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All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

A Novel Mobile Examination and Remediation Facility for On-Site Remediation of Dry Storage Systems - 18389

Most of the spent nuclear fuel (SNF) in dry storage in the United States is safely and securely stored in dual-purpose (i.e., storage and transportation) canisters [1]. However, some of the SNF is stored in storage-only (i.e., non-transportable) packages, or packages that may become nonconforming for continued storage or transportation for a variety of reasons. Currently, the capability to remediate dry storage systems (DSSs) for SNF that are not transportable or that need to be repaired to maintain confinement boundary integrity during storage service does not exist without on-site access to a spent fuel pool. Hence, there is a potential future need for a mobile facility with remediation capabilities that can be used at multiple storage installations. These capabilities would permit needed repairs to an existing DSS to support continued storage or for the dry transfer of fuel to a new conveyance suitable for transport, if necessary. This is a technical paper that does not take into account the contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in canisters to be an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail. A novel design concept for a Mobile Examination and Remediation Facility (MERF) is proposed with functions and capabilities to remediate a DSS. The proposed MERF would be a multi-purpose facility, remotely operated and computer-controlled with semi-automated equipment to allow trained operators to safely perform remediation activities while keeping worker radiation dose as low as reasonably achievable. It would include a shielded rotating platform with support towers and would use commercial off- the-shelf examination and repair equipment. A slip ring arrangement underneath the rotating platform would allow for the power supply without impeding the continuous rotation and would simplify cable management. The support towers would be mounted on the rotating platform, and each tower would have vertical and horizontal positioning capability for axial and radial tool location. The rotating platform would be attached to the main MERF structure. Capabilities would include visual, surface, and volumetric examinations to fully characterize physical conditions on canister surface before and after a repair; repair techniques, e.g., laser beam welding, friction stir welding, or cold spray, to perform needed remediation activities; and associated equipment for surface preparation such as cutting and grinding equipment. The proposed design would be simple, modular, robust, economically viable, and easily transportable and would have a small footprint. The modular design would present several advantages including compatibility with a variety of different DSSs, on-site assembly, and transportability; it also would be built in interchangeable sections to facilitate decontamination. Further, shielding would be supplemented through interface with the existing transfer cask and storage overpack. Robustness and canister stability would be achieved using pneumatic holders and steel support structures designed with vibration absorbers to reduce impacts and/or a 'pendulum' effect during a seismic event. Thus, the proposed MERF would be designed to: (i) minimize handling of an affected DSS to the extent practicable, (ii) provide radiation shielding and minimize dose to operators and the public during remediation activities, and (iii) enable examination, repair, and repackaging. This paper describes the proposed MERF design concept and capabilities, main components, equipment, different modes of operation, and operating procedures. Structural and safety assessments would need to be performed before final design, licensing, and deployment. The MERF is evaluated with respect to regulatory impact (first of a kind, prior experience), operational impact (safety, security, staffing, ongoing operations, constraints), worker dose (low, medium, high), cost (low, medium, high), schedule (duration, time sequence), and technology readiness level. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Spectroscopic and Computational Analysis of Chemical Pollutants Interactions on Steel - 18409

There is limited understanding of the scientific interaction of the chemical contaminants to DOE facility materials which can alter the stability of the materials, affecting the general safety. The resulting studies will have a direct impact towards developing an enhanced cost-effective decontamination method and understanding the negative effects caused by the chemical waste on the structural integrity of the solid surface in contact. First order kinematic models provide a better fit to the experimental data, implying that one layer of Hg adsorbs to the steel surface. Energy dispersive X-ray analysis confirmed that stainless steel samples are mainly composed of Fe, Ni, and Cr, and have a low carbon content. Objective: To determine the physiochemical interactions of Mercury (Hg) and Cesium (Cs) to steel and concrete using spectroscopic, adsorption and computational studies. To develop a multiscale simulation that shows the effects of adsorption and diffusion of alkali metals on the mechanical performance of steel alloys used in nuclear power plants. Approach: Molecular Dynamics (MD) simulations were used to compute the stress gradient generated in steel due to Hg adsorption. Finite element models were developed for simulating the effects of temperature and pressure in a diffusion analysis. Based on experimental data, one layer of Hg was used for the MD simulations. Material properties at the atomic level were determined from the MD simulations The stress gradient from the MD simulations is used as an input parameter in the FE simulation. Conclusions and Future Work: Computational methods such as MD and FE simulations were used to determine the effects of mercury adsorption in stainless steel samples. The adsorption process induces a stress gradient in the facility materials, which can affect the mechanical performance of facility materials under service. Tensile stresses help the diffusion process while compressive have the opposite effect. Temperature has a positive effect in the mass diffusion process. Future work will evaluate the effects of temperature on the material properties at the molecular level using both experimental and MD methods. Future work will evaluate the mechanical performance of facility materials at a wide range of environmental variables.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Management of Metals from Decommissioning of Nuclear Facilities - 18411

