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At least 487 records · Page 27
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Waste management strategy for cost effective and environmentally friendly NPP decommissioning

Decommissioning of nuclear power plants generates large volumes of radioactive or potentially radioactive waste. The proper management of the dismantling waste plays an important role for the time needed for the dismantling phase and thus is critical to the decommissioning cost. An efficient and thorough process for inventorying, characterization and categorization of the waste provides a sound basis for the planning process. As part of comprehensive decommissioning studies for Nordic NPPs, Westinghouse has developed the decommissioning inventories that have been used for estimations of the duration of specific work packages and the corresponding costs. As part of creating the design basis for a national repository for decommissioning waste, the total production of different categories of waste packages has also been predicted. Studsvik has developed a risk based concept for categorization and handling of the generated waste using six different categories with a span from extremely small risk for radiological contamination to high level waste. The two companies have recently joined their skills in the area of decommissioning on selected market in a consortium named ndcon to further strengthen the proposed process. Depending on the risk for radiological contamination or the radiological properties and other properties of importance for waste management, treatment routes are proposed with well-defined and proven methods for on-site or off-site treatment, activity determination and conditioning. The system is based on a graded approach philosophy aiming for high confidence and sustainability, aiming for re-use and recycling where found applicable. The objective is to establish a process where all dismantled material has a pre-determined treatment route. These routes should through measurements, categorization, treatment, conditioning, intermediate storage and final disposal be designed to provide a steady, un-disturbed flow of material to avoid interruptions. Bottle-necks in the process causes increased space requirements and will have negative impact on the project schedule, which increases not only the cost but also the dose exposure to personnel. For these reasons it is critical to create a process that transfers material into conditioned waste ready for disposal as quickly as possible. To a certain extent the decommissioning program should be led by the waste management process. With the objective to reduce time for handling of dismantled material at site and to efficiently and environmental-friendly use waste management methods (clearance for re-use followed by clearance for recycling), the costs for the plant decommissioning could be reduced as well as time needed for performing the decommissioning project. Also, risks for delays would be reduced with a well-defined handling scheme which limits surprises. Delays are a major cost driver for decommissioning projects. (authors)

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

Waste Management Strategy for Dismantling Waste to Reduce Costs for Power Plant Decommissioning - 13543

Decommissioning of nuclear power plants generates large volumes of radioactive or potentially radioactive waste. The proper management of the dismantling waste plays an important role for the time needed for the dismantling phase and thus is critical to the decommissioning cost. An efficient and thorough process for inventorying, characterization and categorization of the waste provides a sound basis for the planning process. As part of comprehensive decommissioning studies for Nordic NPPs, Westinghouse has developed the decommissioning inventories that have been used for estimations of the duration of specific work packages and the corresponding costs. As part of creating the design basis for a national repository for decommissioning waste, the total production of different categories of waste packages has also been predicted. Studsvik has developed a risk based concept for categorization and handling of the generated waste using six different categories with a span from extremely small risk for radiological contamination to high level waste. The two companies have recently joined their skills in the area of decommissioning on selected market in a consortium named 'ndcon' to further strengthen the proposed process. Depending on the risk for radiological contamination or the radiological properties and other properties of importance for waste management, treatment routes are proposed with well-defined and proven methods for on-site or off-site treatment, activity determination and conditioning. The system is based on a graded approach philosophy aiming for high confidence and sustainability, aiming for re-use and recycling where found applicable. The objective is to establish a process where all dismantled material has a pre-determined treatment route. These routes should through measurements, categorization, treatment, conditioning, intermediate storage and final disposal be designed to provide a steady, un-disturbed flow of material to avoid interruptions. Bottle-necks in the process causes increased space requirements and will have negative impact on the project schedule, which increases not only the cost but also the dose exposure to personnel. For these reasons it is critical to create a process that transfers material into conditioned waste ready for disposal as quickly as possible. To a certain extent the decommissioning program should be led by the waste management process. With the objective to reduce time for handling of dismantled material at site and to efficiently and environmental-friendly use waste management methods (clearance for re-use followed by clearance for recycling), the costs for the plant decommissioning could be reduced as well as time needed for performing the decommissioning project. Also, risks for delays would be reduced with a well-defined handling scheme which limits surprises. Delays are a major cost driver for decommissioning projects. (authors)

