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All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Design and Licensing of Irradiated Uranyl Nitrate Liquid Containers and Applicable Loading and Packaging System - 18481

Irradiated highly enriched uranyl nitrate liquid (HEUNL) solutions are typically residual materials resulting from the processing of uranium target materials in the production of radioisotopes. Traditionally, these residual materials have been processed and stored at the site of origin, due to the lack of an option for packaging, transportation and disposition. Modern alternatives to process, stabilize, package and dispose of these materials have been developed, but until recently there has not been a viable Type B package solution that could be used to ship these materials in sufficient quantities to justify the cost of an off-site disposition strategy. For this reason, most facilities that generate irradiated HEUNL materials have implemented on-site storage tanks or containers, resulting in the significant accumulation of HEUNL inventories worldwide. This paper provides an overview of the design and licensing activities for the development of sealed irradiated HEUNL transport containers to permit safe transportation of this material in the NAC Legal Weight Truck (NAC-LWT) cask. A description of the robust HEUNL container, licensing and fabrication is also presented. In addition, NAC has developed a specialized loading system to optimize the process to load, package and un-package HEUNL materials to minimize worker exposure during loading, and execute the packaging process efficiently for safe transport and retrieval. (authors)

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DOE OSTI2020

Retrieval of Transuranic Drums with Deflagration to Detonation Potential from a Vault at the Oak Ridge National Laboratory - 20207

During the final phase of a long-term project, a total of 80 TRU waste drums at ORNL have been retrieved from a below-grade storage vault in which they have remained untouched for about 25 years. The drums, which house material generated from past ORNL operations, are now stored and ready for eventual transfer to the Transuranic Waste Processing Center, located on Highway 95 in Oak Ridge. The Transuranic Waste Processing Center will process and ship the drums to WIPP in Carlsbad, New Mexico, for disposal. Workers lifted the drums one-by-one from the underground storage vault with a 110-ton crane over a one-week period in October 2017. Although the physical work was performed in one week, years of detailed planning preceded the fieldwork. A key obstacle needed to be overcome before the project could safely proceed. This obstacle involved the need to ensure controls were developed for the Deflagration to Detonation Transition (DDT) potential. DDT refers to a phenomenon in ignitable mixtures of a flammable gas and air (or oxygen) when a sudden transition takes place from a deflagration type of combustion to a detonation type of explosion. Basically, these drums, under certain circumstances, have a potential to detonate and overpressurize. Extensive analysis of the drums verified that detonation, while unlikely, was possible. Of the total 103 drums that were retrieved or moved in preparation for retrieval, 18 had DDT potential. Extensive security requirements were also necessary, due to the nature of the material, and this added another layer of complexity to the fieldwork. Prior to retrieving the drums from the underground storage vault, space needed to be made for them within the various above-grade facilities. Inventory and security requirements dictated what could be stored where, and the first step of the field work involved a number of drum movements in preparation for storage of the newly retrieved drums. This effort involved moving 21 drums already in the storage area to facilitate the addition of the newly retrieved ones. The retrieval project required a total of 107 critical lifts using the 110-ton crane, under 3 critical lift plans. The TRU waste storage facility is a Category 2 Nuclear Facility and is managed under an extensive Documented Safety Analysis. The TRU retrieval project was conducted in accordance with DOE Order 425.1D and included three Implementation Verification Reviews, two Management Assessments, one Readiness Assessment, and seven mock-ups in preparation for the field work. Despite being stored for 25 years, the drums were in good shape due to the integrity of the facilities where the drums were stored. Of the 80 drums that were retrieved, 12 were placed in special containers called 'overpacks' based on their content. The drums were removed safely and without incident, thanks to a lot of preparation and the efforts of a highly skilled team. Lessons learned through the DDT control development and the drum retrieval and overpacking process could be beneficial to others having to manage similar drums of TRU waste. (authors)

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DOE OSTI2016

Assessment of Options for the Treatment of Nitrate Salt Wastes at Los Alamos National Laboratory - 16541

