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At least 433 records · Page 24

Anthropogenic Pathways for Modeling and Managing Future Arctic Fires

Wildland fires, including extreme fire events and seasons, are becoming more common in the boreal and Arctic regions due to climate change. Current climate modeling approaches do not include country- or region-specific socioeconomic pathways that specifically address the drivers of and potential mitigation techniques for wildland fires. Forest management, energy extraction, and tourism, together with firefighting capacity and readiness as well as fuels treatment, can have a significant impact on future wildland fire risks and impacts. To assess the impacts of anthropogenic factors and to align with previous work done on shared socioeconomic pathways (SSPs), climate pathways for future wildfires up to 2050 were created for the states that compose the original Arctic Council countries: Canada, the United States, the Kingdom of Denmark, Iceland, Sweden, Norway, and Finland as well as the Russian Federation (with whom the other seven countries withdrew participation from in May 2022 due to the invasion and ongoing war in Ukraine). High and low fire activity and risk pathways for all states comprising the Arctic were made, with expert ‘best guess’ pathways for each state created separately to represent the middle of road. The low activity and low fire risk pathways, named “We Got This”, assume active fire suppression via citizenry participation and official land management, efficient and extensive fuel treatments, and consistent and active wildland firefighting for each new ignition. The high activity and high fire risk pathways, named “Let It Burn”, assume nearly the opposite, due to lack of government and community response and no action on climate change drivers that increase wildland fire risk. The ‘best guess’ pathway, named “The Fire Will Come”, indicates that some countries are currently on the pathway for less fire compared to other Arctic and Boreal states but not a ‘no-fire’ future. For example, in the Nordic countries, human ignition sources from tourism, timber and energy extraction, summer cottages, and expanding wildland-urban intermix due to exurban growth may increase. In North America, these same risks will apply but also may see an expansion of agriculture that increases the likelihood of open burning in croplands. Drier fuel condition and extreme heat events due to climate change create favorable conditions for extreme wildfires from any ignition source. Throughout the Arctic and boreal lightning is expected to increase, increasing the risk of tundra fires in addition to forest fires in hard-to-reach locations that are more difficult to coordinate and execute wildland firefighting. To move the future Arctic fire SSPs forward, several short-term and long-term actions must be completed. Certain data needs are required, like a harmonized pan-Arctic and pan-boreal fuels geospatial product, while also a need to refine and socialize current definitions of fire seasons and fire management – including developing an open-source system to track and share innovation, mitigation, and adaptation strategies across Arctic states.

Arctic↗

In-Field Performance Testing of the Fork Detector for Quantitative Spent Fuel Verification

Expanding spent fuel dry storage activities worldwide are increasing demands on safeguards authorities that perform inspections. The European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) require measurements to verify declarations when spent fuel is transferred to difficult-to-access locations, such as dry storage casks and the repositories planned in Finland and Sweden. EURATOM makes routine use of the Fork detector to obtain gross gamma and total neutron measurements during spent fuel inspections. Data analysis is performed by modules in the integrated Review and Analysis Program (iRAP) software, developed jointly by EURATOM and the IAEA. Under the framework of the US Department of Energy–EURATOM cooperation agreement, a module for automated Fork detector data analysis has been developed by Oak Ridge National Laboratory (ORNL) using the ORIGEN code from the SCALE code system and implemented in iRAP. EURATOM and ORNL recently performed measurements on 30 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel (Clab), operated by the Swedish Nuclear Fuel and Waste Management Company (SKB). The measured assemblies represent a broad range of fuel characteristics. Neutron count rates for 15 measured pressurized water reactor assemblies are predicted with an average relative standard deviation of 4.6%, and gamma signals are predicted on average within 2.6% of the measurement. The 15 measured boiling water reactor assemblies exhibit slightly larger deviations of 5.2% for the gamma signals and 5.7% for the neutron count rates, compared to measurements. These findings suggest that with improved analysis of the measurement data, existing instruments can provide increased verification of operator declarations of the spent fuel and thereby also provide greater ability to confirm integrity of an assembly. These results support the application of the Fork detector as a fully quantitative spent fuel verification technique.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

