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DOE OSTI2015

Lessons learned in the hot cell at the transuranic waste processing center - 15336

The hot cell at the Transuranic Waste Processing Center (TWPC) has been in operation for seven years. Its mission is to empty, process, and repackage the contents Remote Handled (RH) concrete waste casks. The sequence of operation for processing the RH casks includes using a remotely operated manipulator (manipulator) to remove the cask contents and deliver them to a work table where operators process the waste into RH-TRU, CH-TRU, LLW drums using Master Slave Manipulators (MSMs). The manipulator is the only piece of equipment that can remove material from the cask and represents a single point failure that can shut down hot cell processing should it fail. During the first six years of operation the project processing goals were constantly in jeopardy due to the high frequency and extent of manipulator failures that forced unplanned down time in the hot cell to facilitate repairs. In 2013 a hot cell maintenance outage was planned and executed to improve the operability and reduce the equipment failure rate. A new remotely operated manipulator, purchased as a spare before the outage, was installed in the hot cell during the outage. Prior to installing the manipulator into the hot cell several major and minor modifications were made to the manipulator to incorporate improvements manipulator reliability and maintainability based upon operating and maintenance lessons learned from the previous years of operation. Additionally a maintenance glovebox was added to the hot cell for servicing the removable manipulator components. (authors)

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DOE OSTI2015

Treatment of contaminated non-iron metals for clearance - 15361

Iron based metals and concrete make up most of the structures in a nuclear power plants and most other nuclear installations. During the last few decades, a lot of effort has been put into the treatment of steel in order to allow recycling of the metal and to reduce the volume for final disposal. In addition to the iron based metals, there are also other metals in significant amounts that can be treated for recycling and waste minimisation. Aluminium, lead, titanium, brass and copper have all been treated by Studsvik in the metals treatment facility in Sweden over the years. Some of these metals like lead and aluminium causes certain concerns (for reasons not related to their radioactivity) if they were to be disposed as radioactive waste. Other as copper and titanium has a high economic value if recycled outside the nuclear industry. Studsvik experience from treatment of non-iron metals is reported in this paper. (authors)

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DOE OSTI2015

The cementitious barriers partnership experimental programs and software advancing DOE's waste disposal/tank closure efforts - 15436

The U.S. Department of Energy Environmental Management (DOE-EM) Office of Tank Waste Management sponsored Cementitious Barriers Partnership (CBP) is chartered with providing the technical basis for implementing cement-based waste forms and radioactive waste containment structures for long-term disposal. DOE needs in this area include the following to support progress in final treatment and disposal of legacy waste and closure of High-Level Waste (HLW) tanks in the DOE complex: - Long-term performance predictions and - Flow sheet development and flow sheet enhancements - Conceptual designs for new disposal facilities The DOE-EM Cementitious Barriers Partnership is producing software and experimental programs resulting in new methods and data needed for end-users involved with environmental cleanup and waste disposal. Both the modeling tools and the experimental data have already benefited the DOE sites in the areas of performance assessments by increasing confidence backed up with modeling support, leaching methods, and transport properties developed for actual DOE materials. In 2014, the CBP Partnership released the CBP Software Toolbox - 'Version 2.0' which provides concrete degradation models for: 1) sulfate attack, 2) carbonation, 3) chloride initiated rebar corrosion, and includes constituent leaching. These models are applicable and can be used by both DOE and the Nuclear Regulatory Commission (NRC) for service life and long-term performance evaluations and predictions of nuclear and radioactive waste containment structures across the DOE complex (Figure 1) including Future SRS salt-stone and HLW tank performance assessments and special analyses - Hanford site HLW tank closure projects and other projects in which cementitious barriers are required - The Advanced Simulation Capability for Environmental Management (ASCEM) project which requires source terms from cementitious containment structures as input to their flow simulations - Regulatory reviews of DOE performance assessments - Nuclear Regulatory Commission reviews of commercial nuclear power plant (NPP) structures which are part of the overall US Energy Security program to extend the service life of NPPs In addition, the CBP experimental programs have had a significant impact on the DOE complex by providing specific data unique to DOE sodium salt wastes at Hanford and SRS which are not readily available in the literature. Two recent experimental programs on cementitious phase characterization and on technetium (Tc) mobility have provided significant conclusions as summarized below: Recent mineralogy characterization discussed in this paper illustrates that sodium salt waste form matrices are somewhat similar to but not the same as those found in blended cement matrices which to date have been used in long-term thermodynamic modeling and contaminant sequestration as a first approximation. Utilizing the CBP generated data in long-term performance predictions provides for a more defensible technical basis in performance evaluations. - In addition, recent experimental studies related to technetium mobility indicate that conventional leaching protocols may not be conservative for direct disposal of Tc- containing waste forms in vadose zone environments. These results have the potential to influence the current Hanford supplemental waste treatment flow sheet and disposal conceptual design. (authors)

