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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.

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At least 1,279 records · Page 71
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

The fuel integrity analysis of transfer processes for Chinshan dry-storage system by FRAPTRAN/CFD

The dry cask storage operation is separated into 3 stages: handling, transfer, and storage. In this study, the handling and transfer stages were established by CFD (Computational Fluid Dynamics). In order to obtain the burnup information of a single rod in the core, a single rod of Chinshan NPP was modeled by FRAPCON-3.5 and FRAPTRAN-1.5. Then, using the boundary condition from the CFD analysis, the fuel behaviors and performances were analyzed in detail with the FRAPTRAN-1.5 code. Three different thermal loads were considered in this paper: 7 kW, 14 kW, and 21 kW. The simulation results showed that, in all the cases: first, during the handling and transfer stages, the maximum cladding temperature was in the range of 477 to 653 K; secondly, the maximum cladding hoop strain was in the range of 0.0018 to 0.0025; and thirdly, the structural radial gap was almost kept as a constant. Based on these results, the fuel cladding integrity was maintained in the handling and transfer stages. These results help evaluating the safety issue of the dry-storage system for Chinshan NPP. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Simulation methodology for fuel assembly drop accident during handling

Fresh and irradiated fuel assembly handling is a routine part of Nuclear Power Plant activities. Fuel Assembly handling occurs during fresh fuel receipt, core refueling, spent fuel assemblies transfer from spent fuel pool to the storage containers, etc. Westinghouse, ENUSA and KEPCO-NF have collaborated in the technical area of Fuel Assembly Drop Accident (FADA) Simulation to update the approach for estimating the number of failed fuel rods as a consequence of a fresh or irradiated fuel assembly drop, which might occur during a handling accident in a PWR-type reactor. The proposed FADA simulation methodology starts with an assessment of fuel assembly impact velocity at the final stage of the drop in water or air. The vertical, horizontal or inclined drop directions are considered. This velocity is an input to determine a fuel assembly structural response during impact which defines the loading function for the fuel assembly components including spacer grids. The dynamic load on the fuel rod is calculated from the detailed spacer grid model and the corresponding loading function. Finally, the load from the spacer grid spring and dimples are applied to the fuel rod model to determine the cladding parameters for use in compliance to fuel rod cladding failure criteria. A combination of first principal, finite element analysis (FEA) and computational fluid dynamic (CFD) methods has been utilized to create a technically sound methodology. A number of specific and traditional tests for the fuel assembly and its components has been used to confirm that the proposed approaches which describe the physical phenomena related to the fuel assembly drop accident are reasonably conservative.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

DNB evaluation of PWR steam-line break event using Westinghouse and CASL coupled code systems

ANCKVIPRE and the VERA-CS code systems have been successfully applied to predicting PWR core response with respect to departure from nucleate boiling (DNB) at the limiting time step of a postulated hot zero power (HZP) main steam-line break (MSLB) event with and without offsite power available. ANCKVIPRE links the ANC code with the VIPRE-W code. ANC is Westinghouse's multi-dimensional nodal diffusion code for all nuclear core design calculations. It predicts core reactivity, assembly power, rod power, detector thimble flux, and other relevant core characteristics. VIPRE-W is the Westinghouse version of a thermalhydraulic subchannel code developed for light water reactor core design applications. The reactor core simulation was based on input of the core boundary conditions from RETRAN system transient and STAR-CCM+ CFD core inlet distribution calculations. To account for the asymmetric power distribution in the reactor core due to the broken steam pipe in one loop and the stuck RCCA, both ANCKVIPRE and the VERA-CS models were set up for the whole reactor core. The advanced modeling and simulations confirmed that the HZP MSLB high-flow case with offsite power available is more limiting with respect to the DNB acceptance criterion, consistent with the limiting case analyzed in the 4-loop plant Safety Analysis Report (SAR). The results also demonstrated that the currently NRC-approved Westinghouse coupled code system such as ANCKVIPRE remains conservative.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Terminal Sinking Velocity for Waste Packages Falling in a Deep Borehole - 17063

