Dynamic Modeling Strategy for Flow Regime Transition in Gas-Liquid Two-Phase Flows

Directory of Open Access Journals (Sweden)

Xia Wang

2012-12-01

Full Text Available In modeling gas-liquid two-phase flows, the concept of flow regimes has been widely used to characterize the global interfacial structure of the flows. Nearly all constitutive relations that provide closures to the interfacial transfers in two-phase flow models, such as the two-fluid model, are flow regime dependent. Current nuclear reactor safety analysis codes, such as RELAP5, classify flow regimes using flow regime maps or transition criteria that were developed for steady-state, fully-developed flows. As two-phase flows are dynamic in nature, it is important to model the flow regime transitions dynamically to more accurately predict the two-phase flows. The present work aims to develop a dynamic modeling strategy to determine flow regimes in gas-liquid two-phase flows through introduction of interfacial area transport equations (IATEs within the framework of a two-fluid model. The IATE is a transport equation that models the interfacial area concentration by considering the creation of the interfacial area, fluid particle (bubble or liquid droplet disintegration, boiling and evaporation, and the destruction of the interfacial area, fluid particle coalescence and condensation. For flow regimes beyond bubbly flows, a two-group IATE has been proposed, in which bubbles are divided into two groups based on their size and shapes, namely group-1 and group-2 bubbles. A preliminary approach to dynamically identify the flow regimes is discussed, in which discriminators are based on the predicted information, such as the void fraction and interfacial area concentration. The flow regime predicted with this method shows good agreement with the experimental observations.

Oscillating liquid flow ICF Reactor

International Nuclear Information System (INIS)

Petzoldt, R.W.

1990-01-01

Oscillating liquid flow in a falling molten salt inertial confinement fusion reactor is predicted to rapidly clear driver beam paths of residual liquid droplets. Oscillating flow will also provide adequate neutron and x-ray protection for the reactor structure with a short (2-m) fall distance permitting an 8 Hz repetition rate. A reactor chamber configuration is presented with specific features to clear the entire heavy-ion beam path of splashed molten salt. The structural components, including the structure between beam ports, are shielded. 3 refs., 12 figs

Investigation of the transition from forced to natural convection in the research reactor Munich II

International Nuclear Information System (INIS)

Skreba, S.; Adamek, J.; Unger, H.

1999-01-01

The new research reactor Munich II (FRM-II), which is under construction at the Technical University Munich, Germany, makes use of a newly developed compact reactor core consisting of a single fuel element, which is assembled of two concentric pipes. Between the fuel element's inner and outer pipe 113 involutely bent fuel plates are placed rotationally symmetric, forming 113 cooling channels of a constant width of 2.2 mm. After a shut down of the reactor, battery supported cooling pumps are started by the reactor safety system in order to remove the decay heat by a downwards directed forced flow. Three hours after they have been started, the cooling pumps are shut down and so-called 'natural convection flaps' are opened by their own weight. Through a flow path, which is provided by the opening of the natural convection flaps, the decay heat is given off to the water in the reactor pool after the direction of the flow has changed and an upwards directed natural convection flow has developed. At the Department for Nuclear and New Energy Systems of the Ruhr-University Bochum, Germany, a test facility has been built in order to confirm the concept of the decay heat removal in the FRM-II, to acquire data of single and two phase natural convection flows and to detect the dry out in a narrow channel. The thermohydraulics of the FRM-II are simulated by an electrically heated test section, which represents one cooling channel of the fuel element. At first experiments have been performed, which simulated the transition from forced to natural convection in the core of the FRM-II, both at normal operation and at a complete loss of the decay heat removal pumps. In case of normal operation, the transition from forced to natural convection takes place single phased. If a complete loss of the active decay heat removal system occurs, the decay heat removal is ensured by a quasi-steady two phase flow. In a second test series minimum heat flux densities leading to pressure pulsations

Experimental study of the transition from forced to natural circulation in EBR-II at low power and flow

International Nuclear Information System (INIS)

Gillette, J.L.; Singer, R.M.; Tokar, J.V.; Sullivan, J.E.

1979-01-01

A series of tests have been conducted in EBR-II which studied the dynamics of the transition from forced to natural circulation flow in a liquid-metal-cooled fast breeder reactor (LMFBR). Each test was initiated by abruptly tripping an electromagnetic pump which supplies 5 to 6% of the normal full operational primary flow rate. The ensuing flow coast-down reached a minimum value after which the flow increased as natural circulation was established. The effects of secondary system flow through the intermediate heat exchanger and reactor decay power level on the minimum in-core flow rates and maximum in-core temperatures were examined

Numerical simulation of flow field in the China advanced research reactor flow-guide tank

International Nuclear Information System (INIS)

Xu Changjiang

2002-01-01

The flow-guide tank in China advanced research reactor (CARR) acts as a reactor inlet coolant distributor and play an important role in reducing the flow-induced vibration of the internal components of the reactor core. Numerical simulations of the flow field in the flow-guide tank under different conceptual designing configurations are carried out using the PHOENICS3.2. It is seen that the inlet coolant is well distributed circumferentially into the flow-guide tank with the inlet buffer plate and the flow distributor barrel. The maximum cross-flow velocity within the flow-guide tank is reduced significantly, and the reduction of flow-induced vibration of reactor internals is expected

Effect of friction on pebble flow pattern in pebble bed reactor

International Nuclear Information System (INIS)

Li, Yu; Gui, Nan; Yang, Xingtuan; Tu, Jiyuan; Jiang, Shengyao

2016-01-01

Highlights: • A 3D DEM study on particle–wall/particle friction in pebble bed reactor is carried out. • Characteristic values are defined to evaluate features of pebble flow pattern quantitatively. • Particle–wall friction is dominant to determine flow pattern in a specific pebble bed. • Friction effect of hopper part on flow field is more critical than that of cylinder part. • Three cases of 1:1 full scale practical pebble beds are simulated for demonstration. - Abstract: Friction affects pebble flow pattern in pebble-bed high temperature gas-cooled reactor (HTGR) significantly. Through a series of three dimensional DEM (discrete element method) simulations it is shown that reducing friction can be beneficial and create a uniform and consistent flow field required by nuclear engineering. Particle–wall friction poses a decisive impact on flow pattern, and particle–particle friction usually plays a secondary role; relation between particle–wall friction and flow pattern transition is also concluded. Moreover, new criteria are created to describe flow patterns quantitatively according to crucial issues in HTGR like stagnant zone, radial uniformity and flow sequence. Last but not least, it is proved that friction control of hopper part is more important than that of cylinder part in practical pebble beds, so reducing friction between pebbles and hopper surface is the engineering priority.

Neutronics of a mixed-flow gas-core reactor

International Nuclear Information System (INIS)

Soran, P.D.; Hansen, G.E.

1977-11-01

The study was made to investigate the neutronic feasibility of a mixed-flow gas-core reactor. Three reactor concepts were studied: four- and seven-cell radial reactors and a seven-cell scallop reactor. The reactors were fueled with UF 6 (either U-233 or U-235) and various parameters were varied. A four-cell reactor is not practical nor is the U-235 fueled seven-cell radial reactor; however, the 7-cell U-233 radial and scallop reactors can satisfy all design criteria. The mixed flow gas core reactor is a very attractive reactor concept and warrants further investigation

A Novel Dual-Stage Hydrothermal Flow Reactor

DEFF Research Database (Denmark)

Hellstern, Henrik Christian; Becker, Jacob; Hald, Peter

2015-01-01

The dual-stage reactor is a novel continuous flow reactor with two reactors connected in series. It is designed for hydrothermal flow synthesis of nanocomposites, in which a single particle consists of multiple materials. The secondary material may protect the core nanoparticle from oxidation....... The dual-stage reactor combines the ability to produce advanced materials with an upscaled capacity in excess of 10 g/hour (dry mass). TiO2 was synthesized in the primary reactor and reproduced previous results. The dual-stage capability was succesfully demonstrated with a series of nanocomposites incl. Ti...

Nonequilibrium chemical instabilities in continuous flow stirred tank reactors: The effect of stirring

International Nuclear Information System (INIS)

Horsthemke, W.; Hannon, L.

1984-01-01

We present a stochastic model for stirred chemical reactors. In the limiting case of practical interest, i.e., fast stirring, we solve for the characteristic function in steady state and derive expressions for the stationary moments through a perturbation expansion. Moments are explicitly calculated for a generic model of bistable behavior. We find that stirring decreases the area of the bistable region essentially by changing the point of transition from the high reaction rate state to the low reaction rate state. This is in remarkable agreement with the experimental findings of Roux, et al. Our results indicate that stirring should not be considered simply as an ''enhanced diffusion'' process and that nucleation plays only a minor role in transitions between multiple steady states in a continuous flow stirred tank reactor (CSTR)

78 FR 50313 - Physical Protection of Irradiated Reactor Fuel in Transit

Science.gov (United States)

2013-08-19

... Irradiated Reactor Fuel in Transit AGENCY: Nuclear Regulatory Commission. ACTION: Orders; rescission. SUMMARY... the NRC published a final rule, ``Physical Protection of Irradiated Fuel in Transit,'' on May 20, 2013... of Irradiated Reactor Fuel in Transit'' (RIN 3150-AI64; NRC-2009-0163). The final rule incorporates...

Heat and momentum transfer in a gas coolant flow through a circular pipe in a high temperature gas cooled reactor

International Nuclear Information System (INIS)

Ogawa, Masuro

1989-07-01

In Japan Atomic Energy Research Institute (JAERI), a very high temperature gas cooled reactor (VHTR) has been researched and developed with a purpose of attaining a coolant temperature of around 1000degC at the reactor outlet. In order to design VHTR, comprehensive knowledge is required on thermo-hydraulic characteristics of laminar-turbulent transition, of coolant flow with large thermal property variation due to temperature difference, and of heat transfer deterioration. In the present investigation, experimental and analytical studies are made on a gas flow in a circular tube to elucidate the thermo-hydraulic characteristics. Friction factors and heat transfer coefficients in transitional flows are obtained. Influence of thermal property variation on the friction factor is qualitatively determined. Heat transfer deterioration in the turbulent flow subjected to intense heating is experimentally found to be caused by flow laminarization. The analysis based on a k-kL two-equation model of turbulence predicts well the experimental results on friction factors and heat transfer coefficients in flows with thermal property variation and in laminarizing flows. (author)

Assessing the degree of plug flow in oxidation flow reactors (OFRs: a study on a potential aerosol mass (PAM reactor

Directory of Open Access Journals (Sweden)

D. Mitroo

2018-03-01

Full Text Available Oxidation flow reactors (OFRs have been developed to achieve high degrees of oxidant exposures over relatively short space times (defined as the ratio of reactor volume to the volumetric flow rate. While, due to their increased use, attention has been paid to their ability to replicate realistic tropospheric reactions by modeling the chemistry inside the reactor, there is a desire to customize flow patterns. This work demonstrates the importance of decoupling tracer signal of the reactor from that of the tubing when experimentally obtaining these flow patterns. We modeled the residence time distributions (RTDs inside the Washington University Potential Aerosol Mass (WU-PAM reactor, an OFR, for a simple set of configurations by applying the tank-in-series (TIS model, a one-parameter model, to a deconvolution algorithm. The value of the parameter, N, is close to unity for every case except one having the highest space time. Combined, the results suggest that volumetric flow rate affects mixing patterns more than use of our internals. We selected results from the simplest case, at 78 s space time with one inlet and one outlet, absent of baffles and spargers, and compared the experimental F curve to that of a computational fluid dynamics (CFD simulation. The F curves, which represent the cumulative time spent in the reactor by flowing material, match reasonably well. We value that the use of a small aspect ratio reactor such as the WU-PAM reduces wall interactions; however sudden apertures introduce disturbances in the flow, and suggest applying the methodology of tracer testing described in this work to investigate RTDs in OFRs to observe the effect of modified inlets, outlets and use of internals prior to application (e.g., field deployment vs. laboratory study.

Assessing the degree of plug flow in oxidation flow reactors (OFRs): a study on a potential aerosol mass (PAM) reactor

Science.gov (United States)

Mitroo, Dhruv; Sun, Yujian; Combest, Daniel P.; Kumar, Purushottam; Williams, Brent J.

2018-03-01

Oxidation flow reactors (OFRs) have been developed to achieve high degrees of oxidant exposures over relatively short space times (defined as the ratio of reactor volume to the volumetric flow rate). While, due to their increased use, attention has been paid to their ability to replicate realistic tropospheric reactions by modeling the chemistry inside the reactor, there is a desire to customize flow patterns. This work demonstrates the importance of decoupling tracer signal of the reactor from that of the tubing when experimentally obtaining these flow patterns. We modeled the residence time distributions (RTDs) inside the Washington University Potential Aerosol Mass (WU-PAM) reactor, an OFR, for a simple set of configurations by applying the tank-in-series (TIS) model, a one-parameter model, to a deconvolution algorithm. The value of the parameter, N, is close to unity for every case except one having the highest space time. Combined, the results suggest that volumetric flow rate affects mixing patterns more than use of our internals. We selected results from the simplest case, at 78 s space time with one inlet and one outlet, absent of baffles and spargers, and compared the experimental F curve to that of a computational fluid dynamics (CFD) simulation. The F curves, which represent the cumulative time spent in the reactor by flowing material, match reasonably well. We value that the use of a small aspect ratio reactor such as the WU-PAM reduces wall interactions; however sudden apertures introduce disturbances in the flow, and suggest applying the methodology of tracer testing described in this work to investigate RTDs in OFRs to observe the effect of modified inlets, outlets and use of internals prior to application (e.g., field deployment vs. laboratory study).

Chemical-looping combustion in a reverse-flow fixed bed reactor

International Nuclear Information System (INIS)

Han, Lu; Bollas, George M.

2016-01-01

A reverse-flow fixed bed reactor concept for CLC (chemical-looping combustion) is explored. The limitations of conventional fixed bed reactors, as applied to CLC, are overcome by reversing the gas flow direction periodically to enhance the mixing characteristics of the bed, thus improving oxygen carrier utilization and energy efficiency with respect to power generation. The reverse-flow reactor is simulated by a dusty-gas model and compared with an equivalent fixed bed reactor without flow reversal. Dynamic optimization is used to calculate conditions at which each reactor operates at maximum energy efficiency. Several cases studies illustrate the benefits of reverse-flow operation for the CLC with CuO and NiO oxygen carriers and methane and syngas fuels. The results show that periodic reversal of the flow during reduction improves the contact between the fuel and unconverted oxygen carrier, enabling the system to suppress unwanted catalytic reactions and axial temperature and conversion gradients. The operational scheme presented reduces the fluctuations of temperature during oxidation and increases the high-temperature heat produced by the process. CLC in a reverse-flow reactor has the potential to achieve higher energy efficiency than conventional fixed bed CLC reactors, when integrated with a downstream gas turbine of a combined cycle power plant. - Highlights: • Reverse-flow fixed bed CLC reactors for combined cycle power systems. • Dynamic optimization tunes operation of batch and transient CLC systems. • The reverse-flow CLC system provides stable turbine-ready gas stream. • Reverse-flow CLC fixed bed reactor has superior CO 2 capture and thermal efficiency.

Thermohydraulic calculations in rectangular channels for RA-6 type reactors with transition regime

International Nuclear Information System (INIS)

Sillin, N; Vertullo, A.; Masson, V.; Hilal, R

2009-01-01

In August 2000 and within the framework of the RA-6 core conversion from high to low enrichment (20%), a preliminary analysis was performed to evaluate the maximum power that the reactor could operate with the new kernel without makeing substantial changes. This meant keeping intact, for example, the concrete shield of the pool and the nucleus inlet and outlet pipes embedded in the walls. Preliminary results indicated that for these boundary conditions a maximum power of about 3 MWt could be achieved. In August 2005 the project was resumed and new calculations performed taking as a starting point the ECBE plate fuel element(U3O8-Al). A core was developed with cooling channle widths of 2.6 mm for the control fuel elements and 2.7 mm for standard fuel elements. The thermo-hydraulic calculation puts in evidence that coolant flow into the core was in the transitional regime for the vast majority of configurations. While TERMIC code, used for thermo-hydraulic design, has been extensively tested and validated for use in research reactors under turbulent and laminar flows, this is not so for transition conditions. The transition regime is strongly dependent on conditions such as flow inlet characteristics, channel geometry, etc.. and therefore there are no reliable correlations for general use. For this reason we found it convenient to carry out experiments simulating the working conditions in order to adjust the code results with experimental data. In the present work we show the experimental results, the simulation of the experiences using the TERMIC code, and the adjustments made to the correlations used by the code so that it can be applied to the thermo-hydraulic design of the new core. [es

CFD investigation of flow inversion in typical MTR research reactor undergoing thermal-hydraulic transients

International Nuclear Information System (INIS)

Salama, Amgad

2011-01-01

Highlights: → The 3D, CFD simulation of FLOFA accident in the generic IAEA 10 MW research reactor is carried out. → The different flow and heat transfer mechanisms involved in this process were elucidated. → The transition between these mechanisms during the course of FLOFA is discussed and investigated. → The interesting inversion process upon the transition from downward flow to upward flow is shown. → The temperature field and the friction coefficient during the whole transient process were shown. - Abstract: Three dimensional CFD full simulations of the fast loss of flow accident (FLOFA) of the IAEA 10 MW generic MTR research reactor are conducted. In this system the flow is initially downward. The transient scenario starts when the pump coasts down exponentially with a time constant of 1 s. As a result the temperatures of the heating element, the clad, and the coolant rise. When the flow reaches 85% of its nominal value the control rod system scrams and the power drops sharply resulting in the temperatures of the different components to drop. As the coolant flow continues to drop, the decay heat causes the temperatures to increase at a slower rate in the beginning. When the flow becomes laminar, the rate of temperature increase becomes larger and when the pumps completely stop a flow inversion occurs because of natural convection. The temperature will continue to rise at even higher rates until natural convection is established, that is when the temperatures settle off. The interesting 3D patterns of the flow during the inversion process are shown and investigated. The temperature history is also reported and is compared with those estimated by one-dimensional codes. Generally, very good agreement is achieved which provides confidence in the modeling approach.

Reactor coolant flow measurements at Point Lepreau

International Nuclear Information System (INIS)

Brenciaglia, G.; Gurevich, Y.; Liu, G.

1996-01-01

The CROSSFLOW ultrasonic flow measurement system manufactured by AMAG is fully proven as reliable and accurate when applied to large piping in defined geometries for such applications as feedwater flows measurement. Its application to direct reactor coolant flow (RCF) measurements - both individual channel flows and bulk flows such as pump suction flow - has been well established through recent work by AMAG at Point Lepreau, with application to other reactor types (eg. PWR) imminent. At Point Lepreau, Measurements have been demonstrated at full power; improvements to consistently meet ±1% accuracy are in progress. The development and recent customization of CROSSFLOW to RCF measurement at Point Lepreau are described in this paper; typical measurement results are included. (author)

Slug flow transitions in horizontal gas/liquid two-phase flows. Dependence on channel height and system pressure for air/water and steam/water two-phase flows

International Nuclear Information System (INIS)

Nakamura, Hideo

1996-05-01

The slug flow transitions and related phenomena for horizontal two-phase flows were studied for a better prediction of two-phase flows that typically appear during the reactor loss-of-coolant accidents (LOCAs). For better representation of the flow conditions experimentally, two large-scaled facility: TPTF for high-pressure steam/water two-phase flows and large duct test facility for air/water two-phase flows, were used. The visual observation of the flow using a video-probe was performed in the TPTF experiments for good understanding of the phenomena. The currently-used models and correlations based mostly on the small-scale low-pressure experiments were reviewed and improved based on these experimental results. The modified Taitel-Dukler model for prediction of transition into slug flow from wavy flow and the modified Steen-Wallis correlation for prediction of onset of liquid entrainment from the interfacial waves were obtained. An empirical correlation for the gas-liquid interfacial friction factor was obtained further for prediction of liquid levels at wavy flow. The region of slug flow regime that is generally under influences of the channel height and system pressure was predicted well when these models and correlations were applied together. (author). 90 refs

Experimental investigation on flow behavior during start-up of a heating reactor

International Nuclear Information System (INIS)

Jiang Shengyao; Wu Xinxin; Zhang Youjie

1997-09-01

An experimental simulation study on the transition from pressurized to boiling operation of a low-temperature, natural circulation nuclear heating reactor (5 MW) developed by INET of Tsinghua University is presented. The experiment was performed on the test loop (HRTL-5), which simulates the geometry and system design of the 5 MW reactor. The manifestation of different kinds of two-phase flow instabilities, namely geyser instability, flashing instability and low-steam quality density wave instability on the transition from pressurized to boiling operation is described. The mechanism of flashing instability, which has never been studied well on this field, is especially interpreted. It is suggested that the start-up process, from initial condition to boiling operation condition, should consist of three steps: (1) increasing of initial pressure by means of a noncondensable gas (N 2 ), which is a very effective method to eliminate geyser instability and flashing instability at lower pressure. (2)start-up of the reactor at this pressurized condition with a constant heat flux under the limited value of q = 0.15 MW·m -2 , which controls the exit temperature of the heated section below the one of net vapor generation, the low steam quality density wave oscillation can be avoided. (3) transition to a lower pressure, boiling operation. The method of transition with low-heat flux and low-inlet subcooling is proposed: at pressurized operation condition, by reducing the heat flux to its lowest level, releasing the noncondensable gas and increasing the heat flux gradually (dq/dt -2 ·min -1 ), during which the low-steam quality density wave oscillation can be prevented from occurring, then the boiling operation condition can be achieved through adjusting the heat flux and inlet subcooling to their designed value. A stable transition from pressurized to boiling operation of the 5 MW reactor is achieved by careful selection of the thermohydraulic parameters. (7 refs., 7 figs., 1

The flow measurement methods for the primary system of integral reactors

International Nuclear Information System (INIS)

Lee, J.; Seo, J. K.; Lee, D. J.

2001-01-01

It is the common features of the integral reactors that the main components of the primary system are installed within the reactor vessel, and so there are no any flow pipes connecting the reactor coolant pumps or steam generators. Due to no any flow pipes, it is impossible to measure the differential pressure at the primary system of the integral reactors, and it also makes impossible measure the primary coolant flow rate. The objective of the study is to draw up the flow measurement methods for the primary system of integral reactors. As a result of the review, we have made a selection of the flow measurement method by pump speed, bt HBM, and by pump motor power as the flow measurement methods for the primary system of integral reactors. Peculiarly, we did not found out a precedent which the direct pump motor power-flow rate curve is used as the flow measurement method in the existing commercial nuclear power reactors. Therefore, to use this method for integral reactors, it is needed to bear the follow-up measures in mind. The follow-up measures is included in this report

Stresses in transition region of VVER-1000 reactor vessels

International Nuclear Information System (INIS)

Namgung, I.; Nguye, T.L.

2014-01-01

Most of the western PWR reactor's bottom head is hemi-spherical shape, however for Russian designed VVER family of reactor it is ellipsoidal shape. The transition region from shell side to ellipsoidal head and transition top flange to cylindrical shell develop higher stress concentration than western PWR reactor vessel. This region can be modeled as conical shell with varying thickness. The theoretical derivation of stress in the thick-walled conical cylinder with varying thickness was developed and shown in detail. The results is applied to VVER-1000 reactor vessel of which shell to bottom ellipsoidal shell and shell to upper flange were investigated for stress field. The theoretical calculations were also compared with FEM solutions. An axisymmetric 3D model of VVER-1000 reactor vessel (without closure head) FEM model was created and internal hydrostatic pressure boundary condition was applied. The stress results from FEM and theoretical calculation were compared, and the discrepancies and accuracies of the theoretical results were discussed. (author)

Stresses in transition region of VVER-1000 reactor vessels

Energy Technology Data Exchange (ETDEWEB)

Namgung, I. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); Nguye, T.L. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); National Research Inst. of Mechanical Engineering, Hanoi City, Vietnam (China)

2014-07-01

Most of the western PWR reactor's bottom head is hemi-spherical shape, however for Russian designed VVER family of reactor it is ellipsoidal shape. The transition region from shell side to ellipsoidal head and transition top flange to cylindrical shell develop higher stress concentration than western PWR reactor vessel. This region can be modeled as conical shell with varying thickness. The theoretical derivation of stress in the thick-walled conical cylinder with varying thickness was developed and shown in detail. The results is applied to VVER-1000 reactor vessel of which shell to bottom ellipsoidal shell and shell to upper flange were investigated for stress field. The theoretical calculations were also compared with FEM solutions. An axisymmetric 3D model of VVER-1000 reactor vessel (without closure head) FEM model was created and internal hydrostatic pressure boundary condition was applied. The stress results from FEM and theoretical calculation were compared, and the discrepancies and accuracies of the theoretical results were discussed. (author)

Flow distribution of pebble bed high temperature gas cooled reactors using large eddy simulation

International Nuclear Information System (INIS)

Gokhan Yesilyurt; Hassan, Y.A.

2003-01-01

A High Temperature Gas-cooled Reactor (HTGR) is one of the renewed reactor designs to play a role in nuclear power generation. This reactor design concepts is currently under consideration and development worldwide. Since the HTGR concept offers inherent safety, has a very flexible fuel cycle with capability to achieve high burnup levels, and provides good thermal efficiency of power plant, it can be considered for further development and improvement as a reactor concept of generation IV. The combination of coated particle fuel, inert helium gas as coolant and graphite moderated reactor makes it possible to operate at high temperature yielding a high efficiency. In this study the simulation of turbulent transport for the gas through the gaps of the spherical fuel elements (fuel pebbles) will be performed. This will help in understanding the highly three-dimensional, complex flow phenomena in pebble bed caused by flow curvature. Under these conditions, heat transfer in both laminar and turbulent flows varies noticeably around curved surfaces. Curved flows would be present in the presence of contiguous curved surfaces. In the case of a laminar flow and of an appreciable effect of thermogravitional forces, the Nusselt (Nu) number depends significantly on the curvature shape of the surface. It changes with order of 10 times. The flow passages through the gap between the fuel balls have concave and convex configurations. Here the action of the centrifugal forces manifests itself differently on convex and concave parts of the flow path (suppression or stimulation of turbulence). The flow of this type has distinctive features. In such flow there is a pressure gradient, which strongly affects the boundary layer behavior. The transition from a laminar to turbulent flow around this curved flow occurs at deferent Reynolds (Re) numbers. Consequently, noncircular curved flows as in the pebble-bed situation, in detailed local sense, is interesting to be investigated. To the

Analyses of Decrease in Reactor Coolant Flow Rate in SMART

International Nuclear Information System (INIS)

Kim, Hyung Rae; Bae, Kyoo Hwan; Choi, Suhn

2011-01-01

SMART is a small integral reactor, which is under development at KAERI to get the standard design approval by the end of 2011. SMART works like a pressurized light-water reactor in principle though it is more compact than large commercial reactors. SMART houses major components such as steam generators, a pressurizer, and reactor coolant pumps inside the reactor pressure vessel. Due to its compact design, SMART adopts a canned-motor type reactor coolant pump which has much smaller rotational inertia than the ones used in commercial reactors. As a consequence, the reactor coolant pump has very short coastdown time and reactor coolant flow rate decreases more severely compared to commercial reactors. The transients initiated by reduction of reactor coolant flow rate have been analyzed to ensure that SMART can be safely shutdown on such transients. The design basis events in this category are complete loss of flow, single pump locked rotor with loss of offsite power, and single pump shaft break with loss of offsite power

Pulsatile pipe flow transition: Flow waveform effects

Science.gov (United States)

Brindise, Melissa C.; Vlachos, Pavlos P.

2018-01-01

Although transition is known to exist in various hemodynamic environments, the mechanisms that govern this flow regime and their subsequent effects on biological parameters are not well understood. Previous studies have investigated transition in pulsatile pipe flow using non-physiological sinusoidal waveforms at various Womersley numbers but have produced conflicting results, and multiple input waveform shapes have yet to be explored. In this work, we investigate the effect of the input pulsatile waveform shape on the mechanisms that drive the onset and development of transition using particle image velocimetry, three pulsatile waveforms, and six mean Reynolds numbers. The turbulent kinetic energy budget including dissipation rate, production, and pressure diffusion was computed. The results show that the waveform with a longer deceleration phase duration induced the earliest onset of transition, while the waveform with a longer acceleration period delayed the onset of transition. In accord with the findings of prior studies, for all test cases, turbulence was observed to be produced at the wall and either dissipated or redistributed into the core flow by pressure waves, depending on the mean Reynolds number. Turbulent production increased with increasing temporal velocity gradients until an asymptotic limit was reached. The turbulence dissipation rate was shown to be independent of mean Reynolds number, but a relationship between the temporal gradients of the input velocity waveform and the rate of turbulence dissipation was found. In general, these results demonstrated that the shape of the input pulsatile waveform directly affected the onset and development of transition.

Control of reactor coolant flow path during reactor decay heat removal

International Nuclear Information System (INIS)

Hunsbedt, A.N.

1988-01-01

This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

CFD simulation on reactor flow mixing phenomena

International Nuclear Information System (INIS)

Kwon, T.S.; Kim, K.H.

2016-01-01

A pre-test calculation for multi-dimensional flow mixing in a reactor core and downcomer has been studied using a CFD code. To study the effects of Reactor Coolant Pump (RCP) and core zone on the boron mixing behaviors in a lower downcomer and core inlet, a 1/5-scale CFD model of flow mixing test facility for the APR+ reference plant was simulated. The flow paths of the 1/5-scale model were scaled down by the linear scaling method. The aspect ratio (L/D) of all flow paths was preserved to 1. To preserve a dynamic similarity, the ratio of Euler number was also preserved to 1. A single phase water flow at low pressure and temperature conditions was considered in this calculation. The calculation shows that the asymmetric effect driven by RCPs shifted the high velocity field to the failed pump's flow zone. The borated water flow zone at the core inlet was also shifted to the failed RCP side. (author)

Experimental study on flow pattern transitions for inclined two-phase flow

Energy Technology Data Exchange (ETDEWEB)

Kwak, Nam Yee; Lee, Jae Young [Handong Univ., Pohang (Korea, Republic of); Kim, Man Woong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2007-07-01

In this paper, experimental data on flow pattern transition of inclination angles from 0-90 are presented. A test section is constructed 2 mm long and I.D 1inch using transparent material. The test section is supported by aluminum frame that can be placed with any arbitrary inclined angles. The air-water two-phase flow is observed at room temperature and atmospheric condition using both high speed camera and void impedance meter. The signal is sampled with sampling rate 1kHz and is analyzed under fully-developed condition. Based on experimental data, flow pattern maps are made for various inclination angles. As increasing the inclination angels from 0 to 90, the flow pattern transitions on the plane jg-jf are changed, such as stratified flow to plug flow or slug flow or plug flow to bubbly flow. The transition lines between pattern regimes are moved or sometimes disappeared due to its inclined angle.

Flow rate control systems for coolants for BWR type reactors

International Nuclear Information System (INIS)

Igarashi, Yoko; Kato, Naoyoshi.

1981-01-01

Purpose: To increase spontaneous recycling flow rate of coolants in BWR type reactors when the water level in the reactor decreases, by communicating a downcomer with a lower plenum. Constitution: An opening is provided to the back plate disposed at the lower end of a reactor core shroud for communicating a downcomer with a lower plenum, and an ON-OFF valve actuated by an operation rod is provided to the opening. When abnormal water level or pressure in the reactor is detected by a level metal or pressure meter, the operation rod is driven to open the ON-OFF valve, whereby coolants fed from a jet pump partially flows through the opening to increase the spontaneous recycling flow rate of the coolants. This can increase the spontaneous recycling flow rate of the coolants upon spontaneous recycling operation, thereby maintaining the reactor safety and the fuel soundness. (Moriyama, K.)

Flow reactor studies of non-equilibrium plasma-assisted oxidation of n-alkanes.

Science.gov (United States)

Tsolas, Nicholas; Lee, Jong Guen; Yetter, Richard A

2015-08-13

The oxidation of n-alkanes (C1-C7) has been studied with and without the effects of a nanosecond, non-equilibrium plasma discharge at 1 atm pressure from 420 to 1250 K. Experiments have been performed under nearly isothermal conditions in a flow reactor, where reactive mixtures are diluted in Ar to minimize temperature changes from chemical reactions. Sample extraction performed at the exit of the reactor captures product and intermediate species and stores them in a multi-position valve for subsequent identification and quantification using gas chromatography. By fixing the flow rate in the reactor and varying the temperature, reactivity maps for the oxidation of fuels are achieved. Considering all the fuels studied, fuel consumption under the effects of the plasma is shown to have been enhanced significantly, particularly for the low-temperature regime (T<800 K). In fact, multiple transitions in the rates of fuel consumption are observed depending on fuel with the emergence of a negative-temperature-coefficient regime. For all fuels, the temperature for the transition into the high-temperature chemistry is lowered as a consequence of the plasma being able to increase the rate of fuel consumption. Using a phenomenological interpretation of the intermediate species formed, it can be shown that the active particles produced from the plasma enhance alkyl radical formation at all temperatures and enable low-temperature chain branching for fuels C3 and greater. The significance of this result demonstrates that the plasma provides an opportunity for low-temperature chain branching to occur at reduced pressures, which is typically observed at elevated pressures in thermal induced systems. © 2015 The Author(s) Published by the Royal Society. All rights reserved.

Transitioning nuclear fuel cycles with uncertain fast reactor costs

Energy Technology Data Exchange (ETDEWEB)

Phathanapirom, U.B., E-mail: bphathanapirom@utexas.edu; Schneider, E.A.

2016-06-15

This paper applies a novel decision making methodology to a case study involving choices leading to the transition from the current once-through light water reactor fuel cycle to one relying on continuous recycle of plutonium and minor actinides in fast reactors in the face of uncertain fast reactor capital costs. Unique to this work is a multi-stage treatment of a range of plausible trajectories for the evolution of fast reactor capital costs over time, characterized by first-of-a-kind penalties as well as time- and unit-based learning. The methodology explicitly incorporates uncertainties in key parameters into the decision-making process by constructing a stochastic model and embedding uncertainties as bifurcations in the decision tree. “Hedging” strategies are found by applying a choice criterion to select courses of action which mitigate “regrets”. These regrets are calculated by evaluating the performance of all possible transition strategies for every feasible outcome of the uncertain parameter. The hedging strategies are those that preserve the most flexibility for adjusting the fuel cycle strategy in response to new information as uncertainties are resolved.

Transitioning nuclear fuel cycles with uncertain fast reactor costs

International Nuclear Information System (INIS)

Phathanapirom, U.B.; Schneider, E.A.

2016-01-01

This paper applies a novel decision making methodology to a case study involving choices leading to the transition from the current once-through light water reactor fuel cycle to one relying on continuous recycle of plutonium and minor actinides in fast reactors in the face of uncertain fast reactor capital costs. Unique to this work is a multi-stage treatment of a range of plausible trajectories for the evolution of fast reactor capital costs over time, characterized by first-of-a-kind penalties as well as time- and unit-based learning. The methodology explicitly incorporates uncertainties in key parameters into the decision-making process by constructing a stochastic model and embedding uncertainties as bifurcations in the decision tree. “Hedging” strategies are found by applying a choice criterion to select courses of action which mitigate “regrets”. These regrets are calculated by evaluating the performance of all possible transition strategies for every feasible outcome of the uncertain parameter. The hedging strategies are those that preserve the most flexibility for adjusting the fuel cycle strategy in response to new information as uncertainties are resolved.

Predictions of the Bypass Flows in the HTR-PM Reactor Core

International Nuclear Information System (INIS)

Sun Jun; Chen Zhipeng; Zheng Yanhua; Shi Lei; Li Fu

2014-01-01

In the HTR-PM reactor core, the basic structure materials are large amount of graphite reflectors and carbon bricks. Small gaps among those graphite and carbon bricks are widespread in the reactor core so that the cold helium flow may be bypassed and not completely heated. The bypass flows in relative lower temperature would change the flow and temperature distributions in the reactor core, therefore, the accurate prediction of bypass flows need to be carried out carefully to evaluate the influence to the reactor safety. Based on the characteristics of the bypass flow problem, hybrid method of the flow network and the CFD tools was employed to represent the connections and calculate flow distributions of all the main flow and bypass flow paths. In this paper, the hybrid method was described and applied to specific bypass flow problem in the HTR-PM. Various bypass flow paths in the HTR-PM were reviewed, figured out, and modeled by the flow network and the CFD methods, including the axial vertical gaps in the side reflectors, control rod channels, absorber sphere channels and radial gap flow through keys around the hot helium plenum. The bypass flow distributions and its flow rate ratio to the total flow rate in the primary loop were also calculated, discussed and evaluated. (author)

A fast spectrum dual path flow cermet reactor

International Nuclear Information System (INIS)

Anghaie, S.; Feller, G.J.; Peery, S.D.; Parsley, R.C.

1993-01-01

A cermet fueled, dual path fast reactor for space nuclear propulsion applications is conceptually designed. The reactor utilizes an outer annulus core and an inner cylindrical core with radial and axial reflector. The dual path flow minimizes the impact of power peaking near the radial reflector. Basic neutronics and core design aspects of the reactor are discussed. The dual path reactor is integrated into a 25000 lbf thrust nuclear rocket

FFTF and Advanced Reactors Transition Program Resource Loaded Schedule

Energy Technology Data Exchange (ETDEWEB)

GANTT, D.A.

2000-10-31

This Resource Load Schedule (RLS) addresses two missions. The Advanced Reactors Transition (ART) mission, funded by DOE-EM, is to transition assigned, surplus facilities to a safe and compliant, low-cost, stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D&D. Facilities to be transitioned include the 309 Building Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy Legacy facilities. This mission is funded through the Environmental Management (EM) Project Baseline Summary (PBS) RL-TP11, ''Advanced Reactors Transition.'' The second mission, the Fast Flux Test Facility (FFTF) Project, is funded through budget requests submitted to the Office of Nuclear Energy, Science and Technology (DOE-NE). The FFTF Project mission is maintaining the FFTF, the Fuels and Materials Examination Facility (FMEF), and affiliated 400 Area buildings in a safe and compliant standby condition. This mission is to preserve the condition of the plant hardware, software, and personnel in a manner not to preclude a plant restart. This revision of the Resource Loaded Schedule (RLS) is based upon the technical scope in the latest revision of the following project and management plans: Fast Flux Test Facility Standby Plan (Reference 1); Hanford Site Sodium Management Plan (Reference 2); and 309 Building Transition Plan (Reference 4). The technical scope, cost, and schedule baseline is also in agreement with the concurrent revision to the ART Fiscal Year (FY) 2001 Multi-Year Work Plan (MYWP), which is available in an electronic version (only) on the Hanford Local Area Network, within the ''Hanford Data Integrator (HANDI)'' application.

Technical note: Development of a Linear Flow Channel Reactor for ...

African Journals Online (AJOL)

Technical note: Development of a Linear Flow Channel Reactor for sulphur removal ... AFRICAN JOURNALS ONLINE (AJOL) · Journals · Advanced Search ... 000 mg∙ℓ-1 Na2SO4 solution) and the Liner Flow Channel Reactors (surface area ...

Experimental and numerical investigations of high temperature gas heat transfer and flow in a VHTR reactor core

Science.gov (United States)

Valentin Rodriguez, Francisco Ivan

High pressure/high temperature forced and natural convection experiments have been conducted in support of the development of a Very High Temperature Reactor (VHTR) with a prismatic core. VHTRs are designed with the capability to withstand accidents by preventing nuclear fuel meltdown, using passive safety mechanisms; a product of advanced reactor designs including the implementation of inert gases like helium as coolants. The present experiments utilize a high temperature/high pressure gas flow test facility constructed for forced and natural circulation experiments. This work examines fundamental aspects of high temperature gas heat transfer applied to VHTR operational and accident scenarios. Two different types of experiments, forced convection and natural circulation, were conducted under high pressure and high temperature conditions using three different gases: air, nitrogen and helium. The experimental data were analyzed to obtain heat transfer coefficient data in the form of Nusselt numbers as a function of Reynolds, Grashof and Prandtl numbers. This work also examines the flow laminarization phenomenon (turbulent flows displaying much lower heat transfer parameters than expected due to intense heating conditions) in detail for a full range of Reynolds numbers including: laminar, transition and turbulent flows under forced convection and its impact on heat transfer. This phenomenon could give rise to deterioration in convection heat transfer and occurrence of hot spots in the reactor core. Forced and mixed convection data analyzed indicated the occurrence of flow laminarization phenomenon due to the buoyancy and acceleration effects induced by strong heating. Turbulence parameters were also measured using a hot wire anemometer in forced convection experiments to confirm the existence of the flow laminarization phenomenon. In particular, these results demonstrated the influence of pressure on delayed transition between laminar and turbulent flow. The heat

Statistical criterion for Bubbly-slug flow transition

Energy Technology Data Exchange (ETDEWEB)

Zigler, J; Elias, E [Technion-Israel Inst. of Tech., Haifa (Israel). Dept. of Mechanical Engineering

1996-12-01

The investigation of flow pattern transitions is still an interesting problem in the research of multiphase Row. It has been studied theoretically, and experimental confirmation of the models has been found by many investigators. The present paper deals with a statistical approach to bubbly-slug transitions in a vertical upward two phase flow and a new transition criterion is deduced from experimental data (authors).

Renewable Wood Pulp Paper Reactor with Hierarchical Micro/Nanopores for Continuous-Flow Nanocatalysis.

Science.gov (United States)

Koga, Hirotaka; Namba, Naoko; Takahashi, Tsukasa; Nogi, Masaya; Nishina, Yuta

2017-06-22

Continuous-flow nanocatalysis based on metal nanoparticle catalyst-anchored flow reactors has recently provided an excellent platform for effective chemical manufacturing. However, there has been limited progress in porous structure design and recycling systems for metal nanoparticle-anchored flow reactors to create more efficient and sustainable catalytic processes. In this study, traditional paper is used for a highly efficient, recyclable, and even renewable flow reactor by tailoring the ultrastructures of wood pulp. The "paper reactor" offers hierarchically interconnected micro- and nanoscale pores, which can act as convective-flow and rapid-diffusion channels, respectively, for efficient access of reactants to metal nanoparticle catalysts. In continuous-flow, aqueous, room-temperature catalytic reduction of 4-nitrophenol to 4-aminophenol, a gold nanoparticle (AuNP)-anchored paper reactor with hierarchical micro/nanopores provided higher reaction efficiency than state-of-the-art AuNP-anchored flow reactors. Inspired by traditional paper materials, successful recycling and renewal of AuNP-anchored paper reactors were also demonstrated while high reaction efficiency was maintained. © 2017 The Authors. Published by Wiley-VCH Verlag GmbH & Co. KGaA.

Device for preventing cooling water from flowing out of reactor

International Nuclear Information System (INIS)

Chinen, Masanori; Kotani, Koichi; Murase, Michio.

1976-01-01

Object: To provide emergency cooling system, which can prevent cooling water bearing radioactivity from flowing to the outside of the reactor at the time of breakage of feedwater pipe, thus eliminating the possibility of exposure of the fuel rod to provide high reliability and also reducing the possibility of causing radioactive pollution. Structure: The device for preventing cooling water from flowing out from the reactor features a jet nozzle inserted in a feedwater pipe adjacent to the inlet or outlet thereof immediately before the reactor container. The nozzle outlet is provided in the vicinity of the reactor wall and in a direction opposite to the direction of out-flow, and water supplied from a high pressure pump is jetted from it. (Nakamura, S.)

Utilization of Stop-flow Micro-tubing Reactors for the Development of Organic Transformations.

Science.gov (United States)

Toh, Ren Wei; Li, Jie Sheng; Wu, Jie

2018-01-04

A new reaction screening technology for organic synthesis was recently demonstrated by combining elements from both continuous micro-flow and conventional batch reactors, coined stop-flow micro-tubing (SFMT) reactors. In SFMT, chemical reactions that require high pressure can be screened in parallel through a safer and convenient way. Cross-contamination, which is a common problem in reaction screening for continuous flow reactors, is avoided in SFMT. Moreover, the commercially available light-permeable micro-tubing can be incorporated into SFMT, serving as an excellent choice for light-mediated reactions due to a more effective uniform light exposure, compared to batch reactors. Overall, the SFMT reactor system is similar to continuous flow reactors and more superior than batch reactors for reactions that incorporate gas reagents and/or require light-illumination, which enables a simple but highly efficient reaction screening system. Furthermore, any successfully developed reaction in the SFMT reactor system can be conveniently translated to continuous-flow synthesis for large scale production.

Neutron-antineutron transition search at HFIR reactor

International Nuclear Information System (INIS)

Kamyshkov, Yuri A.

