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At least 397 records · Page 22

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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 OSTI2019

Revisit of Phase Separation of Borosilicate Glass for Vitrification of High Level Radioactive Waste - 19166

Phase separation of glasses for vitrification of high level radioactive waste has been experimentally studied to know uncertainty of the phase separation conditions; temperature and glass compositions as well as to propose possibility of recycling of heavy metals on a longer time scale. Our current results at 700 and 750 deg. C for ternary borosilicate glass system showed that the phase separation region at 700 deg. C spreads along with a direction of increasing Na{sub 2}O composition when compared with the data in literature. On the other hand, at the glass composition employed in the reprocessing plant in Japan, no segregation was observed. Concerning the application of phase separation phenomena, we had previously reported recovery of almost all elements from the vitrified borosilicate glass containing simulated fission products through phase separation of the borosilicate glass and this suggests that we could propose that after vitrification of high-level radioactive waste heavy metals may be retrieved as energy or materials resources when they are required to utilize on a longer time scale. From this standpoint, phase separation of lead borosilicate glass was also studied to seek economical and versatile recycling. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Study on release and transport of aerial radioactive materials in reprocessing plants

The release and transport characteristics of radioactive materials at a boiling accident of the high active liquid waste (HALW) in a reprocessing plant have been studied for improving experimental data of source terms of the boiling accident. In the study, a heating test and a thermogravimetry and differential thermal analysis (TG-DTA) test were conducted. In the heating test using a simulated HALW, it was found that ruthenium was mainly released into the air in the form of gas and that non-volatile elements were released into the air in the form of mist. In the TG-DTA test, the rate constants and reaction heat of thermal decomposition of ruthenium nitrosyl nitrate were obtained from TG and DTA curves. (authors)

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2025

Capture of volatile RuO 4 from oxidized simulated used nuclear fuel solutions

Ruthenium is a challenging fission product in used nuclear fuel (UNF) reprocessing due to its complex redox chemistry, variable speciation in nitric acid, partial extractability, and volatility. This work presents a strategy for Ru removal based on the volatilization of RuO 4 using the strong oxidant sodium bismuthate, followed by RuO 2 deposition onto various substrates. Among the materials tested, polymer-based substrates such as polyolefin wax film (Parafilm®), exhibited superior performance, achieving up to near-quantitative Ru removal from solution. After dissolution of the substrate, 99.6% of the Ru was recovered as RuO 2 . The deposition mechanism onto the polyolefin wax film involves both physisorption and chemical reaction through oxidative cleavage of olefinic bonds within the polymer. In contrast, on inorganic substrates such as Al foil, RuO 4 undergoes incomplete surface adsorption and reduction, as confirmed by FTIR and XPS analyses. This approach remains effective under conditions of simulated oxidation of UNF solutions, selectively removing Ru while leaving other fission products in solution. As a result, capture of RuO 4 offers a practical and efficient strategy for ruthenium decontamination and recovery, offering a route for integration into existing UNF reprocessing flowsheets to enhance overall process safety and performance.

Ruthenium↗

DOE OSTI2015

Improvements to neutron detection of Pu in solvent for an aqueous buffer tank, magnox reprocessing plant, Sellafield - 14216

