Axial and Radial Gas Holdup in Bubble Column Reactor

International Nuclear Information System (INIS)

Wagh, Sameer M.; Ansari, Mohashin E Alan; Kene, Pragati T.

2014-01-01

Bubble column reactors are considered the reactor of choice for numerous applications including oxidation, hydrogenation, waste water treatment, and Fischer-Tropsch (FT) synthesis. They are widely used in a variety of industrial applications for carrying out gas-liquid and gas-liquid-solid reactions. In this paper, the computational fluid dynamics (CFD) model is used for predicting the gas holdup and its distribution along radial and axial direction are presented. Gas holdup increases linearly with increase in gas velocity. Gas bubbles tends to concentrate more towards the center of the column and follows a wavy path

Metabolic modeling of synthesis gas fermentation in bubble column reactors.

Science.gov (United States)

Chen, Jin; Gomez, Jose A; Höffner, Kai; Barton, Paul I; Henson, Michael A

2015-01-01

A promising route to renewable liquid fuels and chemicals is the fermentation of synthesis gas (syngas) streams to synthesize desired products such as ethanol and 2,3-butanediol. While commercial development of syngas fermentation technology is underway, an unmet need is the development of integrated metabolic and transport models for industrially relevant syngas bubble column reactors. We developed and evaluated a spatiotemporal metabolic model for bubble column reactors with the syngas fermenting bacterium Clostridium ljungdahlii as the microbial catalyst. Our modeling approach involved combining a genome-scale reconstruction of C. ljungdahlii metabolism with multiphase transport equations that govern convective and dispersive processes within the spatially varying column. The reactor model was spatially discretized to yield a large set of ordinary differential equations (ODEs) in time with embedded linear programs (LPs) and solved using the MATLAB based code DFBAlab. Simulations were performed to analyze the effects of important process and cellular parameters on key measures of reactor performance including ethanol titer, ethanol-to-acetate ratio, and CO and H2 conversions. Our computational study demonstrated that mathematical modeling provides a complementary tool to experimentation for understanding, predicting, and optimizing syngas fermentation reactors. These model predictions could guide future cellular and process engineering efforts aimed at alleviating bottlenecks to biochemical production in syngas bubble column reactors.

Slurry reactors

Energy Technology Data Exchange (ETDEWEB)

Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

1979-08-01

Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

An algorithm for predicting the hydrodynamic and mass transfer parameters in bubble column and slurry bubble column reactors

Energy Technology Data Exchange (ETDEWEB)

Lemoine, Romain; Behkish, Arsam; Sehabiague, Laurent; Heintz, Yannick J.; Morsi, Badie I. [Chemical and Petroleum Engineering Department, University of Pittsburgh, Pittsburgh, PA 15261 (United States); Oukaci, Rachid [Energy Technology Partners, Pittsburgh, PA 15238 (United States)

2008-04-15

A large number of experimental data points obtained in our laboratory as well as from the literature, covering wide ranges of reactor geometry (column diameter, gas distributor type/open area), physicochemical properties (liquid and gas densities and molecular weights, liquid viscosity and surface tension, gas diffusivity, solid particles size/density), and operating variables (superficial gas velocity, temperature and pressure, solid loading, impurities concentration, mixtures) were used to develop empirical as well as Back-Propagation Neural Network (BPNN) correlations in order to predict the hydrodynamic and mass transfer parameters in bubble column reactors (BCRs) and slurry bubble column reactors (SBCRs). The empirical and BPNN correlations developed were incorporated in an algorithm for predicting gas holdups ({epsilon}{sub G}, {epsilon}{sub G-Small}, {epsilon}{sub G-Large}); volumetric liquid-side mass transfer coefficients (k{sub L}a, k{sub L}a{sub -Small,} k{sub L}a{sub -Large}); Sauter mean bubble diameters (d{sub S}, d{sub S-Small}, d{sub S-Large}){sub ;} gas-liquid interfacial areas (a, a{sub Small}, a{sub Large}); and liquid-side mass transfer coefficients (k{sub L}, k{sub L-Large}, k{sub L-Small}) for total, small and large gas bubbles in BCRs and SBCRs. The developed algorithm was used to predict the effects of reactor diameter and solid (alumina) loading on the hydrodynamic and mass transfer parameters in the Fisher-Tropsch (F-T) synthesis for the hydrogenation of carbon monoxide in a SBCR, and to predict the effects of presence of organic impurities (which decrease the liquid surface tension) and air superficial mass velocity in the Loprox process for the wet air oxidation of organic pollutants in a BCR. In the F-T process, the predictions showed that increasing the reactor diameter from 0.1 to 7.0 m and/or increasing the alumina loading from 25 to 50 wt.% significantly decreased {epsilon}{sub G,} k{sub L}a{sub H2} and k{sub L}a{sub CO} and

Catalytic fast pyrolysis of white oak wood in-situ using a bubbling fluidized bed reactor

Science.gov (United States)

Catalytic fast pyrolysis was performed on white oak wood using two zeolite-type catalysts as bed material in a bubbling fluidized bed reactor. The two catalysts chosen, based on a previous screening study, were Ca2+ exchanged Y54 (Ca-Y54) and a proprietary ß-zeolite type catalyst (catalyst M) both ...

The detection of sodium vapor bubble collapse in a liquid metal fast breeder reactor

International Nuclear Information System (INIS)

Carey, W.M.; Gavin, A.P.; Bobis, J.P.; Sheen, S.H.; Anderson, T.T.; Doolittle, R.D.; Albrecht, R.W.

1977-01-01

Sodium boiling detection utilizing the sound pressure emanated during the collapse of a sodium vapour bubble in a subcooled media is discussed in terms of the sound characteristic, the reactor ambient noise background, transmission loss considerations and performance criteria. Data obtained in several loss of flow experiments on Fast Test Reactor Fuel Elements indicate that the collapse of the sodium vapour bubble depends on the presence of a subcooled structure or sodium. The collapse pressure pulse was observed in all cases to be on the order of a kPa, indicating a soft type of cavitational collapse. Spectral examination of the pulses indicates the response function of the test structure and geometry is important. The sodium boiling observed in these experiments was observed to occur at a low ( 0 C) liquid superheat with the rate of occurrence of sodium vapor bubble collapse in the 3 to 30 Hz range. Reactor ambient noise data were found to be due to machinery induced vibrations flow induced vibrations, and flow noise. These data were further found to be weakly stationary enhancing the possibility of acoustic surveillance of an operating Liquid Metal Fast Breeder Reactor. Based on these noise characteristics and extrapolating the noise measurements from the Fast Flux Test Facility Pump (FFTP), one would expect a signal to noise ratio of up to 20 dB in the absence of transmission loss. The requirement of a low false alarm probability is shown to necessitate post detection analysis of the collapse event sequence and the cross correlation with the second derivative of the neutronic boiling detection signal. Sodium boiling detection using the sounds emitted during sodium vapor bubble collapse are shown to be feasible but a need for in-reactor demonstration is necessary. (author)

Discrete bubble modeling for a micro-structured bubble column

NARCIS (Netherlands)

Jain, D.; Lau, Y.M.; Kuipers, J.A.M.; Deen, N.G.

2013-01-01

Gas–liquid flows with solid catalyst particles are encountered in many applications in the chemical, petrochemical, pharmaceutical industries, etc. Most commonly, two reactor types are applied for large scale in the industry. They are slurry bubble column and trickle bed reactors. Both of these

Hydraulic noise in reactor circuits and loops, and its effect on nuclear fuel vibration

International Nuclear Information System (INIS)

Card, D.C.

This paper reports the results of an investigation at WNRE to monitor noise levels in reactor circuits and loops, so as to characterize the systems and establish the importance of this noise on fuel and pressure tube vibration. Some of the techniques necessary for in-reactor installations of pressure transducers have been developed and measurements have been obtained in the vertical fuel channels of a very noisy out-reactor loop as well as in the WR-1 reactor circuits. A very quiet out-reactor loop has been constructed to study the vibration behaviour of 37-element fuel bundles in the horizontal CANDU pressurized-heavy water reactor systems. In this facility various types and levels of hydraulic noise are being generated to study their effect on the fuel bundles and flow tube at flow velocities up to approximately 13 m/s. (author)

Design of Multi Bubble Sonoluminescence Reactor for Low Frequency Pressure Radiation

Energy Technology Data Exchange (ETDEWEB)

Choi, Yun Seok; Lee, Jae Young [Handong Global University, Pohang (Korea, Republic of)

2012-05-15

Sonoluminescence phenomenon has been intensively studied due to its extraordinary capability to produce very high temperature and pressure within micro bubbles exposed by the pressure radiation. In general, it has been widely used for the chemical treatment including dissociation of the toxics and synthesis of functional materials such as nano catalysis. As for the nuclear applications, some of researchers tried to realization of the bubble fusion and change of the decay constant. Unfortunately applications to the nuclear industry are very skeptically accepted in the academic society. In spite of all such skepticism, the studies on sonoluminescence still have the room to be explored. In the present study, we investigated the relation among the reactor size, power and frequency of the pressure radiation. Main motivation of the present study came from some mismatch in the degradation rate of TCE in the multibubble sonoluminescence reactors (MBSL reactor) between Lee et al (2011) and Oh and Lee (2010). Both studies utilized horn type ultrasound source with the frequency of 20 kHz. However, the shape and volume of the reactors were different form each other. In the present study, we simply measured the light emission from the luminol solution in the reactors to evaluate the effectiveness of the MBSL. As noted in the study of Lee et al, the hot spot of MBSL dissociate water molecules into OH radicals which dissociate luminol in the solution to emit radiation. Therefore, the intensity and distribution of the radiation of luminol dissociation in the reactor are key index of the population of hot spots. Results and discussions are made by comparing the light emission intensity with different operating powers

One-loop fluctuation-dissipation formula for bubble-wall velocity

International Nuclear Information System (INIS)

Arnold, P.

1993-01-01

The limiting bubble wall velocity during a first-order electroweak phase transition is of interest in scenarios for electroweak baryogenesis. Khlebnikov has recently proposed an interesting method for computing this velocity based on the fluctuation-dissipation theorem. It is demonstrated that at one-loop order this method is identical to simple, earlier techniques for computing the wall velocity based on computing the friction from particles reflecting off or transmitting through the wall in the ideal gas limit

One-loop fluctuation-dissipation formula for bubble-wall velocity

International Nuclear Information System (INIS)

Arnold, P.

1993-01-01

The limiting bubble-wall velocity during a first-order electroweak phase transition is of interest in scenarios for electroweak baryogenesis. Khlebnikov has recently proposed an interesting method for computing this velocity based on the fluctuation-dissipation theorem. I demonstrate that at one-loop order this method is identical to simple, earlier techniques for computing the wall velocity based on computing the friction from particles reflecting off or transmitting through the wall in the ideal gas (''thin-wall'') limit

Experimental and numerical investigation of bubble column reactors

NARCIS (Netherlands)

Bai, W.

2010-01-01

Due to various advantages, such as simple geometry, ease of operation, low operating and maintenance costs, excellent heat and mass transfer characteristics, bubble column reactors are frequently used in chemical, petrochemical, biochemical, pharmaceutical, metallurgical industries for a variety of

Measuring neutron noise induced by travelling air bubbles in a research reactor

International Nuclear Information System (INIS)

Por, G.; Horanyi, S.

1983-05-01

Travelling air bubble experiments carried out in a research reactor confirm an earlier proposed model. The sink structure could be found experimentally in APSD of neutron signals and was used to determine the bubble velocity. The measurements show that neutron detectors measure the velocity of the travelling bubbles, the thermocouples that of the water flow. (author)

Reactor recirculation pump test loop

International Nuclear Information System (INIS)

Taka, Shusei; Kato, Hiroyuki

1979-01-01

A test loop for a reactor primary loop recirculation pumps (PLR pumps) has been constructed at Ebara's Haneda Plant in preparation for production of PLR pumps under license from Byron Jackson Pump Division of Borg-Warner Corporation. This loop can simulate operating conditions for test PLR pumps with 130 per cent of the capacity of pumps for a 1100 MWe BWR plant. A main loop, primary cooling system, water demineralizer, secondary cooling system, instrumentation and control equipment and an electric power supply system make up the test loop. This article describes the test loop itself and test results of two PLR pumps for Fukushima No. 2 N.P.S. Unit 1 and one main circulation pump for HAZ Demonstration Test Facility. (author)

Recurrence relations for three-loop prototypes of bubble diagrams with a mass

International Nuclear Information System (INIS)

Avdeev, L.V.

1995-01-01

Recurrence relations derived via the Chetyrkin-Tkachev method of integration by parts are applied to reduce scalar three-loop bubble (vacuum) diagrams with a mass to a limited number of master integrals. The reduction is implemented as a package of computer programs for analytic evaluation in FORM. The algorithms are applicable to diagrams with any integer powers on the lines in an arbitrary dimension. A physical application is the evaluation of the three-loop QCD correction to the electroweak ρ-parameter. 8 refs., 1 fig

Steam drum level dynamics in a multiple loop natural circulation system of a pressure-tube type boiling water reactor

International Nuclear Information System (INIS)

Jain, Vikas; Kulkarni, P.P.; Nayak, A.K.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

2011-01-01

Highlights: → We have highlighted the problem of drum level dynamics in a multiple loop type NC system using RELAP5 code. → The need of interconnections in steam and liquid spaces close to drum is established. → The steam space interconnections equalize pressure and liquid space interconnections equalize level. → With this scheme, the system can withstand anomalous conditions. → However, the controller is found to be inevitable for inventory balance. - Abstract: Advanced Heavy Water Reactor (AHWR) is a pressure tube type boiling water reactor employing natural circulation as the mode of heat removal under all the operating conditions. Main heat transport system (MHTS) of AHWR is essentially a multi-loop natural circulation system with all the loops connected to each other. Each loop of MHTS has a steam drum that provides for gravity based steam-water separation. Steam drum level is a very critical parameter especially in multi-loop natural circulation systems as large departures from the set point may lead to ineffective separation of steam-water or may affect the driving head. However, such a system is susceptible to steam drum level anomalies under postulated asymmetrical operating conditions among the different quadrants of the core like feedwater flow distribution anomaly among the steam drums or power anomaly among the core quadrants. Analyses were carried out to probe such scenarios and unravel the underlying dynamics of steam drum level using system code RELAP5/Mod3.2. In addition, a scheme to obviate such problem in a passive manner without dependence on level controller was examined. It was concluded that steam drums need to be connected in the liquid as well as steam space to make the system tolerant to asymmetrical operating conditions.

PG-100 helium loop in the MR reactor

International Nuclear Information System (INIS)

Ponomarev-Stepnoj, N.N.; Yakovlev, V.V.; Tikhonov, N.I.

1983-01-01

Main systems and production equipment units of PG-100 helium loop in the MR reactor are described. Possible long-term synchronizing operation of loop and reactor as well as possibility of carrying out life-time tests of spherical fuel elements and materials are shown. Serviceability of spherical fuel elements under conditions similar to the ones of HTGR-50 operation as well as high serviceability of cleanup system accepted for HTGR are verified. Due to low radiation dose the loop is operated without limits, helium losses in the loop don't exceed 0.5%/24 h, taking account of experimental gas sampling

Thermal performance of plate-type loop thermosyphon at sub-atmospheric pressures

International Nuclear Information System (INIS)

Tsoi, Vadim; Chang, Shyy Woei; Chiang Kuei Feng; Huang, Chuan Chin

2011-01-01

This experimental study examines the thermal performance of a newly devised plate-type two-phase loop thermosyphon with cooling applications to electronic boards of telecommunication systems. The evaporation section is configured as the inter-connected multi channels to emulate the bridging boiling mechanism in pulsating thermosyphon. Two thermosyphon plates using water as the coolant with filling ratios (FR) of 0.22 and 0.32 are tested at sub-atmospheric pressures. The vapor-liquid flow images as well as the thermal resistances and effective spreading thermal conductivities are individually measured for each thermosyphon test plate at various heating powers. The high-speed digital images of the vapor-liquid flow structures reveal the characteristic boiling phenomena and the vapor-liquid circulation in the vertical thermosyphon plate, which assist to explore the thermal physics for this type of loop thermosyphon. The bubble agglomeration and pumping action in the inter-connected boiling channels take place at metastable non-equilibrium conditions, leading to the intermittent slug flows with a pulsation character. Such hybrid loop-pulsating thermosyphon permits the vapor-liquid circulation in the horizontal plate. Thermal resistances and spreading thermal conductivities detected from the present thermosyphon plates; the vapor chamber flat plate heat pipe and the copper plate at free and forced convective cooling conditions with both vertical and horizontal orientations are cross-examined. In most telecommunication systems and units, the electrical boards are vertical so that the thermal performance data on the vertical thermosyphon are most relevant to this particular application. - Highlights: → We examine thermal performances of plate-type loop thermosyphon. → Thermal resistances and spreading conductivities are examined. → Bubble agglomeration in inter-connected boiling channels generates intermittent slug flows with pulsations. → Boiling instability

Studies on modelling of bubble driven flows in chemical reactors

Energy Technology Data Exchange (ETDEWEB)

Grevskott, Sverre

1997-12-31

Multiphase reactors are widely used in the process industry, especially in the petrochemical industry. They very often are characterized by very good thermal control and high heat transfer coefficients against heating and cooling surfaces. This thesis first reviews recent advances in bubble column modelling, focusing on the fundamental flow equations, drag forces, transversal forces and added mass forces. The mathematical equations for the bubble column reactor are developed, using an Eulerian description for the continuous and dispersed phase in tensor notation. Conservation equations for mass, momentum, energy and chemical species are given, and the k-{epsilon} and Rice-Geary models for turbulence are described. The different algebraic solvers used in the model are described, as are relaxation procedures. Simulation results are presented and compared with experimental values. Attention is focused on the modelling of void fractions and gas velocities in the column. The energy conservation equation has been included in the bubble column model in order to model temperature distributions in a heated reactor. The conservation equation of chemical species has been included to simulate absorption of CO{sub 2}. Simulated axial and radial mass fraction profiles for CO{sub 2} in the gas phase are compared with measured values. Simulations of the dynamic behaviour of the column are also presented. 189 refs., 124 figs., 1 tab.

Physical aspects of liquid-impelled loop reactors

NARCIS (Netherlands)

Sonsbeek, van H.

1992-01-01

The liquid-impelled loop reactor (LLR) is a reactor that consists of two parts : the main tube and the circulation tube. Both parts are in open connection at the bottom and at the top. The reactor is filled with a liquid phase: the continuous phase. Another liquid phase is injected in the

The effects of baffles and gas superficial velocity on a bubble fluidized bed reactor's applications

International Nuclear Information System (INIS)

Ghorbanpour, A.; Ghannadi Maragheh, M.; Mallah, M. H.

2008-01-01

Baffles are used for decreasing bubbles diameter in order to increase the conversion rate along the bubbling fluidized bed reactors. The appearance of this phenomenon is due to bursting of the bubbles during the pass of bubbles from baffles. In this work, a computerized modeling and simulation have been performed in order to obtain a fundamental knowledge of the influence of the baffles on the bubble diameter and the specific mass transfer area. The height of the bed is 5 meters and its diameter is 0.3 meter. Baffles are located at 1 and 2 meters from the bottom of the bed. A two phase model together with a comprehensive fluid dynamical description of bubbling fluidized is presented. The effects of baffles and gas superficial velocity on the operating behavior of fluidized bed reactors are considered. The results are compared to the previously reported documents, and the experiments which have been carried out. MATLAB software is used in this simulation

Coolant radiolysis studies in the high temperature, fuelled U-2 loop in the NRU reactor

International Nuclear Information System (INIS)

Elliot, A.J.; Stuart, C.R.

2008-06-01

An understanding of the radiolysis-induced chemistry in the coolant water of nuclear reactors is an important key to the understanding of materials integrity issues in reactor coolant systems. Significant materials and chemistry issues have emerged in Pressurized Water Reactors (PWR), Boiling Water Reactors (BWR) and CANDU reactors that have required a detailed understanding of the radiation chemistry of the coolant. For each reactor type, specific computer radiolysis models have been developed to gain insight into radiolysis processes and to make chemistry control adjustments to address the particular issue. In this respect, modelling the radiolysis chemistry has been successful enough to allow progress to be made. This report contains a description of the water radiolysis tests performed in the U-2 loop, NRU reactor in 1995, which measured the CHC under different physical conditions of the loop such as temperature, reactor power and steam quality. (author)

Modelling and characterization of an airlift-loop bioreactor

NARCIS (Netherlands)

Verlaan, P.

1987-01-01

An airlift-loop reactor is a bioreactor for aerobic biotechnological processes. The special feature of the ALR is the recirculation of the liquid through a downcomer connecting the top and the bottom of the main bubbling section. Due to the high circulation-flow rate, efficient mixing and

Loop-type FBR reactor

International Nuclear Information System (INIS)

Ogura, Kenji; Kimura, Kimitaka; Jinbo, Masaichi; Hirayama, Hiroshi; Taguchi, Junzo; Hirata, Noriaki; Ozaki, Kenji; Maruyama, Shigeki.

1996-01-01

The inside of a vessel of an intermediate heat exchanger is divided vertically by a partition wall into a high temperature plenum region and a low temperature plenum region, a perforated horizontal plate is disposed in a horizontal direction at the upper portion and a flow shroud is disposed so as to surround the upper outside of the intermediate heat exchanger while passing through a lid from a perforated hole of the perforated horizontal plate. In addition, there is disposed a cylinder passing through the partition wall and the horizontal perforated plate for inserting a liquid surface penetrating equipment. The cylinder has an upper end opened above the liquid level of a liquid metal during normal operation and below the liquid level of the liquid metal during shut down of the reactor, and the lower end is opened in a lower plenum region. Vibrations of liquid level due to the high temperature liquid metal inflown from a hot leg pipeline to the inside of the vessel of the intermediate heat exchanger are suppressed by the perforated horizontal plate during reactor operation. On the other hand, upon shut down of the reactor, since the liquid level rises up to the upper portion of the cylinder, the liquid metal at low temperature inflows into the lower plenum region, and the liquid metal at high temperature above the horizontal perforated plate is eliminated in an early stage. (N.H.)

Treatment of sodium spills and leakage detection at loop-type fast reactors

Energy Technology Data Exchange (ETDEWEB)

Foerster, K; Fortmann, M; Lang, H; Moellerfeld, H [Interatom, Bergisch Gladbach (Germany)

1979-03-01

Sodium spills are of great importance in the safety analysis for sodium cooled nuclear plants. Large leakages can lead to a depletion of the heat transfer system and cause the loss of cooling of the reactor. Further the hot sodium may attack structural materials. In areas with air atmosphere large amounts of sodium can burn and cause great damages. Therefore the control of large leakages is an indispensable task in design and construction of sodium cooled reactor systems. Because of the typical arrangement of widespread long pipe systems loop type plants are subject to a gradually greater risk of damage than pool type plants. The sodium catching devices of the SNR-300 are described and their function is illustrated as an example for the treatment of large spills. Since the equipment for the control of large amounts of leaking sodium is very expensive, great efforts are made in order to save costs and to decrease safety problems. It is aimed to minimize the probability of such events to a degree that they no longer are to be considered realistic. The advantageous operating conditions and the favourable material properties support this aim. Under the well known keyword 'leak-before-rupture' criterion this task is pursued. Crack growth measurements are made at structural materials under LMFBR conditions, and leakage detecting systems are being developed. Some test results concerning this task are described. Despite the fact that there are good chances to verify the leak-before-rupture criterion it is assumed that certain hypothetical accidents occur, which are to be considered in the design of the reactor plant. The extremely improbable Bethe-Tait-accident (HCDA) is such an event. It would lead to a super spill, that means to the complete depletion of the reactor tank. For the SNR-300 plant a system is provided that is able to catch this super spill and the core melt. This core catcher must withstand the high temperatures and remove the decay heat. The purpose of this

Treatment of sodium spills and leakage detection at loop-type fast reactors

International Nuclear Information System (INIS)

Foerster, K.; Fortmann, M.; Lang, H.; Moellerfeld, H.

1979-01-01

Sodium spills are of great importance in the safety analysis for sodium cooled nuclear plants. Large leakages can lead to a depletion of the heat transfer system and cause the loss of cooling of the reactor. Further the hot sodium may attack structural materials. In areas with air atmosphere large amounts of sodium can burn and cause great damages. Therefore the control of large leakages is an indispensable task in design and construction of sodium cooled reactor systems. Because of the typical arrangement of widespread long pipe systems loop type plants are subject to a gradually greater risk of damage than pool type plants. The sodium catching devices of the SNR-300 are described and their function is illustrated as an example for the treatment of large spills. Since the equipment for the control of large amounts of leaking sodium is very expensive, great efforts are made in order to save costs and to decrease safety problems. It is aimed to minimize the probability of such events to a degree that they no longer are to be considered realistic. The advantageous operating conditions and the favourable material properties support this aim. Under the well known keyword 'leak-before-rupture' criterion this task is pursued. Crack growth measurements are made at structural materials under LMFBR conditions, and leakage detecting systems are being developed. Some test results concerning this task are described. Despite the fact that there are good chances to verify the leak-before-rupture criterion it is assumed that certain hypothetical accidents occur, which are to be considered in the design of the reactor plant. The extremely improbable Bethe-Tait-accident (HCDA) is such an event. It would lead to a super spill, that means to the complete depletion of the reactor tank. For the SNR-300 plant a system is provided that is able to catch this super spill and the core melt. This core catcher must withstand the high temperatures and remove the decay heat. The purpose of this

Selective purge for hydrogenation reactor recycle loop

Science.gov (United States)

Baker, Richard W.; Lokhandwala, Kaaeid A.

2001-01-01

Processes and apparatus for providing improved contaminant removal and hydrogen recovery in hydrogenation reactors, particularly in refineries and petrochemical plants. The improved contaminant removal is achieved by selective purging, by passing gases in the hydrogenation reactor recycle loop or purge stream across membranes selective in favor of the contaminant over hydrogen.

The radiation safety assessment of the heating loop of district heating reactors

International Nuclear Information System (INIS)

Liu Yuanzhong

1993-01-01

The district heating reactors are used to supply heating to the houses in cities. The concerned problems are whether the radioactive materials reach the heated houses through heating loop, and whether the safety of the dwellers can be ensured. In order to prevent radioactive materials getting into the heated houses, the district heating reactors have three loops, namely, primary loop, intermediate loop, and heating loop. In the paper, the measures of preventing radioactive materials getting into the heating loop are presented, and the possible sources of the radioactivity in the water of the intermediate loop and the heating loop are given. The regulatory aim limit of radioactive concentration in the water of the intermediate loop is put forward, which is 18.5 Bq/l. Assuming that specific radioactivity of the water of contaminated intermediate loop is up to 18.5 Bq/l, the maximum concentration of radionuclides in water of the heating loop is calculated for the normal operation and the accident of district heating reactor. The results show that the maximum possible concentration is 5.7 x 10 -3 Bq/l. The radiation safety assessment of the heating loop is made out. The conclusions are that the district heating reactors do not bring any harmful impact to the dwellers, and the safety of the dwellers can be safeguarded completely

Steam drum level control studies of a natural circulation multi loop reactor

Energy Technology Data Exchange (ETDEWEB)

Kumar, Rajesh; Contractor, A.D.; Srivastava, Abhishek; Lele, H.G. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Reactor Safety Div.; Vaze, K.K. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Reactor Design and Development Group

2013-12-15

The proposed heavy water moderated and light water cooled pressure tube type boiling water reactor works on natural circulation at all power levels. It has parallel inter-connected loops with 452 boiling channels in the main heat transport system configuration. These multiple (four) interconnected loops influence the steam drum level control adversely through the common reactor inlet header. Alternate design studies made earlier for efficient control of SD levels have shown favorable results. This has lead to explore further the present scheme with the compartmentalization of CRIH into four compartments catering to four loops separately. The conventional 3-element level control has been found to be working satisfactorily. The interconnections between ECCS header and inlet header compartments have also increased the safety margin for various LOCA and design basis events. The paper deals with the SD level control aspects for this novel MHT configuration which has been analyzed for various PIEs (Postulated Initiating Events) and found to be satisfactory. (orig.)

Helium bubbles aggravated defects production in self-irradiated copper

Science.gov (United States)

Wu, FengChao; Zhu, YinBo; Wu, Qiang; Li, XinZhu; Wang, Pei; Wu, HengAn

2017-12-01

Under the environment of high radiation, materials used in fission and fusion reactors will internally accumulate numerous lattice defects and bubbles. With extensive studies focused on bubble resolution under irradiation, the mutually effects between helium bubbles and displacement cascades in irradiated materials remain unaddressed. Therefore, the defects production and microstructure evolution under self-irradiation events in vicinity of helium bubbles are investigated by preforming large scale molecular dynamics simulations in single-crystal copper. When subjected to displacement cascades, distinguished bubble resolution categories dependent on bubble size are observed. With the existence of bubbles, radiation damage is aggravated with the increasing bubble size, represented as the promotion of point defects and dislocations. The atomic mechanisms of heterogeneous dislocation structures are attributed to different helium-vacancy cluster modes, transforming from the resolved gas trapped with vacancies to the biased absorption of vacancies by the over-pressured bubble. In both cases, helium impedes the recombination of point defects, leading to the accelerated formation of interstitial loops. The results and insight obtained here might contribute to understand the underlying mechanism of transmutant solute on the long-term evolution of irradiated materials.

PRE design of a molten salt thorium reactor loop

International Nuclear Information System (INIS)

Caire, Jean-Pierre; Roure, Anthony

2007-01-01

This study is a contribution to the 2004 PCR-RSF program of the Centre National de la Recherche Scientifique (CNRS) devoted to research on high temperature thorium molten salt reactors. A major issue of high temperature molten salt reactors is the very large heat duty to be transferred from primary to secondary loop of the reactor with minimal thermal losses. A possible inner loop made of a series of conventional graphite filter plate exchangers, pipes and pumps was investigated. The loop was assumed to use two counter current flows of the same LiF, BeF 2 , ZrF 4 , UF 4 molten salt flowing through the reactor. The 3D model used the coupling of k-ε turbulent Navier-Stokes equations and thermal applications of the Heat Transfer module of COMSOL Multiphysics. For a reactor delivering 2700 MWth, the model required a set of 114 identical exchangers. Each one was optimized to limit the heat losses to 2882 W. The pipes made of a succession of graphite, ceramics, Hastelloy-N alloy and insulating Microtherm layers led to a thermal loss limited to 550 W per linear meter. In such conditions, the global thermal losses represent only 0.013% of the reactor thermal power for elements covered with an insulator only 3 cm thick. (author)

Packed Bed Reactor Technology for Chemical-Looping Combustion

NARCIS (Netherlands)

Noorman, S.; Sint Annaland, van M.; Kuipers, J.A.M.

2007-01-01

Chemical-looping combustion (CLC) has emerged as an alternative for conventional power production processes to intrinsically integrate power production and CO2 capture. In this work a new reactor concept for CLC is proposed, based on dynamically operated packed bed reactors. With analytical

Hydrodynamics of a hybrid circulating fluidized bed reactor with a partitioned loop seal system

Energy Technology Data Exchange (ETDEWEB)

Bae, Dal-Hee; Moon, Jong-Ho; Jin, Gyoung Tae; Shun, Dowon [Korea Institute of Energy Research, Daejeon (Korea, Republic of); Yun, Minyoung; Park, Chan Seung; Norbeck, Joseph M. [University of California, Riverside (United States)

2015-07-15

A circulating fluidized bed (CFB) with a hybrid design has been developed and optimized for steam hydrogasification. The hybrid CFB is composed of a bubbling fluidized bed (BFB) type combustor and a fast fluidized bed (FB) type gasifier. Char is burnt in the combustor and the generated heat is supplied to the gasifier along with the bed materials. Two different types of fluidized beds are connected to each other with a newly developed partitioned loop seal to avoid direct contact between two separate gas streams flowing in each fluidized bed. Gas mixing tests were carried out with Air and Argon in a cold model hybrid CFB to test the loop seal efficiency. Increase in solid inventory in the loop seal can improve the gas separation efficiency. It can be realized at higher gas velocity in fast bed and with higher solid inventory in the loop seal system. In addition, bed hydrodynamics was investigated with varying gas flow conditions and particle sizes in order to obtain a full understanding of changes of solid holdup in the FB. The solid holdup in the FB increased with increasing gas velocity in the BFB. Conversely, increase in gas velocity in the FB contributed to reducing the solid holdup in the FB. It was observed that changing the particle size of bed material does not have a big impact on hydrodynamic parameters.

Hydrodynamics of a hybrid circulating fluidized bed reactor with a partitioned loop seal system

International Nuclear Information System (INIS)

Bae, Dal-Hee; Moon, Jong-Ho; Jin, Gyoung Tae; Shun, Dowon; Yun, Minyoung; Park, Chan Seung; Norbeck, Joseph M.

2015-01-01

A circulating fluidized bed (CFB) with a hybrid design has been developed and optimized for steam hydrogasification. The hybrid CFB is composed of a bubbling fluidized bed (BFB) type combustor and a fast fluidized bed (FB) type gasifier. Char is burnt in the combustor and the generated heat is supplied to the gasifier along with the bed materials. Two different types of fluidized beds are connected to each other with a newly developed partitioned loop seal to avoid direct contact between two separate gas streams flowing in each fluidized bed. Gas mixing tests were carried out with Air and Argon in a cold model hybrid CFB to test the loop seal efficiency. Increase in solid inventory in the loop seal can improve the gas separation efficiency. It can be realized at higher gas velocity in fast bed and with higher solid inventory in the loop seal system. In addition, bed hydrodynamics was investigated with varying gas flow conditions and particle sizes in order to obtain a full understanding of changes of solid holdup in the FB. The solid holdup in the FB increased with increasing gas velocity in the BFB. Conversely, increase in gas velocity in the FB contributed to reducing the solid holdup in the FB. It was observed that changing the particle size of bed material does not have a big impact on hydrodynamic parameters

Chemical looping reactor system design double loop circulating fluidized bed (DLCFB)

Energy Technology Data Exchange (ETDEWEB)

Bischi, Aldo

2012-05-15

Chemical looping combustion (CLC) is continuously gaining more importance among the carbon capture and storage (CCS) technologies. It is an unmixed combustion process which takes place in two steps. An effective way to realize CLC is to use two interconnected fluidized beds and a metallic powder circulating among them, acting as oxygen carrier. The metallic powder oxidizes at high temperature in one of the two reactors, the air reactor (AR). It reacts in a highly exothermic reaction with the oxygen of the injected fluidising air. Afterwards the particles are sent to the other reactor where the fuel is injected, the fuel reactor (FR). There, they transport heat and oxygen necessary for the reaction with the injected fuel to take place. At high temperatures, the particle's oxygen reacts with the fuel producing Co2 and steam, and the particles are ready to start the loop again. The overall reaction, the sum of the enthalpy changes of the oxygen carrier oxidation and reduction reactions, is the same as for the conventional combustion. Two are the key features, which make CLC promising both for costs and capture efficiency. First, the high inherent irreversibility of the conventional combustion is avoided because the energy is utilized stepwise. Second, the Co2 is intrinsically separated within the process; so there is in principle no need either of extra carbon capture devices or of expensive air separation units to produce oxygen for oxy-combustion. A lot of effort is taking place worldwide on the development of new chemical looping oxygen carrier particles, reactor systems and processes. The current work is focused on the reactor system: a new design is presented, for the construction of an atmospheric 150kWth prototype working with gaseous fuel and possibly with inexpensive oxygen carriers derived from industrial by-products or natural minerals. It consists of two circulating fluidized beds capable to operate in fast fluidization regime; this will increase the

Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

Energy Technology Data Exchange (ETDEWEB)

Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-12-15

A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. The various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.

A prediction for bubbling geometries

OpenAIRE

Okuda, Takuya

2007-01-01

We study the supersymmetric circular Wilson loops in N=4 Yang-Mills theory. Their vacuum expectation values are computed in the parameter region that admits smooth bubbling geometry duals. The results are a prediction for the supergravity action evaluated on the bubbling geometries for Wilson loops.

Emergency core cooling system for LMFBR type reactors

International Nuclear Information System (INIS)

Tamano, Toyomi; Fukutomi, Shigeki.

1980-01-01

Purpose: To enable elimination of decay heat in an LMFBR type reactor by securing natural cycling force in any state and securing reactor core cooling capacity even when both an external power supply and an emergency power supply are failed in emergency case. Method: Heat insulating material portion for surrounding a descent tube of a steam drum provided at high position for obtaining necessary flow rate for flowing resistance is removed from heat transmitting surface of a recycling type steam generator to provide a heat sink. That is, when both an external power supply and an emergency power supply are failed in emergency, the heat insulator at part of a steam generator recycling loop is removed to produce natural cycling force between it and the heat transmitting portion of the steam generator as a heat source for the heat sink so as to secure the flow rate of the recycling loop. When the power supply is failed in emergency, the heat removing capacity of the steam generator is secured so as to remove the decay heat produced in the reactor core. (Yoshihara, H.)

Simulation of the Fuel Reactor of a Coal-Fired Chemical Looping Combustor

Science.gov (United States)

Mahalatkar, Kartikeya; O'Brien, Thomas; Huckaby, E. David; Kuhlman, John

2009-06-01

Responsible carbon management (CM) will be required for the future utilization of coal for power generation. CO2 separation is the more costly component of CM, not sequestration. Most methods of capture require a costly process of gas separation to obtain a CO2-rich gas stream. However, recently a process termed Chemical Looping Combustion (CLC) has been proposed, in which an oxygen-carrier is used to provide the oxygen for combustion. This process quite naturally generates a separate exhaust gas stream containing mainly H2O and CO2 but requires two reaction vessels, an Air Reactor (AR) and a Fuel Reactor (FR). The carrier (M for metal, the usual carrier) is oxidized in the AR. This highly exothermic process provides heat for power generation. The oxidized carrier (MO) is separated from this hot, vitiated air stream and transported to the FR where it oxidizes the hydrocarbon fuel, yielding an exhaust gas stream of mainly H2O and CO2. This process is usually slightly endothermic so that the carrier must also transport the necessary heat of reaction. The reduced carrier (M) is then returned to the air reactor for regeneration, hence the term "looping." The net chemical reaction and energy release is identical to that of conventional combustion of the fuel. However, CO2 separation is easily achieved, the only operational penalty being the slight pressure losses required to circulate the carrier. CLC requires many unit operations involving gas-solid or granular flow. To utilize coal in the fuel reactor, in either a moving bed or bubbling fluidized bed, the granular flow is especially critical. The solid coal fuel must be heated by the recycled metal oxide, driving off moisture and volatile material. The remaining char must be gasified by H2O (or CO2), which is recycled from the product stream. The gaseous product of these reactions must then contact the MO before leaving the bed to obtain complete conversion to H2O and CO2. Further, the reduced M particles must be

Steam generating system in LMFBR type reactors

International Nuclear Information System (INIS)

Kurosawa, Katsutoshi.

1984-01-01

Purpose: To suppress the thermal shock loads to the structures of reactor system and secondary coolant system, for instance, upon plant trip accompanying turbine trip in the steam generation system of LMFBR type reactors. Constitution: Additional feedwater heater is disposed to the pipeway at the inlet of a steam generator in a steam generation system equipped with a closed loop extended from a steam generator by way of a gas-liquid separator, a turbine and a condensator to the steam generator. The separated water at high temperature and high pressure from a gas-liquid separator is heat exchanged with coolants flowing through the closed loop of the steam generation system in non-contact manner and, thereafter, introduced to a water reservoir tank. This can avoid the water to be fed at low temperature as it is to the steam generator, whereby the thermal shock loads to the structures of the reactor system and the secondary coolant system can be suppressed. (Moriyama, K.)

Model experiments on simulation of the WWER water-chemical conditions at loop facilities of the MIR reactor

International Nuclear Information System (INIS)

Benderskaya, O.S.; Zotov, E.A.; Kuprienko, V.A.; Ovchinnikov, V.A.

1999-01-01

The experiments on simulation of the WWER type reactors water-chemical conditions have been started at the State Scientific Center RIAR. These experiments are being conducted at the multi-loop research MIR reactor at the PVK-2 loop facility. The dosage stand was created. It allows introduction of boric acid, potassium and lithium hydroxides, ammonia solutions and gaseous hydrogen. Corrosion tests of the Russian E-635 and E-110 alloys are being conducted at the PVK-2 loop under the WWER water-chemical conditions. If necessary, fuel elements are periodically extracted from the reactor to perform visual examination, to measure their length, diameter, to remove the deposits from the claddings, to measure the burnup and to distribute the fission products over the fuel element by gamma-spectrometry. The chemical analytical 'on line' equipment produced by the ORBISPHERE Laboratory (Switzerland) will be commissioned in the nearest future to measure concentration of the dissolved hydrogen and oxygen as well as pH and specific conductivity. The objective of the report is to familiarize the participants of the IAEA Technical Committee with the capabilities of performing the model water-chemical experiments under the MIR reactor loop facility conditions. (author)

A model to estimate volume change due to radiolytic gas bubbles and thermal expansion in solution reactors

International Nuclear Information System (INIS)

Souto, F.J.; Heger, A.S.

2001-01-01

To investigate the effects of radiolytic gas bubbles and thermal expansion on the steady-state operation of solution reactors at the power level required for the production of medical isotopes, a calculational model has been developed. To validate this model, including its principal hypotheses, specific experiments at the Los Alamos National Laboratory SHEBA uranyl fluoride solution reactor were conducted. The following sections describe radiolytic gas generation in solution reactors, the equations to estimate the fuel solution volume change due to radiolytic gas bubbles and thermal expansion, the experiments conducted at SHEBA, and the comparison of experimental results and model calculations. (author)

Modeling of reaction kinetics in bubbling fluidized bed biomass gasification reactor

Energy Technology Data Exchange (ETDEWEB)

Thapa, R.K.; Halvorsen, B.M. [Telemark University College, Kjolnes ring 56, P.O. Box 203, 3901 Porsgrunn (Norway); Pfeifer, C. [University of Natural Resources and Life Sciences, Vienna (Austria)

2013-07-01

Bubbling fluidized beds are widely used as biomass gasification reactors as at the biomass gasification plant in Gussing, Austria. The reactor in the plant is a dual circulating bubbling fluidized bed gasification reactor. The plant produces 2MW electricity and 4.5MW heat from the gasification of biomass. Wood chips as biomass and olivine particles as hot bed materials are fluidized with high temperature steam in the reactor. As a result, biomass undergoes endothermic chemical reaction to produce a mixture of combustible gases in addition to some carbon-dioxide (CO2). The combustible gases are mainly hydrogen (H2), carbon monoxide (CO) and methane (CH4). The gas is used to produce electricity and heat via utilization in a gas engine. Alternatively, the gas is further processed for gaseous or liquid fuels, but still on the process of development level. Composition and quality of the gas determine the efficiency of the reactor. A computational model has been developed for the study of reaction kinetics in the gasification rector. The simulation is performed using commercial software Barracuda virtual reactor, VR15. Eulerian-Lagrangian approach in coupling of gas-solid flow has been implemented. Fluid phase is treated with an Eulerian formulation. Discrete phase is treated with a Lagrangian formulation. Particle-particle and particle-wall interactions and inter-phase heat and mass transfer have been taken into account. Series of simulations have been performed to study model prediction of the gas composition. The composition is compared with data from the gasifier at the CHP plant in Güssing, Austria. The model prediction of the composition of gases has good agreements with the result of the operating plant.

Liquid-metal-gas heat exchanger for HTGR type reactors

International Nuclear Information System (INIS)

Werth, G.

1980-01-01

The aim of this study is to investigate the heat transfer characteristics of a liquid metal heat exchanger (HE) for a helium-cooled high temperature reactor. A tube-type heat exchanger is considered as well as two direct exchangers: a bubble-type heat exchanger and a heat exchanger according to the spray principle. Experiments are made in order to determine the gas content of bubble-type heat exchangers, the dependence of the droplet diameter on the nozzle diameter, the falling speed of the droplets, the velocity of the liquid jet, and the temperature variation of liquid jets. The computer codes developed for HE calculation are structured so that they may be used for gas/liquid HE, too. Each type of HE that is dealt with is designed by accousting for a technical and an economic assessment. The liquid-lead jet spray is preferred to all other types because of its small space occupied and its simple design. It shall be used in near future in the HTR by the name of lead/helium HE. (GL) [de

Making continuous bubble type polyethylene foam incombustible

International Nuclear Information System (INIS)

Kaji, Kanako; Hatada, Motoyoshi; Yoshizawa, Iwao; Komai, Kuniaki; Kohara, Choji.

1989-01-01

Since continuous bubble type plastic foam has excellent compression characteristics and sound absorption characteristics, it has been widely used as cushion material, sealing material, sound insulating material and so on. However, the most part of plastic foam is taken by air, therefore at the time of fires, it becomes a very dangerous material. At present, the material used mostly as the seat cushions for airliners, railroad coaches, automobiles and others is polyurethane foam, but since it contains C-N couples in its molecules, it is feared to generate cyanic gas according to the condition of combustion. As the plastic foam that does not generate harmful gas at the time of fires, there is continuous bubble type polyethylene which is excellent in its weathering property and chemical resistance. A reactive, phosphorus-containing oligomer has large molecular weight and two or more double couplings in a molecule, therefore, it does not enter the inside of polyethylene, and polymerizes and crosslinks on the surfaces of bubble walls in the foam, accordingly it is expected that the apparent graft polymerization is carried out, and it is very effective for making polyethylene foam incombustible. The method of making graft foam, the properties of graft foam and so on are reported. When the graft polymerization of this oligomer to continuous bubble type polyethylene foam was tried, highly incombustible polyethylene foam was obtained. (K.I.)

Primary loop simulation of the SP-100 space nuclear reactor

International Nuclear Information System (INIS)

Borges, Eduardo M.; Braz Filho, Francisco A.; Guimaraes, Lamartine N.F.

