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At least 577 records · Page 32
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Recent Developments in the Management of Cameco Corporation's Fuel Services Division Waste - 13144

Cameco Corporation is a world leader in uranium production. Headquartered in Saskatoon, Saskatchewan our operations provide 16% of the world uranium mine production and we have approximately 435 million pounds of proven and probable uranium reserves. Cameco mining operations are located in Saskatchewan, Wyoming, Nebraska and Kazakhstan. Cameco is also a major supplier of uranium processing services required to produce fuel for the generation of clean energy. These operations are based in Blind River, Cobourg and Port Hope, Ontario and are collectively referred to as the Fuel Services Division. The Fuel Services Division produces uranium trioxide from uranium ore concentrate at the Blind River Refinery. Cameco produces uranium hexafluoride and uranium dioxide at the Port Hope Conversion Facility. Cameco operates a fuel manufacturing facility in Port Hope, Ontario and a metal fabrication facility located in Cobourg, Ontario. The company manufactures fuel bundles utilized in the Candu reactors. Cameco's Fuel Services Division produces several types of low-level radioactively contaminated wastes. Internal processing capabilities at both the Blind River Refinery and Port Hope Conversion Facility are extensive and allow for the recycling of several types of waste. Notwithstanding these capabilities there are certain wastes that are not amenable to the internal processing capabilities and must be disposed of appropriately. Disposal options for low-level radioactively contaminated wastes in Canada are limited primarily due to cost considerations. In recent years, Cameco has started to ship marginally contaminated wastes (<500 ppm uranium) to the United States for disposal in an appropriate landfill. The landfill is owned by US Ecology Incorporated and is located near Grand View, Idaho 70 miles southeast of Boise in the Owyhee Desert. The facility treats and disposes hazardous waste, non-hazardous industrial waste and low-activity radioactive material. The site's arid climate, deep groundwater and favourable geology help ensure permanent waste isolation. Combined with a state of the art multi-layer landfill liner system, the Grand View facility represents an ideal choice to minimize environmental liability. Marginally contaminated wastes from operations within the Fuel Services Division are typically loaded into PacTec IP-2 rated Intermediary Bulk Containers and then transported by road to a nearby rail siding. The Intermediary Bulk Containers are then loaded in US Ecology owned gondola rail-cars. The gondolas are then transported via Canadian Pacific and Union Pacific railroads to the US Ecology Rail Transfer facility located in Mayfield, Idaho. The Intermediary Bulk Containers are unloaded into trucks for transport to the disposal facility located approximately 32 miles away. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Vitrification of Hanford Tank 241-AP-101 Waste and Simulant