This paper summarizes the methodology used to evaluate options for treatment of the remediated nitrate salt waste containers at Los Alamos National Laboratory. The method selected must enable treatment of the waste drums, which consist of a mixture of complex nitrate salts (oxidizer) improperly mixed with sWheat Scoop{sup R}, an organic kitty litter and absorbent (fuel), in a manner that renders the waste safe, meets the specifications of waste acceptance criteria, and is suitable for transport and final disposal in the Waste Isolation Pilot Plant located in Carlsbad, New Mexico. A Core Remediation Team was responsible for comprehensively reviewing the options ensuring a robust, defensible treatment recommendation. The evaluation process consisted of two steps. First, a pre-screening process was conducted to cull the list on the basis for a decision of feasibility of certain potential options with respect to the criteria. Then, the remaining potential options were evaluated and ranked against each of the criteria in a consistent methodology. Numerical scores were established by consensus of the review team. Finally, recommendations were developed based on current information and understanding of the scientific, technical, and regulatory situation. A discussion of the preferred options and documentation of the process used to reach the recommended treatment options are presented. (authors)

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DOE OSTI2016

Prioritizing and Managing Technology Needs to Meet River Protection Project Mission Objectives - 16462

Over 56 million gallons of radioactive waste are stored in 177 large underground tanks at the Hanford Site, located in southeastern Washington State. This nuclear waste resulted from plutonium production for the nation's nuclear defense program. The River Protection Project (RPP) mission is to: - Safeguard and safely manage waste stored in the Hanford tanks, - Treat the waste using the Waste Treatment and Immobilization Plant (WTP), - Ensure safe disposition of the waste to protect the Columbia River and the environment. Each aspect of the mission requires some application of technology in order to meet the mission objectives. The RPP Technology Roadmap document presents a comprehensive, integrated assessment of the technology-related advances needed to ensure successful completion of the Hanford tank waste cleanup mission. This paper describes the development of the RPP Technology Roadmap, summarizes its content, describes integration and collaboration with the Department of Energy's national laboratories to support the work identified, and defines how priorities were identified to ensure near-term RPP mission commitments can be accomplished while preserving the longer term technology development elements necessary for mission completion. The Challenge - Managing a wide variety of technology needs that involve multiple organizations, contractors, and national laboratories presents additional challenges when funding constraints are factored in. A process was needed to ensure the technology needs supporting near term mission priorities were appropriately identified, prioritized, and supported through the fiscal year planning process. Additionally, the national laboratory resources necessary to support many of the technologies require coordination to ensure the most appropriate resources are applied at the appropriate level without duplication or contradiction of effort. The Solution - The RPP Technology Roadmap was developed with the primary purposes to: - Identify technology related items needed to support or enhance the River Protection Project (RPP) mission, - Prioritize technology needs to support fiscal year planning processes, - Organize National laboratory support of the RPP mission and communicate needs to the participating contractors. The Technology Roadmap is consistent with Department of Energy (DOE) Hanford mission priorities and is also in alignment with other Hanford Site technical initiatives and policy and planning documents. The Technology Roadmap highlights the current estimated technology needs and identifies where planning shortfalls exist and where national laboratory participation is needed to support mission success. This information assists fiscal year planning and national laboratory contracting processes to ensure that technology needs are supported. In addition to supporting the RPP baseline mission, the Technology Roadmap provides a basis for strategic planning by identifying opportunities to use technology solutions to enhance mission efficiency. This waste management paper describes the innovative and collaborative processes used to develop, validate, and communicate the results so the project owners and the national laboratories can most effectively plan work scope and resources. A discussion of the following processes is included: - Technology needs and task owner identification, - Functional area grouping and area lead identification, - Needs prioritization, - High priority needs ranking (including validation), - Results compilation and planning gap identification, - Results presentation by project and functional area, - National laboratory coordination and collaboration. Application - The process used to develop the RPP technology roadmap is applicable to any site where multiple national laboratories and contractors are involved in identifying and implementing technology based solutions. This process is also applicable to the DOE HQ role to provide prioritization and coordination of the nationwide technology development activities. (authors)

