Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “RADIOACTIVE WASTE PROCESSING”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 829 records · Page 46
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Molten Salt Reactor Experiment (MSRE) Lay-Up - 19187

UCOR, an AECOM-led partnership with Jacobs, manages the cleanup of the 2,200-acre East Tennessee Technology Park for its client, the U.S. Department of Energy (DOE), in Oak Ridge, Tennessee. The site was contaminated with radioactive, hazardous and industrial wastes generated by more than 40 years of national defense and energy missions. 2019 is the 50. anniversary of the shutdown of the Molten Salt Reactor Experiment (MSRE). MSRE ran a brief four years in the 1960's but earned an enduring legacy as an innovative reactor technology concept which has enjoyed a renewal of global interest in the past several years. In 2018, it was given elevated status among advanced reactor programs within DoE's Nuclear Energy office. The MSRE was a graphite-moderated, liquid fueled reactor that operated from June 1965 through December 1969. Although it was shut down nearly 50 years ago, certain systems within the reactor have continued to operate to maintain the facility in a safe and stable condition until it can be demolished as part of the comprehensive cleanup of the Oak Ridge Reservation. These systems are composed of a complex network of pipes, valves, gauges and associated equipment that control air filtration and pressure buildup resulting from fluorine gas generated through radiolysis of the fluoride salt which once fueled the reactor. In 2016, an MSRE Engineering Evaluation Team was chartered to assess current facility conditions and develop recommendations to maintain the facility in a safe state until deactivation and demolition commences. Plans are now underway to 'lay up' the MSRE facility until deactivation and demolition can occur. The proposed MSRE Facility Layup will minimize maintenance costs, reduce personnel risk of injury and exposure, provide reliable electric service to the facility and require zero personnel occupancy at MSRE. In the proposed layup approach a new source of electrical power to required loads will be routed through the facility, and new gas processing and ventilation systems will be installed that will be optimized for safe processing of the constituent hazardous gasses that exist in the legacy vessels at MSRE. (authors)

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

Computational Fluid Dynamics (CFD) and Safety Factors: Computer modeling of complex processes needs old-fashioned experiments to stay in touch with reality

Computational fluid dynamics (CFD) is recognized as a powerful engineering tool. That is, CFD has advanced over the years to the point where it can now give us deep insight into the analysis of very complex processes. There is a danger, though, that an engineer can place too much confidence in a simulation. If a user is not careful, it is easy to believe that if you plug in the numbers, the answer comes out, and you are done. This assumption can lead to significant errors. As we discovered in the course of a study on behalf of the Department of Energy's Savannah River Site in South Carolina, CFD models fail to capture some of the large variations inherent in complex processes. These variations, or scatter, in experimental data emerge from physical tests and are inadequately captured or expressed by calculated mean values for a process. This anomaly between experiment and theory can lead to serious errors in engineering analysis and design unless a correction factor, or safety factor, is experimentally validated. For this study, blending times for the mixing of salt solutions in large storage tanks were the process of concern under investigation. This study focused on the blending processes needed to mix salt solutions to ensure homogeneity within waste tanks, where homogeneity is required to control radioactivity levels during subsequent processing. Two of the requirements for this task were to determine the minimum number of submerged, centrifugal pumps required to blend the salt mixtures in a full-scale tank in half a day or less, and to recommend reasonable blending times to achieve nearly homogeneous salt mixtures. A full-scale, low-flow pump with a total discharge flow rate of 500 to 800 gpm was recommended with two opposing 2.27-inch diameter nozzles. To make this recommendation, both experimental and CFD modeling were performed. Lab researchers found that, although CFD provided good estimates of an average blending time, experimental blending times varied significantly from the average.

