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Sample records for reactor coolant flow

  1. Use of Distribution Devices for Hydraulic Profiling of Coolant Flow in Core Gas-cooled Reactors

    Directory of Open Access Journals (Sweden)

    A. A. Satin

    2014-01-01

    Full Text Available In setting up a reactor plant for the transportation-power module of the megawatt class an important task is to optimize the path of flow, i.e. providing moderate hydraulic resistance, uniform distribution of the coolant. Significant contribution to the hydraulic losses makes one selected design of the coolant supplies. It is, in particular, hemispherical or semi-elliptical shape of the supply reservoir, which is selected to reduce its mass, resulting in the formation of torusshaped vortex in the inlet manifold, that leads to uneven coolant velocity at the inlet into the core, the flow pulsations, hydraulic losses.To control the flow redistribution in the core according to the level of energy are used the switchgear - deflectors installed in a hemispherical reservoir supplying coolant to the fuel elements (FE of the core of gas-cooled reactor. This design solution has an effect on the structure of the flow, rate in the cooling duct, and the flow resistance of the collector.In this paper we present the results of experiments carried out on the gas dynamic model of coolant paths, deflectors, and core, comprising 55 fuel rod simulators. Numerical simulation of flow in two-parameter model, using the k-ε turbulence model, and the software package ANSYS CFX v14.0 is performed. The paper demonstrates that experimental results are in compliance with calculated ones.The results obtained suggest that the use of switchgear ensures a coolant flow balance directly at the core inlet, thereby providing temperature reduction of fuel rods with a uniform power release in the cross-section. Considered options to find constructive solutions for deflectors give an idea to solve the problem of reducing hydraulic losses in the coolant paths, to decrease pulsation components of flow in the core and length of initial section of flow stabilization.

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

    Science.gov (United States)

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

    2017-07-01

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

  3. Reactor coolant pump flywheel

    Science.gov (United States)

    Finegan, John Raymond; Kreke, Francis Joseph; Casamassa, John Joseph

    2013-11-26

    A flywheel for a pump, and in particular a flywheel having a number of high density segments for use in a nuclear reactor coolant pump. The flywheel includes an inner member and an outer member. A number of high density segments are provided between the inner and outer members. The high density segments may be formed from a tungsten based alloy. A preselected gap is provided between each of the number of high density segments. The gap accommodates thermal expansion of each of the number of segments and resists the hoop stress effect/keystoning of the segments.

  4. Experimental studies into the fluid dynamic performance of the coolant flow in the mixed core of the Temelin NPP VVER-1000 reactor

    Directory of Open Access Journals (Sweden)

    S.M. Dmitriev

    2015-11-01

    Full Text Available The paper presents the results of studies into the interassembly coolant interaction in the Temelin nuclear power plant (NPP VVER-1000 reactor core. An aerodynamic test bench was used to study the coolant flow processes in a TVSA-type fuel assembly bundle. To obtain more detailed information on the coolant flow dynamics, a VVER-1000 reactor core fragment was selected as the test model, which comprised two segments of a TVSA-12 PLUS fuel assembly and one segment of a TVSA-T assembly with stiffening angles and an interassembly gap. The studies into the coolant fluid dynamics consisted in measuring the velocity vector both in representative TVSA regions and inside the interassembly gap using a five-channel pneumometric probe. An analysis into the spatial distribution of the absolute flow velocity projections made it possible to detail the TVSA spacer, mixing and combined spacer grid flow pattern, identify the regions with the maximum transverse coolant flow, and determine the depth of the coolant flow disturbance propagation and redistribution in adjacent TVSA assemblies. The results of the studies into the interassembly coolant interaction among the adjacent TVSA assemblies are used at OKBM Afrikantov to update the VVER-1000 core thermal-hydraulic analysis procedures and have been added to the database for verification of computational fluid dynamics (CFD codes and for detailed cellwise analyses of the VVER-100 reactor cores.

  5. Combined numerical and experimental investigation into the coolant flow hydrodynamics and mass transfer behind the spacer grid in fuel assemblies of the floating power unit reactor

    Directory of Open Access Journals (Sweden)

    S.M. Dmitriev

    2015-09-01

    Full Text Available The results of experimental investigations into the local hydrodynamics and inter-cell mass transfer of the coolant flow in representative zones of the KLT-40C reactor FAs behind the plate-type spacer grid are presented. The investigations were conducted on an aerodynamic rig using the admixture diffusion method (the tracer-gas method. A study into the spatial dispersion of the absolute flow velocity projections and into the distribution of the tracer concentration allowed specify the coolant flow pattern behind the FA plate-type spacer grid of the KLT-40C reactor. The results of measuring the flow friction coefficient in the plate-type spacer grid, depending on the Reynolds number, are presented. Based on the obtained experimental data, recommendations have been provided for updating the procedures to calculate the coolant flow rates for the KLT-40C reactor core by-channel codes. The results of investigating the coolant flow local hydrodynamics and mass transfer in the KLT-40C reactor FAs have been adopted for practical use by Afrikantov OKBM for estimating the heat-engineering reliability of the KLT-40C reactor cores and have been data based for verification of CFD codes and detailed by-channel calculation of the KLT-40C reactor core.

  6. PARTICLE IMAGE VELOCIMETRY MEASUREMENTS IN A REPRESENTATIVE GAS-COOLED PRISMATIC REACTOR CORE MODEL: FLOW IN THE COOLANT CHANNELS AND INTERSTITIAL BYPASS GAPS

    Energy Technology Data Exchange (ETDEWEB)

    Thomas E. Conder; Richard Skifton; Ralph Budwig

    2012-11-01

    Core bypass flow is one of the key issues with the prismatic Gas Turbine-Modular Helium Reactor, and it refers to the coolant that navigates through the interstitial, non-cooling passages between the graphite fuel blocks instead of traveling through the designated coolant channels. To determine the bypass flow, a double scale representative model was manufactured and installed in the Matched Index-of-Refraction flow facility; after which, stereo Particle Image Velocimetry (PIV) was employed to measure the flow field within. PIV images were analyzed to produce vector maps, and flow rates were calculated by numerically integrating over the velocity field. It was found that the bypass flow varied between 6.9-15.8% for channel Reynolds numbers of 1,746 and 4,618. The results were compared to computational fluid dynamic (CFD) pre-test simulations. When compared to these pretest calculations, the CFD analysis appeared to under predict the flow through the gap.

  7. Chimney for enhancing flow of coolant water in natural circulation boiling water reactor

    Science.gov (United States)

    Oosterkamp, Willem Jan; Marquino, Wayne

    1999-01-05

    A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereat access to the fuel assemblies is not obstructed.

  8. Numerical experimentation on convective coolant flow in Ghana ...

    African Journals Online (AJOL)

    Numerical experiments on one dimensional convective coolant flow during steady state operation of the Ghana Research Reactor-1 (GHARR-I) were performed to determine the thermal hydraulic parameters of temperature, density and flow rate. The computational domain was the reactor vessel, including the reactor core.

  9. NGNP Reactor Coolant Chemistry Control Study

    Energy Technology Data Exchange (ETDEWEB)

    Brian Castle

    2010-11-01

    The main focus of this paper is to identify the most desirable ranges of impurity levels in the primary coolant to optimize component life in the primary circuit of the Next Generation Nuclear Plant (NGNP), which will either be a prismatic block or pebble bed reactor.

  10. Application of thermal-hydraulic model of RBMK reactor fuel channel to correction of power and coolant flow measurements

    Science.gov (United States)

    Zagrebaev, A. M.; Trifonenkov, A. V.; Trifonenkov, V. P.

    2017-12-01

    The effectiveness and the security of RBMK reactor operation depends on the accuracy of the control over reactor's parameters and their limitations. The processing of operational parameters archive helps to adjust different mathematical models and significantly widen their field of use. Pressure differential between common pressure header and steam separator is the sum of calculated pressure differential and friction loss on flow control valve. There is known mathematical software, which allows to adapt such model for each fuel channel using the archive. In this research it is suggested not to replace the regular mechanism with such approach, but to use the adapted mathematical model to calculate corrected values of power and flow, which were measured by regular means. Mathematical expressions and procedures for such approach are given.

  11. Corrosion of structural materials by lead-based reactor coolants.

    Energy Technology Data Exchange (ETDEWEB)

    Abraham, D. P.; Leibowitz, L.; Maroni, V. A.; McDeavitt, S. M.; Raraz, A. G.

    2000-11-16

    Advanced nuclear reactor design has, in recent years, focused increasingly on the use of heavy-liquid-metal coolants, such as lead and lead-bismuth eutectic. Similarly, programs on accelerator-based transmutation systems have also considered the use of such coolants. Russian experience with heavy-metal coolants for nuclear reactors has lent credence to the validity of this approach. Of significant concern is the compatibility of structural materials with these coolants. We have used a thermal convection-based test method to allow exposure of candidate materials to molten lead and lead-bismuth flowing under a temperature gradient. The gradient was deemed essential in evaluating the behavior of the test materials in that should preferential dissolution of components of the test material occur we would expect dissolution in the hotter regions and deposition in the colder regions, thus promoting material transport. Results from the interactions of a Si-rich mild steel alloy, AISI S5, and a ferritic-martensitic stainless steel, HT-9, with the molten lead-bismuth are presented.

  12. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T.; Grunwald, G.

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  13. Nuclear reactor downcomer flow deflector

    Science.gov (United States)

    Gilmore, Charles B [Greensburg, PA; Altman, David A [Pittsburgh, PA; Singleton, Norman R [Murrysville, PA

    2011-02-15

    A nuclear reactor having a coolant flow deflector secured to a reactor core barrel in line with a coolant inlet nozzle. The flow deflector redirects incoming coolant down an annulus between the core barrel and the reactor vessel. The deflector has a main body with a front side facing the fluid inlet nozzle and a rear side facing the core barrel. The rear side of the main body has at least one protrusion secured to the core barrel so that a gap exists between the rear side of the main body adjacent the protrusion and the core barrel. Preferably, the protrusion is a relief that circumscribes the rear side of the main body.

  14. Flow boiling test of GDP replacement coolants

    Energy Technology Data Exchange (ETDEWEB)

    Park, S.H. [comp.

    1995-08-01

    The tests were part of the CFC replacement program to identify and test alternate coolants to replace CFC-114 being used in the uranium enrichment plants at Paducah and Portsmouth. The coolants tested, C{sub 4}F{sub 10} and C{sub 4}F{sub 8}, were selected based on their compatibility with the uranium hexafluoride process gas and how well the boiling temperature and vapor pressure matched that of CFC-114. However, the heat of vaporization of both coolants is lower than that of CFC-114 requiring larger coolant mass flow than CFC-114 to remove the same amount of heat. The vapor pressure of these coolants is higher than CFC-114 within the cascade operational range, and each coolant can be used as a replacement coolant with some limitation at 3,300 hp operation. The results of the CFC-114/C{sub 4}F{sub 10} mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C{sub 4}F{sub 10} weight mixture in CFC-114 and the degree of degradation is about 20% from that of CFC-114 boiling heat transfer coefficient. This report consists of the final reports from Cudo Technologies, Ltd.

  15. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems required...

  16. Full reactor coolant system chemical decontamination qualification programs

    Energy Technology Data Exchange (ETDEWEB)

    Miller, P.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1995-03-01

    Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systems leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.

  17. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    Energy Technology Data Exchange (ETDEWEB)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

  18. Application of damage function analysis to reactor coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, D.D. [Center for Electrochemical Science and Technology, Pennsylvania State Univ., University Park, PA (United States)

    2002-07-01

    The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

  19. Expert system for online surveillance of nuclear reactor coolant pumps

    Energy Technology Data Exchange (ETDEWEB)

    Gross, K.C.; Singer, R.M.; Humenik, K.E.

    1992-12-31

    This report describes an expert system for online surveillance of nuclear reactor coolant pumps. This system provides a means for early detection of pump or sensor degradation. Degradation is determined through the use of a statistical analysis technique, sequential probability ratio test, applied to information from several sensors which are responsive to differing physical parameters. The results of sequential testing of the data provide the operator with an early warning of possible sensor or pump failure.

  20. Apparatus for suppressing formation of vortices in the coolant fluid of a nuclear reactor and associated method

    Science.gov (United States)

    Ekeroth, Douglas E.; Garner, Daniel C.; Hopkins, Ronald J.; Land, John T.

    1993-01-01

    An apparatus and method are provided for suppressing the formation of vortices in circulating coolant fluid of a nuclear reactor. A vortex-suppressing plate having a plurality of openings therein is suspended within the lower plenum of a reactor vessel below and generally parallel to the main core support of the reactor. The plate is positioned so as to intersect vortices which may form in the circulating reactor coolant fluid. The intersection of the plate with such vortices disrupts the rotational flow pattern of the vortices, thereby disrupting the formation thereof.

  1. EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2016-01-01

    Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

  2. Thermohydraulic modeling of very high temperature reactors in regimes with loss of coolant using CFD

    Energy Technology Data Exchange (ETDEWEB)

    Moreira, Uebert G.; Dominguez, Dany S. [Universidade Estadual de Santa Cruz (UESC), Ilh´eus, BA (Brazil). Programa de P´os-Graduacao em Modelagem Computacional em Ciencia e Tecnologia; Mazaira, Leorlen Y.R.; Lira, Carlos A.B.O. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear; Hernandez, Carlos R.G., E-mail: uebert.gmoreira@gmail.com, E-mail: dsdominguez@gmail.com, E-mail: leored1984@gmail.com, E-mail: cabol@ufpe.br, E-mail: cgh@instec.cu [Instituto Superior de Tecnologas y Ciencias Aplicadas (InSTEC), La Habana (Cuba)

    2017-07-01

    The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this perspective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core, and advances in computational capacity, CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core, which each pebble bed element is modeled in detail. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Previous works indicate this arrangement as the configuration that obtain higher fuel temperatures inside the core. Parameters considered for reactor design are available in the technical report of benchmark issues by IAEA (TECDOC-1694). Among the results obtained, we obtained the temperature profiles with different mass flow rates for the coolant. In general, the temperature distributions calculated are consistent with phenomenological behaviour. Even without consider the reactivity changes to reduce the reactor power or other safety procedures, the maximum temperatures do not exceed the recommended limits for fuel elements. (author)

  3. A Study on thermal-hydraulic characteristics of the coolant materials for the transmutation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hyun; You, Young Woo; Cho, Jae Seon; Kim, Ju Youl; Kim, Do Hyoung; Kim, Yoon Ik; Yang, Hui Chang [Seoul National University, Taejon (Korea)

    1998-03-01

    The objective of this study is to provide the direction of transmutation reactor design in terms of thermal hydraulics especially through the analysis of thermal hydraulic characteristics of various candidate materials for the transmutation reactor coolant. In this study, the characteristics of coolant materials used in current nuclear power plants and candidate materials for transmutation reactor are analyzed and compared. To evaluate the thermal hydraulic characteristics, the preliminary thermal-hydraulic calculation is performed for the candidate coolant materials of transmutation reactor. An analysis of thermal-hydraulic characteristics of transmutation reactor. An analysis of thermal-hydraulic characteristics of Sodium, Lead, Lead-Bismuth, and Lead-Lithium among the liquid metals considered as the coolant of transmutation reactor is performed by using computational fluid dynamics code FLUENT, and SIMPLER algorithm. (author). 50 refs., 40 figs., 30 tabs.

  4. Cracked shaft detection on large vertical nuclear reactor coolant pump

    Science.gov (United States)

    Jenkins, L. S.

    1985-01-01

    Due to difficulty and radiation exposure associated with examination of the internals of large commercial nuclear reactor coolant pumps, it is necessary to be able to diagnose the cause of an excessive vibration problem quickly without resorting to extensive trial and error efforts. Consequently, it is necessary to make maximum use of all available data to develop a consistent theory which locates the problem area in the machine. This type of approach was taken at Three Mile Island, Unit #1, in February 1984 to identify and locate the cause of a continuously climbing vibration level of the pump shaft. The data gathered necessitated some in-depth knowledge of the pump internals to provide proper interpretation and avoid misleading conclusions. Therefore, the raw data included more than just the vibration characteristics. Pertinent details of the data gathered is shown and is necessary and sufficient to show that the cause of the observed vibration problem could logically only be a cracked pump shaft in the shaft overhang below the pump bearing.

  5. Comparison of THALES and VIPRE-01 Subchannel Codes for Loss of Flow and Single Reactor Coolant Pump Rotor Seizure Accidents using Lumped Channel APR1400 Geometry

    Energy Technology Data Exchange (ETDEWEB)

    Oezdemir, Erdal; Moon, Kang Hoon; Oh, Seung Jong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); Kim, Yongdeog [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    Subchannel analysis plays important role to evaluate safety critical parameters like minimum departure from nucleate boiling ratio (MDNBR), peak clad temperature and fuel centerline temperature. In this study, two different subchannel codes, VIPRE-01 (Versatile Internals and Component Program for Reactors: EPRI) and THALES (Thermal Hydraulic AnaLyzer for Enhanced Simulation of core) are examined. In this study, two different transient cases for which MDNBR result play important role are selected to conduct analysis with THALES and VIPRE-01 subchannel codes. In order to get comparable results same core geometry, fuel parameters, correlations and models are selected for each code. MDNBR results from simulations by both code are agree with each other with negligible difference. Whereas, simulations conducted by enabling conduction model in VIPRE-01 shows significant difference from the results of THALES.

  6. Numerical and experimental investigation of surface vortex formation in coolant reservoirs of reactor safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Pandazis, Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Babcsany, Boglarka [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2016-11-15

    The reliable operation of the emergency coolant pumps and passive gravitational injection systems are an important safety issue during accident scenarios with coolant loss in pressurized water reactors. Because of the pressure drop and flow disturbances surface vortices develops at the pump intakes if the water level decreasing below a critical value. The induced swirling flow and gas entrainment lead to flow limitation and to pump failures and damages. The prediction of the critical submergence to avoid surface vortex building is difficult because it depends on many geometrical and fluid dynamical parameters. An alternative and new method has been developed for the investigation of surface vortices. The method based on the combination of CFD results with the analytical vortex model of Burgers and Rott. For further investigation the small scale experiments from the Institute of Nuclear Techniques of the Budapest University of Technology and Economics are used which were inspired from flow limitation problems during the draining of the bubble condenser trays at a VVER type nuclear power plants.

  7. Experimental and CFD Studies of Coolant Flow Mixing within Scaled Models of the Upper and Lower Plenums of NGNP Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Anand, Nk [Texas A & M Univ., College Station, TX (United States)

    2016-03-30

    A 1/16th scaled VHTR experimental model was constructed and the preliminary test was performed in this study. To produce benchmark data for CFD validation in the future, the facility was first run at partial operation with five pipes being heated. PIV was performed to extract the vector velocity field for three adjacent naturally convective jets at statistically steady state. A small recirculation zone was found between the pipes, and the jets entered the merging zone at 3 cm from the pipe outlet but diverged as the flow approached the top of the test geometry. Turbulence analysis shows the turbulence intensity peaked at 41-45% as the jets mixed. A sensitivity analysis confirmed that 1000 frames were sufficient to measure statistically steady state. The results were then validated by extracting the flow rate from the PIV jet velocity profile, and comparing it with an analytic flow rate and ultrasonic flowmeter; all flow rates lie within the uncertainty of the other two methods for Tests 1 and 2. This test facility can be used for further analysis of naturally convective mixing, and eventually produce benchmark data for CFD validation for the VHTR during a PCC or DCC accident scenario. Next, a PTV study of 3000 images (1500 image pairs) were used to quantify the velocity field in the upper plenum. A sensitivity analysis confirmed that 1500 frames were sufficient to precisely estimate the flow. Subsequently, three (3, 9, and 15 cm) Y-lines from the pipe output were extracted to consider the output differences between 50 to 1500 frames. The average velocity field and standard deviation error that accrued in the three different tests were calculated to assess repeatability. The error was varied, from 1 to 14%, depending on Y-elevation. The error decreased as the flow moved farther from the output pipe. In addition, turbulent intensity was calculated and found to be high near the output. Reynolds stresses and turbulent intensity were used to validate the data by

  8. Production test IP-750 raw water as a reactor coolant. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Frymier, J.W.; Geier, R.G.

    1966-08-10

    Approximately ten years ago single-tube tests demonstrated the feasibility of using unfiltered river water as a reactor coolant from the standpoint of aluminum corrosion and film formation. However, some effluent activity penalty was indicated. Inasmuch as both current water plant operation and the characteristics of Columbia River water have changed, it was deemed appropriate to reinvestigate the use of partially treated water as a reactor coolant. This report summarizes the results of a half-reactor test carried out at F Reactor.

  9. Reactor coolant pump testing using motor current signatures analysis

    Energy Technology Data Exchange (ETDEWEB)

    Burstein, N.; Bellamy, J.

    1996-12-01

    This paper describes reactor coolant pump motor testing carried out at Florida Power Corporation`s Crystal River plant using Framatome Technologies` new EMPATH (Electric Motor Performance Analysis and Trending Hardware) system. EMPATH{trademark} uses an improved form of Motor Current Signature Analysis (MCSA), technology, originally developed at Oak Ridge National Laboratories, for detecting deterioration in the rotors of AC induction motors. Motor Current Signature Analysis (MCSA) is a monitoring tool for motor driven equipment that provides a non-intrusive means for detecting the presence of mechanical and electrical abnormalities in the motor and the driven equipment. The base technology was developed at the Oak Ridge National Laboratory as a means for determining the affects of aging and service wear specifically on motor-operated valves used in nuclear power plant safety systems, but it is applicable to a broad range of electric machinery. MCSA is based on the recognition that an electric motor (ac or dc) driving a mechanical load acts as an efficient and permanently available transducer by sensing mechanical load variations, large and small, long-term and rapid, and converting them into variations in the induced current generated in the motor windings. The motor current variations, resulting from changes in load caused by gears, pulleys, friction, bearings, and other conditions that may change over the life of the motor, are carried by the electrical cables powering the motor and are extracted at any convenient location along the motor lead. These variations modulate the 60 Hz carrier frequency and appear as sidebands in the spectral plot.

  10. Radionuclides in primary coolant of a fluoride salt-cooled high-temperature reactor during normal operation

    National Research Council Canada - National Science Library

    Zhang, Guo-Qing; Wang, Shuai; Zhang, Hai-Qing; Zhu, Xing-Wang; Peng, Chao; Cai, Jun; He, Zhao-Zhong; Chen, Kun

    2017-01-01

    The release of fission products from coated particle fuel to primary coolant, as well as the activation of coolant and impurities, were analysed for a fluoride salt-cooled high-temperature reactor (FHR...

  11. Liquid metal reactor development. Development of LMR coolant technology

    Energy Technology Data Exchange (ETDEWEB)

    Nam, H. Y.; Choi, S. K.; Hwang, J. s.; Lee, Y. B.; Choi, B. H.; Kim, J. M.; Kim, Y. G.; Kim, M. J.; Lee, S. D.; Kang, Y. H.; Maeng, Y. Y.; Kim, T. R.; Park, J. H.; Park, S. J.; Cha, J. H.; Kim, D. H.; Oh, S. K.; Park, C. G.; Hong, S. H.; Lee, K. H.; Chun, M. H.; Moon, H. T.; Chang, S. H.; Lee, D. N.

    1997-07-15

    Following studies have been performed during last three years as the 1.2 phase study of the mid and long term nuclear technology development plan. First, the small scale experiments using the sodium have been performed such as the basic turbulent mixing experiment which is related to the design of a compact reactor, the flow reversal characteristics experiment by natural circulation which is necessary for the analysis of local flow reversal when the electromagnetic pump is installed, the feasibility test of the decay heat removal by wall cooling and the operation of electromagnetic pump. Second, the technology of operation mechanism of sodium facility is developed and the technical analysis and fundamental experiments of sodium measuring technology has been performed such as differential pressure measuring experiment, local flow rate measuring experimenter, sodium void fraction measuring experiment, under sodium facility, the free surface movement experiment and the side orifice pressure drop experiment. A new bounded convection scheme was introduced to the ELBO3D thermo-hydraulic computer code designed for analysis of experimental result. A three dimensional computer code was developed for the analysis of free surface movement and the analysis model of transmission of sodium void fraction was developed. Fourth, the small scale key components are developed. The submersible-in-pool type electromagnetic pump which can be used as primary pump in the liquid metal reactor is developed. The SASS which uses the Curie-point electromagnet and the mock-up of Pantograph type IVTM were manufactured and their feasibility was evaluated. Fifth, the high temperature characteristics experiment of stainless steel which is used as a major material for liquid metal reactor and the material characteristics experiment of magnet coil were performed. (author). 126 refs., 98 tabs., 296 figs.

  12. Estimation of the coolant flow through a natural circulation BWR fuel channel applying and equivalent electrical model

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, DEPFI, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Campus Morelos en IMTA, Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: julfi_ig@yahoo.com.m [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2010-10-15

    This work presents the design and implementation of an advanced controller for a reduced order model of a BWR reactor core cooled by natural circulating water, which allows real time estimates of coolant flows through fuel assemblies about standard neutron flux strings. Nuclear power plants with boiling water reactors control individual fuel assembly coolant flows by forced circulation using external or internal water pumps and different core support plate orifices. These two elements reduce flow dependency on local channel pressure drops. In BWR reactors using only natural circulation coolant flows, these two elements are not available and therefore individual channel coolant flows are highly dependent in local conditions, such as power distributions and local pressure drops. Therefore it is expected that grater uncertainties in these variables be used during safety, fuel management and other analysis, which in turns may lead to increased operation penalties, such as tighter operating limits. The objective of this work is to asses by computer simulations means to reduce uncertainties in the measurement of fuel assembly coolant flows and eventually the associated penalties. During coolant phase transitions, pressure drops and local power may alter local natural circulation through fuel assemblies and flow estimates can be helped or not by control rod moves. This work presents the construction of an optimal controller for a core flow estimator based on a reduced order model of the coolant going though the reactor vessel components and nuclear core. This model is to be driven by plant signals from standard BWR instrumentation in order to estimate the coolant flows in selected fuel assemblies about a LPRM string. For this purpose an equivalent electrical model has been mathematically developed and numerically tested. The power-flow maps of typical BRW are used as steady state references for this equivalent model. Once these were fully reproduced for steady state

  13. ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) simulation of a loss of coolant accident in a space reactor

    Energy Technology Data Exchange (ETDEWEB)

    Roth, P.A.; Shumway, R.W.

    1988-01-01

    The Advanced Thermal Hydraulic Energy Network Analyzer (ATHENA) code was used to simulate a loss-of-coolant accident (LOCA) in a conceptual space reactor design. ATHENA provides the capability of simulating the thermal-hydraulic behavior of the wide variety of systems which are being considered for use in space reactors. Flow loops containing any one of several available working fluids may interact through thermal connections with other loops containing the same or a different working fluid. The code can be used to model special systems such as: heat pipes, point reactor kinetics, plant control systems, turbines, valves, and pumps. This work demonstrates the application of the thermal radiation model which has been recently incorporated into ATHENA and verifies the need for supplemental reactor cooling to prevent reactor fuel damage in the event of a LOCA.

  14. Conceptual design of main coolant pump for integral reactor SMART

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Seok; Kim, Jong In; Kim, Min Hwan [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-12-01

    The conceptual design for MCP to be installed in the integral reactor SMART was carried out. Canned motor pump was adopted in the conceptual design of MCP. Three-dimensional modeling was performed to visualize the conceptual design of the MCP and to check interferences between the parts. The theoretical design procedure for the impeller was developed. The procedures for the flow field and structural analysis of impeller was also developed to assess the design validity and to verify its structural integrity. A computer program to analyze the dynamic characteristics of the rotor shaft of MCP was developed. The rotational speed sensor was designed and its performance test was conducted to verify the possibility of operation. A prototypes of the canned motor was manufactured and tested to confirm the validity of the design concept. The MCP design concept was also investigated for fabricability by establishing the manufacturing procedures. 41 refs., 96 figs., 10 tabs. (Author)

  15. Simulating the corrosion of zirconium alloys in the water coolant of VVER reactors

    Science.gov (United States)

    Kritskii, V. G.; Berezina, I. G.; Motkova, E. A.

    2013-07-01

    A model for predicting the corrosion of cladding zirconium alloys depending on their composition and operating conditions is proposed. Laws of thermodynamics and chemical kinetics of the reactions through which the multicomponent zirconium alloy is oxidized in the reactor coolant constitute the physicochemical heart of the model. The developed version of the model is verified against the results obtained from tests of fuel rod claddings made of commercial-grade and experimental zirconium alloys carried out by different researchers under autoclave and reactor conditions. It is shown that the proposed model adequately describes the corrosion of alloys in coolants used at nuclear power stations. It is determined that, owing to boiling of coolant and its acidification in a VVER-1200 reactor, Zr-1% Nb alloys with additions of iron and oxygen must be more resistant to corrosion than the commercial-grade alloy E110.

  16. Oscillatory flow chemical reactors

    National Research Council Canada - National Science Library

    Slavnić Danijela S; Bugarski Branko M; Nikačević Nikola M

    2014-01-01

    .... However, the reactions of interests for the mentioned industry sectors are often slow, thus continuous tubular reactors would be impractically long for flow regimes which provide sufficient heat...

  17. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

    Energy Technology Data Exchange (ETDEWEB)

    Kryk, Holger, E-mail: h.kryk@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Hoffmann, Wolfgang [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany)

    2014-12-15

    Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products.

  18. Experimental study of local coolant hydrodynamics in TVS-Kvadrat PWR reactor fuel assembly using mixing spacer grids with different types of deflectors

    Directory of Open Access Journals (Sweden)

    S.M. Dmitriev

    2015-12-01

    Full Text Available Results of experimental studies of local hydrodynamic characteristics of coolant flow in fuel assemblies of RWR reactors using different types of mixing spacer grids are presented. Specific features and regularities of coolant flow in fuel pin bundles of TVS-KVADRAT fuel assemblies with different types of mixing spacer grids were revealed in the course of experiments. Analysis of space distribution of projections of absolute flow velocity allowed detailed description of coolant flow beyond the spacer grid with installation of three different types of deflectors. Optimal design of deflector for spacer grid of the TVS-KVADRAT fuel assembly in the standard cell in the area of guiding channels was identified. Results of studies of local hydrodynamics of coolant flow in the TVS-KVADRAT fuel assembly are accepted for subsequent practical application by the JSC Afrikantov Experimental Design Bureau for Mechanical Engineering (OKBM in the evaluations of thermal engineering reliability of PWR reactor cores and were included in the database for verification of computational fluid dynamic codes (CFD-codes and implementation of detailed cell array calculations of PWR reactor cores.

  19. Fuel-coolant interaction (FCI) phenomena in reactor safety. Current understanding and future research needs

    Energy Technology Data Exchange (ETDEWEB)

    Speis, T.P. [Maryland Univ., College Park, MD (United States); Basu, S.

    1998-01-01

    This paper gives an account of the current understanding of fuel-coolant interaction (FCI) phenomena in the context of reactor safety. With increased emphasis on accident management and with emerging in-vessel core melt retention strategies for advanced light water reactor (ALWR) designs, recent interest in FCI has broadened to include an evaluation of potential threats to the integrity of reactor vessel lower head and ex-vessel structural support, as well as the role of FCI in debris quenching and coolability. The current understanding of FCI with regard to these issues is discussed, and future research needs to address the issues from a risk perspective are identified. (author)

  20. Nuclear reactor flow control method and apparatus

    Science.gov (United States)

    Church, J.P.

    1993-03-30

    Method and apparatus for improving coolant flow in a nuclear reactor during accident as well as nominal conditions. The reactor has a plurality of fuel elements in sleeves and a plenum above the fuel and through which the sleeves penetrate. Holes are provided in the sleeve so that coolant from the plenum can enter the sleeve and cool the fuel. The number and size of the holes are varied from sleeve to sleeve with the number and size of holes being greater for sleeves toward the center of the core and less for sleeves toward the periphery of the core. Preferably the holes are all the same diameter and arranged in rows and columns, the rows starting from the bottom of every sleeve and fewer rows in peripheral sleeves and more rows in the central sleeves.

  1. Use of Nitrogen Trifluoride To Purify Molten Salt Reactor Coolant and Heat Transfer Fluoride Salts

    Energy Technology Data Exchange (ETDEWEB)

    Scheele, Randall D.; Casella, Andrew M.; McNamara, Bruce K.

    2017-05-02

    Abstract: The molten salt cooled nuclear reactor is included as one of the Generation IV reactor types. One of the challenges with the implementation of this reactor is purifying and maintaining the purity of the various molten fluoride salts that will be used as coolants. The method used for Oak Ridge National Laboratory’s molten salt experimental test reactor was to treat the coolant with a mixture of H2 and HF at 600°C. In this article we evaluate thermal NF3 treatment for purifying molten fluoride salt coolant candidates based on NF3’s 1) past use to purify fluoride salts, 2) other industrial uses, 3) commercial availability, 4) operational, chemical, and health hazards, 5) environmental effects and environmental risk management methods, 6) corrosive properties, and 7) thermodynamic potential to eliminate impurities that could arise due to exposure to water and oxygen. Our evaluation indicates that nitrogen trifluoride is a viable and safer alternative to the previous method.

  2. Conceptual design of the integral test loop (I): Reactor coolant system and secondary system

    Energy Technology Data Exchange (ETDEWEB)

    Song, Chul Hwa; Lee, Seong Je; Kwon, Tae Soon; Moon, Sang Ki [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    This report describes the conceptual design of the primary coolant system and the secondary system of the Integral Test Loop (ITL) which simulates overall thermal hydraulic phenomena of the primary system of a nuclear power plant during postulated accidents or transients. The design basis for the primary coolant system and secondary system is as follows ; Reference plant: Korean Standard Nuclear Plant (KSNP), Height ratio : 1/1, Volume ratio : 1/200, Power scale : Max. 15% of the scaled nominal power, Temperature, Pressure : Real plant conditions. The primary coolant system includes a reactor vessel, which contains a core simulator, a steam generator, a reactor coolant pump simulator, a pressurizer and piping, which consists of two hot legs, four cold legs and four intermediate legs. The secondary system consists of s steam discharge system, a feedwater supply system and a steam condensing system. This conceptual design report describes general configuration of the reference plant, and major function and operation of each system of the plant. Also described is the design philosophy of each component and system of the ITL, and specified are the design criteria and technical specifications of each component and system of the ITL in the report. 17 refs., 43 figs., 51 tabs. (Author)

  3. Radiogenic lead as coolant, reflector and moderator in advanced fast reactors

    Science.gov (United States)

    Kulikov, E. G.

    2017-01-01

    Main purpose of the study is assessing reasonability for recovery, production and application of radiogenic lead as a coolant, neutron moderator and neutron reflector in advanced fast reactors. When performing the study, thermal, physical and neutron-physical properties of natural and radiogenic lead were analyzed. The following results were obtained: 1. Radiogenic lead with high content of isotope 208Pb can be extracted from thorium or mixed thorium-uranium ores because 208Pb is a final product of 232Th natural decay chain. 2. The use of radiogenic lead with high 208Pb content in advanced fast reactors and accelerator-driven systems (ADS) makes it possible to improve significantly their neutron-physical and thermal-hydraulic parameters. 3. The use of radiogenic lead with high 208Pb content in advanced fast reactors as a coolant opens the possibilities for more intense fuel breeding and for application of well-known oxide fuel instead of the promising but not tested enough nitride fuel under the same safety parameters. 4. The use of radiogenic lead with high 208Pb content in ADS as a coolant can upgrade substantially the level of neutron flux in the ADS blanket, which enables effective transmutation of radioactive wastes with low cross-sections of radiative neutron capture.

  4. Oscillatory flow chemical reactors

    Directory of Open Access Journals (Sweden)

    Slavnić Danijela S.

    2014-01-01

    Full Text Available Global market competition, increase in energy and other production costs, demands for high quality products and reduction of waste are forcing pharmaceutical, fine chemicals and biochemical industries, to search for radical solutions. One of the most effective ways to improve the overall production (cost reduction and better control of reactions is a transition from batch to continuous processes. However, the reactions of interests for the mentioned industry sectors are often slow, thus continuous tubular reactors would be impractically long for flow regimes which provide sufficient heat and mass transfer and narrow residence time distribution. The oscillatory flow reactors (OFR are newer type of tube reactors which can offer solution by providing continuous operation with approximately plug flow pattern, low shear stress rates and enhanced mass and heat transfer. These benefits are the result of very good mixing in OFR achieved by vortex generation. OFR consists of cylindrical tube containing equally spaced orifice baffles. Fluid oscillations are superimposed on a net (laminar flow. Eddies are generated when oscillating fluid collides with baffles and passes through orifices. Generation and propagation of vortices create uniform mixing in each reactor cavity (between baffles, providing an overall flow pattern which is close to plug flow. Oscillations can be created by direct action of a piston or a diaphragm on fluid (or alternatively on baffles. This article provides an overview of oscillatory flow reactor technology, its operating principles and basic design and scale - up characteristics. Further, the article reviews the key research findings in heat and mass transfer, shear stress, residence time distribution in OFR, presenting their advantages over the conventional reactors. Finally, relevant process intensification examples from pharmaceutical, polymer and biofuels industries are presented.

  5. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Thermo- and fluid-dynamic effects

    Energy Technology Data Exchange (ETDEWEB)

    Seeliger, André, E-mail: a.seeliger@hszg.de [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Alt, Sören; Kästner, Wolfgang; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Kryk, Holger; Harm, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany)

    2016-08-15

    Highlights: • Borated coolant supports corrosion at zinc-coated installations in PWR after LOCA. • Dissolved zinc is injected into core by ECCS during sump recirculation phase. • Corrosion products can reach and settle at further downstream components. • Corrosion products can cause head losses at spacers and influence decay heat removal. • Preventive procedures were tested at semi-technical scale facilities. - Abstract: Within the framework of the German reactor safety research, generic experimental investigations were carried out aiming at thermal-hydraulic consequences of physicochemical mechanisms, caused by dissolution of zinc in boric acid during corrosion processes at hot-dip galvanized surfaces of containment internals at lower coolant temperatures and the subsequent precipitation of solid zinc borates in PWR core regions of higher temperature. This constellation can occur during sump recirculation operation of ECCS after LOCA. Hot-dip galvanized compounds, which are installed inside a PWR containment, may act as zinc sources. Getting in contact with boric acid coolant, zinc at their surfaces is released into coolant in form of ions due to corrosion processes. As a long-term behavior resp. over a time period of several days, metal layers of zinc and zinc alloys can dissolve extensively. First fundamental studies at laboratory scale were done at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR). Their experimental results were picked up for the definition of boundary conditions for experiments at semi-technical scale at the Hochschule Zittau/Görlitz (HSZG). Electrical heating rods with zircaloy cladding tubes have been used as fuel rod simulators. As near-plant core components, a 3 × 3 configuration of heating rods (HRC) and a shortened, partially heatable PWR fuel assembly dummy were applied into cooling circuits. The HRC module includes segments of spacers for a suitable representation of a heating channel geometry. Formations of different solid

  6. Investigation of a Coolant Mixing Phenomena within the Reactor Pressure Vessel of a VVER-1000 Reactor with Different Simulation Tools

    Directory of Open Access Journals (Sweden)

    V. Sánchez

    2010-01-01

    Full Text Available The Institute of Neutron Physics and Reactor Technology (INR is involved in the qualification of coupled codes for reactor safety evaluations, aiming to improve their prediction capability and acceptability. In the frame of the VVER-1000 Coolant Transient Benchmark Phase 1, RELAP5/PARCS has been extensively assessed. Phase 2 of this benchmark was focused on both multidimensional thermal hydraulic phenomena and core physics. Plant data will be used to qualify the 3D models of TRACE and RELAP5/CFX, which were coupled for this purpose. The developed multidimensional models of the VVER-1000 reactor pressure vessel (RPV as well as the performed calculations will be described in detail. The predicted results are in good agreement with experimental data. It was demonstrated that the chosen 3D nodalization of the RPV is adequate for the description of the coolant mixing phenomena in a VVER-1000 reactor. Even though only a 3D coarse nodalization is used in TRACE, the integral results are comparable to those obtained by RELAP5/CFX.

  7. Waste Heat Recovery from the Advanced Test Reactor Secondary Coolant Loop

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen

    2012-11-01

    This study investigated the feasibility of using a waste heat recovery system (WHRS) to recover heat from the Advanced Test Reactor (ATR) secondary coolant system (SCS). This heat would be used to preheat air for space heating of the reactor building, thus reducing energy consumption, carbon footprint, and energy costs. Currently, the waste heat from the reactor is rejected to the atmosphere via a four-cell, induced-draft cooling tower. Potential energy and cost savings are 929 kW and $285K/yr. The WHRS would extract a tertiary coolant stream from the SCS loop and pump it to a new plate and frame heat exchanger, from which the heat would be transferred to a glycol loop for preheating outdoor air supplied to the heating and ventilation system. The use of glycol was proposed to avoid the freezing issues that plagued and ultimately caused the failure of a WHRS installed at the ATR in the 1980s. This study assessed the potential installation of a new WHRS for technical, logistical, and economic feasibility.

  8. Failures of the thermal barriers of 900 MWe reactor coolant pumps

    Energy Technology Data Exchange (ETDEWEB)

    Peyrouty, P.

    1996-12-01

    This report describes the anomalies encountered in the thermal barriers of the reactor coolant pumps in French 900 MWe PWR power stations. In addition to this specific problem, it demonstrates how the fortuitous discovery of a fault during a sampling test enabled faults of a generic nature to be revealed in components which were not subject to periodic inspection, the failure of which could seriously affect safety. This example demonstrates the risk which can be associated with the deterioration in areas which are not examined periodically and for which there are no preceding signs which would make early detection of deterioration possible.

  9. Thermal Behavior of the Coolant in the Emergency Cooldown Tank for an Integral Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Joo Hyung; Kim, Seok; Kim, Woo Shik; Jung, Seo Yoon; Kim, Young In [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The Residual Heat Removal System (PRHRS) is one of the passive safety systems which should be activated after an accident to remove the residual heat from the core and the sensible heat of the reactor coolant system (RCS) through the steam generators until the safe shutdown conditions are reached. In the previous study presented at the last KNS Autumn Meeting, transient behavior of the RCS temperature and the cooling performance of the PRHRS were investigated numerically by using newly developed in-house code based on MATLAB software. By using the program, the steady-state and transient (quasi-steady state) characteristics during the operation of the PRHRS had been reported. In this program, the temperature of the coolant in the Emergency Cooldown Tank (ECT) was assumed to be constant at saturated state and pool boiling heat transfer mechanism was applied through the entire time domain. The coolant of the ECT reached at a saturated state in early time. It was revealed that the assumption made in the previous study was reasonable.

  10. Analysis of events related to cracks and leaks in the reactor coolant pressure boundary

    Energy Technology Data Exchange (ETDEWEB)

    Ballesteros, Antonio, E-mail: Antonio.Ballesteros-Avila@ec.europa.eu [JRC-IET: Institute for Energy and Transport of the Joint Research Centre of the European Commission, Postbus 2, NL-1755 ZG Petten (Netherlands); Sanda, Radian; Peinador, Miguel; Zerger, Benoit [JRC-IET: Institute for Energy and Transport of the Joint Research Centre of the European Commission, Postbus 2, NL-1755 ZG Petten (Netherlands); Negri, Patrice [IRSN: Institut de Radioprotection et de Sûreté Nucléaire (France); Wenke, Rainer [GRS: Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH (Germany)

    2014-08-15

    Highlights: • The important role of Operating Experience Feedback is emphasised. • Events relating to cracks and leaks in the reactor coolant pressure boundary are analysed. • A methodology for event investigation is described. • Some illustrative results of the analysis of events for specific components are presented. - Abstract: The presence of cracks and leaks in the reactor coolant pressure boundary may jeopardise the safe operation of nuclear power plants. Analysis of cracks and leaks related events is an important task for the prevention of their recurrence, which should be performed in the context of activities on Operating Experience Feedback. In response to this concern, the EU Clearinghouse operated by the JRC-IET supports and develops technical and scientific work to disseminate the lessons learned from past operating experience. In particular, concerning cracks and leaks, the studies carried out in collaboration with IRSN and GRS have allowed to identify the most sensitive areas to degradation in the plant primary system and to elaborate recommendations for upgrading the maintenance, ageing management and inspection programmes. An overview of the methodology used in the analysis of cracks and leaks related events is presented in this paper, together with the relevant results obtained in the study.

  11. LOSS-OF-COOLANT ACIDENT SIMULATIONS IN THE NATIONAL RESEARCH UNIVERSAL REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, W D; Goodman, R L; Heaberlin, S W; Hesson, G M; Nealley, C; Kirg, L L; Marshall, R K; McNair, G W; Meitzler, W D; Neally, G W; Parchen, L J; Pilger, J P; Rausch, W N; Russcher, G E; Schreiber, R E; Wildung, N J

    1981-02-01

    Pressurized water reactor loss-of-coolant accident (LOCA) phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship among the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. Subsequent experiments establish the fuel rod failure characteristics at selected peak cladding temperatures. Fuel rod cladding pressurization simulates high burnup fission gas pressure levels of modern PWRs. This document contains both an experiment overview of the LOCA simulation program and a review of the safety analyses performed by Pacific Northwest Laboratory (PNL) to define the expected operating conditions as well as to evaluate the worst case operating conditions. The primary intent of this document is to supply safety information required by the Chalk River Nuclear Laboratories (CRNL), to establish readiness to proceed from one test phase to the next and to establish the overall safety of the experiment. A hazards review summarizes safety issues, normal operation and three worst case accidents that have been addressed during the development of the experiment plan.

  12. Selection of sodium coolant for fast reactors in the US, France and Japan

    Energy Technology Data Exchange (ETDEWEB)

    Sakamoto, Yoshihiko, E-mail: sakamoto.yoshihiko@jaea.go.jp [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Ibaraki-ken 311-1393 (Japan); Garnier, Jean-Claude; Rouault, Jacques [CEA, DEN, DER, Centre de Cadarache, 13108 Saint Paul Lez Durance Cedex (France); Grandy, Christopher; Fanning, Thomas; Hill, Robert [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Chikazawa, Yoshitaka; Kotake, Shoji [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency, 4002 Narita-cho, Oarai-machi, Ibaraki-ken 311-1393 (Japan)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer Trilateral study was conducted on coolant selection of fast reactor concept. Black-Right-Pointing-Pointer Fast reactor concepts are vital for nuclear fuel cycle sustainability goals. Black-Right-Pointing-Pointer Sodium, gas and lead cooled fast reactors are capable to achieve the goals. Black-Right-Pointing-Pointer Sodium cooled fast reactor is the most matured technology. Black-Right-Pointing-Pointer Gas and lead cooled fast reactor require long term development. - Abstract: The joint paper presents a common view of fast reactor specific missions in the development of nuclear energy and a cross-analysis of merits and demerits of several Fast Reactors concepts studied worldwide and especially in the Generation-IV International Forum (GIF) framework. The paper provides the context for fast reactors development in the United States, France and Japan and focuses on the comparison on Sodium-cooled Fast Reactor (SFR), Gas-cooled Fast Reactor (GFR), and Lead-cooled Fast Reactor (LFR), i.e. the three fast reactor concepts that have the potential to meet the nuclear fuel cycle sustainability goals. The information provided in the article permits the reader to understand each country's objectives to see that not only the objectives searched for but also the technical orientations are converging. The authors underline that SFR technology evaluation relies significantly on the substantial base technology development programs within each country which is without comparison for the other two fast reactor technologies, e.g., SFR technology has already been developed to commercial or near commercial scale in each country whereas the performance of LFR and GFR technology is still uncertain. The main GFR merits are the potential for high temperatures and the easier possibilities for inspections and repairs. The main challenges are the fuel (fabrication, in-pile behavior), materials for high temperatures, and the implementation of

  13. Modeling of Flow in Nuclear Reactor Fuel Cell Outlet

    Directory of Open Access Journals (Sweden)

    František URBAN

    2010-12-01

    Full Text Available Safe and effective load of nuclear reactor fuel cells demands qualitative and quantitative analysis of relations between coolant temperature in fuel cell outlet temperature measured by thermocouple and middle temperature of coolant in thermocouple plane position. In laboratory at Insitute of thermal power engineering of the Slovak University of Technology in Bratislava was installed an experimental physical fuel cell model of VVER 440 nuclear power plant with V 213 nuclear reactors. Objective of measurements on physical model was temperature and velocity profiles analysis in the fuel cell outlet. In this paper the measured temperature and velocity profiles are compared with the results of CFD simulation of fuel cell physical model coolant flow.

  14. Assessment of Candidate Molten Salt Coolants for the Advanced High Temperature Reactor (AHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D.F.

    2006-03-24

    The Advanced High-Temperature Reactor (AHTR) is a novel reactor design that utilizes the graphite-matrix high-temperature fuel of helium-cooled reactors, but provides cooling with a high-temperature fluoride salt. For applications at temperatures greater than 900 C the AHTR is also referred to as a Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR). This report provides an assessment of candidate salts proposed as the primary coolant for the AHTR based upon a review of physical properties, nuclear properties, and chemical factors. The physical properties most relevant for coolant service were reviewed. Key chemical factors that influence material compatibility were also analyzed for the purpose of screening salt candidates. Some simple screening factors related to the nuclear properties of salts were also developed. The moderating ratio and neutron-absorption cross-section were compiled for each salt. The short-lived activation products, long-lived transmutation activity, and reactivity coefficients associated with various salt candidates were estimated using a computational model. Table A presents a summary of the properties of the candidate coolant salts. Certain factors in this table, such as melting point, vapor pressure, and nuclear properties, can be viewed as stand-alone parameters for screening candidates. Heat-transfer properties are considered as a group in Sect. 3 in order to evaluate the combined effects of various factors. In the course of this review, it became apparent that the state of the properties database was strong in some areas and weak in others. A qualitative map of the state of the database and predictive capabilities is given in Table B. It is apparent that the property of thermal conductivity has the greatest uncertainty and is the most difficult to measure. The database, with respect to heat capacity, can be improved with modern instruments and modest effort. In general, ''lighter'' (low-Z) salts tend to

  15. Optimal estimate of the coolant flow in the assemblies of a BWR of natural circulation in real time; Estimacion optima del flujo de refrigerante en los ensambles de un BWR de circulacion natural en tiempo real

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, Facultad de Ingenieria, Division de Estudios de Posgrado, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, Col. Progreso, 62550 Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: julfi_jg@yahoo.com.mx [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2012-10-15

    The present work exposes the design and the implementation of an advanced controller that allows estimating the coolant flow in the fuel assemblies of a BWR reactor of natural circulation in real time. To be able to reduce the penalizations that are established in the calculations of the operation limits due to the magnitude of the uncertainties in the coolant flows of a natural circulation reactor, is the objective of this research. In this work the construction of the optimal controller that allows estimating the coolant flows in a fuel channels group of the reactor is shown, as well as the operation of this applied to a reduced order model that simulates the dynamics of a natural circulation reactor. The controller design required of an estimator of the valuation variables not directly of the plant and of the estimates use of the local distributions of the coolant flow. The controller construction of the estimator was based mathematically in the filter Kalman whose algorithm allows to be carried out an advanced control of the system. To prove the estimator operation was development a simplified model that reproduces the basic dynamics of the flowing coolant in the reactor, which works as observer of the system, this model is coupled by means of the estimator controller to a detail model of the plant. The results are presented by means of graphics of the interest variables and the estimate flow, and they are documented in the chart that is presented at the end of this article. (Author)

  16. Development of core design and analysis technology for integral reactor; development of coolant activity and dose evaluation program

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Chang Sun; Kim, Byeong Soo; Go, Hyun Seok; Lee, Young Wook; Jang, Mee [Seoul National University, Seoul (Korea)

    2002-03-01

    SMART, small- medium-sized integral reactor, is different from the customary electricity-generation PWR in design concepts and structures. The conventional coolant activity evaluation codes used in customary PWRs cannot be applied to SMART. In this study, SAEP(Specific Activity Evaluation Program) is developed that can be applied to both customary PWR and advanced reactor such as SMART. SAEP uses three methods(SAEP Ver.02, Ver.05, Ver.06) to evaluate coolant activity. They solve inhomogeneous linearly-coupled differential equations generated by considering nuclear system as N sub-components. Coolant activities of customary PWR are evaluated by use of SAEP. The results show good agreement with FSAR data. SAEP is used to evaluate coolant activities for SMART and the results are proposed in this study. These results show that SAEP is able to perform coolant activity evaluation for both customary PWR and advanced reactor such as SMART. In addition, with respect to radiation shielding optimization, conventional optimization methods and their characteristics related to radiation shielding are reviewed and analyzed. Strategies for proper usage of conventional methods are proposed to agree with the shielding design cases. 30 refs., 25 figs., 6 tabs. (Author)

  17. Attenuation of Vane-Rotor Shock Interactions with Pulsating Coolant Flows

    Science.gov (United States)

    2012-03-01

    observed over a range of bleed air mass flows near to the value producing a maximum level of base pressure. Sieverding [15] showed that a higher base...Rotating valve Pulsation in the coolant stream was provided by a perforated rotating disc . When the holes on the disc are facing the inlet and outlet...of the vortex formation location downstream when the highest base pressure is observed [14]. Augmentation in coolant ejection rates brakes down the

  18. Effect of ribbed and smooth coolant cross-flow channel on film cooling

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Wei; Sun, Xiaokai [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China); Jiang, Peixue, E-mail: jiangpx@tsinghua.edu.cn [Key Laboratory for Thermal Science and Power Engineering of Ministry of Educations, Department of Thermal Engineering, Tsinghua University, Beijing 100084 (China); Wang, Jie [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2017-05-15

    Highlights: • Little different for plenum model and the cross-flow model at M = 0.5. • Crossflow model is much better than plenum model at M = 1.0, especially with ribs. • Coolant flow channel with V-shaped ribs has the best adiabatic film cooling. • Film cooling with the plenum model is better at M = 0.5 than at M = 1.0. • Crossflow model is better at M = 0.5 near film hole and at M = 1.0 for downstream. - Abstract: The influence of ribbed and unribbed coolant cross-flow channel on film cooling was investigated with the coolant supply being either a plenum-coolant feed or a coolant cross-flow feed. Validation experiments were conducted with comparison to numerical results using different RANS turbulence models showed that the RNG k–ε turbulence model and the RSM model gave closer predictions to the experimental data than the other RANS models. The results indicate that at a low blowing ratio of M = 0.5, the coolant supply channel structure has little effect on the film cooling. However, at a high blowing ratio of M = 1.0, the adiabatic wall film cooling effectiveness is significantly lower with the plenum feed than with the cross-flow feed, especially for the cases with ribs. The film cooling with the plenum model is better at M = 0.5 than at M = 1.0. The film cooling with the cross-flow model is better at a blowing ratio of M = 0.5 in the near hole region, while further downstream, it is better at M = 1.0. The results also show that the coolant cross-flow channel with V-shaped ribs has the best adiabatic film cooling effectiveness.

  19. Liquid metal reactor development -Studies on safety measure of LMR coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Sung Tae; Choi, Yoon Dong; Park, Jin Hoh; Kwon, Sun Kil; Choi, Jong Hyun; Cho, Byung Ryul; Kim, Tae Joon; Kwon, Sang Woon; Jung, Kyung Chae; Kim, Byung Hoh; Hong, Soon Bok; Jung, Ji Yung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    A study on the safety measures of LMR coolant showed the results as follows; 1. LMR coolant safety measure. A. Analysis and improvement of sodium fire code. B. Analysis of sodium fire phenomena. 2. Sodium fire aerosol characteristics. It was carried out conceptual design and basic design for sodium fire facility of medium size composed of sodium supply tank, sodium reactor vessel, sodium fire aerosol filter system and scrubbing column, and drain tank etc. 3. Sodium purification technology. A. Construction of calibration loop. (1) Design of sodium loop for the calibration of the equipment. (2) Construction of sodium loop including test equipments and other components. B. Na-analysis technology. (1) Oxygen concentration determination by the wet method. (2) Cover gas purification preliminary experiment. 4. The characteristics of sodium-water reaction. A. Analysis of the micro and small leak phenomena. (1) Manufacture of the micro-leak test apparatus. B. Analysis of large leak events. (1) Development of preliminary code for analysis of initial spike pressure. (2) Sample calculation and comparison with previous works. C. Development of test facility for large leak event evaluation. (1) Conceptional and basic design for the water and sodium-water test facility. D. Technology development for water leak detection system. (1) Investigations for the characteristics of active acoustic detection system. (2) Testing of the characteristics of hydrogen leak detection system. 171 figs, 29 tabs, 3 refs. (Author).