2011-01-01

Between 1983 and 1992 the SP-100 space nuclear reactor development project for electric power generation in a range of 100 to 1000 kWh was conducted in the USA. Several configurations were studied to satisfy different mission objectives and power systems. In this reactor the heat is generated in a compact core and refrigerated by liquid lithium, the primary loops flow are controlled by thermoelectric electromagnetic pumps (EMTE), and thermoelectric converters produce direct current energy. To define the system operation point for an operating nominal power, it is necessary the simulation of the thermal-hydraulic components of the space nuclear reactor. In this paper the BEMTE-3 computer code is used to EMTE pump design performance evaluation to a thermalhydraulic primary loop configuration, and comparison of the system operation points of SP-100 reactor to two thermal powers, with satisfactory results. (author)

Production of polygalacturonases by Aspergillus oryzae in stirred tank and internal- and external-loop airlift reactors.

Science.gov (United States)

Fontana, Roselei Claudete; da Silveira, Maurício Moura

2012-11-01

The production of endo- and exo-polygalacturonase (PG) by Aspergillus oryzae was assessed in stirred tank reactors (STRs), internal-loop airlift reactors (ILARs) and external-loop airlift reactors (ELARs). For STR production, we compared culture media formulated with either pectin (WBE) or partially hydrolyzed pectin. The highest enzyme activities were obtained in medium that contained 50% pectin in hydrolyzed form (WBE5). PG production in the three reactor types was compared for WBE5 and low salt WBE medium, with additional salts added at 48, 60 and 72h (WBES). The ELARs performed better than the ILARs in WBES medium where the exo-PG was the same concentration as for STRs and the endo-PG was 20% lower. These results indicate that PG production is higher under experimental conditions that result in higher cell growth with minimum pH values less than 3.0. Copyright © 2012 Elsevier Ltd. All rights reserved.

Cutting bubbles with a single wire

NARCIS (Netherlands)

Baltussen, M.W.; Segers, Q.I.E.; Kuipers, J.A.M.; Deen, N.G.

2017-01-01

Many gas-liquid-solid contactors, such as trickle bed and bubble slurry columns, suffer from heat and mass transfer limitations. To overcome these limitations, new micro-structured bubble column reactor is proposed. In this reactor, a catalyst coated wire mesh is introduced in a bubble column to cut

Biofiltration of Air/Styrene and Air/Styrene/Acetone mixtures in a bubble column reactor

OpenAIRE

Vieira, Ana

2009-01-01

The goal of this work was the treatment of polluted waste gases in a bubble column reactor (BCR), in order to determinate the maximum value of reactor’s efficiency (RE), varying the inlet concentration (C in) of the pollutants. The gaseous mixtures studied were: (i) air with styrene and (ii) air with styrene and acetone. The liquid phase used to contain the biomass in the reactor was a basal salt medium (BSM), fundamental for the microorganisms’ development. The reactor used in this pro...

Investigation of the effects of radiolytic-gas bubbles on the long-term operation of solution reactors for medical-isotope production

Science.gov (United States)

Souto Mantecon, Francisco Javier

One of the most common and important medical radioisotopes is 99Mo, which is currently produced using the target irradiation technology in heterogeneous nuclear reactors. The medical isotope 99Mo can also be produced from uranium fission using aqueous homogeneous solution reactors. In solution reactors, 99Mo is generated directly in the fuel solution, resulting in potential advantages when compared with the target irradiation process in heterogeneous reactors, such as lower reactor power, less waste heat, and reduction by a factor of about 100 in the generation of spent fuel. The commercial production of medical isotopes in solution reactors requires steady-state operation at about 200 kW. At this power regime, the formation of radiolytic-gas bubbles creates a void volume in the fuel solution that introduces a negative coefficient of reactivity, resulting in power reduction and instabilities that may impede reactor operation for medical-isotope production. A model has been developed considering that reactivity effects are due to the increase in the fuel-solution temperature and the formation of radiolytic-gas bubbles. The model has been validated against experimental results from the Los Alamos National Laboratory uranyl fluoride Solution High-Energy Burst Assembly (SHEBA), and the SILENE uranyl nitrate solution reactor, commissioned at the Commissariat a l'Energie Atomique, in Valduc, France. The model shows the feasibility of solution reactors for the commercial production of medical isotopes and reveals some of the important parameters to consider in their design, including the fuel-solution type, 235U enrichment, uranium concentration, reactor vessel geometry, and neutron reflectors surrounding the reactor vessel. The work presented herein indicates that steady-state operation at 200 kW can be achieved with a solution reactor consisting of 120 L of uranyl nitrate solution enriched up to 20% with 235U and a uranium concentration of 145 kg/m3 in a graphite

PROGRESS TOWARDS MODELING OF FISCHER TROPSCH SYNTHESIS IN A SLURRY BUBBLE COLUMN REACTOR

Energy Technology Data Exchange (ETDEWEB)

Donna Post Guillen; Tami Grimmett; Anastasia M. Gandrik; Steven P. Antal

2010-11-01

The Hybrid Energy Systems Testing (HYTEST) Laboratory is being established at the Idaho National Laboratory to develop and test hybrid energy systems with the principal objective to safeguard U.S. Energy Security by reducing dependence on foreign petroleum. A central component of the HYTEST is the slurry bubble column reactor (SBCR) in which the gas-to-liquid reactions will be performed to synthesize transportation fuels using the Fischer Tropsch (FT) process. SBCRs are cylindrical vessels in which gaseous reactants (for example, synthesis gas or syngas) is sparged into a slurry of liquid reaction products and finely dispersed catalyst particles. The catalyst particles are suspended in the slurry by the rising gas bubbles and serve to promote the chemical reaction that converts syngas to a spectrum of longer chain hydrocarbon products, which can be upgraded to gasoline, diesel or jet fuel. These SBCRs operate in the churn-turbulent flow regime which is characterized by complex hydrodynamics, coupled with reacting flow chemistry and heat transfer, that effect reactor performance. The purpose of this work is to develop a computational multiphase fluid dynamic (CMFD) model to aid in understanding the physico-chemical processes occurring in the SBCR. Our team is developing a robust methodology to couple reaction kinetics and mass transfer into a four-field model (consisting of the bulk liquid, small bubbles, large bubbles and solid catalyst particles) that includes twelve species: (1) CO reactant, (2) H2 reactant, (3) hydrocarbon product, and (4) H2O product in small bubbles, large bubbles, and the bulk fluid. Properties of the hydrocarbon product were specified by vapor liquid equilibrium calculations. The absorption and kinetic models, specifically changes in species concentrations, have been incorporated into the mass continuity equation. The reaction rate is determined based on the macrokinetic model for a cobalt catalyst developed by Yates and Satterfield [1]. The

Hardware-in-the-Loop Simulation for the Automatic Power Control System of Research Reactors

International Nuclear Information System (INIS)

Fikry, R.M.; Shehata, S.A.; Elaraby, S.M.; Mahmoud, M.I.; Elbardini, M.M.

2009-01-01

Designing and testing digital control system for any nuclear research reactor can be costly and time consuming. In this paper, a rapid, low-cost proto typing and testing procedure for digital controller design is proposed using the concept of Hardware-In- The-Loop (HIL). Some of the control loop components are real hardware components and thc others are simulated. First, the whole system is modeled and tested by Real- Time Simulation (RTS) using conventional simulation techniques such as MATLAB / SIMULINK. Second the Hardware-in-the-Ioop simulation is tested using Real-Time Windows Target in MATLAB and Visual C++. The control parts are included as hardware components which are the reactor control rod and its drivers. Two kinds of controllers are studied, Proportional derivative (PD) and Fuzzy controller, An experimental setup for the hardware used in HIL concept for the control of the nuclear research reactor has been realized. Experimental results are obtained and compared with the simulation results. The experimental results indicate the validation of HIL method in this domain

ADVANCED DIAGNOSTIC TECHNIQUES FOR THREE-PHASE SLURRY BUBBLE COLUMN REACTORS (SBCR)

Energy Technology Data Exchange (ETDEWEB)

M.H. Al-Dahhan; M.P. Dudukovic; L.S. Fan

2001-07-25

This report summarizes the accomplishment made during the second year of this cooperative research effort between Washington University, Ohio State University and Air Products and Chemicals. The technical difficulties that were encountered in implementing Computer Automated Radioactive Particle Tracking (CARPT) in high pressure SBCR have been successfully resolved. New strategies for data acquisition and calibration procedure have been implemented. These have been performed as a part of other projects supported by Industrial Consortium and DOE via contract DE-2295PC95051 which are executed in parallel with this grant. CARPT and Computed Tomography (CT) experiments have been performed using air-water-glass beads in 6 inch high pressure stainless steel slurry bubble column reactor at selected conditions. Data processing of this work is in progress. The overall gas holdup and the hydrodynamic parameters are measured by Laser Doppler Anemometry (LDA) in 2 inch slurry bubble column using Norpar 15 that mimic at room temperature the Fischer Tropsch wax at FT reaction conditions of high pressure and temperature. To improve the design and scale-up of bubble column, new correlations have been developed to predict the radial gas holdup and the time averaged axial liquid recirculation velocity profiles in bubble columns.

Description of the two-loop RELAP5 model of the L-Reactor at the Savannah River Site

International Nuclear Information System (INIS)

Cozzuol, J.M.; Davis, C.B.

1989-12-01

A two-loop RELAP5 input model of the L-Reactor at the Savannah River Site (SRS) was developed to support thermal-hydraulic analysis of SRS reactors. The model was developed to economically evaluate potential design changes. The primary simplifications in the model were in the number of loops and the detail in the moderator tank. The six loops in the reactor were modeled with two loops, one representing a single loop and the other representing five combined loops. The model has undergone a quality assurance review. This report describes the two-loop model, its limitations, and quality assurance. 29 refs., 18 figs., 10 tabs

A reduced fidelity model for the rotary chemical looping combustion reactor

KAUST Repository

Iloeje, Chukwunwike O.; Zhao, Zhenlong; Ghoniem, Ahmed F.

2017-01-01

The rotary chemical looping combustion reactor has great potential for efficient integration with CO capture-enabled energy conversion systems. In earlier studies, we described a one-dimensional rotary reactor model, and used it to demonstrate

Experimental facilities for PEC reactor design central channel test loop: CPC-1 - thermal shocks loop: CEDI

International Nuclear Information System (INIS)

Calvaresi, C.; Moreschi, L.F.

1983-01-01

PEC (Prova Elementi di Combustibile: Fuel Elements Test) is an experimental fast sodium-cooled reactor with a power of 120 MWt. This reactor aims at studying the behaviour of fuel elements under thermal and neutron conditions comparable with those existing in fast power nuclear facilities. Given the particular structure of the core, the complex operations to be performed in the transfer cell and the strict operating conditions of the central channel, two experimental facilities, CPC-1 and CEDI, have been designed as a support to the construction of the reactor. CPC-1 is a 1:1 scale model of the channel, transfer-cell and loop unit of the channel, whereas CEDI is a sodium-cooled loop which enables to carry out tests of isothermal endurance and thermal shocks on the group of seven forced elements, by simulating the thermo-hydraulic and mechanical conditions existing in the reactor. In this paper some experimental test are briefy discussed and some facilities are listed, both for the CPC-1 and for the CEDI. (Auth.)

Tritium extraction from Pb-17Li by bubble columns

International Nuclear Information System (INIS)

Malara, C.

1995-01-01

Tritium extraction from the Pb-17Li liquid breeder of a fusion reactor can be efficiently carried out by bubble columns. To this aim, a mathematical model describing the complex fluid-dynamics of a bubble extractor is here presented. The model equations are made dimensionless and, together with the proper boundary conditions, numerically solved by the orthogonal collocation technique. Moreover, in order to better understand the role played by the different parameters in determining the performance of a bubble column, a closed solution of the model is obtained by introducing suitable hypotheses. A parametric analysis of the extraction efficiency of a bubble column as a function of the process parameters is carried out and, on this basis, the design of a tritium extraction system from the Pb-17Li breeder of a DEMO-type fusion reactor is proposed. 17 refs., 3 figs., 2 tabs

CFD simulation of fatty acid methyl ester production in bubble column reactor

Science.gov (United States)

Salleh, N. S. Mohd; Nasir, N. F.

2017-09-01

Non-catalytic transesterification is one of the method that was used to produce the fatty acid methyl ester (FAME) by blowing superheated methanol bubbles continuously into the vegetable oil without using any catalyst. This research aimed to simulate the production of FAME from palm oil in a bubble column reactor. Computational Fluid Dynamic (CFD) simulation was used to predict the distribution of fatty acid methyl ester and other product in the reactor. The fluid flow and component of concentration along the reaction time was investigated and the effects of reaction temperature (523 K and 563 K) on the non-catalytic transesterification process has been examined. The study was carried out using ANSYS CFX 17.1. The finding from the study shows that increasing the temperature leads to higher amount of fatty acid methyl ester can be produced in shorter time. On the other hand, concentration of the component such as triglyceride (TG), glycerol (GL) and fatty acid methyl ester (FAME) can be known when reaching the optimum condition.

Molten salt reactors. Synthesis of studies realized between 1973 and 1983. Experimental loop file

International Nuclear Information System (INIS)

1983-03-01

Four test loops were developed for the experimental study of a molten salt reactor with lead salt direct contact. A molten salt loop, completely in graphite, including the pump, showed that this material is convenient for salt containment and circulation. Reactor components like flowmeters, electromagnetic pumps, pressure gauge, valves developed for liquid sodium, were tested with liquid lead. A water-mercury loop was built for lead-molten salt simulation studies. Finally a lead-salt loop (COMPARSE) was built to study the behaviour of salt particles carried by lead in the heat exchanger. [fr

Contribution of nonlinear acoustic to the characterization of micro-bubbles clouds in liquid sodium. Application to the generation IV nuclear reactors

International Nuclear Information System (INIS)

Cavaro, M.

2010-11-01

The SFR system chosen (Sodium Fast Reactor: fast neutron reactors cooled by liquid sodium) by France led to a fourth-generation prototype named ASTRID. The development of this kind of reactors presents several challenges, particularly in terms of improving the safety and monitoring operation. This involves, among other things, characterization of the bubbles presence in liquid sodium. The characterization of the bubbles presence is the subject of this thesis. It involves the determination of void fraction (gas volume fraction) and histogram of the radii of bubbles. The bibliographic work done has shown that linear acoustic techniques for the characterization of bubble clouds are inadequate to achieve this. However promising leads have been identified by studying nonlinear acoustic techniques. This last idea has therefore been explored. An experimental water bench for the generation and optical control of micro-bubbles cloud allowed us to validate finely the reconstruction of histograms of radii through a technique of nonlinear mixing of a high frequency with a low frequency. The potential of the mixing of two high frequencies, more interesting for the industrial point of view has also been demonstrated. Finally, the bases of the transposition of an original technique of nonlinear resonance spectroscopy applied to a bubbles cloud were explored through the introduction of acoustic resonators. The results offer many interesting opportunities, both in terms of industrial applications and for more fundamental understanding of non-linear behavior of a bubble excited by multiple frequencies and of bubbles clouds excited at low frequency. (author)

Simulation test of PIUS-type reactor with large scale experimental apparatus

International Nuclear Information System (INIS)

Tamaki, M.; Tsuji, Y.; Ito, T.; Tasaka, K.; Kukita, Yutaka

1995-01-01

A large scale experimental apparatus for simulating the PIUS-type reactor has been constructed keeping the volumetric scaling ratio to the realistic reactor model. Fundamental experiments such as a steady state operation and a pump trip simulation were performed. Experimental results were compared with those obtained by the small scale apparatus in JAERI. We have already reported the effectiveness of the feedback control for the primary loop pump speed (PI control) for the stable operation. In this paper this feedback system is modified and the PID control is introduced. This new system worked well for the operation of the PIUS-type reactor even in a rapid transient condition. (author)

Safety report content and development for test loop facility on MARIA reactor

International Nuclear Information System (INIS)

Konechko, A.; Shumskij, A.M.; Mikul'ahin, V.E.

1982-01-01

A 600 kW test loop facility for investigatin.o safety problems is realized on MARIA reactor in Poland together with USSR organizations. Safety reports have been developed in two steps at the designstage. The 1st report being essentially a preliminary safety analysis was developed within the scope of the feasibility study. At the engineering design stage the preliminary test loop facility safety report had been prepared considering measures excluding the possibility of the MARIA reactor damage. The test loop facility safety report is fulfilled for normal, transient and emergency operation regimes. Separate safety basing for each group of experiments will be prepared. The report presents the test loop facility safety criteria coordinated by the nuclear safety comission. They contains the preliminary reports on the test loop facility safety. At the final stage of construction and at thecommitioning stage the start-up safety report will be developed which after required correction and adding up the putting into operation data will turn into operation safety report [ru

Influence of ultrasound power on acoustic streaming and micro-bubbles formations in a low frequency sono-reactor: mathematical and 3D computational simulation.

Science.gov (United States)

Sajjadi, Baharak; Raman, Abdul Aziz Abdul; Ibrahim, Shaliza

2015-05-01

This paper aims at investigating the influence of ultrasound power amplitude on liquid behaviour in a low-frequency (24 kHz) sono-reactor. Three types of analysis were employed: (i) mechanical analysis of micro-bubbles formation and their activities/characteristics using mathematical modelling. (ii) Numerical analysis of acoustic streaming, fluid flow pattern, volume fraction of micro-bubbles and turbulence using 3D CFD simulation. (iii) Practical analysis of fluid flow pattern and acoustic streaming under ultrasound irradiation using Particle Image Velocimetry (PIV). In mathematical modelling, a lone micro bubble generated under power ultrasound irradiation was mechanistically analysed. Its characteristics were illustrated as a function of bubble radius, internal temperature and pressure (hot spot conditions) and oscillation (pulsation) velocity. The results showed that ultrasound power significantly affected the conditions of hotspots and bubbles oscillation velocity. From the CFD results, it was observed that the total volume of the micro-bubbles increased by about 4.95% with each 100 W-increase in power amplitude. Furthermore, velocity of acoustic streaming increased from 29 to 119 cm/s as power increased, which was in good agreement with the PIV analysis. Copyright © 2014 Elsevier B.V. All rights reserved.

Hydrogen bubble formation and evolution in tungsten under different hydrogen irradiation conditions

Energy Technology Data Exchange (ETDEWEB)

Hu, Wenhui; Luo, Fengfeng; Shen, Zhenyu [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Guo, Liping, E-mail: guolp@whu.edu.cn [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Zheng, Zhongcheng; Wen, Yongming [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Ren, Yaoyao [Center for Electron Microscopy, Wuhan University, Wuhan 430072 (China)

2015-01-15

Highlights: • Direct and clear observation of hydrogen bubbles evolution by TEM is provided. • The role of temperature playing in bubble formation and evolution is fully explored. • Vacancy trapping mechanism is verified in this experiment. - Abstract: In order to see how hydrogen is behaving in tungsten and to understand the way bubbles form and grow up, specimens were irradiated by hydrogen ions from room temperature to 800 °C to fluence of 2.25 × 10{sup 21} m{sup −2}. Experimental results show that higher temperature helped bubble acquire higher internal pressure, causing interstitial loop punching to happen. In this process bubbles’ size grew and dislocation loops were formed but dislocation loops migrated away at and above 350 °C. And bubble number density reached peak value at 600 °C but then dropped dramatically at 800 °C. Because continuously increasing temperature would cause small bubbles dissolution or leaking out. Besides, high temperature also prevented tiny bubbles growing to be visible under TEM observation by their reaching equilibrium pressure before reaching threshold pressure for loop punching. In the other set of experiments, specimens were irradiated by low hydrogen fluence of 1 × 10{sup 20} m{sup −2} at 600 °C, in which case few hydrogen bubbles appeared. With further increasing irradiation fluence, bubble number density quickly increased. Small bubbles tended to coalesce to become larger visible bubbles. And they continued to grow through loop punching until their internal pressure cannot support their size expansion any more.

Analysis of the CAREM reactor's primary loop stability

International Nuclear Information System (INIS)

Mazzi, R.

1990-01-01

The results obtained to determine the stability conditions of the CAREM reactor's primary loop, using techniques based on the frequential response to the system, are presented. The stability margins were evaluated employing different alternatives; all of them predict a behaviour acceptable to the nominal working conditions. (Author) [es

Closed Loop In-Reactor Assembly (CLIRA): a fast flux test facility test vehicle

International Nuclear Information System (INIS)

Oakley, D.J.

1978-01-01

The Closed Loop In-Reactor Assembly (CLIRA) is a test vehicle for in-core material and fuel experiments in the Fast Flux Test Facility (FFTF). The FFTF is a fast flux nuclear test reactor operated for the Department of Energy (DOE) by Westinghouse Hanford Company in Richland, Washington. The CLIRA is a removable/replaceable part of the Closed Loop System (CLS) which is a sodium coolant system providing flow and temperature control independent of the reactor coolant system. The primary purpose of the CLIRA is to provide a test vehicle which will permit testing of nuclear fuels and materials at conditions more severe than exist in the FTR core, and to isolate these materials from the reactor core

A system for the discharge of gas bubbles from the coolant flow of a nuclear reactor cooled by forced circulation

International Nuclear Information System (INIS)

Markfort, D.; Kaiser, A.; Dohmen, A.

1975-01-01

In a reactor cooled by forced circulation the gas bubbles carried along with the coolant flow are separated before entering the reactor core or forced away into the external zones. For this purpose the coolant is radially guided into a plenum below the core and deflected to a tangential direction by means of flow guide elements. The flow runs spirally downwards. On the bubbles, during their dwell time in this channel, the buoyant force and a force towards the axis of symmetry of the tank are exerted. The major part of the coolant is directed into a radial direction by means of a guiding apparatus in the lower section of the channel and guided through a chimney in the plenum to the center of the reactor core. This inner chimney is enclosed by an outer chimney for the core edge zones through which coolant with a small share of bubbles is taken away. (RW) [de

A reduced fidelity model for the rotary chemical looping combustion reactor

KAUST Repository

Iloeje, Chukwunwike O.

2017-01-11

The rotary chemical looping combustion reactor has great potential for efficient integration with CO capture-enabled energy conversion systems. In earlier studies, we described a one-dimensional rotary reactor model, and used it to demonstrate the feasibility of continuous reactor operation. Though this detailed model provides a high resolution representation of the rotary reactor performance, it is too computationally expensive for studies that require multiple model evaluations. Specifically, it is not ideal for system-level studies where the reactor is a single component in an energy conversion system. In this study, we present a reduced fidelity model (RFM) of the rotary reactor that reduces computational cost and determines an optimal combination of variables that satisfy reactor design requirements. Simulation results for copper, nickel and iron-based oxygen carriers show a four-order of magnitude reduction in simulation time, and reasonable prediction accuracy. Deviations from the detailed reference model predictions range from 3% to 20%, depending on oxygen carrier type and operating conditions. This study also demonstrates how the reduced model can be modified to deal with both optimization and design oriented problems. A parametric study using the reduced model is then applied to analyze the sensitivity of the optimal reactor design to changes in selected operating and kinetic parameters. These studies show that temperature and activation energy have a greater impact on optimal geometry than parameters like pressure or feed fuel fraction for the selected oxygen carrier materials.

Summary of loops in the Chalk River NRX and NRU reactors

International Nuclear Information System (INIS)

Nishimura, D.T.

1980-12-01

The design and operating parameters of the high pressure, high temperature light water loops in the NRX and NRU reactors are presented to assist experimenters reviewing these facilities for their experiments. The NRX and NRU reactor design and operating data of interest to the experimenters are also presented. (author)

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.

Application of a Loop-Type Laboratory Biofilm Reactor to the Evaluation of Biofilm for Some Metallic Materials and Polymers such as Urinary Stents and Catheters

Directory of Open Access Journals (Sweden)

Hideyuki Kanematsu

2016-10-01

Full Text Available A laboratory biofilm reactor (LBR was modified to a new loop-type closed system in order to evaluate novel stents and catheter materials using 3D optical microscopy and Raman spectroscopy. Two metallic specimens, pure nickel and cupronickel (80% Cu-20% Ni, along with two polymers, silicone and polyurethane, were chosen as examples to ratify the system. Each set of specimens was assigned to the LBR using either tap water or an NB (Nutrient broth based on peptone from animal foods and beef extract mainly—cultured solution with E-coli formed over 48–72 h. The specimens were then analyzed using Raman Spectroscopy. 3D optical microscopy was employed to corroborate the Raman Spectroscopy results for only the metallic specimens since the inherent roughness of the polymer specimens made such measurements difficult. The findings suggest that the closed loop-type LBR together with Raman spectroscopy analysis is a useful method for evaluating biomaterials as a potential urinary system.

Bubbles

DEFF Research Database (Denmark)

Dholakia, Nikhilesh; Turcan, Romeo V.

2013-01-01

A goal of our ongoing research stream is to develop a multidisciplinary metatheory of bubbles. In this viewpoint paper we put forward a typology of bubbles by comparing four types of assets – entertainment, commodities, financial securities (stocks), and housing properties – where bubbles could...... and do form occasionally. Cutting across and comparing such varied asset types provides some rich insights into the nature of bubbles – and offers an inductive way to arrive at the typology of bubbles....

Radiation damage structure in irradiated and annealed 440 WWER-Type reactor pressure vessel steels

International Nuclear Information System (INIS)

Kocik, J.; Keilova, E.

1993-01-01

A review of irradiation damages in WWER-type RPV steels based on conventional Transmission Electron Microscopy investigations in a power reactor and a research reactor, is presented; the samples consist in Cr-Mo-V ferritic steel (15Kh2MFA type). The visible part of radiation-induced defects consists of very fine vanadium carbide precipitates, small dislocation loops and black dots (presumably corresponding to clusters and particle embryos formed from vacancies and solute-atoms (vanadium, copper, phosphorus) and carbon associated with vanadium. Radiation-induced defects are concentrated at dislocation substructure during irradiation in a power reactor, revealing the role of radiation-enhanced diffusion in damage structure forming process. Contrarily, the distribution of defects resulting from annealing of specimens irradiated in the research reactor is pre-determined by an homogenous distribution of radiation-induced defects prior to annealing. Increasing the number of re-irradiation and annealing cycles, the amount of dislocation loops among all defects seems to be growing. Simultaneously, the dislocation substructure recovers considerably. (authors). 14 refs., 11 figs., 3 tabs

Radiation damage structure in irradiated and annealed 440 WWER-Type reactor pressure vessel steels

Energy Technology Data Exchange (ETDEWEB)

Kocik, J; Keilova, E [Czech Nuclear Society, Prague (Czech Republic)

1994-12-31

A review of irradiation damages in WWER-type RPV steels based on conventional Transmission Electron Microscopy investigations in a power reactor and a research reactor, is presented; the samples consist in Cr-Mo-V ferritic steel (15Kh2MFA type). The visible part of radiation-induced defects consists of very fine vanadium carbide precipitates, small dislocation loops and black dots (presumably corresponding) to clusters and particle embryos formed from vacancies and solute-atoms (vanadium, copper, phosphorus) and carbon associated with vanadium. Radiation-induced defects are concentrated at dislocation substructure during irradiation in a power reactor, revealing the role of radiation-enhanced diffusion in damage structure forming process. Contrarily, the distribution of defects resulting from annealing of specimens irradiated in the research reactor is pre-determined by an homogenous distribution of radiation-induced defects prior to annealing. Increasing the number of re-irradiation and annealing cycles, the amount of dislocation loops among all defects seems to be growing. Simultaneously, the dislocation substructure recovers considerably. (authors). 14 refs., 11 figs., 3 tabs.

Investigations of the influence of feedback and coupling effects on neutron noise in a nuclear reactor

International Nuclear Information System (INIS)

Vath, W.

1975-01-01

Investigations are described of the influence of a known feedback, i.e., a control loop with known transfer function, on the spectra of neutron chamber signals. Theoretical formulas for the spectra are derived using the point reactor model. These formulas were verified by noise measurements in a zero power reactor. Special attention is given to the noise generated by the control loop, the influence of this feedback noise on the spectra being verified experimentally. In large reactors space dependent transfer functions must be taken into account. As a first approximation to handle the spatial dependence, the afore-mentioned investigations were extended to the two-point reactor model. Corresponding experimental work was done for the Argonaut Reactor Karlsruhe (ARK) with a symmetrical two-slab core loading. Coolant boiling in a BWR was investigated. The boiling must be considered as a feedback mechanism as well as an extrnal reactivity perturbation. In order to simulate the steam bubble content, nitrogen gas was injected into the water moderator of the ARK. By modulating the total gas flow according to the instantaneous reactor power the feedback effect was simulated. The gas flow produced a band-limited, white reactivity noise. The upper break frequency could be used to determine the travelling time of the bubbles through the core. (author)

An estimation of core damage frequency of a pressurized water reactor during mid-loop operation

International Nuclear Information System (INIS)

Chao, C.C.; Chen, C.T.; Lee, M.

2004-01-01

The core damage frequency during mid-loop operation of a Westinghouse designed 3-loop Pressurizer Water Reactor (PWR) due to loss of Residual Heat Removal (RHR) events was assessed. The assessment considers two types of outages (refueling and drained maintenance), and uses failure data collected specifically for shutdown condition. Event trees were developed for five categories of loss of RHR events. Human actions to mitigate the loss of RHR events was identified and human error probabilities were quantified using HCR and THERP model. The result showed that the core damage frequency due to loss of RHR events during mid-loop operation is 3.1x10 -5 per year. The results also showed that the core damage frequency can be reduced significantly by removing a pressurizer safety valve before entering mid-loop operation. The establishment of reflux cooling, i.e. decay heat removal through steam generator secondary side also plays important role in mitigating the loss of RHR events. (author)

A reduced fidelity model for the rotary chemical looping combustion reactor

International Nuclear Information System (INIS)

Iloeje, Chukwunwike O.; Zhao, Zhenlong; Ghoniem, Ahmed F.

2017-01-01

Highlights: • Methodology for developing a reduced fidelity rotary CLC reactor model is presented. • The reduced model determines optimal reactor configuration that meets design and operating requirements. • A 4-order of magnitude reduction in computational cost is achieved with good prediction accuracy. • Sensitivity studies demonstrate importance of accurate kinetic parameters for reactor optimization. - Abstract: The rotary chemical looping combustion reactor has great potential for efficient integration with CO_2 capture-enabled energy conversion systems. In earlier studies, we described a one-dimensional rotary reactor model, and used it to demonstrate the feasibility of continuous reactor operation. Though this detailed model provides a high resolution representation of the rotary reactor performance, it is too computationally expensive for studies that require multiple model evaluations. Specifically, it is not ideal for system-level studies where the reactor is a single component in an energy conversion system. In this study, we present a reduced fidelity model (RFM) of the rotary reactor that reduces computational cost and determines an optimal combination of variables that satisfy reactor design requirements. Simulation results for copper, nickel and iron-based oxygen carriers show a four-order of magnitude reduction in simulation time, and reasonable prediction accuracy. Deviations from the detailed reference model predictions range from 3% to 20%, depending on oxygen carrier type and operating conditions. This study also demonstrates how the reduced model can be modified to deal with both optimization and design oriented problems. A parametric study using the reduced model is then applied to analyze the sensitivity of the optimal reactor design to changes in selected operating and kinetic parameters. These studies show that temperature and activation energy have a greater impact on optimal geometry than parameters like pressure or feed fuel

ADVANCED DIAGNOSTIC TECHNIQUES FOR THREE-PHASE SLURRY BUBBLE COLUMN REACTORS(SBCR)

Energy Technology Data Exchange (ETDEWEB)

M.H. Al-Dahhan; L.S. Fan; M.P. Dudukovic

2002-07-25

This report summarizes the accomplishment made during the third year of this cooperative research effort between Washington University, Ohio State University and Air Products and Chemicals. Data processing of the performed Computer Automated Radioactive Particle Tracking (CARPT) experiments in 6 inch column using air-water-glass beads (150 {micro}m) system has been completed. Experimental investigation of time averaged three phases distribution in air-Therminol LT-glass beads (150 {micro}m) system in 6 inch column has been executed. Data processing and analysis of all the performed Computed Tomography (CT) experiments have been completed, using the newly proposed CT/Overall gas holdup methodology. The hydrodynamics of air-Norpar 15-glass beads (150 {micro}m) have been investigated in 2 inch slurry bubble column using Dynamic Gas Disengagement (DGD), Pressure Drop fluctuations, and Fiber Optic Probe. To improve the design and scale-up of bubble column reactors, a correlation for overall gas holdup has been proposed based on Artificial Neural Network and Dimensional Analysis.

Nuclear reactor types

International Nuclear Information System (INIS)

Jones, P.M.S.

1987-01-01

The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

Numerical study of Taylor bubbles with adaptive unstructured meshes

Science.gov (United States)

Xie, Zhihua; Pavlidis, Dimitrios; Percival, James; Pain, Chris; Matar, Omar; Hasan, Abbas; Azzopardi, Barry

2014-11-01

The Taylor bubble is a single long bubble which nearly fills the entire cross section of a liquid-filled circular tube. This type of bubble flow regime often occurs in gas-liquid slug flows in many industrial applications, including oil-and-gas production, chemical and nuclear reactors, and heat exchangers. The objective of this study is to investigate the fluid dynamics of Taylor bubbles rising in a vertical pipe filled with oils of extremely high viscosity (mimicking the ``heavy oils'' found in the oil-and-gas industry). A modelling and simulation framework is presented here which can modify and adapt anisotropic unstructured meshes to better represent the underlying physics of bubble rise and reduce the computational effort without sacrificing accuracy. The numerical framework consists of a mixed control-volume and finite-element formulation, a ``volume of fluid''-type method for the interface capturing based on a compressive control volume advection method, and a force-balanced algorithm for the surface tension implementation. Numerical examples of some benchmark tests and the dynamics of Taylor bubbles are presented to show the capability of this method. EPSRC Programme Grant, MEMPHIS, EP/K0039761/1.

Primary coolant circuits in FBR type reactors

International Nuclear Information System (INIS)

Kutani, Masushiro.

1985-01-01

Purpose: To eliminate the requirement of a pump for the forcive circulation of primary coolants and avoid the manufacturing difficulty of equipments. Constitution: In primary coolant circuits of an LMFBR type reactor having a recycling path forming a closed loop between a reactor core and a heat exchanger, coolants recycled through the recycling path are made of a magnetic fluid comprising liquid sodium incorporated with fine magnetic powder, and an electromagnet is disposed to the downstream of the heat exchanger. In the above-mentioned structure, since the magnetic fluid as the primary coolants losses its magnetic property when heated in the reactor core but recovers the property at a lower temperature after the completion of the heat exchange, the magnetic fluid can forcively be flown through the recycling path under the effect of the electromagnet disposed to the down stream of the heat exchanger to thereby forcively recycle the primary coolants. (Kawakami, Y.)

Flow patterns in a slurry-bubble-column reactor under reaction conditions

Energy Technology Data Exchange (ETDEWEB)

Toselane, B.A.; Brown, D.M.; Zou, B.S.; Dudukovic, M.P. [Washington Univ., St. Louis, MO (United States)

1995-12-31

The gas and liquid radioactive tracer response curves obtained in an industrial bubble column reactor of height to diameter ratio of 10 are analyzed and the suitability of the axial dispersion model for interpretation of the results is discussed. The relationship between the tracer concentration distribution and measured detector response of the soluble gas tracer (Ar-41) is possibly dominated by the dissolved gas. The one dimensional axial dispersion model cannot match all the experimental observations well and the flow pattern of the undissolved gas cannot be determined with certainty.

The response of the BTI bubble detectors in mixed gamma-neutron workplace fields

International Nuclear Information System (INIS)

Vanhavere, F.; Coeck, M.; Lievens, B.; Reginatto, M.

2005-01-01

Full text: Bubble detectors have become a mature technology and are used as neutron dosemeters in a wide range of applications. At the SCK-CEN and Belgonucleaire they are used as official personal neutron dosemeter for the personnel. Two types are commercially available from Bubble Technology Industries: the BD-PND, which has a neutron energy threshold of around 100 keV, and the BDT, which is mainly sensitive to thermal neutrons. At Belgonucleaire only the BD-PND is worn, and the results are corrected with a site specific factor. At the SCK-CEN both the BD-PND and BDT are worn and a combination of both results is applied for the dose records. In the EC project EVIDOS (Evaluation of Individual Dosimetry in Mixed Neutron and Photon Radiation Fields), a whole range of neutron dosemeters were irradiated in workplace fields in nuclear installations in Europe, including both types of bubble detectors. The bubble detectors were exposed on a phantom with different angles towards the reference directions in the workplace fields. We will report the bubble detectors' results in the simulated workplace fields at Cadarache (CANEL and Sigma), in the workplaces at Kruemmel (boiling water reactor, transport cask), at Mol (Venus research reactor SCK-CEN, MOX-fuel facility Belgonucleaire) and Ringhals (pressurized water reactor, transport cask). The responses of the bubble detectors and the combination of both will be compared to the reference values determined with Bonner Spheres and a novel directional spectrometer. The dosemeter readings were checked for consistency by folding the dosemeter response functions with the corresponding workplace fluence spectra in the same workplace. (author)

Magnetic-bubble devices

International Nuclear Information System (INIS)

Fairholme, R.J.

1978-01-01

Magnetic bubbles were first described only ten years ago when research workers were discussing orthoferrites containing μm diameter bubbles. However, problems of material fabrication limit crystals to a few mm across which severely curtailed device development. Since then materials have changed and rare-earth-iron garnet films can be grown up 3 inches in diameter with bubble diameters down to sizes below 1 μm. The first commercial products have device capacities in the range 64 000 to 100 000 bits with bubble diameters between 4 and 6 μm. Chip capacities of 1 Mbit are presently under development in the laboratory, as are new techniques to use submicrometre bubbles. The operation and fabrication of a bubble device is described using the serial loop devices currently being manufactured at Plessey as models. Chip organization is one important variable which directly affects the access time. A range of access times and capacities is available which offers a wide range of market opportunities, ranging from consumer products to fixed head disc replacements. some of the application areas are described. (author)

Regeneration of barium carbonate from barium sulphide in a pilot-scale bubbling column reactor and utilization for acid mine drainage.

Science.gov (United States)

Mulopo, J; Zvimba, J N; Swanepoel, H; Bologo, L T; Maree, J

2012-01-01

Batch regeneration of barium carbonate (BaCO(3)) from barium sulphide (BaS) slurries by passing CO(2) gas into a pilot-scale bubbling column reactor under ambient conditions was used to assess the technical feasibility of BaCO(3) recovery in the Alkali Barium Calcium (ABC) desalination process and its use for sulphate removal from high sulphate Acid Mine Drainage (AMD). The effect of key process parameters, such as BaS slurry concentration and CO(2) flow rate on the carbonation, as well as the extent of sulphate removal from AMD using the recovered BaCO(3) were investigated. It was observed that the carbonation reaction rate for BaCO(3) regeneration in a bubbling column reactor significantly increased with increase in carbon dioxide (CO(2)) flow rate whereas the BaS slurry content within the range 5-10% slurry content did not significantly affect the carbonation rate. The CO(2) flow rate also had an impact on the BaCO(3) morphology. The BaCO(3) recovered from the pilot-scale bubbling column reactor demonstrated effective sulphate removal ability during AMD treatment compared with commercial BaCO(3).

Mass transfer of ammonia escape and CO2 absorption in CO2 capture using ammonia solution in bubbling reactor

International Nuclear Information System (INIS)

Ma, Shuangchen; Chen, Gongda; Zhu, Sijie; Han, Tingting; Yu, Weijing

2016-01-01

Highlights: • Mass transfer coefficient models of ammonia escape were built. • Influences of temperature, inlet CO 2 and ammonia concentration were studied. • Mass transfer coefficients of ammonia escape and CO 2 absorption were obtained. • Studies can provide the basic data as a reference guideline for process application. - Abstract: The mass transfer of CO 2 capture using ammonia solution in the bubbling reactor was studied; according to double film theory, the mass transfer coefficient models and interface area model were built. Through our experiments, the overall volumetric mass transfer coefficients were obtained, while the interface areas in unit volume were estimated. The volumetric mass transfer coefficients of ammonia escaping during the experiment were 1.39 × 10 −5 –4.34 × 10 −5 mol/(m 3 s Pa), and the volumetric mass transfer coefficients of CO 2 absorption were 2.86 × 10 −5 –17.9 × 10 −5 mol/(m 3 s Pa). The estimated interface area of unit volume in the bubbling reactor ranged from 75.19 to 256.41 m 2 /m 3 , making the bubbling reactor a viable choice to obtain higher mass transfer performance than the packed tower or spraying tower.

Two types of nonlinear wave equations for diffractive beams in bubbly liquids with nonuniform bubble number density.

Science.gov (United States)

Kanagawa, Tetsuya

2015-05-01

This paper theoretically treats the weakly nonlinear propagation of diffracted sound beams in nonuniform bubbly liquids. The spatial distribution of the number density of the bubbles, initially in a quiescent state, is assumed to be a slowly varying function of the spatial coordinates; the amplitude of variation is assumed to be small compared to the mean number density. A previous derivation method of nonlinear wave equations for plane progressive waves in uniform bubbly liquids [Kanagawa, Yano, Watanabe, and Fujikawa (2010). J. Fluid Sci. Technol. 5(3), 351-369] is extended to handle quasi-plane beams in weakly nonuniform bubbly liquids. The diffraction effect is incorporated by adding a relation that scales the circular sound source diameter to the wavelength into the original set of scaling relations composed of nondimensional physical parameters. A set of basic equations for bubbly flows is composed of the averaged equations of mass and momentum, the Keller equation for bubble wall, and supplementary equations. As a result, two types of evolution equations, a nonlinear Schrödinger equation including dissipation, diffraction, and nonuniform effects for high-frequency short-wavelength case, and a Khokhlov-Zabolotskaya-Kuznetsov equation including dispersion and nonuniform effects for low-frequency long-wavelength case, are derived from the basic set.

The Pegase reactor loops; Les boucles du reacteur Pegase

Energy Technology Data Exchange (ETDEWEB)

NONE

1968-07-01

After 4 years operation, experimentation and maintenance of the gas loops built especially for the nuclear fuel testing reactor Pegase, it appears desirable not only to gather together in a single document the essential characteristics and particularities of these devices and of their associated equipment, but also to give the reasons for the technical modifications and the way in which they were carried out; this is done here by the persons themselves who were responsible, day after day, for operating these loops. This essentially practically experience thus complements the careful research and preliminary testing carried out on these loops or on their prototypes. It should be of interest to those who deal with problems concerned with the design or operation of irradiation loops in experimental reactors or of similar equipment. (authors) [French] Apres 4 annees de fonctionnement, d'experimentation et d'entretien sur les boucles a gaz, construites specialement pour le reacteur d'essai des combustibles nucleaires Pegase, il a paru souhaitable non seulement de rassembler dans un meme document les caracteristiques et les particularites essentielles de ces dispositifs et des appareillages qui leur sont associes, mais aussi d'y preciser les raisons et les modalites des mises au point techniques, apportees par ceux qui, jour apres jour pendant cette periode, ont eu la charge de mettre en oeuvre ces boucles. Cette experience essentiellement pratique complete donc les etudes minutieuses et les essais preliminaires de ces boucles ou de leurs prototypes. Elle doit etre de quelque interet pour ceux qui sont confrontes aux problemes de conception ou d'exploitation de boucles d'irradiation dans des reacteurs experimentaux ou des dispositifs analogues. (auteurs)

The NCSU [North Carolina State Univ.] freon PWR [pressurized water reactor] loop

International Nuclear Information System (INIS)

Caves, J.R.; Doster, J.M.; Miller, G.D.; Wehring, B.W.; Turinsky, P.J.

1989-01-01

The nuclear engineering department at North Carolina State University has designed and constructed an operating scale model of a pressurized water reactor (PWR) nuclear steam supply system (NSSS). This facility will be used for education, training, and research. The loop uses electric heaters to simulate the reactor core and Freon as the primary and secondary coolant. Viewing ports at various locations in the loop allow the students to visualize flow regimes in normal and off-normal operating conditions. The objective of the design effort was to scale the thermal-hydraulic characteristics of a two-loop Westinghouse NSSS. Provisions have been made for the simulation of various abnormal occurrences. The model is instrumented in much the same manner as the actual NSSS. Current research projects using the loop include the development of adaptive expert systems to monitor the performance of the facility, diagnose mechanical faults, and to make recommendations to operators for mitigation of accidents. This involves having thermal-hydraulics and core-physics simulators running faster than real time on a mini-supercomputer, with operating parameters updated by communication with the data acquisition and control computer. Further opportunities for research will be investigated as they arise

Detailed modeling of hydrodynamics mass transfer and chemical reactions in a bubble column using a discrete bubble model

NARCIS (Netherlands)

Darmana, D.; Deen, N.G.; Kuipers, J.A.M.

2005-01-01

A 3D discrete bubble model is adopted to investigate complex behavior involving hydrodynamics, mass transfer and chemical reactions in a gas–liquid bubble column reactor. In this model a continuum description is adopted for the liquid phase and additionally each individual bubble is tracked in a

Chemical looping reforming in packed-bed reactors : modelling, experimental validation and large-scale reactor design

NARCIS (Netherlands)

Spallina, V.; Marinello, B.; Gallucci, F.; Romano, M.C.; van Sint Annaland, M.

This paper addresses the experimental demonstration and model validation of chemical looping reforming in dynamically operated packed-bed reactors for the production of H2 or CH3OH with integrated CO2 capture. This process is a combination of auto-thermal and steam methane reforming and is carried

Water hammer characteristics of integral pressurized water reactor primary loop

International Nuclear Information System (INIS)

Zuo, Qiaolin; Qiu, Suizheng; Lu, Wei; Tian, Wenxi; Su, Guanghui; Xiao, Zejun

2013-01-01

Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions

Water hammer characteristics of integral pressurized water reactor primary loop

Energy Technology Data Exchange (ETDEWEB)

Zuo, Qiaolin [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Qiu, Suizheng, E-mail: szqiu@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Lu, Wei; Tian, Wenxi; Su, Guanghui [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Xiao, Zejun [Nuclear Power Institute of China, Chengdu, Sichuan 610041 (China)

2013-08-15

Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions.

Independent CO2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, Task 2.50.05

International Nuclear Information System (INIS)

Stojic, M.; Pavicevic, M.

