up to 300 years was investigated. The concept designs developed demonstrate it is feasible to construct, operate and ventilate disposal vaults to accommodate open 'storage' periods required for transitional thermal decay of high heat generating waste in higher strength and lower strength sedimentary host rock types considered by RWM. This work focused on developing two contrasting disposal concepts which are applicable to both higher strength and lower strength rock environments; both of which illustrate how vault temperature, stability and groundwater inflows can be controlled to delay backfilling and allow MPC package thermal output to decrease. MPC disposal in evaporite rock does not require a cooling period due to its high thermal conductivity and favourable performance at higher temperatures; therefore the concept developed for this rock includes immediate backfilling of disposal tunnels. A clear understanding of the interactions and limitations associated with heat management during the open period was developed by investigation of a 'Heating Ventilation and Air-Conditioning System' (HVAC). It was concluded that removal of heat from the disposal tunnels / vaults is feasible, but introduces significant operational complexity. (authors)

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

Installation and operation of GeoMelt{sup R} in container vitrification in NNL Central Lab active rig hall at Sellafield - 15328

Cleanup and subsequent immobilization of wastes on the UK's NDA estate brings major challenges. With the exception of HLW where vitrification is utilized for immobilization, cement is typically used in the management and immobilization of non-heat generating wastes. However some legacy feeds such as those containing reactive metals can be incompatible with cementation technologies, and the application of engineered cement encapsulating product to raw wastes increases wasteform volume. Cementation has also been developed to be compatible with the repository concept, which provides it with a level of protection from the environment. Potential delays in repository implementation and as yet to be determined repository waste form acceptance criteria for some wastes may require the duration of above ground storage of immobilized wastes to be extended with potential effect on the integrity of the wasteform. Vitrification is a treatment option which destroys organics and hazardous inorganics such as asbestos, reduces wasteform volume, and produces a robust leach resistant wasteform. Kurion's GeoMelt{sup R} In-container Vitrification (ICV){sup TM} is a vitrification process wherein the treatment vessel can serve as an interim-storage, transport, and final long-term storage vessel - thus negating the need to pour glass into a separate container, effectively leading to a much simpler system than traditional waste melters. The inherent volume reduction from vitrification results in a dense product with self-shielding properties, and the GeoMelt treatment process has been demonstrated to homogenize elemental metals and radionuclide contaminants of concern, eliminating hot spots and mitigating criticality concerns. All of which opens up alternatives strategies for storage and disposal. GeoMelt has been proven in the commercial and US DOE waste treatment environment with over 26,000 tonnes of hazardous waste treated to date (half of this radioactive). Kurion and the UK's National Nuclear Laboratory (NNL) plan to demonstrate its application to Sellafield radioactive wastes. A small (engineering) scale demonstrator is being installed in the active rig hall at NNL's Central Laboratory at Sellafield UK with a twofold mission; 1) demonstration of the vitrification of ILW/LLW, and 2) processing of smaller volume orphan wastes that currently have no treatment path. The NNL facility will allow the handling of uranium containing materials which form a substantial component of ILW currently stored on the Sellafield site. Accurate DFs across the melter and off gas system are often difficult to obtain using inactive materials at the low concentrations present in radioactive wastes. Installation in the NNL facility will allow trace quantities of active species such as Cs-137 and Tc-99 to be processed as part of a simulant ILW stream and monitoring will allow accurate decontamination factors (DFs) to be established to support future vitrification facility design. In addition to underpinning melter operation and obtaining DFs across the system, the facility will allow the processing of radioactive waste arisings of low activity which may either be typical of larger waste volumes or may present as orphans without a disposition route. In either case the melter facility will be able to vitrify these wastes into a form that will be suitable for interim storage and disposal. Among the feeds envisaged for processing will be asbestos from power stations, contaminated soils, feeds containing uranium metals, ponds solids, sludge, and miscellaneous active laboratory wastes. The NNL Central Lab capability will be used to analyze the active product to support disposal requirements whether through letters of compliance for the Geological Disposal Facility (GDF) or waste acceptance criteria for the Low Level Waste Repository (LLWR). (authors)

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

Integrating commercially available remote systems to address current D and D challenges - 15399