The EDF group companies within the Cyclife platform have for decades performed treatment and recycling of metals. With different approaches but all considered international Good Practice, as valuable materials are recovered for safe recycling and volumes in the repositories for radioactive waste are saved. For an efficient decommissioning project, a well-defined waste management process is one of the critical paths, not least when it comes to routes for clearance and recycling of materials. It is easy to underestimate the complexity and bottlenecks caused by shortage of logistic space. In order to mitigate these risks, activities such as strengthening categorization, quantification and adopting different routes based on the material properties should be implemented. By categorization of the of waste and materials considered/classified as contaminated or potentially contaminated it is obvious that the management of the metals is a key parameter in nuclear decommissioning. Analyses performed by the EDF companies shows that a majority of the materials, classified as contaminated, from a BWR/PWR plant are metals. Experience gives that a high percentage of these metals from decommissioning can be subject to recycling either directly upon completion of the clearance process or after certain treatment. Non-technical key parameters like the regulatory requirements on applying the waste hierarchy, the impact on repository programs as well as, not at least, the regulatory and public acceptance of recycling of materials from the nuclear industry are, together with costs essential factors when forming the routing of metals and other waste materials from a decommissioning project. Today the EDF group companies are positioned to provide a wide range of services for management of the contaminated metals from decommissioning projects, spanning from industrial management of metals which directly will be subject general clearance after verification measurements up to decontamination and melting of selected reactor internals for recycling to public domain or within the nuclear industry. In this paper the typical waste material composition for BWR and PWR plants, the different ways to treat/manage the radioactive and potentially radioactive metals including a comparison of the different approach commonly used in a technical, logistical and economical perspectives. A few success stories will be resented as well as lessons learned. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

3-D Radiation Mapping and Data Fusion for Environmental Remediation and Cleanup - 18449

Recent developments in radiation detection and imaging in combination with the enormous advances in computer vision and data processing enable new and improved means in the assessment of radiological contamination relevant for environmental remediation. We have developed a new concept called 3-D Scene Data Fusion (SDF) that allows us to map scenes in three dimensions and fuse this scene data with nuclear radiation data in near real time while the instrument is freely moving through the scene. This new technology provides more accurate radio-isotopic mapping at higher speed than currently possible. SDF is designed to be independent of the deployment platform and can be implemented with gamma-ray imaging and non-imaging instruments. We have demonstrated this concept in a large range of configurations on unmanned aerial and ground-based systems, in hand-portable configurations or mounted on cars or manned helicopters. While it can be utilized in a wide range of applications, this paper focuses on the mapping of radiological and nuclear materials in the context of environmental remediation and clean-up specifically in urban environments. We show results from measurements in the evacuated areas in the Fukushima Prefecture to demonstrate the new concept in the mapping of contamination and hot spots and the verification of decontamination activities in urban environments. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