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DOE OSTI2015

Achievements and process improvements in processing transuranic waste at the TRU Waste Processing Center - 15488

Wastren Advantage, Inc., (WAI) is the DOE Prime contractor for the TRU Waste Processing Center (TWPC) located in Oak Ridge, TN. This paper highlights the TWPCs specific achievements and process improvements associated with both Contact (CH) and Remote Handled (RH), Transuranic (TRU) waste receipt processing and repackaging for disposal. The five key improvement areas and achievements include: - The TWPC installed, tested and operated successfully the Drum Venting System (DVS) to vent over-packed CH-TRU waste drums in several configurations including 55-gallon drums over-packed in 79 or 85-gallon drums. - A new Remote Drum Opener (RDO) system was designed, tested and placed in service to vent drums discovered in overpack boxes in the Box Breakdown Area and to vent overpack configurations that could not be handled in the standard DVS (e.g., 110 gallon overpacks and multiple drum configurations). - The Cask Processing Enclosure (CPE) approach to repackaging RH-TRU waste contained in concrete casks was developed and implemented - Hot Cell outage to incorporate planned equipment modifications and improvements to maintain processing rates. - To improve certification and segregation of CH-TRU from drums that meet Low Level Waste criteria for disposal at Nevada the TWPC added additional equipment/facilities for Non-Destructive Assay (NDA) and Real Time Radiography (RTR). (authors)

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DOE OSTI2015

Progress of the enhanced Hanford single shell tank (SST) integrity project - 15497

To improve understanding of the single-shell tanks (SSTs) integrity, Washington River Protection Solutions, LLC (WRPS), the US DOE Hanford Site tank contractor, developed an enhanced Single-Shell Tank Integrity Project (SSTIP) in 2009. An expert panel on SST integrity, consisting of various subject matter experts in industry and academia, was created to provide recommendations supporting the development of the project. This panel developed 33 recommendations in four main areas of interest: structural integrity, liner degradation, leak integrity and prevention, and mitigation of contamination migration. In late 2010, seventeen of these recommendations were used to develop the basis for the M-45-10-1 Change Package for the Hanford Federal Agreement and Compliance Order, which is also known as the Tri-Party Agreement. The change package identified two phases of work for SST integrity. The initial phase was focused on efforts to envelope the integrity of the tanks. The initial phase was divided into two primary areas of investigation: structural integrity and leak integrity. If necessary, based on the outcome from the initial work, second phase would be focused on further definition of the integrity of the tanks and liners. Combined these two phases are designed to support the formal integrity assessment of the Hanford SSTs in 2018 by an Independent Qualified Registered Engineer. As the initial phase of this work completes in early 2015, this paper summarizes the project formation and status the work of the past few years associated the implementation of the Phase one recommendations. Work in the initial phase to further define the DOE's understanding of the structural integrity SSTs involved preparing a modern Analysis of Record (AOR) using finite element analysis. Structural analyses of the SSTs have been conducted since 1957, but these analyses used analog calculations, less rigorous models, or focused on individual structures. As such, an integrated understanding of all of the SSTs has not been developed to modern expectations. In support of this effort, other activities addressed the visual inspection of tank internal conditions and the collection of concrete core samples from the tanks, including a full height tank sidewall core, for analysis of current mechanics properties. The work on the liner leak integrity has examined leaks from 25 tanks with known liner failures. Individual leak assessments were developed for each tank to identify the leak cause and location and estimate historic leak rates. A common cause/failure analysis study was performed to take data from individual tanks to look for trends in the causes of failure. A separate activity is being conducted to examine the propensity for corrosion in select SSTs with aggressive waste layers. (authors)

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DOE OSTI2015

Mock-Up activities for the extraction of samples from a salt-stone disposal unit - 15502