This evaluation of terminal velocity for a waste package sinking in a deep borehole filled with emplacement fluid was done to support concept development for the Deep Borehole Field Test (DBFT). The DBFT objective is to evaluate the safety and feasibility of the deep borehole disposal (DBD) concept for certain radioactive wastes. The DBD concept calls for emplacing waste packages in the lower 2 km of a borehole drilled into crystalline basement rock to a depth of 5 km. Previous studies identified that accidentally dropping one or more waste packages during emplacement could be an important risk, particularly if a waste package is breached. This study evaluated the range of terminal sinking velocity, which can be used to design waste packages and impact limiters to ensure that breach does not occur. Based on an analytical model of a sinking container, a highly turbulent flow is expected to develop in the emplacement fluid. Turbulence is a complex three-dimensional phenomenon and the applicability of analytical models is limited. To backstop the analytical model and to extend predictions to other package dimensions and fluid properties, numerical simulation was performed using the ANSYS Fluent computational fluid dynamics code. Terminal velocity was calculated for downhole temperatures up to 1200 C, and three emplacement fluids (water, NaCl brine, and NaBr brine). Simulation results confirmed that flow around the sinking container is highly turbulent (Reynolds number 5.6 x 10{sup 4} to 7.3 x 10{sup 5}). Turbulence increases with temperature and decreases with density. Calculated terminal velocity ranges from 1.95 (200 C) to 2.13 m/s (1200 C) in water; 1.61 (200 C) to 1.79 m/s (1200 C) in NaCl; and 1.30 (200 C) to 1.46 m/s (1200 C) in NaBr. Terminal velocity is inversely related to fluid density, and the temperature effect on density appears to have a stronger effect on terminal velocity than the temperature effect on viscosity. The foregoing results are for non-perforated casing, but in practice the casing may be perforated to prepare for cementing, and to allow relief of fluid pressure from thermal expansion after emplacement of heat generating waste, and after final borehole closure. To represent the effect from perforations, the analytical solution was modified to include bypass flow through perforations and in the annulus. A workable perforation scheme is identified, consistent with borehole construction and closure requirements, using 2-cm and 5-cm diameter perforations. This scheme is estimated to limit the terminal velocity to 3 m/s, for reference waste packages falling in water at 400 C. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

An Efficient Tool-Set for Modeling Containment Ventilation in Very Complex Facilities - 17215

The applicability of a multi-zone ventilation analysis program, CONTAM, to complex radioactive waste material handling or processing plants is evaluated. Validation results are based on simulations from three-dimensional multi-physics computational fluid dynamics. Insight into the application limits of CONTAM and conditions under which the use of higher fidelity tools are required are also discussed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

A Dissipation-Based Method for Improving the Accuracy of Computational Fluid Dynamics Simulations of High Level Non-Newtonian Waste - 17351

Analysis of shear rate in dissipative scales of turbulence was performed using a quasi-direct numerical simulation (Q-DNS) of flow through a pipe involving Bingham fluid. Dependency of deformation rate on dissipative scales of turbulence was studied and a modification algorithm was developed and implemented in Reynolds-averaged Navier-Stokes (RANS) modeling of the same flow using the standard k-e model. It was shown that the modification was effective in increasing accuracy of the RANS simulations when the velocity profile was modified to match experimental data. In both the Q-DNS and RANS simulations, the Herschel-Bulkley (HB) rheology equation was used to relate shear stress of the non-Newtonian fluid to shear deformation. Additional analysis of dependency of shear on the turbulent dissipation rate (TDR) was performed using various flow conditions in RANS simulations with a modified viscosity model (alpha model). In the past, the alpha model has been shown to produce accurate simulations for the same flow conditions. Results of the analysis shows that a qualitative agreement exists between shear dependency in the RANS-alpha and Q-DNS-HB simulations when the ratio of velocity to length scale is used. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Utilizing a Novel Acoustic Backscatter Array to Characterize Waste Consolidation and Settling in a Horizontal Flow Clarifier - 16051

As part of a collaboration between the University of Leeds, Sellafield Ltd and MMI Engineering Ltd, bespoke acoustic instrumentation and numerical simulation methods have been established to optimize nuclear legacy waste transfers. Specifically, a new ultrasonic measurement system has been developed, allowing suspended particle concentration and consolidation to be measured remotely. This technology has been combined with computational fluid dynamics (CFD) modelling of settling particulate systems, permitting full characterization, simulation and prediction of sludge separation processes. Presented here are experimental and numerical results on the transport and settling behavior of non-active nuclear-analogue suspensions in two flow geometries: gravitational settling of glass dispersions in a 250-liter column; and transport and sedimentation of flocculated calcite in a close-to-full-scale clarifier (9 m{sup 3} total volume and 4.8 m{sup 2} length). In addition, details of the design, construction and calibration of a custom-built Ultrasonic Array Research Platform (UARP) are described. Settling column results suggested that particle concentration changes could be simulated with good accuracy by a CFD model (developed using the open source OpenFOAM framework) when compared to measured samples. Large-scale clarifier trials were used to assess the effectiveness of the UARP controller with an attached in situ acoustic backscatter array (ABA). The ABA consisted of 8 transducers angled down at 20 deg. to the vertical, pulsing at 2-2.5 MHz in cycles (covering 1.8 m total depth). In comparison to a range of sampled concentrations across the length of the clarifier, results confirmed that the UARP-ABA system was highly successful at characterizing these complex flows over several hours of operation, allowing visualization of depth-wise bulk density changes and bed build-up over time. Collectively, this work highlights the capability of these tools to monitor and optimize a number of current waste transfer operations at Sellafield, with further potential applications for solid-liquid separations across waste, water treatment and minerals engineering industries. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