D and D of a Large Waste Treatment Plant in Germany - 18458

With its 40 years of nuclear experience, GNS, the world leading supplier of casks for spent fuel, ILW and HLW, also offers services for conditioning and management of all types of radioactive waste, especially for the German NPP. Since 1985, GNS treats and packages ILW and LLW from the operations and phase-out of German NPP in its Duisburg plant. To enable its permanent safe final disposal, solid wastes are compacted and dried before being shipped to an interim storage facility or - later - the Konrad repository. The total premises cover 20.000 m{sup 2} (215.000 sq ft), of which 9000 m{sup 2} (97.000 sq ft) is controlled area. The Duisburg Plant has three large industrial warehouses (each 25 m x 136 m or 82 ft x 446 ft) with a total capacity of 10.200 m{sup 2} or 110.000 sqft. For the treatment of open contaminated waste, six enclosures were built. These enclosures have a special ventilation system including a specific filtration plant. In one enclosure a high compaction plant was used, in another enclosures a drying facility for drums was in operation. During peak periods, more than 900 tonne of waste were treated annually. One enclosure was mainly used for maintenance, restoration and extension of mobile facilities, which were normally used for campaigns in NPP. Due to the decreasing amount of operational waste during the nuclear phase out in Germany, GNS has decided to close the Duisburg Plant. In the first half-year of 2017, the last waste packages were treated but around 80 20'-containers are still located on site. The D and D started in January 2017. The presentation will show the first experience of decommission processes of the first three enclosures. The following steps including the techniques used will be explained in detail: - Dismantling and removal of the high compaction facility; - Dismantling and removal of the drying facility; - Cleaning and wipe tests of the empty enclosures; - Removal of the stainless steel plates of the floor; - Investigations of the lower floor parts for contamination; - Removal of contaminated underground drainpipes; - Decontamination of the concrete floor; - Dismantling of the ventilation system in three enclosures; - Dismantling of the steel constructions. An important aim during the dismantling is to reduce the amount of nuclear waste and maximize the fraction material for free release. Next to the first three enclosures, in one of the enclosure IV - VI, a 150 tons scrap shear had been installed before the construction of the enclosure building. The scrap shear has to be dismantled and handled by heavy duty lifting equipment. Only afterwards is it possible to start the decommissioning of the last three enclosures. Our goal is to transfer this very large controlled area and the whole plant to the site owner, the Duisburg Port company, after 2 years and 5 months and doing so in a way so that the industrial warehouses can be used again for all possible types of industries. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Decommissioning of the Chernobyl NPP: Problems, Progress, Prospects - 18485

Decommissioning of the Chernobyl NPP provides a special case due to a number of peculiarities: it is prescheduled decommissioning, the facilities undergone a major accident with off-site contamination, no previous experience of decommissioning of RBMK-type nuclear power units, unavailability of proper spent fuel and radioactive waste management infrastructure, special types of waste. The most challenging issue is the final decommissioning of the Unit 4 destroyed by the major accident on 26 April 1986. Neither of four RBMK-1000 nuclear power units was in operation till the end of the design-based lifetime, Units 1 and 3 being in operation for 19 years each, Units 2 and 4 even less. The last operating unit, Unit 3, was permanently shutdown on December 15 2000, however the main obstacle to complete a complex technological and radiological survey was spent fuel left in reactor pools and in the core due to lack of capacity in the existing interim spent fuel storage of wet type (ISF-1). So removal of spent fuel from the core and reactor pools was considered as one of the priority tasks. While construction of ISF-2 (of dry type) was delayed due to different reasons, safety reassessment and modernization of ISF-1 enabled capacity increase and transfer of all non-damaged fuel assemblies into ISF-1. Thus Chernobyl NPP Units were fully vacated from spent fuel. Since that Chernobyl NPP Units are considered as the radioactive waste facility. Dismantlement of Units 1,2 turbine hall equipment is on the way and some progress is achieved in technical solution for fragmentation of long-size equipment. All dismantled equipment after decontamination and characterization is either released from regulatory control or disposed as solid radioactive waste. The details are discussed in the paper. Eventually the comprehensive technological and radiological survey of Chernobyl NPP Units had been completed and detailed decommissioning programme had been endorsed. The decommissioning strategy chosen for the Chernobyl NPP is based on the deferred dismantlement approach, so final dismantlement of all equipment and constructions is supposed after some period of decay storage. However, the issue of the final state of the site and its further use is still open. The main reason for it is the lack of clear picture as for the destroyed Unit 4, as technical solutions concerning its highly damaged and contaminated constructions and so called fuel containing masses (FCM) are not available yet. It could be reasonably predicted also that the internal part of the so called exclusion zone will not be released into non-restricted use in near future due to characteristics of its contamination. Possible implications of these and other factors on further decommissioning activities and some approaches to find solutions are discussed in the paper. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Environmental and Radiological Readiness of Fixatives, Foams, and Intumescent Coatings for D and D Applications - 18553

