Search NASA⌕ Search

SEARCH · Search NASA

Results for “Waste Transfer”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,063 records · Page 59
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

High Performance Computing to Enable Next-Generation Low-Temperature Waste Heat Recovery

US manufacturing industry fails to recover an estimated 900 trillion BTUs of low-temperature waste heat from its processes each year. A grand research challenge has been to develop waste heat recovery technology that can be applied to industrial manufacturing processes and vehicle operations. The Oak Ridge National Laboratory (ORNL)/Eaton research team performed computational research and development to design an innovative direct-contact heat exchanger (DCHE) technology to deliver low-cost, compact, longer-lifetime, high-efficiency waste heat recovery that is optimized for a low-temperature organic Rankine cycle. ORNL resources and expertise in high performance computing (HPC) and multiphase flows were utilized to realize this goal while advancing the fundamental understanding of two-phase, two-immiscible-fluid turbulent flow heat transfer for heat exchangers. In the first stage of the project, a computational fluid dynamics (CFD) model of a DCHE was developed. Verification and limited validation of the numerical solution were performed using experimental data from published literature. To support efforts in verification and validation, two-dimensional CFD models were also developed with the objective of investigating different boundary conditions to reach converged, steady-state solutions.Based on benchmarking efforts, a baseline industry-grade design of a DCHE was developed, which consists of two horizontal pipes of different cross-sectional areas – joined by a converging-diverging nozzle in the middle and relies on a three-phase system with co-current flows for operation. Evaporation of the cold liquid (n-pentane or methanol) by the hot liquid water occurs in the first pipe. The first pipe contains two inlets for liquid water and liquid pentane (or liquid methanol). The second pipe with the larger cross-sectional area serves as a gravity-driven phase-separator. It contains two outlets that are staggered to avoid entrainment of the liquid water phase by the gas phase. This baseline design resulted from several trial-and-error evaluations using two-dimensional CFD models. Confirmatory three-dimensional CFD models were also run for different mass flow rates and inlet temperatures (inlet water temperature ranging from 50°C to 90°C), and efficiency plots were produced to fully characterize the proposed industrial-grade design of the DCHE.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2012

Modeling Ion-Exchange Processing With Spherical Resins For Cesium Removal

The spherical Resorcinol-Formaldehyde and hypothetical spherical SuperLig(r) 644 ion-exchange resins are evaluated for cesium removal from radioactive waste solutions. Modeling results show that spherical SuperLig(r) 644 reduces column cycling by 50% for high-potassium solutions. Spherical Resorcinol Formaldehyde performs equally well for the lowest-potassium wastes. Less cycling reduces nitric acid usage during resin elution and sodium addition during resin regeneration, therefore, significantly decreasing life-cycle operational costs. A model assessment of the mechanism behind ''cesium bleed'' is also conducted. When a resin bed is eluted, a relatively small amount of cesium remains within resin particles. Cesium can bleed into otherwise decontaminated product in the next loading cycle. The bleed mechanism is shown to be fully isotherm-controlled vs. mass transfer controlled. Knowledge of residual post-elution cesium level and resin isotherm can be utilized to predict rate of cesium bleed in a mostly non-loaded column. Overall, this work demonstrates the versatility of the ion-exchange modeling to study the effects of resin characteristics on processing cycles, rates, and cold chemical consumption. This evaluation justifies further development of a spherical form of the SL644 resin.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Feedback from International Decommissioning Planning Projects - 20269

Decommissioning of Nuclear Power Plants is an increasing reality in many countries with nuclear power. Already many years before final shutdown, planning of the decommissioning work is usually started in order to understand for example the costs, waste volume capacity and licensing work needed. In many countries it is required to have a final decommissioning plan before any decommissioning work is allowed to start. Westinghouse has been involved in international decommissioning planning for more than 30 years, where the first 20 years was mostly funding and preliminary plans due to that decommissioning was many years away. However, as many more plants are now moving into decommissioning, the focus has shifted to final decommissioning plans with budgets instead of funding plans, strategic decisions instead of choices, waste volume specifications instead of estimates and a defined approach instead of what ifs. Westinghouse has developed a methodology based on the planning and execution of many decommissioning projects in Europe, America and Asia. That international experience and understanding helps the customers to adopt a systematic planning process where strategy development is done with the end state in mind, where the goals are clearly outlined and the starting point described. The process and steps to reach the desired end state need to be clearly defined and all options/alternatives evaluated from a cost/benefit point of view. This is followed by a thorough and quantitative risk analysis in order to give solid recommendations and credible conclusions. One important feedback has been that real decommissioning experience from reference projects is invaluable for an efficient and timely progress in order not to spend unnecessary time and effort on understanding the barriers and links between different activities in decommissioning. Skilled project managers or engineers can drive the project forward, but without decommissioning experience previously gained, valuable lessons learned are not used and the same mistakes can be made again. Many utilities have invested in knowledge transfer from other decommissioning projects through site visits, exchange programs and supplier meetings, which is a valuable asset and way of implementing lessons learned from performed decommissioning projects into the planning process. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI1996