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

Hot Isostatic Pressing of Radioactive Nuclear Waste: The Calcine at INL - 16348

Hot Isostatic Pressing (HIP) is a technology that has been around for 60+ years. By using high temperature and high gas pressure, dry metal and ceramic powders can be consolidated and a volume decrease can be achieved. This paper presents the efforts using the HIPing process at the Idaho National Laboratory to treat the calcine radioactive waste. Once loaded into collapsible canisters, the HIPing would be used to treat the waste before final disposal. The resulting volume reduction was shown to be 20-70% and the cost ratio vs vitrification is 1:1.74. (authors)

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

The Building and Commissioning of the Batch Pyrolysis Plant in Studsvik, Sweden - 12447

After a sequence of lab scale and bench scale trials the building of a pyrolysis plant could begin at the Studsvik site in Sweden. The facility is primarily aimed at treatment of uranium contaminated organic waste originating at fuel manufacturing plants and other facilities where the main contamination is uranium. The plant is an extension/addition to the already operating incinerator. In order to further widen the waste acceptance criteria the design of the off-gas treatment system does not have the same design as that of the incinerator. The building of the facility began in April 2011, and the first heating of the facility took place in late December, 2011. The site acceptance tests are planned for January, as are the first inactive trials aimed at optimisation of process control. The facility is planned to be operating with radioactive materials from February 2012. The pyrolysis unit is primarily aimed at treatment of uranium contaminated waste, under conditions that facilitate leaching and recovery of the uranium from the ashes. However, a number of other uses are conceivable. The batch fed pyrolysis unit, with its chosen design of the off-gas treatment system, enables treatment of some waste fractions that are difficult to treat in the incinerator. For instance small campaigns, i.e. smaller quantities of waste (typically <5 tonnes), or waste containing high levels of chlorine and sulphur are possible to treat in the pyrolysis unit. The pyrolysis unit is also expected to perform well in the treatment of other 'difficult' waste fractions, for instance waste containing pyrophoric materials, or other types of waste that benefit from the high level of process control, i.e. control of temperature and atmosphere throughout the process, that can be obtained in the pyrolysis unit compared to the incinerator. Furthermore, treatment in the pyrolysis unit minimises the risk of cross contamination between different waste treatment campaigns. This feature is obtained thanks to the low gas flow rates in the vessel, which means that a higher retention of nuclides in the ashes is obtained, but also through a design that facilitates cleaning of the pyrolysis vessel and the system beyond this, or even exchanging the most critical components, such as the pyrolysis vessel itself, if need be. (authors)

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

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

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

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

Advances in Resin Management Using 3R-Scan - 20154

The most important factor underlying optimal waste management is developing a clear picture of the radioactivity content of the waste and its impact on waste disposal cost. For over 35 years since the publishing of 10CFR61, waste characterization has relied on sampling the final waste product after formation. In the days following 10CFR61, the cost of final disposal was marginal with only a small impact on the overall costs. Constraints were added by provisions of the Low Level Waste Policy Act of 1985 leading to increasingly limited access to those disposal sites that remained available. In addition, Nuclear Regulatory Commission (NRC) pressure promoting waste volume reduction led to disposal costs inevitably rising. Despite this, characterization practices in monitoring of waste generation for activity content still center on the same dated processes. This results in a disposal classification on the basis of endpoint sampling without consideration of the homogeneity of the waste mixture. It can also disregard consideration of the representativeness of the single or small sample base. As a minimum effort, a formalized sampling program of a fixed grouping of waste streams can be implemented that could account for more than 95% of all of the activity carried in solid waste products. The sample results for each radionuclide could then be trended as time passes to develop reasonable scaling factors for difficult to measure radionuclides. This process, identified in NRC guidance, has been rigorously followed by a relatively small number of facilities. The trended scaling factors serve to improve accuracy by identifying anomalous results that could otherwise go undetected. Direct monitoring of the accumulation of activity in process streams generating solid radwaste, including demineralizers and filter streams, is a more precise approach. Nearly all of these streams are monitored by plant chemistry on a regular schedule to maintain water quality. This paper discusses viable options for developing the basis for characterization through process monitoring of the accumulation of activity at the point of generation. Special focus is on resin bed tracking and how process knowledge of these streams can be brought together to form a consistent and precise solid waste radioactivity inventory. Some of the specific points covered in this paper include the merger of the fission product release computer program, 3R-STAT, with the radwaste sample analysis computer program, SCAN4 to create 3R-SCAN, the importance of individual waste stream influences on the overall source term, and the use of historic sample data to develop scaling factors using an automated process. (authors)