1964-01-01

This report contains the following volumes V and VI of the Project 'Independent CO 2 loop for cooling the samples irradiated in RA reactor vertical experimental channels': Design project of the dosimetry control system in the independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, and Safety report for the Independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels [sr

Detailed modeling of hydrodynamics mass transfer and chemical reactions in a bubble column using a discrete bubble model

NARCIS (Netherlands)

Darmana, D.; Deen, N.G.; Kuipers, J.A.M.

2005-01-01

A 3D discrete bubble model is adopted to investigate complex behavior involving hydrodynamics, mass transfer and chemical reactions in a gas¿liquid bubble column reactor. In this model a continuum description is adopted for the liquid phase and additionally each individual bubble is tracked in a

DEVELOPMENT OF A COMPUTATIONAL MULTIPHASE FLOW MODEL FOR FISCHER TROPSCH SYNTHESIS IN A SLURRY BUBBLE COLUMN REACTOR

Energy Technology Data Exchange (ETDEWEB)

Donna Post Guillen; Tami Grimmett; Anastasia M. Gribik; Steven P. Antal

2010-09-01

The Hybrid Energy Systems Testing (HYTEST) Laboratory is being established at the Idaho National Laboratory to develop and test hybrid energy systems with the principal objective to safeguard U.S. Energy Security by reducing dependence on foreign petroleum. A central component of the HYTEST is the slurry bubble column reactor (SBCR) in which the gas-to-liquid reactions will be performed to synthesize transportation fuels using the Fischer Tropsch (FT) process. SBCRs are cylindrical vessels in which gaseous reactants (for example, synthesis gas or syngas) is sparged into a slurry of liquid reaction products and finely dispersed catalyst particles. The catalyst particles are suspended in the slurry by the rising gas bubbles and serve to promote the chemical reaction that converts syngas to a spectrum of longer chain hydrocarbon products, which can be upgraded to gasoline, diesel or jet fuel. These SBCRs operate in the churn-turbulent flow regime which is characterized by complex hydrodynamics, coupled with reacting flow chemistry and heat transfer, that effect reactor performance. The purpose of this work is to develop a computational multiphase fluid dynamic (CMFD) model to aid in understanding the physico-chemical processes occurring in the SBCR. Our team is developing a robust methodology to couple reaction kinetics and mass transfer into a four-field model (consisting of the bulk liquid, small bubbles, large bubbles and solid catalyst particles) that includes twelve species: (1) CO reactant, (2) H2 reactant, (3) hydrocarbon product, and (4) H2O product in small bubbles, large bubbles, and the bulk fluid. Properties of the hydrocarbon product were specified by vapor liquid equilibrium calculations. The absorption and kinetic models, specifically changes in species concentrations, have been incorporated into the mass continuity equation. The reaction rate is determined based on the macrokinetic model for a cobalt catalyst developed by Yates and Satterfield [1]. The

Experimental analysis of flowrates distribution features in double-loop reactor channels

International Nuclear Information System (INIS)

Avdeev, E.F.; Chusov, I.A.

2013-01-01

Experimental data on the flowrate distribution in working channels dummies of a research reactor model with double-loop configuration are presented in the paper. The procedures of experiments and received experimental data processing are provided in details [ru

Dynamic Modeling and Control Studies of a Two-Stage Bubbling Fluidized Bed Adsorber-Reactor for Solid-Sorbent CO{sub 2} Capture

Energy Technology Data Exchange (ETDEWEB)

Modekurti, Srinivasarao; Bhattacharyya, Debangsu; Zitney, Stephen E.

2013-07-31

A one-dimensional, non-isothermal, pressure-driven dynamic model has been developed for a two-stage bubbling fluidized bed (BFB) adsorber-reactor for solid-sorbent carbon dioxide (CO{sub 2}) capture using Aspen Custom Modeler® (ACM). The BFB model for the flow of gas through a continuous phase of downward moving solids considers three regions: emulsion, bubble, and cloud-wake. Both the upper and lower reactor stages are of overflow-type configuration, i.e., the solids leave from the top of each stage. In addition, dynamic models have been developed for the downcomer that transfers solids between the stages and the exit hopper that removes solids from the bottom of the bed. The models of all auxiliary equipment such as valves and gas distributor have been integrated with the main model of the two-stage adsorber reactor. Using the developed dynamic model, the transient responses of various process variables such as CO{sub 2} capture rate and flue gas outlet temperatures have been studied by simulating typical disturbances such as change in the temperature, flowrate, and composition of the incoming flue gas from pulverized coal-fired power plants. In control studies, the performance of a proportional-integral-derivative (PID) controller, feedback-augmented feedforward controller, and linear model predictive controller (LMPC) are evaluated for maintaining the overall CO{sub 2} capture rate at a desired level in the face of typical disturbances.

Innovative hybrid pile oscillator technique in the Minerve reactor: open loop vs. closed loop

Science.gov (United States)

Geslot, Benoit; Gruel, Adrien; Bréaud, Stéphane; Leconte, Pierre; Blaise, Patrick

2018-01-01

Pile oscillator techniques are powerful methods to measure small reactivity worth of isotopes of interest for nuclear data improvement. This kind of experiments has long been implemented in the Mineve experimental reactor, operated by CEA Cadarache. A hybrid technique, mixing reactivity worth estimation and measurement of small changes around test samples is presented here. It was made possible after the development of high sensitivity miniature fission chambers introduced next to the irradiation channel. A test campaign, called MAESTRO-SL, took place in 2015. Its objective was to assess the feasibility of the hybrid method and investigate the possibility to separate mixed neutron effects, such as fission/capture or scattering/capture. Experimental results are presented and discussed in this paper, which focus on comparing two measurements setups, one using a power control system (closed loop) and another one where the power is free to drift (open loop). First, it is demonstrated that open loop is equivalent to closed loop. Uncertainty management and methods reproducibility are discussed. Second, results show that measuring the flux depression around oscillated samples provides valuable information regarding partial neutron cross sections. The technique is found to be very sensitive to the capture cross section at the expense of scattering, making it very useful to measure small capture effects of highly scattering samples.

Leak detector for a steam generator in FBR type reactors

International Nuclear Information System (INIS)

Miyaji, Nobuyoshi.

1979-01-01

Purpose: To facilitate maintenance for liquid leak detectors such as exchange of nickel membrane sensors during operation in a sodium-cooled fbr type reactor. Constitution: A pipeway capable of supplying a cover gas such as argon into the cylinder of a hydrogen detector containing a nickel membrane sensor is provided in a liquid leak detector constituting a part of a by-pass loop. The pipeway is also adapted to be evacuated. A pipeway and a small sodium tank for drain use are provided on the side of the by-pass loop near valves. Then, after closing the inlet and outlet valves to disconnect the by-pass loop from the sodium main pipeway, the cover gas is supplied to drive liquid sodium to the drain tank. After the drain of the liquid sodium, the sensor can be replaced. (Ikeda, J.)

Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 1: Reactor Design and Model Development

KAUST Repository

Zhao, Zhenlong; Chen, Tianjiao; Ghoniem, Ahmed F.

2013-01-01

Chemical-looping combustion (CLC) is a novel and promising technology for power generation with inherent CO2 capture. Currently, almost all of the research has been focused on developing CLC-based interconnected fluidized-bed reactors. In this two

Numerical modeling of carbon dioxide chemisorption in sodium hydroxide solution in a micro-structured bubble column

NARCIS (Netherlands)

Jain, D.; Kuipers, J.A.M.; Deen, N.G.

2015-01-01

Gas-liquid flows with solid catalyst particles are encountered in many applications in the chemical, petrochemical, and pharmaceutical industries. Most commonly, two reactor types, slurry bubble column (SBC) and trickle bed (TB) reactors are applied for large scale in the industry. Both of these

Tank type reactor

International Nuclear Information System (INIS)

Otsuka, Fumio.

1989-01-01

The present invention concerns a tank type reactor capable of securing reactor core integrity by preventing incorporation of gases to an intermediate heat exchanger, thgereby improving the reliability. In a conventional tank type reactor, since vortex flows are easily caused near the inlet of an intermediate heat exchanger, there is a fear that cover gases are involved into the coolant main streams to induce fetal accidents. In the present invention, a reactor core is suspended by way of a suspending body to the inside of a reactor vessel and an intermediate heat exchanger and a pump are disposed between the suspending body and the reactor vessel, in which a vortex current preventive plate is attached at the outside near the coolant inlet on the primary circuit of the intermediate heat exchanger. In this way vortex or turbulence near the inlet of the intermediate heata exchanger or near the surface of coolants can be prevented. Accordingly, the cover gases are no more involved, to insure the reactor core integrity and obtain a tank type nuclear reactor of high reliability. (I.S.)

Towards the Characterization of the Bubble Presence in Liquid Sodium of Sodium Cooled Fast Reactors

International Nuclear Information System (INIS)

Cavaro, M.; Jeannot, J.P.; Payan, C.

2013-06-01

In a Sodium cooled Fast Reactors (SFR), different phenomena such as gas entrainment or nucleation can lead to gaseous micro-bubbles presence in the liquid sodium of the primary vessel. Although this free gas presence has no direct impact on the core neutronics, the French Atomic Energy and Alternative Energies Commission (CEA) currently works on its characterization to, among others, check the absence of risk of large gas pocket formation and to assess the induced modifications of the sodium acoustic properties. The main objective is to evaluate the void fraction values (volume fraction of free gas) and the radii histogram of the bubbles present in liquid sodium. Acoustics and electromagnetic techniques are currently developed at CEA: - The low-frequency speed of sound measurement, which allows us to link - thanks to Wood's model - the measured speed of sound to the actual void fraction. - The nonlinear mixing of two frequencies, based on the nonlinear resonance behavior of a bubble. This technique allows knowing the radius histogram associated to a bubble cloud. Two different mixing techniques are presented in this paper: the mixing of two high frequencies and the mixing of a high and a low frequency. - The Eddy-current flowmeter (ECFM), the output signal of which is perturbed by free gas presence and in consequence allows detecting bubbles. For each technique, initial results are presented. Some of them are really promising. So far, acoustic experiments have been led with an air-water experimental set-up. Micro-bubbles clouds are generated with a dissolved air flotation device and monitored by an optical device which provides reference measurements. Generated bubbles have radii range from few micrometers to several tens of micrometers. Present and future air/water experiments are presented. Furthermore, a development plan of in-sodium tests is presented in terms of a device set-up, instrumentation, modeling tools and experiments. (authors)

Innovative hybrid pile oscillator technique in the Minerve reactor: open loop vs. closed loop

Directory of Open Access Journals (Sweden)

Geslot Benoit

2018-01-01

Full Text Available Pile oscillator techniques are powerful methods to measure small reactivity worth of isotopes of interest for nuclear data improvement. This kind of experiments has long been implemented in the Mineve experimental reactor, operated by CEA Cadarache. A hybrid technique, mixing reactivity worth estimation and measurement of small changes around test samples is presented here. It was made possible after the development of high sensitivity miniature fission chambers introduced next to the irradiation channel. A test campaign, called MAESTRO-SL, took place in 2015. Its objective was to assess the feasibility of the hybrid method and investigate the possibility to separate mixed neutron effects, such as fission/capture or scattering/capture. Experimental results are presented and discussed in this paper, which focus on comparing two measurements setups, one using a power control system (closed loop and another one where the power is free to drift (open loop. First, it is demonstrated that open loop is equivalent to closed loop. Uncertainty management and methods reproducibility are discussed. Second, results show that measuring the flux depression around oscillated samples provides valuable information regarding partial neutron cross sections. The technique is found to be very sensitive to the capture cross section at the expense of scattering, making it very useful to measure small capture effects of highly scattering samples.

Tungsten surface evolution by helium bubble nucleation, growth and rupture

International Nuclear Information System (INIS)

Sefta, Faiza; Wirth, Brian D.; Hammond, Karl D.; Juslin, Niklas

2013-01-01

Molecular dynamics simulations reveal sub-surface mechanisms likely involved in the initial formation of nanometre-sized ‘fuzz’ in tungsten exposed to low-energy helium plasmas. Helium clusters grow to over-pressurized bubbles as a result of repeated cycles of helium absorption and Frenkel pair formation. The self-interstitials either reach the surface as isolated adatoms or trap at the bubble periphery before organizing into prismatic 〈1 1 1〉 dislocation loops. Surface roughening occurs as single adatoms migrate to the surface, prismatic loops glide to the surface to form adatom islands, and ultimately as over-pressurized gas bubbles burst. (paper)

Numerical analysis of flow field formed by air bubble dischanging through a sparger

International Nuclear Information System (INIS)

Kim, H. W.; Bae, Y. Y.

2002-01-01

In both a boiling water reactor and an advanced type of pressurized water reactor APR1400 being constructed in Korea, water, air and steam successively discharge into a subcooled water pool through spargers, when a pressure relieving system is in operation. During the discharging processes, the air bubble clouds produce a low-frequency and high-amplitude oscillatory loading, which may result in significant damages to the submerged structures if the resonance between bubble clouds and structures occur. This study deals with a numerical analysis of the flow field due to the oscillation of air bubble clouds by using a commercial thermal hydraulic analysis code FLUENT, version 4.5. The VOF (Volume Of Fluid) model was used to simulate the interface of water, air and steam flows, since it is known to be suitable for the large bubble simulation and it enables to treat air as a compressible fluid. A good agreement between the analysis results and the ABB-Atom test results, which had been performed for the development of BWR sparger, was obtained

MTR loop at the MPR-GA. Siwabessy reactor of Serpong Indonesia for testing of LEU fuel

International Nuclear Information System (INIS)

Arbie, B.; Sunaryadi, D.; Supadi, S.

1991-01-01

The main objective of the MTR-Loop is for testing the specimens of MTR fuel element uprated conditions with respect to the normal conditions of the reactor fuel elements. It is intended to verify the suitability of the fuel elements for operation in a research reactor under preset temperature and pressure conditions. The most important part of the MTR loop is the test section. The fuel elements to be tested are positioned in the test section. For heat removal there is a cooling water flowing through the test section. On this paper the description of the MTR-Loop is described. Installation of the MTR-Loop will be performed in the middle of 1990. In order to facilitate the investigation of fuel behaviour and performance of the new fuel elements the supporting facilities are also already available in the RSG-GAS. (orig.)

Research on water hammer forces caused by rapid growth of bubbles at severe accidents of water cooled reactors

International Nuclear Information System (INIS)

Inasaka, Fujio; Adachi, Masaki; Aya, Izuo

2004-01-01

At severe accidents of Water Cooled Reactors a great deal of gas is expected to be produced in a short time within the water of lower part of nuclear pressure vessel and containment vessel caused by hydrogen production with a metal water reaction and steam explosions with direct contact of melting core and water. Water hammer forces caused by rapid growth of bubbles shall work on the wall of containment vessel and affect its integrity. Coherency of water block movement is not clear, whether simultaneous or in the same direction. Water block behavior and water hammer forces caused by rapid growth of bubbles have been tested using a modified scale model and analyzed to obtain experimental correlated equation to estimate water block's rising distance and velocity from water hammer data. Numerical analysis using RELAP5-3D (Reactor Excursion and Leak Analysis Program) has been conducted to evaluate water hammer forces and makes clear its modifications needed. (T. Tanaka)

Station Blackout Analysis for a 3-Loop Westinghouse PWR Reactor Using Trace

International Nuclear Information System (INIS)

El-Sahlamy, N.M.

2017-01-01

One of the main concerns in the area of severe accidents in nuclear reactors is that of station blackout (SBO). The loss of offsite electrical power concurrent with the unavailability of the onsite emergency alternating current (AC) power system can result in loss of decay heat removal capability, leading to a potential core damage which may lead to undesirable consequences to the public and the environment. To cope with an SBO, nuclear reactors are provided with protection systems that automatically shut down the reactor, and with safety systems to remove the core residual heat. This paper provides a best estimate assessment of the SBO scenario in a 3-loop Westinghouse PWR reactor. The evaluation is performed using TRACE, a best estimate computer code for thermal-hydraulic calculations. Two sets of scenarios for SBO analyses are discussed in the current work. The first scenario is the short term SBO where it is assumed that in addition to the loss of AC power, there is no DC power; i.e., no batteries are available. In the second scenario, a long term SBO is considered. For this scenario, DC batteries are available for four hours. The aim of the current SBO analyses for the 3-loop pressurized water reactor presented in this paper is to focus on the effect of the availability of a DC power source to delay the time to core uncovers and heatup

Development of gas cooled reactors and experimental setup of high temperature helium loop for in-pile operation

Energy Technology Data Exchange (ETDEWEB)

Miletić, Marija, E-mail: marija_miletic@live.com [Czech Technical University in Prague, Prague (Czech Republic); Fukač, Rostislav, E-mail: fuk@cvrez.cz [Research Centre Rez Ltd., Rez (Czech Republic); Pioro, Igor, E-mail: Igor.Pioro@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada); Dragunov, Alexey, E-mail: Alexey.Dragunov@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada)

2014-09-15

Highlights: • Gas as a coolant in Gen-IV reactors, history and development. • Main physical parameters comparison of gas coolants: carbon dioxide, helium, hydrogen with water. • Forced convection in turbulent pipe flow. • Gas cooled fast reactor concept comparisons to very high temperature reactor concept. • High temperature helium loop: concept, development, mechanism, design and constraints. - Abstract: Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuel supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow's needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts is a very high temperature reactor (VHTR). The VHTR concept uses a graphite-moderated core with a once-through uranium fuel cycle, using high temperature helium as the coolant. Because helium is naturally inert and single-phase, the helium-cooled reactor can operate at much higher temperatures, leading to higher efficiency. Current VHTR concepts will use fuels such as uranium dioxide, uranium carbide, or uranium oxycarbide. Since some of these fuels are new in nuclear industry and due to their unknown properties and behavior within VHTR conditions it is very important to address these issues by investigate their characteristics within conditions close to those in VHTRs. This research can be performed in a research reactor with in-pile helium loop designed and constructed in Research Center Rez Ltd. One of the topics analyzed in this article are also physical characteristic and benefits of gas

Spectral analysis of detector signals and the effect of gas and vapor bubbles in the core of the SUR-100 reactor

International Nuclear Information System (INIS)

Song, P.S.

1981-01-01

A series of experiments was performed in the SUR-100 reactor, Hanover, and evaluated by means of statistical analysis methods in order to extend the knowledge about the influence of voids on the neutron flux and facilitate the interpretation of spectra of neutron flux fluctuations measured in power reactors. The investigations were performed in a relatively low frequency band, because the neutron flux spectra generated by air bubbles crossing the reactor core without any essential change in velocity and shape show the typical features of global reactivity effects. A strong relation between the spectra shapes and the transit times of bubbles through the core can be observed. Concerning the experiments with boiling coolant, pronounced neutron flux oscillations were measured originating from periodical flow instabilities in the coolant channel. The neutron flux oscillations depend upon the subcooling of the water and upon the heating power and have evidently the same frequency like the flow oscillations. (orig.) [de

Thermal-Hydraulic Experiment To Test The Stable Operation Of A PIUS Type Reactor

International Nuclear Information System (INIS)

Irianto, Djoko; Kanji, T.; Kukita, Y.

1996-01-01

An advanced type of reaktor concept as the Process Inherent Ultimate Safety (PIUS) reactor was based on intrinsically passive safety considerations. The stable operation of a PIUS type reactor is based on the automation of circulation pump speed. An automatic circulation pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed the PIUS-type reactor. In principle this control system maintains the fluid temperature at the axial center of the lower density lock at average of the fluid temperatures below and above the lower density lock. This control system will prevent the poison water from penetrating into the core during normal operation. The effectiveness of this control system was successfully confirmed by a series of experiments using atmospheric pressure thermal-hydraulic test loop which simulated the PIUS principle. The experiments such as: start-up and power ramping tests for normal operation simulation and loss of feedwater test for an accident condition simulation, carried out in JAERI

Characterizing fluid dynamics in a bubble column aimed for the determination of reactive mass transfer

Science.gov (United States)

Kováts, Péter; Thévenin, Dominique; Zähringer, Katharina

2018-02-01

Bubble column reactors are multiphase reactors that are used in many process engineering applications. In these reactors a gas phase comes into contact with a fluid phase to initiate or support reactions. The transport process from the gas to the liquid phase is often the limiting factor. Characterizing this process is therefore essential for the optimization of multiphase reactors. For a better understanding of the transfer mechanisms and subsequent chemical reactions, a laboratory-scale bubble column reactor was investigated. First, to characterize the flow field in the reactor, two different methods have been applied. The shadowgraphy technique is used for the characterisation of the bubbles (bubble diameter, velocity, shape or position) for various process conditions. This technique is based on particle recognition with backlight illumination, combined with particle tracking velocimetry (PTV). The bubble trajectories in the column can also be obtained in this manner. Secondly, the liquid phase flow has been analysed by particle image velocimetry (PIV). The combination of both methods, delivering relevant information concerning disperse (bubbles) and continuous (liquid) phases, leads to a complete fluid dynamical characterization of the reactor, which is the pre-condition for the analysis of mass transfer between both phases.

Pulse Star Inertial Confinement Fusion Reactor: Heat transfer loop and balance-of-plant considerations

International Nuclear Information System (INIS)

McDowell, M.W.; Blink, J.A.; Curlander, K.A.

1983-01-01

A conceptual heat transfer loop and balance-of-plant design for the Pulse Star Inertial Confinement Fusion Reactor has been investigated and the results are presented. The Pulse Star reaction vessel, a perforated steel bell jar about11 m in diameter, is immersed in Li 17 Pb 83 coolant, which flows through the perforations and forms a 1.5-m-thick plenum of droplets around a 8-m-diameter inner chamber. The bell jar and associated pumps, piping, and steam generators are contained within a 17-m-diameter pool of Li 17 Pb 83 coolant to minimize structural requirements and occupied space, resulting in reduced cost. Four parallel heat transfer loops, each with a flow rate of 5.5 m 3 /s, are necessary to transfer 3300 MWt of power. Liquid metal is pumped to the top of the pool, where it flows downward through eight vertical steam generators. Double-walled tubes are used in the steam generators to assure tritium containment without intermediate heat transfer loops. Each pump is a mixed flow type and has a required NPSH of 3.4 m, a speed of 278 rpm, and an impeller diameter of 1.2 m. The steam generator design was optimized by finding the most cost-effective combination of heat exchanger area and pumping power. The design minimizes the total cost (heat exchanger area plus pumping) for the plant lifetime. The power required for the pumps is 36 MWe. Each resulting steam generator is 12 m high and 1.6 m in diameter, with 2360 tubes. The steam generators and pumps fit easily in the pool between the reactor chamber and the pool wall

Experimental investigation of a pilot-scale jet bubbling reactor for wet flue gas desulphurisation

DEFF Research Database (Denmark)

Zheng, Yuanjing; Kiil, Søren; Johnsson, Jan Erik

2003-01-01

In the present work, an experimental parameter study was conducted in a pilot-scale jet bubbling reactor for wet flue gas desulphurisation (FGD). The pilot plant is downscaled from a limestone-based, gypsum producing full-scale wet FGD plant. Important process parameters, such as slurry pH, inlet...... flue gas concentration of SO2, reactor temperature, and slurry concentration of Cl- have been varied. The degree of desulphurisation, residual limestone content of the gypsum, liquid phase concentrations, and solids content of the slurry were measured during the experimental series. The SO2 removal...... efficiency increased from 66.1% to 71.5% when the reactor slurry pH was changed from 3.5 to 5.5. Addition of Cl(in the form of CaCl2 . 2H(2)O) to the slurry (25 g Cl-/l) increased the degree of desulphurisation to above 99%, due to the onset of extensive foaming, which substantially increased the gas...

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.)

Results of safety analysis on PWR type nuclear power plants with two and three loops

International Nuclear Information System (INIS)

1979-01-01

The results of safety analysis on PWR type nuclear power plants with two and three loops are presented, which was conducted by the Resource and Energy Agency, in June, 1979. This analysis was made simulating the phenomenon relating to the pressurizer level gauge at the time of the TMI accident. The model plants were the Ikata nuclear power plant with two loops and the Takahama No. 1 nuclear power plant with three loops. The premise conditions for this safety analysis were as follows: 1) the main feed water flow is totally lost suddenly at the full power operation of the plants, and the feed water pump is started manually 15 minutes after the accident initiation, 2) the relief valve on the pressurizer is kept open even after the pressure drop in the primary cooling system, and the primary cooling water flows out into the containment vessel through the rupture disc of the pressurizer relief tank, and 3) the electric circuit, which sends out the signal of safety injection at the abnormal low pressure in the reactor vessel, is added from the view-point of starting the operation of the emergency core cooling system as early as possible. Relating to the analytical results, the pressure in the reactor vessels changes less, the water level in the pressurizers can be regulated, and the water level in the steam generators is recovered safely in both two and three-loop plants. It is recognized that the plants with both two- and three loops show the safe transient phenomena, and the integrity of the cores is kept under the premise conditions. The evaluation for each analyzed result was conducted in detail. (Nakai, Y.)

Some particular aspects of control in nuclear power reactors

International Nuclear Information System (INIS)

Vathaire, F. de; Vernier, Ph.; Pascouet, A.

1964-01-01

This paper reviews the experience acquired in France on the question, of reactor safety. Since a special paper is being presented on reactors of the graphite gas type, the safety of the other types studied in France is discussed here: - heavy water-gas reactors, - fast neutron reactors, - water research reactors of the swimming-pool and tank types. The safety rules peculiar to the different types are explained, with emphasis on their influence on the reactor designs and on the power limits they impose. The corresponding safety studies are presented, particular stress being placed on the original work developed in these fields. Special mention is made of the experimental systems constructed for these studies: the reactor CABRI, pile loop for depressurization tests, loops outside the pile, mock-ups etc. (authors) [fr

Thermal and mechanical behaviour of oxygen carrier materials for chemical looping combustion in a packed bed reactor

International Nuclear Information System (INIS)

Jacobs, M.; Van Noyen, J.; Larring, Y.; Mccann, M.; Pishahang, M.; Amini, S.; Ortiz, M.; Galluci, F.; Sint-Annaland, M.V.; Tournigant, D.; Louradour, E.; Snijkers, F.

2015-01-01

Highlights: • Ilmenite-based oxygen carriers were developed for packed-bed chemical looping. • Addition of Mn_2O_3 increased mechanical strength and microstructure of the carriers. • Oxygen carriers were able to withstand creep and thermal cycling up to 1200 °C. • Ilmenite-based granules are a promising shape for packed-bed reactor conditions. - Abstract: Chemical looping combustion (CLC) is a promising carbon capture technology where cyclic reduction and oxidation of a metallic oxide, which acts as a solid oxygen carrier, takes place. With this system, direct contact between air and fuel can be avoided, and so, a concentrated CO_2 stream is generated after condensation of the water in the exit gas stream. An interesting reactor system for CLC is a packed bed reactor as it can have a higher efficiency compared to a fluidized bed concept, but it requires other types of oxygen carrier particles. The particles must be larger to avoid a large pressure drop in the reactor and they must be mechanically strong to withstand the severe reactor conditions. Therefore, oxygen carriers in the shape of granules and based on the mineral ilmenite were subjected to thermal cycling and creep tests. The mechanical strength of the granules before and after testing was investigated by crush tests. In addition, the microstructure of these oxygen particles was studied to understand the relationship between the physical properties and the mechanical performance. It was found that the granules are a promising shape for a packed bed reactor as no severe degradation in strength was noticed upon thermal cycling and creep testing. Especially, the addition of Mn_2O_3 to the ilmenite, which leads to the formation of an iron–manganese oxide, seems to results in stronger granules than the other ilmenite-based granules.

Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 1: Reactor Design and Model Development

KAUST Repository

Zhao, Zhenlong

2013-01-17

Chemical-looping combustion (CLC) is a novel and promising technology for power generation with inherent CO2 capture. Currently, almost all of the research has been focused on developing CLC-based interconnected fluidized-bed reactors. In this two-part series, a new rotary reactor concept for gas-fueled CLC is proposed and analyzed. In part 1, the detailed configuration of the rotary reactor is described. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet and exit. Two purging sectors are used to avoid the mixing between the fuel stream and the air stream. The rotary wheel consists of a large number of channels with copper oxide coated on the inner surface of the channels. The support material is boron nitride, which has high specific heat and thermal conductivity. Gas flows through the reactor at elevated pressure, and it is heated to a high temperature by fuel combustion. Typical design parameters for a thermal capacity of 1 MW have been proposed, and a simplified model is developed to predict the performances of the reactor. The potential drawbacks of the rotary reactor are also discussed. © 2012 American Chemical Society.

Constructing QCD one-loop amplitudes

International Nuclear Information System (INIS)

Forde, D

2008-01-01

In the context of constructing one-loop amplitudes using a unitarity bootstrap approach we discuss a general systematic procedure for obtaining the coefficients of the scalar bubble and triangle integral functions of one-loop amplitudes. Coefficients are extracted after examining the behavior of the cut integrand as the unconstrained parameters of a specifically chosen parameterization of the cut loop momentum approach infinity. Measurements of new physics at the forthcoming experimental program at CERN's Large Hadron Collider (LHC) will require a precise understanding of processes at next-to-leading order (NLO). This places increased demands for the computation of new one-loop amplitudes. This in turn has spurred recent developments towards improved calculational techniques. Direct calculations using Feynman diagrams are in general inefficient. Developments of more efficient techniques have usually centered around unitarity techniques [1], where tree amplitudes are effectively 'glued' together to form loops. The most straightforward application of this method, in which the cut loop momentum is in D = 4, allows for the computation of 'cut-constructible' terms only, i.e. (poly)logarithmic containing terms and any related constants. QCD amplitudes contain, in addition to such terms, rational pieces which cannot be derived using such cuts. These 'missing' rational parts can be extracted using cut loop momenta in D = 4-2 (var e psilon). The greater difficulty of such calculations has restricted the application of this approach, although recent developments [3, 4] have provided new promise for this technique. Recently the application of on-shell recursion relations [5] to obtaining the 'missing' rational parts of one-loop processes [6] has provided an alternative very promising solution to this problem. In combination with unitarity methods an 'on-shell bootstrap' approach provides an efficient technique for computing complete one-loop QCD amplitudes [7]. Additionally

Modeling of radial gas fraction profiles for bubble flow in vertical pipes

Energy Technology Data Exchange (ETDEWEB)

Lucas, D.; Krepper, E.; Prasser, H.-M. [Forschungszentrum Rossendorf e.V., Institute of Safety Research, Dresden (Germany)

2001-07-01

The paper presents a method for the prediction of radial gas fraction profiles from a given bubble size distribution. The method is based on the assumption of the equilibrium of the forces acting on a bubble perpendicularly to the flow direction. Assuming a large number of bubble size classes radial distributions are calculated separately for all bubble classes. The sum of these distributions is the radial profile of the gas fraction. The results of the model are compared with experimental data for a number of gas and liquid volume flow rates. The experiments were performed at a vertical test loop (inner diameter 50 mm) in FZ-Rossendorf using a wire mesh sensor. The sensor enables the determination of void distributions in the cross section of the loop. A special evaluation procedure supplies bubble size distributions as well as local distributions of bubbles within a predefined interval of bubble sizes. There is a good agreement between experimental and calculated data. In particular the change from wall peaking to core peaking is well predicted. (authors)

Modeling of radial gas fraction profiles for bubble flow in vertical pipes

International Nuclear Information System (INIS)

Lucas, D.; Krepper, E.; Prasser, H.-M.

2001-01-01

The paper presents a method for the prediction of radial gas fraction profiles from a given bubble size distribution. The method is based on the assumption of the equilibrium of the forces acting on a bubble perpendicularly to the flow direction. Assuming a large number of bubble size classes radial distributions are calculated separately for all bubble classes. The sum of these distributions is the radial profile of the gas fraction. The results of the model are compared with experimental data for a number of gas and liquid volume flow rates. The experiments were performed at a vertical test loop (inner diameter 50 mm) in FZ-Rossendorf using a wire mesh sensor. The sensor enables the determination of void distributions in the cross section of the loop. A special evaluation procedure supplies bubble size distributions as well as local distributions of bubbles within a predefined interval of bubble sizes. There is a good agreement between experimental and calculated data. In particular the change from wall peaking to core peaking is well predicted. (authors)

The experimental definition of the acoustic standing wave series shapes, formed in the coolant of the primary circuit of VVER-440 type reactor

International Nuclear Information System (INIS)

Bulavin, V.V.; Pavelko, V.I.

1995-01-01

On the basis of pressure fluctuation measurements in some primary circuit loops at 2 nd Unit of Kola NPP with VVER-440 type reactors, the shapes of acoustic standing waves (ASW) were determined at frequencies corresponding to four minimal oscillation eigenfrequencies in the primary circuit coolant. On identification of the ASW modes and properties, experimental results based on six circulating loops in symmetric arrangement allowed determination of the three-dimensional space structure of the wave nodes and antinodes inside and outside of the reactor vessel (RV). As part of this analysis, the geometric features of the primary circuit that caused the formation of these standing waves were identified. Differences in each ASW shape were shown to cause different individual effects on the neutron field in the reactor core and on fuel assembly vibration. This has been partially confirmed by ex-core neutron ionization chamber noise analysis. One type of ASW, possessing an antinode inside the RV, can be used for measurement of the pressure coefficient of reactivity. However, this must be done with care to avoid the potential for incorrect results in some cases. The results presented in this paper can be readily extended to other VVER type reactors with both odd and even number of loops. (author)

Feedback control of primary circulation pump of PIUS-Type reactor

International Nuclear Information System (INIS)

Fujii, Mikiya; Anoda, Yoshinari; Murata, Hideo; Yonomoto, Taisuke; Kukita, Yutaka; Tasaka, Kanji.

1991-05-01

In operating the PIUS-Type reactor, it is required to keep stationary density interfaces between the primary loop hot water and the poison tank cold, borated water by maintaining pressure balance between the primary-loop and the poison-tank. The authors have developed a primary circulation pump speed control system and tested it in small-scale experiments. This control system regulates the pump speed based on measurements of the density lock differential pressure which is proportional to the elevation of the interface in the density lock. This pump speed control facilitated the normal plant operation which included core power changes. However, the elevation of the density interface indicated oscillatory behavior when the pump speed was regulated as a linear function of the density lock differential pressure. The mechanism responsible for such oscillatory behavior was found to be manometric oscillations that could be eliminated by adding a damping term to compensate for the mechanical delay of the primary pump speed. The passive shutdown function of the reactor was retained by setting an upper limit to the pump speed. This was confirmed in a loss-of-feedwater abnormal transient test. (author)

Thermohydraulics in a high-temperature gas-cooled reactor primary loop during early phases of unrestricted core-heatup accidents

International Nuclear Information System (INIS)

Kroeger, P.G.; Colman, J.; Hsu, C.J.

1983-01-01

In High Temperature Gas Cooled Reactor (HTGR) siting considerations, the Unrestricted Core Heatup Accidents (UCHA) are considered as accidents of highest consequence, corresponding to core meltdown accidents in light water reactors. Initiation of such accidents can be, for instance, due to station blackout, resulting in scram and loss of all main loop forced circulation, with none of the core auxiliary cooling system loops being started. The result is a slow but continuing core heatup, extending over days. During the initial phases of such UCHA scenarios, the primary loop remains pressurized, with the system pressure slowly increasing until the relief valve setpoint is reached. The major objectives of the work described here were to determine times to depressurization as well as approximate loop component temperatures up to depressurization

Investigation of fast initialization of spacecraft bubble memory systems

Science.gov (United States)

Looney, K. T.; Nichols, C. D.; Hayes, P. J.

1984-01-01

Bubble domain technology offers significant improvement in reliability and functionality for spacecraft onboard memory applications. In considering potential memory systems organizations, minimization of power in high capacity bubble memory systems necessitates the activation of only the desired portions of the memory. In power strobing arbitrary memory segments, a capability of fast turn on is required. Bubble device architectures, which provide redundant loop coding in the bubble devices, limit the initialization speed. Alternate initialization techniques are investigated to overcome this design limitation. An initialization technique using a small amount of external storage is demonstrated.

Controlling the trajectories of bubble trains at a microfluidic junction

Science.gov (United States)

Parthiban, Pravien; Khan, Saif

2011-11-01

The increasing number of applications facilitated by digital microfluidic flows has resulted in a sustained interest in not only understanding the diverse, interesting and often complex dynamics associated with such flows in microchannel networks but also in developing facile strategies to control them. We find that there are readily accessible flow speeds wherein resistance to flow in microchannels decreases with an increase in the number of confined bubbles present, and exploit this intriguing phenomenon to sort all bubble of a train exclusively into one of the arms of a nominally symmetric microfluidic loop. We also demonstrate how the arm into which the train filters into can be chosen by applying a temporary external stimulus by means of an additional flow of the continuous liquid into one the arms of the loop. Furthermore, we show how by tuning the magnitude and period of this temporary stimulus we can switch controllably, the traffic of bubbles between both arms of the loop even when the loop is asymmetric. The results of this work should aid in developing viable methods to regulate traffic of digital flows in microfluidic networks.

Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

Energy Technology Data Exchange (ETDEWEB)

O' Brien, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yoon, Su -Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States); Housley, Gregory K. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-09-01

This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation

Application of neural network technology to setpoint control of a simulated reactor experiment loop

International Nuclear Information System (INIS)

Cordes, G.A.; Bryan, S.R.; Powell, R.H.; Chick, D.R.

1991-01-01

This paper describes the design, implementation, and application of artificial neural networks to achieve temperature and flow rate control for a simulation of a typical experiment loop in the Advanced Test Reactor (ATR) located at the Idaho National Engineering Laboratory (INEL). The goal of the project was to research multivariate, nonlinear control using neural networks. A loop simulation code was adapted for the project and used to create a training set and test the neural network controller for comparison with the existing loop controllers. The results for the best neural network design are documented and compared with existing loop controller action. The neural network was shown to be as accurate at loop control as the classical controllers in the operating region represented by the training set. 5 refs., 8 figs., 3 tabs

Nuclear reactor container

International Nuclear Information System (INIS)

Kawabe, Ryuhei; Yamaki, Rika.

1989-01-01

Aerosol filters considered so far for nuclear reactor containers in conventional BWR type nuclear power plants make the facility larger and involve a risk of clogging. In view of the above, in the present invention, the diameter of a flow channel of gases entering from a bent pipe to a suppression pool is made smaller thereby decreasing the diameter of gas bubbles in the supperssional pool. Since this reduces the force of surface tension, the diameter of resulted gas bubbles is made remarkably smaller as compared with the case where the gases are released from the lower end of the bent pipe. Since the absorption velocity of bubble-entrained aerosols into water is in proportion to the square of the bubble diameter, the absorption efficiency can be increased remarkably by reducing the diameter of the gas bubbles. Accordingly, it is possible to improve the efficiency of eliminating radioactivity of released gases. (K.M.)

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.

The OPERA loop in the OSIRIS reactor core. Pressurized-water irradiation device to study Advanced Reactor Fuels

International Nuclear Information System (INIS)

Lucot, M.; Roche, M.

1986-09-01

This loop is designed to allow fuel qualification test, i.d. to allow irradiation of representative parts fuel assemblies operating in thermohydraulic and chemical conditions representative of these of present pressurized water reactors or in development. This paper presents the aims of the installation, the general design and the main specifications with a brief detailed description [fr

Actions needed for RA reactor exploitation - I-IV, Part II, Design project VI-SA 1, Experimental loop for testing the EL-4 reactor fuel elements in the central vertical experimental channel of the RA reactor in Vinca

International Nuclear Information System (INIS)

Novakovic, M.

1961-12-01

The objective of installing the VISA-1 loop was testing the fuel elements of the EL-4 reactor. The fuel elements planned for testing are natural UO 2 with beryllium cladding, cooled by CO 2 under nominal pressure of 60 at and temperature 600 deg C. central vertical experimental channel of the RA reactor was chosen for installing a test loop cooled by CO 2 . This report contains the detailed design project of the testing loop with the control system and safety analysis of the planned experiment

A Quasi-Type-1 Phase-Locked Loop Structure

DEFF Research Database (Denmark)

Golestan, Saeed; Fernandez, Francisco Daniel Freijedo; Vidal, Ana

2014-01-01

The grid voltage phase and frequency are crucial information in control of most grid connected power electronic based equipment. Most often, a phase-locked loop (PLL) is employed for this purpose. A PLL is a closed-loop feedback control system that the phase of its output signal is related...... to the phase of its input signal. Arguably, the simplest PLL is a type-1 PLL. The type-1 PLLs are characterized by having only one integrator in their control loop and therefore having a high stability margin. However, they suffer from a serious drawback: they cannot achieve zero average steady-state phase...

Air water loop - an experimental facility to study thermal hydraulics of AHWR steam drum

International Nuclear Information System (INIS)

Bagul, R.K.; Pilkhwal, D.S.; Jain, V.; Vijayan, P.K.

2014-05-01

In the proposed Indian Advanced Heavy Water Reactor (AHWR) the coolant recirculation in the primary system is achieved by two-phase natural circulation. The two-phase steam-water mixture from the reactor core is separated in steam drum by gravity. Gravity separation of phases may lead to undesirable phenomena - carryover and carryunder. Carryover is the entrainment of liquid droplets in the vapor phase.Carryover needs to be minimized to avoid erosion corrosion of turbine blades. Carryunder is the entrainment of vapor bubbles with liquid flowing back to reactor core. Significant carryunder may in turn lead to reduced flow resulting in reduced CHF margin and stability in the coolant channel. An Air-Water Loop (AWL) has been designed to carry out the experiments relevant to AHWR steam drum. The design features and scaling philosophy is described in this report. (author)

Current and prospective fuel test programmes in the MIR reactor

Energy Technology Data Exchange (ETDEWEB)

Izhutov, A.L.; Burukin, A.V.; Iljenko, S.A.; Ovchinnikov, V.A.; Shulimov, V.N.; Smirnov, V.P. [State Scientific Centre of Russia Research Institute of Atomic Reactors, Ulyanovsk region (Russian Federation)

2007-07-01

MIR reactor is a heterogeneous thermal reactor with a moderator and a reflector made of metal beryllium, it has a channel-type design and is placed in a water pool. MIR reactor is mainly designed for testing fragments of fuel elements and fuel assemblies (FA) of different nuclear power reactor types under normal (stationary and transient) operating conditions as well as emergency situations. At present six test loop facilities are being operated (2 PWR loops, 2 BWR loops and 2 steam coolant loops). The majority of current fuel tests is conducted for improving and upgrading the Russian PWR fuel, these tests involve issues such as: -) long term tests of short-size rods with different modifications of cladding materials and fuel pellets; -) further irradiation of power plant re-fabricated and full-size fuel rods up to achieving 80 MW*d/kg U; -) experiments with leaking fuel rods at different burnups and under transient conditions; -) continuation of the RAMP type experiments at high burnup of fuel; and -) in-pile tests with simulation of LOCA and RIA type accidents. Testing of the LEU (low enrichment uranium) research reactor fuel is conducted within the framework of the RERTR programme. Upgrading of the gas cooled and steam cooled loop facilities is scheduled for testing the HTGR fuel and sub-critical water-cooled reactor, correspondingly. The present paper describes the major programs of the WWER high burn-up fuel behavior study in the MIR reactor, capabilities of the applied techniques and some results of the performed irradiation tests. (authors)

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

1989-01-01

In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

Interfacial area transport in a confined Bubbly flow

Energy Technology Data Exchange (ETDEWEB)

Kim, S.; Sun, X.; Ishii, M. [Purdue Univ., Lafayette, IN (United States). School of Nuclear Engineering; Lincoln, F. [Bettis Atomic Power Lab., West Mifflin, Bechtel Bettis, Inc., PA (United States)

2001-07-01

The interfacial area transport equation applicable to the bubbly flow is presented. The model is evaluated against the data acquired in an adiabatic air-water upward two-phase flow loop with a test section of 20 cm in width and 1 cm in gap. In general, a good agreement, within the measurement error of {+-}10%, is observed for a wide range in the bubbly flow regime. The sensitivity analysis on the individual particle interaction mechanisms demonstrates the active interactions between the bubbles and highlights the mechanisms playing the dominant role in interfacial area transport. (author)

Study of essential safety features of a three-loop 1,000 MWe light water reactor (PWR) and a corresponding heavy water reactor (HWR) on the basis of the IAEA nuclear safety standards

International Nuclear Information System (INIS)

1989-02-01

Based on the IAEA Standards, essential safety aspects of a three-loop pressurized water reactor (1,000 MWe) and a corresponding heavy water reactor were studied by the TUeV Baden e.V. in cooperation with the Gabinete de Proteccao e Seguranca Nuclear, a department of the Ministry which is responsible for Nuclear power plants in Portugal. As the fundamental principles of this study the design data for the light water reactor and the heavy water reactor provided in the safety analysis reports (KWU-SSAR for the 1,000 MWe PWR, KWU-PSAR Nuclear Power Plant ATUCHA II) are used. The assessment of the two different reactor types based on the IAEA Nuclear Safety Standards shows that the reactor plants designed according to the data given in the safety analysis reports of the plant manufacturer meet the design requirements laid down in the pertinent IAEA Standards. (orig.) [de

Regeneration of barium carbonate from barium sulphide in a pilot-scale bubbling column reactor and utilization for acid mine drainage

CSIR Research Space (South Africa)

Mulopo, J

2012-01-01

Full Text Available Batch regeneration of barium carbonate (BaCO3) from barium sulphide (BaS) slurries by passing CO2 gas into a pilot-scale bubbling column reactor under ambient conditions was used to assess the technical feasibility of BaCO3 recovery in the Alkali...

Ozonation kinetics of winery wastewater in a pilot-scale bubble column reactor.