Mock-up testing with respect to the core-drilling of simulant salt-stone grout was conducted to support the future retrieval of field-emplaced salt-stone samples from salt-stone Disposal Unit (SDU) Cell 2A. The primary objectives of extracting samples from SDU Cell 2A are to enable measurement of field-emplaced samples with respect to properties that influence long-term performance, such as saturated hydraulic conductivity, and to compare the measured properties for field-emplaced and comparable samples prepared in the laboratory. Initial mock-up activities (Phases 1 and 2) were conducted in Fiscal Years (FY) 2013 and 2014 in order to demonstrate the viability of standard core-drilling techniques that are typically employed for sampling concrete structures. The primary requirement for the core-drilling technique was that it should minimize the potential of physical or chemical sample changes that would artificially degrade key grout properties. Mock-up activities were aimed at optimizing the techniques for core-drilling, and subsequent sample extraction, using a set-up that mimicked, to the extent practical, both the physical characteristics, and operational complexities, of core-drilling salt-stone grout from the roof of the SDU. Samples extracted from the grout monolith were subsequently characterized with respect to saturated hydraulic conductivity and compressive strength, and compared to data for cast samples prepared using the same simulant salt-stone material. The comparison of these properties for the two sample types was expected to highlight potential flaws (e.g., micro-cracking) introduced by the core-drilling process. While the data for saturated hydraulic conductivity were virtually identical (approximately 2 E-09 cm/sec), the compressive strength of the core-drilled samples (approximately 3.5 MPa (500 psi)) was on average half that measured for the cast samples. It was suspected that increased water saturation and surface scoring of the core-drilled samples might account for the disparate compressive Strength data. During Phases 1 and 2 it was determined that cores could be successfully drilled and extracted to a depth of approximately 0.9 meters (3 feet) below the grout surface. Phase 3 (FY 2105) has been initiated and is primarily directed at demonstrating the ability to core-drill through approximately 1.8 meters (6 feet) of overlying grout to retrieve the desired sample from the SDU. The original drill depth of 0.6 - 0.9 meters (2-3 feet) was based on an SDU fill height of 5.6 meters (18.5 feet) and sample location of approximately 4.9 meters (16 feet); recently the allowable fill height in the SDUs has been increased to 6.5 meters (21.25 feet); hence the increased depth of overlying grout above the samples to be retrieved. (authors)

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DOE OSTI2015

Technetium oxidation in slag-based sodium salt waste forms exposed to water and moist Hanford soil - 15514

Several U.S. DOE sites use or plan to use waste forms and/or concrete containment structures for radioactive waste disposal that are designed to have a chemically reducing environment to immobilize selected contaminants such as Tc(VII)O{sub 4}{sup -} and Cr(VI)O{sub 4}{sup 2-}. These waste forms and containment structures are typically deployed in near surface unsaturated oxidizing environments. Consequently, the effect of exposure to air (oxygen) and water containing dissolved oxygen during production, during the period of institutional control, and over the long term period of performance is important for predicting the speciation and mobility of the redox sensitive radioactive and stable contaminants. Both the SRS and Hanford waste streams contain soluble technetium which may require stabilization to meet disposal requirements. Technetium stabilization is a difficult problem because: 1) Tc is soluble and very mobile in the oxidized form typical of near surface environments, and 2) Tc-99 is a long-lived isotope with a half-life of 2.1 E+05 years which places demanding requirements on the engineered barriers and environment to meet current regulatory disposal requirements. A depth-discrete sampling and leaching method approach for measuring contaminant oxidation rate (effective contaminant specific oxidation rate) was used in this study. The method was modified by coating all sides of a cylindrical sample with an impermeable epoxy and cutting a fresh surface 2 to 2.5 cm from the original top surface eliminates sample inhomogeneity as the result of settling as a reason from observed results and provides 1-D soluble ion transport and gas transport information. Soluble Tc was leached from all of the depth-discrete sub-samples from both Tc2-9 and Tc2-10 which strongly suggests that oxygen was present in the entire length of both samples. About 24 mass percent of the Tc in the original sample, was leached (soluble) from sub-samples between 0.8 and 46 mm below the exposed surface of Tc2-9 (exposed to Hanford sediment). The same percent (24%) was leached from the sub-samples between 0.8 and 11 mm below the exposed surface of Tc2-10 (exposed to DI water). This suggests that the rate of oxygen migration into the sample exposed to soil was faster than the rate of migration into the sample exposed to water which is consistent with the more rapid transport of ions through a gas phase as compared to a liquid phase. It is assumed that moisture in the Hanford sediment was not sufficient to completely block the surface pores with respect to gas transport across the soil-waste form boundary. Based on nitrate leaching results for the depth- discrete sub-samples, regions depleted in nitrates were identified from the top surfaces to 9.5 and 3 mm into samples Tc2-9 (exposed to moist Hanford sediment) and Tc2-10 (DI water). Low mass fractions of nitrate were leached from these depth-discrete samples compared to samples further from the exposed surface presumably because a significant portion of the nitrate had already migrated into the soil or water, respectively. Depth-discrete sub-sample leaching results for Na can be interpreted in the same way over the same regions in the two samples tested. In conclusion, leaching monolithic porous cementitious waste forms in water appears to be conservative for non-redox sensitive contaminants. However, leaching data obtained under saturated exposure conditions do not appear to be conservative for redox sensitive contaminants which are easily oxidized. Leaching crushed samples in water still seems to be a conservative approach to estimating the concentrations of soluble contaminants in a waste form. (authors)