During D and D operations of legacy nuclear facilities, the ability to effectively stabilize contamination in large cavities is a significant challenge. For items that must be removed, such as gloveboxes and entry hoods, a method of simply filling the volume to stabilize any residual contamination after contamination efforts have concluded would be of great benefit, and further the As Low As Reasonably Achievable (ALARA) culture of the DOE complex. The development, testing, and characterization of commercial fixatives and foams for applicability towards fixating contamination provides a potential structural reinforcement and fire resiliency in legacy nuclear facility equipment. Much of the equipment, such as gloveboxes, shielded cells, and radiological hoods left over from nuclear processing leave large cavities. Filling these voids after decontamination efforts provides a means of shielding workers from excess radiation exposure while immobilizing residual contamination. As such, we look to continue the development of fixatives/intumescent coatings from the previous year and test commercially available polymer-based foam products for use in filling these cavities, providing a secondary level of protection and shielding during the D and D process This research is funded as part of the In-Situ Decommissioning and Technology Development and sponsored by DOE- Office of Environmental Management Technology Development. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Commissioning and Startup of a New Category 2 Nuclear Facility for RH TRU Sludge Packaging - 18614

For many years the Department of Energy (DOE) has been executing a plan to eliminate all threats of Radiological release to the Columbia River from the Hanford site, in Richland, Washington. While the River Corridor Closure Project is nearing completion, one of the last remaining facilities along the Columbia River is the 100K reactor complex. Since the 1950's the 105KW fuel storage basin has contained varying levels of spent reactor fuel and products of the degradation of aging stored fuel. Having previously removed all of the distinguishable spent fuel components, CH2M HILL Plateau Remediation Company (CH2M) is now in the final stages of readiness review to commence removal of the residual sludge material that has accumulated in the basin throughout the years of operation. There is approximately 27 m3 of sludge in the 105KW basin currently. This sludge material exhibits a very high radiation dose rate with levels reaching 100 R/hr. It is also very difficult to mobilize and transfer due to its physical characteristics. CH2M engineers have developed a novel mechanical system for mobilization and transfer of the sludge to specially designed storage containers. Once transferred to these containers, the sludge will be moved away from the river and taken to T Plant, a decontamination and repair facility on site, for interim storage, while a project to treat and package the material for final disposition is established. The Sludge Removal Project (SRP) is a $309 M capital line item project that was established to achieve the goals of this mission. Part of the project is to develop and install the equipment needed to do the transfer, and part of the project is to construct a new Category 2 sludge packaging facility. Construction of the facility is now complete, all equipment is installed and the project is well into the commissioning phase. Over the final three months of 2017, the project progressed into the final stages of readiness to transition to operations. This will represent the first commissioning of a new Category 2 process and facility at Hanford in over 10 years. The SRP team has crossed over many hurdles in establishing a workable plan for the safety basis for installing, commissioning and transitioning to operations within the boundaries of an existing Facility Safety Analysis Report. The project team has also applied extensive use of a mockup facility in the development and execution of the path to readiness and commencement of operations. At the time of the conference, the project will have completed a DOE Operational Readiness Review and will be very near of just entering full operation. This paper will share the experiences of the project team through these final steps and lessons learned on making the project successful. (author)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Result of Investigation for Inside PCV in 1F-1 by Shape Changing Robot named PMORPH - 18617

In Fukushima Daiichi Nuclear Power Station, before retrieval of molten fuel debris inside PCV, the operation such as decontamination, survey, repair and taking sample of fuel debris are implemented step by step. In case of investigation inside primary containment vessel (PCV), some remotely operated robots have been used for because human can't access to high radiation environment. Because a radiation dose is very high in PCV, therefore it is necessary to prevent the leak of radiation from PCV when a robot is inserted in PCV. Then a pipe of 100 mm inside diameter was let go through by the PCV wall, and the seal box which was completely sealed up attached to the end of pipe. The robot is put in the seal box beforehand and is inserted inside PCV through the pipe. So it is necessary for the robot to be small. In addition, it is necessary to be stable movement on rough plane inside PCV. Therefore shape change type robot 'PMORPH' was developed. This robot consists of three parts of the slim main body and two crawlers attached to the both ends of main body. When this robot is passed in the pipe, three parts lie in a straight line like 'I' shape. And when the robot moves on a rough plane, three parts are formed like 'U' shape. We minimized the use of electric parts in the main body against high radiation environment. And this robot carries a winch to let a sensor unit descend to basement floor. And we carried out the basement floor investigation (B2 investigation) using the robot to obtain information of fuel debris condition in March, 2017. As a result, some information of basement floor condition was acquired. In this paper, it is described the outline of the shape change robot and summary of the B2 investigation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

