WorldWideScience

Sample records for water test loop

  1. Reactor recirculation pump test loop

    International Nuclear Information System (INIS)

    Taka, Shusei; Kato, Hiroyuki

    1979-01-01

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

  2. Performance test of ex-core high temperature and high pressure water loop test equipment (Contract research)

    International Nuclear Information System (INIS)

    Nakano, Hiroko; Uehara, Toshiaki; Takeuchi, Tomoaki; Shibata, Hiroshi; Nakamura, Jinichi; Matsui, Yoshinori; Tsuchiya, Kunihiko

    2016-03-01

    In Japan Atomic Energy Agency, we started research and development so as to monitor the situations in the Nuclear Plant Facilities during a severe accident, such as a radiation-resistant monitoring camera, a radiation-resistant transmission system for conveying the in-core information, and a heat-resistant signal cable. As a part of developments of the heat-resistant signal cable, we prepared ex-core high-temperature and high-pressure water loop test equipment, which can simulate the conditions of BWRs and PWRs, for evaluating reliability and properties of sheath materials of the cable. This equipment consists of autoclave, water conditioning tank, high-pressure metering pump, preheater, heat exchanger and water purification equipment, etc. This report describes the basic design and the performance test results of ex-core high-temperature and high-pressure water loop test equipment. (author)

  3. Water loop for training

    International Nuclear Information System (INIS)

    Moeller, S.V.

    1983-02-01

    The procedures used to operate the water loop of the Institute of Nuclear Enginering (IEN) in Brazil are presented. The aim is to help future operators of the training water loop in the operation technique and in a better comprehension of the phenomena occured during the execution of an experience. (E.G.) [pt

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

    International Nuclear Information System (INIS)

    Gerstner, Douglas M.

    2009-01-01

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

  5. High pressure experimental water loop

    International Nuclear Information System (INIS)

    Grenon, M.

    1958-01-01

    A high pressure experimental water loop has been made for studying the detection and evolution of cladding failure in a pressurized reactor. The loop has been designed for a maximum temperature of 360 deg. C, a maximum of 160 kg/cm 2 and flow rates up to 5 m 3 /h. The entire loop consists of several parts: a main circuit with a canned rotor circulation pump, steam pressurizer, heating tubes, two hydro-cyclones (one de-gasser and one decanter) and one tubular heat exchanger; a continuous purification loop, connected in parallel, comprising pressure reducing valves and resin pots which also allow studies of the stability of resins under pressure, temperature and radiation; following the gas separator is a gas loop for studying the recombination of the radiolytic gases in the steam phase. The preceding circuits, as well as others, return to a low pressure storage circuit. The cold water of the low pressure storage flask is continuously reintroduced into the high pressure main circuit by means of a return pump at a maximum head of 160 kg /cm 2 , and adjusted to the pressurizer level. This loop is also a testing bench for the tight high pressure apparatus. The circulating pump and the connecting flanges (Oak Ridge type) are water-tight. The feed pump and the pressure reducing valves are not; the un-tight ones have a system of leak recovery. To permanently check the tightness the circuit has been fitted with a leak detection system (similar to the HRT one). (author) [fr

  6. Irradiation of UO2 specimens with molten cores in a pressurized water loop. Test X-2-x

    International Nuclear Information System (INIS)

    Bain, A.S.

    1961-08-01

    Two Zircaloy-2 clad specimens containing stoichiometric UO 2 pellets were irradiated in a pressurized water loop for 379 hours at heat ratings sufficient to cause central melting of the UO 2 . There was no appearance of localized overheating or accelerated corrosion of the sheath, but the diametral increases were considerably larger than those observed in loop specimens irradiated at lower heat ratings. The length increases, however, were approximately the same as those measured for specimens at lower ratings. There was a clearly visible demarcation between UO 2 that had been molten and that which had not. The value of ∫ 500 o C Tm kdθ = 74 ± W/cm was essentially the same as that obtained from the short-duration tests in the Hydraulic Rabbit, indicating there is no marked decrease in thermal conductivity of the UO 2 fuel in irradiations up to 379 hours. (author)

  7. Gas Test Loop Booster Fuel Hydraulic Testing

    International Nuclear Information System (INIS)

    Gas Test Loop Hydraulic Testing Staff

    2006-01-01

    The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3

  8. Gas Test Loop Booster Fuel Hydraulic Testing

    Energy Technology Data Exchange (ETDEWEB)

    Gas Test Loop Hydraulic Testing Staff

    2006-09-01

    The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3.

  9. Development of sub-channel/system coupled code and its application to a supercritical water-cooled test loop

    International Nuclear Information System (INIS)

    Liu, X.J.; Yang, T.; Cheng, X.

    2014-01-01

    To analyze the local thermal-hydraulic parameters in the supercritical water reactor-fuel qualification test (SCWR-FQT) fuel bundle with a flow blockage, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code and system code are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the coupled code, the whole system behavior including safety system characteristic is analyzed by system code ATHLET-SC, whereas the local thermal-hydraulic parameters are predicted by the sub-channel code COBRA-SC. Sensitivity analysis are carried out respectively in ATHLET-SC and COBRA-SC code, to identify the appropriate models for description of the flow blockage phenomenon in the test loop. Some measures to mitigate the accident consequence are also trialed to demonstrate their effectiveness. The results indicate that the new developed code has good feasibility to transient analysis of supercritical water-cooled test. And the peak cladding temperature caused by blockage in the fuel assembly can be reduced effectively by the safety measures of SCWR-FQT. (author)

  10. UPTF loop seal tests and their RELAP simulation

    International Nuclear Information System (INIS)

    Tuomainen, M.; Tuunanen, J.

    1997-01-01

    In a pressurized water reactor the loop seals have an effect on the natural circulation. If a loop seal is filled with water it can cause a flow stagnation in the loop during two-phase natural circulation. Also the pressure loss over a filled loop seal is high, which lowers the water level in the core. Tests to investigate the loop seal behaviour were performed on a German Upper Plenum Test Facility (UPTF). The purpose of the tests was to study the amount of water in the loop seal under different steam flow rates. The tests were simulated with RELAP5/MOD3.2. With high steam flow rates the code had problems in simulating the amount of the water remaining in the pump elbow, but in general the agreement between the calculated results and the experimental data was good. (orig.)

  11. A code to study the water flow in a thermal test loop

    International Nuclear Information System (INIS)

    Saunier, Jean-Pierre; Duffourt, Nicole; Lago, Bernard

    1965-01-01

    A first part reports the theoretical and analytical formulation of a flow within a specific circuit used in a thermal test installation. Equations in the different parts of the circuit are developed, and their resolution for integration into a computation code is described, including boundary conditions, constants and input functions (cell characteristics, fluid characteristics, heat transfer, friction, time slicing). The second part reports an extension of this theoretical and analytical development and code development to a two-branch circuit

  12. Dynamic response of IPEN experimental water loop

    International Nuclear Information System (INIS)

    Faya, A.J.G.; Bassel, W.S.

    1982-10-01

    A mathematical model has been developed to analyze the transient thermal response of the I.P.E.N. water loop during change of power operations. The model is capable of estimating the necessary test section power and heat exchanger mass flow rate for a given operating temperature. It can also determine the maximum heating or cooling rate to avoid thermal shocks in pipes and components. (Author) [pt

  13. Analysis of severe accidents on fast reactor test loop

    International Nuclear Information System (INIS)

    Cenerini, R.; Verzelletti, G.; Curioni, S.

    1975-01-01

    The Pec reactor is a sodium cooled fast reactor which is being designed for the primary purpose of accomodating closed sodium cooled test loops for the developmental and proof testing of fast reactor fuel assemblies. The test loops are located in the central test region of reactor. The basic function for which the loop is designed is burn-up to failure testing of fuel under advanced performance conditions. It is therefore necessary to design the loop for failure conditions. Basically two types of accidents can occur within the loops: rupture of gas plenum in the fuel pins and coolant starvation. Explosive tests on Pec loop, whose first set is described in this report, are devoted to investigate the effects of an accidental energy release on loop containment. The loop model reproduces in the test section the prototype dimensions in radial scale 1:1. Using a wire explosive charge of 300mm, the height of test section is sufficient for determining the containment capability of the loop that has a nearly constant deformation in a length of. 3-4 time the diameter. The inertial effects of the coolant column are reproduced by two tubes at the extremities of test section, closed with top plugs. Some tests has been performed by wrapping around the test section four layers of steel wire in order to evaluate the influence on the containment of tungsten wire that is foreseen in prototype loop. The influence of the coolant around the loop was evaluated by inserting the model in water. Dummy sub-assemblies was used and explosive substitutes the central rods. Piezoelectric pressure transducers were mounted on the three plugs and radial deformation was measured directly at different height. From experiments performed it resulted the importance of harmonic wires and inertial reaction of external water on loop containment; maximum containable energy is about 50 Cal with E.1 explosive

  14. Vibration and acoustic signatures of the water circulation pump in the pressurised LMR fuel element test loop at IPEN

    International Nuclear Information System (INIS)

    Holland, L.

    1985-01-01

    This report presents results of vibration and acoustic field measurements made on the water circulating pump in the IPEN - CNEN/Sao Paulo pressurised water loop. The use of such measurements to monitor the vibration of coolant circulating pumps of light water reactors is indicated. Measurements were made for defined water flows and pressures varying between 5 bar/5.22 ls sup(-1) and 40 bar/17,42ls sup(-1). Analyses of various recordings of two accelerometer signals and 1 microphone signal were made principally in the frequency range 0-5 KHz using a Nicolet 660 A Fourier analyser. Results of these analyses indicate that CPSD distributions might be more sensitive indicators of changes in pump operating conditions than the more frequently used PSD distributions. In addition, as an indicador of changing pump conditions the acoustic-vibration signal pair is perhaps a more sensitive indicator than the vibration-vibration signal pair. While coherence distributions are elearly sensitive to changing pump conditions, trends in the change of these distributions were not readily identified. It is recommended that more detailed analyses be made using pattern recognition techniques in conjunction with frequency zooming. (Author) [pt

  15. Technical specification of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y

    1998-03-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. HANARO fuel test loop was designed in accordance with the same code and standards of nuclear power plant because HANARO FTL will be operated the high pressure and temperature same as nuclear power plant operation conditions. The objective of this study is to confirm the operation limit, safety limit, operation condition and checking points of HANARO fuel test loop. This results will become guidances for the planning of irradiation testing and operation of HANARO fuel test loop. (author). 13 refs., 13 tabs., 8 figs.

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  17. An Investigation of Loop Seal Clearings in ATLAS SBLOCA Tests

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeonsik; Cho, Seok; Kang, Kyoungho; Park, Hyunsik; Min, Kyeongho; Choi, Namhyeon; Park, Jonggook; Kim, Bokdeuk; Choi, Kiyong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    In most of the SBLOCA cases, the pressure of the upper-head region will increase mainly owing to the accumulated steam and water inventory in the upper-plenum. This build-up pressure acts as a suppression force to the core water level, and resultantly the core water level will decrease possibly up to and/or below the top of the active core region. Simultaneously, the downcomer water level will increase owing to the evacuated water inventory from the lower part of the core region. This unbalanced hydro-static pressure between the core and downcomer region acts as a potential pushing force to the reactor coolant pump (RCP) side intermediate leg. The potential pushing force will be increased with time to overcome the hydro-static head in the upflow intermediate leg. The unbalanced hydro-static pressure can finally be dissolved with the occurrence of the loop seal clearing. A minimum core collapsed water level, located below the elevation of the loop seal bottom leg in the ATLAS tests, is taken at this time. Since the loop seal bottom leg is located below the core top for typical PWR plants such as an APR1400, the water level depression may uncover the core upper regions until the core water level recovers with the progress of the clearing of the loop seal upflow leg. At this moment, the core temperature may increase to a peak cladding temperature (PCT) owing to an excessive core uncovery by the minimum core collapsed water level. Therefore, the loop seal clearing phenomenon is very important with respect to the PCT occurrence, which is one of the most important parameters to insure the safety of the reactor system. The loop seal clearing behavior seems to be closely related to the break location and break size. Usually, a loop seal in the break loop is cleared first, and the number of loop seal clearings is dependent on the break size. The larger the break size, the more the loop seals that are cleared. An investigation of LSC in the SBLOCA for DVI line and CL breaks

  18. Water chemistry of the JMTR IASCC irradiation loop system

    International Nuclear Information System (INIS)

    Hanawa, Satoshi; Oogiyanagi, Jin; Mori, Yuichiro; Saito, Junichi; Tsukada, Takashi

    2006-01-01

    Irradiation assisted stress corrosion cracking (IASCC) is recognized as an important degradation issue of the core-internal material for aged Boiling Water Reactors (BWRs). Therefore, irradiation loop system has been developed and installed in the Japan Materials Testing Reactor to perform the IASCC irradiation test. In the IASCC irradiation test, water chemistry of irradiation field is one of the most important key parameters because it affects initiation and propagation of cracks. This paper summarizes the measurement and evaluation method of water chemistry of IASCC irradiation loop system. (author)

  19. Hardware-in-the-Loop Testing

    Data.gov (United States)

    Federal Laboratory Consortium — RTC has a suite of Hardware-in-the Loop facilities that include three operational facilities that provide performance assessment and production acceptance testing of...

  20. Accident analysis of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.; Chi, D. Y

    1998-03-01

    Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. The HANARO fuel test loop was designed to match the CANDU and PWR fuel operating conditions. The accident analysis was performed by RELAP5/MOD3 code based on FTL system designs and determined the detail engineering specification of in-pile test section and out-pile systems. The accident analysis results of FTL system could be used for the fuel and materials designer to plan the irradiation testing programs. (author). 23 refs., 20 tabs., 178 figs.

  1. Operation of the hot test loop facilities

    International Nuclear Information System (INIS)

    Cheong, Moon Ki; Park, Choon Kyeong; Won, Soon Yeon; Yang, Sun Kyu; Cheong, Jang Whan; Cheon, Se Young; Song, Chul Hwa; Jeon, Hyeong Kil; Chang, Suk Kyu; Jeong, Heung Jun; Cho, Young Ro; Kim, Bok Duk; Min, Kyeong Ho

    1994-12-01

    The objective of this project is to obtain the available experimental data and to develop the measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics department have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within fuel bundle and to understand the characteristic of pressure drop required for improving the nuclear fuel and to develop the advanced measuring techniques. RCS Loop, which is used to measure the CHF, is presently under design and construction. B and C Loop is designed and constructed to assess the automatic depressurization safety system behavior. 4 tabs., 79 figs., 7 refs. (Author) .new

  2. Design of a pressurized water loop heated by electric resistances

    International Nuclear Information System (INIS)

    Ribeiro, S.V.G.

    1981-01-01

    A pressurized water loop design is presented. Its operating pressure is 420 psi and we seek to simulate qualitatively some thermo-hydraulic phenomena of PWR reactors. The primary circuit simulator consists basically of two elements: 1)the test section housing 16 electric resistences dissipating a total power of 100 Kw; 2)the loop built of SCH40S 304L steel piping, consisting of the pump, a heat exchanger and the pressurizer. (Author) [pt

  3. Liquid Lead-Bismuth Materials Test Loop

    International Nuclear Information System (INIS)

    Tcharnotskaia, Valentina; Ammerman, Curtt; Darling, Timothy; King, Joe; Li, Ning; Shaw, Don; Snodgrass, Leon; Woloshun, Keith

    2002-01-01

    We designed and built the Liquid Lead-Bismuth Materials Test Loop (MTL) to study the materials behavior in a flow of molten lead-bismuth eutectic (LBE). In this paper we present a description of the loop with main components and their functions. Stress distribution in the piping due to sustained, occasional and expansion loads is shown. The loop is designed so that a difference of 100 deg. C can be attained between the coldest and the hottest parts at a nominal flow rate of 8.84 GPM. Liquid LBE flow can be activated by a mechanical sump pump or by natural convection. In order to maintain a self-healing protective film on the surface of the stainless steel pipe, a certain concentration of oxygen has to be maintained in the liquid metal. We developed oxygen sensors and an oxygen control system to be implemented in the loop. The loop is outfitted with a variety of instruments that are controlled from a computer based data acquisition system. Initial experiments include preconditioning the loop, filling it up with LBE, running at uniform temperature and tuning the oxygen control system. We will present some preliminary results and discuss plans for the future tests. (authors)

  4. WWER type reactor primary loop imitation on large test loop facility in MARIA reactor

    International Nuclear Information System (INIS)

    Moldysh, A.; Strupchevski, A.; Kmetek, Eh.; Spasskov, V.P.; Shumskij, A.M.

    1982-01-01

    At present in Poland in cooperation with USSR a nuclear water loop test facility (WL) in 'MARIA' reactor in Sverke is under construction. The program objective is to investigate processes occuring in WWER reactor under emergency conditions, first of all after the break of the mainprimary loop circulation pipe-line. WL with the power of about 600 kW consists of three major parts: 1) an active loop, imitating the undamaged loops of the WWER reactor; 2) a passive loop assignedfor modelling the broken loop of the WWER reactor; 3) the emergency core cooling system imitating the corresponding full-scale system. The fuel rod bundle consists of 18 1 m long rods. They were fabricated according to the standard WWER fuel technology. In the report some general principles of WWERbehaviour imitation under emergency conditions are given. They are based on the operation experience obtained from 'SEMISCALE' and 'LOFT' test facilities in the USA. A description of separate modelling factors and criteria effects on the development of 'LOCA'-type accident is presented (the break cross-section to the primary loop volume ratio, the pressure differential between inlet and outlet reactor chambers, the pressure drop rate in the loop, the coolant flow rate throuh the core etc.). As an example a comparison of calculated flow rate variations for the WWER-1000 reactor and the model during the loss-of-coolant accident with the main pipe-line break at the core inlet is given. Calculations have been carried out with the use of TECH'-M code [ru

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

  6. French nuclear plant safeguard pump qualification testing: EPEC test loop

    International Nuclear Information System (INIS)

    Guesnon, H.

    1985-01-01

    This paper reviews the specifications to which nuclear power plant safeguard pumps must be qualified, and surveys the qualification methods and program used in France to verify operability of the pump assembly and major pump components. The EPEC test loop is described along with loop capabilities and acheivements up to now. This paper shows, through an example, the Medium Pressure Safety Injection Pump designed for service in 1300 MW nuclear power plants, and the interesting possibilities offered by qualification testing

  7. Functional and performance evaluation of 28 bar hot shutdown passive valve (HSPV) at integral test loop (ITL) for advanced heavy water reactor (AHWR)

    International Nuclear Information System (INIS)

    Sapra, M.K.; Kundu, S.; Pal, A.K.; Sharma, B.S.V.G.

    2007-02-01

    During reactor shutdown in advanced heavy water reactor (AHWR), core decay heat is removed by eight isolation condensers (IC) submerged in gravity driven water pool. Passive valves are provided on the down stream of each isolation condenser. On increase in steam drum pressure beyond a set value, these passive valves start opening and establish steam flow by natural circulation between the four steam drums and corresponding isolation condensers under hot shutdown and therefore they are termed as Hot Shut Down Passive Valves (HSPVs). The HSPV is a self acting type valve requiring no external energy, i.e. neither air nor electric supply for actuation. This feature makes the valve functioning independent of external systems such as compressed air supply or electric power supply, thereby providing inherent safety feature in line with reactor design philosophy. The high pressure and high temperature HSPV s for nuclear reactor use, are non-standard valves and therefore not manufactured by the valve industry worldwide. In the process of design and development of a prototype valve for AHWR, a 28 bar HSPV was configured and successfully tested at Integral Test Loop (ITL) at Engineering Hall No.7. During ten continuous experiments spread over 14 days, the HSPV has proved its functional capabilities and its intended use in decay heat removal system. The in-situ pressure setting and calibration aspect of HSPV has also been successfully established during these experiments. This report gives an insight into the HSPV's functional behavior and role in reactor decay heat removal system. The report not only provides the quantitative measure of performance for 28 bar HSPV in terms of valve characteristics, pressure controllability, linearity and hysteresis but also sets qualitative indicators for prototype 80 bar HSPV, being developed for AHWR. (author)

  8. Optimal tests for electroweak loop effects

    International Nuclear Information System (INIS)

    Aoki, Kenichi; Aoyama, Hideaki; Harvard Univ., Cambridge, MA

    1986-01-01

    A statistical analysis is given for the experimental precision necessary for establishing loop effects in the electroweak theory. Cases with three observables, gauge boson masses and the Weinberg angle, is analyzed by an optimised test. An information on the Weinberg angle with even 5% error (+-.01 in sin 2 thetasub(W)) is shown to reduce the requirement for the measurements of gauge boson masses significantly. (orig.)