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

Analysis of Selected Legacy 85 Kr Samples

Legacy samples composed of 85 Kr encapsulated in solid zeolite 5A material and five small metal tubes containing a mixture of the zeolite combined with a glass matrix resulting from hot isostatic pressing have been preserved. The samples were a result of krypton R&D encapsulation efforts in the late 1970s performed at the Idaho Chemical Processing Plant. These samples were shipped to Oak Ridge National Laboratory (ORNL) in mid-FY 2014. Upon receipt the outer shipping package was opened, and the inner package, removed and placed in a radiological hood. The individual capsules were double bagged as they were removed from the inner shipping pig and placed into individual glass sample bottles for further analysis. The five capsules were then x-ray imaged. Capsules 1 and 4 appear intact and to contain an amorphous mass within the capsules. Capsule 2 clearly shows the saw marks on the capsule and a quantity of loose pellet or bead-like material remaining in the capsule. Capsule 3 shows similar bead-like material within the intact capsule. Capsule 5 had been opened at an undetermined time in the past. The end of this capsule appears to have been cut off, and there are additional saw marks on the side of the capsule. X-ray tomography allowed the capsules to be viewed along the three axes. Of most interest was determining whether there was any residual material in the closed end of Capsule 5. The images confirmed the presence of residual material within this capsule. The material appears to be compacted but still retains some of the bead-like morphology. Based on the nondestructive analysis (NDA) results, a proposed path forward was formulated to advance this effort toward the original goals of understanding the effects of extended storage on the waste form and package. Based on the initial NDA and the fact that there are at least two breached samples, it was proposed that exploratory tests be conducted with the breached specimens before opening the three intact capsules. Portions of these would be analyzed to determine the fraction of krypton/xenon remaining in the matrix and the amount of rubidium remaining in the matrix. The inner surface of the breached capsules would be examined for corrosion. The materials contained in Capsules 2 and 5 have been examined. There appears to be a relatively uniform distribution of Kr and Rb throughout the pellets examined. The chemical composition of the pellets appears to be consistent with 5A molecular sieves. The material contained within Capsule 5 showed ~1 at. % lead. The origin of the Pb is currently indeterminate. X-ray diffraction analysis shows a significant shift from the 5A structure, most likely due to the Kr encapsulation / sintering process that occurred when the samples were made. The capsule walls were also examined and showed extensive corrosion throughout. Elemental mapping of the capsule material appeared consistent with carbon steel, while the weld material appeared consistent with a stainless steel. The interior surface of the capsule appeared to have a layer of material containing Al, Si, and Ca similar to the 5A molecular sieve. Analysis for Rb within the corrosion sites was inconclusive.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Building Confidence in LLW Performance Assessments - 13386

The performance assessment process and incorporated input assumptions for four active and one planned DOE disposal sites were analyzed using a systems approach. The sites selected were the Savannah River E-Area Slit and Engineered Trenches, Hanford Integrated Disposal Facility, Idaho Radioactive Waste Management Complex, Oak Ridge Environmental Management Waste Management Facility, and Nevada National Security Site Area 5. Each disposal facility evaluation incorporated three overall system components (1) site characteristics (climate, geology, geochemistry, etc.), (2) waste properties (waste form and package), and (3) engineered barrier designs (cover system, liner system). Site conceptual models were also analyzed to identity the main risk drivers and risk insights controlling performance for each disposal facility. (authors)

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

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

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

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

Case Study: Systems Analysis of Y-12 Waste Sample Collection

The Environmental Management Organization has established a 90-Day Goal for characterizing, packaging, and dispositioning all waste after it has been declared. If adequate process knowledge is not available, a waste sample must be collected and analyzed. The objective of this case study is to reduce the cycle time for waste sample collection from 14.5 calendar days to 10 calendar days 95% of the time.