WOW Technology's Innovative Radioactive Liquid Waste Treatment - 16128

WOW presents its revolutionary technology and equipment for liquid radioactive waste treatment: outperforming ultimate water decontamination and purification process, enhanced sludge concentration, no secondary waste nor consumables, fully automated, remote controlled and self-decontaminating device. The WOW's technology is based upon a never before observed discovery of fluid dynamics science: the possibility of performing a molecular separation between solute and suspended elements and the solvent. The combination of such a molecular separation process with a standard vacuum evaporation improves the abatement performances by thousands of times, with respect to those of the state of the art vacuum evaporators. In addition to this, no secondary waste is produced during the process, as no filters, membranes, resins or additives are used. WOW equipment, automated and remote controlled, self decontaminates after use and can be designed and constructed either tailored to the application needs or with a modular approach for enhanced transportability and application flexibility. After the preliminary verification by CNR, the Italian National Research Center, WOW Technology's decontamination device was tested c/o LENA, the Laboratory of Applied Nuclear Energy of the University of Pavia, Italy with a simulated solution 6000 times more contaminated than the nuclear reactor's cooling water of Fukushima-Daiichi NPP. In addition to that, WOW's technology was also used in a real case at the Radiochemistry laboratory of the Pavia's University Chemistry department. Both the above mentioned contaminated fluids have been successfully decontaminated without production of additional or secondary waste. WOW's technology has already performed on industrial scale c/o the Nuclear Repository of S.S.M. in Saluggia, Italy: 45000 liters of acid radioactive solution have been successfully decontaminated to a Decontamination Factor (DF) of 335000 for Cs-137 by one single evaporation step and without using any additional tool to control steam. [Performances independently certified by: National Physics Laboratory (UK) and Pavia University Radiochemistry Laboratory, Chemistry Department (Italy)]. The high decontamination factors and the high volume reduction of the sludge make this technology suitable for the ultimate treatment of highly contaminated liquid waste such as the Fukushima-Daiichi nuclear reactor's cooling water. The WOW's technology addresses as well to interventions in both, operating and decommissioning nuclear power plants and any other nuclear and radioactive facility whose produced and/or stored radioactive liquid solutions, including acid liquid solutions, can be successfully treated, the waste volume reduced and no secondary waste produced. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Comparisons of Peak Cladding Temperature and Effective Thermal Conductivity: CFD Simulation on PWR Spent Fuel Assemblies - 16278

When spent fuel assemblies from the reactor of nuclear power plants (NPPs) are transported or stored, the assemblies are exposed to a variety of environments that can affect the peak cladding temperature of spent fuel assemblies. There are four methods to calculate the peak cladding temperature of spent fuel assemblies in a cask: Manteufel and Todreas's two-region model, Bahney Lotz's effective thermal conductivity model, Wooton-Epstein correlation, and computational fluid dynamics (CFD) simulation. The CFD simulation based on the FLUENT code is used to calculate the peak cladding temperature and effective thermal conductivity of a spent fuel assemblies used in Republic of Korea NPPs: 14x14, 16x16 ACE7, 16x16 PLUS7 and 17x17 PWR spent fuel assemblies. The CFD simulation results show that the effective thermal conductivity of the 16x16 ACE7 was more conservative than those for the other assemblies. Therefore the effective thermal conductivity using 16x16 ACE7 is used for thermal evaluation of the KORAD's developing cask which can load 14x14, 16x16 ACE7, 16x16 PLUS7 and 17x17 spent fuel assembly. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

High-fidelity simulation-driven model development for coarse-grained computational fluid dynamics