Science.gov (United States)

Lucas, Marco S; Peres, José A; Lan, Bing Yan; Li Puma, Gianluca

2009-04-01

The degradation of organic substances present in winery wastewater was studied in a pilot-scale, bubble column ozonation reactor. A steady reduction of chemical oxygen demand (COD) was observed under the action of ozone at the natural pH of the wastewater (pH 4). At alkaline and neutral pH the degradation rate was accelerated by the formation of radical species from the decomposition of ozone. Furthermore, the reaction of hydrogen peroxide (formed from natural organic matter in the wastewater) and ozone enhances the oxidation capacity of the ozonation process. The monitoring of pH, redox potential (ORP), UV absorbance (254 nm), polyphenol content and ozone consumption was correlated with the oxidation of the organic species in the water. The ozonation of winery wastewater in the bubble column was analysed in terms of a mole balance coupled with ozonation kinetics modeled by the two-film theory of mass transfer and chemical reaction. It was determined that the ozonation reaction can develop both in and across different kinetic regimes: fast, moderate and slow, depending on the experimental conditions. The dynamic change of the rate coefficient estimated by the model was correlated with changes in the water composition and oxidant species.

Dose rates modeling of pressurized water reactor primary loop components with SCALE6.0

International Nuclear Information System (INIS)

Matijević, Mario; Pevec, Dubravko; Trontl, Krešimir

2015-01-01

Highlights: • Shielding analysis of the typical PWR primary loop components was performed. • FW-CADIS methodology was thoroughly investigated using SCALE6.0 code package. • Versatile ability of SCALE6.0/FW-CADIS for deep penetration models was proved. • The adjoint source with focus on specific material can improve MC modeling. - Abstract: The SCALE6.0 simulation model of a typical PWR primary loop components for effective dose rates calculation based on hybrid deterministic–stochastic methodology was created. The criticality sequence CSAS6/KENO-VI of the SCALE6.0 code package, which includes KENO-VI Monte Carlo code, was used for criticality calculations, while neutron and gamma dose rates distributions were determined by MAVRIC/Monaco shielding sequence. A detailed model of a combinatorial geometry, materials and characteristics of a generic two loop PWR facility is based on best available input data. The sources of ionizing radiation in PWR primary loop components included neutrons and photons originating from critical core and photons from activated coolant in two primary loops. Detailed calculations of the reactor pressure vessel and the upper reactor head have been performed. The efficiency of particle transport for obtaining global Monte Carlo dose rates was further examined and quantified with a flexible adjoint source positioning in phase-space. It was demonstrated that generation of an accurate importance map (VR parameters) is a paramount step which enabled obtaining Monaco dose rates with fairly uniform uncertainties. Computer memory consumption by the S N part of hybrid methodology represents main obstacle when using meshes with large number of cells together with high S N /P N parameters. Detailed voxelization (homogenization) process in Denovo together with high S N /P N parameters is essential for precise VR parameters generation which will result in optimized MC distributions. Shielding calculations were also performed for the reduced PWR

Experimental evaluation of reactivity constraints for the closed-loop control of reactor power

International Nuclear Information System (INIS)

Bernard, J.A.; Lanning, D.D.; Ray, A.

1984-01-01

General principles for the closed-loop, digital control of reactor power have been identified, quantitatively enumerated, and experimentally demonstrated on the 5 MWt Research Reactor, MITR-II. The basic concept is to restrict the net reactivity so that it is always possible to make the reactor period infinite at the desired termination point of a transient by reversing the direction of motion of whatever control mechanism is associated with the controller. This capability is formally referred to as ''feasibility of control''. A series of ten experiments have been conducted over a period of eighteen months to demonstrate the efficacy of this property for the automatic control of reactor power. It has been shown that a controller which possesses this property is capable of both raising and lowering power in a safe, efficient manner while using a control rod of varying differential worth, that the reactivity constraints are a sufficient condition for the automatic control of reactor power, and that the use of a controller based on reactivity constraints can prevent overshoots either due to attempts to control a transient with a control rod of insufficient differential worth or due to failure to properly estimate when to commence rod insertion. Details of several of the more significant tests are presented together with a discussion of the rationale for the development of closed-loop control in large commercial power systems. Specific consideration is given to the motivation for designing a controller based on feasibility of control and the associated licensing issues

Reconfiguration of the NRAD delay loop for proposed 1 MW operations

International Nuclear Information System (INIS)

Heidel, C.C.; Richards, W.J.; Pruett, D.P.

1984-01-01

Neutron radiography is provided by the NRAD reactor facility, which is located beneath the HFEF hot cell. The NRAD reactor is a TRIGA reactor and is operated at a steady-state power level of 250 kw solely for neutron radiography and the development of radiography techniques. When the NRAD facility was designed and constructed, an operating power level of 250 kw was considered to be adequate for obtaining radiographs of the type of specimens envisaged at that time. Since that time a second radiography station was installed and the thickness of the specimens being radiographed is greater than was initially envisaged. In order to decrease exposure times, the reactor power level is to be increased to 1 Mw. The present delay loop can not to be used at 1 Mw operation, because the passage way where the primary piping exits the reactor room must be maintained less than 1 MR/hr. To obtain the needed delay before the primary water exits the reactor room using the present internal delay loop system would require two more delay loops of the same size to be placed in series with the present delay loop. Because the NRAD reactor tank is small this is not possible; therefore, the delay must take place external to the reactor tank. The delay loop will have to be located in a shielded area to allow the decay of N 16 . The best location for the delay tank will be in the east radiography

Safety report for the independent CO2 loop for cooling the samples irradiated in the vertical experimental channels of the RA reactor

International Nuclear Information System (INIS)

Pavlovic, A.; Zivkovic, S.; Milosevic, M.; Strugar, P.

1969-01-01

Independent CO 2 loop was designed for cooling the samples irradiated in the vertical experimental channels in the region of heavy water or in the graphite reflector. It is considered as a significant improvement of the RA reactor experimental possibilities. Six 'heads' are placed in vertical experimental channels and connected in parallel with the outer loop which contains out-of-reactor equipment. It is planned to irradiate samples in the 'heads' of the loop for the needs of Hot laboratory and Laboratory for reactor materials. Heat generated during sample irradiation is removed by CO 2 circulation. This report contains detailed description of the main loop, auxiliary systems, calculation results of anti reactivity and activation of construction materials of the low-temperature CO 2 loop. A separate chapter is devoted to control and regulation of temperature and pressure. Testing of the fundamental parameters of the coolant loop showed that it could fulfill more demanding tasks than designed by the project. Annex of this report includes results of leak testing for the loop 'heads' in vertical experimental channels VEK-6 and VEK-8 by helium leak detector [sr

Enhanced CHF with Bubble Cutter and Artificial Flow in Nuclear Plants

International Nuclear Information System (INIS)

Jung, Chan Hee; Suh, Kune Y.

2013-01-01

The main goal of this paper is to body out the notions of forced convection system for enhanced local streams and air bubbles cutting (and/or pushing, breaking) system to explain how CHF can be improved and how those bubble cutter systems are applicable to NPPs. In this paper, the bubble cutter system and an artificial flow system which can cut (and/or push and break) air bubbles is bodied out to drag bubbles. It also make the surface wet condition of heated surfaces and improve heat transfer and prevent on creation of bubbles on the heated surfaces or heat exchangers or reactor cores. Namely, concepts and application methods to increase CHF are presented for NPPs. Enhanced critical heat flux (CHF) is one of our prospective aims for nuclear power plants (NPPs). Previous work has studied the flow boiling CHF enhancement with surfactant solutions under atmospheric pressure because surfactant solutions or surface conditions have an effect on the behavior of occurrence air bubbles on a heated surface. Another possible improvement is to improve efficiency of heat transfer or to body out some types of bubble breaking (and/or pushing, breaking) systems or an artificial flow of fluid that can tear off air bubbles or push hot liquid and bubbles on a surface of heater. During this study, it will be observed that those possible structures can elicit increased CHF by means of maintenance of contact with a coolant such as water

COMPUTATIONAL AND EXPERIMENTAL MODELING OF SLURRY BUBBLE COLUMN REACTORS

Energy Technology Data Exchange (ETDEWEB)

Paul C.K. Lam; Isaac K. Gamwo; Dimitri Gidaspow

2002-05-01

The objective of this study was to develop a predictive experimentally verified computational fluid dynamics (CFD) model for gas-liquid-solid flow. A three dimensional transient computer code for the coupled Navier-Stokes equations for each phase was developed and is appended in this report. The principal input into the model is the viscosity of the particulate phase which was determined from a measurement of the random kinetic energy of the 800 micron glass beads and a Brookfield viscometer. The details are presented in the attached paper titled ''CFD Simulation of Flow and Turbulence in a Slurry Bubble Column''. This phase of the work is in press in a referred journal (AIChE Journal, 2002) and was presented at the Fourth International Conference on Multiphase Flow (ICMF 2001) in New Orleans, May 27-June 1, 2001 (Paper No. 909). The computed time averaged particle velocities and concentrations agree with Particle Image Velocimetry (PIV) measurements of velocities and concentrations, obtained using a combination of gamma-ray and X-ray densitometers, in a slurry bubble column, operated in the bubbly-coalesced fluidization regime with continuous flow of water. Both the experiment and the simulation show a down-flow of particles in the center of the column and up-flow near the walls and nearly uniform particle concentration. Normal and shear Reynolds stresses were constructed from the computed instantaneous particle velocities. The PIV measurement and the simulation produced instantaneous particle velocities. The PIV measurement and the simulation produced similar nearly flat horizontal profiles of turbulent kinetic energy of particles. To better understand turbulence we studied fluidization in a liquid-solid bed. This work was also presented at the Fourth International Conference on Multiphase Flow (ICMF 2001, Paper No. 910). To understand turbulence in risers, measurements were done in the IIT riser with 530 micron glass beads using a PIV

Method of detecting coolant leakages from the pipeways in FBR type reactors

International Nuclear Information System (INIS)

Sonoda, Yukio; Tamaoki, Tetsuo

1986-01-01

Purpose: To detect coolant leakage in the incore pipeways of loop type FBR type reactors in the initial stage at high sensitivity. Constitution: Temperature of the coolants sealed between incore pipeways and the buffle surrounding them is measured by thermocouples and coolant leakage is detected due to fluctuating components. A well-insertion type in which electrode is sealed with argon is used as the thermo-couples. Signals from the thermocouples are once amplified, removed with DC components and then only the fluctuating components are outputted. The fluctuating components are digitalized, passed through an adaptive digital filter and the RMS value as the difference between the output signal and the thermocouple signal is calculated. The calculated value is compared with a threshold value in a comparative calculator. If it exceeds the threshold value, it is judged as abnormal to display an alarm on an alarm display. In this way, the coolant leakage for the pipeways in the FBR type reactor can be detected on real time and at high sensitivity. (Kamimura, M.)

Instrument for detecting gas bubbles in sodium filled pipework

Energy Technology Data Exchange (ETDEWEB)

Stapleton, D

1973-08-01

An instrument employing an electromagnetic sensor is described. Gas bubbles down to 0.5 mm diameter can be detected in sodium filled pipework of 1 in. nominal bore at temperatures up to 400 deg C. Installation of the sensing head involves no break in the integrity of the pipework. Facilities to enable bubble size discrimination, bubble counting and timing pulse generation are provided in the signal processing unit. Initial operating experience has been gained on a sodium loop used for gas entrainment studies. (auth)

Modelling and simulation of a U-loop Reactor for Single Cell Protein Production

DEFF Research Database (Denmark)

Wu, Mengzhe; Huusom, Jakob Kjøbsted; Gernaey, Krist

2016-01-01

In this work, two approaches of modelling a one phase U-loop reactor are presented. A simple CSTR model consisting of first-principles dynamic process equations was implemented in Matlab. The results give a good indication of the basic understanding of the effect of changing operation conditions...

DEVELOPMENT OF A COMPUTATIONAL MULTIPHASE FLOW MODEL FOR FISCHER TROPSCH SYNTHESIS IN A SLURRY BUBBLE COLUMN REACTOR

Energy Technology Data Exchange (ETDEWEB)

Donna Post Guillen; Tami Grimmett; Anastasia M. Gribik; Steven P. Antal

2011-12-01

The Hybrid Energy Systems Testing (HYTEST) Laboratory at the Idaho National Laboratory was established to develop and test hybrid energy systems with the principal objective of reducing dependence on imported fossil fuels. A central component of the HYTEST is the slurry bubble column reactor (SBCR) in which the gas-to-liquid reactions are performed to synthesize transportation fuels using the Fischer Tropsch (FT) process. These SBCRs operate in the churn-turbulent flow regime, which is characterized by complex hydrodynamics, coupled with reacting flow chemistry and heat transfer. Our team is developing a research tool to aid in understanding the physicochemical processes occurring in the SBCR. A robust methodology to couple reaction kinetics and mass transfer into a four-field model (consisting of the bulk liquid, small bubbles, large bubbles and solid catalyst particles) consisting of thirteen species, which are CO reactant, H2 reactant, hydrocarbon product, and H2O product in small bubbles, large bubbles, and the bulk fluid plus catalyst is outlined. Mechanistic submodels for interfacial momentum transfer in the churn-turbulent flow regime are incorporated, along with bubble breakup/coalescence and two-phase turbulence submodels. The absorption and kinetic models, specifically changes in species concentrations, have been incorporated into the mass continuity equation. The reaction rate is based on the macrokinetic model for a cobalt catalyst developed by Yates and Satterfield. The model includes heat generation produced by the exothermic chemical reaction, as well as heat removal from a constant temperature heat exchanger. A property method approach is employed to incorporate vapor-liquid equilibrium (VLE) in a robust manner. Physical and thermodynamic properties as functions of changes in both pressure and temperature are obtained from VLE calculations performed external to the CMFD solver. The novelty of this approach is in its simplicity, as well as its

Preliminary decay heat calculations for the fuel loaded irradiation loop device of the RMB multipurpose Brazilian reactor

Energy Technology Data Exchange (ETDEWEB)

Campolina, Daniel; Costa, Antonio Carlos L. da; Andrade, Edison P., E-mail: campolina@cdtn.br, E-mail: aclp@cdtn.br, E-mail: epa@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (SETRE/CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores

2017-07-01

The structuring project of the Brazilian Multipurpose Reactor (RMB) is responsible for meeting the capacity to develop and test materials and nuclear fuel for the Brazilian Nuclear Program. An irradiation test device (Loop) capable of performing fuel test for power reactor rods is being conceived for RMB reflector. In this work preliminary neutronic calculations have been carried out in order to determine parameters to the cooling system of the Loop basic design. The heat released as a result of radioactive decay of fuel samples was calculated using ORIGEN-ARP and it resulted less than 200 W after 1 hour of irradiation interruption. (author)

Immobilized glucose oxidase--catalase and their deactivation in a differential-bed loop reactor.

Science.gov (United States)

Prenosil, J E

1979-01-01

Glucose oxidase containing catalase was immobilized with a copolymer of phenylenediamine and glutaraldehyde on pumice and titania carrier to study the enzymatic oxidation of glucose in a differential-bed loop reactor. The reaction rate was found to be first order with respect to the concentration of limiting oxygen substrate, suggesting a strong external mass-transfer resistance for all the flow rates used. The partial pressure of oxygen was varied from 21.3 up to 202.6 kPa. The use of a differential-bed loop reactor for the determination of the active enzyme concentration in the catalyst with negligible internal pore diffusion resistance is shown. Catalyst deactivation was studied, especially with respect to the presence of catalase. It is believed that the hydrogen peroxide formed in the oxidation reaction deactivates catalase first; if an excess of catalase is present, the deactivation of glucose oxidase remains small. The mathematical model subsequently developed adequately describes the experimental results.

MTR and PWR/PHWR in-pile loop safety in integration with the operation of multipurpose reactor - GAS

International Nuclear Information System (INIS)

Suharno; Aji, Bintoro; Sugiyanto; Rohman, Budi; Zarkasi, Amin S.; Giarno

1998-01-01

MTR and PWR/PHWR In-Pile Loop safety analysis in integration with the operation of Multipurpose Reactor - Gas has been carried out and completed. The assessment is emphasized on the function of the interface systems from the dependence of the operation and the evaluation to the possibility of leakage or failure of the in-pile part inside the reactor pool and reactor core. The analysis is refers to the logic function of the interface system and the possibility of leakage or failure of the in-pile part inside reactor pool and reactor core to consider the integrity of the core qualitatively. The results show that in normal and in transient conditions , the interface system meet the function requirement in safe integrated operation of in-pile loop and reactor. And the results of the possibility analysis of the leakage shows that the possibility based on mechanically assessment is very low and the impact to core integrity is nothing or can be eliminated. The possible position for leakage is on the flen on which one meter above the top level of the core, therefore no influence of leakage to the core

Power-level regulation and simulation of nonlinear pressurized water reactor core with xenon oscillation using H-infinity loop shaping control

Directory of Open Access Journals (Sweden)

Li Gang

2016-01-01

Full Text Available This investigation is to solve the power-level control issue of a nonlinear pressurized water reactor core with xenon oscillations. A nonlinear pressurized water reactor core is modeled using the lumped parameter method, and a linear model of the core is then obtained through the small perturbation linearization way. The H∞loop shapingcontrolis utilized to design a robust controller of the linearized core model.The calculated H∞loop shaping controller is applied to the nonlinear core model. The nonlinear core model and the H∞ loop shaping controller build the nonlinear core power-level H∞loop shaping control system.Finally, the nonlinear core power-level H∞loop shaping control system is simulatedconsidering two typical load processes that are a step load maneuver and a ramp load maneuver, and simulation results show that the nonlinear control system is effective.

Influences of non-uniform pressure field outside bubbles on the propagation of acoustic waves in dilute bubbly liquids.

Science.gov (United States)

Zhang, Yuning; Du, Xiaoze

2015-09-01

Predictions of the propagation of the acoustic waves in bubbly liquids is of great importance for bubble dynamics and related applications (e.g. sonochemistry, sonochemical reactor design, biomedical engineering). In the present paper, an approach for modeling the propagation of the acoustic waves in dilute bubbly liquids is proposed through considering the non-uniform pressure field outside the bubbles. This approach is validated through comparing with available experimental data in the literature. Comparing with the previous models, our approach mainly improves the predictions of the attenuation of acoustic waves in the regions with large kR0 (k is the wave number and R0 is the equilibrium bubble radius). Stability of the oscillating bubbles under acoustic excitation are also quantitatively discussed based on the analytical solution. Copyright © 2015 Elsevier B.V. All rights reserved.

Modeling of the evolution of bubble size distribution of gas-liquid flow inside a large vertical pipe. Influence of bubble coalescence and breakup models

International Nuclear Information System (INIS)

Liao, Yixiang; Lucas, Dirk

2011-01-01

The range of gas-liquid flow applications in today's technology is immensely wide. Important examples can be found in chemical reactors, boiling and condensation equipments as well as nuclear reactors. In gas-liquid flows, the bubble size distribution plays an important role in the phase structure and interfacial exchange behaviors. It is therefore necessary to take into account the dynamic change of the bubble size distribution to get good predictions in CFD. An efficient 1D Multi-Bubble-Size-Class Test Solver was introduced in Lucas et al. (2001) for the simulation of the development of the flow structure along a vertical pipe. The model considers a large number of bubble classes. It solves the radial profiles of liquid and gas velocities, bubble-size class resolved gas fraction profiles as well as turbulence parameters on basis of the bubble size distribution present at the given axial position. The evolution of the flow along the height is assumed to be solely caused by the progress of bubble coalescence and break-up resulting in a bubble size distribution changing in the axial direction. In this model, the bubble coalescence and breakup models are very important for reasonable predictions of the bubble size distribution. Many bubble coalescence and breakup models have been proposed in the literature. However, some obvious discrepancies exist in the models; for example, the daughter bubble size distributions are greatly different from different bubble breakup models, as reviewed in our previous publication (Liao and Lucas, 2009a; 2010). Therefore, it is necessary to compare and evaluate typical bubble coalescence and breakup models that have been commonly used in the literature. Thus, this work is aimed to make a comparison of several typical bubble coalescence and breakup models and to discuss in detail the ability of the Test Solver to predict the evolution of bubble size distribution. (orig.)

Vibrations measurement in fast and PWR reactor study

International Nuclear Information System (INIS)

Tigeot, Y.; Epstein, A.; Hareux, F.

1975-01-01

In the past severe damages have occured in several nuclear reactors, by structural vibrations induced by the primary cooling flow. To avoid this kind of troubles, the SEMT makes studies for two different types of reactors. For the light pressurized water reactors, some tests have been made on the SAFRAN test loop which is a three loop 1/8 scale internal model of a 900 MWe reactor. This study is actually undertaken jointly with Framatome. Elsewhere, measurements have been made on the Phenix fast breeder sodium reactor, and studies are planned for the Super Phenix reactor [fr

Dynamic of vapor bubble growth in fields of variable pressure

International Nuclear Information System (INIS)

Pedroso, H.K.

1982-01-01

A mathematical model for the description of the growth from an initial nucleus of a vapor bubble imersed in liquid, subjected to a loss of pressure is presented. The model is important for analysing LOCA (Loss of Coolant Acident) in P.W.R. type reactors. Several simplifications were made in the phenomenum governing equations. With such simplifications the heat diffusion equation became the determining factor for the bubble growth, and the problem was reduced to solve the heat diffusion equation for semi infinite solid whose surface temperature is a well known function of time (it is supposed that the surface temperature is equal to the saturation temperature of the liquid at the system pressure at a given moment). The model results in an analytical expression for the bubble radius as a function of time. Comparisons with experimental data and previous models were made, with reasonable agreement. (author) [pt

Reactor core for FBR type reactor

International Nuclear Information System (INIS)

Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

1996-01-01

In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

In-Pile Loop Safety in Integrated with the Multipurpose Reactor in the case of in-Pile Loop Leakage at the Core Position

International Nuclear Information System (INIS)

Suharno; Sugianto; Giarno; Aliq; Widodo, Surip; Aji, Bintoro; Purba, Julwan Hendry; Karyanta, Edy

1999-01-01

In-Pile Loop Safety analysis in integrated with the multipurpose reactor in the case of In-Pile Loop leakage at the core position has been conducted which intended to evaluate the failure of fuel element. By considering design of In-Pile Loop and the highest possibility position of of leakage, the failure of fuel element is emphasized on mechanical aspect. The thermal hydraulic aspect is not taken into account due to the condition that when the leakage occurred the reactor has been in shut down condition. It is determined that the spray attacks the top position of fuel element, and to be calculated the force, of spray that produces 1 cm deflection on the single fuel element. Using that four (4) fuel elements is calculated because in the real condition 4 fuel elements will undergo deflection of 43.8 kg is obtained that producing 1 cm deflection and the force of 1228 kg that causes failure on the bottom of fuel element as shear force is also obtained. Whatever the force, high or low, the damage of fuel element occurred at the bottom part or at the position of grid plate. Therefore there is no damage on the fuel part (uranium meat) and the releasing of radioactive material from fuel plate is not happened

FBR type reactor

International Nuclear Information System (INIS)

Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

1994-01-01

In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

Economic Optimizing Control for Single-Cell Protein Production in a U-Loop Reactor

DEFF Research Database (Denmark)

Drejer, André; Ritschel, Tobias Kasper Skovborg; Jørgensen, Sten Bay

2017-01-01

The production of single-cell protein (SCP) in a U-loop reactor by a methanotroph is a cost efficient sustainable alternative to protein from fish meal obtained by over-fishing the oceans. SCP serves as animal feed. In this paper, we present a mathematical model that describes the dynamics of SCP...

Degradation mechanisms of 4-chlorophenol in a novel gas-liquid hybrid discharge reactor by pulsed high voltage system with oxygen or nitrogen bubbling.

Science.gov (United States)

Zhang, Yi; Zhou, Minghua; Hao, Xiaolong; Lei, Lecheng

2007-03-01

The effect of gas bubbling on the removal efficiency of 4-chlorophenol (4-CP) in aqueous solution has been investigated using a novel pulsed high voltage gas-liquid hybrid discharge reactor, which generates gas-phase discharge above the water surface simultaneously with the spark discharge directly in the liquid. The time for 100% of 4-CP degradation in the case of oxygen bubbling (7 min) was much shorter than that in the case of nitrogen bubbling (25 min) as plenty of hydrogen peroxide and ozone formed in oxygen atmosphere enhanced the removal efficiency of 4-CP. Except for the main similar intermediates (4-chlorocatechol, hydroquinone and 1,4-benzoquinone) produced in the both cases of oxygen and nitrogen bubbling, special intermediates (5-chloro-3-nitropyrocatechol, 4-chloro-2-nitrophenol, nitrate and nitrite ions) were produced in nitrogen atmosphere. The reaction pathway of 4-CP in the case of oxygen bubbling was oxygen/ozone attack on the radical hydroxylated derivatives of 4-CP. However, in the case of nitrogen bubbling, hydroxylation was the main reaction pathway with effect of N atom on degradation of 4-CP.

CFD investigations of natural circulation between the RPV and the cooling pond of VVER-440 type reactors in incidental conditions during maintenance performed with the code CFX-4.3

International Nuclear Information System (INIS)

Legradi, G.; Aszodi, A.

2002-01-01

During the annual maintenance of the VVER-440 type reactors, the RPV, the cooling pond and the transfer pond form a connected flow domain. The reactor is cooled by the natural circulation, which develops in one or two main loops. The cooling pond has its own cooling loops. CFD calculations have been performed with the CFX-4.3 code to investigate whether it is possible to cool the reactor core in case the main loops are lost and other emergency systems are not available. The results point out that the cooling system of the cooling pond is not capable of cooling the reactor core with the present connection. Therefore, modifications of the cooling system are investigated which would make it suitable for removing the remanent heat from the core.(author)

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.)

FBR type reactor

International Nuclear Information System (INIS)

Hayase, Tamotsu.

1991-01-01

The present invention concerns an FBR type reactor in which transuranium elements are eliminated by nuclear conversion. There are loaded reactor core fuels being charged with mixed oxides of plutonium and uranium, and blanket fuels mainly comprising depleted uranium. Further, liquid sodium is used as coolants. As transuranium elements, isotope elements of neptunium, americium and curium contained in wastes taken out from light water reactors or the composition thereof are used. The reactor core comprises a region with a greater mixing ratio and a region with a less mixing ratio of the transuranium elements. The mixing ratio of the transuranium elements is made greater for the fuels in the reactor core region at the boundary with the blanket of great neutron leakage. With such a constitution, since the positive reactivity value at the reactor core central portion is small in the Na void reactivity distribution in the reactor core, the positive reactivity is small upon Na boiling in the reactor core central region upon occurrence of imaginable accident, to attain reactor safety. (I.N.)

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

1987-01-01

Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

Investigation of hydrogen generation in a three reactor chemical looping reforming process

International Nuclear Information System (INIS)

Khan, Mohammed N.; Shamim, Tariq

2016-01-01

Highlights: • Three-reactor based chemical looping reforming system for hydrogen production. • Investigation of operating parameters using a system-level model. • Optimum operating conditions for hydrogen production are identified. • Different operating parameters affect the reactor temperatures differently. - Abstract: Chemical looping reforming (CLR) is a relatively new method to produce hydrogen (H_2) and is also used as an energy conversion method for solid, liquid or gaseous fuels. There are various advantages of this method such as inherent carbon dioxide (CO_2) capture, minimal NOx emissions and the H_2 production. In this process, there is no direct contact between the fuel and oxidizer. This method utilizes oxygen from an oxygen carrier which may be a transition metal. The idea is to split the combustion process into three separate sub-processes by employing three separate reactors: air reactor where the oxygen carrier is oxidized by air, fuel reactor where natural gas is oxidized to produce a stream of CO_2 and H_2O and steam reactor where the steam is reduced to produce H_2. In this study, a thermodynamic model with iron oxides as oxygen carrier has been developed using Aspen Plus by employing conservation of mass and energy for all the components of the CLR system. The developed model was employed to investigate the effect of various operating parameters such as mass flow rates of air, fuel, steam and oxygen carrier and fraction of inert material on H_2 and CO_2 production and key reactor temperatures. The results show that the H_2 production increases with the increase in air, fuel and steam flow rates up to a certain limit and stays constant for higher flow rates. The CO_2 production follows a similar trend. Similarly, the H_2 production also increases with the increase in oxide flow rate and fraction of inert material up to a particular value, but then decrease for higher oxide flow rates and inert fractions. Reactor temperatures were also

Work related to increasing the exploitation and experimental possibilities of the RA reactor, 05. Independent CO2 loop for cooling the samples irradiated in the RA vertical experimental channels (I-IV), Part II, IZ-240-0379-1963, Vol. II Head of the low temperature RA reactor coolant loop

International Nuclear Information System (INIS)

Pavicevic, M.

1963-07-01

The objective of the project was to design the head of the CO 2 coolant loop for cooling the materials during irradiation in the RA reactor. Six heads of coolant loops will be placed in the RA reactor, two in the region of heavy water in the experimental channels VEK-6 and four in the graphite reflector in the channels VEK-G. maximum generated heat in the heads of the coolant loop is 10500 kcal/h and minimum generated heat is 1500 kcal/h. The loops are cooled by CO 2 gas, coolant flow is 420 kg/h, and the pressure is 4.5 atu. There is a need to design and construct the secondary coolant loop for the low temperature coolant loop. This volume includes technical specifications of the secondary CO 2 loop with instructions for construction and testing; needed calculations; specification of materials; cost estimation for materials, equipment and construction; and graphical documentation [sr

DNA loop domain organization in nucleoids from cells of different types.

Science.gov (United States)

Afanasieva, Katerina; Chopei, Marianna; Lozovik, Alexandra; Semenova, Anastasia; Lukash, Lyubov; Sivolob, Andrei

2017-01-29

The loop domain organization of chromatin plays an important role in transcription regulation and thus may be assumed to vary in cells of different types. We investigated the kinetics of DNA loop migration during single cell gel electrophoresis (the comet assay) for nucleoids obtained from human lymphocytes, lymphoblasts and glioblastoma T98G cells. The results confirm our previous observation that there are three parts of DNA in nucleoids: DNA on the nucleoid surface, loops up to ∼150 kb inside the nucleoid, and larger loops that cannot migrate. However, the relative amounts of the three parts were found to be very different for different cell types. The distributions of the loop length up to 150 kb were shown to be exponential, with the distribution parameter, the loop density, to be dependent on the cell type. Copyright © 2016 Elsevier Inc. All rights reserved.

Preliminary Feasibility, Design, and Hazard Analysis of a Boiling Water Test Loop Within the Idaho National Laboratory Advanced Test Reactor National Scientific User Facility

International Nuclear Information System (INIS)

Gerstner, Douglas M.

2009-01-01

The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The ATR and its support facilities are located at the Idaho National Laboratory (INL). A Boiling Water Test Loop (BWTL) is being designed for one of the irradiation test positions within the. The objective of the new loop will be to simulate boiling water reactor (BWR) conditions to support clad corrosion and related reactor material testing. Further it will accommodate power ramping tests of candidate high burn-up fuels and fuel pins/rods for the commercial BWR utilities. The BWTL will be much like the pressurized water loops already in service in 5 of the 9 'flux traps' (region of enhanced neutron flux) in the ATR. The loop coolant will be isolated from the primary coolant system so that the loop's temperature, pressure, flow rate, and water chemistry can be independently controlled. This paper presents the proposed general design of the in-core and auxiliary BWTL systems; the preliminary results of the neutronics and thermal hydraulics analyses; and the preliminary hazard analysis for safe normal and transient BWTL and ATR operation

Return momentum effect on reactor coolant water level distribution during mid-loop conditions

International Nuclear Information System (INIS)

Seo, Jae Kwang; Yang, Jae Young; Park, Goon Cherl

2001-01-01

An accurate prediction of the Reactor Coolant System( RCS) water level is of importance in the determination of the allowable operating range to ensure safety during mid-loop operations. However, complex hydrualic phenomena induced by the Shutdown Cooling System (SCS) return momentum causes different water levels from those in the loop where the water level indicators are located. This was apparently observed at the pre-core cold hydro test of the Younggwang Nuclear Unit 3 (YGN 3) in Korea. In this study, in order to analytically understand the effect of the SCS return momentum on the RCS water level distribution, a model using a one-dimensional momentum and energy conservation for cylindrical channel, hydraulic jump in operating cold leg, water level build-up at the Reactor Vessel (RV) inlet nozzle, Bernoulli constant in downcomer region, and total water volume conservation has been developed. The model predicts the RCS water levels at various RCS locations during the mid-loop conditions and the calculation results were compared with the test data. The analysis shows that the hydraulic jump in the operating cold legs, in conjuction with the pressure drop throughout the RCS, is the main cause creating the water level differences at various RCS locations. The prediction results provide good explanations for the test data and show the significant effect of the SCS return momentum on the RCS water levels

Determination of size distribution of bubbles in a bubbly column two phase flows by ultrasound and neural networks

Energy Technology Data Exchange (ETDEWEB)

Baroni, Douglas B.; Lamy, Carlos A.; Bittencourt, Marcelo S.Q.; Pereira, Claudio M.N.A., E-mail: douglasbaroni@ien.gov.b, E-mail: lamy@ien.gov.b, E-mail: bittenc@ien.gov.b, E-mail: cmnap@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Cunha Filho, Jurandyr S. [Escola Tecnica Estadual Visconde de Maua (ETEVM/RJ), Rio de Janeiro, RJ (Brazil); Motta, Mauricio S., E-mail: mmotta@cefet-rj.b [Centro Federal de Educacao Tecnologica Celso Suckow da Fonseca (CEFET/RJ), Rio de Janeiro, RJ (Brazil)

2011-07-01

The development of advanced nuclear reactor conceptions depends largely on the amount of available data to the designer. Non invasive ultrasonic techniques can contribute to the evaluation of gas-liquid two-phase regimes in the nuclear thermo-hydraulic circuits. A key-point for success of those techniques is the interpretation of the ultrasonic signal. In this work, a methodology based in artificial neural networks (ANN) is proposed to predict size distribution of bubbles in a bubbly flow. To accomplish that, an air feed system control was used to obtain specific bubbly flows in an experimental system utilizing a Plexiglas vertical bubbly column. Four different size distribution of bubbles were generated. The bubbles were photographed and measured. To evaluate the different size distribution of bubbles it was used the ultrasonic reflected echo on the opposite wall of the column. Then, an ANN has been developed for predicting size distribution of bubbles by using the frequency spectra of the ultrasonic signal as input. A trained artificial neural network using ultrasonic signal in the frequency domain can evaluate with a good precision the size distribution of bubbles generated in this system. (author)

Determination of size distribution of bubbles in a bubbly column two phase flows by ultrasound and neural networks

International Nuclear Information System (INIS)

Baroni, Douglas B.; Lamy, Carlos A.; Bittencourt, Marcelo S.Q.; Pereira, Claudio M.N.A.; Cunha Filho, Jurandyr S.; Motta, Mauricio S.

2011-01-01

The development of advanced nuclear reactor conceptions depends largely on the amount of available data to the designer. Non invasive ultrasonic techniques can contribute to the evaluation of gas-liquid two-phase regimes in the nuclear thermo-hydraulic circuits. A key-point for success of those techniques is the interpretation of the ultrasonic signal. In this work, a methodology based in artificial neural networks (ANN) is proposed to predict size distribution of bubbles in a bubbly flow. To accomplish that, an air feed system control was used to obtain specific bubbly flows in an experimental system utilizing a Plexiglas vertical bubbly column. Four different size distribution of bubbles were generated. The bubbles were photographed and measured. To evaluate the different size distribution of bubbles it was used the ultrasonic reflected echo on the opposite wall of the column. Then, an ANN has been developed for predicting size distribution of bubbles by using the frequency spectra of the ultrasonic signal as input. A trained artificial neural network using ultrasonic signal in the frequency domain can evaluate with a good precision the size distribution of bubbles generated in this system. (author)

Non-catalytic alcoholysis process for production of biodiesel fuel by using bubble column reactor

Science.gov (United States)

Hagiwara, S.; Nabetani, H.; Nakajima, M.

2015-04-01

-edible lipids by use of the SMV reactor has not been examined yet. Therefore, this study aims to investigate the productivity of biodiesel produced from waste vegetable oils using the SMV reactor. Biodiesel fuel is a replacement for diesel as a fuel produced from biomass resources. It is generally produced as a FAME derived from vegetable oil by using alkaline catalyzed alcoholysis process. This alkaline method requires deacidification process prior to the reaction process and the alkaline catalyst removal process after the reaction. Those process increases the total cost of biodiesel fuel production. In order to solve the problems in the conventional alkaline catalyzed alcoholysis process, the authors proposed a non-catalytic alcoholysis process called the Superheated Methanol Vapor (SMV) method with bubble column reactor. So, this study aims to investigate the productivity of biodiesel produced from vegetable oils and other lipids using the SMV method with bubble column reactor.

Non-catalytic alcoholysis process for production of biodiesel fuel by using bubble column reactor

International Nuclear Information System (INIS)

Hagiwara, S; Nabetani, H; Nakajima, M

2015-01-01

-edible lipids by use of the SMV reactor has not been examined yet. Therefore, this study aims to investigate the productivity of biodiesel produced from waste vegetable oils using the SMV reactor. Biodiesel fuel is a replacement for diesel as a fuel produced from biomass resources. It is generally produced as a FAME derived from vegetable oil by using alkaline catalyzed alcoholysis process. This alkaline method requires deacidification process prior to the reaction process and the alkaline catalyst removal process after the reaction. Those process increases the total cost of biodiesel fuel production. In order to solve the problems in the conventional alkaline catalyzed alcoholysis process, the authors proposed a non-catalytic alcoholysis process called the Superheated Methanol Vapor (SMV) method with bubble column reactor. So, this study aims to investigate the productivity of biodiesel produced from vegetable oils and other lipids using the SMV method with bubble column reactor

Control of the loop IRENE in the experimental reactor OSIRIS: calculation of the departure from nucleate boiling ratio in real time

International Nuclear Information System (INIS)

Lucot, Michel; Obin, Norbert.

1980-10-01

The loop IRENE is an experimental device to study the behaviour of pressurized light-water fuel pins under neutron flux in the Saclay Nuclear Research Centre reactor OSIRIS. It can take 8 fuel pins and allows irradiations to be carried out under thermodynamic, hydraulic and chemical conditions corresponding to those of a PWR power reactor. A brief outline of the loop specifications is followed by an account of the difficulties raised by operating with saturation temperature at the test channel outlet and it is shown how the microprocessor loop pilots are used to solve these problems [fr

Waves reflected by solid wall and wave interaction in vapour bubbly liquids

International Nuclear Information System (INIS)

Duong, N.H.; Nguyen, V.T.

2004-01-01

The vapour bubbly liquids are met in many natural and industrial processes, including in energy equipment. In the nuclear power plants this kind of medium appears in reactor cores (PWR, BWR and etc.), in turbine generators and in heat transfer loops. Due to some circumstances (for example, a hit caused by detonations or strong collisions) the pressure waves can appear in the bubbly liquid medium contained in those facilities. These waves propagate in the mixtures and interact with themselves and with structures. It is important that what will occur during mentioned above processes. The knowledge of this kind processes will be useful for analysing the different sorts of the processes occurred in the energy facilities where the vapor bubbly liquids are used as working or heat transfer medium, like nuclear power plants, and also useful in finding the measures for prevention of unfavourable phenomena (for example, during wave interactions maybe appear too high pressures, which could lead into damages of facilities and etc.) and safety operating the equipment. From the physical point of view, the waves in this kind of medium are interesting that owing to non-linear, dispersion and dissipation effects the wave patterns in them may be diverse and easy altered. In the paper the investigation results of the waves reflected by solid wall or structure of the moderate intensity shock waves, and the behaviour of pressure in the process of wave interaction in some mixtures of liquid with vapour bubbles (of radium ∼1 mm) are presented. (author)

Work related to increasing the exploitation and experimental possibilities of the RA reactor, 05. Independent CO2 loop for cooling the samples irradiated in the RA vertical experimental channels (IIV), Part I, IZ-240-o379-1963, Vol. I, Head of the low temperature RA reactor coolant loop

International Nuclear Information System (INIS)

Pavicevic, M.

1963-07-01

The objective of the project was to design the head of the CO 2 coolant loop for cooling the materials during irradiation in the RA reactor. Six heads of coolant loops will be placed in the RA reactor, two in the region of heavy water in the experimental channels VEK-6 and four in the graphite reflector in the channels VEK-G. Materials for irradiation are metallurgy and chemical samples. In addition to the project objectives, this volume includes technical specifications of the coolant loop head, thermal calculations, calculations of mechanical stress, antireactivity and activation of the construction materials, cost estimation, scheme of the coolant loop head, diagrams of CO 2 gas temperature, thermal neutron flux distribution, design specifications of two proposed solutions for head of low temperature coolant loop [sr

Neutron transport in irradiation loops (IRENE loop)

International Nuclear Information System (INIS)

Sarsam, Maher.

1980-09-01

This thesis is composed of two parts with different aspects. Part one is a technical description of the loop and its main ancillary facilities as well as of the safety and operational regulations. The measurement methods on the model of the ISIS reactor and on the loop in the OSIRIS reactor are described. Part two deals with the possibility of calculating the powers dissipated by each rod of the fuel cluster, using appropriate computer codes, not only in the reflector but also in the core and to suggest a method of calculation [fr

Some in-reactor loop experiments on corrosion product transport and water chemistry

International Nuclear Information System (INIS)

Balakrishnan, P.V.; Allison, G.M.

1978-01-01

A study of the transport of activated corrosion products in the heat transport circuit of pressurized water-cooled nuclear reactors using an in-reactor loop showed that the concentration of particulate and dissolved corrosion products in the high-temperature water depends on such chemical parameters as pH and dissolved hydrogen concentration. Transients in these parameters, as well as in temperature, generally increase the concentration of suspended corrosion products. The maximum concentration of particles observed is much reduced when high-flow, high-temperature filtration is used. Filtration also reduces the steady-state concentration of particles. Dissolved corrosion products are mainly responsible for activity accumulation on surfaces. The data obtained from this study were used to estimate the rate constants for some of the transfer processes involved in the contamination of the primary heat transport circuit in water-cooled nuclear power reactors

Numerical study on the characteristics of air bubble oscillation in water

International Nuclear Information System (INIS)

Kim, Hwan Yeol; Bae, Yoon Yeong

2003-01-01

In both a boiling water reactor and an advanced type of pressurized water reactor under construction in Korea named APR1400, when a pressure relieving system is in operation, water, air and steam discharge successively into a sub-cooled water pool through spargers. Among the phenomena occurring during the discharging processes, the air bubble clouds with a low-frequency and high-amplitude oscillation may result in significant damage to the submerged structures if the resonance between the bubble clouds and structures occur. The phenomena involved are so complicated that most predictions of frequency and pressure loads have resorted to experimental work and computational approach has been precluded. This study deals with a numerical prediction of the pressure field generated by the oscillation of air bubble. The analysis was performed by using a commercial thermal hydraulic analysis code, FLUENT, version 4.5. The multiphase flows of water, air and steam were simulated by the VOF (Volume Of Fluid) model contained in the code. Unlike the author's previous study, the LRR (Load Reduction Ring) of the sparger is artificially blocked for the investigation of LRR effects on the pressure field. It also includes the effect of air mass and inlet pressure in the piping on the pressure field. (author)

Self-organization of voids, gas bubbles and dislocation patterns under irradiation

International Nuclear Information System (INIS)

Dubinko, V.I.; Turkin, A.A.

1993-01-01

In the present paper three examples of self-organization in solids under irradiation are considered on the basis of original mechanisms, namely, the ordering of voids in void lattices under high temperature irradiation, the alignment of gas bubbles in bubble lattices under low-temperature gas atom implantation, and the formation of superdislocations (one-dimensional pile-ups of dislocation loops) and other dislocation patterns in the regimes of medium and high temperature irradiation. The ordering of cavities (i.e.voids or gas bubbles) is shown to arise due to a dissipative interaction between cavities induced by the interstitial dislocation loop absorption and punching, respectively, which represent anisotropic mechanisms of atomic transport. The dislocation patterning is shown to be driven by the dependence of dislocation bias for absorption of self-interstitial atoms on the dislocation arrangement. (author). 57 refs., 1 tab., 12 figs

MLR reactor

International Nuclear Information System (INIS)

Ryazantsev, E.P.; Egorenkov, P.M.; Nasonov, V.A.; Smimov, A.M.; Taliev, A.V.; Gromov, B.F.; Kousin, V.V.; Lantsov, M.N.; Radchenko, V.P.; Sharapov, V.N.

1998-01-01

The Material Testing Loop Reactor (MLR) development was commenced in 1991 with the aim of updating and widening Russia's experimental base to validate the selected directions of further progress of the nuclear power industry in Russia and to enhance its reliability and safety. The MLR reactor is the pool-type one. As coolant it applies light water and as side reflector beryllium. The direction of water circulation in the core is upward. The core comprises 30 FA arranged as hexagonal lattice with the 90-95 mm pitch. The central materials channel and six loop channels are sited in the core. The reflector includes up to 11 loop channels. The reactor power is 100 MW. The average power density of the core is 0.4 MW/I (maximal value 1.0 MW/l). The maximum neutron flux density is 7.10 14 n/cm 2 s in the core (E>0.1 MeV), and 5.10 14 n/cm 2 s in the reflector (E<0.625 eV). In 1995 due to the lack of funding the MLR designing was suspended. (author)

Bubble inductors: Pneumatic tuning of a stretchable inductor

Science.gov (United States)

Lazarus, Nathan; Bedair, Sarah S.

2018-05-01

From adaptive matching networks in power systems to channel selectable RF filters and circuitry, tunable inductors are fundamental components for circuits requiring reconfigurability. Here we demonstrate a new continuously tunable inductor based on physically stretching the inductor traces themselves. Liquid-metal-based stretchable conductors are wrapped around a pneumatic bubble actuator, allowing the inductor to be collapsed or expanded by application of pressure. In vacuum the bubble collapses, bringing the loop area to nearly zero, while positive pressure brings a dramatic increase in area and loop inductance. Using this approach, the inductor demonstrated in this work was able to achieve a tuning ratio of 2.6 with 1-2 second response time. With conductors available that can stretch by hundreds of percent, this technique is promising for very large tuning ratios in continuously tunable inductors.

