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DOE OSTI2019

Unlocking the Vault - Accelerated Delivery Through the Application of Innovation and Collaboration - 19011

The Pile Fuel Cladding Silo (PFCS) at Sellafield was built in 1949/50 to receive and store fuel cladding from reprocessing of Windscale Pile fuel; it has been used to store this material and other waste from the UK nuclear programme ever since. The Silo is unsuitable for indefinite storage; this along with uncertainty in the make-up and condition of the inventory combine to rank the facility alongside the most hazardous in Western Europe. Sellafield Ltd, began working with its supply chain partners Bechtel Cavendish Nuclear Solutions (BCNS - a joint venture between Bechtel Inc and Cavendish Nuclear Ltd) in 2012, to execute a suite of projects to create an access route to the waste and a capability to retrieve waste from the silo for the first time ever. However, the challenge of providing this capability, while maintaining ongoing safe storage of the waste, and underpinning future performance of both the silo and the waste packages for many hundreds of years, proved to be too much for traditional Sellafield Ltd delivery models. With both cost and schedule forecasts doubling from initial estimates, a new approach was clearly required. The team recognised that success would require all of the experience and expertise of each organisation working together to a common goal. Therefore, significant effort was expended to build a collaborative team environment into which an innovative approach to both the engineering of what was provided and the delivery methods for doing it could be launched. Focussing on collaborative culture in which innovation and fit for purpose thinking were encouraged has completely rebuilt stakeholder confidence. The project is now almost complete; 17 months ahead of schedule and British Pounds 100 m less than the approved cost. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Developing Tomorrows Nuclear Professionals - Apprentice and Graduate Development at PFCS, Sellafield - 19013

The Pile Fuel Cladding Silo (PFCS) at Sellafield, originally built in 1949/50 to receive and store fuel cladding from reprocessing Windscale Pile fuel, has been used to store this material and other waste from the UK nuclear programme ever since. However, the Silo is unsuitable for indefinite storage of the waste; its age, condition and inventory, combined with significant uncertainty in all these factors, rank the facility among the most hazardous in Western Europe. In 2012, Sellafield Ltd began a highly collaborative relationship with its supply chain partners Bechtel Cavendish Nuclear Solutions (BCNS), to undertake a suite of projects to create a capability for retrieving waste. In doing so, the team has discovered a shared passion for graduate student and apprenticed trades development, which has led to the team goal of becoming the number one destination for training secondments within the associated businesses. This paper describes how the team has leveraged the unique challenge of the project, the extensive experience of the delivery team, and the capabilities of the contributing organisations to create a structure secondment experience aimed at maximizing learning and generating enthusiasm for nuclear clean-up. The scheme was developed and implemented, based on some very simple but consistent feedback from the trainee community on what differentiates a good and bad secondment. The paper includes three case studies from individuals who have participated in the programme and the analysis of overall feedback from all programme trainees. Universally, the scheme has been well received with a significant number of trainees coming back to the project to begin their professional careers. The simple differentiators between a good and bad secondment provide a good basis for those wishing to develop similar schemes. (authors)

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DOE OSTI2019

Revisit of Phase Separation of Borosilicate Glass for Vitrification of High Level Radioactive Waste - 19166

Phase separation of glasses for vitrification of high level radioactive waste has been experimentally studied to know uncertainty of the phase separation conditions; temperature and glass compositions as well as to propose possibility of recycling of heavy metals on a longer time scale. Our current results at 700 and 750 deg. C for ternary borosilicate glass system showed that the phase separation region at 700 deg. C spreads along with a direction of increasing Na{sub 2}O composition when compared with the data in literature. On the other hand, at the glass composition employed in the reprocessing plant in Japan, no segregation was observed. Concerning the application of phase separation phenomena, we had previously reported recovery of almost all elements from the vitrified borosilicate glass containing simulated fission products through phase separation of the borosilicate glass and this suggests that we could propose that after vitrification of high-level radioactive waste heavy metals may be retrieved as energy or materials resources when they are required to utilize on a longer time scale. From this standpoint, phase separation of lead borosilicate glass was also studied to seek economical and versatile recycling. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

