Engineering Papers⌕ Search

SEARCH · Engineering Papers

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 613 records · Page 34

Application of evolved gas analysis to cold-cap reactions of melter feeds for nuclear waste vitrification

The melter feed (a mixture of nuclear waste and glass forming and modifying additives) experiences multiple gas-evolving reactions in an electrical glass-melting furnace during the vitrification of nuclear wastes. Foams from the residual gases can significantly alter the melting rate through mass and heat transfers. Here, we employed the thermogravimetry-gas chromatography-mass spectrometry (TGA-GC-MS) combination to perform quantitative evolved gas analysis (EGA) and developed a simple calibration model which correlates the overall mass loss rate with the evolution rates for individual gases. The model parameters are obtained from the least squares analysis, assuming that the gas-evolving reactions are independent. Finally, the EGA adds the ‘chemical identity’ to the reactions indicated by the ‘phenomenological’ kinetic model.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Evaluation of the Potential for Precipitation of Solids during Storage of Non-Aluminum SNF Solutions

Non-aluminum clad spent nuclear fuels (NASNF) stored in the L-Area basin will be dissolved in H-Canyon using the 6.3D electrolytic dissolver. The solutions will be stored in either the hot or warm canyon until the preparation of a sludge batch for the Defense Waste Processing Facility. Spent nuclear fuel solutions could be stored for 1-2 years before transfer to the H-Area Tank Farm depending on the interval between sludge batches. The solution level in the storage tanks will be maintained; therefore, precipitation of solids due to evaporation is not an issue. However, the precipitation of solids from completely dissolved SNF due to solution instabilities has been observed during intermediate storage of solutions generating hydrated oxides.The presence of fissile material in these solids is generally associated with zirconium molybdate, which is known to act as a host lattice for Pu and can carry the actinides upon precipitation. The formation of zirconium molybdate solids which carry fissile material is a potential concern for the storage of NASNF solutions. To address this concern, the Savannah River National Laboratory performed a literature review to identify knowledge gaps which may require experimental work to determine if the formation of solids is a concern during storage of these solutions. Based on the literature review, the precipitation of zirconium molybdate solids from the Campaign 1 NASNF solutions during intermediatestorage is expected. This conclusion is supported by the identification of zirconium molybdate solids found on the H-Canyon 6.1D Dissolver MK-12 insert spacer. The formation of the zirconium molybdate solids is attributed to hydrolysis and radiolytic processes in the nitric acid solution. As the molybdate solids form, U and Pu can substitute for Zr in the crystal lattice resulting in co-precipitation. Generally, the Pu substitutes directly into the crystal lattice during precipitation while the U associated with the molybdate solids more likely absorbs from the solution. The U in the NASNF solutions is present as uranyl nitrate, a 2+ cation which will not substitute as easily into the molybdate crystal lattice for the Zr 4+ ion.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Corrosion Evaluation for Tank 241-AP-106 Re-purposing

Tank AP-106 will be an integral part of the Direct Feed Low Activity Waste (DF-LAW) activity. This tank is an underground, carbon steel, double shell tank (DST). Before it can be utilized, the Cs-137 concentration levels in the waste must be reduced significantly. The AP-106 re-purposing process will manage normal waste tank additions procedures such that waste acceptance criteria objectives for Cs-137 removal are achieved. As this process was being developed, concerns regarding tank integrity during and after the process to Tank AP-106 and associated tanks (i.e., AP-102, AN-101, and AP-108) were addressed. The degradation mechanisms identified include: general and pitting corrosion, caustic stress corrosion cracking, galvanic corrosion, and microbiologically induced corrosion (MIC). Recommendations were provided to minimize the exposure of the tank to corrosive conditions (see Section 6.0). These recommendations included chemistry adjustments at defined steps in the process that will maintain the waste in compliance with the corrosion control program. Additionally, sampling and inspection (see Sections 6.0 and 7.0, respectively) recommendations that will assess the final condition of the tank environment and the tank were presented. The greatest risk for this process involves the handling of large quantities of Columbia River water (i.e., greater than 450,000 gallons) within the AP-106, associated tanks and supporting infrastructure (i.e., transfer line piping). As long as this water is blended into the supernate in a timely fashion, and the waste is adjusted appropriately with corrosion inhibitors, the final condition is expected to be within compliance. Recommendations were also made to address process interruptions that could result in exposure of the tank walls to aggressive conditions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

A heat and mass transfer model for evaluation of damaged cesium chloride radiological sources