The PP Active Feed Tank (PPAFT) is a buffer tank that is not safe by geometry and which is located between the Medium Active (MA) and Plutonium Purification (PP) cycles of the chemical separation process in the Magnox Reprocessing Plant. The PPAFT receives Pu nitrate in acid which has been conditioned ready for the extraction of Pu with TBP/OK (tri-butyl phosphate in kerosene) solvent in the PP Cycle. The criticality safety case for the PPAFT recognises that during both normal operations and potential maloperations solvent which is dissolved (TBP which is present in solution but can be released upon a change in acidity) and entrained (TBP/OK physically carried over but not in solution) will accumulate in the PPAFT. The only practicable route out for Pu in solvent in the PPAFT is the normal process route out from the base of the PPAFT to the PP cycle through intermediate vacuum lift and feed vessels. However, the dished end base of the PPAFT and the intermediate vacuum lift and feed vessels are of particularly challenging geometry. There are five neutron monitors placed around the PPAFT for the detection of Pu in solvent. The original design intent of the neutron detection system was that any single monitor should be able to reveal an accumulation of Pu in solvent before it would potentially be unsafe if fed downstream. However, over time there have been changes to the fuel feed and chemical process of the plant which have had the potential to affect the performance of the neutron monitoring system. The ability of the neutron monitoring system to function to its original intent was questioned. Work has recently been carried out to investigate the capabilities of the PPAFT neutron monitoring system for current plant conditions and to identify potential improvements while making best use of the existing neutron monitors. It was found that moving a neutron monitor from the side to the top of the PPAFT and adding alarms based on the ratio of signals from multiple neutron detectors would be effective in improving detection of Pu in solvent. This paper describes the challenges involved in detecting thin layers of Pu in solvent using neutron monitors, discusses the improvements made to the PPAFT neutron monitors and details the combination of real plant data and modelling with the shielding code MCBEND used to provide confidence in the enhanced neutron monitoring system. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Capacity of ETS-10 supported carbon nanosorbent for removal of multicomponent from off-gas stream

Nuclear fission results in fission and activation products, some of which are volatile under the conditions of used nuclear fuel (UNF) reprocessing. Several volatile radionuclides are released into the atmosphere through offgas stream, which includes {sup 3}H, {sup 14}C, {sup 129}I and noble gases, Kr and Xe. {sup 129}I is a highly mobile and volatile contaminant. It has half-life period about 1.57 x 107 years that tends to bioaccumulate, concentrating in the thyroid glands and affecting bio-metabolism in humans. {sup 85}Kr is a chemically inert radioisotope with a half-life of 10.7 years that continuously accumulate in the environment. Control of emissions of these contaminants is necessary to comply with the stringent regulatory limits placed on their releases to protect human health and environment. Removal of these contaminants has been a challenge since the 1950's, and several methods including cryogenic distillation, fluorocarbon absorption, wet scrubbing, and adsorption have been employed. Currently, adsorption on porous solid sorbent shows the most promise results. Up to now, several sorbents such as silver impregnated mordenite, silver ETS-10, chalcogen based aerogels, and metal organic framework, MOFs have been developed for this purpose. Despite these advances, development of economical and highly active sorbents for complete removal and immobilization of volatile radioactive gases remains a key challenge in realizing advanced fuel cycles. Present research work is focused on the determination of the removal capacity of C'at'ETS-10 sorbent for a multicomponent (iodine and krypton) off-gas stream at ambient temperature. Our previously reported work showed that the 10 wt% C'at'ETS-10 sorbent was active for adsorption of krypton as well as iodine at ambient temperature and pressure. Therefore, 10 wt% C'at'ETS-10 sorbent was synthesized by wet-impregnation technique. The performance evaluation of the prepared nanosorbent was examined for the removal of iodine and krypton from the breakthrough curves obtained using bench scale experimental setup. ETS-10 supported carbon nanoparticles, C'at'ETS-10 nanosorbent was successfully synthesized by wet-impregnation technique. The 1/8 inch pelletized sorbent was used to minimize the pressure drop in adsorption column. The capacity of sorbent was found to be 0.143 mg g{sup -1} for krypton and 24.5 mg g{sup -1} for iodine. This experimental results indicates the C'at'ETS-10 is the effect sorbent for removal of multicomponent, krypton as well as iodine from the off-gas stream at ambient temperature. (authors)

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2015

Initial recovery of the magnox reprocessing plant following a loss of reductant feed in the solvent extraction process - 14429