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DOE OSTI2015

An evaluation of nearest neighbor averaging's improvements to the measurement sensitivity of wide-area radiological scanning instruments - 15560

Through the Sandia National Laboratories (SNL) New Mexico Small Business Assistance (NMSBA) program, Sandia worked with the Environmental Restoration Group (ERG) Inc. to verify and validate a novel algorithm used to determine the scanning Critical Level (Lc) and Minimum Detectable Concentration (MDC) (or Minimum Detectable Areal Activity) for the 102F scanning system. Through the use of Monte Carlo statistical simulations the algorithm mathematically demonstrates accuracy in determining the Lc and MDC when a nearest-neighbor averaging (NNA) technique was used. To empirically validate this approach, SNL prepared several spiked sources and ran a test with the ERG 102F instrument on a bare concrete floor known to have no radiological contamination other than background naturally occurring radioactive material (NORM). The tests conclude that the NNA technique increases the sensitivity (decreases the Lc and MDC) for high-density data maps that are obtained by scanning radiological survey instruments. It is further described in Sandia Laboratories document number SAND2014-16522. (authors)

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DOE OSTI2015

FX2 advanced fogging technology - 15711

Problem statement: The need for an advanced technology that can better address potential airborne contaminants at DOE Savannah River Site 235-F. Current methods of knocking down or fixing this contamination are labor intensive, costly, and sometimes ineffective. Thus, to make exposure ALARA, the advanced fogging effort is being tested using Idaho National Laboratory's FX2 fixative agent. The FX2 fogging agent is a proprietary mixture of water, latex paint (LTX), glycerin (GLY) and sodium lauryl sulfate (SLS). Initial development began in 2006 when a DOE Small Business Innovative Research (SBIR) grant resulted in the development of a conceptual advanced capture coating fog system. This coating is shown to be excellent at reducing airborne contamination and 'fixing' particulates in place as demonstrated by Idaho National Lab. Controlling Potential Airborne Particulates: Testing capacity to knockdown airborne particulates. Evaluating ability to fix to different surfaces (e.g., glass, wood, steel, concrete, brick). Line of sight as well as non-line of sight (shadowed) locations will be tested for coverage. Adhesiveness and Coverage of Surface Area: Contact area will be placed under air pressure to attempt dislodging of dried mixture. Particle analysis utilizing ImageJ will be conducted to determine quantifiable coverage. Ignitability: Observation of the FX2's reactivity to flame and heat sources as it is being applied. Results will provide useful information for the prevention of potential hazards. Determine Shielding Properties Against Alpha Sources: Determine the shielding abilities of the FX2. Fogging agent against an alpha particle emitting point source. Evaluating Critical Properties: These properties will be evaluated to better characterize the FX2 and allow for optimal understanding of results: ASTM E1354 - Burn Rate; NFPA 296-Flashover; ASTM D2196 - Viscosity; ASTM D1331 - Surface Tension; ASTM D792 - Density. Testing Innovations: Developed by the National Institutes of Health, ImageJ is a public domain image processing program that allows for the analysis that this effort requires. Coverage: Through contrast analysis, ImageJ can be used to determine the coverage of the agent upon any of the testing surfaces. Correlations: With the feature of coverage evaluation, a relationship, and thus a graphical representation, can be made between surface coverage and radiation shielding. Future work: Engineering and manufacturing of mobile robotic fogging device. ASTM WK5394: establishing performance specifications for stabilizers to minimize migration of dispersible radioactive contamination.

