Shroud Tube Research Articles

Numerical Analysis of RTD Curves and Inclusions Removal in a Multi-Strand Asymmetric Tundish with Different Configuration of Impact Pad

To effectively remove non-metallic inclusions from the steel during the flowing in a five-strand asymmetric tundish, the novel configuration of the impact pad was developed. For analysis, complex numerical modelling in the programme ANSYS Fluent was used. The Lagrangian Discrete Phase Model of inclusion tracking was applied. The distribution of inclusions, with sizes ranging from 2 µm to 100 µm and density from 2500 to 3500 kg·m−3, was considered only through the shroud tube. The residence time distribution (RTD) curves and inclusion removal efficiency were used for evaluation of steady state steel flow character depending on internal configuration of a tundish with an impact pad in two design modifications (Modification 1—M1, Modification 2—M2). The preliminary results showed that in the case of asymmetric geometry plays a role the computational mesh independency. The assembly method with cut cell approach was satisfactory even when the tundish geometry was changed. The RTD curves with an M1 showed a huge dead volume in the tundish. In the case with an M2, the RTD curves are more or less uniform for all casting strands, and the removal of inclusions to slag increased from about 55% up to 70% in comparison with M1.

Jet breaker plate for suppressing gas entrainment and flow induced vibration of a CSRDM shroud tube assembly inside the control plug shell

Abstract The present study proposes a jet-breaker plate (JBP) assembly near the outlet of the porous shroud tube for suppressing the gas entrainment inside the liquid pool and flow induced vibration of the control plug structure. Several JBP designs have been tested. Accelerometers mounted at different locations of the model are used for measurement of flow induced vibration. Flow visualization techniques, i.e. white light visualization, dye visualization, Shadowgraph imaging have been used to characterize the flow behavior near the JBP assembly. Image processing techniques have been employed to quantify turbulence intensity and free-surface fluctuation characteristics near the JBP. The implementation of JBP leads to reduced free-surface fluctuations, lower turbulence intensity near the free surface (by about 50%), significant reduction in hump size (about 95%), elimination of gas entrainment, and mitigation of flow-induced vibration (about 90%).

Flow structures associated with CSRDM shroud tube and control rod assembly: A combined experimental and simulation study

This paper presents numerical investigation of flow behavior near a submerged porous shroud tube, a protective covering over Control Safety Rod Driving Mechanism (CSRDM) inside liquid sodium coolant pool of a Fast Breeder Reactor (FBR) in synergy with corresponding experimental findings to explicate possible gas entrainment mechanism inside the liquid pool. Fluid entering the shroud tube whose top end is just below the free surface causes swelling of free surface near the axial outlet of the shroud tube that vastly affects the flow behavior nearby and liquid-fall type gas entrainment inside the coolant pool. To numerically capture the effects of various flow rate and free surface height on the flow behavior near the shroud tube, an approximate single phase realizable k−ε turbulence flow model has been implemented. The study is performed assuming Froude number similarity for Reynolds number ranging from 1.3 × 103 to 6.5 × 103 and corresponding Froude number ranging from 0.266 to 1.332. Both computational and experimental studies have been carried out with water as the working fluid. Geometry of hump profile, obtained experimentally, has been incorporated as input to the computation to avoid large computational time associated with free surface modeling. With such a coupled approach, a strong agreement is obtained between numerical prediction and experimental Schlieren and dye visualization results. The simulation study is subsequently used to report the radial and axial velocity and turbulence intensity distribution in the liquid pool. The nature of impinging jet-like velocity profile from the axial exit of the shroud tube and corresponding boundary layer-like growth near the air-water interface is compared with that of the free wall jet, impinging wall jet and impinging confined wall jet. The free surface boundary layer growth follows power law similar to that of the wall jet in low Froude number regime (Fr < 0.666). At higher Froude number, the axial flow along the shroud tube dominates over the radial free surface flow. The power law exponent of the free surface boundary layer growth is lower compared to that of the wall jet due to the lower viscous effect from the adjacent air medium and axial flow due to the hump entrainment.

Porous shroud tube design evaluation of a control plug in a liquid metal cooled reactor

Shroud tube is a porous protective covering over the Control Safety Rod Driving Mechanism assembly vertically submerged just below the free-surface of liquid sodium coolant pool of a Fast Breeder Reactor. Pore size and its arrangement over the shroud tube are important design parameters, which influence the gas entrainment and flow induced vibration inside the liquid coolant pool leading to several reactor operation related risks. The present study focuses on performance evaluation of three shroud tube designs of same porosity but with different pore diameter, thickness of shroud tube and arrangement of pores (staggered and uniform). A parametric investigation with respect to parameters viz. Froude number, free-surface height, and the eccentricity of the control rod has been carried out using white light imaging and high speed laser based Shadowgraphy. The present study effectively demonstrates the complexity in design of porous shroud tube and the role played by several design parameters. The radial flow from the pores and its interaction with the downward flow due to hump motion influence the air-water interface dynamics which influences the gas entrainment process. The flow induced fluctuation of the hump increases with increase in eccentricity of the control rod.

