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At least 739 records · Page 41

Development of a U.S. Rail Transport Capability for Spent Nuclear Fuel and High-Level Waste - 16623

The U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) is laying the groundwork for implementing an integrated nuclear waste management system. This includes preparing for future large-scale transport of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) since transport will be a necessary component of an integrated waste management system. DOE continues to plan and develop options for the design of an integrated nuclear waste management disposition system. A significant destination-independent component of this integrated system is a rail transport capability for SNF and HLW. This paper provides an overview of the current DOE-NE effort related to designing and fabricating prototype rail cars. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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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Development of a Conditioned System-Level Groundwater Model to Evaluate Long-Term Groundwater Impacts within a Performance Assessment - 20115

A preliminary performance assessment (PA) of single-shell tank Waste Management Area (WMA) A-AX located at the U.S. Department of Energy (DOE)'s Hanford Site in southeastern Washington is being conducted to satisfy the requirements of DOE Order 435.1 [1] as it relates to closure of the single-shell radioactive waste tanks in the WMA A-AX tank farms. A PA assesses the fate, transport, and impacts of radionuclides within a low-level radioactive waste disposal facility in its assumed closure configuration and the subsequent potential doses to humans over a 1,000-year compliance period and a 10,000-year performance evaluation period. The WMA A-AX preliminary PA evaluation is structured around the complementary use of process-level and system-level models to calculate the facility performance against established DOE Order 435.1 [1] performance objectives. Process-level models are those that represent a detailed phenomenological representation of processes of concern in the PA. Process models typically only represent one or a few of the components of the PA, such as groundwater flow and transport, and must be integrated with other modeling elements to perform PA calculations. System-level models are those that are abstracted from the process models, retaining the essential features of the process model, while allowing integration of all aspects of the PA in a single modeling framework. System-level models are often characterized by coarser numerical discretization, lower dimensionality, or other similar simplifications compared to the process-level model. Traditionally, a three-dimensional (3-D) process-level model is utilized primarily to evaluate the long-term impact on groundwater and the potential doses to individuals who consume contaminated groundwater. System-level models are also utilized to evaluate the groundwater pathway in PAs. These models typically have reduced dimensionality (1-D) and are conditioned utilizing flow fields (Darcy fluxes) and moisture content distributions that are abstracted from the process level models. The abstraction approach assures that the flow field in both models is consistent for a specific set of input parameters for flow, differing only in the discretization and dimensionality of the two models. The preliminary WMA A-AX PA incorporates a detailed representation of the geological system and hydraulic properties within the 3-D model STOMP{sup C} numerical code so that the effects of relevant features and processes on water flow and radionuclide transport in the subsurface can be evaluated. The complementary system-level model is developed utilizing the GoldSim{sup C} code to implement a simplified, 1-D equivalent model to represent the groundwater pathway. Contaminant transport through the vadose zone and unconfined aquifer for Tc-99 and I-129 were evaluated in each of the models. Adjustments to the saturated portion of the GoldSim{sup C} 1-D model were required to mimic the dispersion effect captured with the 3-D STOMP{sup C} model. Once this conditioning was conducted, highly similar results for the transport of Tc-99 and I-129 were achieved at the point of calculation, located 100 m downgradient from the WMA A-AX fenceline. These radionuclides represent elements that are regarded as primary dose drivers in the PA groundwater pathway analysis. The high degree of conformance between the two models suggests that the use of the equivalent 1-D system model is suitable for evaluating the full suite of radionuclides that will be released from the tank sources within WMA A-AX over the 1,000-year compliance period and over the 10,000-year evaluation period. (authors)

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Colloidal Behavior of Plutonium Oxide in Concentrated Electrolyte Solutions

