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At least 1,603 records · Page 89

DEMONSTRATION OF MIXING AND TRANSFERRING SETTLING COHESIVE SLURRY SIMULANTS IN THE AY-102 TANK

In support of Hanford's waste certification and delivery of tank waste to the Waste Treatment and Immobilization Plant (WTP), Savannah River National Laboratory (SRNL) was tasked by the Washington River Protection Solutions (WRPS) to evaluate the effectiveness of mixing and transferring tank waste in a Double Shell Tank (DST) to the WTP Receipt Tank. The work discussed in this report (Phase III) address the impacts cohesive simulants have on mixing and batch transfer performance. The objective of the demonstrations performed in Phase III was to determine the impact that cohesive particle interactions in the simulants have on tank mixing using 1/22{sup nd} scale mixing system and batch transfer of seed particles. This testing is intended to provide supporting evidence to the assumption that Hanford Small Scale Mixing Demonstration (SSMD) testing in water is conservative. The batch transfers were made by pumping the simulants from the Mixing Demonstration Tank (MDT) to six Receipt Tanks (RTs), and the consistency in the amount of seed particles in each batch was compared. Tests were conducted with non-Newtonian cohesive simulants with Bingham yield stress ranging from 0.3 Pa to 7 Pa. Kaolin clay and 100 {mu}m stainless steel seed particles were used for all the non-Newtonian simulants. To specifically determine the role of the yield stress on mixing and batch transfer, tests were conducted with a Newtonian mixture of glycerol and water with at viscosity of 6.2 cP that was selected to match the Bingham consistency (high shear rate viscosity) of the higher yield stress kaolin slurries. The water/glycerol mixtures used the same 100 {mu}m stainless steel seed particles. For the transfer demonstrations in Phase III, the mixer jet pumps were operated either at 10.0 gpm (28 ft/s nozzle velocity, U{sub o}D=0.63 ft{sup 2}/s) or 8.0 gpm (22.4 ft/s nozzle velocity, U{sub o}D=0.504 ft{sup 2}/s). All batch transfers from the MDT to the RTs were made at 0.58 gpm (MDT suction velocity 3.95 ft/s). The demonstrations that used simulants that ranged from 1.6 Pa to 7 Pa yield stress had the most successful batch transfer of solids to the RTs in terms of the total quantity of seed particles transferred. Testing suggest that when mixing water/seed particles and transferring, water provides the least desired batch transfer of solids based on the total quantity transferred. For the water tests, large dead zones of solids formed in the MDT and fewer solids get transferred to the RTs. For simulants with a yield stress of 0.3 Pa and below, the batch transfer behavior in terms of total transfer of seed particles was slightly higher than water test results. The testing did show somewhat more batch-to-batch variation in the transfer of seed particles with the slurries in comparison to water. A comparison of batch transfers with the kaolin slurries that had Bingham consistencies (viscosities) that wernearly the same as the Newtonian glycerol/water mixtures showed that the kaolin slurries with Bingham yield stresses of 1.6 and 7 Pa gave better batch transfer of seed particles based on the total quantities transferred. Overall, the batch transfer testing results show that testing with water is conservative, since using a simulant with a yield stress and/or elevated viscosity always resulted in a better total transfer of solids.

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Small-Scale Melter Testing With Law Simulants To Assess The Impact Of Higher Temperature Melter Operations

About 50 million gallons of high-level mixed waste is currently in storage in underground tanks at The United States Department of Energy's (DOE's) Hanford site in the State of Washington. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will provide DOE's Office of River Protection (ORP) with a means of treating this waste by vitrification for subsequent disposal. The tank waste will be separated into low- and high-activity fractions, which will then be vitrified respectively into Immobilized Low Activity Waste (ILAW) and Immobilized High Level Waste (IHLW) products. The ILAW product will be disposed of in an engineered facility on the Hanford site while the IHL W product will be directed to the national deep geological disposal facility for high-level nuclear waste. The ILAW and IHLW products must meet a variety of requirements with respect to protection of the environment before they can be accepted for disposal.