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

Use of Automation in Low Level Waste Storage and Management - 19227

The Australian Nuclear Science and Technology Organisation (ANSTO) generates Low Level Waste (LLW) and Intermediate Level Waste (ILW) from various site operations such as the operation of a research Reactor and the manufacture of Molybdenum 99. The Low Level Solid Waste (LLSW) generated is stored at ANSTO before and after conditioning. The current process flow involves the storage of LLSW waste before conditioning on racks, handled using forklifts. When the unconditioned waste is ready for conditioning, it is transferred to another facility, where volume reduction and packaging is carried out. Once conditioned, the waste is placed in further storage, pending dispatch to the proposed Australian National Radioactive Waste Management Facility (NRWMF). A new purpose built facility has been constructed, adjacent to the LLSW processing facility, for the storage and management of LLSW in both unconditioned and conditioned forms. This new facility will streamline the waste management process and significantly enhance the safety and efficiency of the process. In line with international best practises, the new facility utilises automation solutions to minimise the build footprint, and minimise the use of forklifts and other mobile plant. Most importantly, it applies the principle of ALARA (As Low as Reasonably Achievable) to minimise operator activities in areas where radioactive material is stored and eliminates double handling of the waste. The automation for unconditioned waste utilises an Automated Storage and Retrieval System (ASRS), a vertical stacker crane within high density racking, to provide easy storage and retrieval for processing and/or routine inspections. This system allows operators to control it from outside of the storage area whilst utilising integrated software that allows for the required tracking of the unconditioned waste. From the ASRS storage area, unconditioned waste will have a smooth transition to the processing facility which is located adjacent to the ASRS system. Once conditioned, the waste is catalogued and returned to a designated storage area within the new facility using an automated building gantry crane with a target delivery system to eliminate the need for operators in the main storage bay. The waste will be held here in preparation for dispatch to the future repository. The utilisation of the ASRS for the purpose of storing LLSW is an innovative use of automation that will eliminate operators from active areas, reduce dose rates received and integrate with the waste tracking software. Furthermore, the system has significant advantages with operational efficiencies increased, reduced double handling of nuclear waste and maximum use of building footprint. (authors)

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

Biosorption of Uranium by Bone Meal

The treatment of aqueous solution waste containing organic compounds and radionuclides has always been a great concern since the processes available are usually expensive and unfeasible for the low volume of radioactive liquid waste recovered. In view of a more efficient and low cost treatment, biosorption-based technologies for the removal of this hazardous radionuclides, heavy metals, and organic substances from contaminated matrices have been studied as an alternative process. In this context, bone meal is the material of choice, and contains calcium phosphate in the form of hydroxyapatite. Previous studies with hydroxyapatite showed that this material is efficient in removing heavy metals from polluted soils due to its adsorbing ability. In this context, uranium biosorption capacity by using bone meal was evaluated by different experimental conditions. This study aims at analyzing the capacity of bone meal to remove uranium from aqueous solution through biosorption process, considering that this nuclide is radiotoxic and may cause serious environmental problems and risks to human life if not correctly disposed. The biosorption capacity of U by bone meal was evaluated under different experimental conditions. The results suggested that adsorption with this material can be applied in the treatment of aqueous solutions as a low-cost and highly efficient alternative to remove U. Therefore, we believe that our results are useful for other investigators dealing with similar problems. Further work is required to evaluate this material as a potential adsorbent concerning the treatment of liquid organic radioactive waste containing uranium. Next studies are crucial to ascertain the biosorption efficiency, by analyzing other variables, such as adsorbent dosage, pH and temperature.

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

Status of the Development of In-Tank/At-Tank Separations Technologies 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.

High level waste↗
DOE OSTI2016

WIPP Certification of the Retrieval Box Assay System at the Advanced Mixed Waste Treatment Plant, Idaho - 16241