  20. A passively-safe fusion reactor blanket with helium coolant and steel structure

    Energy Technology Data Exchange (ETDEWEB)

    Crosswait, Kenneth Mitchell [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    1994-04-01

    Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accident conditions such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This thesis describes a new helium-cooled tritium breeding blanket concept which performs significantly better during such accidents than current designs. The proposed blanket uses reduced-activation ferritic steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m{sup 2}. The proposed geometry is based on the nested-shell concept developed by Wong, but some novel features are used to reduce the severity of the first wall temperature excursion. These features include the following: (1) A ``beryllium-joint`` concept is introduced, which allows solid beryllium slabs to be used as a thermal conduction path from the first wall to the cooler portions of the blanket. The joint concept allows for significant swelling of the beryllium (10 percent or more) without developing large stresses in the blanket structure. (2) Natural circulation of the coolant in the water-cooled shield is used to maintain shield temperatures below 100 degrees C, thus maintaining a heat sink close to the blanket during the accident. This ensures the long-term passive safety of the blanket.

  1. A study of the tritium behavior in coolant and moderator system of heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. P.; Song, S. S.; Chae, K. S. and others [Chosun Univ., Gwangju (Korea, Republic of)

    1993-12-15

    The objectives of this report is to present a regulatory policy on the environmental impact and personnel exposure by understanding the generation, accumulation, environmental release and management of tritium in heavy water reactors. By estimating the tritium concentration at Wolsong nuclear plant site by estimating and forecasting the generation and accumulation of tritium in coolant and moderator systems at Wolsong unit 1, we will study the management and release of tritium at Wolsong units 3 and 4 which are ready for construction. The major activities of this study are as follows : tritium generation and accumulation in heavy water reactor, a quantitative assessment of the accumulation and release of tritium at Wolsong nuclear plant site, heavy water management at Wolsong nuclear plants. The tritium concentration and accumulation trends in the systems at Wolsong unit 1 was estimated. A quantitative assessment of the tritium accumulation and release for Wolsong 2, 3 and 4 based on data from Wolsong 1 was performed. The tritium removal schemes and its long-term management plan were made.

  2. A probability model: Tritium release into the coolant of a light water tritium production reactor

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, D N

    1992-04-01

    This report presents a probability model of the total amount of tritium that will be released from a core of tritium target rods into the coolant of a light water reactor during a tritium production cycle.The model relates the total tritium released from a core to the release characteristics of an individual target rod within the core. The model captures total tritium release from two sources-release via target rod breach and release via permeation through the target rod. Specifically, under conservative assumptions about the breach characteristics of a target rod, total tritium released from a core is modeled as a function of the probability of a target breach and the mean and standard deviation of the permeation reduction factor (PRF) of an individual target rod. Two dominant facts emerge from the analysis in this report. First, total tritium release cannot be controlled and minimized solely through the PRF characteristics of a target rod. Tritium release via breach must be abated if acceptable tritium production is to be achieved. Second, PRF values have a saturation point to their effectiveness. Specifically, in the presence of any realistic level of PRF variability, increasing PRF values above approximately 1000 wig contribute little to minimizing total tritium release.

  3. STUDY OF DRYING KINETICS OF BUCKWHEAT IN THE DRYER WITH TWISTED COOLANT FLOWS

    Directory of Open Access Journals (Sweden)

    S. T. Antipov

    2014-01-01

    Full Text Available Summary. Drying of buckwheat seeds is one of the most important stages of preparation of this raw material to the subsequent process of its processing. The nutritional value and quality indicators of the finished product depend on the mode of drying and are the result of structural, mechanical, biological and physico-mechanical transformations of substances. Technological modes of drying buckwheat seeds depend on the content of water and have a significant effect on the change of carbohydrates, protein denaturation, oxidation of lipids, changes of vitamins and organic acids. A new method of drying buckwheat and designed and constructed an experimental dryer with adjustable swirling flow of the coolant is proposed. For the study of the experiments and to determine the optimal mode of drying was used central composite rotatable uniforms - planning and selected full factorial experiment. The kinetics of drying and heat buckwheat in the device with twisted coolant flow was investigated. The influence of various parameters on the drying kinetics of buckwheat in the dryer with twisted coolant flow was invesigated. Presents the results of experimental studies, buckwheat drying in the dryer with twisted coolant flows. On the basis of experimental data and their statistical processing was obtained a mathematical model that adequately describes the process of drying buckwheat in the device with twisted coolant flow. The character of changes in the criteria optimization depending on the input factors was determined. The results of the mathematical model will be useful to a wide range of professionals involved in drying buckwheat, as well as for the calculation and design of modern drying - boiler systems.

  4. Assessment of the Use of Nitrogen Trifluoride for Purifying Coolant and Heat Transfer Salts in the Fluoride Salt-Cooled High-Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Scheele, Randall D.; Casella, Andrew M.

    2010-09-28

    This report provides an assessment of the use of nitrogen trifluoride for removing oxide and water-caused contaminants in the fluoride salts that will be used as coolants in a molten salt cooled reactor.

  5. CFD analysis of localized crud effects on the flow of coolant in nuclear rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Cinosi, N., E-mail: n.cinosi@imperial.ac.uk; Walker, S.P.

    2016-08-15

    Highlights: • CDF simulation of PWR sub-channels with crudded rods. • Evaluation of coolant flow variations induced by crud rough surfaces. • Evaluation of coolant temperatures in presence of crudded rods. • Evaluation of crud effects on critical heat flux and DNBR margins. - Abstract: It has been suggested that crud deposits on a number of adjacent fuel rods might reduce coolant flow rates in associated sub-channels. Such reduced flow rates could then worsen thermal-hydraulic conditions, such as margin to saturated boiling, fuel surface temperature, and the DNB ratio. We report the results of a detailed computational fluid dynamics study of the flow pattern in a partially crudded rod bundle. Values obviously depend on, for example, the thickness of crud assumed, but sub-channel flow rate reductions of ∼10% were predicted by this analysis. However, this mass flow rate reduction was found to be more than offset by improved heat transfer induced by the relatively rough surface of the crud. Cladding temperatures were predicted to be essentially unchanged, and the DNBR was similarly little altered. We conclude that such flow reduction and diversion is not likely to be of concern.

  6. Numerical Simulation of the Pressure Distribution in the Reactor Vessel Downcomer Region Fluctuated by the Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Dong Hwa; Jung, Byung Ryul; Jang, Ho Cheol; Yune, Seok Jeong; Kim, Eun Kee [KEPCO EnC, Daejeon (Korea, Republic of)

    2015-10-15

    In this study the numerical simulation of the pressure distribution in the downcomer region resulting from the pressure pulsation by the Reactor Coolant Pump (RCP) is performed using the Finite Difference Method (FDM). Simulation is carried out for the cylindrical shaped 2-dimensional model equivalent to the outer surface of the Core Support Barrel (CSB) of APR1400 and a 1/2 model is adopted based on the bilateral symmetry by the inlet nozzle. The fluid temperature is 555 .deg. F and the forcing frequencies are 120Hz, 240Hz, 360Hz and 480Hz. Simulation results of the axial pressure distributions are provided as the Root Mean Square (RMS) values at the five locations of 0°, 45°, 90°, 135° and 180° in the circumferential direction from the inlet nozzle location. In the study, the numerical simulation of pressure distributions in the downcomer region induced by the RCP was performed using FDM and the results were reviewed. The interference of the waves returned from both boundaries in the axial direction and the source of the sinusoidal wave is shown on the inlet nozzle interface pressure point. It seems that the maximum pressures result from the superposition of the waves reflected from the seating surface and the waves newly arrived from the inlet nozzle interface pressure location.

  7. Experimental and numerical investigation of coolant mixing in a model of reactor pressure vessel down-comer and in cold leg inlets

    Directory of Open Access Journals (Sweden)

    Hutli Ezddin

    2017-01-01

    Full Text Available Thermal fatigue and pressurized thermal shock phenomena are the main problems for the reactor pressure vessel and the T-junctions both of them depend on the mixing of the coolant. The mixing process, flow and temperature distribution has been investigated experimentally using particle image velocimetry, laser induced fluorescence, and simulated by CFD tools. The obtained results showed that the ratio of flow rate between the main pipe and the branch pipe has a big influence on the mixing process. The particle image velocimetry/planar laser-induced fluorescence measurements technologies proved to be suitable for the investigation of turbulent mixing in the complicated flow system: both velocity and temperature distribution are important parameters in the determination of thermal fatigue and pressurized thermal shock. Results of the applied these techniques showed that both of them can be used as a good provider for data base and to validate CFD results.

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

    Energy Technology Data Exchange (ETDEWEB)

    Paul, D.D.

    1976-06-01

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

  9. Thermodynamic consideration of hydrogen injection in BWR coolant. Estimation of potential for SCC control and oxidation-reduction condition of reactor coolant

    Energy Technology Data Exchange (ETDEWEB)

    Miyajima, Kaori; Hirano, Hideo; Domae, Masashi [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab; Kushida, H.

    2001-04-01

    Hydrogen injection into BWR coolant has been carried out in order to reduce stress corrosion cracking (SCC). It was clarified by in-plant test that SCC can be reduced under corrosion potential -0.23 V(v.s.SHE), but the theoretical basis has not been clarified. On the other hand, highly precise water quality analysis of re-circulatory-system water is generally performed. Especially, nitrogen compound changes chemical from to NO{sub 3}{sup -} -> NO{sub 2}{sup -} -> NH{sub 3}, and the NH{sub 3} becomes the cause of the increase of dose rate of the main steamy system in connection with the increase in the amount of hydrogen injection. However, the relation between this chemical form, oxidisation reduction potential, and temperature is not clear: Then, in this paper, these two points were considered by thermodynamics calculation at 25-300degC using the thermodynamics data in the high temperature accumulated in CRIEPI, and calculation results are summarized as follows; (1) the potential of the stainless steel to which the chemical form change to FeCr{sub 2}O{sub 4} from NiFe{sub 2}O{sub 4} is equilibrium is about -0.23 V at 288degC so this change is expected as one of factors for reduction of SCC, (2) the changes of chemical form of nitrogen compounds show oxidation-reduction of reactor coolant, so it can be useful as the index for control of dose rate. (author)

  10. Modular assembly for supporting, straining, and directing flow to a core in a nuclear reactor

    Science.gov (United States)

    Pennell, William E.

    1977-01-01

    A reactor core support arrangement for supporting, straining, and providing fluid flow to the core and periphery of a nuclear reactor during normal operation. A plurality of removable inlet modular units are contained within permanent liners in the lower supporting plate of the reactor vessel lower internals. During normal operation (1) each inlet modular unit directs main coolant flow to a plurality of core assemblies, the latter being removably supported in receptacles in the upper portion of the modular unit and (2) each inlet modular unit may direct bypass flow to a low pressure annular region of the reactor vessel. Each inlet modular unit may include special fluid seals interposed between mating surfaces of the inlet modular units and the core assemblies and between the inlet modular units and the liners, to minimize leakage and achieve an hydraulic balance. Utilizing the hydraulic balance, the modular units are held in the liners and the assemblies are held in the modular unit receptacles by their own respective weight. Included as part of the permanent liners below the horizontal support plate are generally hexagonal axial debris barriers. The axial debris barriers collectively form a bottom boundary of a secondary high pressure plenum, the upper boundary of which is the bottom surface of the horizontal support plate. Peripheral liners include radial debris barriers which collectively form a barrier against debris entry radially. During normal operation primary coolant inlet openings in the liner, below the axial debris barriers, pass a large amount of coolant into the inlet modular units, and secondary coolant inlet openings in the portion of the liners within the secondary plenum pass a small amount of coolant into the inlet modular units. The secondary coolant inlet openings also provide alternative coolant inlet flow paths in the unlikely event of blockage of the primary inlet openings. The primary inlet openings have characteristics which limit the

  11. Hydro-elastic calculations of the dynamic response of a reactor to a sudden loss of coolant

    Energy Technology Data Exchange (ETDEWEB)

    Dienes, J.K.; Hirt, C.W.; Stein, L.R.

    1977-01-01

    In the design of pressurized water reactors (PWRs) it is necessary to assure that a breach of the primary coolant circuit would not lead to failure of the structure supporting the core. Because of the complexity of the problem, designers have used a number of approximations in estimating the structural loadings that are generally believed to lead to conservative designs. The current research effort at the Los Alamos Scientific Laboratory is intended to take advantage of advanced computer codes and the largest computers to analyze simultaneously the fluid motion that accompanies a loss of coolant accident (LOCA), the structural motion resulting from the transient pressure differentials, and the effect of structural motion of the pressure in the fluid. Progress on the study of this problem is described.

  12. The study of ultrasonic reflex-radar waveguide coolant level gage for a nuclear reactor

    Directory of Open Access Journals (Sweden)

    V.I. Mel'nikov

    2016-03-01

    The instrument works reliably and does not require introducing corrections of readings when coolant thermal physical properties change. The measurement instrument is intended for application in heat exchanging equipment in thermal and nuclear power generation.

  13. The study of ultrasonic reflex-radar waveguide coolant level gage for a nuclear reactor

    National Research Council Canada - National Science Library

    Mel'nikov, V.I; Ivanov, V.V; Teplyashin, I.A

    2016-01-01

    Results of experimental study of operation of ultrasonic reflex-radar waveguide level gage in water coolant at elevated parameters with pressure up to 18MPa and temperature up to 350°C are examined...

  14. CFD Analysis of Localised Crud Effects on the Flow of Coolant in Nuclear Rod Bundles

    OpenAIRE

    Cinosi, N; Walker, SP

    2016-01-01

    It has been suggested that crud deposits on a number of adjacent fuel rods might reduce coolant flow rates in associated sub-channels. Such reduced flow rates could then worsen thermal-hydraulic conditions, such as margin to saturated boiling, fuel surface temperature, and the DNB ratio. We report the results of a detailed computational fluid dynamics study of the flow pattern in a partially-crudded rod bundle. Values obviously depend on, for example, the thickness of crud assumed, but sub-ch...

  15. Combustion synthesis continuous flow reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maupin, G.D.; Chick, L.A.; Kurosky, R.P.

    1998-01-06

    The present invention is a reactor for combustion synthesis of inorganic powders. The reactor includes a reaction vessel having a length and a first end and a second end. The reaction vessel further has a solution inlet and a carrier gas inlet. The reactor further has a heater for heating both the solution and the carrier gas. In a preferred embodiment, the reaction vessel is heated and the solution is in contact with the heated reaction vessel. It is further preferred that the reaction vessel be cylindrical and that the carrier gas is introduced tangentially into the reaction vessel so that the solution flows helically along the interior wall of the reaction vessel. As the solution evaporates and combustion produces inorganic material powder, the carrier gas entrains the powder and carries it out of the reactor. 10 figs.

  16. Combustion synthesis continuous flow reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maupin, Gary D. (Richland, WA); Chick, Lawrence A. (West Richland, WA); Kurosky, Randal P. (Maple Valley, WA)

    1998-01-01

    The present invention is a reactor for combustion synthesis of inorganic powders. The reactor includes a reaction vessel having a length and a first end and a second end. The reaction vessel further has a solution inlet and a carrier gas inlet. The reactor further has a heater for heating both the solution and the carrier gas. In a preferred embodiment, the reaction vessel is heated and the solution is in contact with the heated reaction vessel. It is further preferred that the reaction vessel be cylindrical and that the carrier gas is introduced tangentially into the reaction vessel so that the solution flows helically along the interior wall of the reaction vessel. As the solution evaporates and combustion produces inorganic material powder, the carrier gas entrains the powder and carries it out of the reactor.

  17. Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Seon Oh; Cho, Yong Jin [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, Sung Joong [Dept. of Nuclear Engineering, Hanyang University, Seoul (Korea, Republic of)

    2017-08-15

    The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

  18. Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

    Directory of Open Access Journals (Sweden)

    Seon Oh Yu

    2017-08-01

    Full Text Available The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

  19. MATLAB/Simulink Framework for Modeling Complex Coolant Flow Configurations of Advanced Automotive Thermal Management Systems

    Energy Technology Data Exchange (ETDEWEB)

    Titov, Gene; Lustbader, Jason; Leighton, Daniel; Kiss, Tibor

    2016-04-05

    The National Renewable Energy Laboratory's (NREL's) CoolSim MATLAB/Simulink modeling framework was extended by including a newly developed coolant loop solution method aimed at reducing the simulation effort for arbitrarily complex thermal management systems. The new approach does not require the user to identify specific coolant loops and their flow. The user only needs to connect the fluid network elements in a manner consistent with the desired schematic. Using the new solution method, a model of NREL's advanced combined coolant loop system for electric vehicles was created that reflected the test system architecture. This system was built using components provided by the MAHLE Group and included both air conditioning and heat pump modes. Validation with test bench data and verification with the previous solution method were performed for 10 operating points spanning a range of ambient temperatures between -2 degrees C and 43 degrees C. The largest root mean square difference between pressure, temperature, energy and mass flow rate data and simulation results was less than 7%.

  20. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Heams, T J [Science Applications International Corp., Albuquerque, NM (United States); Williams, D A; Johns, N A; Mason, A [UKAEA, Winfrith, (England); Bixler, N E; Grimley, A J [Sandia National Labs., Albuquerque, NM (United States); Wheatley, C J [UKAEA, Culcheth (England); Dickson, L W [Atomic Energy of Canada Ltd., Chalk River, ON (Canada); Osborn-Lee, I [Oak Ridge National Lab., TN (United States); Domagala, P; Zawadzki, S; Rest, J [Argonne National Lab., IL (United States); Alexander, C A [Battelle, Columbus, OH (United States); Lee, R Y [Nuclear Regulatory Commission, Washington, DC (United States)

    1992-12-01

    The VICTORIA model of radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident is described. It has been developed by the USNRC to define the radionuclide phenomena and processes that must be considered in systems-level models used for integrated analyses of severe accident source terms. The VICTORIA code, based upon this model, predicts fission product release from the fuel, chemical reactions involving fission products, vapor and aerosol behavior, and fission product decay heating. Also included is a detailed description of how the model is implemented in VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided.

  1. 3-D slug flow heat transfer analysis of coupled coolant cells in finite LMFBR bundles

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.N.; Wolf, L.

    1978-02-01

    A three-dimensional single region slug flow heat transfer analysis for finite LMFBR rod bundles using a classical analytical solution method has been performed. According to the isolated single cell analysis, the results show that the peripheral clad temperature variation as well as the thermal entrance length are strongly dependent upon the degree of irregularity displayed by various coolant geometries. Since under the present LMFBR conditions, fully-developed temperature fields may hardly be established in such characteristic rod bundle regions, a 3-D heat transfer analysis seems to be mandatory. This implies that the results of fully developed heat transfer analyses are by far too conservative.

  2. Specifics of high-temperature sodium coolant purification technology in fast reactors for hydrogen production and other innovative applications

    Directory of Open Access Journals (Sweden)

    F.A. Kozlov

    2017-03-01

    Full Text Available In creating a large-scale atomic-hydrogen power industry, the resolution of technological issues associated with high temperatures in reactor plants (900°C and large hydrogen concentrations intended as long-term resources takes on particular importance. The paper considers technological aspects of removing impurities from high-temperature sodium used as a coolant in the high-temperature fast reactor (BN-HT 600MW (th. intended for the production of hydrogen as well as other innovative applications. The authors examine the behavior of impurities in the BN-HT circuits associated with the mass transfer intensification at high temperatures (Arrhenius law in different operating modes. Special attention is given to sodium purification from hydrogen, tritium and corrosion products in the BN-HT. Sodium purification from hydrogen and tritium by their evacuation through vanadium or niobium membranes will make it possible to develop compact highly-efficient sodium purification systems. It has been shown that sodium purification from tritium to concentrations providing the maximum permissible concentration of the produced hydrogen (3.6Bq/l according to NRB-99/2009 specifies more stringent requirements to the hydrogen removal system, i.e., the permeability index of the secondary tritium removal system should exceed 140kg/s. Provided that a BN-HN-type reactor meets these conditions, the bulk of tritium (98% will be accumulated in the compact sodium purification system of the secondary circuit, 0.6% (∼ 4·104Bq/s, will be released into the environment and 1.3% will enter the product (hydrogen. The intensity of corrosion products (CPs coming into sodium is determined by the corrosion rate of structural materials: at a high temperature level, a significant amount of corrosion products flows into sodium. The performed calculations showed that, for the primary BN-HT circuit, the amount of corrosion products formed at the oxygen concentration in sodium of 1mln

  3. The study of ultrasonic reflex-radar waveguide coolant level gage for a nuclear reactor

    OpenAIRE

    Mel'Nikov, V.I.; Ivanov, V. V.; Teplyashin, I.A.

    2016-01-01

    Results of experimental study of operation of ultrasonic reflex-radar waveguide level gage in water coolant at elevated parameters with pressure up to 18MPa and temperature up to 350°C are examined. In contrast to the known waveguide level gages, traveltime of acoustic pulses along the waveguide from the radiator to the subsurface layer and back is measured in the level gage under study. Waveguide consists of two acoustically isolated waveguides – the radiating waveguide and the receiving ...

  4. The radionuclides of primary coolant in HANARO and the recent activities performed to reduce the radioactivity or reactor pool water

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minjin [HANARO Research Reactor Centre, Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

    1998-10-01

    In HANARO reactor, there have been activities to identify the principal radionuclides and to quantify them under the normal operation. The purposes of such activities were to establish the measure by which we can reduce the radioactivity of the reactor pool water and detect, in early stage, the abnormal symptoms due to the leakage of radioactive materials from the irradiation sample or the damage of the nuclear fuel, etc. The typical radionuclides produced by the activation of reactor coolant are N{sup 16} and Ar{sup 41}. The radionuclides produced by the activation of the core structural material consist of Na{sup 24}, Mn{sup 56}, and W{sup 187}. Of the various radionuclides, governing the radiation level at the pool surface are Na{sup 24}, Ar{sup 41}, Mn{sup 58}, and W{sup 187}. By establishing the hot water layer system on the pool surface, we expected that the radionuclides such as Ar{sup 41} and Mn{sup 56} whose half-life are relatively short could be removed to a certain extent. Since the content of radioactivity of Na{sup 24} occupies about 60% of the total radioactivity, we assumed that the total radiation level would be greatly reduced if we could decrease the radiation level of Na{sup 24}. However the actual radiation level has not been reduced as much as we expected. Therefore, some experiments have been carried out to find the actual causes afterwards. What we learned through the experiments are that any disturbance in reactor pool water layer causes increase of the pool surface radiation level and even if we maintain the hot water layer well, reactor shutdown will be very much likely to happen once the hot water layer is disturbed. (author)

  5. Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1989-02-01

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs.

  6. Laminar Entrained Flow Reactor (Fact Sheet)

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2014-02-01

    The Laminar Entrained Flow Reactor (LEFR) is a modular, lab scale, single-user reactor for the study of catalytic fast pyrolysis (CFP). This system can be employed to study a variety of reactor conditions for both in situ and ex situ CFP.

  7. Study of Compatibility of Stainless Steel Weld Joints with Liquid Sodium-Potassium Coolants for Fission Surface Power Reactors for Lunar and Space Applications

    Energy Technology Data Exchange (ETDEWEB)

    Grossbeck, Martin [Univ. of Tennessee, Knoxville, TN (United States); Qualls, Louis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-07-31

    To make a manned mission to the surface of the moon or to Mars with any significant residence time, the power requirements will make a nuclear reactor the most feasible source of energy. To prepare for such a mission, NASA has teamed with the DOE to develop Fission Surface Power technology with the goal of developing viable options. The Fission Surface Power System (FSPS) recommended as the initial baseline design includes a liquid metal reactor and primary coolant system that transfers heat to two intermediate liquid metal heat transfer loops. Each intermediate loop transfers heat to two Stirling heat exchangers that each power two Stirling converters. Both the primary and the intermediate loops will use sodium-potassium (NaK) as the liquid metal coolant, and the primary loop will operate at temperatures exceeding 600°C. The alloy selected for the heat exchangers and piping is AISI Type 316L stainless steel. The extensive experience with NaK in breeder reactor programs and with earlier space reactors for unmanned missions lends considerable confidence in using NaK as a coolant in contact with stainless steel alloys. However, the microstructure, chemical segregation, and stress state of a weld leads to the potential for corrosion and cracking. Such failures have been experienced in NaK systems that have operated for times less than the eight year goal for the FSPS. For this reason, it was necessary to evaluate candidate weld techniques and expose welds to high-temperature, flowing NaK in a closed, closely controlled system. The goal of this project was to determine the optimum weld configuration for a NaK system that will withstand service for eight years under FSPS conditions. Since the most difficult weld to make and to evaluate is the tube to tube sheet weld in the intermediate heat exchangers, it was the focus of this research. A pumped loop of flowing NaK was fabricated for exposure of candidate weld specimens at temperatures of 600°C, the expected

  8. Frontier between medium and large break loss of coolant accidents of pressurized water reactor

    Science.gov (United States)

    Kim, Taewan

    2017-10-01

    In order to provide the probabilistic safety assessment with more realistic condition to calculate the frequency of the initiating event, a study on the frontier between medium-break and large-break loss-of-coolant-accidents has been performed by using best-estimate thermal hydraulic code, TRACE. A methodology based on the combination of the essential safety features and system parameter has been applied to the Zion nuclear power plant to evaluate the validity of the frontier utilized for the probabilistic safety assessment. The peak cladding temperature has been chosen as a relevant system parameter that represents the system behavior during the transient. The results showed that the frontier should be extended from 6 in. to 10 in. based on the required safety functions and system response.

  9. Seismic Safety Margins Research Programs. Assessment of potential increases in risk due to degradation of steam generator and reactor coolant pump supports. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bohn, M. P.; Wells, J. E.; Shieh, L. C.; Cover, L. E.; Streit, R. L.

    1983-08-01

    During the NRC licensing review for the North Anna Units 1 and 2 pressurized-water reactors (PWRs), questions were raised regarding the potential for low-fracture toughness of steam-generator and reactor-coolant-pump supports. Because other PWRs may face similar problems, this issue was incorporated into the NRC Program for Resolution of Generic Issues. The work described in this report was performed to provide the NRC with a quantitative evaluation of the value/impact implications of the various options of resolving the fracture-toughness question. This report presents an assessment of the probabilistic risk associated with nil-ductility failures of steam-generator and reactor-coolant-pump structural-support systems during seismic events, performed using the Seismic Safety Margins Research Program codes and data bases.

  10. Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

  11. Particle image velocimetry measurements in a representative gas-cooled prismatic reactor core model for the estimation of bypass flow

    Science.gov (United States)

    Conder, Thomas E.

    Core bypass flow is considered one of the largest contributors to uncertainty in fuel temperature within the Modular High Temperature Gas-cooled Reactor (MHTGR). It refers to the coolant that navigates through the interstitial regions between the graphite fuel blocks instead of traveling through the designated coolant channels. These flows are of concern because they reduce the desired flow rates in the coolant channels, and thereby have significant influence on the maximum fuel element and coolant exit temperatures. Thus, accurate prediction of the bypass flow is important because it directly impacts core temperature, influencing the life and efficiency of the reactor. An experiment was conducted at Idaho National Laboratory to quantify the flow in the coolant channels in relation to the interstitial gaps between fuel blocks in a representative MHTGR core. Particle Image Velocimetry (PIV) was used to measure the flow fields within a simplified model, which comprised of a stacked junction of six partial fuel blocks with nine coolant tubes, separated by a 6mm gap width. The model had three sections: The upper plenum, upper block, and lower block. Model components were fabricated from clear, fused quartz where optical access was needed for the PIV measurements. Measurements were taken in three streamwise locations: in the upper plenum and in the midsection of the large and small fuel blocks. A laser light sheet was oriented parallel to the flow, while velocity fields were measured at millimeter intervals across the width of the model, totaling 3,276 PIV measurement locations. Inlet conditions were varied to incorporate laminar, transition, and turbulent flows in the coolant channels---all which produced laminar flow in the gap and non-uniform, turbulent flow in the upper plenum. The images were analyzed to create vector maps, and the data was exported for processing and compilation. The bypass flow was estimated by calculating the flow rates through the coolant

  12. Investigations for optimal dissolved hydrogen (DH) concentration in reactor coolant system (RCS)

    Energy Technology Data Exchange (ETDEWEB)

    Ishihara, Nobuaki; Tanaka, Muneo [Shikoku Electric Power Co., Inc., Takamatsu (Japan); Nishizawa, Eiichi; Kasahara, Kazuo

    1998-12-31

    Optimal dissolved hydrogen (DH) concentration control is among the most important issues in developing program to maintain plant reliability for aging plant because it is useful in securing material integrity. Also, it is believed to be one of the most promising approaches, following pH control and Zn injection, to radiation exposure source reduction. This work involved collecting data for corrosion products in the coolant, particularly Ni (because the chemical forms of this element, parent element of {sup 58}Co, are affected by DH concentration), during the power operation at Ikata NPP, and determining the relations between DH, crud chemical forms and particle size distributions. In order to determine the optimal DH concentration for exposure source reduction, the results were evaluated in comparison with the findings about crud chemical forms from thermodynamic methods. Regarding DH dependence of crud characteristics, the results of field investigations revealed as follows: In crud chemical form, the ratio of Ni (metal) to total crud increases as the DH concentration increments. {sup 58}Co (Ni (metal) and spinel combined) median particle size grows greater as the DH concentration increments. These findings, together with other obtained findings (e.g., relations between particle size and release/deposition) and the calculations developed using thermodynamic methods, brought us to the following conclusion over the DH concentration control for the radiation exposure source reduction. Provided that the DH concentration should be controlled within the typical value (25 to 35 cc-STP/kg-H{sub 2}O), that concentration should be as close to the lower limit (25 cc-STP/kg-H{sub 2}O) as possible and the variation of DH concentration should be minimized. (J.P.N.)

  13. Investigation of loss of coolant accidents in pressurised water reactors using the ''Dynamic Best-Estimate Safety Analysis'' (DYBESA) method for consideration of uncertainties in TRACE

    Energy Technology Data Exchange (ETDEWEB)

    Sporn, Michael; Hurtado, Antonio [Technische Univ. Dresden (Germany)

    2015-07-01

    Loss of coolant accident must take into account uncertainties with potentially strong effects on the accident sequence prediction. In this paper, the use of the ''Dynamic Best-Estimate Safety Analysis'' (DYBESA) method to quantify the uncertainties in the TRACE thermal-hydraulic programme is demonstrated. For demonstration purposes, loss of coolant accidents with breaks of various types and sizes in a DN 700 reactor coolant pipe are used as an example application.

  14. Novel oscillatory flow reactors for biotechnological applications

    OpenAIRE

    Reis, N.

    2006-01-01

    Tese de Doutoramento em Engenharia Química e Biológica This thesis explores the biotechnological applications of two novel scale-down oscillatory flow reactors (OFRs). A micro-bioreactor (working mostly in batch) and a continuous meso-reactor systems were developed based on a 4.4 mm internal diameter tube with smooth periodic constrictions (SPC), both operating under oscillatory flow mixing (OFM). The first part is dedicated to the flow characterisation in the novel SPC geom...

  15. Nuclear reactor overflow line

    Science.gov (United States)

    Severson, Wayne J.

    1976-01-01

    The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

  16. Effects of a hypothetical loss-of-coolant accident on a Mark I Boiling Water Reactor pressure-suppression system

    Energy Technology Data Exchange (ETDEWEB)

    Pitts, J.H.; McCauley, E.W.

    1977-12-22

    A loss-of-coolant accident (LOCA) in a boiling-water-reactor (BWR) power plant has never occurred. However, because this type of accident could be particularly severe, it is used as a principal theoretical basis for design. A series of consistent, versatile, and accurate air-water tests that simulate LOCA conditions has been completed on a /sup 1///sub 5/-scale Mark I BWR pressure-suppression system. Results from these tests are used to quantify the vertical-loading function and to study the associated fluid dynamics phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variation of hydrodynamic-generated vertical loads with changes in drywell-pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the /sup 1///sub 5/-scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings that are invariant. These groupings show that, if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor and occurs in a time reduced by the square root of the scale factor.

  17. Regulatory instrument review: Aging management of LWR cables, containment and basemat, reactor coolant pumps, and motor-operated valves

    Energy Technology Data Exchange (ETDEWEB)

    Werry, E.V.; Somasundaram, S.

    1995-09-01

    The results of Stage 2 of the Regulatory Instrument Review are presented in this volume. Selected regulatory instruments, such as the Code of Federal Regulations (CFR), US Nuclear Regulatory Commission (NRC), Regulatory Guides, and ASME Codes, were investigated to determine the extent to which these regulations apply aging management to selected safety-related components in nuclear power plants. The Regulatory Instrument Review was funded by the NRC under the Nuclear Plant Aging Research (NPAR) program. Stage 2 of the review focused on four safety-related structures and components; namely, cables, containment and basemat, reactor coolant pumps, and motor-operated valves. The review suggests that the primary-emphasis of the regulatory instruments was on the design, construction, start-up, and operation of a nuclear power plant, and that aging issues were primarily addressed after an aging-related problem was recognized. This Stage 2 review confirms the results of the prior review; (see Regulatory Instrument Review: Management of Aging of LWR Major Safety-Related Components NUREG/CR-5490. The observations indicate that the regulations generally address management of age-related degradation indirectly. Specific age-related degradation phenomena frequently are dealt with in bulletins and notices or through generic issues, letters, etc. The major recommendation of this report, therefore, is that the regulatory instruments should more directly and explicitly address the aging phenomenon and the management of the age-related degradation process.

  18. Phased Array Ultrasonic Examination of Reactor Coolant System (Carbon Steel-to-CASS) Dissimilar Metal Weld Mockup Specimen

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, S. L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cinson, A. D. [US Nuclear Regulatory Commission (NRC), Washington, DC (United States); Diaz, A. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Anderson, M. T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-11-23

    In the summer of 2009, Pacific Northwest National Laboratory (PNNL) staff traveled to the Electric Power Research Institute (EPRI) NDE Center in Charlotte, North Carolina, to conduct phased-array ultrasonic testing on a large bore, reactor coolant pump nozzle-to-safe-end mockup. This mockup was fabricated by FlawTech, Inc. and the configuration originated from the Port St. Lucie nuclear power plant. These plants are Combustion Engineering-designed reactors. This mockup consists of a carbon steel elbow with stainless steel cladding joined to a cast austenitic stainless steel (CASS) safe-end with a dissimilar metal weld and is owned by Florida Power & Light. The objective of this study, and the data acquisition exercise held at the EPRI NDE Center, were focused on evaluating the capabilities of advanced, low-frequency phased-array ultrasonic testing (PA-UT) examination techniques for detection and characterization of implanted circumferential flaws and machined reflectors in a thick-section CASS dissimilar metal weld component. This work was limited to PA-UT assessments using 500 kHz and 800 kHz probes on circumferential flaws only, and evaluated detection and characterization of these flaws and machined reflectors from the CASS safe-end side only. All data were obtained using spatially encoded, manual scanning techniques. The effects of such factors as line-scan versus raster-scan examination approaches were evaluated, and PA-UT detection and characterization performance as a function of inspection frequency/wavelength, were also assessed. A comparative assessment of the data is provided, using length-sizing root-mean-square-error and position/localization results (flaw start/stop information) as the key criteria for flaw characterization performance. In addition, flaw signal-to-noise ratio was identified as the key criterion for detection performance.

  19. NONUNIFORMITIES OF TWO-PHASE COOLANT DISTRIBUTION IN A HEAT GENERATING PARTICLES BED

    Directory of Open Access Journals (Sweden)

    V. V. Sorokin

    2014-01-01

    Full Text Available Sufficient atomic power generation safety increase may be done with microfuel adapting to reactor plants with water coolant. Microfuel particle is a millimeter size grain containing fission material core in a protecting coverage. The coverage protects fuel contact with coolant and provides isolation of fission products inside. Well thermophysical properties of microfuel bed in a direct contact with water coolant excludes fuel overheating when accidents. Microfuel use was suggested for a VVER, а direct flow reactor for superheat steam generation, a reactor with neutron spectra adjustment by the steam partial content varying in the coolant.Nonuniformities of two-phase coolant distribution in a heat generating particles bed are predicted by calculations in this text. The one is due to multiple-valuedness of pressure drop across the bed on the steam quality dependency. The nonuniformity decreases with flow rate and particle size growths absolute pressure diminishing while porosity effect is weak. The worse case is for pressure quality of order of one. Some pure steam filled pores appears parallel to steam water mixture filled pores, latter steam quality is less than the mean of the bed. Considering this regime for the direct flow reactor for superheat steam generation we predict some water drops at the exit flow. The two-phase coolant filtration with subcooled water feed is unstable to strong disturbance effects are found. Uniformity of two-phase coolant distribution is worse than for one-phase in the same radial type reactor.

  20. Helix reactor: great potential for flow chemistry

    NARCIS (Netherlands)

    Geerdink, P.; Runstraat, A. van den; Roelands, C.P.M.; Goetheer, E.L.V.

    2009-01-01

    The Helix reactor is highly suited for precise reaction control based on good hydrodynamics. The hydrodynamics are controlled by the Dean vortices, which create excellent heat transfer properties, approach plug flow and avoid turbulence. The flexibility of this reactor has been demonstrated using a

  1. Safety Design Strategy for the Advanced Test Reactor Primary Coolant Pump and Motor Replacement Project

    Energy Technology Data Exchange (ETDEWEB)

    Noel Duckwitz

    2011-06-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  2. Accelerated Vibration Test of coolant channel components under simulated flow induced excitation

    Energy Technology Data Exchange (ETDEWEB)

    Meher, K.K., E-mail: kkmeher@barc.gov.in; Pandey, J.K., E-mail: jkpandey@barc.gov.in; RamaRao, A., E-mail: arr@barc.gov.in

    2016-04-15

    Highlights: • The present study deals with the issue of loosening of the nut in the Grayloc joint due to flow induced vibration and fret in the feeder pipes in contact due to differential creep in the neighbouring channels. • Accelerated test has been done on the Grayloc joint on simulated flow induced vibration to study the effect of loosening of the nut. • In the present accelerated test, the component has not been led to failure (loosening) and an estimation of its service life has been approached based on the severity of test. • The inverse square law approach based on PSD comparison for severity of test have been used to correlate the actual operational hours and the Laboratory test hours to verify the loosening of the Grayloc nut for the present study. • By inverse power law approach, the minimum number of reactor-hours equivalent to 80 h of testing is 46,080 h (5.26 full power years). - Abstract: The present study outlines the accelerated testing procedure of a Grayloc joint assembly for possible loosening of its nut due to flow induced vibration. The concern of the Grayloc nut getting loosened in the absence of a lock nut due to flow induced vibration and the resulting fretting in the feeder pipes in contact due to differential creep in the neighbouring channels has been addressed here. The severity of the test was decided based on actual site measurement under different operating flow conditions and comparison of power spectral density (PSD). The laboratory test results were extrapolated for estimation of life of the component under operating condition using inverse power law approach. The uniqueness of the accelerated test is that the component under test has not been led to failure for assessing its operating life unlike conventional accelerated testing. From the tests and analysis, it was deduced that 80 h of accelerated laboratory testing was equivalent to 5.26 full power years (46,080 h) of the reactor operating life. The test duration was

  3. MATLAB/Simulink Framework for Modeling Complex Coolant Flow Configurations of Advanced Automotive Thermal Management Systems: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Titov, Eugene; Lustbader, Jason; Leighton, Daniel; Kiss, Tibor

    2016-03-22

    The National Renewable Energy Laboratory's (NREL's) CoolSim MATLAB/Simulink modeling framework was extended by including a newly developed coolant loop solution method aimed at reducing the simulation effort for arbitrarily complex thermal management systems. The new approach does not require the user to identify specific coolant loops and their flow. The user only needs to connect the fluid network elements in a manner consistent with the desired schematic. Using the new solution method, a model of NREL's advanced combined coolant loop system for electric vehicles was created that reflected the test system architecture. This system was built using components provided by the MAHLE Group and included both air conditioning and heat pump modes. Validation with test bench data and verification with the previous solution method were performed for 10 operating points spanning a range of ambient temperatures between -2 degrees C and 43 degrees C. The largest root mean square difference between pressure, temperature, energy and mass flow rate data and simulation results was less than 7%.

  4. Neutronics of a mixed-flow gas-core reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-11-01

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

  5. Modeling of Hydrodynamic Processes at a Large Leak of Water into Sodium in the Fast Reactor Coolant Circuit

    OpenAIRE

    Perevoznikov, Sergey; Shvetsov, Yuriy; Kamayev, Aleksey; Pakhomov, Ilia; Borisov, Viacheslav; Pazin, Gennadiy; Mirzeabasov, Oleg; Korzun, Olga

    2016-01-01

    In this paper, we describe a physicomathematical model of the processes that occur in a sodium circuit with a variable flow cross-section in the case of a water leak into sodium. The application area for this technique includes the possibility of analyzing consequences of this leak as applied to sodium–water steam generators in fast neutron reactors. Hydrodynamic processes that occur in sodium circuits in the event of a water leak are described within the framework of a one-dimensional therma...

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

    Science.gov (United States)

    Fiantini, Rosalina; Umar, Efrizon

    2010-06-01

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

  7. Loss of Coolant Accident Analysis Methodology for SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Bae, K. H.; Lee, G. H.; Yang, S. H.; Yoon, H. Y.; Kim, S. H.; Kim, H. C

    2006-02-15

    The analysis methodology on the Loss-of-coolant accidents (LOCA's) for SMART-P is described in this report. SMART-P is an advanced integral type PWR producing a maximum thermal power of 65.5 MW with metallic fuel. LOCA's are hypothetical accidents that would result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant makeup system, from breaks in pipes in the reactor coolant pressure boundary up to and including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system. Since SMART-P contains the major primary circuit components in a single Reactor Pressure Vessel (RPV), the possibility of a large break LOCA (LBLOCA) is inherently eliminated and only the small break LOCA is postulated. This report describes the outline and acceptance criteria of small break LOCA (SBLOCA) for SMART-P and documents the conservative analytical model and method and the analysis results using the TASS/SMR code. This analysis method is applied in the SBLOCA analysis performed for the ECCS performance evaluation which is described in the section 6.3.3 of the safety analysis report. The prediction results of SBLOCA analysis model of SMART-P for the break flow, system's pressure and temperature distributions, reactor coolant distribution, single and two-phase natural circulation phenomena, and the time of major sequence of events, etc. should be compared and verified with the applicable separate and integral effects test results. Also, it is required to set-up the feasible acceptance criteria applicable to the metallic fueled integral reactor of SMART-P. The analysis methodology for the SBLOCA described in this report will be further developed and validated as the design and licensing status of SMART-P evolves.

  8. Modeling of Hydrodynamic Processes at a Large Leak of Water into Sodium in the Fast Reactor Coolant Circuit

    Directory of Open Access Journals (Sweden)

    Sergey Perevoznikov

    2016-10-01

    Full Text Available In this paper, we describe a physicomathematical model of the processes that occur in a sodium circuit with a variable flow cross-section in the case of a water leak into sodium. The application area for this technique includes the possibility of analyzing consequences of this leak as applied to sodium–water steam generators in fast neutron reactors. Hydrodynamic processes that occur in sodium circuits in the event of a water leak are described within the framework of a one-dimensional thermally nonequilibrium three-component gas–liquid flow model (sodium–hydrogen–sodium hydroxide. Consideration is given to the results of a mathematical modeling of experiments involving steam injection into the sodium loop of a circulation test facility. That was done by means of the computer code in which the proposed model had been implemented.

  9. Prediction of Flow and Temperature Distributions in a High Flux Research Reactor Using the Porous Media Approach

    Directory of Open Access Journals (Sweden)

    Shanfang Huang

    2017-01-01

    Full Text Available High thermal neutron fluxes are needed in some research reactors and for irradiation tests of materials. A High Flux Research Reactor (HFRR with an inverse flux trap-converter target structure is being developed by the Reactor Engineering Analysis Lab (REAL at Tsinghua University. This paper studies the safety of the HFRR core by full core flow and temperature calculations using the porous media approach. The thermal nonequilibrium model is used in the porous media energy equation to calculate coolant and fuel assembly temperatures separately. The calculation results show that the coolant temperature keeps increasing along the flow direction, while the fuel temperature increases first and decreases afterwards. As long as the inlet coolant mass flow rate is greater than 450 kg/s, the peak cladding temperatures in the fuel assemblies are lower than the local saturation temperatures and no boiling exists. The flow distribution in the core is homogeneous with a small flow rate variation less than 5% for different assemblies. A large recirculation zone is observed in the outlet region. Moreover, the porous media model is compared with the exact model and found to be much more efficient than a detailed simulation of all the core components.

  10. Commissioning of the STAR test section for experimental simulation of loss of coolant accident using the EC-208 instrumented fuel assembly of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maprelian, Eduardo; Torres, Walmir M.; Prado, Adelk C.; Umbehaun, Pedro E.; Franca, Renato L.; Santos, Samuel C.; Macedo, Luiz A.; Sabundjian, Gaiane, E-mail: emaprel@ipen.br, E-mail: wmtorres@ipen.br, E-mail: acprado@ipen.br, E-mail: umbehaun@ipen.br, E-mail: rlfranca@ipen.br, E-mail: samuelcs@ipen.br, E-mail: lamacedo@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SO (Brazil)

    2015-07-01

    The three basic safety functions of Research Reactors (RR) are the safe shutdown of the reactor, the proper cooling of the decay heat of the fuel elements and the confinement of radioactive materials. Compared to Nuclear Power Reactors, RR power release is small, yet its three safety functions must be met to ensure the integrity of the reactor. During a loss of coolant accident (LOCA) in pool type RR, partial or complete loss of pool water may occur, with consequent partial or complete uncovering of the fuel assemblies. In such an accident, the decay heat removal safety function must not be compromised. The Test Section for Experimental Simulation of Loss of Coolant Accident (STAR) is in commissioning phase. This test section will provide experimental data on partial and total uncovering of the EC-208 instrumented fuel assembly (IFA) irradiated in the IEA-R1. Experimental results will be useful in validation of computer codes for RR safety analysis, particularly on heat removal efficiency aspects (safety function) in accident conditions. STAR comprises a base on which is installed the IFA, the cylindrical stainless steel hull, the compressed air system for the test section emptying and refilling, and the instrumentation for temperature and level measurements. The commissioning tests or pre-operational check, consist of several preliminary tests to verify experimental procedures, the difficulties during assembling of STAR in the pool, the difficulties in control the emptying and refilling velocities, as well as, the repeatability capacity, tests of equipment, valves and systems and tests of instrumentation and data acquisition system. Safety, accuracy and easiness of operation will be checked. (author)

  11. Determination of the turbulence integral model parameters for a case of a coolant angular flow in regular rod-bundle

    Science.gov (United States)

    Bayaskhalanov, M. V.; Vlasov, M. N.; Korsun, A. S.; Merinov, I. G.; Philippov, M. Ph

    2017-11-01

    Research results of “k-ε” turbulence integral model (TIM) parameters dependence on the angle of a coolant flow in regular smooth cylindrical rod-bundle are presented. TIM is intended for the definition of efficient impulse and heat transport coefficients in the averaged equations of a heat and mass transfer in the regular rod structures in an anisotropic porous media approximation. The TIM equations are received by volume-averaging of the “k-ε” turbulence model equations on periodic cell of rod-bundle. The water flow across rod-bundle under angles from 15 to 75 degrees was simulated by means of an ANSYS CFX code. Dependence of the TIM parameters on flow angle was as a result received.

  12. Thermal fluid dynamic behavior of coolant helium gas in a typical reactor VHTGR channel of prismatic core; Comportamento termofluidodinamico do gas refrigerante helio em um canal topico de reator VHTGR de nucleo prismatico

    Energy Technology Data Exchange (ETDEWEB)

    Belo, Allan Cavalcante

    2016-08-01

    The current studies about the thermal fluid dynamic behavior of the VHTGR core reactors of 4{sup th} generation are commonly developed in 3-D analysis in CFD (computational fluid dynamics), which often requires considerable time and complex mathematical calculations for carrying out these analysis. The purpose of this project is to achieve thermal fluid dynamic analysis of flow of gas helium refrigerant in a typical channel of VHTGR prismatic core reactor evaluating magnitudes of interest such as temperature, pressure and fluid velocity and temperature distribution in the wall of the coolant channel from the development of a computer code in MATLAB considering the flow on one-dimensional channel, thereby significantly reducing the processing time of calculations. The model uses three different references to the physical properties of helium: expressions given by the KTA (German committee of nuclear safety standards), the computational tool REFPROP and a set of constant values for the entire channel. With the use of these three references it is possible to simulate the flow treating the gas both compressible and incompressible. The results showed very close values for the interest quantities and revealed that there are no significant differences in the use of different references used in the project. Another important conclusion to be observed is the independence of helium in the gas compressibility effects on thermal fluid dynamic behavior. The study also indicated that the gas undergoes no severe effects due to high temperature variations in the channel, since this goes in the channel at 914 K and exits at approximately 1263 K, which shows the excellent use of helium as a refrigerant fluid in reactor channels VHTGR. The comparison of results obtained in this work with others in the literature served to confirm the effectiveness of the one-dimensional consideration of method of gas flow in the coolant channel to replace the models made in 3-D for the pressure range

  13. REACTOR COOLING

    Science.gov (United States)

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  14. A Novel Dual-Stage Hydrothermal Flow Reactor

    DEFF Research Database (Denmark)

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

    2015-01-01

    The dual-stage reactor is a novel continuous flow reactor with two reactors connected in series. It is designed for hydrothermal flow synthesis of nanocomposites, in which a single particle consists of multiple materials. The secondary material may protect the core nanoparticle from oxidation...

  15. Development of a Linear Flow Channel Reactor for sulphur removal ...

    African Journals Online (AJOL)

    2013-09-23

    Sep 23, 2013 ... in acid mine wastewater treatment operations ... bed reactor treating a synthetic acid mine water (2 000 mg∙ℓ−1 Na2SO4 solution) and the Liner Flow Channel Reactors (sur- face area 1.1 m2 and ... Keywords: floating sulphur biofilms, acid mine drainage, AMD passive treatment, linear flow channel reactor,.

  16. Oscillating-Coolant Heat Exchanger

    Science.gov (United States)

    Scotti, Stephen J.; Blosser, Max L.; Camarda, Charles J.

    1992-01-01

    Devices useful in situations in which heat pipes inadequate. Conceptual oscillating-coolant heat exchanger (OCHEX) transports heat from its hotter portions to cooler portions. Heat transported by oscillation of single-phase fluid, called primary coolant, in coolant passages. No time-averaged flow in tubes, so either heat removed from end reservoirs on every cycle or heat removed indirectly by cooling sides of channels with another coolant. Devices include leading-edge cooling devices in hypersonic aircraft and "frost-free" heat exchangers. Also used in any situation in which heat pipe used and in other situations in which heat pipes not usable.

  17. Potential for low fracture toughness and lamellar tearing on PWR steam generator and reactor coolant pump supports. Resolution of generic technical activity A-12

    Energy Technology Data Exchange (ETDEWEB)

    Snaider, R.P.; Hodge, J.M.; Levin, H.A.; Zudans, J.J.

    1979-10-01

    This report summarizes work performed by the Nuclear Regulatory Commission staff and its contractor, Sandia Laboratories, in the resolution of Generic Technical Activity A-12, ''Potential for Low Fracture Toughness and Lamellar Tearing in PWR Steam Generator and Reactor Coolant Pump Supports.'' The report describes the technical issues, the technical studies performed by Sandia describes the technical issues, the technical studies performed by Sandia Laboratories, the NRC staff's technical positions based on these studies, and the staff's plan for implementing its technical positions. It also provides recommendations for further work. The complete technical input from Sandia Laboratories is appended to the report.

  18. The study of capability natural uranium as fuel cycle input for long life gas cooled fast reactors with helium as coolant

    Energy Technology Data Exchange (ETDEWEB)

    Ariani, Menik, E-mail: menikariani@gmail.com; Satya, Octavianus Cakra; Monado, Fiber [Department of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University, jl Palembang-Prabumulih km 32 Indralaya OganIlir, South of Sumatera (Indonesia); Su’ud, Zaki [Nuclear and Biophysics Research Division, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, jlGanesha 10, Bandung (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, 2-12-11N1-17 Ookayama, Meguro-Ku, Tokyo (Japan)

    2016-03-11

    The objective of the present research is to assess the feasibility design of small long-life Gas Cooled Fast Reactor with helium as coolant. GCFR included in the Generation-IV reactor systems are being developed to provide sustainable energy resources that meet future energy demand in a reliable, safe, and proliferation-resistant manner. This reactor can be operated without enrichment and reprocessing forever, once it starts. To obtain the capability of consuming natural uranium as fuel cycle input modified CANDLE burn-up scheme was adopted in this system with different core design. This study has compared the core with three designs of core reactors with the same thermal power 600 MWth. The fuel composition each design was arranged by divided core into several parts of equal volume axially i.e. 6, 8 and 10 parts related to material burn-up history. The fresh natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of the region (i) into region (i+1) region after the end of 10 years burn-up cycle. The calculation results shows that for the burn-up strategy on “Region-8” and “Region-10” core designs, after the reactors start-up the operation furthermore they only needs natural uranium supply to the next life operation until one period of refueling (10 years).

  19. Performance Evaluation of AI2O3/Water Nanofluid as Coolant in a Double-Tube Heat Exchanger Flowing under a Turbulent Flow Regime

    Directory of Open Access Journals (Sweden)

    Navid Bozorgan

    2012-01-01

    Full Text Available Nanofluids are expected to be a promising coolant candidate in chemical processes for water waste remediation and heat transfer system size reduction. This paper focuses on the potential mass flowrate reduction in exchanger with a given heat exchange capacity using nanofluids. Al2O3 nanoparticles with diameters of 7 nm dispersed in water with volume concentrations up to 2% are selected as a coolant, and their performance in a horizontal double-tube counterflow heat exchanger under turbulent flow conditions is numerically studied. The results show that the flowrate of nanofluid coolant decreases with the increase of concentration of nanoparticles in the exchanger with a given heat exchange capacity. The mass flowrate of the nanofluid at a volume concentration of 2 vol.% is approximately 24.5% lower than that of pure water (base fluid for given conditions. For the pressure drop, the results show that the pressure drop of nanofluid is slightly higher than water and increases with increase of volume concentrations. In addition, the reduction of wall temperature and heat transfer area is estimated.

  20. NEUTRONIC REACTOR CORE INSTRUMENT

    Science.gov (United States)

    Mims, L.S.

    1961-08-22

    A multi-purpose instrument for measuring neutron flux, coolant flow rate, and coolant temperature in a nuclear reactor is described. The device consists essentially of a hollow thimble containing a heat conducting element protruding from the inner wall, the element containing on its innermost end an amount of fissionsble materinl to function as a heat source when subjected to neutron flux irradiation. Thermocouple type temperature sensing means are placed on the heat conducting element adjacent the fissionable material and at a point spaced therefrom, and at a point on the thimble which is in contact with the coolant fluid. The temperature differentials measured between the thermocouples are determinative of the neutron flux, coolant flow, and temperature being measured. The device may be utilized as a probe or may be incorporated in a reactor core. (AE C)

  1. NUCLEAR REACTOR

    Science.gov (United States)

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

  2. Cooling Characteristics of the V-1650-7 Engine. 1; Coolant-Flow Distribution, Cylinder Temperatures, and Heat Rejections at Typical Operating Conditions

    Science.gov (United States)

    Povolny, John H.; Bogdan, Louis J.

    1947-01-01

    An investigation was conducted to determine the coolant-flow distribu tion, the cylinder temperatures, and the heat rejections of the V-165 0-7 engine . The tests were run a t several power levels varying from minimum fuel consumption to war emergency power and at each power l evel the coolant flows corresponded to the extremes of those likely t o be encountered in typical airplane installations, A mixture of 30-p ercent ethylene glycol and 70-percent water was used as the coolant. The temperature of each cylinder was measured between the exhaust val ves, between the intake valves, in the center of the head, on the exh aust-valve guide, at the top of the barrel on the exhaust side, and o n each exhaust spark-plug gasket. For an increase in engine power fro m 628 to approximately 1700 brake horsepower the average temperature for the cylinder heads between the exhaust valves increased from 437 deg to 517 deg F, the engine coolant heat rejection increased from 12 ,600 to 22,700 Btu. per minute, the oil heat rejection increased from 1030 to 4600 Btu per minute, and the aftercooler-coolant heat reject ion increased from 450 to 3500 Btu -per minute.