1997-01-01

A new experiment to search for neutron-antineutron transitions was recently proposed for High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). In this paper the physics motivation of a new search, the scheme and the discovery potential of the proposed HFIR-based experiment are discussed

Neutron-antineutron transition search at HFIR Reactor

International Nuclear Information System (INIS)

Kamyshkov, Y.A.

1997-01-01

A new experiment to search for neutron-antineutron transitions was recently proposed for High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). In this paper the physics motivation of a new search, the scheme and the discovery potential of the proposed HFIR-based experiment are discussed

75 FR 62695 - Physical Protection of Irradiated Reactor Fuel in Transit

Science.gov (United States)

2010-10-13

... Irradiated Reactor Fuel in Transit AGENCY: Nuclear Regulatory Commission. ACTION: Proposed rule. SUMMARY: The... nuclear fuel in transit? H. Why require a telemetric position monitoring system or an alternative tracking... nuclear fuel in transit. The interim final rule added 10 CFR 73.37, ``Requirements for Physical Protection...

Fiscal year 1999 multi-year work plan, advanced reactors transition program

International Nuclear Information System (INIS)

Gantt, D.A.

1998-01-01

The Advanced Reactors Transition (ART) has two missions. One, funded by DOE-EM is to transition assigned, surplus facilities to a safe and compliant, low-cost stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D and D. Facilities to be transitioned include the 309 Building/Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy (NE) Legacy Facilities. The second mission, funded by DOE-NE, is to maintain the Fast Flux Test Facility (FFTF) and affiliated 400 Area buildings in a safe and compliant standby condition. The condition of the plant hardware, software and personnel is to be preserved in a manner not to preclude a plant restart

Fiscal year 1999 multi-year work plan, advanced reactors transition program

Energy Technology Data Exchange (ETDEWEB)

Gantt, D.A.

1998-09-17

The Advanced Reactors Transition (ART) has two missions. One, funded by DOE-EM is to transition assigned, surplus facilities to a safe and compliant, low-cost stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D and D. Facilities to be transitioned include the 309 Building/Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy (NE) Legacy Facilities. The second mission, funded by DOE-NE, is to maintain the Fast Flux Test Facility (FFTF) and affiliated 400 Area buildings in a safe and compliant standby condition. The condition of the plant hardware, software and personnel is to be preserved in a manner not to preclude a plant restart.

Scale modeling flow-induced vibrations of reactor components

International Nuclear Information System (INIS)

Mulcahy, T.M.

1982-06-01

Similitude relationships currently employed in the design of flow-induced vibration scale-model tests of nuclear reactor components are reviewed. Emphasis is given to understanding the origins of the similitude parameters as a basis for discussion of the inevitable distortions which occur in design verification testing of entire reactor systems and in feature testing of individual component designs for the existence of detrimental flow-induced vibration mechanisms. Distortions of similitude parameters made in current test practice are enumerated and selected example tests are described. Also, limitations in the use of specific distortions in model designs are evaluated based on the current understanding of flow-induced vibration mechanisms and structural response

Nuclear reactor core flow baffling

International Nuclear Information System (INIS)

Berringer, R.T.

1979-01-01

A flow baffling arrangement is disclosed for the core of a nuclear reactor. A plurality of core formers are aligned with the grids of the core fuel assemblies such that the high pressure drop areas in the core are at the same elevations as the high pressure drop areas about the core periphery. The arrangement minimizes core bypass flow, maintains cooling of the structure surrounding the core, and allows the utilization of alternative beneficial components such as neutron reflectors positioned near the core

1-D Two-phase Flow Investigation for External Reactor Vessel Cooling

International Nuclear Information System (INIS)

Kim, Jae Cheol

2007-02-01

During a severe accident, when a molten corium is relocated in a reactor vessel lower head, the RCF(Reactor Cavity Flooding) system for ERVC (External Reactor Vessel Cooling) is actuated and coolants are supplied into a reactor cavity to remove a decay heat from the molten corium. This severe accident mitigation strategy for maintaining a integrity of reactor vessel was adopted in the nuclear power plants of APR1400, AP600, and AP1000. Under the ERVC condition, the upward two-phase flow is driven by the amount of the decay heat from the molten corium. To achieve the ERVC strategy, the two-phase natural circulation in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. Also the natural circulation flow restriction has to be minimized. In this reason, it is needed to review the fundamental structure of insulation. In the existing power plants, the insulation design is aimed at minimizing heat losses under a normal operation. Under the ERVC condition, however, the ability to form the two-phase natural circulation is uncertain. Namely, some important factors, such as the coolant inlet/outlet areas, flow restriction, and steam vent etc. in the flow channel, should be considered for ERVC design. T-HEMES 1D study is launched to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The air injection method was used to simulate the boiling at the external reactor vessel and generate the natural circulation two-phase flow. From the experimental results, the natural circulation flow rate highly depended on inlet/outlet areas and the circulation flow rate increased as the outlet height as well as the supplied water head increased. On the other hand, the simple analysis using the drift

Reactor core flow measurements during plant start-up using non-intrusive flow meter CROSSFLOW

Energy Technology Data Exchange (ETDEWEB)

Kanda, V.; Sharp, B.; Gurevich, A., E-mail: vkanda@amag-inc.com, E-mail: bsharp@amag-inc.com, E-mail: agurevich@amag-inc.com [Advanced Measurement & Analysis Group Inc., Ontario (Canada); Gurevich, Y., E-mail: yuri.gurevich@daystartech.ca [Daystar Technologies Inc., Ontario (Canada); Selvaratnarajah, S.; Lopez, A., E-mail: sselvaratnarajah@amag-inc.com, E-mail: alopez@amag-inc.com [Advanced Measurement & Analysis Group Inc., Ontario (Canada)

2013-07-01

For the first time, direct measurements of the total reactor coolant flow and the flow distribution between the inner reactor zone and the outer zone were conducted using the non-intrusive clamp on ultrasonic cross-correlation flow meter, CROSSFLOW, developed and manufactured by Advanced Measurement & Analysis Group Inc. (AMAG). The measurements were performed at Bruce Power A Unit 1 on the Pump Discharge piping of the Primary Heat Transport (PHT) system during start-up. This paper describes installation processes, hydraulic testing, uncertainty analysis and traceability of the measurements to certified standards. (author)

Low-Rynolds number k-ε turbulence model for calculation of fast-reactor-channel flows

International Nuclear Information System (INIS)

Mikhin, V.I.

2000-01-01

For calculating the turbulent flows in the complex geometry channels typical for the nuclear reactor installation elements the low-Reynolds-number k-ε turbulence model with the model functions not containing the spatial coordinate like y + is proposed. Such spatial coordinate is usually used for modeling the turbulence near the wall correctly. The model completed on the developed flow of the non-viscous incompressible liquid in the plane channel correctly describes the transition from the laminar regime to the turbulent one. The calculated skin friction coefficients obey the well-known Dean and Zarbi - Reynolds laws. The mean velocity distributions are close to that obtained from the empirical three-layer Karman model. (author)

Analyses of fluid flow and heat transfer inside calandria vessel of CANDU-6 reactor using CFD

International Nuclear Information System (INIS)

Yu, Seon Oh; Kim, Man Woong; Kim, Hho Jung

2005-01-01

In a CANDU (CANada Deuterium Uranium) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a Loss Of Coolant Accident (LOCA) with coincident Loss Of Emergency Core Cooling (LOECC). as well as normal operating conditions. This study presents assessments of moderator thermal-hydraulic characteristics in the normal operating conditions and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, using the optimized scheme, analyses of real CANDU-6 in normal operating conditions and the transition condition have been performed. The present model successfully predicted the experimental results and also reasonably assessed the thermal-hydraulic characteristics of a real CANDU-6 with 380 fuel channels. A flow regime map with major parameters representing the flow pattern inside a calandria vessel has also proposed to be used as operational and/or regulatory guidelines

The Cross-Flow Mixing Analysis of Quasi-Static Pebble Flow in Pebble Bed Reactor

International Nuclear Information System (INIS)

Fang Xiang; Liu Zhiyong; Sun Yanfei; Yang Xingtuan; Jiang Shengyao

2014-01-01

In the pebble bed reactor, large number of fuel pebbles’ movement law and moving state can affect the reactor’s design, operation and safety directly. Therefore the pebble flow, which is based on the theory of particle streaming, is one of the most important research subjects of the pebble bed reactor engineering. The in-core pebble flow is a very slow particle flow (or called quasi-static particle flow), which is very different from the usual particle motion. How to accurately describe the characteristics of in-core pebble flow is a central issue for this subject. Due to the presence of random flow, the cross-mixing phenomenon will occur inevitably. In the present paper, the mixing phenomenon of pebble flow is generalized on the basis of experiment results. The pebble flow cross-mixing probability serves as the parameter which describes both the regularity and the randomness of pebble flow. The results are provided in the form of diagrammatic presentation. (author)

A flow reactor for the flow supercritical water oxidation of wastes to mitigate the reactor corrosion problem

International Nuclear Information System (INIS)

Chitanvis, S.M.

1994-01-01

We have designed a flow tube reactor for supercritical water oxidation of wastes that confines the oxidation reaction to the vicinity of the axis of the tube. This prevents high temperatures and reactants as well as reaction products from coming in intimate contact with reactor walls. This implies a lessening of corrosion of the walls of the reactor. We display numerical simulations for a vertical reactor with conservative design parameters that illustrate our concept. We performed our calculations for the destruction of sodium nitrate by ammonium hydroxide In the presence of supercritical water, where the production of sodium hydroxide causes corrosion. We have compared these results with that for a horizontal set-up where the sodium hydroxide created during the reaction ends up on the floor of the tube, implying a higher probability of corrosion

Flow rate analysis of wastewater inside reactor tanks on tofu wastewater treatment plant

Science.gov (United States)

Mamat; Sintawardani, N.; Astuti, J. T.; Nilawati, D.; Wulan, D. R.; Muchlis; Sriwuryandari, L.; Sembiring, T.; Jern, N. W.

2017-03-01

The research aimed to analyse the flow rate of the wastewater inside reactor tanks which were placed a number of bamboo cutting. The resistance of wastewater flow inside reactor tanks might not be occurred and produce biogas fuel optimally. Wastewater from eleven tofu factories was treated by multi-stages anaerobic process to reduce its organic pollutant and produce biogas. Biogas plant has six reactor tanks of which its capacity for waste water and gas dome was 18 m3 and 4.5 m3, respectively. Wastewater was pumped from collecting ponds to reactors by either serial or parallel way. Maximum pump capacity, head, and electrical motor power was 5m3/h, 50m, and 0.75HP, consecutively. Maximum pressure of biogas inside the reactor tanks was 55 mbar higher than atmosphere pressure. A number of 1,400 pieces of cutting bamboo at 50-60 mm diameter and 100 mm length were used as bacteria growth media inside each reactor tank, covering around 14,287 m2 bamboo area, and cross section area of inner reactor was 4,9 m2. In each reactor, a 6 inches PVC pipe was installed vertically as channel. When channels inside reactor were opened, flow rate of wastewater was 6x10-1 L.sec-1. Contrary, when channels were closed on the upper part, wastewater flow inside the first reactor affected and increased gas dome. Initially, wastewater flowed into each reactor by a gravity mode with head difference between the second and third reactor was 15x10-2m. However, head loss at the second reactor was equal to the third reactor by 8,422 x 10-4m. As result, wastewater flow at the second and third reactors were stagnant. To overcome the problem pump in each reactor should be installed in serial mode. In order to reach the output from the first reactor and the others would be equal, and biogas space was not filled by wastewater, therefore biogas production will be optimum.

A Study on the Flow Characterization in the Reactor Cavity

Energy Technology Data Exchange (ETDEWEB)

Lee, Ho Jung; Ko, Kwang Jeok; Kim, Sung Hwan; Kim, Min Gyu; Cho, Yeon Ho; Kim, Hyun Min [KEPCO Engineering and Construction Co. Ltd., Deajeon (Korea, Republic of)

2016-10-15

In this study, the flow characterization of the cooling air in reactor cavity nearby RCPSA has been analyzed by using a 3 dimensional model and the ANSYS CFX software in order to predict the Convective Heat Transfer Coefficient (CHTC) of the RCPSA. The Reactor Cavity is the annular space by the concrete structure, the Reactor Cavity Pool Seal Assembly (RCPSA), which consists of the welded steel and is designed to be installed between the RV and the refueling pool floor, and the Reactor Vessel (RV). For such reason, the RCPSA should be designed to provide the cooling air passage for ventilation to circulate high temperature air passing by the RV during the reactor operation. It means that the RCPSA is influenced by the convection of cooling air and the thermal expansion of the RV. Therefore, the flow characterization at the reactor cavity is one of the factors of the RCPSA design during the reactor operation. The flow distribution of the cooling air in reactor cavity nearby RCPSA has been analyzed using ANSYS CFX software to obtain the CHTC at surface of the RCPSA. 1) The temperature from the RV and the insulation is one of the critical factors for the thermal gradient of the cooling air and the CHTC in the reactor cavity. 2) The rapid change of the CHTC in inner region nearby inner and outer flexure is related to the geometry shape of the RCPSA and velocity of cooling air.

Transition from condensation-induced counter-current flow to dispersed flow

International Nuclear Information System (INIS)

Gale, J.; Tiselj, I.

2004-01-01

Model of transition from the horizontally stratified condensation-induced counter-current flow to slug flow has been analyzed with computer code WAHA and compared to the experimental data obtained in the steamline of the PMK2 test facility of Hungarian Atomic Energy Institute. The experiment was performed in the steamline initially filled with hot vapor that was gradually flooded with cold liquid. Successful simulation of the condensation-induced water hammer that follows the transition, requires accurate description of the horizontally stratified and slug flow regimes and criteria for transition between both flow regimes. Current version of the WAHA code, not verified for the condensation induced type of the water hammer, predicts the water-hammer pressure peak that exceeds 600 bar, while the measured pressure is p m = 170 ± 50 bar. Sensitivity analysis of the inter-phase exchange terms and transition conditions, pointed to the most important closure relations for heat, mass and momentum transfer. The main conclusion of the analysis is large uncertainty of the simulations: minor modification of the crucial correlations can lead to a severe water-hammer in one case, or to the 'calm' transient without pressure peaks in the other case. Large uncertainty is observed in experiments. The same simulation was performed also with RELAP5 code. However, no water hammer was predicted. (author)

Advanced neutron source reactor probabilistic flow blockage assessment

International Nuclear Information System (INIS)

Ramsey, C.T.

1995-08-01

The Phase I Level I Probabilistic Risk Assessment (PRA) of the conceptual design of the Advanced Neutron Source (ANS) Reactor identified core flow blockage as the most likely internal event leading to fuel damage. The flow blockage event frequency used in the original ANS PRA was based primarily on the flow blockage work done for the High Flux Isotope Reactor (HFIR) PRA. This report examines potential flow blockage scenarios and calculates an estimate of the likelihood of debris-induced fuel damage. The bulk of the report is based specifically on the conceptual design of ANS with a 93%-enriched, two-element core; insights to the impact of the proposed three-element core are examined in Sect. 5. In addition to providing a probability (uncertainty) distribution for the likelihood of core flow blockage, this ongoing effort will serve to indicate potential areas of concern to be focused on in the preliminary design for elimination or mitigation. It will also serve as a loose-parts management tool

A CFD Study on Inlet Plenum Flow Field of Pebble Bed Reactor

International Nuclear Information System (INIS)

Kim, Min Hwan; Lee, Won Jae; Chang, Jong Hwa

2005-01-01

High temperature gas cooled reactor, largely divided into two types of PBR (Pebble Bed Reactor) and PMR (Prismatic Modular Reactor), has becomes great interest of researchers in connection with the hydrogen production. KAERI has started a project to develop the gas cooled reactor for the hydrogen production and has been doing in-depth study for selecting the reactor type between PBR and PMR. As a part of the study, PBMR (Pebble Bed Modular Reactor) was selected as a reference PBR reactor for the CFD analysis and the flow field of its inlet plenum was simulated with computational fluid dynamics program CFX5. Due to asymmetrical arrangement of pipes to the inlet plenum, non-uniform flow distribution has been expected to occur, giving rise to non-uniform power distribution at the core. Flow fields of different arrangement of inlet pipes were also investigated, as one of measures to reduce the non-uniformity

Measurement of two phase flow properties using the nuclear reactor instruments

International Nuclear Information System (INIS)

Albrecht, R.W.; Washington Univ., Seattle; Crowe, R.D.; Dailey, D.J.; Kosaly, G.; Damborg, M.J.

1982-01-01

A procedure is introduced for characterizing one dimensional, two phase flow in terms of three properties; propagation, structure, and dynamics. It is shown that all of these properties can be measured by analyzing the response of the reactor neutron field to a two phase flow perturbation. Therefore, a nuclear reactor can be regarded as a two phase flow instrument. (author)

Lattice Boltzmann Simulations in the Slip and Transition Flow Regime with the Peano Framework

KAUST Repository

Neumann, Philipp

2012-01-01

We present simulation results of flows in the finite Knudsen range, which is in the slip and transition flow regime. Our implementations are based on the Lattice Boltzmann method and are accomplished within the Peano framework. We validate our code by solving two- and three-dimensional channel flow problems and compare our results with respective experiments from other research groups. We further apply our Lattice Boltzmann solver to the geometrical setup of a microreactor consisting of differently sized channels and a reactor chamber. Here, we apply static adaptive grids to fur-ther reduce computational costs. We further investigate the influence of using a simple BGK collision kernel in coarse grid regions which are further away from the slip boundaries. Our results are in good agreement with theory and non-adaptive simulations, demonstrating the validity and the capabilities of our adaptive simulation software for flow problems at finite Knudsen numbers.

Optimization of up-flow anaerobic sludge blanket reactor for ...

African Journals Online (AJOL)

Optimization of up-flow anaerobic sludge blanket reactor for treatment of composite ... AFRICAN JOURNALS ONLINE (AJOL) · Journals · Advanced Search ... Granules grown in the bottom part of UASB reactor were more compact and tense ...

A two-phase flow regime map for a MAPLE-type nuclear research reactor fuel channel: Effect of hexagonal finned bundle

International Nuclear Information System (INIS)

Harvel, G.D.; Chang, J.S.

1997-01-01

A two-phase flow regime map is developed experimentally and theoretically for a vertical hexagonal flow channel with and without a 36-finned rod hexagonal bundle. This type of flow channel is of interest to MAPLE-type nuclear research reactors. The flow regime maps are determined by visual observations and observation of waveforms shown by a capacitance-type void fraction meter. The experimental results show that the inclusion of the finned hexagonal bundle shifts the flow regime transition boundaries toward higher water flow rates. Existing flow regime maps based on pipe flow require slight modifications when applied to the hexagonal flow channel with and without a MAPLE-type finned hexagonal bundle. The proposed theoretical model agrees well with experimental results

FFTF scale-model characterization of flow-induced vibrational response of reactor internals

International Nuclear Information System (INIS)

Ryan, J.A.; Julyk, L.J.

1977-01-01

As an integral part of the Fast Test Reactor Vibration Program for Reactor Internals, the flow-induced vibrational characteristics of scaled Fast Test Reactor core internal and peripheral components were assessed under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup. The Hydraulic Core Mockup, a 0.285 geometric scale model, was designed to model the vibrational and hydraulic characteristics of the Fast Test Reactor. Model component vibrational characteristics were measured and determined over a range of 36 percent to 111 percent of the scaled prototype design flow. Selected model and prototype components were shaker tested to establish modal characteristics. The dynamic response of the Hydraulic Core Mockup components exhibited no anomalous flow-rate dependent or modal characteristics, and prototype response predictions were adjudged acceptable

FFTF scale-model characterization of flow induced vibrational response of reactor internals

Energy Technology Data Exchange (ETDEWEB)

Ryan, J A; Julyk, L J [Hanford Engineering Development Laboratory, Richland, WA (United States)

1977-12-01

As an integral part of the Fast Test Reactor Vibration Program for Reactor Internals, the flow-induced vibrational characteristics of scaled Fast Test Reactor core internal and peripheral components were assessed under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup. The Hydraulic Core Mockup, a 0.285 geometric scale model, was designed to model the vibrational and hydraulic characteristics of the Fast Test Reactor. Model component vibrational characteristics were measured and determined over a range of 36% to 111% of the scaled prototype design flow. Selected model and prototype components were shaker tested to establish modal characteristics. The dynamic response of the Hydraulic Core Mockup components exhibited no anomalous flow-rate dependent or modal characteristics, and prototype response predictions were adjudged acceptable. (author)

FFTF scale-model characterization of flow induced vibrational response of reactor internals

International Nuclear Information System (INIS)

Ryan, J.A.; Julyk, L.J.

1977-01-01

As an integral part of the Fast Test Reactor Vibration Program for Reactor Internals, the flow-induced vibrational characteristics of scaled Fast Test Reactor core internal and peripheral components were assessed under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup. The Hydraulic Core Mockup, a 0.285 geometric scale model, was designed to model the vibrational and hydraulic characteristics of the Fast Test Reactor. Model component vibrational characteristics were measured and determined over a range of 36% to 111% of the scaled prototype design flow. Selected model and prototype components were shaker tested to establish modal characteristics. The dynamic response of the Hydraulic Core Mockup components exhibited no anomalous flow-rate dependent or modal characteristics, and prototype response predictions were adjudged acceptable. (author)

Advanced reactors transition fiscal year 1995 multi-year program plan WBS 7.3

International Nuclear Information System (INIS)

Loika, E.F.

1994-01-01

This document describes in detail the work to be accomplished in FY-1995 and the out years for the Advanced Reactors Transition (WBS 7.3). This document describes specific milestones and funding profiles. Based upon the Fiscal Year 1995 Multi-Year Program Plan, DOE will provide authorization to perform the work outlined in the FY 1995 MYPP. Following direction given by the US Department of Energy (DOE) on December 15, 1993, Advanced Reactors Transition (ART), previously known as Advanced Reactors, will provide the planning and perform the necessary activities for placing the Fast Flux Test Facility (FFTF) in a radiologically and industrially safe shutdown condition. The DOE goal is to accomplish the shutdown in approximately five years. The Advanced Reactors Transition Multi-Year Program Plan, and the supporting documents; i.e., the FFTF Shutdown Program Plan and the FFTF Shutdown Project Resource Loaded Schedule (RLS), are defined for the life of the Program. During the transition period to achieve the Shutdown end-state, the facilities and systems will continue to be maintained in a safe and environmentally sound condition. Additionally, facilities that were associated with the Office of Nuclear Energy (NE) Programs, and are no longer required to support the Liquid Metal Reactor Program will be deactivated and transferred to an alternate sponsor or the Decontamination and Decommissioning (D and D) Program for final disposition, as appropriate

Dynamic Time-Resolved Chirped-Pulse Rotational Spectroscopy of Vinyl Cyanide Photoproducts in a Room Temperature Flow Reactor

Science.gov (United States)

Zaleski, Daniel P.; Prozument, Kirill

2017-06-01

Chirped-pulsed (CP) Fourier transform rotational spectroscopy invented by Brooks Pate and coworkers a decade ago is an attractive tool for gas phase chemical dynamics and kinetics studies. A good reactor for such a purpose would have well-defined (and variable) temperature and pressure conditions to be amenable to accurate kinetic modeling. Furthermore, in low pressure samples with large enough number of molecular emitters, reaction dynamics can be observable directly, rather than mediated by supersonic expansion. In the present work, we are evaluating feasibility of in situ time-resolved CP spectroscopy in a room temperature flow tube reactor. Vinyl cyanide (CH_2CHCN), neat or mixed with inert gasses, flows through the reactor at pressures 1-50 μbar (0.76-38 mTorr) where it is photodissociated by a 193 nm laser. Millimeter-wave beam of the CP spectrometer co-propagates with the laser beam along the reactor tube and interacts with nascent photoproducts. Rotational transitions of HCN, HNC, and HCCCN are detected, with ≥10 μs time-steps for 500 ms following photolysis of CH_2CHCN. The post-photolysis evolution of the photoproducts' rotational line intensities is investigated for the effects of rotational and vibrational thermalization of energized photoproducts. Possible contributions from bimolecular and wall-mediated chemistry are evaluated as well.

The influence of the reactor pressure on the hydrodynamics in a cocurrent gas-liquid trickle-bed reactor

NARCIS (Netherlands)

Wammes, W.J.A.; Westerterp, K.R.

1990-01-01

The influence of the reactor pressure on the liquid hold-up in the trickle-flow regime and on the transition between trickle-flow and pulse-flow has been investigated in a trickle-flow column operating up to 6.0 MPa with water, and nitrogen or helium as the gas phase. The effect of the gas velocity

A transit-time flow meter for measuring milliliter per minute liquid flow

DEFF Research Database (Denmark)

Yang, Canqian; Kymmel, Mogens; Søeberg, Henrik

1988-01-01

A transit-time flow meter, using periodic temperature fluctuations as tracers, has been developed for measuring liquid flow as small as 0.1 ml/min in microchannels. By injecting square waves of heat into the liquid flow upstream with a tiny resistance wire heater, periodic temperature fluctuation....... This flow meter will be used to measure and control the small liquid flow in microchannels in flow injection analysis. Review of Scientific Instruments is copyrighted by The American Institute of Physics....... are generated downstream. The fundamental frequency phase shift of the temperature signal with respect to the square wave is found to be a linear function of the reciprocal mean velocity of the fluid. The transit-time principle enables the flow meter to have high accuracy, better than 0.2%, and good linearity...

Flow analysis in a supercritical water oxidation reactor

International Nuclear Information System (INIS)

Oh, C.H.; Kochan, R.J.; Beller, J.M.

1996-01-01

Supercritical water oxidation (SCWO), also known as hydrothermal oxidation (HTO), involves the oxidation of hazardous waste at conditions of elevated temperature and pressure (e.g., 500 C--600 C and 234.4 bar) in the presence of approximately 90% of water and a 10% to 20% excess amount of oxidant over the stoichiometric requirement. Under these conditions, organic compounds are completely miscible with supercritical water, oxygen and nitrogen, and are rapidly oxidized to carbon dioxide and water. The essential part of the process is the reactor. Many reactor designs such as tubular, vertical vessel, and transpiring wall type have been proposed, patented, and tested at both bench and pilot scales. These designs and performances need to be scaled up to a waste throughput 10--100 times that currently being tested. Scaling of this magnitude will be done by creating a numerical thermal-hydraulic model of the smaller reactor for which test data is available, validating the model against the available data, and then using the validated model to investigate the larger reactor performance. This paper presents a flow analysis of the MODAR bench scale reactor (vertical vessel type). These results will help in the design of the reactor in an efficient manner because the flow mixing coupled with chemical kinetics eventually affects the process destruction efficiency

Flow Reactor for studying Physicochemical and aging properties of SOA

Science.gov (United States)

Babar, Z. B.

2016-12-01

Secondary organic aerosols (SOA) have importance in environmental processes such as affecting earth's radiative balance and cloud formation processes. For studying SOA formation large scale environmental batch reactors and laboratory scale flow reactors have been used. In this study application of flow reactor to study physicochemical properties of SOA is also investigated after its characterization. The flow reactor is of cylindrical design (ID 15 cm x L 70 cm) equipped with UV lamps. It is coupled with various instruments such as scanning mobility particle sizer, NOx analyzer, ozone analyzer, VOC analyzer, hygrometer, and temperature sensors for gas and particle phase measurements. OH radicals were generated by custom build ozone generator and relative humidity. The following characterizations were performed: (1) residence time distribution (RTD) measurements, (2) RH and temperature control, (3) OH radical exposure range (atmospheric aging time), (4) gas phase oxidation of SOA precursors such as α-pinene by OH radical. The flow reactor yielded narrow RTDs. In particular, RH and temperature can be controlled effectively between 0-60% and 22-43oC, respectively. OH radical exposure ranges from 6.49x1010 to 3.68x1011 molecules/cm3s (0.49 to 4.91 days). Our initial efforts on OH radical generation using hydrogen peroxide and its quantification by using flourescenet technique will be also be presented.

Quantitative investigation of the transition process in Taylor-Couette flow

International Nuclear Information System (INIS)

Tu, Xin Cheng; Kim, Hyoung Bum Kim; Liu, Dong

2013-01-01

The transition process from circular Couette flow to Taylor vortex flow regime was experimentally investigated by measuring the instantaneous velocity vector fields at the annular gap flow region between two concentric cylinders. The proper orthogonal decomposition method, vorticity calculation, and frequency analysis were applied in order to analyze the instantaneous velocity fields to identify the flow characteristics during the transition process. From the results, the kinetic energy and corresponding reconstructed velocity fields were able to detect the onset of the transition process and the alternation of the flow structure. The intermittency and oscillation of the vortex flows during the transition process were also revealed from the analysis of the instantaneous velocity fields. The results can be a measure of identifying the critical Reynolds number of the Taylor-Couette flow from a velocity measurement method.

Complete Flow Blockage of a Fuel Channel for Research Reactor

International Nuclear Information System (INIS)

Lee, Byeonghee; Park, Suki

2015-01-01

The CHF correlation suitable for narrow rectangular channels are implemented in RELAP5/MOD3.3 code for the analyses, and the behavior of fuel temperatures and MCHFR(minimum critical heat flux ratio) are compared between the original and modified codes. The complete flow blockage of fuel channel for research reactor is analyzed using original and modified RELAP5/MOD3.3 and the results are compared each other. The Sudo-Kaminaga CHF correlation is implemented into RELAP5/MOD3.3 for analyzing the behavior of fuel adjacent to the blocked channel. A flow blockage of fuel channels can be postulated by a foreign object blocking cooling channels of fuels. Since a research reactor with plate type fuel has isolated fuel channels, a complete flow blockage of one fuel channel can cause a failure of adjacent fuel plates by the loss of cooling capability. Although research reactor systems are designed to prevent foreign materials from entering into the core, partial flow blockage accidents and following fuel failures are reported in some old research reactors. In this report, an analysis of complete flow blockage accident is presented for a 15MW pool-type research reactor with plate type fuels. The fuel surface experience different heat transfer regime in the results from original and modified RELAP5/MOD3.3. By the discrepancy in heat transfer mode of two cases, a fuel melting is expected by the modified RELAP5/MOD3.3, whereas the fuel integrity is ensured by the original code

Flow behavior of volume-heated boiling pools: implications with respect to transition phase accident conditions

International Nuclear Information System (INIS)

Ginsberg, T.; Jones, O.C. Jr.; Chen, J.C.

1979-01-01

Observations of two-phase flow fields in single-component volume-heated boiling pools were made. Photographic observations, together with pool-average void fraction measurements, indicate that the churn-turbulent flow regime is stable for superficial vapor velocities up to nearly five times the Kutateladze dispersal limit. Within this range of conditions, a churn-turbulent drift flux model provides a reasonable prediction of the pool-average void fraction data. An extrapolation of the data to transition phase accident conditions suggests that intense boilup could occur where the pool-average void fraction would be >0.6 for steel vaporization rates equivalent to power levels >1% of nominal liquid-metal fast breeder reactor power density. The extended stability of bubbly flow to unusually large vapor fluxes and void fractions, observed in some experiments, is a major unresolved issue

Modelling of non-catalytic reactors in a gas-solid trickle flow reactor: Dry, regenerative flue gas desulphurization using a silica-supported copper oxide sorbent

NARCIS (Netherlands)

Kiel, J.H.A.; Kiel, J.H.A.; Prins, W.; van Swaaij, Willibrordus Petrus Maria

1992-01-01

A one-dimensional, two-phase dispersed plug flow model has been developed to describe the steady-state performance of a relatively new type of reactor, the gas-solid trickle flow reactor (GSTFR). In this reactor, an upward-flowing gas phase is contacted with as downward-flowing dilute solids phase

CFD analysis and flow model reduction for surfactant production in helix reactor

Directory of Open Access Journals (Sweden)

Nikačević N.M.

2015-01-01

Full Text Available Flow pattern analysis in a spiral Helix reactor is conducted, for the application in the commercial surfactant production. Step change response curves (SCR were obtained from numerical tracer experiments by three-dimensional computational fluid dynamics (CFD simulations. Non-reactive flow is simulated, though viscosity is treated as variable in the direction of flow, as it increases during the reaction. The design and operating parameters (reactor diameter, number of coils and inlet velocity are varied in CFD simulations, in order to examine the effects on the flow pattern. Given that 3D simulations are not practical for fast computations needed for optimization, scale-up and control, CFD flow model is reduced to one-dimensional axial dispersion (AD model with spatially variable dispersion coefficient. Dimensionless dispersion coefficient (Pe is estimated under different conditions and results are analyzed. Finally, correlation which relates Pe number with Reynolds number and number of coils from the reactor entrance is proposed for the particular reactor application and conditions.

Development of a detailed core flow analysis code for prismatic fuel reactors

International Nuclear Information System (INIS)

Bennett, R.G.

1990-01-01

The development of a computer code for the analysis of the detailed flow of helium in prismatic fuel reactors is reported. The code, called BYPASS, solves, a finite difference control volume formulation of the compressible, steady state fluid flow in highly cross-connected flow paths typical of the Modular High-Temperature Gas Cooled Reactor (MHTGR). The discretization of the flow in a core region typically considers the main coolant flow paths, the bypass gap flow paths, and the crossflow connections between them. 16 refs., 5 figs

Reverse flow operation with reactor side feeding : analysis, modeling and simulation

NARCIS (Netherlands)

Budhi, Y.W.; Hoebink, J.H.B.J.; Schouten, J.C.

2004-01-01

The novel concept of reverse flow operation with reactor side feeding is studied for selective oxidation of NH3 to produce either N2, N2O, or NO. During normal reverse flow operation, where the feeds are alternately introduced from either end of the reactor, the conversion is always lower when

An objective indicator for two-phase flow pattern transition

International Nuclear Information System (INIS)

Hervieu, E.; Seleghim, P. Jr.

1998-01-01

This work concerns the development of a methodology which objective is to characterize and diagnose two-phase flow regime transitions. The approach is based on the fundamental assumption that a transition flow is less stationary than a flow with an established regime. In a first time, the efforts focused on: the design and construction of an experimental loop, allowing to reproduce the main horizontal two-phase flow patterns, in a stable and controlled way; the design and construction of an electrical impedance probe, providing an imaged information of the spatial phase distribution in the pipe; the systematic study of the joint time-frequency and time-scale analysis methods, which permitted to define an adequate parameter quantifying the unstationarity degree. In a second time, in order to verify the fundamental assumption, a series of experiments were conducted, which objective was to demonstrate the correlation between unstationarity and regime transition. The unstationarity degree was quantified by calculating the Gabor's transform time-frequency covariance of the impedance probe signals. Furthermore, the phenomenology of each transition was characterized by the joint moments and entropy. The results clearly show that the regime transitions are correlated with local time-frequency covariance peaks, which demonstrates that these regime transitions are characterized by a loss of stationarity. Consequently, the time-frequency covariance constitutes an objective two-phase flow regime transition indicator. (author)

Transition and Turbulence Modeling for Blunt-Body Wake Flows

Science.gov (United States)

Nance, Robert P.; Horvath, Thomas J.; Hassan, H. A.

1997-01-01

This study attempts t o improve the modeling and computational prediction of high- speed transitional wake flows. The recently developed kappa - zeta (Enstrophy) turbulence model is coupled with a newly developed transition prediction method and implemented in an implicit flow solver well-suited to hypersonic flows. In this model, transition onset is determined as part of the solution. Results obtained using the new model for a 70- deg blunted cone/sting geometry demonstrate better agreement with experimental heat- transfer measurements when compared to laminar calculations as well as solutions using the kappa - omega model. Results are also presented for the situation where transition onset is preselected. It is shown that, in this case, results are quite sensitive to location of the transition point.

Electrochemical degradation of the chloramphenicol at flow reactor

International Nuclear Information System (INIS)

Rezende, Luis Gustavo P.; Prado, Vania M. do; Rocha, Robson S.; Beati, Andre A.G.F.; Sotomayor, Maria del Pilar T.; Lanza, Marcos R.V.

2010-01-01

This paper reports a study of electrochemical degradation of the chloramphenicol antibiotic in aqueous medium using a flow-by reactor with DSA anode. The process efficiency was monitored by chloramphenicol concentration analysis with liquid chromatography (HPLC) during the experiments. Analysis of Total Organic Carbon (TOC) was performed to estimate the degradation degree and Ion Chromatography (IC) was performed to determinate inorganic ions formed during the electrochemical degradation process. In electrochemical flow-by reactor, 52% of chloramphenicol was degraded, with 12% TOC reduction. IC analysis showed the production of chloride ions (25 mg L -1 ), nitrate ions (6 mg L -1 ) and nitrite ions (4.5 mg L -1 ). (author)

Calorimetric and reactor coolant system flow uncertainty

International Nuclear Information System (INIS)

Bates, L.; McLean, T.

1991-01-01

This paper describes a methodology for the quantification of errors associated with the determination of a feedwater flow, secondary power, and Reactor Coolant System (RCS) flow used at the Trojan Nuclear Plant to ensure compliance with regulatory requirements. The sources of error in Plant indications and process measurement are identified and tracked, using examples, through the mathematical processes necessary to calculate the uncertainty in the RCS flow measurement. An error of approximately 1.4 percent is calculated for secondary power. This error results, along with the consideration of other errors, in an uncertainty of approximately 3 percent in the RCS flow determination

Topological transitions in unidirectional flow of nematic liquid crystal

Science.gov (United States)

Cummings, Linda; Anderson, Thomas; Mema, Ensela; Kondic, Lou

2015-11-01

Recent experiments by Sengupta et al. (Phys. Rev. Lett. 2013) revealed interesting transitions that can occur in flow of nematic liquid crystal under carefully controlled conditions within a long microfluidic channel of rectangular cross-section, with homeotropic anchoring at the walls. At low flow rates the director field of the nematic adopts a configuration that is dominated by the surface anchoring, being nearly parallel to the channel height direction over most of the cross-section; but at high flow rates there is a transition to a flow-dominated state, where the director configuration at the channel centerline is aligned with the flow (perpendicular to the channel height direction). We analyze simple channel-flow solutions to the Leslie-Ericksen model for nematics. We demonstrate that two solutions exist, at all flow rates, but that there is a transition between the elastic free energies of these solutions: the anchoring-dominated solution has the lowest energy at low flow rates, and the flow-dominated solution has lowest energy at high flow rates. NSF DMS 1211713.

Computational fluid dynamics simulations of light water reactor flows

International Nuclear Information System (INIS)

Tzanos, C.P.; Weber, D.P.

1999-01-01

Advances in computational fluid dynamics (CFD), turbulence simulation, and parallel computing have made feasible the development of three-dimensional (3-D) single-phase and two-phase flow CFD codes that can simulate fluid flow and heat transfer in realistic reactor geometries with significantly reduced reliance, especially in single phase, on empirical correlations. The objective of this work was to assess the predictive power and computational efficiency of a CFD code in the analysis of a challenging single-phase light water reactor problem, as well as to identify areas where further improvements are needed

Biological hydrogen production by Clostridium acetobutylicum in an unsaturated flow reactor.

Science.gov (United States)

Zhang, Husen; Bruns, Mary Ann; Logan, Bruce E

2006-02-01

A mesophilic unsaturated flow (trickle bed) reactor was designed and tested for H2 production via fermentation of glucose. The reactor consisted of a column packed with glass beads and inoculated with a pure culture (Clostridium acetobutylicum ATCC 824). A defined medium containing glucose was fed at a flow rate of 1.6 mL/min (0.096 L/h) into the capped reactor, producing a hydraulic retention time of 2.1 min. Gas-phase H2 concentrations were constant, averaging 74 +/- 3% for all conditions tested. H2 production rates increased from 89 to 220 mL/hL of reactor when influent glucose concentrations were varied from 1.0 to 10.5 g/L. Specific H2 production rate ranged from 680 to 1270 mL/g glucose per liter of reactor (total volume). The H2 yield was 15-27%, based on a theoretical limit by fermentation of 4 moles of H2 from 1 mole of glucose. The major fermentation by-products in the liquid effluent were acetate and butyrate. The reactor rapidly (within 60-72 h) became clogged with biomass, requiring manual cleaning of the system. In order to make long-term operation of the reactor feasible, biofilm accumulation in the reactor will need to be controlled through some process such as backwashing. These tests using an unsaturated flow reactor demonstrate the feasibility of the process to produce high H2 gas concentrations in a trickle-bed type of reactor. A likely application of this reactor technology could be H2 gas recovery from pre-treatment of high carbohydrate-containing wastewaters.

Investigation of slightly forced buoyant flow in a training reactor

International Nuclear Information System (INIS)

Legradi, G.; Aszodi, A.; Por, G.

2001-01-01

A measurement based on the temperature noise analysis method was carried out in the Training Reactor of the Budapest University of Technology and Economics. The main goals were the estimation of the flow velocity immediately above the reactor core and investigation of the thermal-hydraulical conditions of the reactor, mainly in the core. Subsequently 2D and 3D computations were carried out with the aid of the code CFX- 4.3. The main objective of the 2D calculation was to clarify the thermal-hydraulical conditions of the whole reactor tank with a reasonable computing demand. It was also necessary to accomplish 3D numerical investigations of the reactor core and the space above since three dimensional effects of the flow could only be studied in this way. In addition, obtaining certain boundary conditions of the 3D computations was another significant aim of the 2D investigations. It is important that the results of the noise analysis and the operational measuring system of the reactor gave us a basis for verifying our computations.(author)

Microbial community composition of a down-flow hanging sponge (DHS) reactor combined with an up-flow anaerobic sludge blanket (UASB) reactor for the treatment of municipal sewage.

Science.gov (United States)

Kubota, Kengo; Hayashi, Mikio; Matsunaga, Kengo; Iguchi, Akinori; Ohashi, Akiyoshi; Li, Yu-You; Yamaguchi, Takashi; Harada, Hideki

2014-01-01

The microbial community composition of a down-flow hanging sponge (DHS) reactor in an up-flow anaerobic sludge blanket (UASB)-DHS system used for the treatment of municipal sewage was investigated. The clone libraries showed marked differences in microbial community composition at different reactor heights and in different seasons. The dominant phylotypes residing in the upper part of the reactor were likely responsible for removing organic matters because a significant reduction in organic matter in the upper part was observed. Quantification of the amoA genes revealed that the proportions of ammonia oxidizing bacteria (AOB) varied along the vertical length of the reactor, with more AOB colonizing the middle and lower parts of the reactor than the top of the reactor. The findings indicated that sewage treatment was achieved by a separation of microbial habitats responsible for organic matter removal and nitrification in the DHS reactor. Copyright © 2013 Elsevier Ltd. All rights reserved.

An objective indicator for two-phase flow pattern transition

International Nuclear Information System (INIS)

Hervieua, E.; Seleghim, P. Jr.

1998-01-01

This work concerns the development of a methodology the objective of which is to characterize and diagnose two-phase flow regime transitions. The approach is based on the fundamental assumption that a transition flow is less stationary than a flow with an established regime. During the first time, the efforts focused on: (1) the design and construction of an experimental loop, allowing to reproduce the main horizontal two-phase flow patterns, in a stable and controlled way; (2) the design and construction of an electrical impedance probe, providing an imaged information of the spatial phase distribution in the pipe; and (3) the systematic study of the joint time-frequency and time-scale analysis methods, which permitted to define an adequate parameter quantifying the unstationarity degree. During the second time, in order to verify the fundamental assumption, a series of experiments were conducted, the objective of which was to demonstrate the correlation between unstationarity and regime transition. The unstationarity degree was quantified by calculating the Gabor's transform time-frequency covariance of the impedance probe signals. Furthermore, the phenomenology of each transition was characterized by the joint moments and entropy. The results clearly show that the regime transitions are correlated with local time-frequency covariance peaks, which demonstrates that these regime transitions are characterized by a loss of stationarity. Consequently, the time-frequency covariance constitutes an objective two-phase flow regime transition indicator. (orig.)

Self-sustained Flow-acoustic Interactions in Airfoil Transitional Boundary Layers

Science.gov (United States)

2015-07-09

AFRL-AFOSR-VA-TR-2015-0235 Self-sustained flow-acoustic interactions in airfoil transitional boundary layers Vladimir Golubev EMBRY-RIDDLE...From - To)      01-04-2012 to 31-03-2015 4.  TITLE AND SUBTITLE Self-sustained flow-acoustic interactions in airfoil transitional boundary layers 5a...complementary experimental and numerical studies of flow-acoustic resonant interactions in transitional airfoils and their impact on airfoil surface

SCDAP/RELAP5 modeling of fluid heat transfer and flow losses through porous debris in a light water reactor

International Nuclear Information System (INIS)

Harvego, E. A.; Siefken, L. J.