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

Requalification and Declaration of High-alpha Cemented TRU-waste from the Late 70's and Early 80's - 20499

KTE and Siemens launched a project to qualify a series of 391 waste drums and 115 concrete containers produced from 1979 to 1984. The waste is characterized by a high content of long-lived alpha-nuclides (TRU-waste - Pu, U, Am in the range of 1 E+11-1 E+12 Bq/ drum. This high radionuclide inventory requires a special packaging according to the waste acceptance criteria (WAC) of the repository Konrad [1] despite the fact, that the waste was treated manually. The raw waste originally was delivered in nitric acid solution and was solidified using ordinary Portland cement. The existing documentation comprises analyses, transfer documentation for the nuclear fuel nuclides and waste treatment documentation yielding information, which waste was treated together in a batch. Additionally all drums were subjected to non-destructive gamma-spectrometry assay. A small number of 7 drums were opened and sampled to support the existing declaration. For the entire series the waste products were attributed to defined batches using the accountancy data supported by the analyses and the gamma-spectrometry. Each batch was produced from a homogeneous stock of waste solution and can therefore be declared and documented together. The declaration concept is based on the gamma spectrometry results performed on the entire drums, which were corrected and extended for the uncertainty of the analyses. The declaration concept is verified by the regulator and its independent expert. The state of the project is in the documentation of the solidification campaigns followed by the planning of the container configuration. (authors)

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

Radiological Characterisation of VLLW Produced at High-Energy Particle Accelerators - 19118

In this paper we describe the characterisation method of very-low-level radioactive waste (VLLW) produced at the European Organization for Nuclear Research (CERN). The method is based on the calculation of a preliminary radionuclide inventory using either analytical or Monte Carlo codes, on the measurement of γ-emitters and on the quantification of pure-α, pure-β and low-energy X-ray emitters, called difficult-to-measure (DTM) radionuclides, using the so-called scaling factor (SF) technique. The first stage of the characterisation process is the calculation of the radionuclide inventory, which is a list of radionuclides that can potentially be produced via the interaction of primary and secondary particles with the surrounding media. Once the preliminary radionuclide inventory is obtained, the γ-emitting radionuclides are measured via γ-ray spectrometry on each package of the waste population. The major γemitter, called key nuclide (KN), is also identified. The scaling factor method estimates the activity of DTM radionuclides by checking for a consistent and repeated relationship between the key nuclide and the activity of the difficult to measure radionuclides from samples. Between 2016 and 2018, several hundred samples were collected at CERN and over 1000 radiochemical analyses were performed to estimate robust statistics to quantify scaling factors for the most common DTM radionuclides, such as H-3, Fe-55 and Ni-63. This paper illustrates the sample collection strategy, the tests performed to choose the most appropriate statistical models describing the behaviour of scaling factors and a complete quantification of the uncertainties associated with the characterisation process. Finally, we present the results obtained during the characterisation of over 2000 m{sup 3} radioactive metals and cables disposed of at the French National repository for VLLW. (authors)

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

Opportunities and limits of Decay storage based on Dismantled Materials in Nuclear Power Plants - 17650

We investigate the potential of decay storage of low and intermediate radioactive waste for release procedures. We present the Safetec software tool HL3 and its application to data from the German nuclear power plant at the site in Stade. In general, during the dismantling process of nuclear power plants, waste of slightly contaminated and/or activated systems is produced in large amounts. These are in particular demolition rubble from concrete structures and ingots from the melting process of metals. If, at the time of disposal, the nuclear inventory is near or slightly above federal clearance limits, the material cannot be released as conventional waste. However, due to the decline of the nuclear inventory over time, a significant amount of that waste could qualify for the release procedure in the future. Although this concept of decay storage is well known, specific predictions of the mass flow to be released is non-trivial. The mass flow over time depends on the composition of known and assumed nuclides (nuclide vector), as well as on the underlying threshold values for release. To handle the waste stream as a function of time, the nuclide composition and specific threshold values, Safetec developed a bespoke software tool named HL3. Its primary function is to visualize and to test/analyze various nuclide vectors and its decompositions over time. For the application within German law, threshold values according to the different disposal routes can be considered. Using the batch interface in the software tool HL3, large amounts of data can be automatically analyzed in a traceable and quick way. To test both, the tool and the benefit for the concept of decay storage, we have applied our code HL3 to material from the dismantling of the pressurized-water reactor at the site in Stade, Germany. We set the focus on concrete debris and metal ingots, which in principle are best suited for decay storage over years. The assignment and the packaging of that waste from Stade were based on application documents for decommissioning of the Year 2000 and according to the state of art conditioning technique at that time. Based on the data of that waste and the appropriate disposal routes, we calculate and discuss the amount of waste releasable after decay storage. As for the situation in Germany, we find that only a negligible amount would be releasable, mainly due to other reasons. These are the deterioration of detectability of radionuclides with time, and other regulatory conditions. However, on an international scale, a different assessment can be drawn. Our results thus motivate to reconsider the option of interim decay storage to reduce radioactive waste in large amounts. (authors)