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Single-faced GRAYQB™: a radiation mapping device

GrayQb{trademark} is a novel technology that has the potential to characterize radioactively contaminated areas such as hot cells, gloveboxes, small and large rooms, hallways, and waste tanks. The goal of GrayQb{trademark} is to speed the process of decontaminating these areas, which reduces worker exposures and promotes ALARA considerations. The device employs Phosphorous Storage Plate (PSP) technology as its primary detector material. PSPs, commonly used for medical applications and non-destructive testing, can be read using a commercially available scanner. The goal of GrayQb{trademark} technology is to locate, quantify, and identify the sources of contamination. The purpose of the work documented in this report was to better characterize the performance of GrayQb{trademark} in its ability to present overlay images of the PSP image and the associated visual image of the location being surveyed. The results presented in this report are overlay images identifying the location of hot spots in both controlled and field environments. The GrayQb{trademark} technology has been mainly tested in a controlled environment with known distances and source characteristics such as specific known radionuclides, dose rates, and strength. The original concept for the GrayQb{trademark} device involved utilizing the six faces of a cube configuration and was designed to be positioned in the center of a contaminated area for 3D mapping. A smaller single-faced GrayQb{trademark}, dubbed GrayQb SF, was designed for the purpose of conducting the characterization testing documented in this report. This lighter 2D version is ideal for applications where entry ports are too small for a deployment of the original GrayQb™ version or where only a single surface is of interest. The shape, size, and weight of these two designs have been carefully modeled to account for most limitations encountered in hot cells, gloveboxes, and contaminated areas. GrayQb{trademark} and GrayQb{trademark} SF share the same fundamental detection system design (e.g., pinhole and PSPs). Therefore, performance tests completed on the single face GrayQB in this report is also applicable to the six- faced GrayQB (e.g., ambient light sensitivity and PSP response). This report details the characterization of the GrayQb{trademark} SF in both an uncontrolled environment; specifically, the Savannah River Site (SRS) Plutonium Fuel Form Facility in Building 235-F (Metallurgical Building) and controlled testing at SRS’s Health Physics Instrument Calibration Facility and SRS’s R&D Engineering Imaging and Radiation Systems Building. In this report, the resulting images from the Calibration Facility were obtained by overlaying the PSP and visual images manually using ImageJ. The resulting images from the Building 235-F tests presented in this report were produced using ImageJ and applying response trends developed from controlled testing results. The GrayQb{trademark} technology has been developed in two main stages at Savannah River National Laboratory (SRNL): 1) the GrayQb{trademark} development was supported by SRNL’s Laboratory Directed Research and Development Program and 2) the GrayQb{trademark} SF development and its testing in Building 235-F were supported by the Office of Deactivation and Decommissioning and Facility Engineering (EM-13), U.S. Department of Energy – Office of Environmental Management.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Critical Skills at SRS – H-Tank Operator

This video series provides a first-hand look at critical skills positions at the Savannah River Site and required qualifications. Meet the people behind the jobs and learn what they do to support important SRS missions.

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Experimental Methods to Estimate Accumulated Solids in Nuclear Waste Tanks - 13313

The Department of Energy has a large number of nuclear waste tanks. It is important to know if fissionable materials can concentrate when waste is transferred from staging tanks prior to feeding waste treatment plants. Specifically, there is a concern that large, dense particles, e.g., plutonium containing, could accumulate in poorly mixed regions of a blend tank heel for tanks that employ mixing jet pumps. At the request of the DOE Hanford Tank Operations Contractor, Washington River Protection Solutions, the Engineering Development Laboratory of the Savannah River National Laboratory performed a scouting study in a 1/22-scale model of a waste tank to investigate this concern and to develop measurement techniques that could be applied in a more extensive study at a larger scale. Simulated waste tank solids and supernatant were charged to the test tank and rotating liquid jets were used to remove most of the solids. Then the volume and shape of the residual solids and the spatial concentration profiles for the surrogate for plutonium were measured. This paper discusses the overall test results, which indicated heavy solids only accumulate during the first few transfer cycles, along with the techniques and equipment designed and employed in the test. Those techniques include: - Magnetic particle separator to remove stainless steel solids, the plutonium surrogate from a flowing stream. - Magnetic wand used to manually remove stainless steel solids from samples and the tank heel. - Photographs were used to determine the volume and shape of the solids mounds by developing a composite of topographical areas. - Laser range finders to determine the volume and shape of the solids mounds. - Core sampler to determine the stainless steel solids distribution within the solids mounds. - Computer driven positioner that placed the laser range finders and the core sampler over solids mounds that accumulated on the bottom of a scaled staging tank in locations where jet velocities were low. These devices and techniques were very effective to estimate the movement, location, and concentrations of the solids representing plutonium and are expected to perform well at a larger scale. The operation of the techniques and their measurement accuracies will be discussed as well as the overall results of the accumulated solids test. (authors)

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The status of LILW disposal facility construction in Korea - 15158

