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Optimised norm waste management taking benefit of 60 years of nuclear waste experience - 15357
Studsvik has more than 60 years of experience of managing radioactive waste from both our own operations as well from international and other domestic waste generators. During last 5 years our services to generators of NORM waste has increased significantly. Building upon the experience of managing nuclear waste the management of NORM waste can be optimised. An optimized NORM waste management strategy should be based upon a graded approach taking into account the actual conditions (technical as well as regulatory). In line with how work is carried out in the nuclear industry it is recommended that material should be treated for reuse or recycling when appropriate as opposed to conditioning for disposal. This paper covers the principals and practices that should be adopted when considering NORM waste management from initial characterisation to the safe disposal or clearance of the material for reuse or recycling. (authors)
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Applying fast calorimetry analysis on spent fuel calorimeter - 15620
Swedish Nuclear Fuel and Waste Management Company (SKB) is responsible for the back-end of the Swedish nuclear industry. This includes transport of spent fuel by boat from the power plants to the Central Interim Storage Facility for Spent Fuel (Clab), and in the future, encapsulation of the spent fuel and further transportation to a geological repository. In order to ensure safety for the planned geological repository, the decay heat has to be well known. For this purpose the SKB has performed calorimetric measurements on over 200 pressurized water reactor (PWR) and boiling water reactor (BWR) assemblies since 2003. The results from earlier analyses have been used to improve computer codes predicting heat as well as to acquire deeper knowledge about the fuel inventory. The work complements research being done as part of the Next Generation of Safeguards Initiative - Spent Fuel project, whose purpose is to develop and test nondestructive assay (NDA) technologies to improve safeguards measurements of spent nuclear fuel. The project is working to accomplish multiple goals, one of which is to determine the heat content of spent fuel assemblies. To shorten the measurement time, the temperature increase method was used instead of the equilibrium temperature method. In the temperature increase method, the temperature inside of the calorimeter is lowered below the surrounding water temperature before the measurement, and the temperature is allowed to increase due to the higher surrounding water temperature during the measurement. The rate of the temperature increase when the temperature is the same as the surrounding temperature is then compared to a calibration curve where an electric heater has been used in the same way. The price for the higher throughput is less accuracy and dependency of fuel geometry since the heat flow will be different for different fuel types, especially between PWR and BWR assemblies. Recently at Los Alamos National Laboratory, more sophisticated analysis methods have been considered for the use of calorimetry for treaty verification. These methods aim to predict the equilibrium temperature based on early data and therefore be able to shorten the measurement time while maintaining good accuracy and eliminate the dependency on fuel geometry. Even though the techniques show promise for future calorimetry of spent fuel, certain design features of the current system have been identified to interfere with the accuracy of equilibrium measurements in general such as changing pool temperatures and heat leakage during calibration. (authors)
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Key Information File for Radioactive Waste Repositories - Preliminary Tests - 16169
The Key Information File (KIF) is a document designed to provide a summary of the existence, location and content of an engineered facility for the permanent disposal of radioactive wastes. It should be recognized as part of an internationally integrated system of records and memory. Its primary function would be to maintain awareness of the site, and long-term confidence in the effectiveness of the disposal system, so that the likelihood of unnecessary human disturbance is minimized. In order to ensure this document is accessible to a large audience, it would need to be limited in size. A draft format for the KIF has been proposed, and testing is being conducted on two cases, that of the 'Centre de stockage de la Manche', the closed French Repository (surface) and that of the planned Swedish Spent Fuel Repository (geological). Preliminary results show the potential for such a document. Comments and questions raised by the 'Records, Knowledge and Memory preservation' (RK and M) initiative, coordinated by the Nuclear Energy Agency (NEA) of OECD, are presented. (authors)
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Treatment of LLW Components for Waste Minimization and Recycling - 16258
The Studsvik Melting Facility has been in operation for almost 30 years. Many turbines, heat exchangers and other retired components, pipes, structures have been treated in the facility during those years. The ambition of all projects is to minimize the waste volumes for final disposal and to maximize the amount of material that can be recycled. Another objective, respecting ALARA, is to keep dose exposition to the personnel at very low levels. The overall concept comprises the whole sequence of preparations from road and sea transports, the management of the metallic LLW by segmentation, decontamination and sorting using specially devised tools and shielded treatment cell, to the decision criteria for recycling of the metals, radiological analyses and conditioning of the residual waste into a final form suitable for customer-related disposal. The records show that the amount of material possible for conditional clearance to open market after treatment is between 95 - 97 % on weight for falling metallic scrap. For large components there is a larger span and especially for heat exchangers. This paper describes the Studsvik technical concept and experience related to treatment of metals from the program for modernization and power upgrade of the Swedish nuclear power plants. (authors)
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Views on Long-Term Safety of Deep Borehole Disposal - 16657