2000-01-01

The SCDAP/RELAP5 code is being developed at the Idaho National Engineering and Environmental Laboratory under the primary sponsorship of the U.S. Nuclear Regulatory Commission (NRC) to provide best-estimate transient simulations of light water reactor coolant systems during severe accidents. This paper describes the modeling approach used in the SCDAP/RELAP5 code to calculate fluid heat transfer and flow losses through porous debris that has accumulated in the vessel lower head and core regions during the latter stages of a severe accident. The implementation of heat transfer and flow loss correlations into the code is discussed, and calculations performed to assess the validity of the modeling approach are described. The different modes of heat transfer in porous debris include: (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, (5) film boiling, and (6) transition from film boiling to convection to vapor. The correlations for flow losses in porous debris include frictional and form losses. The correlations for flow losses were integrated into the momentum equations in the RELAP5 part of the code. Since RELAP5 is a very general non-homogeneous non-equilibrium thermal-hydraulics code, the resulting modeling methodology is applicable to a wide range of debris thermal-hydraulic conditions. Assessment of the SCDAP/RELAP5 debris bed thermal-hydraulic models included comparisons with experimental measurements and other models available in the open literature. The assessment calculations, described in the paper, showed that SCDAP/RELAP5 is capable of calculating the heat transfer and flow losses occurring in porous debris regions that may develop in a light water reactor during a severe accident

Development of a detailed core flow analysis code for prismatic fuel reactors

International Nuclear Information System (INIS)

Bennett, R.G.

1990-01-01

The detailed analysis of the core flow distribution in prismatic fuel reactors is of interest for modular high-temperature gas-cooled reactor (MHTGR) design and safety analyses. Such analyses involve the steady-state flow of helium through highly cross-connected flow paths in and around the prismatic fuel elements. Several computer codes have been developed for this purpose. However, since they are proprietary codes, they are not generally available for independent MHTGR design confirmation. The previously developed codes do not consider the exchange or diversion of flow between individual bypass gaps with much detail. Such a capability could be important in the analysis of potential fuel block motion, such as occurred in the Fort St. Vrain reactor, or for the analysis of the conditions around a flow blockage or misloaded fuel block. This work develops a computer code with fairly general-purpose capabilities for modeling the flow in regions of prismatic fuel cores. The code, called BYPASS solves a finite difference control volume formulation of the compressible, steady-state fluid flow in highly cross-connected flow paths typical of the MHTGR

Scale-model characterization of flow-induced vibrational response of FFTF reactor internals

International Nuclear Information System (INIS)

Ryan, J.A.; Mahoney, J.J.

1980-10-01

Fast Test Reactor core internal and peripheral components were assessed for flow-induced vibrational characteristics under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup as an integral part of the Fast Test Reactor Vibration Program. The Hydraulic Core Mockup was an 0.285 geometric scale model of the Fast Test Reactor internals designed to simulate prototype vibrational and hydraulic characteristics. Using water to simulate sodium coolant, vibrational characteristics were measured and determined for selected model components over the scaled flow range of 36 to 110%. Additionally, in-situ shaker tests were conducted on selected Hydraulic Core Mockup outlet plenum components to establish modal characteristics. Most components exhibited resonant response at all test flow rates; however, the measured dynamic response was neither abnormal nor anomalously flow-rate dependent, and the predicted prototype components' response were deemed acceptable

PRELIMINARY DESIGN OF OSCILLATORY FLOW BIODIESEL REACTOR FOR CONTINUOUS BIODIESEL PRODUCTION FROM JATROPHA TRIGLYCERIDES

Directory of Open Access Journals (Sweden)

AZHARI T. I. MOHD. GHAZI

2008-08-01

Full Text Available The concept of a continuous process in producing biodiesel from jatropha oil by using an Oscillatory Flow Biodiesel Reactor (OFBR is discussed in this paper. It has been recognized that the batch stirred reactor is a primary mode used in the synthesis of biodiesel. However, pulsatile flow has been extensively researcehed and the fundamental principles have been successfully developed upon which its hydrodynamics are based. Oscillatory flow biodiesel reactor offers precise control of mixing by means of the baffle geometry and pulsation which facilitates to continuous operation, giving plug flow residence time distribution with high turbulence and enhanced mass and heat transfer. In conjunction with the concept of reactor design, parameters such as reactor dimensions, the hydrodynamic studies and physical properties of reactants must be considered prior to the design work initiated recently. The OFBR reactor design involves the use of simulation software, ASPEN PLUS and the reactor design fundamentals. Following this, the design parameters shall be applied in fabricating the OFBR for laboratory scale biodiesel production.

A high-pressure plug flow reactor for combustion chemistry investigations

Science.gov (United States)

Lu, Zhewen; Cochet, Julien; Leplat, Nicolas; Yang, Yi; Brear, Michael J.

2017-10-01

A plug flow reactor (PFR) is built for investigating the oxidation chemistry of fuels at up to 50 bar and 1000 K. These conditions include those corresponding to the low temperature combustion (i.e. the autoignition) that commonly occurs in internal combustion engines. Turbulent flow that approximates ideal, plug flow conditions is established in a quartz tube reactor. The reacting mixture is highly diluted by excess air to reduce the reaction rates for kinetic investigations. A novel mixer design is used to achieve fast mixing of the preheated air and fuel vapour at the reactor entrance, reducing the issue of reaction initialization in kinetic modelling. A water-cooled probe moves along the reactor extracting gases for further analysis. Measurement of the sampled gas temperature uses an extended form of a three-thermocouple method that corrects for radiative heat losses from the thermocouples to the enclosed PFR environment. Investigation of the PFR’s operation is first conducted using non-reacting flows, and then with isooctane oxidation at 900 K and 10 bar. Mixing of the non-reacting temperature and species fields is shown to be rapid. The measured fuel consumption and CO formation are then closely reproduced by kinetic modelling using an extensively validated iso-octane mechanism from the literature and the corrected gas temperature. Together, these results demonstrate the PFR’s utility for chemical kinetic investigations.

A high-pressure plug flow reactor for combustion chemistry investigations

International Nuclear Information System (INIS)

Lu, Zhewen; Cochet, Julien; Leplat, Nicolas; Yang, Yi; Brear, Michael J

2017-01-01

A plug flow reactor (PFR) is built for investigating the oxidation chemistry of fuels at up to 50 bar and 1000 K. These conditions include those corresponding to the low temperature combustion (i.e. the autoignition) that commonly occurs in internal combustion engines. Turbulent flow that approximates ideal, plug flow conditions is established in a quartz tube reactor. The reacting mixture is highly diluted by excess air to reduce the reaction rates for kinetic investigations. A novel mixer design is used to achieve fast mixing of the preheated air and fuel vapour at the reactor entrance, reducing the issue of reaction initialization in kinetic modelling. A water-cooled probe moves along the reactor extracting gases for further analysis. Measurement of the sampled gas temperature uses an extended form of a three-thermocouple method that corrects for radiative heat losses from the thermocouples to the enclosed PFR environment. Investigation of the PFR’s operation is first conducted using non-reacting flows, and then with isooctane oxidation at 900 K and 10 bar. Mixing of the non-reacting temperature and species fields is shown to be rapid. The measured fuel consumption and CO formation are then closely reproduced by kinetic modelling using an extensively validated iso-octane mechanism from the literature and the corrected gas temperature. Together, these results demonstrate the PFR’s utility for chemical kinetic investigations. (paper)

Limiting photocurrent analysis of a wide channel photoelectrochemical flow reactor

International Nuclear Information System (INIS)

Davis, Jonathan T; Esposito, Daniel V

2017-01-01

The development of efficient and scalable photoelectrochemical (PEC) reactors is of great importance for the eventual commercialization of solar fuels technology. In this study, we systematically explore the influence of convective mass transport and light intensity on the performance of a 3D-printed PEC flow cell reactor based on a wide channel, parallel plate geometry. Using this design, the limiting current density generated from the hydrogen evolution reaction at a p-Si metal–insulator–semiconductor (MIS) photocathode was investigated under varied reactant concentration, fluid velocity, and light intensity. Additionally, a simple model is introduced to predict the range of operating conditions (reactant concentration, light intensity, fluid velocity) for which the photocurrent generated in a parallel plate PEC flow cell is limited by light absorption or mass transport. This model can serve as a useful guide for the design and operation of wide-channel PEC flow reactors. The results of this study have important implications for PEC reactors operating in electrolytes with dilute reactant concentrations and/or under high light intensities where high fluid velocities are required in order to avoid operation in the mass transport-limited regime. (paper)

Effect of a flow-corrective insert on the flow pattern in a pebble bed reactor

Energy Technology Data Exchange (ETDEWEB)

Li, Yu; Gui, Nan; Yang, Xingtuan [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China); Tu, Jiyuan [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China); School of Aerospace, Mechanical & Manufacturing Engineering, RMIT University, Melbourne 3083, VIC (Australia); Jiang, Shengyao, E-mail: shengyaojiang@sina.com [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China)

2016-04-15

Highlights: • Effect of an insert on improving flow uniformity and eliminating stagnant zone is studied. • Three values concerned with the stagnant zone, radial uniformity and flow sequence are used. • Outlet diameter is a critical parameter that determines balancing mechanism of the insert. • Height/location is varied to let the insert work in unbalanced region and avoid adverse effect. - Abstract: A flow-corrective insert is adopted in the pebble-bed high temperature gas-cooled reactor (HTGR) to improve flow performance of the pebble flow for the first time. 3D discrete element method (DEM) modeling is employed to study this slow and dense granular flow. It is verified that locating a properly designed insert in the bed can help transform unsatisfactory flow field to the preferred flow pattern for pebble bed reactors. Three characteristic values on the stagnant zone, radial uniformity and flow sequence of pebble flow are defined to evaluate uniformity of the overall flow field quantitatively. The results demonstrate that the pebble bed equipped with an insert performs better than normal beds from all these three aspects. Moreover, based on numerical experiments, several universal tips for insert design on height, location and outlet diameter are suggested.

Mass-transfer characterization in a parallel-plate electrochemical reactor with convergent flow

International Nuclear Information System (INIS)

Colli, A.N.; Bisang, J.M.

2013-01-01

Highlights: • A convergent laminar flow enhances and becomes more uniform the mass-transfer rate. • The mass-transfer rate is increased under convergent turbulent flow conditions. • The mass-transfer rate under convergent laminar flow can be theoretically predicted. • A convergent duct improves the reactor behaviour and the concept is easily applicable. -- Abstract: A continuous reduction in the cross-section area is analysed as a means of improving mass-transfer in a parallel-plate electrochemical reactor. Experimental local mass-transfer coefficients along the electrode length are reported for different values of the convergent ratio and Reynolds numbers, using the reduction of ferricyanide as a test reaction. The Reynolds numbers evaluated at the reactor inlet range from 85 to 4600 with interelectrode gaps of 2 and 4 mm. The convergent flow improves the mean mass-transfer coefficient by 10–60% and mass-transfer distribution under laminar flow conditions becomes more uniform. The experimental data under laminar flow conditions are compared with theoretical calculations obtained by a computational fluid dynamics software and also with an analytical simplified model. A suitable agreement is observed between both theoretical treatments and with the experimental results. The pressure drop across the reactor is reported and compared with theoretical predictions

Spray flow-network flow transition of binary Lennard-Jones particle system

KAUST Repository

Inaoka, Hajime

2010-07-01

We simulate gas-liquid flows caused by rapid depressurization using a molecular dynamics model. The model consists of two types of Lennard-Jones particles, which we call liquid particles and gas particles. These two types of particles are distinguished by their mass and strength of interaction: a liquid particle has heavier mass and stronger interaction than a gas particle. By simulations with various initial number densities of these particles, we found that there is a transition from a spray flow to a network flow with an increase of the number density of the liquid particles. At the transition point, the size of the liquid droplets follows a power-law distribution, while it follows an exponential distribution when the number density of the liquid particles is lower than the critical value. The comparison between the transition of the model and that of models of percolation is discussed. The change of the average droplet size with the initial number density of the gas particles is also presented. © 2010 Elsevier B.V. All rights reserved.

Spray flow-network flow transition of binary Lennard-Jones particle system

KAUST Repository

Inaoka, Hajime; Yukawa, Satoshi; Ito, Nobuyasu

2010-01-01

We simulate gas-liquid flows caused by rapid depressurization using a molecular dynamics model. The model consists of two types of Lennard-Jones particles, which we call liquid particles and gas particles. These two types of particles are distinguished by their mass and strength of interaction: a liquid particle has heavier mass and stronger interaction than a gas particle. By simulations with various initial number densities of these particles, we found that there is a transition from a spray flow to a network flow with an increase of the number density of the liquid particles. At the transition point, the size of the liquid droplets follows a power-law distribution, while it follows an exponential distribution when the number density of the liquid particles is lower than the critical value. The comparison between the transition of the model and that of models of percolation is discussed. The change of the average droplet size with the initial number density of the gas particles is also presented. © 2010 Elsevier B.V. All rights reserved.

Study of coolant flow distribution within the PWR type reactor vessel

International Nuclear Information System (INIS)

Eberle, L.M.M.

1983-01-01

The thermohydraulic design of a pressurized water reactor requires the determination of the coolant flow distributions within the reactor vessel, particulary at the core inlet. In this work it is proposed the study of this flow, using potencial flow theory governed by Laplace's equation, nabla 2 φ = O. The solution of the potential field is obtained by the finite element method, which simplifies considerably the treatment of complex geometrical configurations. The equation is solved by the finite element computer code ANSYS, developed and licensed for structural and thermal analysis by using the analogy between steady state heat transfer equation without heat generation, nabla 2 T=O, and Laplace's equation of the velocity potential. The proposed method has been applied to a commercial reactor, and the results are consistent with the available experimental data. (author) [pt

Controlled synthesis of colloidal silver nanoparticles in capillary micro-flow reactor

International Nuclear Information System (INIS)

He Shengtai; Liu Yulan; Maeda, Hideaki

2008-01-01

In this study, using a polytetrafluoroethylene (PTFE) capillary tube as a micro-flow reactor, well-dispersed colloidal silver nanoparticles were controllably synthesized with different flow rates of precursory solution. Scanning transmission electron microscopy images and UV-visible absorbance spectra showed that silver nanoparticles with large size can be prepared with slow flow rate in the PTFE capillary reactor. The effects of tube diameters on the growth of colloidal silver nanoparticles were investigated. Experiment results demonstrated that using tube with small diameter was more propitious for the controllable synthesis of silver nanoparticles with different sizes.

Thermal-hydraulic modeling of flow inversion in a research reactor

International Nuclear Information System (INIS)

Kazeminejad, H.

2008-01-01

The course of loss of flow accident and flow inversion in a pool type research reactor, with scram enabled under natural circulation condition is numerically investigated. The analyses were performed by a lumped parameters approach for the coupled kinetic-thermal-hydraulics, with continuous feedback due to coolant and fuel temperature effects. A modified Runge-Kutta method was adopted for a better solution to the set of stiff differential equations. Transient thermal-hydraulics during the process of flow inversion and establishment of natural circulation were considered for a 10-MW IAEA research reactor. Some important parameters such as the peak temperatures for the hot channel were obtained for both high-enriched and low enriched fuel. The model prediction is also verified through comparison with other computer code results reported in the literature for detailed simulations of loss of flow accidents (LOFA) and the agreement between the results for the peak clad temperatures and key parameters has been satisfactory. It was found that the flow inversion and subsequent establishment of natural circulation keep the peak cladding surface temperature below the saturation temperature to avoid the escalation of clad temperature to the level of onset of nucleate boiling and sub-cooled void formation to ensure the safe operation of the reactor

Jamming transitions induced by an attraction in pedestrian flow

Science.gov (United States)

Kwak, Jaeyoung; Jo, Hang-Hyun; Luttinen, Tapio; Kosonen, Iisakki

2017-08-01

We numerically study jamming transitions in pedestrian flow interacting with an attraction, mostly based on the social force model for pedestrians who can join the attraction. We formulate the joining probability as a function of social influence from others, reflecting that individual choice behavior is likely influenced by others. By controlling pedestrian influx and the social influence parameter, we identify various pedestrian flow patterns. For the bidirectional flow scenario, we observe a transition from the free flow phase to the freezing phase, in which oppositely walking pedestrians reach a complete stop and block each other. On the other hand, a different transition behavior appears in the unidirectional flow scenario, i.e., from the free flow phase to the localized jam phase and then to the extended jam phase. It is also observed that the extended jam phase can end up in freezing phenomena with a certain probability when pedestrian flux is high with strong social influence. This study highlights that attractive interactions between pedestrians and an attraction can trigger jamming transitions by increasing the number of conflicts among pedestrians near the attraction. In order to avoid excessive pedestrian jams, we suggest suppressing the number of conflicts under a certain level by moderating pedestrian influx especially when the social influence is strong.

Jamming transitions induced by an attraction in pedestrian flow.

Science.gov (United States)

Kwak, Jaeyoung; Jo, Hang-Hyun; Luttinen, Tapio; Kosonen, Iisakki

2017-08-01

We numerically study jamming transitions in pedestrian flow interacting with an attraction, mostly based on the social force model for pedestrians who can join the attraction. We formulate the joining probability as a function of social influence from others, reflecting that individual choice behavior is likely influenced by others. By controlling pedestrian influx and the social influence parameter, we identify various pedestrian flow patterns. For the bidirectional flow scenario, we observe a transition from the free flow phase to the freezing phase, in which oppositely walking pedestrians reach a complete stop and block each other. On the other hand, a different transition behavior appears in the unidirectional flow scenario, i.e., from the free flow phase to the localized jam phase and then to the extended jam phase. It is also observed that the extended jam phase can end up in freezing phenomena with a certain probability when pedestrian flux is high with strong social influence. This study highlights that attractive interactions between pedestrians and an attraction can trigger jamming transitions by increasing the number of conflicts among pedestrians near the attraction. In order to avoid excessive pedestrian jams, we suggest suppressing the number of conflicts under a certain level by moderating pedestrian influx especially when the social influence is strong.

10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

Science.gov (United States)

2010-01-01

... fuel in transit. 73.37 Section 73.37 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF PLANTS AND MATERIALS Physical Protection of Special Nuclear Material in Transit § 73.37 Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives. (1...

Anaerobic digestion of cheese whey using up-flow anaerobic sludge blanket reactor

Energy Technology Data Exchange (ETDEWEB)

Yan, J.Q.; Lo, K.V.; Liao, P.H.

1989-01-01

Anaerobic treatment of cheese whey using a 17.5-litre up-flow anaerobic sludge blanket reactor was investigated in the laboratory. The reactor was studied over a range of influent concentration from 4.5 to 38.1 g chemical oxygen demand per litre at a constant hydraulic retention time of 5 days. The reactor start-up and the sludge acclimatization were discussed. The reactor performance in terms of methane production, volatile fatty acids conversion, sludge net growth and chemical oxygen demand reduction were also presented in this paper. Over 97% chemical oxygen demand reduction was achieved in this experiment. At the influent concentration of 38.1 g chemical oxygen demand per litre, an instability of the reactor was observed. The results indicated that the up-flow anaerobic sludge blanket reactor process could treat cheese whey effectively.

An atmospheric pressure flow reactor: Gas phase kinetics and mechanism in tropospheric conditions without wall effects

Science.gov (United States)

Koontz, Steven L.; Davis, Dennis D.; Hansen, Merrill

1988-01-01

A new type of gas phase flow reactor, designed to permit the study of gas phase reactions near 1 atm of pressure, is described. A general solution to the flow/diffusion/reaction equations describing reactor performance under pseudo-first-order kinetic conditions is presented along with a discussion of critical reactor parameters and reactor limitations. The results of numerical simulations of the reactions of ozone with monomethylhydrazine and hydrazine are discussed, and performance data from a prototype flow reactor are presented.

Minnowbrook IV: 2003 Workshop on Transition and Unsteady Aspects of Turbomachinery Flows

Science.gov (United States)

LaGraff, John E. (Editor); Ashpis, David E.

2004-01-01

This Minnowbrook IV 2003 workshop on Transition and Unsteady Aspects of Turbomachinery Flows includes the following topics: 1) Current Issues in Unsteady Turbomachinery Flows; 2) Global Instability and Control of Low-Pressure Turbine Flows; 3) Influence of End Wall Leakage on Secondary Flow Development in Axial Turbines; 4) Active and Passive Flow Control on Low Pressure Turbine Airfoils; 5) Experimental and Numerical Investigation of Transitional Flows as Affected by Passing Wakes; 6) Effects of Freestream Turbulence on Turbine Blade Heat Transfer; 7) Bypass Transition Via Continuous Modes and Unsteady Effects on Film Cooling; 8) High Frequency Surface Heat Flux Imaging of Bypass Transition; 9) Skin Friction and Heat Flux Oscillations in Upstream Moving Wave Packets; 10) Transition Mechanisms and Use of Surface Roughness to Enhance the Benefits of Wake Passing in LP Turbines; 11) Transient Growth Approach to Roughness-Induced Transition; 12) Roughness- and Freestream-Turbulence-Induced Transient Growth as a Bypass Transition Mechanism; 13) Receptivity Calculations as a Means to Predicting Transition; 14) On Streamwise Vortices in a Curved Wall Jet and Their Effect on the Mean Flow; 15) Plasma Actuators for Separation Control of Low Pressure Turbine Blades; 16) Boundary-Layer Separation Control Under Low-Pressure-Turbine Conditions Using Glow-Discharge Plasma Actuators; 17) Control of Separation for Low Pressure Turbine Blades: Numerical Simulation; 18) Effects of Elevated Free-Stream Turbulence on Active Control of a Separation Bubble; 19) Wakes, Calming and Transition Under Strong Adverse Pressure Gradients; 20) Transitional Bubble in Periodic Flow Phase Shift; 21) Modelling Spots: The Calmed Region, Pressure Gradient Effects and Background; 22) Modeling of Unsteady Transitional Flow on Axial Compressor Blades; 23) Challenges in Predicting Component Efficiencies in Turbomachines With Low Reynolds Number Blading; 24) Observations on the Causal Relationship Between

Analytical evaluation of two-phase natural circulation flow characteristics under external reactor vessel cooling

International Nuclear Information System (INIS)

Park, Jong Woon

2009-01-01

This work proposes an analytical method of evaluating the effects of design and operating parameters on the low-pressure two-phase natural circulation flow through the annular shaped gap at the reactor vessel exterior surface heated by corium (molten core) relocated to the reactor vessel lower plenum after loss of coolant accidents. A natural circulation flow velocity equation derived from steady-state mass, momentum, and energy conservation equations for homogeneous two-phase flow is numerically solved for the core melting conditions of the APR1400 reactor. The solution is compared with existing experiments which measured natural circulation flow through the annular gap slice model. Two kinds of parameters are considered for this analytical method. One is the thermal-hydraulic conditions such as thermal power of corium, pressure and inlet subcooling. The others are those for the thermal insulation system design for the purpose of providing natural circulation flow path outside the reactor vessel: inlet flow area, annular gap clearance and system resistance. A computer program NCIRC is developed for the numerical solution of the implicit flow velocity equation.

Numerical simulation of transitional flow on a wind turbine airfoil with RANS-based transition model

Science.gov (United States)

Zhang, Ye; Sun, Zhengzhong; van Zuijlen, Alexander; van Bussel, Gerard

2017-09-01

This paper presents a numerical investigation of transitional flow on the wind turbine airfoil DU91-W2-250 with chord-based Reynolds number Rec = 1.0 × 106. The Reynolds-averaged Navier-Stokes based transition model using laminar kinetic energy concept, namely the k - kL - ω model, is employed to resolve the boundary layer transition. Some ambiguities for this model are discussed and it is further implemented into OpenFOAM-2.1.1. The k - kL - ω model is first validated through the chosen wind turbine airfoil at the angle of attack (AoA) of 6.24° against wind tunnel measurement, where lift and drag coefficients, surface pressure distribution and transition location are compared. In order to reveal the transitional flow on the airfoil, the mean boundary layer profiles in three zones, namely the laminar, transitional and fully turbulent regimes, are investigated. Observation of flow at the transition location identifies the laminar separation bubble. The AoA effect on boundary layer transition over wind turbine airfoil is also studied. Increasing the AoA from -3° to 10°, the laminar separation bubble moves upstream and reduces in size, which is in close agreement with wind tunnel measurement.

Beyond organometallic flow chemistry : the principles behind the use of continuous-flow reactors for synthesis

NARCIS (Netherlands)

Noel, T.; Su, Y.; Hessel, V.; Noël, T.

2015-01-01

Flow chemistry is typically used to enable challenging reactions which are difficult to carry out in conventional batch equipment. Consequently, the use of continuous-flow reactors for applications in organometallic and organic chemistry has witnessed a spectacular increase in interest from the

Effect of ship motions and flow stability in a small marine reactor driven by natural circulation

International Nuclear Information System (INIS)

Yoritsune, Tsutomu; Ishida, Toshihisa

2001-12-01

By using a small reactor as a power source for investigations and developments under sea, widely expanded activity is expectable. In this case, as for a nuclear reactor, small-size and lightweightness, and simplification of a system are needed with the safety. In JAERI, very small reactors for submersible research vessel (Deep-sea Reactor DRX and submersible Compact Reactor SCR) have been designed on the basis of needs investigation of sea research. Although the reactor is a PWR type, self-pressurization and natural circulation system are adopted in a primary system for small size and lightweightness. The fluid flow condition of the reactor core outlet is designed to be the two-phase with a low quality. Although the flow of a primary system is the two-phase flow with a low quality, the density wave oscillation may occur according to operating conditions. Moreover, since there are ship motions of heaving (the vertical direction acceleration) etc., when a submersible research vessel navigates on the sea surface, the circulation flow of the primary system is directly influenced by this external force. In order to maintain stable operations of the reactor, it is necessary to clarify effects of the flow stability characteristic of the primary coolant system and the external force. Until now, as for the flow stability of a nuclear reactor itself, many research reports have been published including the nuclear-coupled thermal oscillation of BWRs such as LaSalle-2, WNP-2 etc. As for the effect of external force, it is reported that the acceleration change based on a seismic wave affects the reactor core flow and the reactor power in a BWR. On the other hand, also in a PWR, since adoption of natural circulation cooling is considered for a generation 4 reactor, it is thought that the margin of the reactor core flow stability becomes an important parameter in the design. The reactor coolant flow mentioned in this report is the two-phase natural circulation flow coupled with

TREATMENT OF METHANOLIC WASTEWATER BY ANAEROBIC DOWN-FLOW HANGING SPONGE (ANDHS) REACTOR AND UASB REACTOR

Science.gov (United States)

Sumino, Haruhiko; Wada, Keiji; Syutsubo, Kazuaki; Yamaguchi, Takashi; Harada, Hideki; Ohashi, Akiyoshi

Anaerobic down-flow hanging sponge (AnDHS) reactor and UASB reactor were operated at 30℃ for over 400 days in order to investigate the process performance and the sludge characteristics of treating methanolic wastewater (2 gCOD/L). The settings OLR of AnDHS reactor and of UASB reactor were 5.0 -10.0 kgCOD/m3/d and 5.0 kgCOD/m3/d. The average of the COD removal demonstrated by both reactors were over 90% throughout the experiment. From the results of methane producing activities and the PCR-DGGE method, most methanol was directly converted to methane in both reactors. The conversion was carried out by different methanogens: one closely related to Methanomethylovorans hollandica in the AnDHS retainted sludge and the other closely related to Methanosarcinaceae and Metanosarciales in the UASB retainted sludge.

Measurement of transitional flow in pipes using ultrasonic flowmeters

Energy Technology Data Exchange (ETDEWEB)

Zheng-Gang, Liu; Guang-Sheng, Du; Zhu-Feng, Shao; Qian-Ran, He; Chun-Li, Zhou, E-mail: lzhenggang@sdu.edu.cn [School of Energy and Power Engineering, Qian-Fo-shan campus, Shandong University, Jinan City 250061, Shandong Province (China)

2014-10-01

The accuracy of an ultrasonic flowmeter depends on the ratio k of average profile velocity of pipe and average velocity of an ultrasonic propagation path. But there is no appropriate method of calculating k for transition flow. In this paper, the velocity field of the transition flow in a pipe is measured by particle image velocimetry. On this basis, the k of U-shaped and V-shaped ultrasonic flowmeter is obtained when Reynolds number is between 2000 and 20 000. It is shown that the k is constant when the Reynolds number is in the range of 2000–2400 and 5400–20 000, and the k decreases with the increasing of Re when the Reynolds number is 2400–5400. The results of study can be used to improve the measurement accuracy of ultrasonic flowmeters when flow is transition flow and can provide help for the study of pipe flow. (paper)

Validation of Reactor Physics-Thermal hydraulics Calculations for Research Reactors Cooled by the Laminar Flow of Water

Energy Technology Data Exchange (ETDEWEB)

Jordan, K. A.; Schubring, D. [Univ. of Florida, Florida (United States); Girardin, G.; Pautz, A. [Swiss Federal Institute of Technology, Zuerich (Switzerland)

2013-07-01

A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics

Validation of Reactor Physics-Thermal hydraulics Calculations for Research Reactors Cooled by the Laminar Flow of Water

International Nuclear Information System (INIS)

Jordan, K. A.; Schubring, D.; Girardin, G.; Pautz, A.

2013-01-01

A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics

Nature and characteristics of pulsing flow in trickle-bed reactors

NARCIS (Netherlands)

Boelhouwer, J.G.; Piepers, H.W.; Drinkenburg, A.A.H.

2002-01-01

Pulsing flow is well known for its advantages in terms of an increase in mass and heat transfer rates, complete catalyst wetting and a decrease in axial dispersion compared to trickle flow. The operation of a trickle-bed reactor in the pulsing flow regime is favorable in terms of a capacity increase

Flow Topology Transition via Global Bifurcation in Thermally Driven Turbulence

Science.gov (United States)

Xie, Yi-Chao; Ding, Guang-Yu; Xia, Ke-Qing

2018-05-01

We report an experimental observation of a flow topology transition via global bifurcation in a turbulent Rayleigh-Bénard convection. This transition corresponds to a spontaneous symmetry breaking with the flow becomes more turbulent. Simultaneous measurements of the large-scale flow (LSF) structure and the heat transport show that the LSF bifurcates from a high heat transport efficiency quadrupole state to a less symmetric dipole state with a lower heat transport efficiency. In the transition zone, the system switches spontaneously and stochastically between the two long-lived metastable states.

Maximum production rate optimization for sulphuric acid decomposition process in tubular plug-flow reactor

International Nuclear Information System (INIS)

Wang, Chao; Chen, Lingen; Xia, Shaojun; Sun, Fengrui

2016-01-01

A sulphuric acid decomposition process in a tubular plug-flow reactor with fixed inlet flow rate and completely controllable exterior wall temperature profile and reactants pressure profile is studied in this paper by using finite-time thermodynamics. The maximum production rate of the aimed product SO 2 and the optimal exterior wall temperature profile and reactants pressure profile are obtained by using nonlinear programming method. Then the optimal reactor with the maximum production rate is compared with the reference reactor with linear exterior wall temperature profile and the optimal reactor with minimum entropy generation rate. The result shows that the production rate of SO 2 of optimal reactor with the maximum production rate has an increase of more than 7%. The optimization of temperature profile has little influence on the production rate while the optimization of reactants pressure profile can significantly increase the production rate. The results obtained may provide some guidelines for the design of real tubular reactors. - Highlights: • Sulphuric acid decomposition process in tubular plug-flow reactor is studied. • Fixed inlet flow rate and controllable temperature and pressure profiles are set. • Maximum production rate of aimed product SO 2 is obtained. • Corresponding optimal temperature and pressure profiles are derived. • Production rate of SO 2 of optimal reactor increases by 7%.

Direct numerical simulation of reactor two-phase flows enabled by high-performance computing

Energy Technology Data Exchange (ETDEWEB)

Fang, Jun; Cambareri, Joseph J.; Brown, Cameron S.; Feng, Jinyong; Gouws, Andre; Li, Mengnan; Bolotnov, Igor A.

2018-04-01

Nuclear reactor two-phase flows remain a great engineering challenge, where the high-resolution two-phase flow database which can inform practical model development is still sparse due to the extreme reactor operation conditions and measurement difficulties. Owing to the rapid growth of computing power, the direct numerical simulation (DNS) is enjoying a renewed interest in investigating the related flow problems. A combination between DNS and an interface tracking method can provide a unique opportunity to study two-phase flows based on first principles calculations. More importantly, state-of-the-art high-performance computing (HPC) facilities are helping unlock this great potential. This paper reviews the recent research progress of two-phase flow DNS related to reactor applications. The progress in large-scale bubbly flow DNS has been focused not only on the sheer size of those simulations in terms of resolved Reynolds number, but also on the associated advanced modeling and analysis techniques. Specifically, the current areas of active research include modeling of sub-cooled boiling, bubble coalescence, as well as the advanced post-processing toolkit for bubbly flow simulations in reactor geometries. A novel bubble tracking method has been developed to track the evolution of bubbles in two-phase bubbly flow. Also, spectral analysis of DNS database in different geometries has been performed to investigate the modulation of the energy spectrum slope due to bubble-induced turbulence. In addition, the single-and two-phase analysis results are presented for turbulent flows within the pressurized water reactor (PWR) core geometries. The related simulations are possible to carry out only with the world leading HPC platforms. These simulations are allowing more complex turbulence model development and validation for use in 3D multiphase computational fluid dynamics (M-CFD) codes.

Effect of 3-D moderator flow configurations on the reactivity of CANDU nuclear reactors

International Nuclear Information System (INIS)

Zadeh, Foad Mehdi; Etienne, Stephane; Chambon, Richard; Marleau, Guy; Teyssedou, Alberto

2017-01-01

Highlights: • 3-D CFD simulations of CANDU-6 moderator flows are presented. • A thermal-hydraulic code using thermal physical fluid properties is used. • The numerical approach and convergence is validated against available data. • Flow configurations are correlated using Richardson’s number. • The interaction between moderator temperatures with reactivity is determined. - Abstract: The reactivity of nuclear reactors can be affected by thermal conditions prevailing within the moderator. In CANDU reactors, the moderator and the coolant are mechanically separated but not necessarily thermally isolated. Hence, any variation of moderator flow properties may change the reactivity. Until now, nuclear reactor calculations have been performed by assuming uniform moderator flow temperature distribution. However, CFD simulations have predicted large time dependent flow fluctuations taking place inside the calandria, which can bring about local temperature variations that can exceed 50 °C. This paper presents robust CANDU 3-D CFD moderator simulations coupled to neutronic calculations. The proposed methodology makes it possible to study not only different moderator flow configurations but also their effects on the reactor reactivity coefficient.

Flow induced vibrational excitation of nuclear reactor structures

International Nuclear Information System (INIS)

Gibert, R.J.

1979-01-01

The pressure fluctuations generated by disturbed flows, encountered in nuclear reactors induce vibrations in the structures. In order to make forecastings for these vibrational levels, it is necessary to know the characteristics of the random pressure fluctuations induced in the walls by the main flow peculiarities of the circuits. This knowledge is essentially provided by experimentation which shows that most of the energy from these fluctuations is in the low frequency area. It is also necessary to determine the transfer functions of the fluid-structure coupled system. Given the frequency range of the excitations, a calculation of the characteristics of the first eigenmodes is generally sufficient. This calculation is carried out by finite element codes, the modal dampings being assessed separately. In this paper, emphasis is placed mainly on the analysis of the sources of excitation due to flow peculiarities. Some examples will also be given of assessments of vibrations in real structures (pipes, reactor internals, etc.) and of comparisons with the experimental results obtained on models or on a site [fr

Fluid-Structure Interaction for Coolant Flow in Research-type Nuclear Reactors

International Nuclear Information System (INIS)

Curtis, Franklin G.; Ekici, Kivanc; Freels, James D.

2011-01-01

The High Flux Isotope Reactor (HFIR), located at the Oak Ridge National Laboratory (ORNL), is scheduled to undergo a conversion of the fuel used and this proposed change requires an extensive analysis of the flow through the reactor core. The core consists of 540 very thin and long fuel plates through which the coolant (water) flows at a very high rate. Therefore, the design and the flow conditions make the plates prone to dynamic and static deflections, which may result in flow blockage and structural failure which in turn may cause core damage. To investigate the coolant flow between fuel plates and associated structural deflections, the Fluid-Structure Interaction (FSI) module in COMSOL will be used. Flow induced flutter and static deflections will be examined. To verify the FSI module, a test case of a cylinder in crossflow, with vortex induced vibrations was performed and validated.

A new oxidation flow reactor for measuring secondary aerosol formation of rapidly changing emission sources

Science.gov (United States)

Simonen, Pauli; Saukko, Erkka; Karjalainen, Panu; Timonen, Hilkka; Bloss, Matthew; Aakko-Saksa, Päivi; Rönkkö, Topi; Keskinen, Jorma; Dal Maso, Miikka

2017-04-01

Oxidation flow reactors (OFRs) or environmental chambers can be used to estimate secondary aerosol formation potential of different emission sources. Emissions from anthropogenic sources, such as vehicles, often vary on short timescales. For example, to identify the vehicle driving conditions that lead to high potential secondary aerosol emissions, rapid oxidation of exhaust is needed. However, the residence times in environmental chambers and in most oxidation flow reactors are too long to study these transient effects ( ˜ 100 s in flow reactors and several hours in environmental chambers). Here, we present a new oxidation flow reactor, TSAR (TUT Secondary Aerosol Reactor), which has a short residence time ( ˜ 40 s) and near-laminar flow conditions. These improvements are achieved by reducing the reactor radius and volume. This allows studying, for example, the effect of vehicle driving conditions on the secondary aerosol formation potential of the exhaust. We show that the flow pattern in TSAR is nearly laminar and particle losses are negligible. The secondary organic aerosol (SOA) produced in TSAR has a similar mass spectrum to the SOA produced in the state-of-the-art reactor, PAM (potential aerosol mass). Both reactors produce the same amount of mass, but TSAR has a higher time resolution. We also show that TSAR is capable of measuring the secondary aerosol formation potential of a vehicle during a transient driving cycle and that the fast response of TSAR reveals how different driving conditions affect the amount of formed secondary aerosol. Thus, TSAR can be used to study rapidly changing emission sources, especially the vehicular emissions during transient driving.

Flow-induced and acoustically induced vibration experience in operating gas-cooled reactors

International Nuclear Information System (INIS)

Halvers, L.J.

1977-03-01

An overview has been presented of flow-induced and acoustically induced vibration failures that occurred in the past in gas-cooled graphite-moderated reactors, and the importance of this experience for the Gas-Cooled Fast-Breeder Reactor (GCFR) project has been assessed. Until now only failures in CO 2 -cooled reactors have been found. No problems with helium-cooled reactors have been encountered so far. It is shown that most of the failures occurred because flow-induced and acoustically induced dynamic loads were underestimated, while at the same time not enough was known about the influence of environmental parameters on material behavior. All problems encountered were solved. The comparison of the influence of the gas properties on acoustically induced and flow-induced vibration phenomena shows that the interaction between reactor design and the thermodynamic properties of the primary coolant precludes a general preference for either carbon dioxide or helium. The acoustic characteristics of CO 2 and He systems are different, but the difference in dynamic loadings due to the use of one rather than the other remains difficult to predict. A slight preference for helium seems, however, to be justified

Dynamic Behavior of Reverse Flow Reactor for Lean Methane Combustion

OpenAIRE

Yogi W. Budhi; M. Effendy; Yazid Bindar; Subagjo

2014-01-01

The stability of reactor operation for catalytic oxidation of lean CH4 has been investigated through modeling and simulation, particularly the influence of switching time and heat extraction on reverse flow reactor (RFR) performance. A mathematical model of the RFR was developed, based on one-dimensional pseudo-homogeneous model for mass and heat balances, incorporating heat loss through the reactor wall. The configuration of the RFR consisted of inert-catalyst-inert, with or without heat ext...

Flow characteristics of Korea multi-purpose research reactor

Energy Technology Data Exchange (ETDEWEB)

Heonil Kim; Hee Taek Chae; Byung Jin Jun; Ji Bok Lee [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-09-01

The construction of Korea Multi-purpose Research Reactor (KMRR), a 30 MW{sub th} open-tank-in-pool type, is completed. Various thermal-hydraulic experiments have been conducted to verify the design characteristics of the KMRR. This paper describes the commissioning experiments to determine the flow distribution of KMRR core and the flow characteristics inside the chimney which stands on top of the core. The core flow is distributed to within {+-}6% of the average values, which is sufficiently flat in the sense that the design velocity in the fueled region is satisfied. The role of core bypass flow to confine the activated core coolant in the chimney structure is confirmed.

Cellular properties of slug flow in vertical co-current gas-liquid flow: slug-churn transition

International Nuclear Information System (INIS)

Lusseyran, Francois

1990-01-01

This research thesis reports the study and description of the structure of a slug flow regime in a co-current vertical cylindrical duct, and the characterization and prediction of its transition towards a slug-churn (de-structured) regime. Flow physical mechanisms highlighted by the measurement of two important dynamics variables (wall friction and thickness of liquid films) are related to hypotheses of cellular models. The author first proposes an overview of slug flow regimes: theoretical steady and one-dimensional analysis, mass assessment equations of cellular models, application to the assessment of the flow rate and of the thickness of the film surrounding the gas slug. In the second part, the author addresses the slug flow regime transition towards the slug-churn regime: assessment of the evolution of flow dynamic properties, use of average wall friction analysis to obtain a relevant transition criterion. The third part presents experimental conditions, and measurement methods: conductometry for thickness measurement, polarography for wall friction measurement, and gas phase detection by using an optic barrier or optic fibres [fr

Neural network modeling of chaotic dynamics in nuclear reactor flows

International Nuclear Information System (INIS)

Welstead, S.T.

1992-01-01

Neural networks have many scientific applications in areas such as pattern classification and time series prediction. The universal approximation property of these networks, however, can also be exploited to provide researchers with tool for modeling observed nonlinear phenomena. It has been shown that multilayer feed forward networks can capture important global nonlinear properties, such as chaotic dynamics, merely by training the network on a finite set of observed data. The network itself then provides a model of the process that generated the data. Characterizations such as the existence and general shape of a strange attractor and the sign of the largest Lyapunov exponent can then be extracted from the neural network model. In this paper, the author applies this idea to data generated from a nonlinear process that is representative of convective flows that can arise in nuclear reactor applications. Such flows play a role in forced convection heat removal from pressurized water reactors and boiling water reactors, and decay heat removal from liquid-metal-cooled reactors, either by natural convection or by thermosyphons

Prediction of flow rates through an orifice at pressures corresponding to the transition between molecular and isentropic flow

International Nuclear Information System (INIS)

DeMuth, S.F.; Watson, J.S.

1985-01-01

A model of compressible flow through an orifice, in the region of transition from free molecular to isentropic expansion flow, has been developed and tested for accuracy. The transitional or slip regime is defined as the conditions where molecular interactions are too many for free molecular flow modeling, yet not great enough for isentropic expansion flow modeling. Due to a lack of literature establishing a well-accepted model for predicting transitional flow, it was felt such work would be beneficial. The model is nonlinear and cannot be satisfactorily linearized for a linear regression analysis. Consequently, a computer routine was developed which minimized the sum of the squares of the residual flow for the nonlinear model. The results indicate an average accuracy within 15% of the measured flow throughout the range of test conditions. Furthermore, the results of the regression analysis indicate that the transitional regime lies between Knudsen numbers of approximately 2 and 45. 4 refs., 3 figs., 1 tab

Development of a Test Facility to Simulate the Reactor Flow Distribution of APR+

International Nuclear Information System (INIS)

Euh, D. J.; Cho, S.; Youn, Y. J.; Kim, J. T.; Kang, H. S.; Kwon, T. S.

2011-01-01

Recently a design of new reactor, APR+, is being developed, as an advanced type of APR1400. In order to analyze the thermal margin and hydraulic characteristics of APR+, quantification tests for flow and pressure distribution with a conservation of flow geometry are necessary. Hetsroni (1967) proposed four principal parameters for a hydraulic model representing a nuclear reactor prototype: geometry, relative roughness, Reynolds number, and Euler number. He concluded that the Euler number should be similar in the prototype and model under the preservation of the aspect ratio on the flow path. The effect of the Reynolds number at its higher values on the Euler number is rather small, since the dependency of the form and frictional loss coefficients on the Reynolds number is seen to be small. ABB-CE has carried out several reactor flow model test programs, mostly for its prototype reactors. A series of tests were conducted using a 3/16 scale reactor model. (see Lee et al., 2001). Lee et al (1991) performed experimental studies using a 1/5.03 scale reactor flow model of Yonggwang nuclear units 3 and 4. They showed that the measured data met the acceptance criteria and were suitable for their intended use in terms of performance and safety analyses. The design of current test facility was based on the conservation of Euler number which is a ratio of pressure drop to dynamic pressure with a sufficiently turbulent region having a high Reynolds number. By referring to the previous study, the APR+ design is linearly reduced to 1/5 ratio with a 1/2 of the velocity scale, which yields a 1/39.7 of Reynolds number scaling ratio. In the present study, the design feature of the facilities, named 'ACOP', in order to investigate flow and pressure distribution are described

Numerical computation of fluid flow in different nonferrous metallurgical reactors

International Nuclear Information System (INIS)

Lackner, A.