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

K-Basins Sludge Treatment and Packaging at the Hanford Site - 13585

Highly radioactive sludge resulting from the storage of degraded spent nuclear fuel has been consolidated in Engineered Containers (ECs) in the 105-K West Storage Basin located on the Hanford site near the Columbia River in Washington State. CH2M Hill Plateau Remediation Company (CHPRC) is proceeding with a project to retrieve the sludge, place it in Sludge Transport and Storage Containers (STSCs) and store those filled containers within the T Plant Canyon facility on the Hanford Site Central Plateau (Phase 1). Retrieval and transfer of the sludge material will enable removal of the 105-K West Basin and allow remediation of the subsurface contamination plumes under the basin. The U.S. Department of Energy (DOE) plans to treat and dispose of this K Basins sludge (Phase 2) as Remote Handled Transuranic Waste (RH TRU) at the Waste Isolation Pilot Plant (WIPP) located in New Mexico. The K Basin sludge currently contains uranium metal which reacts with water present in the stored slurry, generating hydrogen and other byproducts. The established transportation and disposal requirements require the transformation of the K Basins sludge to a chemically stable, liquid-free, packaged waste form. The Treatment and Packaging Project includes removal of the containerised sludge from T Plant, the treatment of the sludge as required, and packaging of all the sludge into a form that is certifiable for transportation to and disposal at WIPP. Completion of this scope will require construction and operation of a Sludge Treatment and Packaging Facility (STPF), which could be either a completely new facility or a modification of an existing Hanford Site facility. A Technology Evaluation and Alternatives Analysis (TEAA) for the STP Phase 2 was completed in 2011. A Request for Technology Information (RFI) had been issued in October 2009 to solicit candidate technologies for use in Phase 2. The RFI also included a preliminary definition of Phase 2 functions and requirements. Potentially applicable technologies were identified through a commercial procurement process, technical workshops, and review of the numerous previous sludge treatment technology studies. The identified technology approaches were screened using the criteria established in the Decision Plan, and focused bench top feasibility testing was conducted. Engineering evaluations of the costs, schedules, and technical maturity were developed and evaluated. Recommendations were developed based on technical evaluations. The criteria used in the evaluation process were as follows: (1) Safety, (2) Regulatory/stakeholder acceptance, (3) Technical maturity, (4) Operability and maintainability, (5) Life cycle cost and schedule, (6) Potential for beneficial integration with ongoing STP-Phase 1 activities, and (7) Integration with Site-wide RH-TRU processing/packaging, planning, schedule, and approach. The TEAA recommended Warm Water Oxidation (WWO) as the baseline treatment technology and two risk reduction enhancement options for further consideration during development of the process - size reduction and chemical oxidation (Fenton's reagent). The enhancement options would potentially allow a useful reduction in the total operating time required to process the K Basins sludge. The U.S. Department of Energy's Richland Field Office (DOE-RL) has approved this recommended technical approach. The baseline process can be broken down into the following main process steps: (1) STSC transfer from T Plant to the Sludge Treatment and Packaging Facility (STPF). (2) Retrieval of sludge from the STSCs and transfer to the Receipt and Reaction Tank (RRT). (3) Preparation for immobilization by oxidation using heated water (i.e., WWO) for those batches that require it and concentration by evaporating water at about atmospheric pressure in the RRT. (4) Immobilization by using additives to eliminate free liquids and packaging of the treated sludge into drums. (5) Inspection and handling of the filled drums prior to transfer to a separate storage and shipping facility. (6) Handling of vapor, condensate, and other waste streams generated by the process. Each of these steps is discussed in the paper, together with the current state of progress in developing the technology and requirements for continued development. A schematic of the recommended baseline WWO treatment process is given below. (authors)