  3. A thermal-hydraulic drift-flux based mixture-fluid model for the description of single- and two-phase flow along a general coolant channel

    Energy Technology Data Exchange (ETDEWEB)

    Alois Hoeld [Bernaysstr. 16A, D-80937 Munich (Germany)

    2005-07-01

    constructed with the aim to relate the mean nodal and nodal boundary function values appearing in the nodalized differential eqs. and to yield necessary gradients at SC boundaries. It represents together with the very thoroughly tested packages for drift flux, heat transfer and single- and two-phase friction factors the central part of the theoretical model and module CCM and guarantees for the exact solution of the mixture-fluid eqs. The module has been derived on the basis of experiences of many years work with the development of an effective theoretical model and corresponding non-linear one-dimensional digital code UTSG-2 for vertical, natural-circulation U-tube steam generators as being used in PWR nuclear power plants. The general module CCM has now been used to replace, in a first step, the special coolant channel elements of the thoroughly tested U-tube steam generator code UTSG-2, constructing thus (among other improvements and additions) the advanced code version UTSG-3. Some of the most characteristic previous calculations with the UTSG-2 code could then be taken as benchmark cases. Post-calculating them showed good agreement, presenting a very convincing argument for the validation of the UTSG-3 code and thus also the underlying code package CCM. This universally applicable thermal-hydraulic code package CCM can and has been taken as a basic element for the simulation of the thermal-hydraulic situation of different complex systems (different types of steam generators, 3D nuclear reactor cores calculating the mass flow distribution into different parallel channels after non-symmetric perturbations), yielding very satisfactory results. The 3-eq. ('separate-region') mixture fluid model can be seen as a counterpart to the at present very much preferred 'separate-phase models' where each phase is treated separately, i.e., the 3 conservation eqs. are splitted into 4, 5 or even 6 field equations, connected by exchange (=closure) terms which represent

  4. Power flow control using distributed saturable reactors

    Science.gov (United States)

    Dimitrovski, Aleksandar D.

    2016-02-13

    A magnetic amplifier includes a saturable core having a plurality of legs. Control windings wound around separate legs are spaced apart from each other and connected in series in an anti-symmetric relation. The control windings are configured in such a way that a biasing magnetic flux arising from a control current flowing through one of the plurality of control windings is substantially equal to the biasing magnetic flux flowing into a second of the plurality of control windings. The flow of the control current through each of the plurality of control windings changes the reactance of the saturable core reactor by driving those portions of the saturable core that convey the biasing magnetic flux in the saturable core into saturation. The phasing of the control winding limits a voltage induced in the plurality of control windings caused by a magnetic flux passing around a portion of the saturable core.

  5. Features of Convective Turbulent Flows Behind the Mixing Spacer Grids of the TVS-Kvadrat of The PWR-Type Reactor

    Science.gov (United States)

    Dmitriev, S. M.; Varentsov, A. V.; Doronkov, D. V.; Legchanov, M. A.; Sorokin, V. D.; Khrobostov, A. E.

    2017-09-01

    This paper presents the results of experimental investigations of the influence of mixing spacer grids with different types of deflectors on the coolant flow in the TVS-Kvadrat fuel assembly of the PWR-type reactor. Investigations were conducted on an aerodynamic stand with the use of a multichannel pneumometric probe. Analysis of the spatial distribution of projections of the absolute flow velocity permitted detailing the coolant flow pattern behind the mixing spacer grid with various types of deflectors of mixing grids. The results obtained in the present paper are used to verify three-dimensional CFD programs, and in applied cell by cell codes they also serve as a database in calculations of thermotechnical reliability of the cores of PWR-type reactors with the TVS-Kvadrat.

  6. Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant; Avaliacao da integridade de varetas combustiveis em reatores PWR a partir da analise espectrometrica da agua do primario

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Iara Arraes

    1999-02-15

    The main objective of this thesis is to provide a better comprehension of the phenomena involved in the transport of fission products, from the fuel rod to the coolant of a PWR reactor. To achieve this purpose, several steps were followed. Firstly, it was presented a description of the fuel elements and the main mechanisms of fuel rod failure, indicating the most important nuclides and their transport mechanisms. Secondly, taking both the kinetic and diffusion models for the transport of fission products as a basis, a simple analytical and semi-empirical model was developed. This model was also based on theoretical considerations and measurements of coolant's activity, according to internationally adopted methodologies. Several factors are considered in the modelling procedures: intrinsic factors to the reactor itself, factors which depend on the reactor's operational mode, isotope characteristic factors, and factors which depend on the type of rod failure. The model was applied for different reactor's operational parameters in the presence of failed rods. The main conclusions drawn from the analysis of the model's output are relative to the variation on the coolant's water activity with the fuel burnup, the linear operation power and the primary purification rate and to the different behaviour of iodine and noble gases. The model was saturated from a certain failure size and showed to be unable to distinguish between a single big fail and many small ones. (author)

  7. Tracking of fuel particles after pin failure in nominal, loss-of-flow and shutdown conditions in the MYRRHA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Buckingham, Sophia; Planquart, Philippe [von Karman Institute, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Van Tichelen, Katrien [SCK- CEN, Boeretang 200, 2400 Mol (Belgium)

    2017-02-15

    Highlights: • Quantification of the design and safety of the MYRRHA reactor in the event of a pin failure. • Simulation of different accident scenarios in both forced and natural convection regime. • The accumulation areas at the free-surface in case of the least dense particles depend on the flow regime. • The densest particles form an important deposit at the bottom of the vessel. • Further study of the risk of core blockage requires a detailed model of the core. - Abstract: This work on fuel dispersion aims at quantifying the design and safety of the MYRRHA nuclear reactor. A number of accidents leading to the release of a secondary phase into the primary coolant loop are investigated. Among these scenarios, an incident leading to the failure of one or more of the fuel pins is simulated while the reactor is operating in nominal conditions, but also in natural convection regime either during accident transients such as loss-of-flow or during the normal shut-down of the reactor. Two single-phase CFD models of the MYRRHA reactor are constructed in ANSYS Fluent to represent the reactor in nominal and natural convection conditions. An Euler–Lagrange approach with one-way coupling is used for the flow and particle tracking. Firstly, a steady state RANS solution is obtained for each of the three conditions. Secondly, the particles are released downstream from the core outlet and particle distributions are provided over the coolant circuit. Their size and density are defined such that test cases represent potential extremes that may occur. Analysis of the results highlights different particle behaviors, depending essentially on gravity forces and kinematic effects. Statistical distributions highlight potential accumulation regions that may form at the free-surfaces, on top of the upper diaphragm plate or at the bottom of the vessel. These results help to localize regions of fuel accumulation in order to provide insight for development of strategies for

  8. Bypass valve and coolant flow controls for optimum temperatures in waste heat recovery systems

    Science.gov (United States)

    Meisner, Gregory P

    2013-10-08

    Implementing an optimized waste heat recovery system includes calculating a temperature and a rate of change in temperature of a heat exchanger of a waste heat recovery system, and predicting a temperature and a rate of change in temperature of a material flowing through a channel of the waste heat recovery system. Upon determining the rate of change in the temperature of the material is predicted to be higher than the rate of change in the temperature of the heat exchanger, the optimized waste heat recovery system calculates a valve position and timing for the channel that is configurable for achieving a rate of material flow that is determined to produce and maintain a defined threshold temperature of the heat exchanger, and actuates the valve according to the calculated valve position and calculated timing.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-29

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

  10. Boiling water neutronic reactor incorporating a process inherent safety design

    Science.gov (United States)

    Forsberg, C.W.

    1985-02-19

    A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (nonborated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two water volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

  11. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Directory of Open Access Journals (Sweden)

    Hwang Bae

    2017-08-01

    Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  12. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Hwang; Ryu, Sung Uk; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Eok [Dept. of Precision Mechanical Engineering, Kyungpook National University, Sangju (Korea, Republic of)

    2017-08-15

    Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  13. Thermal modeling in an engine cooling system to control coolant flow for fuel consumption improvement

    Science.gov (United States)

    Park, Sangki; Woo, Seungchul; Kim, Minho; Lee, Kihyung

    2017-04-01

    The design and evaluation of engine cooling and lubrication systems is generally based on real vehicle tests. Our goal here was to establish an engine heat balance model based on mathematical and interpretive analysis of each element of a passenger diesel engine cooling system using a 1-D numerical model. The purpose of this model is to determine ways of optimizing the cooling and lubrication components of an engine and then to apply these methods to actual cooling and lubrication systems of engines that will be developed in the future. Our model was operated under the New European Driving Cycle (NEDC) mode conditions, which represent the fuel economy evaluation mode in Europe. The flow rate of the cooling system was controlled using a control valve. Our results showed that the fuel efficiency was improved by as much as 1.23 %, cooling loss by 1.35 %, and friction loss by 2.21 % throughout NEDC modes by modification of control conditions.

  14. CONTINOUS MULTI-PHASE FLOW REACTOR FOR SMALL AND LARGE FLOW CAPACITIES THAN L/MIN

    NARCIS (Netherlands)

    Al-Rawashdeh, Ma'moun; Schouten, Jaap; Nijhuis, T. Alexander; Yue, Jun

    2014-01-01

    Multiphase flow processing in flow reactors holds great promises for diverse applications in fine chemicals and materials synthesis primarily due to its precise control over the flow, mixing and reaction inside or between each phase. Even though, flow reactors have shown superior performance, so far

  15. Activation of water coolant in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Khripunov, V.; Santoro, R.T.; Lida, H.; Parker, R.; Janeschitz, G.; Plenteda, R. [ITER Joint Central Team Garching, Muenchen (Germany)

    1998-07-01

    Water as been selected as the main coolant for the ITER blanket and vacuum vessel. Following exposure to DT neutrons, water becomes a source of high energy {sup 16}N-decay photons and energetic ({approx}0.9 MeV) {sup 17}N decay neutrons outside the reactor that lead to shielding problems during both reactor operation and after shutdown. As a result of comprehensive neutronic and hydraulic analyses, corresponding design measures were developed to diminish these effects. The use of a correlation between the {sup 16}N-production rate and the 14.1-MeV neutron flux in flowing water was recommended for determining fusion power by measuring water decay photons behind the radiation shield. (authors)

  16. NEUTRONIC REACTOR SHIELD AND SPACER CONSTRUCTION

    Science.gov (United States)

    Wigner, E.P.; Ohlinger, L.A.

    1958-11-18

    Reactors of the heterogeneous, graphite moderated, fluid cooled type and shielding and spacing plugs for the coolant channels thereof are reported. In this design, the coolant passages extend horizontally through the moderator structure, accommodating the fuel elements in abutting end-to-end relationship, and have access openings through the outer shield at one face of the reactor to facilitate loading of the fuel elements. In the outer ends of the channels which extend through the shields are provided spacers and shielding plugs designed to offer minimal reslstance to coolant fluid flow while preventing emanation of harmful radiation through the access openings when closed between loadings.

  17. Cooling system for a nuclear reactor

    Science.gov (United States)

    Amtmann, Hans H.

    1982-01-01

    A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

  18. Nuclear reactor control column

    Science.gov (United States)

    Bachovchin, Dennis M.

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

  19. Determination of the Coefficients of Heat Transfer and Friction in Supercritical-Pressure Nuclear Reactors with Account of the Intensity and Scale of Flow Turbulence on the Basis of the Theory of Stochastic Equations and Equivalence of Measures

    Science.gov (United States)

    Dmitrenko, A. V.

    2017-11-01

    New dependences of the coefficients of heat transfer and friction have been presented which are used for calculations in the reactor core at a supercritical water pressure by taking account of the parameters of perturbation in the coolant flow: of the intensity and scale of flow turbulence. These solutions have been obtained based on stochastic systems of equations for the turbulence and the equivalence of measures between deterministic (laminar) and random (turbulent) flows.

  20. Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric Richard; Durbin, Samuel G

    2007-04-01

    The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower-power assemblies. Specifically, this program provided data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin

    2013-10-22

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

  2. Chemical reactor modeling multiphase reactive flows

    CERN Document Server

    Jakobsen, Hugo A

    2014-01-01

    Chemical Reactor Modeling closes the gap between Chemical Reaction Engineering and Fluid Mechanics.  The second edition consists of two volumes: Volume 1: Fundamentals. Volume 2: Chemical Engineering Applications In volume 1 most of the fundamental theory is presented. A few numerical model simulation application examples are given to elucidate the link between theory and applications. In volume 2 the chemical reactor equipment to be modeled are described. Several engineering models are introduced and discussed. A survey of the frequently used numerical methods, algorithms and schemes is provided. A few practical engineering applications of the modeling tools are presented and discussed. The working principles of several experimental techniques employed in order to get data for model validation are outlined. The monograph is based on lectures regularly taught in the fourth and fifth years graduate courses in transport phenomena and chemical reactor modeling, and in a post graduate course in modern reactor m...

  3. Three dimensional calculations of the primary coolant flow in a 900 MW PWR vessel. Numerical simulation of the accurate RCP start-up flow rate

    Energy Technology Data Exchange (ETDEWEB)

    Martin, A.; Alvarez, D.; Cases, F.; Stelletta, S. [Electricite de France (EDF), 78 - Chatou (France). Lab. National d`Hydraulique

    1997-06-01

    This report explains the last results about the mixing in the 900 MW PWR vessels. The accurate fluid flow transient, induced by the RCP starting-up, is represented. In a first time, we present the Thermalhydraulic Finite Element Code N3S used for the 3D numerical computations. After that, results obtained for one reactor operation case are given. This case is dealing with the transient mixing of a clear plug in the vessel when one primary pump starts-up. A comparison made between two injection modes; a steady state fluid flow conditions or the accurate RCP transient fluid flow conditions. The results giving the local minimum of concentration and the time response of the mean concentration at the core inlet are compared. The results show the real importance of the unsteadiness characteristics of the fluid flow transport of the clear water plug. (author) 12 refs.

  4. Research of the fluid flow in a radially orientated coolant channel of a turbine blade; Untersuchung der Stroemung in einem radial gerichteten Kuehlkanal eines Turbinenlaufrades

    Energy Technology Data Exchange (ETDEWEB)

    Hein, O.

    1999-07-01

    Due to rotation (Coriolis forces) in a coolant channel a secondary flow is superimposed to the basic flow. This leads to a change in the local heat transfer over the surface of the coolant channel as well as a change in the overall value of the heat transfer. Also the pressure loss over the channel length will change by rotation. By means of computational fluid dynamics (Finite Element Method) it was achieved to figure out the interaction between changing fluid flow and heat transfer. To validate the results obtained by a numerical flow simulation, a new measurement technique was developed. A laser-two-focus velocimeter has been combined with a rotation prism which allows continued measurements in a rotating scaled up channel. (orig.) [German] Bedingt durch die Rotationsbewegung eines Kuehlkanals wird die Grundstroemung von einem Sekundaerwirbel ueberlagert (Corioliskraefte). Durch diese Einfluesse aendert sich sowohl der lokale Waermeuebergang ueber der Kanaloberflaeche als auch die globalen Waermeuebertragungsraten ueber dem gesamten Kanal. Ebenfalls aendert sich durch die Rotation der Druckverlust ueber der Kanallaenge. Durch eine numerische Stroemungssimulation (Finite-Element-Methode) war es moeglich, einen detaillierten Zusammenhang zwischen dem veraenderten Stroemungsverhalten und dem Waermeuebertragungsverhalten darzustellen. Um die numerisch gewonnenen Ergebnisse experimentell abzusichern, wurde eine neuartige Messtechnik entwickelt. Ein Laser-2-Fokus-Velozimeter wurde mit einem Bilddrehprisma kombiniert, und dies erlaubte eine kontinuierliche Messung in einem rotierenden vergroesserten Modellkanal. (orig.)

  5. The Helix reactor in flow chemistry

    NARCIS (Netherlands)

    Geerdink, P.; Bassett, J.M.

    2009-01-01

    In the fine chemicals industry, the batch reactor has been the established workhorse for carrying out reactions. Its main advantage here is being multi-purpose. The chemistry that fits with this kind of equipment is, however, limited to relatively slow and mild reactions because of the limited wall

  6. Development of a Linear Flow Channel Reactor for sulphur removal ...

    African Journals Online (AJOL)

    2013-09-23

    Sep 23, 2013 ... in acid mine wastewater treatment operations. JB Molwantwa1 and PD ... bed reactor treating a synthetic acid mine water (2 000 mg∙ℓ−1 Na2SO4 solution) and the Liner Flow Channel Reactors (sur- face area 1.1 m2 and 2.2 ... structures were reported for tannery wastewater evaporation ponds (Rose et al., ...

  7. Convective wave front locking for a reaction-diffusion system in a conical flow reactor

    DEFF Research Database (Denmark)

    Kuptsov, P.V.; Kuznetsov, S.P.; Knudsen, Carsten

    2002-01-01

    We consider reaction-diffusion instabilities in a flow reactor whose cross-section slowly expands with increasing longitudinal coordinate (cone shaped reactor). Due to deceleration of the flow in this reactor, the instability is convective near the inlet to the reactor and absolute at the downstr......We consider reaction-diffusion instabilities in a flow reactor whose cross-section slowly expands with increasing longitudinal coordinate (cone shaped reactor). Due to deceleration of the flow in this reactor, the instability is convective near the inlet to the reactor and absolute...

  8. Nitrate removal with lateral flow sulphur autotrophic denitrification reactor.

    Science.gov (United States)

    Lv, Xiaomei; Shao, Mingfei; Li, Ji; Xie, Chuanbo

    2014-01-01

    An innovative lateral flow sulphur autotrophic denitrification (LFSAD) reactor was developed in this study; the treatment performance was evaluated and compared with traditional sulphur/limestone autotrophic denitrification (SLAD) reactor. Results showed that nitrite accumulation in the LFSAD reactor was less than 1.0 mg/L during the whole operation. Denitrification rate increased with the increased initial alkalinity and was approaching saturation when initial alkalinity exceeded 2.5 times the theoretical value. Higher influent nitrate concentration could facilitate nitrate removal capacity. In addition, denitrification efficiency could be promoted under an appropriate reflux ratio, and the highest nitrate removal percentage was achieved under reflux ratio of 200%, increased by 23.8% than that without reflux. Running resistance was only about 1/9 of that in SLAD reactor with equal amount of nitrate removed, which was the prominent excellence of the new reactor. In short, this study indicated that the developed reactor was feasible for nitrate removal from waters with lower concentrations, including contaminated surface water, groundwater or secondary effluent of municipal wastewater treatment with fairly low running resistance. The innovation in reactor design in this study may bring forth new ideas of reactor development of sulphur autotrophic denitrification for nitrate-contaminated water treatment.

  9. Flow Reactor for studying Physicochemical and aging properties of SOA

    Science.gov (United States)

    Babar, Z. B.

    2016-12-01

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

  10. An Overview of Reactor Concepts, a Survey of Reactor Designs.

    Science.gov (United States)

    1985-02-01

    implies, this reactor uses gas as the primary coolant . The coolant has a higher exit temperature when leaving the core than the PWR water 6 AFWL-TN-84...nuclear reactors, coolants must be used to ensure material components are not subject to failure due to the temperature exceeding melting points...Reactors that use deu- terium (heavy water) as a coolant can use natural uranium as a fuel. The * Canadian reactor, CANDU , utilizes this concept

  11. Uncertainty analysis and flow measurements in an experimental mock-up of a molten salt reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, Bogdan; Aszodi, Attila [Budapest University of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2016-09-15

    In the paper measurement results from the experimental modelling of a molten salt reactor concept will be presented along with detailed uncertainty analysis of the experimental system. Non-intrusive flow measurements are carried out on the scaled and segmented mock-up of a homogeneous, single region molten salt fast reactor concept. Uncertainty assessment of the particle image velocimetry (PIV) measurement system applied with the scaled and segmented model is presented in detail. The analysis covers the error sources of the measurement system (laser, recording camera, etc.) and the specific conditions (de-warping of measurement planes) originating in the geometry of the investigated domain. Effect of sample size in the ensemble averaged PIV measurements is discussed as well. An additional two-loop-operation mode is also presented and the analysis of the measurement results confirm that without enhancement nominal and other operation conditions will lead to strong unfavourable separation in the core flow. It implies that use of internal flow distribution structures will be necessary for the optimisation of the core coolant flow. Preliminary CFD calculations are presented to help the design of a perforated plate located above the inlet region. The purpose of the perforated plate is to reduce recirculation near the cylindrical wall and enhance the uniformity of the core flow distribution.

  12. Inductive heating with magnetic materials inside flow reactors.

    Science.gov (United States)

    Ceylan, Sascha; Coutable, Ludovic; Wegner, Jens; Kirschning, Andreas

    2011-02-07

    Superparamagnetic nanoparticles coated with silica gel or alternatively steel beads are new fixed-bed materials for flow reactors that efficiently heat reaction mixtures in an inductive field under flow conditions. The scope and limitations of these novel heating materials are investigated in comparison with conventional and microwave heating. The results suggest that inductive heating can be compared to microwave heating with respect to rate acceleration. It is also demonstrated that a very large diversity of different reactions can be performed under flow conditions by using inductively heated flow reactors. These include transfer hydrogenations, heterocyclic condensations, pericyclic reactions, organometallic reactions, multicomponent reactions, reductive cyclizations, homogeneous and heterogeneous transition-metal catalysis. Silica-coated iron oxide nanoparticles are stable under many chemical conditions and the silica shell could be utilized for further functionalization with Pd nanoparticles, rendering catalytically active heatable iron oxide particles. Copyright © 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  13. On-Line Coolant Chemistry Analysis

    Energy Technology Data Exchange (ETDEWEB)

    LM Bachman

    2006-07-19

    Impurities in the gas coolant of the space nuclear power plant (SNPP) can provide valuable indications of problems in the reactor and an overall view of system health. By monitoring the types and amounts of these impurities, much can be implied regarding the status of the reactor plant. However, a preliminary understanding of the expected impurities is important before evaluating prospective detection and monitoring systems. Currently, a spectroscopy system is judged to hold the greatest promise for monitoring the impurities of interest in the coolant because it minimizes the number of entry and exit points to the plant and provides the ability to detect impurities down to the 1 ppm level.

  14. Design and fabrication of magnetic coolant filter

    Science.gov (United States)

    Prashanth, B. N.

    2017-07-01

    Now a day's use of coolants in industry has become dominant because of high production demands. Coolants not only help in speeding up the production but also provide many advantages in the metal working operation. As the consumption of coolants is very high a system is badly in need, so as to recirculate the used coolant. Also the amount of hazardous waste generated by industrial plants has become an increasingly costly problem for the manufactures and an additional stress on the environment. Since the purchase and disposal of the spent cutting fluids is becoming increasingly expensive, fluid recycling is a viable option for minimizing the cost. Separation of metallic chips from the coolants by using magnetic coolant separation has proven a good management and maintenance of the cutting fluid. By removing the metallic chips, the coolant life is greatly extended, increases the machining quality and reduces downtime. Above being the case, a magnetic coolant filter is developed which utilizes high energy permanent magnets to develop a dense magnetic field along a narrow flow path into which the contaminated coolant is directed. The ferromagnetic particles captured and aligned by the dense magnetic field, from the efficient filter medium. This enables the unit to remove ferromagnetic particles from the coolant. Magnetic coolant filters use the principle of magnetic separation to purify the used coolant. The developed magnetic coolant separation has the capability of purifying 40 litres per minute of coolant with the size of the contaminants ranging from 1 µm to 30 µm. The filter will be helpful in saving the production cost as the cost associated with the proposed design is well justified by the cost savings in production. The magnetic field produced by permanent magnets will be throughout the area underneath the reservoir. This produces magnetic field 30mm above the coolant reservoir. Very fine particles are arrested without slip. The magnetic material used will not

  15. Isothermal Reactor for Continuous Flow Microwave-Assisted Chemical Reaction

    Science.gov (United States)

    Matsuzawa, Mitsuhiro; Togashi, Shigenori; Hasebe, Shinji

    An isothermal reactor in which reaction solutions can be controlled at constant temperature under constant microwave irradiation was developed. This is useful for investigating microwave effects on chemical reactions that are not observed under conventional heating conditions. We devised a structure in which a heat-transfer medium with a low dielectric loss factor, which hardly absorbs any microwaves, flowed outside a spiral reaction tube and designed the basic structure of the reactor using electromagnetic simulation to optimize the energy absorption rate. The conditions for increasing the temperature controlling ability of the reactor were also investigated theoretically and experimentally by taking into consideration the influences of three elements: the velocity of the internal fluid, the material for the tube, and the velocity of the external fluid. The velocity of the external fluid had the greatest influence on temperature controlling ability and the material for the tube had the least influence under the experimental conditions. The overall heat transfer coefficient was about 3.9×102 W/(m2·K) when water flowed through the quartz reaction tube at 7.1 mm/s and the external fluid flowed outside the tube at 44 mm/s. We also tested and confirmed that the temperature of water used as internal fluid could be controlled to within ±1.5 K at 309.3 K when microwaves at 26 W were irradiated into the reactor, whereas the temperature of water was over 373 K and boiled without the heat-transfer medium flowing outside the reaction tube using a conventional method of microwave heating. In addition, we investigated microwave effects on Suzuki-Miyaura coupling reaction using the developed isothermal reactor and we confirmed that the temperatures were controlled well in the reactor. The yields obtained by microwave heating were almost the same as that obtained by oil-bath heating.

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

    African Journals Online (AJOL)

    aghomotsegin

    2013-06-05

    Jun 5, 2013 ... Key words: Composite wastewater, up-flow anaerobic sludge blanket (UASB), anaerobic biological treatment, biogas, granulated anaerobic ... collected from the top of the reactor to a flexible water-filled gasholder fitted with .... changes in microbial diversity in biogranules in response to changes in nature ...

  17. Gas and liquid distribution in the monolith film flow reactor

    NARCIS (Netherlands)

    Heibel, A.K.; Vergeldt, F.J.; As, van H.

    2003-01-01

    The gas-liquid distribution in a monolith film flow reactor is investigated in the scope of this work. Magnetic resonance imaging (MRI) and a customized liquid collection method hate been successfully applied to determine the liquid distribution over the monolith cross-section. Using a

  18. Entrained Flow Reactor Test of Potassium Capture by Kaolin

    DEFF Research Database (Denmark)

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

    2015-01-01

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

  19. Detailed Design of the Safety Residual Heat Removal System and a Circulation Pump for the KIJANG Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Hyungi; Seo, Kyoungwoo; Kim, Seonghoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Primary cooling system (PCS) circulates the coolant from the reactor core to the heat exchanger. Therefore the heat generated from the fuel assembly in the reactor core is removed continuously. The PCS is designed based on the required thermal design flow rate of the reactor core, uncertainty of measuring instruments and the safe functions. Primary coolant is generally dumped into the pool and goes to the reactor core through the flow guide. The fission heat generated from the fuel assembly is transferred to the coolant, and then heated coolant goes to the PCS equipment room in order to remove the heat through the heat exchanger. SRHSR is designed based on the required flow rate and system constraints. Centrifugal pump of Case 1 with a non-dimensional specific speed of 0.97 [-] and specific diameter of 3.33 [-] is chosen as the SRHRS pump for the KJRR.

  20. THE CONTROL ALGORITHM OF THE DRYING PROCESS PARTICULATE MATERIALS IN THE APPARATUS WITH THE SWIRLING FLOW OF COOLANT AND MICROWAVE ENERGY SUPPLY

    Directory of Open Access Journals (Sweden)

    S. T. Antipov

    2015-01-01

    Full Text Available The technical task of the process is to improve the drying quality of the final product, increasing the precision and reliability of control, the reduction of specific energy consumption. One of the ways to improve the process is complex and i ts local automation. This paper deals with the problems of development and creation of a new control algorithm drying process of the particulate material. Identified a number of shortcomings of the existing methods of automatic control of the process. As a result, the authors proposed a method for drying particulate materials in the device with swirling flow and the microwave energy supply and its automatic control algorithm. The description of the operating principle of the drying apparatus consists in that the particulate material is wet by using a tangential flow of coolant supplied to the cylinder-drying apparatus which also serves the axial coolant flow, whereby the heat transfer fluid with the particulate material begins to undergo a complex circular movement along the circumference apparatus, thereby increasing its speed and its operation control algorithm. The work of this scheme is carried out at three levels of regulation on the basis of determining the coefficient of efficiency of the dryer, which makes it possible to determine the optimal value of the power equipment and to forecast the cost of electricity. All of the above allows you to get ready for a high quality product while minimizing thermal energy and material costs by optimizing the operating parameters of the drying of the particulate material in the dryer with a combined microwave energy supply and ensure the rational use of heat energy by varying their quantity depending on the characteristics to be dried particulate material and the course of the process.

  1. Containment for the low temperature district nuclear-heating reactor

    Energy Technology Data Exchange (ETDEWEB)

    He Shuyan (Institute of Nuclear Energy Technology, Tsinghua Univ., Beijing (China)); Dong Duo (Institute of Nuclear Energy Technology, Tsinghua Univ., Beijing (China))

    1993-04-01

    An integral arrangement is adopted for the Low Temperature District Nuclear-Heating Reactor. The primary heat exchangers, control rod drives and spent fuel elements are put in the reactor pressure vessel together with the reactor core. The primary coolant flows in natural circulation through the reactor core and the primary heat exchangers. The primary coolant pipes penetrating the wall of the reactor pressure vessel are all of small diameters. The reactor vessel constitutes the main part of the pressure boundary of the primary coolant. Therefore a small sized metallic containment closed to the wall of the reactor vessel can be used for the reactor. Design principles and functions of the containment are the same as for the containment of a PWR. But the adoption of a small sized containment brings about some benefits such as a short period of manufacturing, relatively low cost, and ease for sealing. A loss of primary coolant accident would not be happening during a rupture accident of the primary coolant pressure boundary inside the containment owing to its intrinsic safety. (orig.).

  2. Limiting photocurrent analysis of a wide channel photoelectrochemical flow reactor

    Science.gov (United States)

    Davis, Jonathan T.; Esposito, Daniel V.

    2017-03-01

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

  3. Measured gas and particle temperatures in VTT's entrained flow reactor

    DEFF Research Database (Denmark)

    Clausen, Sønnik; Sørensen, L.H.

    2006-01-01

    Particle and gas temperature measurements were carried out in experiments on VTTs entrained flow reactor with 5% and 10% oxygen using Fourier transform infrared emission spectroscopy (FTIR). Particle temperature measurements were performed on polish coal,bark, wood, straw particles, and bark...... and wood particles treated with additive. A two-color technique with subtraction of the background light was used to estimate particle temperatures during experiments. A transmission-emission technique was used tomeasure the gas temperature in the reactor tube. Gas temperature measurements were in good...

  4. Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

    Directory of Open Access Journals (Sweden)

    Po Hu

    2014-01-01

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

  5. Kinetics and Product Selectivity (Yield) of Second Order Competitive Consecutive Reactions in Fed-Batch Reactor and Plug Flow Reactor

    OpenAIRE

    Selvamony, Subash Chandra Bose

    2013-01-01

    This literature compares the performance of second order competitive consecutive reaction in Fed-Batch Reactor with that in continuous Plug Flow Reactor. In a kinetic sense, this simulation study aims to develop a case for continuous Plug Flow Reactor in pharmaceutical, fine chemical, and related other chemical industries. MATLAB is used to find solutions for the differential equations. The simulation results show that, for certain cases of nonelementary scenario, product selectivity is highe...

  6. Enhanced model predictive control of a catalytic flow reversal reactor

    Energy Technology Data Exchange (ETDEWEB)

    Devals, C.; Bertrand, F.; Perrier, M. [Ecole Polytechnique, Montreal, PQ (Canada). Dept. de Genie Chimique; Fuxman, A.; Forbes, J.F.; Hayes, R.E. [Alberta Univ., Edmonton, AB (Canada). Dept. of Chemical and Materials Engineering

    2009-08-15

    The removal of atmospheric methane by conversion to carbon dioxide has the potential to significantly reduce the greenhouse gas (GHG) effect. Methane can be burned using conventional or catalytic combustion. Different types of reactors can be used for catalytic combustion, including the catalytic flow reversal reactor (CFRR) which has drawn much attention because auto-thermal operation can be achieved for lean low temperature feed. However, the control of CFRR is challenging. This study presented a method to predict the stationary state for the reactor. The method can be incorporated into a model predictive control (MPC) strategy as a terminal constant. The study involved a numerical simulation of the catalytic combustion of lean methane in a CFRR. In particular, the combustion of lean methane air mixtures in a CFRR was examined using a two dimensional heterogeneous continuum model, based on mole and energy balance equations for the solid (the inert and catalytic sections of the reactor) and the fluid phases. Several simulations were performed to study the reactor performance. The results showed the impact on the methane conversion and the maximum temperature in the reactor of key process parameters, such as the methane inlet concentration, the superficial gas velocity, the switching time, and the mass extraction rate. A simple empirical model was created to predict the maximum temperature and conversion of methane in the reactor at stationary state. Simulations revealed an improvement in control performance when adding a constraint for the maximum temperature. The improved results showed better performance in terms of heat extraction and smoothness of operation at low and high inlet concentrations. 23 refs., 4 tabs., 14 figs.

  7. Flow instability in particle-bed nuclear reactors

    Science.gov (United States)

    Kerrebrock, Jack L.

    1993-01-01

    The particle-bed core offers mitigation of some of the problems of solid-core nuclear rocket reactors. Dividing the fuel elements into small spherical particles contained in a cylindrical bed through which the propellant flows radially, may reduce the thermal stress in the fuel elements, allowing higher propellant temperatures to be reached. The high temperature regions of the reactor are confined to the interior of cylindrical fuel assemblies, so most of the reactor can be relatively cool. This enables the use of structural and moderating materials which reduce the minimum critical size and mass of the reactor. One of the unresolved questions about this concept is whether the flow through the particle-bed will be well behaved, or will be subject to destructive flow instabilities. Most of the recent analyses of the stability of the particle-bed reactor have been extensions of the approach of Bussard and Delauer, where the bed is essentially treated as an array of parallel passages, so that the mass flow is continuous from inlet to outlet through any one passage. A more general three dimensional model of the bed is adopted, in which the fluid has mobility in three dimensions. Comparison of results of the earlier approach to the present one shows that the former does not accurately represent the stability at low Re. The more complete model presented should be capable of meeting this deficiency while accurately representing the effects of the cold and hot frits, and of heat conduction and radiation in the particle-bed. It can be extended to apply to the cylindrical geometry of particle-bed reactors without difficulty. From the exemplary calculations which were carried out, it can be concluded that a particle-bed without a cold frit would be subject to instability if operated at the high temperatures desired for nuclear rockets, and at power densities below about 4 megawatts per liter. Since the desired power density is about 40 megawatts per liter, it can be concluded

  8. Comparative analysis of coolants for FBR of future nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I. [State Scientific Center of Russian Federation, Institute for Physics and Power Engineering named after Academician A.I. Leipusky, Kaluga Region (Russian Federation)

    2001-07-01

    Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR.

  9. State of Art Report for the Bypass and Cross Flows in Prismatic Modular High-Temperature Gas-Cooled Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Ji Su; Kim, Min Hwan

    2010-01-15

    The horizontal and vertical gaps between the adjacent fuel blocks occurs significantly in the prismatic modular high-temperature gas-cooled reactor core due to the thermal expansion, irradiation expansion/shrinking, and fuel block column bowing by pressure and gravity during the plant operation, in addition to the initial manufacture/design tolerance of fuel/reflector blocks. The coolant leakage of helium gas occurs through the gaps. These bypass and cross flows highly impact on the effective core cooling flow rate. This report describes the technical state of the bypass and cross flow study, based on ten reference papers, reviewing and summarizing four classified contents in the viewpoints of 'The reactor core fluctuation data for the first identification of the importance of the bypass and cross flow', 'The cross flow test and evaluation data', 'The flow test and evaluation data of the seal mechanism to prevent the leakage flow', and 'The reactor core thermal-fluid analysis and evaluation dat000.

  10. Cyclonic Flows for Reactor Applications with High Engulfment Levels

    OpenAIRE

    Chinnici, Alfonso

    2014-01-01

    This thesis aims at designing, developing and optimizing novel cyclonic vortex reactor configurations in the context of innovative technologies development for increasing the efficiency and reducing the environmental impact of energy systems. Furthermore, not depending on the specific engineering application, the thesis aims at improving our knowledge on cyclonic vortex flows, clarifying several aspects of their aerodynamic. In particular, the interaction mechanism between a jet and the vorte...

  11. Numerical approach to plug-flow activated sludge reactor kinetics.

    Science.gov (United States)

    Muslu, Y

    2000-07-01

    In this study, general relationships applicable to a wide variety of microbial-mediated treatment processes are developed using a number of continuous-flow stirred tank reactors-in-series. Simultaneous equations resulting from mass balance on substrate and biomass are solved numerically taking the longitudinal biomass gradient into account. A relationship between substrate and biomass concentrations is developed as a function of hydraulic residence time. Monod kinetics and mathematical models of reactors-in-series are used to represent the actual conditions resulting from varying degrees of axial dispersion and wastewater quality. Dimensionless quantities are used to reduce the number of parameters to be taken into account. Computer techniques are applied to express the results generally. By means of the numerical approach, the variation of the ratio between the hydraulic residence time of a given reactor and that of an equivalent plug-flow tank with the same inlet and outlet conditions is investigated. Results are expressed in the form of graphs to characterize the plug-flow activated sludge systems and to provide a basis of design. Experimental data reported in the literature are also evaluated to demonstrate the cases where the existing classical solution to the problem differs from the actual results obtained from the computer program and the design graphs given.

  12. Parametric study on maximum transportable distance and cost for thermal energy transportation using various coolants

    Energy Technology Data Exchange (ETDEWEB)

    Su-Jong Yoon; Piyush Sabharwall

    2014-07-01

    The operation temperature of advanced nuclear reactors is generally higher than commercial light water reactors and thermal energy from advanced nuclear reactor can be used for various purposes such as district heating, desalination, hydrogen production and other process heat applications, etc. The process heat industry/facilities will be located outside the nuclear island due to safety measures. This thermal energy from the reactor has to be transported a fair distance. In this study, analytical analysis was conducted to identify the maximum distance that thermal energy could be transported using various coolants such as molten-salts, helium and water by varying the pipe diameter and mass flow rate. The cost required to transport each coolant was also analyzed. The coolants analyzed are molten salts (such as: KClMgCl2, LiF-NaF-KF (FLiNaK) and KF-ZrF4), helium and water. Fluoride salts are superior because of better heat transport characteristics but chloride salts are most economical for higher temperature transportation purposes. For lower temperature water is a possible alternative when compared with He, because low pressure He requires higher pumping power which makes the process very inefficient and economically not viable for both low and high temperature application.

  13. Measurement and Analysis of Structural Integrity of Reactor Core Support Structure in Pressurized Water Reactor (PWR) Plant

    Science.gov (United States)

    Ansari, Saleem A.; Haroon, Muhammad; Rashid, Atif; Kazmi, Zafar

    2017-02-01

    Extensive calculation and measurements of flow-induced vibrations (FIV) of reactor internals were made in a PWR plant to assess the structural integrity of reactor core support structure against coolant flow. The work was done to meet the requirements of the Fukushima Response Action Plan (FRAP) for enhancement of reactor safety, and the regulatory guide RG-1.20. For the core surveillance measurements the Reactor Internals Vibration Monitoring System (IVMS) has been developed based on detailed neutron noise analysis of the flux signals from the four ex-core neutron detectors. The natural frequencies, displacement and mode shapes of the reactor core barrel (CB) motion were determined with the help of IVMS. The random pressure fluctuations in reactor coolant flow due to turbulence force have been identified as the predominant cause of beam-mode deflection of CB. The dynamic FIV calculations were also made to supplement the core surveillance measurements. The calculational package employed the computational fluid dynamics, mode shape analysis, calculation of power spectral densities of flow & pressure fields and the structural response to random flow excitation forces. The dynamic loads and stiffness of the Hold-Down Spring that keeps the core structure in position against upward coolant thrust were also determined by noise measurements. Also, the boron concentration in primary coolant at any time of the core cycle has been determined with the IVMS.

  14. Advanced neutron source reactor probabilistic flow blockage assessment

    Energy Technology Data Exchange (ETDEWEB)

    Ramsey, C.T.

    1995-08-01

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

  15. SIMMER-III applications to fuel-coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    Morita, K.; Kondo, Sa.; Tobita, Y.; Brear, D.J. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-01-01

    The main purpose of the SIMMER-III code is to provide a numerical simulation of complex multiphase, multicomponent flow problems essential to investigate core disruptive accidents in liquid-metal fast reactors (LMFRs). However, the code is designed to be sufficiently flexible to be applied to a variety of multiphase flows, in addition to LMFR safety issues. In the present study, some typical experiments relating to fuel-coolant interactions (FCIs) have been analyzed by SIMMER-III to demonstrate that the code is applicable to such complex and highly transient multiphase flow situations. It is shown that SIMMER-III can reproduce the premixing phase both in water and sodium systems as well as the propagation of steam explosion. It is thus demonstrated the code is basically capable of simulating integral multiphase thermal-hydraulic problems included in FCI experiments. (author)

  16. Characteristics of Butanol Isomers Oxidation in a Micro Flow Reactor

    KAUST Repository

    Bin Hamzah, Muhamad Firdaus

    2017-05-01

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

  17. Numerical simulation of gas volume motion during the gas injection into liquid metal coolant

    Science.gov (United States)

    Usov, E. V.; Lobanov, P. D.; Pribaturin, N. A.; Chuhno, V. I.; Kutlimetov, A. E.; Svetonosov, A. I.

    2017-09-01

    Detailed description of relations and numerical approaches to simulate transport of gas phase in a vertical liquid column is presented in a current paper. These approaches are important to calculate phenomena that take place during steam generator tube rapture in fast reactors with liquid metal coolant. Presented relations determine interphase friction between gas and fluid in different flow regimes of two-phase flow. It is shown that correct definition of interphase friction coefficients determines the correct value of bubble velocity that is very important to simulate two-phase flow in steam generator and reactor core. The paper also contains numerical algorithm to calculate motion of gas volume in fluid flow. Especial attention is paid to describe the algorithms for simulating two-phase flow with sharp edges between phases that are character for slug flow regime. Also some experimental results are presented in the paper. Comparison between experimental data and calculation results has been provided.

  18. Dynamic Behavior of Reverse Flow Reactor for Lean Methane Combustion

    Directory of Open Access Journals (Sweden)

    Yogi W. Budhi

    2014-09-01

    Full Text Available The stability of reactor operation for catalytic oxidation of lean CH4 has been investigated through modeling and simulation, particularly the influence of switching time and heat extraction on reverse flow reactor (RFR performance. A mathematical model of the RFR was developed, based on one-dimensional pseudo-homogeneous model for mass and heat balances, incorporating heat loss through the reactor wall. The configuration of the RFR consisted of inert-catalyst-inert, with or without heat extraction that makes it possible to store the energy released by the exothermic reaction of CH4 oxidation. The objective of this study was to investigate the dynamic behavior of the RFR for lean methane oxidation and to find the optimum condition by exploring a stability analysis of the simple reactor. The optimum criteria were defined in terms of CH4 conversion, CH4 slip, and heat accumulation in the RFR. At a switching time of 100 s, the CH4 conversion reached the maximum value, while the CH4 slip attained its minimum value. The RFR could operate autothermally with positive heat accumulation, i.e. 0.02 J/s. The stability of the RFR in terms of heat accumulation was achieved at a switching time of 100 s.

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

    NARCIS (Netherlands)

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

    1992-01-01

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

  20. Modeling of melt-coolant mixing by bottom injection

    Energy Technology Data Exchange (ETDEWEB)

    Kazachkov, I.V.; Paladino, D.; Sehgal, B.R. [Royal Inst. of Tech., Div. of Nuclear Power Safety, Stockholm (Sweden)

    2001-07-01

    In this paper, the flow characteristics during the coolant injection, with submerged nozzles, at the bottom of a molten pool are studied. The flow pattern developed by the rising coolant is considered for the case of complete coolant vaporization, and the pool-coolant phase distributions are assessed by a modeling approach delivered from literature for a heterogeneous turbulent jet. To calculate the basic characteristics of such flow, integral relationships are proposed for the two-phase boundary layer. The results of numerical computations and approximate solution are compared with the experimental data obtained in the low temperature experiments, conducted in the DECOBI (debris coolability by bottom injection) facility. (authors)

  1. Hydrothermal Processing of Macroalgal Feedstocks in Continuous-Flow Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-02-03

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

  2. Solution of heat removal from nuclear reactors by natural convection

    Directory of Open Access Journals (Sweden)

    Zitek Pavel

    2014-03-01

    Full Text Available This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR.The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-15

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

  4. Treatment of domestic wastewater in an up-flow anaerobic sludge blanket reactor followed by moving bed biofilm reactor

    NARCIS (Netherlands)

    Tawfik, A.; El-Gohary, F.; Temmink, B.G.

    2010-01-01

    The performance of a laboratory-scale sewage treatment system composed of an up-flow anaerobic sludge blanket (UASB) reactor and a moving bed biofilm reactor (MBBR) at a temperature of (22-35 A degrees C) was evaluated. The entire treatment system was operated at different hydraulic retention times

  5. Safe reactor power

    Energy Technology Data Exchange (ETDEWEB)

    Menegus, R.L.; Ring, H.F.; Bernath, L.

    1956-05-15

    The upper limit on reactor operating power is established not only by safety considerations during steady-state operation but also by the requirement that during an accident no permanent damage be inflicted upon the reactor or the fuel charge. Two general categories of accidents are recognized; they are the ``nuclear runaway`` and the ``loss of coolant flow`` incidents. In this memorandum an incident of the latter type is analyzed. It is assumed that the safety rods function normally, and a method is defined for establishing the highest operating power that may be permitted if the postulated accident is to do no damage.

  6. Fluid flow separation in a reactor pressure vessel during an ECC injection. Single phase flow and two phase flow (air-water) experimental results

    Energy Technology Data Exchange (ETDEWEB)

    Thierry Bichet; Alain Martin [EDF - Research and Development Division - Fluid Mechanics and Heat Transfert 6, quai Watier - B.P. 49 - 78401 Chatou CEDEX 01 (France); Frederic Beaud [EDF/ Industry - Basic Design Department., 12-14, Avenue Dutrievoz 69628 Villeurbanne CEDEX (France)

    2005-07-01

    Full text of publication follows: Within the framework of the nuclear power plant lifetime issue, the assessment of the French 900 MWe (3-loops) series reactor pressure vessel (RPV) integrity has been performed. A simplified analysis has shown that the most severe loading conditions are given by the small break loss of coolant accidents due to the pressurized injection of cold water (9 deg. C) into the cold leg and down comer of the RPV. During these transient scenarios, single or two-phase (uncovered cold leg) flows have been shown in the cold leg, depending on the crack size and RPV model (900 MWe or 1300 MWe). An experimental study has been carried out, on the one hand, to consolidate the numerical results obtained with CFD home code (Code-Saturne) which mainly showed the stratified flow in the cold leg and the fluid flow separation and its oscillations in the down comer during a single phase scenario. These physical phenomena are important for the thermal RPV loading assessment. On the other hand, the absence of experimental two-phase data necessitated to carry out an experimental study around the mixing area behavior (free surface, stratified flow) during an ECC injection with an uncovered cold leg. The new EDF R and D mock up, called HYBISCUS, is a facility which is made out of Plexiglas (atmosphere pressure) and represents a half scale CP0 geometry with one cold leg and part of the down comer. The mock up modularity allows us to insert representative ECC nozzles and a thermal shield. In reference to the reactor scenarios, the experimental operating conditions are derived from the conservation of the density effects (Froude number). For that, a heated salted water flow is used to represent the ECC injection whereas water represents the cold leg fluid. This mock up has been defined in order to represent single phase flow (cold leg and down comer full of water) or two-phase flow (uncovered cold leg) ECC scenarios. This paper reports experimental results

  7. Two-phase flow experiments on Counter-Current Flow Limitation in a model of the hot leg of a pressurized water reactor (2015 test series)

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Matthias; Lucas, Dirk; Pietruske, Heiko; Szalinski, Lutz

    2016-12-15

    Counter-Current Flow Limitation (CCFL) is of importance for PWR safety analyses in several accident scenarios connected with loss of coolant. Basing on the experiences obtained during a first series of hot leg tests now new experiments on counter-current flow limitation were conducted in the TOPFLOW pressure vessel. The test series comprises air-water tests at 1 and 2 bar as well as steam-water tests at 10, 25 and 50 bar. During the experiments the flow structure was observed along the hot leg model using a high-speed camera and web-cams. In addition pressure was measured at several positions along the horizontal part and the water levels in the reactor-simulator and steam-generator-simulator tanks were determined. This report documents the experimental setup including the description of operational and special measuring techniques, the experimental procedure and the data obtained. From these data flooding curves were obtained basing on the Wallis parameter. The results show a slight shift of the curves in dependency of the pressure. In addition a slight decrease of the slope was found with increasing pressure. Additional investigations concern the effects of hysteresis and the frequencies of liquid slugs. The latter ones show a dependency on pressure and the mass flow rate of the injected water. The data are available for CFD-model development and validation.

  8. Transverse flow reactor studies of the dynamics of radical reactions

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, R.G. [Argonne National Laboratory, IL (United States)

    1993-12-01

    Radical reactions are in important in combustion chemistry; however, little state-specific information is available for these reactions. A new apparatus has been constructed to measure the dynamics of radical reactions. The unique feature of this apparatus is a transverse flow reactor in which an atom or radical of known concentration will be produced by pulsed laser photolysis of an appropriate precursor molecule. The time dependence of individual quantum states or products and/or reactants will be followed by rapid infrared laser absorption spectroscopy. The reaction H + O{sub 2} {yields} OH + O will be studied.

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

    Science.gov (United States)

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

  10. Modeling phototrophic biofilms in a plug-flow reactor.

    Science.gov (United States)

    Muñoz Sierra, J D; Picioreanu, C; van Loosdrecht, M C M

    2014-01-01

    The use of phototrophic biofilms in wastewater treatment has been recognized as a potential option for development of new reactor configurations. For better understanding of these systems, a numerical model was developed including relevant microbial processes. As a novelty, this model was implemented in COMSOL Multiphysics, a modern computational environment for complex dynamic models. A two-dimensional biofilm model was used to study the spatial distribution of microbial species within the biofilm and along the length of the reactor. The biofilm model was coupled with a one-dimensional plug-flow bulk liquid model. The impact of different operational conditions on the chemical oxygen demand (COD) and ammonia conversions was assessed. The model was tuned by varying two parameters: the half-saturation coefficient for light use by phototrophs and the oxygen mass transfer coefficient. The mass transfer coefficient was found to be determining for the substrate conversion rate. Simulations indicate that heterotrophs would overgrow nitrifiers and phototrophs within the biofilm until a low biodegradable COD value in the wastewater is reached (organic loading rate reactor performance.

  11. 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.

  12. Reclamation and disposal of water-based machining coolants

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, P.A.

    1982-01-01

    The Oak Ridge Y-12 Plant, which is operated by the Union Carbide Corporation, Nuclear Division for the Department of Energy under US government contract W-7405-eng-26, currently uses about 10{sup 6} L/yr (260,000 gal/yr) of water-based coolants in its machining operations. These coolants are disposed of in a 110,000-L (29,000-gal) activated sludge reactor. The reactor has oxidized an average of 38.6 kg of total organic carbon (TOC) per day with an overall efficiency of 90%. The predominant bacteria in the reactor have been identified once each year for the past three years. Six primary types of water-based coolants are currently used in the machine shops. In order to reduce the coolant usage rate, efforts are being made to introduce one universal coolant into the shops. By using a biocide to limit bacterial deterioration and using a filter and centrifuge system to remove dirt and tramp oils from the coolant, the coolant discard rate can be greatly reduced. 1 tab.

  13. Investigation of heat transfer in liquid-metal flows under fusion-reactor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Poddubnyi, I. I., E-mail: poddubnyyii@nikiet.ru [Joint Stock Company Dollezhal Research and Development Institute of Power Engineering (JSC NIKIET) (Russian Federation); Pyatnitskaya, N. Yu.; Razuvanov, N. G.; Sviridov, V. G.; Sviridov, E. V. [Russian Academy of Science, Joint Institute of High Temperatures (Russian Federation); Leshukov, A. Yu. [Joint Stock Company Dollezhal Research and Development Institute of Power Engineering (JSC NIKIET) (Russian Federation); Aleskovskiy, K. V. [National Research University Moscow Power Engineering Institute (MPEI) (Russian Federation); Obukhov, D. M. [Joint Stock Company Efremov Institute of Electrophysical Apparatus (Russian Federation)

    2016-12-15

    The effect discovered in studying a downward liquid-metal flow in vertical pipe and in a channel of rectangular cross section in, respectively, a transverse and a coplanar magnetic field is analyzed. In test blanket modules (TBM), which are prototypes of a blanket for a demonstration fusion reactor (DEMO) and which are intended for experimental investigations at the International Thermonuclear Experimental Reactor (ITER), liquid metals are assumed to fulfil simultaneously the functions of (i) a tritium breeder, (ii) a coolant, and (iii) neutron moderator and multiplier. This approach to testing experimentally design solutions is motivated by plans to employ, in the majority of the currently developed DEMO blanket projects, liquid metals pumped through pipes and/or rectangular channels in a transvers magnetic field. At the present time, experiments that would directly simulate liquid-metal flows under conditions of ITER TBM and/or DEMO blanket operation (irradiation with thermonuclear neutrons, a cyclic temperature regime, and a magnetic-field strength of about 4 to 10 T) are not implementable for want of equipment that could reproduce simultaneously the aforementioned effects exerted by thermonuclear plasmas. This is the reason why use is made of an iterative approach to experimentally estimating the performance of design solutions for liquid-metal channels via simulating one or simultaneously two of the aforementioned factors. Therefore, the investigations reported in the present article are of considerable topical interest. The respective experiments were performed on the basis of the mercury magneto hydrodynamic (MHD) loop that is included in the structure of the MPEI—JIHT MHD experimental facility. Temperature fields were measured under conditions of two- and one-sided heating, and data on averaged-temperature fields, distributions of the wall temperature, and statistical fluctuation features were obtained. A substantial effect of counter thermo gravitational

  14. Lead Coolant Test Facility Technical and Functional Requirements, Conceptual Design, Cost and Construction Schedule

    Energy Technology Data Exchange (ETDEWEB)

    Soli T. Khericha

    2006-09-01

    This report presents preliminary technical and functional requirements (T&FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic. Based on review of current world lead or lead-bismuth test facilities and research need listed in the Generation IV Roadmap, five broad areas of requirements of basis are identified: Develop and Demonstrate Prototype Lead/Lead-Bismuth Liquid Metal Flow Loop Develop and Demonstrate Feasibility of Submerged Heat Exchanger Develop and Demonstrate Open-lattice Flow in Electrically Heated Core Develop and Demonstrate Chemistry Control Demonstrate Safe Operation and Provision for Future Testing. These five broad areas are divided into twenty-one (21) specific requirements ranging from coolant temperature to design lifetime. An overview of project engineering requirements, design requirements, QA and environmental requirements are also presented. The purpose of this T&FRs is to focus the lead fast reactor community domestically on the requirements for the next unique state of the art test facility. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 420oC. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M. It is also estimated that the facility will require two years to be constructed and ready for operation.

  15. Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code

    Directory of Open Access Journals (Sweden)

    Patrícia A. L. Reis

    2015-01-01

    Full Text Available Cases of core coolant flow blockage transient have been simulated and analysed for the TRIGA IPR-R1 research reactor using the RELAP5-MOD3.3 code. The transients are related to partial and to total obstruction of the core coolant channels. The reactor behaviour after the loss of flow was analysed as well as the changes in the coolant and fuel temperatures. The behaviour of the thermal hydraulic parameters from the transient simulations was analysed. For a partial blockage, it was observed that the reactor reaches a new steady state operation with new values for the thermal hydraulic parameters. The total core blockage brings the reactor to an abnormal operation causing increase in core temperature.

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

    NARCIS (Netherlands)

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

    1987-01-01

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

  17. Fast reactor: an experimental study of thermohydraulic processes in different operating regimes

    Science.gov (United States)

    Opanasenko, A. N.; Sorokin, A. P.; Zaryugin, D. G.; Trufanov, A. A.