Irradiation performance of sodium-bonded control rod for the fast breeder reactor

ABSTRACT The lifetime of control rods is limited by the absorber (B4C pellets)–cladding mechanical interaction (ACMI). Therefore, sodium (Na)-bonded control rods were developed for long-life control rods. Na-bonded control rods have been irradiated in the experimental fast breeder reactor, JOYO MK-III, and the diametrical changes of the Na-bonded absorber pins after the irradiation were measured in detail. In this paper, these detailed measurements were compared with the results obtained in helium (He)-bonded control rods with and without the shroud tube in a wide burn-up range. From the comparison, it was found that the diametric changes were smaller in the Na-bonded absorber pins than in the He-bonded ones. It was concluded that the Na-bonded absorber pins are very effective for achieving long-life control rods.

Hydrodynamics around a shroud tube assembly of a fast breeder reactor

Control plug assembly of a liquid sodium cooled fast breeder reactor is partially submerged inside the liquid sodium coolant pool, free surface of which is blanketed with cover gas, mainly Argon to provide a non-reactive environment. It receives high velocity coolant flow from the fuel subassembly which travels through annular gap inside Control Safety Rod Driving Mechanism (CSRDM) assembly leading to swelling of free surface and gas entrainment inside the coolant pool. Entrained gas bubbles travel through different submerged components of the reactor and induces numerous complications including reactor power fluctuation, cavitation in inlet pumps and heat transfer inhibition in heat exchangers. Study of hydrodynamics around the shroud tube is crucial to examine cover gas entrainment inside the pool. The present study reports an experimental study on a 1:9 scale down model of CSRDM shroud tube assembly based on Froude number similarity. Eccentricity of the control rod inside the shroud tube is also an important parameter as temperature gradient along the CSRDM assembly leads to longitudinal thermal stress causing displacement of control rod from its center position in the shroud tube. Flow behavior around shroud tube, hump formation at the axial exit of shroud tube, air entrainment inside coolant pool have been reported for various parameters viz. inlet flow rate (qi), free surface height (hf) and eccentricity of central rod (e). Dye visualization, Schlieren visualization, Bulk visualization have been carried out and digital image processing techniques have been adopted for quantification of hump parameters and entrained air fraction inside the coolant pool.The axial flow from the exit of the shroud tube impinges on the free surface like a wall jet. At low Froude number, the impinging jet flow travels radially outward on the interface like a wall jet boundary layer. The shear of the free surface flow drags the jet like flow from the pores upward leading to mixing between the vortices inside the shear layer and the free vortices from the pores of the shroud tube. At high Froude number, the impinging jet like flow penetrates through the interface leading to hump formation. The downward motion of the hump reduces the radial free surface flow and blankets the radial flow from the pores of the shroud tube. The eccentricity of the central rod leads to asymmetric axial flow from the shroud tube and thus the resulting hump becomes asymmetric due to impingement of asymmetric jet. Therefore, the impingement angle of the downward hump motion becomes different at both sides of the shroud tube with asymmetric flow field and bubble formation process inside the liquid pool. However, the synchronization between the hump unsteadiness and the surface wave motion leads to mixing of air bubbles inside the liquid pool. Average entrained air fraction increases with increase in inlet flow rate and decreases with increase in free surface height.