In this paper, we discuss the experiences during the construction of the first LILW disposal facility in South Korea. In December 2005, the South Korean Government designated Gyeongju-city as a host city of Low- and Intermediate-Level Radioactive Waste (LILW) disposal site through local referendums held in regions whose local governments had applied to host disposal facility in accordance with the site selection procedures. The LILW disposal facility is being constructed in Bongil-ri, Yangbuk-myeon, Gyeongju. The official name of the disposal facility is called 'Wolsong Low and Intermediate Level Radioactive Waste Disposal Center (LILW Disposal Center)'. It can dispose of 800,000 drums of radioactive wastes in a site of 2,100,000 square meters. At the first stage, LILW repository of underground silo type with disposal capacity of 100,000 drums was completed by the end of June 2014. The Wolsong Low and Intermediate Level Radioactive Waste Disposal Center consists of surface facilities and underground facilities. The surface facilities include a receipt and inspection facility, an interim storage facility, a radioactive waste treatment building, and supporting facilities such as main control center, equipment and maintenance shop. The underground facilities consist of a construction tunnel for transport of construction equipment and materials, an operation tunnel for transport of radioactive waste, an access shaft for workers, and six silos for final disposal of radioactive waste. (authors)

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US EPA superfund radiation risk assessment update: new, revised, and upcoming tools - 15350

The U.S. EPA Superfund remedial program is finishing up a significant number of revisions to its guidance for the risk assessment process at radioactively contaminated Superfund sites. The six Preliminary Remediation Goal (PRG) and Dose Compliance Concentration (DCC) internet based calculators for risk and dose assessment at Superfund sites are being revised to reflect better science, revisions to existing exposure scenarios and new scenarios, and changes to match up more closely with the EPA chemical regional screening level calculator. A comprehensive set of revisions to the PRG calculator will be finished in November 2014. A revised version of the 1999 guidance document that provides an overview for the Superfund risk assessment process at radioactively contaminated sites, 'Radiation Risk Assessment At CERCLA Sites: Q and A' was issued on June 13, 2014 reflects Superfund recommended guidance and other technical documents issued over the past 15 years. Also on June 13, 2014 EPA issued a new tool-kit comprised of 22 fact sheets that was developed to help the general public understand more about EPA's risk assessment process used at radioactively contaminated Superfund sites. (authors)

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Analysis of erosion/corrosion data for high-level waste pipelines at Hanford - 15519

Washington River Protection Solutions (WRPS) has implemented a fitness-for-service program which will evaluate the degraded condition of the tank farm waste transfer system. The Tank Farms Waste Transfer System Fitness-for-Service Requirements and Recommendations, includes a requirement to inspect primary piping, encasements, and jumpers for corrosion/erosion. The 242-A Evaporator pump room was upgraded by adding instrumentation to the feed and return jumpers, prior to running the next campaign. As part of this upgrade, several jumpers were removed for disposal and a total of five jumpers were selected for ultrasonic thickness (UT) inspection. The jumpers selected were the following: 18-4, C-4 and 5, J-13A, 13-K, and 19-5. All of these jumpers will be removed permanently except for jumper 19-5 which will be reinstalled for further service. As part of this study, several jumpers from the AW-02E Feed Pit were also removed for disposal and two were selected for UT inspection. The jumpers selected were the 1-4 and B-2 which were packaged and sent to the 222-S Laboratory for UT assessments. This paper includes details of each jumper as well as the estimated remaining useful life (ERUL), if applicable, for its components based on the wall thinning measured in elbows and straight sections. The jumper's ultrasonic thickness measurements are plotted and trends are assessed based on the volume of fluid transferred or time of service. Variability in thickness is generally evaluated in two approaches, radially around the circumference of the pipe or component and longitudinally along the length. The paper also explains the analysis procedure for evaluating the different components. In particular, issues related to thinning from manufacturing different types of elbows (long radius, five diameter bend radius) are presented. Analysis techniques are developed for elbows that differentiate thinning from erosion/corrosion and manufacturing which is critical for this analysis. Results demonstrate that in most components, the thickness measurements are still greater than the manufactured nominal thickness. For the cases in which the thickness is below the nominal thickness, the ERUL well exceeds the life needed for servicing the Waste Treatment Plant. (authors)

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NDA Challenges during Cleanup of the Hanford 618-10 Burial Ground Trenches - 15542