Nuclear Safety [Vol. 37, No. 1, January-March 1996]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 A Decision Support System for Maintenance Management of a Boiling-Water Reactor Power Plant, J. H. Shen, A. Ray, and S. Levine; ACCIDENT ANALYSIS: 12 On Prediction of the Ignition Potential of Uranium Metal and Hydride, M. Epstein, W. Luangdilok, M. G. Plys, and H. K. Fauske; 26 An Overview of the Primary Parameters and Methods for Determining Condensation Heat Transfer to Containment Structures, J. Green and K. Almenas; DESIGN FEATURES: 49 Modem Tornado Design of Nuclear and Other Potentially Hazardous Facilities, J. D. Stevenson and Y. Zaho; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 73 1994 Accident Sequence Precursor Program Results, R. J. Belles, J. W. Cletcher, D. A. Coplnger, B. W. Dolan, J. W. Minarick, and P. D. O'Reilly; ANNOUNCEMENTS: 93 American Institute of Chemical Engineers (AlChE) Spring 1997 National Meeting; 94 European Safety and Reliability Association International Conference on Safety and Reliability ESREL ’97; 95 Criticality Safety Challenges in the Next Decade; 96 21st International Symposium on the Scientific Basis for Nuclear Waste Management; 84 The Authors; 88 Letter to the Editor; 90 Indexes to Nuclear Safety, Volume 36.

05 NUCLEAR FUELS↗
DOE OSTI1996

Nuclear Safety [Vol. 37, No. 1, January-March 1996]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 A Decision Support System for Maintenance Management of a Boiling-Water Reactor Power Plant, J. H. Shen, A. Ray, and S. Levine; ACCIDENT ANALYSIS: 12 On Prediction of the Ignition Potential of Uranium Metal and Hydride, M. Epstein, W. Luangdilok, M. G. Plys, and H. K. Fauske; 26 An Overview of the Primary Parameters and Methods for Determining Condensation Heat Transfer to Containment Structures, J. Green and K. Almenas; DESIGN FEATURES: 49 Modem Tornado Design of Nuclear and Other Potentially Hazardous Facilities, J. D. Stevenson and Y. Zaho; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 73 1994 Accident Sequence Precursor Program Results, R. J. Belles, J. W. Cletcher, D. A. Copinger, B. W. Dolan, J. W. Minarick, and P. D. O'Reilly; ANNOUNCEMENTS: 93 American Institute of Chemical Engineers (AlChE) Spring 1997 National Meeting; 94 European Safety and Reliability Association International Conference on Safety and Reliability ESREL ’97; 95 Criticality Safety Challenges in the Next Decade; 96 21st International Symposium on the Scientific Basis for Nuclear Waste Management; 84 The Authors; 88 Letter to the Editor; 90 Indexes to Nuclear Safety, Volume 36.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Dry Cask Shipments of ATR Spent Nuclear Fuel at INL