Flow induced vibration studies on PFBR control plug components

The construction of Prototype Fast Breeder Reactor (PFBR), a 500MWe liquid sodium cooled reactor, is in progress at Kalpakkam in India. Control plug (CP) is located right above the core subassemblies in the hot pool. Control plug is an important component as many of the critical reactor parameters are sensed and controlled by the components housed in the control plug assembly. In PFBR primary circuit, components are basically thin walled, slender shells with diameter to thickness ratio ranging from 100 to 650. These components are prone to flow induced vibrations. The existence of free liquid (sodium) surfaces, which is the source of sloshing phenomenon and the operation of primary sodium pump in the primary pool are other potential sources of vibration of reactor components.Control plug is a hollow cylindrical shell structure and provides passages and support for 12 absorber rod drive mechanisms (ARDM) which consists of 9 control and safety rods and 3 diverse safety rods, 210 thermo wells to measure the sodium temperature at the exit of various fuel subassemblies, three failed fuel localization modules (FFLM) and acoustic detectors. It consists of a core cover plate (CCP), which forms the bottom end, two intermediate supports plate, i.e. lower stay plate (LSP) and upper stay plate (USP) and an outer shell. The CCP is located at a distance of 1.3m from the core top. With such a gap, there will be long free hanging length of the thermocouple sleeves, Delayed neutron detector (DND) sampling tubes and ARDM shroud tubes and hence they are susceptible to flow induced vibrations. Since control plug is partially immersed in hot sodium, the reactor transients are felt by the components, hence it is very much essential to understand the vibration response of the control plug components. The main vibration excitation mechanisms are vortex shedding, turbulence buffeting, fluid-elastic instability, etc.In order to assess the susceptibility of CP internals against flow induced vibrations (FIVs), to measure structural response and to validate the analytical codes developed for FIV analysis of CP along with the assumptions of supports for various CP internals therein, a flow induced vibration testing program was formulated in 1:4 scale model of PFBR CP. As the first phase of this program, experimental modal analysis of CP internals was carried out in air to estimate the modal parameters. Subsequently, flow induced vibration studies were conducted in the 1:4 scale model of CP in water. In this model, size of the tubes, shell and plates are reduced to ¼ size of the PFBR CP. The stiffness of CP parts is reduced by 4 times and mass decreased by 64 times which results in scaling up of modal frequencies by 4 times. The CP internals in the model were instrumented with accelerometers and strain gages. The studies were conducted in water with flows derived on the basis of velocity similitude. The output signals from the sensors were acquired and analyzed to obtain frequency spectra, overall vibration amplitude and strain values at various locations inside CP. The study carried out confirms the absence of resonance due to flow induced vibration mechanisms for the entire operating range. This paper elucidates the modeling details, similitude criteria, instrumentation employed and experimental results obtained with discussion on results.

Hydraulic design of sparger arrangement for long guide tubes of a fast breeder reactor

In liquid metal cooled fast reactors, the core is submerged in sodium pool by ∼5 m below sodium free surface. This necessitates the control and shutdown of reactor be achieved by long overhanging mechanisms housed inside a control plug. These mechanisms are protected by porous guide tubes with a sparger type arrangement for the sodium flow through them. Comprehensive knowledge of flow distribution of sodium through these guide tubes is essential to assess the risks of flow induced vibration of thin thermowell tubes that pass close to these shroud tubes and entrainment of cover gas due to high free surface velocities. Three dimensional hydraulic analysis of single isolated shroud tube and integrated assembly of shroud tubes have been carried out using CFD tools to acquire this knowledge. The predictions of the CFD models have been validated against experimental predictions. These studies have provided important information regarding critical design parameters. Size of holes in the shroud tube, location of holes in the control plug shell and arrangement for breaking sodium jets emanating from shroud tubes have been optimized to reduce free surface velocity.

Reduction in Degree of Absorber-Cladding Mechanical Interaction by Shroud Tube in Control Rods for the Fast Reactor

Research and development of a long-life control rod for fast reactors is being conducted at Joyo. One of the challenges in developing a long-life control rod is the restraint of absorber-cladding mechanical interaction (ACMI). First, a helium-bonding rod was selected as a control rod for the experimental fast reactor Joyo, which is the first liquid metal fast reactor in Japan. Its lifetime was limited by ACMI, which is induced by the swelling and relocation of B4C pellets. To restrain ACMI, a shroud tube was inserted into the gap between the B4C pellets and the cladding tube. However, once B4C pellets cracked and broke into small fragments, relocation occurred. After this, the narrow gap closed immediately as the degree of B4C pellet swelling increased. To solve this problem, the gap was widened during design, and sodium was selected as the bonding material instead of helium to restrain the increase in pellet temperature. Irradiation testing of the modified sodium-bonding control rod confirmed that ACMI would be restrained by the shroud tube regardless of the occurrence of B4C pellet relocation. As a result of these improvements, the estimated lifetime of the control rod at Joyo was doubled. In this paper, the results of postirradiation examination are reported.

Reduction in Degree of Absorber-Cladding Mechanical Interaction by Shroud Tube in Control Rods for the Fast Reactor

Research and development of a long-life control rod for fast reactors is being conducted at Joyo. One of the challenges in developing a long-life control rod is the restraint of absorber-cladding mechanical interaction (ACMI). First, a helium-bonding rod was selected as a control rod for the experimental fast reactor Joyo, which is the first liquid metal fast reactor in Japan. Its lifetime was limited by ACMI, which is induced by the swelling and relocation of B4C pellets. To restrain ACMI, a shroud tube was inserted into the gap between the B4C pellets and the cladding tube. However, once B4C pellets cracked and broke into small fragments, relocation occurred. After this, the narrow gap closed immediately as the degree of B4C pellet swelling increased. To solve this problem, the gap was widened during design, and sodium was selected as the bonding material instead of helium to restrain the increase in pellet temperature. Irradiation testing of the modified sodium-bonding control rod confirmed that ACMI would be restrained by the shroud tube regardless of the occurrence of B4C pellet relocation. As a result of these improvements, the estimated lifetime of the control rod at Joyo was doubled. In this paper, the results of postirradiation examination are reported.