Nuclear reactor safety analysis requires analysis of a broad range of accident scenarios and determining their consequences. For a nuclear power plant behavior, it is impossible to obtain experimental data at sufficiently large scales to support calibration and validation of various system models. In single-phase flow convective problems, high-resolution methods of Computational Fluid Dynamics (CFD) such as Direct Numerical Simulation (DNS) and Large Eddy Simulation (LES) can provide high fidelity results under conditions where physical measurements are unavailable. However, such high-resolution simulations are computationally expensive and not suitable for simulation of long transient scenarios in nuclear reactor accidents. In this work, we investigate the use of a high fidelity simulation-driven approach to model sub-grid scale (SGS) effect in Coarse Grained Computational Fluid Dynamics CG-CFD. Notably, fine-mesh simulations are used to construct physics-informed statistical surrogate of coarse-mesh SGS model. For an initial analysis, we consider a case of turbulent natural convection in a volumetrically heated fluid layer with a thermally insulated lower boundary and isothermal upper boundary. This scenario of unstable stratification is relevant to turbulent natural convection in a severe nuclear reactor accident, as well as in containment mixing and passive cooling. For this type of flow, the SGS effect is modeled by an added turbulent diffusivity in the energy equation. It is shown that a global correction for the energy equation is sufficient to achieve a significant improvement to the CG-CFD prediction of thermal mixing in the fluid layer. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Time-resolved measurement of subchannel void distribution for boiling two-phase flow in 5 x 5 rod bundle at high-pressure and high-temperature

A subchannel void sensor (SCVS) was improved to acquire time-resolve boiling two-phase flow in a 5-by-5 heated rod bundle at high-pressure and high-temperature. The SCVS consists of 6-by-6 wire electrodes and 5-by-5 rod electrodes. The SCVS can acquire a time series data of local void fraction at 32 points of central subchannel region in addition to 100 points of rod surface regions between rods. The devised sensors are installed in the heated rod bundle at eight height levels to acquire two-phase flow dynamics and its development. The axial and radial power profile of the heated rod bundle are uniform, and eight pairs of sheath thermocouples are embedded on the heated rod to monitor its surface temperature distribution. The paper addresses the axial and cross-sectional distributions of void fraction at various pressure and rod bundle power conditions, which simulates an accidental condition in reactor core and in a spent fuel pool of a boiling water reactor (BWR). The experimental data are suitable to validate the computational multi-fluid dynamics (CMFD) and subchannel codes. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Computational Fluid Dynamics Modeling of Bubbling in a Viscous Fluid for Validation of Waste Glass Melter Modeling

At the Hanford site, radioactive waste stored in underground tanks is slated for vitrification for final disposal. A comprehensive knowledge of the glass batch melting process will be useful in optimizing the process, which could potentially reduce the cost and duration of this multi-billion dollar cleanup effort. We are developing a high-fidelity heat transfer model of a Joule-heated ceramic lined melter to improve the understanding of the complex, inter-related processes occurring with the melter. The glass conversion rates in the cold cap layer are dependent on promoting efficient heat transfer. In practice, heat transfer is augmented by inserting air bubblers into the molten glass. However, the computational simulations must be validated to provide confidence in the solutions. As part of a larger validation procedure, it is beneficial to split the physics of the melter into smaller systems to validate individually. The substitution of molten glass for a simulant liquid with similar density and viscosity at room temperature provides a way to study mixing through bubbling as an isolated effect without considering the heat transfer dynamics. The simulation results are compared to experimental data obtained by the Vitreous State Laboratory at the Catholic University of America using bubblers placed within a large acrylic tank that is similar in scale to a pilot glass waste melter. Comparisons are made for surface area of the rising air bubbles between experiments and CFD simulations for a variety of air flow rates and bubble injection depths. Also, computed bubble rise velocity is compared to a well-accepted expression for bubble terminal velocity (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

CFD Predictions of Severe Accident Natural Circulation Flows in a Combustion Engineering PWR