    2017-05-01

    Results of integrated water model studies of temperature fields and a flow pattern of a nonisothermal primary coolant in the elements of the fast neutron reactor (hereinafter, fast reactor) primary circuit with primary sodium in different regimes, such as forced circulation (FC), transition to the reactor cooldown and emergency cooldown with natural coolant convection, are presented. It is shown that, under the influence of lift forces on the nonisothermal coolant flow in the upper chamber at the periphery of its bottom region over the side shields, a stable cold coolant isothermal zone is formed, whose dimensions increase with increase of total water flowrate. An essential and stable coolant temperature stratification is detected in the peripheral area of the upper (hot) chamber over the side shields, in the pressure and cold side chambers, in the elevator baffle, in the cooling system of the reactor vessel, and in the outlet of intermediate and autonomous heat exchangers in different operating regimes. Large gradients and temperature fluctuations are registered at the interface of stratified and recycling formations. In all of the studied cooldown versions, the coolant outlet temperature at the core fuel assembly is decreased and the coolant temperature in the peripheral zone of the upper chamber is increased compared to the FC. High performance of a passive emergency cooldown system of a fast reactor (BN-1200) with submersible autonomous heat exchangers (AHE) is confirmed. Thus, in a normal operation regime, even in case of malfunction of three submersible AHEs, the temperature of the equipment inside the reactor remains within acceptable limits and decay heat removal from the reactor does not exceed safe operation limits. The obtained results can be used both for computer code verification and for approximate estimate of the reactor plant parameters on the similarity criteria basis.

  18. Embrittlement and Flow Localization in Reactor Structural Materials

    Energy Technology Data Exchange (ETDEWEB)

    Xianglin Wu; Xiao Pan; James Stubbins

    2006-10-06

    Many reactor components and structural members are made from metal alloys due, in large part, to their strength and ability to resist brittle fracture by plastic deformation. However, brittle fracture can occur when structural material cannot undergo extensive, or even limited, plastic deformation due to irradiation exposure. Certain irradiation conditions lead to the development of a damage microstructure where plastic flow is limited to very small volumes or regions of material, as opposed to the general plastic flow in unexposed materials. This process is referred to as flow localization or plastic instability. The true stress at the onset of necking is a constant regardless of the irradiation level. It is called 'critical stress' and this critical stress has strong temperature dependence. Interrupted tensile testes of 316L SS have been performed to investigate the microstructure evolution and competing mechanism between mechanic twinning and planar slip which are believed to be the controlling mechanism for flow localization. Deformation twinning is the major contribution of strain hardening and good ductility for low temperatures, and the activation of twinning system is determined by the critical twinning stress. Phases transform and texture analyses are also discussed in this study. Finite element analysis is carried out to complement the microstructural analysis and for the prediction of materaials performance with and without stress concentration and irradiation.

  19. Effect of flow characteristics on online sterilization of cheese whey in UV reactors.

    Science.gov (United States)

    Singh, J P; Ghaly, A E

    2007-07-01

    An ultraviolet (UV) coil reactor was designed and used for the online sterilization of cheese whey. Its microbial destruction efficiency was compared to that of the conventionally used UV reactor. Both reactors have the same geometry (840 ml volume and 17 mm gap size) and were tested at 11 flow rates of 5, 10, 15, 20, 25, 30, 35, 40, 50, 60, and 70 ml/min. The results obtained from this study showed that despite of its high turbidity, cheese whey could be sterilized using UV radiation if the proper reactor design and flow rate are used. The performances of the UV reactors were governed by the flow rate and the hydraulics of flow inside the reactor. The flow was laminar in both the reactors, as the Reynolds number was in the range of 1.39-20.10. The phenomenon of Dean Flow was observed in the coil reactor and the Dean number was in the range of 1.09-15.41. Dean vortices resulted in higher microbial destruction efficiency in the coil reactor in a shorter retention time. The rate of microbial destruction was found to be exponential in the conventional reactor and polynomial in the coil reactor. Increasing the flow rate from 5 ml/min to 70 ml/min decreased the microbial destruction efficiency of the conventional reactor from 99.40 to 31.58%, while the microbial destruction efficiency in the coil reactor increased from 60.77% at the flow rate of 5 ml/min to 99.98% at the flow rate of 30 ml/min and then decreased with further increases in flow rate reaching 46.2% at the flow rate of 70 ml/min. The maximum effluent temperatures in the conventional and coil reactors were 45.8 and 46.1 degrees C, respectively. Fouling in the coil reactor was significantly less compared to the conventional reactor. The extent of fouling was influenced by flow rate and reactor's hydraulics.

  20. Statistical analysis of the blowdown phase of a loss-of-coolant accident in a pressurized water reactor as calculated by RELAP4/MOD6

    Energy Technology Data Exchange (ETDEWEB)

    Berman, M.; Byers, R.K.; Steck, G.P.

    1979-01-01

    A statistical study is presented of the blowdown phase of a design basis accident (double-ended cold leg guillotine break) in the Zion pressurized water reactor. The response surface method was employed to generate a polynomial approximation of the peak clad temperatures calculated by RELAP4/MOD6. The nodalization was a modification of the RELAP model of Zion developed in the BE/EM study. Twenty one variables were initially selected for the study. These variables, their ranges and distributions resulted from the best engineering judgement of NRC, Sandia, INFL, and other interested and knowledgeable investigators.

  1. Advanced reactor engineering with 3D printing for the continuous-flow synthesis of silver nanoparticles

    OpenAIRE

    Okafor, Obinna; Weilhard, Andreas; Fernandes, Jesum A.; Karjalainen, Erno; Goodridge, Ruth; Sans, Victor

    2017-01-01

    The implementation of advanced reactor engineering concepts employing additive manufacturing is demonstrated. The design and manufacturing of miniaturised continuous flow oscillatory baffled reactors (mCOBR) employing low cost stereolithography based 3D printing is reported for the first time. Residence time distribution experiments have been employed to demonstrate that these small scale reactors offer improved mixing conditions at a millimetre scale, when compared to tubular reactors. Nearl...

  2. COUNTERCURRENT FLOW LIMITATION EXPERIMENTS AND MODELING FOR IMPROVED REACTOR SAFETY

    Energy Technology Data Exchange (ETDEWEB)

    Vierow, Karen

    2008-09-26

    This project is investigating countercurrent flow and “flooding” phenomena in light water reactor systems to improve reactor safety of current and future reactors. To better understand the occurrence of flooding in the surge line geometry of a PWR, two experimental programs were performed. In the first, a test facility with an acrylic test section provided visual data on flooding for air-water systems in large diameter tubes. This test section also allowed for development of techniques to form an annular liquid film along the inner surface of the “surge line” and other techniques which would be difficult to verify in an opaque test section. Based on experiences in the air-water testing and the improved understanding of flooding phenomena, two series of tests were conducted in a large-diameter, stainless steel test section. Air-water test results and steam-water test results were directly compared to note the effect of condensation. Results indicate that, as for smaller diameter tubes, the flooding phenomena is predominantly driven by the hydrodynamics. Tests with the test sections inclined were attempted but the annular film was easily disrupted. A theoretical model for steam venting from inclined tubes is proposed herein and validated against air-water data. Empirical correlations were proposed for air-water and steam-water data. Methods for developing analytical models of the air-water and steam-water systems are discussed, as is the applicability of the current data to the surge line conditions. This report documents the project results from July 1, 2005 through June 30, 2008.

  3. Turbulence coefficients and stability studies for the coaxial flow or dissimiliar fluids. [gaseous core nuclear reactors

    Science.gov (United States)

    Weinstein, H.; Lavan, Z.

    1975-01-01

    Analytical investigations of fluid dynamics problems of relevance to the gaseous core nuclear reactor program are presented. The vortex type flow which appears in the nuclear light bulb concept is analyzed along with the fluid flow in the fuel inlet region for the coaxial flow gaseous core nuclear reactor concept. The development of numerical methods for the solution of the Navier-Stokes equations for appropriate geometries is extended to the case of rotating flows and almost completes the gas core program requirements in this area. The investigations demonstrate that the conceptual design of the coaxial flow reactor needs further development.

  4. Spatial distribution of nanoparticles in PWR nanofluid coolant subjected to local nucleate boiling

    Energy Technology Data Exchange (ETDEWEB)

    Mirghaffari, Reza; Jahanfarnia, Gholamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering

    2016-12-15

    Nanofluids have shown to be promising as an alternative for a PWR reactor coolant or as a safety system coolant to cover the core in the event of a loss of coolant accident. The nanoparticles distribution and neutronic parameters are intensively affected by the local boiling of nanofluid coolant. The main goal of this study was the physical-mathematical modeling of the nanoparticles distribution in the nucleate boiling of nanofluids within the viscous sublayer. Nanoparticles concentration, especially near the heat transfer surfaces, plays a significant role in the enhancement of thermal conductivity of nanofluids and prediction of CHF, Hide Out and Return phenomena. By solving the equation of convection-diffusion for the liquid phase near the heating surface and the bulk stream, the effect of heat flux on the distribution of nanoparticles was studied. The steady state mass conservation equations for liquids, vapors and nanoparticles were written for the flow boiling within the viscous sublayer adjacent the fuel cladding surface. The derived differential equations were discretized by the finite difference method and were solved numerically. It was found out that by increasing the surface heat flux, the concentration of nanoparticles increased.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-10-01

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

  6. Environmentally Friendly Coolant System

    Energy Technology Data Exchange (ETDEWEB)

    David Jackson Principal Investigator

    2011-11-08

    Energy reduction through the use of the EFCS is most improved by increasing machining productivity. Throughout testing, nearly all machining operations demonstrated less land wear on the tooling when using the EFCS which results in increased tool life. These increases in tool life advance into increased productivity. Increasing productivity reduces cycle times and therefore reduces energy consumption. The average energy savings by using the EFCS in these machining operations with these materials is 9%. The advantage for end milling stays with flood coolant by about 6.6% due to its use of a low pressure pump. Face milling and drilling are both about 17.5% less energy consumption with the EFCS than flood coolant. One additional result of using the EFCS is improved surface finish. Certain machining operations using the EFCS result in a smoother surface finish. Applications where finishing operations are required will be able to take advantage of the improved finish by reducing the time or possibly eliminating completely one or more finishing steps and thereby reduce their energy consumption. Some machining operations on specific materials do not show advantages for the EFCS when compared to flood coolants. More information about these processes will be presented later in the report. A key point to remember though, is that even with equivalent results, the EFCS is replacing petroleum based coolants whose production produces GHG emissions and create unsafe work environments.

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

    Science.gov (United States)

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

    2017-08-01

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

  8. Design optimization of metal getter reactors for removing tritium from flowing gas streams

    Energy Technology Data Exchange (ETDEWEB)

    Nobile, A.; Bieniewski, T.; Frame, K.; Little, R.; Fisher, K. [Los Alamos National Lab., NM (United States)

    1995-10-01

    A reaction engineering approach was used to design a SAES St 198 metal getter reactor for a glovebox detritiation system. The detritiation system will be used to decontaminate and decommission an Li(D, T)-contaminated glovebox previously used in the U.S. nuclear weapons program. The approach involved development of a model that calculates reactor breakthrough curves as a function of various reactor physical parameters. Experiments involving flow of deuterium in nitrogen through a small metal getter reactor validated the model. The model was then used to investigate the effects of temperature, getter pellet size, reactor diameter, and reactor volume on the reactor performance. The resulting design was a 7 cm diam. by 40 cm long cylindrical reactor that operates at 250 {degree}C, and is filled with 5 kg of as-received SAES St 198 getter pellets. The reactor handles a flow rate of 100 L/min. An St 909 getter reactor was used upstream of the St 198 reactor for impurity removal and water decomposition. The glovebox cleanup system design and getter reactor mechanical design are discussed. 14 refs., 6 figs.

  9. Regularities pertinent to formation of hydraulic nonuniformities at the outlet from the reactor plant header system

    Science.gov (United States)

    Gabrianovich, B. N.; Del'nov, V. N.

    2014-05-01

    Maintaining the preset distribution of coolant mass flow rate at the outlet from the header system organized as a back turn with lateral admission of coolant to the header and central removal of coolant from it is one of the conditions essential for securing reliable and efficient operation of a reactor plant. In the header system, coolant passes through the annular lateral channel formed by the vessel and central barrel, after which it enters into the header, changes the motion direction in it, and goes out through the channels of the grid placed in the central barrel. The results obtained from experimental investigations of the hydrodynamics in the header system flow paths carried out on models with flat and cylindrical shapes are presented. A previously unknown regularity pertinent to formation of hydraulic irregularities at the header system outlet is revealed, and semi-empirical correlations for determining the coolant mass flow rate distribution at the header system outlet are obtained. The regularity connected with formation of hydrodynamic nonuniformities at the header system outlet lies in the fact that the position of the maximum coolant velocity at the grid outlet coincides with the position of the maximum velocity in the core of the incident jet attacking the grid, and that the maximal average coolant velocity in the grid holes is proportional to the average coolant velocity in the incident jet.

  10. Test facility for investigation of heat transfer of promising coolants for the nuclear power industry

    Science.gov (United States)

    Belyaev, I. A.; Sviridov, V. G.; Batenin, V. M.; Biryukov, D. A.; Nikitina, I. S.; Manchkha, S. P.; Pyatnitskaya, N. Yu.; Razuvanov, N. G.; Sviridov, E. V.

    2017-11-01

    The results are presented of experimental investigations into liquid metal heat transfer performed by the joint research group consisting of specialist in heat transfer and hydrodynamics from NIU MPEI and JIHT RAS. The program of experiments has been prepared considering the concept of development of the nuclear power industry in Russia. This concept calls for, in addition to extensive application of water-cooled, water-moderated (VVER-type) power reactors and BN-type sodium cooled fast reactors, development of the new generation of BREST-type reactors, fusion power reactors, and thermonuclear neutron sources. The basic coolants for these nuclear power installations will be heavy liquid metals, such as lead and lithium-lead alloy. The team of specialists from NRU MPEI and JIHT RAS commissioned a new RK-3 mercury MHD-test facility. The major components of this test facility are a unique electrical magnet constructed at Budker Nuclear Physics Institute and a pressurized liquid metal circuit. The test facility is designed for investigating upward and downward liquid metal flows in channels of various cross-sections in a transverse magnetic field. A probe procedure will be used for experimental investigation into heat transfer and hydrodynamics as well as for measuring temperature, velocity, and flow parameter fluctuations. It is generally adopted that liquid metals are the best coolants for the Tokamak reactors. However, alternative coolants should be sought for. As an alternative to liquid metal coolants, molten salts, such as fluorides of lithium and beryllium (so-called FLiBes) or fluorides of alkali metals (so-called FLiNaK) doped with uranium fluoride, can be used. That is why the team of specialists from NRU MPEI and JIHT RAS, in parallel with development of a mercury MHD test facility, is designing a test facility for simulating molten salt heat transfer and hydrodynamics. Since development of this test facility requires numerical predictions and verification

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

    Science.gov (United States)

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

    2017-03-01

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

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

    Science.gov (United States)

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

    2017-06-22

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

  13. Transition between trickle flow and pulse flow in a cocurrent gas-liquid trickle-bed reactor at elevated pressures

    NARCIS (Netherlands)

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

    1992-01-01

    The effect of reactor pressure in the range of 0.2–2.0 MPa on the transition between the trickle-flow and the pulse-flow regime has been investigated for the non-foaming water—nitrogen and aqueous 40% ethyleneglycol—nitrogen systems. Most models and flow charts which are all based on atmospheric

  14. Hydrothermal upgrading of algae paste in a continuous flow reactor.

    Science.gov (United States)

    Patel, Bhavish; Hellgardt, Klaus

    2015-09-01

    This investigation demonstrates the utility of a novel laboratory scale continuous plug flow reactor for fast Hydrothermal Liquefaction (HTL) of microalgae in a quartz lined chamber. Reactions were carried out between 300 and 380 °C and residence times of 0.5-4 min. Cyclohexane was used as a co-solvent to enhance extraction and prevent char formation. Highest biocrude yield of 38 wt.% was achieved at 380 °C and 30 s as well as Water Soluble Fraction containing up to 60 wt.% matter recovered. Analysis of the biocrude showed that the extent of deoxygenation and denitrogenation after HTL varied and is dependent on the reaction conditions, Fourier Transform Infrared Spectroscopy analysis showed that biocrude contains similar functional moieties with only a small difference observed at different reaction conditions. Conversely, the Simulated Distillation and Size Exclusion Chromatography data showed that harsher conditions produced marginally better biocrude with improved boiling point profile and lower molecular weight compounds, respectively which was confirmed using Gas Chromatography-Mass Spectrometry. Copyright © 2015 Elsevier Ltd. All rights reserved.

  15. A novel reverse flow reactor coupling endothermic and exothermic reactions. Part II: sequential reactor configuration for reversible endothermic reactions

    NARCIS (Netherlands)

    van Sint Annaland, M.; Scholts, H.A.R.; Kuipers, J.A.M.; van Swaaij, Willibrordus Petrus Maria

    2002-01-01

    The new reactor concept for highly endothermic reactions at elevated temperatures with possible rapid catalyst deactivation based on the indirect coupling of endothermic and exothermic reactions in reverse flow, developed for irreversible reactions in Part I, has been extended to reversible

  16. MERCHANT MARINE SHIP REACTOR

    Science.gov (United States)

    Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

    1961-05-01

    A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

  17. Automatically scramming nuclear reactor system

    Science.gov (United States)

    Ougouag, Abderrafi M.; Schultz, Richard R.; Terry, William K.

    2004-10-12

    An automatically scramming nuclear reactor system. One embodiment comprises a core having a coolant inlet end and a coolant outlet end. A cooling system operatively associated with the core provides coolant to the coolant inlet end and removes heated coolant from the coolant outlet end, thus maintaining a pressure differential therebetween during a normal operating condition of the nuclear reactor system. A guide tube is positioned within the core with a first end of the guide tube in fluid communication with the coolant inlet end of the core, and a second end of the guide tube in fluid communication with the coolant outlet end of the core. A control element is positioned within the guide tube and is movable therein between upper and lower positions, and automatically falls under the action of gravity to the lower position when the pressure differential drops below a safe pressure differential.

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

    NARCIS (Netherlands)

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

    2006-01-01

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

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

    NARCIS (Netherlands)

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

    2005-01-01

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

  20. 6-flow reactor : Catalyst testing in a multiphase-parallel packed-microreactor system

    NARCIS (Netherlands)

    Herk, D. van

    2009-01-01

    This thesis deals with the design and operation of a multiple-parallel powder-packed reactor with continuous gas and liquid flows for kinetic tests. The reactor design is discussed in detail. Subsequently, axial dispersion in a multiphase packed microreactor is treated. Also, it is described how to

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

    Energy Technology Data Exchange (ETDEWEB)

    Flach, G.P.

    1999-02-24

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

  2. NEUTRONIC REACTORS

    Science.gov (United States)

    Wigner, E.P.; Young, G.J.

    1958-10-14

    A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

  3. The Caltech Photooxidation Flow Tube reactor: design, fluid dynamics and characterization

    Science.gov (United States)

    Huang, Yuanlong; Coggon, Matthew M.; Zhao, Ran; Lignell, Hanna; Bauer, Michael U.; Flagan, Richard C.; Seinfeld, John H.

    2017-03-01

    Flow tube reactors are widely employed to study gas-phase atmospheric chemistry and secondary organic aerosol (SOA) formation. The development of a new laminar-flow tube reactor, the Caltech Photooxidation Flow Tube (CPOT), intended for the study of gas-phase atmospheric chemistry and SOA formation, is reported here. The present work addresses the reactor design based on fluid dynamical characterization and the fundamental behavior of vapor molecules and particles in the reactor. The design of the inlet to the reactor, based on computational fluid dynamics (CFD) simulations, comprises a static mixer and a conical diffuser to facilitate development of a characteristic laminar flow profile. To assess the extent to which the actual performance adheres to the theoretical CFD model, residence time distribution (RTD) experiments are reported with vapor molecules (O3) and submicrometer ammonium sulfate particles. As confirmed by the CFD prediction, the presence of a slight deviation from strictly isothermal conditions leads to secondary flows in the reactor that produce deviations from the ideal parabolic laminar flow. The characterization experiments, in conjunction with theory, provide a basis for interpretation of atmospheric chemistry and SOA studies to follow. A 1-D photochemical model within an axially dispersed plug flow reactor (AD-PFR) framework is formulated to evaluate the oxidation level in the reactor. The simulation indicates that the OH concentration is uniform along the reactor, and an OH exposure (OHexp) ranging from ˜ 109 to ˜ 1012 molecules cm-3 s can be achieved from photolysis of H2O2. A method to calculate OHexp with a consideration for the axial dispersion in the present photochemical system is developed.

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

    African Journals Online (AJOL)

    aghomotsegin

    2013-06-05

    Jun 5, 2013 ... from The Company of Sugar and Integrated Industries, Hawamdia complex, Giza, Egypt. The final wastewater stream comprises of wastewaters coming from yeast, ethyl alcohol and distillation industries. Bioreactor. The UASB reactor was made of polyvinyl chloride (PVC). The working volume of the reactor ...

  5. Systems and methods for enhancing isolation of high-temperature reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per F.

    2017-09-26

    A high-temperature containment-isolation system for transferring heat from a nuclear reactor containment to a high-pressure heat exchanger is presented. The system uses a high-temperature, low-volatility liquid coolant such as a molten salt or a liquid metal, where the coolant flow path provides liquid free surfaces a short distance from the containment penetrations for the reactor hot-leg and the cold-leg, where these liquid free surfaces have a cover gas maintained at a nearly constant pressure and thus prevent high-pressures from being transmitted into the reactor containment, and where the reactor vessel is suspended within a reactor cavity with a plurality of refractory insulator blocks disposed between an actively cooled inner cavity liner and the reactor vessel.

  6. Wave model for longitudinal dispersion: application to the laminar-flow tubular reactor

    NARCIS (Netherlands)

    Kronberg, Alexandre E.; Benneker, A.H.; Benneker, A.H.; Westerterp, K.R.

    1996-01-01

    The wave model for longitudinal dispersion, published elsewhere as an alternative to the commonly used dispersed plug-flow model, is applied to the classic case of the laminar-flow tubular reactor. The results are compared in a wide range of situations to predictions by the dispersed plug-flow model

  7. THE STRIPPING OF AMMONIA FROM LANDFILL LEACHATE IN PLUG FLOW REACTORS

    National Research Council Canada - National Science Library

    Danuza Costa Campos; Valderi Duarte Leite; Wilton Silva Lopes; Patrícia Carvalho de Aquino Ramos

    2010-01-01

    ... where this chemical will be removed from the leachate by mass transfer of the liquid phase to gas. Therefore, this paper addressed to study the ammonia stripping process in leachates liquids using plug-flow reactors in series...

  8. Safety analysis of switching between reductive and oxidative conditions in a reaction coupling reverse flow reactor.

    NARCIS (Netherlands)

    van Sint Annaland, M.; Kuipers, J.A.M.; van Swaaij, Willibrordus Petrus Maria

    2001-01-01

    A new reverse flow reactor is developed where endothermic reactants (propane dehydrogenation) and exothermic reactants (fuel combustion) are fed sequentially to a monolithic catalyst, while periodically alternating the inlet and outlet positions. Upon switching from reductive to oxidative conditions

  9. From discovery to production: Scale-out of continuous flow meso reactors

    Directory of Open Access Journals (Sweden)

    Peter Styring

    2009-06-01

    Full Text Available A continuous flow parallel reactor system has been developed to provide a rapid and seamless transition from the discovery phase and production phase of chemical synthesis, particularly in low volume-high value pharmaceuticals production. Using a single fixed bed catalytic meso reactor, reactions can be screened on a small discovery scale over short time scales. The intensified process produces sufficient material for a full analysis. By replication of the single reactor in parallel, the same chemistry can be achieved on a larger scale, on a small footprint and without the mass and heat transport limitations of reactor scale-out in batch.

  10. Performance of down-flow hanging sponge (DHS) reactor coupled with up-flow anaerobic sludge blanket (UASB) reactor for treatment of onion dehydration wastewater.

    Science.gov (United States)

    El-Kamah, Hala; Mahmoud, Mohamed; Tawfik, Ahmed

    2011-07-01

    In this study, a promising system consisting of up-flow anaerobic sludge blanket (UASB) reactor followed by down-flow hanging sponge (DHS) reactor was investigated for onion dehydration wastewater treatment. Laboratory experiments were conducted at two different phases, i.e., phase (1) at overall hydraulic retention time (HRT) of 11h (UASB reactor: 6h and DHS reactor: 5h) and phase (2) at overall HRT of 9.4h (UASB reactor: 5.2h and DHS reactor: 4.2h). Long-term operation results of the proposed system showed that its overall TCOD, TBOD, TSS, TKN and NH(4)-N removal efficiencies were 92 ± 5, 95 ± 2, 95 ± 2, 72 ± 6 and 99 ± 1.3%, respectively (phase 1). Corresponding values for the 2nd phase were 85.4 ± 5, 86 ± 3, 87 ± 6, 65 ± 8 and 95 ± 2.8%. Based on the available results, the proposed system could be more viable option for treatment of wastewater generated from onion dehydration industry in regions with tropical or sub-tropical climates and with stringent discharge standards. Copyright © 2011 Elsevier Ltd. All rights reserved.

  11. Efficient [2+2] photocycloadditions under equimolar conditions by employing a continuous UV-flow reactor

    OpenAIRE

    Conradi, Matthias; Junkers, Thomas

    2013-01-01

    The [2 + 2] photocycloaddition between maleimide and various alkenes was optimized in a simple custom-made UV flow reactor. While complete maleimide conversion is only achieved with excesses of at least 10 eq. of alkene and reaction times of 12-24 h in batch, complete reactions with yields in the range >98% could be obtained in 5 min in the flow reactor under strictly equimolar conditions. Functional alkenes carrying allyl, alcohol, amine and ether moieties were successfully conjugated in goo...

  12. A model of microbial growth in a plug flow reactor with wall attachment.

    Science.gov (United States)

    Ballyk, M; Smith, H

    1999-05-01

    A mathematical model of microbial growth for limiting nutrient in a plug flow reactor which accounts for the colonization of the reactor wall surface by the microbes is formulated and studied analytically and numerically. It can be viewed as a model of the large intestine or of the fouling of a commercial bio-reactor or pipe flow. Two steady state regimes are identified, namely, the complete washout of the microbes from the reactor and the successful colonization of both the wall and bulk fluid by the microbes. Only one steady state is stable for any particular set of parameter values. Sharp and explicit conditions are given for the stability of each, and for the long term persistence of the bacteria in the reactor.

  13. A study on the characteristics of alternative coolants

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Ji Young; Kim, B. H.; Kim, T. J.; Jeong, K. C.; Choi, Y. D.; Choi, J. H.; Hwang, S. T

    2000-12-01

    The role of the coolant in liquid metal fast breeder reactor is very important for reasons of system safety. Recently, it has revealed that lead and lead-bismuth alloy show good safety characteristics as a fast reactor coolant compared to the sodium, such as low chemical activity, high boiling temperature and more negative void coefficient. So many countries take interest in these metals. The objectives of this project are to study the characteristics of heavy liquid metals(lead, lead-bismuth alloy) and to provide valuble information useful for the estimate the possibilities of its as the alternative coolant materials. An intensive research was performed into the global development status, basic properties, safety assurance methods, and direction of research in the futures and so on.

  14. Nuclear reactor with internal thimble-type delayed neutron detection system

    Science.gov (United States)

    Gross, Kenny C.; Poloncsik, John; Lambert, John D. B.

    1990-01-01

    This invention teaches improved apparatus for the method of detecting a breach in cladded fuel used in a nuclear reactor. The detector apparatus is located in the primary heat exchanger which conveys part of the reactor coolant past at least three separate delayed-neutron detectors mounted in this heat exchanger. The detectors are spaced apart such that the coolant flow time from the core to each detector is different, and these differences are known. The delayed-neutron activity at the detectors is a function of the delay time after the reaction in the fuel until the coolant carrying the delayed-neutron emitter passes the respective detector. This time delay is broken down into separate components including an isotopic holdup time required for the emitter to move through the fuel from the reaction to the coolant at the breach, and two transit times required for the emitter now in the coolant to flow from the breach to the detector loop and then via the loop to the detector. At least two of these time components are determined during calibrated operation of the reactor. Thereafter during normal reactor operation, repeated comparisons are made by the method of regression approximation of the third time component for the best-fit line correlating measured delayed-neutron activity against activity that is approximated according to specific equations. The equations use these time-delay components and known parameter values of the fuel and of the part and emitting daughter isotopes.

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

    Directory of Open Access Journals (Sweden)

    AZHARI T. I. MOHD. GHAZI

    2008-08-01

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

  16. Comparison of the PLTEMP code flow instability predictions with measurements made with electrically heated channels for the advanced test reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Feldman, E. (Nuclear Engineering Division)

    2011-06-09

    of the fuel element burnout is due to a form of flow instability. Whittle and Forgan provide a formula that predicts when this flow instability will occur. This formula is included in the PLTEMP/ANL code.Error! Reference source not found. Olson has shown that the PLTEMP/ANL code accurately predicts the powers at which flow instability occurs in the Whittle and Forgan experiments. He also considered the electrically heated tests performed in the ANS Thermal-Hydraulic Test Loop at ORNL and report by M. Siman-Tov et al. The purpose of this memorandum is to demonstrate that the PLTEMP/ANL code accurately predicts the Croft and the Waters tests. This demonstration should provide sufficient confidence that the PLTEMP/ANL code can adequately predict the onset of flow instability for the converted MURR. The MURR core uses light water as a coolant, has a 24-inch active fuel length, downward flow in the core, and an average core velocity of about 7 m/s. The inlet temperature is about 50 C and the peak outlet is about 20 C higher than the inlet for reactor operation at 10 MW. The core pressures range from about 4 to about 5 bar. The peak heat flux is about 110 W/cm{sup 2}. Section 2 describes the mechanism that causes flow instability. Section 3 describes the Whittle and Forgan formula for flow instability. Section 4 briefly describes both the Croft and the Waters experiments. Section 5 describes the PLTEMP/ANL models. Section 6 compares the PLTEMP/ANL predictions based on the Whittle and Forgan formula with the Croft measurements. Section 7 does the same for the Waters measurements. Section 8 provides the range of parameters for the Whittle and Forgan tests. Section 9 discusses the results and provides conclusions. In conclusion, although there is no single test that by itself closely matches the limiting conditions in the MURR, the preponderance of measured data and the ability of the Whittle and Forgan correlation, as implemented in PLTEMP/ANL, to predict the onset of flow

  17. A dynamic plug flow reactor model for a vanadium redox flow battery cell

    Science.gov (United States)

    Li, Yifeng; Skyllas-Kazacos, Maria; Bao, Jie

    2016-04-01

    A dynamic plug flow reactor model for a single cell VRB system is developed based on material balance, and the Nernst equation is employed to calculate cell voltage with consideration of activation and concentration overpotentials. Simulation studies were conducted under various conditions to investigate the effects of several key operation variables including electrolyte flow rate, upper SOC limit and input current magnitude on the cell charging performance. The results show that all three variables have a great impact on performance, particularly on the possibility of gassing during charging at high SOCs or inadequate flow rates. Simulations were also carried out to study the effects of electrolyte imbalance during long term charging and discharging cycling. The results show the minimum electrolyte flow rate needed for operation within a particular SOC range in order to avoid gassing side reactions during charging. The model also allows scheduling of partial electrolyte remixing operations to restore capacity and also avoid possible gassing side reactions during charging. Simulation results also suggest the proper placement for cell voltage monitoring and highlight potential problems associated with setting the upper charging cut-off limit based on the inlet SOC calculated from the open-circuit cell voltage measurement.

  18. The state-of-the-art report for flow and pressure measurement techniques in the piping system

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Ho; Lee, Jeong Han; Kim, Bong Soo; Kim, Jong Man; Choi, Jong Hyun

    2002-12-01

    The sodium is widely used as one of the coolants in the liquid metal reactor system since it has important safety features such as a long thermal response time, a large margin to coolant boiling, and operating in near atmospheric pressure, etc. The state-of-the-art on the flow and pressure measurement techniques in the piping system worldwide is investigated and reviewed to utilize it as a basis for developing a new technique applying for the high temperature sodium flow environment.

  19. Modification of the Core Cooling System of TRIGA 2000 Reactor

    Science.gov (United States)

    Umar, Efrizon; Fiantini, Rosalina

    2010-06-01

    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24°C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

  20. SUBSTATIONS OF DISTRICT HEATING SYSTEMS WITH PULSE COOLANT CIRCULATION

    Directory of Open Access Journals (Sweden)

    Andrey N. Makeev

    2017-01-01

    Full Text Available Abstract. Objectives The aim of the study is to generalise the results of the application of technologies and means for organising pulse coolant flow within a district heating system in order to increase its energy efficiency based on the organisation of local hydraulic shocks and the subsequent use of their energy to ensure the purification of heat energy equipment, intensify the heat transfer process and realise the possibility of transforming the available head from one hydraulic circuit to another. Methods Substations connecting the thermal power installations of consumers with heat networks via dependent and independent schemes are analytically generalised. The use of pulse coolant circulation is proposed as a means of overcoming identified shortcomings. Results Principal schemes of substations with pulse coolant circulation for dependent and independent connection of thermal power installations are detailed. A detailed description of their operation is given. The advantages of using pulse coolant circulation in substations are shown. The materials reflecting the results of the technical implementation and practical introduction of this technology are presented. Conclusion Theoretical analysis of the operation of the basic schemes of substations with pulse coolant circulation and the results of their practical application, as well as the materials of scientific works devoted to the use of the energy of a hydraulic impact and the study of the effect of pulse coolant flow on thermal and hydrodynamic processes, have yielded a combination of factors reflecting technical and economic rationality of application of pulse coolant circulation. 

  1. Investigations of the VVER-1000 coolant transient benchmark phase 1 with the coupled code system RELAP5/PARCS

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Espinoza, Victor Hugo

    2008-07-15

    As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during the test and its effects on the

  2. Anaerobic horizontal flow reactor with polyethylene terephthalate as support material

    Directory of Open Access Journals (Sweden)

    Marcelo Muñoz

    2016-06-01

    Full Text Available A pilot anaerobic reactor was installed to remove the organic load of wastewater from dairy industry. It uses a bacterial inoculum previously acclimated to the substrate. It was disposed horizontally and filled with pieces of polyethylene terephthalate (PET, from plastic bottles. The reactor was operated at room temperature, during 100 days, in three phases: 1 the reactor was stabilized with volumetric organic load from 0.013 to 0.500 kg/day.m³; 2 the hydraulic retention time was of 1 day and the volumetric organic load of 3 kg/day.m³; 3 the volumetric organic load was incremented from 4 to 6.6 kg/day.m³ and the hydraulic retention time was 1 day. Organic material removal efficiencies was of 85%, and approximately 75% were obtained in the second and third phase, respectively. The Y value was 0.15, indicating that 0.15 kg of biomass were generated by kg of QDO supplied to the reactor. Finally, the biomass generated inside the reactor was analyzed, obtaining a value of 18868 mg/L, which is a higher value than those of conventional systems.

  3. Calorimetric Analysis to Infer Primary Circuit Flow in Integral and Pool-Type Reactors

    Science.gov (United States)

    Coble, Jamie; Tarver, Ryan; Hines, J. Wesley

    2017-02-01

    Primary system flow rate is a key parameter for monitoring and controlling thermal power in a nuclear power plant. The existing fleet of large light water reactors uses direct measurements of primary flow rate with the application of venturi meters, orifice plates, and magnetic flowmeters in primary loop piping. Integral light water reactors and pool-type advanced reactor designs, however, have largely eliminated primary loop piping to improve the inherent safety characteristics of these reactors. Furthermore, longer operating cycles between maintenance opportunities (typically 4 to 40 years) limit the applicability of these direct measurement methods over the operating period. Methods to infer the primary flow rate based on other, easily measured parameters are needed to ensure the operability of integral and pool-type reactors. Calorimetric analysis across the intermediate heat exchanger was investigated for real-time inference of primary flow rate. Heat balance equations were applied to an experimental forced flow loop to evaluate the efficacy of this approach. When appropriate time delays and heat losses are accounted for, the primary flow rate was inferred with accuracy and 95% prediction variance of 1.57 and 4.80 % mean value, respectively.

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

    Science.gov (United States)

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

    2015-01-01

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

  5. Computation of flow and thermal fields in a model CVD reactor

    Indian Academy of Sciences (India)

    R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

    Computation of flow and thermal fields in a model CVD reactor. VISHWADEEP SAXENA, K MURALIDHAR and V ... been numerically studied. This configuration is encountered during the modelling of flow and heat transfer in CVD .... carried out on a P-III, 600 MHz machine with 512 MB RAM. Typical CPU times required.

  6. Fluid flow structure around the mixer in a reactor with mechanical mixing

    Energy Technology Data Exchange (ETDEWEB)

    Lecheva, A., E-mail: alecheva@uni-ruse.bg [Department of Mathematics, Angel Kanchev University of Rousse, 8 Studentska str., 7017 Rousse (Bulgaria); Zheleva, I., E-mail: izheleva@uni-ruse.bg [Department of Heat Technology, Hydraulics and Ecology, Angel Kanchev University of Rousse, 8 Studentska str., 7017 Rousse (Bulgaria)

    2015-10-28

    Fluid flow structure around the mixer in a cylindrical reactor with mechanical mixing is studied and numerical results are presented in this article. The model area is complex because of the presence of convex corners of the mixer in the fluid flow. Proper boundary conditions for the vorticity calculated on the base of the stream function values near solid boundaries of the examined area are presented. The boundary value problem of motion of swirling incompressible viscous fluid in a vertical tank reactor with a mixer is solved numerically. The calculations are made by a computer code, written in MATLAB. The complex structure of the flow around the mixing disk is described and commented.

  7. Calculation of the VVER-1000 coolant transient benchmark using the coupled code systems DYN3D/RELAP5 and DYN3D/ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Kozmenkov, Y. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Kliem, S. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany)]. E-mail: S.Kliem@fzd.de; Grundmann, U. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Rohde, U. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Weiss, F.-P. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany)

    2007-09-15

    Plant-measured data provided by the OECD/NEA VVER-1000 coolant transient benchmark programme were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to an experiment that was conducted during the commissioning of the Kozloduy NPP Unit 6 in Bulgaria. In this experiment, the fourth main coolant pump was switched on whilst the remaining three were running normal operating conditions. The experiment was conducted at 27.5% of the nominal level of the reactor power. The transient is characterized by a rapid increase in the primary coolant flow through the core, and as a consequence, a decrease of the space-dependent core inlet temperature. The control rods were kept in their original positions during the entire transient. The coupled simulations performed on both DYN3D/RELAP5 and DYN3D/ATHLET were based on the same reactor model, including identical main coolant pump characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has also been performed to evaluate the respective thermal hydraulic models of the system codes RELAP5 and ATHLET.

  8. Numerical analysis of thermal stratification phenomena in upper plenum of a fast breeder reactor (1). Evaluation of thermal stratification phenomena near the region of flow holes

    Energy Technology Data Exchange (ETDEWEB)

    Suda, Kazunori; Muramatu, Toshiharu; Yamaguchi, Akira [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2002-12-01

    Thermal stratification phenomena are observed in an upper plenum of liquid metal fast breeder reactors (LMFBRs) under reactor scram conditions, which give rise to thermal stress on structural components. Therefore it is important to evaluate characteristics of the phenomena in the design on the internal structures in an LMFBR plenum. To evaluate long-term characteristics of thermal stratification phenomena in a typical LMFBR upper plenum, numerical analysis was carried out with a multi-dimensional thermohydraulics code AQUA for a scram event from full power operation condition. Thereafter the numerical results were compared with extrapolated results of measured transient data on the 40% operation condition. From the thermohydraulic analysis by the AQUA code, the following results have been obtained. (Long-term characteristics of thermal stratification phenomena) The cold fluid region near the inside inner barrel was expanded with accumulation of the cold fluid in the lower region of the plenum after 300 seconds from the reactor scram, so that the fluid from core flowed to the lower region of the upper plenum. The characteristics of axial temperature distributions in the upper plenum were similar to them at the 300 seconds. The thermal stratification interface was located initially around intermediate position between upper lower flow holes. And an another thermal stratification interface was formed around the inner barrel support plate after 300 seconds from the scram, so that the cold fluid accumulated in the lower region of the plenum. But the thermal stratification interface around the inner barrel support plate was disappeared by mixture and heat conduction of coolant of circumferential direction. The thermal stratification interface which was located below in the upper flow holes, rose to the upward position of the upper flow holes at the 720 seconds. In annular gap region between the inner barrel and the reactor vessel wall, thermal stratification interface

  9. Aerosol flow in a tube furnace reactor of gas-phase synthesised silver nanoparticles

    Science.gov (United States)

    Mitrakos, D.; Jokiniemi, J.; Backman, U.; Housiadas, C.

    2008-12-01

    In a previous work, gas-phase synthesis of silver nanoparticles through evaporation of silver powder and subsequent particle nucleation by cooling was shown to be a viable method for achieving high purity silver nanoparticles (Backman et al. J Nanopart Res 4:325-335, 2002). In order to control the size of the produced nanoparticles, careful design of the reactor is required with respect to thermal and flow characteristics. In the present work, the silver nanoparticle reactor is rigorously simulated by means of multidimensional computational fluid and particle dynamics. The CFD-computed flow is input for a combined simulation of the vapour field and particle homogeneous nucleation, growth and coagulation. The results are compared with the experimental data and with the predictions from the usually employed simple model of an idealized plug flow reactor. The multidimensional CFD-based analysis is shown to explain and help understand different aspects of the reactor operation and size distribution of the particles produced. Yet the simple plug flow method is found to provide reasonable accuracy when an appropriate correction factor is used for the nucleation rate. Considering its robustness and computational simplicity, the plug flow method can be qualified as adequate from the engineering practical point of view for the case of silver nanoparticle reactors.

  10. Experimental Investigation of Coolant Boiling in a Half-Heated Circular Tube - Final CRADA Report

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Wenhua [Argonne National Lab. (ANL), Argonne, IL (United States); Singh, Dileep [Argonne National Lab. (ANL), Argonne, IL (United States); France, David M. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-11-01

    Coolant subcooled boiling in the cylinder head regions of heavy-duty vehicle engines is unavoidable at high thermal loads due to high metal temperatures. However, theoretical, numerical, and experimental studies of coolant subcooled flow boiling under these specific application conditions are generally lacking in the engineering literature. The objective of this project was to provide such much-needed information, including the coolant subcooled flow boiling characteristics and the corresponding heat transfer coefficients, through experimental investigations.

  11. Rational Basis for Designing Horizontal-Flow Anaerobic Immobilized Sludge (HAIS Reactor for Wastewater Treatment

    Directory of Open Access Journals (Sweden)

    Zaiat M.

    1997-01-01

    Full Text Available The conception and development on a rational basis of a new configuration of anaerobic fixed-bed bioreactor for wastewater treatment, the horizontal-flow anaerobic immobilized sludge (HAIS reactor, is presented. Such a reactor containing immobilized sludge in polyurethane foam matrices was first assayed for treating paper industry wastewater. A very short start-up period was observed and the reactor achieved stable operation by the eighth day. Afterwards, fundamental aspects of the process were investigated in order to obtain a rational basis for HAIS reactor design. A sequence of experiments was carried out for evaluating the cell wash-out from polyurethane foam matrices, the liquid-phase mass transfer coefficient and the intrinsic kinetic parameters, besides the hydrodynamic flow pattern of the reactor. The knowledge of such fundamental phenomena is useful for improving the reactor?s design and operation. Besides, these fundamental studies are essential to provide parameters for simulation and optimization of processes that make use of immobilized biomass

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

    DEFF Research Database (Denmark)

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

    2016-01-01

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

  13. 7Q10 flows for SRS streams

    Energy Technology Data Exchange (ETDEWEB)

    Chen, K.F.

    1996-10-01

    The Environmental Transport Group of the Environmental Technology Section was requested to predict the seven-day ten-year low flow (7Q10 flow) for the SRS streams based on historical stream flow records. Most of the historical flow records for the SRS streams include reactor coolant water discharged from the reactors and process water released from the process facilities. The most straight forward way to estimate the stream daily natural flow is to subtract the measured upstream reactor and/or facility daily effluents from the measured downstream daily flow. Unfortunately, this method does not always work, as indicated by the fact that sometimes the measured downstream volumetric flow rates are lower than the reactor effluent volumetric flow rates. For those cases that cannot be analyzed with the simple subtracting method, an alternative method was used to estimate the stream natural flows by statistically separating reactor coolant and process water flow data. The correlation between the calculated 7Q10 flows and the watershed areas for Four Mile Branch and Pen Branch agrees with that calculated by the USGS for Upper Three Runs and Lower Three Runs Creeks. The agreement between these two independent calculations lends confidence to the 7Q10 flow calculations presented in this report.

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

    Science.gov (United States)

    Khane, Vaibhav B.

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

  15. Characterizing Flow-Induced Vibrations of Fuel Assemblies for Future Liquid Metal Cooled Nuclear Reactors Using Quasi-Distributed Fibre-Optic Sensors

    Directory of Open Access Journals (Sweden)

    Ben De Pauw

    2017-08-01

    Full Text Available Excessive vibration of nuclear reactor components, such as the heat exchanger or the fuel assembly should be avoided as these can compromise the lifetime of these components and potentially lead to safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants. However, identifying adequate sensors or techniques that can be successfully applied to record the vibrations of the components in a flow of liquid metal at elevated temperatures is very challenging. In this paper, we demonstrate the precise measurements of the vibrations of a very representative mock-up of a fuel assembly in a lead-bismuth eutectic cooled installation using quasi-distributed fibre Bragg grating (FBG based sensors. The unique properties of these sensors, in combination with a dedicated integration and mounting approach, allows for accounting of the severe geometrical constraints and allows characterizing the vibration of the fuel assembly elements under nominal operation conditions. To that aim, we instrumented a single fuel pin within the fuel assembly with 84 FBGs, and conducted spectral measurements with an acquisition rate of up to 5000 measurements per second, enabling the monitoring of local strains of a few με. These measurements provide the information required to assess vibration-related safety hazards.

  16. Analysis of Loss-of-Coolant Accidents in the NBSR

    Energy Technology Data Exchange (ETDEWEB)

    Baek J. S.; Cheng L.; Diamond, D.

    2014-05-23

    This report documents calculations of the fuel cladding temperature during loss-of-coolant accidents in the NBSR. The probability of a pipe failure is small and procedures exist to minimize the loss of water and assure emergency cooling water flows into the reactor core during such an event. Analysis in the past has shown that the emergency cooling water would provide adequate cooling if the water filled the flow channels within the fuel elements. The present analysis is to determine if there is adequate cooling if the water drains from the flow channels. Based on photographs of how the emergency water flows into the fuel elements from the distribution pan, it can be assumed that this water does not distribute uniformly across the flow channels but rather results in a liquid film flowing downward on the inside of one of the side plates in each fuel element and only wets the edges of the fuel plates. An analysis of guillotine breaks shows the cladding temperature remains below the blister temperature in fuel plates in the upper section of the fuel element. In the lower section, the fuel plates are also cooled by water outside the element that is present due to the hold-up pan and temperatures are lower than in the upper section. For small breaks, the simulation results show that the fuel elements are always cooled on the outside even in the upper section and the cladding temperature cannot be higher than the blister temperature. The above results are predicated on assumptions that are examined in the study to see their influence on fuel temperature.

  17. Dry fermentation of manure with straw in continuous plug flow reactor: Reactor development and process stability at different loading rates.

    Science.gov (United States)

    Patinvoh, Regina J; Kalantar Mehrjerdi, Adib; Sárvári Horváth, Ilona; Taherzadeh, Mohammad J

    2017-01-01

    In this work, a plug flow reactor was developed for continuous dry digestion processes and its efficiency was investigated using untreated manure bedded with straw at 22% total solids content. This newly developed reactor worked successfully for 230days at increasing organic loading rates of 2.8, 4.2 and 6gVS/L/d and retention times of 60, 40 and 28days, respectively. Organic loading rates up to 4.2gVS/L/d gave a better process stability, with methane yields up to 0.163LCH4/gVSadded/d which is 56% of the theoretical yield. Further increase of organic loading rate to 6gVS/L/d caused process instability with lower volatile solid removal efficiency and cellulose degradation. Copyright © 2016 Elsevier Ltd. All rights reserved.

  18. Vibroacoustic diagnostic monitoring of selected primary coolant circuit components of the Bohunice V-1 nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Bahna, J.; Jaros, I. (Forschungsinstitut fuer Kernkraftwerke, Jaslovske Bohunice (Czechoslovakia))

    1983-02-01

    An information is given about vibroacoustic monitoring systems installed at the two units of the Bohunice V-1 nuclear power plant for oscillation monitoring of the reactor coolant pumps and reactor pressure vessel. Signal processing and analysis techniques have been developed for diagnostic measurements. It is reported on some experience gained from oscillation measurements at the reactor coolant pumps of the V-1 nuclear power plant since commissioning of the plant.

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

    NARCIS (Netherlands)

    Westerterp, K.R.; Kuczynski, M.

    1987-01-01

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

  20. HORIZONTAL BOILING REACTOR SYSTEM

    Science.gov (United States)

    Treshow, M.

    1958-11-18

    Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

  1. Fuel assembly design study for a reactor with supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Hofmeister, J. [RWE Power AG, Huyssenallee 2, D-45128 Essen (Germany); Waata, C. [ANSYS Germany GmbH, Staudenfeldweg 12, D-83624 Otterfing (Germany); Starflinger, J. [Forschungszentrum Karlsruhe GmbH, Institute for Nuclear and Energy Technologies, P.O. Box 3640, D-76021 Karlsruhe (Germany); Schulenberg, T. [Forschungszentrum Karlsruhe GmbH, Institute for Nuclear and Energy Technologies, P.O. Box 3640, D-76021 Karlsruhe (Germany)]. E-mail: thomas.schulenberg@iket.fzk.de; Laurien, E. [University of Stuttgart, Institute for Nuclear Technology and Energy Systems (IKE), Pfaffenwaldring 31, D-70569 Stuttgart (Germany)

    2007-08-15

    The European concept of the High Performance Light Water Reactor (HPLWR) differs from current light water reactors in a higher system pressure beyond the critical point of water, as well as a higher heat-up of the coolant within the core and thus higher core outlet temperatures, leading to a significant increase in turbine power and thermal efficiency of the power plant. The motivation to develop a novel fuel assembly for the HPLWR is caused by the high variation of coolant density in the core by more than a factor of seven. A systematic design study shows that a square fuel assembly with two rows of fuel rods and a central moderator box is best to minimize the structural material, to optimize the moderator to fuel ratio and to reduce differences of fuel rod power. Using neutronic and thermal-hydraulic analyses, a detailed mechanical design of a fuel assembly of the HPLWR has been worked out. Moreover, concepts for the head piece, the foot piece, the steam plenum and the lower mixing plenum, including the lower core plate, have been developed to account for the individual flow paths of this reactor. These allow a leak-tight counter current flow of moderator water and coolant as well as uniform mixing of different mass flows. The assembly design concept can be used as a general key component for any advanced core design of this reactor.

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

    Energy Technology Data Exchange (ETDEWEB)

    Primm III, R.T.C

    1981-02-01

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

  3. Granular flow in pebble-bed nuclear reactors: Scaling, dust generation, and stress

    Energy Technology Data Exchange (ETDEWEB)

    Rycroft, Chris H., E-mail: chr@seas.harvard.edu [Department of Physics, University of California, Berkeley, CA 94720 (United States); Department of Mathematics, University of California, Berkeley, CA 94720 (United States); Department of Mathematics, Lawrence Berkeley Laboratory, Berkeley, CA 94720 (United States); School of Engineering and Applied Sciences, Harvard University, Cambridge, MA 02138 (United States); Dehbi, Abdel, E-mail: abdel.dehbi@psi.ch [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland); Lind, Terttaliisa, E-mail: terttaliisa.lind@psi.ch [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland); Güntay, Salih, E-mail: salih.guentay@psi.ch [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland)

    2013-12-15

    Highlights: • The scaling properties of granular flow in pebble-bed reactors have been analyzed. • Simulations in a full-size reactor are compared to those scaled down by 3:1 and 6:1. • For some features, the scaled-down behavior is complex due to pebble packing effects. • Pebble stresses and dust generation due to pebble wear are analyzed in detail. • Approximate scaling laws for pebble stress and wear are developed. - Abstract: In experimental prototypes of pebble-bed reactors, significant quantities of graphite dust have been observed due to rubbing between pebbles as they flow through the core. At the typical operating conditions in these reactors, which feature high temperatures, pressures, and a helium atmosphere, limited data is available on the frictional properties of the pebble surfaces, and as a result, a conceptual design of a scaled-down version of a pebble-bed reactor has been proposed to investigate this issue in detail. However, this raises general questions about how the granular flow in a scaled facility will emulate that in a full-size reactor. To address this, simulations of granular flow in pebble-bed reactors using the discrete-element method (DEM) have been carried out in a full-size geometry (using 440,000 pebbles) and compared to those in geometries scaled down by factors of 3:1 and 6:1. Differences in velocity profiles, pebble ordering, pebble wear, and stresses are examined, and the effect of friction is discussed. The results show complex behavior due to discrete pebble packing effects, although several simple scaling rules can be derived.

  4. Analysis of Fluid Flow and Heat Transfer Model for the Pebble Bed High Temperature Gas Cooled Reactor

    OpenAIRE

    S. Yamoah; E.H.K. Akaho; Nana G.A. Ayensu; M. Asamoah

    2012-01-01

    The pebble bed type high temperature gas cooled nuclear reactor is a promising option for next generation reactor technology and has the potential to provide high efficiency and cost effective electricity generation. The reactor unit heat transfer poses a challenge due to the complexity associated with the thermalflow design. Therefore to reliably simulate the flow and heat transport of the pebble bed modular reactor necessitates a heat transfer model that deals with radiation as well as ther...

  5. Entrained Flow Reactor Study of KCl Capture by Solid Additives

    DEFF Research Database (Denmark)

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

    been proved to be very promising additives and havereceived extensive studies during the past decades. However, mostprevious studies were carried out in fixed-bed reactors where the reaction conditions are obviously different from that in suspension fired boilers.Detailed knowledge on the reaction...... between K-species and solid additivesunder suspension-fired conditions is still limited. In this study, a water slurrycontaining K-salt and solid additives was introduced into an entrained flowreactor (EFR) to study K-capture at suspension-fired conditions. A model willbe developed based on experimental...

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

    Directory of Open Access Journals (Sweden)

    Nikačević N.M.

    2015-01-01

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

  7. Ultrasound assisted synthesis of performic acid in a continuous flow microstructured reactor.

    Science.gov (United States)

    Jolhe, P D; Bhanvase, B A; Patil, V S; Sonawane, S H; Potoroko, I

    2017-11-01

    The present work establishes in depth study of ultrasound assisted preparation of performic acid (PFA) in a continuous flow microstructured reactor. The influence of various parameters viz. formic acid: hydrogen peroxide molar ratio, flow rate, temperature and catalyst loading on the PFA formation were studied in a continuous flow microstructured reactor. In a continuous microstructured reactor in the presence of ultrasonic irradiation, the formation of PFA was found to be dependent on the molar ratio of formic acid: hydrogen peroxide, flow rate of reactants, temperature and catalyst loading (Amberlite IR-120H). The optimized parameter values are 1:1M ratio, 50mL/h, 40°C and 471mg/cm(3) respectively. Further, the performance of Amberlite IR-120H catalyst was evaluated for three successive cycles in continuous microstructured reactor. The performance of catalyst was found to be decreased with the usage of the catalyst and is attributed to neutralization of the sulfonic acid groups, catalyst shrinkage, or loss in pore sites. The experimental results revealed that, for an ultrasound assisted synthesis of PFA in continuous microstructured reactor the observed reaction time was even less than 10min. The observed intensification in the PFA synthesis process can be attributed to the intense collapse of the cavities formed at low temperature during ultrasonic irradiations, which further improved the heat and mass transfer rates with the formation of H2O2 during the reaction. The combined use of ultrasound and a continuous flow microstructured reactor has proved beneficial process of performic acid synthesis. Copyright © 2017 Elsevier B.V. All rights reserved.

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

    Energy Technology Data Exchange (ETDEWEB)

    Flach, G.P.

    2000-02-11

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

  9. Method for automatically scramming a nuclear reactor

    Science.gov (United States)

    Ougouag, Abderrafi M.; Schultz, Richard R.; Terry, William K.

    2005-12-27

    An automatically scramming nuclear reactor system. One embodiment comprises a core having a coolant inlet end and a coolant outlet end. A cooling system operatively associated with the core provides coolant to the coolant inlet end and removes heated coolant from the coolant outlet end, thus maintaining a pressure differential therebetween during a normal operating condition of the nuclear reactor system. A guide tube is positioned within the core with a first end of the guide tube in fluid communication with the coolant inlet end of the core, and a second end of the guide tube in fluid communication with the coolant outlet end of the core. A control element is positioned within the guide tube and is movable therein between upper and lower positions, and automatically falls under the action of gravity to the lower position when the pressure differential drops below a safe pressure differential.