1996-10-01

Heat, mass and fluid flow phenomena in metallurgical reactor systems such as smelting cyclones or electrolytic cells are complex and intricately linked through the governing equations of fluid flow, chemical reaction kinetics and chemical thermodynamics. The challenges for the representation of flow phenomena in such reactors as well as the transfers of these concepts to non-specialist modelers (e.g. plant operators and management personnel) can be met through scientific flow visualization techniques. In the first example the fluid flow of the gas phase and of concentrate particles in a smelting cyclone for copper production are calculated three dimensionally. The effect of design parameters (length and diameter of reactor, concentrate feeding tangentially or from the top, ..) and operating conditions are investigated. Single particle traces show, how to increase particle retention time before the particles reach the liquid film flowing down the cyclone wall. Cyclone separators are widely used in the metallurgical and chemical industry for collection of large quantities of dust. Most of the empirical models, which today are applied for the design, are lacking in being valid in the high temperature region. Therefore the numerical prediction of the collection efficiency of dust particles is done. The particle behavior close to the wall is considered by applying a particle restitution model, which calculates individual particle restitution coefficients as functions of impact velocity and impact angle. The effect of design parameters and operating are studied. Moreover, the fluid flow inside a copper refining electrolysis cell is modeled. The simulation is based on density variations in the boundary layer at the electrode surface. Density and thickness of the boundary layer are compared to measurements in a parametric study. The actual inhibitor concentration in the cell is calculated, too. Moreover, a two-phase flow approach is developed to simulate the behavior of

Review of leakage-flow-induced vibrations of reactor components

International Nuclear Information System (INIS)

Mulcahy, T.M.

1983-05-01

The primary-coolant flow paths of a reactor system are usually subject to close scrutiny in a design review to identify potential flow-induced vibration sources. However, secondary-flow paths through narrow gaps in component supports, which parallel the primary-flow path, occasionally are the excitation source for significant vibrations even though the secondary-flow rates are orders of magnitude smaller than the primary-flow rate. These so-called leakage flow problems are reviewed here to identify design features and excitation sources that should be avoided. Also, design rules of thumb are formulated that can be employed to guide a design, but quantitative prediction of component response is found to require scale-model testing

Reactor prospects of muon-catalyzed fusion of deuterium and tritium concentrated in transition metals

International Nuclear Information System (INIS)

Stacey, W.M. Jr.

1989-01-01

It is conjectured that the number of fusion events catalyzed by a single muon is orders of magnitude greater for deuterium and tritium concentrated in a transition metal than in gaseous form and that the recent observation of 2.5-MeV neutrons from a D 2 O electrolytic cell with palladium and titanium cathodes can thereby be interpreted in terms of cosmic muon-catalyzed deuterium-deuterium fusion. This suggests a new fusion reactor reactor consisting of deuterium and tritium concentrated in transition metal fuel elements in a fusion core that surrounds an accelerator-produced muon source. The feasibility of net energy production in such a reactor is established in terms of requirements on the number of fusion events catalyzed per muon. The technological implications for a power reactor based on this concept are examined. The potential of such a concept as a neutron source for materials testing and tritium and plutonium production is briefly discussed

3D CFD computations of transitional flows using DES and a correlation based transition model; Wind turbines

Energy Technology Data Exchange (ETDEWEB)

Soerensen, Niels N.

2009-07-15

The report describes the application of the correlation based transition model of Menter et. al. [1, 2] to the cylinder drag crisis and the stalled flow over an DU-96-W-351 airfoil using the DES methodology. When predicting the flow over airfoils and rotors, the laminar-turbulent transition process can be important for the aerodynamic performance. Today, the most widespread approach is to use fully turbulent computations, where the transitional process is ignored and the entire boundary layer on the wings or airfoils is handled by the turbulence model. The correlation based transition model has lately shown promising results, and the present paper describes the application of the model to predict the drag and shedding frequency for flow around a cylinder from sub to super-critical Reynolds numbers. Additionally, the model is applied to the flow around the DU-96 airfoil, at high angles of attack. (au)

Transition of unsteady velocity profiles with reverse flow

OpenAIRE

Das, Debopam; Arakeri, Jaywant H

1998-01-01

This paper deals with the stability and transition to turbulence of wall-bounded unsteady velocity profiles with reverse flow. Such flows occur, for example, during unsteady boundary layer separation and in oscillating pipe flow. The main focus is on results from experiments in time-developing flow in a long pipe, which is decelerated rapidly. The flow is generated by the controlled motion of a piston. We obtain analytical solutions for laminar flow in the pipe and in a two-dimensional channe...

State of art report for critical flow model to analyze a break flow in pressurizer of integral type reactor

Energy Technology Data Exchange (ETDEWEB)

Kang, Yeon Moon; Lee, D. J.; Yoon, J. H.; Kim, J. P.; Kim, H. Y

1999-03-01

At a critical flow condition, the flow rate can't exceed a maximum value for given upstream conditions and the limited flow rate is called as a critical flow rate. The phenomena of critical flow occur at the discharge of a single phase gas or subcooled water through nozzles and pipes. Among the previous researches on critical flow, many accurate correlations on pressure, temperature and flow rate are represented for the single phase gas. However, for the two phase critical flow, the results of previous work showed that there was a large discrepancy between the analytical and experimental data and the data were in agreement for the limited thermodynamic conditions. Thus, further studies are required to enhance the two phase critical flow model. In the integral reactor, the critical flows of nitrogen gas and subcooled water are expected for the break of gas cylinder pipeline connected to the pressurizer. It requires that the inlet shape of the pipe and the nitrogen gas effect should be considered for the critical flow of integral reactor. The nitrogen gas exist in the pressurizer may affect the flow rate of primary coolant, which has been considered only for a few previous researches. Thus, the evaluation of the effect of the nitrogen on the critical flow gas should be preceded for the proper analysis of the critical flow in the integral reactor. In this report, not only the essences of previous work on critical flow were investigated and summarized but also the effect of nitrogen gas and the inlet shape of the pipe on the critical flow were also investigated. (author)

State of art report for critical flow model to analyze a break flow in pressurizer of integral type reactor

International Nuclear Information System (INIS)

Kang, Yeon Moon; Lee, D. J.; Yoon, J. H.; Kim, J. P.; Kim, H. Y.

1999-03-01

At a critical flow condition, the flow rate can't exceed a maximum value for given upstream conditions and the limited flow rate is called as a critical flow rate. The phenomena of critical flow occur at the discharge of a single phase gas or subcooled water through nozzles and pipes. Among the previous researches on critical flow, many accurate correlations on pressure, temperature and flow rate are represented for the single phase gas. However, for the two phase critical flow, the results of previous work showed that there was a large discrepancy between the analytical and experimental data and the data were in agreement for the limited thermodynamic conditions. Thus, further studies are required to enhance the two phase critical flow model. In the integral reactor, the critical flows of nitrogen gas and subcooled water are expected for the break of gas cylinder pipeline connected to the pressurizer. It requires that the inlet shape of the pipe and the nitrogen gas effect should be considered for the critical flow of integral reactor. The nitrogen gas exist in the pressurizer may affect the flow rate of primary coolant, which has been considered only for a few previous researches. Thus, the evaluation of the effect of the nitrogen on the critical flow gas should be preceded for the proper analysis of the critical flow in the integral reactor. In this report, not only the essences of previous work on critical flow were investigated and summarized but also the effect of nitrogen gas and the inlet shape of the pipe on the critical flow were also investigated. (author)

Modeling on bubbly to churn flow pattern transition for vertical upward flows in narrow rectangular channel

International Nuclear Information System (INIS)

Wang Yanlin; Chen Bingde; Huang Yanping; Wang Junfeng

2011-01-01

A theoretical model was developed to predict the bubbly to churn flow pattern transition for vertical upward flows in narrow rectangular channel. The model was developed based on the imbalance theory of Helmholtz and some reasonable assumptions. The maximum ideal bubble in narrow rectangular channel and the thermal hydraulics boundary condition leading to bubbly flow to churn flow pattern transition was calculated. The model was validated by experimental data from previous researches. Comparison between predicted result and experimental result shows a reasonable good agreement. (author)

An estimation of reactor thermal power uncertainty using UFM-based feedwater flow rate in nuclear power plants

International Nuclear Information System (INIS)

Byung Ryul Jung; Ho Cheol Jang; Byung Jin Lee; Se Jin Baik; Woo Hyun Jang

2005-01-01

Most of Pressurized Water Reactors (PWRs) utilize the venturi meters (VMs) to measure the feedwater (FW) flow rate to the steam generator in the calorimetric measurement, which is used in the reactor thermal power (RTP) estimation. However, measurement drifts have been experienced due to some anomalies on the venturi meter (generally called the venturi meter fouling). The VM's fouling tends to increase the measured pressure drop across the meter, which results in indication of increased feedwater flow rate. Finally, the reactor thermal power is overestimated and the actual reactor power is to be reduced to remain within the regulatory limits. To overcome this VM's fouling problem, the Ultrasonic Flow Meter (UFM) has recently been gaining attention in the measurement of the feedwater flow rate. This paper presents the applicability of a UFM based feedwater flow rate in the estimation of reactor thermal power uncertainty. The FW and RTP uncertainties are compared in terms of sensitivities between the VM- and UFM-based feedwater flow rates. Data from typical Optimized Power Reactor 1000 (OPR1000) plants are used to estimate the uncertainty. (authors)

Effects of turbulence modelling on prediction of flow characteristics in a bench-scale anaerobic gas-lift digester.

Science.gov (United States)

Coughtrie, A R; Borman, D J; Sleigh, P A

2013-06-01

Flow in a gas-lift digester with a central draft-tube was investigated using computational fluid dynamics (CFD) and different turbulence closure models. The k-ω Shear-Stress-Transport (SST), Renormalization-Group (RNG) k-∊, Linear Reynolds-Stress-Model (RSM) and Transition-SST models were tested for a gas-lift loop reactor under Newtonian flow conditions validated against published experimental work. The results identify that flow predictions within the reactor (where flow is transitional) are particularly sensitive to the turbulence model implemented; the Transition-SST model was found to be the most robust for capturing mixing behaviour and predicting separation reliably. Therefore, Transition-SST is recommended over k-∊ models for use in comparable mixing problems. A comparison of results obtained using multiphase Euler-Lagrange and singlephase approaches are presented. The results support the validity of the singlephase modelling assumptions in obtaining reliable predictions of the reactor flow. Solver independence of results was verified by comparing two independent finite-volume solvers (Fluent-13.0sp2 and OpenFOAM-2.0.1). Copyright © 2013 Elsevier Ltd. All rights reserved.

Sodium flow measurement in large pipelines of sodium cooled fast breeder reactors with bypass type flow meters

International Nuclear Information System (INIS)

Rajan, K.K.; Jayakumar, T.; Aggarwal, P.K.; Vinod, V.

2016-01-01

Highlights: • Bypass type permanent magnet flow meters are more suitable for sodium flow measurement. • A higher sodium velocity through the PMFM sensor will increase its sensitivity and resolution. • By modifying the geometry of bypass line, higher sodium velocity through sensor is achieved. • With optimized geometry the sensitivity of bypass flow meter system was increased by 70%. - Abstract: Liquid sodium flow through the pipelines of sodium cooled fast breeder reactor circuits are measured using electromagnetic flow meters. Bypass type flow meter with a permanent magnet flow meter as sensor in the bypass line is selected for the flow measurement in the 800 NB main secondary pipe line of 500 MWe Prototype Fast Breeder Reactor (PFBR), which is at the advanced stage of construction at Kalpakkam. For increasing the sensitivity of bypass flow meters in future SFRs, alternative bypass geometry was considered. The performance enhancement of the proposed geometry was evaluated by experimental and numerical methods using scaled down models. From the studies it is observed that the new configuration increases the sensitivity of bypass flow meter system by around 70%. Using experimentally validated numerical tools the volumetric flow ratio for the bypass configurations is established for the operating range of Reynolds numbers.

Lava Fountaining Discharge Regime driven by Slug-to-Churn Flow Transition. (Invited)

Science.gov (United States)

Ripepe, M.; Pioli, L.; Marchetti, E.; Ulivieri, G.

2013-12-01

Lava fountaining episodes at Etna volcano appear characterized by the transition between Strombolian and Hawaiian end-member eruptive styles. There is no evidence for this transition in the seismic (i.e. seismic tremor) signal. However, infrasonic records provide unprecedented evidence on this flow transition. Each eruptive episode is characterized by distinctive common trend in the amplitude, waveform and frequency content of the infrasonic wavefield, which evidences the shift from discrete, and transient, strombolian to sustained, and oscillatory, lava fountain dynamics. Large scale experiments on the dynamics of two-phase flow of basaltic magmas show how the transition between different regimes mainly depends on gas volume flow, which in turn controls pressure distribution within the conduit and also magma vesicularity. In particular, while regular large bubble bursting is associated with slug flow regime, large amplitude and low frequency column oscillations are associated with churn flow. In large pipes, transition from slug to churn flow regime is independent on conduit diameter and it is reached at high superficial gas velocity. Lava fountaining episodes at Etna can be thus interpreted as induced by the transition from the slug (discrete strombolian) to churn flow (sustained lava fountain) regimes that is reflecting an increase in the gas discharge rate. Based on laboratory experiments, we calculate that transition between these two end-member explosive regimes at Etna occurs when gas superficial velocity is 76 m/s for near-the-vent stagnant magma conditions.

Two-phase flow structure in large diameter pipes

International Nuclear Information System (INIS)

Smith, T.R.; Schlegel, J.P.; Hibiki, T.; Ishii, M.

2012-01-01

Highlights: ► Local profiles of various quantities measured in large diameter pipe. ► Database for interfacial area in large pipes extended to churn-turbulent flow. ► Flow regime map confirms previous models for flow regime transitions. ► Data will be useful in developing interfacial area transport models for large pipes. - Abstract: Flow in large pipes is important in a wide variety of applications. In the nuclear industry in particular, understanding of flow in large diameter pipes is essential in predicting the behavior of reactor systems. This is especially true of natural circulation Boiling Water Reactor (BWR) designs, where a large-diameter chimney above the core provides the gravity head to drive circulation of the coolant through the reactor. The behavior of such reactors during transients and during normal operation will be predicted using advanced thermal–hydraulics analysis codes utilizing the two-fluid model. Essential to accurate two-fluid model calculations is reliable and accurate computation of the interfacial transfer terms. These interfacial transfer terms can be expressed as the product of one term describing the potential driving the transfer and a second term describing the available surface area for transfer, or interfacial area concentration. Currently, the interfacial area is predicted using flow regime dependent empirical correlations; however the interfacial area concentration is best computed through the use of the one-dimensional interfacial area transport equation (IATE). To facilitate the development of IATE source and sink term models in large-diameter pipes a fundamental understanding of the structure of the two-phase flow is essential. This understanding is improved through measurement of the local void fraction, interfacial area concentration and gas velocity profiles in pipes with diameters of 0.102 m and 0.152 m under a wide variety of flow conditions. Additionally, flow regime identification has been performed to

Analysis on flow characteristic of nuclear heating reactor

International Nuclear Information System (INIS)

Jiang Shengyao; Wu Xinxin

1997-06-01

The experiment was carried out on the test loop HRTL-5, which simulates the geometry and system design of a 5 MW Nuclear heating reactor. The analysis was based on a one-dimensional two-phase flow drift model with conservation equations for mass, steam mass, energy and momentum. Clausius-Clapeyron equation was used for the calculation of flashing front in the riser. A set of ordinary equation, which describes the behavior of two-phase flow in the natural circulation system, was derived through integration of the above conservation equations in subcooled boiling region, bulk boiling region in the heated section and in the riser. The method of time-domain was used for the calculation. Both static and dynamic results are presented. System pressure, inlet subcooling and heat flux are varied as input parameters. The results show that, firstly, subcooled boiling in the heated section and void flashing in the riser have significant influence on the distribution of the void fraction, mass flow rate and stability of the system, especially at lower pressure, secondly, in a wide range of two-phase flow conditions, only subcooled boiling occurs in the heated section. For the designed two-phase regime operation of the 5 MW nuclear heating reactor, the temperature at the core exit has not reaches its saturation value. Thirdly, the mechanism of two-phase flow oscillation, namely, 'zero-pressure-drop', is described. In the wide range of inlet subcooling (0 K<ΔT<28 K) there exists three regions for system flow condition, namely, (1) stable two-phase flow, (2) bulk and subcooled boiling unstable flow, (3) subcooled boiling and single phase stable flow. The response of mass flow rate, after a small disturbance in the heat flux, is showed in the above inlet subcooling range, and based on it the instability map of the system is given through experiment and calculation. (3 refs., 9 figs.)

Transition of unsteady velocity profiles with reverse flow

Science.gov (United States)

Das, Debopam; Arakeri, Jaywant H.

1998-11-01

This paper deals with the stability and transition to turbulence of wall-bounded unsteady velocity profiles with reverse flow. Such flows occur, for example, during unsteady boundary layer separation and in oscillating pipe flow. The main focus is on results from experiments in time-developing flow in a long pipe, which is decelerated rapidly. The flow is generated by the controlled motion of a piston. We obtain analytical solutions for laminar flow in the pipe and in a two-dimensional channel for arbitrary piston motions. By changing the piston speed and the length of piston travel we cover a range of values of Reynolds number and boundary layer thickness. The velocity profiles during the decay of the flow are unsteady with reverse flow near the wall, and are highly unstable due to their inflectional nature. In the pipe, we observe from flow visualization that the flow becomes unstable with the formation of what appears to be a helical vortex. The wavelength of the instability [simeq R: similar, equals]3[delta] where [delta] is the average boundary layer thickness, the average being taken over the time the flow is unstable. The time of formation of the vortices scales with the average convective time scale and is [simeq R: similar, equals]39/([Delta]u/[delta]), where [Delta]u=(umax[minus sign]umin) and umax, umin and [delta] are the maximum velocity, minimum velocity and boundary layer thickness respectively at each instant of time. The time to transition to turbulence is [simeq R: similar, equals]33/([Delta]u/[delta]). Quasi-steady linear stability analysis of the velocity profiles brings out two important results. First that the stability characteristics of velocity profiles with reverse flow near the wall collapse when scaled with the above variables. Second that the wavenumber corresponding to maximum growth does not change much during the instability even though the velocity profile does change substantially. Using the results from the experiments and the

Experimental Investigation of the Hot Water Layer Effect on Upward Flow Open Pool Reactor Operability

International Nuclear Information System (INIS)

Abou Elmaaty, T.

2014-01-01

The open pool reactor offers a high degree of reliability in the handling and manoeuvring, the replacement of reactor internal components and the suing of vertical irradiation channels. The protection of both the operators and the reactor hall environment against radiation hazards is considered a matter of interest. So, a hot water layer is implemented above many of the research reactors main pool, especially those whose flow direction is upward flow. An experimental work was carried out to ensure the operability of the upward flow open pool research reactor with / without the hot water layer. The performed experiment showed that, the hot water layer is produced an inverse buoyant force make the water to diffuse downward against the ordinary natural circulation from the reactor core. An upward flow - open pool research reactor (with a power greater than 20 M watt) could not wok without a hot water layer. The high temperature of the hot water layer surface could release a considerable amount of water vapour into the reactor hall, so a heat and mass transfer model is built based on the measured hot water layer surface temperature to calculate the amount of released water vapour during the reactor operating period. The effects of many parameters like the ambient air temperature, the reactor hall relative humidity and the speed of the pushed air layer above the top pool end on the evaporation rate is studied. The current study showed that, the hot water layer system is considered an efficient shielding system against Gamma radiation for open pool upward flow reactor and that system should be operated before the reactor start up by a suitable period of time. While, the heat and mass transfer model results showed that, the amount of the released water vapour is increased as a result of both the increase in hot water layer surface temperature and the increase in air layer speed. As the increase in hot water layer surface temperature could produce a good operability

Experimental Investigation of the Hot Water Layer Effect on Upward Flow Open Pool Reactor Operability

International Nuclear Information System (INIS)

Abou Elmaaty, T.

2015-01-01

The open pool reactor offers a high degree of reliability in the handling and manoeuvring, the replacement of reactor internal components and the swing of vertical irradiation channels. The protection of both the operators and the reactor hall environment against radiation hazards is considered a matter of interest. So, a hot water layer implemented above many of the research reactors main pool, especially those whose flow direction is upward flow. An experimental work was carried out to ensure the operability of the upward flow open pool research reactor with / without the hot water layer. The performed experiment showed that, the hot water layer produced an inverse buoyant force making the water to diffuse downward against the ordinary natural circulation from the reactor core. An upward flow-open pool research reactor (with a power greater than 20 Mw) could not wok without a hot water layer. The high temperature of the hot water layer surface could release a considerable amount of water vapour into the reactor hall, so a heat and mass transfer model is built based on the measured hot water layer surface temperature to calculate the amount of released water vapour during the reactor operating period. The effects of many parameters like the ambient air temperature, the reactor hall relative humidity and the speed of the pushed air layer above the top pool end on the evaporation rate is studied. The current study showed that, the hot water layer system is considered an efficient shielding system against gamma radiation for open pool upward flow reactor and that system should be operated before the reactor start up by a suitable period of time. While, the heat and mass transfer model results showed that, the amount of the released water vapour is increased as a result of both the increase in hot water layer surface temperature and the increase in air layer speed. As the increase in hot water layer surface temperature could produce a good operability conditions from

Device for controlling a recirculation flow in a reactor

International Nuclear Information System (INIS)

Shida, Toichi; Tohei, Kazushige; Hirose, Masao; Nakamura, Hideo.

1976-01-01

Object: To provide an emergency cut-off valve in a recirculation system in a reactor to control the recirculation at the time of turbine trip or load cut-off, thereby relieving excessive increase in heat output of fuel. Structure: A recirculation pump is driven through a recirculation pump motor by an AC generator, which is driven by a driving motor through a fluid coupling, so that reactor water passes the emergency cut-off valve and recirculation flow stop valve and then passes a jet pump into the core. At the time of turbine trip or load cut-off, the emergency cut-off valve is closed by a hydraulic circuit, whereby core flow is merely decreased by 20 to 30% in a short period of time to restrain excessive increase in heat output. (Yoshino, Y.)

Computer modeling of flow induced in-reactor vibrations

International Nuclear Information System (INIS)

Turula, P.; Mulcahy, T.M.

1977-01-01

An assessment of the reliability of finite element method computer models, as applied to the computation of flow induced vibration response of components used in nuclear reactors, is presented. The prototype under consideration was the Fast Flux Test Facility reactor being constructed for US-ERDA. Data were available from an extensive test program which used a scale model simulating the hydraulic and structural characteristics of the prototype components, subjected to scaled prototypic flow conditions as well as to laboratory shaker excitations. Corresponding analytical solutions of the component vibration problems were obtained using the NASTRAN computer code. Modal analyses and response analyses were performed. The effect of the surrounding fluid was accounted for. Several possible forcing function definitions were considered. Results indicate that modal computations agree well with experimental data. Response amplitude comparisons are good only under conditions favorable to a clear definition of the structural and hydraulic properties affecting the component motion. 20 refs

Transition to a pair of chaotic symmetric flows

International Nuclear Information System (INIS)

Chen Zhimin; Price, W.G.

2006-01-01

The complexity of transition to chaotic flow is discussed. It is shown that many different bifurcation processes may coexist and join together to excite the chaotic flow. The profile of this nonlinear dynamical behaviour is developed on the basis of a four-mode truncation model

Simulation of the flow obstruction of a jet pump in a BWR reactor with the code RELAP/SCDAPSIM

International Nuclear Information System (INIS)

Cardenas V, J.; Filio L, C.

2016-09-01

This work simulates the flow obstruction of a jet pump in one of the recirculation loops of a nuclear power plant with a reactor of type BWR at 100% of operating power, in order to analyze the behavior of the total flow of the refrigerant passing through the reactor core, the total flow in each recirculation loop of the reactor, together with the 10 jet pumps of each loop. The behavior of the power and the reactivity insertion due to the change of the refrigerant flow pattern is also analyzed. The simulation was carried out using the RELAP/SCDAPSIM version 3.5 code, using a reactor model with 10 jet pumps in each recirculation loop and a core consisting of 6 radial zones and 25 axial zones. The scenario postulates the flow obstruction in a jet pump in a recirculation loop A when the reactor operates at 100% rated power, causing a change in the total flow of refrigerant in the reactor core, leading to a decrease in power. Once the reactor conditions are established to its new power, the operator tries to recover the nominal power using the flow control valve of the recirculation loop A, opening stepwise as a strategy to safely recover the reactor power. In this analysis is assumed that the intention of the nuclear plant operator is to maintain the operation of the reactor during the established cycle. (Author)

Study on mixed convective flow penetration into subassembly from reactor hot plenum in FBRs

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, J.; Ohshima, H.; Kamide, H.; Ieda, Y. [Power Reactor and Nuclear Fuel Development Corporation, Ibaraki (Japan)

1995-09-01

Fundamental experiments using water were carried out in order to reveal the phenomenon of mixed convective flow penetration into subassemblies from a reactor`s upper plenum of fast breeder reactors. This phenomenon appears under a certain natural circulation conditions during the operation of the direct reactor auxiliary cooling system for decay heat removal and might influence the natural circulation head which determines the core flow rate and therefore affects the core coolability. In the experiment, a simplified model which simulates an upper plenum and a subassembly was used and the ultrasonic velocity profile monitor as well as thermocouples were applied for the simultaneous measurement of velocity and temperature distributions in the subassembly. From the measured data, empirical equations related to the penetration flow onset condition and the penetration depth were obtained using relevant parameters which were derived from dimensional analysis.

Simulation of corrosion product activity in pressurized water reactors under flow rate transients

International Nuclear Information System (INIS)

Mirza, Anwar M.; Mirza, Nasir M.; Mir, Imran

1998-01-01

Simulation of coolant activation due to corrosion products and impurities in a typical pressurized water reactor has been done under flow rate transients. Employing time dependent production and losses of corrosion products in the primary coolant path an approach has been developed to calculate the coolant specific activity. Results for 24 Na, 56 Mn, 59 Fe, 60 Co and 99Mo show that the specific activity in primary loop approaches equilibrium value under normal operating conditions fairly rapidly. Predominant corrosion product activity is due to Mn-56. Parametric studies at full power for various ramp decreases in flow rate show initial decline in the activity and then a gradual rise to relatively higher saturation values. The minimum value and the time taken to reach the minima are strong functions of the slope of linear decrease in flow rate. In the second part flow rate coastdown was allowed to occur at different flow half-times. The reactor scram was initiated at 90% of the normal flow rate. The results show that the specific activity decreases and the rate of decrease depends on pump half time and the reactor scram conditions

Flow-induced vibration phenomenon in a Mark III TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, C K; Whittemore, W L; Kim, B S; Lee, J B; Blevins, R D; Burton, T E [Korea Atomic Energy Research Institute, Seoul (Korea, Republic of); General Atomic Company, San Diego, CA (United States)

1976-07-01

The Mark III TRIGA reactor with hexagonal fuel spacing is capable of operating at 2.0 MW. The Mark III at San Diego operated without core cooling problems or vibration at power levels up to 2.0 MW. All Mark III reactors have operated trouble-free up to 1.0 MW. The Mark III TRIGA in Korea was installed in 1972 and operated many months without trouble at 2.0 MW. During this period core changes including addition of new fuel were made. Eighteen months after startup, a coolant flow-induced vibration was observed for the first time at a power of 1.5 MW. A lengthy series of tests showed that it was not possible to establish a core configuration that permitted vibration-free operation for power levels in the range 1.5 - 2.0 MW. Observations during the tests confirmed that standing waves in the reactor tank water coupled the source within the core to the shield structure and surrounding building. Analysis of the data indicates strongly that the source of the vibration is the creation and collapse of bubbles with the core acting as a resonator. A substantially increased flow of coolant through the upper grid plate is expected to eliminate the vibration phenomenon and permit trouble-free operation at power up to 2.0 MW. In an attempt to seek a remedy, both GAC and KAERI have independently developed designs for upper grid plates. KAERI has constructed and installed an interim version of the standard grid plate which was calculated to provide 25% more coolant flow and mounted high so as to provide less restriction to flow around the upper fittings of the fuel elements. A substantial reduction in vibration was observed. No vibration was observed at any power up to 2.0 MW with cooling water at or below 20 C. A slight vibration at 1.8 MW occurred for higher cooling temperatures. The GAC grid plate design provides not only for increasing the flow area but also for streamlining the flow surfaces on the grid plate and possibly also on the top fittings of the fuel elements. It is

Flow-induced vibration phenomenon in a Mark III TRIGA reactor

International Nuclear Information System (INIS)

Lee, C.K.; Whittemore, W.L.; Kim, B.S.; Lee, J.B.; Blevins, R.D.; Burton, T.E.

1976-01-01

The Mark III TRIGA reactor with hexagonal fuel spacing is capable of operating at 2.0 MW. The Mark III at San Diego operated without core cooling problems or vibration at power levels up to 2.0 MW. All Mark III reactors have operated trouble-free up to 1.0 MW. The Mark III TRIGA in Korea was installed in 1972 and operated many months without trouble at 2.0 MW. During this period core changes including addition of new fuel were made. Eighteen months after startup, a coolant flow-induced vibration was observed for the first time at a power of 1.5 MW. A lengthy series of tests showed that it was not possible to establish a core configuration that permitted vibration-free operation for power levels in the range 1.5 - 2.0 MW. Observations during the tests confirmed that standing waves in the reactor tank water coupled the source within the core to the shield structure and surrounding building. Analysis of the data indicates strongly that the source of the vibration is the creation and collapse of bubbles with the core acting as a resonator. A substantially increased flow of coolant through the upper grid plate is expected to eliminate the vibration phenomenon and permit trouble-free operation at power up to 2.0 MW. In an attempt to seek a remedy, both GAC and KAERI have independently developed designs for upper grid plates. KAERI has constructed and installed an interim version of the standard grid plate which was calculated to provide 25% more coolant flow and mounted high so as to provide less restriction to flow around the upper fittings of the fuel elements. A substantial reduction in vibration was observed. No vibration was observed at any power up to 2.0 MW with cooling water at or below 20 C. A slight vibration at 1.8 MW occurred for higher cooling temperatures. The GAC grid plate design provides not only for increasing the flow area but also for streamlining the flow surfaces on the grid plate and possibly also on the top fittings of the fuel elements. It is

Correlation-based Transition Modeling for External Aerodynamic Flows

Science.gov (United States)

Medida, Shivaji

Conventional turbulence models calibrated for fully turbulent boundary layers often over-predict drag and heat transfer on aerodynamic surfaces with partially laminar boundary layers. A robust correlation-based model is developed for use in Reynolds-Averaged Navier-Stokes simulations to predict laminar-to-turbulent transition onset of boundary layers on external aerodynamic surfaces. The new model is derived from an existing transition model for the two-equation k-omega Shear Stress Transport (SST) turbulence model, and is coupled with the one-equation Spalart-Allmaras (SA) turbulence model. The transition model solves two transport equations for intermittency and transition momentum thickness Reynolds number. Experimental correlations and local mean flow quantities are used in the model to account for effects of freestream turbulence level and pressure gradients on transition onset location. Transition onset is triggered by activating intermittency production using a vorticity Reynolds number criterion. In the new model, production and destruction terms of the intermittency equation are modified to improve consistency in the fully turbulent boundary layer post-transition onset, as well as ensure insensitivity to freestream eddy viscosity value specified in the SA model. In the original model, intermittency was used to control production and destruction of turbulent kinetic energy. Whereas, in the new model, only the production of eddy viscosity in SA model is controlled, and the destruction term is not altered. Unlike the original model, the new model does not use an additional correction to intermittency for separation-induced transition. Accuracy of drag predictions are improved significantly with the use of the transition model for several two-dimensional single- and multi-element airfoil cases over a wide range of Reynolds numbers. The new model is able to predict the formation of stable and long laminar separation bubbles on low-Reynolds number airfoils that

Flow reversal power limit for the HFBR

International Nuclear Information System (INIS)

Cheng, L.Y.; Tichler, P.R.

1997-01-01

The High Flux Beam Reactor (HFBR) is a pressurized heavy water moderated and cooled research reactor that began operation at 40 MW. The reactor was subsequently upgraded to 60 MW and operated at that level for several years. The reactor undergoes a buoyancy-driven reversal of flow in the reactor core following certain postulated accidents. Questions which were raised about the afterheat removal capability during the flow reversal transition led to a reactor shutdown and subsequent resumption of operation at a reduced power of 30 MW. An experimental and analytical program to address these questions is described in this report. The experiments were single channel flow reversal tests under a range of conditions. The analytical phase involved simulations of the tests to benchmark the physical models and development of a criterion for dryout. The criterion is then used in simulations of reactor accidents to determine a safe operating power level. It is concluded that the limit on the HFBR operating power with respect to the issue of flow reversal is in excess of 60 MW. Direct use of the experimental results and an understanding of the governing phenomenology supports this conclusion

Flow induced vibrations in liquid metal fast breeder reactors

International Nuclear Information System (INIS)

1989-01-01

Flow induced vibrations are well known phenomena in industry. Engineers have to estimate their destructive effects on structures. In the nuclear industry, flow induced vibrations are assessed early in the design process, and the results are incorporated in the design procedures. In many cases, model testing is used to supplement the design process to ensure that detrimental behaviour due to flow induced vibrations will not occur in the component in question. While these procedures attempt to minimize the probability of adverse performance of the various components, there is a problem in the extrapolation of analytical design techniques and/or model testing to actual plant operation. Therefore, sodium tests or vibrational measurements of components in the reactor system are used to provide additional assurance. This report is a general survey of experimental and calculational methods in this area of structural mechanics. The report is addressed to specialists and institutions in industrialized and developing countries who are responsible for the design and operation of liquid metal fast breeder reactors. 92 refs, 90 figs, 8 tabs

PDBD with continuous liquids flows in a discharge reactor

International Nuclear Information System (INIS)

Rodríguez-Méndez, B G; Gutiérrez-León, D G; López-Callejas, R; Valencia-Alvarado, R; Muñoz-Castro, A E; Mercado-Cabrera, A; Peña-Eguiluz, R; Belman-Flores, J M; De la Piedad-Beneitez, A

2015-01-01

This paper presents the design, construction and testing of a cylindrical pulsed dielectric barrier discharge (PDBD) reactor aimed to microbiological elimination of Escherichia coli ATCC 8739 bacteria. In the reactor, water flowed continuously and to countercurrent an oxygen gas was injected. The water pumping was carried out with a peristaltic pump type, stainless steel and aluminum constructed, and water was recirculated through norprene tubing. The considered parameters in order to promote energetic efficiency were: the residence time of the water contaminated with bacteria, flow rate of the liquid, shape and material used to build electrodes and dielectric, pressure, and gas injection flow rate. The pulsed power supply parameters are featured by 25-30 kV high voltage, 500 Hz frequency and 30 μs width. The outcome elimination of E. coli bacteria at 10 3 , 10 4 and 10 6 CFU/mL concentrations reached an efficiency over 0.5 log-order in absence of oxygen; while >2 log-orders when oxygen gas was injected during the process. (paper)

Thermalhydraulics of flowing particle-bed-type fusion reactor blankets

International Nuclear Information System (INIS)

Nietert, R.E.; Abdelk-Khalik, S.I.

1982-01-01

An experimental investigation has been conducted to determine the heat transfer characteristics of gravity-flowing particle beds using a special heat transfer loop. Glass microspheres were allowed to flow by gravity at controlled rates through an electrically heated stainless steel tubular test section. Values of the local and average convective heat transfer coefficient as a function of the average bed velocity, particle size and heat flux were determined. Such information is necessary for the design of gravity-flowing particle-bed type fusion reactor-blankets and associated tritium recovery systems. (orig.)

Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

Directory of Open Access Journals (Sweden)

Po Hu

2014-01-01

Full Text Available This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal.

Predicting transition in two- and three-dimensional separated flows

International Nuclear Information System (INIS)

Cutrone, L.; De Palma, P.; Pascazio, G.; Napolitano, M.

2008-01-01

This paper is concerned with the numerical prediction of two- and three-dimensional transitional separated flows of turbomachinery interest. The recently proposed single-point transition model based on the use of a laminar kinetic energy transport equation is considered, insofar as it does not require to evaluate any integral parameter, such as boundary-layer thickness, and is thus directly applicable to three-dimensional flows. A well established model, combining a transition-onset correlation with an intermittency transport equation, is also used for comparison. Both models are implemented within a Reynolds-averaged Navier-Stokes solver employing a low-Reynolds-number k-ω turbulence model. The performance of the transition models have been evaluated and tested versus well-documented incompressible flows past a flat plate with semi-circular leading edge, namely: tests T3L2, T3L3, T3L5, and T3LA1 of ERCOFTAC, with different Reynolds numbers and free-stream conditions, the last one being characterized by a non-zero pressure gradient. In all computations, the first model has proven as adequate as or superior to the second one and has been then applied with success to two more complex test cases, for which detailed experimental data are available in the literature, namely: the two- and three-dimensional flows through the T106 linear turbine cascade

Hydrothermal Processing of Macroalgal Feedstocks in Continuous-Flow Reactors

Energy Technology Data Exchange (ETDEWEB)

Elliott, Douglas C.; Hart, Todd R.; Neuenschwander, Gary G.; Rotness, Leslie J.; Roesijadi, Guri; Zacher, Alan H.; Magnuson, Jon K.

2014-02-03

Wet macroalgal slurries have been converted into a biocrude by hydrothermal liquefaction (HTL) in a bench-scale continuous-flow reactor system. Carbon conversion to a gravity-separable oil product of 58.8% was accomplished at relatively low temperature (350 °C) in a pressurized (subcritical liquid water) environment (20 MPa) when using feedstock slurries with a 21.7% concentration of dry solids. As opposed to earlier work in batch reactors reported by others, direct oil recovery was achieved without the use of a solvent, and biomass trace mineral components were removed by processing steps so that they did not cause processing difficulties. In addition, catalytic hydrothermal gasification (CHG) was effectively applied for HTL byproduct water cleanup and fuel gas production from water-soluble organics. Conversion of 99.2% of the carbon left in the aqueous phase was demonstrated. Finally, as a result, high conversion of macroalgae to liquid and gas fuel products was found with low levels of residual organic contamination in byproduct water. Both process steps were accomplished in continuous-flow reactor systems such that design data for process scale-up was generated.

Prediction, analysis and solution of flow inversion phenomenon in a typical MTR reactor with upward core cooling

International Nuclear Information System (INIS)

El-Morshedy, Salah El-Din

2010-01-01

Research reactors of power greater than 20 MW are usually designed to be cooled with upward coolant flow direction inside the reactor core. This is mainly to prevent flow inversion problems following a pump coast down. However, in some designs and under certain operating conditions, flow inversion phenomenon is predicted. In the present work, the best-estimate Material Testing Reactors Thermal-Hydraulic Analysis program (MTRTHA) is used to simulate a typical MTR reactor behavior with upward cooling under a hypothetical case of loss of off-site power. The flow inversion phenomenon is predicted under certain decay heat and/or pool temperature values below the design values. The reactor simulation under loss of off-site power is performed for two cases namely; two-flap valves open and one flap-valve fails to open. The model results for the flow inversion phenomenon prediction is analyzed and a solution of the problem is suggested. (orig.)

Automatic coolant flow control device for a nuclear reactor assembly

Science.gov (United States)

Hutter, Ernest

1986-01-01

A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

Passenger flow analysis of Beijing urban rail transit network using fractal approach

Science.gov (United States)

Li, Xiaohong; Chen, Peiwen; Chen, Feng; Wang, Zijia

2018-04-01

To quantify the spatiotemporal distribution of passenger flow and the characteristics of an urban rail transit network, we introduce four radius fractal dimensions and two branch fractal dimensions by combining a fractal approach with passenger flow assignment model. These fractal dimensions can numerically describe the complexity of passenger flow in the urban rail transit network and its change characteristics. Based on it, we establish a fractal quantification method to measure the fractal characteristics of passenger follow in the rail transit network. Finally, we validate the reasonability of our proposed method by using the actual data of Beijing subway network. It has been shown that our proposed method can effectively measure the scale-free range of the urban rail transit network, network development and the fractal characteristics of time-varying passenger flow, which further provides a reference for network planning and analysis of passenger flow.

Transition to turbulence for flows without linear criticality

International Nuclear Information System (INIS)

Nagata, Masato

2010-01-01

It is well known that plane Couette flow (PCF) and pipe flow (PF) are linearly stable against arbitrary three-dimensional perturbations at any finite Reynolds number, so that transitions from the basic laminar states, if they exist, must be abrupt. Due to this lack of linear criticality, weakly nonlinear analysis does not work in general and numerical approaches must be resorted to. It is only recently that non-trivial nonlinear states for these flows have been discovered numerically at finite Reynolds number as solutions bifurcating from infinity. The onset of turbulence in a subcritical transition is believed to be related to the appearance of steady/travelling wave states caused by disturbances of finite amplitude that take the flows out of the basin of attraction of the laminar state in phase space. In this paper, we introduce other flows that, in a similar way to PCF and PF, exhibit no linear critical point for the laminar states, namely flow in a square duct and sliding Couette flow in an annulus with a certain range of gap ratio. We shall show our recent numerical investigations on these flows where nonlinear travelling wave states are found for the first time by a homotopy approach. We believe that these states constitute the skeleton around which a time-dependent trajectory in the phase space is organized and help in understanding non-equilibrium turbulent processes.

Reactor vessel and core two-phase flow ultrasonic densitometer

International Nuclear Information System (INIS)

Arave, A.E.

1979-01-01

A local ultrasonic density (LUD) detector has been developed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory (INEL) for the Loss-of-Fluid Test (LOFT) reactor vessel and core two-phase flow density measurements. The principle of operating the sensor is the change in propagation time of a torsional ultrasonic wave in a metal transmission line as a function of the density of the surrounding media. A theoretical physics model is presented which represents the total propagation time as a function of the sensor modulus of elasticity and polar moment of inertia. Separate effects tests and two-phase flow tests have been conducted to characterize the detector. Tests show the detector can perform in a 343 0 C pressurized water reactor environment and measure the average density of the media surrounding the sensor

Change of neutron flow sensors effectiveness in the course of reactor experiments

International Nuclear Information System (INIS)

Kurpesheva, A.M.; Kotov, V.M.; Zhotabaev, Zh.R.

2007-01-01

Full text: IGR reactor is a reactor of thermal capacity type. During the operation, uranium-graphite core can be heated up to 1500 deg. C and reactivity can be changed considerably. Core dimensions are comparatively small. Amount of control rods, providing required reactivity, is not big as well. Increasing of core temperature leads to the rise of neutrons path length in its basic material - graphite. Change of temperature is not even. All this causes the non-conservation of neutron flows ratio in irradiated sample and in the place of reactor power sensors installation. Deviations in this ratio were registered during the number of reactor experiments. Empiric corrections can be introduced in order to decrease influence of change of neutron flow effectiveness upon provision of required parameters of investigated matters load. However, dependence of these corrections upon many factors can lead to the increasing of instability of process control. Previous experiment-calculated experiments showed inequality of neutron field in the place of sensors location (up to tens of percent), low effectiveness of experimental works, carried out without access to the individual reactor laying elements. Imperfection during the experiment was an idea of possibility to connect distribution of out of reactor neutron flow and control rods position. Subsequent analysis showed that for the development of representative phenomenon model it is necessary to take into account reactor operation dynamic subject to unevenness of heating of individual laying parts. Elemental calculations showed that temperature laying effects in the change of neutron outer field are great. Algorithm of calculations for the change of outer filed and field of investigated fabrication includes calculation of neutron-physic reactor characteristics interlacing with calculations of thermal-physic reactor characteristics, providing correlation of temperature fields for neutron-physic calculations. In the course of such

Regional groundwater flow model for C, K. L. and P reactor areas, Savannah River Site, Aiken, SC

Energy Technology Data Exchange (ETDEWEB)

Flach, G.P.

