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At least 703 records · Page 39
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Understanding Uranium Binding Mechanisms and Speciation on Stainless Steel - 17462

Stainless steels have been exposed to nitric acid solutions containing natural uranium under similar conditions to those found in reprocessing plants. Solution and solid analysis of the stainless steel samples has been undertaken to describe the contamination phenomenon. General corrosion of the surface was seen to take place over the month-long experiments. Contamination has been shown on all exposed surfaces. Solution analysis indicates that an equilibrium state is reached after 2 weeks exposure to the 12 M HNO{sub 3}. Surface analysis through autoradiography and X-ray absorption near edge structure (XANES) indicate that multiple mechanisms of uptake may occur and consequently concentrated regions of contamination may be produced. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2017

Pile Fuel Cladding Silo Deflector Plate Removal Project Using Long Reach Tooling and Suspended Abrasive Water Jet Cutting System - 17541

The Pile Fuel Cladding Silo (PFCS) at the Sellafield decommissioning, reprocessing and nuclear waste management facility in the UK was constructed in the late 1940-s and used for the storage of Intermediate Level Waste (ILW) arising from the Windscale Piles, Calderhall and Chapelcross Reactors until the 1960's. The PFCS is 21 metres high and comprises 6 identical tall compartments with the waste added to the silo via a charge-hole at the top of each compartment; each compartment is split by a central dividing wall that runs from the base of the silo to a few metres below the underside of the silo roof. A deflector plate assembly (DPA) supported on two horizontal beams is located between the top of the dividing wall and the roof and allowed the waste to spread as it fell into the silo and prevent excessive mounding. Sellafield Ltd is currently engaged in an overall programme of works to remove the waste for storage within modern ILW containers. Prior to the commencement of these works the Deflector Plate Assemblies (DPA) and beam structures within the silos are required to be removed to allow access for the waste retrievals equipment through the side of the silo at the point where the DPAs are situated. James Fisher Nuclear (JFN) and Sellafield Ltd have developed a system using suspended water jet cutting equipment combined with custom long reach tooling, installed through penetrations made in the side of the silo, to enable the remote size reduction of the DPA and beams within each compartment. The use of a low pressure suspended water jet cutting system has enabled a single water/abrasive hose to be routed down a tool deployment pole and allowed long hose lengths up to 80 m from pump to nozzle enabling siting of equipment on a congested and difficult access site. The use of lightweight, yet high strength, carbon fibre poles within the tools has enabled the deployment of tool end effectors at distances of up to 6.5 m from the compartment walls using a relatively small 60 mm diameter pole. This maximised articulation of the pole within the 300 mm wall penetrations to reach all areas of the DPA. Capability of the operators to install the tooling and size reduce DPA and beam assemblies have been demonstrated through the training on a Full Scale Test Rig (FSTR). The FSTR was integral to the development of the equipment and techniques and to providing assurance to the client that the project was technically achievable given the tight building restraints and safety issues. (authors)

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

The Micro-Optical Ring Electrode: A Sensor for Multiple Actinide Ions Monitoring - 17586

We present on the development of the Micro-Optical Ring Electrode (MORE) for the analysis of trans-uranium elements in aqueous mixtures. The MORE is a photo-electrochemical device based on a ring microelectrode that uses the insulator interior to the ring as a light guide. This single device exploits the unique photophysical and electrochemical properties of multiple analytes present in mixtures to quantify them. Our study aims to develop a protocol for the analysis of ions of uranium, neptunium and plutonium, which are most relevant to the nuclear industry, especially in the areas of decommissioning and fuel reprocessing, where speedy and safe identification of radioactive contaminants is essential. The technique offers a number of advantages over traditional methods in the area of radiochemical analysis such as real time monitoring in-situ, reduced radiation dose impact to the analyst as a result of greatly reduced sample handling and preparation, improved analysis times allowing quick decision making, at a much reduced cost (e.g. compared to ICP-MS and Scintillation counting). Here we present results of our study on non-radioactive surrogates for the target elements and discuss the advantages the technique can deliver to the nuclear industry. (authors)