A recent event in the Magnox Reprocessing Plant involved the loss of reductant feed to a contacting vessel and required the plant to be shutdown. The criticality team attended the site out-of-hours to facilitate the production of a recovery plan. This initial recovery plan involved manual dosing of the contactor with reductant but was heavily dependent upon initial information regarding the length of time the maloperation had been occurring. Key assumptions fundamental to maintaining criticality safety during the planned recovery were clearly highlighted in the assessment that was written to support the operation, which allowed informed decisions to be made regarding the adequacy of the plan, once further information emerged suggesting the fault had been ongoing for much longer than originally assumed and therefore the amount of Pu in the contactor was much higher than initially thought. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Design and Test Plan for an Integrated Iodine Scrubber and Polishing Bed System

The capture and subsequent immobilization of four regulated volatile radionuclides ( 3 H, 14 C, 85 Kr, and 129 I) and relevant semivolatile species from the off-gas streams of a used nuclear fuel (UNF) reprocessing facility has been a topic of significant research interest on the part of the US Department of Energy and other international organizations. Significant research and development has been conducted over the past decade. In 2016 an initial engineering evaluation and design of the off-gas abatement systems required for a hypothetical 1000 t/yr UNF reprocessing facility treating 5 yr–cooled, 60 GWd/tIHM UNF was completed. One of the key findings of that report was that the consumption rate of silver-based iodine sorbents in the dissolver off-gas primary iodine capture bed is very high and may warrant the evaluation of alternative methods to capture the bulk of the iodine that could significantly reduce the associated frequent remote handing of the iodine filter beds. This report is intended to describe the design of an experimental system that can be used to examine the use of aqueous scrubbing to remove the bulk of the iodine from the dissolver off-gas stream prior to a silver-based solid sorbent that would be used to provide the final iodine capture or polishing step. This report also provides a description of the initial series of tests that are proposed for this system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Radioactive waste management in France: taking-up present and future challenges - 15120

The reduction of the volume and radiotoxicity of radioactive waste has been one of the key drivers for the implementation of the French nuclear program. Interactions between the nuclear industry, the administration, the competent authorities and other stakeholders including the civil society has led to a continuous evolution of the industrial facilities, practices, solutions and products to move towards waste minimization. Bringing a solution to the management of ultimate radioactive waste is clearly a requirement for the sustainable development of nuclear energy whose crucial role in the world's future energy mix has been confirmed in the past few months by several international organizations. The first question to be answered concerns the definition of radioactive waste. This notion may be seen as self-evident. It is not. For example, used fuel is considered a High Level Waste in Finland or Sweden, but a recyclable material in countries including China, France, India, Japan, Russia, and the Netherlands. In France, a definition is provided by law to clarify this notion. Indeed, the 2006 French law concerning the Sustainable Management of Radioactive Materials and Waste stipulates that 'ultimate radioactive waste shall include any radioactive waste for which no further processing is possible under current technical and economic conditions, notably by extracting their recoverable fraction or by reducing their polluting or hazardous character'. Used fuel is thus reprocessed in order to reduce the final waste volume and long term radiotoxicity, while extracting value from the recovered materials. Even if the general approach and the processes may be common for waste management, it is important to clearly make the distinction between the wastes stemming from used fuel processing or conditioning (when considered as waste), the wastes generated through the operation of nuclear power plants and fuel cycle facilities, and the waste arising from the dismantling of nuclear installations. After a brief overview of the different types of waste stemming from nuclear power generation, this paper will review solutions and processes already implemented for their safe and efficient management. It will also practically illustrate the evolution of these processes and related facilities to further gain in efficiency, notably in terms of volume reduction; the related R and D outcomes will be addressed. It will also describe the upcoming challenges in this field and the envisioned solutions to respond to them; the efficiency, the completeness and consistency of these solutions with the currently considered global framework, based on a stepwise and long term approach from waste generation to disposal, will also be analyzed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2014