MODELING OF THE GROUNDWATER TRANSPORT AROUND A DEEP BOREHOLE NUCLEAR WASTE REPOSITORY

The concept of disposal of high-level nuclear waste in deep boreholes drilled into crystalline bedrock is gaining renewed interest and consideration as a viable mined repository alternative. A large amount of work on conceptual borehole design and preliminary performance assessment has been performed by researchers at MIT, Sandia National Laboratories, SKB (Sweden), and others. Much of this work relied on analytical derivations or, in a few cases, on weakly coupled models of heat, water, and radionuclide transport in the rock. Detailed numerical models are necessary to account for the large heterogeneity of properties (e.g., permeability and salinity vs. depth, diffusion coefficients, etc.) that would be observed at potential borehole disposal sites. A derivation of the FALCON code (Fracturing And Liquid CONvection) was used for the thermal-hydrologic modeling. This code solves the transport equations in porous media in a fully coupled way. The application leverages the flexibility and strengths of the MOOSE framework, developed by Idaho National Laboratory. The current version simulates heat, fluid, and chemical species transport in a fully coupled way allowing the rigorous evaluation of candidate repository site performance. This paper mostly focuses on the modeling of a deep borehole repository under realistic conditions, including modeling of a finite array of boreholes surrounded by undisturbed rock. The decay heat generated by the canisters diffuses into the host rock. Water heating can potentially lead to convection on the scale of thousands of years after the emplacement of the fuel. This convection is tightly coupled to the transport of the dissolved salt, which can suppress convection and reduce the release of the radioactive materials to the aquifer. The purpose of this work has been to evaluate the importance of the borehole array spacing and find the conditions under which convective transport can be ruled out as a radionuclide transport mechanism. Preliminary results show that modeling of the borehole array, including the surrounding rock, predicts convective flow in the system with physical velocities of the order of 10-5 km/yr over 105 years. This results in an escape length on the order of kilometers, which is comparable to the repository depth. However, a correct account of the salinity effects reduces convection velocity and escape length of the radionuclides from the repository.

Modeling↗

A qualitative analysis of the neutron population in fresh and spent fuel assemblies during simulated interrogation using the differential die-away technique

In this study, Monte Carlo simulations were performed for the differential die-away (DDA) technique to analyse the time-dependent behaviour of the neutron population in fresh and spent nuclear fuel assemblies as part of the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Simulations were performed to investigate both a possibly portable as well as a permanent DDA instrument. Taking advantage of a custom made modification to the MCNPX code, the variation in the neutron population, simultaneously in time and space, was examined. The motivation for this research was to improve the design of the DDA instrument, as it is be ing considered for possible deployment at the Central Storage of Spent Nuclear Fuel and Encapsulation Plant in Sweden (Clab), as well as to assist in the interpretation of the both simulated and measured signals.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

The Science of the Deep Underground Neutrino Experiment (DUNE)

This 4-minute animation shows how the international Deep Underground Neutrino Experiment will help scientists understand how the universe works. DUNE will use a huge particle detector a mile underground to embark on a mission with three major science goals: 1.) Study an intense, 1,300-kilometer-long neutrino beam to discover what happened after the big bang: Are neutrinos the reason the universe is made of matter? 2.) Use 70,000 tons of liquid argon to look for proton decay and move closer to realizing Einstein’s dream of a unified theory of matter and energy. 3.) Catch neutrinos from a supernova to watch the formation of neutron stars and black holes in real time. About 1,000 scientists from 160 institutions in 30 countries are working on the Deep Underground Neutrino Experiment, hosted at the Department of Energy’s Fermi National Accelerator Laboratory and South Dakota’s Sanford Underground Research Facility. DUNE collaborators come from institutions in Armenia, Brazil, Bulgaria, Canada, Chile, China, Colombia, Czech Republic, Finland, France, Greece, India, Iran, Italy, Japan, Madagascar, Mexico, Netherlands, Peru, Poland, Romania, Russia, South Korea, Spain, Sweden, Switzerland, Turkey, Ukraine, United Kingdom, and the United States of America.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Development of LWR Fuels with Enhanced Accident Tolerance ATF Feasibility Analysis Report Deliverable for the Westinghouse Accident Tolerant Fuel Program

