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At least 433 records · Page 24

Results for the First, Second, and Third Quarter Calendar Year 2015 Tank 50H WAC slurry samples chemical and radionuclide contaminants

This report details the chemical and radionuclide contaminant results for the characterization of the Calendar Year (CY) 2015 First, Second, and Third Quarter sampling of Tank 50H for the Saltstone Waste Acceptance Criteria (WAC) in effect at that time. Information from this characterization will be used by Defense Waste Processing Facility (DWPF) & Saltstone Facility Engineering (D&S-FE) to support the transfer of low-level aqueous waste from Tank 50H to the Salt Feed Tank in the Saltstone Facility in Z-Area, where the waste will be immobilized. This information is also used to update the Tank 50H Waste Characterization System. Previous memoranda documenting the WAC analyses results have been issued for these three samples.

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Panel Session 57: Nuclear Power Plant Waste Management - LLW Processor Issues

This panel resulted in senior utility managers discussing issues with commercial LLW processors. LLW processors updated utilities/NPPs on new or different services or technologies being offered to assist in their radioactive waste management goals, changes to their facility license, revisions to their acceptance guidelines, and emerging compliance issues. Presentations proposed include: 1 - EnergySolutions Waste Processing Overview (Joe Heckman, EnergySolutions (USA)); 2 - Perma-Fix: A Nuclear Services and Waste Management Company (John Kelly, Perma-Fix Environmental Services (USA)); 3 - UniTech - 'Expanding to Reduce Costs' LLW Processor Update (Gregg Johnstone, UniTech Services Group, Inc. (USA)); 4 - Low - Level Radioactive Materials Processing (Michael Miller, Veolia Nuclear Solutions (USA))

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Virtual and Augmented Realities to Support Development, Training, Deployment, and Operations of Remote Systems - 20516

Throughout the U.S. Department of Energy complex, there is a need to deploy new, unique equipment and systems for cleanup activities, waste processing, infrastructure repair, and decommissioning and demolition in radioactive and hazardous environments. The new equipment often consists of custom systems created by integrating and modifying commercial technologies. In addition, the systems are deployed in unique environments due to limited access, waste properties, and the presence of previously installed equipment or infrastructure that has been left in place. The deployment of custom systems that are 'turnkey' ready for first-time installation and application has its risks. Use of test platforms and mockups for cold (i.e., non-radioactive) testing is one method to mitigate that risk: however, full-scale test mockups for operations such as waste retrieval and tank inspection can be expensive and must be modified to incorporate the uniqueness of each application. An alternative to full-scale mockups is to use three-dimensional (3D) virtual environments and virtual reality (VR) technology to develop interactive virtual test beds. These can be used throughout the down-selection, development, system integration, operational planning, operator training, and mission execution phases of a task. Virtual test beds are readily reconfigured, can be used for multiple tasks, support multiple users in parallel, and can accommodate enhancements and changes not originally planned or conceived in the initial design. Virtual test beds also provide a platform for developing and testing augmented reality (AR) tools for improving operator efficiency. A VR experience has been generated for waste retrieval from a Hanford waste tank to present the potential applications of a virtual test bed. Features of the VR experience are presented, potential applications of VR and AR for remote applications are discussed, and development areas for improving future application are considered. (authors)

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Melt Rate Enhancement for High Aluminum HLW Glass Formulation (Final Report)

The principal objective of the work reported here was to develop and identify HLW glass compositions that maximize waste processing rates for the aluminum limted waste composition specified by ORP while maintaining high waste loadings and acceptable glass properties. This was accomplished through a combination of crucible-scale tests, confirmation tests on the DM100 melter system, and demonstration at pilot scale (DM1200). The DM100-BL unit was selected for these tests since it was used previously with the HLW waste streams evaluated in this study, was used for tests on HLW glass compositions to support subsequent tests on the HLW Pilot Melter, conduct tests to determine the effect of various glass properties (viscosity and conductivity) and oxide concentrations on glass production rates with HLW feed streams, and to assess the volatility of cesium and technetium during the vitrification of an HLW AZ-102 composition.