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DOE OSTI2018

Studies on Workability of Geopolymer Solidification - 18171

In Japan, intermediate-level radioactive waste, such as spent resin, is solidified and set in sealed containers without venting. To reduce the amount of hydrogen generated due to water radiolysis in stored intermediate-level waste, geopolymer solidification has been studied. Geopolymer solidification is an excellent method for conditioning intermediate-level waste from the viewpoint of its high thermal stability and low processing cost. However, it has a problem that the apparent viscosity increases drastically a few minutes after the onset of setting. This may cause some trouble in the process of pouring the geopolymer into drums and mixing it with waste. In this study, in order to control the workability of a geopolymer mixture, we investigated the effects of the compounding ratio and the kinds of chemical agents used. We confirmed that the geopolymer reaction was slowed down by reducing the amount of calcium and adding boron. The uniaxial compressive strength of the geopolymer was 3-80 MPa. (authors)

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DOE OSTI2020

Tank Side Cesium Removal for Hanford Tank Waste - Actual Waste Demonstrations - 20467

The US DOE and its contractors are evaluating several options for treating Hanford tank supernate streams to provide feed for the Low-Activity Waste Facility at the Hanford Waste Treatment and Immobilization Plant. At this time, the intent is to deploy a Tank-Side Cesium Removal system. The key unit operations for this process are filtration and cesium removal by ion exchange. To assess the anticipated performance of the system, tests with actual waste were performed with samples from Hanford tanks AW-102 and AP-107 to provide basis data regarding the performance of the system in treating these materials. The decanted supernate waste samples from tanks AW-102 and AP-107 were received directly by the Radiochemical Processing Laboratory at Pacific Northwest National Laboratory. These samples were then filtered to provide data on dead-end filtration with a Mott{sup TM} filter media grade 5. Subsequently, the samples were run through 10-milliliter ion exchange columns containing crystalline silicotitanate to assess the dynamic ion exchange performance of these feeds. Test results demonstrated that the planned TSCR operations should meet process performance expectations. (authors)

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DOE OSTI2015

Options Assessment Report: Treatment of Nitrate Salt Waste at Los Alamos National Laboratory

This report documents the methodology used to select a method of treatment for the remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The method selected should treat the containerized waste in a manner that renders the waste safe and suitable for transport and final disposal in the Waste Isolation Pilot Plant (WIPP) repository, under specifications listed in the WIPP Waste Acceptance Criteria (DOE/CBFO, 2013). LANL recognizes that the results must be thoroughly vetted with the New Mexico Environment Department (NMED) and that a modification to the LANL Hazardous Waste Facility Permit is a necessary step before implementation of this or any treatment option. Likewise, facility readiness and safety basis approvals must be received from the Department of Energy (DOE). This report presents LANL’s preferred option, and the documentation of the process for reaching the recommended treatment option for RNS and UNS waste, and is presented for consideration by NMED and DOE.

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DOE OSTI2016

Technology Development in Support of Accelerating the Waste Treatment Mission - 16409

The feed for vitrification to produce a nuclear waste glass is a mixture of waste with glass-forming and modifying additives that is charged onto the cold cap that covers 90-95% of the melt surface. The cold cap consists of a layer of reacting molten glass floating on the surface of the melt in an all-electric, continuous glass melter. As the feed moves through the cold cap, it undergoes chemical reactions and phase transitions through which it is converted to molten glass that moves from the cold cap into the melt pool. Multiple overlapping reactions occur within the cold cap. The process involves a series of reactions that generate multiple gases and subsequent mass loss and foaming significantly influence the mass and heat transfers. The rate of glass melting, is greatly influenced by mass and heat transfers, affects the vitrification process and the efficiency of the immobilization of nuclear waste. Therefore, understanding the cold-cap reactions over the temperature range of the conversion process is critical. It also helps to formulate melter feeds for higher production rate, initiation of crystal forming reactions, and chemical reactions that determine the ultimate fate of technetium in the glass melt. Though the flowsheets at each of the two sites is vastly different development and insertion of mature technological chemical processes and mechanical components can have a significant impact on the duration of the treatment mission and hence the total costs. (authors)

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DOE OSTI2020

Lessons Learned from Hanford's Purex Tunnel 2 Structural Stabilization Project - 20509