The Retrieval Box Assay System (RBAS) at the Advanced Mixed Waste Treatment Plant (AMWTP) is an Imaging Passive Active Neutron system supplemented with a High Purity Germanium (HPGe) gamma system for the quantification of Pu-239 Fissile Gram Equivalent (FGE) and transuranic (TRU) alpha activity of waste containers of various sizes up to 249 cm x 149 cm x 198 cm (Length x Width x Height). The system was originally calibrated in 2002 (with a range of surrogate matrices and radioactive sources in a 213 cm x 122 cm x 122 cm FRP box) and has been operational for over a decade. As of August 2015, the RBAS has measured approximately 18,000 boxes (50,000 m3 of waste) which has subsequently been fed into the facility's treatment process. Its success, and the wide envelope of waste that it can characterize, has led AMWTP to extend the RBAS mission to include WIPP certification. As part of this new project, the instrument must be capable of meeting WIPP's Waste Acceptance Criteria (WAC) which include reporting activities and masses of the 10 WIPP-tracked radionuclides. The system will be required to assay sludge and debris waste streams, shredded waste and large metal items. The container size includes Standard Waste Box (SWBs), Fiberglass Reinforced Plywood (FRPs) and large Shredder Boxes. Sludge matrices will be loaded into 210-liter (55-gallon) drums and placed inside an SWB. After assay, boxes will be over-packed into SLB-II's for shipment to WIPP. The target waste streams will contain mixtures of Weapons Grade plutonium, Heat Source plutonium (high in Pu-238 content) and Mixed Oxide (MOX). The activity of the waste is expected to encompass a wide dynamic range of activity from low-level waste (LLW) to TRU with up to several hundred grams of plutonium. Some packages will have high gamma activity associated with fission and activation products such as Cs-137 and Co-60. The existing RBAS hardware (detectors, electronics, assay chamber etc.) was retained for this project. Two lead filters were provided and installed on the existing filter wheels to attenuate the gamma ray flux from the highest activity boxes to avoid overload of the HPGe detectors. The new software suite includes PC-FRAM (Version 5), which is used for the isotopic analysis of the gamma spectra. Integration of FRAM and neutron results is performed using a software package known as RBAS WIPP Analysis Reporting Software (RWARS). This package also performs matrix dependent background adjustments. Other upgrades provided include automated tests that identify when assay parameters fall outside a predetermined range and a new Quality Checking (QC) application. Minimum Detectable Activity (MDA) values were determined for the active and passive neutron modes as a function of measured matrix correction indices. The resulting Minimum Detectable Concentration (MDC) has been demonstrated to be less than 3700 Bq/g alpha activity (100 nCi/g). Modifications were performed to update the system TMU parameters to accommodate assay of non-standard size boxes and a wider range of matrix materials. A series of calibration confirmation, measurements were performed on site in 2014 using a representative range of surrogate boxes and radioactive sources including Pu. The WIPP WAC performance criteria were successfully demonstrated up to 157 g Pu and the system passed its first Performance Demonstration Program (PDP) cycle in the Spring of 2015. (authors)

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

Characterising Encapsulated Nuclear Waste using Cosmic-ray Muon Tomography (MT)

A prototype scintillating-fibre detector system has been developed at the University of Glasgow in collaboration with the UK National Nuclear Laboratory (NNL) for the nondestructive assay of UK legacy nuclear waste containers. This system consists of four tracking modules, two above and two below the container under interrogation. Each module consists of two orthogonal planes of 2 mm-pitch fibres yielding one space point. Per plane, 128 fibres are read out by a single Hamamatsu H8500 64-channel MAPMT with two fibres multiplexed onto each pixel. The configuration allows the reconstruction of the incoming and scattered muon trajectories, thus enabling the container content, with respect to atomic number Z, to be determined. Results are shown from experimental data collected for high-Z objects within an air matrix and within a shielded, concrete-filled container. These reconstructed images show clear discrimination between the low, medium and high-Z materials present, with dimensions and positions determined with sub-centimetre precision. (authors)

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

A Novel and Cost Effective Approach to the Decommissioning and Decontamination of Legacy Glove Boxes - Minimizing TRU Waste and Maximizing LLW Waste - 13634