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Revision of the Branch Technical Position on Concentration Averaging and Encapsulation - 16315

On February 25, 2015, the U.S. Nuclear Regulatory Commission (NRC) issued Revision 1 of the Branch Technical Position on Concentration Averaging and Encapsulation (CA BTP), culminating a six-year process. This revision provides updated guidance on the interpretation of paragraph 61.55(a)(8) of Title 10 of the Code of Federal Regulations, 'Determination of concentrations in wastes', as it applies to the classification of a variety of different types and forms of low-level radioactive waste (LLW) as Class A, B, or C waste. The waste class determines the minimum safety measures to be applied in order to provide reasonable assurance of safe disposal of the waste. The CA BTP is used by thousands of LLW generators, by the waste processing industry, by all operating LLW disposal sites in the U.S. and by Agreement States. The previous version of the CA BTP, published in 1995, was issued before the NRC adopted its risk-informed and performance-based regulatory policy. The revised CA BTP has been informed by that policy. The revised CA BTP also contains new guidance on acceptable methods that can be used to perform concentration averaging of LLW for the purpose of determining its waste class for disposal. This paper provides background information on why the CA BTP was issued and subsequently updated. This paper summarizes the new guidance found in the revised CA BTP. This paper also explains some of the key issues that required NRC's resolution, and the reasoning that underpins the major decisions reflected in Revision 1 of the CA BTP. (authors)

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Saltstone Facility Operating Experience with Mega-Vault Disposal Units - 19497

The Saltstone Production Facility (SPF) has been treating low-level waste aqueous salt solution at the Savannah River Site (SRS) since 1990. Salt solutions are transferred from a large-volume, underground storage tank via underground waste transfer line into the SPF. The waste is received into one of two receipt tanks and then transferred to a twin-shaft mixer where dry feeds are introduced to create a fresh grout slurry. The grout exits the mixer into a hold tank, which has a residence time of approximately two minutes. The slurry is then pumped via peristaltic pump through a grout transfer line to one of several disposal units in the Saltstone Disposal Facility (SDF). The transfer line length varies from 0.4 km to 0.8 km depending on disposal location. With the salt disposal mission at SRS accelerating via the operation of the Modular Caustic Side Solvent Extraction (MCU) facility and the anticipated startup of the Salt Waste Processing Facility (SWPF), it is important that the SPF and SDF operate with minimal outage time and maximum facility availability. To this end, the design of the newest disposal units has incorporated the benefits of commercially available cylindrical concrete tank design on a very large scale to maximize economies of scale. This new design has been placed into radioactive service and the initial operational data will be compared with previous disposal operations along with an overview of how disposal unit design has significantly evolved since the initial construction in the early 1990's. Additionally, several processing facility modifications have either been implemented or developed in recent years to improve operating reliability and ensure facility availability once the SWPF becomes the primary waste generator providing feed for the SPF. These modifications, the issues driving their development, and an assessment of their effectiveness will also be discussed. (author)

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A cost-benefit analysis of landfill mining and material recycling in China

Highlights: • Assessing the economic feasibility of landfill mining. • We applied a cost-benefit analysis model for landfill mining. • Four material cycling and energy recovery scenarios were designed. • We used net present value to evaluate the cost-benefit efficiency. - Abstract: Landfill mining is an environmentally-friendly technology that combines the concepts of material recycling and sustainable waste management, and it has received a great deal of worldwide attention because of its significant environmental and economic potential in material recycling, energy recovery, land reclamation and pollution prevention. This work applied a cost-benefit analysis model for assessing the economic feasibility, which is important for promoting landfill mining. The model includes eight indicators of costs and nine indicators of benefits. Four landfill mining scenarios were designed and analyzed based on field data. The economic feasibility of landfill mining was then evaluated by the indicator of net present value (NPV). According to our case study of a typical old landfill mining project in China (Yingchun landfill), rental of excavation and hauling equipment, waste processing and material transportation were the top three costs of landfill mining, accounting for 88.2% of the total cost, and the average cost per unit of stored waste was 12.7 USD ton{sup −1}. The top three benefits of landfill mining were electricity generation by incineration, land reclamation and recycling soil-like materials. The NPV analysis of the four different scenarios indicated that the Yingchun landfill mining project could obtain a net positive benefit varying from 1.92 million USD to 16.63 million USD. However, the NPV was sensitive to the mode of land reuse, the availability of energy recovery facilities and the possibility of obtaining financial support by avoiding post-closure care.