In 2019, U.S. Department of Energy (DOE) contractor CH2M Hill Plateau Remediation Company (CHPRC) completed structurally stabilizing the Plutonium Uranium Extraction Plant (Purex) waste storage Tunnel 2 with engineered grout. This action significantly reduced risk to Hanford workers, the surrounding community and environment by preventing future collapse of the tunnel and possible release of radioactive materials. Decades of plutonium production at the Purex facility required construction of two large underground tunnels for storage of highly contaminated plutonium processing equipment that had failed during the operation of Purex and other facilities across the Hanford Site. Following the partial collapse of Purex Tunnel 1 on May 9, 2017, DOE and CHPRC took immediate action to safely stabilize the collapsed section of the tunnel. The hole in the tunnel was filled the next day as a temporary stabilization measure, and no contamination was detected as a result of the partial collapse of the tunnel. Workers then stabilized Tunnel 1 in its entirety by filling with engineered grout, which also encapsulated the radioactive equipment and railcars within the tunnel. Grouting eliminated the risk of further collapse and stabilized the tunnel contents, all while not precluding future remedial actions or final closure decisions. While conducting an integrity analysis of Tunnel 1 by using construction drawings, photos and other historical documents, project teams also analyzed the much larger Purex Tunnel 2. The analysis found Tunnel 2 did not meet current codes for structural integrity, and it may not be able to bear the weight of the soil above the tunnel. Therefore, it was determined that Tunnel 2 was also at risk of collapse. DoE's independent panel of experts determined grouting to be the best choice for stabilizing Tunnel 2 because it provides the highest level of stability and protection, and does not preclude future remedial actions. DOE authorized CHPRC to commence work and the Purex Tunnel 2 Structural Stabilization Project was launched in December of 2017. The paper will discuss the process of recognizing the following critical aspects and risks associated with stabilizing Tunnel 2, and the actions taken by the project team to mitigate these risks, including: - Tunnel investigation and data collection to assist with future remediation; - On-site batching and mock-ups to provide the large quantity of grout required for stabilizing the tunnel; - Workforce involvement and innovative actions that mitigated the project's most significant potential hazards, including traffic control; - Project team high level of execution translated into successful project execution. Sharing lessons learned from this critical risk-reduction effort can enhance the safety and efficiency of other complex and hazardous projects across the DOE complex. (authors)

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DOE OSTI2015

SRNL report for the tank waste disposition integrated flowsheet: Corrosion testing

A series of cyclic potentiodynamic polarization (CPP) tests were performed in support of the Tank Waste Disposition Integrated Flowsheet (TWDIF). The focus of the testing was to assess the effectiveness of the SRNL model for predicting the amount of nitrite inhibitor needed to prevent pitting induced by increasing halide concentrations. The testing conditions were selected to simulate the dilute process stream that is proposed to be returned to tank farms from treating the off-gas from the low activity waste melter in the Waste Treatment and Immobilization Plant.

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DOE OSTI2017

Integrated Salt Waste Processing at the Savannah River Site - 17413

Treatment and disposition of salt waste is the critical path to completion of the Savannah River Site (SRS) Liquid Waste (LW) Disposition and Cleanup Program. The new Salt Waste Processing Facility (SWPF) is a critical link in the pre-treatment chain and will process high-curie salt waste removed from various SRS liquid waste storage tanks. Significant infrastructure modifications are underway throughout the existing LW facilities to integrate the SWPF into the LW system. (authors)

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DOE OSTI2019

In-Can Melter for Vitrification of Waste Coming from Decommissioning and Dismantling Operations - 19020

Some waste streams from decommissioning and dismantling operations are identified as an issue in terms of handling, transportation, adapted conditioning matrix or evacuation outlet within the existing regulations. In this context, the consortium of the DEM and MELT project which gathers Orano, the CEA and ECM technologies are currently developing and qualifying a specific vitrification tool designed to match the requirements and constraints of D and D projects. The main characteristics of this In-Can vitrification process are presented hereafter: - Suitable for high or medium level waste; - Flexible enough to accommodate uncertainties on waste composition; - Adequately dimensioned and compact to be implemented and operated inside an existing cell or close to the waste to be treated; - Designed for a short duration of use, as a decommissioning tool, aimed to be dismantled just after the treatment operation; - Designed in order to produce a small amount of secondary waste; - Requiring low investment and operation costs. The In-Can melting process has already been developed by the CEA for an internal application (defense waste vitrification) and can be rated TRL 7. That means it has been tested at full scale at a prototypical demonstration level with a range of simulants in an operational environment. This paper presents the In-Can vitrification process and its performances for a typical D and D waste highly loaded with Cs. The vitrification tests results will be in particular discussed in terms of process parameters, glass properties and cesium volatility. This paper also highlights the methodology and the organization (material, process, technological and long term behavior teams) deployed in order to adapt and qualify the In-Can process for all kind of wastes (liquid, solid, highly loaded with Tc and Cs, waste with no route, ...) from the matrix design definition to the full qualification. (authors)