  10. 76 FR 16842 - Request for a License To Export Reactor Components

    Science.gov (United States)

    2011-03-25

    ... six AP- Construction, China. Mechanical Corporation. coolant pump 1000 (design) maintenance, and systems, related reactors. operation of AP- equipment, and 1000 (design) spare parts. nuclear reactors...

  11. ANALISIS TRANSIEN PADA FIXED BED NUCLEAR REACTOR

    Directory of Open Access Journals (Sweden)

    M. Rizaal

    2015-03-01

    Full Text Available Desain teras Fixed Bed Nuclear Reactor (FBNR yang modular memungkinkan pengendalian daya dapat dilakukan dengan mengatur ketinggian suspended core dan laju aliran massa pendingin. Tujuan penelitian ini adalah mempelajari perubahan daya termal teras sebagai akibat perubahan laju aliran massa pendingin yang masuk ke teras reaktor dan perubahan ketinggian suspended core serta mempelajari karakteristik keselamatan melekat yang dimiliki FBNR saat terjadi kegagalan pelepasan kalor (loss of heat sink. Keadaan neutronik teras dimodelkan pada kondisi tunak dengan menggunakan paket program Standard Reactor Analysis Code (SRAC untuk memperoleh data fluks neutron, konstanta grup, fraksi neutron kasip, konstanta peluruhan prekursor neutron kasip, dan beberapa parameter teras penting lainnya. Selanjutnya data tersebut digunakan pada perhitungan transien sebagai syarat awal. Analisis transien dilakukan pada tiga kondisi, yaitu saat terjadi penurunan laju aliran massa pendingin, saat terjadi penurunan ketinggian suspended core, dan saat terjadi kegagalan sistem pelepasan kalor. Hasil yang diperoleh dari penelitian ini menunjukkan bahwa penurunan laju aliran massa pendingin sebesar 50%, dari kondisi normal, menyebabkan daya termal teras turun 28% dibanding daya sebelumnya. Penurunan ketinggian suspended core sebesar 30% dari ketinggian normal menyebabkan daya termal teras turun 17% dibanding daya sebelumnya. Sementara untuk kondisi kegagalan sistem pelepasan kalor, daya termal teras mengalami penurunan sebesar 76%. Dengan demikian, pengendalian daya pada FBNR dapat dilakukan dengan mengatur laju aliran massa pendingin dan ketinggian suspended core, serta keselamatan melekat yang handal pada kondisi kegagalan sistem pelepasan kalor. Kata kunci: FBNR, transien, daya, laju aliran massa, suspended core Modular in design enables Fixed Bed Nuclear Reactor (FBNR power controlled by the adjustment of suspended core and coolant flow rate. The main purposes of this paper

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

    Science.gov (United States)

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

    2014-01-01

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

  13. Rotor dynamic analysis of main coolant pump

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chong Won; Seo, Jeong Hwan; Kim, Choong Hwan; Shin, Jae Chul; Wang, Lei Tian [Korea Advanced Institute of Science and Technology, Taejon (Korea)

    1999-03-01

    A rotor dynamic analysis program DARBS/MCP, for the main coolant pump of the integral reactor, has been developed. The dynamic analysis model of the main coolant pump includes a vertical shaft, three grooved radial journal bearings and gaps that represent the structure-fluid interaction effects between the rotor and the lubricant fluid. The electromagnetic force from the motor and the hydro-dynamic force induced by impeller are the major sources of vibration that may affect the rotor system stability. DARBS/MCP is a software that is developed to effectively analyze the dynamics of MCP rotor systems effectively by applying powerful numerical algorithms such as FEM with modal truncation and {lambda}-matrix method for harmonic analysis. Main design control parameters, that have much influence to the dynamic stability, have been found by Taguchi's sensitivity analysis method. Design suggestions to improve the stability of MCP rotor system have been documented. The dynamic bearing parameters of the journal bearings used for main coolant pump have been determined by directly solving the Reynolds equation using FDM method. Fluid-structure interaction effect that occurs at the small gaps between the rotor and the stator were modeled as equivalent seals, the electromagnetic force effect was regarded as a linear negative radial spring and the impeller was modeled as a rigid disk with hydrodynamic and static radial force. Although there exist critical speeds in the range of operational speeds for type I and II rotor systems, the amplitude of vibration appears to be less than the vibration limit set by the API standards. Further more, it has been verified that the main design parameters such as the clearance and length of journal bearings, and the static radial force of impeller should be properly adjusted, in order to the improve dynamic stability of the rotor system. (author). 39 refs., 81 figs., 17 tabs.

  14. Prediction of acid hydrolysis of lignocellulosic materials in batch and plug flow reactors.

    Science.gov (United States)

    Jaramillo, Oscar Johnny; Gómez-García, Miguel Ángel; Fontalvo, Javier

    2013-08-01

    This study unifies contradictory conclusions reported in literature on acid hydrolysis of lignocellulosic materials, using batch and plug flow reactors, regarding the influence of the initial liquid ratio of acid aqueous solution to solid lignocellulosic material on sugar yield and concentration. The proposed model takes into account the volume change of the reaction media during the hydrolysis process. An error lower than 8% was found between predictions, using a single set of kinetic parameters for several liquid to solid ratios, and reported experimental data for batch and plug flow reactors. For low liquid-solid ratios, the poor wetting and the acid neutralization, due to the ash presented in the solid, will both reduce the sugar yield. Also, this study shows that both reactors are basically equivalent in terms of the influence of the liquid to solid ratio on xylose and glucose yield. Copyright © 2013 Elsevier Ltd. All rights reserved.

  15. Efficient Synthetic Photocyclization for Phenacenes Using a Continuous Flow Reactor

    National Research Council Canada - National Science Library

    Okamoto, Hideki; Takane, Takamitsu; Gohda, Shin; Kubozono, Yoshihiro; Sato, Kaori; Yamaji, Minoru; Satake, Kyosuke

    2014-01-01

    The continuous flow reaction technique has been applied to the photocyclization of 1,2-diarylethenes, the so-called Mallory reaction, to afford phenacenes in high chemical yields and efficiencies (114–288 mg h−1...

  16. Liquid-liquid Slug Flow in a Microchannel Reactor and its Mass Transfer Properties - A Review

    Directory of Open Access Journals (Sweden)

    Rahul Antony

    2014-10-01

    Full Text Available Mass transfer is a basic phenomenon behind many processes like reaction, absorption, extraction etc. Mass transfer plays a significant role in microfluidic systems where the chemical / biological process systems are shrinkened down to a micro scale. Micro reactor system, with its high compatibility and performance, gains a wide interest among the researchers in the recent years. Micro structured reac-tors holds advantages over the conventional types in chemical processes. The significance of micro re-actor not limited to its scalability but to energy efficiency, on-site / on-demand production, reliability, safety, highly controlled outputs, etc. Liquid-liquid two phase reaction in a microreactor system is highly demandable when both reactants are liquids or when air medium cannot be suitable. This arti-cle overviews various liquid-liquid flow regimes in a microchannel. Discussions on the hydrodynamics of flow in micro scale are made. Considering the importance of mass transfer in liquid-liquid systems and the advantage of slug regime over other regimes, the article focuses especially on the mass trans-fer between two liquid phases in slug flow and the details of experimental studies carried out in this area. The advantages of slug flow over other flow regimes in micro structured reactor applications are showcased. © 2014 BCREC UNDIP. All rights reservedReceived: 31st May 2014; Revised: 6th August 2014; Accepted: 14th August 2014How to Cite: Antony, R., Giri Nandagopal, M.S., Sreekumar, N., Rangabhashiyam, S., Selvaraju, N. (2014. Liquid-liquid Slug Flow in a Microchannel Reactor and its Mass Transfer Properties - A Review. Bulletin of Chemical Reaction Engineering & Catalysis,9(3: 207-223. (doi:10.9767/bcrec.9.3.6977.207-223Permalink/DOI: http://dx.doi.org/10.9767/bcrec.9.3.6977.207-223

  17. Modelling of turbulent flow in a radial reactor with fixed bed

    Science.gov (United States)

    Zhapbasbayev, U. K.; Ramazanova, G. I.; Kenzhaliev, O. B.

    2015-03-01

    The data of the computation of turbulent flow in the CF- π and CP- π configurations of the radial reactor with a fixed bed are presented. The Reynolds motion equations have been solved jointly with the k- ɛ turbulence model. To couple the parameters of flows at the interface free part-fixed bed the classical continuity equations were used. The computational data are obtained for the averaged and turbulent characteristics, and it is shown that the flow in the fixed bed causes the generation of the turbulence kinetic energy and its dissipation rate; the flow in the CF- π configuration is distributed more uniformly as compared to the CP- π configuration of the radial reactor. Computed data are compared with the experimental ones.

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

    DEFF Research Database (Denmark)

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

    2017-01-01

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

  19. Numerical simulations of flow field in the target region of accelerator-driven subcritical reactor system

    CERN Document Server

    Chen Hai Yan

    2002-01-01

    Numerical simulations of flow field were performed by using the PHOENICS 3.2 code for the proposed spallation target of accelerator-driven subcritical reactor system (ADS). The fluid motion in the target is axisymmetric and is treated as a 2-D steady-state problem. A body-fitted coordinate system (BFC) is then chosen and a two-dimensional mesh of the flow channel is generated. Results are presented for the ADS target under both upward and downward flow, and for the target with diffuser plate installed below the window under downward flow

  20. Gas-liquid two-phase flow through packed bed reactors in microgravity

    Science.gov (United States)

    Motil, Brian Joseph

    Experimental results on flow pattern transitions, pressure drop and flow characteristics for cocurrent gas-liquid flow through packed bed reactors in microgravity is presented and analyzed. The pulse flow regime is shown to exist over a much wider range of gas and liquid flow rates when under microgravity conditions. A new model is developed to predict the transition from bubble flow to pulse flow based on the dimensionless Suratman number. The Suratman number is shown to represent the balance of forces at the pore level which determine the conditions necessary for the onset of pulse flow in the column. This model is then extended to normal gravity flows in the downward direction for fixed Bond numbers. A model to predict pressure drop in the absence of gravity is also presented. An additional pressure drop term is developed to extend the applicability of the Ergun equation to gas-liquid flow. This term represents the losses resulting from the dynamic interaction between the two phases and is superposed with the liquid viscous and inertia terms to represent the total pressure loss through a reactor bed in a microgravity environment. The modified two-phase Ergun equation is shown to provide good agreement with the experimental results.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  2. Analysis of first stage ignition delay times of dimethyl ether in a laminar flow reactor

    Science.gov (United States)

    Wada, Tomoya; Sudholt, Alena; Pitsch, Heinz; Peters, Norbert

    2013-10-01

    The combustion chemistry of the first stage ignition and chemistry/flow interactions are studied for dimethyl ether (DME) with a mathematical analysis of two systems: a plug flow reactor study is used to reduce the reaction chemistry systematically. A skeletal reaction mechanism for the low temperature chemistry of DME until the onset of ignition is derived on the basis of the detailed DME mechanism of the Lawrence Livermore National Laboratory - see Curran, Fischer and Dryer, Int. J. Chem. Kinetics, Vol. 32 (2000). It is shown that reasonably good results for ignition delay times can be reached using a simple system of three ordinary differential equations and that the resulting analytical solution depends only on two reaction rates and the initial fuel concentration. The stepwise reduction of the system based on assumptions yields an understanding on why these reactions are so important. Furthermore, the validation of the assumptions yields insight into the influence of the fuel and the oxygen concentration on the temperature during the induction phase. To investigate the influence of chemistry/flow interactions, a 2D model with a laminar Hagen-Poiseuille flow and 2D-polynomial profiles for the radial species concentration is considered. For the 2D model, it is found that only the diffusion coefficients and the reactor radius need to be taken into consideration additionally to describe the system sufficiently. Also, the coupling of flow and chemistry is clarified in the mathematical analysis. The insight obtained from the comparison of the 2D model and the plug flow model is used to establish an average velocity for the conversion of ignition locations to ignition delay times in a laminar flow reactor. Finally, the 2D analytical solution is compared against new experimental data, obtained in such a laminar flow reactor for an undiluted DME/air mixture with an equivalence ratio of φ = 0.835 and a temperature range of 555 to 585 K at atmospheric pressure.

  3. Engineering chemistry: integrating batch and flow reactions on a single, automated reactor platform

    OpenAIRE

    Fitzpatrick,, F.; Ley, Steven Victor

    2016-01-01

    Synthesis chemistry need not be limited to either only batch or only flow; rather, in the future we expect that it will consist of an amalgamation of the best and most appropriate methods. We have therefore devised a single reactor platform to conduct both batch and flow reactions, either singly or in concert, using open source technologies to automate, control and monitor individual processes. We illustrate this concept with the multistep synthesis of 5-methyl-4-propylthiophene-2-carboxylic ...

  4. PEBBLES: A COMPUTER CODE FOR MODELING PACKING, FLOW AND RECIRCULATIONOF PEBBLES IN A PEBBLE BED REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2006-10-01

    A comprehensive, high fidelity model for pebble flow has been developed and embodied in the PEBBLES computer code. In this paper, a description of the physical artifacts included in the model is presented and some results from using the computer code for predicting the features of pebble flow and packing in a realistic pebble bed reactor design are shown. The sensitivity of models to various physical parameters is also discussed.

  5. Ablative fast pyrolysis of biomass in the entrained-flow cyclonic reactor at SERI

    Energy Technology Data Exchange (ETDEWEB)

    Diebold, J.; Scahill, J.

    1982-06-01

    Progress with the entrained flow cyclonic reactor at SERI is detailed. Feedstocks successfully used include wood flour and fairly large sawdust. Preliminary results show that relatively complete vaporization of the biomass is realized and that the yields of tar or gas can be varied over quite a range with trends following first order kinetic concepts.

  6. A high-pressure plug flow reactor for combustion chemistry investigations

    Science.gov (United States)

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

    2017-10-01

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

  7. Automated Determination of Oxygen-Dependent Enzyme Kinetics in a Tube-in-Tube Flow Reactor

    DEFF Research Database (Denmark)

    Ringborg, Rolf Hoffmeyer; Pedersen, Asbjørn Toftgaard; Woodley, John

    2017-01-01

    and limited oxygen supply. Here, we present a novel method for the collection of such kinetic data using a pressurized tube-in-tube reactor, operated in the low-dispersed flow regime to generate time-series data, with minimal material consumption. Experimental development and validation of the instrument...

  8. Air purification by catalytic oxidation in a reactor with periodic flow reversal

    NARCIS (Netherlands)

    van de Beld, L.; van de Beld, Bert; Westerterp, K.R.

    1994-01-01

    The behaviour of an adiabatic packed bed reactor with periodic flow reversal has been studied by means of model calculations. A heterogeneous model as well as a pseudo-homogeneous model have been developed. It is shown that a high degree of conversion can be obtained in an autothermal process even

  9. Novel swirl-flow reactor for kinetic studies of semiconductor photocatalysis

    NARCIS (Netherlands)

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

    1997-01-01

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

  10. A Secondary Flow Effect on the Heat and Mass Transfer Processes in the Finned Rod Bundles of Gas-cooled Reactors

    Directory of Open Access Journals (Sweden)

    A. A. Dunaitsev

    2017-01-01

    Full Text Available In nuclear power engineering a need to justify an operability of products and their components is of great importance. In high-temperature gas reactors, the critical element affecting the facility reliability is the fuel rod cladding, which in turn leads to the need to gain knowledge in the field of gas dynamics and heat transfer in the reactor core and to increase the detail of the calculation results. For the time being, calculations of reactor core are performed using the proven techniques of per-channel calculations, which show good representativeness and count rate. However, these techniques require additional experimental studies to describe correctly the inter-channel exchange, which, being taken into account, largely affects the pattern of the temperature fields in the region under consideration. Increasingly more relevant and demandable are numerical simulation methods of fluid and gas dynamics, as well as of heat exchange, which consist in the direct solution of the system of differential equations of mass balance, kinetic moment, and energy. Calculation of reactor cores or rod bundles according these techniques does not require additional experimental studies and allows us to obtain the local distributions of flow characteristics in the bundle and the flow characteristics that are hard to measure in the physical experiment.The article shows the calculation results and their analysis for an infinite rod lattice of the reactor core. The results were obtained by the technique of modelling one rod of a regular lattice using the periodic boundary conditions, followed by translating the results to the neighbouring rods. In channels of complex shape, there are secondary flows caused by changes in the channel geometry along the flow and directed across the main front of the flow. These secondary flows in the reactor cores with rods spaced by the winding wire lead to a redistribution of the coolant along the channel section, which in turn

  11. Thermal hydraulics modeling of the US Geological Survey TRIGA reactor

    Science.gov (United States)

    Alkaabi, Ahmed K.

    The Geological Survey TRIGA reactor (GSTR) is a 1 MW Mark I TRIGA reactor located in Lakewood, Colorado. Single channel GSTR thermal hydraulics models built using RELAP5/MOD3.3, RELAP5-3D, TRACE, and COMSOL Multiphysics predict the fuel, outer clad, and coolant temperatures as a function of position in the core. The results from the RELAP5/MOD3.3, RELAP5-3D, and COMSOL models are similar. The TRACE model predicts significantly higher temperatures, potentially resulting from inappropriate convection correlations. To more accurately study the complex fluid flow patterns within the core, this research develops detailed RELAP5/MOD3.3 and COMSOL multichannel models of the GSTR core. The multichannel models predict lower fuel, outer clad, and coolant temperatures compared to the single channel models by up to 16.7°C, 4.8°C, and 9.6°C, respectively, as a result of the higher mass flow rates predicted by these models. The single channel models and the RELAP5/MOD3.3 multichannel model predict that the coolant temperatures in all fuel rings rise axially with core height, as the coolant in these models flows predominantly in the axial direction. The coolant temperatures predicted by the COMSOL multichannel model rise with core height in the B-, C-, and D-rings and peak and then decrease in the E-, F-, and G-rings, as the coolant tends to flow from the bottom sides of the core to the center of the core in this model. Experiments at the GSTR measured coolant temperatures in the GSTR core to validate the developed models. The axial temperature profiles measured in the GSTR show that the flow patterns predicted by the COMSOL multichannel model are consistent with the actual conditions in the core. Adjusting the RELAP5/MOD3.3 single and multichannel models by modifying the axial and cross-flow areas allow them to better predict the GSTR coolant temperatures; however, the adjusted models still fail to predict accurate axial temperature profiles in the E-, F-, and G-rings.

  12. Technical specifications, Hanford production reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gilbert, W.D. [comp.

    1962-06-25

    These technical specifications are applicable to the eight operating production reactor facilities, B, C, D, DR, F, H, KE, and KW. Covered are operating and performance restrictions and administrative procedures. Areas covered by the operating and performance restrictions are reactivity, reactor control and safety elements, power level, temperature and heat flux, reactor fuel loadings, reactor coolant systems, reactor confinement, test facilities, code compliance, and reactor scram set points. Administrative procedures include process control procedures, training programs, audits and inspections, and reports and records.

  13. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, Hideki; Kimura, Nobuyuki; Miyakoshi, Hiroyuki [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Nagasawa, Kazuyoshi [Nuclear Energy System Incorporation, Oarai Office, Oarai, Ibaraki (Japan)

    2002-08-01

    Decay heat removal using natural circulation is one of significant functions for a reactor. As the decay heat removal system, a direct reactor auxiliary cooling system has been selected in current designs of fast reactors. In this system, cold sodium is provided in an upper plenum of reactor vessel and it covers the reactor core outlet. The cold sodium can penetrate into the gap region between the subassemblies. This gap flow is referred as inter-wrapper flow (IWF). A numerical estimation method for such phenomena was developed, which modeled each subassembly as a rectangular duct with gap region and also the upper plenum. This numerical simulation method was verified by a sodium test and also a water test. We applied this method to the natural circulation in a 600 MWe class fast reactor. The temperature in the core strongly depended on IWF, flow redistribution in the core, and inter-subassembly heat transfer. (author)

  15. Dynamical Safety Analysis of the SABR Fusion-Fission Hybrid Reactor

    Science.gov (United States)

    Sumner, Tyler; Stacey, Weston; Ghiaassian, Seyed

    2009-11-01

    A hybrid fusion-fission reactor for the transmutation of spent nuclear fuel is being developed at Georgia Tech. The Subcritical Advanced Burner Reactor (SABR) is a 3000 MWth sodium-cooled, metal TRU-Zr fueled fast reactor driven by a tokamak fusion neutron source based on ITER physics and technology. We are investigating the accident dynamics of SABR's coupled fission, fusion and heat removal systems to explore the safety characteristics of a hybrid reactor. Possible accident scenarios such as loss of coolant mass flow (LOFA), of power (LOPA) and of heat sink (LOHSA), as well as inadvertent reactivity insertions and fusion source excursion are being analyzed using the RELAP5-3D code, the ATHENA version of which includes liquid metal coolants.

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

    Science.gov (United States)

    Kalyuzhnyi, Sergey V; Fedorovich, Vyacheslav V; Lens, Piet

    2006-03-01

    A new approach to model upflow anaerobic sludge bed (UASB)-reactors, referred to as a one-dimensional dispersed plug flow model, was developed. This model focusses on the granular sludge dynamics along the reactor height, based on the balance between dispersion, sedimentation and convection using one-dimensional (with regard to reactor height) equations. A universal description of both the fluid hydrodynamics and granular sludge dynamics was elaborated by applying known physical laws and empirical relations derived from experimental observations. In addition, the developed model includes: (1) multiple-reaction stoichiometry, (2) microbial growth kinetics, (3) equilibrium chemistry in the liquid phase, (4) major solid-liquid-gas interactions, and (5) material balances for dissolved and solid components along the reactor height. The integrated model has been validated with a set of experimental data on the start-up, operation performance, sludge dynamics, and solute intermediate concentration profiles of a UASB reactor treating cheese whey [Yan et al. (1989) Biol Wastes 27:289-305; Yan et al. (1993) Biotechnol Bioeng 41:700-706]. A sensitivity analysis of the model, performed with regard to the seed sludge characteristics and the key model parameters, showed that the output of the dispersed plug flow model was most influenced by the sludge settleability characteristics and the growth properties (especially mu(m)) of both protein-degrading bacteria and acetotrophic methanogens.

  17. A Development of Technical Specification of a Research Reactor with Plate Fuels Cooled by Upward Flow

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sujin; Kim, Jeongeun; Kim, Hyeonil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The contents of the TS(Technical Specifications) are definitions, safety limits, limiting safety system settings, limiting conditions for operation, surveillance requirements, design features, and administrative controls. TS for Nuclear Power Plants (NPPs) have been developed since many years until now. On the other hands, there are no applicable modernized references of TS for research reactors with many differences from NPPs in purpose and characteristics. Fuel temperature and Departure from Nuclear Boiling Ratio (DNBR) are being used as references from the thermal-hydraulic analysis point of view for determining whether the design of research reactors satisfies acceptance criteria for the nuclear safety or not. Especially for research reactors using plate-type fuels, fuel temperature and critical heat flux, however, are very difficult to measure during the reactor operation. This paper described the outline of main contents of a TS for open-pool research reactor with plate-type fuels using core cooling through passive systems, where acceptance criteria for nuclear safety such as CHF and fuel temperature cannot be directly measured, different from circumstances in NPPs. Thus, three independent variables instead of non-measurable acceptance criteria: fuel temperature and CHF are considered as safety limits, i.e., power, flow, and flow temperature.

  18. Mitigate Strategy of Very High Temperature Reactor Air-ingress Accident

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Tae Kyu [KHNP CRI, Daejeon (Korea, Republic of); Arcilesi, David J.; Sun, Xiaodong; Christensen, Richard N. [The Ohio State University, Columbus (United States); Oh, Chang H.; Kim, Eung S. [Idaho National Laboratory, Idaho (United States)

    2016-10-15

    A critical safety event of the Very High Temperature Reactor (VHTR) is a loss-of-coolant accident (LOCA). Since a VHTR uses graphite as a core structure, if there is a break on the pressure vessel, the air in the reactor cavity could ingress into the reactor core. The worst case scenario of the accident is initiated by a double-ended guillotine break of the cross vessel that connects the reactor vessel and the power conversion unit. The operating pressures in the vessel and containment are about 7 and 0.1 MPa, respectively. In the VHTR, the reactor pressure vessel is located within a reactor cavity which is filled with air during normal operation. Therefore, the air-helium mixture in the cavity may ingress into the reactor pressure vessel after the depressurization process. In this paper, a commercial computational fluid dynamics (CFD) tool, FLUENT, was used to figure out air-ingress mitigation strategies in the gas-turbine modular helium reactor (GT-MHR) designed by General Atomics, Inc. After depressurization, there is almost no air in the reactor cavity; however, the air could flow back to the reactor cavity since the reactor cavity is placed in the lowest place in the reactor building. The heavier air could flow to the reactor cavity through free surface areas in the reactor building. Therefore, Argon gas injection in the reactor cavity is introduced. The injected argon would prevent the flow by pressurizing the reactor cavity initially, and eventually it prevents the flow by making the gas a heavier density than air in the reactor cavity. The gate opens when the reactor cavity is pressurized during the depressurization and it closes by gravity when the depressurization is terminated so that it can slow down the air flow to the reactor cavity.

  19. Gas-cooled fast breeder reactor. Quarterly progress report, February 1-April 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-01

    Information is presented concerning the reactor vessel; reactivity control mechanisms and instrumentation; reactor internals; primary coolant circuits;core auxiliary cooling system; reactor core; systems engineering; and reactor safety and reliability;

  20. Analysis of granular flow in a pebble-bed nuclear reactor.

    Science.gov (United States)

    Rycroft, Chris H; Grest, Gary S; Landry, James W; Bazant, Martin Z

    2006-08-01

    Pebble-bed nuclear reactor technology, which is currently being revived around the world, raises fundamental questions about dense granular flow in silos. A typical reactor core is composed of graphite fuel pebbles, which drain very slowly in a continuous refueling process. Pebble flow is poorly understood and not easily accessible to experiments, and yet it has a major impact on reactor physics. To address this problem, we perform full-scale, discrete-element simulations in realistic geometries, with up to 440,000 frictional, viscoelastic 6-cm-diam spheres draining in a cylindrical vessel of diameter 3.5m and height 10 m with bottom funnels angled at 30 degrees or 60 degrees. We also simulate a bidisperse core with a dynamic central column of smaller graphite moderator pebbles and show that little mixing occurs down to a 1:2 diameter ratio. We analyze the mean velocity, diffusion and mixing, local ordering and porosity (from Voronoi volumes), the residence-time distribution, and the effects of wall friction and discuss implications for reactor design and the basic physics of granular flow.

  1. Recent advances in AFB biomass gasification pilot plant with catalytic reactors in a downstream slip flow

    Energy Technology Data Exchange (ETDEWEB)

    Aznar, M.P.; Gil, J.; Martin, J.A.; Frances, E.; Olivares, A.; Caballero, M.A.; Perez, P. [Saragossa Univ. (Spain). Dept. of Chemistry and Environment; Corella, J. [Madrid Univ. (Spain)

    1996-12-31

    A new 3rd generation pilot plant is being used for hot catalytic raw gas cleaning. It is based on a 15 cm. i.d. fluidized bed with biomass throughputs of 400-650 kg/h.m{sup 2}. Gasification is performed using mixtures of steam and oxygen. The produced gas is passed in a slip flow by two reactors in series containing a calcined dolomite and a commercial reforming catalyst. Tars are periodically sampled and analysed after the three reactors. Tar conversions of 99.99 % and a 300 % increase of the hydrogen content in the gas are obtained. (author) (2 refs.)

  2. Photo-induced copper-mediated polymerization of methyl acrylate in continuous flow reactors

    OpenAIRE

    Wenn, Benjamin; CONRADI, Matthias; Carreiras, Andre Demetrio; Haddleton, David M.; Junkers, Thomas

    2014-01-01

    Photo-induced copper-mediated radical polymerization of methyl acrylate (MA) is carried out in DMSO at 15 °C in a tubular photo-flow reactor as well as in a glass-chip based microreactor. Polymerization reactions proceed rapidly to approximately 90% monomer conversion within 20 minutes of reactor residence time. Control of reactions is high as evidenced by ideal polymerization kinetics, low dispersities of the obtained polymers (in the range of 1.1) and linear evolution of number average mole...

  3. Entrained Flow Reactor Study of K-Capture by Solid Additives

    DEFF Research Database (Denmark)

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

    2016-01-01

    A method to simulate the reaction between gaseous K-species and solid additives, at suspension fired conditions has been developed, using an entrained flow reactor (EFR). A water slurry containing solid additives (kaolin or coal fly ash) and KCl, is injected into the EFR and the solid products...... of additives, rose when increasing the molar ratio of K/(Al+Si) in the reactants. A change of the reaction temperature, from 1100 °C to 1450 °C, did not significantly influence the extent of the reaction, which is in contradiction to the trend observed in previous fixed-bed reactor studies. The method using...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

  5. Parameter analysis calculation on characteristics of portable FAST reactor

    Energy Technology Data Exchange (ETDEWEB)

    Otsubo, Akira; Kowata, Yasuki [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-06-01

    In this report, we performed a parameter survey analysis by using the analysis program code STEDFAST (Space, TErrestrial and Deep sea FAST reactor-gas turbine system). Concerning the deep sea fast reactor-gas turbine system, calculations with many variable parameters were performed on the base case of a NaK cooled reactor of 40 kWe. We aimed at total equipment weight and surface area necessary to remove heat from the system as important values of the characteristics of the system. Electric generation power and the material of a pressure hull were specially influential for the weight. The electric generation power, reactor outlet/inlet temperatures, a natural convection heat transfer coefficient of sea water were specially influential for the area. Concerning the space reactor-gas turbine system, the calculations with the variable parameters of compressor inlet temperature, reactor outlet/inlet temperatures and turbine inlet pressure were performed on the base case of a Na cooled reactor of 40 kWe. The first and the second variable parameters were influential for the total equipment weight of the important characteristic of the system. Concerning the terrestrial fast reactor-gas turbine system, the calculations with the variable parameters of heat transferred pipe number in a heat exchanger to produce hot water of 100degC for cogeneration, compressor stage number and the kind of primary coolant material were performed on the base case of a Pb cooled reactor of 100 MWt. In the comparison of calculational results for Pb and Na of primary coolant material, the primary coolant weight flow rate was naturally large for the former case compared with for the latter case because density is very different between them. (J.P.N.)

  6. The electrochemistry of IGSCC mitigation in BWR coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, D.D. [Center for Electrochemical Science and Technology, The Pennsylvania State Univ., University Park, PA (United States)

    2002-07-01

    A brief review is presented of the electrochemical mitigation of IGSCC in water-cooled reactor heat transport circuit structural materials. Electrochemical control and mitigation is possible, because of the existence of a critical potential for IGSCC and by the feasibility of modifying the environment to displace the corrosion potential (ECP) to a value that is more negative than the critical value. However, even in cases where the ECP cannot be displaced sufficiently in the negative direction to become more negative than the critical potential, considerable advantage is accrued, because of the roughly exponential dependence of crack growth rate on potential. The most important parameters in affecting electrochemical control over the ECP and crack growth rate are the kinetic parameters (exchange current densities and Tafel constants) for the redox reactions involving the principal radiolysis products of water (O{sub 2}, H{sub 2}, H{sub 2}O{sub 2}), external solution composition (concentrations of O{sub 2}, H{sub 2}O{sub 2}, and H{sub 2}), flow velocity, and the conductivity of the bulk environment. The kinetic parameters for the redox reactions essentially determine the charge transfer impedance of the steel surface, which is shown to be one of the key parameters in affecting the magnitude of the coupling current and hence the crack growth rate. The exchange current densities, in particular, are amenable to control by catalysis or inhibition, with the result that surface modification techniques are highly effective in controlling and mitigating IGSCC in reactor coolant circuit materials. (authors)

  7. One-step synthesis of pyridines and dihydropyridines in a continuous flow microwave reactor.

    Science.gov (United States)

    Bagley, Mark C; Fusillo, Vincenzo; Jenkins, Robert L; Lubinu, M Caterina; Mason, Christopher

    2013-01-01

    The Bohlmann-Rahtz pyridine synthesis and the Hantzsch dihydropyridine synthesis can be carried out in a microwave flow reactor or using a conductive heating flow platform for the continuous processing of material. In the Bohlmann-Rahtz reaction, the use of a Brønsted acid catalyst allows Michael addition and cyclodehydration to be carried out in a single step without isolation of intermediates to give the corresponding trisubstituted pyridine as a single regioisomer in good yield. Furthermore, 3-substituted propargyl aldehydes undergo Hantzsch dihydropyridine synthesis in preference to Bohlmann-Rahtz reaction in a very high yielding process that is readily transferred to continuous flow processing.

  8. Liquid metal cooled nuclear reactor plant system

    Science.gov (United States)

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

  9. Simulation of a large break loss of coolant (LBLOCA), without actuation of the emergency injection systems (ECCS) for a BWR-5; Simulacion de un escenario de perdida de refrigerante grande (LBLOCA), sin actuacion de los sistemas de inyeccion de emergencia (ECCS) para un reactor BWR-5

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

    In this paper the analysis of scenario for the loss of coolant case was realized with break at the bottom of a recirculation loop of a BWR-5 with containment type Mark II and a thermal power of 2317 MWt considering that not have coolant injection. This in order to observe the speed of progression of the accident, the phenomenology of the scenario, the time to reach the limit pressure of containment venting and the amount of radionuclides released into the environment. This simulation was performed using the MELCOR code version 2.1. The scenario posits a break in one of the shear recirculation loops. The emergency core cooling system (ECCS) and the reactor core isolation cooling (Rcic) have not credit throughout the event, which allowed achieve greater severity on scenario. The venting of the primary containment was conducted via valve of 30 inches instead of the line of 24 inches of wet well, this in order to have a larger area of exhaust of fission products directly to the reactor building. The venting took place when the pressure in the primary containment reached the 4.5 kg/cm{sup 2} and remained open for the rest of the scenario to maximize the amount released of radionuclides to the atmosphere. The safety relief valves were considered functional they do not present mechanical failure or limit their ability to release pressure due to the large number of performances in safety mode. The results of the analysis covers about 48 hours, time at which the accident evolution was observed; behavior of level, pressure in the vessel and the fuel temperature profile was analyzed. For progression of the scenario outside the vessel, the pressure and temperature of the primary containment, level and temperature of the suppression pool, the hydrogen accumulation in the container and the radionuclides mass released into the atmosphere were analyzed. (Author)

  10. Some Specific CASL Requirements for Advanced Multiphase Flow Simulation of Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    R. A. Berry

    2010-11-01

    Because of the diversity of physical phenomena occuring in boiling, flashing, and bubble collapse, and of the length and time scales of LWR systems, it is imperative that the models have the following features: • Both vapor and liquid phases (and noncondensible phases, if present) must be treated as compressible. • Models must be mathematically and numerically well-posed. • The models methodology must be multi-scale. A fundamental derivation of the multiphase governing equation system, that should be used as a basis for advanced multiphase modeling in LWR coolant systems, is given in the Appendix using the ensemble averaging method. The remainder of this work focuses specifically on the compressible, well-posed, and multi-scale requirements of advanced simulation methods for these LWR coolant systems, because without these are the most fundamental aspects, without which widespread advancement cannot be claimed. Because of the expense of developing multiple special-purpose codes and the inherent inability to couple information from the multiple, separate length- and time-scales, efforts within CASL should be focused toward development of a multi-scale approaches to solve those multiphase flow problems relevant to LWR design and safety analysis. Efforts should be aimed at developing well-designed unified physical/mathematical and high-resolution numerical models for compressible, all-speed multiphase flows spanning: (1) Well-posed general mixture level (true multiphase) models for fast transient situations and safety analysis, (2) DNS (Direct Numerical Simulation)-like models to resolve interface level phenmena like flashing and boiling flows, and critical heat flux determination (necessarily including conjugate heat transfer), and (3) Multi-scale methods to resolve both (1) and (2) automatically, depending upon specified mesh resolution, and to couple different flow models (single-phase, multiphase with several velocities and pressures, multiphase with single

  11. Experimental studies into the dependences of the axial lead coolant pump performance on the impeller cascade parameters

    Directory of Open Access Journals (Sweden)

    A.V. Beznosov

    2017-06-01

    Full Text Available The paper presents results of experimental studies into the dependences of the axial lead coolant pump performance (delivery, head, efficiency on the impeller cascade parameters, including the number of blades, the cascade blade angle and the cascade solidity. The studies were conducted as applied to conditions of small and medium sized plants based on lead cooled fast neutron reactors with horizontal steam generators. The designs of such plants are now in the process of elaboration at Nizhny Novgorod State Technical University (NNSTU. The studies were conducted at NNSTU's FT-4 test facility at a lead coolant temperature of 440–500°C. In the process of investigations, the number of blades in the form of flat plates was 3, 4, 6 and 8, the cascade blade angle was in a range of 9–43°, and the cascade solidity (0.6–1.2 was changed by changing the blade section chord length. The shaft speed of the NNSTU's NSO-01 pump, onto which changeable impellers were installed, was changed in steps of 100 rev/min in an interval of 600–1100 rev/min. The blade diameter was about 200mm, and the maximum lead coolant flow rate in the course of the tests reached ∼2000t/h. The performance of 27 impellers was investigated. It is recommended that the investigation results should be used in design of axial HLMC pumps.

  12. Comparison of heavy metal toxicity in continuous flow and batch reactors

    Science.gov (United States)

    Sengor, S. S.; Gikas, P.; Moberly, J. G.; Peyton, B. M.; Ginn, T. R.

    2009-12-01

    The presence of heavy metals may significantly affect microbial growth. In many cases, small amounts of particular heavy metals may stimulate microbial growth; however, larger quantities may result in microbial growth reduction. Environmental parameters, such as growth pattern may alter the critical heavy metal concentration, above which microbial growth stimulation turns to growth inhibition. Thus, it is important to quantify the effects of heavy metals on microbial activity for understanding natural or manmade biological reactors, either in situ or ex situ. Here we compare the toxicity of Zn and Cu on Arthrobacter sp., a heavy metal tolerant microorganism, under continuous flow versus batch reactor operations. Batch and continuous growth tests of Arthrobacter sp. were carried out at various individual and combined concentrations of Zn and Cu. Biomass concentration (OD) was measured for both the batch and continuous reactors, whereas ATP, oxygen uptake rates and substrate concentrations were additionally measured for the continuous system. Results indicated that Cu was more toxic than Zn under all conditions for both systems. In batch reactors, all tested Zn concentrations up to 150 uM showed a stimulatory effect on microbial growth. However, in the case of mixed Zn and Cu exposures, the presence of Zn either eliminated (at the 50 uM level both Zn and Cu) or reduced by ~25% (at the 100 and 150 uM levels both Zn and Cu) the Cu-induced inhibition. In the continuous system, only one test involved combined Cu (40uM) and Zn (125uM) and this test showed similar results to the 40uM Cu continuous test, i.e., no reduction in inhibition. The specific ATP concentration, i.e., ATP/OD, results for the continuous reactor showed an apparent recovery for both Cu-treated populations, although neither the OD nor glucose data showed any recovery. This may reflect that the individual microorganisms that survived after the addition of heavy metals, kept maintaining the usual ATP

  13. Development of a plant dynamics computer code for analysis of a supercritical carbon dioxide Brayton cycle energy converter coupled to a natural circulation lead-cooled fast reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Moisseytsev, A.; Sienicki, J. J.

    2007-03-08

    STAR-LM is a lead-cooled pool-type fast reactor concept operating under natural circulation of the coolant. The reactor core power is 400 MWt. The open-lattice core consists of fuel pins attached to the core support plate, (the does not consist of removable fuel assemblies). The coolant flows outside of the fuel pins. The fuel is transuranic nitride, fabricated from reprocessed LWR spent fuel. The cladding material is HT-9 stainless steel; the steady-state peak cladding temperature is 650 C. The coolant is single-phase liquid lead under atmospheric pressure; the core inlet and outlet temperatures are 438 C and 578 C, respectively. (The Pb coolant freezing and boiling temperatures are 327 C and 1749 C, respectively). The coolant is contained inside of a reactor vessel. The vessel material is Type 316 stainless steel. The reactor is autonomous meaning that the reactor power is self-regulated based on inherent reactivity feedbacks and no external power control (through control rods) is utilized. The shutdown (scram) control rods are used for startup and shutdown and to stop the fission reaction in case of an emergency. The heat from the reactor is transferred to the S-CO{sub 2} Brayton cycle in in-reactor heat exchangers (IRHX) located inside the reactor vessel. The IRHXs are shell-and-tube type heat exchangers with lead flowing downwards on the shell side and CO{sub 2} flowing upwards on the tube side. No intermediate circuit is utilized. The guard vessel surrounds the reactor vessel to contain the coolant, in the very unlikely event of reactor vessel failure. The Reactor Vessel Auxiliary Cooling System (RVACS) implementing the natural circulation of air flowing upwards over the guard vessel is used to cool the reactor, in the case of loss of normal heat removal through the IRHXs. The RVACS is always in operation. The gap between the vessels is filled with liquid lead-bismuth eutectic (LBE) to enhance the heat removal by air by significantly reducing the thermal

  14. The effects of acoustic flow and mechanical flow on the sonochemical efficiency in a rectangular sonochemical reactor.

    Science.gov (United States)

    Kojima, Yoshihiro; Asakura, Yoshiyuki; Sugiyama, Genki; Koda, Shinobu

    2010-08-01

    Visualization of cavitation behavior in a rectangular sonochemical reactor at 490 kHz was carried out by a laser sheet technique and the distribution of liquid flow was measured by a laser Doppler velocimeter. The pattern of liquid flow and distribution of acoustic pressure of the rectangular sonochemical reactor were investigated as a function of the input power from 10 to 50 W. The liquid moved upward above the transducer at every power. As increasing the input power, the random flow out side the cylindrical part above the transducer changed into the convective one and the region of the visualized standing wave which was formed in the cylindrical part changed with the input power. The position showing the sonochemical luminescence exists inside or near the region where the standing wave was visualized. Introduction of a stirrer resulted in disturbance of liquid flow and expanded the position showing the sonochemical luminescence, but the luminescence intensity was weakened. The sonochemical efficiency was enhanced by about twice by introduction of the stirrer. From these results, we discussed the effects of liquid flow on sonochemical efficiency with and without a stirrer. Copyright 2009 Elsevier B.V. All rights reserved.

  15. Treatment of domestic wastewater in an up-flow anaerobic sludge blanket reactor followed by moving bed biofilm reactor.

    Science.gov (United States)

    Tawfik, A; El-Gohary, F; Temmink, H

    2010-02-01

    The performance of a laboratory-scale sewage treatment system composed of an up-flow anaerobic sludge blanket (UASB) reactor and a moving bed biofilm reactor (MBBR) at a temperature of (22-35 degrees C) was evaluated. The entire treatment system was operated at different hydraulic retention times (HRT's) of 13.3, 10 and 5.0 h. An overall reduction of 80-86% for COD(total); 51-73% for COD(colloidal) and 20-55% for COD(soluble) was found at a total HRT of 5-10 h, respectively. By prolonging the HRT to 13.3 h, the removal efficiencies of COD(total), COD(colloidal) and COD(soluble) increased up to 92, 89 and 80%, respectively. However, the removal efficiency of COD(suspended) in the combined system remained unaffected when increasing the total HRT from 5 to 10 h and from 10 to 13.3 h. This indicates that, the removal of COD(suspended) was independent on the imposed HRT. Ammonia-nitrogen removal in MBBR treating UASB reactor effluent was significantly influenced by organic loading rate (OLR). 62% of ammonia was eliminated at OLR of 4.6 g COD m(-2) day(-1). The removal efficiency was decreased by a value of 34 and 43% at a higher OLR's of 7.4 and 17.8 g COD m(-2) day(-1), respectively. The mean overall residual counts of faecal coliform in the final effluent were 8.9 x 10(4) MPN per 100 ml at a HRT of 13.3 h, 4.9 x 10(5) MPN per 100 ml at a HRT of 10 h and 9.4 x 10(5) MPN per 100 ml at a HRT of 5.0 h, corresponding to overall log(10) reduction of 2.3, 1.4 and 0.7, respectively. The discharged sludge from UASB-MBBR exerts an excellent settling property. Moreover, the mean value of the net sludge yield was only 6% in UASB reactor and 7% in the MBBR of the total influent COD at a total HRT of 13.3 h. Accordingly, the use of the combined UASB-MBBR system for sewage treatment is recommended at a total HRT of 13.3 h.

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

    KAUST Repository

    Hatzell, Marta C.

    2014-12-09

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

  17. Physics Study of Canada Deuterium Uranium Lattice with Coolant Void Reactivity Analysis

    Directory of Open Access Journals (Sweden)

    Jinsu Park

    2017-02-01

    Full Text Available This study presents a coolant void reactivity analysis of Canada Deuterium Uranium (CANDU-6 and Advanced Canada Deuterium Uranium Reactor-700 (ACR-700 fuel lattices using a Monte Carlo code. The reactivity changes when the coolant was voided were assessed in terms of the contributions of four factors and spectrum shifts. In the case of single bundle coolant voiding, the contribution of each of the four factors in the ACR-700 lattice is large in magnitude with opposite signs, and their summation becomes a negative reactivity effect in contrast to that of the CANDU-6 lattice. Unlike the coolant voiding in a single fuel bundle, the 2 × 2 checkerboard coolant voiding in the ACR-700 lattice shows a positive reactivity effect. The neutron current between the no-void and voided bundles, and the four factors of each bundle were analyzed to figure out the mechanism of the positive coolant void reactivity of the checkerboard voiding case. Through a sensitivity study of fuel enrichment, type of burnable absorber, and moderator to fuel volume ratio, a design strategy for the CANDU reactor was suggested in order to achieve a negative coolant void reactivity even for the checkerboard voiding case.

  18. Physics study of Canada deuterium uranium lattice with coolant void reactivity analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Su; Lee, Hyun Suk; Tak, Tae Woo; Lee, Deok Jung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Shin, Ho Cheol [Korea Hydro and Nuclear Power Central Research Institute (KHNP-CRI), Daejeon (Korea, Republic of)

    2017-02-15

    This study presents a coolant void reactivity analysis of Canada Deuterium Uranium (CANDU)-6 and Advanced Canada Deuterium Uranium Reactor-700 (ACR-700) fuel lattices using a Monte Carlo code. The reactivity changes when the coolant was voided were assessed in terms of the contributions of four factors and spectrum shifts. In the case of single bundle coolant voiding, the contribution of each of the four factors in the ACR-700 lattice is large in magnitude with opposite signs, and their summation becomes a negative reactivity effect in contrast to that of the CANDU-6 lattice. Unlike the coolant voiding in a single fuel bundle, the 2 x 2 checkerboard coolant voiding in the ACR-700 lattice shows a positive reactivity effect. The neutron current between the no-void and voided bundles, and the four factors of each bundle were analyzed to figure out the mechanism of the positive coolant void reactivity of the checkerboard voiding case. Through a sensitivity study of fuel enrichment, type of burnable absorber, and moderator to fuel volume ratio, a design strategy for the CANDU reactor was suggested in order to achieve a negative coolant void reactivity even for the checkerboard voiding case.

  19. Biodegradation of redox dye Methylene Blue by up-flow anaerobic sludge blanket reactor.

    Science.gov (United States)

    Ong, Soon-An; Toorisaka, Eiichi; Hirata, Makoto; Hano, Tadashi

    2005-09-30

    The objective of this study is to evaluate the decolorization of Methylene Blue (MB) by an up-flow anaerobic sludge blanket (UASB) reactor. The UASB reactor was operated under batch condition with total treatment volume of 3l and operation time of 24 h per batch. It was found that the color of MB disappeared within a few minutes after entering into the UASB reactor due to reduction by anaerobic biomass. However, the reduced MB was re-oxidized again by air after discharged from the reactor and thus caused low color removal efficiency. The presence of suitable amount of organic content (sucrose and peptone) as an electron donor played an important factor for color removal. It was observed that more than 90% of color removal efficiency was achieved in the UASB reactor with 0.627 mmoll(-1) of MB concentration and the presence of low amount of organic content (<0.45 g COD/(ld)). Biological dye reduction kinetics depends on the concentration of dye and reducing equivalents. The kinetic behavior of MB biodegradation by microbes was also investigated to determine the model involved in the process.

  20. Anaerobic digestion of pineapple pulp and peel in a plug-flow reactor.

    Science.gov (United States)

    Namsree, Pimjai; Suvajittanont, Worakrit; Puttanlek, Chureerat; Uttapap, Dudsadee; Rungsardthong, Vilai

    2012-11-15

    The objective of this research was to study the production of biogas by using pineapple pulp and peel, the by-products from fruit processing plants, in a plug-flow reactor (17.5 L total volume). The effects of feed concentration, total solids (TS) and hydraulic retention time (HRT) on degradation of the waste were investigated. The increase of pineapple pulp and peel of 2% (wt/vol) at HRT 7 d to 4% (wt/vol) at HRT 10 d showed increases in biogas production rate, biogas yield and methane yield - from 0.12 v/v-d, 0.26 m(3)/kg COD removed and 0.11 m(3)/kg COD removed, with COD removal at 64.1%, to 0.25 v/v-d, 0.43 m(3)/kg COD removed and 0.14 m(3)/kg COD removed, with COD removal at 60.41%. The methanogenic fermentation was more active in the middle and final parts of the reactor. The recirculation of fermentation effluent at 40% (vol/vol) of the working volume into the reactor could increase the biogas production rate and biogas yield up to 52% and 12%, respectively. The results showed technological potential for waste treatment of pineapple pulp and peel in a plug-flow reactor. Copyright © 2012 Elsevier Ltd. All rights reserved.

  1. Measurement of Turbulent Fluxes of Swirling Flow in a Scaled Up Multi Inlet Vortex Reactor

    Science.gov (United States)

    Olsen, Michael; Hitimana, Emmanual; Hill, James; Fox, Rodney

    2017-11-01

    The multi-inlet vortex reactor (MIVR) has been developed for use in the FlashNanoprecipitation (FNP) process. The MIVR has four identical square inlets connected to a central cylindrical mixing chamber with one common outlet creating a highly turbulent swirling flow dominated by a strong vortex in the center. Efficient FNP requires rapid mixing within the MIVR. To investigate the mixing, instantaneous velocity and concentration fields were acquired using simultaneous stereoscopic particle image velocimetry and planar laser-induced fluorescence. The simultaneous velocity and concentration data were used to determine turbulent fluxes and spatial cross-correlations of velocity and concentration fluctuations. The measurements were performed for four inlet flow Reynolds numbers (3250, 4875, 6500, and 8125) and at three measurement planes within the reactor. A correlation between turbulent fluxes and vortex strength was found. For all Reynolds numbers, turbulent fluxes are maximum in the vortex dominated central region of the reactor and decay away from the vortex. Increasing Reynolds number increased turbulent fluxes and subsequently enhanced mixing. The mixing performance was confirmed by determining coefficients of concentration variance within the reactor.

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

    Science.gov (United States)

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

    2015-08-01

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

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

    Science.gov (United States)

    Motil, Brian J.; Balakotaiah, Vemuri

    2001-01-01

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

  4. Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

    Science.gov (United States)

    Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

    2017-01-01

    In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

  5. On Reachability and Observability of a Plug-Flow Reactor Diffusion Equation

    Science.gov (United States)

    Sano, Hideki; Nakagiri, Shin-Ichi

    This paper is concerned with the problems of reachability and observability of a plug-flow reactor equation with diffusion term. It is shown that the system with boundary inputs is formulated as an well-defined boundary control system, and that it is approximately reachable at any time through a Fourier series expansion of the solution. It is also shown that the boundary control system associated with only one boundary input is approximately reachable except for special values of diffusion and reaction constants, and the unreachable subspace is determined for the exceptional boundary control systems. Furthermore, the results concerning observability are derived through a concrete expression of the solution to the system of plug-flow reactor diffusion equation with the output equation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-02-01

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

  7. Biomass Gasification Behavior in an Entrained Flow Reactor: Gas Product Distribution and Soot Formation

    DEFF Research Database (Denmark)

    Qin, Ke; Jensen, Peter Arendt; Lin, Weigang

    2012-01-01

    Biomass gasification and pyrolysis were studied in a laboratory-scale atmospheric pressure entrained flow reactor. Effects of operating parameters and biomass types on the syngas composition were investigated. In general, the carbon conversion during biomass gasification was higher than 90......% at the optimal conditions of 1400 °C with steam addition. The biomass carbon that was not converted to gas in the gasification process only appeared as soot particles in the syngas in all of the experiments, except for the two experiments performed at 1000 °C, where a very small amount of char was also left....... In comparison to pyrolysis, lower yields of soot, H2, and CO were produced during gasification. The yield of soot could be reduced by a longer residence time, larger feeder air flow, lower oxygen concentration, higher excess air ratio, higher steam/carbon ratio, and higher reactor temperature. Changes...

  8. Large-Scale Production of CdSe Nanocrystal by a Continuous Flow Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kawa, Manabu, E-mail: 508532@cc.m-kagaku.co.jp; Morii, Hidekazu; Ioku, Atau; Saita, Soichiro [MCC-Group Science and Technology Research Center, Mitsubishi Chemical Corporation (Japan); Okuyama, Kikuo [Hiroshima University, Department of Chemical Engineering, Graduate School of Engineering (Japan)

    2003-04-15

    Organically capped CdSe nanocrystals were successfully produced by a continuous flow reactor in 13 g/h rate as isolated CdSe nanocrystal, using trioctylphosphine oxide (TOPO) both as the capping organic reagent and the high-temperature reaction solvent. Relatively high reaction temperature (e.g. 350 deg. C) was necessary for matured crystal growth. The quality of TOPO (i.e. impurity composition such like phosphonic acids) was also influential on the quality of the resulting CdSe nanocrystal. The continuous flow reactor was able to produce highly-luminescence, monodispersed CdSe nanocrystals, confirmed by transmission electron microscope observation. The production rate was stable at least 1 h to allow over 10 g production.

  9. Electrochemical oxidation of wastewater containing aromatic amines using a flow electrolytic reactor

    OpenAIRE

    de Moraes, PB; Bertazzoli, R.; Bidoia, ED

    2013-01-01

    Aromatic amines are environmental pollutants and represent one of the most important classes of industrial and natural chemicals. Some types of complex effluents containing these chemical species, mainly those originated from chemicals plants are not fully efficiently treated by conventional processes. In this work, the use of electrochemical technology through an electrolytic pilot scale flow reactor is considered for treatment of wastewater of a chemical industry manufacturer of antioxidant...

  10. Synthesis of quantum dot nanocrystals and plasmonic nanoparticles using a segmented flow reactor

    Science.gov (United States)

    Mbwahnche, R. C.; Matyushkin, L. B.; Ryzhov, O. A.; Aleksandrova, O. A.; Moshnikov, V. A.

    2017-01-01

    The purpose of this research is to develop an automated method of synthesizing quantum dot nanocrystals and plasmonic nanoparticles using segmented flow rector synthesis as a new alternative to the batch method of synthesizing nanoparticles. A reactor was successfully applied to the synthesis of colloidal solutions of semiconductor (CdSe) and metal (Ag) nanoparticles. This instrument is applicable in both material science laboratories and industry.

  11. Operation of a catalytic reverse flow reactor for the purification of air contamined with volatile organic compounds

    NARCIS (Netherlands)

    van de Beld, L.; van de Beld, L.; Westerterp, K.R.

    1997-01-01

    Catalytic oxidation in a reverse flow reactor is an attractive process for the decontamination of air polluted with volatile organic compounds (VOCs). In this paper several aspects of operating this type of reactor for air purification under strongly varying conditions will be discussed. For a

  12. Experimental and Numerical Evaluation of the By-Pass Flow in a Catalytic Plate Reactor for Hydrogen Production

    DEFF Research Database (Denmark)

    Sigurdsson, Haftor Örn; Kær, Søren Knudsen

    2011-01-01

    Numerical and experimental study is performed to evaluate the reactant by-pass flow in a catalytic plate reactor with a coated wire mesh catalyst for steam reforming of methane for hydrogen generation. By-pass of unconverted methane is evaluated under different wire mesh catalyst width to reactor...

  13. Highly Efficient Photocatalysts and Continuous-Flow Photocatalytic Reactors for Degradation of Organic Pollutants in Wastewater.