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Material Properties and Temperature Effects Quantified for Slurry Wear - 16141

Literature indicates that the hardness of solid particles relative to the target material hardness and temperature influence erosion. Investigations of temperature effects on erosion also show that erosion resistance is a function of target material's homologous temperature to melting point temperature. While relative hardness effects on erosion and temperature effects on erosion have been investigated independently, their combined effect has not been evaluated. This paper summarizes a review of relevant literature on relative hardness and temperature effects on erosion and via probabilistic calculations, quantifies the combined effect of relative hardness and temperature on liquid-particle induced wear, or slurry wear. These insights can assist material selection and plant operations at waste processing facilities. (authors)

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Effects Influencing Plutonium-Absorber Interactions and Distributions in Routine and Upset Waste Treatment Plant Operations

This report is the third in a series of analyses written in support of a plan to revise the Hanford Waste Treatment and Immobilization Plant (WTP) Preliminary Criticality Safety Evaluation Report (CSER) that is being implemented at the request of the U.S. Department of Energy (DOE) Criticality Safety Group. A report on the chemical disposition of plutonium in Hanford tank wastes was prepared as Phase 1 of this plan (Delegard and Jones 2015). Phase 2 is the provision of a chemistry report to describe the potential impacts on criticality safety of waste processing operations within the WTP (Freer 2014). In accordance with the request from the Environmental and Nuclear Safety Department of the WTP (Miles and Losey 2012), the Phase 2 report assessed the potential for WTP process conditions within and outside the range of normal control parameters to change the ratio of fissile material to neutron-absorbing material in the waste as it is processed with an eye towards potential implications for criticality safety. The Phase 2 study also considered the implications should WTP processes take place within the credible range of chemistry upset conditions. In the present Phase 3 report, the 28 phenomena described in the Phase 2 report were considered with respect to the disposition of plutonium and various absorber elements. The phenomena identified in the Phase 2 report are evaluated in light of the Phase 1 report and other resources to determine the impacts these phenomena might have to alter the plutonium/absorber dispositions and ratios. The outcomes of the Phase 3 evaluations then can be used to inform subsequent engineering decisions and provide reasonable paths forward to mitigate or overcome real or potential criticality concern in plant operations.

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Progress updates for the Hanford waste treatment and immobilization plant project - 15593

The Waste Treatment and Immobilization Plant (WTP) Project is vital to the U.S. Department of Energy's (DOE) mission to clean up radioactive waste at the Hanford Site in southeastern Washington State. The overall WTP Project objective is to design, build, and commission the facilities and systems that will treat and immobilize approximately 211 983 m{sup 3} (56 million gallons) of radioactive waste stored in 177 underground storage tanks. The combination of technologies in the WTP makes it a first-of-a-kind facility at a scale larger than ever built for radioactive waste processing. In recent years, the project has encountered issues that affected project design completion, especially in the Pretreatment and High-Level Waste Facilities. These included technical issues, changes in requirements, changes in funding, weaknesses in the nuclear supply chain, identification of new hazards, and new information about the tank waste. The combination of these issues led to a suspension of construction of the Pretreatment Facility and a partial suspension of construction of the High-Level Waste Facility in mid-2012. Since then, the WTP Project has made significant progress in establishing and implementing a strategy for resolving these issues. This strategy focuses on resuming full production work in a sequence that fully supports the DOE plan for a phased approach for starting treatment of the Hanford tank waste, beginning with the low-activity waste streams. Concurrent with the technical and programmatic issues, the DOE Office of River Protection has made numerous changes in the processes it uses to manage and oversee the WTP Project. These changes have had a beneficial impact on the progress the project has made in establishing a path forward for resuming production activities, and on planning for the necessary contract and baseline changes needed to complete the project. (authors)

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Analysis of Water Recovery Rate from the Heat Melt Compactor