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DOE OSTI2019

Transport and Treatment of Chapelcross Gas Ducts for Clearance and Recycling - 19662

In the summer of 2018, Cyclife, the EDF group's specialist radioactive waste management and decommissioning company, transported sixteen gas ducts and associated equipment from the shutdown Chapelcross nuclear power plant to treatment facilities in the UK and Sweden. The client requested a full removal, treatment and disposal service for approximately 820 te of low level metallic waste comprising Upper Gas Duct Assemblies (including bellows), Upper Gas Duct elbows, Filter Pots and Convoluted pipework. The complex scope involved heavy lifting, transport by road and sea, processing and recycling on a large scale, thereby necessitating involvement of a wide range of suppliers at different tiers all managed by Cyclife. The purpose of the entire project was to safely handle and treat all of the components for clearance and recycling into the public domain for the manufacture of new steel products. Most of the treatment took place in Cyclife's nuclear licensed metal treatment facility in Sweden. Cyclife has a long and unique experience of arranging safe transport of large contaminated components on road and sea. The metal treatment facility in Sweden can, without special arrangements, receive components up to 30 meters long with a weight of up to 400 tonnes for decontamination, size reduction and melting. The actual treatment of the components consisted of the following steps: a) Separation of a gas duct from the gantry, b) size reduction, radiological verification measurements and clearance of the gantry, c) stepwise segmentation of the gas duct and elbow including removal of asbestos, d) decontamination, e) melting, f) radiological assessment and clearance of the metal ingots, g) management of the secondary waste for return to UK. The result of the project, upon completed treatment of all the material, demonstrates a high recycling rate of the material. 5-10% of the original mass of material will have to disposed as radioactive waste. This number include all residues, untreatable material (asbestos, mineral wool, filters etc.) and two melt batches with a too high activity content for clearance. The paper takes the reader through the entire project from the initial planning to the clearance of the metal and the return of the secondary waste to UK for disposal. It will also present and discuss pros and cons for alternative management approaches in several perspectives. (authors)

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DOE OSTI2020

Mercury Speciation of Hanford 241-AP-107 Tank Waste Samples - 20384

The concentration of key mercury species in the Hanford tank waste has not been previously measured. Based on process knowledge, an estimated inventory of 2,100 kg of mercury is assumed and was distributed in varied amounts across 149 single-shell tanks (SSTs) and 28 double-shell tanks (DSTs). Limited analyses of individual SSTs and DSTs for total mercury were all below detection, yielding upper limits in concentration estimates based solely on instrument detection limits, resulting in an uncertain mercury inventory for Hanford tanks. Accurate and reliable chemical speciation and quantification is desired to anticipate, predict, and abate the mass movements of various mercury species through the Hanford waste processing flowsheet. Highly sensitive separation and quantification methods for total, elemental, and monomethyl mercury species, in tank waste, have been developed at the Pacific Northwest National Laboratory's (PNNL) Radiochemical Processing Laboratory (RPL). A non-radiological environmental method to quantify mercury species has been adapted for application to Hanford tank 241-AP-107 raw supernatant feed, filtered feed, and cesium-decontaminated samples. These samples represent feed inventory and pretreatment process effluents prior to immobilization by vitrification. This method separates and preconcentrates total, elemental, and monomethyl mercury onto selective solid substrates with little-to-no radioactive background. The solid samples can then be released from strict radiological control, allowing for analytical work to be conducted in low-level radiological facilities. Separation of mercury species was demonstrated at PNNL's RPL in Richland, Washington before transport and analytical quantification using cold vapor atomic fluorescence spectrometry (CVAFS) at PNNL's Marine Science Laboratory (MSL) in Sequim, Washington. Additional studies were conducted to assess the stability of monomethyl mercury in distilled matrices and preconcentrated onto the Carbotrap{sup R} solid substrate to assess sample degradation over time during transport. (authors)