Effects of a Shroud Tube on Flow Field and Particle Behavior Inside a Bag-Filter Vessel

A shroud tube was used to decrease the amount of particles toward the bag filters from whole particles entering a filter vessel. The effects of the shroud tube on the flow field and particle behavior inside the vessel were studied. The air mixed with dust particles enters the vessel through a tangential inlet duct. Some of the particles are deposited on the inside wall of the vessel and the surface of the shroud tube. The other ones are collected on the filter surface or passed through it. The particles deposited on the wall surfaces fall into a hopper by gravity, and those collected on filters are removed by back pulse-jet flow. Computational simulation was performed to know the prereduction rate of particles by deposition on the wall surfaces for the different shroud tubes. The experiment was accomplished with some shroud tubes suggested by the results of computational simulation, and the experimental results were compared qualitatively with the computational results. The shroud tube blocked the direct transport of particles toward the bag filters and reduced the particle loading onto the filters. The particle loading was reduced when the upper region of the vessel was not blocked by the shroud tube more than when the vessel was blocked wholly with the filters from the upper end wall. However, the re-entrainment of the particles removed from the filters by the back pulse cleaning increased when the upper region of the vessel was not blocked by the shroud tube more than when the vessel was blocked wholly with the filters from the upper end wall.

THE EFFECT OF VARIOUS CONSTRAINT CONDITIONS IN THE FREQUENCY RESPONSE MODEL OF THERMAL STRIPING

Abstract— Thermal fatigue damage may be caused when incompletely mixed hot and cold fluid streams pass over the surface of a component or structure. This kind of thermal fatigue, known as thermal striping, is of particular concern in various types of nuclear reactors and for rapid shut‐downs of hot plant. A computer code, known as TBL, is an engineering design tool which has been developed to assess thermal striping damage for fully constrained plates. This model is further developed to account for various plate‐constraint conditions. It is compared with two other methods of assessing thermal striping damage. Good agreement is found for the cases when the component is (a) a plate, constrained in various ways and exposed to a sinusoidally varying temperature time striping history and (b) a shroud tube found in the above‐core region of a fast reactor pressure vessel and striped with a random temperature time history.

Analysis method for calculating vibration characteristics of beams with intermediate supports

Several reactor system components, such as fuel pins, heat exchanger tubes, control rods, and various instrumentation and shroud tubes, are beam-like components with intermediate supports along their length. Basic to any flow induced vibration or seismic analysis of these components is the need to determine their vibrational characteristics including natural frequencies and modes. This paper presents a simple solution method for calculating the natural frequencies and modes of beams with any of the classical boundary conditions and with unlimited intermediate supports. An exact solution is obtained using the method of stationary potential energy in which the modal displacement is expanded in a Fourier series. Lagrange multipliers are utilized to enforce intermediate constraint conditions and geometric boundary conditions that are not identically satisfied by the assumed series. Stokes' transformation is exploited to obtain the correct series expressions for the derivatives of the Fourier series.

General features of CHF of forced convection boiling in vertical concentric annuli with a uniformly heated rod and zero inlet subcooling

A graphical method is employed to give a generalized bird's-eye view of the existing data for critical heat flux (CHF) in internally and uniformly heated vertical annuli with zero inlet subcooling. 301 data points collected from 25 sources are used for this purpose, including 7 different fluids (water, R-12, R-114, acetone, toluene, monoisopropyl-biphenyl, and sodium), axial length of heated rod from 0.0762 to 8.84 m, outer diameter of heated rod from 0.00500 to 0.0964 m, inner diameter of unheated shroud tube from 0.0127 to 0.101 m, and vapor/liquid density ratio from 0.0000580 to 0.160.

Dispersion-Strengthened Ferritic Steels as Fast-Reactor Structural Materials

Dispersion-strengthened ferritic steels are being studied for possible use as canning material for sodium-cooled fast reactors. The basic alloy chosen contains nominally Fe—13% Cr—1.5% Mo— 3.5% Ti to which 2% TiO2 or 1% Y2O3 is added by a powder metallurgy technique.At 700°C, the alloys studied can favorably be compared in stress rupture tests (up to 12 000 h) to the best austenitic steels. Corrosion tests in dynamic sodium at 700°C showed that after 4 000 h the affected zones remained narrow and had no significant influence on the mechanical resistance at high temperature.Neutron irradiation of these alloys demonstrated their remarkable resistance to embrittlement in mechanical tests at 700°C. Comparison with other alloys showed that they had the highest elongation to rupture after irradiation. Simulation tests by 1-MeV electrons gave almost zero swelling in the temperature range of 475 to 700°C.The combined properties of dispersion-strengthened ferritic alloys make them excellent candidates not only for canning material but also for shroud tubes for fast-reactor fuel elements.