The Hanford Site in Washington State manages legacy high-level radioactive waste streams that display major chemistry and engineering challenges, including the high salt levels and pH values that correspond to conditions under which many classical concepts describing chemical reactivity cannot be applied. One particular challenge that needs to be tackled at Hanford is that Pu concentrations, [Pu], in the soluble phases of the tank wastes are higher than expected based on the solubility of crystalline PuO2, which is widely accepted to be caused by the formation of PuO2 colloid, consisting of nano- to submicron-sized particles (PuO2 NPs). Fundamental research underpinning the behavior of PuO2 NPs under conditions not only relevant to the Hanford tank waste but at high ionic strength in general is needed to reliably predict the chemical reactivity of PuO2 NPs and develop engineering solutions to safely and efficiently process high-level radioactive waste into forms suitable for long-term storage. In this work, we study the behavior of PuO2 NPs (particle size ~100 nm) under high ionic strength conditions by reacting it with highly concentrated (up to 5 M) salt solutions. We explore different electrolyte compositions to elucidate the impact of different anions (NO3-, Cl-, ClO4-, SO42-, C2O42-, CO32-) on the stability of PuO2 NP in the acidic and alkaline pH regime. PuO2 NP aggregation and precipitation as function electrolyte concentration is tracked by a combination of liquid scintillation counting, dynamic light scattering for determination of particle size distributions, and zeta potentials as a proxy for particle charge. At acidic pH, electrolytes containing non-coordinating anions, such as NaNO3, NaCl, and NaClO4 mostly stabilize PuO2 NPs over a large electrolyte concentration range, showing only subtle differences in their reactivity. Other electrolyte anions show a more pronounced effect on the PuO2 NP stability: SO42-, binds directly to the particles’ surface, reverses the particle charge, and precipitates the PuO2 NPs efficiently even at intermediate sulfate concentrations (>0.1 M). In contrast, C2O42- is found to lead to high [Pu] in solution, in the milli-molar range, even at mildly acidic pH (~4). Thermodynamic modeling of the dissolved Pu concentrations using PHREEQC is unable to predict the observed [Pu] in the acidic pH regime, supporting the influence of colloids in maintaining elevated [Pu]. It is noteworthy that the current thermodynamic databases do not include constants for colloidal Pu phases and cannot accurately predict many of the high ionic strength solutions relevant to this work. The mechanisms and models responsible for these observations will need further investigation in the future. At high pH values (~12), PuO2 NPs exhibits classical sol-gel chemistry, meaning that upon destabilization of the colloidal sol, for example by addition of concentrated NaOH, highly porous and viscid PuO2 coagulates are formed that consist of a three-dimensional network likely held together by physical interactions. The PuO2 NP coagulate shows no significant reversibility of the aggregation when contacted with concentrated brines; however, PuO2 NPs can be efficiently resuspended in solution by addition of diluted electrolytes, alkaline solutions containing high amounts of carbonate, or simple addition of water. Especially carbonate is shown to stabilize PuO2 NPs in solution at high pH, characterized by stable colloidal suspensions that are resistant against sedimentation during centrifugation. Thermodynamic modeling of the carbonate system was able to predict an increasing dissolved Pu concentration with increasing carbonate concentration. However, the model was profoundly sensitive to the fixed redox potential and does not include any thermodynamic constants for colloidal Pu species.

Neumann, Julia↗

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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Extending Spent Fuel Storage until Transport for Reprocessing or Disposal

Spent fuel (SF) must be stored until an end point such as reprocessing or geologic disposal is imple-mented. Selection and implementation of an end point for SF depends upon future funding, legisla-tion, licensing and other factors that cannot be predicted with certainty. Past presumptions related to the availability of an end point have often been wrong and resulted in missed opportunities for properly informing spent fuel management policies and strategies. For example, dry cask storage systems were originally conceived to free up needed space in reactor spent fuel pools and also to provide SFS of up to 20 years until reprocessing and/or deep geological disposal became available. Hundreds of dry cask storage systems are now employed throughout the world and will be relied upon well beyond the originally envisioned design life. Given present and projected rates for the use of nuclear power coupled with projections for SF repro-cessing and disposal capacities, one concludes that SF storage will be prolonged, potentially for several decades. The US Nuclear Regulatory Commission has recently considered 300 years of storage to be appropriate for the characterization and prediction of ageing effects and ageing management issues associated with extending SF storage and subsequent transport. This paper encourages addressing the uncertainty associated with the duration of SF storage by de-sign – rather than by default. It suggests ways that this uncertainty may be considered in design, li-censing, policy, and strategy decisions and proposes a framework for safely extending spent fuel storage until SF can be transported for reprocessing or disposal – regardless of how long that may be. The paper however is not intended to either encourage or facilitate needlessly extending spent fuel storage durations. Its intent is to ensure a design and safety basis with sufficient margin to accommodate the full range of potential future scenarios. Although the focus is primarily on storage of SF from commercial operation, the principles described are equally applicable to SF from research and production reactors as well as high-level radioactive waste.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Instrumentation: Nondestructive Examination for Verification of Canister and Cladding Integrity. FY2014 Status Update

This report documents FY14 efforts for two instrumentation subtasks under storage and transportation. These instrumentation tasks relate to developing effective nondestructive evaluation (NDE) methods and techniques to (1) verify the integrity of metal canisters for the storage of used nuclear fuel (UNF) and to (2) verify the integrity of dry storage cask internals.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Panel Session 111: Nuclear Power Plant Onsite SNF storage, ISFSIs and Failed Fuel Handling Issues