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DOE OSTI2016

Utilization and Application of Kurion Mobile Processing System in Remediation of Radioactively Contaminated Water - 16370

Kurion{sup R} Mobile Processing Systems (KMPS) were deployed in Fukushima as first-of-a- kind at-tank isotope removal systems to remediate Sr from reverse osmosis concentrated salt water. The systems were successful in their efforts and achieved their decontamination and availability goals. The use of KMPS has been considered for other applications where a modular system is of benefit. The use of different Kurion Ion Specific Media (ISM) allows for precision removal of hazardous ions from a variety of waste streams. This paper describes a range of projects in which KMPS modularity will benefit the nuclear industry. The KMPS is a flexible system that can be deployed quickly and efficiently. One treatment system option being considered is the utilization of a KMPS for the removal of contamination from groundwater. Similar to its use in Japan, the KMPS can be sent to remote locations for the direct treatment of ground water. In particular, there are several sites across the world where uranium contamination is a problem. Uranium treatability testing was performed on collected groundwater samples from a radiologically contaminated site. The tests were used to help provide designs for KMPS modifications. These designs included the addition of chemical treatment module for the control of pH and other chemistry, the addition of an ultraviolet module for the destruction of biological material, among others. The tests determined the loading capacities of uranium capture resins (UCR) and media (UCM) for this specific waste stream. Based on uranium loading capacities of 37.3 mg/L (end of test C/C{sub 0} = 0.5) and 54.6 mg/L (extrapolated to C/C{sub 0} = 0.98), and the longevity of the resin (45,911 BV for C/C{sub 0} = 0.5, and 80,556 BV for C/C{sub 0} = 0.98) UCR-1 was selected as the media/resin to be proposed for a uranium treatment system. (authors)

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DOE OSTI2019

Measurements of Hydrogen Generation Rates using a Sealed Reactor System - 19428

Sealed reactor tests were performed to measure the hydrogen generation rate (HGR) from stored radioactive waste at the Savannah River Site. HGR may be very low, on the order of 7.5 x 10{sup -13} (m{sup 3}/h)/L [that is, cubic meters of H{sub 2} generated per hour per liter of radioactive liquid waste, or 10{sup -10} (ft{sup 3}/h)/gal]. Because the estimated HGR limit of detection for an open, continuous flow, system is approximately 7.5 x 10{sup -10} (m{sup 3}/h)/L [or 10{sup -7} (ft{sup 3}/h)/gal], a more sensitive method was needed. By accumulating generated gasses in a sealed system without a continuous purge gas, the resulting concentrations can increase the sensitivity of measurement, which is based on time of heating. In 2017, Savannah River Remediation declared a Potential Inadequacy in the Safety Analysis related to the potential thermal decomposition of organics that generate hydrogen in each of three Liquid Waste facilities: Concentration, Storage and Transfer Facility, Saltstone Production Facility, and the Defense Waste Processing Facility. The research for this study concerns low activity waste of Tank 50, which feeds the Saltstone Production Facility. This study evaluated the impact of organics including formate on the generation of H{sub 2}. Samples of SRS Tank 50 waste, which is a nominal 5 molar sodium salt solution, were sealed in stainless steel reactors and held at varying temperatures, e.g., 50 deg. C, 75 deg. C, 100 deg. C, and 120 deg. C, to determine if organic compounds are decomposing into H2. The experiments included trace solids used to stabilize waste in grout, e.g., grout-premix and Blast Furnace Slag, to determine if any solid phase reactions are present, which would accelerate or inhibit the generation of H{sub 2} at these temperatures. This paper describes the sealed reactors designed and constructed at Savannah River National Laboratory and the radioactive tests that were performed to measure HGR in both oxic and anoxic atmospheres. The shakedown tests showed the sealed reactors to be leak tight and hydrogen concentrations from calibrated standards could be accurately measured. Previous simulant tests helped to perfect the test methods and better design the radioactive tests. The radioactive results showed that the rate of hydrogen generated from Tank 50 waste, over a range of temperatures from 30 deg. C up to 120 deg. C, is bounded by an empirical relation (Hu Equation) that can be used to approximate HGRs in Tank 50 radioactive waste. (authors)