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

Drying Simulated Advanced Gas Reactor Fuel: Proving the Concept - 17637

Spent Nuclear Fuel (SNF) reprocessing is due to cease in the UK in 2018 and a new interim storage method may be required if it is found that pond storage is not viable beyond the 25 year period which is currently understood to be safe. Dry storage has been used elsewhere in the world for zircaloy and aluminium clad fuel and has been proposed for use in the UK, however since the UK's fuel is predominantly stainless steel (SS) clad, testing will need to be carried out to ensure that drying methods are suitable. This will require representative samples for macro scale drying tests. To date, little long term stored fuel has been examined. As a result, work has been carried out to try and ascertain whether the corrosion associated with long term storage leads to significant bound water and whether through wall cracks can be formed in intact lengths of SS tube. It was found that unlike an aluminium simulant, corroded SS showed no sign of water loss as a result of oxide dehydration during controlled drying by thermogravimetric analysis, and consequently bound water could be disregarded. It was also found that through wall cracks could be produced in an intact length of SS tube which would be suitable for macro scale drying tests. Some initial macro scale drying tests have been carried out using an alternative test piece with preformed cracks, and three distinct drying regimes to complete dryness have been detected. (authors)

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

Selective separation of Sr by potassium titanates from solution of nitrate - 15248

The radioactive waste treatment is planned in LWTF (Low-level radioactive Waste Treatment Facility, JAEA) for the low-level liquid wastes generated from Tokai-reprocessing facility [1]. The target liquid wastes mainly consist of highly concentrated sodium nitrate (4.7 M NaNO{sub 3}) containing low level radioactive {sup 137}Cs and {sup 90}Sr, while the conventional adsorbent (Titanic acid-PAN) has some problems in the adsorption performance and irradiation resistance. In this study, the adsorption properties of Sr{sup 2+} ions, leaching properties of Sr{sup 2+} and irradiation properties were clarified by using the novel adsorbent of potassium titanates (KT-1). Relatively large uptake percentage (99%) and maximum uptake capacity (0.37 mmol/g) were obtained for the KT-1 compared to that of Titanic acid-PAN. As for the breakthrough properties, KT-1 had the smoother breakthrough carve and larger 5% breakthrough capacity compared to Titanic acid-PAN. From the results of batch and column experiments, KT-1 was effective for the uptake of Sr{sup 2+} ions in the presence of highly concentrated Na{sup +} ions. The leached percentages of Sr{sup 2+} ions form calcined products of Sr-KT-1 in pure water and 0.1 M NaOH were less than the detection limit of ICP-AES, and those in 0.1 M HCl tended to markedly decrease with calcining temperature; the formation of SrTiO{sub 3} phase above 800 deg. deg. C was effective for the lowering of leachability. The irradiation stability of the adsorbent was examined by instrumental analyses and the evaluation for the uptake percentage of Sr{sup 2+} ions by batch method. The growth of surface crystal was observed by SEM, but no significant change in X-ray diffraction peak was observed, and Sr{sup 2+} uptake (%) was almost constant. The novel adsorbent of KT-1 had excellent characteristics for Sr selectivity, uptake capacity, leachability and irradiation resistance. KT-1 is thus effective for the treatment and disposal in LWTF. (authors)