System Integration for Remote Systems at DOE Hanford Nuclear Site - 18712

The deployment of remotely operated technology that functions in environments too hazardous for worker entry are often needed at the Hanford Site to perform functions like infrastructure inspection; decommissioning and decontamination; characterization; environmental sampling; and nuclear waste sampling, mobilization, and retrieval. The development and production of Hanford remote systems typically requires the integration of robotic components and sensor technologies that have been engineered or adapted to meet function and performance requirements. Pacific Northwest National Laboratory has led the engineering and integration of robotic and sensor/manipulator technology for several remote system deployments at Hanford and for national security sponsors. This paper describes the system integration process for ensuring that the mechanical, electrical, and software aspects of each component are successfully brought together to produce a remote system that meets function and performance requirements under extreme environments and limited-access operational demands. A brief summary of previous and current remote system projects is also provided. (authors)

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

Tritium Distribution and Capture Requirements for UNF Reprocessing

Four radionuclides have been identified as being sufficiently volatile in reprocessing of nuclear fuel that their gaseous release needs to be controlled to meet U.S. regulatory requirements. These radionuclides are {sup 3}H, {sup 14}C, {sup 85}Kr, and {sup 129}I. Previous studies have discussed the need for treating off-gas streams to control {sup 3}H, {sup 14}C, {sup 85}Kr, and {sup 129}I from the reprocessing of used nuclear fuel (UNF) and the regulations that set the limits for releases from such a facility. These regulations ultimately prescribe the decontamination factor (DF) that must be met in the facility. Jubin et al. examined the role of off-gas pathways since the gaseous radionuclides can partition to different reprocessing off-gas streams. Based on this report and subsequent analysis, the required DF for {sup 129}I could be 1000 to 3000. The required DF for {sup 3}H could be as high as 200, depending on the age of the fuel processed. The DF for {sup 85}Kr could be up to ∼35, also depending on fuel age. The required DF for {sup 14}C is 1 in many cases (no treatment required) but could be higher in specific cases. While iodine has been the focus of a number of studies due to the high DF requirements on multiple escape pathways, it is clear from the studies noted above that the control of tritium will also be a significant issue for future reprocessing plants, especially if processing relatively short-cooled fuels (<10 yr post reactor discharge). Tritium contained in a UNF matrix will be released into the liquid and off-gas streams during fuel reprocessing. The amount released into the process streams is a function of the type and burnup of the fuel processed, and the chemical processes employed within the reprocessing plant. Previous studies have examined the tritium capture requirements for the reprocessing facility assuming that all of the tritium is volatilized in the head-end operations, i.e., tritium pretreatment is employed and tritium is completely released from the cladding. This is clearly a worst case scenario from the standpoint of the amount of tritium that could be released, but the scenario tends to limit the release pathway to the head-end off-gas stream. In this analysis, the volume of HTO and the HTO to H{sub 2}O ratio for each of the three plant-wide tritium DFs were determined. The overall amount of HTO that had to be captured from the off-gas streams was virtually the same for the three DF cases evaluated; 98%, 99% and 99.5% of the tritium in the fuel entering the plant needed to be recovered for overall plant DFs of 50, 100 and 200, respectively. The use of the 'tritium barrier' to limit the spread of HTO from the head-end systems to the balance of the plant results in an increase in tritiated waste concentration in the head-end recycle acid systems which results in a decrease in the total amount of HTO + H{sub 2}O that must recovered from the gaseous effluent from the plant. The use of the 'tritium barrier' decreases the waste volume due to the higher HTO:H{sub 2}O ratio in the treated off-gas stream. The results also show that increasing the acid recycle percentage increases the concentration of HTO within the UNF reprocessing plant. This is then manifested in less HTO + H{sub 2}O that must be recovered from the gaseous effluents. Finally, the application of tritium recovery on the recycle acid reduces the HTO concentration in the recycle acid stream and generates a separate HTO + H{sub 2}O stream that must be managed. However, the resulting volume of tritium-containing waste generated plant-wide is a function of the selectivity of the tritium recovery system and the ability of the tritium recovery system to limit the accumulation of HTO in the recycle acid such that the HTO + H{sub 2}O recovered from gaseous effluents does not contribute significantly to the overall tritium waste volume. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Different Approaches for Clearance of Metals - a Comparison - 19079