Hanford tank 241-AP-101 (referred to herein as AP-101) is the second Hanford radioactive tank waste planned to be processed and vitrified. A simulant version of AP 101 waste was formulated from the best-basis inventory (BBI) for the Hanford Tank 241-AP-101 liquid with an assumed target dilution of the waste from the BBI sodium molarity of 8.61 M to the desired 5.5 M Na. After the addition of glass-forming chemicals (GFCs), the simulant melter feed was processed in a non-radioactive, continuous laboratory-scale melter (CLSM) system. The AP-101 simulant melter feed was charged into the CLSM for 6.11 h of processing, which produced 6.55 kg of glass, for an average glass production rate of 2275 kg m 2 d -1 . Since there were no processing issues with the AP-101 simulant melter feed, AP-101 melter feed made with actual waste was then processed in a CLSM system built into a contamination area in a radioactive environment. The melting behavior characteristics appeared similar for both the simulant and waste melter feeds. The AP-101 waste melter feed was charged into the CLSM for 12.14 h of processing, which produced 8.75 kg of glass, for an average glass production rate of 1530 kg m 2 d -1 . During the AP-101 waste melter feed charging, the pump used to move the feed reached a maximum and it is believed that if the pump had a greater capacity, a greater average glass production rate could have been achieved. A constituent of interest present in low quantities in the AP-101 waste is 99 Tc or its non-radioactive surrogate, Re, added to the AP-101 simulant. Analysis for the quantities of 99 Tc and Re in the AP-101 glass product resulted in an average single-pass retention from the melter feed during relative chemical steady state of 55 ± 2 % for 99 Tc and 45 ± 2 % for Re. Compared to the processing of other melter feeds, the retention of 99 Tc in the AP-101 glass was greater than in both AP-107 and AP-105 glass, while the retention of Re in the AP-101 was less than in the AP-107 glass, but greater than in the AP-105 glass. A spike of I was added into the AP-101 melter feed that could be detected above the analysis detection limits. However, the iodine was only detectable above the ~6 ppm limit in one glass pour: the pour immediately following the burn off of the cold cap, where the I level reached ~30 ppm. This event was significant because the glass was poured immediately after burn off and thus it is presumed that the iodine had yet to volatilize from the glass melt while idling. It is recommended to perform future tests with I spikes at greater levels so that it can be detected in additional glass pours to determine if the expected 50 % retention of I used in the Kim et al. glass models can be confirmed. Offgas liquid samples were analyzed for acetonitrile, which was present at greater concentrations in CLSM liquids than in other scaled melter systems. This result was expected based on unique conditions with the CLSM system including a small plenum space leading to low residence time for offgas and the rapidity of offgas cooling upon exiting the CLSM vessel due to the location and environment. About 90 % of the total acetonitrile captured during both the AP-101 simulant and waste CLSM runs was found in the offgas condensate and demister liquids, thus it is recommended that only those liquids be sent for analysis if future testing to study the presence of acetonitrile in offgas products is desired.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Evaluation of Alternative Filter Media for the Rotary Microfilter

The Savannah River Site is currently developing and testing several processes to treat high level radioactive liquid waste. Each of these processes has a solid-liquid separation process that limits its throughput. Savannah River National Laboratory researchers identified and tested the rotary microfilter as a technology to increase solid-liquid separation throughput. The authors believe the rotary microfilter throughput can be improved by using a better filter membrane. Previous testing showed that asymmetric filters composed of a ceramic membrane on top of a stainless steel support produced higher filter flux than 100% stainless steel symmetric filters in crossflow filter tests. Savannah River National Laboratory and Oak Ridge National Laboratory are working together to develop asymmetric ceramic ? stainless steel composite filters and asymmetric 100% stainless steel filters to improve the throughput of the rotary microfilter. The Oak Ridge National Laboratory Inorganic Membrane Group fabricated samples of alternative filter membranes. In addition, Savannah River National Laboratory obtained samples of filter membranes from Pall, Porvair, and SpinTek. They tested these samples in a static test cell with feed slurries containing monosodium titanate and simulated sludge.

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DOE OSTI2020

Computationally-Aided Design of a Small-Scale Radioactive Waste Glass Melter

To provide mission support to the Hanford Waste Immobilization and Treatment Plant for the vitrification of legacy nuclear tank waste, a reduced-scale vitrification pilot system is being designed to process simulated and actual radioactive tank wastes. The tank waste will be separated into high-level waste (HLW) and low-activity waste (LAW) fractions, where the majority by mass (~90%) is LAW and by activity (>95%) is HLW. The first tank waste to be processed at the WTP will be LAW. Since Hanford tank waste and the resultant melter feed compositions are known to vary widely, pilot-scale operations are essential to identify potential problems, provide needed data, confirm assumptions, and determine impacts to full-scale melters and off-gas systems from these different feeds. In addition, the reduced-scale melter system must be capable of quickly providing results to operations without the typically high costs of radioactive operations and provide an engineering platform in close proximity to local operations staff. The objective is to produce similar process conditions to those encountered in the full-scale LAW melters and minimize the volume of radioactive waste necessary for the evaluations. To this end, melter design features are being explored to increase production without significantly increasing surface area, melter glass volumes or compromising data quality. In support of these objectives, a set of computational fluid dynamic models have been developed to provide insight into the flow patterns within these small melters and evaluate design features to increase production without significantly increasing glass inventory. Computed velocities were evaluated at the interface between the cold cap and the molten glass pool and within the bulk glass pool. A method was developed to correlate results to an estimated melt rate and down-select design features that produce the highest gains.