The High Load Charger (HLC) cask will be replacing the Advanced Test Reactor (ATR) spent fuel transfer cask that has been used at Idaho National Laboratory (INL) to wet-transfer spent ATR fuel elements from ATR to Idaho Nuclear Technology and Engineering Center (INTEC) in order to support future plans at both facilities. INTEC intends to discontinue use of the CPP-666 storage pools in compliance with the Idaho Settlement Agreement and the basins will no longer be available to store spent ATR fuel elements. In order to support the core-internals change out in 2021 for ATR, spent ATR fuel elements must be removed from the canal and shipped to INTEC. INL contractors will use the HLC cask for safe, dry transfer of spent Advanced Test Reactor (ATR) fuel elements between the ATR Complex managed by Battelle Energy Alliance, LLC and the Idaho Nuclear Technology and Engineering Center (INTEC) managed by Fluor Idaho, LLC. The HLC cask (also known as the High Load Charging Cask or High Load Charger) is a lead-shielded, stainless steel transfer container originally designed in 1954 by the Idaho Nuclear Corporation to meet the radiation and contamination controls requirements of the time for shipment of spent nuclear fuel. The cask has been modified over the years and is currently used for wet-to-dry transfers at INTEC. It is approved only for out-of-commerce transfers. The cask will arrive at the ATR where eight spent ATR fuel elements will be loaded into two, newly-designed buckets. Each bucket is a quarter section, right-cylinder shape that has four compartments. The cask will be lifted out of the ATR canal allowing most of the water to drain from the cask. Processing the 12,500 kg cask will be completed in ATR and then transported on a tractor trailer to INTEC where the loaded buckets will be removed from the cask. The loaded buckets will undergo a drying process before placing the four buckets into a canister for long-term storage in the Irradiated Fuel Storage Facility (IFSF) at INTEC. A new transportation plan that was submitted to, and approved by, the U.S. Department of Energy to provide the cask safety basis, which describes the hazard analysis and accident analysis for the cask during transport. Supporting accident analysis include criticality, thermal, dose, and drop analyses. Of particular interest for this cask was the drop analysis from which was derived the speed limit for the transport. The drop analysis addressed drops in air onto an unyielding surface and in water onto an energy-absorbing pad for the cask in the ATR facility. These analyses are typically applied as the bounding drop scenarios during transport; however, this resulted in a safety basis speed limit of 12.9 kph. In order to more accurately represent the in-transit accident scenario, LS-DYNA software was procured and the model was updated to more closely represent the potential drop heights and impact surfaces of asphalt and soil for the cask while in transit. This analysis resulted in a safety basis speed limit of 48 kph. This safety basis speed limit is more practical to implement and control.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Sample Results from MCU Solids Outage

Savannah River National Laboratory (SRNL) has received several solid and liquid samples from MCU in an effort to understand and recover from the system outage starting on April 6, 2014. SRNL concludes that the presence of solids in the Salt Solution Feed Tank (SSFT) is the likely root cause for the outage, based upon the following discoveries: A solids sample from the extraction contactor #1 proved to be mostly sodium oxalate; A solids sample from the scrub contactor#1 proved to be mostly sodium oxalate; A solids sample from the Salt Solution Feed Tank (SSFT) proved to be mostly sodium oxalate; An archived sample from Tank 49H taken last year was shown to contain a fine precipitate of sodium oxalate; A solids sample from ; A liquid sample from the SSFT was shown to have elevated levels of oxalate anion compared to the expected concentration in the feed. Visual inspection of the SSFT indicated the presence of precipitated or transferred solids, which were likely also in the Salt Solution Receipt Tank (SSRT). The presence of the solids coupled with agitation performed to maintain feed temperature resulted in oxalate solids migration through the MCU system and caused hydraulic issues that resulted in unplanned phase carryover from the extraction into the scrub, and ultimately the strip contactors. Not only did this carryover result in the Strip Effluent (SE) being pushed out of waste acceptance specification, but it resulted in the deposition of solids into several of the contactors. At the same time, extensive deposits of aluminosilicates were found in the drain tube in the extraction contactor #1. However it is not known at this time how the aluminosilicate solids are related to the oxalate solids. The solids were successfully cleaned out of the MCU system. However, future consideration must be given to the exclusion of oxalate solids into the MCU system. There were 53 recommendations for improving operations recently identified. Some additional considerations or additional details are provided below as recommendations. From this point on, IC-Anions analyses of the DSSHT should be part of the monthly routine analysis in order to spot negative trends in the oxalate leaving the MCU system. Care must be taken to monitor the oxalate content to watch for sudden precipitation of oxalate salts in the system; Conduct a study to optimize the cleaning strategy at ARP-MCU through decreasing the concentration or entirely eliminating the oxalic acid; The contents of the SSFT should remain unagitated. Routine visual observation should be maintained to ensure there is not a large buildup of solids. As water with agitation provided sufficient removal of the solids in the feed tank, it should be considered as a good means for dissolving oxalate solids if they are found in the future; Conduct a study to improve prediction of oxalate solubility in salt batch feed materials. As titanium and mercury have been found in various solids in this report, evaluate if either element plays a role in oxalate solubility during processing; Salt batch characterization focuses primarily on characterization and testing of unaltered Tank 21H material; however, non-typical feeds are developed through cleaning, washing, and/or sump transfers. As these solutions are processed through MCU, they may precipitate solids or reduce performance. Salt batch characterization and testing should be expanded to encompass a broader range of feeds that may be processed through ARPMCU.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRSDate not supplied