  9. Gas Test Loop Functional and Technical Requirements

    International Nuclear Information System (INIS)

    Glen R. Longhurst; Soli T. Khericha; James L. Jones

    2004-01-01

    This document defines the technical and functional requirements for a gas test loop (GTL) to be constructed for the purpose of providing a high intensity fast-flux irradiation environment for developers of advanced concept nuclear reactors. This capability is needed to meet fuels and materials testing requirements of the designers of Generation IV (GEN IV) reactors and other programs within the purview of the Advanced Fuel Cycle Initiative (AFCI). Space nuclear power development programs may also benefit by the services the GTL will offer. The overall GTL technical objective is to provide developers with the means for investigating and qualifying fuels and materials needed for advanced reactor concepts. The testing environment includes a fast-flux neutron spectrum of sufficient intensity to perform accelerated irradiation testing. Appropriate irradiation temperature, gaseous environment, test volume, diagnostics, and access and handling features are also needed. This document serves to identify those requirements as well as generic requirements applicable to any system of this kind

  10. Gas Test Loop Functional and Technical Requirements

    Energy Technology Data Exchange (ETDEWEB)

    Glen R. Longhurst; Soli T. Khericha; James L. Jones

    2004-09-01

    This document defines the technical and functional requirements for a gas test loop (GTL) to be constructed for the purpose of providing a high intensity fast-flux irradiation environment for developers of advanced concept nuclear reactors. This capability is needed to meet fuels and materials testing requirements of the designers of Generation IV (GEN IV) reactors and other programs within the purview of the Advanced Fuel Cycle Initiative (AFCI). Space nuclear power development programs may also benefit by the services the GTL will offer. The overall GTL technical objective is to provide developers with the means for investigating and qualifying fuels and materials needed for advanced reactor concepts. The testing environment includes a fast-flux neutron spectrum of sufficient intensity to perform accelerated irradiation testing. Appropriate irradiation temperature, gaseous environment, test volume, diagnostics, and access and handling features are also needed. This document serves to identify those requirements as well as generic requirements applicable to any system of this kind.

  11. Conceptural design of multipurpose sodium test loop

    International Nuclear Information System (INIS)

    Kim, W.C.; Lee, Y.W.; Nam, H.Y.; Chun, S.Y.; Kim, J.; Yuh, M.W.

    1982-01-01

    This report describes the conceptural design of the multipurpose sodium test loop (MSTL). This MSTL consists mainly of impurity control and measurement system, corrosion and masstransfer system and heat transfer system. Problems associated with liquid sodium coolant will be studied and operating experiences will be obtained by the use of this facility. This technology will be used to evaluate safety and reliability of large sodium facility in the future. The total cost excluding the cost of building construction is estimated to 175 thousand dollars. (Author)

  12. In-pile loop experiments in water chemistry and corrosion

    International Nuclear Information System (INIS)

    Kysela, J.; Jindrich, K.; Masarik, V.; Fric, Z.; Chotivka, V.; Hamerska, H.; Vsolak, R.; Erben, O.

    1986-08-01

    Methods and techniques used were as follows: (a) Method of polarizing resistance for remote monitoring of instantaneous rate of uniform corrosion. (b) Out-of-pile loop at the temperature 350 degC, pressure 19 MPa, circulation 20 kgs/h, testing time 1000 h. (c) High temperature electromagnetic filter with classical solenoid and ball matrix for high pressure filtration tests. (d) High pressure and high temperature in-pile water loop with coolant flow rate 10 000 kgs/h, neutron flux in active channel 7x10 13 n/cm 2 .s, 16 MPa, 330 degC. (e) Evaluation of experimental results by chemical and radiochemical analysis of coolant, corrosion products and corrosion layer on surface. The results of measurements carried out in loop facilities can be summarized into the following conclusions: (a) In-pile and out-of-pile loops are suitable means of investigating corrosion processes and mass transport in the nuclear power plant primary circuit. (b) In studying transport phenomena in the loop, it is necessary to consider the differences in geometry of the loop and the primary circuit, mainly the ratio of irradiated and non-irradiated surfaces and volumes. (c) In the experimental facility simulating the WWER-type nuclear power plant primary circuit, solid suspended particles of a chemical composition corresponding most frequently to magnetite or nickel ferrite, though with non-stoichiometric composition Me x 2+ Fe 3-x 3+ O 4 , were found. (d) Continuous filtration of water by means of an electromagnetic filter removing large particles of corrosion products leads to a decrease in radioactivity of the outer epitactic layer only. The effect of filtration on the inner topotactic layer is negligible

  13. Tests in the ATLE loop on the PIUS design

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  14. Water hammer characteristics of integral pressurized water reactor primary loop

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  15. Water hammer characteristics of integral pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-08-15

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

  16. Summary of ALSEP Test Loop Solvent Irradiation Testing

    Energy Technology Data Exchange (ETDEWEB)

    Peterman, Dean Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Olson, Lonnie Gene [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    Separating the minor actinide elements (americium and curium) from the fission product lanthanides is an important step in closing the nuclear fuel cycle. Isolating the minor actinides will allow transmuting them to short lived or stable isotopes in fast reactors, thereby reducing the long-term hazard associated with these elements. The Actinide Lanthanide Separation Process (ALSEP) is being developed by the DOE-NE Material Recovery and Waste Form Development Campaign to accomplish this separation with a single process. To develop a fundamental understanding of the solvent degradation mechanisms for the ALSEP Process, testing was performed in the INL Radiolysis/Hydrolysis Test Loop for the extraction section of the ALSEP flowsheet. This work culminated in the completion of the level two milestone (M2FT-16IN030102021) "Complete ALSEP test loop solvent irradiation test.” This report summarizes the testing performed and the impact of radiation on the ALSEP Process performance as a function of dose.

  17. High Temperature Fluoride Salt Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Aaron, Adam M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cunningham, Richard Burns [Univ. of Tennessee, Knoxville, TN (United States); Fugate, David L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holcomb, David Eugene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kisner, Roger A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peretz, Fred J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yoder, Jr, Graydon L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    with 3 cm diameter graphite-based fuel pebbles slowly circulating up through the core. Molten salt coolant (FLiBe) at 700°C flows concurrently (at significantly higher velocity) with the pebbles and is used to remove heat generated in the reactor core (approximately 1280 W/pebble), and supply it to a power conversion system. Refueling equipment continuously sorts spent fuel pebbles and replaces spent or damaged pebbles with fresh fuel. By combining greater or fewer numbers of pebble channel assemblies, multiple reactor designs with varying power levels can be offered. The PB-AHTR design is discussed in detail in Reference [1] and is shown schematically in Fig. 1. Fig. 1. PB-AHTR concept (drawing taken from Peterson et al., Design and Development of the Modular PB-AHTR Proceedings of ICApp 08). Pebble behavior within the core is a key issue in proving the viability of this concept. This includes understanding the behavior of the pebbles thermally, hydraulically, and mechanically (quantifying pebble wear characteristics, flow channel wear, etc). The experiment being developed is an initial step in characterizing the pebble behavior under realistic PB-AHTR operating conditions. It focuses on thermal and hydraulic behavior of a static pebble bed using a convective salt loop to provide prototypic fluid conditions to the bed, and a unique inductive heating technique to provide prototypic heating in the pebbles. The facility design is sufficiently versatile to allow a variety of other experimentation to be performed in the future. The facility can accommodate testing of scaled reactor components or sub-components such as flow diodes, salt-to-salt heat exchangers, and improved pump designs as well as testing of refueling equipment, high temperature instrumentation, and other reactor core designs.

  18. NOMAGE4 activities 2011. Part II, Supercritical water loop

    Energy Technology Data Exchange (ETDEWEB)

    Vierstraete, P. (Ecole Nationale Superieure des mines, Paris (France)); Van Nieuwenhove, R. (Institutt for Energiteknikk, OECD Halden Reactor Project (HRP), Kjeller (Norway)); Lauritzen, B. (Technical Univ. of Denmark, Risoe National Lab. for Sustainable Energy, Roskilde (Denmark))

    2012-01-15

    The supercritical water reactor (SCWR) is one of the six different reactor technologies selected for research and development under the Generation IV program. Several countries have shown interest to this concept but up to now, there exist no in-pile facilities to perform the required material and fuel tests. Working on this direction, the Halden Reactor Project has started an activity in collaboration with Risoe-DTU (with Mr. Rudi Van Nieuwenhove as the project leader) to study the feasibility of a SCW loop in the Halden Reactor, which is a Heavy Boiling Water Reactor (HBWR). The ultimate goal of the project is to design a loop allowing material and fuel test studies at significant mass flow with in-core instrumentation and chemistry control possibilities. The present report focusses on the main heat exchanger required for such a loop in the Halden Reactor. The goal of this heat exchanger is to assure a supercritical flow state inside the test section (the core side) and a subcritical flow state inside the pump section. The objective is to design the heat exchanger in order to optimize the efficiency of the heat transfer and to respect several requirements as the room available inside the reactor hall, the maximal total pressure drop allowed and so on. (Author)

  19. NOMAGE4 activities 2011, Part II, Supercritical water loop

    DEFF Research Database (Denmark)

    Vierstraete, Pierre; Van Nieuwenhove, Rudi; Lauritzen, Bent

    The supercritical water reactor (SCWR) is one of the six different reactor technologies selected for research and development under the Generation IV program. Several countries have shown interest to this concept but up to now, there exist no in-pile facilities to perform the required material...... and fuel tests. Working on this direction, the Halden Reactor Project has started an activity in collaboration with Risoe-DTU (with Mr. Rudi Van Nieuwenhove as the project leader) to study the feasibility of a SCW loop in the Halden Reactor, which is a Heavy Boiling Water Reactor (HBWR). The ultimate goal...

  20. Review of design criteria and safety analysis of safety class electric building for fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. HANARO fuel test loop was designed for CANDU and PWR fuel testing. Safety related system of Fuel Test Loop such as emergency cooling water system, component cooling water system, safety ventilation system, high energy line break mitigation system and remote control room was required 1E class electric supply to meet the safety operation in accordance with related code. Therefore, FTL electric building was designed to construction and install the related equipment based on seismic category I. The objective of this study is to review the design criteria and analysis the safety function of safety class electric building for fuel test loop, and this results will become guidance for the irradiation testing in future. (author). 10 refs., 6 tabs., 30 figs.

  1. Flow loop studies with AMAX coal-water mixtures

    Energy Technology Data Exchange (ETDEWEB)

    Wildman, D.J.; Ekmann, J.M.

    1984-03-01

    The coal-water mixtures (CWM) with a stabilizer and the CWM without stabilizers were successfully transported through a flow loop facility under a variety of conditions. The handling characteristics of both CWM were reasonable. The mix tank mixer was not needed during nontesting hours to prevent settling of either material. After several days of transporting the nonstabilized material in the loop facility, the viscosity-reducing agent became ineffective. It was necessary to increase the concentration of the viscosity-reducing agent. The material with stabilizer could not be transported through the loop facility at mass flow rates greater than 209 lb/min until overnight shearing of the CWM in the tank. The CWM without a stabilizer appeared to be slightly shear-thickening, whereas the stabilized CWM initially exhibited shear-thinning behavior. The pressure losses measured for the nonstabilized material were similar to the pressure losses measured for CWM prepared at PETC with three or four percent higher concentration of Pittsburgh seam coal. Tests performed with the stabilized CWM experienced pressure losses similar to CWM prepared at PETC with Pittsburgh seam coal of five to seven percent higher concentration. Tests 1A, 2A, 1B, and 2B were not included in the comparison of in-house-prepared CWM due to differences in pretest handling procedures. 1 figure, 2 tables.

  2. Thermal Interface Evaluation of Heat Transfer from a Pumped Loop to Titanium-Water Thermosyphons

    Science.gov (United States)

    Jaworske, Donald A.; Sanzi, James L.; Gibson, Marc A.; Sechkar, Edward A.

    2009-01-01

    Titanium-water thermosyphons are being considered for use in the heat rejection system for lunar outpost fission surface power. Key to their use is heat transfer between a closed loop heat source and the heat pipe evaporators. This work describes laboratory testing of several interfaces that were evaluated for their thermal performance characteristics, in the temperature range of 350 to 400 K, utilizing a water closed loop heat source and multiple thermosyphon evaporator geometries. A gas gap calorimeter was used to measure heat flow at steady state. Thermocouples in the closed loop heat source and on the evaporator were used to measure thermal conductance. The interfaces were in two generic categories, those immersed in the water closed loop heat source and those clamped to the water closed loop heat source with differing thermal conductive agents. In general, immersed evaporators showed better overall performance than their clamped counterparts. Selected clamped evaporator geometries offered promise.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  4. Helium Loop for the HCPB Test Blanket Module

    International Nuclear Information System (INIS)

    Neuberger, H.; Boccaccini, L.V.; Ghidersa, B. E.; Jin, X.; Meyder, R.

    2006-01-01

    In the frame of the activities of the EU Breeder Blanket Programme and of the Test Blanket Working Group, the Helium loop for the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) in ITER has been investigated with regard to the layout definition, selection of components, control, dimensioning and integration. This paper presents the status of development. The loop design for the HCPB-TBM in ITER will mainly base on the experience gained from Helium Loop Karlsruhe (HELOKA) which is currently developed at the FZK for experiments under ITER relevant conditions. The ITER loop will be equipped with similar components like HELOKA and will mainly consist of a circulator with variable speed drive, a recuperator, an electric heater, a cooler, a dust filter and auxilary components e.g. pipework and valves. A Coolant Purification System (CPS) and a Pressure Control System (PCS) are foreseen to meet the requirements on coolant conditioning. To prepare a TBM for a new experimental campaign, a succession of operational states like '' cold maintenance '', '' baking '' and '' cold standby '' is required. Before a pulse operation, a '' hot stand-by '' state should be achieved providing the TBM with inlet coolant at nominal conditions. This operation modus is continued in the dwell time waiting for the successive pulse. A '' tritium out-gassing '' will be also required after several TBM-campaigns to remove the inventory rest of T in the beds for measurement purpose. The dynamic circuit behaviour during pulses, transition between different operational states as well as the behaviour in accident situations are investigated with RELAP. The main components of the loop will be accommodated inside the Tokamak Cooling Water System(TCWS)- vault from where the pipes require connection to the TBM which is attached to port 16 of the vacuum vessel. Therefore pipes across the ITER- building of about 110 m in length (each) are required. Additional equipment is also located in the port cell

  5. Supercritical CO2 test loop operation and first test results

    International Nuclear Information System (INIS)

    Wright, Steven A.; Pickard, Paul S.

    2009-01-01

    The DOE Office of Nuclear Energy is investigating advanced Brayton cycles for use with next generation nuclear power plants. The focus of this work is on the supercritical CO 2 Brayton cycle which has the potential for high efficiency, and for reduced capital costs due to very compact turbomachinery. Sandia has fabricated and is operating a supercritical CO 2 (S-CO 2 ) test loop to investigate the key technology issues associated with this cycle. This loop is part of a multi-year phased development program to develop a megawatt (MW) class closed S-CO 2 Brayton cycle to demonstrate the applicability of this cycle for DOE Gen-IV program. The current loop has been configured as both a compression loop and as simple heated but unrecuperated Brayton cycle. A second split-flow or re-compression Brayton cycle is currently under development that will use approximately 1 MW of heat to run the Brayton cycle. Early configurations of this split-flow Brayton cycle will be operational later this fiscal year. The key issues for this cycle include the fundamental issues of compressor fluid performance and system control near the critical point, but also the supporting technology issues of bearings, sealing technologies, and rotor windage losses which are also essential to achieving efficiency and cost objectives. These tests are providing the first measurements and information on these key supercritical CO 2 power conversion systems questions. Important data for all these issues has been obtained. This report presents the major results of the testing by showing and comparing the measured compressor performance map with the predicted performance. The compression loop uses a ∼50 kWe motor driven compressor to spin a 37 mm OD compressor at design speeds up to 75,000 rpm with a pressure ratio of 1.8 and a flow rate of 3.53 kg/s for a compressor inlet condition of 305.3 K and 7690 kPa. The most recent configuration of this loop has added a small turbine and 260 kW of heater power is

  6. Xenon oscillation tests in four-loop PWR cores

    International Nuclear Information System (INIS)

    Aoki, Norihiko; Osaka, Kenichi; Shimada, Shoichiro; Tochihara, Hiroshi; Machii, Seigo

    1980-01-01

    The Kansai Electric Power Co.'s OHI Unit 1 and 2 are the first 4-loop PWRs in Japan which use 17 x 17 fuel assemblies and have essentially the same plant parameters. A 4-loop core has larger core radius and higher power density than those of 2- or 3-loop cores, and is less stable for Xe oscillation. It is therefore important to confirm that Xe oscillations in radial direction are sufficiently stable in a 4-loop core. Radial and axial Xe oscillation tests were performed during the startup physics tests of OHI Unit 1 and 2; Xe oscillation was induced by perturbation of control rods and the Xe effect on power distribution observed periodically. The test results show that the transverse Xe oscillation in the 4-loop core is sufficiently stable and that the agreement between the measurement and the calculated prediction is good. (author)

  7. Theory of energy level and its application in water-loop heat pump system

    International Nuclear Information System (INIS)

    Yu, Qi Dong

    2017-01-01

    Highlights: • Novel theory of saving energy and its application in water loop heat pump. • Reverse energy caused by units to water loop and its solution. • New method for determining the energy-saving range of water loop heat pump. • Capacity model of auxiliary heat source and its size for all building types. • Advice for reducing total energy consumption of water loop heat pump. - Abstract: It is a difficult problem to how to determine the reverse energy caused by units to water loop when a water-loop heat pump (WLHP) is in cooling and heating simultaneous mode, which not only has a great impact on energy-saving rate but also decides the use of auxiliary heat source in winter. This paper presents a theory of energy level to improve the research on WLHP system by using the relationship among building, circulating water and units. In this theory, the circulating water replaces building load as a new method to convert the reverse energy into energy change of circulating water and the equation of energy level also is built to determine the energy-saving range of WLHP system and report the capacity model of auxiliary heat source for all building types. An office building with different auxiliary powers is tested to analyze system operation characteristic and the effect of auxiliary heat source on unit and system and the results validate previous conclusions and suggest that an energy balance should be considered between units and auxiliary power to improve overall operation.

  8. Design of Test Loops for Forced Convection Heat Transfer Studies at Supercritical State

    Science.gov (United States)

    Balouch, Masih N.

    Worldwide research is being conducted to improve the efficiency of nuclear power plants by using supercritical water (SCW) as the working fluid. One such SCW reactor considered for future development is the CANDU-Supercritical Water Reactor (CANDU-SCWR). For safe and accurate design of the CANDU-SCWR, a detailed knowledge of forced-convection heat transfer in SCW is required. For this purpose, two supercritical fluid loops, i.e. a SCW loop and an R-134a loop are developed at Carleton University. The SCW loop is designed to operate at pressures as high as 28 MPa, temperatures up to 600 °C and mass fluxes of up to 3000 kg/m2s. The R-134a loop is designed to operate at pressures as high as 6 MPa, temperatures up to 140 °C and mass fluxes in the range of 500-6000 kg/m2s. The test loops designs allow for up to 300 kW of heating power to be imparted to the fluid. Both test loops are of the closed-loop design, where flow circulation is achieved by a centrifugal pump in the SCW loop and three parallel-connected gear pumps in the R-134a loop, respectively. The test loops are pressurized using a high-pressure nitrogen cylinder and accumulator assembly, which allows independent control of the pressure, while simultaneously dampening pump induced pressure fluctuations. Heat exchangers located upstream of the pumps control the fluid temperature in the test loops. Strategically located measuring instrumentation provides information on the flow rate, pressure and temperature in the test loops. The test loops have been designed to accommodate a variety of test-section geometries, ranging from a straight circular tube to a seven-rod bundle, achieving heat fluxes up to 2.5 MW/m2 depending on the test-section geometry. The design of both test loops allows for easy reconfiguration of the test-section orientation relative to the gravitational direction. All the test sections are of the directly-heated design, where electric current passing through the pressure retaining walls of the

  9. Preliminary Design of the Liquid Lead Corrosion Test Loop

    International Nuclear Information System (INIS)

    Cho, Chung Ho; Cha, Jae Eun; Cho, Choon Ho; Song, Tae Yung; Kim, Hee Reyoung

    2005-01-01

    Recently, Lead-Bismuth Eutectic (LBE) or Lead has newly attracted considerable attraction as a coolant to get the more inherent safety. Above all, LBE is preferred as the coolant and target material for an Accelerator-Driven System (ADS) due to its high production rate of neutrons, effective heat removal, and good radiation damage properties. But, the LBE or Lead as a coolant has a challenging problem that the LBE or Lead is more corrosive to the construction materials and fuel cladding material than the sodium because the solubility of Ni, Cr and Fe is high. After all, the LBE or Lead corrosion has been considered as an important design limit factor of ADS and Liquid Metal cooled Fast Reactors (LMFR). The Korea Atomic Energy Research Institute (KAERI) has been developing an ADS called HYPER. HYPER is designed to transmute Transuranics (TRU), Tc-99 and I-129 coming from Pressurized Water Reactors (PWRs) and uses an LBE as a coolant and target material. Also, an experimental apparatuses for the compatibility of fuel cladding and structural material with the LBE or Lead are being under the construction or design. The main objective of the present paper is introduction of Lead corrosion test loop which will be built the upside of the LBE corrosion test loop by the end of October of 2005

  10. ac power control in the Core Flow Test Loop

    International Nuclear Information System (INIS)

    McDonald, D.W.