Radiological sources are vital to many applications, and typically contain byproducts from waste streams which decay over time to form complex mixtures of elements. Daughter products resulting from these decay processes can introduce complicating factors to the integrity and safety of the vessel that contains the radiological material. Sources containing cesium 137 ( 137 Cs) are of particular interest, because the decay of this isotope produces barium metal which reacts readily and exothermically with oxygen. This work employs physics-based numerical tools to examine the thermal response of a radiological source containing a mixture of cesium and barium in the event that vessel walls are damaged and atmospheric gases contact source material. A parametric study was conducted to determine the sensitivity of the response to various factors, including vessel geometry, source material age, the degree of vessel damage, and other parameters. Here, it was found that the peak temperatures that occurred within the source material strongly depend on these parameters, particularly vessel geometry and age, which determine whether a breach would be a relatively minor accident or a catastrophic incident. Finally, the model’s sensitivity to uncertain thermophysical properties is discussed.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Reduced-Order CFD Modeling to Support Waste Loading Optimization in Hanford WTP Vitrification

The U.S. DOE Hanford Site stores over 56 million gallons of radioactive liquid tank waste that must be treated and immobilized for long-term disposal The Waste Treatment and Immobilization Plant (WTP) will vitrify this waste by feeding it into Joule-heated melters, where it is incorporated into a stable borosilicate glass Computational fluid dynamics (CFD) simulations of glass melters can provide insight into the maximum achievable waste loading under varying melter operating conditions Fully resolved VOF multiphase simulations were used as the reference model to capture bubble-driven convection in the melter, including bubble formation, rise behavior, and induced glass melt circulation Effective bubble column diameter and rise velocity were extracted from the resolved simulations, compared with empirical correlations, and refit across relevant viscosity and gas flow rate conditions Explicit gas–liquid interface tracking was replaced with a single-phase momentum source term model, enabling faster steady-state CFD simulations while preserving the dominant hydrodynamic effects of bubbling New empirical correlations were developed for effective bubble column diameter and bubble rise velocity by fitting resolved simulation data across expected melter viscosity and gas flow rate ranges, providing improved inputs for the momentum source term model compared with existing literature correlations The momentum source term model reduced fluid-domain mesh size by 89% and achieved an 8.4× computational speedup relative to resolved bubbling simulations The validated momentum source term approach enables prediction of process-relevant heat transfer behavior in the integrated melter model, including heat transfer from the molten glass to the cold cap, plenum, refractory walls, and surrounding structural regions under varying melter operating conditions

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗

Tank 48 Chemical Destruction - 13237

Small tank copper-catalyzed peroxide oxidation (CCPO) is a potentially viable technology to facilitate the destruction of tetraphenylborate (TPB) organic solids contained within the Tank 48H waste at the Savannah River Site (SRS). A maturation strategy was created that identified a number of near-term development activities required to determine the viability of the CCPO process, and subsequent disposition of the CCPO effluent. Critical activities included laboratory scale validation of the process and identification of forward transfer paths for the CCPO effluent. The technical documentation and the successful application of the CCPO process on simulated Tank 48 waste confirm that the CCPO process is a viable process for the disposition of the Tank 48 contents. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Tank 48 - Chemical Destruction

Small tank copper-catalyzed peroxide oxidation (CCPO) is a potentially viable technology to facilitate the destruction of tetraphenylborate (TPB) organic solids contained within the Tank 48H waste at the Savannah River Site (SRS). A maturation strategy was created that identified a number of near-term development activities required to determine the viability of the CCPO process, and subsequent disposition of the CCPO effluent. Critical activities included laboratory-scale validation of the process and identification of forward transfer paths for the CCPO effluent. The technical documentation and the successful application of the CCPO process on simulated Tank 48 waste confirm that the CCPO process is a viable process for the disposition of the Tank 48 contents.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Evaluation of Pu Solubility in Glass for Sludge Batch 11

Stainless-steel clad Pu from Japan’s Fast Critical Assembly (FCA) reactor is currently being dispositioned at the Savannah River Site. The electrolytic dissolver, operated by Savannah River Nuclear Solutions in H-Canyon, is being utilized to dissolve the material. The resulting solutions are transferred to the Concentration, Storage, and Transfer Facilities, operated by Savannah River Mission Completion (SRMC), for subsequent vitrification at the Defense Waste Processing Facility (DWPF). In support of the FCA mission startup, a preliminary evaluation was conducted by the Savannah River National Laboratory to assess the impact of the FCA discards on the liquid waste system.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Skylab Trash Airlock

The Skylab Trash Airlock (TAL) used throughout the Skylab mission to transfer trash materials that could support microbial growth from the pressurized cabin to the unpressurized waste tank is described. The TAL, which uses several basic mechanisms, was successfully operated daily for the 170 days of manned missions for a total of 637 cycles.