The 618-10 Burial Ground on the Hanford Reservation was an early trench waste site used for the disposal of waste from the 300 Area at Hanford. Much of this waste consisted of residues from evaluations of failed fuel as well as various process and separations studies associated with the Department of Energy (DOE) mission which were performed in the 300 Area between March 1954 and September 1963. Much of this waste contained significant concentrations of fission products as well as Transuranic (TRU) radionuclides. Waste forms consist of drums and boxes which were disposed of in the trenches. Many of the drums are shielded containers, cemented sludge, or cemented debris. There are no identifying marks on the containers to correlate them to a specific waste stream. In addition there is some waste in the trenches which came from other locations outside of the 300 area. Cleanup of the 618-10 burial ground is covered under the Washington Closure Hanford (WCH) contract. Due to the proximity of the 618-10 site to the Columbia River, waste is being removed from the 618-10 site, stabilized, and disposed of into either the Environmental Restoration Disposal Facility (ERDF) if the containers are determined to be LLW or transferred to the Hanford Central Waste Complex (CWC) if the containers are suspected to be TRU waste. In order to determine the correct disposal path Non Destructive Assay (NDA) must be performed on all of the containers excavated from the trenches. The NDA of these containers is performed using germanium gamma spectroscopy with an In Situ Object Counting System (ISOCS) and with a neutron slab counter. Due to the challenges associated with these difficult waste streams, matrices, and lack of information to associate these containers with a particular process, the NDA program has had to develop a more robust process to ensure the correct disposition of the waste. (authors)

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Plan for resolution of technical issues at the Hanford waste treatment and immobilization plant project - 15596

The Waste Treatment and Immobilization Plant (WTP) Project has a number of first-of-a-kind technical challenges that have necessitated expert analysis and testing activities to resolve. Since 2006, the U.S. Department of Energy (DOE) chartered two major WTP technical reviews. Most recently, in 2012, a number of top scientists and engineers conducted a review focused on the design and operability of WTP black cell and hard-to-reach areas. To provide leadership and oversight for resolution of the technical issues, DOE assembled a Design Completion Team composed of personnel from the DOE Office of River Protection (ORP), DOE contractors, and the national laboratories. The Design Completion Team established five technical sub-teams to address specific focus areas. The work of these sub-teams has evolved such that the project is now focused on eight technical issues referred to as T1 to T8. The WTP contractor has developed, and DOE has approved, plans for resolution of these issues. These plans form the basis for the Pretreatment (PT) Facility engineering work that will be performed over the coming years. This work, along with a corresponding update to the PT Facility safety basis, will result in an authorization to resume the engineering work necessary to complete the PT Facility design. (authors)

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Qualification of the First ICS-3000 ION Chromatograph for use at the Defense Waste Processing Facility

The ICS-3000 Ion Chromatography (IC) system installed in 221-S M-13 has been qualified for use. The qualification was a head to head comparison of the ICS-3000 with the currently used DX-500 IC system. The crosscheck work included standards for instrument calibration and calibration verifications and standards for individual anion analysis, where the standards were traceable back to the National Institute of Standards and Technology (NIST). In addition the crosscheck work included the analysis of simulated Sludge Receipt Adjustment Tank (SRAT) Receipt, SRAT Product, and Slurry Mix Evaporator (SME) samples, along with radioactive Sludge Batch 5 material from the SRAT and SME tanks. Based upon the successful qualification of the ICS-3000 in M-13, it is recommended that this task proceed in developing the data to qualify, by a head to head comparison of the two ICS-3000 instruments, a second ICS-3000 to be installed in M-14. The Defense Waste Processing Facility (DWPF) requires the analysis of specific anions at various stages of its processing of high level waste (HLW). The anions of interest to the DWPF are fluoride, formate, chloride, nitrite, nitrate, sulfate, oxalate, and phosphate. The anion analysis is used to evaluate process chemistry including formic acid/nitric acid additions to establish optimum conditions for mercury stripping, reduction-oxidation (REDOX) chemistry for the melter, nitrite destruction, organic acid constituents, etc. The DWPF Laboratory (Lab) has been using Dionex DX-500 ion chromatography (IC) systems since 1998. The vendor informed DWPF in 2006 that the instruments would no longer be supported by service contracts after 2008. DWPF purchased three new ICS-3000 systems in September of 2006. The ICS-3000 instruments are (a) designed to be more stable using an eluent generator to make eluent, (b) require virtually no daily chemical handling by the analysts, (c) require less line breaks in the hood, and (d) generally require less maintenance due to the pump configuration only using water versus the current system where the pump uses various hydroxide concentrations. The ICS-3000 instruments also allow the DWPF to maintain current service contracts, which support routine preventive maintenance and emergency support for larger problems such as component failure. One of the three new systems was set up in the DWPF Lab trailers in January of 2007 to be used for the development of methods and procedures. This system will continue to be used for training, new method development and potential improvements to current methods. The qualification of the other two ICS-3000 instruments is to be a phased effort. This effort is to be supported by the Applied Computational Engineering and Statistical (ACES) group of the Savannah River National Laboratory (SRNL) as authorized by the Technical Task Request (TTR) and as directed by the corresponding Task Technical and Quality Assurance (TT&QA) plan. The installation of the first 'rad' system into the M-13 Lab module required modifications to both the Lab module and to the radiohood. The installation was completed in July 2008. The testing of this system was conducted as directed by the TT&QA plan. The purpose of this technical report is to provide a review of the data generated by these tests that will lead to the recommendation for the qualification of the M-13 ICS-3000 instrument. With the successful qualification of this first ICS-3000, plans will be developed for the installation of the second 'rad' system in the M-14 Lab module later in fiscal year 2009. When the second 'rad' ICS-3000 system is installed, the DX-500 systems will be removed and retired from service.