La Hague STE2 Sludge Retrieval and Conditioning Strategy - 19229

About 9000 m{sup 2} of Long-Lived Intermediate Level Waste LL-ILW STE2 sludges have been generated by the reprocessing operations of UNGG and later UOx fuel in La Hague (UP2-400) and subsequent liquid effluent treatment. They were placed into a set of adjoining tanks, called the STE2 silos. Today, the STE2 Sludges have to be retrieved and conditioned for final disposal in the deep geological repository under development in France (Cigeo). To this end, since the early 2000's, different retrieval and conditioning scenarios have been evaluated. This paper tracks the history of the technical and non-technical key factors having led to the choice of the baseline scenario developed until the end of 2016, describes the technical achievements and makes a state-of the art of the current scenario consisting in retrieving and temporarily storing STE2 sludges and, meanwhile, studying, researching and testing final conditioning processes and technologies in the aim of achieving a technical and economic optimum. (authors)

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DOE OSTI2019

ANSTO Synroc Wasteforms for the Immobilization of a Range of Mo-99 Production Wastes - 19326

Tc-99m, the most commonly used radioisotope in nuclear medicine, is produced from the decay of Mo-99 which is predominantly produced in reactors by irradiation of enriched uranium targets. This results in three waste streams, one of which is a uranium filter cake containing fission products. This filter cake could itself be treated as a waste stream, or processed to retrieve the enriched uranium for further use, thus producing additional waste streams during reprocessing. All wastes are considered intermediate-level wastes and require an appropriate wasteform design for termination. ANSTO is collaborating with NECSA (Nuclear Energy Corporation of South Africa) to develop wasteform and processing solutions for the treatment of intermediate level wastes (ILW) from the production of Mo-99. The work is supported by the U.S. Department of Energy's (DOE) National Nuclear Security Administration (NNSA) with technical support from Argonne National Laboratory (ANL). Here we highlight some of the outcomes of this collaboration which has seen the investigation and performance testing of a range of waste treatment technologies. Among the treatment options for the various waste streams, including the enriched uranium filter cake itself, were glass, glass-ceramic, ceramic and geopolymer wasteforms. All wasteforms were characterized (e.g., by X-ray diffraction (XRD) and scanning electron microscopy (SEM)) and their performance characteristics, for example compressive strength and aqueous durability, determined by standard methods and compared. Consideration has been given to the most suitable, practical and cost effective solutions for the immobilization of these Mo-99 wastes. This paper will discuss the down-selection of wasteforms and process technologies for Mo-99 wastes and compare these with regards to wasteform performance and process technology readiness. (authors)

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DOE OSTI2019

Treatment of Foreign Radioactive Waste at the BELGOPROCESS Plant - 19345

Belgium has used a large number of nuclear technologies during the last 60 years. Power generation, reprocessing, fuel fabrication, isotope production and research activities have produced a large variety of radioactive waste streams. BELGOPROCESS is the operator of the centralised waste treatment facilities located in the Campine region in Flanders. Over the years BELGOPROCESS has developed the know-how and facilities to treat each one of these waste streams. BELGOPROCESS is a subsidiary of NIRAS, the governmental institution responsible for the safe management of radioactive waste and enriched fissile materials in Belgium. A number of these waste streams are only produced during a short period of time. After processing the existing stockpile the treatment facility is put into operational standby and eventually prepared for decommissioning. In order to treat the stockpile of some types of legacy waste within an acceptable time frame, the treatment facilities needed to be designed at a certain treatment capacity. Since the majority of the legacy waste is processed nowadays, a significant amount of treatment capacity becomes available. In June 2006 the Belgian government decided that this know-how and facilities should become available to foreign waste producers. This would allow BELGOPROCESS to fill up the available capacity of its facilities and consequently lower the cost for the Belgian waste producers. Nowadays two facilities are open to accept foreign radioactive waste. The CILVA-facility can be used for sorting, cutting, incineration, precompaction, supercompaction and cementation of low level solid waste. The PAMELA-facility - a former vitrification plant - can be used for the treatment of medium level and alpha-bearing waste. In this paper we will discuss the technical characteristics of the CILVA-incinerator and a showcase for the treatment of foreign waste at our facilities. In 2016 BELGOPROCESS started a cooperation with MATE-FIN for the incineration of low level burnable waste from the Cernavoda NPP in Romania. (authors)