    1980-01-01

    This work represents a status report on a development effort to design an ac power controller for the Core Flow Test Loop. The Core Flow Test Loop will be an engineering test facility which will simulate the thermal environment of a gas-cooled fast-breeder reactor. The problems and limitations of using sinusoidal ac power to simulate the power generated within a nuclear reactor are addressed. The transformer-thyristor configuration chosen for the Core Flow Test Loop power supply is presented. The initial considerations, design, and analysis of a closed-loop controller prototype are detailed. The design is then analyzed for improved performance possibilities and failure modes are investigated at length. A summary of the work completed to date and a proposed outline for continued development completes the report

  11. Two-loop feed water control system in BWR plants

    International Nuclear Information System (INIS)

    Omori, Takashi; Watanabe, Takao; Hirose, Masao.

    1982-01-01

    In the process of the start-up and shutdown of BWR plants, the operation of changing over feed pumps corresponding to plant output is performed. Therefore, it is necessary to develop the automatic changeover system for feed pumps, which minimizes the variation of water level in reactors and is easy to operate. The three-element control system with the water level in reactors, the flow rate of main steam and the flow rate of feed water as the input is mainly applied, but long time is required for the changeover of feed pumps. The two-loop feed control system can control simultaneously two pumps being changed over, therefore it is suitable to the automatic changeover control system for feed pumps. Also it is excellent for the control of the recirculating valves of feed pumps. The control characteristics of the two-loop feed water control system against the external disturbance which causes the variation of water level in reactors were examined. The results of analysis by simulation are reported. The features of the two-loop feed water control system, the method of simulation and the evaluation of the two-loop feed water control system are described. Its connection with a digital feed water recirculation control system is expected. (Kako, I.)

  12. ECP measurements in the BWR-1 water loop relative to water composition changes

    Energy Technology Data Exchange (ETDEWEB)

    Kus, P.; Vsolak, R.; Kysela, J., E-mail: ksp@ujv.cz [Nuclear Research Inst. Rez plc, Husinec - Rez (Czech Republic); Hanawa, S.; Nakamura, T.; Uchida, S., E-mail: hanawa.satoshi@jaea.go.jp [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan)

    2010-07-01

    The goal of this study is to investigate the usage of ECP sensors in nuclear power plants. ECP sensors were tested using the LVR-15 reactor at the Nuclear Research Institute Rez plc (NRI) in the Czech Republic. The experiment took place on the BWR-1 loop, which was designed for investigating the behaviour of structural materials and radioactivity transport under BWR conditions. The BWR-1 loop facilitates irradiation experiments within a wide range of operating parameters (max. pressure of 10 MPa, max. temperature of 573 K and a neutron flux of 1.0* 10{sup 18} n/m{sup 2}s). This study involves the measurement of electrochemical potential (ECP). Corrosion potential is the main parameter for monitoring of water composition changes in nuclear power plants (NPP). The electrochemical potentials of stainless steel were measured under high temperatures in a test loop (BWR-1) under different water composition conditions. Total neutron flux was ∼10{sup -3} to ∼10{sup 12} n/cm{sup 2}s (>0.1 MeV) at a temperature of 560K, neutral pH, and water resistivity of 18.2 MOhm. ECP sensor response related to changes in water composition was monitored. Switching from NWC (normal water conditions) to HWC (hydrogen water conditions) was controlled using oxygen dosage. Water chemistry was monitored approx. 50 meters from the active channel. The active channel temperature was maintained within a range of 543 - 561 K from the start of irradiation for the entire duration of the experiment. A total of 24 reference electrodes composed of platinum (Pt), silver/silver chloride (Ag/AgCl) and a zircon membrane containing silver oxide (Ag{sub 2}O) powder were installed inside the active channel of the LVR-15 test reactor. The active channel (Field tube) was divided into four zones, with each zone containing six sensors. A mathematical radiolysis code model was created in cooperation with the Japan Atomic Energy Agency. (author)

  13. Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

    International Nuclear Information System (INIS)

    Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

    1997-01-01

    A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs

  14. Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

    Energy Technology Data Exchange (ETDEWEB)

    Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

    1997-01-01

    A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs.

  15. Smart Home Hardware-in-the-Loop Testing

    Energy Technology Data Exchange (ETDEWEB)

    Pratt, Annabelle

    2017-07-12

    This presentation provides a high-level overview of NREL's smart home hardware-in-the-loop testing. It was presented at the Fourth International Workshop on Grid Simulator Testing of Energy Systems and Wind Turbine Powertrains, held April 25-26, 2017, hosted by NREL and Clemson University at the Energy Systems Integration Facility in Golden, Colorado.

  16. Storm water monitoring along loop 202 and Salt River.

    Science.gov (United States)

    2010-10-01

    A comprehensive research program for the characterization of storm water runoff from an Arizona : highway was conducted from January through December 2007. The study area covered a portion of : the Loop 202 freeway west of Mesa Drive to a retention b...

  17. Design and Evaluation of a Water Recirculation Loop Maintenance Device for the Advanced Spacesuit Water Membrane Evaporator

    Science.gov (United States)

    Steele, John W.; Rector, Tony; Bue, Grant C.; Campbell, Colin; Makinen, Janice

    2012-01-01

    A dual-bed device to maintain the water quality of the Advanced Spacesuit Water Membrane Evaporation (SWME) water recirculation loop has been designed and is undergoing testing. The SWME is a heat rejection device under development at the NASA Johnson Space Center to perform thermal control for advanced spacesuits. One advantage to this technology is the potential for a significantly greater degree of tolerance to contamination when compared to the existing sublimator technology. The driver for the development of a water recirculation maintenance device is to further enhance this advantage through the leveraging of fluid loop management lessons learned from the International Space Station (ISS). A bed design that was developed for a Hamilton Sundstrand military application, and considered for a potential ISS application with the Urine Processor Assembly, provides a low pressure drop means for water maintenance in a recirculation loop. The bed design is coupled with high-capacity ion exchange resins, organic adsorbents, and a cyclic methodology developed for the Extravehicular Mobility Unit Transport Water Loop. The bed design further leverages a sorbent developed for the ISS that introduces a biocide in a microgravity-compatible manner for the Internal Active Thermal Control System. The leveraging of these water maintenance technologies to the SWME recirculation loop is a unique demonstration of applying the valuable lessons learned on the ISS to the next generation of crewed spaceflight Environmental Control and Life Support System hardware.

  18. Performance of a Water Recirculation Loop Maintenance Device and Process for the Advanced Spacesuit Water Membrane Evaporator

    Science.gov (United States)

    Steele, John W.; Rector, Tony; Bue, Grant C.; Campbell, Colin; Makinen, Janice

    2013-01-01

    A dual-bed device to maintain the water quality of the Advanced Spacesuit Water Membrane Evaporation (SWME) water recirculation loop has been designed and is undergoing testing. The SWME is a heat rejection device under development at the NASA Johnson Space Center to perform thermal control for advanced spacesuits. One advantage to this technology is the potential for a significantly greater degree of tolerance to contamination when compared to the existing Sublimator technology. The driver for the development of a water recirculation maintenance device is to further enhance this advantage through the leveraging of fluid loop management lessons-learned from the International Space Station (ISS). A bed design that was developed for a Hamilton Sundstrand military application, and considered for a potential ISS application with the Urine Processor Assembly, provides a low pressure drop means for water maintenance in a recirculation loop. The bed design is coupled with high capacity ion exchange resins, organic adsorbents, and a cyclic methodology developed for the Extravehicular Mobility Unit (EMU) Transport Water loop. The bed design further leverages a sorbent developed for ISS that introduces a biocide in a microgravity-compatible manner for the Internal Active Thermal Control System (IATCS). The leveraging of these water maintenance technologies to the SWME recirculation loop is a unique demonstration of applying the valuable lessons learned on the ISS to the next generation of manned spaceflight Environmental Control and Life Support System (ECLSS) hardware.

  19. Design and Evaluation of a Water Recirculation Loop Maintenance Device for the Advanced Spacesuit Water Membrane Evaporator

    Science.gov (United States)

    Steele, John W.; Rector, Tony; Bue, Grant C.; Campbell, Colin; Makinen, Janice

    2011-01-01

    A dual-bed device to maintain the water quality of the Advanced Spacesuit Water Membrane Evaporation (SWME) water recirculation loop has been designed and is undergoing testing. The SWME is a heat rejection device under development at the NASA Johnson Space Center to perform thermal control for advanced spacesuits. One advantage to this technology is the potential for a significantly greater degree of tolerance to contamination when compared to the existing Sublimator technology. The driver for the development of a water recirculation maintenance device is to further enhance this advantage through the leveraging of fluid loop management lessons-learned from the International Space Station (ISS). A bed design that was developed for a Hamilton Sundstrand military application, and considered for a potential ISS application with the Urine Processor Assembly, provides a low pressure drop means for water maintenance in a recirculation loop. The bed design is coupled with high capacity ion exchange resins, organic adsorbents, and a cyclic methodology developed for the Extravehicular Mobility Unit (EMU) Transport Water loop. The bed design further leverages a sorbent developed for ISS that introduces a biocide in a microgravity-compatible manner for the Internal Active Thermal Control System (IATCS). The leveraging of these water maintenance technologies to the SWME recirculation loop is a clear demonstration of applying the valuable lessons learned on the ISS to the next generation of manned spaceflight Environmental Control and Life Support System (ECLSS) hardware.

  20. A probabilistic safety assessment of in-pile test loop in HWRR

    International Nuclear Information System (INIS)

    Cao Xuewu; Li Zhaohuan

    1991-07-01

    The PSA methodology has been applied to the in-pile test loop which is installed in the Heavy Water Research Reactor (HWRR). This loop is designed and operated for fuel assembly testing of the Qinshan PWR plant. This analysis is to assess the safety and to evaluate the design of this operating loop. The procedure and models are similar to a PSA on nuclear power plant. The major contents in the analysis consist of the familiarization of the object, the investigation and selection of accident initiators, setting events and fault trees, data collections, quantitative calculations, qualitative and result analyses and final conclusion. This analysis is only limited to the initiators of in-pile loop itself and possible errors made by operators during normal operation. The accident occurence is less than 10 -4 a -1 which may be recommended as an acceptance risk for safety operation of an in-pile test loop. Finally, suggestions have been raised to improve the design of test loop, especially in reducing operation errors by local operators

  1. Development of water chemistry diagnosis system for BWR primary loop

    International Nuclear Information System (INIS)

    Nagase, Makoto; Asakura, Yamato; Sakagami, Masaharu; Uchida, Shunsuke; Ohsumi, Katsumi.

    1988-01-01

    The prototype of a water chemistry diagnosis system for BWR primary loop has been developed. Its purposes are improvement of water chemistry control and reduction of the work burden on plant chemistry personnel. It has three main features as follows. (1) Intensifying the observation of water chemistry conditions by variable sampling intervals based on the on-line measured data. (2) Early detection of water chemistry data trends using a second order regression curve which is calculated from the measured data, and then searching the cause of anomaly if anything (3) Diagnosis of Fe concentration in feedwater using model simulations, in order to lower the radiation level in the primary system. (author)

  2. Modeling a forced to natural convection boiling test with the program LOOP-W

    International Nuclear Information System (INIS)

    Carbajo, J.J.

    1984-01-01

    Extensive testing has been conducted in the Simulant Boiling Flow Visualization (SBFV) loop in which water is boiled in a vertical transparent tube by circulating hot glycerine in an annulus surrounding the tube. Tests ranged from nonboiling forced convection to oscillatory boiling natural convection. The program LOOP-W has been developed to analyze these tests. This program is a multi-leg, one-dimensional, two-phase equilibrium model with slip between the phases. In this study, a specific test, performed at low power where non-boiling forced convection was changed to boiling natural convection and then to non-boiling again, has been modeled with the program LOOP-W

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  4. CLOSED LOOP AOCS TESTING OF AN AUTONOMOUS STAR TRACKER

    DEFF Research Database (Denmark)

    Jørgensen, John Leif

    1999-01-01

    not even a high quality star pattern generator may be able to pass the outlier rejection filtering of the ASC thus efficiently precluding artificial stimuli during AIT tests. In order to circumvent this impasse, the ASC has a series of build-in features enabling simple, yet comprehensive, closed loop...

  5. A one-loop test of string duality

    International Nuclear Information System (INIS)

    Vafa, C.

    1995-01-01

    We test Type IIA-heterotic string duality in six dimensions by showing that the sigma model anomaly of the heterotic string is generated by a combination of a tree level and a string one-loop correction on the Type IIA side. (orig.)

  6. Development of Aerosol Scrubbing Test Loop for Containment Filtered Venting System

    International Nuclear Information System (INIS)

    Lee, Doo Yong; Jung, Woo Young; Lee, Hyun Chul; Lee, Jong Chan; Kim, Gyu Tae

    2016-01-01

    The scrubber tank is filled with scrubbing water with the chemical additives. The droplet separator based on a cyclone is installed above the scrubbing water pool to remove the large droplets that may clog a metal fiber filter installed at the upper section of the scrubber tank. The outlet piping is connected from the scrubber tank to the molecular sieve to chemically remove the gaseous iodine. The aerosol as a particle is physically captured in the scrubbing water pool passing through the scrubbing nozzle as well as the metal fiber filter. The gaseous iodine such as molecular iodine as well as organic iodide is chemically removed in the scrubbing water pool and molecular sieve. The thermal-hydraulic as well as scrubbing performance for the CFVS should be verified with the experiments. The experiment can be divided into the filtration component based experiment and whole system based one. In this paper, the aerosol scrubbing test loop developed to test the thermal-hydraulic and aerosol scrubbing performance of the scrubbing nozzle with the scrubbing water pool is introduced. The aerosol scrubbing test loop has been developed as a part of the Korean CFVS project. In this loop, the filtration components such as the scrubbing nozzle submerged in the scrubbing water pool as well as the cyclone as droplet separator can be tested under the CFVS operating conditions. The aerosol scrubbing performance of the filtration components including pool scrubbing behavior can be tested with the aerosol generation and feeding system and aerosol measurement system.

  7. Development of Aerosol Scrubbing Test Loop for Containment Filtered Venting System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Doo Yong; Jung, Woo Young; Lee, Hyun Chul; Lee, Jong Chan; Kim, Gyu Tae [FNC Technology, Yongin (Korea, Republic of)

    2016-05-15

    The scrubber tank is filled with scrubbing water with the chemical additives. The droplet separator based on a cyclone is installed above the scrubbing water pool to remove the large droplets that may clog a metal fiber filter installed at the upper section of the scrubber tank. The outlet piping is connected from the scrubber tank to the molecular sieve to chemically remove the gaseous iodine. The aerosol as a particle is physically captured in the scrubbing water pool passing through the scrubbing nozzle as well as the metal fiber filter. The gaseous iodine such as molecular iodine as well as organic iodide is chemically removed in the scrubbing water pool and molecular sieve. The thermal-hydraulic as well as scrubbing performance for the CFVS should be verified with the experiments. The experiment can be divided into the filtration component based experiment and whole system based one. In this paper, the aerosol scrubbing test loop developed to test the thermal-hydraulic and aerosol scrubbing performance of the scrubbing nozzle with the scrubbing water pool is introduced. The aerosol scrubbing test loop has been developed as a part of the Korean CFVS project. In this loop, the filtration components such as the scrubbing nozzle submerged in the scrubbing water pool as well as the cyclone as droplet separator can be tested under the CFVS operating conditions. The aerosol scrubbing performance of the filtration components including pool scrubbing behavior can be tested with the aerosol generation and feeding system and aerosol measurement system.

  8. Detection Test for Leakage of CO2 into Sodium Loop

    International Nuclear Information System (INIS)

    Park, Sun Hee; Wi, Myung-Hwan; Min, Jae Hong

    2015-01-01

    This report is about the facility for the detection test for leakage of CO 2 into sodium loop. The facility for the detection test for leakage of CO 2 into sodium loop was introduced. The test will be carried out. Our experimental results are going to be expected to be used for approach methods to detect CO 2 leaking into sodium in heat exchangers. A sodium-and-carbon dioxide (Na-CO 2 ) heat exchanger is one of the key components for the supercritical CO 2 Brayton cycle power conversion system of sodium-cooled fast reactors (SFRs). A printed circuit heat exchanger (PCHE) is considered for the Na-CO 2 heat exchanger, which is known to have potential for reducing the volume occupied by the exchangers compared to traditional shell-and-tube heat exchangers. Among various issues about the Na- CO 2 exchanger, detection of CO 2 leaking into sodium in the heat exchanger is most important thing for its safe operation. It is known that reaction products from sodium and CO 2 such as sodium carbonate (Na 2 CO 3 ) and amorphous carbon are hardly soluble in sodium, which cause plug sodium channels. Detection technique for Na 2 CO 3 in sodium loop has not been developed yet. Therefore, detection of CO 2 and CO from reaction of sodium and CO 2 are proper to detect CO 2 leakage into sodium loop

  9. The Analysis of Loop Seal Purge Time for the KHNP Pressurizer Safety Valve Test Facility Using the GOTHIC Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Ae; Kim, Chang Hyun; Kweon, Gab Joo; Park, Jong Woon [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2007-10-15

    The pressurizer safety valves (PSV) in Pressurized Water Reactors are required to provide the overpressure protection for the Reactor Coolant System (RCS) during the overpressure transients. Korea Hydro and Nuclear Power Company (KHNP) plans to build the PSV test facility for the purpose of providing the PSV pop-up characteristics and the loop seal dynamics for the new safety analysis. When the pressurizer safety valve is mounted in a loop seal configuration, the valve must initially pass the loop seal water prior to popping open on steam. The loop seal in the upstream of PSV prevents leakage of hydrogen gas or steam through the safety valve seat. This paper studies on the loop seal clearing dynamics using GOTHIC-7.2a code to verify the effects of loop seal purge time on the reactor coolant system overpressure.

  10. Seismic proving test of BWR primary loop recirculation system

    International Nuclear Information System (INIS)

    Sato, H.; Shigeta, M.; Karasawa, Y.

    1987-01-01

    The seismic proving test of BWR Primary Loop Recirculation system is the second test to use the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory. The purpose of this test is to prove the seismic reliability of the primary loop recirculation system (PLR), one of the most important safety components in the BWR nuclear plants, and also to confirm the adequacy of seismic analysis method used in the current seismic design. To achieve the purpose, the test was conducted under conditions and scale as near as possible to actual systems. The strength proving test was carried out with the test model mounted on the vibration table in consideration of basic design earthquake ground motions and other conditions to confirm the soundness of structure and the strength against earthquakes. Detailed analysis and analytic evaluation of the data obtained from the test was conducted to confirm the adequacy of the seismic analysis method and earthquake response analysis method used in the current seismic design. Then, on the basis of the results obtained, the seismic safety and reliability of BWR primary loop recirculation of the actual plants was fully evaluated

  11. First domestic primary loop recircuration pump for boiling water reactor

    International Nuclear Information System (INIS)

    Fukuda, Minoru; Taka, Shusei; Kato, Hiroyuki

    1981-01-01

    Two primary loop recirculation (PLR) pumps for the second unit of the Fukushima No. 2 Nuclear Power Station of the Tokyo Electric Power Co., Inc., have been manufactured by Ebara Corporation. They are the first domestically produced pumps for commercial power plants and were manufactured under license from Byron Jackson Pump Division of Borg Warner Corporation. This article describes the special features of pump design and stress analysis, and the results of the 700 hours of factory loop tests, which are all essential for the PLR pump. (author)

  12. Analysis of the UPTF Separate Effects Test 11 (steam-water counter-current flow in the broken loop hot leg) using RELAP5/MOD2

    International Nuclear Information System (INIS)

    Dillistone, M.J.