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

Computationally-Aided Design of a Small-Scale Radioactive Waste Glass Melter

To provide mission support to the Hanford Waste Immobilization and Treatment Plant for the vitrification of legacy nuclear tank waste, a reduced-scale vitrification pilot system is being designed to process simulated and actual radioactive tank wastes. The tank waste will be separated into high-level waste (HLW) and low-activity waste (LAW) fractions, where the majority by mass (~90%) is LAW and by activity (>95%) is HLW. The first tank waste to be processed at the WTP will be LAW. Since Hanford tank waste and the resultant melter feed compositions are known to vary widely, pilot-scale operations are essential to identify potential problems, provide needed data, confirm assumptions, and determine impacts to full-scale melters and off-gas systems from these different feeds. In addition, the reduced-scale melter system must be capable of quickly providing results to operations without the typically high costs of radioactive operations and provide an engineering platform in close proximity to local operations staff. The objective is to produce similar process conditions to those encountered in the full-scale LAW melters and minimize the volume of radioactive waste necessary for the evaluations. To this end, melter design features are being explored to increase production without significantly increasing surface area, melter glass volumes or compromising data quality. In support of these objectives, a set of computational fluid dynamic models have been developed to provide insight into the flow patterns within these small melters and evaluate design features to increase production without significantly increasing glass inventory. Computed velocities were evaluated at the interface between the cold cap and the molten glass pool and within the bulk glass pool. A method was developed to correlate results to an estimated melt rate and down-select design features that produce the highest gains.

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DOE OSTI2018

Comparative Study of the Standards and Guides of a Management System Applicable to Radioactive Waste - Paper 18292

A Management System may be defined as a set of standardized and interrelated practices that aim to improve effectiveness and produce better results. This system should be structured in such a way as to allow essential information to flow properly within the organization, including technical, administrative, production and human resources information, among others, within an adopted compliance model. There are, currently, several standards and guidelines for structuring a compliance model related to a management system. A system that covers various aspects of management in a single system is usually called integrated. Only the International Atomic Energy Agency - IAEA has a guide with this approach for radioactive waste: GS-G-3.3, 'The Management System for the Processing, Handling and Storage of Radioactive Waste' and GS-G-3.4, 'The Management System for the Disposal of Radioactive Waste'. The objective of this work is to present a comparison between existing standards, even if not specific for the management of radioactive waste, but which are widely used by companies. In due course, the result of this work will be object of proposal for specific standard to be presented to regulatory agency for this type of activity. This study should be useful for countries where there are no specific guidelines by the regulatory agency, adopting the best standard or guide as a conformity standards, creating and establishing a hybrid structure best suited to the safety and security aspects of the local culture. (authors)

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DOE OSTI2025

Selected Chemical Engineering Applications in Nuclear-Waste Processing at the Savannah River Site

The Savannah River Site has been successfully processing and immobilizing nuclear waste since 1996. However, recent developments in both the scientific understanding of chemical principles and the engineering of immobilizing nuclear-waste systems demand a review of the state of the art. These recent advances have significance to other locations that immobilize nuclear waste. Here, the subject matter of this review may find special applicability to chemical engineers interested in hazardous chemical processes (such as processing toxic and radioactive nuclear waste) and to those in the nuclear industry curious about current research in nuclear-waste processing at a site that has eclipsed the quarter-century mark of large-scale (136 million L total) nuclear-waste processing.