The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program utilizes chromium (Cr) coated zirconium alloy cladding with uranium silicide (U 3 Si 2 ) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. The lead test assembly (LTA) program will use both SiC/SiC composites and Cr coated zirconium alloy claddings with the high density/high thermal conductivity U 3 Si 2 pellet which will begin in 2022. Over the past several years, Westinghouse has tested the Cr coated zirconium (Zr) and silicon carbide (SiC) claddings in autoclaves and in the Massachusetts Institute of Technology (MIT) reactor and the U 3 Si 2 pellets have been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA have been carried out to determine the time and temperature limits for the SiC and Cr coated zirconium claddings. Fuel rod and assembly design in preparation for the LTR and LTA programs is underway as well as licensing efforts with the Nuclear Regulatory Commission (NRC). Finally, accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. Cr coated zirconium alloy claddings showed low corrosion rates in both autoclave and in-reactor tests and will be fabricated using commercial processes for the first LTRs for insertion in 2019. Current autoclave results for SiC composite claddings indicate that a corrosion rate of fewer than 2 micrometers per year which meets corrosion requirements may be achieved under normal operating conditions. Modular Accident Analysis Program 5 (MAAP5) calculations indicate that solid fission products can be contained within SiC cladding for up to two hours longer than current Zr based cladding in a station blackout scenario. This additional time is due to the much higher temperature capability of SiC (~2000°C). These two hours can be used to implement additional response options instituted through the FLEX program by reactor operators. While the Cr coated Zr alloy option offers more modest ATF gains (~200°C higher than bare Zr) in similar situations, the coatings do delay the runaway oxidation encountered by uncoated zirconium cladding. Both ATF cladding options also reduce hydrogen production which dramatically reduces primary system and containment pressure and the risk of fission product release beyond containment in the unlikely event of an accident. The lower pressure in the system allows more time to feed cooling water to the core, resulting in the avoidance of fuel melting. The coping time is extended indefinitely as long as the water flow continues. SiC exhibits the highest accident tolerance of any cladding and significant advances in understanding the design, manufacture and accident behavior of SiC cladding have been made. Department of Energy support is needed to continue the on-going experimental work addressing the interaction between U 3 Si 2 and SiC at temperatures >1200°C, the effect of SiC manufacturing methods and manufacturing variability on physical properties, and the ability to predict the mechanical properties of SiC composites based on the composite design. Additionally, experimental work on methods to rapidly fabricate SiC composite structures with high density and reduce the fabrication price of SiC fibers while maintaining a high level of performance is needed. Minimal (<1%) swelling of U 3 Si 2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U 3 Si 2 fuel in ATR to determine these properties at 40 to 50 MWd/kgU are underway. These experiments will provide the data needed to license fuel rod design codes. Research is continuing on production methods for making U 3 Si 2 from UF 6 without going through the U metal step. Oxidation tests of fuel pellets and powder in steam and synthetic air indicate that U 3 Si 2 has a lower oxidation reaction initiation temperature than UO 2 . Options to reduce the U 3 Si 2 oxidation rate are being explored. A hybrid fuel design using pellets with the weight ratio of 30%/70% U 3 Si 2 /UN is currently undergoing testing in the ATR. This approach is being explored to further increase the positive economics of the U 3 Si 2 fuel. In addition to the work supported by the USDOE, research and testing activities are being carried on in a world-wide effort funded by many countries such as Sweden, United Kingdom, Belgium, Netherlands, Spain, Germany, Japan and France. This work is being facilitated through the Westinghouse led Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) program. Annual meetings are organized by Westinghouse as a venue for presentation of this work and to provide for the cross-fertilization of ideas among the many outstanding researchers in the ATF area. Since SiC, coated cladding and high density fuel options are not currently used in the nuclear industry, support from the government and industry members is needed to further the significant effort of setting new standards. The same is true for the Nuclear Regulatory Commission (NRC) which must review and approve the commercial use of these new fuels since all current regulations are oriented toward Zr/ UO 2 fuel. There are no showstoppers for accident tolerant fuel technologies when considering supply chain, fabrication, enrichment suppliers, production, and licensing. Due to the cost savings from higher density fuel and the accident tolerance of coated claddings, Exelon has identified potential commercial plants to host the LTRs and LTAs. Planning is underway to implement this schedule.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Highlights of the Quad Multilateral Nuclear Arms Control Research Initiative LETTERPRESS Simulation