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DOE OSTI2015

Hanford building 324 REC hot cell decommissioning approach - 15484

The 324 Building is a facility located at the Hanford Site and is undergoing decontamination and demolition (D and D) activities by Washington Closure Hanford (WCH). In October 1986, an in-cell leak of approximately 1.3 million curies occurred with a substantial fraction being lost into the subsurface soil beneath the floor of a large hot cell complex. Sampling under the hot cell floor revealed hot spots in the soil and cobble reading in excess of 10,000 R. In January of 2014, WCH contracted an AREVA lead team with Kurion as primary design agent to remove the B-Cell floor and highly contaminated underlying soils. The objective of this project is to remove the contaminated debris and soil and repackage them into the existing adjacent hot cells for final disposition as monoliths. The approach to this project uses an innovate combination of existing cell equipment, commercial equipment, and special tools to perform a series of completely remote operations. Also, a novel waste handling approach allows all the highly radioactive waste to remain within the existing cells. The remote handling approach to this project is to use customized off-the-shelf excavators which are mounted and operated remotely inside the cell. These can be moved and replaced remotely with a -Plug-and-Play- connection method. All the cables and utilities for the excavator and tools are managed through this cell connection with connections in a worker-accessible gallery space. The remote excavators are wall-mounted, which frees up the floor for the excavator work. The design also includes a transfer barrier which allows the B-Cell door to remain open during waste transfers while preventing spread of contamination between cells. The project balances the use of existing cell equipment such as cranes and shield doors, with the use of new remote equipment Existing equipment will be removed from the floor of the hot cell to adjacent cells. The concrete floor and stainless steel liner will then be remotely cut and removed, revealing the contaminated soil beneath the cell. The contaminated soil will be mixed with grout in the cell using the excavators and then placed in rigid bins. These bins will then be placed in the adjacent cells using existing overhead cranes. The waste containers allow optimal utilization of the limited adjacent hot cell space and improve the characteristics of the final waste monolith package. The system will use each bin once which significantly decreases the risk of contamination spread. The project includes a full-scale mockup facility in which all installation and operational activities will be tested and proven prior to hot cell construction activities. The mockup facility will also be used for operator training before actual hot work starts. This facility will remain available for use during operations with a full set of spares for redundancy and to support operations modeling during construction and remediation activities. (authors)

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DOE OSTI2016

Nuclide Release Pathways in and Around a Trench Repository for Radioactive Waste - 16014

An effective model for a safety assessment of a conceptual trench repository system, in which low-and intermediate level radioactive waste (LILW) is disposed of has been developed. The repository system is assumed to be located in the surface area of a mountain hill near the seashore. An easy-to-use computer program based on this model, developed as a GoldSim template, has been extended based on the previous one in consideration of the rather possible complex pathways in and around the trench for its performance assessment, which is able to evaluate a nuclide release from the repository and farther transport into the geosphere and biosphere under various normal, disruptive natural, and manmade events, and scenarios that can occur after a failure of waste drums with associated uncertainty. To demonstrate the nuclide release and transport behavior through these various possible pathways in the near- and far-fields of the repository system under certain scenarios, some illustrative evaluations and a simple sensitivity investigation with C-14 for the concrete barriers of the trench were made. (authors)

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DOE OSTI2016

Advances in Decommissioning Using Laser Cutting - 16035

This paper describes advancements made in laser cutting in the UK Collaborative Project 'LaserSnake2'. In particular it will focus on the use of special laser cutting heads and optics to enhance cutting performance, whilst at the same time, reducing the effects of the residual laser beam passing through the cut. The benefits of using lasers for applications in decommissioning include the high speeds available, the tolerance of the process, the lightness of the cutting head, the lack of a reaction force with the part being cut and the ease of automation of the laser cutting operation. Laser cutting, however, is a thermal process and a potential detriment, is that the residual laser beam passing through the kerf, might damage or indeed set fire to, something positioned behind the part being cut. This paper describes the use of a diffractive optical element in the laser beam forming optics, designed to extend the depth of focus of the system, without increasing the physical focal length of the optic. In this way, for cutting thick materials, the goal is to achieve the cutting performance of a long focal length lens, with the beam divergence of a short focal length lens. A design of diffractive optical element is presented, which when used with a laser power of 5 kW, in conjunction with a set of conventional beam forming optics, including a 250 mm focal length focusing lens, is able to cut 40 mm thick C-Mn steel plate at 50 mm/min, while varying the nozzle tip to plate stand-off distance by 100 mm. For the same laser power in the focusing beam, this performance can be equaled by removing the diffractive optic and 250 mm lens and replacing these with just a 500 mm lens. However, when using the diffractive optical element, the power density in the residual beam has been reduced by a factor four when compared to that from the optical arrangement using the 500 mm lens. As well as the cutting performance, the effect of the two different power density residual beams on concrete will be presented. (authors)