    1989-08-01

    RELAP5/MOD2 predictions of countercurrent flow limitation in the UPTF hot leg separate effects Test (test 11) are compared with the experimental data. The code underestimates, by a factor of more than three, the gas flow necessary to prevent liquid runback from the steam generator, and this is shown to be due to an oversimplified flow-regime map which does not allow the possibility of stratified flow in the hot leg riser. The predicted countercurrent flow is also shown to depend, wrongly, on the depth of liquid in the steam generator plenum. The same test is also modelled using a version of the code in which stratified flow in the riser is made possible. The gas flow needed to prevent liquid runback is then predicted quite well, but at all lower gas flows the code predicts that the flow is completely unrestricted - i.e. liquid flows between full flow and zero flow are not predicted. This is shown to happen because the code cannot calculate correctly the liquid level in the hot leg, mainly because of a numerical effect of upwind donoring in the momentum flux terms of the code's basic equations. It is also shown that the code cannot model the considerable effect of the ECCS injection pipe (which runs inside the hot leg) on the liquid level. (author)

  13. Detail design of test loop for FIV in fuel bundle and preliminary test

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Woo Gunl; Lee, Wan Young; Kim, Sung Won [Hannam University, Taejeon (Korea)

    2002-04-01

    It is urgent to develop the analytical model for the structural/mechanical integrity of fuel rod. In general, it is not easy to develop a pure analytical model. Occasionally, experimental results have been utilized for the model.Because of this reason, it is required to design proper test loop. Using the optimized test loop, With the optimized test loop, the dynamic behaviour of the rod will be evaluated and the critical flow velocity, which the rod loses the stability in, will be measured for the design of the rod. To verify the integrity of the fuel rod, it is required to evaluate the dynamic behaviour and the critical flow velocity with the test loop. The test results will be utilized to the design of the rod. Generally, the rod has a ground vibration due to turbulence in wide range of flow velocity and the amplitude of vibration becomes larger by the resonance, in a range of the velocity where occurs vortex. The rod loses stability in critical flow velocity caused by fluid-elastic instability. For the purpose of the present work to perform the conceptional design of the test loop, it is necessary (1) to understand the mechanism of the flow-induced vibration and the related experimental coefficients, (2) to evaluate the existing test loops for improving the loop with design parameters and (3) to decide the design specifications of the major equipments of the loop. 35 refs., 14 figs., 4 tabs. (Author)

  14. Two-phase natural circulation experiments in a pressurized water loop with CANDU geometry

    Energy Technology Data Exchange (ETDEWEB)

    Ardron, K.H.; Krishnan, V.S.; McGee, G.R.; Anderson, J.W.D.; Hawley, E.H.

    1984-07-01

    A series of tests has been performed in the RD-12 loop, a 10-MPa pressurized-water loop containing two active boilers, two pumps, and two, or four, heated horizontal channels arranged in a symmetrical figure-of-eight configuration characteristic of the CANDU reactor primary heat-transport system. In the tests, single-phase natural circulation was established in the loop and void was introduced by controlled draining, with the surge tank (pressurizer) valved out of the system. Results indicate that a stable, two-phase, natural circulation flow can usually be established. However, as the void fraction in the loop is increased, large-amplitude flow oscillations can occur. The initial flow oscillations in the two halves of the loop are usually very nearly 180/sup 0/ out-of-phase. However, as the loop inventory is further decreased, an in-phase oscillation component is observed. In tests with two parallel, heated channels in each half-loop, oscillations associated with mass transfer between the channel pairs are also observed. Although flow oscillations can lead to intermittent dryout of the upper elements of the heater-rod assemblies in the horizontal channels, natural circulation cooling appears to be effective until about 50% of the loop inventory is drained; sustained flow stratification then occurs in the heated channels, leading to heater temperature excursions. The paper reviews the experimental results obtained and describes the evolution of natural circulation flow in particular cases as voidage is progressively increased. The stability behavior is discussed briefly with reference to a simple stability model.

  15. High temperature, high pressure gas loop - the Component Flow Test Loop (CFTL)

    International Nuclear Information System (INIS)

    Gat, U.; Sanders, J.P.; Young, H.C.

    1984-01-01

    The high-pressure, high-temperature, gas-circulating Component Flow Test Loop located at Oak Ridge National Laboratory was designed and constructed utilizing Section III of the ASME Boiler and Pressure Vessel Code. The quality assurance program for operating and testing is also based on applicable ASME standards. Power to a total of 5 MW is available to the test section, and an air-cooled heat exchanger rated at 4.4 MW serves as heat sink. The three gas-bearing, completely enclosed gas circulators provide a maximum flow of 0.47 m 3 /s at pressures to 10.7 MPa. The control system allows for fast transients in pressure, power, temperature, and flow; it also supports prolonged unattended steady-state operation. The data acquisition system can access and process 10,000 data points per second. High-temperature gas-cooled reactor components are being tested

  16. Development of 3-Pin Fuel Test Loop and Utilization Technology

    International Nuclear Information System (INIS)

    Lee, Chung Young; Sim, B. S.; Lee, C. Y.

    2007-06-01

    The principal contents of this project are to design, fabricate and install the steady-state fuel test loop in HANARO for nuclear technology development. Procurement and, fabrication of main equipment, licensing and installation for fuel test loop have been performed. Following contents are described in the report. 1. Design - Design of the In-pile system and Out pile system 2. Fabrication and procurement of the equipment - Fabrication of the In-pile system and In-pool piping - Fabrication and procurement of the equipment of the out-pile system 3. Acquisition of the license - Preparation of the safety analysis report and acquisition of the license - Pre-service inspection of the facility 4. Installation and commissioning - Installation of the FTL - Development of the commissioning procedure

  17. Disassembly and removal of sodium instrumentation test loop

    International Nuclear Information System (INIS)

    Ishikawa, Okinobu; Onojima, Takamitu; Nagai, Keiichi

    2000-07-01

    In 1999, the Sodium Instrumentation Test Loop was disassembled and removed. This report describes the tasks and experiences obtained in removing sodium from a storage tank, disassembling, and cleansing components and related activities. Overall the disassembly, handling and cleansing tasks proceeded as planned and the activities were carried out efficiently and safely. Documentation of the process is meant to establish not only a procedure, but also a guideline for future similar tasks. (author)

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

    International Nuclear Information System (INIS)

    Calvaresi, C.; Moreschi, L.F.

    1983-01-01

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

  19. Mechanisms Engineering Test Loop - Phase 1 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Kultgen, D. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Hvasta, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Lisowski, D. [Argonne National Lab. (ANL), Argonne, IL (United States); Toter, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Borowski, A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-09-01

    This report documents the current status of the Mechanisms Engineering Test Loop (METL) as of the end of FY2016. Currently, METL is in Phase I of its design and construction. Once operational, the METL facility will test small to intermediate-scale components and systems in order to develop advanced liquid metal technologies. Testing different components in METL is essential for the future of advanced fast reactors as it will provide invaluable performance data and reduce the risk of failures during plant operation.

  20. In-pile loop experiments in water chemistry and corrosion

    International Nuclear Information System (INIS)

    Kysela, J.

    1986-09-01

    Results on the study of Zr-1% Nb alloy corrosion, in out-of and in-pile loops simulating the working conditions of the VVER-440 reactor (Soviet, PWR type), covered the time period May 1982-April 1986 were reported, as well as, results on transport and filtration of corrosion products. Methods and techniques used in the study included remote measurement of corrosion rate by polarizing resistance, out-of-pile loop at the temperature 350 deg. C, pressure 19 MPa, circulation 20 kgs/h and in-pile water loop with constant flow rate 10,000 kgs/h, pressure 16 MPa, temperature 330 deg. C and neutron flux 7x10 13 n/cm 2 .s. It was shown that solid suspended particles with chemical composition corresponding most frequently to magnetite or nickelous ferrite, though with non-stoichiometric composition Me x 2+ Fe 3- x 3+ O 4 were found. Continuous filtration of water by means of electromagnetic filter leads to a decrease of radioactivity of the outer epitactic layer only. Effect of filtration on the inner topotactic layer is negligible. The corrosion rates for the above-mentioned parameters are given

  1. Two-phase natural circulation experiments in a pressurized water loop with CANDU geometry

    International Nuclear Information System (INIS)

    Ardron, K.H.; Krishnan, V.S.; McGee, G.R.; Anderson, J.W.D.; Hawley, E.H.

    1984-07-01

    To provide information on two-phase natural circulation in a CANDU-type coolant circuit a series of tests has been performed in the RD-12 loop at the Whiteshell Nuclear Research Establishment. RD-12 is a 10-MPa pressurized-water loop containing two active boilers, two pumps, and two, or four, heated horizontal channels arranged in a symmetrical figure-of-eight configuration characteristic of the CANDU reactor primary heat-transport system. In the tests, single-phase natural circulation was established in the loop and void was introduced by controlled draining, with the surge tank (pressurizer) valved out of the system. The paper reviews the experimental results obtained and describes the evolution of natural circulation flow in particular cases as voidage is progressively increased. The stability behaviour is discussed briefly with reference to a simple stability model

  2. Performance of Water Recirculation Loop Maintentance Components for the Advanced Spacesuit Water Membrane Evaporator

    Science.gov (United States)

    Rector, Tony; Peyton, Barbara; Steele, John W.; Bue, Grant C.; Campbell, Colin; Makinen, Janice

    2014-01-01

    Water loop maintenance components to maintain the water quality of the Advanced Spacesuit Water Membrane Evaporation (SWME) water recirculation loop have undergone a comparative performance evaluation with a second SWME water recirculation loop with no water quality maintenance. Results show the benefits of periodic water maintenance. The SWME is a heat rejection device under development at the NASA Johnson Space Center to perform thermal control for advanced spacesuits. One advantage to this technology is the potential for a significantly greater degree of tolerance to contamination when compared to the existing Sublimator technology. The driver for the evaluation of water recirculation maintenance components was to further enhance this advantage through the leveraging of fluid loop management lessonslearned from the International Space Station (ISS). A bed design that was developed for a UTAS military application, and considered for a potential ISS application with the Urine Processor Assembly, provided a low pressure drop means for water maintenance in a recirculation loop. The bed design is coupled with high capacity ion exchange resins, organic adsorbents, and a cyclic methodology developed for the Extravehicular Mobility Unit (EMU) Transport Water loop. The maintenance cycle included the use of a biocide delivery component developed for ISS to introduce a biocide in a microgravity-compatible manner for the Internal Active Thermal Control System (IATCS). The leveraging of these water maintenance technologies to the SWME recirculation loop is a unique demonstration of applying the valuable lessons learned on the ISS to the next generation of manned spaceflight Environmental Control and Life Support System (ECLSS) hardware.

  3. Performance of Water Recirculation Loop Maintenance Components for the Advanced Spacesuit Water Membrane Evaporator

    Science.gov (United States)

    Rector, Tony; Peyton, Barbara M.; Steele, John W.; Makinen, Janice; Bue, Grant C.; Campbell, Colin

    2014-01-01

    Water loop maintenance components to maintain the water quality of the Advanced Spacesuit Water Membrane Evaporation (SWME) water recirculation loop have undergone a comparative performance evaluation with a second SWME water recirculation loop with no water quality maintenance. Results show the benefits of periodic water maintenance. The SWME is a heat rejection device under development at the NASA Johnson Space Center to perform thermal control for advanced spacesuits. One advantage to this technology is the potential for a significantly greater degree of tolerance to contamination when compared to the existing Sublimator technology. The driver for the evaluation of water recirculation maintenance components was to further enhance this advantage through the leveraging of fluid loop management lessons learned from the International Space Station (ISS). A bed design that was developed for a UTAS military application, and considered for a potential ISS application with the Urine Processor Assembly, provided a low pressure drop means for water maintenance in a recirculation loop. The bed design is coupled with high capacity ion exchange resins, organic adsorbents, and a cyclic methodology developed for the Extravehicular Mobility Unit (EMU) Transport Water loop. The maintenance cycle included the use of a biocide delivery component developed for ISS to introduce a biocide in a microgravity compatible manner for the Internal Active Thermal Control System (IATCS). The leveraging of these water maintenance technologies to the SWME recirculation loop is a unique demonstration of applying the valuable lessons learned on the ISS to the next generation of manned spaceflight Environmental Control and Life Support System (ECLSS) hardware.

  4. Counter-part Test and Code Analysis of the Integral Test Loop, SNUF

    International Nuclear Information System (INIS)

    Park, Goon Cherl; Bae, B. U.; Lee, K. H.; Cho, Y. J.

    2007-02-01

    The thermal-hydraulic phenomena of Direct Vessel Injection (DVI) line Small-Break Loss-of-Coolant Accident (SBLOCA) in pressurized water reactor, APR1400, were investigated. The reduced-height and reduced-pressure integral test loop, SNUF (Seoul National University Facility), was constructed with scaling down the prototype. For the appropriate test conditions in the experiment of SNUF, the energy scaling methodology was suggested as scaling the coolant mass inventory and thermal power for the reduced-pressure condition. From the MARS code analysis, the energy scaling methodology was confirmed to show the reasonable transient when ideally scaled-down SNUF model was compared to the prototype model. In the experiments according to the conditions determined by energy scaling methodology, the phenomenon of downcomer seal clearing had a dominant role in decrease of the system pressure and increase of the coolant level of core. The experimental results was utilized to validate the calculation capability of MARS

  5. Hanford Tank Farms Waste Certification Flow Loop Test Plan

    Energy Technology Data Exchange (ETDEWEB)

    Bamberger, Judith A.; Meyer, Perry A.; Scott, Paul A.; Adkins, Harold E.; Wells, Beric E.; Blanchard, Jeremy; Denslow, Kayte M.; Greenwood, Margaret S.; Morgen, Gerald P.; Burns, Carolyn A.; Bontha, Jagannadha R.

    2010-01-01

    A future requirement of Hanford Tank Farm operations will involve transfer of wastes from double shell tanks to the Waste Treatment Plant. As the U.S. Department of Energy contractor for Tank Farm Operations, Washington River Protection Solutions anticipates the need to certify that waste transfers comply with contractual requirements. This test plan describes the approach for evaluating several instruments that have potential to detect the onset of flow stratification and critical suspension velocity. The testing will be conducted in an existing pipe loop in Pacific Northwest National Laboratory’s facility that is being modified to accommodate the testing of instruments over a range of simulated waste properties and flow conditions. The testing phases, test matrix and types of simulants needed and the range of testing conditions required to evaluate the instruments are described

  6. Loop corrections and a new test of inflation

    CERN Document Server

    Tasinato, Gianmassimo; Nurmi, Sami; Wands, David

    2013-01-01

    Inflation is the leading paradigm for explaining the origin of primordial density perturbations and the observed temperature fluctuations of the cosmic microwave background. However many open questions remain, in particular whether one or more scalar fields were present during inflation and how they contributed to the primordial density perturbation. We propose a new observational test of whether multiple fields, or only one (not necessarily the inflaton) generated the perturbations. We show that our test, relating the bispectrum and trispectrum, is protected against loop corrections at all orders, unlike previous relations.

  7. MES lead bismuth forced circulation loop and test results

    International Nuclear Information System (INIS)

    Ono, Mikinori; Mine, Tatsuya; Kitano, Teruaki; Kamata, Kin-ya

    2003-01-01

    Liquid lead-bismuth is a promising material as future reactor coolant or intensive neutron source material for accelerator driven system (ADS). Mitsui Engineering and Shipbuilding Co., Ltd. (MES) completed lead-bismuth coolant (LBC) forced circulation loop in May 2001 and acquired engineering data on economizer, electro magnetic pump, electro magnetic flow meter and so on. For quality control of LBC, oxygen sensor and filtering element are developing using some hydrogen and moisture mixed gases. Structural materials corrosion test for accelerator driver system (ADS) will start soon. And thermal hydraulic test for ADS will start in tree years. (author)

  8. Design criteria and fabrication in-pile test section of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1997-10-01

    Safety state fuel test loop will be equipped in HANARO to obtain the development and betterments of advanced fuel and materials through the irradiation tests. The objective of this study is to determine the design criteria and technical specification of in-pile test section and to specify the manufacturing requirements of in-pile test section. HANARO fuel test loop was designed to meet the CANDU and PWR fuel testing and in-pile section will be manufactured and installed in HANARO. The design criteria and technical specification of in-pile test section could be used the fuel and materials design with for irradiation testing IPS of HANARO fuel test loop. This results will become guidances for the planning and programming of irradiation testing. (author). 12 refs., tabs., figs.

  9. New in-pile water loop facility for IASCC studies at JMTR

    International Nuclear Information System (INIS)

    Tsukada, T.; Tsuji, H.; Nakajima, H.; Komori, Y.; Ito, H.

    2002-01-01

    Irradiation assisted stress corrosion cracking (IASCC) is caused by the synergistic effects of neutron and gamma radiation, residual and applied stresses and high temperature water environment on the structural materials of vessel internals. IASCC has been studied since the beginning of the 1980's and the phenomenological knowledge on IASCC is accrued extensively. However, mainly due to the experimental difficulties, data for the mechanistic understanding and prediction of failures of the specific in-vessel components are still insufficient and further well-controlled experiments are needed [1]. In recent years, efforts to perform the in-pile materials test for IASCC study have been made at some research reactors [2-4]. At JAERI, a high temperature water loop facility was designed to install at the Japan Materials Testing Reactor (JMTR) to carry out the in-core IASCC testing. This report describes an overview of design and specification of the loop facility. (authors)

  10. Systematic Unit Testing in a Read-eval-print Loop

    DEFF Research Database (Denmark)

    Nørmark, Kurt

    2010-01-01

    .  The process of collecting the expressions and their results imposes only little extra work on the programmer.  The use of the tool provides for creation of test repositories, and it is intended to catalyze a much more systematic approach to unit testing in a read-eval-print loop.  In the paper we also discuss...... how to use a test repository for other purposes than testing.  As a concrete contribution we show how to use test cases as examples in library interface documentation.  It is hypothesized---but not yet validated---that the tool will motivate the Lisp programmer to take the transition from casual...

  11. Solar cooling in the hardware-in-the-loop test; Solare Kuehlung im Hardware-in-the-Loop-Test

    Energy Technology Data Exchange (ETDEWEB)

    Lohmann, Sandra; Radosavljevic, Rada; Goebel, Johannes; Gottschald, Jonas; Adam, Mario [Fachhochschule Duesseldorf (Germany). Erneuerbare Energien und Energieeffizienz E2

    2012-07-01

    The first part of the BMBF-funded research project 'Solar cooling in the hardware-in-the-loop test' (SoCool HIL) deals with the simulation of a solar refrigeration system using the simulation environment Matlab / Simulink with the toolboxes Stateflow and Carnot. Dynamic annual simulations and DoE supported parameter variations were used to select meaningful system configurations, control strategies and dimensioning of components. The second part of this project deals with hardware-in-the-loop tests using the 17.5 kW absorption chiller of the company Yazaki Europe Limited (Hertfordshire, United Kingdom). For this, the chiller is operated on a test bench in order to emulate the behavior of other system components (solar circuit with heat storage, recooling, buildings and cooling distribution / transfer). The chiller is controlled by a simulation of the system using MATLAB / Simulink / Carnot. Based on the knowledge on the real dynamic performance of the chiller the simulation model of the chiller can then be validated. Further tests are used to optimize the control of the chiller to the current cooling load. In addition, some changes in system configurations (for example cold backup) are tested with the real machine. The results of these tests and the findings on the dynamic performance of the chiller are presented.

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  13. LOCA simulation tests in the RD-12 loop with multiple heat channels

    International Nuclear Information System (INIS)

    Ardron, K.H.; McGee, G.R.; Hawley, E.H.

    1985-11-01

    A series of tests has been performed in the RD-12 loop to study the bahaviour of a CANDU-type, primary heat transport system (PHTS) during the blowdown and injection phases of a loss-of-coolant accident (LOCA). Specifically, the tests were used to investigate flow stagnation and refilling of the core following a LOCA. RD-12 is a pressurized water loop with the basic geometry of a CANDU reactor PHTS, but at approximately 1/125 volume scale. The loop consists of U-tube steam generators, pumps, headers, feeders, and heated channels arranged in the symmetrical figure-of-eight configuration of the CANDU PHTS. In the LOCA simulation tests, the loop contained four horizontal heated channels, each containing a seven-element assembly of indirectly heated, fuel-rod simulators. The channels were nominally identical, and were arranged in parallel pairs between the headers in each half-circuit. Tests were carried out using various restricting orifices to represent pipe breaks of different sizes. The break sizes were specifically chosen such that stagnation conditions in the heated channels would be likely to occur. In some tests, the primary pumps were programmed to run down over a 100-s period to simulate a LOCA with simultaneous loss of pump power. Test results showed that, for certain break sizes, periods of low flow occurred in the channels in one half of the loop, leading to flow stratification and sheath temperature excursions. This report reviews the results of two of the tests, and discusses possible mechanisms that may have led to the low channel flow conditions observed in some cases. Plans for future experiments in the larger scale RD-14 facility are outlined. 5 refs

  14. High-pressure test loop design and application

    International Nuclear Information System (INIS)

    Burnette, R.D.; Graves, J.N.; Blair, P.G.; Baldwin, N.L.

    1980-07-01

    A high-pressure test loop (HPTL) has been constructed for the purpose of performing a number of chemistry experiments at simulated HTGR conditions of temperature, pressure, flow, and impurity content. The HPTL can be used to develop, modify, and verify computer codes for a variety of chemical processes involving gas phase transport in the reactor. Processes such as graphite oxidation, fission product transport, fuel reactions, purification systems, and dust entrainment can be studied at high pressure, which would largely eliminate difficulties in correlating existing laboratory data and reactor conditions

  15. SRF cavity testing using a FPGA Self Excited Loop

    CERN Document Server

    Ben-Zvi, Ilan

    2018-01-01

    This document provides a detailed description of procedures for very-high precision calibration and testing of superconducting RF cavities using digital Low-Level RF (LLRF) electronics based on Field Programmable Gate Arrays (FPGA). The use of a Self-Excited Loop with an innovative procedure for fast turn-on allows the measurement of the forward, reflected and transmitted power from a single port of the directional coupler in front of the cavity, thus eliminating certain measurement errors. Various procedures for measuring the quality factor as a function of cavity fields are described, including a single RF pulse technique. Errors are estimated for the measurements.