The US, UK, Sweden and Norway have formed the Quad Nuclear Verification Partnership and have been working together as the ‘Quad’ to ensure non-nuclear weapon states (NNWS) and nuclear weapon states (NWS) participate to overcome the challenge of nuclear arms control verification. The Quad aims to complement and inform the work of other multinational initiatives such as the International Partnership for Nuclear Disarmament Verification (IPNDV). The initial engagement of the Quad was the development and execution of a role-playing, in-field verification simulation, named LETTERPRESS. A benefit of role-playing events is the capacity they have to allow researchers to engage critically with a challenge from multiple perspectives. This paper will introduce LETTERPRESS and the process the Quad followed to develop and execute the multilateral simulation. The paper will highlight the in-play objectives of the verification regime, the site hosting the on-site inspection, and the verification technologies and their roles with respect to the objectives. The paper will include lessons learned from LETTERPRESS, focusing on the in-play objectives and technologies and their contribution to a successful monitoring result. The paper will conclude by highlighting the in-play success of a multilateral verification body and the ability of both NNWS and NWS to contribute to the verification process within LETTERPRESS.

arms control, Quad, LETTERPRESS↗

Geothermal Energy and Resilience in Arctic Countries

The eight Arctic countries have diverse energy systems but can be split into two distinct groups based on energy characteristics. The first group includes countries which are heavily gridconnected (Iceland, Norway, Sweden, and Finland); the second group includes countries with some grids as well as an abundance of remote microgrids, particularly in their more northern regions (Canada, Russia, the United States [Alaska], and Greenland). The primary energy source for both heat and power in remote communities is almost exclusively diesel. Geothermal energy is currently used in all eight Arctic countries, providing heat and sometimes electricity at utility scales and at the microgrid scale. However, the availability of geothermal resources is poorly defined in Arctic countries. We reframe geothermal heat and power as integrated energy systems, asking the question: are integrated geothermal energy systems - where available and economic - resilient solutions for communities in Arctic countries? Resilience attributes of integrated geothermal energy systems are identified, with a focus on microgrids and small-scale applications.

arctic energy↗

Integrated transient modeling of gas turbine and sCO2 power cycle for exhaust heat recovery application

For distributed generation installations, the transient response of the power generation system is critical to meet power quality (voltage and frequency) requirements. Transient models for gas turbine and supercritical carbon dioxide (sCO2) power cycle were integrated for an exhaust heat recovery bottoming cycle system for a distributed generation application. The transient model for the SGT-750 Siemens gas turbine was a ‘black-box’ functional mockup interface (FMI) model developed by Siemens Industrial Turbomachinery in $Finsp\mathring{a}ng$, Sweden. The SGT-750 is a twin-shaft gas turbine that produces 40 MW electricity with an efficiency of about 40% at ISO conditions. At 100% gas turbine throttle (load), the SGT-750 has average exhaust conditions of 114.6 kg/s and 469.8 °C.

20 FOSSIL-FUELED POWER PLANTS↗

Global Energy Policy to Achieve the United States’ Nationally Determined Contributions: Potential Pathway

In 2021, the United States declared its Nationally Determined Contribution (NDC) in preparation for the COP26 climate summit, setting goals to achieve zero emission electricity by 2035, and reducing net emissions by 50–52% compared to 2005 levels. The current U.S. energy policies were found to not be adequate to meet these goals, thus it is necessary to draft a pathway to achieving these goals through policy changes. This study focused on foreign energy policies to analyze the most effective policies in a host of sectors. By identifying the most ambitious and effective policies from around the world in each sector, this study was able to suggest a policy pathway that can potentially achieve both of goals set in the NDC. In guiding the development of the proposed policy pathway, this study analyzed policies from Australia, Canada, California, China, Denmark, France, Germany, Japan, Russia, South Korea, Sweden, the United Kingdom, and the United States. This study found that the most drastic CO2 reductions came from low-carbon electricity policy, but contributions were also made from carbon capture, reforestation, hydrogen, energy efficiency, and climate smart agriculture.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗

Inspection of a LOCA test rod at the Halden reactor project using gamma emission tomography