Tank type LMFBR type reactors

International Nuclear Information System (INIS)

Shimizu, Hiroshi

1985-01-01

Purpose: To detect the abnormality in the suspended body or reactor core supporting structures thereby improve the safety and reliability of tank type LMFBR reactors. Constitution: Upon inspection during reactor operation period, the top end of the gripper sensing rod of a fuel exchanger is abutted against a supporting bed and the position of the reactor core supporting structures from the roof slab is measured by a stroke measuring device. Then, the sensing rod is pulled upwardly to abut against the arm portion and the position is measured by the stroke measuring device. The measuring procedures are carried out for all of the sensing rods and the measured values are compared with a previously determined value at the initial stage of the reactor operation. As a result, it is possible to detect excess distortions and abnormal deformation in the suspended body or reactor core supporting structures. Furthermore, integrity of the suspended body against thermal stresses can be secured by always measuring the coolant liquid level by the level measuring sensor. (Kamimura, M.)

MULTI-LOOP CONTROL DESIGN IN MULTIVARIABLE (2X2 CONTINUOUS STIRRED TANK REACTOR

Directory of Open Access Journals (Sweden)

Abdul Wahid

2015-06-01

Full Text Available With this study, the design and tuning of multi-loop for multivariable (2x2 CSTR will be made in order to achieve optimum CSTR control performance. This study used Bequette model reactor and MATLAB software and is expected to be able to cope with disturbances in the reactor so that the reactor system is able to stabilize quickly despite the distractions. In this study, the design will be made using multi-loop approach, along with PI controller as the next step. Then, BLT and auto-tune tuning method will be used in PI controller and given disturbances to both of tuning method. The controller performances are then compared. Results of the study are then analyzed for discussions and conclusions. Results from this study have shown that in terms of disturbance rejection, BLT is better than auto-tune based on comparison between both of controller performances. For IAE for the case of temperature, BLT is 30% better than auto-tune, but it is almost the same for the case of concentration. For settling time for the case of concentration, BLT is 30% better than auto-tune, and for the case of temperature, BLT is 18% better than auto-tune. For rise time for the case of concentration and temperature, BLT is 30% better than auto-tune.

A numerical study of cutting bubbles with a wire mesh

NARCIS (Netherlands)

Baltussen, M.W.; Kuipers, J.A.M.; Deen, N.G.

2017-01-01

Gas-liquid-solid flows are frequently encountered in chemical, petrochemical and biochemical industries. To overcome the heat and mass transfer limitations in trickle bed reactors and bubble slurry columns, respectively, a micro-structured bubble column (MSBC) can serve as an attractive alternative.

BWR type nuclear reactors

International Nuclear Information System (INIS)

Yamamoto, Toru.

1987-01-01

Purpose: To obtain reactor core characteristics with less changes in the excess reactivity due to fuel burnup even when the operation period varies. Constitution: In a BWR type reactor where fuel assemblies containing fuel rods incorporated with burnable poisons are arranged, the fuel assemblies are grouped into first fuel assemblies and second fuel assemblies. Then, the number of fuel rods incorporated with burnable poisons within the first fuel assemblies is made greater than that of the second fuel rods, while the concentration of the burnable poisons in the fuel rods incorporated with the burnable poisons in the first fuel assemblies is made lower than that of the fuel rods incorporated with the burnable poisons in the second fuel assemblies. In the BWR type reactor constituted in this way, the reactor core characteristics can be improved by changing the ratio between the first fuel assemblies and the second fuel assemblies charged to the reactor core, thereby decreasing the changes in the burnup of the excess reactivity. (Kamimura, M.)

The structure and dynamics of bubble-type vortex breakdown

Science.gov (United States)

Spall, R. E.; Ash, R. L.; Gatski, T. B.

1990-01-01

A unique discrete form of the Navier-Stokes equations for unsteady, three-dimensional, incompressible flow has been used to study vortex breakdown numerically. A Burgers-type vortex was introduced along the central axis of the computational domain, and allowed to evolve in space and time. By varying the strength of the vortex and the free stream axial velocity distribution, using a previously developed Rossby number criterion as a guide, the location and size of the vortex breakdown region was controlled. While the boundaries of the vortex breakdown bubble appear to be nominally symmetric, the internal flow field is not. Consequently, the mechanisms for mixing and entrainment required to sustain the bubble region are different from those suggested by earlier axisymmetric models. Results presented in this study, for a Reynolds number of 200, are in good qualitative agreement with higher Reynolds number experimental observations, and a variety of plots have been presented to help illuminate the fluid physics.

Improvements to PWR type reactors

International Nuclear Information System (INIS)

Ailloud, Jean; Monteil, Marcel.

1978-01-01

Improvements to pressurized water nuclear reactors are described, where the core coolant, called primary fluid, flows under the effect of a circulating pump in a primary loop between a steam generator and a pressure vessel containing the reactor core. The steam generator includes a bundle of tubes through which flows the primary fluid which exchanges calories with a secondary fluid, generally water, entering the generator as a liquid and issuing from it as steam. After expansion in turbines and recovery in a condenser, this steam is returned to the inside of the generator. Each primary fluid circulating pump is powered by a back-pressure turbine located in parallel with the high pressure section of the main turbine and hence fed with steam taken directly from the steam generator or the main steam pipe outside it [fr

Natural circulation of integrated-type marine reactor at inclined attitude

International Nuclear Information System (INIS)

Iyori, Isao; Aya, Izuo; Murata, Hiroyuki; Kobayashi, Michiyuki; Nariai, Hideki

1987-01-01

A steady-state single-phase natural circulation test was performed to clarify the effect of inclination by using a model of an integrated-type marine reactor. It was found that several types of flow pattern occur in the natural circulation loop corresponding to the range of inclination angle. Stable flow rates are sustained up to near 90 0 because of the occurrence of a driving force arising from those sections of the facility which were horizontal before the inclination. It was found that the temperature distribution in the steam generator at inclined attitude depends essentially only on the elevation z. The applicability of a one-dimensional analytical model was examined. It was clarified that employment of detailed U-turn flow paths, their correlation, and temperature-distribution function of core is essential for improvement. (orig.)

An equation oriented approach to steady state flowsheeting of methanol synthesis loop

International Nuclear Information System (INIS)

Fathikalajahi, J.; Baniadam, M.; Rahimpour, M.R.

2008-01-01

An equation-oriented approach was developed for steady state flowsheeting of a commercial methanol plant. The loop consists of fixed bed reactor, flash separator, preheater, coolers, and compressor. For steady sate flowsheeting of the plant mathematical model of reactor and other units are needed. Reactor used in loop is a Lurgi type and its configuration is rather complex. Previously reactor and flash separator are modeled as two important units of plant. The model is based on mass and energy balances in each equipment and utilizing some auxiliary equations such as rate of reaction and thermodynamics model for activity coefficients of liquid. In order to validate the mathematical model for the synthesis loop, some simulation data were performed using operating conditions and characteristics of the commercial plant. The good agreement between the steady state simulation results and the plant data shows the validity of the model

Closed-loop digital control of nuclear reactors characterized by spatial dynamics

International Nuclear Information System (INIS)

Bernard, J.A.; Henry, A.F.; Lanning, D.D.; Meyer, J.E.

1991-03-01

This report describes the theoretical development and the evaluation via both simulation and, to a lesser degree, experiment of a digital method for the closed-loop control of power and temperature in reactors characterized by spatial dynamics. The major conclusions of the research are that (1) the sophistication of advanced reactor physics and thermal-hydraulic nodal methods is now such that accurate, real-time models of spatially-dependent, heterogeneous reactor cores can be run on present-generation minicomputers; (2) operation of both present-day commercial reactors as well as the multi-modular reactors now being considered for construction in the United States could be significantly improved by incorporating model-generated information on in-core conditions in a digital controller; and (3) digital controllers for spatially-dependent reactors should have a hierarchical or multi-tiered structure consisting of supervisory algorithms that preclude challenges to the safety system, global control laws designed to provide an optimal response to temperature and power perturbations, and local control laws that maintain parameters such as the margin to departure from nucleate boiling within specification. The technology described is appropriate to present-day pressurized water reactors and to the proposed multi-modular designs. The end-product of this research was a (near) real-time analytic plant-estimation code that was given the acronym POPSICLE for POwer Plant SImulator and ControlLEr. POPSICLE's core neutronics model is based on a quasi-static transient solution of the analytic nodal diffusion equations. 126 refs., 159 figs., 17 tabs

Closed-loop digital control of nuclear reactors characterized by spatial dynamics

Energy Technology Data Exchange (ETDEWEB)

Bernard, J.A. (Massachusetts Inst. of Tech., Cambridge, MA (USA). Nuclear Reactor Lab.); Henry, A.F.; Lanning, D.D.; Meyer, J.E. (Massachusetts Inst. of Tech., Cambridge, MA (USA). Dept. of Nuclear Engineering)

1991-03-01

This report describes the theoretical development and the evaluation via both simulation and, to a lesser degree, experiment of a digital method for the closed-loop control of power and temperature in reactors characterized by spatial dynamics. The major conclusions of the research are that (1) the sophistication of advanced reactor physics and thermal-hydraulic nodal methods is now such that accurate, real-time models of spatially-dependent, heterogeneous reactor cores can be run on present-generation minicomputers; (2) operation of both present-day commercial reactors as well as the multi-modular reactors now being considered for construction in the United States could be significantly improved by incorporating model-generated information on in-core conditions in a digital controller; and (3) digital controllers for spatially-dependent reactors should have a hierarchical or multi-tiered structure consisting of supervisory algorithms that preclude challenges to the safety system, global control laws designed to provide an optimal response to temperature and power perturbations, and local control laws that maintain parameters such as the margin to departure from nucleate boiling within specification. The technology described is appropriate to present-day pressurized water reactors and to the proposed multi-modular designs. The end-product of this research was a (near) real-time analytic plant-estimation code that was given the acronym POPSICLE for POwer Plant SImulator and ControlLEr. POPSICLE's core neutronics model is based on a quasi-static transient solution of the analytic nodal diffusion equations. 126 refs., 159 figs., 17 tabs.

A master-follower type distributed scheme for reactor inlet temperature control

International Nuclear Information System (INIS)

Garcia, H.E.; Dean, E.M.; Vilim, R.B.

1995-01-01

This paper describes the implementation of a computer-based controller for regulating reactor inlet temperature in a pool-type power plant. The elements of the control system are organized in a master-follower hierarchical architecture that takes advantage of existing in-plant hardware and software to minimize the need for plant modifications. Low level control algorithms are executed on existing local digital controllers (followers) with the high level algorithms executed on a new plant supervisory computer (master). A distributed computing strategy provides integration of the existing and additional computer platforms. The control system operates by having the master controller first estimate the secondary sodium flow needed to achieve a given reactor inlet temperature. The estimated flow is then used as a setpoint by the follower controller to regulate sodium flow using a motor-generator pump set. The control system has been implemented in a Hardware-In-the-Loop (FM) setup and qualified for operation in the Experimental Breader reactor 11 of Argonne National Laboratory. Some HIL results are provided

Experimental study of steam bubble velocities and dimensions in the draught trunk of the AST-500 reactor simulator

International Nuclear Information System (INIS)

Shanin, V.K.; Drobkov, V.P.; Kulakov, I.V.; Khalmeh, M.V.

1988-01-01

Local characteristics for two-phase steam water flow in the vertical channel with 0.45 m diameter and 2 m length, which is the draught trunk of the AST-500 reactor simulator, are investigated. Steam bubble velocities and dimensions were determined by the time-of-flight method using the twinned conductometric transducers. The data obtained testify to the existance of unstable circulation flows in the trunk peripheral region. These flows effect considerably the steam phase motion in the trunk middle part. At the same time the circulation flows to a lesser degree affect steam bubble motion in the trunk low peripheral part and to the lesser degree affect the steam phase in the axial zone near the outlet from the heating section. So the data obtained confirm the conclusion, made earlier, about steam-water flow acceleration in the draught trunk central part

Small scale thermal-hydraulic experiment for stable operation of a pius-type reactor

International Nuclear Information System (INIS)

Tasaka, K.; Tamaki, M.; Imai, S.; Irianto, I.D.; Tsuji, Y.; Kukita, Y.

1994-01-01

Thermal-hydraulic experiments using a small-scale atmospheric pressure test loop have been performed for the Process Inherent Ultimate Safety (PIUS)-type reactor to develop the new pump speed feedback control system. Three feedback control systems based on the measurement of flow rate, differential pressure, and fluid temperature distribution in the lower density lock have been proposed and confirmed by a series of experiments. Each of the feedback control systems had been verified in the simulation experiment such as a start-up simulation test. The automatic pump speed control based on the fluid temperature at the lower density lock was quite effective to maintain the stratified interface between primary water and borated pool water for stable operation of the reactor. (author)

Process and kinetics of the fundamental radiation-electrochemical reactions in the primary coolant loop of nuclear reactors

International Nuclear Information System (INIS)

Kozomara-Maic, S.

1987-06-01

In spite of the rather broad title of this report, its major part is devoted to the corrosion problems at the RA reactor, i.e. causes and consequences of the reactor shutdown in 1979 and 1982. Some problems of reactor chemistry are pointed out because they are significant for future reactor operation. The final conclusion of this report is that corrosion processes in the primary coolant circuit of the nuclear reactor are specific and that radiation effects cannot be excluded when processes and reaction kinetics are investigated. Knowledge about the kinetics of all the chemical reactions occurring in the primary coolant loop are of crucial significance for safe and economical reactor operation [sr

Experimental Analysis of a Bubble Wake Influenced by a Vortex Street

Directory of Open Access Journals (Sweden)

Sophie Rüttinger

2018-01-01

Full Text Available Bubble column reactors are ubiquitous in engineering processes. They are used in waste water treatment, as well as in the chemical, pharmaceutical, biological and food industry. Mass transfer and mixing, as well as biochemical or chemical reactions in such reactors are determined by the hydrodynamics of the bubbly flow. The hydrodynamics of bubbly flows is dominated by bubble wake interactions. Despite the fact that bubble wakes have been investigated intensively in the past, there is still a lack of knowledge about how mass transfer from bubbles is influenced by bubble wake interactions in detail. The scientific scope of this work is to answer the question how bubble wakes are influenced by external flow structures like a vortex street behind a cylinder. For this purpose, the flow field in the vicinity of a single bubble is investigated systematically with high spatial and temporal resolution. High-speed Particle Image Velocimetry (PIV measurements are conducted monitoring the flow structure in the equatorial plane of the single bubble. It is shown that the root mean square (RMS velocity profiles downstream the bubble are influenced significantly by the interaction of vortices. In the presence of a vortex street, the deceleration of the fluid behind the bubble is compensated earlier than in the absence of a vortex street. This happens due to momentum transfer by cross-mixing. Both effects indicate that the interaction of vortices enhances the cross-mixing close to the bubble. Time series of instantaneous velocity fields show the formation of an inner shear layer and coupled vortices. In conclusion, this study shows in detail how the bubble wake is influenced by a vortex street and gives deep insights into possible effects on mixing and mass transfer in bubbly flows.

On reactor type comparisons for the next generation of reactors

International Nuclear Information System (INIS)

Alesso, H.P.; Majumdar, K.C.

1991-01-01

In this paper, we present a broad comparison of studies for a selected set of parameters for different nuclear reactor types including the next generation. This serves as an overview of key parameters which provide a semi-quantitative decision basis for selecting nuclear strategies. Out of a number of advanced reactor designs of the LWR type, gas cooled type, and FBR type, currently on the drawing board, the Advanced Light Water Reactors (ALWR) seem to have some edge over other types of the next generation of reactors for the near-term application. This is based on a number of attributes related to the benefit of the vast operating experience with LWRs coupled with an estimated low risk profile, economics of scale, degree of utilization of passive systems, simplification in the plant design and layout, modular fabrication and manufacturing. 32 refs., 1 fig., 3 tabs

CFD Investigation of the effects of bubble aerator layouts on hydrodynamics of an activated sludge channel reactor.

Science.gov (United States)

Hreiz, Rainier; Potier, Olivier; Wicks, Jim; Commenge, Jean-Marc

2018-03-08

In this paper, computational fluid dynamics (CFD) simulations are employed to characterize the effects of bubble aerator layouts (i.e. spatial arrangement) on the hydrodynamics in activated sludge (AS) reactors. The first configuration considered is a channel reactor with aerators placed alongside one lateral wall, for which velocity measurements are available in literature. CFD results were in good agreement with experimental data, which proves that the model is sufficiently accurate and predictive. Accordingly, simulations and numerical residence time distribution tests were conducted for different aerator layouts to determine their effects on the reactor hydrodynamics. The results revealed that the flow characteristics are extremely sensitive to the aerators arrangement given the high gas flow rates used in AS processes. Among the layouts investigated, the one where diffusers are placed all over the reactor floor has led to the least dispersive flow, i.e. which characteristics best tend toward that of an ideal plug flow reactor. Indeed, this flow field presented the lowest average turbulent diffusion and the most uniform axial velocity and turbulence fields. Such a flow behaviour is expected to be highly beneficial for biological treatment since it reduces pollutant dilution by axial diffusion and limits raw wastewater channelling to the outlet.

Studies on the closed-loop digital control of multi-modular reactors

International Nuclear Information System (INIS)

Bernard, J.A.; Henry, A.F.; Lanning, D.D.; Meyer, J.E.

1992-11-01

This report describes the theoretical development and the evaluation via both experiment and simulation of digital methods for the closed-loop control of power, temperature, and steam generator level in multi-modular reactors. The major conclusion of the research reported here is that the technology is currently available to automate many aspects of the operation of multi-modular plants. This will in turn minimize the number of required personnel and thus contain both operating and personnel costs, allow each module to be operated at a different power level thereby staggering the times at which refuelings would be needed, and maintain the competitiveness of US industry relative to foreign vendors who are developing and applying advanced control concepts. The technology described in this report is appropriate to the proposed multi-modular reactor designs and to present-generation pressurized water reactors. Its extension to boiling water reactors is possible provided that the commitment is made to create a real-time model of a BWR. The work reported here was performed by the Massachusetts Institute of Technology (MIT) under contract to the Oak Ridge National Laboratory (ORNL) and to the United States Department of Energy (Division of Industry and University Programs, Contract No. DE-FG07-90ER12930.)

Active acoustic leak detection in steam generator units of fast reactors

International Nuclear Information System (INIS)

Oriol, L.; Journeau, Ch.

1996-01-01

Steam generators (SG) of Fast Reactors can be subject to water leakage into the sodium secondary circuit, causing an exothermic chemical reaction with potential serious damage to plant. Within the framework of the European Fast Reactor project, the CEA has developed an active acoustic detection technique which, when used in parallel with passive acoustic detection, will lead to effective leak detection results in terms of reliability and false alarm rates. Whilst the passive method is based on the increase in acoustic noise generated by the reaction, the active method takes advantage of the acoustic attenuation by the hydrogen bubbles produced. The method has been validated: in water, during laboratory testing at the Centre d'Etudes de Cadarache; in sodium, at the ASB loop at Bensberg (Germany) and at AEA Dounreay (Scotland). Full analysis of the tests carried out on the SG of the Prototype Fast Reactor in 1994 during end-of-life testing should lead to reactor validation on the method. (authors)

TESTING OF GAS REACTOR MATERIALS AND FUEL IN THE ADVANCED TEST REACTOR

International Nuclear Information System (INIS)

Grover, S.B.

2004-01-01

The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

Testing of Gas Reactor Materials and Fuel in the Advanced Test Reactor

International Nuclear Information System (INIS)

S. Blaine Grover

2004-01-01

The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

A study of bubbly flow characteristics in a vertical tube using wire mesh tomography

International Nuclear Information System (INIS)

Wangjiraniran, Weerin; Motegi, Yuichi; Kikura, Hiroshige; Aritomi, Masanori; Richter, Steffen; Yamamoto, Kazuhiko

2003-01-01

For the development of nuclear reactors and the assessment of their safety features, the development of computer code with the high quantity database from the measurement as well as the understanding of the multiphase flow physics are necessary. In this study, the characteristics of bubbly flow in a vertical tube are investigated using Wire Mesh Tomography (WMT). Void fraction is detected from the dependency of electrical conductivity on the local void fraction. The developed sensor is a circular type with two parallel measuring planes to have the capability of gas velocity and bubble size evaluation. The experiment is conducted in a 50 mm ID tube at the fully developed condition (93D). The mean bubble size is treated as a constant parameter independent from the superficial gas and liquid velocity by using the bubble generator with a water sub flow. The result shows the capability of WMT for bubbly flow characteristic study. The effects of superficial gas and liquid velocity and the additional bubble intensity on the void fraction distribution are presented. These effects are supposed to change the lateral lift force in both magnitude and directions which induce the bubble migrated toward to or depart from the wall. (author)

The SPHINX reactor for engineering tests

International Nuclear Information System (INIS)

Adamov, E.O.; Artamkin, K.N.; Bovin, A.P.; Bulkin, Y.M.; Kartashev, E.F.; Korneev, A.A.; Stenbok, I.A.; Terekhov, A.S.; Khmel'Shehikov, V.V.; Cherkashov, Y.M.

1990-01-01

A research reactor known as SPHINX is under development in the USSR. The reactor will be used mainly to carry out tests on mock-up power reactor fuel assemblies under close-to-normal parameters in experimental loop channels installed in the core and reflector of the reactor, as well as to test samples of structural materials in ampoule and loop channels. The SPHINX reactor is a channel-type reactor with light-water coolant and moderator. Maximum achievable neutron flux density in the experimental channels (cell composition 50% Fe, 50% H 2 O) is 1.1 X 10 15 neutrons/cm 2 · s for fast neutrons (E > 0.1 MeV) and 1.7 X 10 15 for thermal neutrons at a reactor power of 200 MW. The design concepts used represent a further development of the technical features which have met with approval in the MR and MIR channel-type engineering test reactors currently in use in the USSR. The 'in-pond channel' construction makes the facility flexible and eases the carrying out of experimental work while keeping discharges of radioactivity into the environment to a low level. The reactor and all associated buildings and constructions conform to modern radiation safety and environmental protection requirements

Cap Bubble Drift Velocity in a Confined Test Section

International Nuclear Information System (INIS)

Xiaodong Sun; Seungjin Kim; Mamoru Ishii; Lincoln, Frank W.; Beus, Stephen G.

2002-01-01

In the two-group interfacial area transport equation, bubbles are categorized into two groups, i.e., spherical/distorted bubbles as group 1 and cap/slug/churn-turbulent bubbles as group 2. The bubble rise velocities for both groups of bubbles may be estimated by the drift flux model by applying different distribution parameters and drift velocities for both groups. However, the drift velocity for group 2 bubbles is not always applicable (when the wall effect becomes important) as in the current test loop of interest where the flow channel is confined by two parallel flat walls, with a dimension of 200-mm in width and 10-mm in gap. The previous experiments indicated that no stable slug flow existed in this test section, which was designed to permit visualization of the flow patterns and bubble characteristics without the distortion associated with curved surfaces. In fact, distorted cap bubbly and churn-turbulent flow was observed. Therefore, it is essential to developed a correlation for cap bubble drift velocity in this confined flow channel. Since the rise velocity of a cap bubble depends on its size, a high-speed movie camera is used to capture images of cap bubbles to obtain the bubble size information. Meanwhile, the rise velocity of cap and elongated bubbles (called cap bubbles hereafter) is investigated by examining the captured images frame by frame. As a result, the conventional correlation of drift velocity for slug bubbles is modified and acceptable agreements between the measurements and correlation estimation are achieved

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

Numerical simulation on single bubble rising behavior in liquid metal using moving particle semi-implicit method

International Nuclear Information System (INIS)

Zuo Juanli; Tian Wenxi; Qiu Suizheng; Chen Ronghua; Su Guanghui

2011-01-01

The gas-lift pump in liquid metal cooling fast reactor (LMFR) is an innovational conceptual design to enhance the natural circulation ability of reactor core. The two-phase flow character of gas-liquid metal makes significant improvement of the natural circulation capacity and reactor safety. In present basic study, the rising behavior of a single nitrogen bubble in five kinds of liquid metals (lead bismuth alloy, liquid kalium, sodium, potassium sodium alloy and lithium lead alloy) was numerically simulated using moving particle semi-implicit (MPS) method. The whole growing process of single nitrogen bubble in liquid metal was captured. The bubble shape and rising speed of single nitrogen bubble in each liquid metal were compared. The comparison between simulation results using MPS method and Grace graphical correlation shows a good agreement. (authors)

Design study on sodium-cooled large-scale reactor

International Nuclear Information System (INIS)

Shimakawa, Yoshio; Nibe, Nobuaki; Hori, Toru

2002-05-01

In Phase 1 of the 'Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)', an advanced loop type reactor has been selected as a promising concept of sodium-cooled large-scale reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase 2 of the F/S, it is planed to precede a preliminary conceptual design of a sodium-cooled large-scale reactor based on the design of the advanced loop type reactor. Through the design study, it is intended to construct such a plant concept that can show its attraction and competitiveness as a commercialized reactor. This report summarizes the results of the design study on the sodium-cooled large-scale reactor performed in JFY2001, which is the first year of Phase 2. In the JFY2001 design study, a plant concept has been constructed based on the design of the advanced loop type reactor, and fundamental specifications of main systems and components have been set. Furthermore, critical subjects related to safety, structural integrity, thermal hydraulics, operability, maintainability and economy have been examined and evaluated. As a result of this study, the plant concept of the sodium-cooled large-scale reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowing down candidate concepts at the end of Phase 2. (author)

Bubble coalescence in a Newtonian fluid

Science.gov (United States)

Garg, Vishrut; Basaran, Osman

2017-11-01

Bubble coalescence plays a central role in the hydrodynamics of gas-liquid systems such as bubble column reactors, spargers, and foams. Two bubbles approaching each other at velocity V coalesce when the thin film between them ruptures, which is often the rate-limiting step. Experimental studies of this system are difficult, and recent works provide conflicting results on the effect of V on coalescence times. We simulate the head-on approach of two bubbles of equal radii R in an incompressible Newtonian fluid (density ρ, viscosity μ, and surface tension σ) by solving numerically the free boundary problem comprised of the Navier Stokes and continuity equations. Simulations are made challenging by the existence of highly disparate lengthscales, i.e. film thickness and drop radii, which are resolved by using the method of elliptic mesh generation. For a given liquid, the bubbles are shown to coalesce for all velocities below a critical value. The effects of Ohnesorge number Oh = μ /√{ ρσR } on coalescence time and critical velocity are also investigated.

Sodium cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Hokkyo, N; Inoue, K; Maeda, H

1968-11-21

In a sodium cooled fast neutron reactor, an ultrasonic generator is installed at a fuel assembly hold-down mechanism positioned above a blanket or fission gas reservoir located above the core. During operation of the reactor an ultrsonic wave of frequency 10/sup 3/ - 10/sup 4/ Hz is constantly transmitted to the core to resonantly inject the primary bubble with ultrasonic energy to thereby facilitate its growth. Hence, small bubbles grow gradually to prevent the sudden boiling of sodium if an accident occurs in the cooling system during operation of the reactor.

Reactor

International Nuclear Information System (INIS)

Toyama, Masahiro; Kasai, Shigeo.

1978-01-01

Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

Interaction mechanism of double bubbles in hydrodynamic cavitation

Science.gov (United States)

Li, Fengchao; Cai, Jun; Huai, Xiulan; Liu, Bin

2013-06-01

Bubble-bubble interaction is an important factor in cavitation bubble dynamics. In this paper, the dynamic behaviors of double cavitation bubbles driven by varying pressure field downstream of an orifice plate in hydrodynamic cavitation reactor are examined. The bubble-bubble interaction between two bubbles with different radii is considered. We have shown the different dynamic behaviors between double cavitation bubbles and a single bubble by solving two coupling nonlinear equations using the Runge-Kutta fourth order method with adaptive step size control. The simulation results indicate that, when considering the role of the neighbor smaller bubble, the oscillation of the bigger bubble gradually exhibits a lag in comparison with the single-bubble case, and the extent of the lag becomes much more obvious as time goes by. This phenomenon is more easily observed with the increase of the initial radius of the smaller bubble. In comparison with the single-bubble case, the oscillation of the bigger bubble is enhanced by the neighbor smaller bubble. Especially, the pressure pulse of the bigger bubble rises intensely when the sizes of two bubbles approach, and a series of peak values for different initial radii are acquired when the initial radius ratio of two bubbles is in the range of 0.9˜1.0. Although the increase of the center distance between two bubbles can weaken the mutual interaction, it has no significant influence on the enhancement trend. On the one hand, the interaction between two bubbles with different radii can suppress the growth of the smaller bubble; on the other hand, it also can enhance the growth of the bigger one at the same time. The significant enhancement effect due to the interaction of multi-bubbles should be paid more attention because it can be used to reinforce the cavitation intensity for various potential applications in future.

Coupled hydrodynamic-structural analysis of an integral flowing sodium test loop in the TREAT reactor

International Nuclear Information System (INIS)

Zeuch, W.R.; A-Moneim, M.T.

1979-01-01

A hydrodynamic-structural response analysis of the Mark-IICB loop was performed for the TREAT (Transient Reactor Test Facility) test AX-1. Test AX-1 is intended to provide information concerning the potential for a vapor explosion in an advanced-fueled LMFBR. The test will be conducted in TREAT with unirradiated uranium-carbide fuel pins in the Mark-IICB integral flowing sodium loop. Our analysis addressed the ability of the experimental hardware to maintain its containment integrity during the reference accident postulated for the test. Based on a thermal-hydraulics analysis and assumptions for fuel-coolant interaction in the test section, a pressure pulse of 144 MPa maximum pressure and pulse width of 1.32 ms has been calculated as the reference accident. The response of the test loop to the pressure transient was obtained with the ICEPEL and STRAW codes. Modelling of the test section was completed with STRAW and the remainder of the loop was modelled by ICEPEL

Heat transfer between immiscible liquids enhanced by gas bubbling

International Nuclear Information System (INIS)

Greene, G.A.; Schwarz, C.E.; Klages, J.; Klein, J.

1982-08-01

The phenomena of core-concrete interactions impact upon containment integrity of light water reactors (LWR) following postulated complete meltdown of the core by containment pressurization, production of combustible gases, and basemat penetration. Experiments have been performed with non-reactor materials to investigate one aspect of this problem, heat transfer between overlying immiscible liquids whose interface is disturbed by a transverse non-condensable gas flux emanating from below. Hydrodynamic studies have been performed to test a criterion for onset of entrainment due to bubbling through the interface and subsequent heat transfer studies were performed to assess the effect of bubbling on interfacial heat transfer rates, both with and without bubble induced entrainment. Non-entraining interfacial heat transfer data with mercury-water/oil fluid pairs were observed to be bounded from below within a factor of two to three by the Szekeley surface renewal heat transfer model. However heat transfer data for fluid pairs which are found to entrain (water-oil), believed to be characteristic of molten reactor core-concrete conditions, were measured to be up to two orders of magnitude greater than surface renewal predictions and are calculated by a simple entrainment heat transfer model

LMFBR type reactor

International Nuclear Information System (INIS)

Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

1988-01-01

Purpose: To flatten the power distribution while maintaining the flattening in the axial power distribution in LMFBR type reactors. Constitution: Main system control rods are divided into control rods used for the operation and starting rods used for the starting of the reactor, and the starting rods are disposed in the radial periphery of the reactor core, while the control rods are disposed to the inside of the starting rods. With such a constitution, adjusting rods can be disposed in the region where the radial power peaking is generated to facilitate the flattening of the power distribution even in such a design that the ratio of the number of control rods to that of fuel assemblies is relatively large. That is, in this reactor, the radial power peaking is reduced by about 10% as compared with the conventional reactor core. As a result, the maximum linear power density during operation is reduced by about 10% to increase the thermal margin of the reactor core. If the maximum linear power density is set identical, the number of the fuel assemblies can be decreased by about 10%, to thereby reduce the fuel production cost. (K.M.)

Power to Fuels: Dynamic Modeling of a Slurry Bubble Column Reactor in Lab-Scale for Fischer Tropsch Synthesis under Variable Load of Synthesis Gas

Directory of Open Access Journals (Sweden)

Siavash Seyednejadian

2018-03-01

Full Text Available This research developed a comprehensive computer model for a lab-scale Slurry Bubble Column Reactor (SBCR (0.1 m Dt and 2.5 m height for Fischer–Tropsch (FT synthesis under flexible operation of synthesis gas load flow rates. The variable loads of synthesis gas are set at 3.5, 5, 7.5 m3/h based on laboratory adjustments at three different operating temperatures (483, 493 and 503 K. A set of Partial Differential Equations (PDEs in the form of mass transfer and chemical reaction are successfully coupled to predict the behavior of all the FT components in two phases (gas and liquid over the reactor bed. In the gas phase, a single-bubble-class-diameter (SBCD is adopted and the reduction of superficial gas velocity through the reactor length is incorporated into the model by the overall mass balance. Anderson Schulz Flory distribution is employed for reaction kinetics. The modeling results are in good agreement with experimental data. The results of dynamic modeling show that the steady state condition is attained within 10 min from start-up. Furthermore, they show that step-wise syngas flow rate does not have a detrimental influence on FT product selectivity and the dynamic modeling of the slurry reactor responds quite well to the load change conditions.

Advanced Intermediate Heat Transport Loop Design Configurations for Hydrogen Production Using High Temperature Nuclear Reactors

International Nuclear Information System (INIS)

Chang Oh; Cliff Davis; Rober Barner; Paul Pickard

2005-01-01

The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood. A number of possible configurations for a system that transfers heat between the nuclear reactor and the hydrogen and/or electrical generation plants were identified. These configurations included both direct and indirect cycles for the production of electricity. Both helium and liquid salts were considered as the working fluid in the intermediate heat transport loop. Methods were developed to perform thermal-hydraulic evaluations and cycle-efficiency evaluations of the different configurations and coolants. The thermal-hydraulic evaluations estimated the sizes of various components in the intermediate heat transport loop for the different configurations. The relative sizes of components provide a relative indication of the capital cost associated with the various configurations. Estimates of the overall cycle efficiency of the various

Hydrodynamic effects of air sparging on hollow fiber membranes in a bubble column reactor.

Science.gov (United States)

Xia, Lijun; Law, Adrian Wing-Keung; Fane, Anthony G

2013-07-01

Air sparging is now a standard approach to reduce concentration polarization and fouling of membrane modules in membrane bioreactors (MBRs). The hydrodynamic shear stresses, bubble-induced turbulence and cross flows scour the membrane surfaces and help reduce the deposit of foulants onto the membrane surface. However, the detailed quantitative knowledge on the effect of air sparging remains lacking in the literature due to the complex hydrodynamics generated by the gas-liquid flows. To date, there is no valid model that describes the relationship between the membrane fouling performance and the flow hydrodynamics. The present study aims to examine the impact of hydrodynamics induced by air sparging on the membrane fouling mitigation in a quantitative manner. A modelled hollow fiber module was placed in a cylindrical bubble column reactor at different axial heights with the trans-membrane pressure (TMP) monitored under constant flux conditions. The configuration of bubble column without the membrane module immersed was identical to that studied by Gan et al. (2011) using Phase Doppler Anemometry (PDA), to ensure a good quantitative understanding of turbulent flow conditions along the column height. The experimental results showed that the meandering flow regime which exhibits high flow instability at the 0.3 m is more beneficial to fouling alleviation compared with the steady flow circulation regime at the 0.6 m. The filtration tests also confirmed the existence of an optimal superficial air velocity beyond which a further increase is of no significant benefit on the membrane fouling reduction. In addition, the alternate aeration provided by two air stones mounted at the opposite end of the diameter of the bubble column was also studied to investigate the associated flow dynamics and its influence on the membrane filtration performance. It was found that with a proper switching interval and membrane module orientation, the membrane fouling can be effectively

Fluid dynamics of bubbly flows

International Nuclear Information System (INIS)

Ziegenhein, Thomas

2016-01-01

Bubbly flows can be found in many applications in chemical, biological and power engineering. Reliable simulation tools of such flows that allow the design of new processes and optimization of existing one are therefore highly desirable. CFD-simulations applying the multi-fluid approach are very promising to provide such a design tool for complete facilities. In the multi-fluid approach, however, closure models have to be formulated to model the interaction between the continuous and dispersed phase. Due to the complex nature of bubbly flows, different phenomena have to be taken into account and for every phenomenon different closure models exist. Therefore, reliable predictions of unknown bubbly flows are not yet possible with the multi-fluid approach. A strategy to overcome this problem is to define a baseline model in which the closure models including the model constants are fixed so that the limitations of the modeling can be evaluated by validating it on different experiments. Afterwards, the shortcomings are identified so that the baseline model can be stepwise improved without losing the validity for the already validated cases. This development of a baseline model is done in the present work by validating the baseline model developed at the Helmholtz-Zentrum Dresden-Rossendorf mainly basing on experimental data for bubbly pipe flows to bubble columns, bubble plumes and air-lift reactors that are relevant in chemical and biological engineering applications. In the present work, a large variety of such setups is used for validation. The buoyancy driven bubbly flows showed thereby a transient behavior on the scale of the facility. Since such large scales are characterized by the geometry of the facility, turbulence models cannot describe them. Therefore, the transient simulation of bubbly flows with two equation models based on the unsteady Reynolds-averaged Navier-Stokes equations is investigated. In combination with the before mentioned baseline model these

Fluid dynamics of bubbly flows

Energy Technology Data Exchange (ETDEWEB)

Ziegenhein, Thomas

2016-07-08

Bubbly flows can be found in many applications in chemical, biological and power engineering. Reliable simulation tools of such flows that allow the design of new processes and optimization of existing one are therefore highly desirable. CFD-simulations applying the multi-fluid approach are very promising to provide such a design tool for complete facilities. In the multi-fluid approach, however, closure models have to be formulated to model the interaction between the continuous and dispersed phase. Due to the complex nature of bubbly flows, different phenomena have to be taken into account and for every phenomenon different closure models exist. Therefore, reliable predictions of unknown bubbly flows are not yet possible with the multi-fluid approach. A strategy to overcome this problem is to define a baseline model in which the closure models including the model constants are fixed so that the limitations of the modeling can be evaluated by validating it on different experiments. Afterwards, the shortcomings are identified so that the baseline model can be stepwise improved without losing the validity for the already validated cases. This development of a baseline model is done in the present work by validating the baseline model developed at the Helmholtz-Zentrum Dresden-Rossendorf mainly basing on experimental data for bubbly pipe flows to bubble columns, bubble plumes and air-lift reactors that are relevant in chemical and biological engineering applications. In the present work, a large variety of such setups is used for validation. The buoyancy driven bubbly flows showed thereby a transient behavior on the scale of the facility. Since such large scales are characterized by the geometry of the facility, turbulence models cannot describe them. Therefore, the transient simulation of bubbly flows with two equation models based on the unsteady Reynolds-averaged Navier-Stokes equations is investigated. In combination with the before mentioned baseline model these

The growth of intra-granular bubbles in post-irradiation annealed UO2 fuel

International Nuclear Information System (INIS)

White, R.J.

2001-01-01

Post-irradiation examinations of low temperature irradiated UO 2 reveal large numbers of very small intra-granular bubbles, typically of around 1 nm diameter. During high temperature reactor transients these bubbles act as sinks for fission gas atoms and vacancies and can give rise to large volumetric swellings, sometimes of the order of 10%. Under irradiation conditions, the nucleation and growth of these bubbles is determined by a balance between irradiation-induced nucleation, diffusional growth and an irradiation induced re-solution mechanism. This conceptual picture is, however, incomplete because in the absence of irradiation the model predicts that the bubble population present from the pre-irradiation would act as the dominant sink for fission gas atoms resulting in large intra-granular swellings and little or no fission gas release. In practice, large fission gas releases are observed from post-irradiation annealed fuel. A recent series of experiments addressed the issue of fission gas release and swelling in post-irradiation annealed UO 2 originating from Advanced Gas Cooled Reactor (AGR) fuel which had been ramp tested in the Halden Test reactor. Specimens of fuel were subjected to transient heating at ramp rates of 0.5 deg. C/s and 20 deg. C/s to target temperatures between 1600 deg. C and 1900 deg. C. The release of fission gas was monitored during the tests. Subsequently, the fuel was subjected to post-irradiation examination involving detailed Scanning Electron Microscopy (SEM) analysis. Bubble-size distributions were obtained from seventeen specimens, which entailed the measurement of nearly 26,000 intra-granular bubbles. The analysis reveals that the bubble densities remain approximately invariant during the anneals and the bubble-size distributions exhibit long exponential tails in which the largest bubbles are present in concentrations of 10 4 or 10 5 lower than the concentrations of the average sized bubbles. Detailed modelling of the bubble

Safety problems of nuclear power plants with channel-type graphite boiling water reactors

International Nuclear Information System (INIS)

Emel'yanov, I.Ya.; Vasilevskij, V.P.; Volkov, V.P.; Gavrilov, P.A.; Kramerov, A.Ya.; Kuznetsov, S.P.; Kunegin, E.P.; Rybakov, N.Z.

1977-01-01

Construction of nuclear power plants in a highly populated region near large industrial centres necessitates to pay a special attention to their nuclear and radiation safety. Safety problems of nuclear reactor operation are discussed, in particular, they are: reliable stoppage of fission chain reaction at any emergency cases; reliable core cooling with failure of various equipment; emergency core cooling with breached pipes of a circulating circuit; and prevention of radioactive coolant release outside the nuclear power plant in amount exceeding the values adopted. Channel-type water boiling reactors incorporate specific features requiring a new approach to safety operation of a reactor and a nuclear power plant. These include primarily a rather large steam volume in the coolant circuit, large amount of accumulated heat, void reactivity coefficient. Channel-type reactors characterized by fair neutron balance and flexible fuel cycle, have a series of advantages alleviating the problem of ensuring their safety. The possibility of reliable control over the state of each channel allows to replace failed fuel elements by the new ones, when operating on-load, to increase the number of circulating loops and reduce the diameter of main pipelines, simplifies significantly the problem of channel emergency cooling and localization of a radioactive coolant release from a breached circuit. The concept of channel-type reactors is based on the solution of three main problems. First, plant safety should be assured in emergency switch off of separate units and, if possible, energy conditions should be maintained, this is of particular importance considering the increase in unit power. Second, the system of safety and emergency cooling should eliminate a great many failures of fuel elements in case of potential breaches of any tube in the circulating circuit. Finally, rugged boxes and localizing devices should be provided to exclude damage of structural elements of the nuclear power

Bubbly flows around a two-dimensional circular cylinder

Science.gov (United States)

Lee, Jubeom; Park, Hyungmin

2016-11-01

Two-phase cross flows around a bluff body occur in many thermal-fluid systems like steam generators, heat exchangers and nuclear reactors. However, our current knowledge on the interactions among bubbles, bubble-induced flows and the bluff body are limited. In the present study, the gas-liquid bubbly flows around a solid circular cylinder are experimentally investigated while varying the mean void fraction from 5 to 27%. The surrounding liquid (water) is initially static and the liquid flow is only induced by the air bubbles. For the measurements, we use the high-speed two-phase particle image velocimetry techniques. First, depending on the mean void fraction, two regimes are classified with different preferential concentration of bubbles in the cylinder wake, which are explained in terms of hydrodynamic force balances acting on rising bubbles. Second, the differences between the two-phase and single-phase flows (while matching their Reynolds numbers) around a circular cylinder will be discussed in relation to effects of bubble dynamics and the bubble-induced turbulence on the cylinder wake. Supported by a Grant (MPSS-CG-2016-02) through the Disaster and Safety Management Institute funded by Ministry of Public Safety and Security of Korean government.

Metallic sodium as a coolant of high speed nuclear reactors, (2)

International Nuclear Information System (INIS)

Atsumo, Hideo

1975-01-01

Tables are given on all the sodium loops in Japan and most of the sodium loops all over the world. Name and purpose of the loops, time of establishment, highest temperature, amount of sodium, flow rate, the materials used for the construction of the loops, and the diameter of the main pipings are given. Also, the problems related with these loops are discussed. For example, the high temperature sodium facility at HEDL-WADCO was made for the FFTF component test and instrument test, and uses 50,000 gallons of metallic sodium. The highest temperature is 590 0 C. The sodium flows at the rate of 60 g/m. The body is made of Type 304 stainless steel. Main data of existing sodium-cooled reactors in the world are also tabulated. The data include thermal output, electric output, the structure of the reactor cores, the dimensions of the cores, fuel used, the highest temperature in the reactors, the temperature of sodium at the inlet and outlet, the rate of multiplication, the amount of sodium used, number of control rods, number of heat exchangers, and the pressure of steam. The Monju type nuclear reactor in Japan uses 1,800 ton of sodium. (Fukutomi, T.)

Research activity on thermohydraulic problems of PWR reactors

International Nuclear Information System (INIS)

Szabados, L.

1976-06-01

The general review of the experimental and theoretical research works on thermohydraulic investigation of pressurized water type reactors being done in the Central Research Institute for Physics is given. The main results of the theoretical and theoretical-numerical research are summarized. The most important result of the past years is the construction of the High Pressure Water Cooled Loop (NVH) thermohydraulic loop. Another significant achievement was the development of the reactor thermohydraulic program system. (Sz.N.Z.)