Human space missions generate trash with a substantial amount of plastic (20% or greater by mass). The trash also contains water trapped in food residue and paper products and other trash items. The Heat Melt Compactor (HMC) under development by NASA Ames Research Center (ARC) compresses the waste, dries it to recover water and melts the plastic to encapsulate the compressed trash. The resulting waste disk or puck represents an approximately ten-fold reduction in the volume of the initial trash loaded into the HMC. In the current design concept being pursued, the trash is compressed by a piston after it is loaded into the trash chamber. The piston face, the side walls of the waste processing chamber and the end surface in contact with the waste can be heated to evaporate the water and to melt the plastic. Water is recovered by the HMC in two phases. The first is a pre-process compaction without heat or with the heaters initially turned on but before the waste heats up. Tests have shown that during this step some liquid water may be expelled from the chamber. This water is believed to be free water (i.e., not bound with or absorbed in other waste constituents) that is present in the trash. This phase is herein termed Phase A of the water recovery process. During HMC operations, it is desired that liquid water recovery in Phase A be eliminated or minimized so that water-vapor processing equipment (e.g., condensers) downstream of the HMC are not fouled by liquid water and its constituents (i.e., suspended or dissolved matter) exiting the HMC. The primary water recovery process takes place next where the trash is further compacted while the heated surfaces reach their set temperatures for this step. This step will be referred to herein as Phase B of the water recovery process. During this step the waste chamber may be exposed to different selected pressures such as ambient, low pressure (e.g., 0.2 atm), or vacuum. The objective for this step is to remove both bound and any remaining free water in the trash by evaporation. The temperature settings of the heated surfaces are usually kept above the saturation temperature of water but below the melting temperature of the plastic in the waste during this step to avoid any encapsulation of wet trash which would reduce the amount of recovered water by blocking the vapor escape. In this paper, we analyze the water recovery rate during Phase B where the trash is heated and water leaves the waste chamber as vapor, for operation of the HMC in reduced gravity. We pursue a quasi-one-dimensional model with and without sidewall heating to determine the water recovery rate and the trash drying time. The influences of the trash thermal properties, the amount of water loading, and the distribution of the water in the trash on the water recovery rates are determined.

Heat Melt Compactor↗

Environmental cleanup of the East Tennessee Technology Park - Year three - A status report - 15522

On August 1, 2011, URS - CH2M Oak Ridge LLC (UCOR) began its five-year, $1.4 billion cleanup of the East Tennessee Technology Park (ETTP), the former Oak Ridge Gaseous Diffusion Plant, located on the U.S. Department of Energy's (DOE) Oak Ridge Reservation (Reservation) in Tennessee. Under the contract, the base period scope of work to be completed by 2016 includes demolition of both the K-25 and K-31 gaseous diffusion buildings and all demolition waste dispositioned, initiation of demolition on the remaining K-27 gaseous diffusion building, safe storage of transuranic (TRU) waste for eventual transfer to the Transuranic Waste Processing Center, previously designated -No-Path-To-Disposition Waste- dispositioned to the extent possible, and completion of design and engineering of a mercury treatment system to be installed at the Y-12 National Security Complex (Y-12). UCOR will have also managed DOE Oak Ridge Office of Environmental Management (OREM)-owned facilities at ETTP, Oak Ridge National Laboratory (ORNL), and Y-12 in a safe and cost-effective manner. Since assuming responsibilities as the ETTP and Oak Ridge Reservation cleanup contractor, UCOR has completed its life-cycle Performance Measurement Baseline; received its Earned Value Management System (EVMS) certification; completed the deactivation and demolition (D and D) of the 44-acre K-25 gaseous diffusion building; removed and disposed of a contaminated underground storage tank (Tank W-1A) at ORNL; dispositioned mercury-contaminated storage tanks at Y-12; completed closure of the Toxic Substances Control Act Incinerator; disposed of more than 158,000 cubic yards of cleanup waste while managing the on-site Environmental Management Waste Management Facility (EMWMF); and conducted operations, surveillance, and maintenance activities at 180 DOE EM facilities at ETTP, ORNL, and Y-12 in a safe and cost-effective manner. In 2014, UCOR completed the removal and disposal of all of the K-25 demolition debris; completed the removal of asbestos-laden transite siding on the K-31 gaseous diffusion building and began demolition of that 17-acre building; advanced the characterization and deactivation of the K-27 gaseous diffusion building, readying the building for the start of demolition in 2015; advanced the design and engineering of a Mercury Treatment Facility to be located at Y-12; continued the cost-effective and safe disposition of all cleanup wastes; and readied ETTP for Voluntary Protection Program Star Status Assessment in 2015. Project performance as of December 31, 2014, continued to be excellent: - Cost Performance Index - 1.12 - Schedule Performance Index - 1.02 At the same time, since safety is the foundation of all cleanup work, UCOR's safety record goes hand-in-hand with its excellent project performance. Through calendar year 2014, UCOR's recordable injury rate was 1.04 occurrences per 200,000 work hours, and the lost workday case rate was 0.52. This performance continues to be due, in large part, to the people and processes URS and CH2M HILL, the parent companies of UCOR, brought to the project. Key enablers included: - Selected and deployed experienced staff in key leadership positions throughout the organization, continually re-energizing the organization with new leaders - Implemented previously successful safety programs and initiatives from other DOE sites, such as DOE's Voluntary Protection Program - Brought fresh thinking and innovation to tackling tough project issues, such as the deactivation of the Technetium-99-contaminated areas of K-27, applying lessons learned from the D and D of K-25; and implementing a -waste factory- approach to cleanup - Maintained continuing alignment between UCOR's contract with DOE and its Performance Measurement Baseline - rapidly addressing contract changes when they arose - Maintained a readiness both from a process as well as experienced personnel perspective to support DOE OREM in performing other significant Oak Ridge Reservation cleanup activities. (authors)