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DOE OSTI2019

A Python Package for Fuel Cycle and Waste Management Analysis - 19545

The future of nuclear power utilization in the United States may involve a transition from the once-through light water reactor fuel cycle to a so-called advanced fuel cycle. An advanced fuel cycle may offer improvements in natural resource utilization, reactor safety, and environmental impact. In addition, adoption of an advanced fuel cycle will result in changes in the amount, composition, and disposal of the nuclear wastes they produce. Characteristics intrinsic to the wastes themselves, such as mass and radioactivity at specific times, were used as waste management metrics in the Department of Energy 2014 Fuel Cycle Evaluation and Screening (FCES) study. Binning of these metrics informed fuel cycle according to relative potential benefit. However, waste properties alone cannot qualify management strategy. Therefore, comparing fuel cycles from the perspective of waste management can provide additional insights. To that end, an object-oriented Python package has been developed to perform fuel cycle waste management calculations. The package, named nwpy (short for Nuclear Waste analysis in Python), mimics the mass flow operations reported in the FCES study and interfaces with the ORIGEN-S code to assess metric data as reported in the FCES. In addition, this Python package determines the loading of waste streams into forms suitable for final disposal in a geological repository and uses such data for repository performance assessment. This work describes the Python package methodology, assumptions, and capabilities. As input data, the package requires information about the fuel cycle such as the mass of fuel discharged, the cooling time between reactor discharge and subsequent operations, and the separations process efficiencies and waste forms it produces. In addition, the isotopic composition of the reactor discharge stream, such as that available from the FCES database, is required. With that, it can perform prescribed mass flow and decay operations throughout the back end of the fuel cycle, the latter carried out by the ORIGEN-S code distributed with the SCALE6.2 software suite. The result is time-dependent data characterizing the waste on a per-package basis. Package loading for waste streams depends on the stream form (i.e. UNF vs. HLW, or glass HLW from aqueous reprocessing vs. ceramic HLW from electrochemical processing) and is based on literature review and engineering assumption. As an example of the application of the code, calculations for the areal footprint of direct-contact repositories for different fuel cycles is shown. Development is ongoing, and the package is expected to grow incorporating capabilities for waste management analyses such as waste transportation and radionuclide transport through the biosphere. (authors)

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DOE OSTI2019

Progress Towards Treating the Sludges Produced During Spent Fuel Processing - 19546

As a consequence of over 50 years of Magnox reprocessing, Sellafield site is in possession of significant quantities of Magnox sludge. This sludge material is the by-product of corrosion of the magnesium based Magnox metal that clads the Magnox fuel elements. Retrievals and processing of this material poses a significant challenge as corrosion of Magnox metal has occurred under a variety of conditions and over a significant period of time, in a variety of locations. Devising processes to treat heterogeneous materials offers a significant challenge to identify technologies with robust envelopes of acceptance, to effectively operate together. This is further compounded by difficulties defining test programmes and materials for testing that will provide assurance that nonradioactive testing is suitably representative or bounding. An overview of the significant body of work undertaken to process Sellafield's Magnox sludge is provided, from the thousands of commercially sensitive reports generated over the past five decades. This covers the sampling, characterisation, strategy and process evolution, technology development and test materials. (authors)

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DOE OSTI2019

Developments in the Regulatory Framework of Nuclear Waste Management in Germany - Licensing Procedures for storage of SNF and Public Involvement - 19612