Computational fluid dynamics (CFD) is used to predict natural circulation flows between a reactor vessel upper plenum and a steam generator during specific severe accidents in a pressurized-water reactor (PWR). The predictions are based on a model for a simplified Combustion Engineering (CE) design. This work is part of a larger study to investigate the potential for severe accident induced steam generator (SG) tube failures in PWRs being conducted by the U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research. The CFD model builds upon the lessons learned from previous studies of a generic Westinghouse PWR under similar conditions. The CE SG is significantly larger than the Westinghouse design and geometric differences are expected to reduce the mixing and increase the potential thermal challenge to the SG tubes. The CFD predictions are utilized to improve the basis for simplified flow and mixing models applied in one-dimensional severe accident codes. Parameters are determined to support system code modeling of hot leg and SG inlet plenum mixing, hot leg and SG tube bundle flow rates, and tube inlet temperatures. This updated modeling uses boundary conditions obtained from system analysis code predictions of severe accident behavior in a CE PWR. Model refinements include an improved tube bundle representation and improvements to the thermal boundary conditions. Results indicate less mixing and higher tube inlet temperatures for the CE design compared to previous Westinghouse studies. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Particle-image Velocimetry Sensitivity through Automatic Differentiation

A commonly encountered difficulty in modern experimental fluid dynamics experiments is the need for full-flow visualization and measurement; traditional techniques such as hot-wire probes and laser Doppler velocimetry cannot provide such information without difficult scanning procedures. Particle-image velocimetry (PIV) is a measurement technique which can overcome this spatial limitation to provide real-time three-dimensional fluid velocity vectors for an entire plane within the flow, as opposed to a single point that is provided by other measurement techniques. To accomplish this, a PIV system seeds the fluid flow with tracer particles; these must have nearly the same density as that of the fluid and a high drag coefficient to ensure that they follow the fluid flow without deviation. A laser is sent through optics to expand its beam into a sheet which illuminates the plane of interest in the fluid. The particles reflect more light than the fluid, and their image is recorded using a high speed camera. Due to limitations in bandwidth, video cannot be taken fast enough to be useful, and instead pairs of images are captured with a known separation in time. The illuminated particles will move between the two images taken by the camera, and their relative motion can be used to compute the fluid velocity at that point. When a single camera is used, only two-dimensional velocity vectors can be computed, but with the addition of another camera this can be improved to full three-dimensional vectors. The flow's time evolution can be measured by taking series of image pairs and processing each pair to get the velocity field at the time between each image pair. Only the two-dimensional version will be considered in this work. To compute the velocity vectors of the flow, the images are broken up into interrogation regions, usually in a uniform grid. In the most basic PIV algorithm, corresponding regions in the image pairs are cross-correlated to compute the most likely displacement from one image to the next. This is used since the cross correlation will achieve a maximum value when the two interrogation regions line up the best. The location of the peak in the correlation map and scaling information for the optical setup is used to compute the particle displacement from one image to the next. This process is similar to holding two semi-transparent images and moving one until they match as well as possible. Sub-pixel displacements between images can be obtained by fitting a Gaussian function to the correlation peak. Knowing the displacement vector for each set of interrogation regions in the image pair, it is a simple matter to divide by the time gap between the two images to compute the fluid velocity at the point in time and space. Since there are many interrogation regions in each image pair, the velocity field for a sheet is sampled according to the arrangement of the interrogation regions. The sensitivity of the final computed vector to the input images is difficult to compute, and has led most considerations of uncertainty in the PIV field to apply Monte Carlo methods and correlations to quantify the systematic and random errors which result from application of various PIV algorithms. In this work, automatic differentiation (AD) has been used to trace the dependence of velocity vectors computed using the correlation-peak PIV algorithm back to the input values given to an artificial image generator. These initial results form the foundation for continued work in the application of AD to algorithms in PIV. They demonstrate that AD methods can be successfully applied to PIV algorithms without undue complexity. The derivatives of each individual vector with regard to the coefficients of the velocity field used to create the artificial images can be computed and the statistics of them considered collectively, or correlated between each other to quantify the effectiveness of a PIV algorithm. With the initial implementation complete, adding extra algorithms is relatively simple and considerable data can be generated for analysis. It is hoped that application of these methods can lead to improved PIV algorithms which demonstrate better sensitivity to small variations in particle displacement. While these methods can be used as an analysis tool for refinement of new algorithms, computation of actual resultant vector sensitivities to input images' individual pixels is the next logical step of this work. Such information will be analyzed with experimental data to produce a better understanding of how output vectors' uncertainty relates to the actual image pairs, and lead to automatic bad vector removal. (authors)

42 ENGINEERING↗
DOE OSTI2019

Development of Transient Coupled Chemical-Thermal-Fluids Simulation for Unsealed Aluminum-Clad Spent Nuclear Fuel Storage Canister - 19273