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

Milestone report - M4FT-15OR03120210 - Multi-lab Analytical Plan for Analysis of Legacy 85 Kr Samples

Legacy samples composed of 85 Kr encapsulated in solid zeolite 5A material and five small metal tubes containing a mixture of the zeolite combined with a glass matrix resulting from hot isostatic pressing have been preserved. The samples were a result of krypton R&D encapsulation efforts in the late 1970s performed at the Idaho Chemical Processing Plant. These samples were shipped to Oak Ridge National Laboratory (ORNL) in mid-FY 2014. Upon receipt the outer shipping package was opened and the inner package removed and placed in a radiation hood. The individual capsules were double bagged as they were removed from the inner shipping pig and placed into individual glass sample bottles for further analysis. The five capsules were then x-ray imaged. Capsules 1 and 4 appear to contain an amorphous mass within the capsules. Capsule 2 clearly shows the saw marks on the capsule and a quantity of loose pellet or bead-like material remaining in the capsule. Capsule 3 shows similar bead-like material within the intact capsule. Capsule 5 was previously opened. The end of this capsule appears to have been cut off and there are additional saw marks on the side of the capsule. X-ray tomography allowed the capsules to be viewed along the three axes. Of most interest was determining whether there was any residual material in the closed end of Capsule 5. The images confirmed the presence of residual material within this capsule. The material appears to be compacted but still retains some of the bead-like morphology. A conference call was held on February 19, 2015, with participation from ORNL, Idaho National Laboratory, Sandia National Laboratories, and Pacific Northwest National Laboratory Sigma Team members. The nondestructive analysis (NDA) results were discussed, and a proposed path forward was formulated to advance this effort toward the original goals of understanding the effects of extended storage on the waste form and package. Based on the initial NDA and the fact that there are at least two breached samples it was proposed that exploratory tests be conducted with the breached specimens before opening the three intact capsules. Portions of these would be analyzed to determine the fraction of krypton/xenon remaining in the matrix and the amount of rubidium remaining in the matrix. The inner surface of the breached capsules would be examined for corrosion.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
NASA NTRS1979

Food system galley for Space Shuttle

The Galley, a 42 man-day meal preparation facility (with flexibility to accomodate 210 man-days) is described. The facility is designed for minimum meal preparation and clean-up time in zero g and ease of servicing, maintenance, and removal in one gravity. The Galley provides a centralized location for performing all of the food-related functions (except dining) within the orbiter. Consideration is given to the oven-water heater, personal hygiene station, water dispensers, and water supply subsystem. The Galley is positioned in the orbiting mid deck, interfacing with rehydratable food packages and the waste collector subsystem.