Price, L. R.↗

An assessment of gas-side fouling in cement plants

The cement industry is the most energy-intensive industry in the United States in terms of energy cost as a percentage of the total product cost. An assessment of gas-side fouling in cement plants with special emphasis on heat recovery applications is provided. In the present context, fouling is defined as the buildup of scale on a heat-transfer surface which retards the transfer of heat and includes the related problems of erosion and corrosion. Exhaust gases in the cement industry which are suitable for heat recovery range in temperature from about 100 to 1300 K, are generally dusty, may be highly abrasive, and are often heavily laden with alkalies, sulfates, and chlorides. Particulates in the exhaust streams range in size from molecular to about 100 micrometers in diameter and come from both the raw feed as well as the ash in the coal which is the primary fuel used in the cement industry. The major types of heat-transfer equipment used in the cement industry include preheaters, gas-to-air heat exchangers, waste heat boilers, and clinker coolers. At the present time, the trend in this country is toward suspension preheater systems, in which the raw feed is heated by direct contact with the hot kiln exit gases, and away from waste heat boilers as the principal method of heat recovery. The most important gas-side fouling mechanisms in the cement industry are those due to particulate, chemical reaction, and corrosion fouling.

Marner, W. J.↗

Long-term Performance of Engineered Barrier Systems (PEBS) - An International EURATOM Project on the Study and Testing of Engineered Barriers for the Final Disposal of HAW Using PEBS as an Example - 13299

The main aim of the PEBS project is to evaluate the sealing and barrier performance of the EBS over time, through development of a comprehensive approach involving experiments, model development, and consideration of the potential impacts on long-term safety functions. The experiments and models cover the full range of conditions, from initial emplacement of wastes (high heat generation and EBS re-saturation) through to later stage establishment of near steady state conditions, i.e. full re-saturation and thermal equilibrium with the host rock. These aspects will be integrated in a manner that will lead to greater certainty and thus greater confidence regarding the development from the initial transient state of the EBS to its long-term state, which provides the required isolation of the wastes. The work proposed within the project builds on existing knowledge and experience generated during recent years and supported by ongoing national and EC research programs. The project aims to provide a more complete description of the THM and THMC (thermo-hydro-mechanical- chemical) evolution of the EBS system, a more quantitative basis for relating the evolutionary behavior to the safety functions of the system, and a further clarification of the significance of residual uncertainties for long-term performance assessment. The importance of uncertainties arising from potential disagreement between the process models and the laboratory and in-situ experiments to be performed within PEBS, and their implications for an extrapolation of the results, will be reviewed, with particular emphasis on possible impacts on safety functions. In addition to the scientific-technical aims, the consortium will disseminate the basic findings to the broad scientific community within the EU, China and Japan, use expertise gained for public information purposes, and promote knowledge and technology transfer through training. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Ion-Exchange Modeling of Crystalline Silicotitanate for Cesium Removal - 20283

The Tank Closure Cesium Removal (TCCR) system is a Savannah River Site (SRS) demonstration 'at-tank' process designed to remove {sup 137}Cs from the high-level aqueous tank waste so that the decontaminated solution can be disposed as low-level waste. Cesium is removed by ion exchange (IX) columns using engineered IONSIV{sup TM} R9120-B form of the Crystalline Silicotitanate (CST) media. The TCCR system is deployed at Tanks 10 and Tank 11 in the SRS H Tank Farm. Water is added to the salt-cake in Tank 10 H to dissolve it. The dissolved salt solution waste is pumped out of Tank 10H (feed tank), through filters and IX columns. The decontaminated salt solution is transferred to Tank 11 (receipt tank), and on to Tank 50H for final disposal in the Saltstone Production Facility. The current TCCR can accommodate lead-lag (two-column) or lead-lag-guard (three-column) configurations to optimize media utilization and achieve the target decontamination. To assist the TCCR operations, a parametric study was conducted to evaluate the impact of different parameters (e.g., column configurations (single column, two columns or three columns in series), waste characteristics, operating temperature, process flow rate, CST average particle size) on the IX column performance including CST bed utilization. The initial results indicate that the IX column performance is improved at slower process flow rate, at lower operating temperature, and with smaller CST average particle size. Multi-column configurations are recommended, because the single-column configuration does not utilize CST bed effectively. This paper demonstrates the versatility of the ion exchange modeling to evaluate the effects of CST characteristics and operational parameters on IX column performances. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Coupled Thermal-Hydrologic-Chemical Coupled Model for In-Drift Disposal Test