This study examines a preliminary three-dimensional (3D) multi-physics computational fluid dynamics (CFD) model of the unsealed canisters containing aluminum-clad spent nuclear fuel from INL's ATR in the irradiated fuel storage facility. This 3D CFD model considers fully resolved fuel plates in an unsealed canister with small mass exchange with surrounding air as time evolves in the simulation and contains a network of chemical reactions to consider due to several of the side reactions occurring in the presence of water vapor, nitrogen, oxygen and carbon dioxide rather than in helium backfilled DOE standard canisters. Prior to the CFD modeling, the full set of chemical reactions under these conditions was analyzed, reduced, and refit to a more manageable number of equations for the computational efficiency of the 3D CFD simulation. To deal with the long time-scale of interest for tracking the possible buildup of trace species such as nitric acid, NOx, and hydrogen, a segregated time stepping algorithm is used, wherein the velocity field is only updated periodically, and assumed to be in a pseudo steady state. This study first examines the theory of the model, workflow to establish radiolysis reaction network, and initial simulations of the evolutions of thermal fields, hydrogen gas and nitric acid concentrations within unsealed canisters in conditions relevant to the INL INTEC CPP-603 facility over a long period of time. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Melter Modeling over a Range of Scales to Support Vitrification at the WTP - 19664

Computational modeling is a powerful complement to laboratory experiments and testing of systems at various scales. When properly applied and validated, computational fluid dynamics (CFD) can provide information not readily obtainable by physical testing, especially in cases where the operating environment does not permit visual observation or the insertion of typical data acquisition instrumentation as in the case of waste glass vitrification. A series of CFD and heat transfer models over a range of scales are being developed to support melter operations at the Hanford Waste Treatment and Immobilization Plant. This paper provides a description of the models and the resulting information they provide that can be used to improve melter throughput, as well as aid in evaluating and resolving operating plant issues or upsets. This work is important for the WTP operation by providing a potential for model-predictive control of the melter operation. Increases in the rate of glass production and avoiding idling can significantly decrease the duration and cost of the vitrification campaign, potentially saving billions of dollars. (authors)

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

Computationally Aided Design of a Small-Scale Waste-Glass Melter - 20452

To provide mission support to the WTP for the vitrification of legacy nuclear tank waste at the Hanford site, a reduced-scale vitrification pilot system is being designed to process simulated and actual radioactive tank wastes. The waste stored in underground tanks will be separated into HLW and LAW fractions, where the majority is LAW by mass (approximately 90%) and HLW by activity (greater than 95%). The first tank waste to be processed at the WTP will be LAW. Because Hanford tank waste and the resultant melter feed compositions are known to vary widely, pilot-scale operations are essential to identify potential problems, provide needed data, confirm assumptions, and determine impacts to full-scale melters and off-gas systems from these different feeds. In addition, the reduced-scale melter system must be capable of quickly providing results to operations without the typically high cost of full-scale radioactive operations and to provide an engineering platform in close proximity to local operations staff. The objectives of this computational modeling effort are to produce process conditions like those encountered in the WTP LAW melters and to minimize the volume of radioactive waste necessary for the evaluations. To this end, melter design features are being explored to increase production without significantly increasing surface area or melter-glass volumes or compromising data quality. In support of these objectives, a set of computational fluid dynamic models was developed to provide insight into the flow patterns within these small melters and evaluate design features to increase production without significantly increasing glass inventory. Computed velocities were evaluated at the interface between the cold cap and the molten-glass pool and within the bulk-glass pool. Since melt rate correlates with the velocity underneath the cold cap, this information can be used to select design features that produce the highest gains. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Numerical simulation of break flows occurring in the reactor cavity cooling system of a very high temperature reactor using a system CFD approach - 18437

To avoid accidents due to human error or the failure of components, the reactor cavity cooling system (RCCS) of a high temperature reactor must be able to operate naturally or passively safe. The system under consideration in this investigation is an air-cooled system which consists of an annular downcomer which is connected via a bottom-header to riser tubes. The downcomer and riser tubes are connected to the atmosphere via inlet and outlet manifold systems respectively. In this study the entire system is simulated using the 1D network codes GAMMA+ and Flownex. The study investigates the effect that breaks, causing a short circuit between one half of the inlet manifold and the corresponding half of the outlet manifold, will have on the performance of the cooling system. The agreement between the results obtained with the two codes is good. When the break size exceeds a certain value the flow may reverse in the relevant half of the outlet manifold. The critical break size is dependent on the elevation at which the break occurs. The reversal in the flow results in a drop of 10% in the heat removal capacity of the RCCS. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