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

Solutions for management of defective fuel assemblies - 15318

Defective fuel management is a major challenge for nuclear operators when the time comes for finding a long-term solution for their used (or spent) fuel assemblies. This question of defective fuel management can be raised in several cases: - For optimizing the remaining operating costs of shutdown reactors, the spent fuel pools must be emptied as soon as possible; a solution needs therefore to be found for evacuating all the spent fuel assemblies, defective fuel assemblies and in particular defective fuel rods included; - Some utilities are today facing a saturation of their reactor spent fuel pool; evacuating the defective fuel assemblies and extracted defective fuel rods allows optimizing the space in the pool. Whereas the solutions for managing intact used fuel assemblies are well-known, a specific technical solution must be implemented for managing the defective spent fuel assemblies as well. The purpose of this paper is to describe the existing solutions for the management of all types of defective fuel assemblies. Interim Dry Storage Solution Interim dry storage of defective spent fuel assemblies is being implemented for decades in the USA. Specific operations are performed for preparing the defective fuel assemblies for interim dry storage and they are then stored with intact used fuel assemblies in the dry storage cask. The advantage of such a solution is its short-term cost effectiveness. The paper presents the existing dry interim storage solutions for defective used fuel rods, the way they are implemented, the associated operational experience. Long-term Dry Storage Solution A long-term dry storage solution for defective fuel assemblies is being developed, compliant with direct disposal requirements. The solution is based on the encapsulation of each defective fuel rod into a fuel rod capsule and an associated specific dry storage licensing approach: this is the best available technology regarding long term storage safety requirements. The paper presents this unique capsule technology, the transport and storage casks that can be used for transportation of defective fuel rods to the storage facility, and the associated transport licensing approach currently used in Europe. Treatment Solution. The most definitive way to get rid of defective fuel assemblies is their treatment (or reprocessing) in a treatment plant. There is then no need to assess the behaviour of these specific used fuel assemblies as for an interim storage facility or a final repository. The advantage of such a solution is its sustainable approach. The paper presents the existing transport and treatment solutions, the way they are implemented, the associated operational experience. (authors)

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

Preliminary performance assessment for the Waste Management Area C at the Hanford site in southeast Washington - 15331

A performance assessment of Single-Shell Tank Waste Management Area C located at the U.S. Department of Energy's Hanford Site in southeastern Washington is being conducted to satisfy the requirements of the Hanford Federal Facility Agreement and Consent Order, as well as other Federal requirements and State-approved closure plans and permits. The Waste Management Area C performance assessment assesses the fate, transport, and impacts of radionuclides and hazardous chemicals within residual wastes left in tanks and ancillary equipment and facilities in their assumed closed configuration and the subsequent risks to humans into the far future. The part of the performance assessment focused on radiological impacts is being developed to meet the requirements for a closure authorization under U.S. Department of Energy Order 435.1 that include a waste incidental to reprocessing determination for residual wastes remaining in tanks, ancillary equipment, and facilities. An additional part of the performance assessment will evaluate human health and environmental impacts from hazardous chemical inventories in residual wastes remaining in Waste Management Area C tanks, ancillary equipment, and facilities needed to meet the requirements for permitted closure under the Resource Conservation and Recovery Act of 1976. Preliminary results of the performance assessment are evaluating the groundwater and air pathways to determine radiation exposure once radionuclides are released to the environment due to degradation of the engineered barriers. Per DOE Order 435.1, results of the preliminary analyses are being compared to the performance objectives during the 1,000-year post-closure period of compliance. Per the Order, performance assessment results are also being compared to the performance measures related to acute and chronic inadvertent intruder analyses and groundwater protection. (authors)

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

Operational plant lifetime assessment - UK national nuclear laboratory experience - 15366

Inspection and monitoring of the condition of key plant assets at the Sellafield Site is an important task in helping to enable the continued operation of the reprocessing and waste management facilities. This takes place within a range of environments, from non-active areas to highly active black cells where no man-access is possible. NNL staff work closely with the plant operators and other contractors in carrying out inspection and monitoring tasks, and have a particular expertise in the black cell environments where increased challenges exist. As plants age more work is required in order to underpin ongoing operations and understand the operational lifetime of key assets in order to achieve maximum throughput and utilisation. This paper discusses how expertise in a variety of different technical disciplines is combined in order to both understand and extend the operational life of the Highly Active (HA) evaporators at Sellafield, and a small number of further similar examples are also discussed. (authors)

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

Submersible ROVs pioneer decommissioning of a legacy nuclear fuel storage pond - 15368