The thermal performance of a loop-type heat pipe for passively removing residual heat from spent fuel pool

Energy Technology Data Exchange (ETDEWEB)

Xiong, Zhenqin [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Gu, Hanyang, E-mail: guhanyang@stu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Wang, Minglu [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Cheng, Ye [Shanghai Nuclear Engineering Research and Design Institute, Shanghai 200233 (China)

2014-12-15

Highlights: • Feasibility of applying loop-type heat pipes for SFP is studied. • The heat transfer rate of the heat pipes was tested. • The heat transfer coefficient was between 200 and 490 W/m{sup 2}/s. • The effect of the water temperature is dominant. • Three kinds of the filling ratio 27%, 21% and 14% are compared. - Abstract: Heat pipe is an efficient heat transfer device without electrically driven parts. Therefore large-scale loop type heat pipe systems have potential uses for passively removing heat from spent fuel pools and reactor cores under the accidental conditions to improve the safety of the nuclear power station. However, temperature difference between the hot water in the spent fuel pool and the ambient air which is the heat sink is small, in the range of 20–60 °C. To understand and predict the heat removal capacity of such a large scale loop type heat pipe in the situation similar to the accidental condition of the spent fuel pool (SFP) for the design purpose, a loop-type heat pipe with a very high and large evaporator has been fabricated and was tested using ammonia as the working fluid. The evaporator with inner diameter of 65 mm and length of 7.6 m is immersed in a hot water tube which simulate the spent fuel pool. The condenser of the loop-type heat pipe is cooled by the air. The tests were performed with the velocity of the hot water in the tube in the range of 0.7–2.1 × 10{sup −2} m/s, the hot water inlet temperature between 50 and 90 °C and the air velocity ranging from 0.5 m/s to 2.5 m/s. Three kinds of the ammonia volumetric filling ratio in the heat pipe were tested, i.e. 27%, 21% and 14%. It is found that the heat transfer rate was in the range of 1.5–14.9 kW, and the heat transfer coefficient of evaporator was between 200 and 490 W/m{sup 2}/s. It is feasible to use the large scale loop type heat pipe to passively remove the residual heat from SFP. Furthermore, the effect of air velocity, air temperature, water flow

The thermal performance of a loop-type heat pipe for passively removing residual heat from spent fuel pool

International Nuclear Information System (INIS)

Xiong, Zhenqin; Gu, Hanyang; Wang, Minglu; Cheng, Ye

2014-01-01

Highlights: • Feasibility of applying loop-type heat pipes for SFP is studied. • The heat transfer rate of the heat pipes was tested. • The heat transfer coefficient was between 200 and 490 W/m 2 /s. • The effect of the water temperature is dominant. • Three kinds of the filling ratio 27%, 21% and 14% are compared. - Abstract: Heat pipe is an efficient heat transfer device without electrically driven parts. Therefore large-scale loop type heat pipe systems have potential uses for passively removing heat from spent fuel pools and reactor cores under the accidental conditions to improve the safety of the nuclear power station. However, temperature difference between the hot water in the spent fuel pool and the ambient air which is the heat sink is small, in the range of 20–60 °C. To understand and predict the heat removal capacity of such a large scale loop type heat pipe in the situation similar to the accidental condition of the spent fuel pool (SFP) for the design purpose, a loop-type heat pipe with a very high and large evaporator has been fabricated and was tested using ammonia as the working fluid. The evaporator with inner diameter of 65 mm and length of 7.6 m is immersed in a hot water tube which simulate the spent fuel pool. The condenser of the loop-type heat pipe is cooled by the air. The tests were performed with the velocity of the hot water in the tube in the range of 0.7–2.1 × 10 −2 m/s, the hot water inlet temperature between 50 and 90 °C and the air velocity ranging from 0.5 m/s to 2.5 m/s. Three kinds of the ammonia volumetric filling ratio in the heat pipe were tested, i.e. 27%, 21% and 14%. It is found that the heat transfer rate was in the range of 1.5–14.9 kW, and the heat transfer coefficient of evaporator was between 200 and 490 W/m 2 /s. It is feasible to use the large scale loop type heat pipe to passively remove the residual heat from SFP. Furthermore, the effect of air velocity, air temperature, water flow rate and

FBR type reactor core

International Nuclear Information System (INIS)

Tamiya, Tadashi; Kawashima, Katsuyuki; Fujimura, Koji; Murakami, Tomoko.

1995-01-01

Neutron reflectors are disposed at the periphery of a reactor core fuel region and a blanket region, and a neutron shielding region is disposed at the periphery of them. The neutron reflector has a hollow duct structure having a sealed upper portion, a lower portion opened to cooling water, in which a gas and coolants separately sealed in the inside thereof. A driving pressure of a primary recycling pump is lowered upon reduction of coolant flow rate, then the liquid level of coolants in the neutron reflector is lowered due to imbalance between the driving pressure and a gas pressure, so that coolants having an effect as a reflector are eliminated from the outer circumference of the reactor core. Therefore, the amount of neutrons leaking from the reactor core is increased, and negative reactivity is charged to the reactor core. The negative reactivity of the neutron reflector is made greater than a power compensation reactivity. Since this enables reactor scram by using an inherent performance of the reactor core, the reactor core safety of an LMFBR-type reactor can be improved. (I.N.)

Simplified Models for Analysis and Design of the Control System Main Loops of CAREM Reactor

International Nuclear Information System (INIS)

Etchepareborda, Andres; Flury, Celso

2000-01-01

The target of this work is to show a few models developed for control analysis and design of the reactor CAREM's main control loops within a broad range of power (between 40 % and 100%).By one side, it is shown the main features of a analytic model programed in MATLAB.This model is based on fitting steady state points at different power levels of the CAREM's RETRAN model.By the other side, it is shown linear models of black-box type denoting the perturbed behavior of the system for each level power point.These models are identified from temporal responses of CAREM's RETRAN model to perturbed input signals over the different steady power level points.Then the dynamics of these models are verified contrasting the temporal responses of the RETRAN model versus the responses of the MATLAB model and the identified models, in each steady power level point.Also are contrasting the frequency response of the linearization of MATLAB model versus the frequency response of the identified models, in each steady power level point.Either the MATLAB model as the identified models are good enough for the control analysis and design of the three main control loops.The MATLAB model has a few differences against the RETRAN model in the primary pressure output variable, that it must be taken into account in the design of this control loop if this model is used.The aim of these models is to represent in a satisfactory way the dynamics of the plant for a later control analysis and design of the control loops in a frequency range between 0.01 rad/seg and 0.3 rad/seg, and a power range between 40 % and 100 %

Status of the French research in the field of molten salt nuclear reactors

International Nuclear Information System (INIS)

Hery, M.; Israel, M.; Fauger, P.; Lecocq, A.

1977-01-01

The research program of the CEA in the field of molten salt nuclear reactors has been concerned with MSBR type reactors (Molten Salt Breeder Reactor). The papers written after having performed the theoretical analysis are entitled: core, circuits, chemistry and economy; they include some criticisms and suggestions. The experimental studies consisted in: graphite studies, chemical studies of the salt, metallic materials, the salt loop and the lead loop [fr

Full instantaneous traversal rupture of the primary loop pipeline

International Nuclear Information System (INIS)

Baytelesov, S.A.; Kungurov, F.R.

2010-01-01

Accident, reflecting full immediate cross rupture of primary loop pipe of WWR-SM research reactor of INP AS RUz is observed in this paper. Calculations for accident situation and analysis for different reactor cores, formed from fully IRT-3M type high enriched fuel (36% enrichment on 235 U), first mixed core, compiled from 16 IRT-3M fuel assemblies and 4 IRT-4M type fuel assemblies with low enriched fuel (19,7% enrichment on 235 U) and the core fully formed from low enriched fuel are carried out

Pulse*Star Inertial Confinement Fusion Reactor: heat transfer loop and balance of plant considerations

International Nuclear Information System (INIS)

McDowell, M.W.; Murray, K.A.

1984-01-01

A conceptual heat transfer loop and balance of plant design for the Pulse*Star Inertial Confinement Fusion Reactor has been investigated and results are presented. The Pulse*Star reaction vessel, a perforated steel bell jar approximately 11 m in diameter, is immersed in Li 17 Pb 83 coolant which flows through the perforations and forms a 1.5 m thick plenum of droplets around an 8 m diameter inner chamber. The reactor and associated pumps, piping, and steam generators are contained within a 17 m diameter pool of Li 17 Pb 83 coolant to minimize structural requirements and occupied space, resulting in reduced cost. Four parallel heat transfer loops with flow rates of 5.5 m 3 /s each are necessary to transfer 3300 MWt of power. The steam generator design was optimized by finding the most cost-effective combination of heat exchanger area and pumping power. Power balance calculations based on an improved electrical conversion efficiency revealed a net electrical output of 1260 MWe to the bus bar and a resulting net efficiency of 39%. Suggested balance-of-plant layouts are also presented

Cluster Dynamics Modeling with Bubble Nucleation, Growth and Coalescence

Energy Technology Data Exchange (ETDEWEB)

de Almeida, Valmor F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Blondel, Sophie [Univ. of Tennessee, Knoxville, TN (United States); Bernholdt, David E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wirth, Brian D. [Univ. of Tennessee, Knoxville, TN (United States)

2017-06-01

The topic of this communication pertains to defect formation in irradiated solids such as plasma-facing tungsten submitted to helium implantation in fusion reactor com- ponents, and nuclear fuel (metal and oxides) submitted to volatile ssion product generation in nuclear reactors. The purpose of this progress report is to describe ef- forts towards addressing the prediction of long-time evolution of defects via continuum cluster dynamics simulation. The di culties are twofold. First, realistic, long-time dynamics in reactor conditions leads to a non-dilute di usion regime which is not accommodated by the prevailing dilute, stressless cluster dynamics theory. Second, long-time dynamics calls for a large set of species (ideally an in nite set) to capture all possible emerging defects, and this represents a computational bottleneck. Extensions beyond the dilute limit is a signi cant undertaking since no model has been advanced to extend cluster dynamics to non-dilute, deformable conditions. Here our proposed approach to model the non-dilute limit is to monitor the appearance of a spatially localized void volume fraction in the solid matrix with a bell shape pro le and insert an explicit geometrical bubble onto the support of the bell function. The newly cre- ated internal moving boundary provides the means to account for the interfacial ux of mobile species into the bubble, and the growth of bubbles allows for coalescence phenomena which captures highly non-dilute interactions. We present a preliminary interfacial kinematic model with associated interfacial di usion transport to follow the evolution of the bubble in any number of spatial dimensions and any number of bubbles, which can be further extended to include a deformation theory. Finally we comment on a computational front-tracking method to be used in conjunction with conventional cluster dynamics simulations in the non-dilute model proposed.

Strategies of development of reactor types

International Nuclear Information System (INIS)

Bacher, P.

2004-01-01

The development of nuclear energy in the coming decades will depend on the goals followed, on the available technologies and on the strategies implemented in the world in agreement with public acceptation. This article is limited to the technical aspects of the strategies of development of reactor types: 1 - objectives; 2 - common constraints to all reactor types: safety and terrorism risks, wastes, non-proliferation, economics; 3 - different reactor types: general considerations, proven technologies (PWR, BWR, Candu), non-proven technologies but having an important experience, technologies at the design stage; 4 - energy systems and 'Generation IV forum': systems based on thermal neutron reactors and low enrichment, systems for the valorization of 238 U, systems for Pu burning, systems allowing the destruction of minor actinides, thorium-based systems, the Gen IV international forum; 5 - conclusion. (J.S.)

Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 2: Base Case and Sensitivity Analysis

KAUST Repository

Zhao, Zhenlong; Chen, Tianjiao; Ghoniem, Ahmed F.

2013-01-01

Part 1 (10.1021/ef3014103) of this series describes a new rotary reactor for gas-fueled chemical-looping combustion (CLC), in which, a solid wheel with microchannels rotates between the reducing and oxidizing streams. The oxygen carrier (OC) coated

Scalar one-loop integrals for QCD

International Nuclear Information System (INIS)

Ellis, R. Keith; Zanderighi, Giulia

2008-01-01

We construct a basis set of infra-red and/or collinearly divergent scalar one-loop integrals and give analytic formulas, for tadpole, bubble, triangle and box integrals, regulating the divergences (ultra-violet, infra-red or collinear) by regularization in D = 4-2ε dimensions. For scalar triangle integrals we give results for our basis set containing 6 divergent integrals. For scalar box integrals we give results for our basis set containing 16 divergent integrals. We provide analytic results for the 5 divergent box integrals in the basis set which are missing in the literature. Building on the work of van Oldenborgh, a general, publicly available code has been constructed, which calculates both finite and divergent one-loop integrals. The code returns the coefficients of 1/ε 2 ,1/ε 1 and 1/ε 0 as complex numbers for an arbitrary tadpole, bubble, triangle or box integral

Closed loop auto control system software for Miniature Neutron Source Reactors (MNSRs)

International Nuclear Information System (INIS)

Iqbal, Masood; Qadir, Javed

2009-01-01

A closed loop auto control system has been developed and implemented at the Pakistan Research Reactor-2 (PARR-2). For interfacing of the site signals with a computer, a data acquisition card was employed. The system utilizes nine signals and, on their basis, carries out continuous on-line analysis, maintaining safe operation. On-line reactivity measurement is included in this software. Also, it generates control signals for keeping the reactor at the desired power level in auto-mode round-the-clock without human intervention. In case of abnormal conditions for either power or temperature or radiation level, alarms are initiated and if their levels reach beyond prescribed safe operation regime, automatic shutdown sequence is initiated. The control system has been thoroughly tested for various postulated scenarios. The test results have been in good agreement with the expected response. (orig.)

Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

Energy Technology Data Exchange (ETDEWEB)

Rest, J. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)], E-mail: jrest@anl.gov; Hofman, G.L.; Kim, Yeon Soo [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

2009-04-15

An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than {approx}7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

Science.gov (United States)

Rest, J.; Hofman, G. L.; Kim, Yeon Soo

2009-04-01

An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than ˜7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

Numerical investigation of the coolant mixing during fast deboration transients for VVER-440 type reactors

International Nuclear Information System (INIS)

Hoehne, T.; Rhode, U.

2000-01-01

The VVER-440 (440 MW) V-230 was considered for analyzing the flow field and mixing processes. The V-230 has no elliptical sieve plate in the lower plenum. Previously, the 3D flow distribution in the downcomer and the lower plenum of the VVER-440 reactor have been calculated by means of the CFD code CFX-4 for operational conditions. The CFX-calculations were compared with the experimental data and the analytical mixing model. In this paper, CFD calculations for the start-up of the first main coolant pump in a VVER-440 type reactor are reported about. This scenario is important in case that there is a plug of lower borated water in one of the primary coolant loops. (orig.)

Water feeding method upon reactor isolation

International Nuclear Information System (INIS)

Sasaki, Koichi; Takahara, Kuniaki; Hamamura, Kenji; Arakawa, Masahiro.

1990-01-01

The present invention concerns a method of feeding water upon reactor isolation in a plural loop type reactor having a plurality of reactor cooling systems. Water can be injected to a plurality of pools even if the pressure between the pools is not balanced and the water level in the reactor cooling system is optimally controlled. That is, water can be injected in accordance with the amount required for each of the pools by setting the opening of a turbine inlet steam control valve to somewhat higher than the cooling system pressure of the highest pressure loop. Water feeding devices upon reactor isolation were required by the same number as that for the reactor cooling systems. Whereas since pumps and turbines are used in common without worsening the water injection controllability to each of the loops according to the method of this invention and, accordingly, the cost performance can be improved. Further, since the opening degree of the turbine inlet steam control valve is controlled while making the difference pressure constant between the turbine inlet pressure and the pump exhaust pressure, the amount of the turbine exhausted steams can be reduced and, further, water injection controllability of the flow rate control valve in the injection line is improved. (I.S.)

Carbon Capture via Chemical-Looping Combustion and Reforming

Energy Technology Data Exchange (ETDEWEB)

Johansson, Marcus; Mattisson, Tobias; Ryden, Magnus; Lyngfelt, Anders

2006-10-15

Chemical-looping combustion (CLC) is a combustion technology with inherent separation of the greenhouse gas CO{sub 2}. The technique involves the use of a metal oxide as an oxygen carrier which transfers oxygen from combustion air to the fuel, and hence a direct contact between air and fuel is avoided. Two inter-connected fluidized beds, a fuel reactor and an air reactor, are used in the process. In the fuel reactor, the metal oxide is reduced by the reaction with the fuel and in the air reactor; the reduced metal oxide is oxidized with air. The outlet gas from the fuel reactor consists of CO{sub 2} and H{sub 2}O, and almost pure stream of CO{sub 2} is obtained when water is condensed. Considerable research has been conducted on CLC in the last decade with respect to oxygen carrier development, reactor design, system efficiencies and prototype testing. The technique has been demonstrated successfully with both natural gas and syngas as fuel in continuous prototype reactors based on interconnected fluidized beds within the size range 0.3 - 50 kW, using different types of oxygen carriers based on the metals Ni, Co, Fe, Cu and Mn. From these tests it can be established that almost complete conversion of the fuel can be obtained and 100% CO{sub 2} capture is possible. Further, two different types of chemical-looping reforming (CLR) have been presented in recent years. CLR is a technology to produce hydrogen with inherent CO{sub 2} capture. This paper presents an overview of the research performed on CLC and CLR highlights the current status of the technology.

p-Nitrophenol degradation by Fenton's oxidation in a bubble column reactor.

Science.gov (United States)

Rodrigues, Carmen S D; Borges, Ricardo A C; Lima, Vanessa N; Madeira, Luis M

2018-01-15

This paper reports on a study of the oxidation of p-nitrophenol (PNP) in a bubble column reactor (BCR). The use of the air stream aimed to provide perfect mixing in the liquid phase, which was successfully achieved and checked experimentally; there were no concentration gradients along the column, even at the lowest air flow rate used (Q = 1 mL/min at room temperature and atmospheric pressure). The effect of the operating variables was assessed, and a total reduction of PNP was reached, as well as mineralization of 49.2%, oxidant consumption of 90.3%, and with an efficiency of use - η H2O2 - of 0.09 mg C/mg H 2 O 2 , under the best operating conditions found - Q = 1 mL/min, [H 2 O 2 ] = 1.6 g/L, [Fe 2+ ] = 80 mg/L, pH = 3.0 and T = 22-24 °C - (after 120 min of reaction). Following this, various strategies were developed for improving the mineralization rate; it was found that the addition of H 2 O 2 every 5 min and readjusting the pH after 30 min of reaction allow the attainment of a much higher TOC removal (75.1%) and efficiency of oxidant use (η H2O2  = 0.17 mg C/mg H 2 O 2 ) with less oxidant. A reaction mechanism was proposed, based on intermediates identified that include p-nitrocatechol - PNC, p-benzoquinone - PB, hydroquinone - HQ - and carboxylic acids (oxalic, maleic and fumaric). Since the performance achieved in the BCR was good, and very similar to that obtained in a conventional batch reactor, it was possible to verify the efficacy of carrying out the Fenton process in this reactor configuration, which in our future work will focus on the treatability of industrial effluents. Copyright © 2017 Elsevier Ltd. All rights reserved.

Two-phase flow patterns recognition and parameters estimation through natural circulation test loop image analysis

Energy Technology Data Exchange (ETDEWEB)

Mesquita, R.N.; Libardi, R.M.P.; Masotti, P.H.F.; Sabundjian, G.; Andrade, D.A.; Umbehaun, P.E.; Torres, W.M.; Conti, T.N.; Macedo, L.A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Nuclear Engineering Center], e-mail: rnavarro@ipen.br

2009-07-01

Visualization of natural circulation test loop cycles is used to study two-phase flow patterns associated with phase transients and static instabilities of flow. Experimental studies on natural circulation flow were originally related to accidents and transient simulations relative to nuclear reactor systems with light water refrigeration. In this regime, fluid circulation is mainly caused by a driving force ('thermal head') which arises from density differences due to temperature gradient. Natural circulation phenomenon has been important to provide residual heat removal in cases of 'loss of pump power' or plant shutdown in nuclear power plant accidents. The new generation of compact nuclear reactors includes natural circulation of their refrigerant fluid as a security mechanism in their projects. Two-phase flow patterns have been studied for many decades, and the related instabilities have been object of special attention recently. Experimental facility is an all glass-made cylindrical tubes loop which contains about twelve demineralized water liters, a heat source by an electrical resistor immersion heater controlled by a Variac, and a helicoidal heat exchanger working as cold source. Data is obtained through thermo-pairs distributed over the loop and CCD cameras. Artificial intelligence based algorithms are used to improve (bubble) border detection and patterns recognition, in order to estimate and characterize, phase transitions patterns and correlate them with the periodic static instability (chugging) cycle observed in this circuit. Most of initial results show good agreement with previous numerical studies in this same facility. (author)

Two-phase flow patterns recognition and parameters estimation through natural circulation test loop image analysis

International Nuclear Information System (INIS)

Mesquita, R.N.; Libardi, R.M.P.; Masotti, P.H.F.; Sabundjian, G.; Andrade, D.A.; Umbehaun, P.E.; Torres, W.M.; Conti, T.N.; Macedo, L.A.

2009-01-01

Visualization of natural circulation test loop cycles is used to study two-phase flow patterns associated with phase transients and static instabilities of flow. Experimental studies on natural circulation flow were originally related to accidents and transient simulations relative to nuclear reactor systems with light water refrigeration. In this regime, fluid circulation is mainly caused by a driving force ('thermal head') which arises from density differences due to temperature gradient. Natural circulation phenomenon has been important to provide residual heat removal in cases of 'loss of pump power' or plant shutdown in nuclear power plant accidents. The new generation of compact nuclear reactors includes natural circulation of their refrigerant fluid as a security mechanism in their projects. Two-phase flow patterns have been studied for many decades, and the related instabilities have been object of special attention recently. Experimental facility is an all glass-made cylindrical tubes loop which contains about twelve demineralized water liters, a heat source by an electrical resistor immersion heater controlled by a Variac, and a helicoidal heat exchanger working as cold source. Data is obtained through thermo-pairs distributed over the loop and CCD cameras. Artificial intelligence based algorithms are used to improve (bubble) border detection and patterns recognition, in order to estimate and characterize, phase transitions patterns and correlate them with the periodic static instability (chugging) cycle observed in this circuit. Most of initial results show good agreement with previous numerical studies in this same facility. (author)

Heat removing device for reactor container

International Nuclear Information System (INIS)

Hisamochi, Kohei; Matsumoto, Tomoyuki; Matsumoto, Masayoshi; Sato, Ken-ichi.

1996-01-01

A recycling loop for reactor water is disposed in a reactor pressure vessel of a BWR type reactor. Extracted reactor water from the recycling loop passes through a extracted reactor water pipeline and flows into a reactor coolant cleanup system. A pipeline for connecting the extracted reactor water pipeline and a suppression pool is disposed, and a discharged water pressurizing pump is disposed to the pipeline. Upon occurrence of emergency, discharged water from the suppression pool is pressurized by a discharged water pressurizing pump and sent to a reactor coolant cleanup system. The discharged water is cooled while passing through a sucking water cooling portion of a regenerative heat exchanger and a non-regenerative heat exchanger. Then, it is sent to a feed water pipeline passing a bypass line of a filtering desalter and a bypass line of the sucked water cooling portion of the regenerative heat exchanger, injected to the inside of the pressure vessel to cool the reactor core and remove after-heat. Then, it cools the inside of the reactor container together with coolants flown out of the pressure vessel and then returns to the suppression pool. (I.N.)

United Kingdom and USSR reactor types

International Nuclear Information System (INIS)

Lewins, Jeffery

1988-01-01

The features of the RBMK reactor operated at Chernobyl are compared with reactor types pertinent to the UK. The UK reactors covered are in three classes: the commercial reactors now built and operated or in commission (Magnox and Advanced Gas-cooled Reactor (AGR)); the prototype Steam Generating Heavy Water Reactor (SGHWR) and Prototype Fast Reactor (PFR) that have comparable performance to commercial reactors; and the proposed Pressurised Water Reactor (PWR) or Sizewell 'B' design which, it will be recollected, is different in detail from PWRs built elsewhere. We do not include research and test reactors nor the Royal Navy PWRs. The appendices explain resonances, Doppler and Xenon effects, the reactor physics of Chernobyl and positive void coefficients all of which are relevant to the comparisons. (author)

Application of a bistable convection loop to LMFBR [liquid metal fast breeder reactor] emergency core cooling

International Nuclear Information System (INIS)

Anand, G.; Christensen, R.N.

1990-01-01

The concept of passive safety features for nuclear reactors has been developed in recent years and has gained wide acceptance. A literature survey of current reactors with passive features indicates that these reactors have some passive features but still do not fully meet the design objectives. Consider a current liquid-metal reactor design like PRISM. During normal operation, liquid sodium enters the reactor at ∼395 degree C and exits at ∼550 degree C. In the event of loss of secondary cooling with or without scram, the primary coolant (liquid sodium) initially acts as a heat sink and its temperature increases. For events without scram, the negative reactivity induced by the increase in temperature shuts the reactor down. When the average temperature of the sodium reaches ∼600 to 650 degree C, it overflows from the reactor vessel, activating the auxiliary cooling system. The auxiliary cooling system uses natural circulation of air around the reactor guard vessel. An alternative to the current design incorporates a bistable convection loop (BCL). The incorporation of the BCL concept remarkably improves the safety of the nuclear reactors. Application of the BCL concept to liquid-metal fast breeder reactors is described in this paper

Design study on sodium-cooled middle-scale modular reactor

International Nuclear Information System (INIS)

Shimakawa, Yoshio; Nibe, Nobuaki; Hori, Toru

2002-05-01

In Phase 1 of the 'Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)', an advanced loop type reactor has been selected as a promising concept of sodium-cooled middle-scale modular reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase 2 of the F/S, it is planed to precede a preliminary conceptual design of a sodium-cooled middle-scale modular reactor based on the design of the advanced loop type reactor. Through the design study, it is intended to construct such a plant concept that can show its attraction and competitiveness as a commercialized reactor. This report summarizes the results of the design study on the sodium-cooled middle-scale modular reactor performed in JFY2001, which is the first year of Phase 2. As the construction cost of the sodium-cooled middle-scale modular reactor, which has been constructed in Phase 1, was about 10% higher than that of the sodium-cooled large-scale reactor, a new concept of the middle-scale modular reactor, which is expected to be equal to the large-scale reactor from a viewpoint of economic competitiveness, has been re-constructed based on the design of the advanced loop type reactor. After that, fundamental specifications of main systems and components for the new concept have been set, and critical subjects related to safety, structural integrity, thermal hydraulics, operability, maintainability and economy have been examined and evaluated. As a result of this study, the plant concept of the sodium-cooled middle-scale modular reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000 yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowing down candidate concepts at the end of Phase 2. (author)

Universe out of a breathing bubble

International Nuclear Information System (INIS)

Guendelman, Eduardo I.; Sakai, Nobuyuki

2008-01-01

We consider the model of a false-vacuum bubble with a thin wall where the surface energy density is composed of two different components, 'domain-wall' type and 'dust' type, with opposite signs. We find stably oscillating solutions, which we call 'breathing bubbles'. By decay to a lower mass state, such a breathing bubble could become either (i) a child universe or ii) a bubble that 'eats up' the original universe, depending on the sign of the surface energy of the domain-wall component. We also discuss the effect of the finite-thickness corrections to the thin-wall approximation and possible origins of the energy contents of our model

Thermal-hydraulic analyses for in-pile SCWR fuel qualification test loops and SCWR material loop

Energy Technology Data Exchange (ETDEWEB)

Vojacek, A.; Mazzini, G.; Zmitkova, J.; Ruzickova, M. [Research Centre Rez (Czech Republic)

2014-07-01

One of the R&D directions of Research Centre Rez is dedicated to the supercritical water-cooled reactor concept (SCWR). Among the developed experimental facilities and infrastructure in the framework of the SUSEN project (SUStainable ENergy) is construction and experimental operation of the supercritical water loop SCWL focusing on material tests. At the first phase, this SCWL loop is assembled and operated out-of-pile in the dedicated loop facilities hall. At this out-of-pile operation various operational conditions are tested and verified. After that, in the second phase, the SCWL loop will be situated in-pile, in the core of the research reactor LVR-15, operated at CVR. Furthermore, it is planned to carry out a test of a small scale fuel assembly within the SuperCritical Water Reactor Fuel Qualification Test (SCWR-FQT) loop, which is now being designed. This paper presents the results of the thermal-hydraulic analyses of SCWL loop out-of-pile operation using the RELAP5/MOD3.3. The thermal-hydraulic modeling and the performed analyses are focused on the SCWL loop model validation through a comparison of the calculation results with the experimental results obtained at various operation conditions. Further, the present paper focuses on the transient analyses for start-up and shut-down of the FQT loop, particularly to explore the ability of system codes ATHLET 3.0A to simulate the transient between subcritical conditions and supercritical conditions. (author)

Advanced nuclear reactor types and technologies

International Nuclear Information System (INIS)

Ignatiev, V.; Devell, L.

1995-01-01

The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary

Advanced nuclear reactor types and technologies

Energy Technology Data Exchange (ETDEWEB)

Ignatiev, V [ed.; Feinberg, O; Morozov, A [Russian Research Centre ` Kurchatov Institute` , Moscow (Russian Federation); Devell, L [Studsvik Eco and Safety AB, Nykoeping (Sweden)

1995-07-01

The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary.

Pressure tube type reactors

International Nuclear Information System (INIS)

Komada, Masaoki.

1981-01-01

Purpose: To increase the safety of pressure tube type reactors by providing an additional ECCS system to an ordinary ECCS system and injecting heavy water in the reactor core tank into pressure tubes upon fractures of the tubes. Constitution: Upon fractures of pressure tubes, reduction of the pressure in the fractured tubes to the atmospheric pressure in confirmed and the electromagnetic valve is operated to completely isolate the pressure tubes from the fractured portion. Then, the heavy water in the reactor core tank flows into and spontaneously recycles through the pressure tubes to cool the fuels in the tube to prevent their meltdown. By additionally providing the separate ECCS system to the ordinary ECCS system, fuels can be cooled upon loss of coolant accidents to improve the safety of the reactors. (Moriyama, K.)

Slowing down bubbles with sound

Science.gov (United States)

Poulain, Cedric; Dangla, Remie; Guinard, Marion

2009-11-01

We present experimental evidence that a bubble moving in a fluid in which a well-chosen acoustic noise is superimposed can be significantly slowed down even for moderate acoustic pressure. Through mean velocity measurements, we show that a condition for this effect to occur is for the acoustic noise spectrum to match or overlap the bubble's fundamental resonant mode. We render the bubble's oscillations and translational movements using high speed video. We show that radial oscillations (Rayleigh-Plesset type) have no effect on the mean velocity, while above a critical pressure, a parametric type instability (Faraday waves) is triggered and gives rise to nonlinear surface oscillations. We evidence that these surface waves are subharmonic and responsible for the bubble's drag increase. When the acoustic intensity is increased, Faraday modes interact and the strongly nonlinear oscillations behave randomly, leading to a random behavior of the bubble's trajectory and consequently to a higher slow down. Our observations may suggest new strategies for bubbly flow control, or two-phase microfluidic devices. It might also be applicable to other elastic objects, such as globules, cells or vesicles, for medical applications such as elasticity-based sorting.

Separated type nuclear superheating reactor

International Nuclear Information System (INIS)

Hida, Kazuki.

1993-01-01

In a separated type nuclear superheating reactor, fuel assemblies used in a reactor core comprise fuel rods made of nuclear fuel materials and moderator rods made of solid moderating materials such as hydrogenated zirconium. Since the moderating rods are fixed or made detachable, high energy neutrons generated from the fuel rods are moderated by the moderating rods to promote fission reaction of the fuel rods. Saturated steams supplied from the BWR type reactor by the fission energy are converted to high temperature superheated steams while passing through a steam channel disposed between the fuel rods and the moderating rods and supplied to a turbine. Since water is not used but solid moderating materials sealed in a cladding tube are used as moderation materials, isolation between superheated steams and water as moderators is not necessary. Further, since leakage of heat is reduced to improve a heat efficiency, the structure of the reactor core is simplified and fuel exchange is facilitated. (N.H.)

An open loop equilibrator for continuous monitoring of radon at the groundwater-surface water interface

International Nuclear Information System (INIS)

Kil Yong Lee; Yoon Yeol Yoon; Soo Young Cho; Eunhee Lee; Sang-Ho Moon; Dong-Chan Koh; Kyoochul Ha; Yongcheol Kim; Kyung-Seok Ko

2015-01-01

A continuous monitoring system (CMS) using an open loop equilibrator for assessment of 222 Rn at the groundwater-surface water interface was developed and tested. For the characterization and validation of the system, three air loops (open loop, closed loop, and open bubble loop) were tested in relation to high and precise count rates, rapid response, and equilibration of radon. The water and air stream is fed to the equilibrator by an experimental setup with a commercial submersible water pump and the internal pump with built-in radon-in-air detector. Efficiency calibration of the CMS is done by simultaneous determination of a groundwater sample using liquid scintillation counting, and the RAD7 accessories RAD-H 2 O, BigBottle RAD-H 2 O. The higher count rates are provided by the closed loop. However, the open loop with bubbler (open bubble loop) provides the best precision count rates, rapid response, and equilibration time. The CMS allows radon determination in discrete water samples as well as continuous water streams. (author)

Chemical-looping combustion - status of development

Energy Technology Data Exchange (ETDEWEB)

Lyngfelt, Anders; Johansson, Marcus; Mattisson, Tobias

2008-05-15

Chemical-looping combustion (CLC) is a combustion technology with inherent separation of the greenhouse gas CO{sub 2}. The technique involves the use of a metal oxide as an oxygen carrier which transfers oxygen from combustion air to the fuel, and hence a direct contact between air and fuel is avoided. Two inter-connected fluidized beds, a fuel reactor and an air reactor, are used in the process. In the fuel reactor, the metal oxide is reduced by the reaction with the fuel and in the air reactor; the reduced metal oxide is oxidized with air. The outlet gas from the fuel reactor consists of CO{sub 2} and H{sub 2}O, and almost pure stream of CO{sub 2} is obtained when water is condensed. Considerable research has been conducted on CLC in the last years with respect to oxygen carrier development, reactor design, system efficiencies and prototype testing. In 2002 the process was a paper concept, albeit with some important but limited laboratory work on oxygen carrier particles. Today more than 600 materials have been tested and the technique has been successfully demonstrated in chemical-looping combustors in the size range 0.3 - 50 kW, using different types of oxygen carriers based on the metals Ni, Co, Fe, Cu and Mn. The total time of operational experience is more than a thousand hours. From these tests it can be established that almost complete conversion of the fuel can be obtained and 100% CO{sub 2} capture is possible. Most work so far has been focused on gaseous fuels, but the direct application to solid fuels is also being studied. Moreover, the same principle of oxygen transfer is used in chemical-looping reforming (CLR), which involves technologies to produce hydrogen with inherent CO{sub 2} capture. This paper presents an overview of the research performed on CLC and CLR highlights the current status of the technology

Effect of Spray System on Fission Product Distribution in Containment During a Severe Accident in a Two-Loop Pressurized Water Reactor

Directory of Open Access Journals (Sweden)

Mehdi Dehjourian

2016-08-01

Full Text Available The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

Activity build-up on the circulation loops of boiling water reactors: Basics for modelling of transport and deposition processes

International Nuclear Information System (INIS)

Covelli, B.; Alder, H.P.

1988-03-01

In the past 20 years the radiation field of nuclear power plant loops outside the core zone was the object of investigations in many countries. In this context test loops were built and basic research done. At our Institute PSI the installation of a LWR-contamination loop is planned for this year. This experimental loop has the purpose to investigate the complex phenomena of activity deposition from the primary fluid of reactor plants and to formulate analytical models. From the literature the following conclusions can be drawn: The principal correlations of the activity build-up outside the core are known. The plant specific single phenomena as corrosion, crud-transport, activation and deposit of cobalt in the oxide layer are complex and only partially understood. The operational experience of particular plants with low contaminated loops (BWR-recirculation loops) show that in principle the problem is manageable. The reduction of the activity build-up in older plants necessitates a combination of measures to modify the crud balance in the primary circuit. In parallel to the experimental work several simulation models in the form of computer programs were developed. These models have the common feature that they are based on mass balances, in which the exchange of materials and the sedimentation processes are described by global empirical transport coefficients. These models yield satisfactory results and allow parameter studies; the application however is restricted to the particular installation. All programs lack models that describe the thermodynamic and hydrodynamic mechanisms on the surface of deposition layers. Analytical investigations on fouling of process equipment led to models that are also applicable to the activity build-up in reactor loops. Therefore it seems appropriate to combine the nuclear simulation models with the fundamental equations for deposition. 10 refs., 18 figs., 3 tabs

In pile helium loop ''Comedie''

International Nuclear Information System (INIS)

Blanchard, R.J.

1985-01-01

The loop is located in the SILOE reactor at Centre d'Etudes Nucleaires de Grenoble. The purpose and objectives are divided into two groups, principal and secondary. The primary objective was to provide basic data on the deposition behavior of important condensable fission products on a variety of steel surfaces, i.e. temperature (sorption isotherms) and mass transfer (physical adsorption) dependencies; to provide information concerning the degree of penetration of important fission products into the metals comprising the heat exchanger-recuperator tubes as a function of alloy type and/or metal temperature; to provide complementary information on the reentrainment (liftoff) of important fission and activation products by performing out-of-pile blowdown experiments on tube samples representative of the alloy types used in the heat exchanger-recuperator and of the surface temperatures experienced during plateout. The secondary objective was to provide information concerning the migration of important fission products through graphite. To this end, concentration profiles in the web between the fuel rods containing the fission product source and the coolant channels and in the graphite diffusion sample will be measured to study the corrosion of metallic specimens placed in the conditions of high temperature gas cooled reactor. The first experiment SRO enables to determine the loop characteristics and possibilities related to thermal, thermodynamic, chemical and neutronic properties. The second experiment has been carried out in high temperature gas cooled reactor operating conditions. It enables to determine in particular the deposition axial profile of activation and fission products in the plateout section constituting the heat exchanger, the fission products balance trapped in the different filter components, and the cumulated released fraction of solid fission products. The SR1 test permits to demonstrate in particular the Comedie loop operation reliability, either

Hydrodynamics and mass transfer in slurry bubble columns : scale and pressure effects

NARCIS (Netherlands)

Chilekar, V.P.

2007-01-01

Slurry bubble columns (SBC) are widely used in the chemical industry as a multiphase reactor. Applications include oxidation and hydrogenation reactions, fermentation, Fischer-Tropsch synthesis, and waste water treatment. The advantages of a SBC over other multiphase reactors are the simple

Screening of external hazards for NPP with bank type reactor. Modeling of safety related systems and equipment for RBMK. Probabilistic assessment of NPP safety on aircraft impact. Progress report

International Nuclear Information System (INIS)

Kostarev, V.

1999-01-01

This progress report was produced within the frame of IAEA research project on screening the hazards for NPP with bank type reactor. It covers the following tasks; development of the model for the primary loop system of RBMK; developing the models for safety related equipment of RBMK; developing of models for safety related models of EGP-6 type reactor (Bilibinskaya Nuclear Co-generated heat and Power Plant); and probabilistic assessment of NPP safety on aircraft impact

Bubble dynamics and bubble-induced turbulence of a single-bubble chain

Science.gov (United States)

Lee, Joohyoung; Park, Hyungmin

2016-11-01

In the present study, the bubble dynamics and liquid-phase turbulence induced by a chain of bubbles injected from a single nozzle have been experimentally investigated. Using a high-speed two-phase particle image velociemtry, measurements on the bubbles and liquid-phase velocity field are conducted in a transparent tank filled with water, while varying the bubble release frequency from 0.1 to 35 Hz. The tested bubble size ranges between 2.0-3.2 mm, and the corresponding bubble Reynolds number is 590-1100, indicating that it belongs to the regime of path instability. As the release frequency increases, it is found that the global shape of bubble dispersion can be classified into two regimes: from asymmetric (regular) to axisymmetric (irregular). In particular, at higher frequency, the wake vortices of leading bubbles cause an irregular behaviour of the following bubble. For the liquid phase, it is found that a specific trend on the bubble-induced turbulence appears in a strong relation to the above bubble dynamics. Considering this, we try to provide a theoretical model to estimate the liquid-phase turbulence induced by a chain of bubbles. Supported by a Grant funded by Samsung Electronics, Korea.

Numerical modelling of inert gas bubble rising in liquid metal pool

International Nuclear Information System (INIS)

Pradeep, Arjun; Sharma, Anil Kumar; Ponraju, D.; Nashine, B K.

2016-01-01

Two-phase flow finds several applications in safe operation of Sodium-cooled Fast Reactor (SFR). Numerical modelling of bubble rise dynamics in liquid metal pool of SFR is essential for the evaluation of residence time and shape changes, which are of utmost importance for simulating associated heat and mass transfer processes involved in reactor safety. A numerical model has been developed based on OpenFOAM for the evaluation of two-dimensional inert gas bubble rise dynamics in stagnant liquid metal pool. The governing model equations are discretized and solved using the Volume of Fluid based solver available in OpenFOAM with appropriate initial and boundary conditions. The model has been validated with available numerical benchmark results for laminar transient two-phase flow. The model has been used to evaluate velocity and rise trajectory of argon gas bubble with different diameters through a pool of liquid sodium. (author)

Process/Equipment Co-Simulation on Syngas Chemical Looping Process

Energy Technology Data Exchange (ETDEWEB)

Zeng, Liang; Zhou, Qiang; Fan, Liang-Shih

2012-09-30

The chemical looping strategy for fossil energy applications promises to achieve an efficient energy conversion system for electricity, liquid fuels, hydrogen and/or chemicals generation, while economically separate CO{sub 2} by looping reaction design in the process. Chemical looping particle performance, looping reactor engineering, and process design and applications are the key drivers to the success of chemical looping process development. In order to better understand and further scale up the chemical looping process, issues such as cost, time, measurement, safety, and other uncertainties need to be examined. To address these uncertainties, advanced reaction/reactor modeling and process simulation are highly desired and the modeling efforts can accelerate the chemical looping technology development, reduce the pilot-scale facility design time and operating campaigns, as well as reduce the cost and technical risks. The purpose of this work is thus to conduct multiscale modeling and simulations on the key aspects of chemical looping technology, including particle reaction kinetics, reactor design and operation, and process synthesis and optimization.

Defluoridation of drinking water by electrocoagulation/electroflotation in a stirred tank reactor with a comparative performance to an external-loop airlift reactor

International Nuclear Information System (INIS)

Essadki, A.H.; Gourich, B.; Vial, Ch.; Delmas, H.; Bennajah, M.

2009-01-01

Defluoridation using batch electrocoagulation/electroflotation (EC/EF) was carried out in two reactors for comparison purpose: a stirred tank reactor (STR) close to a conventional EC cell and an external-loop airlift reactor (ELAR) that was recently described as an innovative reactor for EC. The respective influences of current density, initial concentration and initial pH on the efficiency of defluoridation were investigated. The same trends were observed in both reactors, but the efficiency was higher in the STR at the beginning of the electrolysis, whereas similar values were usually achieved after 15 min operation. The influence of the initial pH was explained using the analyses of sludge composition and residual soluble aluminum species in the effluents, and it was related to the prevailing mechanisms of defluoridation. Fluoride removal and sludge reduction were both favored by an initial pH around 4, but this value required an additional pre-treatment for pH adjustment. Finally, electric energy consumption was similar in both reactors when current density was lower than 12 mA/cm 2 , but mixing and complete flotation of the pollutants were achieved without additional mechanical power in the ELAR, using only the overall liquid recirculation induced by H 2 microbubbles generated by water electrolysis, which makes subsequent treatments easier to carry out.

Non-electric applications of pool-type nuclear reactors

International Nuclear Information System (INIS)

Adamov, E.O.; Cherkashov, Yu.M.; Romenkov, A.A.

1997-01-01

This paper recommends the use of pool-type light water reactors for thermal energy production. Safety and reliability of these reactors were already demonstrated to the public by the long-term operation of swimming pool research reactors. The paper presents the design experience of two projects: Apatity Underground Nuclear Heating Plant and Nuclear Sea-Water Desalination Plant. The simplicity of pool-type reactors, the ease of their manufacturing and maintenance make this type of a heat source attractive to the countries without a developed nuclear industry. (author). 6 figs, 1 tab

Artificial neural network for bubbles pattern recognition on the images

International Nuclear Information System (INIS)

Poletaev, I E; Pervunin, K S; Tokarev, M P

2016-01-01

Two-phase bubble flows have been used in many technological and energy processes as processing oil, chemical and nuclear reactors. This explains large interest to experimental and numerical studies of such flows last several decades. Exploiting of optical diagnostics for analysis of the bubble flows allows researchers obtaining of instantaneous velocity fields and gaseous phase distribution with the high spatial resolution non-intrusively. Behavior of light rays exhibits an intricate manner when they cross interphase boundaries of gaseous bubbles hence the identification of the bubbles images is a complicated problem. This work presents a method of bubbles images identification based on a modern technology of deep learning called convolutional neural networks (CNN). Neural networks are able to determine overlapping, blurred, and non-spherical bubble images. They can increase accuracy of the bubble image recognition, reduce the number of outliers, lower data processing time, and significantly decrease the number of settings for the identification in comparison with standard recognition methods developed before. In addition, usage of GPUs speeds up the learning process of CNN owning to the modern adaptive subgradient optimization techniques. (paper)

Scenario for commercialization of fast breeder reactors

International Nuclear Information System (INIS)

Kumaoka, Yoshio; Sato, Morihiko

1989-01-01

To realize the commercialization of fast breeder reactors (FBRs), it is essential to reduce construction costs to the same level as those for the current light water reactors. For this target to be attained, a highly important factor is to reduce to the lowest-levels possible the quantities of materials and volume of the buildings required for the primary and secondary sodium loops of the FBR. In this direction, an innovative compact FBR plant concept which holds promise for commercialization has been developed by introducing the pooltype reactor concept with the shortest possible secondary sodium loops, realized by coupling electromagnetic pumps with the steam generators. In comparison with the French Super Phenix reactor, for example, the construction of this 1,300-MWe FBR plant could be achieved with half the material quantities and plant volume required by the former type. (author)

Bubble and boundary layer behaviour in subcooled flow boiling

Energy Technology Data Exchange (ETDEWEB)

Maurus, Reinhold; Sattelmayer, Thomas [Lehrstuhl fuer Thermodynamik, Technische Universitaet Muenchen, 85747 Garching (Germany)

2006-03-15

Subcooled flow boiling is a commonly applied technique for achieving efficient heat transfer. In the study, an experimental investigation in the nucleate boiling regime was performed for water circulating in a closed loop at atmospheric pressure. The horizontal orientated test-section consists of a rectangular channel with a one side heated copper strip and good optical access. Various optical observation techniques were applied to study the bubble behaviour and the characteristics of the fluid phase. The bubble behaviour was recorded by the high-speed cinematography and by a digital high resolution camera. Automated image processing and analysis algorithms developed by the authors were applied for a wide range of mass flow rates and heat fluxes in order to extract characteristic length and time scales of the bubbly layer during the boiling process. Using this methodology, the bubbles were automatically analysed and the bubble size, bubble lifetime, waiting time between two cycles were evaluated. Due to the huge number of observed bubbles a statistical analysis was performed and distribution functions were derived. Using a two-dimensional cross-correlation algorithm, the averaged axial phase boundary velocity profile could be extracted. In addition, the fluid phase velocity profile was characterised by means of the particle image velocimetry (PIV) for the single phase flow as well as under subcooled flow boiling conditions. The results indicate that the bubbles increase the flow resistance. The impact on the flow exceeds by far the bubbly region and it depends on the magnitude of the boiling activity. Finally, the ratio of the averaged phase boundary velocity and of the averaged fluid velocity was evaluated for the bubbly region. (authors)

Bifurcation scenarios for bubbling transition.