The simulation work presented in this report supports DOE-NE Used Fuel Disposition Campaign (UFDC) goals related to the development of drift scale in-situ field testing of heat-generating nuclear waste (HGNW) in salt formations. Numerical code verification and validation is an important part of the lead-up to field testing, allowing exploration of potential heater emplacement designs, monitoring locations, and perhaps most importantly the ability to predict heat and mass transfer around an evolving test. Such predictions are crucial for the design and location of sampling and monitoring that can be used to validate our understanding of a drift scale test that is likely to span several years.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Saltstone 3QCY11 TCLP Results

A Saltstone waste form was prepared in the Savannah River National Laboratory (SRNL) from a Tank 50H sample and Z-Area premix material for the third quarter of calendar year 2011 (3QCY11). After the prescribed 32 day cure, samples of the saltstone were collected, and the waste form was shown to meet the South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24 and R.61-79.268.48(a) requirements for a nonhazardous waste form with respect to RCRA metals and underlying hazardous constituents. These analyses met all quality assurance specifications of USEPA SW-846. The Saltstone Production Facility (SPF) receives waste from Tank 50H for treatment. In the third quarter of the 2011 calendar year (3QCY11), Tank 50H accepted transfers of approximately 20 kgal from the Effluent Treatment Project (ETP), approximately 236 kgal from the Actinide Removal Process/Modular Caustic Side Solvent Extraction Unit (ARP/MCU) Decontaminated Salt Solution Hold Tank (DSS-HT), and approximately 25 kgal from other sources. The Saltstone Grout Sampling plan provides the South Carolina Department of Health and Environmental Control (SCDHEC) with the chemical and physical characterization strategy for the salt solution which is to be disposed of in the Z-Area Solid Waste Landfill (SWLF). During operation, samples were collected from Tank 50H and grout samples prepared to determine the non-hazardous nature of the grout to meet the requirements of the South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24(b) and R.61-79.268.48(a). Savannah River National Laboratory (SRNL) was asked to prepare saltstone from samples of Tank 50H obtained July 7, 2011 during 3QCY11 to determine the non-hazardous nature of the grout. The samples were cured and shipped to Babcock & Wilcox Technical Services Group-Radioisotope and Analytical Chemistry Laboratory (B&W TSG-RACL) to perform the Toxic Characteristic Leaching Procedure (TCLP) and subsequent extract analysis on saltstone samples for the analytes required for the quarterly analysis saltstone sample. In addition to the eight toxic metals - arsenic, barium, cadmium, chromium, mercury, lead, selenium and silver - analytes included the underlying hazardous constituents (UHC) antimony, beryllium, nickel, and thallium which could not be eliminated from analysis by process knowledge. B&W TSG-RACL provided subsamples to GEL Laboratories, LLC for analysis for the UHCs benzene, phenols and total and amenable cyanide.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

In-situ X-ray and visual observation of foam morphology and behavior at the batch-melt interface during melting of simulated waste glass

We report to attain a basic understanding of the primary foam structure and behavior, which affects the heat and mass transfer and the efficiency of the glass melting process, we investigated the primary foam layer under the glass batch floating on molten glass. The primary foam affects mass transfer during batch melting, in turn affecting the melting process. The recently performed direct in-situ three-dimensional X-ray computed tomography of the batch melting in a laboratory-scale melter vessel allowed us to visualize the features of the reacting batch layer and the foam that develops at its bottom, though with an insufficient resolution of images. In this study, we obtained better temporal and spatial resolution using the two-dimensional X-ray radiography and visual observation of the structure and behavior of transient primary foam as it formed and decayed. As soon as the batch was charged onto the melt surface, foam bubbles began to evolve, grow, and coalesce, forming a primary foam layer, 5-10 mm thick, within tens of seconds. This foam layer was sustained by ongoing gas evolving reactions counterbalanced by bubble coalescence into cavities that moved sideways and escaped to the atmosphere. Eventually, the entire remaining batch turned into foam that gradually decayed at the melt surface. The decay rate agreed with literature observations of surface foam produced by secondary foaming.

36 MATERIALS SCIENCE↗

BEAM COLLIMATION IN THE PIP-II LINAC TO BOOSTER TRANSFER LINE

The new PIP-II superconducting linac will deliver a 2 mA average H⁻ beam to the existing Booster synchrotron. The injected beam is accumulated by charge exchange over approximately 300 turns; phase space painting is used to mitigate space charge effects. To limit the power load on the internal waste beam absorber from the transverse tails of the H⁻ distribution missing the foil, the beam will be collimated in both planes in the linac to Booster transfer line using compact collimators of a novel design. Both the number of parasitic hits and the fraction of the beam missing the foil are sensitive functions of the H⁻ beam centroid position with respect to the edge of the foil. The positioning of the collimation is constrained by the availability of suitable space in the transfer line lattice, by specifics of the collimator design, by the phase space orientation at the collimator, and by the betatron phase advance to the foil needed to achieve proper orientation of the spatial distribution at the injection point. In this contribution, we describe the procedure by which collimator positions were optimized. We then discuss the expected performance of the overall system.