  16. Feasibility study on the transient fuel test loop installation

    International Nuclear Information System (INIS)

    Kim, J. Y.; Lee, C. Y.

    1997-02-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. The objective of this study is to investigate and analyze the test capsules and loops in research reactors of the other countries and to design preliminarily the eligible transient fuel test facility to be installed in HANARO. The principle subjects of this study are to analyze the contents, kinds and scopes of the irradiation test facilities for nuclear technology development. The guidances for the basic and detail design of the transient fuel test facility in the future are presented. The investigation and analysis of various kinds of test facilities that are now in operation at the research reactors of nuclear advanced countries are carried out. Based on the design data of HANARO the design materials for an eligible transient fuel test facility comprises two pacts : namely, in pile test fuel in reactor core site, and out of pile system regulates the experimental conditions in the in pile test section. Especially for power ramping and cycling selection of the eligible power variation equipment in HANARO is carried out. (author). 13 refs., 4 tabs., 46 figs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

  18. The economics of supplying the supplementary heat in a closed loop water source heat pump system

    International Nuclear Information System (INIS)

    Johnson, R.P.; Bartkus, V.E.; Singh, J.B.

    1993-01-01

    The paper describes the details of a research and demonstration project that will be completed in August 1992 at a healthcare facility in northeastern Pennsylvania. The purpose of the project is to compare the economics of several methods of supplying the supplementary heating in a facility served by a closed loop water source heat pump system. The systems being tested include a storage hot water tank with electric resistance heaters and three air source heat pumps that have the ability to supply the same heat during on-peak hours as well as off-peak hours. The paper compares the projected operating costs of the following: (1) Gas boiler supplying the supplementary heat. (2) Stored hot water supplying the supplementary heat which is generated and stored during off-peak hours using resistance heat on PP ampersand L's offpeak rate. (3) Stored hot water supplying the supplementary heat generated during off-peak hours using the air source heat pumps on PP ampersand L's off-peak rate. (4) Hot water generated by the air source heat pumps supplying the supplementary loop heating on PP ampersand L's general service and time-of-day electric rates. It is generally known in the HVAC industry that a closed loop water source heat pump system can provide one of the most efficient means of space conditioning to a building with high internal gains by transferring the excess heat available in one part of the building to another part of the building where it may be needed for heating. The following flow diagram depicts the relationship of the air source heat pumps with the storage tanks and the building closed water loop

  19. The design of in-pile test section for fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, K. N.; Lee, J. M.; Shim, B. S.; Zee, D. Y.; Park, S. H.; Ahn, S. H.; Lee, J. Y.; Kim, Y. J. [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    As an equipment for nuclear fuel's general performance irradiation test in HANARO, Fuel Test Loop(FTL) has been developed that can irradiate the pin to the maximum number of 3 at the core irradiation hole(IR1 hole) by considering for it's utility and user's irradiation requirement. 3-Pin FTL consists of In-Pile Test Section (IPS) and Out-of-Pile System (OPS). IPS consists for IPS Vessel assembly, In-Pool Piping, IPS Support, In-Pool Piping Support etc. Design that such IPS considers interference item consisted to do not bear in existing facilities by one. IVA that is connected to the OPS are controlled and regulated by means of system pressure, system temperature and the water quality. IPS Vessel assembly is consisted of outer pressure vessel, inner pressure vessel, IPS head, inner assembly and test fuel carrier. After 3-Pin FTL development which is expected to be finished by the 2006, FTL will be used for the irradiation test of the new PWR-type fuel and can maximize the usage of HANARO.

  20. System Description of the Electrical Power Supply System for the ATLAS Integral Test Loop

    International Nuclear Information System (INIS)

    Moon, S. K.; Park, J. K.; Kim, Y. S.; Song, C. H.; Baek, W. P.

    2007-02-01

    An integral effect test loop for pressurized water reactors (PWRs), the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), is constructed by Thermal-Hydraulics Safety Research Team in Korea Atomic Energy Research Institute (KAERI). The ATLAS facility has been designed to have the length scale of 1/2 and area scale of 1/144 compared with the reference plant, APR1400. This report describes the design and technical specifications of the electrical power supply system which supplies the electrical powers to core heater rods, other heaters, various pumps and other systems. The electrical power supply system had acquired the final approval on the operation from the Korea Electrical Safety Corporation. During performance tests for the operation and control, the electrical power supply system showed completely acceptable operation and control performance

  1. Experiments in a natural circulation loop with supercritical water at low powers

    International Nuclear Information System (INIS)

    Pilkhwal, D.S.; Sharma, Manish; Jana, S.S.; Vijayan, P.K.

    2013-05-01

    Earlier, 1/2 ″ uniform diameter Supercritical Pressure Natural Circulation Loop (SPNL) was set-up in hall-7, BARC for carrying out experiments related to supercritical fluids. The loop is a rectangular loop having two heaters and two coolers. Experiments were carried out with CO 2 under supercritical conditions for various pressures and different combinations of heater and cooler orientations. Since, the design conditions are more severe for supercritical water (SCW) experiments, the loop was modified for SCW by installing new test sections, pressurizer and power supply for operation with supercritical water. Experimental data were generated on steady state, heat transfer and stability under natural circulation conditions for the horizontal heater and horizontal cooler (HHHC) orientation with SCW up to a heater power of 8.5 kW. The flow rate data and instability data were compared with the predictions of in-house developed 1-D code NOLSTA, which showed reasonable agreement. The heat transfer coefficient data were also compared with the predictions of various correlations exhibit peak at bulk temperature lower than that obtained in the experiments. Most of these correlations predicted experimental data well in the pseudo-critical region. However, all correlations are matching well with experimental data beyond the pseudo-critical region. The details of the experimental facility, Experiments carried out and the results presented in this report. (author)

  2. Molten Salt Test Loop (MSTL) system customer interface document.

    Energy Technology Data Exchange (ETDEWEB)

    Gill, David Dennis; Kolb, William J.; Briggs, Ronald D.

    2013-09-01

    The National Solar Thermal Test Facility at Sandia National Laboratories has a unique test capability called the Molten Salt Test Loop (MSTL) system. MSTL is a test capability that allows customers and researchers to test components in flowing, molten nitrate salt. The components tested can range from materials samples, to individual components such as flex hoses, ball joints, and valves, up to full solar collecting systems such as central receiver panels, parabolic troughs, or linear Fresnel systems. MSTL provides realistic conditions similar to a portion of a concentrating solar power facility. The facility currently uses 60/40 nitrate %E2%80%9Csolar salt%E2%80%9D and can circulate the salt at pressure up to 40 bar (600psi), temperature to 585%C2%B0C, and flow rate of 44-50kg/s(400-600GPM) depending on temperature. The purpose of this document is to provide a basis for customers to evaluate the applicability to their testing needs, and to provide an outline of expectations for conducting testing on MSTL. The document can serve as the basis for testing agreements including Work for Others (WFO) and Cooperative Research and Development Agreements (CRADA). While this document provides the basis for these agreements and describes some of the requirements for testing using MSTL and on the site at Sandia, the document is not sufficient by itself as a test agreement. The document, however, does provide customers with a uniform set of information to begin the test planning process.

  3. Study and Optimization of Design Parameters in Water Loop Heat Pump Systems for Office Buildings in the Iberian Peninsula

    Directory of Open Access Journals (Sweden)

    Francisco Javier Fernández

    2017-11-01

    Full Text Available Water loop heat pump (WLHP air conditioning systems use heat pumps connected to a common water circuit to fulfill the energy demands of different thermal zones in a building. In this study, the energy consumption was analyzed for the air conditioning of an office building in the typical climate of four important cities of the Iberian Peninsula. The energy consumption of one water loop heat pump system was compared with a conventional water system. Two design parameters, the range in the control temperatures and the water loop thermal storage size, were tested. Energy redistribution is an important advantage of the WLHP system, but significant savings came from high efficiency parameters in the heat pumps and minor air flow rates in the cooling tower. The low thermal level in the water loop makes this technology appropriate to combine with renewable sources. Using natural gas as the thermal energy source, a mean decrease in CO2 emissions of 8.1% was reached. Simulations showed that the installation of big thermal storage tanks generated small energy savings. Besides, the total annual consumption in buildings with high internal loads can be reduced by keeping the water loop as cool as possible.

  4. Conceptual Design for a High-Temperature Gas Loop Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    James B. Kesseli

    2006-08-01

    This report documents an early-stage conceptual design for a high-temperature gas test loop. The objectives accomplished by the study include, (1) investigation of existing gas test loops to determine ther capabilities and how the proposed system might best complement them, (2) development of a preliminary test plan to help identify the performance characteristics required of the test unit, (3) development of test loop requirements, (4) development of a conceptual design including process flow sheet, mechanical layout, and equipment specifications and costs, and (5) development of a preliminary test loop safety plan.

  5. Gas Test Loop Facilities Alternatives Assessment Report Rev 1

    International Nuclear Information System (INIS)

    William J. Skerjanc; William F. Skerjanc

    2005-01-01

    An important task in the Gas Test Loop (GTL) conceptual design was to determine the best facility to serve as host for this apparatus, which will allow fast-flux neutron testing in an existing nuclear facility. A survey was undertaken of domestic and foreign nuclear reactors and accelerator facilities to arrive at that determination. Two major research reactors in the U.S. were considered in detail, the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR), each with sufficient power to attain the required neutron fluxes. HFIR routinely operates near its design power limit of 100 MW. ATR has traditionally operated at less than half its design power limit of 250 MW. Both of these reactors should be available for at least the next 30 years. The other major U.S. research reactor, the Missouri University Research Reactor, does not have sufficient power to reach the required neutron flux nor do the smaller research reactors. Of the foreign reactors investigated, BOR-60 is perhaps the most attractive. Monju and BN 600 are power reactors for their respective electrical grids. Although the Joyo reactor is vigorously campaigning for customers, local laws regarding transport of radioactive material mean it would be very difficult to retrieve test articles from either Japanese reactor for post irradiation examination. PHENIX is scheduled to close in 2008 and is fully booked until then. FBTR is limited to domestic (Indian) users only. Data quality is often suspect in Russia. The only accelerator seriously considered was the Fuel and Material Test Station (FMTS) currently proposed for operation at Los Alamos National Laboratory. The neutron spectrum in FMTS is similar to that found in a fast reactor, but it has a pronounced high-energy tail that is atypical of fast fission reactor spectra. First irradiation in the FMTS is being contemplated for 2008. Detailed review of these facilities resulted in the recommendation that the ATR would be the best host for the GTL

  6. Closing the brain-to-brain loop in laboratory testing.

    Science.gov (United States)

    Plebani, Mario; Lippi, Giuseppe

    2011-07-01

    Abstract The delivery of laboratory services has been described 40 years ago and defined with the foremost concept of "brain-to-brain turnaround time loop". This concept consists of several processes, including the final step which is the action undertaken on the patient based on laboratory information. Unfortunately, the need for systematic feedback to improve the value of laboratory services has been poorly understood and, even more risky, poorly applied in daily laboratory practice. Currently, major problems arise from the unavailability of consensually accepted quality specifications for the extra-analytical phase of laboratory testing. This, in turn, does not allow clinical laboratories to calculate a budget for the "patient-related total error". The definition and use of the term "total error" refers only to the analytical phase, and should be better defined as "total analytical error" to avoid any confusion and misinterpretation. According to the hierarchical approach to classify strategies to set analytical quality specifications, the "assessment of the effect of analytical performance on specific clinical decision-making" is comprehensively at the top and therefore should be applied as much as possible to address analytical efforts towards effective goals. In addition, an increasing number of laboratories worldwide are adopting risk management strategies such as FMEA, FRACAS, LEAN and Six Sigma since these techniques allow the identification of the most critical steps in the total testing process, and to reduce the patient-related risk of error. As a matter of fact, an increasing number of laboratory professionals recognize the importance of understanding and monitoring any step in the total testing process, including the appropriateness of the test request as well as the appropriate interpretation and utilization of test results.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  9. Physics evaluation for testino. of RAPS and TAPS fuel pins in CIRUS pressurised water loop

    International Nuclear Information System (INIS)

    John, Benjamin; Paul, O.P.K.

    1976-01-01

    Relevant calculations carried out to assess the reactivity effect, heat generation and other parameters for testing of RAPS and TAPS fuel pins in the Cirus pressurised water loop are summarised. The Cirus neutron flux level being low, in order to simulate the RAPS design heat rating of ∫ Kdtheta = 40 w/cm, the required plutonium enrichment in mixed plutonium uranium oxide fuel pin was worked out. The results showed that a PuO 2 enrichment of 1.5 wt percent would be necessary to meet the above requirement. The analysis for the TAPS pin indicated that the desired heat flux of 115w/cm 2 cannot be obtained in the Cirus loop with either a 7 pin cluster geometry, or with a single pin with the enrichment level as used in TAPS pin. Lattice code DUMLAC and the core simulation code AECLHEX were used for these studies. (author)

  10. Commissioning of an Integral Effect Test Loop for SMART

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyunsik; Bae, Hwang; Kim, Dongeok; Min, Kyoungho; Shin, Yongcheol; Ko, Yungjoo; Yi, Sungjae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    An integral-effect test loop for SMART, SMART-ITL (or FESTA), has been constructed at KAERI. Its height was preserved and its flow area and volume were scaled down to 1/49 compared with the prototype plant, SMART. The ratio of the hydraulic diameter is 1/7. The SMART is a 330 MW thermal power reactor, and its core exit temperature and PZR pressure are 323 .deg. C and 15 MPa during a normal working condition, respectively. The maximum power of the core heater in the SMART-ITL is 30% of the scaled full power. As shown in Fig. 1, the SMART-ITL consists of a primary system including a reactor pressure vessel with a pressurizer, four steam generators and four main coolant pumps, a secondary system, a safety system, and an auxiliary system. The SMART-ITL facility will be used to investigate the integral performance of the inter-connected components and possible thermal-hydraulic phenomena occurring in the SMART design, to validate its safety for various design basis events and broad transient scenarios, and to validate the related thermal-hydraulic models of the safety analysis codes. The scenarios include small-break loss-of coolant accident (SBLOCA) scenarios, complete loss of RCS flowrate (CLOF), steam generator tube rupture (SGTR), feedwater line break (FLB), and main steam line break (MSLB). The role of SMART-ITL will be extended to examine and verify the normal, abnormal, and emergency operating procedures required during the construction and export phases of SMART. After an extensive series of commissioning tests in 2012, the SMART-ITL facility is now in operation. In this paper, the major test results acquired during the commissioning tests will be discussed.

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

    International Nuclear Information System (INIS)

    Oakley, D.J.

    1978-01-01

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

  12. Efforts to Reduce International Space Station Crew Maintenance for the Management of the Extravehicular Mobility Unit Transport Loop Water Quality

    Science.gov (United States)

    Steele, John W.; Etter, David; Rector, Tony; Boyle, Robert; Vandezande, Christopher

    2013-01-01

    The EMU (Extravehicular Mobility Unit) contains a semi-closed-loop re-circulating water circuit (Transport Loop) to absorb heat into a LCVG (Liquid Coolant and Ventilation Garment) worn by the astronaut. A second, single-pass water circuit (Feed-water Loop) provides water to a cooling device (Sublimator) containing porous plates, and that water sublimates through the porous plates to space vacuum. The cooling effect from the sublimation of this water translates to a cooling of the LCVG water that circulates through the Sublimator. The quality of the EMU Transport Loop water is maintained through the use of a water processing kit (ALCLR Airlock Cooling Loop Remediation) that is used to periodically clean and disinfect the water circuit. Opportunities to reduce crew time associated with on-orbit ALCLR operations include a detailed review of the historical water quality data for evidence to support an extension to the implementation cycle. Furthermore, an EMU returned after 2-years of use on the ISS (International Space Station) is being used as a test bed to evaluate the results of extended and repeated ALCLR implementation cycles. Finally, design, use and on-orbit location enhancements to the ALCLR kit components are being considered to allow the implementation cycle to occur in parallel with other EMU maintenance and check-out activities, and to extend the life of the ALCLR kit components. These efforts are undertaken to reduce the crew-time and logistics burdens for the EMU, while ensuring the long-term health of the EMU water circuits for a post-Shuttle 6-year service life.

  13. Microcomputer-controlled flow meter used on a water loop

    International Nuclear Information System (INIS)

    Haniger, L.

    1982-01-01

    The report describes a microcomputer-controlled instrument intended for operational measurement on an experimental water loop. On the basis of pressure and temperature input signals the instrument calculates the specific weight, and for ten operator-selectable measuring channels it calculates the mass flow G(kp/s), or the voluminal flow Q(m 3 /h). On pressing the appropriate push-buttons the built-in display indicates the values of pressure (p) and temperature (t), as well as the values of specific weight γ calculated therefrom. For ten individually selectable channels the instrument displays either the values of the pressure differences of the measuring throttling elements (√Δpsub(i)), or the values of Gsub(i) or Qsub(i) as obtained by calculation. In addition, on pressing the Σ-push-button it summarizes the values of Gsub(i) and Qsub(i) for the selected channels. The device is controlled by an 8085 microprocessor, the analog unit MP 6812 being used as the A/D convertor. The instrument algorithm indicates some possible errors which may concern faults of input signals or mistakes in calculation. (author)

  14. Dynamic modelling and hardware-in-the-loop testing of PEMFC

    Energy Technology Data Exchange (ETDEWEB)

    Vath, Andreas; Soehn, Matthias; Nicoloso, Norbert; Hartkopf, Thomas [Technische Universitaet Darmstadt/Institut fuer Elektrische Energie wand lung, Landgraf-Georg-Str. 4, D-64283 Darmstadt (Germany); Lemes, Zijad; Maencher, Hubert [MAGNUM Automatisierungstechnik GmbH, Bunsenstr. 22, D-64293 Darmstadt (Germany)

    2006-07-03

    Modelling and hardware-in-the-loop (HIL) testing of fuel cell components and entire systems open new ways for the design and advance development of FCs. In this work proton exchange membrane fuel cells (PEMFC) are dynamically modelled within MATLAB-Simulink at various operation conditions in order to establish a comprehensive description of their dynamic behaviour as well as to explore the modelling facility as a diagnostic tool. Set-up of a hardware-in-the-loop (HIL) system enables real time interaction between the selected hardware and the model. The transport of hydrogen, nitrogen, oxygen, water vapour and liquid water in the gas diffusion and catalyst layers of the stack are incorporated into the model according to their physical and electrochemical characteristics. Other processes investigated include, e.g., the membrane resistance as a function of the water content during fast load changes. Cells are modelled three-dimensionally and dynamically. In case of system simulations a one-dimensional model is preferred to reduce computation time. The model has been verified by experiments with a water-cooled stack. (author)

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

    International Nuclear Information System (INIS)

    Lucot, M.; Roche, M.

    1986-09-01

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

  16. The analysis of SCS return momentum effects on the RCS water level during mid-loop operations

    Energy Technology Data Exchange (ETDEWEB)

    swang Seo, J.; Young Yang, J.; Tack Hwang, S. [Seoul National Univ. (Korea, Republic of)

    1995-09-01

    An accurate prediction of Reactor Coolant System (RCS) water levels is of importance in the determination of allowable operating range to ensure the safety during the mid-loop operations. However, complex hydraulic phenomena induced by Shutdown Cooling System (SCS) return momentum cause different water levels from those in the loop where the water level indicators are located. This was apparantly observed at the pre-core cold hydro test of the Younggwang Nuclear Unit 3 (YGN 3) in Korea. In this study, in order to analytically understand the effect of the SCS return momentum on the RCS water level and its general trend, a model using one-dimensional momentum equation, hydraulic jump, Bernoulli equation, flow resistance coefficient, and total water volume conservation has been developed to predict the RCS water levels at various RCS locations during the mid-loop conditions and the simulation results were compared with the test data. The analysis shows that the hydraulic jump in the operating cold legs in conjunction with the momentum loss throughout the RCS is the main cause creating the water level differences at various RCS locations. The prediction results provide good explanations for the test data and show the significant effect of the SCS return momentum on the RCS water levels.

  17. The analysis of SCS return momentum effects on the RCS water level during mid-loop operations

    International Nuclear Information System (INIS)

    swang Seo, J.; Young Yang, J.; Tack Hwang, S.