Murray, R. W.↗
DOE OSTI2012

Final Demolition and Disposition of 209-E Critical Mass Laboratory - 12267

The 209-E Critical Mass Laboratory was constructed in 1960 to provide a heavy shielded reactor room where quantities of plutonium or uranium in solution could be brought to near-critical configurations under carefully controlled and monitored conditions. In the late 1980's, the responsible contractor, Pacific Northwest National Laboratory (PNNL), was directed by the Department of Energy (DOE) to prepare the facility for unoccupied status. The facility was demolished under a Removal Action Work Plan pursuant to the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA). The funding for this project was provided by the American Recovery and Reinvestment Act (ARRA). The primary rooms of concern with regards to contamination in 209-E facility, which is over 9,000 square feet, are the criticality assembly room (CAR), the mix room, and the change room. The CAR contained two reactor hoods (HO-140 and HO-170), which each had a high efficiency particulate air (HEPA) filter system. The CAR contained 13 tanks ranging from 38 L (10 gal) to 401 L (106 gal). Tanks TK-109 and TK-110 are below grade, and were removed as part of this demolition and disposition remedy. Nonradiological and radiological hazardous substances were removed, decontaminated, or fixed in place, prior to demolition. Except for the removal of below grade tanks TK-109 and TK-110, the facility was demolished to slab-on-grade. PNNL performed stabilization and deactivation activities that included removal of bulk fissile material and chemicals, flushing tanks, stabilizing contamination within gloveboxes and hoods, and packaging and removing waste. The removal of the contaminated plutonium equipment and materials from the 209E facility presented a number of challenges similar in nature to those associated with the inventory reduction and cleanup activities at the Plutonium Finishing Plant. Although there were no bulk fissile materials or chemicals within the facility, there were residual radiological materials (isotopes of plutonium and americium) in the tanks and hoods. The complexity of the remedy was present because of the various configurations of the tanks and hoods, combined with the residual contaminants. Because of the weight and dimensional configuration, size reduction of the slab tanks, as well as removal and disposal of the different material used for moderation and absorption, were two examples of challenges that were resolved to complete the remedy. One of the key methods developed and implemented at the facility was the design and construction of a shroud to allow the cutting of the Pu contaminated tanks. The shroud design, development and implementation at the 209E Project was an example of enhanced work planning and task hazards analysis with worker involvement. This paper will present the lessons learned from the 209E facility inventory reduction activities including the shroud and other methodologies used. The initial Lessons Learned discussion for this project was scheduled for late January 2012. This facility is the first open-air demolition of a highly contaminated plutonium-contaminated facility accomplished by CH2M Hill under the Plateau Remediation Contract. The demolition was completed without spread of contamination to the workers and the surrounding area. As with any project of this complexity, there are significant accomplishments, as well as experience that can be applied to future demolition of plutonium-contaminated facilities on the Hanford Site. These experiences will be documented at a later date. (authors)

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

Inerting of Casks Cavity Atmospheres - The GNS Approach to Treat Hydrogen Generation in Sealed ILW Packages - 16043

A new approach to deal with the issue of hydrogen generation in sealed packages for radioactive waste with organic content is presented. The so-called 'inerting' of the cask cavity atmosphere can be either executed after loading or after interim storage via a new facility introduced by GNS. This new approach is approved by the German competent authority and yield to the recent prolongation of the Type B(U) certificate of approval of the GNS MOSAIK{sup R} cask. (authors)

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

lNAC's OPTIMUS{sup TM}-L Package - the Next Transformation - 19396

NAC has developed the OPTIMUS{sup TM}-L package, the newest addition to the OPTIMUS{sup TM} packaging product line. OPTIMUS{sup TM}-L is a Type B(U)F-96 transportation package designed for maximum flexibility and cost-efficiency to support shipment of a wide range of challenging low-activity wastes. The OPTIMUS{sup TM}-L package utilizes the same cask containment vessel (CCV) as its predecessor, OPTIMUS{sup TM}-H, packaged inside lightweight outer packaging. The large cavity size of the CCV, which can accommodate a 110-gallon drum, combined with the small size and modularity of the OPTIMUS{sup TM}-L package provides unmatched flexibility for the user. OPTIMUS{sup TM}-L can accommodate more radioactive waste in each drum (up to a Fissile Gram Equivalent (FGE) limit of 395 g Pu-239) than larger packages can in 10 drums. Furthermore, because of its small size and weight, up to ten (10) OPTIMUS{sup TM}-L packages can be accommodated per legal-weight truck shipment. Thus, one Legal Weight Truck (LWT) shipment of 10 OPTIMUS{sup TM}-L packages can accommodate more than 10 times the FGE as a large package containing 10 drums. OPTIMUS{sup TM}-L contents include low level waste, mixed low-level waste, intermediate level waste and contact handled transuranic (CH-TRU) waste. Furthermore, all contents can include aerosol cans with compressed gas and liquefied gas propellants and standard Department of Transportation (DOT) 3E lecture bottles for shipment to treatment facilities. The OPTIMUS{sup TM}-L is also designed to accommodate irradiated fuel debris and other fissile material waste, which may be packaged inside shield inserts if required for supplemental radiation shielding. In this paper, NAC provides a technical overview of the OPTIMUS{sup TM}-L package and identifies the design features and technology advancements that makes the OPTIMUS{sup TM} product line a readily adaptable and flexible solution for packaging DOE and other orphan wastes and enables the integration of these wastes into the waste management system process and disposition infrastructure. (authors)