Johnson, David E.↗

Safety aspects of nuclear waste disposal in space

Safety issues involved in the disposal of nuclear wastes in space as a complement to mined geologic repositories are examined as part of an assessment of the feasibility of nuclear waste disposal in space. General safety guidelines for space disposal developed in the areas of radiation exposure and shielding, containment, accident environments, criticality, post-accident recovery, monitoring systems and isolation are presented for a nuclear waste disposal in space mission employing conventional space technology such as the Space Shuttle. The current reference concept under consideration by NASA and DOE is then examined in detail, with attention given to the waste source and mix, the waste form, waste processing and payload fabrication, shipping casks and ground transport vehicles, launch site operations and facilities, Shuttle-derived launch vehicle, orbit transfer vehicle, orbital operations and space destination, and the system safety aspects of the concept are discussed for each component. It is pointed out that future work remains in the development of an improved basis for the safety guidelines and the determination of the possible benefits and costs of the space disposal option for nuclear wastes.

Rice, E. E.↗

Sorption Modeling and Verification for Off-Gas Treatment

The project has made progress toward developing a comprehensive modeling capability for the capture of target species in off gas evolved during the reprocessing of nuclear fuel. The effort has integrated experimentation, model development, and computer code development for adsorption and absorption processes. For adsorption, a modeling library has been initiated to include (a) equilibrium models for uptake of off-gas components by adsorbents, (b) mass transfer models to describe mass transfer to a particle, diffusion through the pores of the particle and adsorption on the active sites of the particle, and (c) interconnection of these models to fixed bed adsorption modeling which includes advection through the bed. For single-component equilibria, a Generalized Statistical Thermodynamic Adsorption (GSTA) code was developed to represent experimental data from a broad range of isotherm types; this is equivalent to a Langmuir isotherm in the two-parameter case, and was demonstrated for Kr on INL-engineered sorbent HZ PAN, water sorption on molecular sieve A sorbent material (MS3A), and Kr and Xe capture on metal-organic framework (MOF) materials. The GSTA isotherm was extended to multicomponent systems through application of a modified spreading pressure surface activity model and generalized predictive adsorbed solution theory; the result is the capability to estimate multicomponent adsorption equilibria from single-component isotherms. This advance, which enhances the capability to simulate systems related to off-gas treatment, has been demonstrated for a range of real-gas systems in the literature and is ready for testing with data currently being collected for multicomponent systems of interest, including iodine and water on MS3A. A diffusion kinetic model for sorbent pellets involving pore and surface diffusion as well as external mass transfer has been established, and a methodology was developed for determining unknown diffusivity parameters from transient uptake data. Two parallel approaches have been explored for integrating the kernels described above into a mass-transport model for adsorption in fixed beds. In one, the GSTA isotherm kernel has been incorporated into the MOOSE framework; in the other approach, a focused finite-difference framework and PDE kernels have been developed. Issues, including oscillatory behavior in MOOSE solutions to advection-diffusion problems, and opportunities have been identified for each approach, and a path forward has been identified toward developing a stronger modeling platform. Experimental systems were established for collection of microscopic kinetics and equilibria data for single and multicomponent uptake of gaseous species on solid sorbents. The systems, which can operate at ambient temperature to 250°C and dew points from -69 to 17°C, are useful for collecting data needed for modeling performance of sorbents of interest. Experiments were conducted to determine applicable models and parameters for isotherms and mass transfer for water and/or iodine adsorption on MS3A. Validation experiments were also conducted for water adsorption on fixed beds of MS3A. For absorption, work involved modeling with supportive experimentation. A dynamic model was developed to simulate CO 2 absorption with chemical reaction using high alkaline content water solutions. A computer code was developed to implement the model based upon transient mass and energy balances. Experiments were conducted in a laboratory-scale column to determine model parameters. The influence of geometric parameters and operating variables on CO 2 absorption was studied over a wide range of conditions. This project has resulted in 7 publications, with 3 manuscripts in preparation. Also, 15 presentations were given at national meetings of ANS and AIChE and at Material Recovery and Waste Forms Campaign Working Group meetings.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