This panel session provided updates on several issues seen in developing, operating independent spent fuel storage installations (ISFSIs), Damaged Fuel and aging management recent activities. Presentations proposed include: 1 - Spent Nuclear Fuel Management Activities at Holtec (Myron Kaczmarsky, Holtec International (USA)); 2 - INL Spent Fuel Management: Challenges and Path Forward for Dry Storage (Josh Jarrell, Idaho National Laboratory (USA))

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Concoction and Appraisal of Radiation Hardened Polyurethane Foams for DOE Decontamination and Decommissioning (D and D) Operations

Since World War 2, a number of nuclear facilities have been relocating their nuclear waste to long term storage, or final disposition. These facilities face monumental challenges in cleaning up the fixated transferrable radioactive waste that is left in these facilities after initial cleaning operations have ended. SRNL has identified a solution that is utilizing two-part epoxy foams with heavy metal additives for fixating and shielding transferable radioactive material that has accumulated in non uniform volumes. The purpose of this experiment is to make a library of data of physical characteristics of each foam and additive combination that will have data concerning how the material cures and performs in different environments with varying temperatures and how it will act and perform after being irradiated. Materials Being Tested: - Foam iT 7FR, Foam iT 23FR, 3M Fire Barrier foam, and Hilti CP620 foam; - Additives Bi, Bi2O{sub 3}, Na{sub 2}WO{sub 3}, and WO{sub 3} at 12.5% and 25%. What's being investigated: - Expansion data from each foam; - Radiation shielding data for each foam combination; - Peak temperatures for each foam. Utilized a FLIR thermal camera to find peak temperature and time it took the foam to reach it. No noticeable difference in volume or weight after blank foam samples sat in the environmental chamber for 24 hours except the 3M foam at the extreme relative humidity of 90% highlighted above. The more additives used in the foam, the more they were able to shield. The foams begin to saturate around 100% wt% where more additive used will not increase the shielding capability of the foam. Similar thermal dispersion in blanks and in samples with Bi additive. Time taken to peak temperature is very similar between the two. Similar peak temperatures in blanks and in Bi samples. Hilti had slightly elevated temperatures. All foams maintained their physical properties post environmental chamber except the 3M foam. With the increased sponginess from the high relative humidity's, 3M may be disqualified for further testing if its mechanical properties also suffer. While curing all foams had their peak temperature occur between 40 deg. C and 70 deg. C except the Hilti which occurred above 90 deg. C with and without additive. With this information, no foams would be disqualified. After radiation shielding studies, it was found that IT 14 with Bismuth and Bismuth Oxide Additives provided the best shielding of Co-60 with over 9% attenuation.

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Cement Solidification of Residual Metal Oxide from Radioactive Waste

A lot of uranium contaminated soil and concrete waste is generated from the dismantlement of uranium conversion facilities. There are several disposal methods for radioactive materials such as regulation exemption, decontamination and long-term storage. It is necessary for us to achieve the permanent disposal of radioactive waste. The methods for solidifying radioactive waste include cement solidification, asphalt solidification, and polymer solidification. Cement solidification has several advantages such as well-known materials and technologies, various applications, and reasonable cost. On the contrary, it has certain disadvantages such as a large volume expansion of solidification for radioactive waste. Because cement solidification has excellent structural strength and a shielding effect, it is widely used for the permanent disposal of radioactive waste. Cement solidification operates as a barrier against the leaching of nuclides during radioactive waste disposal. To evaluate the criteria of KORAD (Korea Radioactive Waste Agency), integrity tests on the solidification of radioactive waste need to be performed. In this study, to inspect the criteria of KORAD, we carried out the cement solidification of residual metal oxide from radioactive waste. An integrity inspection such as the pressure strength, inundation, and leaching test on cement solidification was performed. We also analyzed the experimental results according to the criteria of KORAD. The calculation of the leachability index on the primary and secondary cement solidification for leaching detection of nuclides is carried out. The integrity of primary and secondary cement solidification is evaluated. (authors)

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Efficient and compact mobile equipment based on the new RADEON-NWM technology to process liquid radioactive wastes resulted from the accidents of the nuclear installations