Many current D and D challenges across the DOE complex require some form of remote handling expertise and equipment. For example, high visibility projects such as Building 324 D and D and the recovery of wastes from the Alpha-Caissons and the 618 10-11 caissons at Hanford all require the development and application of advanced remote systems. Additionally although alternate solutions and options are being sought, robotic/remote systems will likely be required for completion of D and D at Savannah River Site's Building 235-F, at the West Valley Demonstration Project and many other DOE facilities that are scheduled for D and D. Similar systems are also needed to support the UK cleanup program on projects such as the decommissioning of the Pile Fuel Cladding Silo. As a result, there are a significant number of R and D and Technology Development projects underway in order to address the growing need for remote systems in D and D. In addition to these significant R and D efforts, there are also numerous remote systems in use in multiple applications in both nuclear and non-nuclear industries and the use of remote systems at the Fukushima Daiichi plant has been extensive in the past 12-24 months. In many cases, the systems developed are customized for their given application but there are also multiple systems in use which could be applied to the worldwide D and D challenge either directly or with some minor adaptation. During FY14, the Office of D and D and Facility Engineering (EM13) funded a review/assessment of the current state-of-the-art of robotics and remote technologies in order to generate a baseline of what is available, what has been used where and for what applications and what has been successful and what hasn't. Where possible, the assessment was extended to identify remote systems that have been used in non-nuclear industries but which, with limited modification, may have application in the nuclear industry. The work also included a review of the most recent publicly available TBuRD (Technical Baseline and Underpinning R and D) activities being undertaken by the UK NDA and their Site Licensed Companies (SLCs) in the area of remote systems in order to compile a searchable database of both available, commercial-off-the-shelf (COTS) remote systems and developing technologies. Other suitably qualified organizations in the US and the UK (such as UK National Nuclear Laboratory, SRNL, PNNL ORNL, OC Robotics and others) were also engaged to identify other ongoing technology development efforts which may be of interest and benefit to the decommissioning community. In all, of the order of 500 remote systems were identified and cataloged in both hard copy and as a searchable database. These systems were categorized in terms of their main application e.g. manipulator, wheeled land-based platform, track land-based platform, water-based platform, air-based platform, end effectors, sensor, crawler etc to clearly demonstrate the depth and breadth of available systems in each category. In addition, single page summaries of each system were produced and used, in a workshop setting, to identify technologies which, when integrated with other currently-available technologies, could provide a workable system to address existing needs. An initial workshop was held with CH2 Washington Idaho (CWI) and DOE Idaho in August 2014 to explore how the available systems could be integrated to address current and future D and D needs at that site. (authors)

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

Development of disposal container designs for disposal of high level waste and legacy spent fuel in the UK - 15471

Radioactive Waste Management Limited (RWM), a fully owned subsidiary of the Nuclear Decommissioning Authority (NDA), is responsible for the implementation of geological disposal in the UK. Three illustrative concepts of a geological disposal facility (GDF), corresponding to three generic geological environments - higher strength rock, lower strength sedimentary rock and evaporite rock - have been developed to demonstrate the viability of geological disposal of intermediate level waste, high level waste and spent fuel. It is envisaged that HLW which has been encapsulated in Waste Vitrification Plant (WVP) canisters, advanced gas-cooled reactor (AGR) SF which has been stripped of their spacers and packaged into slotted fuel cans, and PWR SF will be packaged into disposal containers for disposal. A set of schematic designs of these disposal containers was developed in the late 2000's. The design concept for the higher strength rock environment was based on the joint Posiva/SKB disposal container design, and the design concept for the lower strength sedimentary rock and evaporate rock illustrative environments was based on the disposal container concept from Switzerland's National Cooperative for the Disposal of Radioactive Waste (Nagra). The size, mass and dimensions of the disposal containers in the scheme designs vary significantly. The construction method and the lid closing method also vary between the disposal containers for the different waste contents. A design study has indicated that optimization of the number of HLW WVP canisters and the number of AGR slotted fuel cans in the disposal containers, standardizing construction method and closure arrangement across the containers for each geology type, providing handling features at both ends of the disposal container, and balancing of shielding around the containers, could bring significant advantage to the overall GDF programme by reducing the number of disposal containers, reducing GDF footprint, and reducing the extent of excavations, spoil generated and materials that are required for construction. A project to develop the scheme designs to conceptual level was carried in 2011 to 2013. In this project, the above changes were taken into account, technical underpinning of the designs was carried out and manufacturing method was considered. Two disposal container variants for different host rock environments have been developed: - Variant 1, for compatibility with the higher strength rock illustrative disposal system design. This is constructed from copper, with a cast iron insert and is based on the joint Posiva/SKB disposal container design. - Variant 2, for compatibility with the lower strength sedimentary rock and evaporite rock illustrative disposal system designs. This is a steel container based on the NAGRA disposal container design. Each variant has three internal configurations to accommodate 3 HLW vitrified canisters, 16 consolidated cans of AGR SF and 4 PWR spent fuel assemblies. These designs have been shown to bring significant benefit to the GDF designs, bringing a significant reduction to the operating programme and GDF programme cost, including reducing the number of disposal containers by 33% for HLW, 50% for AGR SF and reducing the size of the GDF footprint by 25%. (authors)

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

Achievements and process improvements in processing transuranic waste at the TRU Waste Processing Center - 15488