According to article 6 of the German Atomic Energy Act (AtG), storage of Spent Nuclear Fuel requires a license. The competent licensing authority in this field is the Federal Office for Safety of Nuclear Waste Management (BfE). According to the Act for selecting a repository site for heat-generating radioactive waste (HAW) (Site Selection Act - StandAG) dating from 2017, BfE has been assigned the task of developing, conducting, and evaluating public participation procedures in the process of site search and selection for a deep geological repository for radioactive waste and SNF. Article 6 AtG applies to interim storage of SNF. If the legally defined licensing conditions by article 6 (2) AtG are met, the storage license applied for has to be granted. For on-site facilities, interim storage has been limited to 40 years. The licensing procedures for interim storage of SNF are carried out in accordance with the general administrative procedural rules. In certain cases due to the Environmental Impact Assessment Act (EIA Act), an EIA has to be carried out. In these cases, a formalized licensing procedure including an involvement of the public is conducted. The Nuclear Licensing Procedure Ordinance (AtVfV) originating from 1977 and reformulated in 1995 defines formalized rules for the public participation procedure which have to be followed strictly. These include a public hearing. The risen objections are evaluated by the licensing authority and taken account of in the licensing decision. The process of repository site selection defined by the StandAG exceeds these formalized procedures significantly: The objective of this process is to achieve a broad social consensus about the site selection procedure and about a repository site that will offer the best possible safety for one million years. Therefore, the process has to be particularly transparent, and involvement of the public - including newly developed forms of public involvement - is an essential part of it. BfE as competent authority in both the fields of repository site search and selection and licensing of interim storage of SNF has recently developed new strategies of involving the public concerning interim storage of SNF. Supplementing the standard licensing procedures, BfE presently informs and involves the public about topical issues regarding the present state and recent developments in the field of interim storage of SNF. Current central topics are for example the return of HAW and MAW from reprocessing of SNF in France and UK as well as perspective longer interim storage periods on the interface with final disposal. Regarding longer interim storage, host communities of interim storage facilities are concerned that interim storage might be extended indefinitely. Additionally, there is a strong fear of transports of radioactive waste and SNF from one place to another. (authors)

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DOE OSTI2019

Hanford Tank Side Cesium Removal System: Process Description and Performance Metrics - 19613

The Tank Side Cesium Removal (TSCR) System is a modular, quickly-deployable and passively safe system for efficiently removing solids and cesium from alkaline liquid waste raffinate created by spent fuel reprocessing on the Hanford site. It further describes TSCR component design features, system performance metrics and operational sequences used for liquid waste treatment, as well as spent ion exchange column (IXC) replacement and storage. This presentation should be of great interest to parties responsible for treatment and disposal of liquid waste containing high concentrations of cesium. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Start-up of Silo 130 Waste Retrieval at La Hague: first Step Towards Reducing Legacy Inventory - 20020

On June 25, 2019, the first grapple of waste and debris was lifted from Silo 130, one of the most challenging legacy waste storage silos at La Hague. This was the achievement of more than ten years of studies and efforts focused on solving some very unique technical and safety challenges, and on designing an innovative, efficient solution. The waste is from de-cladding materials from past Gas-Cooled Reactor fuel reprocessing operations. It is an unsorted mixture of magnesium cladding pieces, graphite debris, and various other metallic waste types. These materials can be chemically reactive and therefore need specific dispositions for handling and storing. The building is more than 40 years old and does not meet modern standards. French safety authorities have required the silo to be emptied. After about three years of operation, the bulk of this waste will be removed and placed in safe storage, pending the selection of a final conditioning solution. Removal of bottom debris will then be possible. When finished, this operation will bring a conclusion to a significant legacy-related hazard on the site. The silo will then be available for final cleaning and de-activation. This paper will describe the technical, safety, and economical challenges, and the various decisions and optimizations that led to the definition of the process and technology to be implemented. Several innovative features were developed and implemented, in all sections of the project: project organization, progressive approach for the waste disposition strategy, light building design, customized and robust retrieval technology, processing technologies involving the most modern robotic and sorting techniques, storage method... Several of these first-of-a-kind innovative features have been developed and qualified specifically for this application, using the detailed qualification procedure specified for all La Hague projects. The facility was built in less than five years, overcoming various difficulties along the line. Commissioning is now finished with the combined efforts of the project, engineering, start-up, and operating teams assisted by dedicated suppliers. The facility is now ramping-up. The paper will also report on the commissioning and start-up of the facility and describe the first operational results. (authors)

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DOE OSTI2020

Alternative Treatment of Defense Waste Processing Facility Recycle via Reuse of Existing Liquid Waste Facilities - 20097