Voxel Based Three-Dimensional Topology Optimization of Heat Exchanger Fins

Increasing interest in novel aircraft propulsion systems and potential for unwanted heat generation, or capture and reuse of waste heat, may require increasingly lightweight and high performing heat exchangers. Advances in manufacturing technologies have shown potential to create complex designs, but design tools need more flexibility. This study utilizes genetic algorithm-driven topology optimization to develop high performance heat exchanger fins for critical applications such as aerospace. The solid domain is generated using voxel representation where a voxel value of 1 indicates the solid domain and a voxel value of 0 refers to the fluid domain. The use of voxel representation somewhat matches the digitization of a model that is required to fabricate using additive manufacturing, and also allows for a highly unconstrained geometry. To test the topology optimization approach, a three-dimensional(3D)baseline offset strip fin geometry in steady laminar flow(Reynolds number = 215)with conjugate heat transfer(simultaneous solution of solid and fluid temperature fields)is optimized. New designs are generated using the genetic algorithm (GA) and sent to evaluation by the CFD software OpenFOAM; then the GA sorts and selects the reproduction pool for the following generation. This process is repeated for 60 generations. The study also investigates the effect off in material on the performance of the GA and the resulting designs. The results show that the optimal designs have overall performance improvement of 18% relative to the baseline. Additionally, a fin constructed of a lower conductivity material (such as an Inconel super alloy that might be necessary for waste heat recovery applications)results in lower overall performance improvement (11%)and optimal designs with higher pressure drop relative to their baseline, and relative to optimal designs produced using higher conductivity materials.

3D Optimization↗
DOE OSTI2026

Effect of polymer structure and material properties on mechanochemical reaction environments for polymer recycling

Plastic waste accumulation necessitates innovative recycling approaches to achieve sustainability goals. Mechanochemical depolymerization offers a solvent-free, energy-efficient route to convert polymers into valuable monomers. In addition to their chemical properties, the way that polymers absorb kinetic energy is a key parameter of any mechanochemical process. This perspective explores the principles underpinning mechanochemical recycling, emphasizing how deformation and localized transient heating mediate energy transfer between impacts and localized excitations. Key factors such as polymer crystallinity, molecular weight, viscoelasticity, and thermal effects are analyzed to elucidate their role in energy transfer mechanisms during ball milling. This work establishes a foundational framework for the design and optimization of mechanochemical recycling by connecting polymer response to mechanical energy with the intention to improve depolymerization efficiency. Future research opportunities are outlined to advance the integration of polymer science and mechanochemistry for scalable, sustainable plastic upcycling.

ball mill↗
DOE OSTI2021

Small to Full-Height Scale Comparisons of Cesium Ion Exchange Performance with Crystalline Silicotitanate

Crystalline silicotitanate (CST) ion exchanger is planned to be used to remove cesium (137Cs) from the aqueous phases of Hanford tank wastes in preparation for vitrification at the Waste Treatment and Immobilization Plant (WTP) Pretreatment Facility. Column scale up testing was conducted to evaluate performance of Cs exchange onto the CST. Batch contact testing was conducted to assess exchange kinetics of Cs exchange at four different CST PSDs. Column testing was conducted at three column sizes, small (2.5% full height), medium (12% full height), and full height, to assess Cs load performance behavior. Testing at the small scale was compared to actual Hanford tank waste testing to verify the validity of simulant tests to accurately represent full height column performance. A change in CST particle size was essential in scaling the small column dynamics up to full scale. Based on these results, a determination of intraparticle and film diffusion impacts on overall mass transfer coefficients will allow future modeling of the breakthrough performance at a range of process conditions.