2000-02-11

A regional groundwater flow model encompassing approximately 100 mi2 surrounding the C, K, L, and P reactor areas has been developed. The reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department. The model provides a quantitative understanding of groundwater flow on a regional scale within the near surface aquifers and deeper semi-confined to confined aquifers. The model incorporates historical and current field characterization data up through Spring 1999. Model preprocessing is automated so that future updates and modifications can be performed quickly and efficiently. The CKLP regional reactor model can be used to guide characterization, perform scoping analyses of contaminant transport, and serve as a common base for subsequent finer-scale transport and remedial/feasibility models for each reactor area.

Regional groundwater flow model for C, K. L. and P reactor areas, Savannah River Site, Aiken, SC

International Nuclear Information System (INIS)

Flach, G.P.

2000-01-01

A regional groundwater flow model encompassing approximately 100 mi2 surrounding the C, K, L, and P reactor areas has been developed. The reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department. The model provides a quantitative understanding of groundwater flow on a regional scale within the near surface aquifers and deeper semi-confined to confined aquifers. The model incorporates historical and current field characterization data up through Spring 1999. Model preprocessing is automated so that future updates and modifications can be performed quickly and efficiently. The CKLP regional reactor model can be used to guide characterization, perform scoping analyses of contaminant transport, and serve as a common base for subsequent finer-scale transport and remedial/feasibility models for each reactor area

Falling liquid film flow along cascade-typed first wall of laser-fusion reactor

International Nuclear Information System (INIS)

Kunugi, T.; Nakai, T.; Kawara, Z.

2007-01-01

To protect from high energy/particle fluxes caused by nuclear fusion reaction such as extremely high heat flux, X rays, Alpha particles and fuel debris to a first wall of an inertia fusion reactor, a 'cascade-typed' first wall with a falling liquid film flow is proposed as the 'liquid wall' concept which is one of the reactor chamber cooling and wall protection schemes: the reactor chamber can protect by using a liquid metal film flow (such as Li 17 Pb 83 ) over the wall. In order to investigate the feasibility of this concept, we conducted the numerical analyses by using the STREAM code and also conducted the flow visualization experiments. The numerical results suggested that the cascade structure design should be improved, so that we redesigned the cascade-typed first wall and performed the flow visualization as a POP (proof-of-principle) experiment. In the numerical analyses, the water is used as the working liquid and an acrylic plate as the wall. These selections are based on two reasons: (1) from the non-dimensional analysis approach, the Weber number (We=ρu 2 δ/σ: ρ is density, u is velocity, δ is film thickness, σ is surface tension coefficient) should be the same between the design (Li 17 Pb 83 flow) and the model experiment (water flow) because of the free-surface instability, (2) the SiC/SiC composite would be used as the wall material, so that the wall may have the less wettability: the acrylic plate has the similar feature. The redesigned cascade-typed first wall for one step (30 cm height corresponding to 4 Hz laser duration) consists of a liquid tank having a free-surface for keeping the constant water-head located at the backside of the first wall, and connects to a slit which is composed of two plates: one plate is the first wall, and the other is maintaining the liquid level. This design solved the trouble of the previous design. The test section for the flow visualization has the same structure and the same height as the reactor design

Dynamic Transitions and Baroclinic Instability for 3D Continuously Stratified Boussinesq Flows

Science.gov (United States)

Şengül, Taylan; Wang, Shouhong

2018-02-01

The main objective of this article is to study the nonlinear stability and dynamic transitions of the basic (zonal) shear flows for the three-dimensional continuously stratified rotating Boussinesq model. The model equations are fundamental equations in geophysical fluid dynamics, and dynamics associated with their basic zonal shear flows play a crucial role in understanding many important geophysical fluid dynamical processes, such as the meridional overturning oceanic circulation and the geophysical baroclinic instability. In this paper, first we derive a threshold for the energy stability of the basic shear flow, and obtain a criterion for local nonlinear stability in terms of the critical horizontal wavenumbers and the system parameters such as the Froude number, the Rossby number, the Prandtl number and the strength of the shear flow. Next, we demonstrate that the system always undergoes a dynamic transition from the basic shear flow to either a spatiotemporal oscillatory pattern or circle of steady states, as the shear strength of the basic flow crosses a critical threshold. Also, we show that the dynamic transition can be either continuous or catastrophic, and is dictated by the sign of a transition number, fully characterizing the nonlinear interactions of different modes. Both the critical shear strength and the transition number are functions of the system parameters. A systematic numerical method is carried out to explore transition in different flow parameter regimes. In particular, our numerical investigations show the existence of a hypersurface which separates the parameter space into regions where the basic shear flow is stable and unstable. Numerical investigations also yield that the selection of horizontal wave indices is determined only by the aspect ratio of the box. We find that the system admits only critical eigenmodes with roll patterns aligned with the x-axis. Furthermore, numerically we encountered continuous transitions to multiple

Transition of Gas-Liquid Stratified Flow in Oil Transport Pipes

Directory of Open Access Journals (Sweden)

D. Lakehal

2011-12-01

Full Text Available Large-Scale Simulation results of the transition of a gas-liquid stratified flow to slug flow regime in circular 3D oil transport pipes under turbulent flow conditions expressed. Free surface flow in the pipe is treated using the Level Set method. Turbulence is approached via the LES and VLES methodologies extended to interfacial two-phase flows. It is shown that only with the Level Set method the flow transition can be accurately predicted, better than with the two-fluid phase-average model. The transition from stratified to slug flow is found to be subsequent to the merging of the secondary wave modes created by the action of gas shear (short waves with the first wave mode (high amplitude long wave. The model is capable of predicting global flow features like the onset of slugging and slug speed. In the second test case, the model predicts different kinds of slugs, the so-called operating slugs formed upstream that fill entirely the pipe with water slugs of length scales of the order of 2-4 D, and lower size (1-1.5 D disturbance slugs, featuring lower hold-up (0.8-0.9. The model predicts well the frequency of slugs. The simulations revealed important parameter effects on the results, such as two-dimensionality, pipe length, and water holdup.

Transition from slug to annular flow in horizontal air-water flow

International Nuclear Information System (INIS)

Reismann, J.; John, H.; Seeger, W.

1981-11-01

The transition from slug to annular flow in horizontal air-water and steam-water flow was investigated. Test sections of 50; 66.6 and 80 mm ID were used. The system pressure was 0.2 and 0.5 MPa in the air-water experiments and 2.5; 5; 7.5 and 10 MPa in the steam-water experiments. For flow pattern detection local impedance probes were used. This method was compared in a part of the experiments with differential pressure and gamma-beam measurements. The flow regime boundary is shifting strongly to smaller values of the superficial gas velocity with increasing pressure. Correlations from literature fit unsatisfactorily the experimental results. A new correlation is presented. (orig.) [de

Identification of the two-phase flow in the upper part of a boiling water reactor core using reactor noise analysis

International Nuclear Information System (INIS)

Miteff, L.

1983-08-01

The starting point of this work were neutron flux correlation measurements in the core of the Muehleberg boiling water reactor. During these measurements a new effect was observed i.e. that the cross power spectral density (CPSD) phases could be approximated by two straight lines of different slopes. The two linear domains of the CPSD-phases are separated along the frequency axis by a transition interval. It was supposed that such CPSD-phases could account for the propagation in the axial direction of two different thermohydraulic perturbations in the upper part of the Muehleberg BWR-core. It was also assumed that the second linear domain of these CPSD-phases could be related to a characteristic property of an annular flow regime of the steam-water two-phase flows in the bundles around the neutron detectors. The author attempts to give an explanation for the existence of a second transport phenomenon. This was achieved by in-core and out-core correlation measurements as well as by theoretical work. The out-core measurements were performed on an water-air simulation loop by use of laser beams. (Auth.)

Hydrogeological and Groundwater Flow Model for C, K, L, and P Reactor Areas, Savannah River Site, Aiken, South Carolina

International Nuclear Information System (INIS)

Flach, G.P.

1999-01-01

A regional groundwater flow model encompassing approximately 100 mi 2 surrounding the C, K. L. and P reactor areas has been developed. The Reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department

Transient performance of flow in PWR reactor circuits

International Nuclear Information System (INIS)

Hirdes, V.R.T.R.; Carajilescov, P.

1988-12-01

Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

Evolution of weighted complex bus transit networks with flow

Science.gov (United States)

Huang, Ailing; Xiong, Jie; Shen, Jinsheng; Guan, Wei

2016-02-01

Study on the intrinsic properties and evolutional mechanism of urban public transit networks (PTNs) has great significance for transit planning and control, particularly considering passengers’ dynamic behaviors. This paper presents an empirical analysis for exploring the complex properties of Beijing’s weighted bus transit network (BTN) based on passenger flow in L-space, and proposes a bi-level evolution model to simulate the development of transit routes from the view of complex network. The model is an iterative process that is driven by passengers’ travel demands and dual-controlled interest mechanism, which is composed of passengers’ spatio-temporal requirements and cost constraint of transit agencies. Also, the flow’s dynamic behaviors, including the evolutions of travel demand, sectional flow attracted by a new link and flow perturbation triggered in nearby routes, are taken into consideration in the evolutional process. We present the numerical experiment to validate the model, where the main parameters are estimated by using distribution functions that are deduced from real-world data. The results obtained have proven that our model can generate a BTN with complex properties, such as the scale-free behavior or small-world phenomenon, which shows an agreement with our empirical results. Our study’s results can be exploited to optimize the real BTN’s structure and improve the network’s robustness.

Fiscal year 1998 multi-year work plan. Advanced reactors transition program

International Nuclear Information System (INIS)

Gantt, D.A.

1997-01-01

The mission of the Advanced Reactors Transition program is two-fold. First, the program is to maintain the Fast Flux Test Facility (FFTF) and the Fuels and Materials Examination Facility (FMEF) in Standby to support a possible future role in the tritium production strategy. Secondly, the program is to continue deactivation activities which do not conflict with the Standby directive. On-going deactivation activities include the processing of non-usable, irradiated, FFTF components for storage or disposal; deactivation of Nuclear Energy legacy test facilities; and deactivation of the Plutonium Recycle Test Reactor (PRTR) facility, 309 Building

The concept of the sodium cooled small fast reactor 4S and the analyses of the loss of flow events

International Nuclear Information System (INIS)

Nishi, Yoshihisa; Ueda, Nobuyuki; Koga, Tomonari; Matsumiya, Hisato

2007-01-01

CRIEPI has been developing the 4S reactor (Super Safe, Small and Simple reactor) for application in dispersed energy supply and multipurpose use, in conjunction with Toshiba Corporation. The 4S is sodium cooled fast reactor and their electrical output has two options of 10MWe and 50MWe. In this paper, 10MWe 4S (4S-10M) was proposed. 4S-10M has some unique features. It employs a burn-up control system with annular reflector in place of the control rod that requires the frequent maintenance service. The core life time of the 4S-10M is 30 years and the fuel transport is not required during core life time. All temperature feedback coefficients are negative during core life time. In the latest design for 4S-10M, a pool and tall type reactor design was selected to reduce the construction cost. Two types of decay heat removal system (Reactor Vessel Auxiliary Cooling System; RVACS, Intermediate Reactor Auxiliary Cooling System; IRACS) using natural convection power were adopted. It is necessary to confirm that these two heat removal system can operate appropriately. The transition analyses were executed by the CERES code to evaluate the design feasibility and the thermal hydraulic characteristics of the 4S-10M. CERES is a multi-dimensional plant dynamics simulation code for liquid metal reactors developed by the CRIEPI. CERES can perform simulations ranging from forced circulation (full/partial power operation) to natural circulation. Components (pumps, IHXs, SGs, pipings, etc.) of the reactor are modeled as one-dimensional. Multi-dimensional plena are connected to such components. Two loss-of-flow accident sequences are considered. In the first case, it is assumed that the primary and the secondary pump were stopped by the total station black out. The reactor shut down system was assumed to be success. This sequence is referred to as the protected loss-of-flow accident (PLOF). In the second case, it is assumed that the reactor shut down systems fail to operate and the

Calculation of pressure drop and flow redistribution in the core of LMFBR type reactors

International Nuclear Information System (INIS)

Botelho, D.A.; Morgado, O.J.

1985-01-01

It is studied the flow redistribution through of fuel elements to the pressure drop calculation in the core of sodium cooled reactors. Using the quasi-static formulation of equations of the conservation of mass, energy and momentum, it was developed a computer program to flow redistribution calculations and pressure drop for different power levels and total flow simulating an arbitrary number of channels for sodium flowing . An optimization of the number of sufficient channels for calculations of this nature is done. The method is applied in studies of transients in the same reactor. (M.C.K.) [pt

Study on natural circulation flow under reactor cavity flooding condition in advanced PWRs

International Nuclear Information System (INIS)

Tao Jun; Yang Jiang; Cao Jianhua; Lu Xianghui; Guo Dingqing

2015-01-01

Cavity flooding is an important severe accident management measure for the in-vessel retention of a degraded core by external reactor vessel cooling in advanced PWRs. A code simulation study on the natural circulation flow in the gap between the reactor vessel wall and insulation material under cavity flooding condition is performed by using a detailed mechanistic thermal-hydraulic code package RELAP 5. By simulating of an experiment carried out for studying the natural circulation flow for APR1400 shows that the code is applicable for analyzing the circulation flow under this condition. The analysis results show that heat removal capacity of the natural circulation flow in AP1000 is sufficient to prevent thermal failure of the reactor vessel under bounding heat load. Several conclusions can be drawn from the sensitivity analysis. Larger coolant inlet area induced larger natural circulation flow rate. The outlet should be large enough and should not be submerged by the cavity water to vent the steam-water mixture. In the implementation of cavity flooding, the flooding water level should be high enough to provide sufficient natural circulation driven force. (authors)

Recycling flow rate control device in BWR type reactor

International Nuclear Information System (INIS)

Fujiwara, Tadashi; Koda, Yasushi

1988-01-01

Purpose: To reduce the recycling pump speed if the pressure variation width and the variation ratio in the nuclear reactor exceed predetermined values, to thereby avoid the shutdown of the plant. Constitution: There has been proposed a method of monitoring the neutron flux increase thereby avoiding unnecessary plant shutdown, but it involves a problems of reactor scram depending on the state of the plant and the set values. In view of the above, in the plant using internal pumps put under the thyristor control and having high response to recycling flow rate, the reactor pressure is monitored and the speed of the internal pump is rapidly reduced when the pressure variation width and variation ratio exceed predetermined values to reduce the reactor power and avoid the plant shutdown. This can reduce the possibility of unnecessary power reduction due to neutron flux noises or the possibility of plant shutdown under low power conditions. Further, since the reactor operation can be continued without stopping the recycling pump, the operation upon recovery can be made rapid. (Horiuchi, T.)

Hydrogeological and Groundwater Flow Model for C, K, L, and P Reactor Areas, Savannah River Site, Aiken, South Carolina

Energy Technology Data Exchange (ETDEWEB)

Flach, G.P.

1999-02-24

A regional groundwater flow model encompassing approximately 100 mi{sup 2} surrounding the C, K. L. and P reactor areas has been developed. The Reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department.

Accidents of loss of flow for the ETTR-2 reactor; deterministic analysis

International Nuclear Information System (INIS)

El-Messiry, A.M.

2000-01-01

The main objective for reactor safety is to keep the fuel in a thermally safe condition with adequate safety margins during all operational modes (normal-abnormal and accidental states). To achieve this purpose an accident analysis of different design base accident (DBA) as loss of flow accident (LOFA), is required assessing reactor safety. The present work concerns this transients applied to Egypt Test and Research Reactor ETRR-3 (new reactor). An accident analysis code FLOWTR is developed to investigate the thermal behaviour of the core during such flow transients. The active core is simulated by two channels: 1 - hot channel (HC), and 2 - average channel (AC) representing the remainder of the core. Each channel is divided into four axial sections. The external loop, core plenums, and core chimney are simulated by different dynamic loops. The code includes modules for pump cast down, flow regimes, decay heat, temperature distributions, and feedback coefficients. FLOWTR is verified against results from RETRAN code, THERMIC code and commissioning tests for null transient case. The comparison shows a good agreement. The study indicates that for LOFA transients, provided the scram system is available, the core is shutdown safely by low flow signal (496.6 kg/s) at 1.4 s were the HC temperature reaches the maximum value, 45.64 o C after shutdown. On the other hand, if the scram system is unavailable, and at t = 47.33 s, the core flow decreases to 67.41 kg/s, the HC temperature increases to 164.02 o C, and the HC clad surface heat flux exceeds its critical value of 400.00 W/cm 2 resulting of fuel burnout. (author)

Fluid flow and heat transfer investigation of pebble bed reactors using mesh-adaptive LES

International Nuclear Information System (INIS)

Pavlidis, Dimitrios; Lathouwers, Danny

2013-01-01

The very high temperature reactor is one of the designs currently being considered for nuclear power generation. One its variants is the pebble bed reactor in which the coolant passes through complex geometries (pores) at high Reynolds numbers. A computational fluid dynamics model with anisotropic mesh adaptivity is used to investigate coolant flow and heat transfer in such reactors. A novel method for implicitly incorporating solid boundaries based on multi-fluid flow modelling is adopted. The resulting model is able to resolve and simulate flow and heat transfer in randomly packed beds, regardless of the actual geometry, starting off with arbitrarily coarse meshes. The model is initially evaluated using an orderly stacked square channel of channel-height-to-particle diameter ratio of unity for a range of Reynolds numbers. The model is then applied to the face-centred cubical geometry. coolant flow and heat transfer patterns are investigated

Fluid Flow Characteristic Simulation of the Original TRIGA 2000 Reactor Design Using Computational Fluid Dynamics Code

International Nuclear Information System (INIS)

Fiantini, Rosalina; Umar, Efrizon

2010-01-01

Common energy crisis has modified the national energy policy which is in the beginning based on natural resources becoming based on technology, therefore the capability to understanding the basic and applied science is needed to supporting those policies. National energy policy which aims at new energy exploitation, such as nuclear energy is including many efforts to increase the safety reactor core condition and optimize the related aspects and the ability to build new research reactor with properly design. The previous analysis of the modification TRIGA 2000 Reactor design indicates that forced convection of the primary coolant system put on an effect to the flow characteristic in the reactor core, but relatively insignificant effect to the flow velocity in the reactor core. In this analysis, the lid of reactor core is closed. However the forced convection effect is still presented. This analysis shows the fluid flow velocity vector in the model area without exception. Result of this analysis indicates that in the original design of TRIGA 2000 reactor, there is still forced convection effects occur but less than in the modified TRIGA 2000 design.

Unsteady flow characteristic analysis of turbine based combined cycle (TBCC inlet mode transition

Directory of Open Access Journals (Sweden)

Jun Liu

2015-09-01

Full Text Available A turbine based combined cycle (TBCC propulsion system uses a turbine-based engine to accelerate the vehicle from takeoff to the mode transition flight condition, at which point, the propulsion system performs a “mode transition” from the turbine to ramjet engine. Smooth inlet mode transition is accomplished when flow is diverted from one flowpath to the other, without experiencing unstart or buzz. The smooth inlet mode transition is a complex unsteady process and it is one of the enabling technologies for combined cycle engine to become a functional reality. In order to unveil the unsteady process of inlet mode transition, the research of over/under TBCC inlet mode transition was conducted through a numerical simulation. It shows that during the mode transition the terminal shock oscillates in the inlet. During the process of inlet mode transition mass flow rate and Mach number of turbojet flowpath reduce with oscillation. While in ramjet flowpath the flow field is non-uniform at the beginning of inlet mode transition. The speed of mode transition and the operation states of the turbojet and ramjet engines will affect the motion of terminal shock. The result obtained in present paper can help us realize the unsteady flow characteristic during the mode transition and provide some suggestions for TBCC inlet mode transition based on the smooth transition of thrust.

Transition from annular flow to plug/slug flow in condensation of steam in microchannels

Energy Technology Data Exchange (ETDEWEB)

Quan, Xiaojun; Cheng, Ping; Wu, Huiying [School of Mechanical and Power Engineering, Shanghai Jiaotong University, 800 Dong Chuan Road, Shanghai 200240 (China)

2008-02-15

A visualization study has been conducted to investigate the transition from annular flow to plug/slug flow in the condensation of steam in two different sets of parallel microchannels, having hydraulic diameters of 90 {mu}m and 136 {mu}m, respectively. The steam in the parallel microchannels was cooled on the bottom by forced convection of water and by natural convection of air from the top. It is found that the location, where the transition from annular flow to plug/slug flow takes place, depends on mass flux and cooling rate of steam. The effects of mass flux and cooling rate on the occurrence frequency of the injection flow in a single microchannel, having a hydraulic diameter of 120 {mu}m and 128 {mu}m, respectively, are investigated. It is found that two different shapes of injection flow occur in the smooth annular flow in microchannels: injection flow with unsteady vapor ligament occurring at low mass flux (or high cooling rate) and injection flow with steady vapor ligament occurring at high mass flux (or low cooling rate). It is also found that increase of steam mass flux, decrease of cooling rate, or decrease of the microchannel diameter tends to enhance instability of the condensate film on the wall, resulting in occurrence of the injection flow further toward the outlet with an increase in occurrence frequency. (author)

Advances in transitional flow modeling applications to helicopter rotors

CERN Document Server

Sheng, Chunhua

2017-01-01

This book provides a comprehensive description of numerical methods and validation processes for predicting transitional flows based on the Langtry–Menter local correlation-based transition model, integrated with both one-equation Spalart–Allmaras (S–A) and two-equation Shear Stress Transport (SST) turbulence models. A comparative study is presented to combine the respective merits of the two coupling methods in the context of predicting the boundary-layer transition phenomenon from fundamental benchmark flows to realistic helicopter rotors. The book will of interest to industrial practitioners working in aerodynamic design and the analysis of fixed-wing or rotary wing aircraft, while also offering advanced reading material for graduate students in the research areas of Computational Fluid Dynamics (CFD), turbulence modeling and related fields.

Characteristics of Butanol Isomers Oxidation in a Micro Flow Reactor

KAUST Repository

Bin Hamzah, Muhamad Firdaus

2017-05-01

Ignition and combustion characteristics of n-butanol/air, 2-butanol.air and isobutanol/air mixtures at stoichiometric (ϕ = 1) and lean (ϕ = 0.5) conditions were investigated in a micro flow reactor with a controlled temperature profile from 323 K to 1313 K, under atmospheric pressure. Sole distinctive weak flame was observed for each mixture, with inlet fuel/air mixture velocity set low at 2 cm/s. One-dimensional computation with comprehensive chemistry and transport was conducted. At low mixture velocities, one-stage oxidation was confirmed from heat release rate profiles, which was broadly in agreement with the experimental results. The weak flame positions were congruent with literature describing reactivity of the butanol isomers. These weak flame responses were also found to mirror the trend in Anti-Knock Indexes of the butanol isomers. Flux and sensitivity analyses were performed to investigate the fuel oxidation pathways at low and high temperatures. Further computational investigations on oxidation of butanol isomers at higher pressure of 5 atm indicated two-stage oxidation through the heat release rate profiles. Low temperature chemistry is accentuated in the region near the first weak cool flame for oxidation under higher pressure, and its impact on key species – such as hydroxyl radical, hydrogen peroxide and carbon monoxide – were considered. Both experimental and computational findings demonstrate the advantage of employing the micro flow reactor in investigating oxidation processes in the temperature region of interest along the reactor channel. By varying physical conditions such as pressure, the micro flow reactor system is proven to be highly beneficial in elucidating oxidation behavior of butanol isomers in conditions in engines such as those that mirror HCCI operations.

Geometrical effects on the airfoil flow separation and transition

KAUST Repository

Zhang, Wei; Cheng, Wan; Gao, Wei; Qamar, Adnan; Samtaney, Ravi

2015-01-01

We present results from direct numerical simulations (DNS) of incompressible flow over two airfoils, NACA-4412 and NACA-0012-64, to investigate the effects of the airfoil geometry on the flow separation and transition patterns at Re=104 and 10

A method and programme (BREACH) for predicting the flow distribution in water cooled reactor cores

International Nuclear Information System (INIS)

Randles, J.; Roberts, H.A.

1961-03-01

The method presented here of evaluating the flow rate in individual reactor channels may be applied to any type of water cooled reactor in which boiling occurs The flow distribution is calculated with the aid of a MERCURY autocode programme, BREACH, which is described in detail. This programme computes the steady state longitudinal void distribution and pressure drop in a single channel on the basis of the homogeneous model of two phase flow. (author)

A method and programme (BREACH) for predicting the flow distribution in water cooled reactor cores

Energy Technology Data Exchange (ETDEWEB)

Randles, J; Roberts, H A [Technical Assessments and Services Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1961-03-15

The method presented here of evaluating the flow rate in individual reactor channels may be applied to any type of water cooled reactor in which boiling occurs The flow distribution is calculated with the aid of a MERCURY autocode programme, BREACH, which is described in detail. This programme computes the steady state longitudinal void distribution and pressure drop in a single channel on the basis of the homogeneous model of two phase flow. (author)

Geometrical effects on the airfoil flow separation and transition

KAUST Repository

Zhang, Wei

2015-04-25

We present results from direct numerical simulations (DNS) of incompressible flow over two airfoils, NACA-4412 and NACA-0012-64, to investigate the effects of the airfoil geometry on the flow separation and transition patterns at Re=104 and 10 degrees incidence. The two chosen airfoils are geometrically similar except for maximum camber (respectively 4%C and 0 with C the chord length), which results in a larger projection area with respect to the incoming flow for the NACA-4412 airfoil, and a larger leeward surface curvature at the leading edge for the NACA-0012-64 airfoil. The governing equations are discretized using an energy conservative fourth-order spatial discretization scheme. An assessment on the two-point correlation indicates that a spanwise domain size of 0.8C is sufficiently large for the present simulations. We discuss flow separation at the airfoil leading edge, transition of the separated shear layer to three-dimensional flow and subsequently to turbulence. Numerical results reveal a stronger adverse pressure gradient field in the leading edge region of the NACA-0012-64 airfoil due to the rapidly varying surface curvature. As a result, the flow experiences detachment at x/C=0.08, and the separated shear layer transition via Kelvin-Helmholtz mechanism occurs at x/C=0.29 with fully developed turbulent flow around x/C=0.80. These flow development phases are delayed to occur at much downstream positions, respectively, observed around x/C=0.25, 0.71 and 1.15 for the NACA-4412 airfoil. The turbulent intensity, measured by the turbulent fluctuations and turbulent Reynolds stresses, are much larger for NACA-0012-64 from the transition onset until the airfoil trailing edge, while turbulence develops significantly downstream of the trailing edge for the NACA-4412 airfoil. For both airfoils, our DNS results indicate that the mean Reynolds stress u\\'u\\'/U02 reaches its maximum value at a distance from the surface approximately equal to the displacement

Experimental and computational investigation of flow of pebbles in a pebble bed nuclear reactor

Science.gov (United States)

Khane, Vaibhav B.

The Pebble Bed Reactor (PBR) is a 4th generation nuclear reactor which is conceptually similar to moving bed reactors used in the chemical and petrochemical industries. In a PBR core, nuclear fuel in the form of pebbles moves slowly under the influence of gravity. Due to the dynamic nature of the core, a thorough understanding about slow and dense granular flow of pebbles is required from both a reactor safety and performance evaluation point of view. In this dissertation, a new integrated experimental and computational study of granular flow in a PBR has been performed. Continuous pebble re-circulation experimental set-up, mimicking flow of pebbles in a PBR, is designed and developed. Experimental investigation of the flow of pebbles in a mimicked test reactor was carried out for the first time using non-invasive radioactive particle tracking (RPT) and residence time distribution (RTD) techniques to measure the pebble trajectory, velocity, overall/zonal residence times, flow patterns etc. The tracer trajectory length and overall/zonal residence time is found to increase with change in pebble's initial seeding position from the center towards the wall of the test reactor. Overall and zonal average velocities of pebbles are found to decrease from the center towards the wall. Discrete element method (DEM) based simulations of test reactor geometry were also carried out using commercial code EDEM(TM) and simulation results were validated using the obtained benchmark experimental data. In addition, EDEM(TM) based parametric sensitivity study of interaction properties was carried out which suggests that static friction characteristics play an important role from a packed/pebble beds structural characterization point of view. To make the RPT technique viable for practical applications and to enhance its accuracy, a novel and dynamic technique for RPT calibration was designed and developed. Preliminary feasibility results suggest that it can be implemented as a non

The Rolling Transition in a Granular Flow along a Rotating Wall

Directory of Open Access Journals (Sweden)

Aurélie Le Quiniou

2011-11-01

Full Text Available The flow of a dry granular material composed of spherical particles along a rotating boundary has been studied by the discrete element method (DEM. This type of flow is used, among others, as a process to spread particles. The flow consists of several phases. A compression phase along the rotating wall is followed by an elongation of the flow along the same boundary. Eventually, the particles slide or roll independently along the boundary. We show that the main motion of the flow can be characterized by a complex deformation rate of traction/compression and shear. We define numerically an effective friction coefficient of the flow on the scale of the continuum and show a strong decrease of this effective friction beyond a certain critical friction coefficient μ*. We correlate this phenomenon with the apparition of a new transition from a sliding regime to a rolling without sliding regime that we called the rolling transition; this dynamic transition is controlled by the value of the friction coefficient between the particle and the wall. We show that the spherical shape for the particles may represent an optimum for the flow in terms of energetic.

CFD analysis and flow model reduction for surfactant production in helix reactor

NARCIS (Netherlands)

Nikačević, N.M.; Thielen, L.; Twerda, A.; Hof, P.M.J. van den

2014-01-01

Flow pattern analysis in a spiral Helix reactor is conducted, for the application in the commercial surfactant production. Step change response curves (SCR) were obtained from numerical tracer experiments by three-dimensional computational fluid dynamics (CFD) simulations. Non-reactive flow is

Mode transition in bubbly Taylor-Couette flow measured by PTV

International Nuclear Information System (INIS)

Yoshida, K; Tasaka, Y; Murai, Y; Takeda, T

2009-01-01

The drag acting to the inner cylinder in Taylor-Couette flow system can be reduced by bubble injection. In this research, relationship between drag reduction and change of vortical structure in a Taylor-Couette flow is investigated by Particle Tracking Velocimetry (PTV). The velocity vector field in the r-z cross section and the bubble concentration in the front view (z-θ plane) are measured. This paper describes the change of vortical structures with bubbles, and the mode transition that is sensitively affected by the bubbles is discussed. The bubbles accumulate in the three parts relative to vortex position by the interaction between bubbles and vortices. The status of bubble's distribution is different depending on position. This difference affects mode transition as its trigger significantly. The presence of bubbles affects the transition from toroidal mode to spiral mode but does not induce the transition from spiral mode to toroidal mode. Further we found that Taylor vortex bifurcates and a pair of vortices coalesces when the flow switches between spiral mode and toroidal mode.

Study of the Transition Flow Regime using Monte Carlo Methods

Science.gov (United States)

Hassan, H. A.

1999-01-01

This NASA Cooperative Agreement presents a study of the Transition Flow Regime Using Monte Carlo Methods. The topics included in this final report are: 1) New Direct Simulation Monte Carlo (DSMC) procedures; 2) The DS3W and DS2A Programs; 3) Papers presented; 4) Miscellaneous Applications and Program Modifications; 5) Solution of Transitional Wake Flows at Mach 10; and 6) Turbulence Modeling of Shock-Dominated Fows with a k-Enstrophy Formulation.

Gas-Liquid Two-Phase Flows Through Packed Bed Reactors in Microgravity

Science.gov (United States)

Motil, Brian J.; Balakotaiah, Vemuri

2001-01-01

The simultaneous flow of gas and liquid through a fixed bed of particles occurs in many unit operations of interest to the designers of space-based as well as terrestrial equipment. Examples include separation columns, gas-liquid reactors, humidification, drying, extraction, and leaching. These operations are critical to a wide variety of industries such as petroleum, pharmaceutical, mining, biological, and chemical. NASA recognizes that similar operations will need to be performed in space and on planetary bodies such as Mars if we are to achieve our goals of human exploration and the development of space. The goal of this research is to understand how to apply our current understanding of two-phase fluid flow through fixed-bed reactors to zero- or partial-gravity environments. Previous experiments by NASA have shown that reactors designed to work on Earth do not necessarily function in a similar manner in space. Two experiments, the Water Processor Assembly and the Volatile Removal Assembly have encountered difficulties in predicting and controlling the distribution of the phases (a crucial element in the operation of this type of reactor) as well as the overall pressure drop.

A plug flow reactor model of a vanadium redox flow battery considering the conductive current collectors

Science.gov (United States)

König, S.; Suriyah, M. R.; Leibfried, T.

2017-08-01

A lumped-parameter model for vanadium redox flow batteries, which use metallic current collectors, is extended into a one-dimensional model using the plug flow reactor principle. Thus, the commonly used simplification of a perfectly mixed cell is no longer required. The resistances of the cell components are derived in the in-plane and through-plane directions. The copper current collector is the only component with a significant in-plane conductance, which allows for a simplified electrical network. The division of a full-scale flow cell into 10 layers in the direction of fluid flow represents a reasonable compromise between computational effort and accuracy. Due to the variations in the state of charge and thus the open circuit voltage of the electrolyte, the currents in the individual layers vary considerably. Hence, there are situations, in which the first layer, directly at the electrolyte input, carries a multiple of the last layer's current. The conventional model overestimates the cell performance. In the worst-case scenario, the more accurate 20-layer model yields a discharge capacity 9.4% smaller than that computed with the conventional model. The conductive current collector effectively eliminates the high over-potentials in the last layers of the plug flow reactor models that have been reported previously.

Modeling on bubbly to churn flow pattern transition in narrow rectangular channel

International Nuclear Information System (INIS)

Wang Yanlin; Chen Bingde; Huang Yanping; Wang Junfeng

2012-01-01

A theoretical model based on some reasonable concepts was developed to predict the bubbly flow to churn flow pattern transition in vertical narrow rectangular channel under flow boiling condition. The maximum size of ideal bubble in narrow rectangular channel was calculated based on previous literature. The thermal hydraulics boundary condition of bubbly to churn flow pattern transition was exported from Helmholtz and maximum size of ideal bubble. The theoretical model was validated by existent experimental data. (authors)

Analysis of loss of flow events on Brazilian multipurpose reactor by RELAP5 code

International Nuclear Information System (INIS)

Soares, Humberto V.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Aronne, Ivan D.; Rezende, Guilherme P.

2011-01-01

The Brazilian Multipurpose Reactor (BMR) is currently being projected and analyzed. It will be a 30 MW open pool multipurpose research reactor with a compact core using Materials Testing Reactor (MTR) type fuel assembly, with planar plates. BMR will be cooled by light water and moderated by beryllium and heavy water. This work presents the calculations of steady state operation of BMR using the RELAP5 model and also three transient cases of loss of flow accident (LOFA), in the primary cooling system. A LOFA may arise through failures associated with the primary cooling system pumps or through events resulting in a decrease in the primary coolant flow with the primary cooling system pumps functioning normally. The cases presented in this paper are: primary cooling system pump shaft seizure, failure of one primary cooling system pump motor and failure of both primary cooling system pump motors. In the shaft seizure case, the flow reduction is sudden, with the blocking of the flow coast down The motor failure cases, deal with the failure of one or two pump motor due to, for example, malfunction or interruption of power and differently of the shaft seizure it can be observed the flow coast down provided by the pump inertia. It is shown that after all initiating events the reactor reaches a safe new steady state keeping the integrity of the fuel elements. (author)

Analysis of loss of flow events on Brazilian multipurpose reactor by RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Soares, Humberto V.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria Auxiliadora F., E-mail: antonella@nuclear.ufmg.br, E-mail: laubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br [Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, UFMG, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores, CNPq (Brazil); Aronne, Ivan D.; Rezende, Guilherme P., E-mail: aroneid@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte (Brazil).

2011-07-01

The Brazilian Multipurpose Reactor (BMR) is currently being projected and analyzed. It will be a 30 MW open pool multipurpose research reactor with a compact core using Materials Testing Reactor (MTR) type fuel assembly, with planar plates. BMR will be cooled by light water and moderated by beryllium and heavy water. This work presents the calculations of steady state operation of BMR using the RELAP5 model and also three transient cases of loss of flow accident (LOFA), in the primary cooling system. A LOFA may arise through failures associated with the primary cooling system pumps or through events resulting in a decrease in the primary coolant flow with the primary cooling system pumps functioning normally. The cases presented in this paper are: primary cooling system pump shaft seizure, failure of one primary cooling system pump motor and failure of both primary cooling system pump motors. In the shaft seizure case, the flow reduction is sudden, with the blocking of the flow coast down The motor failure cases, deal with the failure of one or two pump motor due to, for example, malfunction or interruption of power and differently of the shaft seizure it can be observed the flow coast down provided by the pump inertia. It is shown that after all initiating events the reactor reaches a safe new steady state keeping the integrity of the fuel elements. (author)

Experimental study of transition from laminar to turbulent flow in vertical narrow channel

International Nuclear Information System (INIS)

Wang Chang; Gao Puzhen; Wang Zhanwei; Tan Sichao

2012-01-01

Highlights: ► The effect of wall heating on the laminar to turbulent transition is experimentally studied. ► The flow characteristic demonstrates that heating leads to the delay of transition from laminar to turbulent regimes. ► The heat transfer characteristics also indicates that heating leads to the delay of flow regime transition. - Abstract: Experimental investigation of flow and heat transfer characteristics of a vertical narrow channel with uniform heat flux condition are conducted to analysis the effect of wall heating on the laminar to turbulent transition. The friction factor in the heating condition is compared with that in the adiabatic condition and the results show that wall heating leads to the delay of laminar to turbulent transition. In addition, the heat transfer characteristic indicates that the critical Reynolds number at the point of laminar flow breakdown increases with the increase of fluid temperature difference, and the local Nusselt number at the point of laminar breakdown increases with the increase of the inlet Reynolds number. The analyses of the flow and heat transfer characteristics both indicate that the heating has a stabilizing effect on the water flow at present experimental scale.

Reactor mass flow data base prepared for the nonproliferation alternative systems assessment program

International Nuclear Information System (INIS)

Primm III, R.T.C.

1981-02-01

This report presents charge and discharge mass flow data for reactors judged to have received sufficient technical development to enable them to be demonstrated or commercially available by the year 2000. Brief descriptions of the reactors and fuel cycles evaluated are presented. A discussion of the neutronics methods used to produce the mass flow data is provided. Detailed charge and discharge fuel isotopics are presented. U 3 O 8 , separative work, and fissile material requirements are computed and provided for each fuel cycle

Study of core flow distribution for small modular natural circulation lead or lead-alloy cooled fast reactors

International Nuclear Information System (INIS)

Chen, Zhao; Zhao, Pengcheng; Zhou, Guangming; Chen, Hongli

2014-01-01

Highlights: • A core flow distribution calculation code for natural circulation LFRs was developed. • The comparison study between the channel method and the CFD method was conducted. • The core flow distribution analysis and optimization design for a 10MW natural circulation LFR was conducted. - Abstract: Small modular natural circulation lead or lead-alloy cooled fast reactor (LFR) is a potential candidate for LFR development. It has many attractive advantages such as reduced capital costs and inherent safety. The core flow distribution calculation is an important issue for nuclear reactor design, which will provide important input parameters to thermal-hydraulic analysis and safety analysis. The core flow distribution calculation of a natural circulation LFR is different from that of a forced circulation reactor. In a forced circulation reactor, the core flow distribution can be controlled and adjusted by the pump power and the flow distributor, while in a natural circulation reactor, the core flow distribution is automatically adjusted according to the relationship between the local power and the local resistance feature. In this paper, a non-uniform heated parallel channel flow distribution calculation code was developed and the comparison study between the channel method and the CFD method was carried out to assess the exactness of the developed code. The core flow distribution analysis and optimization design for a 10MW natural circulation LFR was conducted using the developed code. A core flow distribution optimization design scheme for a 10MW natural circulation LFR was proposed according to the optimization analysis results

Physics of Transitional Shear Flows Instability and Laminar–Turbulent Transition in Incompressible Near-Wall Shear Layers

CERN Document Server

Boiko, Andrey V; Grek, Genrih R; Kozlov, Victor V

2012-01-01

Starting from fundamentals of classical stability theory, an overview is given of the transition phenomena in subsonic, wall-bounded shear flows. At first, the consideration focuses on elementary small-amplitude velocity perturbations of laminar shear layers, i.e. instability waves, in the simplest canonical configurations of a plane channel flow and a flat-plate boundary layer. Then the linear stability problem is expanded to include the effects of pressure gradients, flow curvature, boundary-layer separation, wall compliance, etc. related to applications. Beyond the amplification of instability waves is the non-modal growth of local stationary and non-stationary shear flow perturbations which are discussed as well. The volume continues with the key aspect of the transition process, that is, receptivity of convectively unstable shear layers to external perturbations, summarizing main paths of the excitation of laminar flow disturbances. The remainder of the book addresses the instability phenomena found at l...

The role of heater thermal response in reactor thermal limits during oscillartory two-phase flows

Energy Technology Data Exchange (ETDEWEB)

Ruggles, A.E.; Brown, N.W. [Univ. of Tennessee, Knoxville, TN (United States); Vasil`ev, A.D. [Nuclear Safety Institute, Moscow, (Russian Federation); Wendel, M.W. [Oak Ridge National Lab., TN (United States)

1995-09-01

Analytical and numerical investigations of critical heat flux (CHF) and reactor thermal limits are conducted for oscillatory two-phase flows often associated with natural circulation conditions. It is shown that the CHF and associated thermal limits depend on the amplitude of the flow oscillations, the period of the flow oscillations, and the thermal properties and dimensions of the heater. The value of the thermal limit can be much lower in unsteady flow situations than would be expected using time average flow conditions. It is also shown that the properties of the heater strongly influence the thermal limit value in unsteady flow situations, which is very important to the design of experiments to evaluate thermal limits for reactor fuel systems.

Mercury adsorption characteristics of HBr-modified fly ash in an entrained-flow reactor.

Science.gov (United States)

Zhang, Yongsheng; Zhao, Lilin; Guo, Ruitao; Song, Na; Wang, Jiawei; Cao, Yan; Orndorff, William; Pan, Wei-ping

2015-07-01

In this study, the mercury adsorption characteristics of HBr-modified fly ash in an entrained-flow reactor were investigated through thermal decomposition methods. The results show that the mercury adsorption performance of the HBr-modified fly ash was enhanced significantly. The mercury species adsorbed by unmodified fly ash were HgCl2, HgS and HgO. The mercury adsorbed by HBr-modified fly ash, in the entrained-flow reactor, existed in two forms, HgBr2 and HgO, and the HBr was the dominant factor promoting oxidation of elemental mercury in the entrained-flow reactor. In the current study, the concentration of HgBr2 and HgO in ash from the fine ash vessel was 4.6 times greater than for ash from the coarse ash vessel. The fine ash had better mercury adsorption performance than coarse ash, which is most likely due to the higher specific surface area and longer residence time. Copyright © 2015. Published by Elsevier B.V.

Flow-rate measurement using radioactive tracers and transit time method

International Nuclear Information System (INIS)

Turtiainen, Heikki

1986-08-01

The transit time method is a flow measurement method based on tracer techniques. Measurement is done by injecting to the flow a pulse of tracer and measuring its transit time between two detection positions. From the transit time the mean flow velosity and - using the pipe cross section area - the volume flow rate can be calculated. When a radioisotope tracer is used the measurement can be done from outside the pipe and without disturbing the process (excluding the tracer injection). The use of the transit time method has been limited because of difficulties associated with handling and availability of radioactive tracers and lack of equipment suitable for routine use in industrial environments. The purpose of this study was to find out if these difficulties may be overcome by using a portable isotope generator as a tracer source and automating the measurement. In the study a test rig and measuring equipment based on the use of a ''1''3''7Cs/''1''3''7''''mBa isotope generator were constructed. They were used to study the accuracy and error sources of the method and to compare different algorithms to calculate the transit time. The usability of the method and the equipment in industrial environments were studied by carrying out over 20 flow measurements in paper and pulp mills. On the basis of the results of the study, a project for constructing a compact radiatracer flowmeter for industrial use has been started. The application range of this kind of meter is very large. The most obvious applications are in situ calibration of flowmeters, material and energy balance studies, process equipment analyses (e.g. pump efficiency analyses). At the moment tracer techniques are the only methods applicable to these measurements on-line and with sufficient accuracy

Methanol synthesis in a countercurrent gas-solid-solid trickle flow reactor. An experimental study

NARCIS (Netherlands)

Kuczynski, M.; Oyevaar, M.H.; Pieters, R.T.; Westerterp, K.R.