Large quantities of contaminated and potentially contaminated metals are generated during the lifetime of nuclear installations, especially when they are dismantled and demolished. In many countries, clearance for recycling is considered Best Practice for the management of metals which are contaminated or potentially contaminated. In some countries, such as Sweden, it has become a legal requirement to recycle metals and other materials when it is safe to do so and the cost is reasonable. Therefore, multiple management routes are available with the aim of establishing a graded approach. These include general (unconditional) clearance, conditional (specific) clearance of surface contaminated metals, conditional clearance of metal ingots after treatment in a licensed facility and in some cases, recycling inside the nuclear sector. Kemakta Konsult AB recently derived rounded clearance limits for the conditional clearance of scrap metal for re-melting from the decommissioning of Swedish nuclear sites, on behalf of the Swedish Radiation Safety Authority (SSM). The calculations used the RP89 exposure scenarios, but the parameters were adjusted to reflect normal practice in the Swedish scrap metal and recycling industries. The study found that, for most of the 39 radionuclides judged to be relevant, the rounded clearance levels were the same as or higher than those in RP 89. However, longer transport times in Sweden compared to the ones used for RP 89 resulted in lower rounded clearance levels for three gamma-emitting radionuclides, including Co-60, the most important radionuclide in terms of dose exposure. Cyclife Sweden AB has, through more than 40 years of operation of the metal treatment facility at the Studsvik industrial site, significant experience of the practical implementation of clearing metal ingots as well as metallic strap, in total about 50,000 tonnes. A previous study confirms very low radiological impact to the receivers of the ingots and to the general public under the current procedures and conditions for clearance. The production of steel ingots for recycling in the Cyclife plant involves pre-treatment (segmentation and decontamination) followed by melting. Certain nuclides, like Cs-137 and alpha nuclides, are transferred to the secondary waste (slag and filter dust) in the melting process. The non-separated nuclides are homogenized in the ingot. The radionuclides that are present in the ingots are expected to remain in the metal during re-melting at metal works. The general approach of [1] together with specific information for typical ingots has been used to further verify the radiological implications of applying conditional clearance of ingots. This paper provides a background, present the studies performed, give an overview of the findings and the overall result including a comparison between different approaches. (authors)

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

Cameco Corporation's Approach to Legacy and Ongoing Waste Management Within the Fuel Services Division - 19124

Since 2011, a Fuel Services Division (FSD) employee has been dedicated to oversee divisional waste management with the goal of accelerating the recycle/disposal of accumulated legacy radioactive wastes stored at its Ontario operations in Blind River and Port Hope. The Blind River Refinery (BRR) and Port Hope Conversion Facility (PHCF) were previously owned by the federal Crown Corporation, Eldorado Nuclear, becoming part of Cameco in 1988 and Cameco Fuel Manufacturing (CFM), previously known as Zircatec Precision Industries, becoming part of Cameco in 2006. Accumulated legacy radioactive waste is stored at each of the facilities. This radioactive waste has accumulated from operations beginning in the 1930's in Port Hope and the 1980's in Blind River. Due to minimal commercially available outlets for the disposal and/or treatment of these wastes, the waste was managed onsite by storage in underutilized space in buildings, trailers and/or in the secured yard. This relatively low-cost practice meant that over decades, the inventory increased at an unsustainable rate until available space became a limiting factor. During this time, the labelling, inventory records, characterization details and other history of many of the waste materials was not maintained adequately. Due to the volume of waste that had accumulated, campaigns to address these deficiencies in information were not considered feasible, sustaining the issue of accumulation of low-level radioactive waste. Since 2011, FSD has completed annual campaigns to address the accumulated legacy waste, which has included processing of material for decontamination to meet the unrestricted release criteria under the Canadian Nuclear Substance and Radiation Devices Regulations (NSRDR), disposal of marginally contaminated wastes (<500 ppm uranium) at a permitted landfill in the United States, processing of contaminated materials to reduce the volume of low-level radioactive waste and disposal at a designated facility in Port Hope. In addition FSD has undertaken significant changes to radioactive waste management practices at each of the facilities. Multiple characterization techniques, including visual observation, review of process and other historical records, laboratory analysis and radiation scanning have been utilized in campaign planning to identify the scope and appropriate waste management technique(s) to reduce the inventory of stored material. This divisional waste management program has resulted in a significant change to how radioactive waste is generated, inventoried, characterized and managed within FSD and has increased the visibility within the leadership team of the opportunities and challenges of waste management, such as alignment with decommissioning planning, project planning and budgeting. The heightened awareness has led to better financial and physical resource planning to execute waste projects with a current projected elimination of the legacy waste within FSD by 2025. (authors)

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

Dismantling of the Spent Resins Hot Compaction Unit at Tihange NPP, Belgium - 19139