    Science.gov (United States)

    Chang, Sujie; Yang, Xiaoqiu; Sang, Yuanhua; Liu, Hong

    2016-09-06

    One of the most important applications for photocatalysis is engineered water treatment that photodegrades organic pollutants in wastewater at low cost. To overcome the low efficiency of batch degradation methods, continuous-flow photocatalytic reactors have been proposed and have become the most promising method for mass water treatment. However, most commercial semiconductor photocatalysts are granular nanoparticles with low activity and a narrow active light wavelength band; this creates difficulties for direct use in continuous-flow photocatalytic reactors. Therefore, a high-performance photodegradation photocatalyst with proper morphology or structure is key for continuous photocatalytic degradation. Moreover, a well-designed photocatalytic device is another important component for continuous-flow photocatalysis and determines the efficiency of photocatalysis in practical water treatment. This review describes the basic design principles and synthesis of photocatalysts with excellent performance and special morphologies suitable for a filtering photocatalysis process. Certain promising continuous photodegradation reactors are also categorized and summarized. Additionally, selected scientific and technical problems that must be urgently solved are suggested. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

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

    Science.gov (United States)

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

    2015-07-01

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

  15. Assessment of the heat carrier movement in the primary coolant circuit by its own momentum

    Energy Technology Data Exchange (ETDEWEB)

    Kadalev, Stoyan, E-mail: kadalev@inrne.bas.bg

    2014-10-15

    Highlights: • We model the heat carrier flow alteration after the circulation pump(s) stop. • The general mathematical model used is described in details. • The model is adapted and applied to a particular example research reactor. • Assessment is presented in detail, step by step with references. • The information provided is enough to apply calculations to another facility. - Abstract: In the presented paper is considered the approach to an assessment of the heat carrier flow alteration in the primary water–water reactor coolant circuit after the circulation pump(s) stop. This topic is highly relevant trough advanced and increased nuclear safety requirements because such a process is observed in case of black-out accident or damaged pump(s). The general mathematical model used is described; enabling preparation of this evaluation adapted and applied to a particular example facility namely a pool type research reactor. The factors influencing to the heat carrier movement by its own momentum are examined. The evaluation measures and includes the factors influencing the heat carrier flow rate from the moment the pump(s) stops down to a negligible value. Assessment is presented in detail, step by step and where needed with references to specific data and/or formulae from reference books to allow repetition of the calculations and/or apply to another facility. The calculations are presented utilizing all necessary data according to the design and technological documentation. No account is given to the pressure of the natural circulation caused by the residual heat generation in the fuel after the reactor scram system extinction of the fission reaction.

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

    Directory of Open Access Journals (Sweden)

    Christian H. Hornung

    2017-01-01

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

  17. Diels-Alder reactions of myrcene using intensified continuous-flow reactors.

    Science.gov (United States)

    Hornung, Christian H; Álvarez-Diéguez, Miguel Á; Kohl, Thomas M; Tsanaktsidis, John

    2017-01-01

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

  18. Optimization of hydrogen dispersion in thermophilic up-flow reactors for ex situ biogas upgrading

    DEFF Research Database (Denmark)

    Bassani, Ilaria; Kougias, Panagiotis; Treu, Laura

    2017-01-01

    This study evaluates the efficiency of four novel up-flow reactors for ex situ biogas upgrading converting externally provided CO2 and H2 to CH4, via hydrogenotrophic methanogenesis. The gases were injected through stainless steel diffusers combined with alumina ceramic sponge or through alumina...... ceramic membranes. Pore size, input gas loading and gas recirculation flow rate were modulated to optimize gas-liquid mass transfer, and thus methanation efficiency. Results showed that larger pore size diffusion devices achieved the best kinetics and output-gas quality converting all the injected H2...

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

    Science.gov (United States)

    Álvarez-Diéguez, Miguel Á; Kohl, Thomas M; Tsanaktsidis, John

    2017-01-01

    This work describes the Diels–Alder reaction of the naturally occurring substituted butadiene, myrcene, with a range of different naturally occurring and synthetic dienophiles. The synthesis of the Diels–Alder adduct from myrcene and acrylic acid, containing surfactant properties, was scaled-up in a plate-type continuous-flow reactor with a volume of 105 mL to a throughput of 2.79 kg of the final product per day. This continuous-flow approach provides a facile alternative scale-up route to conventional batch processing, and it helps to intensify the synthesis protocol by applying higher reaction temperatures and shorter reaction times. PMID:28228853

  20. Power up-grading study for the first Egyptian research reactor

    Energy Technology Data Exchange (ETDEWEB)

    El-Sawy Temraz, H.; Ashoub, N. E-mail: nageeb@pcn.aea.sci.eg; Fathallah, A

    2001-09-01

    In the present work, power up-grading study is performed, for the first Egyptian Research Reactor (ET-RR-1), using the present fuel basket with 4x4 fuel rods, (17.5 mm pitch), and a proposed fuel basket with 5x5 fuel rods, (14.0 mm pitch), without violating the thermal hydraulic safety criteria. These safety criteria are; fuel centerline temperature (fuel melting), clad surface temperature (surface boiling), outlet coolant temperature, and maximum heat flux (critical heat flux ratio). Different thermal reactor powers (2-10 MW) and different core coolant flow rates (450, 900, 1350 m{sup 3} h{sup -1}) are considered. The thermal hydraulic analysis was performed using the subchannel code COBRA-IIIC for the estimation of temperatures, coolant velocities and critical heat flux. The neutronic calculations were performed using WIMS-D4 code with 5-group neutron cross section library. These cross sections were adapted to use in the two-dimensional (2-D) diffusion code DIXY for core calculations. The study concluded that ET-RR-1 power can be upgraded safely up to 4 MW with the present 4x4-fuel basket and with the proposed 5x5-fuel basket up to 5 MW with the present coolant flow rate (900 m{sup 3} h{sup -1}). With the two fuel arrays, the reactor power can be upgraded to 6 MW with coolant flow rate of 1350 m{sup 3} h{sup -1} without violating the safety criterion. It is also concluded that, loading the ET-RR-1 core with the proposed fuel basket (5x5) increases the excess reactivity of the reactor core than the present 4x4 fuel matrix with equal U-235 mass load and gave better fuel economy of fuel utilization.

  1. Benchmark exercise for fluid flow simulations in a liquid metal fast reactor fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Merzari, E., E-mail: emerzari@anl.gov [Mathematics and Computer Science Division, Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Fischer, P. [Mathematics and Computer Science Division, Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Yuan, H. [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL (United States); Van Tichelen, K.; Keijers, S. [SCK-CEN, Boeretang 200, Mol (Belgium); De Ridder, J.; Degroote, J.; Vierendeels, J. [Ghent University, Ghent (Belgium); Doolaard, H.; Gopala, V.R.; Roelofs, F. [NRG, Petten (Netherlands)

    2016-03-15

    Highlights: • A EUROTAM-US INERI consortium has performed a benchmark exercise related to fast reactor assembly simulations. • LES calculations for a wire-wrapped rod bundle are compared with RANS calculations. • Results show good agreement for velocity and cross flows. - Abstract: As part of a U.S. Department of Energy International Nuclear Energy Research Initiative (I-NERI), Argonne National Laboratory (Argonne) is collaborating with the Dutch Nuclear Research and consultancy Group (NRG), the Belgian Nuclear Research Centre (SCK·CEN), and Ghent University (UGent) in Belgium to perform and compare a series of fuel-pin-bundle calculations representative of a fast reactor core. A wire-wrapped fuel bundle is a complex configuration for which little data is available for verification and validation of new simulation tools. UGent and NRG performed their simulations with commercially available computational fluid dynamics (CFD) codes. The high-fidelity Argonne large-eddy simulations were performed with Nek5000, used for CFD in the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) suite. SHARP is a versatile tool that is being developed to model the core of a wide variety of reactor types under various scenarios. It is intended both to serve as a surrogate for physical experiments and to provide insight into experimental results. Comparison of the results obtained by the different participants with the reference Nek5000 results shows good agreement, especially for the cross-flow data. The comparison also helps highlight issues with current modeling approaches. The results of the study will be valuable in the design and licensing process of MYRRHA, a flexible fast research reactor under design at SCK·CEN that features wire-wrapped fuel bundles cooled by lead-bismuth eutectic.

  2. BTEX removal in a horizontal-flow anaerobic immobilized biomass reactor under denitrifying conditions.

    Science.gov (United States)

    Ribeiro, Rogers; de Nardi, Ivana Ribeiro; Fernandes, Bruna Soares; Foresti, Eugenio; Zaiat, Marcelo

    2013-04-01

    Because benzene, toluene, ethylbenzene, and xylenes (BTEX) and ethanol are important contaminants present in Brazilian gasoline, it is essential to develop technology that can be used in the bioremediation of gasoline-contaminated aquifers. This paper evaluates the performance of a horizontal-flow anaerobic immobilized biomass (HAIB) reactor fed with water containing gasoline constituents under denitrifying conditions. Two HAIB reactors filled with polyurethane foam matrices (5 mm cubes, 23 kg/m(3) density and 95 % porosity) for biomass attachment were assayed. The reactor fed with synthetic substrate containing protein, carbohydrates, sodium bicarbonate and BTEX solution in ethanol, at an Hydraulic retention time (HRT) of 13.5 h, presented hydrocarbon removal efficiencies of 99 % at the following initial concentrations: benzene 6.7 mg/L, toluene 4.9 mg/L, m-xylene and p-xylene 7.2 mg/L, ethylbenzene 3.7 mg/L, and nitrate 60 mg N/L. The HAIB reactor fed with gasoline-contaminated water at an HRT of 20 h showed hydrocarbon removal efficiencies of 96 % at the following initial concentrations: benzene, 4.9 mg/L; toluene, 7.2 mg/L; m-xylene, 3.7 mg/L; and nitrate 400 mg N/L. Microbiological observations along the length of the HAIB reactor fed with gasoline-contaminated water confirmed that in the first segment of the reactor, denitrifying metabolism predominated, whereas from the first sampling port on, the metabolism observed was predominantly methanogenic.

  3. Synthesis of Core@Shell Nanostructures in a Continuous Flow Droplet Reactor: Controlling Structure through Relative Flow Rates.

    Science.gov (United States)

    Santana, Joshua S; Koczkur, Kallum M; Skrabalak, Sara E

    2017-06-20

    Bimetallic nanostructures are primarily synthesized in small volume batches. However, droplet-based reactors are receiving attention due to their ability to maintain thermal and compositional equilibrium within and between droplets, enabling flow operations for inline analyses and the scale-up of nanomaterial syntheses. Here, the syntheses of shape-controlled core@shell Au@Pd nanostructures with variable shell thicknesses are reported through control of the relative flow rates of reagents within the microreactor. Specifically, Pd shells were grown on cubic or octahedral Au seeds, selected as a model system. In batch reactions, shell thickness is determined by precursor concentration; however, as shown here, precursor feedstock concentration can be held constant, with the precursor concentration within the droplets being controlled through relative flow rates. This approach allows process conditions to be modified inline rather than from batch to batch to achieve particles with different shell thicknesses, and this procedure should be applicable to other multicomponent systems.

  4. The effects of flow multiplicity on GaN deposition in a rotating disk CVD reactor

    Science.gov (United States)

    Gkinis, P. A.; Aviziotis, I. G.; Koronaki, E. D.; Gakis, G. P.; Boudouvis, A. G.

    2017-01-01

    The effect of gas flow multiplicity, i.e. the possibility of two very different flow regimes prevailing at random in a rotating disk metalorganic chemical vapor deposition (MOCVD) reactor, on the deposited GaN film is investigated. A transport model coupled with a system of chemical reactions in the gas phase and on the wafer where the film is formed, is implemented in the parameter regions where multiple flows are possible. In the region of multiplicity where either plug flow, imposed by forced convection, or buoyancy-dominated flow is possible, the results in the latter case indicate high deposition rate and decreased uniformity. In the former case, increasing the pressure and the rotation rate has a favorable effect on the deposition rate without sacrificing uniformity. In the parameter window of multiplicity where either rotation or combined rotation/buoyancy may prevail, the effects of buoyancy lead to higher deposition rate at the center of the wafer and reduced uniformity. The Arrhenius plots in the regions of multiplicity for exactly the same operating conditions reveal that the system operates in a diffusion-limited regime in the plug flow and in the rotation-dominated flow, in the first and second region of multiplicity respectively. In contrast, in the buoyancy-dominated flow and the combined rotation/buoyancy flow (first and second region of multiplicity respectively) the process shifts into the kinetics-limited regime.

  5. Biogas potential of high strength municipal wastewater treatment in laboratory scale up-flow anaerobic slugde blanket (UASB) reactors

    OpenAIRE

    Safitri, Anissa Sukma

    2016-01-01

    The main focus of this study is investigating the effectiveness of anaerobic treatment of municipal wastewater for converting organic matter to methane production in anaerobic granular sludge reactors. In-house designed laboratory scale, up-flow anaerobic sludge blanket (UASB) reactor systems were set up for treating of high strength municipal wastewater treatment i.e. below 1200 mg COD/l under mesophilic condition (20 – 25 °C). Three UASB reactors were set up in the study; one reactor (React...

  6. Steel slag carbonation in a flow-through reactor system: the role of fluid-flux.

    Science.gov (United States)

    Berryman, Eleanor J; Williams-Jones, Anthony E; Migdisov, Artashes A

    2015-01-01

    Steel production is currently the largest industrial source of atmospheric CO2. As annual steel production continues to grow, the need for effective methods of reducing its carbon footprint increases correspondingly. The carbonation of the calcium-bearing phases in steel slag generated during basic oxygen furnace (BOF) steel production, in particular its major constituent, larnite {Ca2SiO4}, which is a structural analogue of olivine {(MgFe)2SiO4}, the main mineral subjected to natural carbonation in peridotites, offers the potential to offset some of these emissions. However, the controls on the nature and efficiency of steel slag carbonation are yet to be completely understood. Experiments were conducted exposing steel slag grains to a CO2-H2O mixture in both batch and flow-through reactors to investigate the impact of temperature, fluid flux, and reaction gradient on the dissolution and carbonation of steel slag. The results of these experiments show that dissolution and carbonation of BOF steel slag are more efficient in a flow-through reactor than in the batch reactors used in most previous studies. Moreover, they show that fluid flux needs to be optimized in addition to grain size, pressure, and temperature, in order to maximize the efficiency of carbonation. Based on these results, a two-stage reactor consisting of a high and a low fluid-flux chamber is proposed for CO2 sequestration by steel slag carbonation, allowing dissolution of the slag and precipitation of calcium carbonate to occur within a single flow-through system. Copyright © 2014. Published by Elsevier B.V.

  7. Effects of Parallel Channel Interactions on Two-Phase Flow Split in ...

    African Journals Online (AJOL)

    The tests would aid the development of a realistic transient computer model for tracking the distribution of two-phase flows into the multiple parallel channels of a Nuclear Reactor, during Loss of Coolant Accidents (LOCA), and were performed at the General Electric Nuclear Energy Division Laboratory, California. The test ...

  8. Disinfection effect of a continuous-flow ultrasound/ultraviolet baffled reactor at a pilot scale.

    Science.gov (United States)

    Zhou, Xiaoqin; Yan, Yichang; Li, Zifu; Yin, Jie

    2017-07-01

    An ultrasound/ultraviolet (US/UV) baffled reactor was developed to fill the gap in ultraviolet (UV) disinfection associated with disinfection efficiency. According to the previously selected operational condition, a continuous-flow US/UV baffled reactor was continuously operated in a wastewater treatment plant at a pilot scale for nearly three months, and the disinfection influent and effluent were analyzed, including fecal coliforms, Escherichia coli, and fecal streptococci. The US/UV baffled reactor could guarantee a high effluent disinfection performance in terms of fecal coliforms removal even with the fluctuation of the secondary effluent. All the disinfected effluents satisfied the requirement of the "Pollutants Discharge Standard of Municipal Wastewater Treatment Plant in China" (fecal coliforms below 1000CFU/L for class 1A), and 87% of the tested fecal coliforms concentration in the disinfected effluent was below 100CFU/L, nearly eliminating all fecal coliforms. Further analysis of the E. coli and fecal streptococci showed the broad disinfection ability and high disinfection efficiency of the US/UV baffled reactor. The flexibility of the specific energy consumption for the disinfection system depends on the water quality. Copyright © 2017 Elsevier B.V. All rights reserved.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, H.; Kimura, N.; Miyakoshi, H. [Japan Nuclear Cycle Development Institute, Reactor Engineering Group, O-arai Engineering Center, Ibaraki (Japan); Nagasawa, K. [Nuclear Energy System Incorporation, O-arai Office, Ibaraki (Japan)

    2001-07-01

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

  10. Study thermal characteristics of millet grain, dried in the machine with the swirling flow of the coolant and the microwave energy supply

    Directory of Open Access Journals (Sweden)

    S. T. Antipov

    2016-01-01

    Full Text Available The article discusses the problems of determining the thermal characteristics of millet. The choice of the research object. The paper presents the principle of operation of the plant and the parameters of the standard, organic glass for measurements. Method was to study millet grains and organic glass, which are brought into contact on a common plane. The heater is brought into contact with the product and passed the constant heat flow, which passed through a layer of millet grain at different speeds. As a result, the temperature in the contact plane of the changed and recorded on the chart of the potentiometer in the form of the curve, by which you can determine the time and temperature change. The thermal diffusivity and thermal conductivity determined by empirical formulas obtained by solving a system of differential equations, made up for the system of two bodies, one of which includes the unknown thermal characteristics. Test two bodies in contact on a common plane, resulting in mathematical physics principles constitute two differential equations with uniform initial and boundary conditions of the first kind, due to the parameters of ongoing experience. It is a plot of thermal performance of the temperature and humidity. Revealed linear dependence of the physical thermal characteristics, showing that with increasing temperature the thermal diffusivity value decreases, and the thermal conductivity and specific heat capacity are increasing character. Revealed the equations describing the thermal characteristics of millet grain with a humidity in the range of 13.6–35.1% and the temperature range 293–373 K.

  11. Secondary organic aerosol from VOC mixtures in an oxidation flow reactor

    Science.gov (United States)

    Ahlberg, Erik; Falk, John; Eriksson, Axel; Holst, Thomas; Brune, William H.; Kristensson, Adam; Roldin, Pontus; Svenningsson, Birgitta

    2017-07-01

    The atmospheric organic aerosol is a tremendously complex system in terms of chemical content. Models generally treat the mixtures as ideal, something which has been questioned owing to model-measurement discrepancies. We used an oxidation flow reactor to produce secondary organic aerosol (SOA) mixtures containing oxidation products of biogenic (α-pinene, myrcene and isoprene) and anthropogenic (m-xylene) volatile organic compounds (VOCs). The resulting volume concentration and chemical composition was measured using a scanning mobility particle sizer (SMPS) and a high-resolution time-of-flight aerosol mass spectrometer (HR-ToF-AMS), respectively. The SOA mass yield of the mixtures was compared to a partitioning model constructed from single VOC experiments. The single VOC SOA mass yields with no wall-loss correction applied are comparable to previous experiments. In the mixtures containing myrcene a higher yield than expected was produced. We attribute this to an increased condensation sink, arising from myrcene producing a significantly higher number of nucleation particles compared to the other precursors. Isoprene did not produce much mass in single VOC experiments but contributed to the mass of the mixtures. The effect of high concentrations of isoprene on the OH exposure was found to be small, even at OH reactivities that previously have been reported to significantly suppress OH exposures in oxidation flow reactors. Furthermore, isoprene shifted the particle size distribution of mixtures towards larger sizes, which could be due to a change in oxidant dynamics inside the reactor.

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

    Science.gov (United States)

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

    2013-08-01

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

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

    Directory of Open Access Journals (Sweden)

    Josef Schachtner

    2016-08-01

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

  14. Plasma flow reactor for steady state monitoring of physical and chemical processes at high temperatures

    Science.gov (United States)

    Koroglu, Batikan; Mehl, Marco; Armstrong, Michael R.; Crowhurst, Jonathan C.; Weisz, David G.; Zaug, Joseph M.; Dai, Zurong; Radousky, Harry B.; Chernov, Alex; Ramon, Erick; Stavrou, Elissaios; Knight, Kim; Fabris, Andrea L.; Cappelli, Mark A.; Rose, Timothy P.

    2017-09-01

    We present the development of a steady state plasma flow reactor to investigate gas phase physical and chemical processes that occur at high temperature (1000 < T < 5000 K) and atmospheric pressure. The reactor consists of a glass tube that is attached to an inductively coupled argon plasma generator via an adaptor (ring flow injector). We have modeled the system using computational fluid dynamics simulations that are bounded by measured temperatures. In situ line-of-sight optical emission and absorption spectroscopy have been used to determine the structures and concentrations of molecules formed during rapid cooling of reactants after they pass through the plasma. Emission spectroscopy also enables us to determine the temperatures at which these dynamic processes occur. A sample collection probe inserted from the open end of the reactor is used to collect condensed materials and analyze them ex situ using electron microscopy. The preliminary results of two separate investigations involving the condensation of metal oxides and chemical kinetics of high-temperature gas reactions are discussed.

  15. Rapid Determination of Optimal Conditions in a Continuous Flow Reactor Using Process Analytical Technology

    Directory of Open Access Journals (Sweden)

    Michael F. Roberto

    2013-12-01

    Full Text Available Continuous flow reactors (CFRs are an emerging technology that offer several advantages over traditional batch synthesis methods, including more efficient mixing schemes, rapid heat transfer, and increased user safety. Of particular interest to the specialty chemical and pharmaceutical manufacturing industries is the significantly improved reliability and product reproducibility over time. CFR reproducibility can be attributed to the reactors achieving and maintaining a steady state once all physical and chemical conditions have stabilized. This work describes the implementation of a smart CFR with univariate physical and multivariate chemical monitoring that allows for rapid determination of steady state, requiring less than one minute. Additionally, the use of process analytical technology further enabled a significant reduction in the time and cost associated with offline validation methods. The technology implemented for this study is chemistry and hardware agnostic, making this approach a viable means of optimizing the conditions of any CFR.

  16. Coolant mixing in the HPLWR upper plenum

    Energy Technology Data Exchange (ETDEWEB)

    Wank, Alexander; Schulenberg, Thomas; Class, Andreas G. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Kern- und Energietechnik

    2008-08-15

    The High Performance Light Water Reactor (HPLWR) is a new type of reactor cooled with supercritical water. The cooling water is heated in 3 stages. In the first stage, the evaporator, the water flows to the top, then flows down in superheater 1 before again flowing upward in superheater II. To prevent peak temperatures and enthalpies being passed on from one heating stage to the next, and to homogenize temperature, the water is mixed in two mixing chambers between the heating stages. Mixing in the upper plenum was computed as a reference case by the STAR-CD CFD code. For quantitative evaluation of mixing, passive scalars were added to the flow and evaluated. Further studies will be conducted to improve mixing by appropriate design measures. (orig.)

  17. THE STRIPPING OF AMMONIA FROM LANDFILL LEACHATE IN PLUG FLOW REACTORS

    Directory of Open Access Journals (Sweden)

    Danuza Costa Campos

    2010-10-01

    Full Text Available The leachate is a highly liquid polluting, considering, besides high concentrations of ammonia nitrogen. Treating leachate is a very difficult task once its composition is very complex. A king of treatment that is being developed is the stripping of ammonia where this chemical will be removed from the leachate by mass transfer of the liquid phase to gas. Therefore, this paper addressed to study the ammonia stripping process in leachates liquids using plug-flow reactors in series. In order to accomplish the experimental part four plugflow reactors in series were build, with an average height of 50 centimeters, without forced air supplier and without leachate pH adjustment. The leachate sample used was from the metropolitan landfill of Joao Pessoa city, which was collected and transported to the EXTRABES laboratory and made the physicochemical characterization. The experimental monitoring system consisted of four distinct phases, with applied superficial loads of 450, 500, 600 and 700 kg NH4+ ha-1 day-1, hydraulic detention time equal to 65, 60, 50 and 38 days, and it was obtained an average efficiency of ammonia nitrogen removal around 96.1%, 99.7%, 99.5% and 98.5% respectively. It was found that the removal efficiency of NH3 was satisfactory in all phases, however, the kinetic study showed that the higher the surface charge applied to the reactors, the higher the rate constant for removal of NH3 stripping process. Thus we can conclude that the process of stripping ammonia from landfill leachate may be done in plug-flow reactors in series, with an average depth of 50 cm, due to provide high removal efficiency of ammonia nitrogen, with low operating costs.

  18. Investigation of Multiphase Flow in a Packed Bed Reactor Under Microgravity Conditions

    Science.gov (United States)

    Lian, Yongsheng; Motil, Brian; Rame, Enrique

    2016-01-01

    In this paper we study the two-phase flow phenomena in a packed bed reactor using an integrated experimental and numerical method. The cylindrical bed is filled with uniformly sized spheres. In the experiment water and air are injected into the bed simultaneously. The pressure distribution along the bed will be measured. The numerical simulation is based on a two-phase flow solver which solves the Navier-Stokes equations on Cartesian grids. A novel coupled level set and moment of fluid method is used to construct the interface. A sequential method is used to position spheres in the cylinder. Preliminary experimental results showed that the tested flow rates resulted in pulse flow. The numerical simulation revealed that air bubbles could merge into larger bubbles and also could break up into smaller bubbles to pass through the pores in the bed. Preliminary results showed that flow passed through regions where the porosity is high. Comparison between the experimental and numerical results in terms of pressure distributions at different flow injection rates will be conducted. Comparison of flow phenomena under terrestrial gravity and microgravity will be made.

  19. New Configurations of Micro Plate-Fin Heat Sink to Reduce Coolant Pumping Power

    DEFF Research Database (Denmark)

    Kolaei, Alireza Rezania; Rosendahl, Lasse

    2012-01-01

    The thermal resistance of heat exchangers has a strong influence on the electric power produced by a thermoelectric generator (TEG). In this work, a real TEG device is applied to three configurations of micro plate-fin heat sink. The distance between certain microchannels is varied to find...... the optimum heat sink configuration. The particular focus of this study is to reduce the coolant mass flow rate by considering the thermal resistances of the heat sinks and, thereby, to reduce the coolant pumping power in the system. The threedimensional governing equations for the fluid flow and the heat...... heat sink configurations reduces the coolant pumping power in the system....

  20. Performance of Helical Coil Heat Recovery Exchanger using Nanofluid as Coolant

    Directory of Open Access Journals (Sweden)

    Navid Bozorgan

    2015-07-01

    Full Text Available Nanofluids are expected to be a promising coolant condidate in chemical processes for heat transfer system size reduction. This paper focuses on reducing the number of turns in a helical coil heat recovery exchanger with a given heat exchange capacity in a biomass heating plant using γ-Al2O3/n-decane nanofluid as coolant. The nanofluid flows through the tubes and the hot n-hexane flows through the shell. The numerical results show that using nanofluid as coolant in a helical coil heat exchanger can reduce the manufacturing cost of the heat exchanger and pumping power by reducing the number of turns of the coil.

  1. Hydrodynamic analysis of a three-fluidized bed reactor cold flow model for chemical looping hydrogen generation. Pressure characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Xue, Zhipeng; Xiang, Wenguo; Chen, Shiyi; Wang, Dong [Southeast Univ., Nanjing (China). School of Energy and Environment

    2013-07-01

    Chemical looping hydrogen generation (CLHG) can produce pure hydrogen with inherent separation of CO{sub 2} from fossils fuel. The process involves a metal oxide, as an oxygen carrier, such as iron oxide. The CLHG system consists of three reactors: a fuel reactor (FR), a steam reactor (SR) and an air reactor (AR). In the FR, the fuel gases react with iron oxides (hematite Fe{sub 2}O{sub 3}, magnetite Fe{sub 3}O{sub 4}, wuestite FeO), generating reduced iron oxides (FeO or even Fe), and with full conversion of gaseous fuels, pure CO{sub 2} can be obtained after cooling the flue gas from the fuel reactor; in the SR, FeO and Fe reacts with steam to generate magnetite (Fe{sub 3}O{sub 4}) and H{sub 2}, the latter representing the final target product of the process; in the AR, the magnetite is oxidized back to hematite which is used in another cycle. A cold flow model of three-fluidized bed for CLHG corresponding to 50 KW hot units has been built. A major novelty of this facility is the compact fuel reactor, which integrates a bubble and a fast fluidized bed to avoid the incomplete conversion of the fuel gas caused by the thermodynamics equilibrium. In order to study the pressure characteristics and the solids concentration of the system, especially in the fuel reactor, the gas velocity of three reactors, gas flow of L-type value, total solids inventory (TSI) and the secondary air of fuel reactor were varied. Results show that the pressure and the solids concentration are strongly influenced by the fluidizing-gas velocity of three reactors. Moreover, the entrainment of the upper part of fuel reactor increases as the total solids inventory increases, and the operating range of the FR can be changed by introducing secondary air or increasing the total solids inventory.

  2. Bioremoval of trivalent chromium using Bacillus biofilms through continuous flow reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sundar, K.; Sadiq, I. Mohammed; Mukherjee, Amitava [Centre for Nanobiotechnology, Nano Bio-Medicine Laboratory School of Bio Sciences and Technology VIT University, Vellore - 632014 (India); Chandrasekaran, N., E-mail: nchandrasekaran@vit.ac.in [Centre for Nanobiotechnology, Nano Bio-Medicine Laboratory School of Bio Sciences and Technology VIT University, Vellore - 632014 (India)

    2011-11-30

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

  3. Demonstrating electromagnetic control of free-surface, liquid-metal flows relevant to fusion reactors

    Science.gov (United States)

    Hvasta, M. G.; Kolemen, E.; Fisher, A. E.; Ji, H.

    2018-01-01

    Plasma-facing components (PFC’s) made from solid materials may not be able to withstand the large heat and particle fluxes that will be produced within next-generation fusion reactors. To address the shortcomings of solid PFC’s, a variety of liquid-metal (LM) PFC concepts have been proposed. Many of the suggested LM-PFC designs rely on electromagnetic restraint (Lorentz force) to keep free-surface, liquid-metal flows adhered to the interior surfaces of a fusion reactor. However, there is very little, if any, experimental data demonstrating that free-surface, LM-PFC’s can actually be electromagnetically controlled. Therefore, in this study, electrical currents were injected into a free-surface liquid-metal that was flowing through a uniform magnetic field. The resultant Lorentz force generated within the liquid-metal affected the velocity and depth of the flow in a controllable manner that closely matched theoretical predictions. These results show the promise of electromagnetic control for LM-PFC’s and suggest that electromagnetic control could be further developed to adjust liquid-metal nozzle output, prevent splashing within a tokamak, and alter heat transfer properties for a wide-range of liquid-metal systems.

  4. UV reactor flow visualization and mixing quantification using three-dimensional laser-induced fluorescence.

    Science.gov (United States)

    Gandhi, Varun; Roberts, Philip J W; Stoesser, Thorsten; Wright, Harold; Kim, Jae-Hong

    2011-07-01

    Three-dimensional laser-induced fluorescence (3DLIF) was applied to visualize and quantitatively analyze mixing in a lab-scale UV reactor consisting of one lamp sleeve placed perpendicular to flow. The recirculation zone and the von Karman vortex shedding that commonly occur in flows around bluff bodies were successfully visualized. Multiple flow paths were analyzed by injecting the dye at various heights with respect to the lamp sleeve. A major difference in these pathways was the amount of dye that traveled close to the sleeve, i.e., a zone of higher residence time and higher UV exposure. Paths away from the center height had higher velocities and hence minimal influence by the presence of sleeve. Approach length was also characterized in order to increase the probability of microbes entering the region around the UV lamp. The 3DLIF technique developed in this study is expected to provide new insight on UV dose delivery useful for the design and optimization of UV reactors. Copyright © 2011 Elsevier Ltd. All rights reserved.

  5. A novel reverse flow reactor coupling endothermic and exothermic reactions. Part I: comparison of reactor configurations for irreversible endothermic reactions

    NARCIS (Netherlands)

    van Sint Annaland, M.; Scholts, H.A.R.; Kuipers, J.A.M.; van Swaaij, Willibrordus Petrus Maria

    2002-01-01

    A new reactor concept is studied for highly endothermic heterogeneously catalysed gas phase reactions at high temperatures with rapid but reversible catalyst deactivation. The reactor concept aims to achieve an indirect coupling of energy necessary for endothermic reactions and energy released by

  6. EOIL power scaling in a 1-5 kW supersonic discharge-flow reactor

    Science.gov (United States)

    Davis, Steven J.; Lee, Seonkyung; Oakes, David B.; Haney, Julie; Magill, John C.; Paulsen, Dwane A.; Cataldi, Paul; Galbally-Kinney, Kristin L.; Vu, Danthu; Polex, Jan; Kessler, William J.; Rawlins, Wilson T.

    2008-02-01

    Scaling of EOIL systems to higher powers requires extension of electric discharge powers into the kW range and beyond with high efficiency and singlet oxygen yield. We have previously demonstrated a high-power microwave discharge approach capable of generating singlet oxygen yields of ~25% at ~50 torr pressure and 1 kW power. This paper describes the implementation of this method in a supersonic flow reactor designed for systematic investigations of the scaling of gain and lasing with power and flow conditions. The 2450 MHz microwave discharge, 1 to 5 kW, is confined near the flow axis by a swirl flow. The discharge effluent, containing active species including O II(a1Δ g, b1Σ g +), O( 3P), and O 3, passes through a 2-D flow duct equipped with a supersonic nozzle and cavity. I2 is injected upstream of the supersonic nozzle. The apparatus is water-cooled, and is modular to permit a variety of inlet, nozzle, and optical configurations. A comprehensive suite of optical emission and absorption diagnostics is used to monitor the absolute concentrations of O II(a), O II(b), O( 3P), O 3, I II, I(2P 3/2), I(2P 1/2), small-signal gain, and temperature in both the subsonic and supersonic flow streams. We discuss initial measurements of singlet oxygen and I* excitation kinetics at 1 kW power.

  7. Investigation of Abnormal Heat Transfer and Flow in a VHTR Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Kawaji, Masahiro [City College of New York, NY (United States); Valentin, Francisco I. [City College of New York, NY (United States); Artoun, Narbeh [City College of New York, NY (United States); Banerjee, Sanjoy [City College of New York, NY (United States); Sohal, Manohar [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schultz, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States); McEligot, Donald M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-12-21

    The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spots in the VHTR core.

  8. Modelingof Acetylene Pyrolysis under Steel Vacuum Carburizing Conditions in a Tubular Flow Reactor

    OpenAIRE

    Rainer Reimert; Frank Graf; Siegfried Bajohr; Rafi Ullah Khan

    2007-01-01

    In the present work, the pyrolysis of acetylene was studied under steel vacuumcarburizing conditions in a tubular flow reactor. The pyrolysis temperature ranged from650 °C to 1050 °C. The partial pressure of acetylene in the feed mixture was 10 and 20mbar, respectively, while the rest of the mixture consisted of nitrogen. The total pressureof the mixture was 1.6 bar. A kinetic mechanism which consists of seven species andnine reactions has been used in the commercial computational fluid...

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

    DEFF Research Database (Denmark)

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

    of the mixture was varied from oxidizing to reducing conditions. Moreover, a series of experiments in an oxygen atmosphere instead of a nitrogen atmosphere has been done. A reaction mechanism based on a recent work by Burke et al. has been developed. In addition to modeling of the present experiments......, ignition occurs at the temperature of 775–800 K. In general, the present model provides a good agreement with the measurements in the flow reactor and with recent data on laminar burning velocity and ignition delay time....

  10. 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.

  11. Flow synthesis using gaseous ammonia in a Teflon AF-2400 tube-in-tube reactor: Paal-Knorr pyrrole formation and gas concentration measurement by inline flow titration.

    Science.gov (United States)

    Cranwell, Philippa B; O'Brien, Matthew; Browne, Duncan L; Koos, Peter; Polyzos, Anastasios; Peña-López, Miguel; Ley, Steven V

    2012-08-14

    Using a simple and accessible Teflon AF-2400 based tube-in-tube reactor, a series of pyrroles were synthesised in flow using the Paal-Knorr reaction of 1,4-diketones with gaseous ammonia. An inline flow titration technique allowed measurement of the ammonia concentration and its relationship to residence time and temperature.

  12. Titer plate formatted continuous flow thermal reactors for high throughput applications: fabrication and testing

    Science.gov (United States)

    Sang-Won Park, Daniel; Chen, Pin-Chuan; You, Byoung Hee; Kim, Namwon; Park, Taehyun; Lee, Tae Yoon; Datta, Proyag; Desta, Yohannes; Soper, Steven A.; Nikitopoulos, Dimitris E.; Murphy, Michael C.

    2010-05-01

    A high throughput, multi-well (96) polymerase chain reaction (PCR) platform, based on a continuous flow (CF) mode of operation, was developed. Each CFPCR device was confined to a footprint of 8 × 8 mm2, matching the footprint of a well on a standard micro-titer plate. While several CFPCR devices have been demonstrated, this is the first example of a high-throughput multi-well continuous flow thermal reactor configuration. Verification of the feasibility of the multi-well CFPCR device was carried out at each stage of development from manufacturing to demonstrating sample amplification. The multi-well CFPCR devices were fabricated by micro-replication in polymers, polycarbonate to accommodate the peak temperatures during thermal cycling in this case, using double-sided hot embossing. One side of the substrate contained the thermal reactors and the opposite side was patterned with structures to enhance thermal isolation of the closely packed constant temperature zones. A 99 bp target from a λ-DNA template was successfully amplified in a prototype multi-well CFPCR device with a total reaction time as low as ~5 min at a flow velocity of 3 mm s-1 (15.3 s cycle-1) and a relatively low amplification efficiency compared to a bench-top thermal cycler for a 20-cycle device; reducing the flow velocity to 1 mm s-1 (46.2 s cycle-1) gave a seven-fold improvement in amplification efficiency. Amplification efficiencies increased at all flow velocities for 25-cycle devices with the same configuration.

  13. On-site evaluation of the performance of a full-scale down-flow hanging sponge reactor as a post-treatment process of an up-flow anaerobic sludge blanket reactor for treating sewage in India.

    Science.gov (United States)

    Okubo, Tsutomu; Onodera, Takashi; Uemura, Shigeki; Yamaguchi, Takashi; Ohashi, Akiyoshi; Harada, Hideki

    2015-10-01

    A down-flow hanging sponge (DHS) reactor is a novel, unaerated, aerobic, biofilm reactor that is used to polish effluent received from an up-flow anaerobic sludge blanket (UASB) reactor for treating municipal sewage. A full-scale DHS reactor was constructed for post-treatment of a full-scale UASB reactor at a municipal sewage treatment plant in India. Performance of the DHS reactor was evaluated with respect to organic removal over 1800 days of continuous operation. The UASB+DHS system consistently produced effluent with chemical oxygen demand (COD), biochemical oxygen demand (BOD), and suspended solids (SS) values of 37, 6.0 and 19 mg L(-1), on average, respectively. The sludge yield of the DHS reactor was estimated to be 0.04 kg SS kg(-1) COD removed or 0.12 kg SS kg(-1) BOD removed, which is considerably lower than other aerobic treatment methods that have been employed for polishing UASB effluent. Copyright © 2015 Elsevier Ltd. All rights reserved.

  14. Simple and rapid hydrogenation of p-nitrophenol with aqueous formic acid in catalytic flow reactors

    Directory of Open Access Journals (Sweden)

    Rahat Javaid

    2013-06-01

    Full Text Available The inner surface of a metallic tube (i.d. 0.5 mm was coated with a palladium (Pd-based thin metallic layer by flow electroless plating. Simultaneous plating of Pd and silver (Ag from their electroless-plating solution produced a mixed distributed bimetallic layer. Preferential acid leaching of Ag from the Pd–Ag layer produced a porous Pd surface. Hydrogenation of p-nitrophenol was examined in the presence of formic acid simply by passing the reaction solution through the catalytic tubular reactors. p-Aminophenol was the sole product of hydrogenation. No side reaction occurred. Reaction conversion with respect to p-nitrophenol was dependent on the catalyst layer type, the temperature, pH, amount of formic acid, and the residence time. A porous and oxidized Pd (PdO surface gave the best reaction conversion among the catalytic reactors examined. p-Nitrophenol was converted quantitatively to p-aminophenol within 15 s of residence time in the porous PdO reactor at 40 °C. Evolution of carbon dioxide (CO2 was observed during the reaction, although hydrogen (H2 was not found in the gas phase. Dehydrogenation of formic acid did not occur to any practical degree in the absence of p-nitrophenol. Consequently, the nitro group was reduced via hydrogen transfer from formic acid to p-nitrophenol and not by hydrogen generated by dehydrogenation of formic acid.

  15. Low-temperature anaerobic digestion of swine manure in a plug-flow reactor.

    Science.gov (United States)

    Massé, Daniel I; Gilbert, Yan; Saady, N M C; Liu, Charle

    2013-01-01

    A low-temperature (25 degrees C) anaerobic eight-compartment (PF01 to PF08) cascade reactor simulating a plug-flow reactor (PFR) treating pig manure was monitored for a year. The bioreactor was fed at an average loading rate of 2.4 +/- 0.2 g of total chemical oxygen demand (TCOD) per litre of reactor per day for a theoretical hydraulic retention time (HRT) of 67 +/- 7 d. An average of 79% of TCOD was removed from pig manure (converted into biogas and in sediments), whereas specific methane yields ranging from 397 to 482 NL CH4 kg(-1) VS (148.6 to 171.4 NL CH4 kg(-1) TCOD) were obtained. After 150 d, fluctuating performances of the process were observed, associated with solids accumulation in the upstream compartments, preventing the complete anaerobic digestion of swine manure in the compartments PF01 to PF04. Low-temperature anaerobic PFR represents an interesting alternative for the treatment of pig manure and recovery of green energy. Further investigations regarding a modified design, with better accumulating solids management, are needed to optimize the performance of this low-temperature PFR treating pig manure.

  16. A new concept of high flow rate non-thermal plasma reactor for air treatment

    Energy Technology Data Exchange (ETDEWEB)

    Goujard, V.; Tatibouet, J.M. [Univ. de Poitiers, Poitiers (France). Centre national de la recherche scientifique, Laboratoire de Catalyse en Chimie Organique

    2010-07-01

    Although several non-thermal plasma reactors have been tested for air treatment at the laboratory scale, up-scaling to pilot or industrial scale remains a challenge because several parameters must be considered, such as hydrodynamic behaviour, maximum voltage in an industrial environment, and maintenance of the system. This paper presented a newly developed reactor which consists to a DBD plasma generated on individual supports that could be directly inserted in gas pipes where air flow must be treated. Elimination of 40 percent of 15 ppm of propene was obtained with a energy density as low as 10 J/L. The propene conversion increased when a manganese oxide based catalyst was used because the ozone produced by the plasma was used as an as an oxidant. A simple model of the plasma-catalyst reactor behaviour showed that more than 90 percent of propene conversion can be expected for an input energy density of 10 J/L and residual ozone concentration less than 100 ppb.

  17. Study of the fluid flow pattern in a bubble column reactor for biodiesel production

    Science.gov (United States)

    Suhaimi, A. A.; Nasir, N. F.

    2017-09-01

    The applications of bubble column reactor (BCR) are very important as multiphase reactors in process industry. In biodiesel production, one of the reactors that are used is BCR. The advantages of BCR are low operating cost and maintenance due to the compactness and no moving parts. It is important to understand the nature of hydrodynamics and operational parameters to characterize their operation, including gas superficial velocity and bubble rise velocity to make the design and scale-up process. Computational fluid dynamics (CFD) can be used to evaluate the performance of BCR at lower cost compared to the experimental setup for biodiesel production. In this work, a commercial CFD software, FLUENT 14.0 was used for modeling of gas-liquid flow in a BCR for biodiesel production. Multiphase simulations were performed using an Eulerian-Eulerian two-fluid model. In this study, the simulation was conducted by using three different temperature, which are 523 K, 543 K and 563 K. The CFD result predicts the turbulent kinetic energy, gas hold-up and the liquid velocity were fairly good, although the results seem to suggest that further improvement on the interface exchange models and possibly further refine the two-fluid modeling approaches are necessary especially for the liquid velocity and turbulent kinetic energy.

  18. Development of LMR Coolant Technology - Development of a submersible-in-pool electromagnetic pump

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sang Hi; Kim, Hee Reyoung; Lee, Sang Don; Seo, Joon Ho [Seoul National University, Seoul (Korea, Republic of); Cho, Su Won [Kyoungki University, Suwon (Korea, Republic of)

    1997-07-15

    A submersible-in-pool type annular linear induction pumps of 60 l/min and 200 l/min, and 600 deg C has been designed with optimum geometrical and operating values found from MHD and circuit analyses reflecting the high-temperature characteristics of pump materials. Through the characteristics analyses inside the narrow flow channel of electromagnetic pump, the distribution of the time-varying flow field is calculated, and magnetic flux and force density are evaluated by end effects of linear induction electromagnetic pump and the instability analyses are carried out introducing one-dimensional linear perturbation. Testing the pump with the flow rate of 60 l/min in the suitably manufactured loop system shows a flow rate of 58 l/min at an input power of 1,377 VA with 60Hz. The design of a scaled-up pump is further taken into account LMR coolant system requiring increased capacity, and a basic analysis is carried out on the pump of 40,000 l/min for KALIMER. The present project contributes to the further design of engineering prototype electromagnetic pumps with higher capacity and to the development of liquid metal reactor with innovative simplicity. 89 refs., 8 tabs., 45 figs. (author)

  19. 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)

  20. Conducting water chemistry of the secondary coolant circuit of VVER-based nuclear power plant units constructed without using copper containing alloys

    Science.gov (United States)

    Tyapkov, V. F.

    2014-07-01

    The secondary coolant circuit water chemistry with metering amines began to be put in use in Russia in 2005, and all nuclear power plant units equipped with VVER-1000 reactors have been shifted to operate with this water chemistry for the past seven years. Owing to the use of water chemistry with metering amines, the amount of products from corrosion of structural materials entering into the volume of steam generators has been reduced, and the flow-accelerated corrosion rate of pipelines and equipment has been slowed down. The article presents data on conducting water chemistry in nuclear power plant units with VVER-1000 reactors for the secondary coolant system equipment made without using copper-containing alloys. Statistical data are presented on conducting ammonia-morpholine and ammonia-ethanolamine water chemistries in new-generation operating power units with VVER-1000 reactors with an increased level of pH. The values of cooling water leaks in turbine condensers the tube system of which is made of stainless steel or titanium alloy are given.

  1. Flow chemistry: intelligent processing of gas-liquid transformations using a tube-in-tube reactor.

    Science.gov (United States)

    Brzozowski, Martin; O'Brien, Matthew; Ley, Steven V; Polyzos, Anastasios

    2015-02-17

    reactive gas in a given reaction mixture. We have developed a tube-in-tube reactor device consisting of a pair of concentric capillaries in which pressurized gas permeates through an inner Teflon AF-2400 tube and reacts with dissolved substrate within a liquid phase that flows within a second gas impermeable tube. This Account examines our efforts toward the development of a simple, unified methodology for the processing of gaseous reagents in flow by way of development of a tube-in-tube reactor device and applications to key C-C, C-N, and C-O bond forming and hydrogenation reactions. We further describe the application to multistep reactions using solid-supported reagents and extend the technology to processes utilizing multiple gas reagents. A key feature of our work is the development of computer-aided imaging techniques to allow automated in-line monitoring of gas concentration and stoichiometry in real time. We anticipate that this Account will illustrate the convenience and benefits of membrane tube-in-tube reactor technology to improve and concomitantly broaden the scope of gas/liquid/solid reactions in organic synthesis.

  2. Techniques in Gas-Phase Thermolyses. Part 6. Pulse Pyrolysis: Gas Kinetic Studies in an Inductively Heated Flow Reactor

    DEFF Research Database (Denmark)

    Egsgaard, Helge; Bo, P.; Carlsen, Lars

    1985-01-01

    A prototype of an inductively heated flow reactor for gas kinetic studies is presented. The applicability of the system, which is based on a direct coupling between the reactor and the ion source of a mass spectrometer, is illustrated by investigations of a series of simple bond fission reactions....... The method permits the direct determination of low-pressure rate constants, the transformation to high-pressure values, and correspondingly evaluation of activation parameters, being derived by means of an empirical effective temperature approach.......A prototype of an inductively heated flow reactor for gas kinetic studies is presented. The applicability of the system, which is based on a direct coupling between the reactor and the ion source of a mass spectrometer, is illustrated by investigations of a series of simple bond fission reactions...

  3. System Study: High-Pressure Coolant Injection 1998-2014

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-12-01

    This report presents an unreliability evaluation of the high-pressure coolant injection system (HPCI) at 25 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCI results.

  4. DETERMINATION OF THE 129I IN PRIMARY COOLANT OF PWR

    Directory of Open Access Journals (Sweden)

    KE CHON CHOI

    2013-02-01

    In this report, the effect of the boron content in a pressurized-water reactor primary coolant on the separation process of 129I was examined, as was the effect of 3H on the measurement of the activity of iodine. As a result, no influence of the boron content and of the simultaneous 3H presence was found with activity concentrations of 3H lower than 50 Bq/mL, and with a boron concentration of less than 2,000 μg/mL.

  5. System Study: High-Pressure Coolant Injection 1998–2013

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-01-31

    This report presents an unreliability evaluation of the high-pressure coolant injection system (HPCI) at 25 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCI results.

  6. Experimental and Analytic Study on the Core Bypass Flow in a Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard Schultz

    2012-04-01

    Core bypass flow has been one of key issues in the very high temperature reactor (VHTR) design for securing core thermal margins and achieving target temperatures at the core exit. The bypass flow in a prismatic VHTR core occurs through the control element holes and the radial and axial gaps between the graphite blocks for manufacturing and refueling tolerances. These gaps vary with the core life cycles because of the irradiation swelling/shrinkage characteristic of the graphite blocks such as fuel and reflector blocks, which are main components of a core's structure. Thus, the core bypass flow occurs in a complicated multidimensional way. The accurate prediction of this bypass flow and counter-measures to minimize it are thus of major importance in assuring core thermal margins and securing higher core efficiency. Even with this importance, there has not been much effort in quantifying and accurately modeling the effect of the core bypass flow. The main objectives of this project were to generate experimental data for validating the software to be used to calculate the bypass flow in a prismatic VHTR core, validate thermofluid analysis tools and their model improvements, and identify and assess measures for reducing the bypass flow. To achieve these objectives, tasks were defined to (1) design and construct experiments to generate validation data for software analysis tools, (2) determine the experimental conditions and define the measurement requirements and techniques, (3) generate and analyze the experimental data, (4) validate and improve the thermofluid analysis tools, and (5) identify measures to control the bypass flow and assess its performance in the experiment.

  7. Using reactive tracers to detect flow field anomalies in water treatment reactors.

    Science.gov (United States)

    Gresch, Markus; Braun, Daniel; Gujer, Willi

    2011-02-01

    The hydraulics of water and wastewater treatment reactors has a major impact on their performance and control. The residence time distribution as a measure for the hydraulics represents macroscopic mixing in an integrated way with no spatial information. However, with regard to optimal sensor location for process control and for process optimisation measures, spatial information about macro-mixing is helpful. Spatially distributed measurements of reactive tracers can provide this information. In this paper we generally discuss how reactive tracers can be used to detect and characterize distinct large scale flow structures. It is shown that tracer substances are particularly suited if their reaction time scale is similar to the time scale of the large scale flow structure. For nitrifying activated sludge systems, ammonium is identified to be a suitable tracer. In a comprehensive experimental study at a real aeration tank, two distinct large scale flow features were identified by distributed ammonium measurements. Flow velocity measurements using acoustic Doppler velocimetry clearly supported the nature of these flow field anomalies. Ion-selective electrodes are a well suited device for ammonium measurements providing the temporal resolution that is needed for such an analysis. Copyright © 2010 Elsevier Ltd. All rights reserved.

  8. A 1D plug flow reactor as validation tool for reactive transport simulations

    Science.gov (United States)

    Battaïa, G.; Garcia, D.

    2012-04-01

    Predictions in CO2 geological sequestration involve a broad range of earth sciences linked in complex models. Amongst the processes commonly described, fluid-rock interactions are both a central issue and a source of discomfort for modelers since it has to deal with 1) kinetics data obtained through experimental procedures that dramatically differ from natural systems and 2) reactive surface model that are very diverse and often empirical. This study presents a new type of plug flow reactor developed to provide an experimental validation of reactive transport simulations. This is a 1D pressurized packed-bed plug-flow reactor containing a granular mixture as a porous medium. This mixture is composed of a reactive solids and unreactive quartz used to set an adequate ratio between fluid and reactive mineral to control the front velocity. A seven sampling valve unit allows concentration profiles of the reacting fluid to be captured at any time. One the one side, a low reaction rate (diopside, HNO3, pH 2) produces linear profile resulting from a constant dissolution rate along the reactor length. But on the other side, when performing the reaction of CO2 saturated solutions (5 bar) at 40°C with dolomite it gives rise to dissolution fronts migrating downstream. A proper projection of experimental data reveals a dynamic steady state of front shape is reached. Texture of the mineral recovered at the end of the experiment is quantified by Hg-porosimetry and these results are linked to SEM observations. Altogether, this provides a robust way for the parameterization of a reactive surface area model.

  9. [Continuous dry fermentation of pig manure using up plug-flow type anaerobic reactor].

    Science.gov (United States)

    Chen, Chuang; Deng, Liang-Wei; Xin, Xin; Zheng, Dan; Liu, Yi; Kong, Chui-Xue

    2012-03-01

    To solve the problems of ammonia inhibition and discharging difficulty in continuous dry fermentation of pig manure, under the experimental conditions of temperature of (25 +/- 2) degrees C and organic loading rate (TS) of 4.44 g x (L x d) (-1), a lab-scale up plug-flow type anaerobic reactor (UPAR) was setup to investigate biogas production, ammonia inhibition, effluent liquidity, and the feasibility of continuous dry fermentation of pig manure using up plug-flow type anaerobic reactor. The experiment was operated for 160 days using the pig manure with four different TS mass fractions (20%, 25%, 30%, 35%) as feeding. Results showed that the feeding TS mass fraction exerted a significant influence on the dry fermentation of pig manure; the stable volumetric biogas production rates of four different feeding TS mass fractions were 2.40, 1.73, 0.89, and 0.62 L x (L x d)(-1), respectively; the biogas producing efficiencies of the reactors with feeding TS mass fractions of 20%, 25% and 30% were obviously superior to that with feeding TS of 35%. With feeding TS mass fraction increased from 20% to 35%, obvious inhibition to biogas producing occurred when concentration of ammonia nitrogen reached more than 2 300 mg x L(-1). When the feeding TS mass fraction was 35%, the concentration of ammonia nitrogen could accumulate to 3 800 mg x L(-1) but biogas production rate decreased 74.1% of that with feeding TS of 20%. Additionally, while the feeding TS mass fraction was 35%, the effluent TS mass fraction achieved 17.1%, and the velocity of effluent was less than 0.002 m x s(-1) the effluent of UPAR could not be smoothly discharged.

  10. Continuous Polyol Synthesis of Metal and Metal Oxide Nanoparticles Using a Segmented Flow Tubular Reactor (SFTR).

    Science.gov (United States)

    Testino, Andrea; Pilger, Frank; Lucchini, Mattia Alberto; Quinsaat, Jose Enrico Q; Stähli, Christoph; Bowen, Paul

    2015-06-08

    Over the last years a new type of tubular plug flow reactor, the segmented flow tubular reactor (SFTR), has proven its versatility and robustness through the water-based synthesis of precipitates as varied as CaCO3, BaTiO3, Mn(1-x)NixC2O4·2H2O, YBa oxalates, copper oxalate, ZnS, ZnO, iron oxides, and TiO2 produced with a high powder quality (phase composition, particle size, and shape) and high reproducibility. The SFTR has been developed to overcome the classical problems of powder production scale-up from batch processes, which are mainly linked with mass and heat transfer. Recently, the SFTR concept has been further developed and applied for the synthesis of metals, metal oxides, and salts in form of nano- or micro-particles in organic solvents. This has been done by increasing the working temperature and modifying the particle carrying solvent. In this paper we summarize the experimental results for four materials prepared according to the polyol synthesis route combined with the SFTR. CeO2, Ni, Ag, and Ca3(PO4)2 nanoparticles (NPs) can be obtained with a production rate of about 1-10 g per h. The production was carried out for several hours with constant product quality. These findings further corroborate the reliability and versatility of the SFTR for high throughput powder production.