The LOCA test series IFA-650 conducted at the OECD Halden Reactor Project (HRP) has provided unique data on the performance of fuel rods during LOCA transients. One focus of the current investigations is the performance of the fuel in the ballooning stage of the LOCA transient. In this stage, relocation of fuel material is a possibility, in which case pellet fragments fall down to fill the void introduced by the increased volume of the ballooned cladding. This increases the heat load in that region, further promoting corrosion of the cladding. A special concern in the case of high-burnup fuels is the increasing number of small fuel fragments, which may be expected to cause a higher packing fraction in the ballooned region. In this work, a novel technique is presented for assessing the average density of the fuel material in the ballooned region of LOCA test rods. The investigation is based on non-destructive gamma emission tomography measurements, using the dedicated instrument recently developed at the HRP in collaboration with Westinghouse (Sweden) and Uppsala University. In this approach, the gamma radiation field surrounding the test rod has been measured with a narrowly collimated HPGe detector. Tomographic reconstruction of the data was performed, providing the radial gamma-ray source distribution within the measured volume, which reveals the fuel fragment distribution. From this, the density of the fuel in the measured volume (i.e., the packing fraction) may be calculated. The technique has been used to investigate a LOCA test rod of the Halden Reactor Project LOCA series. The LOCA experiment was carried out about one month prior to the gamma tomography examination. The results show that the distribution of the relocated fuel can be imaged using gamma rays from fission products. The reconstructions of the 662 keV rays from {sup 137}Cs and 1596 keV from {sup 140}Ba/La are demonstrated. In addition, the peaks of activation products offer valuable information on the location of the test rig structures, which may be utilized in a quantitative tomographic reconstruction to assess the spatially resolved packing fraction. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Treatment of Radioactive Sludge From Waste Water Treatment for Weight and Volume Reduction with FreezeTec-Technique - 17010

At the Studsvik site in Sweden, Studsvik Nuclear AB operates a water treatment facility that treat all the waste water from the nuclear facilities at the site. The facility contains several pools for waste water treatment. The standard procedure for treating the water consist of adding FeCl{sub 3} to precipitate any particulates. Historically this waste has been mixed with concrete and grouted into barrels which will increase the volume needed for disposal in final repository. The sludge that is generated from the water treatment facility has been subject to treatment with the Freeze-Tec method where the sludge is freeze dewatered and after thawing generating a solid phase with high dry substance and a clear water phase. The radioactive particles/nuclides tend to stay in the solid phase while the water phase has very low concentration of radioactive particles. 99.9% of the Co-60 and Cs-137 activity end up in the solids and a dry solid with smaller volume and weight is obtained after treatment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

End of the (Fuel) Cycle: What can South Australia learn from the world? - 17047

This project consists of a study of five countries' nuclear waste policy histories, a review of the presence and role of bureaucracy, how these factors have influenced their citizens' attitudes towards nuclear waste and progress made towards the planning, siting and construction of nuclear waste repositories. A survey was sent to a selected stratified sub-group of the community questioning their attitude towards the location of a potential nuclear waste repository in South Australia and what the most influential factor responsible for their opinion was. France is currently in the process of converting its underground nuclear waste research facility in Bure into the country's first high-level nuclear waste repository. Sweden is also constructing an underground high level nuclear waste repository, after decades of planning, community engagement and candidate site selection processes. The Nuclear Energy Agency estimates that Switzerland will construct a low and intermediate level nuclear waste repository by 2035 and a high level nuclear waste repository by 2045. Finland will be the first country in the world to successfully construct and operate a high-level nuclear waste repository and it is expected to be operational in 2023. The project summarises the most influential factors affecting the success of five case-study countries in managing their nuclear and radioactive waste and apply them to an Australian context. It also highlights the outcome of the surveys and connects the result to those previously conducted in the five case-study countries. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Westinghouse Quiver - A simple and safe system for the complete handling and storing of failed BWR and PWR fuel rods and fragments - 17049