Westesen, Amy M.↗
NASA NTRS2000

Microreactor System Design for a NASA In Situ Propellant Production Plant on Mars

The NASA In Situ Resource Utilization (ISRU) program is planning near-term missions to Mars that will include chemical processes for converting the carbon dioxide (CO2) and possibly water from the Martian environment to propellants, oxygen, and other useful chemicals. The use of indigenous resources reduces the size and weight of the payloads from Earth significantly, representing enormous cost savings that make human exploration of Mars affordable. Extraterrestrial chemical processing plants will need to be compact, lightweight, highly efficient under reduced gravity, and extraordinarily reliable for long periods. Microchemical and thermal systems represent capability for dramatic reduction in size and weight, while offering high reliability through massive parallelization. In situ propellant production (ISPP), one aspect of the ISRU program, involves collecting and pressurizing atmospheric CO2, conversion reactions, chemical separations, heat exchangers, and cryogenic storage. A preliminary system design of an ISPP plant based on microtechnology has demonstrated significant size, weight, and energy efficiency gains over the current NASA baseline. Energy management is a strong driver for Mars-based processes, not only because energy is a scarce resource, but because heat rejection is problematic; the low pressure environment makes convective heat transfer ineffective. Energy efficiency gains are largely achieved in the microchemical plant through extensive heat recuperation and energy cascading, which has a small size and weight penalty because the added micro heat exchangers are small. This leads to additional size and weight gains by reducing the required area of waste heat radiators. The microtechnology-based ISPP plant is described in detail, including aspects of pinch analysis for optimizing the heat exchanger network. Three options for thermochemical compression Of CO2 from the Martian atmosphere, adsorption, absorption, and cryogenic freezing, are presented, as well as three options for water decomposition, low temperature electrolysis, high temperature electrolysis, and thermochemical decomposition. Other elements of the plant include Sabatier and reverse water gas shift reactors, water recovery, chemical separations, and cryogenic storage. Data are presented supporting preliminary sizing of components, and results of the system design are compared to the existing NASA baseline that is based on conventional technologies.

TeGrotenhuis, W. E.↗
NASA NTRS1996

Significance of Plant Root Microorganisms in Reclaiming Water in CELSS

Since many microorganisms demonstrate the ability to quickly break down complex mixtures of waste and environmental contaminants, examining their potential use for water recycling in a closed environment is appealing. Water contributes approximately 90 percent of the life sustaining provisions in a human space habitat. Nearly half of the daily water requirements will be used for personal hygiene and dish washing. The primary contaminants of the used "gray" water will be the cleansing agents or soaps used to carry out these functions. Reclaiming water from the gray water waste streams is one goal of the NASA program, Controlled Ecological Life Support Systems (CELSS). The microorganisms of plane roots are well documented to be of a beneficial effect to promote plant growth. Most plants exhibit a range of bacteria and fungi which can be highly plant-specific. In our investigations with lettuce grown in hydroponic culture, we identified a microflora of normal rhizosphere. When the roots were exposed to an anionic surfactant, the species diversity changed, based on morphological characteristics, with the numbers of species being reduced from 7 to 2 after 48 hours of exposure. In addition, the species that became dominant in the presence of the anionic surfactant also demonstrated a dramatic increase in population density which corresponded to the degradation of the surfactant in the root zone. The potential for using these or other rhizosphere bacteria as a primary or secondary waste processor is promising, but a number of issues still warrant investigation; these include but are not limited to: (1) the full identification of the microbes, (2) the classes of surfactants the microbes will degrade, (3) the environmental conditions required for optimal processing efficiency and (4) the ability of transferring the microbes to a non-living solid matrix such as a bioreactor.