This paper describes the process of decommissioning two gloveboxes at the Argonne National Laboratory (ANL) that were employed for work with plutonium and other radioactive materials. The decommissioning process involved an initial phase of clearing tools and materials from the glove boxes and disconnecting them from the laboratory infrastructure. The removed materials, assessed as Transuranic (TRU) waste, were packaged into 55 gallon (200 litre) drums and prepared for ultimate disposal at the Waste Isolation Pilot Plant (WIPP) at Carlsbad New Mexico. The boxes were then sampled to determine the radioactive contents by means of smears that were counted with alpha and beta detectors to determine the residual surface contamination, especially in terms of alpha particle emitters that are an indicator of TRU activity. Paint chip samples were also collected and sent for laboratory analysis in order to ascertain the radioactive contamination contributing to the TRU activity as a fixed contamination. The investigations predicted that it may be feasible to reduce the residual surface contamination and render the glovebox structure low level waste (LLW) for disposal. In order to reduce the TRU activity a comprehensive decontamination process was initiated using chemical compounds that are particularly effective for lifting and dissolving radionuclides that adhere to the inner surfaces of the gloveboxes. The result of the decontamination process was a reduction in the TRU surface activity on the inner surfaces of the gloveboxes by four orders of magnitude in terms of disintegrations per unit area (DPA). The next phase of the process involved a comprehensive assay of the gloveboxes using a combination of passive neutron and gamma ray scintillation detectors and a shielded and collimated high purity Germanium (HPGe) gamma ray detector. The HPGe detector was used to obtain gamma ray spectra for a variety of measurement positions within the glovebox. The spectra were used to determine the TRU content of the boxes by assessing the activity of Am-241 (59 keV) and Pu-241 (414 keV). Using the data generated it was possible for qualified subject matter experts (SME) to assess that the gloveboxes could be consigned for disposition as LLW and not as TRU. Once this determination was assessed and accepted the gloveboxes were prepared for final disposition to the Nevada National Security Site (NNSS) - formerly the Nevada Test Site (NTS). This preparation involved fixing any remaining radioactive contamination within the gloveboxes by filling them with a foam compound, prior to transportation. Once the remaining contamination was fixed the gloveboxes were removed from the laboratory and prepared for transported by road to NNSS. This successful glovebox decontamination and decommissioning process illustrates the means by which TRU waste generation has been minimized, LLW generation has been maximized, and risk has been effectively managed. The process minimizes the volume of TRU waste and reduced the decommissioning time with significant cost savings as the result. (authors)

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

Hanford Site Composite Analysis Technical Approach Description: Radionuclide Inventory and Waste Site Selection Process.

The updated Hanford Site Composite Analysis will provide an all-pathways dose projection to a hypothetical future member of the public from all planned low-level radioactive waste disposal facilities and potential contributions from all other projected end-state sources of radioactive material left at Hanford following site closure. Its primary purpose is to support the decision-making process of the U.S. Department of Energy (DOE) under DOE O 435.1-1, Radioactive Waste Management (DOE, 2001), related to managing low-level waste disposal facilities at the Hanford Site.

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

Special Analyses for the Hanford Integrated Disposal Facility Performance Assessment - 20102

In 2014, the Department of Energy (DOE) Office of River Protection and its contractors began to develop a performance assessment for the near-surface disposal of low-level and mixed low-level waste at the Hanford Site's Integrated Disposal Facility (IDF). The IDF is a doubly-lined landfill that was constructed between 2004 and 2006 to be the disposal facility for the vitrified low-activity waste that will be produced at the Waste Treatment and Immobilization Plant (WTP). IDF is also expected to receive solid secondary waste produced at the WTP and other solid wastes from site activities. The IDF has been in a preoperational state awaiting authorization from DOE and a RCRA permit modification from the State of Washington Department of Ecology to receive waste. Both the Disposal Authorization Statement and permit modification require a performance assessment demonstrating that the system of engineered and natural features will limit releases of radionuclides and hazardous chemicals from the IDF and be protective of human health and the environment. The simulated duration is 10,000 years. Based on the analyses presented in the 2017 Integrated Disposal Facility Performance Assessment, DOE issued a conditional Operating Disposal Authorization Statement for the IDF in June 2018. The long-term performance of the IDF to be protective of human health and the environment was evaluated under the requirements of DOE Order 435.1, Radioactive Waste Management. Computer simulations were performed to evaluate whether or not the IDF would comply with DOE requirements. In the time that has passed since the performance assessment was approved by DOE, new information has been discovered that had not been considered in the performance assessment. Since this new information has not been evaluated, the potential impact of the changes have not been taken into consideration in DoE's disposal authorization. DOE and its contractors follow a change control process to screen and, when necessary, evaluate new information that could potentially impact the conclusions of the completed performance assessment. This paper will describe the change control process and provide two examples of evaluations performed following the change control process. The first example evaluates a new waste form for liquid secondary waste that was not evaluated in the performance assessment. In the performance assessment, liquid secondary waste was assumed to be solidified with grout. A new recommendation to dispose of the liquid secondary waste after drying it to a powder was evaluated. The second example evaluates inventory implications from changes to the flow sheet that estimates the feed composition to the low-activity waste vitrification facility. The changes result in higher strontium concentrations in the vitrified waste stream. (authors)