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

Development of Waste Disposal Strategy for RH LLW Stored in the ATR Canal - 19453

The Advanced Test Reactor (ATR), located at Idaho National Laboratory (INL), is 250 megawatt research reactor designed and operated for the United States Department of Energy for irradiation testing of advanced nuclear fuels and material samples. The pressurized light-water moderated and cooled reactor achieved criticality in 1967. It is designed to allow high neutron flux irradiation of structural materials and nuclear fuels to replicate long-term exposure of materials in traditional lower-flux reactor cores. The ATR also produces rare isotopes for the use in the medical and other industries. The ATR is one of only four reactors in the world with this high-flux irradiation capability. Adjacent to the ATR is a water-filled canal which is used to handle casks, prepare experiments, perform irradiation experiments, and store irradiated components generated from reactor operations. The canal is used to store irradiated (activated) metals that are subsequently determined to be waste. In many instances, these items are too radioactive to remove from the canal and remain in temporary storage until a sufficient decay period has elapsed to where they can be safely removed. The more highly-radioactive remote-handled (RH) waste items requiring hundreds of years to decay have been in storage in the canal since the first ATR core internal changeout (CIC) which was conducted in the 1970's. The INL has recently completed construction and startup of a new Remote-Handled Low-Level Waste (RH LLW) Disposal Facility, which will serve as the final disposal capability for legacy RH (greater than 200 mR/hr on contact) metal waste currently stored in the canal. The ATR will need to develop a method for packaging, characterizing, removing, and shipping this material to the new disposal facility. Based on the configuration of the new disposal cells at the RH LLW Disposal Facility, the ATR must produce an entire new process for handing and packaging the legacy RH waste and preparing it for transport and final disposal. The ATR Canal RH LLW Cleanup Project scope is to develop the processes, tools, and equipment for safely sizing, packaging, and removing RH LLW from the canal. This paper presents highlights and challenges the project faces to develop and implement this capability: - Designing new tools, fabricating new equipment, and implementing new processes at an active Hazard Category-I nuclear facility. - Ensuring new equipment and tools are compatible with the existing equipment and processes at both the generator and the disposal facilities. - Sizing legacy waste under water using remote tools and equipment to minimize void space in waste disposal canisters while ensuring as low as reasonably achievable (ALARA) goals are met. - Loading waste containers underwater and preparing them to meet the liquid-free criteria of the disposal facility waste acceptance criteria (WAC). - Providing a new shielded transport container to transfer waste from ATR to the new on-site RH LLW Disposal Facility. (authors)

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

Remote Dismantling and Packaging of the RPV and Thermal Shield at the Obrigheim NPP - 16073