Bubenheim, David L.↗
NASA NTRS2000

Near-Net-Shape Processing of Sintered Fibrous Ceramics Achieved

A variety of sintered fibrous ceramic (SFC) materials have been developed over the last 50 years as thermal barrier materials for reentry applications. SFC materials typically exhibit very low thermal conductivities combined with low densities and good thermal stability up to 2500 F. These materials have flown successfully on the space shuttle orbiters since the 1960's. More recently, the McDonnell Douglas Corporation successfully used SFC tiles as a heat shield on the underside of its DC X test vehicle. For both of these applications, tiles are machined from blocks of a specific type of SFC called an alumina-enhanced thermal barrier (AETB). The sizes of these blocks have been limited by the manufacturing process. In addition, as much as 80 to 90 percent of the material can be lost during the machining of tiles with significant amounts of curvature. To address these problems, the NASA Glenn Research Center at Lewis Field entered a cooperative contract with the Boeing Company to develop a vacuum-assisted forming process that can produce large (approximately 4 square feet), severely contoured panels of AETB while saving costs in comparison to the conventional cast-and-machine billet process. For shuttle use, AETB is slurry cast, drained, and fired to form square billets conforming to the shape of the filtration box. The billets are then cut into tiles of the appropriate size for thermally protecting the space shuttle. Processing techniques have limited the maximum size of AETB billets to 21.5 square inches by 6.5-in. thick, but the space shuttles use discrete heat shield tiles no more than 8 to 12 square inches. However, in other applications, large, complex shapes are needed, and the tiling approach is undesirable. For such applications, vacuum-assisted forming can produce large parts with complex shapes while reducing machining waste and eliminating cemented joints between bonded billets. Because it allows contoured shapes to be formed, material utilization is inherently high. Initial estimates show that the amount of material lost during machining can be reduced by 50 percent or more. In addition, a fiber alignment favorable for minimum heat transfer is maintained for all panel shapes since the fibers are aligned parallel to the contoured surface of the forming tool or mold. The vacuum-assisted forming process can complete the entire forming operation in a matter of minutes and can produce multiple parts whose size is limited only by the size of the forming tool. To date, panels as large as 2 square feet have been demonstrated The vacuum-assisted forming process starts with the fabrication of a permeable forming tool, or mold, with the proper part contour. This reusable tool is mounted over an internal rib support structure, as depicted in the diagram, such that a vacuum can be pulled on the bottom portion of the tool. AETB slurry is then poured over and around the tool, liquid is drawn from the slurry, and the part forms over the tool surface. The part is then dried, fired, and finished machined. Future plans include an evaluation of the need for additional coatings and surface-toughness treatments to extend the durability and performance of this material.

Angel, Paul W.↗
DOE OSTI2016

The Importance of Experience Based Decommissioning Planning - 16242

Decommissioning of a nuclear facility is an extensive and multidisciplinary task which involves the management and technical actions associated with ceasing operation and thereafter the step by step transfer of the facility from an operating plant to an object under decommissioning. The decommissioning phase includes dismantling of systems and components, decontamination and clearance, demolition of buildings, remediation of any contaminated ground and finally a survey of the site. Several of these activities generate radioactive or potentially radioactive waste which has to be managed properly prior to clearance or disposal. What makes decommissioning of nuclear installations unique is to a large extent the radioactive waste management. No other industries have that complex regulatory frameworks for waste management. If a decommissioning project in the nuclear industry does not consider the waste aspects to the extent required, there is a large risk of failure causing a reduced trust by the regulators and other stakeholders as well as cost and schedule overruns. This paper will give an overview of important aspects and findings gathered during decades of planning and conducting decommissioning and nuclear facility modernization projects. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

ENRESA's Spent Nuclear Fuel and Special Waste Database - 18115

Empresa Nacional de Residuos Radiactivos S.A. (ENRESA, by its initials in Spanish) Spent Nuclear Fuel (SNF) and Special Waste (SW) database development was initiated in a common effort between ENRESA and UNESA (the Spanish Association of the Electricity Industry) as a consequence of new regulatory requirements in the frame of radioactive waste management plans. Although these requirements were only mandatory for operators of nuclear power plants, ENRESA deemed necessary to have a centralized database which collected the data from the plant operators. The most relevant features of this database are: - The information collected covers not only the general characteristics of the SNF and SW but also the operation details needed for their characterization. - An application, named Gestion de Elementos Combustibles y Residuos Especiales (GECYRE, by its initials in Spanish) has been implemented for easing the queries and the generation of predefined reports. - An agile data transfer scheme, based on XML files, is used for an effective communication between ENRESA and the operators. The database is currently in production phase since 2014. During this phase it has demonstrated its effectiveness as an information repository to address the SNF and SW interim storage projects in Spain foreseen in the near future. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Characterization of Tank 9H Salt Dissolution Batch 2A in Support of Tank Closure Cesium Removal (TCCR) 1A Batch 2 Preparations