Background: Extracted leaking fuel rods from fuel assemblies have been stored in the spent fuel pools at the Nuclear Power Plants for many years. Sending those fuel rods to an intermediate storage has been very difficult due to the uncertainty of the technical requirements. The Westinghouse Quiver is a container that can be used from the first moment when the leaking fuel rod is extracted from the Fuel Assembly. The leaking fuel rod can be placed in the Westinghouse Quiver. When the Quiver is filled, it can be sealed and dried and then treated as a regular fuel assembly during intermediate storage and final repository. Purpose of System: The purpose of the system is to store leaking fuel rods, broken rods and fuel rod fragments leak tight in the fuel pool and after vacuum drying and permanent sealing at an intermediate storage facility. The Quiver is designed to be fully compatible with a regular fuel assembly. That means all handling of the Quiver in a Nuclear Power Plant, during transportation or storage is also identical to that of a regular fuel assembly. Fulfilling these criteria together with functional testing, the Quiver is considered acceptable for long time storage of leaking fuel rods, both in dry or wet condition. The Quiver has been qualified in Switzerland for use in the spent fuel pool, intermediate storage, and for a final repository. The system will be implemented as well in Sweden during 2016-2019. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Development of an Integrated System for the Long-term Preservation of Records, Knowledge and Memory relating to Radioactive Waste Disposal Facilities - 17099

The Nuclear Energy Agency (NEA) has co-ordinated an international initiative to develop a common approach to the preservation of records, knowledge and long-term memory (RK and M), focused on radioactive waste disposals. The work programme has run in two phases, starting in 2011, with the current phase of the project due to be completed in 2018. Preparations for final reporting are now underway. The detailed implementation of an approach to RK and M preservation for any disposal facility will always remain a national decision. However, the NEA project has produced a 'menu' of tools and techniques that can be accessed and adapted to suit national needs. The common source of these menu components will help to ensure that the memory of the repository is kept alive, that messages to future populations can be clearly understood, and that evidence derived from the disposal environment can be properly interpreted. This paper describes the vision for implementation of the RK and M initiative within national programmes. Practical testing of some components of the menu of tools and techniques is currently underway in France, Sweden and USA, with encouraging results. Other disposal facilities, both under development or in planning, are invited to 'road test' the products and provide feedback. Refinement and maintenance of the tools remains important, and the NEA is committed to ensuring the continued accessibility and availability of the project outputs into the future. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Pilot Plant Test of Removal of Co-60 and Cs-137 from the Evaporator Concentrates of Forsmark unit 3 - 17158

A pilot plant test has been conducted to treat evaporator concentrates at Forsmark BWR NPP unit 3 in Sweden using Fortum's selective ion exchanger technology. The concentrate contains an array of radionuclides of which C-60 and Cs-137 are radiologically the most important ones. Concentrate also contained a high salt content ∼ 20 g/l and high organic carbon loading ∼ 1 g/l The pilot system consisted of pre-treatment (mechanical filters, activated carbon, UV oxidation) and Cs and Co selective ion exchangers. CsTreat ion exchanger was very effective in removing Cs from the evaporator concentrate: {sup 137}Cs activity was lowered from the initial value 280 MBq/m{sup 3} to below detection limit 550 Bq/m{sup 3}. Removing Co was more complicated: results suggest that Co is bound to organic chelating compounds, thus fairly compete oxidation of organic carbon was required in order to achieve good ion exchange results. Finally, after total organic carbon (TOC) was treated to below 50 ppm, {sup 60}Co activity was lowered from the initial value of 57 MBq/m{sup 3} to 2.7 kBq/m{sup 3}. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Strategic Aspects on Waste Management in Decommissioning - 17224

A team composed of the facility owner Oskarshamn Kraftgrupp AB (OKG) specialists and external experts was appointed to develop a basis for decision on an overall strategy for the management of the material and waste arising from the decommissioning of two BWR NPPs at the Oskarshamn site in Sweden. To be able to provide a good basis for decision the full waste management chain from generation to disposition had to be assessed, categorized, quantified and analysed with regards to costs, environmental impact and risks. A systematic approach was applied taking benefit of the existing decommissioning studies, well recognized decommissioning concepts and the combined knowledge and experience in the project team. In total four different waste management concepts were compared individually and in combinations. This paper discusses the important aspects in the work to develop a state of the art waste management in decommissioning. It also contains reflections and gives certain recommendations for decommissioning planning in general as well as an overview of some specific findings, results and recommendations from the actual project at Oskarshamn. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