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DOE OSTI2011

STATUS OF THE DEVELOPMENT OF IN-TANK/AT-TANK SEPARATIONS TECHNOLOGIES FOR FOR HIGH-LEVEL WASTE PROCESSING FOR THE U.S. DEPARTMENT OF ENERGY

Within the U.S. Department of Energy's (DOE) Office of Technology Innovation and Development, the Office of Waste Processing manages a research and development program related to the treatment and disposition of radioactive waste. At the Savannah River (South Carolina) and Hanford (Washington) Sites, approximately 90 million gallons of waste are distributed among 226 storage tanks (grouped or collocated in 'tank farms'). This waste may be considered to contain mixed and stratified high activity and low activity constituent waste liquids, salts and sludges that are collectively managed as high level waste (HLW). A large majority of these wastes and associated facilities are unique to the DOE, meaning many of the programs to treat these materials are 'first-of-a-kind' and unprecedented in scope and complexity. As a result, the technologies required to disposition these wastes must be developed from basic principles, or require significant re-engineering to adapt to DOE's specific applications. Of particular interest recently, the development of In-tank or At-Tank separation processes have the potential to treat waste with high returns on financial investment. The primary objective associated with In-Tank or At-Tank separation processes is to accelerate waste processing. Insertion of the technologies will (1) maximize available tank space to efficiently support permanent waste disposition including vitrification; (2) treat problematic waste prior to transfer to the primary processing facilities at either site (i.e., Hanford's Waste Treatment and Immobilization Plant (WTP) or Savannah River's Salt Waste Processing Facility (SWPF)); and (3) create a parallel treatment process to shorten the overall treatment duration. This paper will review the status of several of the R&D projects being developed by the U.S. DOE including insertion of the ion exchange (IX) technologies, such as Small Column Ion Exchange (SCIX) at Savannah River. This has the potential to align the salt and sludge processing life cycle, thereby reducing the Defense Waste Processing Facility (DWPF) mission by 7 years. Additionally at the Hanford site, problematic waste streams, such as high boehmite and phosphate wastes, could be treated prior to receipt by WTP and thus dramatically improve the capacity of the facility to process HLW. Treatment of boehmite by continuous sludge leaching (CSL) before receipt by WTP will dramatically reduce the process cycle time for the WTP pretreatment facility, while treatment of phosphate will significantly reduce the number of HLW borosilicate glass canisters produced at the WTP. These and other promising technologies will be discussed.

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

Unreviewed Disposal Question Evaluation: Disposal of the Transuranic Waste Processing Center Low-Level Waste at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation assesses whether the Transuranic Waste Processing Center (TWPC) Low-Level waste stream is suitable for shallow land burial at the Area 5 Radioactive Waste Management Site on the Nevada National Security Site. Disposal of the TWPC Low-Level Waste meets all DOE M 435.1-1 performance objectives. The TWPC Low-Level Waste stream is recommended for acceptance without conditions.

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DOE OSTI2020

Developing Materials for Extreme Environments and Their Use in Nuclear Waste Canisters and Reactors - 20559

The paper focuses on the development of materials which can be utilized in extreme industrial environments. The paper discusses the advancement of materials and their application in components which are utilized in extreme industrial environments as such that of the nuclear field and especially in the nuclear waste field. Nuclear power plants have become an indispensable part of the electricity generation, medical, food processing and many other industries. Though burning small amount of fuel as compared to conventional power plants, they also create waste which is not easy to handle as other types of wastes. The nuclear waste sitting at the bottom of spent fuel tanks around the globe has raised public concerns on how it will be dealt with. Europe has started to give an answer by working on geological repositories. Though till yet, waste has been cooled in spent fuel tanks and after processing, is precisely placed in nuclear waste canisters and casks and transported to sites across the country for further cooling and storage. Though in all this process there had been a great deal of advancement on the type of casks and canisters used and the processes for transporting nuclear material and waste. Often reports about radioactive material leakage, moisture content inside barrels etc. has surfaced. Be it official or unofficial report, there seems a technical failure on how the nuclear waste has been stored and placed at facilities around the globe. The other aspect of this material development is to be beneficial for nuclear reactors. As the technology gets advanced, nuclear reactors change their titles from Gen II to Gen III and so on. The new fleet of upcoming reactors will be exceptionally safer, reliable and more efficient. But these benefits will also bring drawbacks with them. As these reactors will operate at high temperature (above 800 degree Celsius) they also ask for advancement in materials used in the reactor core, fuel used, reactor control etc. The paper tries to answer such concerning questions. The work shows the study done on different material that can be better candidates for extreme environments. (authors)