    1995-01-01

    An accurate prediction of Reactor Coolant System (RCS) water levels is of importance in the determination of allowable operating range to ensure the safety during the mid-loop operations. However, complex hydraulic phenomena induced by Shutdown Cooling System (SCS) return momentum cause different water levels from those in the loop where the water level indicators are located. This was apparantly observed at the pre-core cold hydro test of the Younggwang Nuclear Unit 3 (YGN 3) in Korea. In this study, in order to analytically understand the effect of the SCS return momentum on the RCS water level and its general trend, a model using one-dimensional momentum equation, hydraulic jump, Bernoulli equation, flow resistance coefficient, and total water volume conservation has been developed to predict the RCS water levels at various RCS locations during the mid-loop conditions and the simulation results were compared with the test data. The analysis shows that the hydraulic jump in the operating cold legs in conjunction with the momentum loss throughout the RCS is the main cause creating the water level differences at various RCS locations. The prediction results provide good explanations for the test data and show the significant effect of the SCS return momentum on the RCS water levels

  18. Antimony measurement in high pressure reactor water loop

    International Nuclear Information System (INIS)

    Svarc, V.; Dudjakova, K.; Martykan, M.; Sus, F.; Kysela, J.

    2005-01-01

    RVS 3 loop is highly contaminated due to antimony main circulating pump bearings. Antimony is one of the major contaminant in PWR units. Different technologies to remove Sb from the systems have been tried. (N.T.)

  19. Prediction of the Long Term Cooling Performance for the 3-Pin Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, H. R

    2005-12-15

    In the long term cooling phase that the emergency cooling water injection ends, the performance of the residual heat removal for the 3-pin fuel test loop has been predicted by a simplified heat transfer model. In the long term cooling phase the residual heat is 1323W for PWR fuel test mode and 1449W for CANDU fuel test mode. The each residual heat is assumed as 2% of the fission power of the test fuel used in the anticipated operational occurrence and design basis accident analyses. The each fission power used for the analyses is 105% of the rated fission power in the normal operation. In the long term cooling phase the residual heat is removed to the HANARO pool through the double pressure vessels of the in-pile test section. Saturate pooling boiling is assumed on the test fuel and condensation heat transfer is expected on the inner wall of the fuel carrier and the flow divider. Natural convection heat transfer on a heated vertical wall is also assumed on the outer wall of the outer pressure vessel. The conduction heat transfer is only considered in the gap between the double pressure vessels charged with neon gas and in the downcomer filled with coolant. The heat transfer rate between the coolant temperature of 152 .deg. C in the in-pile test section and the water temperature of 45 .deg. C in the HANARO pool is predicted as about 1666W. The 152 .deg. C is the saturate temperature of the coolant pressure predicted from the MARS code. The cooling capacity of 1666W is greater than the residual heats of 1323W and 1449W. Consequently the long term cooling performance of the 3-pin fuel test loop is sufficient for the anticipated operational occurrences and design basis accidents.

  20. Results from tests of TFL Hydragard sampling loop

    International Nuclear Information System (INIS)

    Steimke, J.L.

    1995-03-01

    When the Defense Waste Processing Facility (DWPF) is operational, processed radioactive sludge will be transferred in batches to the Slurry Mix Evaporator (SME), where glass frit will be added and the contents concentrated by boiling. Batches of the slurry mixture are transferred from the SME to the Melter Feed Tank (MFT). Hydragard reg-sign sampling systems are used on the SME and the MFT for collecting slurry samples in vials for chemical analysis. An accurate replica of the Hydragard sampling system was built and tested in the thermal Fluids Laboratory (TFL) to determine the hydragard accuracy. It was determined that the original Hydragard valve frequently drew a non-representative sample stream through the sample vial that ranged from frit enriched to frit depleted. The Hydragard valve was modified by moving the plunger and its seat backwards so that the outer surface of the plunger was flush with the inside diameter of the transfer line when the valve was open. The slurry flowing through the vial accurately represented the composition of the slurry in the reservoir for two types of slurries, different dilution factors, a range of transfer flows and a range of vial flows. It was then found that the 15 ml of slurry left in the vial when the Hydragard valve was closed, which is what will be analyzed at DWPF, had a lower ratio of frit to sludge as characterized by the lithium to iron ratio than the slurry flowing through it. The reason for these differences is not understood at this time but it is recommended that additional experimentation be performed with the TFL Hydragard loop to determine the cause

  1. Cabri - water loop a new IPSN-OECD international research program

    International Nuclear Information System (INIS)

    Anon.

    2001-01-01

    In 1993, the Institut de Protection et de Surete Nucleaire (IPSN, the French Nuclear Safety and Radiation Protection Institute) working with EDF (electric utilities) and backed by the NRC (Nuclear Regulatory Commission, USA) launched a research program, dubbed Cabri REP Na addressing uranium oxide-based fuels and MOX fuels. So far twelve tests have been conducted including eight on UO 2 fuel and four on MOX fuel. More testing is now required to determine fuel performance at higher specific burn-up levels in typical PWR (Pressurized Water Reactor) conditions, the purpose being to determine the acceptance criteria for tomorrow's fuels. IPSN has defined a new research program for the Cabri reactor. The OECD's Nuclear Energy Agency is quarterbacking the international program called 'Cabri-Water Loop'. (authors)

  2. UF6 test loop for evaluation and implementation of international enrichment plant safeguards

    International Nuclear Information System (INIS)

    Cooley, J.N.; Fields, L.W.; Swindle, D.W. Jr.

    1987-06-01

    A functional test loop capable of simulating UF 6 flows, pressures, and pipe deposits characteristic of gas centrifuge enrichment plant piping has been designed and fabricated by the Enrichment Safeguards Program of Martin Marietta Energy Systems, Inc., for use by International Atomic Energy Agency (IAEA) at its Safeguards Analytical Laboratory in Seibersdorf, Austria. Purpose of the test loop is twofold: (1) to enable the IAEA to evaluate and to calibrate enrichment safeguards measurement instrumentation to be used in limited frequency-unannounced access (LFUA) inspection strategy measurements at gas centrifuge enrichment plants and (2) to train IAEA inspectors in the use of such instrumentation. The test loop incorporates actual sections of cascade header pipes from the centrifuge enrichment plants subject to IAEA inspections. The test loop is described, applications for its use by the IAEA are detailed, and results from an initial demonstration session using the test loop are summarized

  3. Analysis of th SBLOCAs in the room 1 for the 3-pin fuel test loop

    International Nuclear Information System (INIS)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, Y. J.

    2004-10-01

    Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Small Break Loss of Coolant Accidents (SBLOCAs) in the Room 1 for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the SBLOCAs. The location of the pipe break is assumed at the downstream of the main cooling water pump and the upstream of the main cooler in the room 1. The break size is also assumed less than 20% of the cross section area of the pipe. The test fuels are heated up when the cold leg break occur. However, they are not heated up when the hot leg break occur. The maximum Peak Cladding Temperature (PCT) is predicted to be about 931.4 .deg. C for the cold leg break accident in PWR fuel test mode and 931.6 .deg. C in CANDU fuel test mode respectively. The critical break size is about the 8% of the cross section area of the pipe for PWR fuel test mode and the 10% for CANDU fuel test mode. The PCTs meet the design criterion of commercial PWR fuel that the maximum PCT is lower than 1204 .deg. C

  4. Analysis of the LBLOCAs in the HANARO pool for the 3-pin fuel test loop

    International Nuclear Information System (INIS)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, Y. J.

    2004-12-01

    The Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Large Break Loss of Coolant Accidents (LBLOCAs) in HANARO pool for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the LBLOCAs. The location of the pipe break is assumed at the hill taps connecting the cold and hot legs in HANARO pool to the inlet and outlet nozzles of the In-Pile test Section (IPS). Double ended guillotine break is assumed for the large break loss of coolant accidents. The discharge coefficients of 0.1, 0.33, 0.67, 1.0 are investigated for the LBLOCAs. The test fuels for PWR and CANDU test modes are not heated up for the LBLOCAs caused by the double ended guillotine break in the HANARO pool. The reason is that the sufficient emergency cooling water to cool down the test fuels is supplied continuously to the in-pile test section. Therefore the PCTs for the LBLOCAs in the HANARO pool meet the design criterion of commercial PWR fuel that maximum PCT is lower than 1204 .deg. C

  5. A mathematical model for the simulation of thermal transients in the water loop of IPEN

    International Nuclear Information System (INIS)

    Pontedeiro, A.C.

    1980-01-01

    A mathematical model for simulation of thermal transients in the water loop at the Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo, Brasil, is developed. The model is based on energy equations applied to the components of the experimental water loop. The non-linear system of first order diferencial equations and of non-linear algebraic equations obtained through the utilization of the IBM 'System/360-Continous System Modeling Program' (CSMP) is resolved. An optimization of the running time of the computer is made and a typical simulation of the water loop is executed. (Author) [pt

  6. Application of the X-in-the-Loop Testing Method in the FCV Hybrid Degree Test

    Directory of Open Access Journals (Sweden)

    Haiyu Gao

    2018-02-01

    Full Text Available With the development of fuel cell vehicle technology, an effective testing method that can be applied to develop and verify the fuel cell vehicle powertrain system is urgently required. This paper presents the X-in-the-Loop (XiL testing method in the fuel cell vehicle (FCV hybrid degree test to resolve the first and key issues for the powertrain system design, and the test process and scenarios were designed. The hybrid degree is redefined into the static hybrid degree for system architecture design and the dynamic hybrid degree for vehicle control strategy design, and an integrated testing platform was introduced and a testing application was implemented by following the designed testing flowchart with two loops. Experimental validations show that the sizing of the FCE (Fuel Cell Engine, battery pack, and traction motor with the powertrain architecture can be determined, the control strategy can be evaluated seamlessly, and a systematic powertrain testing solution can be achieved through the whole development process. This research has developed a new testing platform and proposed a novel testing method on the fuel cell vehicle powertrain system, which will be a contribution to fuel cell vehicle technology and its industrialization.

  7. In-pile loop OWL-2 and irradiation tests done with it

    International Nuclear Information System (INIS)

    Suzuki, Shinobu; Ikeshima, Yoshiaki; Kawano, Masakatsu; Watanabe, Hiroyuki; Sato, Hitoshi; Tanaka, Isao

    1990-11-01

    The OWL-2 which was built in the JMTR as the biggest water loop in Japan has been operating for irradiation service since February 1972. The desired objective of the OWL-2, contributing to the development of various nuclear fuels and materials for the light water power reactor and to reactor engineering, has been so fully achieved that the OWL-2 is planned to be dismantled. After the dismantling, a loop, needed for the research and development of the breeding blanket for the fusion reactor, is going to be installed in place of the OWL-2 as a part of the JMTR Modification Program. This paper deals with the history of the OWL-2 with an emphasis on the technical affairs taken into consideration when designing the OWL-2, the irradiation tests, development of the turbine flowmeter, results of the surveillance test of the material of the in-reactor tube, the knowledge gained in the course of the investigation into the cause of transgranular stress corrosion cracking (TGSCC) which developed in the wall of the in-reactor tube, and countermeasures taken to prevent TGSCC from recurring. (author)

  8. The Phillips Laboratory capillary pumped loop test facility

    Science.gov (United States)

    Gluck, Donald F.; Kaylor, Marc C.

    1996-03-01

    An ammonia capillary pumped loop (CPL) test facility has been designed, fabricated, subject to acceptance tests, and assembled at Phillips Laboratory. Its intent is to support a wide range of Air Force programs, bringing CPL technology to flight readiness for operational systems. The facility provides a high degree of modularity and flexibility with several heating and cooling options, and capability for elevation (+/- 15 in.), tilt (+/-60°) and transport length variation. It has a 182 by 44 by 84 inch envelope, an expected heat load capability of 2500 W, and a temperature range of 0 to 50 °C. The evaporator section has two plates with four capillary pumps (CPs) each, with a starter pump on one plate. The CPs are 5/8 in., with TAG aluminum 6063-T6 casing and UHMW polyethylene wicks. The active lengths are 15 and 30 inch with both 10 and 15 micron wicks. The individual CPs have thermal and hydraulic isolation capability, and are removable. The transport section consists of stainless steel lines in a serpentine configuration, a 216 in3 free volume reservoir, and a mechanical pump. The vapor transport line contains a capillary device (which can be bypassed) for vapor blockage during startup. The condenser consists of two separately valved, parallel cold plates each with a downstream noncondensible gas trap. Cooling of up to 1500 W at -50 °C is provided by an FTS Systems chiller using Flourinert FC-72. An enclosure/exhaust system is provided for safety and emergency venting of ammonia. An ammonia charge station performs or supports the functions of proof pressure, flushing with ammonia, purging with gaseous nitrogen, evacuation of all or part of the CPL to 20 microns, and charging. Instrumentation consists of over 116 thermocouples, five of which are internal; one absolute and six differential pressure transducers; eleven watt transducers, and a reservoir load cell. The data acquisition system consists of a temperature scanner, Bernoulli drive, and two Macintosh

  9. Development of Start-up and Shutdown Procedure for the HANARO Fuel Test Loop

    International Nuclear Information System (INIS)

    Park, S. K.; Sim, B. S.; Chi, D. Y.; Lee, J. M.; Lee, C. Y.; Ahn, S. H.

    2009-06-01

    A start-up and shutdown procedure for the HANARO fuel test loop has been developed. This is a facility for fuel and material irradiation tests. The facility provides experimental conditions similar to the normal operational pressures and temperatures of commercial PWR and CANDU plants. The normal operation modes of the HANARO fuel test loop are classified into loop shutdown, cold stand-by 1, cold stand-by 2, hot stand-by, and hot operation. The operation modes depend on the fission power of test fuels and the coolant temperature at the inlet of the in-pile test section. The HANARO must maintain a shutdown mode if the HANARO fuel test loop is loop shutdown, cold stand-by 1, cold stand-by 2, or hot stand-by. As the HANARO becomes power operation mode, the operation mode of the HANARO fuel test loop comes to hot operation from hot stand-by. The procedure for the HANARO fuel test loop consists of four main parts such as check of initial conditions, stat-up operation procedure, shutdown operation procedure, and check lists for operations. Several hot test operations ensure that the procedure is appropriate

  10. Water Hammer Test

    Science.gov (United States)

    2008-01-01

    [figure removed for brevity, see original site] Click on the image for the animation This video shows the propulsion system on an engineering model of NASA's Phoenix Mars Lander being successfully tested. Instead of fuel, water is run through the propulsion system to make sure that the spacecraft holds up to vibrations caused by pressure oscillations. The test was performed very early in the development of the mission, in 2005, at Lockheed Martin Space Systems, Denver. Early testing was possible because Phoenix's main structure was already in place from the 2001 Mars Surveyor program. The Phoenix Mission is led by the University of Arizona, Tucson, on behalf of NASA. Project management of the mission is by NASA's Jet Propulsion Laboratory, Pasadena, Calif. Spacecraft development is by Lockheed Martin Space Systems, Denver.

  11. Analysis of the LBLOCAs in the room 1 for the 3-pin fuel test loop

    International Nuclear Information System (INIS)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, Y. J.

    2004-12-01

    Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Large Break Loss of Coolant Accidents (LBLOCAs) in the Room 1 for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the LBLOCAs. The location of the pipe break is assumed at the downstream of the main cooling water pump and the upstream of the main cooler in the room 1. Double ended guillotine break is assumed for the large break loss of coolant accidents. The discharge coefficients of 0.1, 0.33, 0.67, 1.0 are investigated for the LBLOCAs. The maximum Peak Cladding Temperature (PCT) is predicted to be about 734.7 .deg. C for the PWR fuel test mode and 850.4 .deg. C for the CANDU fuel test mode respectively. The maximum peak cladding temperatures meet the design criterion of commercial PWR fuel that the maximum PCT is lower than 1204 .deg. C

  12. Investigation of Loop Seal Clearing Phenomena for the ATLAS SBLOCA Long Term Cooling Test using TRACE and MARS-KS

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Min Jeong; Park, M. H.; Marigomen Ralph; Sim, S. K. [Environment and Energy Technology, Daejeon (Korea, Republic of)

    2016-10-15

    During Design Certificate(DC) review of the APR1400, USNRC raised a long term cooling safety issue on the effect of loop seal clearing during cold leg Small Break Loss Of Coolant Accident(SBLOCA) due to relatively deep cross-over loop compared to the US PWRs. The objective of this study is thus to investigate the loop seal clearing phenomena during cold leg slot break SBLOCA long term cooling and resolve the safety issue on the SBLOCA long term cooling related to the APR1400 DC. TRACE and MARS-KS were used to predict the test results and to perform sensitivity studies for the SBLOCA loop seal clearing phenomena. The calculation shows that the TRACE code well predict the sequence of Test LTC-CL-04R. However, compared to the experiment, the TRACE over predicts the primary pressure due to smaller break flow prediction. MARS-KS well predicts major thermal hydraulic parameters during the transient with reasonable agreement. MARS-KS better predicts ATLAS LTC-CL-04R test data with a good agreement than the TRACE due to better prediction of the break flow. Overall, compared to the experiment, the TRACE and MARS-KS Codes show a discrepancy in predicting the loop seal clearing and reformation time. Both TRACE and MARS-KS correctly predicts core water level and fuel cladding temperatures. From this study, it can be said that even though APR1400 cross-over leg design has slightly deeper loop seals, the effect on the safety of the SBLOCA long term cooling is minimal compared to the SBLOCA cladding failure criteria. Further study on the SBLOCA loop seal clearing phenomena is needed.

  13. Performance of a Water Recirculation Loop Maintenance Device and Process for the Advanced Spacesuit Water Membrane Evaporator

    Science.gov (United States)

    Rector, Tony; Steele, John W.; Bue, Grant C.; Campbell, Colin; Makinen, Janice

    2012-01-01

    A water loop maintenance device and process to maintain the water quality of the Advanced Spacesuit Water Membrane Evaporation (SWME) water recirculation loop has been undergoing a performance evaluation. The SWME is a heat rejection device under development at the NASA Johnson Space Center to perform thermal control for advanced spacesuits. One advantage to this technology is the potential for a significantly greater degree of tolerance to contamination when compared to the existing Sublimator technology. The driver for the water recirculation maintenance device and process is to further enhance this advantage through the leveraging of fluid loop management lessons-learned from the International Space Station (ISS). A bed design that was developed for a Hamilton Sundstrand military application, and considered for a potential ISS application with the Urine Processor Assembly, provides a low pressure drop means for water maintenance in a recirculation loop. The bed design is coupled with high capacity ion exchange resins, organic adsorbents, and a cyclic methodology developed for the Extravehicular Mobility Unit (EMU) Transport Water loop. The maintenance process further leverages a sorbent developed for ISS that introduces a biocide in a microgravity-compatible manner for the Internal Active Thermal Control System (IATCS). The leveraging of these water maintenance technologies to the SWME recirculation loop is a unique demonstration of applying the valuable lessons learned on the ISS to the next generation of manned spaceflight Environmental Control and Life Support System (ECLSS) hardware. This

  14. Production circulator fabrication and testing for core flow test loop. Final report, Phase III

    Energy Technology Data Exchange (ETDEWEB)

    1981-05-01

    The performance testing of two production helium circulators utilizing gas film lubrication is described. These two centrifugal-type circulators plus an identical circulator prototype will be arranged in series to provide the helium flow requirements for the Core Flow Test Loop which is part of the Gas-Cooled Fast Breeder Reactor Program (GCFR) at the Oak Ridge National Laboratory. This report presents the results of the Phase III performance and supplemental tests, which were carried out by MTI during the period of December 18, 1980 through March 19, 1981. Specific test procedures are outlined and described, as are individual tests for measuring the performance of the circulators. Test data and run descriptions are presented.

  15. Production circulator fabrication and testing for core flow test loop. Final report, Phase III

    International Nuclear Information System (INIS)

    1981-05-01

    The performance testing of two production helium circulators utilizing gas film lubrication is described. These two centrifugal-type circulators plus an identical circulator prototype will be arranged in series to provide the helium flow requirements for the Core Flow Test Loop which is part of the Gas-Cooled Fast Breeder Reactor Program (GCFR) at the Oak Ridge National Laboratory. This report presents the results of the Phase III performance and supplemental tests, which were carried out by MTI during the period of December 18, 1980 through March 19, 1981. Specific test procedures are outlined and described, as are individual tests for measuring the performance of the circulators. Test data and run descriptions are presented

  16. A two-loop test of M(atrix) theory

    International Nuclear Information System (INIS)

    Becker, K.

    1997-01-01

    We consider the scattering of two Dirichlet zero-branes in M(atrix) theory. Using the formulation of M(atrix) theory in terms of ten-dimensional super Yang-Mills theory dimensionally reduced to (0+1) dimensions, we obtain the effective (velocity-dependent) potential describing these particles. At one loop we obtain the well-known result for the leading order of the effective potential V eff ∝v 4 /r 7 , where v and r are the relative velocity and distance between the two zero-branes, respectively. A calculation of the effective potential at two loops shows that no renormalizations of the v 4 term of the effective potential occur at this order. (orig.)

  17. Testing of cobalt-free alloys for valve applications using a special test loop

    International Nuclear Information System (INIS)

    Benhamou, C.