Submersible Remote Handling Vehicles (ROVs) are being used in innovative ways at the Sellafield decommissioning, reprocessing and nuclear waste management facility in the UK in a legacy fuel storage pond in support of pond decommissioning. A wide range of sensors and tooling attached to submersible ROVs are being used to characterise the contents of the pond and perform a wide range of maintenance and material retrieval tasks. Visual surveys, sludge and liquor sampling and radiological surveys are just some of the characterisation projects undertaken with the help of the ROVs. The ROVs have also been deployed with tooling including manipulators which can retrieve debris from the pond floor. A variety of mini and work class ROVs are used in the ponds and these can access areas of the pond that the traditional technologies such as long reach tooling and manipulators from overhead cranes cannot reach safely, efficiently and cost effectively. The ROVs are controlled at a distance from the pond to minimise operator exposure to radiation and avoid potential radiological contamination. (authors)

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

Methods for radiological assessment of C-14 in air effluents - 15419

The purpose of this paper is to report on methods used for the assessment of -dose significant- C-14 being discharged into the air from nuclear facilities. GEL laboratories, LLC and GEL Engineering LLC, (GEL) undertook the task of developing new and improved methods to better identify the release of C-14 in the various chemical forms released from nuclear reactors and fuel reprocessing facilities. To completely understand the environmental releases and impact to the public a two-step approach was selected. The first step was to develop a sampling method that would enable an accurate assessment of the ratio of C-14 discharged as carbon dioxide compared to organic gases like methane. This is very important to dose modeling calculations to members of the public. Once this ratio was understood, the second step was to develop a method to track inorganic forms of C-14 with standard environmental air monitoring equipment and achieve a detection limit that was in line with NRC 10CFR Appendix I dose modeling requirements. GEL development efforts focused on using a solid sample cartridge that would fit into standard air sampling equipment used at nuclear plants. The cartridge and sorbent needed to be capable of running 7 days at 24 hours per day and not saturate with stable carbon dioxide. The process developed for analyzing the sorbent involves chemically removing the carbon dioxide and ultimately counting by liquid scintillation counting. The method yields a detection limit for C-14 of 0.02 Bq/m{sup 3} which provides a dose calculation of approximately 0.1 mSv/yr when applied to the child model. This paper will describe the methods and results used to determine field environmental monitoring for C-14 using GEL cartridges also known as -GELsorb C14-. (authors)

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

Advanced dismantling techniques applied in an industrial project for the dismantling of highly activated equipment - 15446

Commissioned in 1958 on the CEA nuclear site of Marcoule (France) and shut down in 1997, the former UP1 plant was dedicated to reprocess the spent fuel from G1, G2 and G3 GCR reactors. Since 1998, a wide program for cleaning-up and dismantling the plant has been developed in order to dismantle the process equipment, eliminate any radiological risk in all buildings concerned and remove all the resulting waste. Major technical challenges are related to the wide range of complex components and equipment in the facilities such as chemical dissolvers, extractor batteries, evaporators, tanks; and to a high risk of exposure, due to the presence of fission products. To address these challenges, CEA led R and D programs to develop advanced tools for complex dismantling projects: immersive virtual reality, laser cutting techniques, special remote-controlled arms combining dexterity and high resistance to irradiation The purpose of this paper is to provide an overview of a practical implementation and to show first results of these advanced techniques on a real case: the dismantling of dissolver's equipment from the dissolution workshop of UP1 plant. ONET Technologies has been chosen by CEA to design, build and operate the overall dismantling and waste treatment system for this project. ONET Technologies designed two main subsystems to meet all the requirements for this project: first, the 'Remotely-Operated Handling and Cutting Unit' and second the 'Waste Extraction and Conditioning Unit' and the combining innovative technologies. During the design phase, the 'Immersive Virtual Reality' has been used to qualify the kinematics in the final environment and the preliminary dismantling scenario. Following a first R and D program of CEA regarding the laser cutting technique applied to nuclear decommissioning, ONET Technologies and CEA realized elementary inactive tests based on a 6 kW Nd-Yag laser source to qualify the cutting performance according to material and thickness. Main objective was to meet the safety criteria and to check the physical characteristics of waste resulting from the laser cutting operation. The Remotely-Operated Handling and Cutting Unit was equipped with a MAESTRO manipulator, resulting from ten years of research and development cooperation between CEA and Cybernetix. This manipulator, a 6 axis hydraulic manipulator-arm specifically designed to work in harsh environment, entered in its industrial phase. ONET Technologies was responsible for the overall system integration, confinement and performance. A comprehensive qualification program in real conditions has been driven during 2014. The dismantling scenario was tested with a mockup at a full scale of dissolver equipment and confirmed at this stage the expected level of performance and safety. (authors)