The Obrigheim Nuclear Power Plant (KWO) is located in the Federal State of Baden- Wuerttemberg in Germany and was operated by the EnKK-KWO. The Nuclear Power Plant - a light water pressurized water reactor with 357 MW of electrical power - was operated from 1968 to 2005. In June 2009 the Energiewerke Nord GmbH (EWN) was charged with the dismantling planning and dismantling of the reactor. According to the German nuclear and radiation protection law, several licenses will be required for the realisation of this task. This also applies to the disposal of the waste produced by dismantling, which must finally be stored in the German final repository for low and medium-radioactive waste (underground repository 'Konrad'). In spring 2010 EWN completed all application documents for the dismantling of the reactor pressure vessel (RPV) and the RPV internals and EnKK-KWO delivered them to the responsible Environmental Ministry of Baden-Wuerttemberg. Since then it has been worked intensively on the design, construction and manufacturing of the facilities for dismantling. With a separate notification of amendment, it was applied for the license to install and commission the facilities for the dismantling of the RPV and the RPV internals at the Environmental Ministry of Baden-Wuerttemberg in October 2010. This notification of amendment was confirmed in summer 2012. In September 2011 the work process and test plan and the technical information about the campaign for the removal of the radioactive waste resulting from the RPV with internals, peripheral components and secondary waste with the aim of final disposal in the Federal Repository 'Konrad' was submitted to the Federal Office for Radiation Protection. The evaluation and assessment of the work process and test plan was concluded in August 2013, when it was audited by TUV Nord EnSys without any further measures and indications. On September 10 the work process and test plan was released by the Federal Office for radiation protection. The dismantling of the RPV internals upper core and lower core structure started in September 2013 in the wet cutting area in room 01.106/306 and were finished in October 2014. In the presentation 'Dismantling of the Obrigheim NPP Reactor and Waste Management' held on the WM Symposia in Phoenix, 2014, an overview about the remote cutting and packaging of the upper and lower core structure was given. In today's presentation the subsequent dismantling activities will be outlined and evaluated. The presentation will start with the short description of the basic frame conditions for the nuclear licensing procedure, important procedural principles, and necessary plant-specific conditions. After the remote cutting and packing of the reactor internals of the upper and lower core structure was completed, the wet cutting area 'reactor room 01.106/306' was cleaned and the reactor room rebuilt into a cutting area for peripheral components. After the peripheral components were dismantled, all preconditions were set to lift the RPV from its mounting position and to arrange it in the upper reactor room 01.306. After these activities were completed, the lifting of the RPV was prepared and the cutting area 'Dismantling of the RPV insulation in the reactor room 01.106/306' was set up. To carry out this task, a strand jack unit with connection block was installed on the bridge of the reactor building crane. The opening created by the dismantling of the pool plate was closed with a flap door. The manipulators used for the wet cutting of the RPV internals were also used for the remote dismantling of the RPV insulation and mounted in the cutting area. The remote facilities were handled from a control centre and/or control stand. The air-technical housing which has already been used for the remote dismantling of the RPV internals was designed for the dismantling of the RPV insulation and was extended for this dismantling scope with appropriate ventilation systems. A special feature is the fixation of the RPV insulation at the RPV corpus. The RPV insulation consists of insulation wool, wire mesh and metal-sheet insulation. This 'Sandwich Construction' is detached with a steel corset at the RPV corpus and partly tightened. The insulation of the RPV bottom dome is tightened at the supporting frame. Due to the activation of the material in the core area and the resulting high ambient dose rate the RPV insulation can only be dismantled remotely. Additionally, as asbestos fibers are expected to be found in the insulation the dismantling has to be performed according to the Technical Rules for hazardous substances. The demolition, restoration or maintenance works have to be performed remotely according to the Technical Rules for hazardous substances when the RPV is located in the cutting area. But in case of dismantling of a power reactor no such experience can be found. Beginning with the explanation of the technical, radiological and substantial boundary conditions this presentation will not only impart knowledge about the dismantling of the RPV insulation but also about the remote controlled packaging of the metal sheet insulation and the manual conditioning of the RPV insulation. Subsequently, reports on the remotely controlled cutting and packaging of the RDB are given. Here is particularly interesting that cutting methods have been used and how the activity determination of the segments was carried out. On the basis of the evaluation of the realisation the subsequent dismantling activities will be lined out. In parallel to the dismantling tasks, the waste package documentation for the final repository was issued, checked and confirmed for the containers loaded with parts of the RDB internals. The most important stages are exemplified in this presentation. (authors)

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DOE OSTI2015

Characterization of crystalline ceramic waste form

The research conducted in this work package is to collect data in support of future development of ceramic waste forms and to help develop the next generation high performance waste management technologies. This work package collaborates closely with the work conducted under “Ceramics - Savannah River National Laboratory (SRNL) - (FT-15SR030702)” and the NEUP project “Alternative High-Performance Ceramic Waste Forms” – Alfred University (AU) – (DE-AC07-05ID14517)”. The Los Alamos National Laboratory (LANL) is focusing on characterization and testing procedure on the reference crystalline ceramics to further develop the crystalline ceramic waste form to incorporate combined HLW raffinate stream and/or TRU stream from fuels fabrication processes. In Fiscal Year 2015, we are mainly working on cold crucible induction melter (CCIM) test multi-phase crystalline ceramic waste form samples and single phase hollandite materials. Various characterization techniques (including X-ray diffraction, transmission electron microscopy, scanning electron microscopy) were used to characterize CCIM samples. For the radiation stability test, in situ radiation techniques were employed to simulate self-radiation in waste form samples.

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