Science.gov (United States)

Zimin, Aleksey V; Hunt, Brian R; Ott, Edward

2003-01-01

Dynamical systems with chaos on an invariant submanifold can exhibit a type of behavior called bubbling, whereby a small random or fixed perturbation to the system induces intermittent bursting. The bifurcation to bubbling occurs when a periodic orbit embedded in the chaotic attractor in the invariant manifold becomes unstable to perturbations transverse to the invariant manifold. Generically the periodic orbit can become transversely unstable through a pitchfork, transcritical, period-doubling, or Hopf bifurcation. In this paper a unified treatment of the four types of bubbling bifurcation is presented. Conditions are obtained determining whether the transition to bubbling is soft or hard; that is, whether the maximum burst amplitude varies continuously or discontinuously with variation of the parameter through its critical value. For soft bubbling transitions, the scaling of the maximum burst amplitude with the parameter is derived. For both hard and soft transitions the scaling of the average interburst time with the bifurcation parameter is deduced. Both random (noise) and fixed (mismatch) perturbations are considered. Results of numerical experiments testing our theoretical predictions are presented.

FBR type reactors

International Nuclear Information System (INIS)

Nakamura, Tsugio.

1986-01-01

Purpose: To ensure the thermal integrity of a reactor vessel in FBR type reactors by preventing sodium vapors or the likes from intruding into a shielding chamber and avoiding spontaneous convection thereof. Constitution: There are provided a shielding plug for shielding the upper opening of a reactor container, an annular thermal member disposed to the circumferential side in the container, a shielding member for shielding upper end of the shielding chamber and a plurality of convection preventive plates suspended from the thermal member into the shielding chamber, and the shielding chamber is communicated by way of the relatively low temperature portion of the container with a gas communication pipe. That is, by closing the upper end of the shielding chamber with the shielding member, coolant vapors, etc. can be prevented from intruding into the shielding chamber. Further, the convection preventive plates prevent the occurrence of spontaneous convection in the shielding chamber. Further, the gas communication pipe absorbs the expansion and contraction of gases in the shielding chamber to effectively prevent the deformation or the like for each of the structural materials. In this way, the thermal integrity of the reactor container can surely be maintained. (Horiuchi, T.)

RAPK-7. code for calculating mass transfer and corrosion products activation in the circulation loops of water-cooled reactors

International Nuclear Information System (INIS)

Mikhaylov, A.V.; Moryakov, A.V.; Nikitin, A.V.

2012-09-01

The RAPK-7 code was developed to simulate formation of non-irradiated and activated corrosion products, their transport and deposition on inner surfaces of primary components and in primary coolant of water-cooled reactors during their operation on power and after shutdown. The key feature of this code is its particular emphasis on the contamination of circulation loops by radioactive corrosion products of reactor which operates on variable modes. Such reactors typically are: research reactors and their experimental loops, naval nuclear power systems, etc. It's typical for such reactors to have repeated (over the campaign) and frequent variations in power (activating neutron fluxes), thermal-physical, hydrodynamic and other parameters of coolant, intensive water mass exchange between the circulation loop and the pressuriser, etc. The processes of mass-transfer are described by the RAPK-7 code with the use of models similar to those employed by the COTRAN and PACTOLE codes. The circulation circuit is broken down into computation areas. The user will then set the concentrations of water chemistry adjusting additives (alkali, boric acid, ammonia, hydrogen), as well as parameters in each area, such as wall temperature, coolant flow core temperature, pressure, flow rate, velocity, the radial component of coolant flowrate and activating neutron flux density. All the above parameters can be set as time-dependent step functions (bar charts), with independent time steps for each of them. The number of computation areas, the number of time dependencies and the level of detail in their description are limited by computer capabilities only. A 'brake' mode with a single-step change of the required set of parameters is provided to allow for jump-type events, such as replacement of contaminated components with clean ones during core refueling or repairs, emergency injection of boric acid, water mass exchange between the circulation circuit and the pressuriser, etc

Numerical simulation of bubble motion about a grid spacer in a rod bundle

International Nuclear Information System (INIS)

Zhang, Zheng; Hosokawa, Shigeo; Hayashi, Kosuke; Tomiyama, Akio

2009-01-01

Numerical simulations based on a three-dimensional two-way bubble tracking method are carried out to predict bubble motions in a square duct with an obstacle and in a two-by-three rod bundle with a grid spacer. Comparisons between measured and predicted bubble motions demonstrate that the two-way bubble tracking method gives good predictions for trajectories of small bubbles in the upstream side of the grid spacer in the rod bundle geometry. The predicted bubble trajectories clearly show that bubbles are apt to migrate toward the rod surface in the vicinity of the bottom of the grid spacer. Analysis of forces acting on the bubbles confirms that pressure gradient force induced by the presence of the spacer is the main cause of the bubble lateral migration toward the rod surface. Motions of steam bubbles at a nominal operating condition of a pressurized water reactor (PWR) are also predicted by using the bubble tracking method, which indicates that steam bubbles also migrate toward the rod surface at the upstream side of the spacer due to the spacer-induced pressure gradient force. (author)

The analysis for inventory of experimental reactor high temperature gas reactor type

International Nuclear Information System (INIS)

Sri Kuntjoro; Pande Made Udiyani

2016-01-01

Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it

Radioactive waste management at WWER type reactors

International Nuclear Information System (INIS)

1993-05-01

This report was prepared within the framework of the Technical Assistance Regional Project on Advice on Waste Management at WWER Type Reactors, which was initiated by the IAEA in 1991. The Regional Project is an integral part of the IAEA's activities directed towards improvement of the safety and reliability of nuclear power plants with WWER type reactors (Soviet designed PWRs). Forty-five WWER type units are currently in operation and twenty-five are under construction in Bulgaria, Czechoslovakia, Finland, Hungary and the former USSR. The idea of regional collaboration between eastern European countries under the auspices of the IAEA was discussed for the first time during the last meeting of the Council for Mutual Economic Assistance (CMEA) on spent fuel and radioactive waste management, held in Rez, Czechoslovakia, in October 1990. Since then, the CMEA and some of its former Member States have ceased to exist. However, there are many reasons for eastern European countries to continue their regional collaboration at a higher level. The USSR, the designer and supplier of WWER type reactors in eastern European countries, participated in the first phase of the project. The majority of WWER type reactors are situated in States of the former USSR (Russia and Ukraine). The main results of the first phase of the Regional Project are: (i) Re-establishment of communication channels among eastern European countries operating WWER type reactors by incorporating the IAEA's technical assistance; (ii) Identification of common waste management problems (administrative and technical) requiring resolution; (iii) Familiarization with radioactive waste management systems at nuclear power plants with WWER type reactors - Paks (Hungary), Loviisa (Finland), Jaslovske Bohunice (Czechoslovakia) and Novovoronezh (Russian Federation). Tabs

Experimental Study of Thermal Crisis in Connection with Tokamak Reactor High Heat Flux Components

International Nuclear Information System (INIS)

Gallo, D.; Giardina, M.; Castiglia, F.; Celata, G.P.; Mariani, A.; Zummo, G.; Cumo, M.

2000-01-01

The results of an experimental research on high heat flux thermal crisis in forced convective subcooled water flow, under operative conditions of interest to the thermal-hydraulic design of TOKAMAK fusion reactors, are here reported. These experiments, carried out in the framework of a collaboration between the Nuclear Engineering Department of Palermo University and the National Institute of Thermal - Fluid Dynamics of the ENEA - Casaccia (Rome), were performed on the STAF (Scambio Termico Alti Flussi) water loop and consisted, essentially, in a high speed photographic study which enabled focusing several information on bubble characteristics and flow patterns taking place during the burnout phenomenology

Annihilation of interstitial-type dislocation loops in {alpha}-Fe during He irradiation

Energy Technology Data Exchange (ETDEWEB)

Xu, Q., E-mail: xu@rri.kyoto-u.ac.jp [Research Reactor Institute, Kyoto University, Osaka 590-0494 (Japan); Wang, Y.X. [Institute of Modern Physics, Fudan University, Shanghai 200433 (China); Katakabe, Y. [Research Institute for Applied Mechanics, Kyushu University, Fukuoka 816-8580 (Japan); Iwakiri, H. [Faculty of Education, University of the Ryukyus, Okinawa 903-0213 (Japan); Yoshida, N. [Research Institute for Applied Mechanics, Kyushu University, Fukuoka 816-8580 (Japan); Sato, K.; Yoshiie, T. [Research Reactor Institute, Kyoto University, Osaka 590-0494 (Japan)

2011-10-01

Interstitial-type dislocation loops were formed in Fe-9Cr alloys on irradiation with 1-MeV He ions at 673 K. However, with increasing irradiation dose, the dislocation loops shrunk. A molecular dynamics simulation was used to elucidate the mechanism of this unexpected phenomenon. The simulation shows that, although the binding energy of a self-interstitial atom to a dislocation loop is normally greater than that of a vacancy, the energy hierarchy is reversed when He atoms decorate the loop. This may indicates preferential absorption of vacancies, causing loop shrinkage at high doses, consistent with experimental observation.

Annihilation of interstitial-type dislocation loops in α-Fe during He irradiation

International Nuclear Information System (INIS)

Xu, Q.; Wang, Y.X.; Katakabe, Y.; Iwakiri, H.; Yoshida, N.; Sato, K.; Yoshiie, T.

2011-01-01

Interstitial-type dislocation loops were formed in Fe-9Cr alloys on irradiation with 1-MeV He ions at 673 K. However, with increasing irradiation dose, the dislocation loops shrunk. A molecular dynamics simulation was used to elucidate the mechanism of this unexpected phenomenon. The simulation shows that, although the binding energy of a self-interstitial atom to a dislocation loop is normally greater than that of a vacancy, the energy hierarchy is reversed when He atoms decorate the loop. This may indicates preferential absorption of vacancies, causing loop shrinkage at high doses, consistent with experimental observation.

Measuring the diameter of rising gas bubbles by means of the ultrasound transit time technique

Energy Technology Data Exchange (ETDEWEB)

Richter, T., E-mail: Thomas.Richter6@tu-dresden.de; Eckert, K., E-mail: Kerstin.Eckert@tu-dresden.de; Yang, X.; Odenbach, S.

2015-09-15

Highlights: • Ultrasound transit time technique (UTTT) is applied to the zig-zag raise of gas bubble. • Comparison of bubble diameter and tilt, measured by UTTT, with high-speed imaging. • Uncertainty in the determination of the bubble diameter by UTTT is less than 7%. • UTTT is able to measure dynamic changes in bubble size in opaque liquids and vessels. • UTTT can be applied to liquid metal loops. - Abstract: This study presents ultrasound transit time technique (UTTT) measurements of the diameter variations of single argon bubbles rising in a zig-zag trajectory in water. Simultaneous size measurements with a high-speed camera show that UTTT resolves both the apparent diameter and the tilt of the bubble axis with an accuracy of better than 7%. This qualifies UTTT for the measurement of bubble sizes in opaque liquids, such as liquid metals, or vessels.

Numerical model for the breakdown of a molecular F{sub 2} bubble in molten FLiBe

Energy Technology Data Exchange (ETDEWEB)

Seto, K., E-mail: kelvin.seto@uoit.ca [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada)

2014-07-01

A novel one-dimensional numerical model of the breakdown for a fluorine bubble due to break-up and chemical reactions with dissolved UF{sub 4} and PuF{sub 4} in the molten salt reactor (MSR) volatilization process was developed. The results show that an initial bubble diameter of 4.0 cm would be required in order for it to reach the top of the volatilization reactor. Chemical reactions were found to be the primary cause of bubble breakdown. Physical characteristics (density and surface tension) were found to have a negligible effect on the breakdown process, as determined by a sensitivity analysis comparing molten Li to FLiBe. (author)

Reactor scram device for FBR type reactor

International Nuclear Information System (INIS)

Kumasaka, Katsuyuki; Arashida, Genji; Itooka, Satoshi.

1991-01-01

In a control rod attaching structure in a reactor scram device of an FBR type reactor, an anti-rising mechanism proposed so far against external upward force upon occurrence of earthquakes relies on the engagement of a mechanical structure but temperature condition is not taken into consideration. Then, in the present invention, a material having curie temperature characteristics and which exhibits ferromagnetism only under low temperature condition and a magnet device are disposed to one of a movable control rod and a portion secured to the reactor. Alternatively, a bimetal member or a shape memory alloy which actuates to fix to the mating member only under low temperature condition is secured. The fixing device is adapted to operate so as to secure the control rods when the low temperature state is caused depending on the temperature condition. With such a constitution, when the control rods are separated from a driving device, they are prevented from rising even if they undergo external upward force due to earthquakes and so on, which can improve the reactor safety. (N.H.)

CFD analysis of bubble microlayer and growth in subcooled flow boiling

Energy Technology Data Exchange (ETDEWEB)

Owoeye, Eyitayo James, E-mail: msgenius10@ufl.edu; Schubring, DuWanye, E-mail: dlschubring@ufl.edu

2016-08-01

Highlights: • A new LES-microlayer model is introduced. • Analogous to the unresolved SGS in LES, analysis of bubble microlayer was performed. • The thickness of bubble microlayer was computed at both steady and transient states. • The macroscale two-phase behavior was captured with VOF coupled with AMR. • Numerical validations were performed for both the micro- and macro-region analyses. - Abstract: A numerical study of single bubble growth in turbulent subcooled flow boiling was carried out. The macro- and micro-regions of the bubble were analyzed by introducing a LES-microlayer model. Analogous to the unresolved sub-grid scale (SGS) in LES, a microlayer analysis was performed to capture the unresolved thermal scales for the micro-region heat transfer by deriving equations for the microlayer thickness at steady and transient states. The phase change at the macro-region was based on Volume-of-Fluid (VOF) interface tracking method coupled with adaptive mesh refinement (AMR). Large Eddy Simulation (LES) was used to model the turbulence characteristics. The numerical model was validated with multiple experimental data from the open literature. This study includes parametric variations that cover the operating conditions of boiling water reactor (BWR) and pressurized water reactor (PWR). The numerical model was used to study the microlayer thickness, growth rate, dynamics, and distortion of the bubble.

Three-loop massive tadpoles and polylogarithms through weight six

Energy Technology Data Exchange (ETDEWEB)

Kniehl, B.A.; Pikelner, Andrey F.; Veretin, Oleg L. [Hamburg Univ. (Germany). 2. Inst. fuer Theoretische Physik

2017-05-15

We evaluate the three-loop massive vacuum bubble diagrams in terms of polylogarithms up to weight six. We also construct the basis of irrational constants being harmonic polylogarithms of arguments e{sup kiπ/3}.

Design of slurry bubble column reactors: novel technique for optimum catalyst size selection contractual origin of the invention

Science.gov (United States)

Gamwo, Isaac K [Murrysville, PA; Gidaspow, Dimitri [Northbrook, IL; Jung, Jonghwun [Naperville, IL

2009-11-17

A method for determining optimum catalyst particle size for a gas-solid, liquid-solid, or gas-liquid-solid fluidized bed reactor such as a slurry bubble column reactor (SBCR) for converting synthesis gas into liquid fuels considers the complete granular temperature balance based on the kinetic theory of granular flow, the effect of a volumetric mass transfer coefficient between the liquid and the gas, and the water gas shift reaction. The granular temperature of the catalyst particles representing the kinetic energy of the catalyst particles is measured and the volumetric mass transfer coefficient between the gas and liquid phases is calculated using the granular temperature. Catalyst particle size is varied from 20 .mu.m to 120 .mu.m and a maximum mass transfer coefficient corresponding to optimum liquid hydrocarbon fuel production is determined. Optimum catalyst particle size for maximum methanol production in a SBCR was determined to be in the range of 60-70 .mu.m.

Fuel assembly for FBR type reactor and reactor core thereof

International Nuclear Information System (INIS)

Kobayashi, Kaoru.

1998-01-01

The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

Multi-loop PWR modeling and hardware-in-the-loop testing using ACSL

International Nuclear Information System (INIS)

Thomas, V.M.; Heibel, M.D.; Catullo, W.J.

1989-01-01

Westinghouse has developed an Advanced Digital Feedwater Control System (ADFCS) which is aimed at reducing feedwater related reactor trips through improved control performance for pressurized water reactor (PWR) power plants. To support control system setpoint studies and functional design efforts for the ADFCS, an ACSL based model of the nuclear steam supply system (NSSS) of a Westinghouse (PWR) was generated. Use of this plant model has been extended from system design to system testing through integration of the model into a Hardware-in-Loop test environment for the ADFCS. This integration includes appropriate interfacing between a Gould SEL 32/87 computer, upon which the plant model executes in real time, and the Westinghouse Distributed Processing family (WDPF) test hardware. A development program has been undertaken to expand the existing ACSL model to include capability to explicitly model multiple plant loops, steam generators, and corresponding feedwater systems. Furthermore, the program expands the ADFCS Hardware-in-Loop testing to include the multi-loop plant model. This paper provides an overview of the testing approach utilized for the ADFCS with focus on the role of Hardware-in-Loop testing. Background on the plant model, methodology and test environment is also provided. Finally, an overview is presented of the program to expand the model and associated Hardware-in-Loop test environment to handle multiple loops

Nanosecond Discharge in Bubbled Liquid n-Heptane: Effects of Gas Composition and Water Addition

KAUST Repository

Hamdan, Ahmad

2016-08-30

Recently, an aqueous discharge reactor was developed to facilitate reformation of liquid fuels by in-liquid plasma. To gain a microscopic understanding of the physical elements behind this aqueous reactor, we investigate nanosecond discharges in liquid n-heptane with single and double gaseous bubbles in the gap between electrodes. We introduce discharge probability (DP) to characterize the stochastic nature of the discharges, and we investigate the dependence of DP on the gap distance, applied voltage, gaseous bubble composition, and the water content in n-heptane/distilled-water emulsified mixtures. Propagation of a streamer through the bubbles indicates no discharges in the liquids. DP is controlled by the properties of the gaseous bubble rather than by the composition of the liquid mixture in the gap with a single bubble; meanwhile, DP is determined by the dielectric permittivity of the liquid mixture in the gap with double bubbles, results that are supported by static electric field simulations. We found that a physical mechanism of increasing DP is caused by an interaction between bubbles and an importance of the dielectric permittivity of a liquid mixture on the local enhancement of field intensity. We also discuss detailed physical characteristics, such as plasma lifetime and electron density within the discharge channel, by estimating from measured emissions with a gated-intensified charge-coupled device and by using spectroscopic images, respectively. © 1973-2012 IEEE.

Detail analysis of tritium permeation in the metal liquid channels of the regenerating sheaths of a fusion reactor in presence of helium bubbles

International Nuclear Information System (INIS)

Banet, L.; Mas de les Valls, E.; Sedano, L. A.

2012-01-01

Inside the channels of liquid metal of the fusion reactor regenerative wrappers, the possible existence of nucleated helium bubbles is not remote. Helium is formed joined the tritium in the escaped neutrons of plasma with lithium. The accumulation of helium in the contact surfaces, between the structure and ML, lead a reduction of heat transfer, at the same time a reduction in the permeation of tritium. The coexistence of three phases in touch: metal liquid, helium and structural material, makes the transport of heat and tritium in a complex phenomenon. To enrich tritium transport studies conducted in the past, there is now a detail analysis of the helium bubble environment adhered to the channel ML wall of a regenerative wrap. For the study we used a CFD tool development on free code OpenFOAM.

Fuel assembly for FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki.

1995-01-01

Ordinary sodium bond-type fuel pins using nitride fuels, carbide fuels or metal fuels and pins incorporated with hydride moderators are loaded in a wrapper tube at a ratio of from 2 to 10% based on the total number of fuel pins. The hydride moderators are sealed in the hydride moderator incorporated pins at the position only for a range from the upper end to a reactor core upper position of substantially 1/4 of the height of the reactor core from the upper end of the reactor core as a center. Then, even upon occurrence of ULOF (loss of flow rate scram failure phenomenon), it gives characteristic of reducing the power only by a doppler coefficient and not causing boiling of coolant sodium but providing stable cooling to the reactor core. Therefore, a way of thinking on the assurance of passive safety is simplified to make a verification including on the reactor structure unnecessary. In an LMFBR type reactor using the fuel assembly, a critical experiment for confirming accuracy of nuclear design is sufficient for the item required for study and development, which provides a great economical effect. (N.H.)

Method of generating energy by acoustically induced cavitation fusion and reactor therefor

International Nuclear Information System (INIS)

Flynn, H.G.

1982-01-01

Two different cavitation fusion reactors (cfr's) are disclosed. Each comprises a chamber containing a liquid (host) metal such as lithium or an alloy thereof. Acoustical horns in the chamber walls operate to vary the ambient pressure in the liquid metal, creating therein small bubbles which are caused to grow to maximum sizes and then collapse violently in two steps. In the first stage the bubble contents remain at the temperature of the host liquid, but in the second stage the increasing speed of collapse causes an adiabatic compression of the bubble contents, and of the thin shell of liquid surrounding the bubble. Application of a positive pressure on the bubble accelerates this adiabatic stage, and causes the bubble to contract to smaller radius, thus increasing maximum temperatures and pressures reached within the bubble. At or near its minimum radius the bubble generates a very intense shock wave, creating high pressures and temperatures in the host liquid. These extremely high pressures and temperatures occur both within the bubbles and in the host liquid, and cause hydrogen isotopes in the bubbles and liquid to undergo thermonuclear reactions. In one type of cfr the thermonuclear reaction is generated by cavitation within the liquid metal itself, and in the other type the reaction takes place primarily within the bubbles. The fusion reactions generate energy that is absorbed as heat by the liquid metal, and this heat is removed from the liquid by conduction through the acoustical horns to an external heat exchanger, without any pumping of the liquid metal

Safety design study of fast breeder reactors in Japan

International Nuclear Information System (INIS)

Miura, M.; Inagaki, T.

1992-01-01

This paper reports on two fast breeder reactor (FBR) concepts, the tank type and the loop type, that have been studied as possible reactor designs to be used for a demonstration FBR (DFBR). The basic principle fo the DFBR design is to ensure plant safety through a defense-in-depth methodology. Improvements in the seismic and thermal stress designs have been attempted for both reactor concepts. The system design study strives to maximize the reliability of the safety-related systems and to rationalize commercialization of the plant

Inertial manipulation of bubbles in rectangular microfluidic channels.

Science.gov (United States)

Hadikhani, Pooria; Hashemi, S Mohammad H; Balestra, Gioele; Zhu, Lailai; Modestino, Miguel A; Gallaire, François; Psaltis, Demetri

2018-03-27

Inertial microfluidics is an active field of research that deals with crossflow positioning of the suspended entities in microflows. Until now, the majority of the studies have focused on the behavior of rigid particles in order to provide guidelines for microfluidic applications such as sorting and filtering. Deformable entities such as bubbles and droplets are considered in fewer studies despite their importance in multiphase microflows. In this paper, we show that the trajectory of bubbles flowing in rectangular and square microchannels can be controlled by tuning the balance of forces acting on them. A T-junction geometry is employed to introduce bubbles into a microchannel and analyze their lateral equilibrium position in a range of Reynolds (1 < Re < 40) and capillary numbers (0.1 < Ca < 1). We find that the Reynolds number (Re), the capillary number (Ca), the diameter of the bubble (D[combining macron]), and the aspect ratio of the channel are the influential parameters in this phenomenon. For instance, at high Re, the flow pushes the bubble towards the wall while large Ca or D[combining macron] moves the bubble towards the center. Moreover, in the shallow channels, having aspect ratios higher than one, the bubble moves towards the narrower sidewalls. One important outcome of this study is that the equilibrium position of bubbles in rectangular channels is different from that of solid particles. The experimental observations are in good agreement with the performed numerical simulations and provide insights into the dynamics of bubbles in laminar flows which can be utilized in the design of flow based multiphase flow reactors.

Atomistic simulations of thermodynamic properties of Xe gas bubbles in U10Mo fuels

Energy Technology Data Exchange (ETDEWEB)

Hu, Shenyang, E-mail: shenyang.hu@pnnl.gov; Setyawan, Wahyu; Joshi, Vineet V.; Lavender, Curt A.

2017-07-15

Xe gas bubble superlattice formation is observed in irradiated uranium–10 wt% molybdenum (U10Mo) fuels. However, the thermodynamic properties of the bubbles (the relationship among bubble size, equilibrium Xe concentration, and bubble pressure) and the mechanisms of bubble superlattice formation are not well known. In this work, the molecular dynamics (MD) method is used to study these properties and mechanisms. The results provide important inputs for quantitative mesoscale models of gas bubble evolution and fuel performance. In the MD simulations, the embedded-atom method (EAM) potential of U10Mo-Xe [1] is employed. Initial gas bubbles with a low Xe concentration (underpressured) are generated in a body-centered cubic (bcc) U10Mo single crystal. Then Xe atoms are sequentially added into the bubbles one by one, and the evolution of pressure and dislocation emission around the bubbles is analyzed. The relationship between pressure, equilibrium Xe concentration, and radius of the bubbles is established. It was found that an overpressured gas bubble emits partial dislocations with a Burgers vector along the <111> direction and a slip plane of (11-2). Meanwhile, dislocation loop punch out was not observed. The overpressured bubble also induces an anisotropic stress field. A tensile stress was found along <110> directions around the bubble, favoring the nucleation and formation of a face-centered cubic bubble superlattice in bcc U10Mo fuels.

Atomistic simulations of thermodynamic properties of Xe gas bubbles in U10Mo fuels

Science.gov (United States)

Hu, Shenyang; Setyawan, Wahyu; Joshi, Vineet V.; Lavender, Curt A.

2017-07-01

Xe gas bubble superlattice formation is observed in irradiated uranium-10 wt% molybdenum (U10Mo) fuels. However, the thermodynamic properties of the bubbles (the relationship among bubble size, equilibrium Xe concentration, and bubble pressure) and the mechanisms of bubble superlattice formation are not well known. In this work, the molecular dynamics (MD) method is used to study these properties and mechanisms. The results provide important inputs for quantitative mesoscale models of gas bubble evolution and fuel performance. In the MD simulations, the embedded-atom method (EAM) potential of U10Mo-Xe [1] is employed. Initial gas bubbles with a low Xe concentration (underpressured) are generated in a body-centered cubic (bcc) U10Mo single crystal. Then Xe atoms are sequentially added into the bubbles one by one, and the evolution of pressure and dislocation emission around the bubbles is analyzed. The relationship between pressure, equilibrium Xe concentration, and radius of the bubbles is established. It was found that an overpressured gas bubble emits partial dislocations with a Burgers vector along the direction and a slip plane of (11-2). Meanwhile, dislocation loop punch out was not observed. The overpressured bubble also induces an anisotropic stress field. A tensile stress was found along directions around the bubble, favoring the nucleation and formation of a face-centered cubic bubble superlattice in bcc U10Mo fuels.

Analysis on small long life reactor using thorium fuel for water cooled and metal cooled reactor types

International Nuclear Information System (INIS)

Permana, Sidik

2009-01-01

Long-life reactor operation can be adopted for some special purposes which have been proposed by IAEA as the small and medium reactor (SMR) program. Thermal reactor and fast reactor types can be used for SMR and in addition to that program the utilization of thorium fuel as one of the candidate as a 'partner' fuel with uranium fuel which can be considered for optimizing the nuclear fuel utilization as well as recycling spent fuel. Fissile U-233 as the main fissile material for thorium fuel shows higher eta-value for wider energy range compared with other fissile materials of U-235 and Pu-239. However, it less than Pu-239 for fast energy region, but it still shows high eta-value. This eta-value gives the reactor has higher capability for obtaining breeding condition or high conversion capability. In the present study, the comparative analysis on small long life reactor fueled by thorium for different reactor types (water cooled and metal cooled reactor types). Light water and heavy water have been used as representative of water-cooled reactor types, and for liquid metal-cooled reactor types, sodium-cooled and lead-bismuth-cooled have been adopted. Core blanket arrangement as general design configuration, has been adopted which consist of inner blanket region fueled by thorium oxide, and two core regions (inner and out regions) fueled by fissile U-233 and thorium oxide with different percentages of fissile content. SRAC-CITATION and JENDL-33 have been used as core optimization analysis and nuclear data library for this analysis. Reactor operation time can reaches more than 10 years operation without refueling and shuffling for different reactor types and several power outputs. As can be expected, liquid metal cooled reactor types can be used more effective for obtaining long life reactor with higher burnup, higher power density, higher breeding capability and lower excess reactivity compared with water-cooled reactors. Water cooled obtains long life core operation

Defluoridation of drinking water by electrocoagulation/electroflotation in a stirred tank reactor with a comparative performance to an external-loop airlift reactor

Energy Technology Data Exchange (ETDEWEB)

Essadki, A.H., E-mail: essadki@hotmail.com [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Gourich, B. [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Vial, Ch. [Laboratoire de Genie Chimique et Biochimique, LGCB-UBP/ENSCCF, 24 avenue des Landais, BP 206, 63174 Aubiere Cedex (France); Delmas, H. [Laboratoire de Genie Chimique, ENSIACET-INPT, 5 rue Paulin Talabot, 31106 Toulouse (France); Bennajah, M. [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Laboratoire de Genie Chimique, ENSIACET-INPT, 5 rue Paulin Talabot, 31106 Toulouse (France)

2009-09-15

Defluoridation using batch electrocoagulation/electroflotation (EC/EF) was carried out in two reactors for comparison purpose: a stirred tank reactor (STR) close to a conventional EC cell and an external-loop airlift reactor (ELAR) that was recently described as an innovative reactor for EC. The respective influences of current density, initial concentration and initial pH on the efficiency of defluoridation were investigated. The same trends were observed in both reactors, but the efficiency was higher in the STR at the beginning of the electrolysis, whereas similar values were usually achieved after 15 min operation. The influence of the initial pH was explained using the analyses of sludge composition and residual soluble aluminum species in the effluents, and it was related to the prevailing mechanisms of defluoridation. Fluoride removal and sludge reduction were both favored by an initial pH around 4, but this value required an additional pre-treatment for pH adjustment. Finally, electric energy consumption was similar in both reactors when current density was lower than 12 mA/cm{sup 2}, but mixing and complete flotation of the pollutants were achieved without additional mechanical power in the ELAR, using only the overall liquid recirculation induced by H{sub 2} microbubbles generated by water electrolysis, which makes subsequent treatments easier to carry out.

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)

Bubble formation in irradiated Li2O

International Nuclear Information System (INIS)

Verrall, R.A.; Rose, D.H.; Miller, J.M.; Hastings, I.J.; MacDonald, D.S.

1991-01-01

Lithium oxide, irradiated to a burnup of 1 at% (total lithium) at temperatures between 400 and 850deg C with on-line tritium recovery and measurement, has been examined out-reactor. Residual tritium content ranged from 2.4 to 16 mCi/g, but, conservatively, averaged less than 10 mCi/g or 1 wppm. Scanning electron microscopy showed bubble formation in the ceramic which is thought to be due to helium formed from the in-reactor 6 Li(n, α) 3 H reaction. (orig.)

Design project of the dosimetry control system in the independent CO2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, Vol. V

International Nuclear Information System (INIS)

1964-01-01

Design project of the dosimetry control system in the independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels includes the following: calculations of CO 2 gas activity, design of the dosimetry control system, review of the changes that should be done in the RA reactor building for installing the independent CO 2 loop, specification of the materials with cost estimation, engineering drawings of the system [sr

PWR type reactor plant

International Nuclear Information System (INIS)

Matsuoka, Tsuyoshi.

1993-01-01

A water chamber of a horizontal U-shaped pipe type steam generator is partitioned to an upper high temperature water chamber portion and a lower low temperature water chamber portion. An exit nozzle of a reactor container containing a reactor core therein is connected to a suction port of a coolant pump by way of first high temperature pipelines. The exit port of the coolant pump is connected to the high temperature water chamber portion of the steam generator by way of second high temperature pipelines. The low temperature water chamber portion of the steam generator is connected to an inlet nozzle of the reactor container by way of the low temperature pipelines. The low temperature water chamber portion of the steam generator is positioned lower than the high temperature water chamber portion, but upper than the reactor core. Accordingly, all of the steam generator for a primary coolant system, coolant pumps as well as high temperature pipelines and low temperature pipelines connecting them are disposed above the reactor core. With such a constitution, there is no worry of interrupting core cooling even upon occurrence of an accident, to improve plant safety. (I.N.)

Numerical analysis of coolant mixing in the pressure vessel of WWER-440 type nuclear reactors

International Nuclear Information System (INIS)

Boros, I.; Aszodi, A.

2003-01-01

The precise description of the coolant mixing processes taking place in the reactor pressure vessel (RPV) of pressurized water nuclear reactors has an essential importance during power operation, as well as in case of incidental or accidental conditions. In this paper the detailed CFD model of the pressure vessel of a WWER-440 type reactor and calculations performed with this RPV model are presented. The CFD model of the pressure vessel contains all the important internal structural elements of the RPV. Sensitivity study on the effect of these elements was also carried out. Both steady-state and transient calculation were performed using the CFD code CFX-5.5.1. The results of the steady-state calculations give the so called mixing factors, i.e. the effect of each single primary loop at the core inlet. The mixing factors can be given for nominal circumstances (i.e. all main coolant pumps are working) or in case of less than six working MCPs. In order to validate the model the calculated mixing factors are compared with the values measured in the Paks NPP (Authors)

Design study of ship based nuclear power reactor

International Nuclear Information System (INIS)

Su'ud, Zaki; Fitriyani, Dian

2002-01-01

Preliminary design study of ship based nuclear power reactors has been performed. In this study the results of thermohydraulics analysis is presented especially related to behaviour of ship motion in the sea. The reactors are basically lead-bismuth cooled fast power reactors using nitride fuels to enhance neutronics and safety performance. Some design modification are performed for feasibility of operation under sea wave movement. The system use loop type with relatively large coolant pipe above reactor core. The reactors does not use IHX, so that the heat from primary coolant system directly transferred to water-steam loop through steam generator. The reactors are capable to be operated in difference power level during night and noon. The reactors however can also be used totally or partially to produce clean water through desalination of sea water. Due to the influence of sea wave movement the analysis have to be performed in three dimensional analysis. The computation time for this analysis is speeded up using Parallel Virtual Machine (PVM) Based multi processor system

Consequences of pipe ruptures in metal fueled, liquid metal cooled reactors

International Nuclear Information System (INIS)

Dunn, F.E.

1990-01-01

The capability to simulate pipe ruptures has recently been added to the SASSYS-1 LMR systems analysis code. Using this capability, the consequences of severe pipe ruptures in both loop-type and pool-type reactors using metal fuel were investigated. With metal fuel, if the control rods scram then either type of reactor can easily survive a complete double-ended break of a single pipe; although, as might be expected, the consequences are less severe for a pool-type reactor. A pool-type reactor can even survive a protected simultaneous breaking of all of its inlet pipes without boiling of the coolant or melting of the fuel or cladding. 2 refs., 16 figs., 1 tab

Contributions of Cu-rich clusters, dislocation loops and nanovoids to the irradiation-induced hardening of Cu-bearing low-Ni reactor pressure vessel steels

Energy Technology Data Exchange (ETDEWEB)

Bergner, F., E-mail: f.bergner@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstr. 400, 01328 Dresden (Germany); Gillemot, F. [Centre for Energy Research of the Hungarian Academy of Sciences, 29-33 Konkoly-Thege street, 1121 Budapest XII (Hungary); Hernández-Mayoral, M.; Serrano, M. [Division of Materials, CIEMAT, Avenida Complutense 22, 28040 Madrid (Spain); Török, G. [Wigner Research Center for Physics of the Hungarian Academy of Sciences, 29-33 Konkoly-Thege street, 1121 Budapest XII (Hungary); Ulbricht, A.; Altstadt, E. [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstr. 400, 01328 Dresden (Germany)

2015-06-15

Highlights: • TEM and SANS were applied to estimate mean size and number density of loops, nanovoids and Cu-rich clusters. • A three-feature dispersed-barrier hardening model was applied to estimate the yield stress increase. • The values and errors of the dimensionless obstacle strength were estimated in a consistent way. • Nanovoids are stronger obstacles for dislocation glide than dislocation loops, loops are stronger than Cu-rich clusters. • For reactor-relevant conditions, Cu-rich clusters contribute most to hardening due to their high number density. - Abstract: Dislocation loops, nanovoids and Cu-rich clusters (CRPs) are known to represent obstacles for dislocation glide in neutron-irradiated reactor pressure vessel (RPV) steels, but a consistent experimental determination of the respective obstacle strengths is still missing. A set of Cu-bearing low-Ni RPV steels and model alloys was characterized by means of SANS and TEM in order to specify mean size and number density of loops, nanovoids and CRPs. The obstacle strengths of these families were estimated by solving an over-determined set of linear equations. We have found that nanovoids are stronger than loops and loops are stronger than CRPs. Nevertheless, CRPs contribute most to irradiation hardening because of their high number density. Nanovoids were only observed for neutron fluences beyond typical end-of-life conditions of RPVs. The estimates of the obstacle strength are critically compared with reported literature data.

Integration of a turbine expander with an exothermic reactor loop--Flow sheet development and application to ammonia production

International Nuclear Information System (INIS)

Greeff, I.L.; Visser, J.A.; Ptasinski, K.J.; Janssen, F.J.J.G.

2003-01-01

This paper investigates the direct integration of a gas turbine power cycle with an ammonia synthesis loop. Such a loop represents a typical reactor-separator system with a recycle stream and cold separation of the product from the recycle loop. The hot reaction products are expanded directly instead of raising steam in a waste heat boiler to drive a steam turbine. Two new combined power and chemicals production flow sheets are developed for the process. The flow sheets are simulated using the flow sheet simulator AspenPlus (licensed by Aspen Technology, Inc.) and compared to a simulated conventional ammonia synthesis loop. The comparison is based on energy as well as exergy analysis. It was found that the pressure ratio over the turbine expander plays an important role in optimisation of an integrated system, specifically due to the process comprising an equilibrium reaction. The inlet temperature to the reactor changes with changing pressure ratio, which in turn determines the conversion and consequently the heat of reaction that is available to produce power. In terms of the minimum work requirement per kg of product a 75% improvement over the conventional process could be obtained. The work penalty due to refrigeration needed for separation was also accounted for. Furthermore this integrated flow sheet also resulted in a decrease in exergy loss and the loss was more evenly distributed between the various unit operations. A detailed exergy analysis over the various unit operations proved to be useful in explaining the overall differences in exergy loss between the flow sheets

Post shut-down decay heat removal from nuclear reactor core by natural convection loops in sodium pool

Energy Technology Data Exchange (ETDEWEB)

Rajamani, A. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Sundararajan, T., E-mail: tsundar@iitm.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Prasad, B.V.S.S.S. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Parthasarathy, U.; Velusamy, K. [Nuclear Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

2016-05-15

Highlights: • Transient simulations are performed for a worst case scenario of station black-out. • Inter-wrapper flow between various sub-assemblies reduces peak core temperature. • Various natural convection paths limits fuel clad temperatures below critical level. - Abstract: The 500 MWe Indian pool type Prototype Fast Breeder Reactor (PFBR) has a passive core cooling system, known as the Safety Grade Decay Heat Removal System (SGDHRS) which aids to remove decay heat after shut down phase. Immediately after reactor shut down the fission products in the core continue to generate heat due to beta decay which exponentially decreases with time. In the event of a complete station blackout, the coolant pump system may not be available and the safety grade decay heat removal system transports the decay heat from the core and dissipates it safely to the atmosphere. Apart from SGDHRS, various natural convection loops in the sodium pool carry the heat away from the core and deposit it temporarily in the sodium pool. The buoyancy driven flow through the small inter-wrapper gaps (known as inter-wrapper flow) between fuel subassemblies plays an important role in carrying the decay heat from the sub-assemblies to the hot sodium pool, immediately after reactor shut down. This paper presents the transient prediction of flow and temperature evolution in the reactor subassemblies and the sodium pool, coupled with the safety grade decay heat removal system. It is shown that with a properly sized decay heat exchanger based on liquid sodium and air chimney stacks, the post shutdown decay heat can be safely dissipated to atmospheric air passively.

Investigation of hydrogen bubbles behavior in tungsten by high-flux hydrogen implantation

Science.gov (United States)

Zhao, Jiangtao; Meng, Xuan; Guan, Xingcai; Wang, Qiang; Fang, Kaihong; Xu, Xiaohui; Lu, Yongkai; Gao, Jun; Liu, Zhenlin; Wang, Tieshan

2018-05-01

Hydrogen isotopes retention and bubbles formation are critical issues for tungsten as plasma-facing material in future fusion reactors. In this work, the formation and growing up behavior of hydrogen bubbles in tungsten were investigated experimentally. The planar TEM samples were implanted by 6.0keV hydrogens to a fluence of 3.38 ×1018 H ṡ cm-2 at room temperature, and well-defined hydrogen bubbles were observed by TEM. It was demonstrated that hydrogen bubbles formed when exposed to a fluence of 1.5 ×1018 H ṡ cm-2 , and the hydrogen bubbles grew up with the implantation fluence. In addition, the bubbles' size appeared larger with higher beam flux until saturated at a certain flux, even though the total fluence was kept the same. Finally, in order to understand the thermal annealing effect on the bubbles behavior, hydrogen-implanted samples were annealed at 400, 600, 800, and 1000 °C for 3 h. It was obvious that hydrogen bubbles' morphology changed at temperatures higher than 800 °C.

Determination of 16N and 19O activities in loop water of swimming pool reactor

International Nuclear Information System (INIS)

Ding Shengyao; Xu Kun; Yu Baosheng; Ling Yude

2006-01-01

Measurements of activities for 16 N and 19 O nuclei in the loop water of swimming pool reactor at China Institute of Atomic Energy were carried out. In order to verify the experiment results, a calculation for same purpose was also performed. The results show their coincidence is well in uncertainty range. The evaluated recommendation data for 18 O(n, γ) 19 O reaction cross sections are also given in the paper. (authors)

BWR type reactors

International Nuclear Information System (INIS)

Watanabe, Shoichi

1986-01-01

Purpose: To enable to remove water not by way of mechanical operation in a reactor core and improve the fuel economy in BWR type reactors. Constitution: A hollow water removing rod of a cross-like profile made of material having a smaller neutron absorption cross section than the moderator is disposed to the water gap for each of unit structures composed of four fuel assemblies, and water is charged and discharged to and from the water removing rod. Water is removed from the water removing rod to decrease the moderators in the water gap to carry out neutron spectrum shift operation from the initial to the medium stage of reactor core cycles. At the final stage of the cycle, airs in the water removing rod are extracted and the moderator is introduced. The moderator is filled and the criticality is maintained with the accumulated nuclear fission materials. The neutron spectrum shift operation can be attained by eliminating hydrothermodynamic instability and using a water removing rod of a simple structure. (Horiuchi, T.)

Novel Magnetically Fluidized Bed Reactor Development for the Looping Process: Coal to Hydrogen Production R&D

Energy Technology Data Exchange (ETDEWEB)

Mei, Renwei; Hahn, David; Klausner, James; Petrasch, Jorg; Mehdizadeh, Ayyoub; Allen, Kyle; Rahmatian, Nima; Stehle, Richard; Bobek, Mike; Al-Raqom, Fotouh; Greek, Ben; Li, Like; Chen, Chen; Singh, Abhishek; Takagi, Midori; Barde, Amey; Nili, Saman

2013-09-30

The coal to hydrogen project utilizes the iron/iron oxide looping process to produce high purity hydrogen. The input energy for the process is provided by syngas coming from gasification process of coal. The reaction pathways for this process have been studied and favorable conditions for energy efficient operation have been identified. The Magnetically Stabilized Porous Structure (MSPS) is invented. It is fabricated from iron and silica particles and its repeatable high performance has been demonstrated through many experiments under various conditions in thermogravimetric analyzer, a lab-scale reactor, and a large scale reactor. The chemical reaction kinetics for both oxidation and reduction steps has been investigated thoroughly inside MSPS as well as on the surface of very smooth iron rod. Hydrogen, CO, and syngas have been tested individually as the reducing agent in reduction step and their performance is compared. Syngas is found to be the most pragmatic reducing agent for the two-step water splitting process. The transport properties of MSPS including porosity, permeability, and effective thermal conductivity are determined based on high resolution 3D CT x-ray images obtained at Argonne National Laboratory and pore-level simulations using a lattice Boltzmann Equation (LBE)-based mesoscopic model developed during this investigation. The results of those measurements and simulations provide necessary inputs to the development of a reliable volume-averaging-based continuum model that is used to simulate the dynamics of the redox process in MSPS. Extensive efforts have been devoted to simulate the redox process in MSPS by developing a continuum model consist of various modules for conductive and radiative heat transfer, fluid flow, species transport, and reaction kinetics. Both the Lagrangian and Eulerian approaches for species transport of chemically reacting flow in porous media have been investigated and verified numerically. Both approaches lead to correct

Decommissioning of TRIGA Mark II type reactor

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dooseong; Jeong, Gyeonghwan; Moon, Jeikwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-10-15

The first research reactor in Korea, KRR 1, is a TRIGA Mark II type with open pool and fixed core. Its power was 100 kWth at its construction and it was upgraded to 250 kWth. Its construction was started in 1957. The first criticality was reached in 1962 and it had been operated for 36,000 hours. The second reactor, KRR 2, is a TRIGA Mark III type with open pool and movable core. These reactors were shut down in 1995, and the decision was made to decommission both reactors. The aim of the decommissioning activities is to decommission the KRR 2 reactor and decontaminate the residual building structures and site, and to release them as unrestricted areas. The KRR 1 reactor was decided to be preserve as a historical monument. A project was launched for the decommissioning of these reactors in 1997, and approved by the regulatory body in 2000. A total budget for the project was 20.0 million US dollars. It was anticipated that this project would be completed and the site turned over to KEPCO by 2010. However, it was discovered that the pool water of the KRR 1 reactor was leaked into the environment in 2009. As a result, preservation of the KRR 1 reactor as a monument had to be reviewed, and it was decided to fully decommission the KRR 1 reactor. Dismantling of the KRR 1 reactor takes place from 2011 to 2014 with a budget of 3.25 million US dollars. The scope of the work includes licensing of the decommissioning plan change, removal of pool internals including the reactor core, removal of the thermal and thermalizing columns, removal of beam port tubes and the aluminum liner in the reactor tank, removal of the radioactive concrete (the entire concrete structure will not be demolished), sorting the radioactive waste (concrete and soil) and conditioning the radioactive waste for final disposal, and final statuses of the survey and free release of the site and building, and turning over the site to KEPCO. In this paper, the current status of the TRIGA Mark-II type reactor

Vibration-proof FBR type reactor

International Nuclear Information System (INIS)

Kawamura, Yutaka.