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

Overcoming challenges in delivering 5 Avantech water treatment systems at Fukushima - 15544

The Great East Japan Earthquake has left the Fukushima Daiichi Nuclear Station with many challenges related to water treatment. This paper will provide information on three systems provided for the treatment of [1] accumulated water associated with the Circulated Water Cooling System, [2] sub-drain water removed from wells adjacent to the Unit 1-4 Reactor and Turbine Buildings, and [3] wash water from decontamination of vehicles and equipment. Specific information will be presented on work that was completed to overcome process, installation and logistical challenges associated with these water streams including the following: - Selection of technology and equipment for removal of specific radionuclides - Performance results of highly selective adsorbents used for removing radionuclides in the presence of varying concentrations of chlorides and other natural constituents - Description of process flowsheets and equipment configuration - Discussion of the post-accident work environment and the related work site restrictions Results show that the customized solutions will enable the utility to clean-up the wastewater for cooling, reuse and potential discharge. The technologies used for the three water treatment systems are directly applicable to the treatment of similarly challenging water streams at commercial nuclear power plants, fuel reprocessing plants and governmental nuclear facilities worldwide. The use of these technologies will aid in the lowering of radioactivity released to the environment, worker radiation exposure and the amount of secondary waste requiring long term storage and/ or disposal. (authors)

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

Zircaloy Cladding as Starting Precursor for Ceramic Waste-Forms - 15562

Zircaloy cladding of spent fuel is regular waste-stream related to the process of reprocessing. It remains in the apparatus after spent fuel dissolution. The most common technique of zircaloy cladding treatment is cementation and disposal in metallic containers. In view of the high cost of the cladding material (mainly zirconium), there have been a number of proposals for this material to be recovered and re-used. However, the residual contamination of the claddings following the dissolution of the fuel and the chemical stability of the zirconium militate against any proposal for a recycling process that might be economically reasonable. At the same time there is a lot of information available about synthesis of mineral-like materials based on zirconium for immobilization of long-lived radionuclides. The object of this work was to study experimentally electrochemical dissolution of fuel rod cladding for further obtaining starting precursor for the synthesis of ceramic-waste forms. The results of experiments on electrochemical dissolution of zirconium and zirconium alloy (fuel rod cladding) in nitric acid solutions are presented. The peak dissolution rate of specimens is shown to be observed in dilute nitric acid at elevated temperatures. The rate of dissolution rises with temperature, reaching ∼40-70 mg/cm{sup 2}.h at 90 deg. C. The current yield at this point is 0.6-0.7 g/A.h. The results of analysis of the surface oxide film on specimens and residues formed during dissolution are also presented. An electrochemical dissolution of zircaloy claddings is the common technique for preparation of zirconium solutions to be used for ceramic waste form synthesis. (authors)

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

Preparation of Waste Fingerprints for the Miscellaneous Beta Gamma Waste Feeds to the Box Encapsulation Plant at Sellafield - 16080