  11. A Reactor Development Scenario for the FuZE Sheared-Flow Stabilized Z-pinch

    Science.gov (United States)

    McLean, Harry S.; Higginson, D. P.; Schmidt, A.; Tummel, K. K.; Shumlak, U.; Nelson, B. A.; Claveau, E. L.; Forbes, E. G.; Golingo, R. P.; Stepanov, A. D.; Weber, T. R.; Zhang, Y.

    2017-10-01

    We present a conceptual design, scaling calculations, and development path for a pulsed fusion reactor based on a flow-stabilized Z-pinch. Experiments performed on the ZaP and ZaP-HD devices have largely demonstrated the basic physics of sheared-flow stabilization at pinch currents up to 100 kA. Initial experiments on the FuZE device, a high-power upgrade of ZaP, have achieved 20 usec of stability at pinch current 100-200 kA and pinch diameter few mm for a pinch length of 50 cm. Scaling calculations based on a quasi-steady-state power balance show that extending stable duration to 100 usec at a pinch current of 1.5 MA and pinch length of 50 cm, results in a reactor plant Q 5. Future performance milestones are proposed for pinch currents of: 300 kA, where Te and Ti are calculated to exceed 1-2 keV; 700 kA, where DT fusion power would be expected to exceed pinch input power; and 1 MA, where fusion energy per pulse exceeds input energy per pulse. This work funded by USDOE ARPA-E and performed under the auspices of Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344. LLNL-ABS-734770.

  12. A mixed flow reactor method to synthesize amorphous calcium carbonate under controlled chemical conditions.

    Science.gov (United States)

    Blue, Christina R; Rimstidt, J Donald; Dove, Patricia M

    2013-01-01

    This study describes a new procedure to synthesize amorphous calcium carbonate (ACC) from well-characterized solutions that maintain a constant supersaturation. The method uses a mixed flow reactor to prepare ACC in significant quantities with consistent compositions. The experimental design utilizes a high-precision solution pump that enables the reactant solution to continuously flow through the reactor under constant mixing and allows the precipitation of ACC to reach steady state. As a proof of concept, we produced ACC with controlled Mg contents by regulating the Mg/Ca ratio of the input solution and the carbonate concentration and pH. Our findings show that the Mg/Ca ratio of the reactant solution is the primary control for the Mg content in ACC, as shown in previous studies, but ACC composition is further regulated by the carbonate concentration and pH of the reactant solution. The method offers promise for quantitative studies of ACC composition and properties and for investigating the role of this phase as a reactive precursor to biogenic minerals. © 2013 Elsevier Inc. All rights reserved.

  13. Continuous Polyol Synthesis of Metal and Metal Oxide Nanoparticles Using a Segmented Flow Tubular Reactor (SFTR

    Directory of Open Access Journals (Sweden)

    Andrea Testino

    2015-06-01

    Full Text Available Over the last years a new type of tubular plug flow reactor, the segmented flow tubular reactor (SFTR, has proven its versatility and robustness through the water-based synthesis of precipitates as varied as CaCO3, BaTiO3, Mn(1−xNixC2O4·2H2O, YBa oxalates, copper oxalate, ZnS, ZnO, iron oxides, and TiO2 produced with a high powder quality (phase composition, particle size, and shape and high reproducibility. The SFTR has been developed to overcome the classical problems of powder production scale-up from batch processes, which are mainly linked with mass and heat transfer. Recently, the SFTR concept has been further developed and applied for the synthesis of metals, metal oxides, and salts in form of nano- or micro-particles in organic solvents. This has been done by increasing the working temperature and modifying the particle carrying solvent. In this paper we summarize the experimental results for four materials prepared according to the polyol synthesis route combined with the SFTR. CeO2, Ni, Ag, and Ca3(PO42 nanoparticles (NPs can be obtained with a production rate of about 1–10 g per h. The production was carried out for several hours with constant product quality. These findings further corroborate the reliability and versatility of the SFTR for high throughput powder production.

  14. METHOD AND APPARATUS FOR CONTROL OF A NUCLEAR REACTOR

    Science.gov (United States)

    Cawley, W.E.

    1962-12-11

    A method and apparatus are described for controlling an overmoderated nuclear reactor containing columns of fuel elements aligned in a plurality of coolant tubes in a stream of coolant water. The invention includes means for adjusting the distance between halves of the fuel element column to vary the relative proportion of fuel and moderator at the center of the reactor. (AEC)

  15. VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS

    Science.gov (United States)

    Furgerson, W.T.

    1963-12-17

    A hollow, porous-walled fuel element filled with fissionable fuel and provided with an outlet port through its wall is described. In operation in a gas-cooled reactor, the element is connected, through its outlet port, to the vacuum side of a pump that causes a portion of the coolant gas flowing over the exterior surface of the element to be drawn through the porous walls thereof and out through the outlet port. This continuous purging gas flow sweeps away gaseous fission products as they are released by the fissioning fuel. (AEC) A fuel element for a nuclear reactor incorporating a body of metal of melting point lower than the temperature of operation of the reactor and a nuclear fuel in finely divided form dispersed in the body of metal as a settled slurry is presented. (AEC)

  16. Coupled neutronic-thermal-hydraulics analysis in a coolant subchannel of a PWR using CFD techniques

    Energy Technology Data Exchange (ETDEWEB)

    Ribeiro, Felipe P.; Su, Jian, E-mail: sujian@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2017-07-01

    The high capacity of Computational Fluid Dynamics code to predict multi-dimensional thermal-hydraulics behaviour and the increased availability of capable computer systems are making that method a good tool to simulate phenomena of thermal-hydraulics nature in nuclear reactors. However, since there are no neutron kinetics models available in commercial CFD codes to the present day, the application of CFD in the nuclear reactor safety analyses is still limited. The present work proposes the implementation of the point kinetics model (PKM) in ANSYS - Fluent to predict the neutronic behaviour in a Westinghouse Sequoyah nuclear reactor, coupling with the phenomena of heat conduction in the rod and thermal-hydraulics in the cooling fluid, via the reactivity feedback. Firstly, a mesh convergence and turbulence model study was performed, using the Reynolds-Average Navier-Stokes method, with square arrayed rod bundle featuring pitch to diameter ratio of 1:32. Secondly, simulations using the k-! SST turbulence model were performed with an axial distribution of the power generation in the fuel to analyse the heat transfer through the gap and cladding, and its in fluence on the thermal-hydraulics behaviour of the cooling fluid. The wall shear stress distribution for the centre-line rods and the dimensionless velocity were evaluated to validate the model, as well as the in fluence of the mass flow rate variation on the friction factor. The coupled model enabled to perform a dynamic analysis of the nuclear reactor during events of insertion of reactivity and shutdown of primary coolant pumps. (author)

  17. Crack stability analysis of low alloy steel primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T.; Kameyama, M. [Kansai Electric Power Company, Osaka (Japan); Urabe, Y. [Mitsubishi Heavy Industries, Ltd., Takasago (Japan)] [and others

    1997-04-01

    At present, cast duplex stainless steel has been used for the primary coolant piping of PWRs in Japan and joints of dissimilar material have been applied for welding to reactor vessels and steam generators. For the primary coolant piping of the next APWR plants, application of low alloy steel that results in designing main loops with the same material is being studied. It means that there is no need to weld low alloy steel with stainless steel and that makes it possible to reduce the welding length. Attenuation of Ultra Sonic Wave Intensity is lower for low alloy steel than for stainless steel and they have advantageous inspection characteristics. In addition to that, the thermal expansion rate is smaller for low alloy steel than for stainless steel. In consideration of the above features of low alloy steel, the overall reliability of primary coolant piping is expected to be improved. Therefore, for the evaluation of crack stability of low alloy steel piping to be applied for primary loops, elastic-plastic future mechanics analysis was performed by means of a three-dimensioned FEM. The evaluation results for the low alloy steel pipings show that cracks will not grow into unstable fractures under maximum design load conditions, even when such a circumferential crack is assumed to be 6 times the size of the wall thickness.

  18. Modification of reference temperature program in reactor regulating system

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Sung Sik; Lee, Byung Jin; Kim, Se Chang; Cheong, Jong Sik [Korea Power Engineering Company, Inc., Seoul (Korea, Republic of); Kim, Ji In; Doo, Jin Yong [Korea Electric Power Cooperation, Yonggwang (Korea, Republic of)

    1998-12-31

    In Yonggwang nuclear units 3 and 4 currently under commercial operation, the cold temperature was very close to the technical specification limit of 298 deg C during initial startup testing, which was caused by the higher-than-expected reactor coolant system flow. Accordingly, the reference temperature (Tref) program needed to be revised to allow more flexibility for plant operations. In this study, the method of a specific test performed at Yonggwang nuclear unit 4 to revise the Tref program was described and the test results were discussed. In addition, the modified Tref program was evaluated on its potential impacts on system performance and safety. The methods of changing the Tref program and the associated pressurizer level setpoint program were also explained. Finally, for Ulchin nuclear unit 3 and 4 currently under initial startup testing, the effects of reactor coolant system flow rate on the coolant temperature were evaluated from the thermal hydraulic standpoint and an optimum Tref program was recommended. 6 refs., 4 figs., 2 tabs. (Author)

  19. Light water reactor fuel response during reactivity initiated accident experiments

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, P. E.; McCardell, R. K.; Martinson, Z. R.; Seiffert, S. L.

    1979-01-01

    Experimental results from six recent Power Burst Facility (PBF) reactivity initiated accident (RIA) tests are compared with data from previous Special Power Excursion Reactor Test (SPERT), and Japanese Nuclear Safety Research Reactor (NSRR) tests. The RIA fuel behavior experimental program recently started in the PBF is being conducted with coolant conditions typical of hot-startup conditions in a commercial boiling water reactor. The SPERT and NSRR test programs investigated the behavior of single or small clusters of light water reactor (LWR) type fuel rods under approximate room temperature and atmospheric pressure conditions in capsules containing stagnant water. As observed in the SPERT and NSRR tests, energy deposition, and consequent enthalpy increase in the PBF test fuel, appears to be the single most important variable. However, the consequences of failure at boiling water hot-startup system conditions appear to be more severe than previously observed in either the stagnant capsule SPERT or NSRR tests. Metallographic examination of both previously unirradiated and irradiated PBF fuel rod cross sections revealed extensive variation in cladding wall thicknesses (involving considerable plastic flow) and fuel shattering along grain boundaries in both restructured and unrestructured fuel regions. Oxidation of the cladding resulted in fracture at the location of cladding thinning and disintegration of the rods during quench. In addition,swelling of the gaseous and potentially volatile fission products in previously irradiated fuel resulted in volume increases of up to 180% and blockage of the coolant channels within the flow shrouds surrounding the fuel rods.

  20. Electrochemistry of Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

  1. Coolant Compatibility Studies for Fusion and Fusion-Fission Hybrid Reactor Concepts: Corrosion of Oxide Dispersion Strengthened Iron-Chromium Steels and Tantalum in High Temperature Molten Fluoride Salts

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, Joseph [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); El-dasher, Bassem [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ferreira, James [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Caro, Magdalena Serrano de [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kimura, Akihiko [Kyoto Univ. (Japan). Inst. of Advanced Energy

    2010-05-04

    Alloys such as 12YWT & 14YWT have exceptional high-temperature strength at temperatures greater than 550 C. This class of materials has also demonstrated relatively little radiation induced swelling at damage levels of at least 75 dpa in sodium-cooled fast reactors. However, corrosion of oxide dispersion strengthened (ODS) steels in high temperature molten fluoride salts may limit the life of advanced reactor systems, including some fusion and fusionfission hybrid systems that are now under consideration. This paper reports corrosion studies of ODS steel in molten fluoride salts at temperatures ranging from 600 to 900 C. Electrochemical impedance spectroscopy (EIS) was used to measure the temperature dependence of charge transfer kinetics in situ, while an environmental electron microscope (ESEM) equipped with energy dispersive spectroscopy (EDS) was used for postexposure examination of test samples. ODS steel experienced corrosion in the molten fluoride salts at 550 to 900 C, even in carefully controlled glove-box environments with very low levels of oxygen and moisture. The observed rate of attack was found to accelerate dramatically at temperatures above 800 C. Tantalum and tantalum-based alloys such as Ta-1W and Ta-10W have exceptional high temperature strength, far better than ODS steels. Unlike ODS steels, tantalum has been found to exhibit some immunity to corrosive attack by molten fluoride salts at temperatures as high as 900 C, though there is some indication that grain boundary attack may have occurred. Unfortunately, tantalum alloys are known to become brittle during irradiation and exposure to hydrogen, both of which are important in fusion applications.

  2. Simulations of flow behavior of fuel particles in a conceptual helium-cooled spout fluidized bed nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wang Shuyan; Li Xiang [School of Energy Science and Engineering, Harbin Institute of Technology, Harbin 150001 (China); Lu Huilin [School of Energy Science and Engineering, Harbin Institute of Technology, Harbin 150001 (China)], E-mail: huilin@hit.edu.cn; Bouillard, Jacques [INERIS, Parc Technologique Alata, BP2, Verneuil-en-Halatte 60550 (France); Sun Qiaoqun; Wang Shuai [School of Energy Science and Engineering, Harbin Institute of Technology, Harbin 150001 (China)

    2009-01-15

    Hydrodynamics of helium and fuel particles are simulated in a conceptual helium-cooled spout fluidized bed nuclear reactor. The conceptual reactor consists of an axis-symmetric column with a sharp cone inside which the fuel particles are fluidized by helium. An isothermal gas-solid two-fluid flow model is presented. The kinetic-frictional constitutive model for dense assemblies of solids is incorporated. The kinetic stress is modeled using the kinetic theory of granular flow, while the friction stress is from the normal frictional stress model proposed by (Johnson, P.C., Nott, P., Jackson, R., 1990. Frictional-collisional equations of motion for particulate flows and their application to chutes. Journal of Fluid Mechanics 210, 501-535). Detailed spatial/temporal concentration and velocity profiles have been obtained in a conceptual spout fluidized bed nuclear reactor. The influence of inlet spouting jet velocity and conical angles on flow behavior of fluid and fuel particles is analyzed. The numerical simulations show that the unique mixing ability of the spout fluidized bed nuclear reactor gives rise, as expected, to uniform particle distributions. This uniformity enhances the heat transfer and therefore the power produced by the reactor.

  3. Analysis of small break loss of coolant accident for Chinese CPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ju Youl [FNC Technology Co., Yongin (Korea, Republic of); Cilier, Anthonie [North-West University, Mahikeng (South Africa); Poc, Li-chi Cliff [Micro-Simulation Technology, Montville (United States)

    2016-05-15

    This research analyses the small break loss of coolant accident (LOCA) on a Chinese CPR1000 type reactor. LOCA accident is used as benchmark for the PCTRAN/CPR1000 code by comparing the effects and results to the Manshaan FSAR accident analysis. LOCA is a design basis accident in which a guillotine break is postulated to occur in one of the cold legs of a pressurized water reactor (PWR). Consequently, the primary system pressure would drop and almost all the reactor coolant would be discharged into the reactor containment. The drop in pressure would activate the reactor protection system and the reactor would trip. The simulation of a 3-inch small break loss of coolant accident using the PCTRAN/CPR1000 has revealed this code's effectiveness as well as weaknesses in specific simulation applications. The code has the ability to run at 16 times real time and produce very accurate results. The results are consistently producing the same trends as licensed codes used in Safety Assessment Reports. It is however able to produce these results in a fraction of the time and also provides a whole plant simulation coupling the various thermal, hydraulic, chemical and neutronic systems together with a plant specific control system.

  4. Heat-transfer characteristics of flowing and stationary particle-bed-type fusion-reactor blankets

    Energy Technology Data Exchange (ETDEWEB)

    Nietert, R.E.

    1983-02-01

    The heat-transfer characteristics of flowing and stationary packed-particle beds have recently become of interest in connection with conceptual designs of fusion reactor blankets. A detailed literature survey has shown that the processes taking place in such beds are not fully understood despite their widespread use in the chemical industry and other engineering disciplines for more than five decades. In this study, two experimental investigations were pursued. In the first, a heat-transfer loop was constructed through which glass microspheres were allowed to flow by rgravity at controlled rates through an electrically heated stainless steel tubular test section. In the second, an annular packed bed was constructed in which heat was applied through the outer wall by electric heating of a stainless steel tube. Cooling occurred at the inner wall of the annular bed by flowing air through the central tube. A second air stream was allowed to flow through the voids of the packed bed. An error-minimization technique was utilized in order to obtain the two-dimensional one-parameter effective conductivity for the bed by comparing the experimental and theoretically predicted temperature profiles. Experiments were conducted for various modified Reynolds numbers less than ten.

  5. Serrated Flow Behavior of Aisi 316l Austenitic Stainless Steel for Nuclear Reactors

    Science.gov (United States)

    Li, Qingshan; Shen, Yinzhong; Han, Pengcheng

    2017-10-01

    AISI 316L austenitic stainless steel is a candidate material for Generation IV reactors. In order to investigate the influence of temperature on serrated flow behavior, tensile tests were performed at temperatures ranging from 300 to 700 °C at an initial strain rate of 2×10-4 s-1. Another group of tensile tests were carried out at strain rates ranging from 1×10-4 to 1×10-2 s-1 at 600 °C to examine the influence of strain rates on serrated flow behavior. The steel exhibited serrated flow, suggesting the occurrence of dynamic strain ageing at 450-650°C. No plateau of yield stresses of the steel was observed at an initial strain rate of 2×10-4 s-1. The effective activation energy for serrated flow occurrence was calculated to be about 254.72 kJ/mol-1. Cr, Mn, Ni and Mo solute atoms are expected to be responsible for dynamic strain ageing at high temperatures of 450-650 °C in the steel.

  6. Modeling of Heat Transfer in the Helical-Coil Heat Exchanger for the Reactor Facility "UNITERM"

    Directory of Open Access Journals (Sweden)

    V. I. Solonin

    2014-01-01

    Full Text Available Circuit heat sink plays an important role in the reactor system. Therefore it imposes high requirements for quality of determining thermal-hydraulic parameters. This article is aimed at modeling of heat exchange process of the helical-coil heat exchanger, which is part of the heat sink circuit of the reactor facility "UNITERM."The simulation was performed using hydro-gas-dynamic software package ANSYS CFX. Computational fluid dynamics of this package allows us to perform calculations in a threedimensional setting, giving an idea of the fluid flow nature. The purpose of the simulation was to determine the parameters of the helical-coil heat exchanger (temperature, velocity at the outlet of the pipe and inter-tubular space, pressure drop, and the nature of the fluid flow of primary and intermediate coolants. Geometric parameters of the model were determined using the preliminary calculations performed by the criterion equations. In calculations Turbulence models k-ε RNG, Shear Stress Transport (SST are used. The article describes selected turbulence models, and considers relationship with wall function.The calculation results allow us to give the values obtained for thermal-hydraulic parameters, to compare selected turbulence models, as well as to show distribution patterns of the coolant temperature, pressure, and velocity at the outlet of the intermediate cooler.Calculations have shown that:- maximum values of primary coolant temperature at the outlet of the heat exchanger surface are encountered in the space between the helical-coil tubes;- higher temperatures of intermediate coolant at the outlet of the coils (in space of helicalcoil tubes are observed for the peripheral row;- primary coolant movement in the inter-tubular space of helical-coil surface is formed as a spiral flow, rather than as a in-line tube bank cross flow.

  7. Azo dye removal in a membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cui, Dan; Guo, Yu-Qi; Cheng, Hao-Yi; Liang, Bin; Kong, Fan-Ying [State Key Laboratory of Urban Water Resource and Environment, Harbin Institute of Technology, No. 202 Haihe Road, Harbin 150090 (China); Lee, Hyung-Sool [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West Waterloo, Ontario, Canada N2L 3G1 (Canada); Wang, Ai-Jie, E-mail: waj0578@hit.edu.cn [State Key Laboratory of Urban Water Resource and Environment, Harbin Institute of Technology, No. 202 Haihe Road, Harbin 150090 (China)

    2012-11-15

    Highlights: Black-Right-Pointing-Pointer A membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor was developed. Black-Right-Pointing-Pointer Alizarin Yellow R as the mode of azo dyes was efficiently converted to p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA). Black-Right-Pointing-Pointer PPD and 5-ASA were further oxidized in a bio-contact oxidation reactor. Black-Right-Pointing-Pointer The mechanism of UBER for azo dye removal was discussed. - Abstract: Azo dyes that consist of a large quantity of dye wastewater are toxic and persistent to biodegradation, while they should be removed before being discharged to water body. In this study, Alizarin Yellow R (AYR) as a model azo dye was decolorized in a combined bio-system of membrane-free, continuous up-flow bio-catalyzed electrolysis reactor (UBER) and subsequent aerobic bio-contact oxidation reactor (ABOR). With the supply of external power source 0.5 V in the UBER, AYR decolorization efficiency increased up to 94.8 {+-} 1.5%. Products formation efficiencies of p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA) were above 90% and 60%, respectively. Electron recovery efficiency based on AYR removal in cathode zone was nearly 100% at HRTs longer than 6 h. Relatively high concentration of AYR accumulated at higher AYR loading rates (>780 g m{sup -3} d{sup -1}) likely inhibited acetate oxidation of anode-respiring bacteria on the anode, which decreased current density in the UBER; optimal AYR loading rate for the UBER was 680 g m{sup -3} d{sup -1} (HRT 2.5 h). The subsequent ABOR further improved effluent quality. Overall the Chroma decreased from 320 times to 80 times in the combined bio-system to meet the textile wastewater discharge standard II in China.

  8. Operation of a full-scale pumped flow biofilm reactor (PFBR) under two aeration regimes.

    Science.gov (United States)

    O'Reilly, E; Rodgers, M; Clifford, E

    2011-01-01

    A novel technology suitable for centralised and decentralised wastewater treatment has been developed, extensively tested at laboratory-scale, and trialled at a number of sites for populations ranging from 15 to 400 population equivalents (PE). The two-reactor-tank pumped flow biofilm reactor (PFBR) is characterised by: (i) its simple construction; (ii) its ease of operation and maintenance; (iii) low operating costs; (iv) low sludge production; and (v) comprising no moving parts or compressors, other than hydraulic pumps. By operating the system in a sequencing batch biofilm reactor (SBBR) mode, the following treatment can be achieved: 5-day biochemical oxygen demand (BOD5), chemical oxygen demand (COD) and total suspended solids (TSS) reduction; nitrification and denitrification. During a 100-day full-scale plant study treating municipal wastewater and operating at 165 PE and 200 PE (Experiments 1 and 2, respectively), maximum average removals of 94% BOD5, 86% TSS and 80% ammonium-nitrogen (NH4-N) were achieved. During the latter part of Experiment 2, effluent concentrations averaged: 14 mg BOD5/l; 32 mg COD(filtered)/l; 14 mg TSS/l; 4.4 mg NH4-N/l; and 4.0 mg NO3-N/l (nitrate-nitrogen). The average energy consumption was 0.46-0.63 kWh/m3(treated) or 1.25-1.76 kWh/kg BOD5 removed. No maintenance was required during these experiments. The PFBR technology offers a low energy, minimal maintenance technology for the treatment of municipal wastewater.

  9. Hydrogen production enhancement and the effect of passive mixing using flow disturbers in a steam-reforming reactor

    Science.gov (United States)

    Liao, Chang-Hsien

    This study investigates the influence of changing the flow pathway inside a methanol steam reformer by introduction of flow disturbers in the flow field. In a reforming reaction, it is known that fuel conversion from hydrocarbons to hydrogen can be limited by chemical kinetics, which is a function of local temperature. For a typical cylindrical reactor, large thermal gradients inside the packed bed result from insufficient heat and mass transfer. This causes a non-ideal condition for complete conversion to hydrogen. Active mixing methods in critical fluid pathways have been proven to improve heat and mass transfer inside reforming reactors. A new method of passive mixing in the fluid pathways by introducing flow disturbers inside the packed catalyst bed is presented. A principle of characteristic time of a 1st order reaction is also presented and studied. The reactor output parameters of fuel conversion, temperature profile, characteristic time, pressure drop, reactor efficiency, and power demand are analyzed and compared to quantify the influence of the passive mixing technique. Input variables in this study are packing density of flow disturbers, space velocity and catalyst size. This study is expected to provide a basic analysis and contribute to the improvement of reformer design for better fuel processing system performance.

  10. Development of a Reduced-Order Three-Dimensional Flow Model for Thermal Mixing and Stratification Simulation during Reactor Transients

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Rui

    2017-09-03

    Mixing, thermal-stratification, and mass transport phenomena in large pools or enclosures play major roles for the safety of reactor systems. Depending on the fidelity requirement and computational resources, various modeling methods, from the 0-D perfect mixing model to 3-D Computational Fluid Dynamics (CFD) models, are available. Each is associated with its own advantages and shortcomings. It is very desirable to develop an advanced and efficient thermal mixing and stratification modeling capability embedded in a modern system analysis code to improve the accuracy of reactor safety analyses and to reduce modeling uncertainties. An advanced system analysis tool, SAM, is being developed at Argonne National Laboratory for advanced non-LWR reactor safety analysis. While SAM is being developed as a system-level modeling and simulation tool, a reduced-order three-dimensional module is under development to model the multi-dimensional flow and thermal mixing and stratification in large enclosures of reactor systems. This paper provides an overview of the three-dimensional finite element flow model in SAM, including the governing equations, stabilization scheme, and solution methods. Additionally, several verification and validation tests are presented, including lid-driven cavity flow, natural convection inside a cavity, laminar flow in a channel of parallel plates. Based on the comparisons with the analytical solutions and experimental results, it is demonstrated that the developed 3-D fluid model can perform very well for a wide range of flow problems.

  11. A novel ion transport membrane reactor for fundamental investigations of oxygen permeation and oxy-combustion under reactive flow conditions

    KAUST Repository

    Kirchen, Patrick

    2013-01-01

    Ion transport membrane (ITM) reactors present an attractive technology for combined air separation and fuel conversion in applications such as syngas production, oxidative coupling or oxy-combustion, with the promise of lower capital and operating costs, as well higher product selectivities than traditional technologies. The oxygen permeation rate through a given ITM is defined by the membrane temperature and oxygen chemical potential difference across it. Both of these parameters can be strongly influenced by thermochemical reactions occurring in the vicinity of the membrane, though in the literature they are often characterized in terms of the well mixed product stream at the reactor exit. This work presents the development of a novel ITM reactor for the fundamental investigation of the coupling between fuel conversion and oxygen permeation under well defined fluid dynamic and thermodynamic conditions, including provisions for spatially resolved, in-situ investigations. A planar, finite gap stagnation flow reactor with optical and probe access to the reaction zone is used to facilitate in-situ measurements and cross-validation with detailed numerical simulations. Using this novel reactor, baseline measurements are presented to elucidate the impact of the sweep gas fuel (CH4) fraction on the oxygen permeation and fuel conversion. In addition, the difference between well-mixed gas compositions measured at the reactor outlet and those measured in the vicinity of the membrane surface are discussed, demonstrating the unique utility of the reactor. © 2012 The Combustion Institute. Published by Elsevier Inc. All rights reserved.

  12. Characteristics-based model predictive control of a catalytic flow reversal reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fuxman, A.M.; Forbes, J.F.; Hayes, R.E. [Alberta Univ., Edmonton, AB (Canada). Dept. of Chemical and Materials Engineering

    2007-08-15

    A model-based controller for a catalytic flow reversal reactor (CFRR) was presented. The characteristics-based model predictive control (CBMPC) was used to provide greater accuracy in the prediction of process output variables as well as to ensure the maintenance of safe operating temperatures. Performance of the CBMPC was simulated in order to evaluate combustion of lean methane streams for the reduction of greenhouse gas (GHG) emissions. Dynamics of the CFRR were described using partial differential equations (PDEs) derived from mass and energy balances. The PDEs were then transformed into an equivalent lumped parameter model, which was in turn used to design the non-linear predictive controller. The prediction horizon was divided into Hp intervals during each half cycle. A constrained quadratic program was then solved to obtain an optimal input sequence. The strategy was then evaluated by applying it to a simple CFRR plant, as well as a more complex plant modelled by a dynamical-dimensional heterogenous model that incorporated the effect of a large insulation layer needed to reduce heat loss from the reactor. Results of the simulations suggested that mass extraction in a CBMPC scheme can be used to maintain safe operating conditions. It was concluded that the strategy provided good control performance for regulation and set point tracking in the presence of inlet disturbances and other changes in operating conditions. 18 refs., 1 tab., 10 figs.

  13. A Reactor Development Scenario for the FUZE Shear-flow Stabilized Z-pinch

    Science.gov (United States)

    McLean, H. S.; Higginson, D. P.; Schmidt, A.; Tummel, K. K.; Shumlak, U.; Nelson, B. A.; Claveau, E. L.; Golingo, R. P.; Weber, T. R.

    2016-10-01

    We present a conceptual design, scaling calculations, and a development path for a pulsed fusion reactor based on the shear-flow-stabilized Z-pinch device. Experiments performed on the ZaP device have demonstrated stable operation for 40 us at 150 kA total discharge current (with 100 kA in the pinch) for pinches that are 1cm in diameter and 100 cm long. Scaling calculations show that achieving stabilization for a pulse of 100 usec, for discharge current 1.5 MA, in a shortened pinch 50 cm, results in a pinch diameter of 200 um and a reactor plant Q 5 for reasonable assumptions of the various system efficiencies. We propose several key intermediate performance levels in order to justify further development. These include achieving operation at pinch currents of 300 kA, where Te and Ti are calculated to exceed 1 keV, 700 kA where fusion power exceeds pinch input power, and 1 MA where fusion energy per pulse exceeds input energy per pulse. This work funded by USDOE ARPAe ALPHA Program and performed under the auspices of Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344. LLNL-ABS-697801.

  14. 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

  15. Efficient H2O2/CH3COOH oxidative desulfurization/denitrification of liquid fuels in sonochemical flow-reactors.

    Science.gov (United States)

    Calcio Gaudino, Emanuela; Carnaroglio, Diego; Boffa, Luisa; Cravotto, Giancarlo; Moreira, Elizabeth M; Nunes, Matheus A G; Dressler, Valderi L; Flores, Erico M M

    2014-01-01

    The oxidative desulfurization/denitrification of liquid fuels has been widely investigated as an alternative or complement to common catalytic hydrorefining. In this process, all oxidation reactions occur in the heterogeneous phase (the oil and the polar phase containing the oxidant) and therefore the optimization of mass and heat transfer is of crucial importance to enhancing the oxidation rate. This goal can be achieved by performing the reaction in suitable ultrasound (US) reactors. In fact, flow and loop US reactors stand out above classic batch US reactors thanks to their greater efficiency and flexibility as well as lower energy consumption. This paper describes an efficient sonochemical oxidation with H2O2/CH3COOH at flow rates ranging from 60 to 800 ml/min of both a model compound, dibenzotiophene (DBT), and of a mild hydro-treated diesel feedstock. Four different commercially available US loop reactors (single and multi-probe) were tested, two of which were developed in the authors' laboratory. Full DBT oxidation and efficient diesel feedstock desulfurization/denitrification were observed after the separation of the polar oxidized S/N-containing compounds (S≤5 ppmw, N≤1 ppmw). Our studies confirm that high-throughput US applications benefit greatly from flow-reactors. Copyright © 2013 Elsevier B.V. All rights reserved.

  16. Tar Production from Biomass Pyrolysis in a Fluidized Bed Reactor: A Novel Turbulent Multiphase Flow Formulation

    Science.gov (United States)

    Bellan, J.; Lathouwers, D.

    2000-01-01

    A novel multiphase flow model is presented for describing the pyrolysis of biomass in a 'bubbling' fluidized bed reactor. The mixture of biomass and sand in a gaseous flow is conceptualized as a particulate phase composed of two classes interacting with the carrier gaseous flow. The solid biomass is composed of three initial species: cellulose, hemicellulose and lignin. From each of these initial species, two new solid species originate during pyrolysis: an 'active' species and a char, thus totaling seven solid-biomass species. The gas phase is composed of the original carrier gas (steam), tar and gas; the last two species originate from the volumetric pyrolysis reaction. The conservation equations are derived from the Boltzmann equations through ensemble averaging. Stresses in the gaseous phase are the sum of the Newtonian and Reynolds (turbulent) contributions. The particulate phase stresses are the sum of collisional and Reynolds contributions. Heat transfer between phases, and heat transfer between classes in the particulate phase is modeled, the last resulting from collisions between sand and biomass. Closure of the equations must be performed by modeling the Reynolds stresses for both phases. The results of a simplified version (first step) of the model are presented.

  17. Optimal design of radioactive particle tracking experiments for flow mapping in opaque multiphase reactors.

    Science.gov (United States)

    Roy, Shantanu; Larachi, Faical; Al-Dahhan, M H; Duduković, M P

    2002-03-01

    In the past decade, radioactive particle tracking techniques have emerged in the field of chemical engineering and have become increasingly popular for non-invasive flow mapping of the hydrodynamics in multiphase reactors. Based on gamma-ray sensitization of an array of scintillation detectors, the Computer Automated Radioactive Particle Tracking (CARPT) technique measures flow fields by monitoring the actual motion path of a single discrete radioactive flow follower which has the physical properties of the phase whose motion is being followed. A limitation to the accuracy of CARPT lies in the error associated with the reconstruction of the tracer particle position which affects the space-resolution capability of the technique. It is of interest, therefore, to minimize this error by choosing wisely the best hardware and an optimal configuration of CARPT detectors' array. Such choices are currently based on experience, without firm scientific basis. In this paper, through theoretical modeling and simulation, we describe how the accuracy of a radioactive particle tracking setup may be assessed a priori. Through an example of a proposed implementation of CARPT on a gas-solids riser, we demonstrate how this knowledge can be used for choosing the hardware required for the experiment. Finally, we show how the optimal arrangement of detectors can be effected for maximum accuracy for a given amount of monetary investment for the experiment.

  18. VALIDATION OF NUMERICAL METHODS TO CALCULATE BYPASS FLOW IN A PRISMATIC GAS-COOLED REACTOR CORE

    Directory of Open Access Journals (Sweden)

    NAM-IL TAK

    2013-11-01

    Full Text Available For thermo-fluid and safety analyses of a High Temperature Gas-cooled Reactor (HTGR, intensive efforts are in progress in the developments of the GAMMA+ code of Korea Atomic Energy Research Institute (KAERI and the AGREE code of the University of Michigan (U of M. One of the important requirements for GAMMA+ and AGREE is an accurate modeling capability of a bypass flow in a prismatic core. Recently, a series of air experiments were performed at Seoul National University (SNU in order to understand bypass flow behavior and generate an experimental database for the validation of computer codes. The main objective of the present work is to validate the GAMMA+ and AGREE codes using the experimental data published by SNU. The numerical results of the two codes were compared with the measured data. A good agreement was found between the calculations and the measurement. It was concluded that GAMMA+ and AGREE can reliably simulate the bypass flow behavior in a prismatic core.

  19. Computational fluid dynamics of a cylindrical nucleation flow reactor with detailed cluster thermodynamics.

    Science.gov (United States)

    Panta, Baradan; Glasoe, Walker A; Zollner, Juliana H; Carlson, Kimberly K; Hanson, David R

    2012-10-18

    Particle formation and growth with H(2)SO(4) molecules in an axially symmetric flow reactor was simulated with computational fluid dynamics. A warm (~310 K) gas containing H(2)SO(4) flows into a cooled section (296 K) that induces particle formation. The fluid dynamics gives flow fields, temperatures, and reactant and cluster distributions. Particle formation and growth are simulated with detailed H(2)SO(4) cluster kinetics with chemistry based on measured small cluster thermodynamics and on bulk thermodynamics for large clusters. Results show that particle number densities have power law dependencies on sulfuric acid of ~7, in accord with the thermodynamics of the cluster chemistry. The region where particle formation rates are largest has a temperature that is within 3 K of the wall. Additional simulations show that the H(2)SO(4) concentration in this region is 5 to 10 times greater than the measured H(2)SO(4): this information allows for direct comparisons of experiment and theory. Experiments where ammonia was added as a third nucleating species were simulated with a three-dimensional model. Ammonia was dispersed quickly and particle formation during this mixing was seen to be low. Downstream of the initial mixing region, however, ammonia greatly affected particle formation.

  20. Oxygen Transfer Model for a Flow-Through Hollow-Fiber Membrane Biofilm Reactor

    DEFF Research Database (Denmark)

    Gilmore, K. R.; Little, J. C.; Smets, Barth F.

    2009-01-01

    A mechanistic oxygen transfer model was developed and applied to a flow-through hollow-fiber membrane-aerated biofilm reactor. Model results are compared to conventional clean water test results as well as performance data obtained when an actively nitrifying biofilm was present on the fibers......-liquid interface was the most accurate of the predictive models (overpredicted by a factor of 1.1) while a coefficient determined by measuring bulk liquid dissolved oxygen underpredicted the oxygen transfer by a factor of 3. The mechanistic model was found to be an adequate tool for design because it used....... With the biofilm present, oxygen transfer efficiencies between 30 and 55% were calculated from the measured data including the outlet gas oxygen concentration, ammonia consumption stoichiometry, and oxidized nitrogen production stoichiometry, all of which were in reasonable agreement. The mechanistic model...

  1. A full-scale plug-flow reactor for biological sludge ozonation.

    Science.gov (United States)

    Gardoni, Davide; Ficara, Elena; Vergine, Pompilio; Canziani, Roberto

    2015-01-01

    The reduction of biological excess sludge production using ozone is a well-known technology and is applied in several full-scale plants around the world. Nevertheless, optimisation of the process is not yet adequately documented in the literature. Operational parameters are usually chosen by assuming a direct proportionality between ozone dose and excess sludge reduction. This paper investigates the role of ozone concentration on process efficiency and demonstrates the (non-linear) inverse relationship between ozone dose and specific particulate chemical oxygen demand solubilisation in plug-flow contact reactors. The influence of total suspended solids concentration is also studied and described. No short-term lethal effects on heterotrophic biomass have been observed.

  2. Modeling of acetylene pyrolysis under steel vacuum carburizing conditions in a tubular flow reactor.

    Science.gov (United States)

    Khan, Rafi Ullah; Bajohr, Siegfried; Graf, Frank; Reimert, Rainer

    2007-03-02

    In the present work, the pyrolysis of acetylene was studied under steel vacuum carburizing conditions in a tubular flow reactor. The pyrolysis temperature ranged from 650 degrees C to 1050 degrees C. The partial pressure of acetylene in the feed mixture was 10 and 20 mbar, respectively, while the rest of the mixture consisted of nitrogen. The total pressure of the mixture was 1.6 bar. A kinetic mechanism which consists of seven species and nine reactions has been used in the commercial computational fluid dynamics (CFD) software Fluent. The species transport and reaction model of Fluent was used in the simulations. A comparison of simulated and experimental results is presented in this paper.

  3. Modelingof Acetylene Pyrolysis under Steel Vacuum Carburizing Conditions in a Tubular Flow Reactor

    Directory of Open Access Journals (Sweden)

    Rainer Reimert

    2007-03-01

    Full Text Available In the present work, the pyrolysis of acetylene was studied under steel vacuumcarburizing conditions in a tubular flow reactor. The pyrolysis temperature ranged from650 °C to 1050 °C. The partial pressure of acetylene in the feed mixture was 10 and 20mbar, respectively, while the rest of the mixture consisted of nitrogen. The total pressureof the mixture was 1.6 bar. A kinetic mechanism which consists of seven species andnine reactions has been used in the commercial computational fluid dynamics (CFDsoftware Fluent. The species transport and reaction model of Fluent was used in thesimulations. A comparison of simulated and experimental results is presented in thispaper.

  4. Modeling flow stress constitutive behavior of SA508-3 steel for nuclear reactor pressure vessels

    Science.gov (United States)

    Sun, Mingyue; Hao, Luhan; Li, Shijian; Li, Dianzhong; Li, Yiyi

    2011-11-01

    Based on the measured stress-strain curves under different temperatures and strain rates, a series of flow stress constitutive equations for SA508-3 steel were firstly established through the classical theories on work hardening and softening. The comparison between the experimental and modeling results has confirmed that the established constitutive equations can correctly describe the mechanical responses and microstructural evolutions of the steel under various hot deformation conditions. We further represented a successful industrial application of this model to simulate a forging process for a large conical shell used in a nuclear steam generator, which evidences its practical and promising perspective of our model with an aim of widely promoting the hot plasticity processing for heavy nuclear components of fission reactors.

  5. Determination of hydrogen peroxide in reactor moderator solutions by flow injection analysis

    Energy Technology Data Exchange (ETDEWEB)

    Whitaker, M.J.

    1991-12-31

    A flow injection analysis (FIA) method for the determination of hydrogen peroxide in reactor moderator water was developed and installed at the Savannah River Site (SRS) Water Quality Laboratory. The technique has an analytical range of 0.10 to 2.50 ppm (ug/mL) with a sampling rate of 40 samples per hour. The calibration curve is linear with a correlation coefficient of 0.999, and the precision is excellent with relative standard deviations at the 0.50% level for both 0.10 and 2.50 ppm standards. When the automated FIA procedure is compared to the manual method it demonstrates a twenty minute reduction in analysis time per sample, and the total liquid waste generated per sample analyzed is reduced by roughly 95 mL. 6 refs., 10 figs., 6 tabs.

  6. Determination of hydrogen peroxide in reactor moderator solutions by flow injection analysis

    Energy Technology Data Exchange (ETDEWEB)

    Whitaker, M.J.

    1991-01-01

    A flow injection analysis (FIA) method for the determination of hydrogen peroxide in reactor moderator water was developed and installed at the Savannah River Site (SRS) Water Quality Laboratory. The technique has an analytical range of 0.10 to 2.50 ppm (ug/mL) with a sampling rate of 40 samples per hour. The calibration curve is linear with a correlation coefficient of 0.999, and the precision is excellent with relative standard deviations at the 0.50% level for both 0.10 and 2.50 ppm standards. When the automated FIA procedure is compared to the manual method it demonstrates a twenty minute reduction in analysis time per sample, and the total liquid waste generated per sample analyzed is reduced by roughly 95 mL. 6 refs., 10 figs., 6 tabs.

  7. An asymptotic observer-based monitoring scheme for a class of plug flow reactors.

    Science.gov (United States)

    Aguilar-Garnica, Efrén; García-Sandoval, Juan Paulo; Aceves-Lara, César Arturo; Escalante, Froylán Mario E

    2012-01-01

    In this paper a monitoring tool is designed for a class of plug flow reactors whose mathematical model is described by a set of first-order partial differential equations with different coefficients in the convective terms. The infinite dimensional structure of such a tool is derived according to the methodology established in the design of the well-known asymptotic observer. As a consequence, it preserves the robustness of the aforementioned observer against the lack of information of the nonlinear terms involved in the model. The original structure of the estimator is then represented as a couple of integral equations by means of the method of characteristics and its behaviour is analyzed through simulation experiments. These simulations show that the mean square observation error is 0.58 when the proposed observer is implemented in a solid-waste anaerobic digestion process to estimate the evolution of biomass concentration.

  8. Propellant actuated nuclear reactor steam depressurization valve

    Science.gov (United States)

    Ehrke, Alan C.; Knepp, John B.; Skoda, George I.

    1992-01-01

    A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

  9. Performance Analysis of Thermoelectric Based Automotive Waste Heat Recovery System with Nanofluid Coolant

    Directory of Open Access Journals (Sweden)

    Zhi Li

    2017-09-01

    Full Text Available Output performance of a thermoelectric-based automotive waste heat recovery system with a nanofluid coolant is analyzed in this study. Comparison between Cu-Ethylene glycol (Cu-EG nanofluid coolant and ethylene glycol with water (EG-W coolant under equal mass flow rate indicates that Cu-EG nanofluid as a coolant can effectively improve power output and thermoelectric conversion efficiency for the system. Power output enhancement for a 3% concentration of nanofluid is 2.5–8 W (12.65–13.95% compared to EG-Water when inlet temperature of exhaust varies within 500–710 K. The increase of nanofluid concentration within a realizable range (6% has positive effect on output performance of the system. Study on the relationship between total area of thermoelectric modules (TEMs and output performance of the system indicates that optimal total area of TEMs exists for maximizing output performance of the system. Cu-EG nanofluid as coolant can decrease optimal total area of TEMs compared with EG-W, which will bring significant advantages for the optimization and arrangement of TEMs whether the system space is sufficient or not. Moreover, power output enhancement under Cu-EG nanofluid coolant is larger than that of EG-W coolant due to the increase of hot side heat transfer coefficient of TEMs.

  10. Exhaust temperature analysis of four stroke diesel engine by using MWCNT/Water nanofluids as coolant

    Science.gov (United States)

    Muruganandam, M.; Mukesh Kumar, P. C.

    2017-10-01

    There has been a continuous improvement in designing of cooling system and in quality of internal combustion engine coolants. The liquid engine coolant used in early days faced many difficulties such as low boiling, freezing points and inherently poor thermal conductivity. Moreover, the conventional coolants have reached their limitations of heat dissipating capacity. New heat transfer fluids have been developed and named as nanofluids to try to replace traditional coolants. Moreover, many works are going on the application of nanofluids to avail the benefits of them. In this experimental investigation, 0.1, 0.3 and 0.5% volume concentrations of multi walled carbon nanotube (MWCNT)/water nanofluids have been prepared by two step method with surfactant and is used as a coolant in four stroke single cylinder diesel engine to assess the exhaust temperature of the engine. The nanofluid prepared is characterized with scanning electron microscope (SEM) to confirm uniform dispersion and stability of nanotube with zeta potential analyzer. Experimental tests are performed by various mass flow rate such as 270 300 330 LPH (litre per hour) of coolant nanofluids and by changing the load in the range of 0 to 2000 W and by keeping the engine speed constant. It is found that the exhaust temperature decreases by 10-20% when compared to water as coolant at the same condition.

  11. Sensitivity of hydrodynamic parameters' distributions in VVER-1000 reactor pressure vessel (RPV) with respect to uncertainty of the local hydraulic resistance coefficients

    Energy Technology Data Exchange (ETDEWEB)

    Pasichnyk, I.; Velkov, K. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany); Nikonov, S. [NRC ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2014-08-15

    The paper presents an uncertainty and sensitivity (U and S) study of the VVER-1000 reactor hydraulic properties. It is based on the OECD/NEA coolant transient Benchmark (K-3) on measured data at Kalinin-3 Nuclear Power Plant (NPP). The novelty of the work consists of taking into consideration all hydraulic uncertainty parameters used in the modeling of the reactor pressure vessel (RPV) internals. A detailed parallel channel ATHLET model of the RPV is developed. It consists of ca. 26 600 control volumes most of them connected with junctions for cross flows. The specific geometry of the gap between upper part of the baffle and upper part of fuel assembly and also a fuel assembly head is taken explicitly into account The influence of the input parameters on the sensitivity and uncertainty of the RPV outlet and inlet temperatures and mass flows as well assembly-wise mass flow and coolant temperature axial distributions is shown.

  12. Composite polymer/oxide hollow fiber contactors: versatile and scalable flow reactors for heterogeneous catalytic reactions in organic synthesis.

    Science.gov (United States)

    Moschetta, Eric G; Negretti, Solymar; Chepiga, Kathryn M; Brunelli, Nicholas A; Labreche, Ying; Feng, Yan; Rezaei, Fateme; Lively, Ryan P; Koros, William J; Davies, Huw M L; Jones, Christopher W

    2015-05-26

    Flexible composite polymer/oxide hollow fibers are used as flow reactors for heterogeneously catalyzed reactions in organic synthesis. The fiber synthesis allows for a variety of supported catalysts to be embedded in the walls of the fibers, thus leading to a diverse set of reactions that can be catalyzed in flow. Additionally, the fiber synthesis is scalable (e.g. several reactor beds containing many fibers in a module may be used) and thus they could potentially be used for the large-scale production of organic compounds. Incorporating heterogeneous catalysts in the walls of the fibers presents an alternative to a traditional packed-bed reactor and avoids large pressure drops, which is a crucial challenge when employing microreactors. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  13. Development of neutron measurement techniques in reactor diagnostics and determination of water content and water flow

    Energy Technology Data Exchange (ETDEWEB)

    Avdic, Senada

    2000-09-01

    The present thesis deals with three comparatively different topics in neutron physics research. These topics are as follows: construction and experimental investigation of a new detector, capable of measuring the neutron current, and investigation of the possibility to use it for the localisation of a neutron source in a simple experimental arrangement; execution of neutron transmission measurements based on a stationary neutron generator, and the study of their suitability for determining the volume porosity of geological samples; study of the possibility for improving the accuracy of water flow measurements based on the pulsed neutron activation technique. The first subject of this thesis concerns the measurement of the neutron current by a newly constructed detector. The motivation for this work stems from a recent suggestion that the performance of core monitoring methods could be enhanced if, in addition to the scalar neutron flux, also the neutron current was measured. To this end, a current detector was based on a scintillator mounted on a fibre and a Cd layer on one side of the detector. The measurements of the 2-D neutron current were performed in an experimental system by using this detector. The efficiency of the detector in reactor diagnostics was illustrated by demonstrating that the position of a neutron source can be determined by measuring the scalar neutron flux and the neutron current in one spatial point. The results of measurement and calculation show both the suitability of the detector construction for the measurement of the neutron current vector and the use of the current in diagnostics and monitoring. The second subject of this thesis concerns fast neutron transmission measurements, based on a stationary neutron generator, for determining the volume porosity of a sample in a model experiment. Such a technique could be used in field measurements with obvious advantages in comparison with thermal neutron transmission techniques, which can

  14. Investigation of the Air-Argon-Steel-Slag Flow in an Industrial RH Reactor with VOF-DPM Coupled Model

    Science.gov (United States)

    Chen, Gujun; He, Shengping; Li, Yugang

    2017-08-01

    A coupled three-dimensional volume of fluid method-discrete phase model (VOF-DPM) is developed to investigate the air-argon-steel-slag flow in an industrial Rheinsahl-Heraeus (RH) reactor while considering the expansion of argon bubbles. The simulated results of mixing time and recirculation flow rate of molten steel, and the flow pattern and local velocity of water agree well with the measured results reported in the literature. Comparison of the results with and without consideration of the expansion of bubbles indicates that the expansion of bubbles has an enormous impact on the multiphase flow in the industrial RH reactor. The proposed mathematical model presents a more realistic free surface in the RH vacuum vessel.

  15. CFD Modeling of Sodium-Oxide Deposition in Sodium-Cooled Fast Reactor Compact Heat Exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Tatli, Emre; Ferroni, Paolo; Mazzoccoli, Jason

    2015-09-02

    The possible use of compact heat exchangers (HXs) in sodium-cooled fast reactors (SFR) employing a Brayton cycle is promising due to their high power density and resulting small volume in comparison with conventional shell-and-tube HXs. However, the small diameter of their channels makes them more susceptible to plugging due to Na2O deposition during accident conditions. Although cold traps are designed to reduce oxygen impurity levels in the sodium coolant, their failure, in conjunction with accidental air ingress into the sodium boundary, could result in coolant oxygen levels that are above the saturation limit in the cooler parts of the HX channels. This can result in Na2O crystallization and the formation of solid deposits on cooled channel surfaces, limiting or even blocking coolant flow. The development of analysis tools capable of modeling the formation of these deposits in the presence of sodium flow will allow designers of SFRs to properly size the HX channels so that, in the scenario mentioned above, the reactor operator has sufficient time to detect and react to the affected HX. Until now, analytical methodologies to predict the formation of these deposits have been developed, but never implemented in a high-fidelity computational tool suited to modern reactor design techniques. This paper summarizes the challenges and the current status in the development of a Computational Fluid Dynamics (CFD) methodology to predict deposit formation, with particular emphasis on sensitivity studies on some parameters affecting deposition.

  16. Methanol oxidation in a flow reactor: Implications for the branching ratio of the CH3OH+OH reaction

    DEFF Research Database (Denmark)

    Rasmussen, Christian Lund; Wassard, K.H.; Dam-Johansen, Kim

    2008-01-01

    The oxidation of methanol in a flow reactor has been studied experimentally under diluted, fuel-lean conditions at 650-1350 K, over a wide range of O-2 concentrations (1%-16%), and with and without the presence of nitric oxide. The reaction is initiated above 900 K, with the oxidation rate...

  17. Effects of Swirl Bubble Injection on Mass Transfer and Hydrodynamics for Bubbly Flow Reactors: A Concept Paper

    Directory of Open Access Journals (Sweden)

    Farooqi Ahmad Salam

    2017-01-01

    Full Text Available Bubble flow reactors (BFR are commonly used for various industrial processes in the field of oil and gas production, pharmaceutical industries, biochemical and environmental engineering etc. The operation and performance of these reactors rely heavily on a range of hydrodynamic parameters; prominent among them are geometric configurations including gas injection geometry, operating conditions, mass transfer etc. A huge body of literature is available to describe the optimum design and performance of bubbly flow reactors with conventional bubble injection. Attempts were made to modify gas injection for improved efficiency of BFR’s. However, here instead of modifying the geometry of the gas injection, an attempt has been made to generate swirl bubbles for gaining larger mass transfer between gas and liquid. Here an exceptionally well thought strategies have been used in our numerical simulations towards the design of swirl injection mechanism, whose paramount aspect is to inhibit the rotary liquid motion but facilitates the swirl movement for bubbles in nearly stationary liquid. Our comprehension here is that the swirl motion can strongly affect the performance of bubbly reactor by identifying the changes in hydrodynamic parameters as compared to the conventional bubbly flows. In order to achieve this bubbly flow, an experimental setup has been designed as well as computational fluid dynamic (CFD code was used with to highlight a provision of swirl bubble injection by rotating the sparger plate.

  18. Ignition of DME and DME/CH4 at High Pressure: Flow Reactor Experiments and Kinetic Modeling

    DEFF Research Database (Denmark)

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

    The pyrolysis and oxidation of dimethyl ether (DME) and its mixtures with methane were investigated at high pressures (50 and 100 bar) and intermediate temperatures (450―900 K) in a laminar flow reactor. DME pyrolysis started at 825 K (at 50 bar). The onset of DME reaction was detected at 525―550 K...

  19. 75 FR 8412 - Office of New Reactors: Interim Staff Guidance on Assessing Ground Water Flow and Transport of...

    Science.gov (United States)

    2010-02-24

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Office of New Reactors: Interim Staff Guidance on Assessing Ground Water Flow and Transport of... Interim Staff Guidance (ISG) DC/COL-ISG-014 (Agencywide Documents Access and Management System (ADAMS...

  20. 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.

  1. Isothermal flow measurement using planar PIV in the 1/4 scaled model of CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Im, Sunghyuk; Sung, Hyung Jin [KAIST, Daejeon (Korea, Republic of); Seo, Han; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Kim, Hyoung Tae [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The local temperature of the moderator is a key parameter in determining the available subcooling. To predict the flow field and local temperature distribution in the calandria, Korea Atomic Energy Research Institute (KAERI) started the experimental research on moderator circulation as one of a national R and D research programs from 2012. This research program includes the construction of the Moderator Circulation Test (MCT) facility, production of the validation data for self-reliant CFD tools, and development of optical measurement system using the Particle Image Velocimetry (PIV) and Laser Induced Fluorescence (LIF) techniques. Small-scale 1/40 and 1/8 small-scale model tests were performed prior to installation of the main MCT facility to identify the potential problems of the flow visualization and measurement expected in the 1/4 scale MCT facility. In the 1/40 scale test, a flow field was measured with a PIV measurement technique under an iso-thermal state, and the temperature field was visualized using a LIF technique. In this experiment, the key point was to illuminate the region of interest as uniformly as possible since the velocity and temperature fields in the shadow regions were distorted and unphysical. In the 1/8 scale test, the flow patterns from the inlet nozzles to the top region of the tank were investigated using PIV measurement at two different positions of the inlet nozzle. For each position of laser beam exposure the measurement sections were divided to 7 groups to overcome the limitation of the laser power to cover the relatively large test section. The MCT facility is the large-scale facility designed to reproduce the important characteristics of moderator circulation in a CANDU6 calandria under a range of operating conditions. It is reduced in a 1/4 scale and a moderator test vessel is built to the specifications of the CANDU6 reactor design, where a working fluid is sub-cooled water with atmospheric pressure. Previous studies were

  2. Development of a novel heterogeneous flow reactor -- Soot formation and nanoparticle catalysis

    Science.gov (United States)

    Camacho, Joaquin

    The development of novel experimental approaches to investigate fundamental surface kinetics is presented. Specifically, fundamental soot formation and surface catalysis processes are examined in isolation from other competing processes. In terms of soot formation, two experimental techniques are presented: the Burner Stabilized Stagnation (BSS) flame configuration is extended to isolate the effect of the parent fuel structure on soot formation and the fundamental rate of surface oxidation for nascent soot is measured in a novel aerosol flow reactor. In terms of nanoparticles, the physical and chemical properties of freely suspended nanoparticles are investigated in a novel aerosol flow reactor for methane oxidation catalyzed by palladium. The role of parent fuel structure within soot formation is examined by following the time resolved formation nascent soot from the onset of nucleation to later growth stages for premixed BSS flames. Specifically, the evolution of the detailed particle size distribution function (PSDF) is compared for butanol, butane and C6 hydrocarbons in two separate studies where the C/O ratio and temperature are fixed. Under this constraint, the overall sooting process were comparable as evidenced by similar time resolved bimodal PSDF. However, the nucleation time and the persistence of nucleation with time is strongly dependent upon the structure of the parent fuel. For the C6 hydrocarbon fuels, the fastest onset of soot nucleation is observed in cyclohexane and benzene flames and this may be due to significant aromatic formation that is predicted in the pre-flame region. In addition, the evolution of the PSDF shows that nucleation ends sooner in cylclohexane and benzene flames and this may be due to relatively quick depletion of soot precursors such as acetylene and benzene. Interestingly,within the butanol fuels studied the effect of the branched chain in i-butanol and i-butane was more significant than the presence of fuel bound oxygen. A

  3. TACT 1: A computer program for the transient thermal analysis of a cooled turbine blade or vane equipped with a coolant insert. 2. Programmers manual

    Science.gov (United States)

    Gaugler, R. E.