The Swedish Nuclear Fuel and Waste Management Co., SKB, has since the end of the 1980s evaluated Deep Borehole Disposal (DBD), of spent nuclear fuel (SNF). This has been done in parallel with the development of a mined repository concept based on the KBS-3V method. The most recent evaluation involved a broad comparison between the KBS-3V method and DBD. The comparison highlights differences between the two methods in the whole chain of the SNF handling including premises for siting, licensing, repository construction, safety during the operational phase, post-closure safety, physical protection, nuclear safeguards, retrieval of the deposited SNF, time planning, need for development, costs and nontechnical project risks. The ambition has been to make an as objective comparison between the two methods as possible, bearing in mind that there are big differences between the two concepts in terms of technical maturity as well as in terms of both quantity and quality of available data. While the primary safety function of the KBS-3V method is long-term containment in a corrosion resistant copper canister, the safety of DBD relies on stagnant groundwater at depth. It is argued in the aforementioned study that this groundwater stagnancy may be jeopardised because of heat and gas evolution in the disposal boreholes, and that deformation of the boreholes make efficient sealing of the deposition boreholes difficult. SKB has taken the position that the DBD method is associated with too many question marks to make it interesting as an alternative method in the Swedish programme. (authors)
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Criticality effect of long-term changes in material compositions and geometry in disposal canisters
For the Swedish final repository the basic requirement concerning margin to criticality is that k{sub eff} should be less than 0.95. This criterion is always fulfilled if the disposal canister is dry. If it is assumed that a canister is filled with water it is shown that the criticality criterion is fulfilled using burnup-credit for PWR fuel and credit for burnable poison for BWR fuel. If water is assumed to leak in, long term chemical processes will change the material compositions and geometries in the canister during the one million year analysis period. The main process expected to affect significantly the chemical composition of the materials in the canister is the corrosion of cast iron and carbon steel of the insert, the corrosion of the stainless steel, Inconel or Zircaloy in the fuel element structural material, the corrosion of Zircaloy cladding and finally the corrosion of spent fuel itself. During metal corrosion, various anoxic corrosion products are formed, usually with lower density than the metal itself, causing thus a volume increase in the canister, accompanied by geometrical changes. In this paper the reactivity effects of these changes are analysed and results presented. It is shown that the largest reactivity increase is due to the decrease of the water gap between the fuel and the fuel channel, due to the formation of Magnetite. Other process that possibly will increase the reactivity is if Siderite is produced besides from Magnetite, the production of Hydrogen, corrosion of the fuel and radial movements of fuel rods. (authors)
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Transition From Operation to Decommissioning with a Sudden Decision to Permanent Shutdown and with Parallel Operation On-site - 18559
In a short time the Swedish nuclear operator OKG have to shift from a operator of three nuclear power plants at one site to a company where both decommissioning and operation will take place in parallel for numerous of years as a result of a decision to permanently shut down two of the three reactors. The paper will be focusing on the challenges to change organization and mind-set from operations to decommissioning within parts of a company while still maintaining and developing the highest standard for the plant still in operation, and this in the combination with a decision of permanent shutdown earlier than expected. The paper will also address how OKG managed to create and maintain momentum in essential processes like permits, defueling and project planning before, during and short after the organization was changed and some of the employees had to leave the company. (authors)
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Use of tsunami in the validation of scale 6.1 for the Swedish spent fuel repository - 14381
In March 2011 SKB applied to the Swedish Radiation Safety Authority (SSM) for a permit to build a final repository for spent Fuel and an encapsulation plant (Clink). The application contained a criticality safety analysis for the canisters to be used in Clink and the copper canister to be used in the final repository. In 2012 SSM requested several updates to the application. One of SSM's concerns was the validation of the codes. SSM emphasized that SKB must improve the method to select critical experiments to the validation suite. To satisfy this demand, SKB used the tool Tsunami in the Scale 6.1 code package to select experiments from the handbook of criticality experiment (IHECSBE). With help of sensitivity files SKB evaluated almost 700 experiments from the LEU-COMP-THERM library. SKB then picked the experiments that most resembled our safety cases. The sensitivity coefficient, ck, from Tsunami has been used to evaluate the similarity between the experiments and the applications. It has been considered that a ck over 0.8 provides a 'good enough' similarity between the experiment and our evaluated safety cases. For the different safety cases between 50-600 experiments satisfied the 0.8 criteria giving a good base of data for statistical analysis. It is worth noticing that PWR-fuel generally receives a higher ck than BWR fuel. Initial analysis shows that the differences in results between BWR and PWR can be explained by a larger sensitivity for the U{sup 238} n,n' part of the cross section in the PWR cases. To make sure the validation covers all materials and required physical parameters a gap analysis was made using engineering judgement by criticality safety experts. From this exercise additional experiments were added to the suite. This paper will discuss the Tsunami results for different safety cases and the reasons for differences between them. It will also present the final results of the validation and discuss experiments that were excluded from the validation suite. (authors)
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