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DOE OSTI2016

The Enhanced Sealing Project, Monitoring a Full-Scale Shaft Seal at Canada's Underground Research Laboratory (2009-2015) - 16070

A full-scale shaft seal was designed and installed in the 5-m-diameter main shaft at Canada's Underground Research Laboratory (URL) in 2009. Located where the main shaft intersects a water-bearing fracture zone at a depth of approximately 273 m, a 6-m-thick clay component was sandwiched between two 3-m-thick concrete components. The clay component was 40:60 mixture of bentonite and quartz/feldspar sand and spans approximately 1 m beyond the maximum identified vertical extent of the fracture zone. Its purpose is to limit the mixing of saline groundwater from the deeper region with the fresher, near-surface groundwater. Its construction was part of Canada's Nuclear Legacy Liabilities Program (NLLP). The 1. phase of the Enhanced Sealing Project (ESP), a jointly-funded Research, Development and Demonstration (RD and D) project was started in 2009 to monitor the saturation of the shaft seal until the end of 2013. The 2. phase, extended monitoring of the ESP is supported by Posiva (Finland), Andra (France) and CNL (Canada) and this will run until the end of 2016. The slow evolution of the seal and its surroundings and complete removal of the surface facilities at the URL site in 2014, has meant that the ESP monitoring has evolved to a stand-alone, lower frequency data collection program. Information gained from the ESP monitoring is applicable to support the development of sealing technology to safely close a deep geological repository for the permanent disposal of radioactive wastes. (authors)

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DOE OSTI2016

Selection and Validation of Alternate Slag Sources for Saltstone Production - 16145

To stabilize low activity salt solution produced at the Savannah River Site (SRS) the salt waste is mixed with a combination of fly ash (FA), ground granulated blast furnace slag (GGBFS), and ordinary Portland cement (OPC) to form a grout termed saltstone. Upon curing, the low activity waste is encapsulated within the hardened cementitious matrix. While the sources of the dry feed components have remained unchanged for several years, SRS was recently notified that its domestic source of GGBFS would no longer be available due to the shutdown of the steel making plant that produced the ferrous slag as a by-product. This shut-down follows a trend in the reduction of domestic ferrous slag sources. Alternate GGBFS vendors were sought and a program developed to validate the alternate sources for use in saltstone. GGBFS is primarily utilized in the construction industry as a replacement for OPC and the requirements of a GGBFS source, based on an American Society for Testing and Materials (ASTM) standard, are predominantly limited to demonstrating that the GGBFS replacement will provide equivalent strength in comparison to the baseline OPC-based concrete. This affords the construction industry significant flexibility with respect to the GGBFS source that can be used. In contrast, GGBFS has a role of significantly more importance when utilized in saltstone. In particular, the reducing blast furnace environment under which GGBFS is produced yields reducing species, such as sulfide (S{sup 2-}) and ferrous iron (Fe{sup 2+}), which are believed to facilitate the long-term immobilization of redox sensitive contaminants (e.g. technetium-99 ({sup 99}Tc)). In addition, a change in GGBFS source has the potential to impact the rheological properties and the reactive heat generation of the grout, which may adversely affect the ability to pump the grout to a Saltstone Disposal Unit (SDU) and the temperatures attained within an SDU, respectively. As such, alternate GGBFS sources for the processing of saltstone require significantly more scrutiny than is prescribed by the ASTM standard. Four alternate GGBFS sources were evaluated. All four sources met the ASTM requirements, only one source was considered viable for use in saltstone based on preliminary evaluation. The other three sources will require a more intensive assessment to determine their potential applicability. (authors)