During the operation of nuclear reactors important volume of liquid and solid radioactive wastes are generated, which, in normal conditions, becomes processed by stationary equipment by different methods to minimize their volume and then sent to specially constructed storages. The cases of accidents of Chernobyl and Fukushima showed that the localization of rejected big quantity of radioactive wastes is a prior problem for their further processing by stationary equipment. In this regard it is very important the processing of radioactive wastes on the contaminated areas to localize them by mobile equipment based on the efficient technologies. RADEONNWM new technology allows resolving this problem. This technology is compact, completely automated, which makes possible to assemble it on a standard 40-ft by 7-ft trailer driven by heavy-duty truck. The new technology is fully elaborated, the necessary tests are conducted. (authors)

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Experiences with Mobile Units for Liquid RAW Management in Slovak and Czech Nuclear Power Plants - 20089

In Nuclear Power Plants VVER 440, Russian type of reactor there was designed auxiliary building where according to basic design all operational liquid radioactive waste (RAW) should be stored in large capacity tanks during the whole operational period. That's why there are in the auxiliary building several types of large storage tanks with different internal volume from 110 m{sup 3}, 460 m{sup 3} to 550 m{sup 3}. Storage tanks are usually situated in rooms with limited access and minimal handling space. Operational waste was here transported by the system of pipes from the reactor building. There is possibility to handle with the waste and set the pipeline in to the chosen storage tanks. Nobody takes care about the quality of waste in the storage tanks and waste waits for the period of decommissioning when it should be treated. The storage capacity was designed for the whole lifetime of nuclear power plant. Currently the life time of Nuclear Power Plant's (NPP) is extended and there is necessity to retrieve and treat the RAW in advance. Another reason why the stakeholder, who operates the NPPs, decides to treat the operational RAW during the operational period of NPP, is that there are still new requirements from safety point of view which need to have free available space for example because of potential accident. So treating the operational RAW during the life time of NPP is present issue. Because in basic design nobody cares how the RAW will be retrieved and treated in the future, there is usually no technology for that in existing NPP. Some of the NPPs build an additional building with suitable technologies for treatment or hire the company which offer mobile technology for retrieving the waste from the storage tanks. Wood Nuclear Slovakia PLC provides all activities within liquid RAW management projects by its own staff including management, design, construction and deployment of equipment, engineering support, as well as all RAW retrieval and solidification activities. We offer RAW management as a service without any significant changes in existing NPP's buildings and technology. It is always a technical challenge to conduct successful retrieval, pumping and treatment of RAW. The complex RAW retrieval system can be designed as tailored solution for any RAW storage according to known boundary conditions and specific needs. These systems are based on years of experience with RAW manipulation, processing and treatment and the whole RAW management knowledge by Wood Nuclear Slovakia LTD. (authors)

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The Cires facility in Morvilliers: from a VLL waste disposal facility to the development of industrial activities - 15252

The disposal facility of CIRES for very low level wastes was designed to provide safe and cost effective services to waste generators, in particular for decommissioning works. The facility could significantly increase waste annual deliveries to meet waste generators' needs and accommodate large disused components as PWR steam generators after decontamination. However an anticipated saturation of its capacity is foreseen as needs for disposal has doubled since its start-up. The geometry of the cells was optimized and the technical capacity of the repository is presently by 40% above its licensed capacity. Saturation has been delayed; however a new facility will be necessary by 2030 and other waste management options are to be considered as recycling of metallic wastes or crushed concrete rubble. CIRES facility also became Andra's industrial platform for the management of institutional wastes. A grouping and a storage facility have been built and a new processing facility should start up in 2016. (authors)

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Knowledge Management in the Field of Nuclear Energy - How to Bridge the Gap between Retiring Workforce and Next Generations - 17101

The current phase-out decision of the German Federal Government and the 'Energiewende' - resulting from the Fukushima Daiichi Accident in 2011 - changed the nuclear market significantly. Together with the change in political and social attitude towards nuclear energy since 2002 it has contributed to a decline in technical expertise. This as well as the generation change in the nuclear technology lead to the risk of losing accumulated knowledge and experience which still are essential for the future - not only in Germany but in other countries - even though the focus in Germany will shift from operating to decommissioning, dismantling and radioactive waste management. A long-term strategy in order to preserve technical expertise to solve the forthcoming tasks in Germany was required. This strategy should include e.g. the implementation of technical programs not only for internal training of the staff but also for authorities. Together with authorities TUV NORD developed individual training concepts for a broad range of customers to preserve a transfer of know-how with the state-of-the-art in science and technology and even support international projects, especially in countries with less experience in the nuclear fields. Additionally, special data bases were developed in order to ensure the accessibility of long-term nuclear data storage. Furthermore technical conferences in different fields like decommissioning and radioactive waste disposal were established as an opportunity for knowledge exchange. Based on given examples it is being demonstrated how technical expertise can be preserved even though the field of nuclear energy in Germany has changed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Disposal of a pretreated oil sludge by the cementation method