Savannah River Remediation (SRR) is currently preparing the second batch of material to be processed through the Tank Closure Cesium Removal (TCCR) 1A system. The feed for TCCR 1A consists of dissolved saltcake from Tank 9H. The first batch of salt to make up Batch 2 (Batch 2A) has been dissolved in Tank 9H and will later be transferred to Tank 10H. Savannah River National Laboratory (SRNL) received samples from the batch of dissolved salt for characterization.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Management of the Cs/Sr Capsule Project at the Hanford Site. Technology Readiness Assessment Report

The Federal Project Director (FPD) for the U.S. Department of Energy (DOE), Richland Operations Office (RL) Waste Management and D&D Division (WMD) requested a Technology Readiness Assessment (TRA) for the Management of the Cesium/Strontium Capsule Storage Project (MCSCP) at the Waste Encapsulation and Storage Facility (WESF) on the Hanford Site in Washington State. The MCSCP CD-1 TRA was performed by a team selected in collaboration between the Office of Environmental Management (EM) Chief Engineer (EM-3.3) and RL, WMD FPD. The TRA Team included subject matter and technical experts having experience in cask storage, process engineering, and system design who were independent of the MCSCP, and the team was led by the Director of Operations and Processes from the EM Chief Engineer's Office (EM-3.32). Movement of the Cs/Sr capsules to dry storage, based on information from the conceptual design, involves (1) capsule packaging, (2) capsule transfer, and (3) capsule storage. The project has developed a conceptual process, described in 30059-R-02, "NAC Conceptual Design Report for the Management of the Cesium and Strontium Capsules Project", which identifies the five major activities in the process to complete the transfer from storage pool to pad-mounted cask storage. The process, shown schematically in Figure 1, is comprised of the following process steps: (1) loading capsules into the UCS; (2) UCS processing; (3) UCS insertion into the TSC Basket; (4) cask transport from WESF to CSA and (5) extended storage at the CSA.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2010

Cryogenic Transport of High-Pressure-System Recharge Gas

A method of relatively safe, compact, efficient recharging of a high-pressure room-temperature gas supply has been proposed. In this method, the gas would be liquefied at the source for transport as a cryogenic fluid at or slightly above atmospheric pressure. Upon reaching the destination, a simple heating/expansion process would be used to (1) convert the transported cryogenic fluid to the room-temperature, high-pressure gaseous form in which it is intended to be utilized and (2) transfer the resulting gas to the storage tank of the system to be recharged. In conventional practice for recharging high-pressure-gas systems, gases are transported at room temperature in high-pressure tanks. For recharging a given system to a specified pressure, a transport tank must contain the recharge gas at a much higher pressure. At the destination, the transport tank is connected to the system storage tank to be recharged, and the pressures in the transport tank and the system storage tank are allowed to equalize. One major disadvantage of the conventional approach is that the high transport pressure poses a hazard. Another disadvantage is the waste of a significant amount of recharge gas. Because the transport tank is disconnected from the system storage tank when it is at the specified system recharge pressure, the transport tank still contains a significant amount of recharge gas (typically on the order of half of the amount transported) that cannot be used. In the proposed method, the cryogenic fluid would be transported in a suitably thermally insulated tank that would be capable of withstanding the recharge pressure of the destination tank. The tank would be equipped with quick-disconnect fluid-transfer fittings and with a low-power electric heater (which would not be used during transport). In preparation for transport, a relief valve would be attached via one of the quick-disconnect fittings (see figure). During transport, the interior of the tank would be kept at a near-ambient pressure far below the recharge pressure. As leakage of heat into the tank caused vaporization of the cryogenic fluid, the resulting gas would be vented through the relief valve, which would be set to maintain the pressure in the tank at the transport value. Inasmuch as the density of a cryogenic fluid at atmospheric pressure greatly exceeds that of the corresponding gas in a practical high-pressure tank at room temperature, a tank for transporting a given mass of gas according to the proposed method could be smaller (and, hence, less massive) than is a tank needed for transporting the same mass of gas according to the conventional method.

Ungar, Eugene K,↗