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Requirements Development Issues for Advanced Life Support Systems: Solid Waste Management

Long duration missions pose substantial new challenges for solid waste management in Advanced Life Support (ALS) systems. These possibly include storing large volumes of waste material in a safe manner, rendering wastes stable or sterilized for extended periods of time, and/or processing wastes for recovery of vital resources. This is further complicated because future missions remain ill-defined with respect to waste stream quantity, composition and generation schedule. Without definitive knowledge of this information, development of requirements is hampered. Additionally, even if waste streams were well characterized, other operational and processing needs require clarification (e.g. resource recovery requirements, planetary protection constraints). Therefore, the development of solid waste management (SWM) subsystem requirements for long duration space missions is an inherently uncertain, complex and iterative process. The intent of this paper is to address some of the difficulties in writing requirements for missions that are not completely defined. This paper discusses an approach and motivation for ALS SWM requirements development, the characteristics of effective requirements, and the presence of those characteristics in requirements that are developed for uncertain missions. Associated drivers for life support system technological capability are also presented. A general means of requirements forecasting is discussed, including successive modification of requirements and the need to consider requirements integration among subsystems.

Levri, Julie A.↗

Analysis and Testing of Parsons NGS Solvent Formulation

Three drums containing solvent used during the Next Generation Solvent Test (NGST) conducted by Parsons in 2014-2015 were received at SRNL. After homogenization, samples from each drum were removed, and a composite sample prepared. This composite was analyzed to determine current SRNL characterization method efficacy, and to determine if the 1,3-dicyclohexyl-2- (isotridecyl) guanidine (DCiTG) suppressor could be selectively removed. The results of this work indicate that the titration analytical method currently employed at SRNL is effective at quantitating the DCiTG. The nuclear magnetic resonance (NMR) method at SRNL can detect the DCiTG, as well its amine and urea decomposition products. However, further development work will be required to allow the 1 H NMR method to quantitate these species. A series of three washing tests were performed on composite samples of the solvent. The results show that up to 67% of the DCiTG was removed from the solvent by simple washing, at multiple ratios of solvent: aqueous phase. This may indicate that a simple pathway exists to wash out the DCiTG and reclaim the solvent (~150 gallons) for future use at the Salt Waste Processing Facility (SWPF).

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TEX 2.0: DOE-EM NCS Experimental Needs [Slides]

This presentation touches on the needs presented from recent work with Hanford Tank Farms (HTF) and Savannah River Site (SRS) Liquid Waste Process (LWP). The Meetings were held with other sites by presenting information about TEX 2.0 and having discussions on what the current challenges were for criticality safety validation.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Characterization of SuperLig® 639 rhenium and technetium resin with batch contact and column tests

Two new (2013) lots of SuperLig® 639 ion exchange resin (IBC Advanced Technologies, American Forks, UT) were tested for the first time above typical sodium concentrations (7.8 M sodium, along with typical 5 M concentration) in highly alkaline solutions. Batch contact and ion exchange column tests characterized rhenium (perrhenate) adsorption as a surrogate for pertechnetate. The work supports technetium removal options for Supplemental Low Activity Waste processing at the Hanford River Protection Project Waste Treatment Plant (WTP). The current work found that the resin performs well in the 7.8 M sodium simulant despite complete floating of the beads. A notable difference in performance between the two new resin lots was found. As a result, resin loading overall versus temperature, potassium concentration, and rhenium/nitrate ratios is consistent with previous data and expectations despite the high sodium concentration and floating of the resin beads.