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DOE OSTI2016

Managing Low-Level Waste at Dounreay - 16269

Dounreay Site Restoration Limited (DSRL) is the site licence company responsible for the clean-up and demolition of Britain's former centre of fast reactor research and development. The Dounreay site requires facilities to dispose of up to 175,000 cubic metres of low level waste (LLW) that has been and will continue to be generated during the decommissioning and restoration of the site. In May 2014, construction of two new vaults for the disposal of LLW at Dounreay was completed, including the necessary roads, infrastructures and services, followed by an encapsulation plant in September 2014. The vaults are located immediately to the northeast of the Dounreay licensed site, on NDA-owned land. The facilities are for disposal of the solid LLW generated at Dounreay and the adjacent Vulcan site only. In 2004, following consultation with stakeholders and members of the public, Dounreay issued the 'Best Practicable Environmental Option (BPEO) Report' on the management of Dounreay LLW. This report assessed potential management options for LLW against a range of criteria, including technical, environmental, cost, health and safety issues. The study was conducted in line with best practice and involved consultation with a range of people, including the local community and organisations with an interest in radioactive waste management. The recommendations and conclusions from the study provided the basis for the 'Dounreay Solid LLW Overall Strategy', which was published in March 2005. A fundamental component of this strategy was the development of new below-ground disposal vaults for LLW at Dounreay. Siting the vaults on land at Dounreay avoids any need to transport the LLW away from Dounreay on public roads. Disposal at Dounreay therefore satisfies the proximity principle of managing the waste at source. Disposal at Dounreay is also consistent with UK Government Policy on LLW management, the Scottish Government's Radioactive Waste Management Policy and the Nuclear Decommissioning Authority's (NDA's) strategy on management of LLW from the UK nuclear industry. A planning application was initially lodged with the Highland Council in June 2006. The period for the determination of this application was extended to allow the Scottish Environment Protection Agency (SEPA) sufficient time to assess fully the supporting Environmental Safety Case (ESC). The planning application was re-activated in May 2008, and planning permission was granted by Highland Council in April 2009. The planning permission included 26 conditions on DSRL, covering a wide range of environmental issues such as construction noise/dust, traffic, visual impact and also a condition to develop a community fund aimed at socio-economic development of the area around the Dounreay site. After undertaking the necessary detailed design work, construction of the phase one vaults commenced in November 2011 by principal contractor, Graham Construction. Work was undertaken on both vaults simultaneously, excavating around 243,000 m{sup 3} of rock. When excavation was complete, steel reinforcement was placed and concrete poured to construct the floor slabs. This was followed by the construction of the reinforced walls and then the installation of the steel support structure and materials that form the roof. Mechanical and electrical installation then followed, in parallel with the construction of the vault apron, ramps and access roads. A pumping system and control building was also constructed to remove groundwater from around the vaults to allow the vaults to be kept dry while they are being filled with waste. A layer of excavated material, approximately 3 m thick, was placed across the existing ground surface between the vaults and the coast, to ensure the water table lies well below the new ground surface after closure of the vaults, and that any future trace contaminated groundwater flows out to the sea rather than to the soil zone. This layer is referred to in project documents as the 'enhanced geosphere'. After disposals are completed and the vaults are closed, this layer will be joined to the capping over the vaults so that the ensemble merges visually with the surrounding landscape. The original vegetation and soil mix will be returned as far as is practicable. An encapsulation plant has been constructed to fill the LLW containers with grout prior to disposal in the vault, similar to the process used at the national Low-Level Waste Repository facility (LLWR), near Drigg in Cumbria. The plant has the capacity to grout up to four containers per day. The grout fills voids in the container and helps provide chemical conditions in the vault that limit the release of radionuclides in the far future, after the vault has been penetrated by groundwater. On April 24 2015, active commissioning began at the Encapsulation Plant and LLW Vaults. Half-height ISO containers packed with low level waste are currently being filled with grout and disposed of in the LLW Vault. Also packaged Demolition Low Level Waste (DLLW) is being processed and disposed of in the Demolition Low Level Waste Vault. To date over 2500 packages of DLLW and 70 containers LLW have been disposed of. Completion of construction and the start of operating the Phase One vaults and encapsulation plant are fundamental to the successful and efficient decommissioning of the Dounreay site, upon which many millions of pounds of investment and hundreds of jobs rely.