    1979-01-01

    A computer program to calculate transient and steady state temperatures, pressures, and coolant flows in a cooled axial flow turbine blade or vane with an impingement insert is described. Coolant-side heat transfer coefficients are calculated internally in the program, with the user specifying either impingement or convection heat transfer at each internal flow station. Spent impingement air flows in a chordwise direction and is discharged through the trailing edge and through film cooling holes. The ability of the program to handle film cooling is limited by the internal flow model. Input to the program includes a description of the blade geometry, coolant-supply conditions, outside thermal boundary conditions, and wheel speed. The blade wall can have two layers of different materials, such as a ceramic thermal barrier coating over a metallic substrate. Program output includes the temperature at each node, the coolant pressures and flow rates, and the coolant-side heat transfer coefficients.

  4. Solar disinfection for the post-treatment of greywater by means of a continuous flow reactor.

    Science.gov (United States)

    Pansonato, Natália; Afonso, Marcos V G; Salles, Carlos A; Boncz, Marc A; Paulo, Paula L

    2011-01-01

    SODIS (solar disinfection) is a low-cost alternative for water decontamination. The method is based on the exposure of water, contained in PET bottles, to direct sunlight, and mainly its UV-A and infrared components. The present research studied SODIS as a low cost alternative for the inactivation of Escherichia coli (E. coli) in treated greywater, aiming at its reuse for more noble applications. Experiments were performed in (i) batch mode (2 L PET-bottles), testing the effect of turbidity on system efficiency and, (ii) in a continuous pilot-scale reactor prototype (51 L, using interconnected 2 L-PET bottles), testing hydraulic retention times (HRT) of 18 and 24 h. Samples were exposed to an average solar radiation intensity of 518 W/m2. The results obtained indicate that the SODIS system has potential for total coliforms and E. coli inactivation in the pre-treated greywater, reaching 2.1 log units E. coli inactivation in batch experiments for low turbidity samples (21 NTU), and > 2 log units inactivation of total coliforms (and E. coli, when present) for the 24 h HRT-continuous prototype. The continuous flow prototype needs more testing and structural improvements to cope with the difficulties posed by algae growth, as they complicate maintaining conditions of constant flow and make frequent maintenance inevitable.

  5. Thermal transfer structures coupling electronics card(s) to coolant-cooled structure(s)

    Science.gov (United States)

    David, Milnes P; Graybill, David P; Iyengar, Madhusudan K; Kamath, Vinod; Kochuparambil, Bejoy J; Parida, Pritish R; Schmidt, Roger R

    2014-12-16

    Cooling apparatuses and coolant-cooled electronic systems are provided which include thermal transfer structures configured to engage with a spring force one or more electronics cards with docking of the electronics card(s) within a respective socket(s) of the electronic system. A thermal transfer structure of the cooling apparatus includes a thermal spreader having a first thermal conduction surface, and a thermally conductive spring assembly coupled to the conduction surface of the thermal spreader and positioned and configured to reside between and physically couple a first surface of an electronics card to the first surface of the thermal spreader with docking of the electronics card within a socket of the electronic system. The thermal transfer structure is, in one embodiment, metallurgically bonded to a coolant-cooled structure and facilitates transfer of heat from the electronics card to coolant flowing through the coolant-cooled structure.

  6. Adaptive LQ Cascade Control of a Tubular Chemical Reactor

    Directory of Open Access Journals (Sweden)

    Petr Dostal

    2016-01-01

    Full Text Available The paper deals with adaptive LQ cascade control design of a tubular chemical reactor with an exothermic consecutive reaction. The control is performed in primary and secondary control-loops where the primary controlled output of the reactor is the concentration of a main reaction product and the secondary output is the mean temperature of the reactant. A common control input is the coolant flow rate. The controller in the primary control-loop is a nonlinear P-controller with the gain calculated using simulated or measured steady-state characteristics of the reactor. The controller in the secondary control-loop is a LQ adaptive controller. The proposed method is verified by control simulations.

  7. A comparison of coal char reactivity determined from thermogravimetric and laminar flow reactor experiments

    Energy Technology Data Exchange (ETDEWEB)

    Zolin, A.; Jensen, A.; Pedersen, L.S.; Dam-Johansen, K.; Toerslev, P. [Technical University of Denmark, Lyngby (Denmark). Dept. of Chemical Engineering

    1998-03-01

    The reactivity of nine different coals ranking from subbituminous to low-volatile bituminous has been studied by thermogravimetric analysis (TGA). At a standard set of conditions a qualitative fuel reactivity classification (ranking) with respect to one of the coals, Cerrejon, is presented. Particle reaction rates per unit external surface area and a normalized reactivity index based on raw experimental data were used as reactivity parameters to compare the fuels. The TGA chars were prepared at 900{degree}C with 15 min holding time and then combusted in a 20 mol % O{sub 2} environment at several temperatures in the range 450-650{degree}C. TGA reaction rate data were adequately interpreted by a random pore model. However, at 650{degree}C it is believed that particle ignition gave rise to a char reaction rate behavior that the model was incapable of describing properly. Except for two Southern Hemisphere coals, the reactivity ranking obtained with the TGA apparatus at a combustion temperature of 550{degree}C agrees well with a corresponding classification based on experiments carried out in another study with a laminar flow reactor (LFR) at {approximately} 1400{degree}C. The maximum difference in reaction rates based on external surface area between the coal chars in the low-temperature TGA experiments was 1 order of magnitude higher than in the high-temperature LFT experiments, due to the increasing effect of pore diffusion and thermal annealing of the coal chars in the LFR tests. The similarity in the reactivity ranking obtained for the Northern Hemisphere coals from both reactor systems indicates that a ranking can be performed by thermogravimetric analysis. This provides a simple means for determining a fuel reactivity ranking that could be applied to full scale suspension fired plants. 28 refs., 8 figs., 4 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Efanov, A.D.; Kuznetzov, Y.N.; Kaliakin, S.G.; Lisitza, F.D.; Remizov, O.V.; Serdun, N.P. [Institute of Physics and Power Engineering State Scientific Center, Obninsk (Russian Federation)

    2001-07-01

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

  9. Tube-in-tube reactor as a useful tool for homo- and heterogeneous olefin metathesis under continuous flow mode.

    Science.gov (United States)

    Skowerski, Krzysztof; Czarnocki, Stefan J; Knapkiewicz, Paweł

    2014-02-01

    A tube-in-tube reactor was successfully applied in homo- and heterogeneous olefin metathesis reactions under continuous flow mode. It was shown that the efficient removal of ethylene facilitated by connection of the reactor with a vacuum pump significantly improves the outcome of metathesis reactions. The beneficial aspects of this approach are most apparent in reactions performed at low concentration, such as macrocyclization reactions. The established system allows achievement of both improved yield and selectivity, and is ideal for industrial applications. Copyright © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  10. TACT1- TRANSIENT THERMAL ANALYSIS OF A COOLED TURBINE BLADE OR VANE EQUIPPED WITH A COOLANT INSERT

    Science.gov (United States)

    Gaugler, R. E.

    1994-01-01

    As turbine-engine core operating conditions become more severe, designers must develop more effective means of cooling blades and vanes. In order to design reliable, cooled turbine blades, advanced transient thermal calculation techniques are required. The TACT1 computer program was developed to perform transient and steady-state heat-transfer and coolant-flow analyses for cooled blades, given the outside hot-gas boundary condition, the coolant inlet conditions, the geometry of the blade shell, and the cooling configuration. TACT1 can analyze turbine blades, or vanes, equipped with a central coolant-plenum insert from which coolant-air impinges on the inner surface of the blade shell. Coolant-side heat-transfer coefficients are calculated with the heat transfer mode at each station being user specified as either impingement with crossflow, forced convection channel flow, or forced convection over pin fins. A limited capability to handle film cooling is also available in the program. The TACT1 program solves for the blade temperature distribution using a transient energy equation for each node. The nodal energy balances are linearized, one-dimensional, heat-conduction equations which are applied at the wall-outer-surface node, at the junction of the cladding and the metal node, and at the wall-inner-surface node. At the mid-metal node a linear, three-dimensional, heat-conduction equation is used. Similarly, the coolant pressure distribution is determined by solving the set of transfer momentum equations for the one-dimensional flow between adjacent fluid nodes. In the coolant channel, energy and momentum equations for one-dimensional compressible flow, including friction and heat transfer, are used for the elemental channel length between two coolant nodes. The TACT1 program first obtains a steady-state solution using iterative calculations to obtain convergence of stable temperatures, pressures, coolant-flow split, and overall coolant mass balance. Transient

  11. Effects of Cooling Fluid Flow Rate on the Critical Heat Flux and Flow Stability in the Plate Fuel Type 2 MW TRIGA Reactor

    Directory of Open Access Journals (Sweden)

    H. P. Rahardjo

    2017-12-01

    Full Text Available The conversion program of the 2 MW TRIGA reactor in Bandung consisted of the replacement of cylindrical fuel (produced by General Atomic with plate fuel (produced by BATAN. The replacement led into the change of core cooling process from upward natural convection type to downward forced convection type, and resulted in different thermohydraulic safety criteria, such as critical heat flux (CHF limit, boiling limit, and cooling fluid flow stability. In this paper, a thermohydraulic safety analysis of the converted TRIGA reactor is presented by considering the Dynamic Nucleate Boiling Ratio (DNBR criterion, Onset Nucleate Boiling Ratio (ONBR limit, and cooling fluid flow stability at various cooling fluid flow rate.The numerical analyses were performed using the HEATHYD program on the hottest channels of reactor core.The combination of heat transfer and fluid flow analysis were conducted for reactor operation at 2 MW with 20 fuel element bundles and four control rod bundles. Incoming fluid flow to the cooling channel was fixed at 44.5 °C temperature and 1.9970 bar pressure, and its flow rate was varied from 1.25 to 3.5 m3/h. By inputting these values, as well as the total power of fuel elements per bundle, the wall temperature distribution of the plate fuel element, cooling fluid temperature distribution, and pressure losses in the channels were obtained for the analysis of CHF limit, boiling limit, and flow stability. It was shown that no boiling occurred for the cooling fluid flow rate range of 2.4 to 3.5 m3/h, and even at the cooling fluid flow rate of 1.25 m3/h where some bubbles occurred, the DNBR was higher than the critical limit (more than 23 while the flow stability criterion in some channels were slightly less than 1 (unstable. At the cooling fluid flow rate of 1.4 m3/h, however, the flow became stable in all channel. System design description for GCFR-core flow test loop

    Energy Technology Data Exchange (ETDEWEB)

    Huntley, W.R.; Grindell, A.G.

    1980-12-01

    The Core Flow Test Loop is a high-pressure, high-temperature, out-of-reactor helium circulation system that is being constructed to permit detailed study of the thermomechanical and thermal performance at prototypic steady-state and transient operating conditions of simulated segments of core assemblies for a GCFR Demonstration Plant, as designed by General Atomic Company. It will also permit the expermental verification of predictive analytical models of the GCFR core assemblies needed to reduce operational and safety uncertainties of the GCFR. Full-sized blanket assemblies and segments of fuel rod and control rod fuel assemblies will be simulated with test bundles of electrically powered fuel rod or blanket rod simulators. The loop will provide the steady-state and margin test requirements of bundle power and heat removal, and of helium coolant flow rate, pressure, and temperature for test bundles having up to 91 rods; these requirements set the maximum power, coolant helium flow, and thermal requirements for the loop. However, the size of the test vessel that contains the test bundles will be determined by the bundles that simulate a full-sized GCFR blanket assembly. The loop will also provide for power and coolant transients to simulate transient operation of GCFR core assemblies, including the capability for rapid helium depressurization to simulate the depressurization class of GCFR accidents. In addition, the loop can be used as an out-of-reactor test bed for characterizing in-reactor test bundle configurations.

  12. Guidelines to achieve seals with minimal leak rates for HWR-NPR coolant system components

    Energy Technology Data Exchange (ETDEWEB)

    Finn, P.A.

    1991-03-01

    Seal design practices that are acceptable in pressurized-water and boiling-water reactors in the United States are not usable for the Heavy Water Reactor-New Production Reactor (HWR-NPR) because of the stringent requirement on tritium control for the atmosphere within its containment building. To maintain an atmosphere in which workers do not need protective equipment, the components of the coolant system must have a cumulative leak rate less than 0.00026 L/s. Existing technology for seal systems was reviewed with regard to flange, elastomer, valve, and pump design. A technology data base for the designers of the HWR-NPR coolant system was derived from operating experience and seal development work on reactors in the United States, Canada, and Europe. This data base was then used to generate guidelines for the design of seals and/or joints for the HWR-NPR coolant system. Also discussed are needed additional research and development, as well as the necessary component qualification tests for an effective quality control program. 141 refs., 21 figs., 14 tabs.

  13. Bioaugmentation of an acetate-oxidising anaerobic consortium in up-flow sludge blanket reactor subjected to high ammonia loads

    DEFF Research Database (Denmark)

    Fotidis, Ioannis; Karakashev, Dimitar Borisov; Angelidaki, Irini

    . in association with Methanoculleus spp. strain MAB1), is an acetate oxidising methanogenic consortium that can produce methane (CH4) at high ammonia levels. In the current study the bioaugmentation of the SAO culture in a mesophilic up-flow anaerobic sludge blanket (UASB) reactor subjected to high ammonia loads...... was tested. The co-cultivation in fed-batch of a fast-growing hydrogenotrophic methanogen (i.e. Methanoculleus bourgensis) with the SAO culture was also investigated. Results obtained clearly demonstrated that bioaugmentation of SAO culture in a UASB reactor was not possible most probably due to the slow...... growth of the culture. The incubation period (duration of lag+exponential phase) of SAO culture was reduced more than 30% when it was cocultivated with Methanoculleus bourgensis, in fed-batch reactors. Therefore, the bioaugmentation of the SAO culture along with Methanoculleus bourgensis in a UASB...

  14. Lessons Learned From Gen I Carbon Dioxide Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    David E. Shropshire

    2004-04-01

    This paper provides a review of early gas cooled reactors including the Magnox reactors originating in the United Kingdom and the subsequent development of the Advanced Gas-cooled Reactors (AGR). These early gas cooled reactors shared a common coolant medium, namely carbon dioxide (CO2). A framework of information is provided about these early reactors and identifies unique problems/opportunities associated with use of CO2 as a coolant. Reactor designers successfully rose to these challenges. After years of successful use of the CO2 gas cooled reactors in Europe, the succeeding generation of reactors, called the High Temperature Gas Reactors (HTGR), were designed with Helium gas as the coolant. Again, in the 21st century, with the latest reactor designs under investigation in Generation IV, there is a revived interest in developing Gas Cooled Fast Reactors that use CO2 as the reactor coolant. This paper provides a historical perspective on the 52 CO2 reactors and the reactor programs that developed them. The Magnox and AGR design features and safety characteristics were reviewed, as well as the technologies associated with fuel storage, reprocessing, and disposal. Lessons-learned from these programs are noted to benefit the designs of future generations of gas cooled nuclear reactors.

  15. Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 2. Numerical investigations on proprieties of R/V upper plenum separation

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu; Yamaguchi, Akira [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Murakami, Satoshi [Customer System Co. Ltd., Tokai, Ibaraki (Japan)

    2002-03-01

    A large-scale sodium-cooled fast breeder reactor investigated in the feasibility studies on commercialized fast reactors has a tendency of consideration of thorough simplified and compacted system designs to realize drastic economical improvements. Therefore, special attention should be paid to thermohydraulic designs for gas entrainment behavior from free surfaces, a flow-induced vibration of in-vessel components, thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. In-vessel thermohydraulic analyses were carried out using a multi-dimensional code AQUA to contribute to a planning of water scaled-model experiments for the separated upper plena of the reactor vessel. >From the analysis, the following results were obtained. (1) It can be considered that there is little effect of volumetric flow rate between the upper plenum and the free surface plenum on hydrodynamic characteristics in the whole upper plenum. Because the affected area is limited in the neighborhood of gaps of hot/cold leg pipes. (2) Change in flow velocity components is limited around the outer wall of hot leg pipes if the gap width of the dipped plate and penetrating components changes. From the above results, it was concluded that the assumption seems to be valid that the free surface plenum hydrodynamic characteristics is independent of the upper plenum flows. However it is necessary to consider jet effects due to the hot leg intake flows affecting vortex concentrations in the upper plenum. (author)

  16. Alkylation of Benzene with Propylene in a Flow-Through Membrane Reactor and Fixed-Bed Reactor: Preliminary Results

    Directory of Open Access Journals (Sweden)

    Sibele Pergher

    2012-05-01

    Full Text Available Benzene alkylation with propylene was studied in the gas phase using a catalytic membrane reactor and a fixed-bed reactor in the temperature range of 200–300 °C and with a weight hourly space velocity (WHSV of 51 h−1. β-zeolite was prepared by hydrothermal synthesis using silica, aluminum metal and TEAOH as precursors. The membrane’s XRD patterns showed good crystallinity for the β-zeolite film, while scanning electron microscopy SEM results indicated that its random polycrystalline film was approximately 1 μm thick. The powders’ specific area was determined to be 400 m2×g−1 by N2 adsorption/desorption, and the TPD results indicated an overall acidity of 3.4 mmol NH3×g−1. Relative to the powdered catalyst, the catalytic membrane showed good activity and product selectivity for cumene.

  17. Superheated Water-Cooled Small Modular Underwater Reactor Concept

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-12-01

    Full Text Available A novel fully passive small modular superheated water reactor (SWR for underwater deployment is designed to produce 160 MWe with steam at 500ºC to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF. The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and then the steam, instead of exiting the reactor pressure vessel, turns around and flows downward in the central channel of some IXAF fuel rods within each assembly and then flows upward through the rest of the IXAF pins in the assembly and exits the reactor pressure vessel as superheated steam. In this study, new cladding material to withstand high temperature steam in addition to the fuel mechanical and safety behavior is investigated. The steam temperature was found to depend on the thermal and mechanical characteristics of the fuel. The SWR showed a very different transient behavior compared with a boiling water reactor. The inter-play between the inner and outer channels of the IXAF was mainly beneficial except in the case of sudden reactivity insertion transients where additional control consideration is required.

  18. Measurements of Flow Mixing at Subchannels in a Wire-Wrapped 37-Rod Bundle for a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyungmo; Bae, Hwang; Chang, Seok-Kyu; Choi, Sun Rock; Lee, Dong Won; Ko, Yung Joo; Choi, Hae Seob; Euh, Dong-Jin; Lee, Hyeong-Yeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    For a safety analysis in a core thermal design of a sodium-cooled fast reactor (SFR), flow mixing characteristics at subchannels in a wire-wrapped rod bundle are very important. Wrapped wires make a cross flow in a around the fuel rod) of the fuel rod, and this effect lets flow be mixed. Experimental results of flow mixing can be meaningful for verification and validation of thermal mixing correlation in a reactor core thermo-hydraulic design code. A wire mesh sensing technique can be useful method for measuring of flow mixing characteristics. A wire mesh sensor has been traditionally used to measure the void fraction of a two-phase flow field, i.e. gas and liquid. However, it has been recently reported that the wire mesh sensor can be used successfully to recognize the flow field in liquid phase by injecting a tracing liquid with a different level of electric conductivity. This can be powerfully adapted to recognize flow mixing characteristics by wrapped wires in SFR core thermal design. In this work, we conducted the flow mixing experiments using a custom designed wire mesh sensor. To verify and validate computer codes for the SFR core thermal design, mixing experiments were conducted at a hexagonally arrayed 37-pin wire-wrapped fuel rod bundle test section. The well-designed wire mesh sensor was used to measure flow mixing characteristics. The developed post-processing method has its own merits, and flow mixing results were reasonable. In addition, by uncertainty analysis, the system errors and the random error were estimated in experiments. Therefore, the present results and methods can be used for design code verification and validation.

  19. Distributions and activities of ammonia oxidizing bacteria and polyphosphate accumulating organisms in a pumped-flow biofilm reactor.

    Science.gov (United States)

    Wu, Guangxue; Nielsen, Michael; Sorensen, Ketil; Zhan, Xinmin; Rodgers, Michael

    2009-10-01

    The spatial distributions and activities of ammonia oxidizing bacteria (AOB) and polyphosphate accumulating organisms (PAOs) were investigated for a novel laboratory-scale sequencing batch pumped-flow biofilm reactor (PFBR) system that was operated for carbon, nitrogen and phosphorus removal. The PFBR comprised of two 16.5l tanks (Reactors 1 and 2), each with a biofilm module of 2m(2) surface area. To facilitate the growth of AOB and PAOs in the reactor biofilms, the influent wastewater was held in Reactor 1 under stagnant un-aerated conditions for 6 h after feeding, and was then pumped over and back between Reactors 1 and 2 for 12 h, creating aerobic conditions in the two reactors during this period; as a consequence, the biofilm in Reactor 2 was in an aerobic environment for almost all the 18.2 h operating cycle. A combination of micro-sensor measurements, molecular techniques, batch experiments and reactor studies were carried out to analyse the performance of the PFBR system. After 100 days operation at a filtered chemical oxygen demand (COD(f)) loading rate of 3.46 g/m(2) per day, the removal efficiencies were 95% COD(f), 87% TN(f) and 74% TP(f). While the PFBR microbial community structure and function were found to be highly diversified with substantial AOB and PAO populations, about 70% of the phosphorus release potential and almost 100% of the nitrification potential were located in Reactors 1 and 2, respectively. Co-enrichment of AOB and PAOs was realized in the Reactor 2 biofilm, where molecular analyses revealed unexpected microbial distributions at micro-scale, with population peaks of AOB in a 100-250 microm deep sub-surface zone and of PAOs in the 0-150 microm surface zone. The micro-distribution of AOB coincided with the position of the nitrification peak identified during micro-sensor analyses. The study demonstrates that enrichment of PAOs can be realized in a constant or near constant aerobic biofilm environment. Furthermore, the findings suggest

  1. Numerical Simulation of Particle Flow Motion in a Two-Dimensional Modular Pebble-Bed Reactor with Discrete Element Method

    Directory of Open Access Journals (Sweden)

    Guodong Liu

    2013-01-01

    Full Text Available Modular pebble-bed nuclear reactor (MPBNR technology is promising due to its attractive features such as high fuel performance and inherent safety. Particle motion of fuel and graphite pebbles is highly associated with the performance of pebbled-bed modular nuclear reactor. To understand the mechanism of pebble’s motion in the reactor, we numerically studied the influence of number ratio of fuel and graphite pebbles, funnel angle of the reactor, height of guide ring on the distribution of pebble position, and velocity by means of discrete element method (DEM in a two-dimensional MPBNR. Velocity distributions at different areas of the reactor as well as mixing characteristics of fuel and graphite pebbles were investigated. Both fuel and graphite pebbles moved downward, and a uniform motion was formed in the column zone, while pebbles motion in the cone zone was accelerated due to the decrease of the cross sectional flow area. The number ratio of fuel and graphite pebbles and the height of guide ring had a minor influence on the velocity distribution of pebbles, while the variation of funnel angle had an obvious impact on the velocity distribution. Simulated results agreed well with the work in the literature.

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

    Science.gov (United States)

    Zaleski, Daniel P.; Prozument, Kirill

    2017-06-01

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

  3. Flow pattern analysis of a full-scale expanded granular sludge bed-type reactor under different organic loading rates.

    Science.gov (United States)

    Zheng, M X; Wang, K J; Zuo, J E; Yan, Z; Fang, H; Yu, J W

    2012-03-01

    The hydraulic characteristics of a lab-scale and a full-scale (275 m(3)) expanded granular sludge bed (EGSB)-type reactor under different organic loading rates varying from 10 kg COD m(-3)d(-1) to 45 kg COD m(-3)d(-1) were investigated. A modified combined model composed of two completely mixing regions and a plug flow region was sufficient for simulating the flow pattern of a full-scale EGSB-type reactor. Moreover, the outputs fitted the measured tracer distribution results well. The simplified model structure was in very good agreement with the physical structure of a full-scale EGSB-type reactor. The upflow (liquid+gas) velocity, high concentration of granular sludge, and gas hold-up effect may contribute to the generation of dead spaces (maximum of 19.5%). The bed expansion characteristics indicated that the sludge bed of the EGSB-type reactor performed as a suspended bed, in which the bed expansion was controlled between 20% and 30%, rather than the usually considered expanded bed. Copyright © 2011 Elsevier Ltd. All rights reserved.

  4. Combining UASB and the "fourth generation" down-flow hanging sponge reactor for municipal wastewater treatment.

    Science.gov (United States)

    Tandukar, M; Uemura, S; Ohashi, A; Harada, H

    2006-01-01

    A "fourth generation" down-flow hanging sponge (DHS) Reactor has been developed and proposed as an improved variant of post-treatment system for UASB treating domestic wastewater. This paper evaluates the potential of the proposed combination of UASB and DHS as a sewage treatment system, especially for developing countries. A pilot-scale UASB (1.15 m3) and DHS (0.38 m3; volume of sponge) was installed in a municipal sewage treatment site and constantly monitored for 2 years. UASB was operated at an HRT of 6 h corresponding to an organic load of 2.15 kg-COD/m3 per day. Subsequently, the organic load in DHS was 2.35 kg-COD/m3 per day, operated at an HRT of 2 h. Organic removal by the whole system was satisfactory, accomplishing 96% of unfiltered BOD removal and 91% of unfiltered COD removal. However, nitrification decreased from 56% during the startup period to 28% afterwards. Investigation on DHS sludge was made by quantifying it and evaluating oxygen uptake rates with various substrates. Average concentration of trapped biomass was 26 g-VSS/L of sponge volume, increasing the SRT of the system to 100-125 d. Removal of coliforms obtained was 3-4 log10 with the final count of 10(3) to 10(4) MPN/100 ml in DHS effluent.

  5. Chromate reduction by waste iron from electroplating wastewater using plug flow reactor.

    Science.gov (United States)

    Chen, Shiao-Shing; Hsu, Bao-Chrung; Hung, Li-Wei

    2008-04-15

    Waste iron was used to treat high concentration chromate (534 mg/L as Cr) from electroplating wastewater by plug flow reactor (PFR) due to the following reasons: (1) two wastes are treated simultaneously, (2) low pH of the electroplating wastewater ( approximately 2) benefits the reaction between these two wastes, (3) effluent pH is elevated in the PFR, reducing the base requirement to meet the pH discharge standard for wastewater (pH 6-9). Complete chromate reductions were achieved at pH 1.7 for hydraulic retention time (HRT) of 98 min, pH 1.5 for HRT of 40 min and pH 1.3 for HRT of 20 min. Consequently, optimum HRT for complete chromate reduction was obtained for different pHs. Although more acids were used to lower influent pH to reduce HRT, effluent pH was higher due to more hydrogen ion reacting with chromate. Eventually, fewer bases are required to fulfill the discharge pH requirement of wastewater. Effluent pH 3-5 was observed with high turbidity, indicating the precipitations of chromium oxide and hydroxide were enhanced by the dissolved iron coagulation. X-ray diffraction was conducted to examine the remaining species. Other than chromium oxide and hydroxide species, an iron-chromium complex (Cr2FeO4) was also observed.

  6. A horizontal plug-flow baffled bioelectrocatalyzed reactor for the reductive decolorization of Alizarin Yellow R.

    Science.gov (United States)

    Sun, Qian; Li, Zhiling; Wang, Youzhao; Cui, Dan; Liang, Bin; Thangavel, Sangeetha; Chung, Jong Shik; Wang, Aijie

    2015-11-01

    An application-oriented membrane-free, continuous plug-flow baffled bioelectrocatalyzed reactor (PFB-BER), was designed and testified for the decolorization of Alizarin Yellow R. Decolorization efficiency (DE) with an external power source of 0.5 V was higher than without electrolysis, i.e. 93.4% versus 73.6% (HRT of 24 h). Product formation efficiencies of p-phenylenediamine and 5-aminosalicylic acid were above 95% and 50%, respectively. When HRT decreased to 8 h and 4 h, DE reduced to 69.9% and 44.9%, respectively. An additional electrode assembly improved DE to 96.4% (HRT of 8 h) and 80% (HRT of 4 h), while energy consumption (HRT of 4 h) was lower than that of HRT of 12 h with single electrode assembly under comparable DE. The PFB-BER with higher removal capacity, lower internal resistance and energy consumption provides a new solution to treat the high loading azo dye-containing wastewaters. Copyright © 2015. Published by Elsevier Ltd.

  7. GoAmazon2014/15. Oxidation Flow Reactor Final Campaign Report

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, J. L. [Univ. of Colorado, Boulder, CO (United States); Day, D. A. [Univ. of Colorado, Boulder, CO (United States); Hu, W. [Univ. of Colorado, Boulder, CO (United States); Palm, B. B. [Univ. of Colorado, Boulder, CO (United States); Campuzano-Jost, P. [Univ. of Colorado, Boulder, CO (United States)

    2016-03-01

    The primary goal of the Green Ocean Amazon (GoAmazon2014/5) field campaign was to measure and mechanistically understand the formation of particle number and mass in a region affected by large tropical rainforest biogenic emissions and sometimes anthropogenic influence from a large urban center. As part of the two intensive operational periods (IOPs) and in collaboration with Pacific Northwest National Laboratory (PNNL) and Harvard, the Jimenez Group proposed to deploy a High-Resolution Time-of-Flight Aerosol Mass Spectrometer (HR-ToF-AMS), Thermal Denuder (TD), Scanning Mobility Particle Size (SMPS), two oxidation flow reactors (OFR; including supporting O3, CO/CO2/CH4, RH analyzers), and a high volume filter sampler (MCV) for the measurement of gas and aerosol chemical, physicochemical, and volatility properties. The two IOPs were conducted during the wet season (February to March, 2014) and dry season (August to October, 2014). This proposal was part of a collaborative proposal involving other university and government laboratories.

  8. An investigation of sulfate production in clouds using a flow-through chemical reactor model approach

    Science.gov (United States)

    Hong, M. S.; Carmichael, G. R.

    1983-01-01

    A flow-through chemical reactor model is developed to describe the mass transfer and chemical processes that atmospheric gases undergo in clouds. The model includes the simultaneous absorption of SO2, NH3, O3, NO(x), HNO3, CO2 and H2O2, the accompanying dissociation and oxidation reactions in cloud water, considers electrical neutrality, and includes qualitative parameterization of cloud microphysics. The model is used to assess the importance of the oxidation reactions H2O2-S(IV), O3-S(IV), and S(IV)-Mn(2+) catalysis, and the effects of cloud parameters such as drop size, rain intensity, liquid water content, and updraft velocity. Both precipitating and nonprecipitating clouds are studied. Model results predict sulfate production rates varying from 3 percent/hr to 230 percent/hr. The actual rate is highly dependent on the chemical composition of the uptake air and the physical conditions of the cloud. Model results also show that both the H2O2 and the O3 oxidation reactions can be significant.

  9. Time-Resolved Kinetic Chirped-Pulse Rotational Spectroscopy in a Room-Temperature Flow Reactor.

    Science.gov (United States)

    Zaleski, Daniel P; Harding, Lawrence B; Klippenstein, Stephen J; Ruscic, Branko; Prozument, Kirill

    2017-12-11

    Chirped-pulse Fourier transform millimeter-wave spectroscopy is a potentially powerful tool for studying chemical reaction dynamics and kinetics. Branching ratios of multiple reaction products and intermediates can be measured with unprecedented chemical specificity; molecular isomers, conformers, and vibrational states have distinct rotational spectra. Here we demonstrate chirped-pulse spectroscopy of vinyl cyanide photoproducts in a flow tube reactor at ambient temperature of 295 K and pressures of 1-10 μbar. This in situ and time-resolved experiment illustrates the utility of this novel approach to investigating chemical reaction dynamics and kinetics. Following 193 nm photodissociation of CH2CHCN, we observe rotational relaxation of energized HCN, HNC, and HCCCN photoproducts with 10 μs time resolution and sample the vibrational population distribution of HCCCN. The experimental branching ratio HCN/HCCCN is compared with a model based on RRKM theory using high-level ab initio calculations, which were in turn validated by comparisons to Active Thermochemical Tables enthalpies.

  10. Improved nonlinear fault detection strategy based on the Hellinger distance metric: Plug flow reactor monitoring

    KAUST Repository

    Harrou, Fouzi

    2017-03-18

    Fault detection has a vital role in the process industry to enhance productivity, efficiency, and safety, and to avoid expensive maintenance. This paper proposes an innovative multivariate fault detection method that can be used for monitoring nonlinear processes. The proposed method merges advantages of nonlinear projection to latent structures (NLPLS) modeling and those of Hellinger distance (HD) metric to identify abnormal changes in highly correlated multivariate data. Specifically, the HD is used to quantify the dissimilarity between current NLPLS-based residual and reference probability distributions obtained using fault-free data. Furthermore, to enhance further the robustness of these methods to measurement noise, and reduce the false alarms due to modeling errors, wavelet-based multiscale filtering of residuals is used before the application of the HD-based monitoring scheme. The performances of the developed NLPLS-HD fault detection technique is illustrated using simulated plug flow reactor data. The results show that the proposed method provides favorable performance for detection of faults compared to the conventional NLPLS method.

  11. Recovery of partial nitrification in a down-flow hanging sponge reactor by salt shock loading.

    Science.gov (United States)

    Matsubayashi, Miri; Harada, Hideki; Okubo, Tsutomu; Uemura, Shigeki

    2016-01-01

    Partial nitrification of ammonium-containing artificial wastewater was achieved using a down-flow hanging sponge (DHS) reactor with a hydraulic retention time (HRT) of 2 h by adjusting the influent salinity to 25 g Cl L(-1) with NaCl. The effect of HRT on partial nitrification was examined by varying HRT from 1 to 4 h. Extending HRT from 2 to 4 h had the effect of decreasing nitrite production and increasing nitrates. Since partial nitrification was not completely recovered after returning the HRT to 2 h, we examined the effect of salt shock loading on the recovery of partial nitrification. Salt shock loading with 150 gCl L(-1) for 72 h resulted in the fraction of NO2-N to total inorganic nitrogen in the effluent reaching 83.0% as much as 83 days after returning the salinity to the original level. Thus, despite the time required for the restoration of partial nitrification, the effectiveness of salt shock loading to achieve this aim was verified.

  12. Scaling the Shear-flow Stabilized Z-pinch to Reactor Conditions

    Science.gov (United States)

    McLean, H. S.; Schmidt, A.; Shumlak, U.; Nelson, B. A.; Golingo, R. P.; Cleveau, E.

    2015-11-01

    We present a conceptual design along with scaling calculations for a pulsed fusion reactor based on the shear-flow-stabilized Z-pinch device. Experiments performed on the ZaP device, at the University of Washington, have demonstrated stable operation for durations of 20 usec at ~100kA discharge current for pinches that are ~1 cm in diameter and 100 cm long. The inverse of the pinch diameter and plasma energy density scale strongly with pinch current and calculations show that maintaining stabilization durations of ~7 usec for increased discharge current (~15x) in a shortened pinch (10 cm) results in a pinch diameter of ~200 um and plasma conditions that approach those needed to support significant fusion burn and energy gain (Ti ~ 30keV, density ~ 3e26/m3, ntau ~1.4e20 sec/m3). Compelling features of the concept include operation at modest discharge current (1.5 MA) and voltage (40kV) along with direct adoption of liquid metals for at least one electrode--technological capabilities that have been proven in existing, commercial, pulse power devices such as large ignitrons. LLNL-ABS-674920. This work performed under the auspices of the U.S. Department of Energy ARPAe ALPHA Program by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

  13. Preliminary study on aerobic granular biomass formation with aerobic continuous flow reactor

    Science.gov (United States)

    Yulianto, Andik; Soewondo, Prayatni; Handajani, Marissa; Ariesyady, Herto Dwi

    2017-03-01

    A paradigm shift in waste processing is done to obtain additional benefits from treated wastewater. By using the appropriate processing, wastewater can be turned into a resource. The use of aerobic granular biomass (AGB) can be used for such purposes, particularly for the processing of nutrients in wastewater. During this time, the use of AGB for processing nutrients more reactors based on a Sequencing Batch Reactor (SBR). Studies on the use of SBR Reactor for AGB demonstrate satisfactory performance in both formation and use. SBR reactor with AGB also has been applied on a full scale. However, the use use of SBR reactor still posses some problems, such as the need for additional buffer tank and the change of operation mode from conventional activated sludge to SBR. This gives room for further reactor research with the use of a different type, one of which is a continuous reactor. The purpose of this study is to compare AGB formation using continuous reactor and SBR with same operation parameter. Operation parameter are Organic Loading Rate (OLR) set to 2,5 Kg COD/m3.day with acetate as substrate, aeration rate 3 L/min, and microorganism from Hospital WWTP as microbial source. SBR use two column reactor with volumes 2 m3, and continuous reactor uses continuous airlift reactor, with two compartments and working volume of 5 L. Results from preliminary research shows that although the optimum results are not yet obtained, AGB can be formed on the continuous reactor. When compared with AGB generated by SBR, then the characteristics of granular diameter showed similarities, while the sedimentation rate and Sludge Volume Index (SVI) characteristics showed lower yields.

  14. Coolant and ambient temperature control for chillerless liquid cooled data centers

    Energy Technology Data Exchange (ETDEWEB)

    Chainer, Timothy J.; David, Milnes P.; Iyengar, Madhusudan K.; Parida, Pritish R.; Simons, Robert E.

    2017-08-29

    Cooling control methods and systems include measuring a temperature of air provided to one or more nodes by an air-to-liquid heat exchanger; measuring a temperature of at least one component of the one or more nodes and finding a maximum component temperature across all such nodes; comparing the maximum component temperature to a first and second component threshold and comparing the air temperature to a first and second air threshold; and controlling a proportion of coolant flow and a coolant flow rate to the air-to-liquid heat exchanger and the one or more nodes based on the comparisons.

  15. Experimental investigation of thermoelectric power generation versus coolant pumping power in a microchannel heat sink

    DEFF Research Database (Denmark)

    Kolaei, Alireza Rezania; Rosendahl, Lasse; Andreasen, Søren Juhl

    2012-01-01

    The coolant heat sinks in thermoelectric generators (TEG) play an important role in order to power generation in the energy systems. This paper explores the effective pumping power required for the TEGs cooling at five temperature difference of the hot and cold sides of the TEG. In addition......, the temperature distribution and the pressure drop in sample microchannels are considered at four sample coolant flow rates. The heat sink contains twenty plate-fin microchannels with hydraulic diameter equal to 0.93 mm. The experimental results show that there is a unique flow rate that gives maximum net...

  16. Operating limits Hanford Production Reactors. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Owsley, G.F. [comp.

    1963-05-20

    This report is applicable to the eight operating production reactors, B, C, D, DR, F, H, KE, and KW. It covers the following: operating parameter limitations; reactivity limitations; control and safety systems; reactor fuel loading; coolant requirements with irradiated fuel in reactor; reactor confinement; test facilities; code compliance; safety instrumentation and set points; and control criteria. Also discussed are administrative procedures for process control, training, audits and inspection, and reports and records.

  17. An introduction to the engineering of fast nuclear reactors

    CERN Document Server

    Judd, Anthony M

    2014-01-01

    An invaluable resource for both graduate-level engineering students and practising nuclear engineers who want to expand their knowledge of fast nuclear reactors, the reactors of the future! This book is a concise yet comprehensive introduction to all aspects of fast reactor engineering. It covers topics including neutron physics; neutron flux spectra; flux distribution; Doppler and coolant temperature coefficients; the performance of ceramic and metal fuels under irradiation, structural changes, and fission-product migration; the effects of irradiation and corrosion on structural materials, irradiation swelling; heat transfer in the reactor core and its effect on core design; coolants including sodium and lead-bismuth alloy; coolant circuits; pumps; heat exchangers and steam generators; and plant control. The book includes new discussions on lead-alloy and gas coolants, metal fuel, the use of reactors to consume radioactive waste, and accelerator-driven subcritical systems.

  18. Refurbishment of the IEAR1 primary coolant system piping supports

    Energy Technology Data Exchange (ETDEWEB)

    Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Mattar Neto, Miguel, E-mail: gfainer@ipen.br, E-mail: afaloppa@ipen.br, E-mail: calberto@ipen.br, E-mail: mmattar@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    A partial replacement of the IEA-R1 piping system was concluded in 2014. This paper presents the study and the structural analysis of the IEA-R1 primary circuit piping supports, considering all the changes involved in the replacement. The IEA-R1 is a nuclear reactor for research purposes designed by Babcox-Willcox that is operated by IPEN since 1957. The reactor life management and modernization program is being conducted for the last two decades and already resulted in a series of changes, especially on the reactor coolant system. This set of components, divided in primary and secondary circuit, is responsible for the circulation of water into the core to remove heat. In the ageing management program that includes regular inspection, some degradation was observed in the primary piping system. As result, the renewing of the piping system was conducted in 2014. Moreover the poor condition of some original piping supports gave rise to the refurbishment of all piping supports. The aim of the present work is to review the design of the primary system piping supports taking into account the current conditions after the changes and refurbishment. (author)

  19. Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

    Science.gov (United States)

    2006-12-01

    geothermal , are far from being viable options due to limited supply and economic considerations. For these reasons, it is reasonable to expect that...1 Solution Attack 5 Corrosion Erosion 2 Direct Alloying 6 Thermal Gradient Transfer 3 Inter-angular Penetration 7 Concentration Gradient Transfer...Fundamental Issues In LBE Corrosion”, 2001, Los Alamos, New Mexico , USA: Los Alamos National Laboratory, LA-UR- 04-0869. 71 INITIAL DISTRIBUTION

  20. Development of Reactor Coolant Pump for APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Sang-Youn; Chu, Sung-Min; Chang, Jin-Young [Doosan Heavy Industries and Construction, Changwon (Korea, Republic of)

    2015-10-15

    The development was focused on the performance requirements for APR1400 and to achieve the goals of the safety, reliability and adaptability for APR1400 system design. In addition, APR1400 RCP design was customized considering convenience of installation, operation and maintainability. This paper describes the details of the development process, improved design feature and type test results. Based on development of core technology of RCP, DOOSAN supplies the localized and improved APR1400 RCP to Shin-Hanul 1 and 2 Project. This would be good experience that the RCP core technology can break foreign monopoly in supplying the domestic nuclear industry. Also, there expect APR1400 RCP can be sustainable revenue models in nuclear industry. Moreover, development of RCP will be a catalyst to enhance design capacity for equipment and system of nuclear power plant as well as evaluation and verification skills of Korean nuclear industry.

  1. Transesterification by lipase entrapped in electrospun poly(vinyl alcohol) fibers and its application to a flow-through reactor.

    Science.gov (United States)

    Sakai, Shinji; Antoku, Koichi; Yamaguchi, Tetsu; Kawakami, Koei

    2008-06-01

    We entrapped lipase in electrospun poly(vinyl alcohol) fibers of approximately 1 mum in diameter and evaluated the transesterification activity by converting (s)-glycidol to glycidyl n-butyrate with vinyl n-butyrate. The initial transesterification rate of the entrapped lipase was 5.2-fold faster than that of non-treated lipase. The fibrous membrane could be used as a component of a flow-through reactor for continuous transesterification.

  2. Continuous flowing micro-reactor for aqueous reaction at temperature higher than 100 °C

    OpenAIRE

    Xie, Fei; Wang, Baojun; Wang, Wei; Dong, Tian; Tong, Jianhua; Xia, Shanhong; Wu, Wengang; Li, Zhihong

    2013-01-01

    Some aqueous reactions in biological or chemical fields are accomplished at a high temperature. When the reaction temperature is higher than 100 °C, an autoclave reactor is usually required to elevate the boiling point of the water by creating a high-pressure environment in a closed system. This work presented an alternative continuous flowing microfluidic solution for aqueous reaction with a reaction temperature higher than 100 °C. The pressure regulating function was successfully fulfilled ...

  3. A novel reverse flow reactor coupling endothermic and exothermic reactions: an experimental study

    NARCIS (Netherlands)

    van Sint Annaland, M.; Nijssen, R.C.

    2002-01-01

    A new reactor concept is studied for highly endothermic heterogeneously catalysed gas phase reactions at high temperatures with rapid but reversible catalyst deactivation. The reactor concept aims to achieve an indirect coupling of energy necessary for endothermic reactions and energy released by

  4. THE IMPACT OF POWER COEFFICIENT OF REACTIVITY ON CANDU 6 REACTORS

    OpenAIRE

    D. KASTANYA; S. BOYLE; Hopwood, J; Park, Joo Hwan

    2013-01-01

    The combined effects of reactivity coefficients, along with other core nuclear characteristics, determine reactor core behavior in normal operation and accident conditions. The Power Coefficient of Reactivity (PCR) is an aggregate indicator representing the change in reactor core reactivity per unit change in reactor power. It is an integral quantity which captures the contributions of the fuel temperature, coolant void, and coolant temperature reactivity feedbacks. All nuclear reactor design...

  5. Neutronic analysis of a dual He/LiPb coolant breeding blanket for DEMO

    OpenAIRE

    Catalán, J.P.; Ogando Serrano, Francisco; Sanz Gonzalo, Javier; Palermo, I.; Veredas, G.; Gómez Ros, J.M.; Sedano, L.

    2010-01-01

    A conceptual design of a DEMO fusion reactor is being developed under the Spanish Breeding Blanket Technology Programme: TECNO_FUS based on a He/LiPb dual coolant blanket as reference design option. The following issues have been analyzed to address the demonstration of the neutronic reliability of this conceptual blanket design: power amplification capacity of the blanket, tritium breeding capability for fuel self-sufficiency, power deposition due to nuclear heating in superconducting coils ...

  6. A Well-Posed Two Phase Flow Model and its Numerical Solutions for Reactor Thermal-Fluids Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kadioglu, Samet Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Berry, Ray [Idaho National Lab. (INL), Idaho Falls, ID (United States); Martineau, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    A 7-equation two-phase flow model and its numerical implementation is presented for reactor thermal-fluids applications. The equation system is well-posed and treats both phases as compressible flows. The numerical discretization of the equation system is based on the finite element formalism. The numerical algorithm is implemented in the next generation RELAP-7 code (Idaho National Laboratory (INL)’s thermal-fluids code) built on top of an other INL’s product, the massively parallel multi-implicit multi-physics object oriented code environment (MOOSE). Some preliminary thermal-fluids computations are presented.

  7. WATER BOILER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1960-11-22

    As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

  8. One pass core design of a super fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Qingjie; Oka, Yoshiaki [Cooperative Major in Nuclear Energy, Waseda University, Tokyo 169-8555 (Japan)

    2013-07-01

    One pass core design for Supercritical-pressure light water-cooled fast reactor (Super FR) is proposed. The whole core is cooled with upward flow in one through flow pattern like PWR. Compared with the previous two pass core design; this new flow pattern can significantly simplify the core concept. Upper core structure, coolant flow scheme as well as refueling procedure are as simple as in PWR. In one pass core design, supercritical-pressure water is at approximately 25.0 MPa and enters the core at 280 C. degrees and is heated up in one through flow pattern upwardly to the average outlet temperature of 500 C. degrees. Great density change in vertical direction can cause significant axial power offset during the cycle. Meanwhile, Pu accumulated in the UO{sub 2} fuel blanket assemblies also introduces great power increase during cycle, which requires large amount of flow for heat removal and makes the outlet temperature of blanket low at the beginning of equilibrium cycle (BOEC). To deal with these issues, some MOX fuel is applied in the bottom region of the blanket assembly. This can help to mitigate the power change in blanket due to Pu accumulation and to increase the outlet temperature of the blanket during cycle. Neutron transport and thermohydraulics coupled calculation shows that this design can satisfy the requirement in the Super FR principle for both 500 C. degrees outlet temperature and negative coolant void reactivity. (authors)

  9. ENVIRONMENTALLY REDUCING OF COOLANTS IN METAL CUTTING

    Directory of Open Access Journals (Sweden)

    Veijo KAUPPINEN

    2012-11-01

    Full Text Available Strained environment is a global problem. In metal industries the use of coolant has become more problematic in terms of both employee health and environmental pollution. It is said that the use of coolant forms approximately 8 - 16 % of the total production costs.The traditional methods that use coolants are now obviously becoming obsolete. Hence, it is clear that using a dry cutting system has great implications for resource preservation and waste reduction. For this purpose, a new cooling system is designed for dry cutting. This paper presents the new eco-friendly cooling innovation and the benefits gained by using this method. The new cooling system relies on a unit for ionising ejected air. In order to compare the performance of using this system, cutting experiments were carried out. A series of tests were performed on a horizontal turning machine and on a horizontal machining centre.

  10. Performance of staged and non-staged up-flow anaerobic sludge bed (USSB and UASB) reactors treating low strength complex wastewater.

    Science.gov (United States)

    Sevilla-Espinosa, Susana; Solórzano-Campo, Maricela; Bello-Mendoza, Ricardo

    2010-09-01

    The use of anaerobic processes to treat low-strength wastewater has been increasing in recent years due to their favourable performance-costs balance. For optimal results, it is necessary to identify reactor configurations that are best suited for this kind of application. This paper reports on the comparative study carried out with two high-rate anaerobic reactor systems with the objective of evaluating their performances when used for the treatment of low-strength, complex wastewater. One of the systems is the commonly used up-flow anaerobic sludge blanket (UASB) reactor. The other is the up-flow staged sludge bed (USSB) system in which the reactor was divided longitudinally into 3, 5 and 7 compartments by the use of baffles. The reactors (9 l) were fed with a synthetic, soluble and colloidal waste (chemical oxygen demand (COD) flow hydraulics, between plug-flow and completely-mixed, in the UASB and 7 stages USSB reactors allowed efficient degradation of substrates with minimum effluent concentrations. Low number of compartments in the USSB reactors increased the levels of short-circuiting thus reducing substrate removal efficiencies. All reactors showed high COD removal efficiencies (93-98%) and thus can be regarded as suitable for the treatment of low strength, complex wastewater. Staged anaerobic reactors can be a good alternative for this kind of application provided they are fitted with a large enough (> or =7) number of compartments to fully take advantage of their strengths. Scale factors seem to have influenced importantly on the comparison between one and multi staged sludge-bed reactors and, therefore, observations made here could change at larger reactor volumes.

  11. Ethanol and toluene removal in a horizontal-flow anaerobic immobilized biomass reactor in the presence of sulfate.

    Science.gov (United States)

    Cattony, E B M; Chinalia, F A; Ribeiro, R; Zaiat, M; Foresti, E; Varesche, M B A

    2005-07-20

    In this study it is reported the operation of a horizontal-flow anaerobic immobilized biomass (HAIB) reactor under sulfate-reducing condition which was also exposed to different amounts of ethanol and toluene. The system was inoculated with sludge taken from up-flow anaerobic sludge blanket (UASB) reactors treating refuses from a poultry slaughterhouse. The HAIB reactor comprised of an immobilized biomass on polyurethane foam and ferrous and sodium sulfate solutions were used (91 and 550 mg/L, respectively), to promote a sulfate-reducing environment. Toluene was added at an initial concentration of 2.0 mg/L followed by an increased range of different amendments (5, 7, and 9 mg/L). Ethanol was added at an initial concentration of 170 mg/L followed by an increased range of 960 mg/L. The reactor was operated at 30(+/-2) degrees C with hydraulic detention time of 12 h. Organic matter removal efficiency was close to 90% with a maximum toluene degradation rate of 0.06 mg(toluene)/mg(vss)/d. Sulfate reduction was close to 99.9% for all-nutritional amendments. Biofilm microscopic characterization revealed a diversity of microbial morphologies and DGGE-profiling showed a variation of bacterial and sulfate reducing bacteria (SRB) populations, which were significantly associated with toluene amendments. Diversity of archaea remained unaltered during the different phases of this experiment. Thus, this study demonstrates that compact units of HAIB reactors, under sulfate reducing conditions, are a potential alternative for in situ aromatics bioremediation.

  12. New experimental device for VHTR structural material testing and helium coolant chemistry investigation - High Temperature Helium Loop in NRI Rez

    Energy Technology Data Exchange (ETDEWEB)

    Berka, Jan, E-mail: bej@cvrez.cz [Research Centre Rez, Ltd, Husinec-Rez 130, 25068 Rez (Czech Republic); Institute of Chemical Technology Prague, Technicka 1905, 16628 Prague 6 (Czech Republic); Matecha, Josef, E-mail: josef.matecha@ujv.cz [Nuclear Research Institute Rez plc., Husinec-Rez 130, 25068 Rez (Czech Republic); Cerny, Michal [Institute of Chemical Technology Prague, Technicka 1905, 16628 Prague 6 (Czech Republic); Viden, Ivan, E-mail: ivan.viden@vscht.cz [Institute of Chemical Technology Prague, Technicka 1905, 16628 Prague 6 (Czech Republic); Sus, Frantisek [Research Centre Rez, Ltd, Husinec-Rez 130, 25068 Rez (Czech Republic); Nuclear Research Institute Rez plc., Husinec-Rez 130, 25068 Rez (Czech Republic); Hajek, Petr [Nuclear Research Institute Rez plc., Husinec-Rez 130, 25068 Rez (Czech Republic)

    2012-10-15

    The High Temperature Helium Loop (HTHL) is an experimental device for simulation of VHTR helium coolant conditions. The purpose of the HTHL is structural materials testing and helium coolant chemistry investigation. In the HTHL pure helium will be used as working medium and its main physical parameters are 7 MPa, max. temperature in the test section 900 Degree-Sign C and flow rate 37.8 kg/h. The HTHL consists of an active channel, the helium purification system, the system of impurities dosage (e.g. CO, CO{sub 2}, H{sub 2}, H{sub 2}O, O{sub 2}, N{sub 2}, and CH{sub 4}) and the helium chemistry monitoring system (sampling and on-line analysis and determination of impurities in the helium flow). The active channel is planned to be placed into the core of the experimental reactor LVR-15 which will serve as a neutron flux source (max. 2.5 Multiplication-Sign 10{sup 18} n/m{sup 2} s for fast neutrons). The HTHL is now under construction. Some of its main parts are finished, some are still being produced (active channel internals, etc.), some should be improved to work correctly (the helium circulatory compressor); certain sub-systems are planned to be integrated to the loop (systems for the determination of moisture and other impurities in helium, etc.). The start of the HTHL operation is expected during 2011 and the integration of the active channel into the LVR-15 core during 2012.

  13. Continuous flowing micro-reactor for aqueous reaction at temperature higher than 100 °C.

    Science.gov (United States)

    Xie, Fei; Wang, Baojun; Wang, Wei; Dong, Tian; Tong, Jianhua; Xia, Shanhong; Wu, Wengang; Li, Zhihong

    2013-01-01

    Some aqueous reactions in biological or chemical fields are accomplished at a high temperature. When the reaction temperature is higher than 100 °C, an autoclave reactor is usually required to elevate the boiling point of the water by creating a high-pressure environment in a closed system. This work presented an alternative continuous flowing microfluidic solution for aqueous reaction with a reaction temperature higher than 100 °C. The pressure regulating function was successfully fulfilled by a small microchannel based on a delicate hydrodynamic design. Combined with micro heater and temperature sensor that integrated in a single chip by utilizing silicon-based microfabrication techniques, this pressure regulating microchannel generated a high-pressure/high-temperature environment in the upstream reaction zone when the reagents continuously flow through the chip. As a prel