1987-01-01

The synthesis of methanol from CO and H2 was executed in a gas-solid-solid trickle flow reactor. The reactor consisted of three tubular reactor sections with cooling sections in between. The catalyst was Cu on alumina, the adsorbent was a silica-alumina powder and the experimental range 498–523 K,

CFD simulation of flow pattern in a bubble column reactor for forming aerobic granules and its development.

Science.gov (United States)

Fan, Wenwen; Yuan, LinJiang; Li, Yonglin

2018-06-04

The flow pattern is considered to play an important role in the formation of aerobic granular sludge in a bubble column reactor; therefore, it is necessary to understand the behavior of the flow in the reactor. A three-dimensional computational fluid dynamics (CFD) simulation for bubble column reactor was established to visualize the flow patterns of two-phase air-liquid flow and three-phase air-liquid-sludge flow under different ratios of height to diameter (H/D ratio) and superficial gas upflow velocities (SGVs). Moreover, a simulation of the three-phase flow pattern at the same SGV and different characteristics of the sludge was performed in this study. The results show that not only SGV but also properties of sludge involve the transformation of flow behaviors and relative velocity between liquid and sludge. For the original activated sludge floc to cultivate aerobic granules, the flow pattern has nothing to do with sludge, but is influenced by SGV, and the vortices is occurred and the relative velocity is increased with an increase in SGV; the two-phase flow can simplify the three-phase flow that predicts the flow pattern development in bubble column reactor (BCR) for aerobic granulation. For the aerobic granules, the liquid flow behavior developed from the symmetrical circular flow to numbers and small-size vortices with an increase in the sludge diameter, the relative velocity is amount up to u r =5.0, it is 29.4 times of original floc sludge.

Entrained Flow Reactor Test of Potassium Capture by Kaolin

DEFF Research Database (Denmark)

Wang, Guoliang; Jensen, Peter Arendt; Wu, Hao

2015-01-01

In the present study a method to simulate the reaction between gaseous KCl and kaolin at suspension fired condition was developed using a pilot-scale entrained flow reactor (EFR). Kaolin was injected into the EFR for primary test of this method. By adding kaolin, KCl can effectively be captured...

Flow velocity calculation to avoid instability in a typical research reactor core

International Nuclear Information System (INIS)

Oliveira, Carlos Alberto de; Mattar Neto, Miguel

2011-01-01

Flow velocity through a research reactor core composed by MTR-type fuel elements is investigated. Core cooling capacity must be available at the same time that fuel-plate collapse must be avoided. Fuel plates do not rupture during plate collapse, but their lateral deflections can close flow channels and lead to plate over-heating. The critical flow velocity is a speed at which the plates collapse by static instability type failure. In this paper, critical velocity and coolant velocity are evaluated for a typical MTR-type flat plate fuel element. Miller's method is used for prediction of critical velocity. The coolant velocity is limited to 2/3 of the critical velocity, that is a currently used criterion. Fuel plate characteristics are based on the open pool Australian light water reactor. (author)

Churn-annular flow pattern transition in a vertical upward gas-liquid two-phase flow in various conduits

International Nuclear Information System (INIS)

Takenaka, Nobuyuki; Ueda, Tadanobu; Asano, Hitoshi

2008-01-01

Void fraction was measured by neutron radiography for a vertical upward gas-water two-phase flow in a concentric annular tube with and with out a spacer, 4x4 rod bundle with and without a spacer and a tight rod bundle with and without a wrapping wire for various gas and liquid flow rates. The flow patterns of these two-phase flows were determined by the Mishima-Ishii flow pattern map and void fraction was calculated by the Ishii's drift flux model. The predicted values were compared with the experimental results. The void fraction was well predicted by the Mishima-Ishii flow pattern map and the Ishii's drift flux model except the annular flow region with void fraction lower than 0.8 for conduits with small equivalent diameter. A new churn-annular flow pattern transition condition of the void fraction equal to 0.8 was added. The void fraction for the present experimental condition was successful predicted with the new transition model. (author)

Investigation on the Core Bypass Flow in a Very High Temperature Reactor

Energy Technology Data Exchange (ETDEWEB)

Hassan, Yassin

2013-10-22

Uncertainties associated with the core bypass flow are some of the key issues that directly influence the coolant mass flow distribution and magnitude, and thus the operational core temperature profiles, in the very high-temperature reactor (VHTR). Designers will attempt to configure the core geometry so the core cooling flow rate magnitude and distribution conform to the design values. The objective of this project is to study the bypass flow both experimentally and computationally. Researchers will develop experimental data using state-of-the-art particle image velocimetry in a small test facility. The team will attempt to obtain full field temperature distribution using racks of thermocouples. The experimental data are intended to benchmark computational fluid dynamics (CFD) codes by providing detailed information. These experimental data are urgently needed for validation of the CFD codes. The following are the project tasks: • Construct a small-scale bench-top experiment to resemble the bypass flow between the graphite blocks, varying parameters to address their impact on bypass flow. Wall roughness of the graphite block walls, spacing between the blocks, and temperature of the blocks are some of the parameters to be tested. • Perform CFD to evaluate pre- and post-test calculations and turbulence models, including sensitivity studies to achieve high accuracy. • Develop the state-of-the art large eddy simulation (LES) using appropriate subgrid modeling. • Develop models to be used in systems thermal hydraulics codes to account and estimate the bypass flows. These computer programs include, among others, RELAP3D, MELCOR, GAMMA, and GAS-NET. Actual core bypass flow rate may vary considerably from the design value. Although the uncertainty of the bypass flow rate is not known, some sources have stated that the bypass flow rates in the Fort St. Vrain reactor were between 8 and 25 percent of the total reactor mass flow rate. If bypass flow rates are on the

Experimental scaling law for the subcritical transition to turbulence in plane Poiseuille flow.

Science.gov (United States)

Lemoult, Grégoire; Aider, Jean-Luc; Wesfreid, José Eduardo

2012-02-01

We present an experimental study of the transition to turbulence in a plane Poiseuille flow. Using a well-controlled perturbation, we analyze the flow by using extensive particle image velocimetry and flow visualization (using laser-induced fluorescence) measurements, and use the deformation of the mean velocity profile as a criterion to characterize the state of the flow. From a large parametric study, four different states are defined, depending on the values of the Reynolds number and the amplitude of the perturbation. We discuss the role of coherent structures, such as hairpin vortices, in the transition. We find that the minimal amplitude of the perturbation triggering transition scales asymptotically as Re(-1).

Investigation on flow patterns and transition characteristics in a tube-bundle channel

International Nuclear Information System (INIS)

Xiang Wenyuan; Lu Yonghong; Zhao Guisheng

2012-01-01

Tube-bundle channels have been widely used in condenser-evaporator and other industrial heat-exchange equipment. The characteristics of two-phase flow patterns and their transitions for refrigerant R-113 through a vertical tube-bundle channel are experimentally investigated using high-speed camera. Experiments show that there are four main flow patterns in the tube-bundle channel, which are bubbly flow, bubbly-churn flow, churn flow and annular flow. And in the same cross-section of tube- bundle channels, it is shown that there might be different flow patterns in different sub-channels. The flow pattern transitions exhibit unsynchronized in different sub-channels. On the basis of experimental research, the flow pattern map is drawn and analyses are made on the comparison of differences between boiling flow patterns in a circular tube and those in a tube-bundle channel. (authors)

Dynamics of baroclinic wave pattern in transition zones between different flow regimes

International Nuclear Information System (INIS)

Larcher, Thomas von; Egbers, Christoph

2005-01-01

Baroclinic waves, both steady and time-dependent, are studied experimentally in a differentially heated rotating cylindrical gap with a free surface, cooled from within. Water is used as working fluid. We focus especially on transition zones between different flow regimes, where complex flow pattern like mixed-mode states are found. The transition from steady wave regime to irregular flow is also of particular interest. The surface flow is observed with visualisation techniques. Velocity time series are measured with the optical laser-Doppler-velocimetry technique. Thermographic measurements are applied for temperature field visualisations

Removal of natural organic matter and arsenic from water by electrocoagulation/flotation continuous flow reactor.

Science.gov (United States)

Mohora, Emilijan; Rončević, Srdjan; Dalmacija, Božo; Agbaba, Jasmina; Watson, Malcolm; Karlović, Elvira; Dalmacija, Milena

2012-10-15

The performance of the laboratory scale electrocoagulation/flotation (ECF) reactor in removing high concentrations of natural organic matter (NOM) and arsenic from groundwater was analyzed in this study. An ECF reactor with bipolar plate aluminum electrodes was operated in the horizontal continuous flow mode. Electrochemical and flow variables were optimized to examine ECF reactor contaminants removal efficiency. The optimum conditions for the process were identified as groundwater initial pH 5, flow rate=4.3 l/h, inter electrode distance=2.8 cm, current density=5.78 mA/cm(2), A/V ratio=0.248 cm(-1). The NOM removal according to UV(254) absorbance and dissolved organic matter (DOC) reached highest values of 77% and 71% respectively, relative to the raw groundwater. Arsenic removal was 85% (6.2 μg As/l) relative to raw groundwater, satisfying the drinking water standards. The specific reactor electrical energy consumption was 17.5 kWh/kg Al. The specific aluminum electrode consumption was 66 g Al/m(3). According to the obtained results, ECF in horizontal continuous flow mode is an energy efficient process to remove NOM and arsenic from groundwater. Copyright © 2012 Elsevier B.V. All rights reserved.

Discontinuous jamming transitions in soft materials: coexistence of flowing and jammed states

International Nuclear Information System (INIS)

Dennin, Michael

2008-01-01

Many systems in nature exhibit transitions between fluid-like states and solid-like states, or 'jamming transitions'. There is a strong theoretical foundation for understanding equilibrium phase transitions that involve solidification, or jamming. Other jamming transitions, such as the glass transition, are less well understood. The jamming phase diagram has been proposed to unify the description of equilibrium phase transitions, the glass transitions, and other nonequilibrium jamming transitions. As with equilibrium phase transitions, which can either be first order (discontinuous in a relevant order parameter) or second order (continuous), one would expect that generalized jamming transitions can be continuous or discontinuous. In studies of flow in complex fluids, there is a wide range of evidence for discontinuous transitions, mostly in the context of shear localization, or shear banding. In this paper, I review the experimental evidence for discontinuous transitions. I focus on systems in which there is a discontinuity in the rate of strain between two, coexisting states: one in which the material is flowing and the other in which it is solid-like. (topical review)

Comparison of reactivity in a flow reactor and a single cylinder engine

Energy Technology Data Exchange (ETDEWEB)

Natelson, Robert H.; Johnson, Rodney O.; Kurman, Matthew S.; Cernansky, Nicholas P.; Miller, David L. [Department of Mechanical Engineering and Mechanics, Drexel University, 3141 Chestnut Street, Philadelphia, PA 19104-2875 (United States)

2010-10-15

The relative reactivity of 2:1:1 and 1:1:1 mixtures of n-decane:n-butylcyclohexane:n-butylbenzene and an average sample of JP-8 were evaluated in a single cylinder engine and compared to results obtained in a pressurized flow reactor. At compression ratios of 14:1, 15:1, and 16:1, inlet temperature of 500 K, inlet pressure of 0.1 MPa, equivalence ratio of 0.23, and engine speed of 800 RPM, the autoignition delay times were, from shortest to longest, the 2:1:1, followed by the 1:1:1, and then the JP-8. This order corresponded with recent results in a pressurized flow reactor, where the preignition oxidation chemistry was monitored at temperatures of 600-800 K, 0.8 MPa pressure, and an equivalence ratio of 0.30, and where the preignition reactivity from highest to lowest was the 2:1:1, followed by the 1:1:1, and the JP-8. This shows that the relative reactivity at low temperatures in the flow reactor tracks the autoignition tendencies in the engine for these particular fuels. (author) the computed experimental error. (author)

Optimum cadmium reactor designs for colorimetric determination of nitrate with flow injection and gas-segmented continuous flow analyzers

International Nuclear Information System (INIS)

Patton, C.J.

1989-01-01

Cadmium reactor types can be grouped into four categories: packed bed; filamentous; open tubular; and planar. Packed bed cadmium reactors, in the form of cadmium filings, granules, powder, or electrolytically precipitated needles packed into glass or polymeric tubes, are by far the most widely used for both FIA and CFA methods. Surprisingly, filamentous cadmium reactors, in the form of cadmium wire slipped into flexible polymeric tubing, have been reported for CFA applications only. Open tubular cadmium reactors, in the form of small diameter cadmium tubing coiled into a helix, have been fully characterized and described for CFA applications. A preliminary description of planar cadmium reactors, in the form of cadmium foil sandwiched between continuous flow dialyzer blocks has also been reported. In this presentation, each reactor type is evaluated in terms of cost, ease of use, reduction efficiency, and long-term stability. Factors that make some reactors more applicable to FIA than to CFA (or the reverse) are also discussed, and experimental data are presented

Characteristics of Butanol Isomers Oxidation in a Micro Flow Reactor

KAUST Repository

Bin Hamzah, Muhamad Firdaus

2017-01-01

Ignition and combustion characteristics of n-butanol/air, 2-butanol.air and isobutanol/air mixtures at stoichiometric (ϕ = 1) and lean (ϕ = 0.5) conditions were investigated in a micro flow reactor with a controlled temperature profile from 323 K

K-capture by Al-Si based Additives in an Entrained Flow Reactor

DEFF Research Database (Denmark)

Wang, Guoliang; Jensen, Peter Arendt; Wu, Hao

2016-01-01

A water slurry, consisting of KCl and Al-Si based additives (kaolin and coal fly ash) was fed into an entrained flow reactor (EFR) to study the K-capturing reaction of the additives at suspension-fired conditions. Solid products collected from the reactor were analysed with respect to total...... of KCl to K-aluminosilicate decreased. When reaction temperature increased from 1100 °C to 1450 °C, the conversion of KCl does not change significantly, which differs from the trend observed in fixed-bed reactor....

Evaluating the consequences of loss of flow accident for a typical VVER-1000 nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Mirvakili, S.M.; Safaei, S. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Dept. of Nuclear Engineering, School of Mechanical Engineering; Faghihi, F. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Safety Research Center

2010-07-01

The loss of coolant flow in a nuclear reactor can result from a mechanical or electrical failure of the coolant pump. If the reactor is not tripped promptly, the immediate effect is a rapid increase in coolant temperature, decrease in minimum departure from nucleate boiling ratio (DNBR) and fuel damage. This study evaluated the shaft seizure of a reactor coolant pump in a VVER-1000 nuclear reactor. The locked rotor results in rapid reduction of flow through the affected reactor coolant loop and in turn leads to an increase in the primary coolant temperature and pressure. The analysis was conducted with regard for superimposing loss of power to the power plant at the initial accident moment. The required transient functions of flow, pressure and power were obtained using system transient calculations applied in COBRA-EN computer code in order to calculate the overall core thermal-hydraulic parameters such as temperature, critical heat flux and DNBR. The study showed that the critical period for the locked rotor accident is the first few seconds during which the maximum values of pressure and temperature are reached. 10 refs., 1 tab., 3 figs.

Modelling of sludge blanket height and flow pattern in UASB reactors treating municipal wastewater

International Nuclear Information System (INIS)

Singh, K.S.; Viraraghavan, T.

2002-01-01

Two upflow anaerobic sludge blanket (UASB) reactors were started-up and operated for approximately 900 days to examine the feasibility of treating municipal wastewater under low temperature conditions. A modified solid distribution model was formulated by incorporating the variation of biogas production rate with a change in temperature. This model was used to optimize the sludge blanket height of UASB reactors for an effective operation of gas-liquid-solid (GLS) separation device. This model was found to simulate well the solid distribution as confirmed experimental observation of solid profile along the height of the reactor. Mathematical analysis of tracer curves indicated the presence of a mixed type of flow pattern in the sludge-bed zone of the reactor. It was found that the dead-zone and by-pass flow fraction were impacted by the change in operating temperatures. (author)

On natural circulation in High Temperature Gas-Cooled Reactors and pebble bed reactors for different flow regimes and various coolant gases

International Nuclear Information System (INIS)

Melesed'Hospital, G.

1983-01-01

The use of CO 2 or N 2 (heavy gas) instead of helium during natural circulation leads to improved performance in both High Temperature Gas-Cooled Reactors (HTGR) and in Pebble Bed Reactors (PBR). For instance, the coolant temperature rise corresponding to a coolant pressure level and a rate of afterheat removal could be only 18% with CO 2 as compared to He, for laminar flow in HTGR; this value would be 40% in PBR. There is less difference between HTGR and PBR for turbulent flows; CO 2 is found to be always better than N 2 . These types of results derived from relationships between coolant properties, coolant flow, temperature rise, pressure, afterheat levels and core geometry, are obtained for HTGR and PBR for various flow regimes, both within the core and in the primary loop

Secondary flows in the cooling channels of the high-performance light-water reactor

Energy Technology Data Exchange (ETDEWEB)

Laurien, E.; Wintterle, Th. [Stuttgart Univ., Institute for Nuclear Technolgy and Energy Systems (IKE) (Germany)

2007-07-01

The new design of a High-Performance Light-Water Reactor (HPLWR) involves a three-pass core with an evaporator region, where the compressed water is heated above the pseudo-critical temperature, and two superheater regions. Due to the strong dependency of the supercritical water density on the temperature significant mass transfer between neighboring cooling channels is expected if the temperature is unevenly distributed across the fuel element. An inter-channel flow is then superimposed to the secondary flow vortices induced by the non-isotropy of turbulence. In order to gain insight into the resulting flow patterns as well as into temperature and density distributions within the various subchannels of the fuel element CFD (Computational Fluid Dynamics) calculations for the 1/8 fuel element are performed. For simplicity adiabatic boundary conditions at the moderator box and the fuel element box are assumed. Our investigation confirms earlier results obtained by subchannel analysis that the axial mass flux is significantly reduced in the corner subchannel of this fuel element resulting in a net mass flux towards the neighboring subchannels. Our results provide a first estimation of the magnitude of the secondary flows in the pseudo-critical region of a supercritical light-water reactor. Furthermore, it is demonstrated that CFD is an efficient tool for investigations of flow patterns within nuclear reactor fuel elements. (authors)

Micro reactor and flow chemistry for industrial applications in drug discovery and development

NARCIS (Netherlands)

Tambarussi Baraldi, P.; Hessel, V.

2012-01-01

In this review, case studies focused on syntheses of active pharmaceutical ingredients, intermediates and lead compounds are reported employing micro reactors and continuous flow technology in areas such as medicinal chemistry, chemical development and manufacturing. The advantages of flow

A flow reactor setup for photochemistry of biphasic gas/liquid reactions

Directory of Open Access Journals (Sweden)

Josef Schachtner

2016-08-01

Full Text Available A home-built microreactor system for light-mediated biphasic gas/liquid reactions was assembled from simple commercial components. This paper describes in full detail the nature and function of the required building elements, the assembly of parts, and the tuning and interdependencies of the most important reactor and reaction parameters. Unlike many commercial thin-film and microchannel reactors, the described set-up operates residence times of up to 30 min which cover the typical rates of many organic reactions. The tubular microreactor was successfully applied to the photooxygenation of hydrocarbons (Schenck ene reaction. Major emphasis was laid on the realization of a constant and highly reproducible gas/liquid slug flow and the effective illumination by an appropriate light source. The optimized set of conditions enabled the shortening of reaction times by more than 99% with equal chemoselectivities. The modular home-made flow reactor can serve as a prototype model for the continuous operation of various other reactions at light/liquid/gas interfaces in student, research, and industrial laboratories.

Partial nitrification using aerobic granules in continuous-flow reactor: rapid startup.

Science.gov (United States)

Wan, Chunli; Sun, Supu; Lee, Duu-Jong; Liu, Xiang; Wang, Li; Yang, Xue; Pan, Xiangliang

2013-08-01

This study applied a novel strategy to rapid startup of partial nitrification in continuous-flow reactor using aerobic granules. Mature aerobic granules were first cultivated in a sequencing batch reactor at high chemical oxygen demand in 16 days. The strains including the Pseudoxanthomonas mexicana strain were enriched in cultivated granules to enhance their structural stability. Then the cultivated granules were incubated in a continuous-flow reactor with influent chemical oxygen deamnad being stepped decreased from 1,500 ± 100 (0-19 days) to 750 ± 50 (20-30 days), and then to 350 ± 50 mg l(-1) (31-50 days); while in the final stage 350 mg l(-1) bicarbonate was also supplied. Using this strategy the ammonia-oxidizing bacterium, Nitrosomonas europaea, was enriched in the incubated granules to achieve partial nitrification efficiency of 85-90% since 36 days and onwards. The partial nitrification granules were successfully harvested after 52 days, a period much shorter than those reported in literature. Copyright © 2013 Elsevier Ltd. All rights reserved.

Progress in the Development of Compressible, Multiphase Flow Modeling Capability for Nuclear Reactor Flow Applications

Energy Technology Data Exchange (ETDEWEB)

R. A. Berry; R. Saurel; F. Petitpas; E. Daniel; O. Le Metayer; S. Gavrilyuk; N. Dovetta

2008-10-01

In nuclear reactor safety and optimization there are key issues that rely on in-depth understanding of basic two-phase flow phenomena with heat and mass transfer. Within the context of multiphase flows, two bubble-dynamic phenomena – boiling (heterogeneous) and flashing or cavitation (homogeneous boiling), with bubble collapse, are technologically very important to nuclear reactor systems. The main difference between boiling and flashing is that bubble growth (and collapse) in boiling is inhibited by limitations on the heat transfer at the interface, whereas bubble growth (and collapse) in flashing is limited primarily by inertial effects in the surrounding liquid. The flashing process tends to be far more explosive (and implosive), and is more violent and damaging (at least in the near term) than the bubble dynamics of boiling. However, other problematic phenomena, such as crud deposition, appear to be intimately connecting with the boiling process. In reality, these two processes share many details.

Using Flow Electrodes in Multiple Reactors in Series for Continuous Energy Generation from Capacitive Mixing

KAUST Repository

Hatzell, Marta C.

2014-12-09

Efficient conversion of “mixing energy” to electricity through capacitive mixing (CapMix) has been limited by low energy recoveries, low power densities, and noncontinuous energy production resulting from intermittent charging and discharging cycles. We show here that a CapMix system based on a four-reactor process with flow electrodes can generate constant and continuous energy, providing a more flexible platform for harvesting mixing energy. The power densities were dependent on the flow-electrode carbon loading, with 5.8 ± 0.2 mW m–2 continuously produced in the charging reactor and 3.3 ± 0.4 mW m–2 produced in the discharging reactor (9.2 ± 0.6 mW m–2 for the whole system) when the flow-electrode carbon loading was 15%. Additionally, when the flow-electrode electrolyte ion concentration increased from 10 to 20 g L–1, the total power density of the whole system (charging and discharging) increased to 50.9 ± 2.5 mW m–2.

Student-Fabricated Microfluidic Devices as Flow Reactors for Organic and Inorganic Synthesis

Science.gov (United States)

Feng, Z. Vivian; Edelman, Kate R.; Swanson, Benjamin P.

2015-01-01

Flow synthesis in microfluidic devices has been rapidly adapted in the pharmaceutical industry and in many research laboratories. Yet, the cost of commercial flow reactors is a major factor limiting the dissemination of this technology in the undergraduate curriculum. Here, we present a laboratory activity where students design and fabricate…

Technical and QA plan: Boiling behavior during flow instability

International Nuclear Information System (INIS)

Coutts, D.A.

1991-01-01

The coolant flow in a nuclear reactor core under normal operating conditions is kept as a subcooled liquid. This coolant is evenly distributed throughout the multiple flow channels with a uniform pressure profile across each coolant flow channel. If the coolant flow is reduced, the flow through individual channels will also decrease. A decrease in coolant flow will result in higher coolant temperatures if the heat flux is not reduced. When flow is significantly decreased, localized boiling may occur. This localized boiling can restrict coolant flow and the ability to transfer heat out of the reactor system. The maximum operating power for the reactor may be limited by how the coolant system reacts to a flow instability. One of the methods to assure safe operation during a reducing flow transient, is to operate at a power level below that necessary to initiate a flow excursion. Several correlations have been used to predict the conditions which will proceed a flow excursion. These correlations rely on the steady state behavior of the coolant and are based on steady-state testing. There are two significant points which this project will try to identify. The first is when vapor first forms on the channel surface. This might be designated as the Nucleate Vapor Transition. (Steady state equivalent is ONB). The second is when the vapor formation rate is large enough to lead to flow instability and thermal excursion. This point might be designated as the Significant Vapor Transition. (Steady state equivalent is OSV). A correlation will be developed to relate established steady state relations with the behavior of transient systems

Enlargement of the pulsing flow regime by periodic operation of a trickle-bed reactor.

NARCIS (Netherlands)

Boelhouwer, J.G.; Piepers, H.W.; Drinkenburg, A.A.H.

1999-01-01

Potential advantages of pulsing flow in trickle-bed reactors include capacity increase and elimination of hot spots through the enhanced mass and heat transfer rates. A disadvantage of naturally occurring pulsing flow is the necessity of relatively high gas and liquid flow rates, especially at

Investigation of fluid flow in various geometries related to nuclear reactor using PIV system

International Nuclear Information System (INIS)

Kansal, A.K.; Maheshwari, N.K.; Singh, R.K.; Vijayan, P.K.; Saha, D.; Singh, R.K.; Joshi, V.M.

2011-01-01

Particle Image Velocimetry (PIV) is a non-intrusive technique for simultaneously measuring the velocities at many points in a fluid flow. The PIV system used is comprised of Nd:YAG laser source, CCD (Charged Coupled Device) camera, timing controller (to control the laser and camera) and software used for analyzing the flow velocities. Several case studies related to nuclear reactor were performed with the PIV system. Some of the cases like flow in circular tube, submerged jet, natural convection in a water pool, flow field of moderator inlet diffuser of 500 MWe Pressurised Heavy Water Reactor (PHWR) and fluidic flow control device (FFCD) used in advanced accumulator of Emergency Core Cooling System (ECCS) have been studied using PIV system. Theoretical studies have been performed and comparisons with PIV results are also given in the present studies. (author)

Numerical Simulation of a Coolant Flow and Heat Transfer in a Pebble Bed Reactor

International Nuclear Information System (INIS)

In, Wang-Kee; Kim, Min-Hwan; Lee, Won-Jae

2008-01-01

Pebble Bed Reactor(PBR) is one of the very high temperature gas cooled reactors(VHTR) which have been reviewed in the Generation IV International Forum as potential sources for future energy needs, particularly for a hydrogen production. The pebble bed modular reactor(PBMR) exhibits inherent safety features due to the low power density and the large amount of graphite present in the core. PBR uses coated fuel particles(TRISO) embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the PBR core during a reactor operation and the coolant flows around randomly distributed spheres. For the reliable operation and the safety of the PBR, it is important to understand the coolant flow structure and the fuel pebble temperature in the PBR core. There have been few experimental and numerical studies to investigate the fluid and heat transfer phenomena in the PBR core. The objective of this paper is to predict the fluid and heat transfer in the PBR core. The computational fluid dynamics (CFD) code, STAR-CCM+(V2.08) is used to perform the CFD analysis using the design data for the PBMR400

Depression of the Superfluid Transition Temperature in 4He by a Heat Flow

International Nuclear Information System (INIS)

Yin Liang; Qi Xin; Lin Peng

2014-01-01

The depression of the superfluid transition temperature T λ in 4 He by a heat flow Q is studied. A small sealed cell with a capillary is introduced and a stable and flat superfluid transition temperature plateau is easily obtained by controlling the temperature of the variable-temperature platform and the bottom chamber of the sealed cell. Owing to the depression effect of the superfluid transition temperature by the heat flow, the heat flow through the capillary is changed by the temperature control to obtain multiple temperature plateaus of different heat flows. The thermometer self-heating effect, the residual heat leak of the 4.2 K environment, the temperature difference on the He II liquid column, the Kapiza thermal resistance between the liquid helium and the copper surface of the sealed cell, the temperature gradient of the sealed cell, the static pressure of the He II liquid column and other factors have influence on the depression effect and the influence is analyzed in detail. Twenty experiments of the depression of the superfluid transition temperature in 4 He by heat flow are made with four sealed cells in one year. The formula of the superfluid transition temperature pressured by the heat flow is T λ (Q) = −0.00000103Q + 2.1769108, and covers the range 229 ≤ Q ≤ 6462 μW/cm 2

A coupled overlapping domain method for the computation of transitional flow through artificial heart valves

NARCIS (Netherlands)

Verkaik, A.C.; Bogaerds, A.C.B.; Storti, F.; Van De Vosse, F.N.

2012-01-01

When blood is pumped through the aortic valves, it has a time dependent flow with a relatively high speed, resulting in Reynolds numbers between 1500 and 3000. Hence, flow is in the transitional regime between laminar and turbulent flow. Transitional flow contains small scale fluctuations, see

Computer simulation of two-phase flow in nuclear reactors

International Nuclear Information System (INIS)

Wulff, W.

1993-01-01

Two-phase flow models dominate the requirements of economic resources for the development and use of computer codes which serve to analyze thermohydraulic transients in nuclear power plants. An attempt is made to reduce the effort of analyzing reactor transients by combining purpose-oriented modelling with advanced computing techniques. Six principles are presented on mathematical modeling and the selection of numerical methods, along with suggestions on programming and machine selection, all aimed at reducing the cost of analysis. Computer simulation is contrasted with traditional computer calculation. The advantages of run-time interactive access operation in a simulation environment are demonstrated. It is explained that the drift-flux model is better suited than the two-fluid model for the analysis of two-phase flow in nuclear reactors, because of the latter's closure problems. The advantage of analytical over numerical integration is demonstrated. Modeling and programming techniques are presented which minimize the number of needed arithmetical and logical operations and thereby increase the simulation speed, while decreasing the cost. (orig.)

Evaluation method for core thermohydraulics during natural circulation in fast reactors numerical predictions of inter-wrapper flow

International Nuclear Information System (INIS)

Kamide, H.; Kimura, N.; Miyakoshi, H.; Nagasawa, K.

2001-01-01

Decay heat removal using natural circulation is one of the important functions for the safety of fast reactors. As a decay heat removal system, direct reactor auxiliary cooling system has been selected in current designs of fast reactors. In this design, dumped heat exchanger provides cold sodium and it covers the reactor core outlet. The cold sodium can penetrate into the gap region between the subassemblies. This gap flow is referred as inter-wrapper flow (IWF). A numerical estimation method for such natural circulation phenomena in a reactor core has been developed, which models each subassembly as a rectangular duct with gap region between the subassemblies and also the upper plenum in a reactor vessel. This numerical simulation method was verified based on experimental data of a sodium test using 7- subassembly core model and also a water test which simulates IWF using the 1/12 sector model of a reactor core. We applied the estimation method to the natural circulation in a 600 MW class fast reactor. The temperature in the core strongly depended on IWF, flow redistribution in the core, and inter-subassembly heat transfer. It is desired for prediction methods on the natural circulation to simulate these phenomena. (author)

Analysis of transition to fuel cycle system with continuous recycling in fast and thermal reactors - 5060

International Nuclear Information System (INIS)

Passereini, S.; Feng, B.; Fei, T.; Kim, T.K.; Taiwo, T.A.; Brown, N.R.; Cuadra, A.

2015-01-01

A recent Evaluation and Screening study of nuclear fuel cycle options identified a few groups of options as most promising. One of these most promising Evaluation Groups (EGs) is characterized by the continuous recycling of uranium (U) and transuranics (TRU) with natural uranium feed in both fast and thermal critical reactors. This evaluation group, designated as EG30, is represented by an example fuel cycle option that employs a two-technology, two-stage fuel cycle system. The first stage involves the continuous recycling of co-extracted U/TRU in Sodium-cooled Fast Reactors (SFRs) with metallic fuel and breeding ratio greater than 1. The second stage involves the use of the surplus TRU in Mixed Oxide (MOX) fuel in Pressurized Water Reactors that are MOX-capable (MOX-PWRs). This paper presents and discusses preliminary fuel cycle analysis results from the fuel cycle codes VISION and DYMOND for the transition to this fuel cycle option from the current once-through cycle in the United States (U.S.) that consists of Light Water Reactors (LWRs) that only use conventional UO 2 fuel. The analyses in this paper are applicable for a constant 100 GWe capacity, roughly the size of the U.S. nuclear fleet. Two main strategies for the transition to EG30 were analyzed: 1) deploying both SFRs and MOX-PWRs in parallel or 2) deploying them in series with the SFR fleet first. With an estimated retirement schedule for the existing LWRs, an assumed reactor lifetime of 60 years, and no growth, the nuclear system fully transitions to the new fuel cycle within 100 years for both strategies without SFR fuel shortages. Compared to the once-through cycle, transition to the SFR/MOX-PWR fleet with continuous recycle was shown to offer significant reductions in uranium consumption and waste disposal requirements. In addition, these initial calculations revealed a few notable modeling and strategy questions regarding how recycled resources are allocated, reactors that can switch between

Novel swirl-flow reactor for kinetic studies of semiconductor photocatalysis

NARCIS (Netherlands)

Ray, A.K; Beenackers, A.A C M

1997-01-01

A new two-phase swirl-flow monolithic-type reactor was designed to study the kinetics of heterogeneous photocatalytic processes on immobilized semiconductor catalysts. True kinetic rate constants for destruction of a textile dye were measured as a function of wavelength of light intensity and angle

Calibration of a γ- Re θ transition model and its application in low-speed flows

Science.gov (United States)

Wang, YunTao; Zhang, YuLun; Meng, DeHong; Wang, GunXue; Li, Song

2014-12-01

The prediction of laminar-turbulent transition in boundary layer is very important for obtaining accurate aerodynamic characteristics with computational fluid dynamic (CFD) tools, because laminar-turbulent transition is directly related to complex flow phenomena in boundary layer and separated flow in space. Unfortunately, the transition effect isn't included in today's major CFD tools because of non-local calculations in transition modeling. In this paper, Menter's γ- Re θ transition model is calibrated and incorporated into a Reynolds-Averaged Navier-Stokes (RANS) code — Trisonic Platform (TRIP) developed in China Aerodynamic Research and Development Center (CARDC). Based on the experimental data of flat plate from the literature, the empirical correlations involved in the transition model are modified and calibrated numerically. Numerical simulation for low-speed flow of Trapezoidal Wing (Trap Wing) is performed and compared with the corresponding experimental data. It is indicated that the γ- Re θ transition model can accurately predict the location of separation-induced transition and natural transition in the flow region with moderate pressure gradient. The transition model effectively imporves the simulation accuracy of the boundary layer and aerodynamic characteristics.

A microcatalytic flow reactor for the study of heterogeneous catalytic reactions at elevated pressures

Energy Technology Data Exchange (ETDEWEB)

Belyi, A S; Fomichev, Yu V; Duplyakin, V K; Alfeev, V S

1977-07-01

A microcatalytic flow reactor for the study of heterogeneous catalytic reactions at elevated pressures (i.e., up to 40 atm) and nearly isothermal conditions up to 600/sup 0/C was designed for the conversion of small quantities of petrochemical feeds or feed mixtures at uniform, controllable flow rates of 0.5-5.0 cc/hr, for direct gas-chromatographic analysis of product samples at the reactor outlet, and for continuous monitoring of the degree of conversion in processes that evolve or absorb hydrogen. The device includes a feed injection system with a unique sealing feature that ensures a constant flow of liquid from a feed buret under positive displacement by a counterweight piston at very low rates into a tubular reactor of the perfect mixing type, a highly efficient vaporizer-mixer, and a two-channel sampler leading to the chromatograph. The apparatus has proved reliable, accurate, and convenient in two years of regular use. Diagrams.

Mean shear flows, zonal flows, and generalized Kelvin-Helmholtz modes in drift wave turbulence: A minimal model for L→H transition

International Nuclear Information System (INIS)

Kim, Eun-jin; Diamond, P.H.

2003-01-01

The dynamics of and an interplay among structures (mean shear flows, zonal flows, and generalized Kelvin-Helmholtz modes) are studied in drift wave turbulence. Mean shear flows are found to inhibit the nonlinear generation of zonal flows by weakening the coherent modulation response of the drift wave spectrum. Based on this result, a minimal model for the L→H (low- to high-confinement) transition is proposed, which involves the amplitude of drift waves, zonal flows, and the density gradient. A transition to quiescent H-mode sets in as the profile becomes sufficiently steep to completely damp out drift waves, following an oscillatory transition phase where zonal flows regulate drift wave turbulence. The different roles of mean flows and zonal flows are elucidated. Finally, the effect of poloidally nonaxisymmetric structures (generalized Kelvin-Helmholtz mode) on anomalous transport is investigated, especially in reference to damping of collisionless zonal flows. Results indicate that nonlinear excitation of this structure can be potentially important in enhancing anomalous transport as well as in damping zonal flows

Critical heat flux and flow instability in an advanced light water reactor

International Nuclear Information System (INIS)

Dae-Hyun Hwang; Kyong-Won Seo; Chung-Chan Lee; Sung-Kyun Zee

2005-01-01

Full text of publication follows: An advanced light water reactor concept has been continuously studied in KAERI with an output in the range of about 60 to 300 MW th . The reactor is purposed to be utilized as an energy source for seawater desalination as well as small scale power generation. In order to achieve the intrinsic safety and enhanced operational flexibility, some specific design considerations such as low power density and soluble boron free operation have been incorporated in the multiple-parallel-channel type reactor core. The low power density can be achieved by adopting fuel assemblies with tightly spaced non-square lattice rod array. The allowable core operating region should be primarily limited by the two design parameters; the critical heat flux(CHF) and the flow instabilities in the multiple parallel fuel assembly channels. The characteristics of CHF and flow instability have been investigated through experimental and analytical works. The CHF prediction model was established on the basis of experimental data obtained from 19-rod test bundles. The CHF experiments have been conducted for various test bundles with different heated lengths, uniform and non-uniform radial and axial power distributions, water and Freon as the working fluids, and different number of unheated rods. The parametric ranges of CHF experiments covers the pressure from 6 to 18 MPa, the mass flux from 150 to 2000 kg/m 2 /s, and the inlet subcooling from 10 to 120 deg. C. The flow instabilities due to density wave oscillations were investigated by conducting experiments with two parallel channels under the pressure ranges from 6 to 16 MPa. The parametric behavior of flow instability was examined for the test sections with different lengths of adiabatic risers, different axial power shapes, different inlet restrictions, and different channel cross sections. The stability boundary was experimentally determined by increasing channel inlet temperature or reducing the flow rate

A New Approach for Accurate Prediction of Liquid Loading of Directional Gas Wells in Transition Flow or Turbulent Flow

Directory of Open Access Journals (Sweden)

Ruiqing Ming

2017-01-01

Full Text Available Current common models for calculating continuous liquid-carrying critical gas velocity are established based on vertical wells and laminar flow without considering the influence of deviation angle and Reynolds number on liquid-carrying. With the increase of the directional well in transition flow or turbulent flow, the current common models cannot accurately predict the critical gas velocity of these wells. So we built a new model to predict continuous liquid-carrying critical gas velocity for directional well in transition flow or turbulent flow. It is shown from sensitivity analysis that the correction coefficient is mainly influenced by Reynolds number and deviation angle. With the increase of Reynolds number, the critical liquid-carrying gas velocity increases first and then decreases. And with the increase of deviation angle, the critical liquid-carrying gas velocity gradually decreases. It is indicated from the case calculation analysis that the calculation error of this new model is less than 10%, where accuracy is much higher than those of current common models. It is demonstrated that the continuous liquid-carrying critical gas velocity of directional well in transition flow or turbulent flow can be predicted accurately by using this new model.

Flow Driven by an Archimedean Helical Permanent Magnetic Field. Part I: Flow Patterns and Their Transitions

Science.gov (United States)

Wang, Bo; Wang, Xiaodong; Etay, Jacqueline; Na, Xianzhao; Zhang, Xinde; Fautrelle, Yves

2016-04-01

In this study, an Archimedean helical permanent magnetic field was constructed and its driving effects on liquid metal were examined. A magnetic stirrer was constructed using a series of arc-like magnets. The helical distribution of its magnetic field, which was confirmed via Gauss probe measurements and numerical simulations, can be considered a combination of rotating and traveling magnetic fields. The characteristics of the flow patterns, particularly the transitions between the meridian secondary flow (two vortices) and the global axial flow (one vortex), driven by this magnetic field were quantitatively measured using ultrasonic Doppler velocimetry. The transient and modulated flow behaviors will be presented in a companion article. The D/ H dimension ratio was used to characterize the transitions of these two flow patterns. The results demonstrated that the flow patterns depend on not only the intrinsic structure of the magnetic field, e.g., the helix lead angle, but also the performance parameters, e.g., the dimensional ratio of the liquid bulk. The notable opposing roles of these two flow patterns in the improvement of macrosegregations when imposing such magnetic fields near the solidifying front were qualitatively addressed.

Removal of natural organic matter and arsenic from water by electrocoagulation/flotation continuous flow reactor

International Nuclear Information System (INIS)

Mohora, Emilijan; Rončević, Srdjan; Dalmacija, Božo; Agbaba, Jasmina; Watson, Malcolm; Karlović, Elvira; Dalmacija, Milena

2012-01-01

Highlights: ► A continuous electrocoagulation/flotation reactor was designed built and operated. ► Highest NOM removal according to UV 254 was 77% relative to raw groundwater. ► Highest NOM removal accordance to DOC was 71%, relative to raw groundwater. ► Highest As removal archived was 85% (6.2 μg/l), relative to raw groundwater. ► Specific reactor energy and electrode consumption was 1.7 kWh/m 3 and 66 g Al/m 3 . - Abstract: The performance of the laboratory scale electrocoagulation/flotation (ECF) reactor in removing high concentrations of natural organic matter (NOM) and arsenic from groundwater was analyzed in this study. An ECF reactor with bipolar plate aluminum electrodes was operated in the horizontal continuous flow mode. Electrochemical and flow variables were optimized to examine ECF reactor contaminants removal efficiency. The optimum conditions for the process were identified as groundwater initial pH 5, flow rate = 4.3 l/h, inter electrode distance = 2.8 cm, current density = 5.78 mA/cm 2 , A/V ratio = 0.248 cm −1 . The NOM removal according to UV 254 absorbance and dissolved organic matter (DOC) reached highest values of 77% and 71% respectively, relative to the raw groundwater. Arsenic removal was 85% (6.2 μg As/l) relative to raw groundwater, satisfying the drinking water standards. The specific reactor electrical energy consumption was 17.5 kWh/kg Al. The specific aluminum electrode consumption was 66 g Al/m 3 . According to the obtained results, ECF in horizontal continuous flow mode is an energy efficient process to remove NOM and arsenic from groundwater.

Biological nitrogen and carbon removal in a gravity flow biomass concentrator reactor for municipal sewage treatment.

Science.gov (United States)

Scott, Daniel; Hidaka, Taira; Campo, Pablo; Kleiner, Eric; Suidan, Makram T; Venosa, Albert D

2013-01-01

A novel membrane system, the Biomass Concentrator Reactor (BCR), was evaluated as an alternative technology for the treatment of municipal wastewater. Because the BCR is equipped with a membrane whose average poresize is 20 μm (18-28 μm), the reactor requires low-pressure differential to operate (gravity). The effectiveness of this system was evaluated for the removal of carbon and nitrogen using two identical BCRs, identified as conventional and hybrid, that were operated in parallel. The conventional reactor was operated under full aerobic conditions (i.e., organic carbon and ammonia oxidation), while the hybrid reactor incorporated an anoxic zone for nitrate reduction as well as an aerobic zone for organic carbon and ammonia oxidation. Both reactors were fed synthetic wastewater at a flow rate of 71 L d(-1), which resulted in a hydraulic retention time of 9 h. In the case of the hybrid reactor, the recycle flow from the aerobic zone to the anoxic zone was twice the feed flow rate. Reactor performance was evaluated under two solids retention times (6 and 15 d). Under these conditions, the BCRs achieved nearly 100% mixed liquor solids separation with a hydraulic head differential of less than 2.5 cm. The COD removal efficiency was over 90%. Essentially complete nitrification was achieved in both systems, and nitrogen removal in the hybrid reactor was close to the expected value (67%). Copyright © 2012 Elsevier Ltd. All rights reserved.