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

Geochemical Modeling of F Area Seepage Basin Composition and Variability

From the 1950s through 1989, the F Area Seepage Basins at the Savannah River Site (SRS) received low level radioactive wastes resulting from processing nuclear materials. Discharges of process wastes to the F Area Seepage Basins followed by subsequent mixing processes within the basins and eventual infiltration into the subsurface resulted in contamination of the underlying vadose zone and downgradient groundwater. For simulating contaminant behavior and subsurface transport, a quantitative understanding of the interrelated discharge-mixing-infiltration system along with the resulting chemistry of fluids entering the subsurface is needed. An example of this need emerged as the F Area Seepage Basins was selected as a key case study demonstration site for the Advanced Simulation Capability for Environmental Management (ASCEM) Program. This modeling evaluation explored the importance of the wide variability in bulk wastewater chemistry as it propagated through the basins. The results are intended to generally improve and refine the conceptualization of infiltration of chemical wastes from seepage basins receiving variable waste streams and to specifically support the ASCEM case study model for the F Area Seepage Basins. Specific goals of this work included: (1) develop a technically-based 'charge-balanced' nominal source term chemistry for water infiltrating into the subsurface during basin operations, (2) estimate the nature of short term and long term variability in infiltrating water to support scenario development for uncertainty quantification (i.e., UQ analysis), (3) identify key geochemical factors that control overall basin water chemistry and the projected variability/stability, and (4) link wastewater chemistry to the subsurface based on monitoring well data. Results from this study provide data and understanding that can be used in further modeling efforts of the F Area groundwater plume. As identified in this study, key geochemical factors affecting basin chemistry and variability included: (1) the nature or chemistry of the waste streams, (2) the open system of the basins, and (3) duration of discharge of the waste stream types. Mixing models of the archetype waste streams indicated that the overall basin system would likely remain acidic much of the time. Only an extended periods of predominantly alkaline waste discharge (e.g., >70% alkaline waste) would dramatically alter the average pH of wastewater entering the basins. Short term and long term variability were evaluated by performing multiple stepwise modeling runs to calculate the oscillation of bulk chemistry in the basins in response to short term variations in waste stream chemistry. Short term (1/2 month and 1 month) oscillations in the waste stream types only affected the chemistry in Basin 1; little variation was observed in Basin 2 and 3. As the largest basin, Basin 3 is considered the primary source to the groundwater. Modeling showed that the fluctuation in chemistry of the waste streams is not directly representative of the source term to the groundwater (i.e. Basin 3). The sequence of receiving basins and the large volume of water in Basin 3 'smooth' or nullify the short term variability in waste stream composition. As part of this study, a technically-based 'charge-balanced' nominal source term chemistry was developed for Basin 3 for a narrow range of pH (2.7 to 3.4). An example is also provided of how these data could be used to quantify uncertainty over the long term variations in waste stream chemistry and hence, Basin 3 chemistry.

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

Electro-Mechanical Manipulator for Use in the Remote Equipment Decontamination Cell at the Defense Waste Processing Facility, Savannah River Site - 12454

One of the legacies of the cold war is millions of liters of radioactive waste. One of the locations where this waste is stored is at the Savannah River Site (SRS) in South Carolina. A major effort to clean up this waste is on-going at the defense waste processing facility (DWPF) at SRS. A piece of this effort is decontamination of the equipment used in the DWPF to process the waste. The remote equipment decontamination cell (REDC) in the DWPF uses electro-mechanical manipulators (EMM) arms manufactured and supplied by PaR Systems to decontaminate DWPF process equipment. The decontamination fluid creates a highly corrosive environment. After 25 years of operational use the original EMM arms are aging and need replacement. To support continued operation of the DWPF, two direct replacement EMM arms were delivered to the REDC in the summer of 2011. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

UNREVIEWED DISPOSAL QUESTION EVALUATION: Disposal of the Transuranic Waste Processing Center Low-Level Waste at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation (UDQE) assesses whether the Transuranic Waste Processing Center (TWPC) Low-Level Waste (LLW), FWORCHLLW0102, Revision 12 (TWPC 2021), is suitable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the TWPC LLW meets all performance objectives of DOE Manual DOE M 435.1-1, Radioactive Waste Management Manual, Chapter IV, Section P (DOE 1999). The TWPC LLW waste stream is recommended for acceptance without conditions.