47 OTHER INSTRUMENTATION↗

The Pretreatment and Immobilization Test on The Greater Than Class C Spent Ion Exchange Resins from Taiwan Research Reactor - 17177

In this work, the experience on the removal of greater than Class C bead ion exchange resins from Taiwan Research Reactor spent fuel pool was reported. All of the spent ion exchange resins had been flushed, dewatered, and packed into lead-shield containers for 60 drums and pending for stabilization. The meta-kaolin type of geo-polymer was selected to immobilize the ion exchange resin. According to the 'Regulations on Final Disposal of Low Level Radioactive Waste and Safety Management of the Facilities' Act in Taiwan, the immobilization waste form shall meet seven testing items. Herein, several immobilize samples were prepared with the following amounts of ion exchange resin, 6, 12 and 18 wt.%. The compressive strength of the samples ranged from 2.9 to 18.2 MPa. The Cs and Sr ions were effectively trapped by ion exchange resins. After the resins incorporated with the geo-polymer matrix, the captured ions were expected difficult to leach out. However, it is found that at the higher contents of ion exchange resins in the immobilize samples exhibited poor water resistance due to the swelling effect. Several parameters such as the ratios of SiO{sub 2}/Na{sub 2}O and powder/solution also the blast furnace slag dosage were investigated to improve processibility and improve performance. (authors)

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High Waste Loading Glass Formulation Development for High-Zr HLW

This report describes work performed to develop and test new glass and feed formulations in order to increase waste loadings and glass melting rates for Hanford HLW with high concentrations of zirconium. Testing was designed to identify glass and melter feed formulations that optimize waste loading and waste processing rate while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts to evaluate the effects of waste loading on glass properties and the maximum waste loadings that can be achieved. The testing included an assessment of melt rate using a Vertical Gradient Furnace (VGF) system to develop new formulations with both high waste loadings and high melt rates. The results from crucible-scale testing supported subsequent DM100 tests designed to further assess the enhanced glass and feed formulations with respect to processing rate and product quality.

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COMSOL MULTIPHYSICS MODEL FOR DWPF CANISTER FILLING, REVISION 1

This revision is an extension of the COMSOL Multiphysics model previously developed and documented to simulate the temperatures of the glass during pouring a Defense Waste Processing Facility (DWPF) canister. In that report the COMSOL Multiphysics model used a lumped heat loss term derived from experimental thermocouple data based on a nominal pour rate of 228 lbs./hr. As such, the model developed using the lumped heat loss term had limited application without additional experimental data. Therefore, the COMSOL Multiphysics model was modified to simulate glass pouring and subsequent heat input which, replaced the heat loss term in the initial model. This new model allowed for changes in flow geometry based on pour rate as well as the ability to increase and decrease flow and stop and restart flow to simulate varying process conditions. A revised COMSOL Multiphysics model was developed to predict temperatures of the glass within DWPF canisters during filling and cooldown. The model simulations and experimental data were in good agreement. The largest temperature deviations were {approx} 40 C for the 87 inch thermocouple location at 3000 minutes and during the initial cool down at the 51 inch location occurring at approximately 600 minutes. Additionally, the model described in this report predicts the general temperature trends during filling and cooling as observed experimentally. The revised model incorporates a heat flow region corresponding to the glass pouring down the centerline of the canister. The geometry of this region is dependent on the flow rate of the glass and can therefore be used to see temperature variations for various pour rates. The equations used for this model were developed by comparing simulation output to experimental data from a single pour rate. Use of the model will predict temperature profiles for other pour rates but the accuracy of the simulations is unknown due to only a single flow rate comparison.