Hydrogen/Oxygen Reactions at High Pressures and Intermediate Temperatures: Flow Reactor Experiments and Kinetic Modeling

DEFF Research Database (Denmark)

Hashemi, Hamid; Christensen, Jakob Munkholt; Glarborg, Peter

A series of experimental and numerical investigations into hydrogen oxidation at high pressures and intermediate temperatures has been conducted. The experiments were carried out in a high pressure laminar flow reactor at 50 bar pressure and a temperature range of 600–900 K. The equivalence ratio......, the mechanism is used to simulate published data on ignition delay time and laminar burning velocity of hydrogen. The flow reactor results show that at reducing, stoichiometric, and oxidizing conditions, conversion starts at temperatures of 750–775 K, 800–825 K, and 800–825 K, respectively. In oxygen atmosphere......, ignition occurs at the temperature of 775–800 K. In general, the present model provides a good agreement with the measurements in the flow reactor and with recent data on laminar burning velocity and ignition delay time....

A numerical study of boiling flow instability of a reactor thermosyphon system

International Nuclear Information System (INIS)

Nayak, A.K.; Lathouwers, D.; Hagen, T.H.J.J. van der; Schrauwen, Frans; Molenaar, Peter; Rogers, Andrew

2006-01-01

A numerical study has been carried out to investigate the boiling flow instability of a reactor thermosyphon system. The numerical model solves the conservation equations of mass, momentum and energy applicable to a two-fluid and three-field steam-water system using a finite difference technique. The computer code MONA was used for this purpose. The code was applied to the thermosyphon system of an EO (ethylene oxide) chemical reactor in which the heat released by a catalytic reaction is carried by boiling water under natural circulation conditions. The steady-state characteristics of the reactor thermosyphon system were predicted using the MONA code and conventional two-phase flow models in order to understand the model applicability for this type of thermosyphon system. The two-fluid model was found to predict the flow closest to the measured value of the plant. The stability behaviour of the thermosyphon system was investigated for a wide range of operating conditions. The effects of power, subcooling, riser length and riser diameter on the boiling flow instability were determined. The system was found to be unstable at higher power conditions which is typical for a Type II instability. However, with an increase in riser diameter, oscillations at low power were observed as well. These are classified as Type I instabilities. Stability maps were predicted for both Type I and Type II instabilities. Methods of improving the stability of the system are discussed

A numerical study of boiling flow instability of a reactor thermosyphon system

Energy Technology Data Exchange (ETDEWEB)

Nayak, A.K.; Lathouwers, D.; Hagen, T.H.J.J. van der [Interfaculty Reactor Institute, Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands); Schrauwen, Frans; Molenaar, Peter; Rogers, Andrew [Shell Research and Technology Centre, Badhuisweg 3, 1031 CM Amsterdam (Netherlands)

2006-04-01

A numerical study has been carried out to investigate the boiling flow instability of a reactor thermosyphon system. The numerical model solves the conservation equations of mass, momentum and energy applicable to a two-fluid and three-field steam-water system using a finite difference technique. The computer code MONA was used for this purpose. The code was applied to the thermosyphon system of an EO (ethylene oxide) chemical reactor in which the heat released by a catalytic reaction is carried by boiling water under natural circulation conditions. The steady-state characteristics of the reactor thermosyphon system were predicted using the MONA code and conventional two-phase flow models in order to understand the model applicability for this type of thermosyphon system. The two-fluid model was found to predict the flow closest to the measured value of the plant. The stability behaviour of the thermosyphon system was investigated for a wide range of operating conditions. The effects of power, subcooling, riser length and riser diameter on the boiling flow instability were determined. The system was found to be unstable at higher power conditions which is typical for a Type II instability. However, with an increase in riser diameter, oscillations at low power were observed as well. These are classified as Type I instabilities. Stability maps were predicted for both Type I and Type II instabilities. Methods of improving the stability of the system are discussed. [Author].

Optimal conditions in direct dimethyl ether synthesis from syngas utilizing a dual-type fluidized bed reactor

International Nuclear Information System (INIS)

Yousefi, Ahmad; Eslamloueyan, Reza; Kazerooni, Nooshin Moradi

2017-01-01

Concerns over environmental pollution and ever-increasing energy demand have urged the global community to tap clean-burning fuels among which dimethyl ether is a promising candidate for contribution in the transportation sector. Direct dimethyl ether synthesis from syngas, in which methanol production and dehydration take place simultaneously, is arguably the preferred route for large scale production. In this study, direct dimethyl ether synthesis is proposed in an industrial dual-type fluidized bed reactor. This configuration involves two fluidized bed reactors operating in different conditions. In the first catalytic reactor (water-cooled reactor), the synthesis gas is partly converted to methanol after being preheated by the reaction heat in the second reactor (gas-cooled reactor). A two-phase generalized comprehensive reactor model, comprised of the flow in three different regimes is applied and a smooth transition between flow regimes is provided based on the probabilistic averaging approach. The optimal operating conditions are sought by employing differential evolution algorithm as a robust optimization strategy. The dimethyl ether mole fraction is considered as the objective function during the optimization. The results show considerable dimethyl ether enhancement by 16% and 14% compared to the conventional direct dimethyl ether synthesis reactor and dual-type fixed bed dimethyl ether reactor arrangements, respectively. - Highlights: • Dual-type catalytic fluidized bed reactors for dimethyl ether synthesis is studied. • A two-phase comprehensive model comprised of flow in three regimes is used. • Probabilistic averaging approach is applied for smooth transitions between regimes. • Differential evolution method is employed to determine optimal operating conditions. • Production capacity is remarkably enhanced compared to conventional reactor.

Hydrodynamically induced dryout and post dryout important to heavy water reactors: A yearly progress report

International Nuclear Information System (INIS)

Ishii, M.; Revankar, S.T.; Babelli, I.; Lele, S.

1992-06-01

Recently, the safety of low pressure liquid cooled nuclear reactors has become a very important issue with reference to the operation of the heavy water reactors at Savannah River Plant. Under accident conditions such as loss-of-flow or loss-of-coolant, these reactors typically encounter unstable two-phase flow which may lead to the occurrence of dryout and subsequent fuel failure. An analytical study using the one-dimensional drift flux model was carried out to investigate the two-phase flow instability for Westinghouse Savannah River Site reactor. The analysis indicates that the first and higher order instabilities exist in the possible transient operational conditions. The instabilities are encountered at higher heat fluxes or lower flow rates. The subcooling has a stabilizing effect except at very low subcooling. An experimental loop has been designed and constructed to study the CBF induced by various flow instabilities. Details of this test loop are presented. Initial test results have been presented. The two-phase flow regimes and hydrodynamic behaviors in the post dryout region have been studied under propagating rewetting conditions. The effect of subcooling and inlet velocity on flow transition as well as on the quench front propagation was investigated. The test liquid was Freon 113 which was introduced into the bottom of the quartz test section whose walls were maintained well above the film boiling temperature of the test liquid, via a transparent heat transfer fluid. The flow regimes observed down stream of the upward moving quench front were the rough wavy, the agitated, and the dispersed droplet/ligaments. A correlation for the flow regime transition between the inverted annular and the dispersed droplet/ligament flow patterns was developed. The correlation showed a marked dependence on the void fraction at the CBF location and hence on the flow regime encountered in the pre-CBF region

Computational Fluid Dynamics simulation of hydrothermal liquefaction of microalgae in a continuous plug-flow reactor.

Science.gov (United States)

Ranganathan, Panneerselvam; Savithri, Sivaraman

2018-06-01

Computational Fluid Dynamics (CFD) technique is used in this work to simulate the hydrothermal liquefaction of Nannochloropsis sp. microalgae in a lab-scale continuous plug-flow reactor to understand the fluid dynamics, heat transfer, and reaction kinetics in a HTL reactor under hydrothermal condition. The temperature profile in the reactor and the yield of HTL products from the present simulation are obtained and they are validated with the experimental data available in the literature. Furthermore, the parametric study is carried out to study the effect of slurry flow rate, reactor temperature, and external heat transfer coefficient on the yield of products. Though the model predictions are satisfactory in comparison with the experimental results, it still needs to be improved for better prediction of the product yields. This improved model will be considered as a baseline for design and scale-up of large-scale HTL reactor. Copyright © 2018 Elsevier Ltd. All rights reserved.

Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

International Nuclear Information System (INIS)

Aji, Indarta Kuncoro; Waris, Abdul; Permana, Sidik

2015-01-01

Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF 2 -ThF 4 - 233 UF 4 respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155

Investigation of flow dynamics of liquid phase in a pilot-scale trickle bed reactor using radiotracer technique

International Nuclear Information System (INIS)

Pant, H.J.; Sharma, V.K.

2016-01-01

A radiotracer investigation was carried out to measure residence time distribution (RTD) of liquid phase in a trickle bed reactor (TBR). The main objectives of the investigation were to investigate radial and axial mixing of the liquid phase, and evaluate performance of the liquid distributor/redistributor at different operating conditions. Mean residence times (MRTs), holdups (H) and fraction of flow flowing along different quadrants were estimated. The analysis of the measured RTD curves indicated radial non-uniform distribution of liquid phase across the beds. The overall RTD of the liquid phase, measured at the exit of the reactor was simulated using a multi-parameter axial dispersion with exchange model (ADEM), and model parameters were obtained. The results of model simulations indicated that the TBR behaved as a plug flow reactor at most of the operating conditions used in the investigation. The results of the investigation helped to improve the existing design as well as to design a full-scale industrial TBR for petroleum refining applications. - Highlights: • Residence time distributions of liquid phase were measured in a trickle bed reactor. • Bromine-82 as ammonium bromide was used as a radiotracer. • Mean residence times, holdups and radial distribution of liquid phase were quantified. • Axial dispersion with exchange model was used to simulate the measured data. • The trickle bed reactor behaved as a plug flow reactor.

Liquid-Gas-Like Phase Transition in Sand Flow Under Microgravity

Science.gov (United States)

Huang, Yu; Zhu, Chongqiang; Xiang, Xiang; Mao, Wuwei

2015-06-01

In previous studies of granular flow, it has been found that gravity plays a compacting role, causing convection and stratification by density. However, there is a lack of research and analysis of the characteristics of different particles' motion under normal gravity contrary to microgravity. In this paper, we conduct model experiments on sand flow using a model test system based on a drop tower under microgravity, within which the characteristics and development processes of granular flow under microgravity are captured by high-speed cameras. The configurations of granular flow are simulated using a modified MPS (moving particle simulation), which is a mesh-free, pure Lagrangian method. Moreover, liquid-gas-like phase transitions in the sand flow under microgravity, including the transitions to "escaped", "jumping", and "scattered" particles are highlighted, and their effects on the weakening of shear resistance, enhancement of fluidization, and changes in particle-wall and particle-particle contact mode are analyzed. This study could help explain the surface geology evolution of small solar bodies and elucidate the nature of granular interaction.

Markov transition probability-based network from time series for characterizing experimental two-phase flow

International Nuclear Information System (INIS)

Gao Zhong-Ke; Hu Li-Dan; Jin Ning-De

2013-01-01

We generate a directed weighted complex network by a method based on Markov transition probability to represent an experimental two-phase flow. We first systematically carry out gas—liquid two-phase flow experiments for measuring the time series of flow signals. Then we construct directed weighted complex networks from various time series in terms of a network generation method based on Markov transition probability. We find that the generated network inherits the main features of the time series in the network structure. In particular, the networks from time series with different dynamics exhibit distinct topological properties. Finally, we construct two-phase flow directed weighted networks from experimental signals and associate the dynamic behavior of gas-liquid two-phase flow with the topological statistics of the generated networks. The results suggest that the topological statistics of two-phase flow networks allow quantitative characterization of the dynamic flow behavior in the transitions among different gas—liquid flow patterns. (general)

A reverse flow catalytic membrane reactor for the production of syngas: an experimental study

NARCIS (Netherlands)

Smit, J.; Bekink, G.J.; van Sint Annaland, M.; Kuipers, J.A.M.

2005-01-01

In this paper experimental results are presented for a demonstration unit of a recently proposed novel integrated reactor concept (Smit et. al., 2005) for the partial oxidation of natural gas to syngas (POM), namely a Reverse Flow Catalytic Membrane Reactor (RFCMR). Natural gas has great potential

Flow instability tests for a particle bed reactor nuclear thermal rocket fuel element

Science.gov (United States)

Lawrence, Timothy J.

1993-05-01

Recent analyses have focused on the flow stability characteristics of a particle bed reactor (PBR). These laminar flow instabilities may exist in reactors with parallel paths and are caused by the heating of the gas at low Reynolds numbers. This phenomena can be described as follows: several parallel channels are connected at the plenum regions and are stabilized by some inlet temperature and pressure; a perturbation in one channel causes the temperature to rise and increases the gas viscosity and reduces the gas density; the pressure drop is fixed by the plenum regions, therefore, the mass flow rate in the channel would decrease; the decrease in flow reduces the ability to remove the energy added and the temperature increases; and finally, this process could continue until the fuel element fails. Several analyses based on different methods have derived similar curves to show that these instabilities may exist at low Reynolds numbers and high phi's ((Tfinal Tinitial)/Tinitial). These analyses need to be experimentally verified.

Parametric study of natural circulation flow in molten salt fuel in molten salt reactor

Energy Technology Data Exchange (ETDEWEB)

Pauzi, Anas Muhamad, E-mail: Anas@uniten.edu.my [Centre of Nuclear Energy, Universiti Tenaga Nasional (UNITEN), Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia); Cioncolini, Andrea; Iacovides, Hector [School of Mechanical, Aerospace, and Civil Engineering (MACE), University of Manchester, Oxford Road, M13 9PL Manchester (United Kingdom)

2015-04-29

The Molten Salt Reactor (MSR) is one of the most promising system proposed by Generation IV Forum (GIF) for future nuclear reactor systems. Advantages of the MSR are significantly larger compared to other reactor system, and is mainly achieved from its liquid nature of fuel and coolant. Further improvement to this system, which is a natural circulating molten fuel salt inside its tube in the reactor core is proposed, to achieve advantages of reducing and simplifying the MSR design proposed by GIF. Thermal hydraulic analysis on the proposed system was completed using a commercial computation fluid dynamics (CFD) software called FLUENT by ANSYS Inc. An understanding on theory behind this unique natural circulation flow inside the tube caused by fission heat generated in molten fuel salt and tube cooling was briefly introduced. Currently, no commercial CFD software could perfectly simulate natural circulation flow, hence, modeling this flow problem in FLUENT is introduced and analyzed to obtain best simulation results. Results obtained demonstrate the existence of periodical transient nature of flow problem, hence improvements in tube design is proposed based on the analysis on temperature and velocity profile. Results show that the proposed system could operate at up to 750MW core power, given that turbulence are enhanced throughout flow region, and precise molten fuel salt physical properties could be defined. At the request of the authors and the Proceedings Editor the name of the co-author Andrea Cioncolini was corrected from Andrea Coincolini. The same name correction was made in the Acknowledgement section on page 030004-10 and in reference number 4. The updated article was published on 11 May 2015.

The gas-solid trickle-flow reactor for the catalytic oxidation of hydrogen sulphide: a trickle-phase model

NARCIS (Netherlands)

Verver, A.B.; van Swaaij, Willibrordus Petrus Maria

1987-01-01

The oxidation of H2S by O2 producing elemental sulphur has been studied at temperatures of 100–300°C and at atmospheric pressure in a laboratory-scale gas-solid trickle-flow reactor. In this reactor one of the reaction products, i.e. sulphur, is removed continuously by flowing solids. A porous,

The Effect of the Holes Size Change of Lower-Support-Structure-Bottom Plate on the Reactor Core-Inlet Flow-Distribution

International Nuclear Information System (INIS)

Lee, Gong Hee; Bang, Young Seok; Cheong, Ae Ju

2015-01-01

Complex thermal-hydraulic phenomena exist inside PWR because reactor interiors include a fuel assembly, control rod assembly, ICI (In-Core Instrumentation), and other internal structures. Because changes to reactor design may influence interior, thermal-hydraulic characteristics, licensing applicants commonly conduct a flow-distribution test and use test results (e.g., core-inlet flow-rate distribution) as the input data for a core thermal-margin analysis program. Because the APR+ (Advanced Power Reactor Plus) had more fuel assemblies (241EA → 257EA) and the design of some internal structures was changed (from those of APR1400), the core-inlet flow-rate distribution for a 1/5 scaled-down reactor model was measured and high flow-rates were found especially near the outer region of the reactor core. In this study, to examine the effect of the holes size change (i.e. smaller diameter) in the outer region of the LSSBP, not a 50% blockage of the flow holes, on the reactor core-inlet flow-distribution, simulations were conducted with the commercial CFD (Computational Fluid Dynamics) software, ANSYS CFX R.14. The predicted results were compared with those of the original LSSBP. In this study, to examine the effect of the holes size change (smaller diameter) in the outer region of the LSSBP on the reactor core-inlet flow-distribution, simulations were conducted with the commercial CFD software, ANSYS CFX R.14. The predicted results were compared with those of the original LSSBP. Through these comparisons it was concluded that a more uniform distribution of the mass-flow rate at the core-inlet plane could be obtained by reducing the holes size in the outer region of the LSSBP

Experimental analysis of upward vertical two-phase flow in four-cusp channels simulating the conditions of a typical nuclear reactor channel, degraded by a loss of coolant accident

International Nuclear Information System (INIS)

Assad, A.C.A.

1984-01-01

The present work deals with an experimental analysis of upward vertical two-phase flow in channels with circular and four-cusp cross-sections. The latter simulates the conditions of a typical nuclear reactor channel, degraded by a loss of coolant accident. Simultaneous flow of air and water has been employed to simulate adiabatic steam-water flow. The installation of air-water separators helped eliminate instabilities during pressure-drop measurements. The gamma ray attenuation was utilized for the void fraction determination. For the four-cusp geommetry, new criteria for two-phase flow regime transitions have been determined, as well as new correlatins for pressure drop and void fraction, as function of the Lockhart-Martinelli factor and vapour mass-fraction, respectively. (Author) [pt

CFD Modeling of Flow and Ion Exchange Kinetics in a Rotating Bed Reactor System

DEFF Research Database (Denmark)

Larsson, Hilde Kristina; Schjøtt Andersen, Patrick Alexander; Byström, Emil

2017-01-01

A rotating bed reactor (RBR) has been modeled using computational fluid dynamics (CFD). The flow pattern in the RBR was investigated and the flow through the porous material in it was quantified. A simplified geometry representing the more complex RBR geometry was introduced and the simplified...... model was able to reproduce the main characteristics of the flow. Alternating reactor shapes were investigated, and it was concluded that the use of baffles has a very large impact on the flows through the porous material. The simulations suggested, therefore, that even faster reaction rates could...... be achieved by making the baffles deeper. Two-phase simulations were performed, which managed to reproduce the deflection of the gas–liquid interface in an unbaffled system. A chemical reaction was implemented in the model, describing the ion-exchange phenomena in the porous material using four different...

Investigation of cascade-typed falling liquid film flow along first wall of laser-fusion reactor

International Nuclear Information System (INIS)

Kunugi, Tomoaki; Nakai, Tadakatsu; Kawara, Zensaku

2007-01-01

To protect from high energy/particle fluxes caused by nuclear fusion reaction such as extremely high heat flux, X rays, Alpha particles and fuel debris to a first wall of an inertia fusion reactor, a ''cascade-typed'' falling liquid film flow is proposed as the ''liquid wall'' concept which is one of the reactor chamber cooling and wall protection schemes: the reactor chamber can protect by using a liquid metal film flow (such as Li 17 Pb 83 ) over the wall. In order to investigate the feasibility of this concept, we conducted the numerical analyses by using the commercial code (STREAM: unsteady three-dimensional general purpose thermofluid code) and also conducted the flow visualization experiments. The numerical results suggested that the cascade structure design should be improved, so that we redesigned the cascade-typed first wall and performed the flow visualization as a POP (proof-of-principle) experiment. In the numerical analyses, the water is used as the working liquid and an acrylic plate as the wall. These selections are based on two reasons: (1) from the non-dimensional analysis approach, the Weber number (We=ru 2 d/s: r is density, u is velocity, d is film thickness, s is surface tension coefficient) should be the same between the design (Li 17 Pb 83 flow) and the model experiment (water flow) because of the free-surface instability, (2) the SiC/SiC composite would be used as the wall material, so that the wall may have the less wettability: the acrylic plate has the similar feature. The redesigned cascade-typed first wall for one step (30 cm height corresponding to 4 Hz laser duration) consists of a liquid tank having a free-surface for keeping the constant waterhead located at the backside of the first wall, and connects to a slit which is composed of two plates: one plate is the first wall, and the other is maintaining the liquid level. This design solved the trouble of the previous design. The test section for the flow visualization has the same

Removal of natural organic matter and arsenic from water by electrocoagulation/flotation continuous flow reactor

Energy Technology Data Exchange (ETDEWEB)

Mohora, Emilijan, E-mail: emohora@ifc.org [University of Novi Sad Faculty of Sciences, Department of Chemistry, Biochemistry and Environmental Protection, Trg D. Obradovica 3, 21000 Novi Sad (Serbia); Roncevic, Srdjan; Dalmacija, Bozo; Agbaba, Jasmina; Watson, Malcolm; Karlovic, Elvira; Dalmacija, Milena [University of Novi Sad Faculty of Sciences, Department of Chemistry, Biochemistry and Environmental Protection, Trg D. Obradovica 3, 21000 Novi Sad (Serbia)

2012-10-15

Highlights: Black-Right-Pointing-Pointer A continuous electrocoagulation/flotation reactor was designed built and operated. Black-Right-Pointing-Pointer Highest NOM removal according to UV{sub 254} was 77% relative to raw groundwater. Black-Right-Pointing-Pointer Highest NOM removal accordance to DOC was 71%, relative to raw groundwater. Black-Right-Pointing-Pointer Highest As removal archived was 85% (6.2 {mu}g/l), relative to raw groundwater. Black-Right-Pointing-Pointer Specific reactor energy and electrode consumption was 1.7 kWh/m{sup 3} and 66 g Al/m{sup 3}. - Abstract: The performance of the laboratory scale electrocoagulation/flotation (ECF) reactor in removing high concentrations of natural organic matter (NOM) and arsenic from groundwater was analyzed in this study. An ECF reactor with bipolar plate aluminum electrodes was operated in the horizontal continuous flow mode. Electrochemical and flow variables were optimized to examine ECF reactor contaminants removal efficiency. The optimum conditions for the process were identified as groundwater initial pH 5, flow rate = 4.3 l/h, inter electrode distance = 2.8 cm, current density = 5.78 mA/cm{sup 2}, A/V ratio = 0.248 cm{sup -1}. The NOM removal according to UV{sub 254} absorbance and dissolved organic matter (DOC) reached highest values of 77% and 71% respectively, relative to the raw groundwater. Arsenic removal was 85% (6.2 {mu}g As/l) relative to raw groundwater, satisfying the drinking water standards. The specific reactor electrical energy consumption was 17.5 kWh/kg Al. The specific aluminum electrode consumption was 66 g Al/m{sup 3}. According to the obtained results, ECF in horizontal continuous flow mode is an energy efficient process to remove NOM and arsenic from groundwater.

Flow Through a Rectangular-to-Semiannular Diffusing Transition Duct

Science.gov (United States)

Foster, Jeff; Wendt, Bruce J.; Reichert, Bruce A.; Okiishi, Theodore H.

1997-01-01

Rectangular-to-semiannular diffusing transition ducts are critical inlet components on supersonic airplanes having bifucated engine inlets. This paper documents measured details of the flow through a rectangular-to-semiannular transition duct having an expansion area ratio of 1.53. Three-dimensional velocity vectors and total pressures at the exit plane of the diffuser are presented. Surface oil-flow visualization and surface static pressure data are shown. The tests were conducted with an inlet Mach number of 0.786 and a Reynolds number based on the inlet centerline velocity and exit diameter of 3.2 x 10(exp 6). The measured data are compared with previously published computational results. The ability of vortex generators to reduce circumferential total pressure distortion is demonstrated.

Numerical simulation study on the air/water countercurrent flow limitation in nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Morghi, Youssef; Mesquita, Amir Z., E-mail: ssfmorghi@gmail.com, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Puente, Jesus, E-mail: jpuente720@gmail.com [Centro Federal de Educaçao Tecnologica Celso Suckowda Fonseca (CEFET), Angra dos Reis, RJ (Brazil); Baliza, Ana R., E-mail: baliza@eletronuclear.gov.br [Eletrobras Eletronuclear Angra dos Reis, RJ (Brazil)

2017-07-01

After a loss-of-coolant accident (LOCA) in a Pressurized Water Reactor (PWR), the temperature of the fuel elements cladding increases dramatically due to the heat produced by the fission products decay, which is not adequately removed by the vapor contained in the core. In order to avoid this sharp rise in temperature and consequent melting of the core, the Emergency Core Cooling System is activated. This system initially injects borated water from accumulator tanks of the reactor through the inlet pipe (cold leg) and the outlet pipe (hot leg), or through the cold leg only, depending on the plant manufacturer. Some manufacturers add to this, direct injection into the upper plenum of the reactor. The penetration of water into the reactor core is a complex thermo fluid dynamic process because it involves the mixing of water with the vapor contained in the reactor, added to that generated in the contact of the water with the still hot surfaces in various geometries. In some critical locations, the vapor flowing in the opposite direction of the water can control the penetration of this into the core. This phenomenon is known as Countercurrent Flow Limitation (CCFL) or Flooding, and it is characterized by the control that a gas exerts in the liquid flow in the opposite direction. This work presents a proposal to use a CFD to simulate the CCFL phenomenon. Numerical computing can provide important information and data that is difficult or expensive to measure or test experimentally. Given the importance of computational science today, it can be considered a third and independent branch of science on an equal footing with the theoretical and experimental sciences. (author)

Numerical modeling of turbulent swirling flow in a multi-inlet vortex nanoprecipitation reactor using dynamic DDES

Science.gov (United States)

Hill, James C.; Liu, Zhenping; Fox, Rodney O.; Passalacqua, Alberto; Olsen, Michael G.

2015-11-01

The multi-inlet vortex reactor (MIVR) has been developed to provide a platform for rapid mixing in the application of flash nanoprecipitation (FNP) for manufacturing functional nanoparticles. Unfortunately, commonly used RANS methods are unable to accurately model this complex swirling flow. Large eddy simulations have also been problematic, as expensive fine grids to accurately model the flow are required. These dilemmas led to the strategy of applying a Delayed Detached Eddy Simulation (DDES) method to the vortex reactor. In the current work, the turbulent swirling flow inside a scaled-up MIVR has been investigated by using a dynamic DDES model. In the DDES model, the eddy viscosity has a form similar to the Smagorinsky sub-grid viscosity in LES and allows the implementation of a dynamic procedure to determine its coefficient. The complex recirculating back flow near the reactor center has been successfully captured by using this dynamic DDES model. Moreover, the simulation results are found to agree with experimental data for mean velocity and Reynolds stresses.

Numerical methods and transition investigation of transient flows around a pitching hydrofoil

International Nuclear Information System (INIS)

Wu, Q; Wang, G Y; Huang, B

2013-01-01

The numerical simulations for a NACA66 hydrofoil are performed by using the standard k-ω SST turbulence model and revised γ-Re θ transition model respectively. The simulation results are compared with the experimental results, and the hydrodynamic property and the fluid structure during the pitching process is studied. It is revealed that, compared with the standard k-ω SST turbulence model, the revised γ-Re θ transition model is able to present the hydrodynamic property and the fluid structure of the transient flow around a pitching hydrofoil more accurately, and better predict the separation and transition process in the boundary layer. The transient flow process around a pitching hydrofoil can be divided into 5 parts. At small angle of attack, transition is observed at the leading edge of the foil, resulting in the inflection of dynamic property curves. As the angle of attack increases, a clockwise trailing edge vortex expands toward the leading edge of the foil. At high angles of attack, large-scale load fluctuations are observed due to the stall caused by separation of the leading edge vortex. The flow transitions back to laminar during the downward pitching process

FLODIS: a computer model to determine the flow distribution and thermal response of the Fort St. Vrain reactor

Energy Technology Data Exchange (ETDEWEB)

Paul, D.D.

1976-06-01

FLODIS is a combined heat transfer and fluid flow analysis calculation written specifically for the core of the Fort St. Vrain reactor. It is a lumped-node representation of the 37 refueling regions in the active core. Heat conduction to the coolant and in the axial direction is represented; however, the effect of conduction between refueling regions is not included. The calculation uses the specified operating conditions for the reactor at power to determine appropriate loss coefficients for the variable orifices in each refueling region. Flow distributions following reactor trip and a reduction in coolant pressure and flow are determined assuming that the orifice coefficients remain constant. Iterative techniques are used to determine the distribution of coolant flow as a function of time during the transient. Results are presented for the evaluation of the transient for the Fort St. Vrain reactor following depressurization and cooling with two circulators operating at 8000 rpm.

FLODIS: a computer model to determine the flow distribution and thermal response of the Fort St. Vrain reactor

International Nuclear Information System (INIS)

Paul, D.D.

1976-06-01

FLODIS is a combined heat transfer and fluid flow analysis calculation written specifically for the core of the Fort St. Vrain reactor. It is a lumped-node representation of the 37 refueling regions in the active core. Heat conduction to the coolant and in the axial direction is represented; however, the effect of conduction between refueling regions is not included. The calculation uses the specified operating conditions for the reactor at power to determine appropriate loss coefficients for the variable orifices in each refueling region. Flow distributions following reactor trip and a reduction in coolant pressure and flow are determined assuming that the orifice coefficients remain constant. Iterative techniques are used to determine the distribution of coolant flow as a function of time during the transient. Results are presented for the evaluation of the transient for the Fort St. Vrain reactor following depressurization and cooling with two circulators operating at 8000 rpm

Reactor water level control device

International Nuclear Information System (INIS)

Utagawa, Kazuyuki.

1993-01-01

A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

Reactive flow analysis with fluorine thermal dissociation in a FLUOREX flame reactor

International Nuclear Information System (INIS)

Ohtsuka, Masaya; Tagawa, Hisato; Sasahira, Akira; Hoshino, Kuniyoshi; Kawamura, Fumio; Homma, Shunji; Amano, Osamu

2004-01-01

A reactive flow analysis method for flame reactors of the FLUOREX (Hybrid Process of Fluoride Volatility and Solvent Extraction) method was been developed. Transport equations for UO 2 /PuO 2 mixed particles were formulated in the Lagrangian framework and several fluid/particles interactions were modeled using mass, momentum and energy exchanges through surface chemical reactions, forces and heat transfers. The coal combustion model was modified without devolatilization and the char burnout model was replaced by the UO 2 /PuO 2 fluorination model. Overall reaction rates were calculated using the combined model of the surface reaction rate and the diffusion rate of F2 and F. Fluid flows were modeled through incompressible flows using the k-ε turbulent model in the Euler framework. A cylindrical flame reactor (φ 80 mm x 500mm was analyzed where 99%UO 2 +1%PuO 2 mixed particles were injected with Ar and 5% excess F 2 flow. The average particle diameter was 4 μm and the flow rate was 300 g/h. The fluorination reaction of PuO 2 was limited through fluorine molecular reaction but was accelerated due to fluorine thermal dissociation. The simulated corresponded to the experimental result in that both UO 2 and PuO 2 were almost completely fluorinated. (author)

Development of the test facilities for the measurement of core flow and pressure distribution of SMART reactor

International Nuclear Information System (INIS)

Ko, Y.J.; Euh, D.J.; Youn, Y.J.; Chu, I.C.; Kwon, T.S.

2011-01-01

A design of SMART reactor has been developed, of which the primary system is composed of four internal circulation pumps, a core of 57 fuel assemblies, eight cassettes of steam generators, flow mixing head assemblies, and other internal structures. Since primary design features are very different from conventional reactors, the characteristics of flow and pressure distribution are expected to be different accordingly. In order to analyze the thermal margin and hydraulic design characteristics of SMART reactor, design quantification tests for flow and pressure distribution with a preservation of flow geometry are necessary. In the present study, the design feature of the test facility in order to investigate flow and pressure distribution, named “SCOP” is described. In order to preserve the flow distribution characteristics, the SCOP is linearly reduced with a scaling ratio of 1/5. The core flow rate of each fuel assembly is measured by a venturi meter attached in the lower part of the core simulator having a similarity of pressure drop for nominally scaled flow conditions. All the 57 core simulators and 8 S/G simulators are precisely calibrated in advance of assembling in test facilities. The major parameters in tests are pressures, differential pressures, and core flow distribution. (author)

Cyclic electron flow is redox-controlled but independent of state transition.

Science.gov (United States)

Takahashi, Hiroko; Clowez, Sophie; Wollman, Francis-André; Vallon, Olivier; Rappaport, Fabrice

2013-01-01

Photosynthesis is the biological process that feeds the biosphere with reduced carbon. The assimilation of CO2 requires the fine tuning of two co-existing functional modes: linear electron flow, which provides NADPH and ATP, and cyclic electron flow, which only sustains ATP synthesis. Although the importance of this fine tuning is appreciated, its mechanism remains equivocal. Here we show that cyclic electron flow as well as formation of supercomplexes, thought to contribute to the enhancement of cyclic electron flow, are promoted in reducing conditions with no correlation with the reorganization of the thylakoid membranes associated with the migration of antenna proteins towards Photosystems I or II, a process known as state transition. We show that cyclic electron flow is tuned by the redox power and this provides a mechanistic model applying to the entire green lineage including the vast majority of the cases in which state transition only involves a moderate fraction of the antenna.

Laminar/transition sweeping flow-mixing model for wire-wrapped LMFBR assemblies

International Nuclear Information System (INIS)

Burns, K.F.; Rohsenow, W.M.; Todreas, N.E.

1980-07-01

Recent interest in analyzing the thermal hydraulic characteristics of LMFBR assemblies operating in the mixed convection regime motivates the extension of the aforementioned turbulent sweeping flow model to low Reynolds number flows. The accuracy to which knowledge of the mixing parameters is required has not been well determined, due to the increased influence of conduction and buoyancy effects with respect to energy transport at low Reynolds numbers. This study represents a best estimate attempt to correlate the existing low Reynolds number sweeping flow data. The laminar/transition model which is presented is expected to be useful in anayzing mixed convection conditions. However, the justification for making additional improvemements is contingent upon two factors. First, the ability of the proposed laminar/transition model to predict additional low Reynolds number sweeping flow data for other geometries needs to be investigated. Secondly, the sensitivity of temperature predictions to uncertainties in the values of the sweeping flow parameters should be quantified

Performance Assessment of Turbulence Models for the Prediction of the Reactor Internal Flow in the Scale-down APR+

International Nuclear Information System (INIS)

Lee, Gonghee; Bang, Youngseok; Woo, Swengwoong; Kim, Dohyeong; Kang, Minku

2013-01-01

The types of errors in CFD simulation can be divided into the two main categories: numerical errors and model errors. Turbulence model is one of the important sources for model errors. In this study, in order to assess the prediction performance of Reynolds-averaged Navier-Stokes (RANS)-based two equations turbulence models for the analysis of flow distribution inside a 1/5 scale-down APR+, the simulation was conducted with the commercial CFD software, ANSYS CFX V. 14. In this study, in order to assess the prediction performance of turbulence models for the analysis of flow distribution inside a 1/5 scale-down APR+, the simulation was conducted with the commercial CFD software, ANSYS CFX V. 14. Both standard k-ε model and SST model predicted the similar flow pattern inside reactor. Therefore it was concluded that the prediction performance of both turbulence models was nearly same. Complex thermal-hydraulic characteristics exist inside reactor because the reactor internals consist of fuel assembly, control rod assembly, and the internal structures. Either flow distribution test for the scale-down reactor model or computational fluid dynamics (CFD) simulation have been conducted to understand these complex thermal-hydraulic features inside reactor

A model for a countercurrent gas—solid—solid trickle flow reactor for equilibrium reactions. The methanol synthesis

NARCIS (Netherlands)

Westerterp, K.R.; Kuczynski, M.

1987-01-01

The theoretical background for a novel, countercurrent gas—solid—solid trickle flow reactor for equilibrium gas reactions is presented. A one-dimensional, steady-state reactor model is developed. The influence of the various process parameters on the reactor performance is discussed. The physical

Transient performance of flow in circuits of PWR type reactors

International Nuclear Information System (INIS)

Hirdes, V.R.; Carajilescov, P.

1988-09-01

Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which could cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

Flow Orientation Analysis for Major Activity Regions Based on Smart Card Transit Data

Directory of Open Access Journals (Sweden)

Parul Singh

2017-10-01

Full Text Available Analyzing public movement in transportation networks in a city is significant in understanding the life of citizen and making improved city plans for the future. This study focuses on investigating the flow orientation of major activity regions based on smart card transit data. The flow orientation based on the real movements such as transit data can provide the easiest way of understanding public movement in the complicated transportation networks. First, high inflow regions (HIRs are identified from transit data for morning and evening peak hours. The morning and evening HIRs are used to represent major activity regions for major daytime activities and residential areas, respectively. Second, the directional orientation of flow is then derived through the directional inflow vectors of the HIRs to show the bias in directional orientation and compare flow orientation among major activity regions. Finally, clustering analysis for HIRs is applied to capture the main patterns of flow orientations in the city and visualize the patterns on the map. The proposed methodology was illustrated with smart card transit data of bus and subway transportation networks in Seoul, Korea. Some remarkable patterns in the distribution of movements and orientations were found inside the city. The proposed methodology is useful since it unfolds the complexity and makes it easy to understand the main movement patterns in terms of flow orientation.

Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

Energy Technology Data Exchange (ETDEWEB)

Aji, Indarta Kuncoro, E-mail: indartaaji@s.itb.ac.id [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia); Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Permana, Sidik [Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia)

2015-09-30

Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.

Flexible fuel cycle initiative for the transition period from current reactors to next generation reactors

International Nuclear Information System (INIS)

Yamashita, Junichi; Fukasawa, Tetsuo; Hoshino, Kuniyoshi; Kawamura, Fumio; Shiina, Kouji; Sasahira, Akira

2005-01-01

A sustainable electricity supply by fast breeder reactors (FBRs) is essential to ensure energy security and prevent global warming. Transition from light water reactors (LWRs) to FBRs and establishment of an FBR cycle are indispensable, which requires plutonium (Pu) for the introduction of FBRs. The authors propose advanced system called 'Flexible Fuel Cycle Initiative (FFCI)' which can respond flexibly the future expected technical and social uncertainties, can hold no surplus Pu, and can achieve an economical FBR cycle. In the new concept of FFCI, 2nd LWR reprocessing which would succeed Rokkasho Reprocessing Plant is a simple facility to carry out only uranium (U) removal and residual 'recycle material' is stored or utilized. According to FBRs introduction status, recycle material is immediately treated in an FBR reprocessing to fabricate FBR fuel or temporarily stored for the utilization in FBRs at necessary timing. FFCI has high flexibility by having several options for future uncertainties by the introduction of recycle material as a buffer material between LWR and FBR cycles. (author)

Deleterious Thermal Effects due to Randomized Flow Paths in Pebble Bed, and Particle Bed Style Reactors

Science.gov (United States)

Moran, Robert P.

2013-01-01

Reactor fuel rod surface area that is perpendicular to coolant flow direction (+S) i.e. perpendicular to the P creates areas of coolant stagnation leading to increased coolant temperatures resulting in localized changes in fluid properties. Changes in coolant fluid properties caused by minor increases in temperature lead to localized reductions in coolant mass flow rates leading to localized thermal instabilities. Reductions in coolant mass flow rates result in further increases in local temperatures exacerbating changes to coolant fluid properties leading to localized thermal runaway. Unchecked localized thermal runaway leads to localized fuel melting. Reactor designs with randomized flow paths are vulnerable to localized thermal instabilities, localized thermal runaway, and localized fuel melting.

Investigation of two-phase flow structure in model of draught pipe of water boiling reactor VK-300

International Nuclear Information System (INIS)

Efanov, A.D.; Kuznetzov, Y.N.; Kaliakin, S.G.; Lisitza, F.D.; Remizov, O.V.; Serdun, N.P.

2001-01-01

VK-300 reactor represents a vessel-type boiling reactor with integral arrangement of assemblies and in-vessel steam separation at one-circuit scheme. The circuit consists of core, draught pipes, and separation facilities. The vessel of VK-300 reactor is chosen on the base of the dimensions of that of VVER-1000 reactor. The following thermal-hydraulic parameters of nuclear power plant (NPP) were investigated experimentally: dependence of void fraction upon the steam quality in mixing chamber (on the draught section input); pressure losses at different, specific zones of up-flow and down-flow sections of the circuit with free circulation; degree of steam separation in the separating chamber (at the first step of phase separation) and its dependence upon steam quality; structure of steam-water flow in draught pipes (distribution of phases over the draught pipe cross- section); presence of steam hovering and height of this hovering in inter-pipe space of draught section. (author)

Experimental and numerical investigation of the flow field in the gradual transition of rectangular to trapezoidal open channels

Directory of Open Access Journals (Sweden)

Adel Asnaashari

2016-01-01

Full Text Available Transitions are structures that can change geometry and flow velocity through varying the cross-sections of their channels. Under subcritical flow and steady flow conditions, it is necessary to reduce the flow velocity gradually due to increasing water pressure and adverse pressure gradients. Due to the separation of flow and subsequent eddy formation, a significant energy loss is incurred along the transition. This study presents the results of experimental investigations of the subcritical flow along the expansive transition of rectangular to trapezoidal channels. A numerical simulation was developed using a finite volume of fluid (VOF method with a Reynolds stress turbulence model. Water surface profiles and velocity distributions of flow through the transition were measured experimentally and compared with the numerical results. A good agreement between the experimental and numerical model results showed that the Reynolds model and VOF method are capable of simulating the hydraulic flow in open channel transitions. Also, the efficiency of the transition and coefficient of energy head loss were calculated. The results show that with an increasing upstream Froude number, the efficiency of the transition and coefficient of energy head loss decrease and increase, respectively. The results also show the ability of numerical simulation to simulate the flow separation zones and secondary current along the transition for different inlet discharges.

DNS Studies of Transitional Hypersonic Reacting Flows Over 3-D Hypersonic Vehicles

National Research Council Canada - National Science Library

Zhong, Xiaolin

2003-01-01

The objectives of this research project are to develop CFD techniques and to conduct DNS studies of fundamental flow physics leading to boundary-layer instability and transition in hypersonic flows...

Characterisation of cerebral blood flow via determining the vascular mean transit time

International Nuclear Information System (INIS)

Lindner, P.; Thelen, M.

1987-01-01

By using a recently developed algorithm it is possible to quantify the dynamic information of a DSA sequence of the brain. The theory of algorithm allows to calculate vascular mean transit from time density curves. The algorithm minimizes the problems of densitometry with regard to 'quantitative DSA'. There is a strong correlation between vascular mean transit times and cerebral blood flow values, and therefore the results for mean transit times also correspond to the results obtained for cerebral blood flow. By computerized postprocessing of DSA-images it is possible to generate functional images of the brain with a spatial resolution that had not been attainable so far. The images represent the distribution pattern of reverse vascular mean transit times. The results from 36 patients with proven stenoses of the cervical vessels are reported. (orig.) [de

A CFD method to evaluate the integrated influence of leakage and bypass flows on the PBMR Reactor Unit

International Nuclear Information System (INIS)

Janse van Rensburg, J.J.; Kleingeld, M.

2010-01-01

Research highlights: → Research and analysis to identify and rank different leakage flow paths in a HTR. → Development of integrated CFD methodology for the prediction of leakage flows. → Development of a methodology to simulate flow resistances in above CFD model. → Validation of predicted flow results against different numerical methodology. → Illustration of the significant improvement achieved through this methodology. - Abstract: An area that has been identified as significantly important in the development of a High Temperature Reactor (HTR) is the prediction of leakage and bypass flows through such a reactor. It is therefore essential to understand the causes of bypass flows and to determine the effect on the predicted fuel and component temperatures. This paper discusses the identification of leakage flows that are applicable to the Pebble Bed Modular Reactor (Pty) Ltd. (PBMR) design and the ranking of these leakage flows. The modeling methodology and results are also discussed. Similar to previous HTR's, it was found that leakage and bypass flows are important parameters to consider for safe and efficient operation of the PBMR. Through a focused approach, it is shown that PBMR is able to improve the understanding of this phenomenon and quantify the flows and subsequent influence on the operation of the system. This has resulted in a reduction of leakage and bypass from approximately 46% to 20%. The improved understanding of leakage and bypass flows allows PBMR to address this issue during the design phase of the project, which subsequently results in a vast improvement over historical HTR designs. This gives PBMR a distinct advantage over previous High Temperature Reactors.