Adsorption Isotherms for Xenon and Krypton using INL HZ-PAN and AgZ-PAN Sorbents

The generation of adsorption isotherms compliments the scale-up of off-gas processes used to control the emission of encapsulated radioactive volatile fission and activation products released during Used Nuclear Fuel (UNF) reprocessing activities. A series of experiments were conducted to obtain capacity results for varying Kr and Xe gas concentrations using HZ-PAN and AgZ-PAN engineered form sorbents. Gas compositions for Kr ranged from 150-40,000 ppmv and 250-5020 ppmv for Xe in a helium balance. The experiments were all performed at 220 K at a flowrate of 50 sccm. Acquired capacities were then respectively fit to the Langmuir equation using the Langmuir linear regression method to obtain the equilibrium parameters Qmax and Keq. Generated experimental adsorption isotherms were then plotted with the Langmuir predicted isotherms to illustrate agreement between the two. The Langmuir parameters were provided for input into the OSPREY model to predict breakthrough of single component adsorption of Kr and Xe on HZ-PAN and AgZ-PAN sorbents at the experimental conditions tested. Kr and Xe capacities resulting from model breakthrough predictions were then compared to experimental capacities for model validation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Alternative Materials Literature Review: A Review of Alternative Materials for Xenon Capture in Comparison to AgZ-PAN

A collaborative literature review between Idaho National Laboratory (INL) and Pacific National Laboratory (PNNL) was conducted to research alternative metals/materials for xenon adsorption in gaseous effluents generated from used nuclear fuel (UNF) reprocessing. The review was conducted to research sorbents that could outperform silver mordenite (AgZ) at ambient temperature. The use of silver requires complex and expensive waste disposal processes due to it being an RCRA metal, therefore alternative materials need to be researched and evaluated. The report details two different types of sorbent material: alternative zeolite materials and non-zeolite-based materials (i.e., MOFs, COFs, etc.). This report documents a brief history of the materials, what testing has been conducted on them, and what results suggest the possibility of outperforming the silver-based materials. Sorbent performance characteristics such as durability, capacity, and operating temperatures were evaluated. The simplicity of the sorbents was also evaluated, meaning how they are acquired (synthesized vs. naturally abundant). Recommendations were made for the most promising sorbents based on available research in the subject area and these recommended sorbents will be tested in the upcoming fiscal year. The recommended sorbents that are zeolite based are CHA zeolites, ZIF-69, and Na-ETS-10. The recommended sorbents that are non-zeolite based are SBMOF-1 and CC3.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Alternative Materials Literature Review: A Review of Alternative Materials for Xenon Capture in Comparison to AgZ-PAN

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Determination of Desorbed Species During Heating of AgI-Mordenite Provided by ORNL

This study is focused on describing the desorbed off gases due to heating of the AgIMordenite (MOR) produced at ORNL for iodine (I 2 ) gas capture from nuclear fuel aqueous reprocessing. In particular, the interest is for the incorporation of the AgI-MOR into a waste form, which might be the Sandia developed, low temperature sintering, Bi-Si oxide based, Glass Composite Material (GCM). The GCM has been developed as a waste form for the incorporation any oxide based getter material. In the case where iodine may be released during the sintering process of the GCM, additional Ag flake is added as further insurance in total iodine capture and retention. This has been the case for the incorporated ORNL developed AgIMOR. Thermal analysis studies were carried out to determine off gasing processes of ORNL AgIMOR. Independent of sample size, ~7wt% of total water is desorbed by 225°C. This includes both bulk surface and occluded water, and are monitored as H2O and OH. Of that total, ~5.5wt% is surface water which is removed by 125°C, and 1.5wt% is occluded (in zeolite pore) water. Less than ~1 wt% total water continues to desorb, but is completely removed by 500°C. Above 300°C, the detectable remaining desorbing species observed are iodine containing compounds, including I and I 2 .