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DOE OSTI2016

Implementation Guide for the Revised Branch Technical Position on Concentration Averaging and Encapsulation - 16021

As a result of the research begun in 2007 to better understand the bases of the Branch Technical Position (BTP) on 'Concentration Averaging and Encapsulation' and 10 CFR 61 'Licensing Requirements for Land Disposal of Radioactive Waste', EPRI identified three regulatory initiatives that could be pursued to enhance disposal options for the industry. These initiatives directly follow the NRC strategic assessment. The first was to revise the BTP to allow broader blending of compatible waste types. The second was to examine how 10 CFR 61.58 'Alternative requirements for waste classification and characteristics' could be used to allow for an alternative disposal criterion based on site-specific hydro-geological characteristics and end land use scenarios. And, the third is to develop a technical basis for risk informed disposal regulations in anticipation of a 10CFR 61 rulemaking. By the summer of 2015 these initiatives had been broadly addressed and resulted in the NRC issuing Revision 1 to the BTP in February of 2015. Revision 1 of the BTP (BTP 2015) is a significant re-write of the draft issued in May 2012 and a significant improvement over the 1995 BTP (BTP 1995). The NRC has altered many of the positions in the interest of clarification. The development of an Implementation Guide for applying BTP 2015 to industry-specific waste streams that is vetted and supported by a EPRI sponsored working group consisting of industry participants (including utilities, disposal sites, NRC and Agreement State representatives and industry experts) will ensure compliant and consistent application during waste generation, packaging, processing and disposal activities. The Implementation Guide will provide for a comparison to previous BTP guidance and provide clarification and common understanding of the key concepts contained within the revised BTP. These include averaging constraints, hot spots, waste streams, waste types, discrete items, blending, solidification, encapsulation and alternative approaches. EPRI plans to publish the BTP Implementation Guide in the second quarter of 2016 and it will be available to the public. (authors)

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DOE OSTI2019

Development of a Testbed for Pipeline Flushing - 19409

Pipelines which carry radioactive waste within the DOE complex should be properly flushed, particularly in the cases where stationary or moving beds of solid sediment occur, or lines are prone to hydrogen gas buildup. Present guidelines establish a minimum for the amount of water used for post-transfer flushing operations to achieve satisfactory cleaning of pipelines. However, further studies are needed to limit the amount of water used for flushing operations, thus minimizing the consequent downstream waste. These efforts are significantly helpful to DOE waste remediation sites by preserving tank storage, preventing additional waste processing, and minimizing dilution and changes in waste chemistry. An experimental pipe loop designed for study of non-Newtonian slurry flushing is currently being developed at the Florida International University (FIU). The testbed is designed to investigate parameters that effect the efficiency of flushing operations. The objective is to find flush velocity values/modes which lead to satisfactory cleaning of transport lines with a minimum amount of water usage. The developed methodologies will be streamlined according to the demands from both Hanford and Savannah River sites. A test matrix containing parameters such as pipeline length, flow rate, and simulant composition (solid concentration, rheology, particle size and density) has been developed. In the initial phase of testing, material composition and loop length will be fixed to a mixture of 17 wt% kaolin in water (to attain density, viscosity, and yield stress in the applicable range to operations at Savannah River and Hanford sites) and 165 feet, respectively. This matrix provides a parametric study on the flush velocity. Prescribed velocities in this test matrix comply with the maximum and minimum recommended flush velocity values by the flushing guidelines for Hanford waste (RPP-RPT-59600, Rev. 0, SRNL-STI-2015-00014, TFC-ENG-STD-26, REV C-1). Efforts will focus on creating repeatable sediment beds inside the pipeline in both fully-flooded and gravity-drained systems. Bed characterization (i.e., measurement of sediment height, solids concentration, and rheology) will be conducted to ensure consistency of initial conditions between tests. A combination of Pulse-Echo technique, Coriolis meter, and optical inspections will be used to accomplish initial bed characterization (ensuring consistency between tests), control/monitor density variations (initial bed creation, flush velocity control, and post-flush evaluations), and serve as reference (validation of PulseEcho and pipe surface cleanness evaluation), respectively. This paper will present the details of the development of the test loop and the test matrix. (authors)

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