The Defense Waste Processing Facility (DWPF) at Savannah River Site (SRS) has been immobilizing high level waste since 1996. The chemical process within DWPF generates a large volume of condensate, which is recycled to the SRS H-Area Tank Farm. The recycle stream includes a small quantity of sludge solids, as well as soluble cesium that is volatilized during melter operation. The recycle waste is currently received into a large, underground waste tank that separates insoluble solids via decanting. The supernate is treated by an evaporator with a concentrate stream that is stored for future processing. The evaporator overheads are collected and sent to the Effluent Treatment Project (ETP) for final polishing and testing prior to discharge to local surface water. Recycle storage and treatment as described above complicates the overall mission within the tank farms, which are primarily engaged in waste retrieval and preparation activities that support sludge and salt disposition, as well as tank characterization and closure. The need to devote a portion of available storage space to recycle treatment limits operational flexibility, and ultimately the DWPF recycle stream must be diverted to fully close all the SRS waste tanks. An alternative treatment process is being explored to decouple the recycle stream from the tank farm. The proposed treatment process will accomplish solids separation via crossflow filtration and will utilize a wiped film evaporator to volume-reduce the filtrate stream. Evaporator overheads will continue to be further processed in ETP while the solids stream and evaporator concentrate stream will be returned for reprocessing with the DWPF and Salt Waste Processing Facility (SWPF). This process will utilize existing facilities within DWPF and the tank farm that were previously dedicated to Interim Salt Disposition (ISD), but no longer have an identified mission with the startup of SWPF. The reuse of these facilities will remove several constraints from the current Liquid Waste (LW) System Plan without expanding the current footprint of Department of Energy Environmental Management (EM) infrastructure within LW. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Sampling and Analysis Plan for Investigation of Undeclared Nuclear Activities - 20104

The nuclear-centered energy policy in South Korea is moving towards a nuclear-free era. It also facilitates an action or progress on the denuclearization of the Korean peninsula. The Nuclear Safety and Security Commission (NSSC) launched a new R and D project which attempts to verify nuclear activities via analyzing environmental samples. Environmental samples possibly contain significant radiological information relevant to past undeclared activities. In order to precisely verify them, it is essential to provide 1) the concrete information of target facilities, 2) appropriate methodologies and procedures, and 3) well-suited instrumentation and equipment corresponding to sample signatures. It is also important to provide scientific reliability or validity in any process such as sample collection, handling, transportation and laboratory analysis. Therefore, the information of nuclear facilities (fuel fabrication plant, graphite moderated reactor, reprocessing plant) in North Korea are reviewed in order to determine parameters for appropriate sampling methods representing a facility's characteristics. Secondly, a sampling procedure for radioactive elements is being developed based on international and domestic standards such as ASTM (American Society for Testing and Materials), KS (Korean industrial Standards), and International Atomic Energy Agency (IAEA) manuals. The procedure contains some specific information including sampling process, sampling tools, amount of sample, sample identification, and other related information. Developing the qualified national guidelines is most important since sampling and analyzing results can be used as evidences of nuclear activities. The guideline developed through this work will clearly define various methodologies such as document verification, sampling procedures, collection methods, sample storage criteria, chain of custody, and transportation conditions. The final objective of this work is to establish the infrastructure and regulatory framework for assessment of nuclear activities, and provide a national guideline for sampling and analysis. This will bring confidence in the results and play a critical role in complete denuclearization progress in South Korea. (authors)