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DOE OSTI2016

Obtaining a Special Administrative License Exemption to Ship the West Valley Melter Box - 16327

In order to ship the West Valley Melter Package (WVMP) package to support its disposal, the U.S. Department of Energy (DOE) requested a one-time shipment license exemption from the U.S. Nuclear Regulatory Commission (NRC). Based on content, the previously constructed WVMP, containing residual borosilicate glass and Low Density Cellular Concrete was shown to require a Type B Package. Therefore, Title 10 of the United States Code of Federal Regulations, Part 71 (10 CFR 71) required that radioactive releases under Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) would remain less than 1 E-7 A{sub 2}/hour and 1 E-1 A{sub 2}/week, respectively. Although the NRC normal method for demonstrating conformance to the 10CFR71 requirement is through performance of leak rate testing per American National Standards Institute (ANSI) standard N14.5, the WVMP does not contain a pressure vessel; therefore, performance of leak rate testing was not applicable/possible. Instead, an alternate method, adapted from DOE Handbook 3010 was used to show that the maximum allowed 10 CFR 71 radionuclide release rates for NCT and HAC would not be exceeded. Summarizing the effort, this paper provides the technical synopsis for requesting the exemption. (authors)

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DOE OSTI2016

Development of the Licensing Procedure and Regulatory Framework for the Spent Fuel Storage facility in Korea - 16423

The Public Engagement Commission on SNF management (PECOS) published the recommendation report on the spent fuel management of Korea in June, 2015. Based on this report, the future national policy for spent fuel management of Korea has been established, taking into consideration the national/international trends on policy and technology development. Regardless of the recommendation report of PECOS, Spent fuel storage casks are necessary in Korea. Therefore, various types of spent fuel storage casks are now being developed by Korea vendors. For example, Concrete storage casks and dual purpose metal casks are now under development by the KORAD (Korea Radioactive waste agency) for future use in the interim storage facility. Unfortunately, there is no independent licensing procedure for spent fuel storage casks in Nuclear Safety Act (NSA) because is the casks are considered an integral part of the interim storage facility safety equipment. Furthermore, the definition of radioactive waste in the NSA was changed recently. Now, spent fuel is separated from the definition of 'radioactive waste'. Spent fuel can only become radioactive waste by the decision through the meeting of the relative ministers of Korea government. For that reason, the licensing procedure for interim spent fuel storage facilities was separated from the current 'disposal facility, etc.' in accordance with the article 63 of NSA. The aim of this study is to develop the licensing procedure for the storage cask, and also to develop a revised draft of interim storage facility procedure within the NSA. Once this independent licensing procedure on the spent fuel storage cask is introduced to the NSA, it is expected that developers or operators can develop and commercialize the storage cask in Korea. In this study, an independent licensing procedure for the spent fuel storage facility was developed including the storage cask licensing procedure for the NSA. The procedure was composed of the permit for construction and operation of interim storage facility, criteria for permit, inspection criteria, periodic safety review, and decommissioning criteria, etc. The detailed contents of these requirements were also developed to introduce the regulatory framework for the interim storage facility. Particularly, periodic safety review and decommissioning criteria were developed in accordance with the recommendation of IRRS (International Regulatory Review Service) of IAEA. The developed licensing procedure for spent fuel storage facility will be introduced in NSA in response to the revised definition of NSA and the recommendation report of PECOS. (authors)

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DOE OSTI2016

Use of Multiple Remote Systems for Calcine Retrieval Operations at Idaho - 16592

There are approximately 5000 m{sup 3} of radioactive calcine material stored in six Calcine Solids Storage Facilities (CSSFs) at the DOE Idaho site. Each CSSF contains from three to twelve stainless steel tanks surrounded by a concrete vault. In accordance with the Idaho Settlement Agreement (ISA), this waste has to be retrieved and immobilized by December 31 2035 although no final decision has yet been made on how the calcine wastes are going to be retrieved from the CSSF's. One of the primary issues with the retrieval is that, in accordance with the ISA, any existing risers into the tanks have to be used 'to the maximum extent possible' in the retrieval process. Since these risers are only 6-8'' in diameter, this limits the types of systems that can be used for the retrieval. In addition, each bin containing the calcine has a series of obstructions which need to be navigated around and so inflexible systems such as conventional hose-based vacuum type systems are not feasible. In addition, the retrieval system selected has to be able to manage multiple forms of calcine (powder, granular, agglomerated, chunks etc) and operate in a very highly radioactive environment. This paper describes the progress made to date in developing an approach based on the use of multiple remote system platforms to gain entry into the bin-sets and to safely and cost-effectively retrieve the calcine wastes. Much of the work has focused on the application of the OC Robotic snake-arm robot system although significant effort has also been allocated to understanding the deployment mechanism of the snake-arm and to the ancillary supporting systems such as those for contamination control, cutting and welding new risers into the bin-sets and vacuum hose management within the bin-set. Future work will include the construction of a scaled mock up to demonstrate the field implementation of the system. (authors)

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