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

The Start Of Ebullition In Quiescent, Yield-Stress Fluids

Non-Newtonian rheology is typical for the high-level radioactive waste (HLW) slurries processed in the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Hydrogen and other flammable gases are generated in the aqueous phase by radiolytic and chemical reactions. HLW slurries have a capacity for retaining gas characterized by the shear strength holding the bubbles still. The sizes and degassing characteristics of flammable gas bubbles in the HLW slurries expected to be processed by the WTP are important considerations for designing equipment and operating procedures. Slurries become increasingly susceptible to degassing as the bubble concentration increases. This susceptibility and the process of ebullitive bubble enlargement are described here. When disturbed, the fluid undergoes localized flow around neighboring bubbles which are dragged together and coalesce, producing an enlarged bubble. For the conditions considered in this work, bubble size increase is enough to displace the weight required to overcome the fluid shear strength and yield the surroundings. The buoyant bubble ascends and accumulates others within a zone of influence, enlarging by a few orders of magnitude. This process describes how the first bubbles appear on the surface of a 7 Pa shear strength fluid a few seconds after being jarred.

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

Progress on TRANSCEND Theme 1: Integrated Waste Management - 20194

The management of radioactive waste includes mobilisation, processing, packaging, storage, transport and final disposal. Research in the Integrated Waste Management theme of TRANSCEND focuses on underpinning science and engineering in areas of relevance to hazard reduction and decommissioning and the objectives are to: (1) develop new materials for effluent decontamination; (2) demonstrate predictive modelling of sludge/slurry removal, underpinned by experimental data; (3) transform understanding of radiation-driven processes in nuclear waste sludges; and (4) provide underpinning studies to develop innovative processes and wasteform materials. This paper gives overviews of the 16 research projects that will be done over the next 3 years in this theme. (authors)

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

Full-Scale Retrieval and Transfer System Mockup for Pneumatic Conveyance of High-Level Calcine Waste - 19563

The Idaho National Laboratory reprocessed spent nuclear fuel from 1953 to 1994. The reprocessing used solvent extraction to separate uranium, resulting in a high-level radioactive liquid waste. A calcination process was used to convert the liquid waste to a high-level radioactive granular solid, referred to as calcine. The calcine is stored in stainless-steel bins contained within six concrete vaults called the Calcined Solids Storage Facilities (CSSFs). The Calcine Disposition Project is tasked with demonstrating the ability to retrieve the calcine from the bins of CSSF 1 and transferring it to the bins of CSSF 6. CSSF 1 was constructed to be a final calcine repository, with no forethought of retrieval. It represents the greatest technical hurdle to calcine extraction of all the CSSFs. The challenges inherent with accessing the calcine, safely extracting it, and securely transporting it are copious and complex. To overcome those challenges in a timely and objective manner, a full-scale mockup will be constructed. The mockup includes a full-size model of a concentric group of bins from CSSF 1 and a storage vessel representing CSSF 6. Mockup equipment will include an air compressor, filtration, blower/vacuum, and retrieval devices. In addition to defining process parameters, the mockup testing scope will demonstrate the amount of residual calcine that can be expected to be in the bins following extraction. This will be key to preparing for closure of CSSF 1 following removal of the calcine. Once testing is complete, the mockup will constitute a model of the calcine extraction and transport system and will be prepared to fulfill a second advantageous objective: the operators that will carry out the calcine transfer from CSSF 1 to CSSF 6 will be trained using the mockup. The control system will be identical, the major pieces of equipment will be representative, and the procedures will be analogous; hence, high-fidelity training can take place at the mockup. Once radiological operations commence at CSSF 1, the mock-up's commission will still not be completed. If any problems arise during calcine transport, the mockup will be ready to serve as a simulator to more effectively reproduce dilemmas and deduce solutions. The mock-up's mission spans the duration of the project. Its usefulness begins with helping to identify process criteria. Operations personnel will benefit greatly from its use as a training tool. And finally, the immediate availability of a simulator to reproduce problems encountered during radiological extraction and transfer will be invaluable. The knowledge gained during this process will also be of continued benefit as future work is done to remove various types of calcine from the remaining CSSFs. The methodology of safe and secure pneumatic transfer of radioactive material may be of benefit to others throughout the U.S. Department of Energy complex. (authors)