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2020

The TRANSCEND University Consortium: Theme 4: Nuclear Materials - 20431

The safe and secure management of Pu is a matter of international concern, with ∼250 t of separated Pu currently stockpiled worldwide. The UK's civil inventory of nuclear materials contains significant stocks of separated Pu from the reprocessing of Magnox and AGR spent fuels. The preferred option for the 138.5 tonnes of Pu is re-use as mixed oxide (MOx) fuel, although 5% is not suitable for re-use and is recommended for direct disposal. However, it will take more than 15 years to implement re-use, requiring that the Pu be kept in interim storage in its current state for that period, i.e. as PuO{sub 2} powder within inert steel storage cans at Sellafield. The focus of the work presented here is thus plutonium storage and the direct disposal of plutonium. The Research and Development needs of both are now pressing: in the case of storage due to it being the current default; in the case of immobilization and disposal because of a comparative lack of Research and Development on Pu conditioning and packaging due to policy uncertainty as to whether it would be disposed of in a Geologic Disposal Facility (GDF). Addressing these needs is complicated by Pu's high radioactivity, decay heat and radiotoxicity, criticality, nuclear safeguard requirements and, for some UK Pu contaminated materials targeted for disposal, poor inventory. Thus, there is also a critical requirement for underpinning research on Pu bearing materials in these two contexts. In response to these needs, the TRANSCEND Consortium (Transformative Science and Engineering for Nuclear Decommissioning, a multi-disciplinary collaboration of 11 universities and 8 key industry partners from across the UK's civil nuclear sector) is seeking to provide technical underpinning to ongoing option development for the UK's civil Pu stockpile. Whilst understanding the behaviour of plutonium during its re-use as MOx is beyond the scope of the TRANSCEND Consortium work plan, the main objectives of the work are: (1) For interim storage: to understand how the surface structure and properties of pristine and radiation damaged PuO{sub 2} change with time in the absence and presence of water; and (2) For immobilization and disposal: to understand the mechanisms of incorporation of Pu into ceramic and glass-ceramic waste-forms, as well as the effect on these of self-induced radiation damage. Each objective is being addressed through separate work packages, the details of which are discussed in this paper. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

A Decade of Innovative Approaches to Treating and Dispositioning Radioactive Waste at the Savannah River Site - 20620

This year marks the 70. anniversary of the U.S. Department of Energy's (DOE) Savannah River Site (SRS), located near Aiken, South Carolina. SRS is a key industrial complex responsible for environmental stewardship, environmental cleanup, waste management, and disposition of nuclear materials. SRS also continues to have a role in critical defense-related activities and the reprocessing of used reactor fuel. The SRS encompasses 803 square kilometers (310 square miles) in parts of Aiken, Barnwell, and Allendale counties. The Liquid Waste Mission starts with the safe receipt and storage of radioactive liquid waste in the waste tanks, which principally includes Cold War legacy waste, but also waste from support of National Aeronautics Space Administration missions, medical isotope production, and research activities. This 133,000 cubic meters (m{sup 3}) (35 million gallons [Mgal]) of high-level radioactive waste (HLW) is currently held in 43 large underground waste tanks. The capacity of each of these tanks range from 2,800 m{sup 3} to 5,000 m{sup 3} (0.75 to 1.33 Mgal) and were placed in operation between 1954 and 1986. In July 2009, the DOE awarded a stand-alone contract to execute the Liquid Waste Mission at SRS. The Liquid Waste Contractor selected was Savannah River Remediation LLC1 (SRR), which is responsible for receipt, storage, retrieval and treatment of all HLW, disposal of the decontaminated low-activity waste fraction, as well as operationally closing cleaned HLW tanks. Since July 2009, SRR has successfully grouted and operationally closed six large underground tanks, poured 1,476 canisters of vitrified HLW (a stable glass waste form), treated approximately 34,200 m{sup 3} (9.0 Mgal) of HLW, and dispositioned over 41,200 m{sup 3} (10.9 Mgal) of decontaminated salt solution as low-level waste (LLW) into the Saltstone Disposal Units (SDUs) being constructed by SRR on site. The innovative work performed by SRR in the LLW portion of its mission has earned it the Richard S. Hodes Award. The work conducted by SRR is the same work championed by the late Richard S. Hodes, a respected physician, statesmen, and Chairman of the Southeast Compact Commission for Low-Level Radioactive Waste Management. (authors)

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DOE OSTI2020

Evaluation of Material Balance Approaches for Hanford Direct-Feed Low Activity Waste Processing - 20022