09 BIOMASS FUELS↗

DOE OSTI2013

Know-how and experience of the decommissioning of the Eurochemic reprocessing plant

Belgoprocess started the industrial decommissioning of the main process building of the former Eurochemic reprocessing plant in 1987. In 2008 the eastern part was torn down and in 2010 the central part of the building was removed. The demolition of the third and last part is planned in 2014. At that point the area will be restored to brown field conditions. During the 25 years of decommissioning the Eurochemic installations a lot of effort has been invested in improving the efficiency. The Belgian decommissioning policy is based on maximizing effort for decontamination with the goal of unconditional release and minimization of nuclear waste. As a result a number of installations have been implemented that reduce the processing costs, increase the output of decontamination and increase the speed of release measurements. - The abrasive blasting installation is used to remove a thin layer from metal components and concrete objects. - The concrete crushing and sampling installation is used as the final measurement before unconditional release of concrete debris. - The concrete spec installation is used to separate the contaminated part from the non contaminated part of concrete debris. As the end of the decommissioning project is approaching the time is right to look back and evaluate the course of the project. Looking back what were the key decisions during the project, what were important milestones and what could or should have been done differently knowing what we know now. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Thermodynamic Assessment of LiCl-KCl-PuCl{sub 3} Ternary System

Pyroprocessing in LiCl-KCl molten salt is being considered as an alternative to the PUREX for the used nuclear fuel treatment. As more and more spent fuel is reprocessed, the composition of molten salt electrolyte changes, which alters the thermodynamic properties of the system and subsequently requires the adjustment of operation conditions. In order to optimize the system design and carefully operate it, thermodynamic assessment of actinides and lanthanides in LiCl-KCl molten salt is needed. Even though several systems have been assessed recently, such as LiCl-KCl-CeCl{sub 3} and LiCl-KCl-UCl{sub 3}, as one of the major actinides in pyro-processing, PuCl{sub 3} in LiCl-KCl hasn't been calculated. In present work, LiCl- KCl-PuCl{sub 3} is being considered and studied. Sub-binary systems of LiCl-PuCl{sub 3} and LiCl-KCl system was calculated firstly by using two-sublattice model based on the CALPHAD (Calculation of Phase Diagram) theory. The result indicates both of LiCl-PuCl{sub 3} and LiCl-KCl show simple eutectic properties and no solid solution is found. This research will help to predict the properties in electro-refiner when the concentration of elements in LiCl- KCl molten salt electrolyte changes during pyro-processing. Currently, binary system of KCl-PuCl{sub 3} is also being assessed and after that, phase diagram of ternary system of LiCl-KCl-PuCl{sub 3} will be calculated to provide a whole sight of the properties of PuCl{sub 3} in LiCl-KCl eutectic salt. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Recovery from loss of reductant feed to a U/Pu separation contactor, magnox reprocessing plant, Sellafield - 14217

Following a revealed sustained loss of reductant feed to the Primary Separation 4 (PS4) U/Pu separation contactor in the Magnox Reprocessing Plant in 2014 the solvent extraction cycles were safely shutdown. However, initial mass balance calculations indicated that over four times the typical mass of Pu could be present in PS4. Due to the loss of reductant feed, much of this Pu was expected to be in the solvent phase of Tributyl Phosphate and Odourless Kerosene (TBP/OK) where it would be difficult to observe with the neutron monitors located under PS4. Samples were taken of the solvent in PS4 which confirmed the presence of Pu in solvent; normally almost all of the Pu in PS4 is in the aqueous phase. The volume ratio of solvent to aqueous in the PS4 contactor is high. This meant that simply re-dosing PS4 with reductant could not be supported as a recovery option due to the potential to realise high Pu concentrations in the aqueous phase. This paper details the recovery options which were considered and the ultimate decision to recover PS4 using Gadolinium (Gd) poisoning and Gd-dosed reductant feeds. The recovery included: - The need to consider the impact of Gd poisoning on monitoring of the recovery with the PS4 neutron monitors; - Close support from criticality safety personnel on the plant at all times during the recovery; - Production of operator aids and short fit for purpose safety documentation to assist with the recovery. The recovery of PS4 proceeded safely and smoothly. There were no losses of Pu to raffinate streams despite the prolonged shutdown. The plant was then safely rundown to flush out residual Pu and Gd. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Evaluation of performance indicators applied to a material recovery facility fed by mixed packaging waste