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

A Study on the Comparison of International WTS for Integrated Management of Radioactive Waste - 19034

Domestic radioactive waste is generated by nuclear power plants, the Korea Atomic Energy Research Institute and KEPCO Nuclear Fuel Company. These nuclear energy agencies have their own waste tracking system. To track waste more efficiently, it is necessary to build an integrated tracking system from the viewpoint of the final disposer. As for the korea Atomic Energy Research Institute, the decommissioning waste due to the decommissioning of research reactors No. 1 and 2, and the low and intermediate-level radioactive waste, generated by the operation of the High Flux Advanced Neutron Application Reactor(HANARO), account for the most. The solid waste from each facility and laboratory is sorted and collected according to radiation levels and the contents. For starters, according to the contents, the waste is divided into combustible, noncombustible compressible and non-compressible waste. The combustible waste, generated by the Korea Atomic Energy Research Institute, is treated according to the low and intermediate-level radioactive waste delivery regulation of the Nuclear Safety and Security Commission and the acceptance criteria of the disposal facility. KEPCO Nuclear Fuel Company is making the nuclear fuel used in nuclear power plants. Radioactive waste is generated in the nuclear fuel manufacturing process, and it is classified into 8 types of waste according to the solid waste management procedure. Most of the radioactive waste, generated while KEPCO Nuclear Fuel Company processes the light-water reactor and heavy-water reactor nuclear fuel, is classified as low and intermediate-level radioactive waste. The radioactive liquid waste, generated in the radiation controlled area in nuclear fuel processing facilities, is divided into low-concentration liquid waste and high-concentration liquid waste. The miscellaneous solids, generated in the Hanbit NPP of Korea Hydro and Nuclear Power Co., Ltd., are classified in consideration of the physical properties, density and amount of the waste, and substances subject to restricted disposal. This paper investigated information on the structure, functions and maintenance of the systems implemented and operated by individual countries and IAEA, and the types of managed data, and analyzed the detailed configuration of the system and the characteristics of DBMS (Data Base Management System), and discovered during the analysis process the information on the waste management in each country or institution. This paper investigated the classification standards and treatment methods according to the characteristics of the radioactive substances handled by each agency. These classification and treatment methods can be applied to the database when the radioactive waste tracking system is implemented in the future. Also, this paper analyzed the radioactive waste tracking system of each country. Most overseas waste tracking systems are configured based on the ORACLE DBMS. The ORACLE DBMS is highly secure, and can manage large-scale databases. Also, Germany's ReVK and US IWTS are using the Bar-Code system that can track and manage all waste data, generated during decommissioning or transportation, in real time, and the real-time tracking system of web-based US CID can simply create and output reports quickly. If the advantages of overseas waste tracking systems are added to the domestic integrated waste tracking system, more efficient and safer management of radioactive waste will be possible.

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DOE OSTI2013

XRD studies of cement embedding matrices containing complexing reagents from decontamination practices

The study of embedding matrix of radioactive waste is an important issue concerning the need to assure the radiological safety requirements for final disposal. The conditioning of the radioactive wastes by cementation process involves also a structural investigation by X-ray diffraction (XRD) of the samples prepared with cement and complexing agents at different concentration. This paper gives useful information about the influence of complexing agents related to damages produced in concrete microstructures. The experimental data obtained on samples kept in real repository environment for a period up to 15 years, offer structural information regarding the interaction between the different decontamination agents with the major components of the matrix. (authors)

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