Design, construction and commissioning of the Spent Ion Exchange Resins Hot Compaction Unit at Tihange Nuclear Power Plant took place from 2005 till 2011. In this conditioning process, batches of radioactive resin beads were first dried to less than 4% moisture by a conical vacuum dryer. Resins were then poured in compactable 200-liter drums and super-compacted. Resulting pellets were emplaced in standard 400-liter drums, the packages qualified for final disposal in Belgium, and eventually grouted. The Belgian Waste Management Agency (ONDRAF/NIRAS) was consulted before initiating the project and issued a statement of no objection. The technology turned out to be used for the first time in the particular case of bead shaped resins (unlike powdered resins) and with a grouting requirement. Finally, commissioning tests ended in a safe and reproducible waste form, compliant with prevailing Waste Acceptance Criteria. Volume Reduction Factors of about 2 were recorded after drying and compaction. New tests were carried out to support the Qualification File of the process, which was submitted to ONDRAF/NIRAS. After performing many trials using inactive resins, the Operator dried two batches of radioactive resins and super-compacted four drums: this was a planned qualification step, aiming at the validation of process parameters. In the meantime, ONDRAF/NIRAS made parallel progress in the design of the future Near Surface Repository, where conditioned resins will be disposed of, and found some incompatibility between encapsulated dried resin beads and water intrusion hazards inside the repository. Based on the risk that dried resin beads will swell if they absorb water, causing mechanical stresses likely to compromise the integrity of the disposal site, ONDRAF/NIRAS informed Tihange NPP that the Hot Compaction Unit would not be qualified as spent resins conditioning process. Further to this irrevocable decision, the Operator undertook three actions: - The first one was the qualification and purchase of waste containers to provide additional temporary storage capacity on site for its operational spent resins; - The second one is the selection of a new spent resins conditioning process and the performance of the related qualification tests to demonstrate it meets the updated Waste Acceptance Criteria; - The last one is the dismantling of the Hot Compaction Unit to clear the space and enable the installation of new equipment. This latter task was carried out in 2018, involving decontamination and free release measurements of all the materials, a very limited fraction of it having to be further processed as radioactive waste. Nevertheless, Tihange NPP wishes to keep the super-compactor and its hydraulic unit for a possible resale or reuse for the treatment of mixed waste. (authors)

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

Win-Win Regulatory Strategies in Idaho - 19241

During the course of any waste management or environmental remediation project, the earliest goals must include the creation and maintenance of mutual trust, respect, and cooperation with regulatory agencies and stakeholders. The heavy equipment and treatment systems remain idled until bounding characterization and risk analysis is complete, technical and functional requirements are identified, required controls and limits are established, and the proper permits and approvals are issued. Once a project receives approval to start up or break ground, the regulatory process has only just begun. Process modifications and improvements, as well as unexpected events, will continue to necessitate further reviews, discussions, and approvals. Building healthy professional relationships with the regulatory community and consistently meeting or exceeding their expectations, while operating within controls, is a cornerstone to jointly finding cost-effective solutions to complex issues, and streamlining the path to success. Navigating the optimal routes to compliance through the myriad regulatory waypoints is essential to avoiding schedule delays, staying within budget, and keeping clients continuously satisfied and delighted. Two complex case studies from the Idaho Cleanup Project Core will examine regulatory communication and integration strategies that have proven effective to date in obtaining timely approvals and maintaining compliance: (1) conducting Decontamination and Decommissioning as non-time critical removal actions (NTCRA) under CERCLA, and (2) preparing the integrated documentation necessary to obtain DOE, EPA, DEQ, and NRC approvals to close calcine storage units at the Idaho Nuclear Technology and Engineering Center (INTEC). (author)