1992-01-01

In a reactor container in an FBR type reactor, an outer building and upper and lower portions of a reactor container are connected by a load transmission device made of a laminated material of rubber and steel plates. Each of the reactor container and the outer building is disposed on a lower raft disposed on a rock by way of a vibration-proof device made of a laminated material of rubber and steel plates. Vibration-proof elements for providing vertical eigen frequency of the vibration-proof system comprising the reactor building and the vibration-proof device within a range of 3Hz to 5Hz are used. That is, the peak of designed acceleration for response spectrum in the horizontal direction of the reactor structural portions is shifted to side of shorter period from the main frequency region of the reactor structure. Alternatively, rigidity of the vibration-proof elements is decreased to shift the peak to the side of long period from the main frequency region. Designed seismic force can be greatly reduced both horizontally and vertically, to reduce the wall thickness of the structural members, improve the plant economy and to ensure the safety against earthquakes. (N.H.)

Three-dimensional reactor dynamics code for VVER type nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

1995-10-01

A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

Three-dimensional reactor dynamics code for VVER type nuclear reactors

International Nuclear Information System (INIS)

Kyrki-Rajamaeki, R.

1995-10-01

A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.)

Chemical surveillance of commercial fast breeder reactors

International Nuclear Information System (INIS)

Stamm, H.H.; Stade, K.Ch.

1988-01-01

After BN-600 (USSR) and SUPERPHENIX (France) were started succesfully, the international development of LMFBRs is standing at the doorstep of commercial use. For commercial use of LMFBRs cost reductions for construction and operation are highly desirable and necessary. Several nations developing breeder reactors have joined in a common effort in order to reach this aim by standardization and harmonization. On the base of more than 20 years of operation experience of experimental reactors (EBR-II, FFTF, RAPSODIE, DFR, BR-5/BR-10, BOR-60, JOYO, KNK-II) and demonstration plants (PHENIX, PFR, BN-350), possibilities for standardization in chemical surveillance of commercial breeder reactors without any loss of availability, reliability and reactor safety will be discussed in the following chapters. Loop-type reactors will be considered as well as pool-type reactors, although all commercial plants under consideration so far (SUPERPHENIX II, BN-800, BN-1600, CFBR, SNR-2, EFR) include pool-type reactors only. Table 1 gives a comparison of the Na inventories of test reactors, prototype plants and commercial LMFBRs

Acoustic emission events from sodium vapour bubble collapsing: a stochastic model

Energy Technology Data Exchange (ETDEWEB)

Colombino, A; Dentico, G; Pacilio, N; Papalia, B; Taglienti, S; Tosi, V; Vigo, A [Comitato Nazionale per l' Energia Nucleare, Casaccia (Italy). Centro di Studi Nucleari; Galli, C [Rome Univ. (Italy). Ist. di Matematica

1981-01-01

The forward Kolomogorov equation method has been applied to a zero-dimensional model which describes the time distribution of acoustic emissions from sodium vapour bubble collapsing. Processes taken into account as components for outlining the upstated phenomenon are: energy generation, energy dissipation, bubble creation, acoustic emission and energy release from bubble collapsing. Processes involve affect or are induced by a population of particles (bubbles, acoustic pulses) and pseudoparticles (energetic units). A formulation is obtained for the expected values of some stochastic indicators, i.e., factorial moments and cumulants, autocorrelation functions, waiting time distribution between contiguous events, of the time series consisting of acoustic emission pulses as detected by a suitable sensor. Preliminary, but promising, validation of the model and a sound prelude to effective boiling regime diagnosing is obtained by processing data from the out-of-pile CFNa loop in Grenoble, France. Data are collected from a piezoelectric accelerometer located nearby the circuit.

Determination of the Clean Air Delivery Rate (CADR of Photocatalytic Oxidation (PCO Purifiers for Indoor Air Pollutants Using a Closed-Loop Reactor. Part I: Theoretical Considerations

Directory of Open Access Journals (Sweden)

Éric Dumont

2017-03-01

Full Text Available This study demonstrated that a laboratory-scale recirculation closed-loop reactor can be an efficient technique for the determination of the Clean Air Delivery Rate (CADR of PhotoCatalytic Oxidation (PCO air purification devices. The recirculation closed-loop reactor was modeled by associating equations related to two ideal reactors: one is a perfectly mixed reservoir and the other is a plug flow system corresponding to the PCO device itself. Based on the assumption that the ratio between the residence time in the PCO device and the residence time in the reservoir τP/τR tends to 0, the model highlights that a lab closed-loop reactor can be a suitable technique for the determination of the efficiency of PCO devices. Moreover, if the single-pass removal efficiency is lower than 5% of the treated flow rate, the decrease in the pollutant concentration over time can be characterized by a first-order decay model in which the time constant is proportional to the CADR. The limits of the model are examined and reported in terms of operating conditions (experiment duration, ratio of residence times, and flow rate ranges.

Experience and prospects for developing research reactors of different types

International Nuclear Information System (INIS)

Kuatbekov, R.P.; Tretyakov, I.T.; Romanov, N.V.; Lukasevich, I.B.

2015-01-01

NIKIET has a 60-year experience in the development of research reactors. Altogether, there have been more than 25 NIKIET-designed plants of different types built in Russia and 20 more in other countries, including pool-type water-cooled and water moderated research reactors, tank-type and pressure-tube research reactors, pressurized high-flux, heavy-water, pulsed and other research reactors. Most of the research reactors were designed as multipurpose plants for operation at research centers in a broad range of applications. Besides, unique research reactors were developed for specific application fields. Apart from the experience in the development of research reactor designs and the participation in the reactor construction, a unique amount of knowledge has been gained on the operation of research reactors. This makes it possible to use highly reliable technical solutions in the designs of new research reactors to ensure increased safety, greater economic efficiency and maintainability of the reactor systems. A multipurpose pool-type research reactor of a new generation is planned to be built at the Center for Nuclear Energy Science & Technology (CNEST) in the Socialist Republic of Vietnam to be used to support a spectrum of research activities, training of skilled personnel for Vietnam nuclear industry and efficient production of isotopes. It is exactly the applications a research reactor is designed for that defines the reactor type, design and capacity, and the selection of fuel and components subject to all requirements of industry regulations. The design of the new research reactor has a great potential in terms of upgrading and installation of extra experimental devices. (author)

The water desalination complex based on ABV-type reactor plant

International Nuclear Information System (INIS)

Panov, Yu.K.; Fadeev, Yu.P.; Vorobiev, V.M.; Baranaev, Yu.D.

1997-01-01

A floating nuclear desalination complex with two barges, one for ABV type reactor plant, with twin reactor 2 x 6 MW(e), and one for reverse osmosis desalination plant, was described. The principal specifications of the ABV type reactor plant and desalination barge were given. The ABV type reactor has a traditional two-circuit layout using an integral type reactor vessel with all mode natural convection of primary coolant. The desalted water cost was estimated to be around US $0.86 per cubic meter. R and D work has been performed and preparations for commercial production are under way. (author)

Development of toroid-type HTS DC reactor series for HVDC system

Energy Technology Data Exchange (ETDEWEB)

Kim, Kwangmin, E-mail: kwangmin81@gmail.com [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Lee, Sangjin [Uiduk University, Gyeongju 780-713 (Korea, Republic of); Oh, Yunsang [Vector Fields Korea Inc., Pohang 790-834 (Korea, Republic of); Park, Minwon; Yu, In-Keun [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of)

2015-11-15

Highlights: • The authors developed the 400 mH, 400 A class toroid-type HTS DC reactor system. • The target temperature, inductance and operating current are under 20 K at magnet, 400 mH and 400 A, respectively. All target performances of the HTS DC reactor were achieved. • The HTS DC reactor was conducted through the interconnection operation with a LCC type HVDC system. • Now, the authors are studying the 400 mH, 1500 A class toroid-type HTS DC reactor for the next phase HTS DC reactor. - Abstract: This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

Development of toroid-type HTS DC reactor series for HVDC system

International Nuclear Information System (INIS)

Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

2015-01-01

Highlights: • The authors developed the 400 mH, 400 A class toroid-type HTS DC reactor system. • The target temperature, inductance and operating current are under 20 K at magnet, 400 mH and 400 A, respectively. All target performances of the HTS DC reactor were achieved. • The HTS DC reactor was conducted through the interconnection operation with a LCC type HVDC system. • Now, the authors are studying the 400 mH, 1500 A class toroid-type HTS DC reactor for the next phase HTS DC reactor. - Abstract: This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

Theory calculation of combination of 'embryo' bubble growing-up visible bubble in bubble chamber

International Nuclear Information System (INIS)

Ye Zipiao; Sheng Xiangdong; Dai Changjiang

2004-01-01

By aid of island combination theory of 'embryo' bubble, it is resolved well the question which 'embryo' bubble grows up a visible bubble in the bubble chamber. Through theory calculation it is shown that radius of the big' embryo' bubble combinated not only relates with work matter such as surface tension coefficient, saturation vapour pressure and boiling point of liquid, but also does absorbing quantity of heat and the numbers of 'embryo' bubbles combination. It is explained reasonably that the radius of bubbles in bubble chamber is different for the same energies of neutrons and proton. The track of neutron in bubble chamber is long and thin, and the track of proton in bubble chamber is wide and short. It is also explained reasonably that the bubble radius of the incident particles with more charges which there are the same energies will be wider than that of the incident particles with less charges in the track. (author)

Flow Components in a NaK Test Loop Designed to Simulate Conditions in a Nuclear Surface Power Reactor

Science.gov (United States)

Polzin, Kurt A.; Godfroy, Thomas J.

2008-01-01

A test loop using NaK as the working fluid is presently in use to study material compatibility effects on various components that comprise a possible nuclear reactor design for use on the lunar surface. A DC electromagnetic (EM) pump has been designed and implemented as a means of actively controlling the NaK flow rate through the system and an EM flow sensor is employed to monitor the developed flow rate. These components allow for the matching of the flow rate conditions in test loops with those that would be found in a full-scale surface-power reactor. The design and operating characteristics of the EM pump and flow sensor are presented. In the EM pump, current is applied to a set of electrodes to produce a Lorentz body force in the fluid. A measurement of the induced voltage (back-EMF) in the flow sensor provides the means of monitoring flow rate. Both components are compact, employing high magnetic field strength neodymium magnets thermally coupled to a water-cooled housing. A vacuum gap limits the heat transferred from the high temperature NaK tube to the magnets and a magnetically-permeable material completes the magnetic circuit. The pump is designed to produce a pressure rise of 5 psi, and the flow sensor's predicted output is roughly 20 mV at the loop's nominal flow rate of 0.5 GPM.

WWER-440 type reactor core

International Nuclear Information System (INIS)

Mizov, J.; Svec, P.; Rajci, T.

1987-01-01

Assemblies with patly spent fuel of enrichment within 5 and 36 MWd/kg U or lower than the maximum enrichment of freshly charged fuel are placed in at least one of the peripheral positions of each hexagonal sector of the WWER-440 reactor type core. This increases fuel availability and reduces the integral neutron dose to the reactor vessel. The duration is extended of the reactor campaign and/or the mean fuel enrichment necessary for the required duration of the period between refuellings is reduced. Thus, fuel costs are reduced by 1 up to 3%. The results obtained in the experiment are tabulated. (J.B.). 1 fig., 3 tabs

Improvement of the bubble rise velocity model in the pressurizer using ALMOD 3 computer code to calculate evaporation

International Nuclear Information System (INIS)

Madeira, A.A.

1985-01-01

It's studied the improvement for the calculation of bubble rise velocity, by adding two different ways to estimate this velocity, one of which more adequate to pressures normally found in the Reactor Cooling System. Additionally, a limitation in bubble rise velocity growth was imposed, to account for the actual behavior of bubble rise in two-phase mixtures. (Author) [pt

Tests in the ATLE loop on the PIUS design

International Nuclear Information System (INIS)

Bredolt, U.; Babala, D.; Kemppainen, J.

1992-01-01

This paper describes experimental demonstration of the self-protective features of Process Inherent Ultimate Safety (PIUS) design in a large scale test loop in ABB Atoms engineering laboratories. The loop employs real time simulation of core power as a function of coolant conditions in an electrically heated fuel assembly model. System responses to various severe transients were studied. Comparisons were made with predictions of the RIGEL code, which has been developed specifically for study of PIUS type reactors. A comparison between test results and calculated results was made for main state variables such as pressure, temperatures, concentrations, heat fluxes and mass flow rates. The tests have demonstrated the self-protective thermal-hydraulics of pressurized water reactor primary systems designed according to the PIUS principle and verified the capability of the RIGEL code to predict their behavior during severe accidents and in normal operation transients

Study on ex-vessel cooling of RPV (behavior of coalesced bubbles and trigger condition of critical heat flux on inclined plate)

International Nuclear Information System (INIS)

Ohtake, H.; Koizumi, Y.; Takano, K.I.

2001-01-01

The Ex-vessel cooling of Reactor-Pressure-Vessel in Light-Water-Reactor at the severe accident have been proposed for future nuclear reactors. The estimation of Critical-Heat-Flux on a downward-facing curvilinear surface, like a hemisphere, is important to the assessment of the cooling. In this study, the CHFs on inclined surfaces were examined experimentally focusing on orientation of the heating surface. In order to discuss detailed mechanism of the CHF, the behaviors of coalesced bubbles near the heating surface were investigated through visual observations. The critical heat flux obtained in the present experiments increased with the inclined angle over the present experimental range. The dependence of the inclined angle on the critical heat flux was q CHF,R-113 [q] = f (q 0.33 ) for the present experimental results. The effect of the surface orientation on the critical heat flux was roughly explained by using the simple analytical model based on the macro-layer model and Kelvin-Helmholtz instability. From visual observations for behavior of bubbles near the heating surface, whereas the coalesced bubble covered over the heating surface for the inclined angle of 0 degree, the coalesced bubble moved upward to avoid packing the bubble on the surface above 5 degree. As the inclined angle increased, the velocity of the coalesced bubble was high, the period covered the heater and the bubble length were small. The results suggested that the CHF was closely related to forming the coalesced bubble and the behavior of the bubble. (author)

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)

Effects of surface orientation on lifetime of near-surface nanoscale He bubble in tungsten

Energy Technology Data Exchange (ETDEWEB)

Cui, Jiechao; Fu, Baoqin; Wu, Zhangwen; Hou, Qing, E-mail: qhou@scu.edu.cn

2017-02-15

In multiscale modeling of the morphological evolution of plasma facing materials in nuclear fusion reactors, the knowledge of the timescales of the involved physical processes is important. In the present study, a new method based on molecular dynamics simulations was developed to extract the lifetime of helium bubbles near tungsten surfaces. It was found that the lifetime of a helium bubble can be described by the Arrhenius equation. However, the lifetime of a helium bubble depends on the thickness of tungsten film above the helium bubble in the substrate and the bubble size. The influence of surface orientations on the lifetime of helium bubbles was also observed, and the performance of helium bubbles on the (1 1 1) surface is very different from on the (0 0 1) and (0 1 1) surfaces. The role of the helium bubble lifetime in other simulation techniques, such as in kinetic Monte Carlo methods and rate theory, is discussed.

Methodology on the sparger development for Korean next generation reactor

International Nuclear Information System (INIS)

Kim, Hwan Yeol; Hwang, Y.D.; Kang, H.S.; Cho, B.H.; Park, J.K.

1999-06-01

In case of an accident, the safety depressurization system of Korean Next Generation Reactor (KNGR) efficiently depressurize the reactor pressure by directly discharge steam of high pressure and temperature from the pressurizer into the in-containment refuelling water storage tank (IRWST) through spargers. This report was generated for the purpose of developing the sparger of KNGR. This report presents the methodology on application of ABB-Atom. Many thermal hydraulic parameters affecting the maximum bubble could pressure were obtained and the maximum bubble cloud pressure transient curve so called forcing function of KNGR was suggested and design inputs for IRWST (bubble cloud radius vs. time, bubble cloud velocity vs. time, bubble cloud acceleration vs. time, etc.) were generated by the analytic using Rayleigh-Plesset equation. (author). 17 refs., 6 tabs., 27 figs

Methodology on the sparger development for Korean next generation reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Hwan Yeol; Hwang, Y.D.; Kang, H.S.; Cho, B.H.; Park, J.K

1999-06-01

In case of an accident, the safety depressurization system of Korean Next Generation Reactor (KNGR) efficiently depressurize the reactor pressure by directly discharge steam of high pressure and temperature from the pressurizer into the in-containment refuelling water storage tank (IRWST) through spargers. This report was generated for the purpose of developing the sparger of KNGR. This report presents the methodology on application of ABB-Atom. Many thermal hydraulic parameters affecting the maximum bubble could pressure were obtained and the maximum bubble cloud pressure transient curve so called forcing function of KNGR was suggested and design inputs for IRWST (bubble cloud radius vs. time, bubble cloud velocity vs. time, bubble cloudacceleration vs. time, etc.) were generated by the analytic using Rayleigh-Plesset equation. (author). 17 refs., 6 tabs., 27 figs.

Prevention device for rapid reactor core shutdown in BWR type reactors

International Nuclear Information System (INIS)

Koshi, Yuji; Karatsu, Hiroyuki.

1986-01-01

Purpose: To surely prevent rapid shutdown of a nuclear reactor upon partial load interruption due to rapid increase in the system frequency. Constitution: If a partial load interruption greater than the sum of the turbine by-pass valve capacity and the load setting bias portion is applied in a BWR type power plant, the amount of main steams issued from the reactor is decreased, the thermal input/output balance of the reactor is lost, the reactor pressure is increased, the void is collapsed, the neutron fluxes are increased and the reactor power rises to generate rapid reactor shutdown. In view of the above, the turbine speed signal is compared with a speed setting value in a recycling flowrate control device and the recycling pump is controlled to decrease the recycling flowrate in order to compensate the increase in the neutron fluxes accompanying the reactor power up. In this way, transient changes in the reactor core pressure and the neutron fluxes are kept within a setting point for the rapid reactor shutdown operation thereby enabling to continue the plant operation. (Horiuchi, T.)

Hydrogen injection device in BWR type reactor

International Nuclear Information System (INIS)

Takagi, Jun-ichi; Kubo, Koji.

1988-01-01

Purpose: To reduce the increasing ratio of main steam system dose rate due to N-16 activity due to excess hydrogen injection in the hydrogen injection operation of BWR type reactors. Constitution: There are provided a hydrogen injection mechanism for injecting hydrogen into primary coolants of a BWR type reactor, and a chemical injection device for injecting chemicals such as methanol, which makes nitrogen radioisotopes resulted in the reactor water upon hydrogen injection non-volatile, into the pressure vessel separately from hydrogen. Injected hydrogen and the chemicals are not reacted in the feedwater system, but the reaction proceeds due to the presence of radioactive rays after the injection into the pressure vessel. Then, hydrogen causes re-combination in the downcomer portion to reduce the dissolved oxygen concentration. Meanwhile, about 70 % of the chemicals is supplied by means of a jet pump directly to the reactor core, thereby converting the chemical form of N-16 in the reactor core more oxidative (non-volatile). (Kawakami, Y.)

Evaluation of very low frequencies of ATWS and PLOHS in a loop-type FBR plant by making use of inherently safe features

International Nuclear Information System (INIS)

Sakata, K.; Koyama, K.; Aoi, S.; Simonelli, R.B.; Wallace, I.T.

1987-01-01

Frequencies of ATWS (Anticipated Transient Without Scram) and PLOHS (Protected Loss of Heat Sink) for a large loop-type FBR plant were evaluated by applying PSA methodologies. The frequencies were found to be so low that ATWS and PLOHS could be excluded from candidates of the design basis events. Furthermore, the inherently safe features introduced to the system design were verified to be very effective for reduction of the Probability of CCF (Common Cause Failure), which deteriorates reliability of both the reactor shutdown and the decay heat removal systems. (orig.)

Post irradiation examination of type 316 stainless steels for in-pile Oarai water loop No.2 (OWL-2)

International Nuclear Information System (INIS)

Shibata, Akira; Kimura, Tadashi; Nagata, Hiroshi; Aoyama, Masashi; Kanno, Masaru; Ohmi, Masao

2010-11-01

The Oarai water loop No.2 (OWL-2) was installed in JMTR in 1972 for the purpose of irradiation experiments of fuel element and component material for light water reactors. Type 316 stainless steels (SSs) were used for tube material of OWL-2 in the reactor. But data of mechanical properties of highly irradiated Type 316 SSs has been insufficient since OWL-2 was installed. Therefore surveillance tests of type 316 SSs which were irradiated up to 3.4x10 25 n/m 2 in fast neutron fluence (>1 MeV) were performed. Meanwhile type 316 stainless steel (SS) is widely used in JMTR such as other irradiation apparatus and irradiation capsule, and additional data of type 316 SSs irradiated higher is required. Therefore post irradiation examinations of surveillance specimens made of type 316 SSs which were irradiated up to 1.0x10 26 n/m 2 in fast neutron fluence were performed and reported in this paper. In this result of surveillance tests of type 316 SSs irradiated up to 1.0x10 26 n/m 2 , tensile strength increase with increase of Neutron fluence and total elongation decreased with increase of Neutron fluence compared to unirradiated specimens and specimens irradiated up to 3.4x10 25 n/m 2 . This tendency has good agreement with results of 10 24 - 10 25 n/m 2 in fast neutron fluence. More than 37% in total elongation was confirmed in all test conditions. It was confirmed that type 316 SS irradiated up to 1.0x10 26 n/m 2 in fast neutron fluence has enough ductility as structure material. (author)

Acoustic detection of boiling in the Sodium Loop Safety Facility in-reactor experiment P1

International Nuclear Information System (INIS)

Carey, W.M.; Anderson, T.T.; Bobis, J.P.

1976-06-01

Acoustic data were obtained from two high-temperature lithium niobate microphones on the loop background noise and transient pressure pulses during the Sodium Loop Safety Facility (SLSF) P1 in-reactor experiment. This experiment simulated an LMFBR loss-of-piping-integrity (LOPI) transient on a nineteen element, end-of-life, enriched-UO 2 fuel assembly. The microphones were exposed to liquid sodium at a distance 4.85 meters above the reactor core at temperatures between 315 0 and 590 0 C. The distance and location of the microphones in the P1 Test Train provided an attenuative transmission path which was undesirable for optimum acoustic detection of sodium boiling and fuel failure. The data gathered on the loop background noise was observed to be dominated by pump and electrical noise at frequencies below 1.5 KHz and appeared to be dominated by flow induced local turbulence noise at higher frequencies. During the period of time that the sodium in the fuel assembly was at its saturation temperature 943 0 C (1730 0 F), as indicated by the wire wrap thermocouples, several discrete pulses were observed with peak-to-peak pressure between 3.3 kPa and 7.9 kPa and center frequencies between 360 and 550 Hz. The pulses occurred at two separate gradually increasing repetition rates. These observations appear to be consistent with the result of an impulsive forcing function interacting with a band passed Helmholtz resonator. These data are consistent with the hypothesis that sodium boiling occurred in the P1 fuel assembly, resulting in the formation of individual voids that collapsed upon reaching the subcooled sodium. These data provide pertinent information regarding the feasibility of sodium boiling detection and may provide additional insight into the dynamics of the void behavior

A complete closed loop control system for MNSR

International Nuclear Information System (INIS)

Wang Liyu

1988-01-01

The MNSR (Miniature Neutron Source Reactor), a 27 kW reactor, was built in 1984 in the Institute of Atomic Energy, Beijing, China. A microcomputer closed-loop control system has been developed for the MNSR using an IBM PC-XT. This system not only controls the reactor, but it is also a data acquisition system and a reactivity monitor for the reactor. A closed-loop control system can be used for reactors such as the MNSR, in which the power is self-limited and the reactivity is limited to less than 5mk (∼0.7$). As a data acquisition system or a reactivity monitor, it is applicable for any nuclear reactor. This paper describes the hardware and software in detail. 4 refs, 11 figs

Inhomogeneous MUSIG Model - a population balance approach for polydispersed bubbly flows

International Nuclear Information System (INIS)

Frank, T.; Zwart, P.J.; Shi, J.; Krepper, E.; Lucas, D.; Rohde, U.

2005-01-01

Many flow regimes in Nuclear Reactor Safety (NRS) Research are characterized by multiphase flows, with one phase being a continuous liquid and the other phase consisting of gas or vapour of the liquid phase. In the range of low to intermediate volume fraction of the gaseous phase the multiphase flow under consideration is a bubbly or slug flow, where the disperse phase is characterized by an evolving bubble size distribution due to bubble breakup and coalescence processes. The paper presents a generalized inhomogeneous Multiple Size Group (MUSIG) Model. Within this model the disperse gaseous phase is divided into N inhomogeneous velocity groups (phases) and each of these groups is subdivided into M bubble size classes. Bubble breakup and coalescence processes between all bubble size classes are taken into account by appropriate models. The derived inhomogeneous MUSIG model has been validated against experimental data from the TOPFLOW test facility at the Research Center Rossendorf (FZR). Comparisons of gas volume fraction and velocity profiles with TOPFLOW-074 test case data are provided, showing the applicability and accuracy of the model for polydispersed bubbly flow in large diameter vertical pipe flow. (author)

Development of toroid-type HTS DC reactor series for HVDC system

Science.gov (United States)

Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

2015-11-01

This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

PREDICTION OF GAS HOLD-UP IN A COMBINED LOOP AIR LIFT FLUIDIZED BED REACTOR USING NEWTONIAN AND NON-NEWTONIAN LIQUIDS

Directory of Open Access Journals (Sweden)

Sivakumar Venkatachalam

2011-09-01

Full Text Available Many experiments have been conducted to study the hydrodynamic characteristics of column reactors and loop reactors. In this present work, a novel combined loop airlift fluidized bed reactor was developed to study the effect of superficial gas and liquid velocities, particle diameter, fluid properties on gas holdup by using Newtonian and non-Newtonian liquids. Compressed air was used as gas phase. Water, 5% n-butanol, various concentrations of glycerol (60 and 80% were used as Newtonian liquids, and different concentrations of carboxy methyl cellulose aqueous solutions (0.25, 0.6 and 1.0% were used as non-Newtonian liquids. Different sizes of spheres, Bearl saddles and Raschig rings were used as solid phases. From the experimental results, it was found that the increase in superficial gas velocity increases the gas holdup, but it decreases with increase in superficial liquid velocity and viscosity of liquids. Based on the experimental results a correlation was developed to predict the gas hold-up for Newtonian and non-Newtonian liquids for a wide range of operating conditions at a homogeneous flow regime where the superficial gas velocity is approximately less than 5 cm/s

DynMo: Dynamic Simulation Model for Space Reactor Power Systems

International Nuclear Information System (INIS)

El-Genk, Mohamed; Tournier, Jean-Michel

2005-01-01

A Dynamic simulation Model (DynMo) for space reactor power systems is developed using the SIMULINK registered platform. DynMo is modular and could be applied to power systems with different types of reactors, energy conversion, and heat pipe radiators. This paper presents a general description of DynMo-TE for a space power system powered by a Sectored Compact Reactor (SCoRe) and that employs off-the-shelf SiGe thermoelectric converters. SCoRe is liquid metal cooled and designed for avoidance of a single point failure. The reactor core is divided into six equal sectors that are neutronically, but not thermal-hydraulically, coupled. To avoid a single point failure in the power system, each reactor sector has its own primary and secondary loops, and each loop is equipped with an electromagnetic (EM) pump. A Power Conversion assembly (PCA) and a Thermoelectric Conversion Assembly (TCA) of the primary and secondary EM pumps thermally couple each pair of a primary and a secondary loop. The secondary loop transports the heat rejected by the PCA and the pumps TCA to a rubidium heat pipes radiator panel. The primary loops transport the thermal power from the reactor sector to the PCAs for supplying a total of 145-152 kWe to the load at 441-452 VDC, depending on the selections of the primary and secondary liquid metal coolants. The primary and secondary coolant combinations investigated are lithium (Li)/Li, Li/sodium (Na), Na-Na, Li/NaK-78 and Na/NaK-78, for which the reactor exit temperature is kept below 1250 K. The results of a startup transient of the system from an initial temperature of 500 K are compared and discussed

Study of rare gases behavior in uranium dioxide: diffusion and bubble nucleation and growth mechanisms

International Nuclear Information System (INIS)

Michel, A.

2011-01-01

During in-reactor irradiation of the nuclear fuel, fission gases, mainly xenon and krypton, are generated that are subject to several phenomena: diffusion and precipitation. These phenomena can have adverse consequences on the fuel physical and chemical properties and its in-reactor behavior. The purpose of this work is to better understand the behavior of fission gases by identifying diffusion, bubble nucleation and growth mechanisms. To do this, studies involving separate effects have been established coupling ion irradiations/implantations with fine characterizations on Large Scale Facilities. The influence of several parameters such as gas type, concentration and temperature has been identified separately. Interpretation of the Thermal Desorption Spectrometry (TDS) measurements has enabled us to determine xenon and krypton diffusion coefficients in uranium dioxide. A heterogeneous nucleation mechanism on defects was determined by means of experiments on the JANNuS platform in Orsay that consists of a coupling of an implantor, an accelerator and a Transmission Electron Microscope (TEM). Finally, TEM and X-ray Absorption Spectroscopy characterizations of implanted and annealed samples put in relieve a bubble growth mechanism by atoms and vacancies capture. (author) [fr

Catalytic wet air oxidation of coke-plant wastewater on ruthenium-based eggshell catalysts in a bubbling bed reactor.

Science.gov (United States)

Yang, M; Sun, Y; Xu, A H; Lu, X Y; Du, H Z; Sun, C L; Li, C

2007-07-01

Catalytic wet air of coke-plant wastewater was studied in a bubbling bed reactor. Two types of supported Ru-based catalysts, eggshell and uniform catalysts, were employed. Compared with the results in the wet air oxidation of coke-plant wastewater, supported Ru uniform catalysts showed high activity for chemical oxygen demand (COD) and ammonia/ammonium compounds (NH3-N) removal at temperature of 250 degrees C and pressure of 4.8 MPa, and it has been demonstrated that the catalytic activity of uniform catalyst depended strongly on the distribution of active sites of Ru on catalyst. Compared to the corresponding uniform catalysts with the same Ru loading (0.25 wt.% and 0.1 wt.%, respectively), the eggshell catalysts showed higher activities for CODcr removal and much higher activities for NH3-N degradation. The high activity of eggshell catalyst for treatment of coke-plant wastewater can be attributed to the higher density of active Ru sites in the shell layer than that of the corresponding uniform catalyst with the same Ru loading. It has been also evidenced that the active Ru sites in the internal core of uniform catalyst have very little or no contribution to CODcr and NH3-N removal in the total oxidation of coke-plant wastewater.

Hydrodynamics of the continuously filtering slurry reactor. Influence of load of solids and particle size distribution.

NARCIS (Netherlands)

Huizenga, P.; Kuipers, J.A.M.; van Swaaij, Willibrordus Petrus Maria

1997-01-01

Internal filtration in slurry bubble columns offers a possible solution to the filtration problems related to this reactor type. The applicability of the concept has already been demonstrated at full-scale for waste water treatment. Theoretical description of internal filtration is lacking, however.

Hydrodynamics of the continuously filtering slurry reactor. Influence of load of solids and particle size distribution

NARCIS (Netherlands)

Huizenga, P.; Kuipers, J.A.M.; van Swaaij, W.P.M.

1997-01-01

Internal filtration in slurry bubble columns offers a possible solution to the filtration problems related to this reactor type. The applicability of the concept has already been demonstrated at full-scale for waste water treatment. Theoretical description of internal filtration is lacking, however.

Reactor Power Meter type SG-8

Energy Technology Data Exchange (ETDEWEB)

Glowacki, S W

1981-01-01

The report describes the principle and electronic circuits of the Reactor Power Meter type SG-8. The gamma radiation caused by the activity of the reactor first cooling circuit affectes the ionization chamber being the detector of the instrument. The output detector signal direct current is converted into the frequency of electric pulses by means of the current-to-frequency converter. The output converter frequency is measured by the digital frequency meter: the number of measured digits in time unit is proportional to the reactor power.

A physiological model of the interaction between tissue bubbles and the formation of blood-borne bubbles under decompression

International Nuclear Information System (INIS)

Chappell, M A; Payne, S J

2006-01-01

Under decompression, bubbles can form in the human body, and these can be found both within the body tissues and the bloodstream. Mathematical models for the growth of both types of bubbles have previously been presented, but they have not been coupled together. This work thus explores the interaction between the growth of tissue and blood-borne bubbles under decompression, specifically looking at the extent to which they compete for the common resource of inert gas held in solution in the tissues. The influence of tissue bubbles is found to be significant for densities as low as 10 ml -1 for tissues which are poorly perfused. However, the effects of formation of bubbles in the blood are not found until the density of bubble production sites reaches 10 6 ml -1 . From comparison of the model predictions with experimental evidence for bubbles produced in animals and man under decompression, it is concluded that the density of tissue bubbles is likely to have a significant effect on the number of bubbles produced in the blood. However, the density of nucleation sites in the blood is unlikely to be sufficiently high in humans for the formation of bubbles in the blood to have a significant impact on the growth of the bubbles in the tissue

Power control system in BWR type reactors

International Nuclear Information System (INIS)

Nishizawa, Yasuo.

1980-01-01

Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

Thermal-hydraulic experiment for safe and stable operation of a PIUS-type reactor

International Nuclear Information System (INIS)

Tasaka, K.; Imai, S.; Masaoka, H.; Irianto, I.D.; Kohketsu, H.; Tamaki, M.; Anoda, Y.; Murata, H.; Kukita, Y.

1992-01-01

A new automatic pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed for the PIUS-type reactor. This control system maintains the fluid temperature at the axial center of the lower density lock at the average of the fluid temperatures below and above the density lock in order to prevent the poison water from penetrating into the core during normal operation. The effectiveness of this control system was successfully confirmed by a series of experiments such as start-up and power ramping tests for normal operation simulation and a loss of feedwater test for an accident condition simulation, using a small scale atmospheric pressure test loop which simulated the PIUS principle. (author)

Dynamic behaviors of cavitation bubble for the steady cavitating flow

Science.gov (United States)

Cai, Jun; Huai, Xiulan; Li, Xunfeng

2009-12-01

In this paper, by introducing the flow velocity item into the classical Rayleigh-Plesset dynamic equation, a new equation, which does not involve the time term and can describe the motion of cavitation bubble in the steady cavitating flow, has been obtained. By solving the new motion equation using Runge-Kutta fourth order method with adaptive step size control, the dynamic behaviors of cavitation bubble driven by the varying pressure field downstream of a venturi cavitation reactor are numerically simulated. The effects of liquid temperature (corresponding to the saturated vapor pressure of liquid), cavitation number and inlet pressure of venturi on radial motion of bubble and pressure pulse due to the radial motion are analyzed and discussed in detail. Some dynamic behaviors of bubble different from those in previous papers are displayed. In addition, the internal relationship between bubble dynamics and process intensification is also discussed. The simulation results reported in this work reveal the variation laws of cavitation intensity with the flow conditions of liquid, and will lay a foundation for the practical application of hydrodynamic cavitation technology.

Bubbles generated from wind-steepened breaking waves: 1. Bubble plume bubbles

NARCIS (Netherlands)

Leifer, I.; Leeuw, G. de

2006-01-01

Measurements of bubble plumes from paddle-amplified, wind stress breaking waves were made in a large wind-wave channel during the LUMINY experiment in fresh (but not clean) water. Bubble plumes exhibited considerable variability with respect to dynamics, bubble size distribution, and physical

Electron microscopy observations of helium bubble-void transition effects in nimonic PE16 alloys

International Nuclear Information System (INIS)

Mazey, D.J.; Nelson, R.S.

1980-01-01

High-nickel alloys based on the Nimonic PE16 composition have been injected at temperatures of 525 0 C and 625 0 C with 1000 ppm helium to produce a high gas-bubble concentration and subsequently irradiated with 36 MeV nickel ions. Extensive heterogeneous nucleation of bubbles is observed on faulted interstitial loops and dislocations. Evidence is found in standard PE16 alloy for bimodal bubble plus void distributions which persist during nickel-ion irradiation to 30 and 60 dpa at 625 0 C and result in a low void volume swelling of approximately 1%. The observations can be correlated with the critical bubble/void transition radius which is calculated from theory to be approximately 4.4 nm. Pre-injection of helium into a 'matrix' PE16 (low Si, Ti and Al) alloy produced an initial bubble population whose average size was above the calculated transition radius such that all bubbles eventually grew as voids during subsequent nickel-ion irradiation up to 60 dpa at 625 0 C where the void volume swelling reached approximately 12%. The observations are discussed briefly and related to theoretical predictions of the bubble/void transition radius. (author)

Reactor noise analysis of experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Ohtani, Hideji; Yamamoto, Hisashi

1980-01-01

As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

Method of operating FBR type reactors

International Nuclear Information System (INIS)

Arie, Kazuo.

1984-01-01

Purpose: To secure the controlling performance and the safety of FBR type reactors by decreasing the amount of deformation due to the difference in the heat expansion of a control rod guide tube. Method: The reactor is operated while disposing reactor core fuel assemblies of a same power at point-to-point symmetrical positions relative to the axial center for the control rod assembly. This can eliminate the temperature difference between opposing surfaces of the control rod guide tube and eliminate the difference in the thermal expansion. (Yoshino, Y.)

Thermal hydraulic test for reactor safety system; a visualization study on flow boiling and bubble behavior

Energy Technology Data Exchange (ETDEWEB)

Chang, Soon Heung; Baek, Won Pil; Ban, In Cheol [Korea Advanced Institute of Science and Technology, Taejeon (Korea)

2002-03-01

The project contribute to understand and to clarify the physical mechanism of flow nucleate boiling and CHF phenomena through the visualization experiments. the results are useful in the development of the enhancement device of heat transfer and to enhance nuclear fuel safety 1. Visual experimental facility 2. Application method of visualization Technique 3. Visualization results of flow nucleate boiling regime - Overall Bubble Behavior on the Heated Surface - Bubble Behavior near CHF Condition - Identification of Flow Structure - Three-layer flow structure 4. Quantifying of bubble parameter through a digital image processing - Image Processing Techniques - Classification of objects and measurements of the size - Three dimensional surface plot with using the luminance 5. Development and estimation of a correlation between bubble diameter and flow parameter - The effect of system parameter on bubble diameter - The development of a bubble diameter correlation . 49 refs., 42 figs., 7 tabs. (Author)

Anti-Bubbles

Science.gov (United States)

Tufaile, Alberto; Sartorelli, José Carlos

2003-08-01

An anti-bubble is a striking kind of bubble in liquid that seemingly does not comply the buoyancy, and after few minutes it disappears suddenly inside the liquid. Different from a simple air bubble that rises directly to the liquid surface, an anti-bubble wanders around in the fluid due to its slightly lesser density than the surrounding liquid. In spite of this odd behavior, an anti-bubble can be understood as the opposite of a conventional soap bubble in air, which is a shell of liquid surrounding air, and an anti-bubble is a shell of air surrounding a drop of the liquid inside the liquid. Two-phase flow has been a subject of interest due to its relevance to process equipment for contacting gases and liquids applied in industry. A chain of bubbles rising in a liquid formed from a nozzle is a two-phase flow, and there are certain conditions in which spherical air shells, called anti-bubbles, are produced. The purpose of this work is mainly to note the existence of anti-bubbling regime as a sequel of a bubbling system. We initially have presented the experimental apparatus. After this we have described the evolution of the bubbling regimes, and emulated the effect of bubbling coalescence with simple maps. Then is shown the inverted dripping as a consequence of the bubble coalescence, and finally the conditions for anti-bubble formation.

Consideration of BORAX-type reactivity accidents applied to research reactors

International Nuclear Information System (INIS)

Couturier, Jean; Meignen, Renaud; Bourgois, Thierry; Biaut, Guillaume; Mireau, Jean-Pierre; Natta, Marc

2011-01-01

Most of the research reactors discussed in this document are pool-type reactors in which the reactor vessel and some of the reactor coolant systems are located in a pool of water. These reactors generally use fuel in plate assemblies formed by a compact layer of uranium (or U 3 Si 2 ) and aluminium particles, sandwiched between two thin layers of aluminium serving as cladding. The fuel melting process begins at 660 deg. C when the aluminium melts, while the uranium (or U 3 Si 2 ) particles may remain solid. The accident that occurred in the American SL-1 reactor in 1961, together with tests carried out in the United States as of 1954 in the BORAX-1 reactor and then, in 1962, in the SPERT-1 reactor, showed that a sudden substantial addition of reactivity in this type of reactor could lead to explosive mechanisms caused by degradation, or even fast meltdown, of part of the reactor core. This is what is known as a 'BORAX-type' accident. The aim of this document is first to briefly recall the circumstances of the SL-1 reactor accident, the lessons learned, how this operational feedback has been factored into the design of various research reactors around the world and, second, to describe the approach taken by France with regard to this type of accident and how, led by IRSN, this approach has evolved in the last decade. (authors)

Processes and instruments for detecting bubbles in a medium as a liquid metal

International Nuclear Information System (INIS)

1977-01-01

This invention concerns processes and apparatuses for detecting bubbles in a medium containing them and, particularly although not exclusively, bubbles in a liquid metal used in the cooling system of a fast nuclear reactor. The process consists in seeing that a relative movement is produced between the bubbles and a receiving device, in emitting a collimated ultrasonic signal, beamed at the bubble, by means of a transmitter at a frequency equal to or greater than the resonance frequency of the bubble and in detecting a Doppler signal emitted by the bubble and received by the receiving device so as to detect the bubble. Preferably the diffusion due to the Doppler effect is such that a received diffused Doppler signal has a pulse shape having a peak amplitude proportional to the radius of the bubble and appears as a lateral asymmetrical band with respect to the ultrasonic signal. Preferably the diffusion due to the Doppler effect is brought about by the movement of the bubbles. According to another of its characteristics, the invention concerns an apparatus for detecting a bubble in a medium containing it where a relative movement is produced between the apparatus and the bubble. This apparatus includes a device for emitting an ultrasonic signal beamed at the bubble, a device for receiving an ultrasonic signal in return, a Doppler signal diffused by the bubble and a device for detecting the diffused Doppler signal received by the receiving device so as to detect the bubble [fr

Wiener's Loop Filter for PLL-Based Carrier Recovery of OQPSK and MSK-Type Modulations

Directory of Open Access Journals (Sweden)

Arnaldo Spalvieri

2008-01-01

Full Text Available This letter considers carrier recovery for offset quadrature phase shift keying (OQPSK and minimum shift keying-type (MSK-type modulations based on phase-lock loop (PLL. The concern of the letter is the optimization of the loop filter of the PLL. The optimization is worked out in the light of Wiener's theory taking into account the phase noise affecting the incoming carrier, the additive white Gaussian noise that is present on the channel, and the self-noise produced by the phase detector. Delay in the loop, which may affect the numerical implementation of the PLL, is also considered. Closed-form expressions for the loop filter and for the mean-square error are given for the case where the phase noise is characterized as a first-order process.

Improved lignin pyrolysis for phenolics production in a bubbling bed reactor--Effect of bed materials.

Science.gov (United States)

Li, Dongbing; Briens, Cedric; Berruti, Franco

2015-01-01

Lignin pyrolysis was studied in a bubbling fluidized bed reactor equipped with a fractional condensation train, using nitrogen as the fluidization gas. The effect of different bed materials (silica sand, lignin char, activated lignin char, birch bark char, and foamed glass beads) on bio-oil yield and quality was investigated for a pyrolysis temperature of 550 °C. Results how that a bed of activated lignin char is preferable to the commonly used silica sand: pyrolysis of Kraft lignin with a bed of activated lignin char not only provides a pure char product, but also a higher dry bio-oil yield (with a relative increase of 43%), lower pyrolytic water production, and better bio-oil quality. The bio-oil obtained from Kraft lignin pyrolysis with a bed of activated lignin char has a lower average molecular weight, less tar, more phenolics, and less acidity than when sand is used as bed material. Copyright © 2015 Elsevier Ltd. All rights reserved.

Bubble properties of heterogeneous bubbly flow in a square bubble column

NARCIS (Netherlands)

Bai, Wei; Deen, Niels G.; Kuipers, J.A.M.

2010-01-01

The present work focuses on the measurements of bubble properties in heterogeneous bubbly flows in a square bubble column. A four-point optical fibre probe was used for this purpose. The accuracy and intrusive effect of the optical probe was investigated first. The results show that the optical

Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Bodey, Isaac T [ORNL

2014-05-01

Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

BWR type reactor