Wastren Advantage, Inc., (WAI) is the DOE Prime contractor for the TRU Waste Processing Center (TWPC) located in Oak Ridge, TN. This paper highlights the TWPCs specific achievements and process improvements associated with both Contact (CH) and Remote Handled (RH), Transuranic (TRU) waste receipt processing and repackaging for disposal. The five key improvement areas and achievements include: - The TWPC installed, tested and operated successfully the Drum Venting System (DVS) to vent over-packed CH-TRU waste drums in several configurations including 55-gallon drums over-packed in 79 or 85-gallon drums. - A new Remote Drum Opener (RDO) system was designed, tested and placed in service to vent drums discovered in overpack boxes in the Box Breakdown Area and to vent overpack configurations that could not be handled in the standard DVS (e.g., 110 gallon overpacks and multiple drum configurations). - The Cask Processing Enclosure (CPE) approach to repackaging RH-TRU waste contained in concrete casks was developed and implemented - Hot Cell outage to incorporate planned equipment modifications and improvements to maintain processing rates. - To improve certification and segregation of CH-TRU from drums that meet Low Level Waste criteria for disposal at Nevada the TWPC added additional equipment/facilities for Non-Destructive Assay (NDA) and Real Time Radiography (RTR). (authors)

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

Applying fast calorimetry analysis on spent fuel calorimeter - 15620

Swedish Nuclear Fuel and Waste Management Company (SKB) is responsible for the back-end of the Swedish nuclear industry. This includes transport of spent fuel by boat from the power plants to the Central Interim Storage Facility for Spent Fuel (Clab), and in the future, encapsulation of the spent fuel and further transportation to a geological repository. In order to ensure safety for the planned geological repository, the decay heat has to be well known. For this purpose the SKB has performed calorimetric measurements on over 200 pressurized water reactor (PWR) and boiling water reactor (BWR) assemblies since 2003. The results from earlier analyses have been used to improve computer codes predicting heat as well as to acquire deeper knowledge about the fuel inventory. The work complements research being done as part of the Next Generation of Safeguards Initiative - Spent Fuel project, whose purpose is to develop and test nondestructive assay (NDA) technologies to improve safeguards measurements of spent nuclear fuel. The project is working to accomplish multiple goals, one of which is to determine the heat content of spent fuel assemblies. To shorten the measurement time, the temperature increase method was used instead of the equilibrium temperature method. In the temperature increase method, the temperature inside of the calorimeter is lowered below the surrounding water temperature before the measurement, and the temperature is allowed to increase due to the higher surrounding water temperature during the measurement. The rate of the temperature increase when the temperature is the same as the surrounding temperature is then compared to a calibration curve where an electric heater has been used in the same way. The price for the higher throughput is less accuracy and dependency of fuel geometry since the heat flow will be different for different fuel types, especially between PWR and BWR assemblies. Recently at Los Alamos National Laboratory, more sophisticated analysis methods have been considered for the use of calorimetry for treaty verification. These methods aim to predict the equilibrium temperature based on early data and therefore be able to shorten the measurement time while maintaining good accuracy and eliminate the dependency on fuel geometry. Even though the techniques show promise for future calorimetry of spent fuel, certain design features of the current system have been identified to interfere with the accuracy of equilibrium measurements in general such as changing pool temperatures and heat leakage during calibration. (authors)

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

Thermal treatment of plutonium contaminated materials (PCM) Waste - 15714

Considering the high over all costs of waste disposal and the growing requirements for improved quality of the final waste form, the benefits offered by thermal processing become very significant. Key drivers for the application of thermal treatment processes include the reduced volume, improved passive safety, and superior long term stability, of the vitrified waste form products. The project has demonstrated these benefits are derived from the oxidation of the metallic waste fraction, destruction of organic components, and evaporation of entrained water, combined with simultaneous immobilization of radioactive and chemo-toxic elements with in a glass or slag (i.e. a partially crystallised glass) material. Currently a fundamental lack of scientific knowledge and understanding significantly hinders the uptake of thermal treatment processes for the immobilization and disposal of plutonium contaminated material waste. The project will contribute to accelerating the acquisition of knowledge and experience required to support the NDA in deploying thermal technologies as a national asset for ILW treatment. (authors)