    1992-01-01

    Considering that use of cobalt alloys should be avoided as far as possible in PWR components, a programme aimed at establishing the performance of cobalt-free alloys has been performed for valve applications, where cobalt alloys are mainly used. Referring to past work, two types of cobalt-free alloys were selected: Ni-Cr-B-Si and Ni-Cr-Fe alloys. Cobalt-free valves' behaviour has been evaluated comparatively with cobalt valves by implementation of a programme in a special PWR test loop. At the issue of the loop test programme, which included endurance, thermal shock and erosion tests, cobalt-free alloys candidate to replace cobalt alloys are proposed in relation with valve type (globe valve and swing check valve). The following was established: (i) Colmonoy 4-26 (Ni-Cr-B-Si alloy) and Cenium Z20 (Ni-Cr-Fe alloy) deposited by plasma arc process were found suitable for use in 3inch swing check valves; (ii) for integral parts acting as guide rings, Nitronic 60 and Cesium Z20/698 were tested successfully; (iii) for small-bore components such as 2inch globe valves, no solution can yet be proposed; introduction of cobalt-free alloys is dependent on the development of automatic advanced arc surfacing techniques applied to small-bore components

  18. Alternative water chemistry for the primary loop of PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Henzel, N [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

    1997-02-01

    Advanced fuel element concepts (longer cycles, higher burnup, increased rod power) call for more reactivity binding capacity and, moreover, might produce higher void fractions, particularly in the hot channel. Thus, on the one hand, more alcalizing agent is needed to maintain a high coolant pH according to the approved ``modified boron-lithium mode of operation`` in the presence of more boric acid (chemical shim); on the other hand, increasing enrichment of coolant constituents due to local boiling (higher void fraction), which must not result in accelerated corrosion of fuel cladding and structural materials, imposes enhanced requirements on both, materials technology and water chemistry. At present, the use of boric acid enriched in B10 (the isotope effective in terms of reactivity control) appears to advantageously compromise in capturing more neutrons with less total boron while maintaining or even slightly reducing lithium concentrations at the same time. There is no feasible alternative for boric acid used as the chemical shim and for hydrogen gas as the reducing agent used to suppress oxygen formation by water radiolysis. Systematic screening as performed in phase 1 of a recent project proved potassium hydroxide to be the only potential candidate to favourably replace lithium 7 hydroxide as an alcalizing agent. Unfortunately, the results of pertinent comparative corrosion tests are not unambiguous, and available operational experience with potassium hydroxide in WWER plants is not readily applicable to western world-type PWR plants. Therefore, a switch-over from lithium to potassium can be envisaged only subsequent to a comprehensive qualification program which is planned to be the objective of phase 2 of the project. This program should also comprise zinc addition tests in order to confirm the alleged positive impact of this element on corrosion rates and activity buildup. (Abstract Truncated)

  19. Theoretical investigation of the performance of a novel loop heat pipe solar water heating system for use in Beijing, China

    International Nuclear Information System (INIS)

    Zhao Xudong; Wang Zhangyuan; Tang Qi

    2010-01-01

    A novel loop heat pipe (LHP) solar water heating system for typical apartment buildings in Beijing was designed to enable effective collection of solar heat, distance transport, and efficient conversion of solar heat into hot water. Taking consideration of the heat balances occurring in various parts of the loop, such as the solar absorber, heat pipe loop, heat exchanger and storage tank, a computer model was developed to investigate the thermal performance of the system. With the specified system structure, the efficiency of the solar system was found to be a function of its operational characteristics - working temperature of the loop heat pipe, water flow rate across the heat exchanger, and external parameters, including ambient temperature, temperature of water across the exchanger and solar radiation. The relationship between the efficiency of the system and these parameters was established, analysed and discussed in detail. The study suggested that the loop heat pipe should be operated at around 72 deg. C and the water across the heat exchanger should be maintained at 5.1 l/min. Any variation in system structure, i.e., glazing cover and height difference between the absorber and heat exchanger, would lead to different system performance. The glazing covers could be made using either borosilicate or polycarbonate, but borosilicate is to be preferred as it performs better and achieves higher efficiency at higher temperature operation. The height difference between the absorber and heat exchanger in the design was 1.9 m which is an adequate distance causing no constraint to heat pipe heat transfer. These simulation results were validated with the primary testing results.

  20. Dynamic Characterization of a Low Cost Microwave Water-Cut Sensor in a Flow Loop

    KAUST Repository

    Karimi, Muhammad Akram

    2017-03-31

    Inline precise measurement of water fraction in oil (i.e. water-cut [WC]) finds numerous applications in oil and gas industry. This paper presents the characterization of an extremely low cost, completely non-intrusive and full range microwave water-cut sensor based upon pipe conformable microwave T-resonator. A 10″ microwave stub based T-resonator has been implemented directly on the pipe surface whose resonance frequency changes in the frequency band of 90MHz–190MHz (111%) with changing water fraction in oil. The designed sensor is capable of detecting even small changes in WC with a resolution of 0.07% at low WC and 0.5% WC at high WC. The performance of the microwave WC sensor has been tested in an in-house flow loop. The proposed WC sensor has been characterized over full water-cut range (0%–100%) not only in vertical but also in horizontal orientation. The sensor has shown predictable response in both orientations with huge frequency shift. Moreover, flow rate effect has also been investigated on the proposed WC sensor’s performance and it has been found that the sensor’s repeatability is within 2.5% WC for variable flow rates.

  1. Results from Carbon Dioxide Washout Testing Using a Suited Manikin Test Apparatus with a Space Suit Ventilation Test Loop

    Science.gov (United States)

    Chullen, Cinda; Conger, Bruce; McMillin, Summer; Vonau, Walt; Kanne, Bryan; Korona, Adam; Swickrath, Mike

    2016-01-01

    NASA is developing an advanced portable life support system (PLSS) to meet the needs of a new NASA advanced space suit. The PLSS is one of the most critical aspects of the space suit providing the necessary oxygen, ventilation, and thermal protection for an astronaut performing a spacewalk. The ventilation subsystem in the PLSS must provide sufficient carbon dioxide (CO2) removal and ensure that the CO2 is washed away from the oronasal region of the astronaut. CO2 washout is a term used to describe the mechanism by which CO2 levels are controlled within the helmet to limit the concentration of CO2 inhaled by the astronaut. Accumulation of CO2 in the helmet or throughout the ventilation loop could cause the suited astronaut to experience hypercapnia (excessive carbon dioxide in the blood). A suited manikin test apparatus (SMTA) integrated with a space suit ventilation test loop was designed, developed, and assembled at NASA in order to experimentally validate adequate CO2 removal throughout the PLSS ventilation subsystem and to quantify CO2 washout performance under various conditions. The test results from this integrated system will be used to validate analytical models and augment human testing. This paper presents the system integration of the PLSS ventilation test loop with the SMTA including the newly developed regenerative Rapid Cycle Amine component used for CO2 removal and tidal breathing capability to emulate the human. The testing and analytical results of the integrated system are presented along with future work.

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  3. Conceptual Design of Forced Convection Molten Salt Heat Transfer Testing Loop

    Energy Technology Data Exchange (ETDEWEB)

    Manohar S. Sohal; Piyush Sabharwall; Pattrick Calderoni; Alan K. Wertsching; S. Brandon Grover

    2010-09-01

    This report develops a proposal to design and construct a forced convection test loop. A detailed test plan will then be conducted to obtain data on heat transfer, thermodynamic, and corrosion characteristics of the molten salts and fluid-solid interaction. In particular, this report outlines an experimental research and development test plan. The most important initial requirement for heat transfer test of molten salt systems is the establishment of reference coolant materials to use in the experiments. An earlier report produced within the same project highlighted how thermophysical properties of the materials that directly impact the heat transfer behavior are strongly correlated to the composition and impurities concentration of the melt. It is therefore essential to establish laboratory techniques that can measure the melt composition, and to develop purification methods that would allow the production of large quantities of coolant with the desired purity. A companion report describes the options available to reach such objectives. In particular, that report outlines an experimental research and development test plan that would include following steps: •Molten Salts: The candidate molten salts for investigation will be selected. •Materials of Construction: Materials of construction for the test loop, heat exchangers, and fluid-solid corrosion tests in the test loop will also be selected. •Scaling Analysis: Scaling analysis to design the test loop will be performed. •Test Plan: A comprehensive test plan to include all the tests that are being planned in the short and long term time frame will be developed. •Design the Test Loop: The forced convection test loop will be designed including extensive mechanical design, instrument selection, data acquisition system, safety requirements, and related precautionary measures. •Fabricate the Test Loop. •Perform the Tests. •Uncertainty Analysis: As a part of the data collection, uncertainty analysis will

  4. High pressure experimental water loop; Cellule experimentale a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Grenon, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Terny, M [Societe Grenobloise d' Etudes et d' Applications Hydrauliques, 38 (France)

    1958-07-01

    A high pressure experimental water loop has been made for studying the detection and evolution of cladding failure in a pressurized reactor. The loop has been designed for a maximum temperature of 360 deg. C, a maximum of 160 kg/cm{sup 2} and flow rates up to 5 m{sup 3}/h. The entire loop consists of several parts: a main circuit with a canned rotor circulation pump, steam pressurizer, heating tubes, two hydro-cyclones (one de-gasser and one decanter) and one tubular heat exchanger; a continuous purification loop, connected in parallel, comprising pressure reducing valves and resin pots which also allow studies of the stability of resins under pressure, temperature and radiation; following the gas separator is a gas loop for studying the recombination of the radiolytic gases in the steam phase. The preceding circuits, as well as others, return to a low pressure storage circuit. The cold water of the low pressure storage flask is continuously reintroduced into the high pressure main circuit by means of a return pump at a maximum head of 160 kg /cm{sup 2}, and adjusted to the pressurizer level. This loop is also a testing bench for the tight high pressure apparatus. The circulating pump and the connecting flanges (Oak Ridge type) are water-tight. The feed pump and the pressure reducing valves are not; the un-tight ones have a system of leak recovery. To permanently check the tightness the circuit has been fitted with a leak detection system (similar to the HRT one). (author)Fren. [French] Une cellule experimentale a eau sous pression a ete realisee pour l'etude de l'evolution et de la detection de ruptures de gaines dans le cas d'un reacteur pressurise. Les performances de ce circuit sont: 350 deg. C maximum pour la temperature et 160 kg/cm{sup 2} maximum pour la pression, le debit pouvant atteindre 5 m{sup 3}/h. Le circuit complet est constitue de plusieurs parties: un circuit principal avec pompe de circulation a rotor immerge, preiseur a vapeur, tubes chaudieres

  5. UF/sub 6/ test loop for evaluation and implementation of international enrichment plant safeguards

    International Nuclear Information System (INIS)

    Cooley, J.N.; Fields, L.W.; Swindle, D.W. Jr.

    1987-01-01

    A functional test loop capable of simulating UF/sub 6/ flows, pressures, and pipe deposits characteristic of gas centrifuge enrichment plant piping has been designed and fabricated by the Enrichment Safeguards Program of Martin Marietta Energy Systems, Inc., for use by the International Atomic Energy Agency (IAEA) at its Safeguards Analytical Laboratory in Seibersdorf, Austria. The purpose of the test loop is twofold: (1) to enable the IAEA to evaluate and to calibrate enrichment safeguards measurement instrumentation to be used in limited frequency-unannounced access (LFUA) inspection strategy measurements at gas centrifuge enrichment plants and (2) to train IAEA inspectors in the use of such instrumentation. The test loop incorporates actual sections of cascade header pipes from the centrifuge enrichment plants subject to IAEA inspections. The test loop is described, applications for its use by the IAEA are detailed, and results from an initial demonstration session using the test loop are summarized. By giving the IAEA the in-house capability to evaluate LFUA inspection strategy approaches, to develop inspection procedures, to calibrate instrumentation, and to train inspectors, the UF/sub 6/ cascade header pipe test loop will contribute to the IAEA's success in implementing LFUA strategy inspections at gas centrifuge enrichment facilities subject to international safeguards inspections

  6. Test of Flow Characteristics in Tubular Fuel Assembly I - Establishment of test loop and measurement validation test

    International Nuclear Information System (INIS)

    Park, Jong Hark; Chae, H. T.; Park, C.; Kim, H.

    2005-12-01

    Tubular type fuel has been developed as one of candidates for Advanced HANARO Reactor(AHR). It is necessary to test the flow characteristics such as velocity in each flow channels and pressure drop of tubular type fuel. A hydraulic test-loop to examine the hydraulic characteristics for a tubular type fuel has been designed and constructed. It consists of three parts; a) piping-loop including pump and motor, magnetic flow meter and valves etc, b) test-section part where a simulated tubular type fuel is located, and 3) data acquisition system to get reading signals from sensors or instruments. In this report, considerations during the design and installation of the facility and the selection of data acquisition sensors and instruments are described in detail. Before doing the experiment to measure the flow velocities in flow channels, a preliminary tests have been done for measuring the coolant velocities using pitot-tube and for validating the measurement accuracy as well. Local velocities of the radial direction in circular tubes are measured at regular intervals of 60 degrees by three pitot-tubes. Flow rate inside the circular flow channel can be obtained by integrating the velocity distribution in radial direction. The measured flow rate was compared to that of magnetic flow meter. According to the results, two values had a good agreement, which means that the measurement of coolant velocity by using pitot-tube and the flow rate measured by the magnetic flow meter are reliable. Uncertainty analysis showed that the error of velocity measurement by pitot-tube is less than ±2.21%. The hydraulic test-loop also can be adapted to others such as HANARO 18 and 36 fuel, in-pile system of FTL(Fuel Test Loop), etc

  7. Water Pipeline Network Analysis Using Simultaneous Loop Flow ...

    African Journals Online (AJOL)

    2013-03-01

    Mar 1, 2013 ... *Department of Mechanical Engineering, Michael Okpara University of Agriculture, Umudike, Nigeria ... significant fluid acceleration, the behavior of a network can be ... world water day centers on water and food security as.

  8. Coral mucus fuels the sponge loop in warm- and cold-water coral reef ecosystems.

    Science.gov (United States)

    Rix, Laura; de Goeij, Jasper M; Mueller, Christina E; Struck, Ulrich; Middelburg, Jack J; van Duyl, Fleur C; Al-Horani, Fuad A; Wild, Christian; Naumann, Malik S; van Oevelen, Dick

    2016-01-07

    Shallow warm-water and deep-sea cold-water corals engineer the coral reef framework and fertilize reef communities by releasing coral mucus, a source of reef dissolved organic matter (DOM). By transforming DOM into particulate detritus, sponges play a key role in transferring the energy and nutrients in DOM to higher trophic levels on Caribbean reefs via the so-called sponge loop. Coral mucus may be a major DOM source for the sponge loop, but mucus uptake by sponges has not been demonstrated. Here we used laboratory stable isotope tracer experiments to show the transfer of coral mucus into the bulk tissue and phospholipid fatty acids of the warm-water sponge Mycale fistulifera and cold-water sponge Hymedesmia coriacea, demonstrating a direct trophic link between corals and reef sponges. Furthermore, 21-40% of the mucus carbon and 32-39% of the nitrogen assimilated by the sponges was subsequently released as detritus, confirming a sponge loop on Red Sea warm-water and north Atlantic cold-water coral reefs. The presence of a sponge loop in two vastly different reef environments suggests it is a ubiquitous feature of reef ecosystems contributing to the high biogeochemical cycling that may enable coral reefs to thrive in nutrient-limited (warm-water) and energy-limited (cold-water) environments.

  9. Reflooding Experiment on BETA Test Loop: The Effects of Inlet Temperature on the Rewetting Velocity

    International Nuclear Information System (INIS)

    Khairul H; Anhar R Antariksawan; Edy Sumarno; Kiswanta; Giarno; Joko P; Ismu Handoyo

    2003-01-01

    Loss of Coolant Accident (LOCA) on Nuclear Reactor Plant is an important topic because this condition is a severe accident that can be postulated. The phenomenon of LOCA on Pressurized Water Reactor (PWR) can be divided in three stages, e.g.: blowdown, refill and reflood. In the view of Emergency Coolant System evaluation, the reflood is the most important stage. In this stage, an injection of emergency water coolant must be done in a way that the core can be flooded and the overheating can be avoid. The experiment of rewetting on BETA Test Loop had been conducted. The experiment using one heated rod of the test section to study effects of inlet temperature on the wetting velocity. Results of the series of experiments on 2,5 lt/min flow rate and variable of temperature : 28 o C, 38 o C, 50 o C, 58 o C it was noticed that for 58 o C inlet temperature of test section and 572 o C rod temperature the rewetting phenomenon has been observed. The time of refill was 32.81 sec and time of rewetting was 42.87 sec. (author)

  10. Determination of total flow rate and flow rate of every operating branch in commissioning of heavy water loop for ARR-2

    International Nuclear Information System (INIS)

    Han Yan

    1997-01-01

    The heavy water loop (i,e, RCS) for ARR-2 in Algeria is a complex loop. Flow regulating means are not provided by the design in order to operate the reactor safely and simplify operating processes. How to determine precisely the orifice diameters of resistance parts for the loop is a key point for decreasing deviation between practical and design flow rates. Commissioning tests shall ensure that under every one of combined operating modes for the pumps, total coolant flow rate is about the same (the number of pumps operating in parallel is the same) and is consistent with design requirement, as well as the distribution of coolant flow rate to every branch is uniform. The flow Determination is divided into two steps. First and foremost, corresponding resistance part at each pump outlet is determined in commissioning test of shorted heavy water loop with light water, so that the problem about uniform distribution of the flow rate to each branch is solved, Secondly, resistance part at the reactor inlet is determined in commissioning test of heavy water loop connected with the vessel, so that the problem about that total heavy water flow rate is within optimal range is solved. According to practical requirements of the project, a computer program of hydraulic calculation and analysis for heavy water loop has been developed, and hydraulic characteristics test for a part of loop has been conducted in order to correct calculation error. By means of program calculation combining with tests in site, orifice diameters of 9 resistance parts has been determined rapidly and precisely and requirements of design and operation has been met adequately

  11. PWR type reactor equipped with a primary circuit loop water level gauge

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro.

    1990-01-01

    The time of lowering a water level to less than the position of high temperature side pipeway nozzle has been rather delayed because of the swelling of mixed water level due to heat generation of the reactor core. Further, there has been a certain restriction for the installation, maintenance and adjustment of a water level gauge since it is at a position under high radiation exposure. Then, a differential pressure type water level gauge with temperature compensation is disposed at a portion below a water level gauge of a pressurizer and between the steam generator exit plenum and the lower end of the loop seal. Further, a similar water level system is disposed to all of the loops of the primary circulation circuits. In a case that the amount of water contained in a reactor container should decreased upon occurrence of loss of coolant accidents caused by small rupture and stoppage of primary circuit pumps, lowering of the water level preceding to the lowering of the water level in the reactor core is detected to ensure the amount of water. Since they are disposed to all of the loops and ensure the excess margin, reliability for the detection of the amount of contained water can be improved by averaging time for the data of the water level and averaging the entire systems, even when there are vibrations in the fluid or pressure in the primary circuit. (N.H.)

  12. Test results of reliable and very high capillary multi-evaporators / condenser loop

    Energy Technology Data Exchange (ETDEWEB)

    Van Oost, S; Dubois, M; Bekaert, G [Societe Anonyme Belge de Construction Aeronautique - SABCA (Belgium)

    1997-12-31

    The paper present the results of various SABCA activities in the field of two-phase heat transport system. These results have been based on a critical review and analysis of the existing two-phase loop and of the future loop needs in space applications. The research and the development of a high capillary wick (capillary pressure up to 38 000 Pa) are described. These activities have led towards the development of a reliable high performance capillary loop concept (HPCPL), which is discussed in details. Several loop configurations mono/multi-evaporators have been ground tested. The presented results of various tests clearly show the viability of this concept for future applications. Proposed flight demonstrations as well as potential applications conclude this paper. (authors) 7 refs.

  13. Test results of reliable and very high capillary multi-evaporators / condenser loop

    Energy Technology Data Exchange (ETDEWEB)

    Van Oost, S.; Dubois, M.; Bekaert, G. [Societe Anonyme Belge de Construction Aeronautique - SABCA (Belgium)

    1996-12-31

    The paper present the results of various SABCA activities in the field of two-phase heat transport system. These results have been based on a critical review and analysis of the existing two-phase loop and of the future loop needs in space applications. The research and the development of a high capillary wick (capillary pressure up to 38 000 Pa) are described. These activities have led towards the development of a reliable high performance capillary loop concept (HPCPL), which is discussed in details. Several loop configurations mono/multi-evaporators have been ground tested. The presented results of various tests clearly show the viability of this concept for future applications. Proposed flight demonstrations as well as potential applications conclude this paper. (authors) 7 refs.