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DOE OSTI2019

Stabilization of Residual Contamination with Alternative Materials - 19404

Since 2015, Savannah River National Laboratory (SRNL) in conjunction with Florida International University (FIU) have been investigating the applicability of intumescent coatings as radiological contamination fixatives for stabilization of residual materials after decontamination efforts have concluded. Intumescent coatings are designed to prevent heat transfer to a protected substrate and are typically used in both the aerospace industry to protect components from thermal shock and in building design to protect building joints that may fail in the event of a fire. For this work, these materials are examined for their ability to reduce worker risk by fixating residual contamination and lower the impact of a release in a scenario that breaches containment of a contaminated building and subsequently induces a large-scale fire by crediting their fire-retardant attributes. Initial studies of intumescent coatings and commercial fixatives included performance in direct flame, varied environmental conditions (20-110 F or -6.67 C, 10-95% relative humidity), adhesion studies to stainless steel after curing in varied environmental conditions, and radiological exposure. Of all the materials tested, many were found to be inadequate in one or more of these areas. These materials failed in one or more of three characteristic ways: 1) through delamination at reduced temperatures (≤40 F or 4.44 C) or from even moderate handling, 2) melting off the substrate at elevated temperatures (≥90 F or 32.22 C), 3) proving to be flammable and promote smoke propagation. Failure in these ways results in the potential for a subsequent release of the fixated contamination in the event of water presence/absence or direct flame. Using these results as a platform for the best in class identified material that did not experience failure in any of these tests, SRNL and FIU targeted the plutonium-238 contaminated hot cells within Savannah River Site (SRS) Building 235-F as a characteristic test bed to benchmark an intumescent coating used as a contaminant fixative. Through discussion with the SRS 235-F Risk Reduction personnel, appropriate hazards and performance metrics were discussed and a cold demonstration was performed in a hot cell mockup located at SRS. The application methods were proven functional in the cold mock up, and these intumescent coatings were subsequently utilized in the Savannah River Site Building 235-F's Plutonium Fuel Form (PuFF) Facility to aid in stabilization of plutonium-238 in both a contaminated hot cell and Entry Hood in September 2018. Presented in this work is the overview of the application process, associated working considerations for introduction of equipment and materials into the facility, lessons learned, and lifetime monitoring data to date. (authors)

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DOE OSTI2019

Panel Session 53: Japan Fukushima Daiichi Decommissioning Update (1/2)

This panel focused on up-dating the activities at Fukushima Daiichi. Panelists include Japanese representatives from Tokyo Electric Power Company Holdings, Inc. (TEPCO) and invited US Institute specialists. Presentations proposed include: 1 - Fukushima Daiichi Decontamination and Decommissioning: Current Status and Challenges (Akira Ono, TEPCO Holdings (Japan)); 2 - SRNL Engagement in Japan Recovery: From Clean-up to Decommissioning and Reclamation (Andrew Fellinger, Savannah River National Laboratory (USA)); 3 - Fukushima Daiichi Decommissioning Progress (Adrian Simper, Nuclear Decommissioning Authority UK (United Kingdom)); 4 - Stable Storage of Secondary Waste from Contaminated Water Treatment in Fukushima Daiichi (Kosuke Sato, Tokyo Electric Power Company Holdings, Inc. (Japan))

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DOE OSTI2019

Panel Session 68: Small Business Procurement and Contracting Opportunities Within the DOE Complex and USACE FUSRAP Program

This panel session was held in the Panel Zone in the Exhibit Hall and focused on small business procurement and contracting opportunities within the DOE Complex as well as the US Army Corps of Engineers (USACE) National Formerly Utilized Sites Remedial Action Program (FUSRAP) projects. Each panelist provided a general overview of their respective organization, small business goals and the success in meeting those goals. They further discussed recently awarded procurement, upcoming opportunities and provided additional contact information for individual small business program managers within their organization. This panel was open to all WM participants, including Exhibitors. This panel session focused on small business procurement and contracting opportunities within the DOE Complex and the USACE FUSRAP Program. The DOE and USACE contract directly with small businesses and encourage prime contractors to obtain goods and services from small and disadvantaged businesses (Tier 2) and often tie these goals to contract award fee. Goods and services being considered for these procurements include decontamination and decommissioning, remediation, transportation and disposal of radioactive waste, health physics, emergency response planning and training, lab services, R and D products, waste treatment, maintenance, A/E and professional consulting. This panel complemented Session 21 and Session 48, US DOE Spring Quarterly Forum, and Session 88 on Doing Business with the DOE. Presentations proposed include: 1 - Small Business Procurement and Contracting Opportunities Within the DOE Complex (Aaron Deckard, US DOE (USA)); 2 - Small Business Procurement and Contracting Opportunities Within the DOE Complex (William Badger, AECOM (USA)); 3 - Engineering the Extraordinary (Lisa Tribuce-Leoung Tat, Bechtel National, Inc. (USA)); 4 - Fluor Small Business Procurement and Contracting Opportunities (Debra Sampson, Fluor Government Group (USA)); 5 - HII Technical Solutions Division Small Business Opportunities (Kela Lofton, HII Nuclear Inc. (USA)); 6 - U. S. Army Corps of Engineers Radiological Remediation Work (Nicki Fatherly, USACE (USA))

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