The Hanford Site has accumulated millions of gallons of tank waste from reprocessing spent fuel to recover plutonium, uranium, cesium, and strontium. The supernatant from the accumulated tank waste will be treated using a Direct Feed Low-Activity Waste approach. The supernatant will be treated to remove solids and cesium in the Tank-Side Cesium Removal process in the Hanford tank farm, then vitrified in a semi-batch process in the Hanford Waste Treatment and Immobilization Plant (WTP). Currently, each batch of feed is sampled at three locations prior to being fed to the melter: the feed qualification tank in the Hanford tank farm as well as the concentrate receipt vessel (CRV) and melter feed preparation vessel (MFPV) in the WTP. The feed qualification sample is taken from a large batch of accumulated feed, only two to three samples are expected each year. Approximately 275 samples from the CRVs and 1100 samples from the MFPV are expected each year. An evaluation was performed to determine if a material balance based on the feed qualification sample could replace most of the sampling in the CRVs and MFPVs. The evaluation consisted of three elements: (1) determination of the practicality of using a material balance to estimate the stream composition of the CRV and MFPV contents, (2) evaluation of whether the material balance could be automated using the existing process control system, and (3) determination of the uncertainty in glass composition using the material balance approach. It is assumed that periodic sampling at the CRV and MFPV would be performed periodically to re-baseline the material balance, evaluations are in progress to determine the frequency of this periodic sampling. Process sample locations downstream of the melter were reviewed as well, but the partitioning of semi-volatile species in the melter was determined to preclude extending the material balance approach past the melter. It was determined that replacement of the CRV sample location was feasible and did not increase process uncertainty or significantly impact waste loading. Replacement of the MFPV sample was also determined to be feasible, but that measurement of the glass former chemical addition may be needed prior to addition of these chemicals to the MFPV. This measurement could be performed by an in situ laser-induced breakdown spectroscopy (LIBS) system. Limited tests were performed to evaluate LIBS for direct measurements of the low-activity waste melter feed. The use of a material balance would eliminate over 1200 samples each year if only 10% of the CRV and MFPV batches are sampled and could likely allow the WTP laboratory to operate on days only versus 24/7 operation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Progress towards the POCO of UK Highly Active Storage Tanks - 20112

In the UK Highly Active (HA) waste is stored in stainless steel Highly Active Storage Tanks (HASTs) prior to encapsulation in glass in one of three operating vitrification plants. Significant progress has been made towards the reduction in the volume of HA Liquor (HAL) stored in these HASTs and several tanks currently operate at heel levels. In recent years a strategy has been developed towards the Post Operational Clean Out (POCO) of the tanks and significant progress has been made in confirming this strategy since it was last presented at WMS [1]. For a variety of reasons changes to associated operational assumptions have also been made. This paper will highlight recent progress and changes associated with the strategy, including a description of some of the challenges that have been overcome and associated technical work undertaken. It will also detail current plans to POCO the HASTs, as quickly as possible, while minimising the risk and continuing to support other ongoing reprocessing and clean-up programmes as part of the overall UK clean-up mission. (authors)

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DOE OSTI2020

Estimation of Sorption Behavior of Europium to Calcium Silicate Hydrate by Thermal Analysis - 20172

High-level radioactive waste generated from reprocessing processes will be disposed in a geological disposal that is deep underground. Calcium silicate hydrate is formed as a secondary mineral in the repository via the reaction of calcium ions leaching from the cement and the silicic acid dissolved from the host rocks. In this study, the sorption behavior of europium, which is also known as a chemical analog of americium, on calcium silicate hydrate was examined. The solid phase was analyzed via thermogravimetry- differential thermal analysis, differential scanning calorimetry, Raman spectroscopy, fluorescence spectroscopy, and fluorescence lifetime analysis. To understand the behavior of calcium silicate hydrate under conditions saturated with groundwater, the study samples were analyzed using the above methods, except for deferential scanning calorimetry, without a drying process. The results of the differential scanning calorimetry analyses showed that the temperatures for the endothermic peaks of calcium silicate hydrate for the lower Ca/Si ratios were larger than those for the higher Ca/Si ratios for coexistent conditions of europium ions. Hence, the relationships of the temperatures of the endothermic peaks and the Ca/Si ratios were reversed by the addition of europium ions. These study results suggest that calcium silicate hydrate can immobilize europium ions, thus preventing the production of colloidal forms. (authors)

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