Radioactive oil sludge is generated in oil production rigs worldwide and in large quantities. However, in Brazil, there is no disposal route or well-defined treatment option for this waste. This sludge is classified as TENORM due to the presence of Potassium-40 and the radionuclides of the natural decay chains of thorium and uranium. The management of this radioactive waste has been in situ packaging on the offshore platforms, followed by its transport to storage facilities onshore. Long term storage is problematic due to the high content of hydrocarbons, sulfur and water, which generates the toxic gas H{sub 2}S and the corrosive sulfuric acid. Ozonation, an advanced oxidation process (AOP) can be considered an attractive option of treatment, and followed by cementation can improve safety in long-term storage. This work aims at investigating if the ozone process was able to enhance the immobilization capacity of oil sludge. Oil sludge from different drums and mixed oil sludge (obtained by mixing equal parts of each drum) were analyzed by ICP. Solidification of oil sludge with Portland cement enhances safety of the waste disposal. In addition, ozonation treatment is efficient in the degradation of organic compounds, increasing the resistance of the solidified material.

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Innovative Waste Packaging and Associated Venting/Hydrogen Management - 16594

The original strategy for managing intermediate level waste (ILW) was developed in the 1980s by the UK Nuclear Industry Radioactive Waste Executive (NIREX), which is now the Radioactive Waste Management Limited (RWM), a wholly owned subsidiary of the Nuclear Decommissioning Authority (NDA). The strategy involved retrieving ILW, sorting and encapsulating it in cement based grout within thin walled stainless steel containers. The containers would then be transferred to a large purpose-built shielded ILW store on site, where they would be stored until the planned final deep geological disposal facility becomes available. In recognition of the high up-front costs of such a strategy, which requires remote facilities for the encapsulation and storage, and coupled with the long programme duration for the design, construction and commissioning of these facilities, waste management organisations responsible for managing the clean-up of legacy facilities within the UK have, in recent years, sought innovative solutions which could potentially both accelerate clean-up and hazard reduction on sites, and with a lower overall programme lifecycle cost. Many of the recent innovative solutions are focused around a new family of packaging designs known as robust shielded containers (RSCs). These are self-shielded containers manufactured from materials such as ductile cast iron or fabricated steel. These containers offer the potential to package wastes for disposal without encapsulation and avoid the need for a highly engineered remote handled shielded store for interim storage. The paper provides a high level overview of the generic concepts of RSCs developed by Croft and focuses on the specific challenges of wasteform evolution on waste package integrity and performance. In particular, the integrity and performance of the filtered vent system required to manage internal pressurisation and to ensure sufficient venting capacity of hydrogen and other evolved gases. The paper summarises the analysis by ARUP undertaken to substantiate the RSC integrity under detonation/deflagration explosion loadings arising from hydrogen accumulation. Analyses of the structural response of the filter assembly were undertaken for both detonation and deflagration explosion scenarios, with the loadings applied to the filter housing rim and to the filter vent cover. The paper also presents evidence of the mechanical performance of the filter medium in terms of pressure transients during deflagration events and the degree to which filtration efficiency is maintained. (authors)

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Application of polymers for the emergency damage control and remediation at nuclear and chemical industry sites - 15259

From the standpoint of possible accidents and the following complex emergency damage control the nuclear and chemical industry sites are of the primary concern. Among the possible causes of accidents and emergency situations are technological and waste solution spills. Polymers can be used for this problem resolution and the spilled solutions could be effectively isolated. The apparent advantage of polymer materials is the possibility of solidifying solutions with rather high acidity and high salt bearing. To date the situations with the RW storage tank decommissioning in Russia and in the United States have much in common: the lack of the waste management strategy developed at the initial stage of production, the end of the storage tank lifetime. Development of the in-situ solidification technique for the mobile (liquid) components of heterogeneous waste will provide a good option to contain the radio nuclides within the storage site boundaries. The presented work was aimed at the search for the simple engineering decisions that would provide the waste immobilization in case of the accidental liquid spills. The application of polymers for the purpose indicated would allow in case of emergency to reduce the radioactive aerosol concentration in the air and prevent the distribution of aerosols on the premises and in the environment. Another promising area of the polymer application could be the technical challenges associated with the decommissioning of NPPs and other objects and facilities including the large volume LRW storage tanks. (authors)

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