There are a number of legacy pond and silo storage facilities on Sellafield Site that contain a broad range of ILW arising from the early Windscale Pile and Magnox Fuel Reprocessing Programmes. The waste comprises of solid Miscellaneous Beta Gamma Waste (MBGW) from operations (including filters, containerised waste, thermocouple wire, support struts, pumps, soft wastes) and associated sludges that have arisen from the corrosion of the fuel, Magnox Swarf cladding and the MBGW. It is intended to retrieve and package this waste to meet the requirements of interim storage and final disposal in the Box Encapsulation Plant (BEP) at Sellafield. As part of this process each package must be assigned a realistic and justifiable radionuclide and material inventory. These inventories can either be assigned by radiometric measurements including the associated errors or developed by fundamental data underpinning these assessments and a 'chain of custody' approach of knowing where the waste has been and how it has evolved. The subject of this paper is concerned with the development of the waste fingerprints for these legacy waste streams. As the wastes have arisen over a ∼70 year period there have been an assortment of different working practices and developments in equipment design that have led to significant variations in the waste condition and its composition. During storage the waste items have become mixed, corroded and sludge deposits have arisen. Because of this, and that early consignment records are either incomplete or missing, the task of assigning material and radionuclide fingerprints has been very challenging. In addition, to comply with regulatory requirements the inventory assignment has to be sufficiently robust to meet the processing envelope of the encapsulation process i.e. to allow controls to be made on the amounts of reactive materials (e.g. aluminium, Magnox and uranium) that can be processed within each product and allow identification of excluded items from being consigned from the donor plants. Also, as the waste comprises actinides, fission and activation products it was noted that whilst the UK standard inventory code FISPIN10 predicts the actinides and fission products well, there was a significant discrepancy in the prediction of activation products when compared with those predicted by the specialised fusion code FISPACT. It should be noted that the use of FISPACT to predict actinides and fission products is not recommended. To address these issues, - position statements were produced for each of the waste streams that would be consigned to BEP; - gaps within the data were assessed and where possible inactive trial work was identified to address it; - material and radionuclide fingerprint justification documents were developed from the position statements based upon available data, the experimental work and justified assumptions; - where FISPIN runs were required, these were run with a modified activation data library based on the European Activation File 2010 (EAF2010) to model the activation products. This paper summarises the BEP process, discusses examples of how the material and radionuclide waste stream fingerprints have been justified by revisiting the sample data, original consignment records, engineering drawings and recent remotely operated vehicle (ROV) pond surveys and making reasoned assumptions based on operational practices and non-active experimental work. Incorporation of the EAF2010 library within the FISPIN code has offered significant benefits to BEP and indeed other plants processing mixed fission and activated wastes as it allows comparable predictions of activation products compared with that made by FISPACT and for the plant to maintain one data processing stream rather than two if it had to utilise both modelling codes. This work will underpin the inventory submission in support of the Interim Letter of Compliance (ILoC) to Radioactive Waste Management Limited (RWM). The process of examining the original records, plant operation schedules and engineering drawings has enabled removal of some of the pessimisms that have been made in previous assessments. (authors)

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

The INFODEM Information Centre at CEA Marcoule: A Unique Information Tool in Europe About Nuclear Decommissioning - 16217

In September 2014, the French Alternative Energies and Atomic Energy Commission (CEA) inaugurated a unique information center focused on decommissioning of its nuclear facilities. This facility, named INFODEM information center, is inside the main building of the G1 nuclear reactor at Marcoule, a nuclear site in the south of France with world-recognized expertise in decommissioning. Constructed in the 1950's, the G1 reactor was the first industrial-scale reactor for plutonium production in France; it was also used to produce electricity. Operations ceased in 1968, and it was progressively dismantled (through removal of fuel and main components) and placed in monitoring status. In the 1980s, mockups of fuel cycle process equipment were built under a temporary mission to support development activities for the La Hague reprocessing plant. When that ended, further dismantling occurred, and during 1996 through 2010, the site established itself as a reference example for worldwide nuclear dismantling activities. In 2012, the project to establish an information center at the reactor began, targeting three main audiences: the public, agency personnel and politicians, and technical professionals. Considerable thought went into the facility design, to address the interests and information needs of these three audiences. The INFODEM concept extended over 800 square meters (about 960 square yards), to host an informative tour addressing several major themes: CEA's D and D strategy, programs, skills, tools and methodologies and 'everyday life' on decommissioning projects. Illustrating its unique value and considerable success, INFODEM received the Public Information Prize from the French Nuclear Energy Society (SFEN) in June 2015. (authors)