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

Remote Retrieval, Disruption and Processing of Legacy Nuclear Waste - 16179

The UK National Nuclear Laboratory (NNL) is currently undertaking, in conjunction with Sellafield Ltd, a series of high profile trials of robotic equipment on a full scale test facility at their Workington Laboratory in England. These trials are designed to underpin and substantiate the processes proposed for the remote retrieval, disruption and processing of legacy Miscellaneous Beta Gamma Waste (MBGW) from a number of different donor plants on the Sellafield site, including some of the older fuel storage ponds and storage silos. These processes will be conducted inside the new Box Encapsulation Plant (BEP) which is due to be commissioned at Sellafield by 2020. These processes will subsequently enable Sellafield Ltd to safely recover legacy Intermediate Level Waste (ILW) from the older storage facilities on site and process it into a safer form which is more suitable for long term storage and then ultimately final disposal in a Geological Disposal Facility. As part of this legacy waste processing within the BEP plant, MBGW waste items will have to be remotely removed from a variety of different types of import skips. This waste will then be transferred, using robots, into the export container, which is a 3 m{sup 3} box liner, where the processed waste will be flood grouted to produce an encapsulated waste package. As part of this processing, any MBGW items consisting of sealed containers will have to be remotely disrupted before being grout encapsulated. This disruption is in order to (a) release any trapped liquors or gases from the containers and (b) ensure full grout infiltration of the container contents, thus minimising the amount of free liquors and voids within the final waste product. The NNL are conducting this technical work to confirm the viability of the proposed concept, processes and equipment. The developed solution consists of a linear layout of processing equipment with a pair of heavy duty, standard industrial, 'commercially off the shelf' (COTS) robots positioned around a Waste Handling Table. These robots have a 500 kg payload capacity and are operated remotely to handle all of the various types of MBGW waste, including the heavy and bulky items. The robots can be operated in manual tele-operation mode as well as pre-programmed automatic mode and have a range of primary and secondary tooling in order to perform the waste processing and disruption tasks as well as undertake general house-keeping and cleaning duties within the cell. The main tools include two types of hydraulic grab for the main waste recovery and handling tasks and a hydraulic shear for the waste disruption activities, but a number of other tools have also been designed and built for ancillary activities, such as a set of general purpose gripper jaws, drilling tool, bolt removal tool, pump tool, spray tool and a number of different housekeeping tools etc. The robot trials were started in November 2014, and have attracted a significant degree of high level stakeholder interest, including the UK Nuclear Decommissioning Authority (NDA), Office for Nuclear Regulation (ONR), Environment Agency (EA) and Radioactive Waste Management Limited (RWML) witnessing a number of the critical demonstration trials. Robots of the type selected are readily available, robust, reliable and easy to use. The trials have been very successful in proving the retrieval, processing and disruption operations, reducing the burden in the legacy waste facilities. Utilising robots to perform these tasks will lower the probability of producing non-compliant waste packages. In addition the robots can increase the consolidation of incoming waste skips into outgoing 3 m{sup 3} box liners, significantly increasing packing efficiencies and reducing the overall number of export containers produced thus lowering the decommissioning and waste storage/disposal costs. The trials have also enabled an assessment of the optimum way to operate the robots, as well as their reliability and maintainability issues. Engineering work associated with recovering failed robots and removing them from a typical Sellafield plant environment and then subsequently replacing them with functioning robots has also been undertaken and further work is now planned. The success of the trials, completed safely to a very challenging timescale, has enabled the use of robots in BEP to be endorsed as a viable option by the key stakeholders. Following a robust decision making process, this was a clear cut decision. The technical development and testing of the robot technology is currently continuing in order to investigate areas such as robot reliability, failure recoveries, human factors, ease of waste identification, optimisation of throughput rates, nuclearisation modifications etc. (authors)

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

Cementitious Grouts for ILW Encapsulation - Hydration and Continuity of Supply within the UK - 16184

In the UK the preferred route to disposal for the majority of intermediate level nuclear waste (ILW) streams is encapsulation within a cementitious matrix, accounting for 83% of waste conditioned to date (NDA, 2013). The method is well-defined and produces a waste package that is both chemically and physically stable, providing a multi-layered barrier to the release of the waste species contained within. Due to the process of treatment, facilitated via remote operation, it is necessary for the grouting matrices to meet strict performance limits in terms of their flow, setting time and avoidance of the formation of bleed water or segregation. To this end, the limits on composition of the anhydrous materials (a mixture of ordinary Portland cement [OPC] and blast furnace slag [BFS]) are tight to ensure conformity. Whilst these limits on composition are well understood, their impact upon grout hydration, and hence performance, is poorly defined. This study investigates the effects of modifying the physical and chemical composition of both the OPC and the BFS components of the matrix in order to identify the sensitivity or otherwise of the system to ensure that a continuity of supply is maintained. A number of techniques, scanning electron microscopy - image analysis (SEM-IA), chemical shrinkage (measured via dilatometry), isothermal conduction calorimetry (ICC), thermogravimetric analysis (TGA) and X-ray diffraction (XRD), have been applied to assess the rate and degree of hydration (DoH) of a range of grouts prepared with OPC and BFS with a range of chemical and physical compositions. This has been coupled with subsequent engineering performance testing, with additional analysis to understand the impact of a fire scenario within a storage or disposal facility, to provide an indication towards the resilience of the waste packages to a potential catastrophic event. (authors)

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