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A Simple Criterion to Estimate Performance of Pulse Jet Mixed Vessels

Pulse jet mixed process vessels comprise a key element of the U.S. Department of Energy’s strategy to process millions of gallons of legacy nuclear waste slurries. Slurry suctioned into a pulse jet mixer (PJM) tube at the end of one pulse is pneumatically driven from the PJM toward the bottom of the vessel at the beginning of the next pulse, forming a jet. The jet front traverses the distance from nozzle outlet to the bottom of the vessel and spreads out radially. Varying numbers of PJMs are typically arranged in a ring configuration within the vessel at a selected radius and operated concurrently. Centrally directed radial flows from neighboring jets collide to create a central upwell that elevates the solids in the center of the vessel when the PJM tubes expel their contents. An essential goal of PJM operation is to elevate solids to the liquid surface to minimize stratification. Solids stratification may adversely affect throughput of the waste processing plant. Unacceptably high slurry densities at the base of the vessel may plug the pipeline through which the slurry exits the vessel. Additionally, chemical reactions required for processing may not achieve complete conversion. To avoid these conditions, a means of predicting the elevation to which the solids rise in the central upwell that can be used during vessel design remains essential. In this paper we present a simple criterion to evaluate the extent of solids elevation achieved by a turbulent upwell jet. The criterion asserts that at any location in the central upwell the local velocity must be in excess of a cutoff velocity to remain turbulent. We find that local velocities in excess of 0.6 m/s are necessary for turbulent jet flow through both Newtonian and yield stress slurries. By coupling this criterion with the free jet velocity equation relating the local velocity to elevation in the central upwell, we estimate the elevation at which turbulence fails, and consequently the elevation at which the upwell fails to further lift the slurry. Comparing this elevation to the vessel fill level predicts whether the jet flow will achieve the full vertical extent of the vessel at the center. This simple local-velocity criterion determines a minimum PJM nozzle velocity at which the full vertical extent of the central upwell in PJM vessels will be turbulent. The criterion determines a minimum because flow in regions peripheral to the central upwelling jet may not be turbulent, even when the center of the vessel in the upwell is turbulent, if the jet pulse duration is too short. The local-velocity criterion ensures only that there is sufficient wherewithal for the turbulent jet flow to drive solids to the surface in the center of the vessel in the central upwell.

mixing↗

Determination Of Reportable Radionuclides For DWPF Sludge Batch 7B (Macrobatch 9)

The Waste Acceptance Product Specifications (WAPS) 1.2 require that “The Producer shall report the inventory of radionuclides (in Curies) that have half-lives longer than 10 years and that are, or will be, present in concentrations greater than 0.05 percent of the total inventory for each waste type indexed to the years 2015 and 3115”. As part of the strategy to comply with WAPS 1.2, the Defense Waste Processing Facility (DWPF) will report for each waste type, all radionuclides (with half-lives greater than 10 years) that have concentrations greater than 0.01 percent of the total inventory from time of production through the 1100 year period from 2015 through 3115. The initial listing of radionuclides to be included is based on the design-basis glass as identified in the Waste Form Compliance Plan (WCP) and Waste Form Qualification Report (WQR). However, it is required that this list be expanded if other radionuclides with half-lives greater than 10 years are identified that may meet the greater than 0.01% criterion for Curie content. Specification 1.6 of the WAPS, International Atomic Energy Agency (IAEA) Safeguards Reporting for High Level Waste (HLW), requires that the ratio by weights of the following uranium and plutonium isotopes be reported: U-233, U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, and Pu-242. Therefore, the complete set of reportable radionuclides must also include this set of U and Pu isotopes. The DWPF is receiving radioactive sludge slurry from HLW Tank 40. The radioactive sludge slurry in Tank 40 is a blend of the heel from Sludge Batch 7a (SB7a) and Sludge Batch 7b (SB7b) that was transferred to Tank 40 from Tank 51. The blend of sludge in Tank 40 is also referred to as Macrobatch 9 (MB9). This report develops the list of reportable radionuclides and associated activities as a function of time. The DWPF will use this list and the activities as one of the inputs for the development of the Production Records that relate to radionuclide inventory.

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