Highlights: • Material Flows assessment of a Material Recovery Facility has been realized. • Performance Indicators for a generic MRF have been proposed. • A Dashboard has been utilized to measure the facility performance. • Experimental values of incoming waste and output products composition are reported. - Abstract: Most of the integrated systems for municipal solid waste management aim to increase the recycling of secondary materials by means of physical processes including sorting, shredding and reprocessing. Several restrictions prevent from reaching a very high material recycling efficiency: the variability of the composition of new-marketed materials used for packaging production and its shape and complexity are critical issues. The packaging goods are in fact made of different materials (aluminium, polymers, paper, etc.), possibly assembled, having different shape (flat, cylindrical, one-dimensional, etc.), density, colours, optical properties and so on. These aspects limit the effectiveness and efficiency of the sorting and reprocessing plants. The scope of this study was to evaluate the performance of a large scale Material Recovery Facility (MRF) by utilizing data collected during a long period of monitoring. The database resulted from the measured data has been organized in four sections: (1) data related to the amount and type of inlet waste; (2) amount and composition of output products and waste; (3) operating data (such as worked hours for shift, planned and unscheduled maintenance time, setting parameters of the equipment, and energy consumption for shift); (4) economic data (value of each product, disposal price for the produced waste, penalty for non-compliance of products and waste, etc.). A part of this database has been utilized to build an executive dashboard composed by a set of performance indicators suitable to measure the effectiveness and the efficiency of the MRF operations. The dashboard revealed itself as a powerful tool to support managers and engineers in their decisions in respect to the market demand or compliance regulation variation as well as in the designing of the lay-out improvements. The results indicated that the 40% of the input waste was recovered as valuable products and that a large part of these (88%) complied with the standards of the recycling companies. The evaluation of the indicators led to the decision to modify the layout to improve the interception of some polymers for which the performance indicators were poor. In particular, two additional optical sorters have to be inserted to increase the yield indicator and to the overall performance of the facility. Definitely, the results of the work allowed to: increase the yield and purity of products’ flows; ensure the compliance of waste flows; increase the workability.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Exploiting Knowledge of Technetium in the Subsurface to Develop Remediation Strategies - 18223

Technetium-99 (Tc) contamination remains a major environmental problem at legacy nuclear reprocessing sites, such as the Hanford Site in south-central Washington State, where the total inventory of released Tc is estimated at 700 curies (Ci). The environmental risk associated with Tc is due to its long half-life (213,000 years) and the high mobility of the oxidized anionic pertechnetate species (Tc(VII)O{sub 4}{sup -}). Thus, there is a critical need to resolve scientific and technical issues underpinning the ability to safely, cost-effectively, and efficiently remediate Tc-contaminated vadose zone sediments and groundwater. Developing enhanced attenuation strategies for released Tc that reduce contaminant migration, while mitigating risk to human health and the environment, requires characterization of any retardation processes and Tc mass flux. The current conceptual model for Tc transport in the vadose zone assumes that Tc is predominantly present in its oxidized form as TcO{sub 4}{sup -}, which is highly soluble and mobile. However, under certain conditions, Tc can become entrained in solid phases. Tc-containing mineral phases from sediment cores extracted from various Hanford waste disposal sites were characterized to determine Tc present in vadose zone sediments as (i) TcO{sub 4} - that remains in water films associated with fine-grained sediments; (ii) Tc co-precipitated with other lanthanides in phosphate mineral phases; and (iii) Tc present as TcO{sub 2}.nH{sub 2}O in localized reducing zones. These results will be used to develop improved long-term Tc remediation strategies that are most likely to succeed during field application, through stimulation of conditions that enhance Tc attenuation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