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

Treatment of irradiated graphite from French Bugey reactor

In 2008, following the general French plan for nuclear waste management, Electricite de France attempted to find for irradiated graphite an alternative solution to direct storage at the low-activity long-life storage center in France managed by the national agency for wastes (ANDRA). EDF management requested that its engineering arm, EDF CIDEN, study the graphite treatment alternatives to direct storage. In mid-2008, this study revealed the potential advantage for EDF to use a steam reforming process known as Thermal Organic Reduction, 'THOR' (owned by Studsvik, Inc., USA), to treat or destroy the graphite matrix and limit the quantity of secondary waste to be stored. In late 2009, EDF began a test program with Studsvik to determine if the THOR steam reforming process could be used to destroy the graphite. The program also sought to determine if the graphite could be treated to release the bulk of activity while minimizing the gasification of the bulk mass of the graphite. In October 2009, tests with non-irradiated graphite were completed and demonstrated destruction of a graphite matrix by the THOR process at satisfactory rates. After gasifying the graphite, focus shifted to the effect of roasting graphite at high temperatures in inert gases with low concentrations of oxidizing gases to preferentially remove volatile radionuclides while minimizing the graphite mass loss to 5%. A radioactive graphite sleeve was imported from France to the US for these tests. Completed in April 2010, 'Phase I' of testing showed that the process removed >99% of H-3 and 46% of C-14 with <6% mass loss. Completed in September 2011, 'Phase II' testing achieved increased removals as high as 80% C-14. During Phase II, it was also discovered that roasting in a reducing atmosphere helped to limit the oxidation of the graphite. Future work seeks to explore the effects of reducing gases to limit the bulk oxidation of graphite. If the graphite could be decontaminated of long-lived radionuclides up to 95% for C-14 while minimizing mass loss to <5%, this would minimize the volume of any secondary waste streams and potentially lower the waste class of the larger bulk of graphite. Alternatively, if up to 95% decontamination of C-14 is achieved, the graphite may be completely gasified which could result in lower disposal. (authors)

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

Main results of the french R and D on minor actinides partitioning, a significant improvement towards nuclear waste reduction - 15618

In the framework of the successive 1991 and 2006 Waste Management Act, French government supported a very significant R and D program on partitioning and transmutation of minor actinides (MA). This program aims to study potential solutions for still minimizing the quantity and the hazardousness of final waste, by MA recycling. Indeed, MA recycling can reduce the heat load and the half-life of most of the waste to be buried to a couple of hundred years, overcoming the concerns of the public related to the long-life of the waste. Within this framework, this paper aims to present the most recent progress obtained in CEA on the development of innovative actinide partitioning hydrometallurgical processes in support of their recycling, either in an homogeneous mode (MA are recycled at low concentration in all the standard reactor fuel) or in an heterogeneous mode (MA are recycled at higher concentration in specific targets, at the periphery of the reactor core). Recovery performances obtained on recent tests in high active conditions of the so-called GANEX and DIAMEX-SANEX process will be presented and discussed in light of the potential P and T scenarios. Finally, recent developments regarding the recycling of the sole Am will be presented as well as the results obtained on highly active solutions for this so-called EXAM process. This set of results gives to the French government a portfolio of potential recycling processes which could be separately and progressively implemented if decided. (authors)

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

Road Map for Developing Iron Phosphate Waste Forms for Salt Wastes

In this report, issues that must be addressed to advance the technology readiness level of phosphate glass waste forms being developed to immobilize high-level radioactive salt waste streams are identified, the states of understanding various technical aspects of formulation, processing, and performance are summarized, and approaches supporting further development are recommended. Processing results in dehalogenation of the waste salt, capture of the gaseous halide-bearing species, and immobilization of the residual salt components in a phosphate glass waste form. The approach is suitable for high-level salt waste from electrochemical reprocessing and molten salt reactors. The technology has been demonstrated for chloride-based salts and may also be suitable for the treatment and immobilization of fluoride-based and iodide-bearing waste salts. Aspects of the process requiring further development are identified and approaches recommended.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