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Process and technological aspects of municipal solid waste gasification. A review

Highlights: Black-Right-Pointing-Pointer Critical assessment of the main commercially available MSW gasifiers. Black-Right-Pointing-Pointer Detailed discussion of the basic features of gasification process. Black-Right-Pointing-Pointer Description of configurations of gasification-based waste-to-energy units. Black-Right-Pointing-Pointer Environmental performance analysis, on the basis of independent sources data. - Abstract: The paper proposes a critical assessment of municipal solid waste gasification today, starting from basic aspects of the process (process types and steps, operating and performance parameters) and arriving to a comparative analysis of the reactors (fixed bed, fluidized bed, entrained bed, vertical shaft, moving grate furnace, rotary kiln, plasma reactor) as well as of the possible plant configurations (heat gasifier and power gasifier) and the environmental performances of the main commercially available gasifiers for municipal solid wastes. The analysis indicates that gasification is a technically viable option for the solid waste conversion, including residual waste from separate collection of municipal solid waste. It is able to meet existing emission limits and can have a remarkable effect on reduction of landfill disposal option.

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Licensing of the national repository for LILW waste in Bulgaria - 15609

Radioactive waste arises from the operation and decommissioning of the Bulgarian nuclear power plants and from other sources in research, medical applications and industry. The need for building the National Disposal Facility (NDF) is recognized by the Bulgarian government, and described and justified in the Environment Impact Assessment Decision. The Bulgarian State Enterprise for Radioactive Waste Management (SERAW) is endeavoring to build a near-surface repository for short-lived low and intermediate level radioactive waste (SL-LILW). The European Union finances the establishment of the NDF through the Kozloduy International Decommissioning Support Fund. The Bulgarian government has defined the NDF as a Project of National Importance. At the conclusion of an international tendering process, SERAW awarded the contract for development of the NDF design, including Intermediate Safety Analysis Report (ISAR), to the Consortium of Westinghouse Electric Spain S.A.U, DBE Technology GmbH of Germany, and ENRESA, the Spanish National Waste Management Agency. Two Bulgarian companies also participate in the project as subcontractors;KK-Project and EQE-Bulgaria. The project 'Technical Design and ISAR Preparation for the National Disposal Facility at Bulgaria' is identified within SERAW as 'R-Project 5'. Since obtaining SERAW's approval for the preferred repository design variant, R Project 5 has focused on the development of the Technical Design and on preparation of the ISAR, which demonstrates the safety and suitability of the proposed NDF design. The Consortium submitted to SERAW the Technical Design for revision on March 2013. SERAW's experts immediately started the review of the design. In addition, SERAW invited independent, highly qualified Bulgarian experts and scientists to participate in the review. In November 2013, SERAW approved the Technical Design. The Consortium submitted the Intermediate Safety Analysis Report to SERAW for review following a similar process in July 2014. Both documents were meanwhile handed over to the appropriate Bulgarian licensing authorities. SERAW's goal is to receive a license for construction of the NDF in early 2015. (authors)

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FX Hg Fogging Fixative Deployment for Mercury Vapor Suppression

Idaho National Laboratory’s (INL) FX Hg fixative solution was deployed at the Y-12 Complex in Oak Ridge, Tennessee to support disposal of mercury-contaminated metal debris. The fixative was dispensed via fogging. Fogging infiltrates non-line-of-sight areas improving fixative coating on complex geometries such as debris piles. FX Hg is the mercury vapor controlling derivative of INL’s FX2 fixative. FX2 was jointly developed with the National Nuclear Laboratory of the United Kingdom, and INL developed the FX Hg derivative to suppress mercury vapor generation. The Y-12 deployment was performed in concert with cleanup contractor UCOR. A dumpster filled with debris was fogged with FX Hg. The material was acceptable as municipal landfill waste, save for the mercury vapor levels measured. If the mercury vapor generation rate could be sufficiently reduced, disposal costs for this waste would be dramatically reduced. FX Hg had previously proven effective at significantly reducing mercury vapor generation rates in bench scale testing at INL. This deployment was the first field scale deployment of the method. Efficacy results were underwhelming, but interpretation of the outcome is complicated by a paucity of hard data. Methods for improving data capture and analysis are analyzed. Operational difficulties associated with scaling the process up and process improvements for future deployments are discussed.

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