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

Preferred Concepts for Disposal of the UK Inventory of Depleted, Natural and Low Enriched Uranium - 16382

Depleted, natural and low enriched uranium (DNLEU) forms a significant quantity of the UK inventory of higher activity radioactive materials. The great majority comprises uranic materials produced from uranium enrichment operations and from reprocessing of spent nuclear fuels. The most recent estimate of the total quantity (stocks plus arisings) of UK-owned DNLEU from civil nuclear operations is 170,000 tU. Current UK Government strategy is for DNLEU to be stored indefinitely in existing or new facilities, as required. However, Government strategy for long-term management of uranic materials is under review and a change could potentially cause this material to be reclassified as a waste. An Integrated Project Team (IPT) was formed in 2012 with the objective of identifying and addressing the issues associated with the potential need to dispose of DNLEU, in order to inform wider decision-making on the long-term management of UK DNLEU. Key technical issues considered by the IPT relate to the following: 1. The need for an improved understanding of the UK DNLEU inventory. 2. The identification of preferred chemical and physical form(s), and preferred packaging and disposal concepts for DNLEU - the main subject of this paper. 3. The approach to safety assessment of DNLEU disposal. This paper summarises the disposal concepts under consideration by the IPT, and the rationale for identification of preferred concepts for disposal of DNLEU inventory components. The conclusions will be of interest to other programmes and countries with a requirement to manage large quantities of similar materials. (authors)

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

Application of Modular Design in Abatement of Fuel Element Debris (FED) Liquid Waste Stream - 16635

This paper provides a description of the processes applied within the Aqueous Discharge Abatement Plant (ADAP) at Bradwell nuclear power plant in the UK. This incorporates the Fuel Element Debris (FED) abatement plant and the Modular Active Effluent Treatment Plant (MAETP). These plants are housed in adjacent, but separate, structures and are inter-linked and share common services and control systems. The FED abatement plant, which has a shorter mission than MAETP, can be removed for decommissioning or redeployment without any impact to MAETP operation. ADAP treats aqueous waste streams generated during the ongoing decontamination and decommissioning (D and D) operations for release from the Bradwell Power Station. The plant is designed to receive and treat two major classes of liquids: 1. Slightly contaminated water from various D and D sources and collected rainwater from the power station, referred to as Aqueous Effluent (AE). 2. Dissolved fuel element debris solution. This comprises of contaminated magnesium alloy (generated during the stripping of external structures from spent Magnox fuel elements prior to shipment for reprocessing) which has been dissolved in nitric acid. All received wastes are treated to remove heavy metals and radioactive constituents prior to isolation, sampling, verification, and ultimately release to the Blackwater Estuary. The treatment consists of multiple processes including precipitation, filtration, adsorption and ion exchange carefully designed to remove targeted radionuclides and heavy metals. (authors)

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

Use of advanced optimization algorithms for the design of critical experiments

The need for validating molybdenum and rhodium cross sections in the thermal energy range has arisen for many years for reprocessing activities and burnup credit calculations. An Integral Experimental Request (IER) sheet was submitted by IRSN to the NCSP (Nuclear Criticality Safety Program) in 2015. A preliminary design was therefore proposed for review. In order to define experiments that meet the requirements and that exhibit optimized sensitivities of k{sub eff} to the capture cross sections of {sup 95}Mo and {sup 103}Rh, advanced algorithms of the PROMETHEE workbench, developed at IRSN, were used. The methodology as well as the main results are presented in this paper, which focuses only on the molybdenum experiments design. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