Calibration of the 7—Equation Transition Model for High Reynolds Flows at Low Mach

Science.gov (United States)

Colonia, S.; Leble, V.; Steijl, R.; Barakos, G.

2016-09-01

The numerical simulation of flows over large-scale wind turbine blades without considering the transition from laminar to fully turbulent flow may result in incorrect estimates of the blade loads and performance. Thanks to its relative simplicity and promising results, the Local-Correlation based Transition Modelling concept represents a valid way to include transitional effects into practical CFD simulations. However, the model involves coefficients that need tuning. In this paper, the γ—equation transition model is assessed and calibrated, for a wide range of Reynolds numbers at low Mach, as needed for wind turbine applications. An aerofoil is used to evaluate the original model and calibrate it; while a large scale wind turbine blade is employed to show that the calibrated model can lead to reliable solutions for complex three-dimensional flows. The calibrated model shows promising results for both two-dimensional and three-dimensional flows, even if cross-flow instabilities are neglected.

Experimental investigation of flow and slip transition in nanochannels

Science.gov (United States)

Li, Zhigang; Li, Long; Mo, Jingwen

2014-11-01

Flow slip in nanochannels is sought in many applications, such as sea water desalination and molecular separation, because it can enhance fluid transport, which is essential in nanofluidic systems. Previous findings about the slip length for simple fluids at the nanoscale appear to be controversial. Some experiments and simulations showed that the slip length is independent of shear rate, which agrees with the prediction of classic slip theories. However, there is increasing work showing that slip length is shear rate dependent. In this work, we experimentally investigate the Poiseuille flows in nanochannels. It is found that the flow rate undergoes a transition between two linear regimes as the shear rate is varied. The transition indicates that the non-slip boundary condition is valid at low shear rate. When the shear rate is larger than a critical value, slip takes place and the slip length increases linearly with increasing shear rate before approaching a constant value. The results reported in this work can help advance the understanding of flow slip in nanochannels. This work was supported by the Research Grants Council of the Hong Kong Special Administrative Region under Grant Nos. 615710 and 615312. J. Mo was partially supported by the Postgraduate Scholarship through the Energy Program at HKUST.

Tokamak turbulence in self-regulated differentially rotating flow and L-H transition dynamics

International Nuclear Information System (INIS)

Terry, P.W.; Carreras, B.A.; Sidikman, K.

1992-01-01

An analytical study of turbulence in the presence of turbulently generated differentially rotating flow is presented as a paradigm for fluctuation dynamics in L- and H-mode plasmas. Using a drift wave model, the role of both flow shear and flow curvature (second radial derivative of the poloidal ExB flow) is detailed in linear and saturated turbulence phases. In the strong turbulence saturated state, finite amplitude-induced modification of the fluctuation structure near low order rational surfaces strongly inhibits flow shear suppression. Suppression by curvature is not diminished, but it occurs through a frequency shift. A description of L-H mode transition dynamics based on the self-consistent linking of turbulence suppression by differentially rotating flow and generation of flow by turbulent momentum transport is presented. In this model, rising edge temperature triggers a transition characterized by spontaneous generation of differentially rotating flow and decreasing turbulence intensity

Reactive turbulent flow CFD study in supercritical water oxidation process: application to a stirred double shell reactor

International Nuclear Information System (INIS)

Moussiere, S.

2006-12-01

Supercritical water oxidation is an innovative process to treat organic liquid waste which uses supercritical water properties to mix efficiency the oxidant and the organic compounds. The reactor is a stirred double shell reactor. In the step of adaptation to nuclear constraints, the computational fluid dynamic modeling is a good tool to know required temperature field in the reactor for safety analysis. Firstly, the CFD modeling of tubular reactor confirms the hypothesis of an incompressible fluid and the use of k-w turbulence model to represent the hydrodynamic. Moreover, the EDC model is as efficiency as the kinetic to compute the reaction rate in this reactor. Secondly, the study of turbulent flow in the double shell reactor confirms the use of 2D axisymmetric geometry instead of 3D geometry to compute heat transfer. Moreover, this study reports that water-air mixing is not in single phase. The reactive turbulent flow is well represented by EDC model after adaptation of initial conditions. The reaction rate in supercritical water oxidation reactor is mainly controlled by the mixing. (author)

Impact of VOC Composition and Reactor Conditions on the Aging of Biomass Cookstove Emissions in an Oxidation Flow Reactor

Science.gov (United States)

Oxidation flow reactor (OFR) experiments in our lab have explored secondary organic aerosol (SOA) production during photochemical aging of emissions from cookstoves used by billions in developing countries. Previous experiments, conducted with red oak fuel under conditions of hig...

Experimental Study on EHD Flow Transition in a Small Scale Wire-plate ESP

Directory of Open Access Journals (Sweden)

Wang Chuan

2016-06-01

Full Text Available The electrohydrodynamic (EHD flow induced by the corona discharge was experimentally investigated in an electrostatic precipitator (ESP. The ESP was a narrow horizontal Plexiglas box (1300 mm×60 mm×60 mm. The electrode set consisted of a single wire discharge electrode and two collecting aluminum plate electrodes. Particle Image Velocimetry (PIV method was used to visualize the EHD flow characteristics inside the ESP seeded with fine oil droplets. The influence of applied voltage (from 8 kV to 10 kV and primary gas flow (0.15 m/s, 0.2 m/s, 0.4 m/s on the EHD flow transition was elucidated through experimental analysis. The formation and transition of typical EHD flows from onset to the fully developed were described and explained. Experimental results showed that the EHD flow patterns change depends on the gas velocity and applied voltage. EHD flow starts with flow streamlines near collecting plates bending towards the wire electrode, forming two void regions. An oscillating jet forming the downstream appeared and moved towards the wire electrode as voltage increased. For higher velocities (≥0.2 m/s, the EHD transition became near wire phenomenon with a jet-like flow structure near the wire, forming a void region behind the wire and expanding as voltage increased. Fully developed EHD secondary flow in the form of counter-rotating vortices appeared upstream with high applied voltage.

Investigation of transition scenarios in boundary-layer flows

Energy Technology Data Exchange (ETDEWEB)

Stolte, A.

1999-11-01

Laminar-turbulent transition mechanisms triggered by crossflow instability in three-dimensional, accelerated boundary-layer flows are investigated using numerical methods of stability analysis. The investigations are based on the DLR swept plate experiment, where stationary and traveling crossflow modes can be selectively introduced into the flow field. Nonlinear instability analyses employing the parabolized stability equations (PSE) show that unique saturation amplitudes do neither exist for stationary crossflow vortices nor for traveling crossflow waves. This phenomenon is explained by means of a spatial bifurcation model. Using Floquet theory, temporal secondary instability analyses are then performed for the mean flow distorted by primary disturbances. In these analyses, secondary high-frequency disturbances with high growth rates are found. The location of these disturbances correlates well with regions of high shear in the primarily distorted flow field, especially on the back of the primary crossflow vortices. (orig.)

Continuous-flow stirred-tank reactor 20-L demonstration test: Final report

International Nuclear Information System (INIS)

Lee, D.D.; Collins, J.L.

2000-01-01

One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test using the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required

Continuous-flow stirred-tank reactor 20-L demonstration test: Final report

Energy Technology Data Exchange (ETDEWEB)

Lee, D.D.; Collins, J.L.

2000-02-01

One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test using the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required.

Simulation of the flow obstruction of a jet pump in a BWR reactor with the code RELAP/SCDAPSIM; Simulacion de la obstruccion de flujo de una bomba jet en un reactor BWR con el codigo RELAP/SCDAPSIM

Energy Technology Data Exchange (ETDEWEB)

Cardenas V, J.; Filio L, C., E-mail: jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose M. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

2016-09-15

This work simulates the flow obstruction of a jet pump in one of the recirculation loops of a nuclear power plant with a reactor of type BWR at 100% of operating power, in order to analyze the behavior of the total flow of the refrigerant passing through the reactor core, the total flow in each recirculation loop of the reactor, together with the 10 jet pumps of each loop. The behavior of the power and the reactivity insertion due to the change of the refrigerant flow pattern is also analyzed. The simulation was carried out using the RELAP/SCDAPSIM version 3.5 code, using a reactor model with 10 jet pumps in each recirculation loop and a core consisting of 6 radial zones and 25 axial zones. The scenario postulates the flow obstruction in a jet pump in a recirculation loop A when the reactor operates at 100% rated power, causing a change in the total flow of refrigerant in the reactor core, leading to a decrease in power. Once the reactor conditions are established to its new power, the operator tries to recover the nominal power using the flow control valve of the recirculation loop A, opening stepwise as a strategy to safely recover the reactor power. In this analysis is assumed that the intention of the nuclear plant operator is to maintain the operation of the reactor during the established cycle. (Author)

Effect of polymer additives on transition in pipe flow

Energy Technology Data Exchange (ETDEWEB)

Castro, W; Squire, W

1967-09-01

Small amounts of long-chain, water-soluble polymers have a marked effect on turbulent flow resulting in appreciable reduction of turbulent friction. The maximum reduction in pipe flow resistance is obtained at such low concentrations that the density and viscosity are not altered appreciably. The minimum friction curve varies as Re-2/3 and appears to be the same for all effective additives tested. The transition process is affected by these additives. Quantitative results are presented showing a reduction in the intensity of the turbulent flashes and the fraction of the time the flow is turbulent at a given Reynolds number. (13 refs.)

Multiphase flow problems on thermofluid safety for fusion reactors

International Nuclear Information System (INIS)

Takase, Kazuyuki

2003-01-01

As the thermofluid safety study for the International Thermonuclear Experimental Reactor (ITER), thermal-hydraulic characteristics of Tokamak fusion reactors under transient events were investigated experimentally and analyzed numerically. As severe transient events an ingress-of-coolant event (ICE) and a loss-of-vacuum event (LOVA) were considered. An integrated ICE test facility was constructed to demonstrate that the ITER safety design approach and parameters are adequate. Water-vapor two-phase flow behavior and performance of the ITER pressure suppression system during the ICE were clarified by the integrated ICE experiments. The TRAC was modified to specify the two-phase flow behavior under the ICE. The ICE experimental results were verified using the modified TRAC code. On the other hand, activated dust mobilization and air ingress characteristics in the ITER vacuum vessel during the LOVA were analyzed using a newly developed analysis code. Some physical models on the motion of dust were considered. The rate of dust released from the vacuum vessel through breaches to the outside was characterized quantitatively. The predicted average pressures in the vacuum vessel during the LOVA were in good agreement with the experimental results. Moreover, direct-contact condensation characteristics between water and vapor inside the ITER suppression tank were observed visually and simulated by the direct two-phase flow analysis. Furthermore, chemical reaction characteristics between vapor and ITER plasma-facing component materials were predicted numerically in order to obtain qualitative estimation on generation of inflammable gases such as hydrogen and methane. The experimental and numerical results of the present studies were reflected in the ITER thermofluid safety design. (author)

Integrated flow reactor that combines high-shear mixing and microwave irradiation for biodiesel production

International Nuclear Information System (INIS)

Choedkiatsakul, I.; Ngaosuwan, K.; Assabumrungrat, S.; Tabasso, S.; Cravotto, G.

2015-01-01

A new simple flow system which is made up of a multi-rotor high-shear mixer connected to a multimode microwave reactor has been assembled. This simple loop reactor has been successfully used in the NaOH-catalyzed transesterification of refined palm oil in methanol. Thanks to optimal mass/heat transfer, full conversion was achieved within 5 min (biodiesel yield of 99.80%). High-quality biodiesel was obtained that is in accordance with international specifications and analytical ASTM standards. The procedure's high efficiency and low energy consumption should pave the way for process scale up. - Highlights: • The combination of HSM-MW flow system for biodiesel production has been proposed. • Highly efficient mass and heat transfer in transesterification reaction. • The hybrid reactor enables a complete conversion in 5 min reaction time. • The new system halved the energy consumption of conventional processes

Dynamics of a two-phase flow through a minichannel: Transition from churn to slug flow

Science.gov (United States)

Górski, Grzegorz; Litak, Grzegorz; Mosdorf, Romuald; Rysak, Andrzej

2016-04-01

The churn-to-slug flow bifurcations of two-phase (air-water) flow patterns in a 2mm diameter minichannel were investigated. With increasing a water flow rate, we observed the transition of slugs to bubbles of different sizes. The process was recorded by a digital camera. The sequences of light transmission time series were recorded by a laser-phototransistor sensor, and then analyzed using the recurrence plots and recurrence quantification analysis (RQA). Due to volume dependence of bubbles velocities, we observed the formation of periodic modulations in the laser signal.

Estimation on the Flow Phenomena and the Pressure Loss for the Inlet Part of a Research Reactor Vessel

International Nuclear Information System (INIS)

Seo, Kyoung Woo; Oh, Jae Min; Seo, Jae Kwang; Yoon, Ju Hyeon; Lee, Doo Jeong

2009-01-01

For a research reactor, a conceptual primary cooling system (PCS) was designed for an adequate cooling to the reactor core. The developed primary cooling circuit consisted of decay tanks, pumps, heat exchangers, vacuum breakers, some isolation and check valves, connection piping, and instruments. The main function of the primary cooling pumps (PCPs) of the PCS was to circulate the reactor coolant through the fuel core and the heat exchangers during a normal operation. The head according to the design flow rate which was determined by the thermal hydraulic design analysis for the core should be estimated to design the PCPs in the fluid system. The pressure loss in the PCS can be calculated by the dimensional analysis of the pipe flow and the head loss coefficient of the components. However, it is insufficient to estimate the pressure loss for 3-dimensional flow phenomena such as the flow path in the reactor with the theoretical dimensional analysis based on experimental data. The purpose of this research is to evaluate the pressure loss of the part of a research reactor vessel. For evaluating the pressure loss, the commercially available CFD computer model, FLUENT, was employed. First, for validating the application of FLUENT to the pressure loss, a simple case was calculated and compared with the Idelchik empirical correlation. Secondly, several cases for the inlet part of a research reactor vessel were estimated by a FLUENT 3- dimensional calculation

Effect of the design change of the LSSBP on core flow distribution of APR+ Reactor

International Nuclear Information System (INIS)

Kim, Kihwan; Euh, Dong-Jin; Choi, Hae-Seob; Kwon, Tae-Soon

2014-01-01

The uniform core inlet flow distribution of an Advanced Power Reactor Plus (APR+) is required to prevent the failure rate of the HIPER fuel assembly and improve the core thermal margin. KEPCO-E and C and KAERI proposed a design change of the Lower Support Structure Bottom Plate (LSSBP), since the core flow rates were intense near the outer region of the intact LSSBP in a previous study. In this study, an experiment was carried out to evaluate the effect of the design change of the LSSBP on the core flow distribution using the APR+ Core Flow and Pressure (ACOP) test facility. The results showed great improvement on the core flow distribution under a 4-pump balanced flow condition. Under the 4-pump balanced flow condition, fifteen tests were repeated using the ACOP test facility to verify the effect of the 50% blocked flow area at the outer region of the LSSBP on the core inlet flow distribution. The profiles of the core inlet mass flow rates were analyzed using ensemble averaged values, and compared with that of the intact LSSBP. The results showed great improvement for the overall core region. The change in design of the LSSBP is expected to improve the hydraulic performance of an APR+ reactor

Sterilization of E. coli bacterium in a flowing N2-O2 post-discharge reactor

International Nuclear Information System (INIS)

Villeger, S; Cousty, S; Ricard, A; Sixou, M

2003-01-01

Effective destruction of Escherichia coli (E. coli) bacteria has been obtained in a flowing N 2 -O 2 microwave post-discharge reactor. The sterilizing agents are the O atoms and the UV emissions of NOβ which are produced by N and O atoms recombination in the reactor. In the following plasma conditions: pressure 5 Torr, flow rate 1 L n min -1 , microwave power of 100 W in a quartz tube of 5 mm, an O atom density of 2.5x10 15 cm -3 is measured by NO titration in the post-discharge reactor with UV emission in a N 2 -(5%)O 2 gas mixture. Full destruction of 10 13 cfu ml -1 E. coli is observed after a treatment time of 25 min. (rapid communication)

Computational and Experimental Investigations of the Coolant Flow in the Cassette Fissile Core of a KLT-40S Reactor

Science.gov (United States)

Dmitriev, S. M.; Varentsov, A. V.; Dobrov, A. A.; Doronkov, D. V.; Pronin, A. N.; Sorokin, V. D.; Khrobostov, A. E.

2017-07-01

Results of experimental investigations of the local hydrodynamic and mass-exchange characteristics of a coolant flowing through the cells in the characteristic zones of a fuel assembly of a KLT-40S reactor plant downstream of a plate-type spacer grid by the method of diffusion of a gas tracer in the coolant flow with measurement of its velocity by a five-channel pneumometric probe are presented. An analysis of the concentration distribution of the tracer in the coolant flow downstream of a plate-type spacer grid in the fuel assembly of the KLT-40S reactor plant and its velocity field made it possible to obtain a detailed pattern of this flow and to determine its main mechanisms and features. Results of measurement of the hydraulic-resistance coefficient of a plate-type spacer grid depending on the Reynolds number are presented. On the basis of the experimental data obtained, recommendations for improvement of the method of calculating the flow rate of a coolant in the cells of the fissile core of a KLT-40S reactor were developed. The results of investigations of the local hydrodynamic and mass-exchange characteristics of the coolant flow in the fuel assembly of the KLT-40S reactor plant were accepted for estimating the thermal and technical reliability of the fissile cores of KLT-40S reactors and were included in the database for verification of computational hydrodynamics programs (CFD codes).

Analysis of impact of mixing flow on the pebble bed high temperature reactor

International Nuclear Information System (INIS)

Hao Chen; Li Fu; Guo Jiong

2014-01-01

The impact of the mixing flow in the pebble flow on pebble bed high temperature gas cooled reactor (HTR) was analyzed in the paper. New code package MFVSOP which can simulate the mixing flow was developed. The equilibrium core of HTR-PM was selected as reference case, the impact of the mixing flow on the core parameters such as core power peak factor, power distribution was analyzed with different degree of mixing flow, and uncertainty analysis was carried out. Numerical results showed that the mixing flow had little impact on key parameters of pebble bed HTR, and the multiple-pass-operation-mode in pebble bed HTR can reduce the uncertainty arouse from the mixing flow. (authors)

Calculation of the flow distribution for the new core of the RA-6 reactor

International Nuclear Information System (INIS)

Garcia, J.C.; Delmastro, Dario F.

2007-01-01

In this work the pressure drop, the flow distribution, effective cooling flow rate and the velocity in the subchannels that cool fuel plates for the new core of RA-6 research reactor were calculated. These calculations were performed for a flow of 340 m 3 /hr and water temperatures of 12 C degrees, of 35 C degrees and 42 C degrees. The flow distribution was calculated without considering either safety factors or geometric changes. All the calculations were performed considering the flow as isothermal. (author) [es

Internal fluid flow management analysis for Clinch River Breeder Reactor Plant sodium pumps

International Nuclear Information System (INIS)

Cho, S.M.; Zury, H.L.; Cook, M.E.; Fair, C.E.

1978-12-01

The Clinch River Breeder Reactor Plant (CRBRP) sodium pumps are currently being designed and the prototype unit is being fabricated. In the design of these large-scale pumps for elevated temperature Liquid Metal Fast Breeder Reactor (LMFBR) service, one major design consideration is the response of the critical parts to severe thermal transients. A detailed internal fluid flow distribution analysis has been performed using a computer code HAFMAT, which solves a network of fluid flow paths. The results of the analytical approach are then compared to the test data obtained on a half-scale pump model which was tested in water. The details are presented of pump internal hydraulic analysis, and test and evaluation of the half-scale model test results

FBR type reactors

International Nuclear Information System (INIS)

Maemoto, Junko.

1985-01-01

Purpose: To moderate abrupt temperature change near the inner walls of a suspended body thereby prevent thermal shocks and thermal deformations to structural materials. Constitution: High temperature coolants during ordinary operation of the nuclear reactor flow from the reactor core through the flow holes of the suspended body and from the upper plenum into an intermediate heat exchanger. The temperature of the coolants is lowered with heat exchanging effect with secondary coolants in the heat exchange and the coolants are then flow through the lower plenum into the reactor core and heated again. Upon generation of reactor scram, the temperature of the coolants at the exit of the reactor core is reduced abruptly and the flow rate is lowered due to the pump coast down. However, mixing of the coolants in the suspended body is accelerated by the coolants at high temperature flowing out of the flow holes and the coolants at the low temperature flowing from the flow hole group, to reduce the temperature difference and moderate the stratification flow forming an abrupt temperature slope. (Yoshihara, H.)

Transient thermal hydraulic analysis of the IAEA 10 MW MTR reactor during Loss of Flow Accident to investigate the flow inversion

International Nuclear Information System (INIS)

AL-Yahia, Omar S.; Albati, Mohammad A.; Park, Jonghark; Chae, Heetaek; Jo, Daeseong

2013-01-01

Highlights: • Transient analyses of a slow and fast LOFA were investigated. • A reactor kinetic and thermal hydraulic coupled model was developed. • Based on force balance, the flow rate during flow inversion was determined. • Flow inversion in a hot channel occurred earlier than in an average channel. • Two temperature peaks were observed during both slow and fast LOFA. - Abstract: Transient analyses of the IAEA 10 MW MTR reactor are investigated during a fast and slow Loss of Flow Accident (LOFA) with a neutron kinetic and thermal hydraulic coupling model. A spatial-dependent thermal hydraulic technique is adopted for analyzing the local thermal hydraulic parameters and hotspot location during a flow inversion. The flow rate through the channel is determined in terms of a balance between driving and preventing forces. Friction and buoyancy forces act as resistance of the flow before a flow inversion while buoyancy force becomes the driving force after a flow inversion. By taking into account the buoyancy effect to determine the flow rate, the difference in the flow inversion time between hot and average channels is investigated: a flow inversion occurs earlier in the hot channel than in an average channel. Furthermore, the movement of the hotspot location before and after a flow inversion is investigated for a slow and fast LOFA. During a flow inversion, two temperature peaks are observed: (1) the first temperature peak is at the initiation of the LOFA, and (2) the second temperature peak is when a flow inversion occurs. The maximum temperature of the cladding is found at the second temperature peak for both LOFA analyses, and is lower than the saturation temperature

Synthesis of Struvite using a Vertical Canted Reactor with Continuous Laminar Flow Process

Science.gov (United States)

Sutiyono, S.; Edahwati, L.; Muryanto, S.; Jamari, J.; Bayuseno, A. P.

2018-01-01

Struvite is a white crystalline that is chemically known as magnesium ammonium phosphorus hexahydrate (MgNH4PO4·6H2O). It can easily dissolve in acidic conditions and slightly soluble in neutral and alkaline conditions. In industry, struvite forms as a scale deposit on a pipe with hot flow fluid. However, struvite can be used as fertilizer because of its phosphate content. A vertical canted reactor is a promising technology for recovering phosphate levels in wastewater through struvite crystallization. The study was carried out with the vertical canted reactor by mixing an equimolar stock solution of MgCl2, NH4OH, and H3PO4 in 1: 1: 1 ratio. The crystallization process worked with the flow rate of three stock solution entering the reactor in the range of 16-38 ml/min, the temperature in the reactor is worked on 20°, 30°, and 40°C, while the incoming air rate is kept constant at 0.25 liters/min. Moreover, pH was maintained at a constant value of 9. The struvite crystallization process run until the steady state was reached. Then, the result of crystal precipitates was filtered and dried at standard temperature room for 48 hours. After that, struvite crystals were stored for the subsequent analysis by Scanning Electron Microscope (SEM) and XRD (X-Ray Diffraction) method. The use of canted reactor provided the high pure struvite with a prismatic crystal morphology.

Flooding in counter-current two-phase flow

International Nuclear Information System (INIS)

Ragland, W.A.; Ganic, E.N.

1982-01-01

Flooding is a phenomenon which is best described as the transition from counter-current to co-current flow. Early notice was taken of this phenomenon in the chemical engineering industry. Flooding also plays an important role in the field of two-phase heat transfer since it is a limit for many systems involving counter-current flow. Practical applications of flooding limited processes include wickless thermosyphons and the emergency core cooling system (ECCS) of pressurized water nuclear reactors. The phenomenon of flooding also is involved in the behavior of nuclear reactor core materials during severe accident conditions where flooding is one of the mechanisms governing the motion of the molten fuel pin cladding

Flooding in counter-current two-phase flow

Energy Technology Data Exchange (ETDEWEB)

Ragland, W.A.; Ganic, E.N.

1982-01-01

Flooding is a phenomenon which is best described as the transition from counter-current to co-current flow. Early notice was taken of this phenomenon in the chemical engineering industry. Flooding also plays an important role in the field of two-phase heat transfer since it is a limit for many systems involving counter-current flow. Practical applications of flooding limited processes include wickless thermosyphons and the emergency core cooling system (ECCS) of pressurized water nuclear reactors. The phenomenon of flooding also is involved in the behavior of nuclear reactor core materials during severe accident conditions where flooding is one of the mechanisms governing the motion of the molten fuel pin cladding.

CFD analysis of flow distribution of reactor core and temperature rise of coolant in fuel assembly for VVER reactor

International Nuclear Information System (INIS)

Du Daiquan; Zeng Xiaokang; Xiong Wanyu; Yang Xiaoqiang

2015-01-01

Flow field of VVER-1000 reactor core was investigated by using computational fluid dynamics code CFX, and the temperature rise of coolant in hot assembly was calculated. The results show that the maximum value of flow distribution factor is 1.12 and the minimum value is 0.92. The average value of flow distribution factor in hot assembly is 0.97. The temperature rise in hot assembly is higher than current warning limit value ΔT t under the deviated operation condition. The results can provide reference for setting ΔT t during the operation of nuclear power plant. (authors)

Modelling of flow stabilization by the swirl of a peripheral flow as applied to plasma reactors

International Nuclear Information System (INIS)

Volchkov, E.P.; Lebedev, V.P.; Terekhov, V.I.; Shishkin, N.E.

2000-01-01

The gas-swirl stabilization of plasma jets is one of effective methods of its retention in the near-axial area of channels in generators of low-temperature plasma. Except the effect of gas-dynamic compression, the peripheral swirl allows to solve another urgent problem - to protect the reactor walls from the heat influence of the plasma jet. Swirl flows are also used for the flow structure formation and control of the heat and gas-dynamic characteristics of different power devices and apparatuses, using high-temperature working media: in swirl furnaces and burners, in aviation engines, etc. Investigations show that during swirl stabilization the gas-dynamic structure of the flow influences significantly the spatial stability of the plasma column and its characteristics

Analysis Of Core Management For The Transition Cores Of RSG-GAS Reactor To Full-Silicide Core

International Nuclear Information System (INIS)

Malem Sembiring, Tagor; Suparlina, Lily; Tukiran

2001-01-01

The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 g U/cc is still doing. At the end of 2000, the reactor has been operated for 3 transition cores which is the mixed core of oxide-silicide. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 10 transition cores to achieve a full-silicide core. The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters such as excess reactivity and shutdown margin. The measurement of the core parameters was carried out using the method of compensation of couple control rods. The experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safety to a full-silicide core

Bioremoval of trivalent chromium using Bacillus biofilms through continuous flow reactor

International Nuclear Information System (INIS)

Sundar, K.; Sadiq, I. Mohammed; Mukherjee, Amitava; Chandrasekaran, N.

2011-01-01

Highlights: ► Effective bioremoval of Cr(III) using bacterial biofilms. ► Simplified bioreactor was fabricated for the biofilm development and Cr(III) removal. ► Economically feasible substrate like coarse sand and pebbles were used. - Abstract: Present study deals with the applicability of bacterial biofilms for the bioremoval of trivalent chromium from tannery effluents. A continuous flow reactor was designed for the development of biofilms on different substrates like glass beads, pebbles and coarse sand. The parameters for the continuous flow reactor were 20 ml/min flow rate at 30 °C, pH4. Biofilm biomass on the substrates was in the following sequence: coarse sand > pebbles > glass beads (4.8 × 10 7 , 4.5 × 10 7 and 3.5 × 10 5 CFU/cm 2 ), which was confirmed by CLSM. Biofilms developed using consortium of Bacillus subtilis and Bacillus cereus on coarse sand had more surface area and was able to remove 98% of Cr(III), SEM-EDX proved 92.60% Cr(III) adsorption on biofilms supported by coarse sand. Utilization of Bacillus biofilms for effective bioremoval of Cr(III) from chrome tanning effluent could be a better option for tannery industry, especially during post chrome tanning operation.

Characterisation of two-phase horizontal flow regime transition by the application of time-frequency analysis methods

International Nuclear Information System (INIS)

Seleghim, Paulo

1996-01-01

This work concerns the development of a methodology which objective is to characterize and diagnose two-phase flow regime transitions. The approach is based on the fundamental assumption that a transition flow is less stationary than a flow with an established regime. In a first time, the efforts focused on: 1) the design and construction of an experimental loop, allowing to reproduce the main horizontal two-phase flow patterns, in a stable and controlled way, 2) the design and construction of an electrical impedance probe, providing an imaged information of the spatial phase distribution in the pipe, the systematic study of the joint time-frequency and time-scale analysis methods, which permitted to define an adequate parameter quantifying the un-stationary degree. In a second time, in order to verify the fundamental assumption, a series of experiments were conducted, which objective was to demonstrate the correlation between un-stationary and regime transition. The un-stationary degree was quantified by calculating the Gabor's transform time-frequency covariance of the impedance probe signals. Furthermore, the phenomenology of each transition was characterized by the joint moments and entropy. The results clearly show that the regime transitions are correlated with local-time frequency covariance peaks, which demonstrates that these regime transitions are characterized by a loss of stationarity. Consequently, the time-frequency covariance constitutes an objective two-phase flow regime transition indicator. (author) [fr

Calculation of gas-flow in plasma reactor for carbon partial oxidation

Science.gov (United States)

Bespala, Evgeny; Myshkin, Vyacheslav; Novoselov, Ivan; Pavliuk, Alexander; Makarevich, Semen; Bespala, Yuliya

2018-03-01

The paper discusses isotopic effects at carbon oxidation in low temperature non-equilibrium plasma at constant magnetic field. There is described routine of experiment and defined optimal parameters ensuring maximum enrichment factor at given electrophysical, gas-dynamic, and thermodymanical parameters. It has been demonstrated that at high-frequency generator capacity of 4 kW, supply frequency of 27 MHz and field density of 44 mT the concentration of paramagnetic heavy nuclei 13C in gaseous phase increases up to 1.78 % compared to 1.11 % for natural concentration. Authors explain isotopic effect decrease during plasmachemical separation induced by mixing gas flows enriched in different isotopes at the lack of product quench. With the help of modeling the motion of gas flows inside the plasma-chemical reactor based on numerical calculation of Navier-Stokes equation authors determine zones of gas mixing and cooling speed. To increase isotopic effects and proportion of 13C in gaseous phase it has been proposed to use quench in the form of Laval nozzle of refractory steel. The article represents results on calculation of optimal Laval Nozzle parameters for plasma-chemical reactor of chosen geometry of. There are also given dependences of quench time of products on pressure at the diffuser output and on critical section diameter. Authors determine the location of quench inside the plasma-chemical reactor in the paper.

Characteristics of a novel nanosecond DBD microplasma reactor for flow applications

Science.gov (United States)

Elkholy, A.; Nijdam, S.; van Veldhuizen, E.; Dam, N.; van Oijen, J.; Ebert, U.; de Goey, L. Philip H.

2018-05-01

We present a novel microplasma flow reactor using a dielectric barrier discharge (DBD) driven by repetitive nanosecond high-voltage pulses. Our DBD-based geometry can generate a non-thermal plasma discharge at atmospheric pressure and below in a regular pattern of micro-channels. This reactor can work continuously up to about 100 min in air, depending on the pulse repetition rate and operating pressure. We here present the geometry and main characteristics of the reactor. Pulse energies of 1.46 and 1.3 μJ per channel at atmospheric pressure and 50 mbar, respectively, have been determined by time-resolved measurements of current and voltage. Time-resolved optical emission spectroscopy measurements have been performed to calculate the relative species concentrations and temperatures (vibrational and rotational) of the discharge. The effects of the operating pressure and flow velocity on the discharge intensity have been investigated. In addition, the effective reduced electric field strength {(E/N)}eff} has been obtained from the intensity ratio of vibronic emission bands of molecular nitrogen at different operating pressures and different locations. The derived {(E/N)}eff} increases gradually from about 550 to 4600 Td when decreasing the pressure from 1 bar to 100 mbar. Below 100 mbar, further pressure reduction results in a significant increase in {(E/N)}eff} up to about 10000 Td at 50 mbar.

FBR type reactors

International Nuclear Information System (INIS)

Suzuoki, Akira; Yamakawa, Masanori.

1985-01-01

Purpose: To enable safety and reliable after-heat removal from a reactor core. Constitution: During ordinary operation of a FBR type reactor, sodium coolants heated to a high temperature in a reactor core are exhausted therefrom, collide against the reactor core upper mechanisms to radially change the flowing direction and then enter between each of the guide vanes. In the case if a main recycling pump is failed and stopped during reactor operation and the recycling force is eliminated, the swirling stream of sodium that has been resulted by the flow guide mechanism during normal reactor operation is continuously maintained within a plenum at a high temperature. Accordingly, the sodium recycling force in the coolant flow channels within the reactor vessel can surely be maintained for a long period of time due to the centrifugal force of the sodium swirling stream. In this way, since the reactor core recycling flow rate can be secured even after the stopping of the main recycling pump, after-heat from the reactor core can safely and surely be removed. (Seki, T.)

Optimal Homogenization of Perfusion Flows in Microfluidic Bio-Reactors: A Numerical Study

DEFF Research Database (Denmark)

Okkels, Fridolin; Dufva, Martin; Bruus, Henrik

2011-01-01

In recent years, the interest in small-scale bio-reactors has increased dramatically. To ensure homogeneous conditions within the complete area of perfused microfluidic bio-reactors, we develop a general design of a continually feed bio-reactor with uniform perfusion flow. This is achieved...... by introducing a specific type of perfusion inlet to the reaction area. The geometry of these inlets are found using the methods of topology optimization and shape optimization. The results are compared with two different analytic models, from which a general parametric description of the design is obtained...... and tested numerically. Such a parametric description will generally be beneficial for the design of a broad range of microfluidic bioreactors used for, e. g., cell culturing and analysis and in feeding bio-arrays....

Dispersed plug flow model for upflow anaerobic sludge bed reactors with focus on granular sludge dynamics

NARCIS (Netherlands)

Kalyuzhnyi, S.V.; Fedorovich, V.V.; Lens, P.N.L.

2006-01-01

A new approach to model upflow anaerobic sludge bed (UASB)-reactors, referred to as a one-dimensional dispersed plug flow model, was developed. This model focusses on the granular sludge dynamics along the reactor height, based on the balance between dispersion, sedimentation and convection using

The development of NRTM-turbine flow meter and measurement of the coolant flow rate in-core of 5 MW heating reactor

International Nuclear Information System (INIS)

Zha Meisheng; Wang Xiuqin; Ni Mengchen

1995-01-01

In order to measure the coolant flow rate in-core of 5 MW Heating Reactor the special turbine flowmeter of the type of NRTM has been developed. It consists of a body, a turbine with long screw blade and six pieces of Alnico magnets, and a coil mounted on the body. The advantage of this turbine flowmeter is of low resistance and long working-life. Another advantage is that when the turbine is working or not working its factor of resistance is about the same. It is very important for a natural circulation heating reactor. Because the cable, which is welded to the coil assembly, is long enough to extend out of the reactor vessel to the control room, the signal of flow rate is easy to be disturbed by noise in the case. The traditional method of counting the frequency of the A-C voltage which is induced in the coil has a poor ability for resisting noise. The method of the frequency-spectrum analysis of the frequency of the A-C voltage is used to make sure the accuracy of the measurement of the turbine flow meter. Compared with the method of the count it has a good ability for resisting noise. After three years operation a lot of valuable data were obtained

Experimental Investigation of Flow Resistance in a Coal Mine Ventilation Air Methane Preheated Catalytic Oxidation Reactor

OpenAIRE

Zheng, Bin; Liu, Yongqi; Liu, Ruixiang; Meng, Jian; Mao, Mingming

2015-01-01

This paper reports the results of experimental investigation of flow resistance in a coal mine ventilation air methane preheated catalytic oxidation reactor. The experimental system was installed at the Energy Research Institute of Shandong University of Technology. The system has been used to investigate the effects of flow rate (200 Nm3/h to 1000 Nm3/h) and catalytic oxidation bed average temperature (20°C to 560°C) within the preheated catalytic oxidation reactor. The pressure drop and res...

Device for measuring flow rate in a nuclear reactor core

International Nuclear Information System (INIS)

Hamano, Jiro.

1980-01-01

Purpose: To always calculate core flow rate automatically and accurately in BWR type nuclear power plants. Constitution: Jet pumps are provided to the recycling pump and to the inside of the pressure vessel of a nuclear reactor. The jet pumps comprise a plurality of calibrated jet pumps for forcively convecting the coolants and a plurality of not calibrated jet pumps in order to cool the heat generated in the reactor core. The difference in the pressures between the upper and the lower portions in both of the jet pumps is measured by difference pressure transducers. Further, a thermo-sensitive element is provided to measure the temperature of recycling water at the inlet of the recycling pump. The output signal from the difference pressure transducer is inputted to a process computer, calculated periodically based on predetermined calculation equations, compensated for the temperature by a recycling water temperature signal and outputted as a core flow rate signal to a recoder. The signal is also used for the power distribution calculation in the process computer and the minimum limit power ratio as the thermal limit value for the fuels is outputted. (Furukawa, Y.)

Korea advanced liquid metal reactor development - Development of measuring techniques of the sodium two-phase flow

Energy Technology Data Exchange (ETDEWEB)

Kim, Moo Hwan; Cha, Jae Eun [Pohang University of Science and Technology, Pohang (Korea)

2000-04-01

The technology which models and measures the behavior of bubble in liquid sodium is very important to insure the safety of the liquid metal reactor. In this research, we designed/ manufactured each part and loop of experimental facility for sodium two phase flow, and applied a few possible methods, measured characteristic of two phase flow such as bubbly flow. A air-water loop similar to sodium loop on each measuring condition was designed/manufactured. This air-water loop was utilized to acquire many informations which were necessary in designing the two phase flow of sodium and manufacturing experimental facility. Before the manufacture of a electromagnetic flow meter for sodium, the experiment using each electromagnetic flow mete was developed and the air-water loop was performed to understand flow characteristics. Experiments for observing the signal characteristics of flow were performed by flowing two phase mixture into the electromagnetic flow mete. From these experiments, the electromagnetic flow meter was designed and constructed by virtual electrode, its signal processing circuit and micro electro magnet. It was developed to be applicable to low conductivity fluid very successfully. By this experiment with the electromagnetic flow meter, we observed that the flow signal was very different according to void fraction in two phase flow and that probability density function which was made by statistical signal treatment is also different according to flow patterns. From this result, we confirmed that the electromagnetic flow meter could be used to understand the parameters of two phase flow of sodium. By this study, the experimental facility for two phase flow of sodium was constricted. Also the new electromagnetic flow meter was designed/manufactured, and experimental apparatus for two phase flow of air-water. Finally, this study will be a basic tool for measurement of two phase flow of sodium. As the fundamental technique for the applications of sodium at

Patterned Roughness for Cross-flow Transition Control at Mach 6

Science.gov (United States)

Arndt, Alexander; Matlis, Eric; Semper, Michael; Corke, Thomas

2017-11-01

Experiments are performed to investigate patterned discrete roughness for transition control on a sharp right-circular cone at an angle of attack at Mach 6.0. The approach to transition control is based on exciting less-amplified (subcritical) stationary cross-flow (CF) modes that suppress the growth of the more-amplified (critical) CF modes, and thereby delay transition. The experiments were performed in the Air Force Academy Ludwieg Tube which is a conventional (noisy) design. The cone model is equipped with a motorized 3-D traversing mechanism that mounts on the support sting. The traversing mechanism held a closely-spaced pair of fast-response total pressure Pitot probes. The model utilized a removable tip to exchange between different tip-roughness conditions. Mean flow distortion x-development indicated that the transition Reynolds number increased by 25% with the addition of the subcritical roughness. The energy in traveling disturbances was centered in the band of most amplified traveling CF modes predicted by linear theory. The spatial pattern in the amplitude of the traveling CF modes indicated a nonlinear (sum and difference) interaction between the stationary and traveling CF modes that might explain differences in Retrans between noisy and quiet environments. Air Force Grant FA9550-15-1-0278.

Startup and oxygen concentration effects in a continuous granular mixed flow autotrophic nitrogen removal reactor.

Science.gov (United States)

Varas, Rodrigo; Guzmán-Fierro, Víctor; Giustinianovich, Elisa; Behar, Jack; Fernández, Katherina; Roeckel, Marlene

2015-08-01

The startup and performance of the completely autotrophic nitrogen removal over nitrite (CANON) process was tested in a continuously fed granular bubble column reactor (BCR) with two different aeration strategies: controlling the oxygen volumetric flow and oxygen concentration. During the startup with the control of oxygen volumetric flow, the air volume was adjusted to 60mL/h and the CANON reactor had volumetric N loadings ranging from 7.35 to 100.90mgN/Ld with 36-71% total nitrogen removal and high instability. In the second stage, the reactor was operated at oxygen concentrations of 0.6, 0.4 and 0.2mg/L. The best condition was 0.2 mgO2/L with a total nitrogen removal of 75.36% with a CANON reactor activity of 0.1149gN/gVVSd and high stability. The feasibility and effectiveness of CANON processes with oxygen control was demonstrated, showing an alternative design tool for efficiently removing nitrogen species. Copyright © 2015 Elsevier Ltd. All rights reserved.

Experimental study of flow field characteristics on bed configurations in the pebble bed reactor

International Nuclear Information System (INIS)

Jia, Xinlong; Gui, Nan; Yang, Xingtuan; Tu, Jiyuan; Jia, Haijun; Jiang, Shengyao

2017-01-01

Highlights: • PTV study of flow fields of pebble bed reactor with different configurations are carried out. • Some criteria are proposed to quantify vertical velocity field and flow uniformity. • The effect of different pebble bed configurations is also compared by the proposed criteria. • The displacement thickness is used analogically to analyze flow field characteristics. • The effect of mass flow variation in the stagnated region of the funnel flow is measured. - Abstract: The flow field characteristics are of fundamental importance in the design work of the pebble bed high temperature gas cooled reactor (HTGR). The different effects of bed configurations on the flow characteristics of pebble bed are studied through the PTV (Particle Tracking Velocimetry) experiment. Some criteria, e.g. flow uniformity (σ) and mass flow level (α), are proposed to estimate vertical velocity field and compare the bed configurations. The distribution of the Δθ (angle difference between the individual particle velocity and the velocity vector sum of all particles) is also used to estimate the resultant motion consistency level. Moreover, for each bed configuration, the thickness of displacement is analyzed to measure the effect of the funnel flow zone based on the boundary layer theory. Detailed information shows the quantified characteristics of bed configuration effects on flow uniformity and other characteristics; and the sequence of levels of each estimation criterion is obtained for all bed configurations. In addition, a good design of the pebble bed configuration is suggested and these estimation criteria can be also applied and adopted in testing other geometry designs of pebble bed.

Properties of Phase Transition of Traffic Flow on Urban Expressway Systems with Ramps and Accessory Roads

International Nuclear Information System (INIS)

Mei Chaoqun; Liu Yejin

2011-01-01

In this paper, we develop a cellular automaton model to describe the phase transition of traffic flow on urban expressway systems with on-off-ramps and accessory roads. The lane changing rules are given in detailed, the numerical results show that the main road and the accessory road both produce phase transitions. These phase transitions will often be influenced by the number of lanes, lane changing, the ramp flow, the input flow rate, and the geometry structure. (interdisciplinary physics and related areas of science and technology)

Diels–Alder reactions of myrcene using intensified continuous-flow reactors

Directory of Open Access Journals (Sweden)

Christian H. Hornung

2017-01-01

Full Text Available This work describes the Diels–Alder reaction of the naturally occurring substituted butadiene, myrcene, with a range of different naturally occurring and synthetic dienophiles. The synthesis of the Diels–Alder adduct from myrcene and acrylic acid, containing surfactant properties, was scaled-up in a plate-type continuous-flow reactor with a volume of 105 mL to a throughput of 2.79 kg of the final product per day. This continuous-flow approach provides a facile alternative scale-up route to conventional batch processing, and it helps to intensify the synthesis protocol by applying higher reaction temperatures and shorter reaction times.

Direct In Situ Quantification of HO2 from a Flow Reactor.