  14. Optimal Pipe Size Design for Looped Irrigation Water Supply System Using Harmony Search: Saemangeum Project Area

    Science.gov (United States)

    Lee, Ho Min; Sadollah, Ali

    2015-01-01

    Water supply systems are mainly classified into branched and looped network systems. The main difference between these two systems is that, in a branched network system, the flow within each pipe is a known value, whereas in a looped network system, the flow in each pipe is considered an unknown value. Therefore, an analysis of a looped network system is a more complex task. This study aims to develop a technique for estimating the optimal pipe diameter for a looped agricultural irrigation water supply system using a harmony search algorithm, which is an optimization technique. This study mainly serves two purposes. The first is to develop an algorithm and a program for estimating a cost-effective pipe diameter for agricultural irrigation water supply systems using optimization techniques. The second is to validate the developed program by applying the proposed optimized cost-effective pipe diameter to an actual study region (Saemangeum project area, zone 6). The results suggest that the optimal design program, which applies an optimization theory and enhances user convenience, can be effectively applied for the real systems of a looped agricultural irrigation water supply. PMID:25874252

  15. Optimal Pipe Size Design for Looped Irrigation Water Supply System Using Harmony Search: Saemangeum Project Area

    Directory of Open Access Journals (Sweden)

    Do Guen Yoo

    2015-01-01

    Full Text Available Water supply systems are mainly classified into branched and looped network systems. The main difference between these two systems is that, in a branched network system, the flow within each pipe is a known value, whereas in a looped network system, the flow in each pipe is considered an unknown value. Therefore, an analysis of a looped network system is a more complex task. This study aims to develop a technique for estimating the optimal pipe diameter for a looped agricultural irrigation water supply system using a harmony search algorithm, which is an optimization technique. This study mainly serves two purposes. The first is to develop an algorithm and a program for estimating a cost-effective pipe diameter for agricultural irrigation water supply systems using optimization techniques. The second is to validate the developed program by applying the proposed optimized cost-effective pipe diameter to an actual study region (Saemangeum project area, zone 6. The results suggest that the optimal design program, which applies an optimization theory and enhances user convenience, can be effectively applied for the real systems of a looped agricultural irrigation water supply.

  16. Water loop heat pump and its characteristics%水环热泵及其特点

    Institute of Scientific and Technical Information of China (English)

    冯润娣

    2011-01-01

    This paper introduces the definition and working principle of water loop beat pump, and describes its development condition at home and abroad. Through analyzing the characteristics of water loop heat pump, it points out the water loop heat pump is a kind of economic, energy saving and environment protection air-conditioning system, and has great application prospect.%阐述了水环热泵的定义及工作原理,介绍了水环热泵在国内外的发展概况,通过分析水环热泵的特点,指出其是一种经济、节能、环保的空调系统,有着广阔的应用前景。

  17. Review of application code and standards for mechanical and piping design of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. HANARO fuel test loop was designed in accordance with the same code and standards of nuclear power plant because HANARO FTL will be operated the high pressure and temperature same as nuclear power plant operation conditions. The objective of this study is to confirm the propriety of application code and standards for mechanical and piping of HANARO fuel test loop and to decide the technical specification of FTL systems. (author). 18 refs., 8 tabs., 6 figs.

  18. ECP measurements under neutron and gamma ray in in-pile loop and their data evaluation by water radiolysis calculations

    Energy Technology Data Exchange (ETDEWEB)

    Hanawa, S.; Nakamura, T.; Uchida, S. [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan); Kus, P.; Vsolak, R.; Kysela, J. [Nuclear Research Inst. Rez plc, Husinec - Rez (Czech Republic)

    2010-07-01

    In order to establish reliable electrochemical corrosion potential (ECP) sensors for applying in reactor core peripherals of power plants, performance tests of sensors under irradiation were carried out in the in-pile loop of the experimental reactor, LVR-15, at the Nuclear Research Institute (NRI) in Czech Republic. Responses of different kinds of sensors under neutron and gamma irradiation conditions have been compared each other. Corrosive conditions along the in-pile loop were calculated by water radiolysis calculation code, WRAC-J and calculated corrosive conditions were compared with the measured results. As a result of the evaluation, it was confirmed that the ECP sensors could be applied to irradiation conditions of reactor peripherals, while the water radiolysis model could be also applied for evaluation of corrosive conditions of reactor peripherals. (author)

  19. Determination of air-loop volume and radon partition coefficient for measuring radon in water sample.

    Science.gov (United States)

    Lee, Kil Yong; Burnett, William C

    A simple method for the direct determination of the air-loop volume in a RAD7 system as well as the radon partition coefficient was developed allowing for an accurate measurement of the radon activity in any type of water. The air-loop volume may be measured directly using an external radon source and an empty bottle with a precisely measured volume. The partition coefficient and activity of radon in the water sample may then be determined via the RAD7 using the determined air-loop volume. Activity ratios instead of absolute activities were used to measure the air-loop volume and the radon partition coefficient. In order to verify this approach, we measured the radon partition coefficient in deionized water in the temperature range of 10-30 °C and compared the values to those calculated from the well-known Weigel equation. The results were within 5 % variance throughout the temperature range. We also applied the approach for measurement of the radon partition coefficient in synthetic saline water (0-75 ppt salinity) as well as tap water. The radon activity of the tap water sample was determined by this method as well as the standard RAD-H 2 O and BigBottle RAD-H 2 O. The results have shown good agreement between this method and the standard methods.

  20. Determination of air-loop volume and radon partition coefficient for measuring radon in water sample

    International Nuclear Information System (INIS)

    Kil Yong Lee; Burnett, W.C.

    2013-01-01

    A simple method for the direct determination of the air-loop volume in a RAD7 system as well as the radon partition coefficient was developed allowing for an accurate measurement of the radon activity in any type of water. The air-loop volume may be measured directly using an external radon source and an empty bottle with a precisely measured volume. The partition coefficient and activity of radon in the water sample may then be determined via the RAD7 using the determined air-loop volume. Activity ratios instead of absolute activities were used to measure the air-loop volume and the radon partition coefficient. In order to verify this approach, we measured the radon partition coefficient in deionized water in the temperature range of 10-30 deg C and compared the values to those calculated from the well-known Weigel equation. The results were within 5 % variance throughout the temperature range. We also applied the approach for measurement of the radon partition coefficient in synthetic saline water (0-75 ppt salinity) as well as tap water. The radon activity of the tap water sample was determined by this method as well as the standard RAD-H 2 O and BigBottle RAD-H 2 O. The results have shown good agreement between this method and the standard methods. (author)

  1. Facility for generating crew waste water product for ECLSS testing

    Science.gov (United States)

    Buitekant, Alan; Roberts, Barry C.

    1990-01-01

    An End-use Equipment Facility (EEF) has been constructed which is used to simulate water interfaces between the Space Station Freedom Environmental Control and Life Support Systems (ECLSS) and man systems. The EEF is used to generate waste water to be treated by ECLSS water recovery systems. The EEF will also be used to close the water recovery loop by allowing test subjects to use recovered hygiene and potable water during several phases of testing. This paper describes the design and basic operation of the EEF.

  2. Summary of TRUEX Radiolysis Testing Using the INL Radiolysis Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Dean R. Peterman; Lonnie G. Olson; Rocklan G. McDowell; Gracy Elias; Jack D. Law

    2012-03-01

    The INL radiolysis and hydrolysis test loop has been used to evaluate the effects of hydrolytic and radiolytic degradation upon the efficacy of the TRUEX flowsheet for the recovery of trivalent actinides and lanthanides from acidic solution. Repeated irradiation and subsequent re-conditioning cycles did result in a significant decrease in the concentration of the TBP and CMPO extractants in the TRUEX solvent and a corresponding decrease in americium and europium extraction distributions. However, the build-up of solvent degradation products upon {gamma}-irradiation, had little impact upon the efficiency of the stripping section of the TRUEX flowsheet. Operation of the TRUEX flowsheet would require careful monitoring to ensure extraction distributions are maintained at acceptable levels.

  3. Loop capabilities in Rez for water chemistry and corrosion control of cladding and in-core components

    International Nuclear Information System (INIS)

    Kysela, J.; Zmitko, M.; Srank, J.; Vsolak, R.

    1999-01-01

    Main characteristics of LVR-15 research reactor and its irradiation facilities are presented. For testing of cladding, internals and RPV materials specialised loop are used. There are now five high pressure loops modelling PWR, WWER or BWR water environment and chemistry. Loops can be connected with instrumented in-pile channels enable slow strain rate testing, 1CT or 2CT specimens loading and electrically heated rods exposition. Reactor dosimetry including neutronic parameters measurements and calculations and mock-up experiments are used. Water chemistry control involves gas (O 2 , H 2 ) dosing system, Orbisphere H 2 /O 2 measurement, electrochemical potential (ECP) measurements and specialised analytical chemistry laboratory. For cladding corrosion studies in-pile channels with four electrically heated rods with heat flux up to 100 W/cm 2 , void fraction 5 % at the outlet, inlet temperature 320 deg. C and flow velocity 3 m/s were development and tested. For corrosion layer investigation there is eddy current measurements and PIE techniques which use crud thickness measurement, chemical analyses of the crud, optical metallography, hydrogen analysis, SEM and TEM. (author)

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  5. Thermal analysis of lithium cooled natural circulation loop module for fuel rod testing in the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Eyler, L.L.; Kim, D.; Stover, R.L.; Beaver, T.R.

    1987-01-01

    Maximum heat removal capability of a lithium cooled natural circulation fuel rod test module design is determined. Loop geometry is optimized within limitations of design specifications for nominal operation temperatures, materials, and test module environment. Results provide test module operation limits and range of potential uncertainties. 3 refs., 12 figs

  6. Portland, Oregon Test Data Set Arterial Loop Detector Data

    Data.gov (United States)

    Department of Transportation — This set of data files was acquired under USDOT FHWA cooperative agreement DTFH61-11-H-00025 as one of the four test data sets acquired by the USDOT Data Capture and...

  7. Portland, Oregon Test Data Set Freeway Loop Detector Data

    Data.gov (United States)

    Department of Transportation — This set of data files was acquired under USDOT FHWA cooperative agreement DTFH61-11-H-00025 as one of the four test data sets acquired by the USDOT Data Capture and...

  8. Validity and Reliability of Orthodontic Loops between Mechanical Testing and Computer Simulation: An Finite Element Method Study

    Directory of Open Access Journals (Sweden)

    Gaurav Sepolia

    2014-01-01

    Full Text Available The magnitude and direction of orthodontic force is one of the essential concerns of orthodontic tooth movements. Excessive force may cause root resorption and mobility of the tooth, whereas low force level may results in prolonged treatment. The addition of loops allows the clinician to more accurately achieve the desired results. Aims and objectives: The purpose of the study was to evaluate the validity and reliability of orthodontic loops between mechanical testing and computer simulation. Materials and methods: Different types of loops were taken and divided into four groups: The Teardrop loop, Opus loop, L loop and T loop. These were artificially activated for multiple lengths and studied using the FEM. Results: The Teardrop loop showed the highest force level, and there is no significant difference between mechanical testing and computer simulation.

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  10. Unique rod lens/video system designed to observe flow conditions in emergency core coolant loops of pressurized water reactors

    International Nuclear Information System (INIS)

    Carter, G.W.

    1979-01-01

    Techniques and equipment are described which are used for video recordings of the single- and two-phase fluid flow tests conducted with the PKL Spool Piece Measurement System designed by Lawrence Livermore Laboratory and EG and G Inc. The instrumented spool piece provides valuable information on what would happen in pressurized water reactor emergency coolant loops should an accident or rupture result in loss of fluid. The complete closed-circuit television video system, including rod lens, light supply, and associated spool mounting fixtures, is discussed in detail. Photographic examples of test flows taken during actual spool piece system operation are shown

  11. Water conservation and improved production efficiency using closed-loop evaporative cooling systems

    Energy Technology Data Exchange (ETDEWEB)

    Marchetta, C. [Niagara Blower Co., Buffalo, NY (United States)

    2009-07-01

    This paper described wet surface air coolers (WSAC) that can be used in refineries and hydrocarbon processing plants to address water use issues. These closed-loop evaporative cooling systems are a cost-effective technology for both heat transfer and water conservation. WSACs can help deliver required cooling water temperatures and improve plant performance while using water streams currently considered to be unusable with conventional towers and heat exchangers. WSACs are versatile and can provide solutions to water use, water quality, and outlet temperature. The benefits of the WSAC include capital cost savings, reduced system pressures, lower carbon footprint, and the ability to use poor quality water as makeup. Water makeup can be blowdown from other equipment, plant effluent, reclaimed water, produced water, flue gas desulphurization (FGD) wastewater, and even seawater. Units can be manufactured with a wide variety of materials depending on water quality, water treatment, and cycles of concentration. This paper also provided comparisons to other alternative technologies, capital and operating cost savings, and site specific case studies. Two other system designs can accommodate closed-loop heat transfer applications, notably an open tower with a heat exchanger and a dry, air-cooled system. A WSAC system is an efficient and effective heat rejection technology for several reasons. The WSAC cooler or condenser utilizes latent cooling, which is far more efficient than sensible cooling. This means that a WSAC system can cool the same heat load with a smaller footprint than all-dry systems. 6 figs.

  12. Field Test of Boiler Primary Loop Temperature Controller

    Energy Technology Data Exchange (ETDEWEB)

    Glanville, P. [Partnership for Advanced Residential Retrofit, Des Plaines, IL (United States); Rowley, P. [Partnership for Advanced Residential Retrofit, Des Plaines, IL (United States); Schroeder, D. [Partnership for Advanced Residential Retrofit, Des Plaines, IL (United States); Brand, L. [Partnership for Advanced Residential Retrofit, Des Plaines, IL (United States)

    2014-09-01

    Beyond these initial system efficiency upgrades are an emerging class of Advanced Load Monitoring (ALM) aftermarket controllers that dynamically respond to the boiler load, with claims of 10% to 30% of fuel savings over a heating season. For hydronic boilers specifically, these devices perform load monitoring, with continuous measurement of supply and, in some cases, return water temperatures. Energy savings from these ALM controllers are derived from dynamic management of the boiler differential, where a microprocessor with memory of past boiler cycles prevents the boiler from firing for a period of time, to limit cycling losses and inefficient operation during perceived low load conditions. These differ from OTR controllers, which vary boiler setpoint temperatures with ambient conditions while maintaining a fixed differential.

  13. Design of Supercritical Carbon Dioxide Compressor Testing Loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seong Gu; Lee, Jeong Ik; Ahn, Yoon Han; Lee, Je Kyoung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Yoon, Ho Joon; Addad, Yacine [KUSTAR, Abu Dhabi (United Arab Emirates)

    2012-05-15

    For small and medium-sized reactors and Generation IV reactors such as sodium-cooled fast reactor are recently under development actively. The supercritical CO{sub 2} Brayton cycle is considered as an attractive cycle for the above mentioned nuclear systems. This is because the supercritical CO{sub 2} Brayton cycle (S-CO{sub 2} cycle) is especially effective to reduce the volume of power generation system, which occupies 1.5{approx}2 times more space than the primary nuclear system in general. Comparing to traditional water-vapor cycle and helium cycle, the S-CO{sub 2} system has relatively much less volume and component size. Therefore, S-CO{sub 2} cycle can be used for many purposes such as nuclear ship propulsion where volume requirement is strict, or a small nuclear reactor when it is constructed on geographically limited area

  14. SRF cavity testing using a FPGA Self Excited Loop

    Energy Technology Data Exchange (ETDEWEB)

    Ben-Zvi, I. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2017-08-30

    Various authors have previously studied the theory and practice of cavity testing, notably an extensive treatment by Powers [1] and Padamsee [2]. The advent of the digital Low Level RF (LLRF) electronics based on Field Programmable Logic Arrays (FPGA) provides various improvements over the rather complex systems used in the past as well as enabling new measurement techniques.In this document we reintroduce a technique that seems to have fallen out of practice in recent times, that is obtaining the coupling constant β through measurements from just one port, the reflected power port, of the directional coupler placed in front of the cavity.

  15. Field Test of Boiler Primary Loop Temperature Controller

    Energy Technology Data Exchange (ETDEWEB)

    Glanville, P.; Rowley, P.; Schroeder, D.; Brand, L.

    2014-09-01

    Beyond these initial system efficiency upgrades are an emerging class of Advanced Load Monitoring (ALM) aftermarket controllers that dynamically respond to the boiler load, with claims of 10% to 30% of fuel savings over a heating season. For hydronic boilers specifically, these devices perform load monitoring, with continuous measurement of supply and in some cases return water temperatures. Energy savings from these ALM controllers are derived from dynamic management of the boiler differential, where a microprocessor with memory of past boiler cycles prevents the boiler from firing for a period of time, to limit cycling losses and inefficient operation during perceived low load conditions. These differ from OTR controllers, which vary boiler setpoint temperatures with ambient conditions while maintaining a fixed differential. PARR installed and monitored the performance of one type of ALM controller, the M2G from Greffen Systems, at multifamily sites in the city of Chicago and its suburb Cary, IL, both with existing OTR control. Results show that energy savings depend on the degree to which boilers are over-sized for their load, represented by cycling rates. Also savings vary over the heating season with cycling rates, with greater savings observed in shoulder months. Over the monitoring period, over-sized boilers at one site showed reductions in cycling and energy consumption in line with prior laboratory studies, while less over-sized boilers at another site showed muted savings.

  16. Sensitivity Study of the Peak Cladding Temperature for the Pipe Break Accidents of the 3-Pin Fuel Test Loop

    International Nuclear Information System (INIS)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, H. R.

    2005-12-01

    The effect of the thermal hydraulic operation parameters, the stroke times of safety-related valves, the node number of test fuel for MARS modeling, and the axial power distribution on the peak cladding temperature (PCT) has been investigated for the loss of coolant accident of the 3-pin fuel test loop. The thermal hydraulic operation parameters investigated are the thermal power of the fuel test loop and the flow rate, temperature, and pressure of the main cooling water. The effect of the thermal power and the coolant temperature on the peak cladding temperature is dominant as compared with that of the coolant flow rate and pressure. The maximum PCT increases up to about 34.3K for the room 1 LOCA when the thermal power increase by 5% of the normal operation power and decreases up to about 38.9K for the room 1 LOCA when the coolant temperature decrease by 2% of the normal operation temperature. The effect of the stroke time of the loop isolation valves on the PCT is also dominant. However the effect of the stroke time of the safety injection valves and depressurization vent valves are negligible. Especially the maximum PCT increases up to 25.7K with the increase of the design stroke time of the cold leg loop isolation valve by 13% and decreases up to 25.1K with the decrease of the design stroke time by 13%. The maximum PCT increases by 3.3K as the number of nodes increases from 7 to 14 for the MARS model of test fuel. Three different axial power distributions are also investigated. The maximum PCT occurs for the room 1 LOCA in case the peak power is shifted to the downstream by 20cm

  17. Complex astrophysical experiments relating to jets, solar loops, and water ice dusty plasma

    Science.gov (United States)

    Bellan, P. M.; Zhai, X.; Chai, K. B.; Ha, B. N.

    2015-10-01

    > Recent results of three astrophysically relevant experiments at Caltech are summarized. In the first experiment magnetohydrodynamically driven plasma jets simulate astrophysical jets that undergo a kink instability. Lateral acceleration of the kinking jet spawns a Rayleigh-Taylor instability, which in turn spawns a magnetic reconnection. Particle heating and a burst of waves are observed in association with the reconnection. The second experiment uses a slightly different setup to produce an expanding arched plasma loop which is similar to a solar corona loop. It is shown that the plasma in this loop results from jets originating from the electrodes. The possibility of a transition from slow to fast expansion as a result of the expanding loop breaking free of an externally imposed strapping magnetic field is investigated. The third and completely different experiment creates a weakly ionized plasma with liquid nitrogen cooled electrodes. Water vapour injected into this plasma forms water ice grains that in general are ellipsoidal and not spheroidal. The water ice grains can become quite long (up to several hundred microns) and self-organize so that they are evenly spaced and vertically aligned.

  18. Design criteria of out-pile system of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1997-07-01

    The objective of HANARO aims at the development and localization of nuclear technologies through the engineering tests. Thus it is very important the design and installation of the irradiation test facilities to be installed at the irradiation hole for verification test of the fuel performance are in connection with maximization of the utilization of HANARO. The principle subjects of this study are to presend and informed the detail design criteria and technical specification of out-pile system of HANARO fuel test loop for the developing of the fuel and reactor material. This results will become guidance for the planning of the irradiation testing using the HANARO fuel test loop. (author). 16 refs., 31 tabs., 9 figs.

  19. Efforts to Reduce International Space Station Crew Maintenance Time in the Management of the Extravehicular Mobility Unit Transport Loop Water Quality

    Science.gov (United States)

    Etter,David; Rector, Tony; Boyle, robert; Zande, Chris Vande

    2012-01-01

    The EMU (Extravehicular Mobility Unit) contains a semi-closed-loop re-circulating water circuit (Transport Loop) to absorb heat into a LCVG (Liquid Coolant and Ventilation Garment) worn by the astronaut. A second, single-pass water circuit (Feed-water Loop) provides water to a cooling device (Sublimator) containing porous plates, and that water sublimates through the porous plates to space vacuum. The cooling effect from the sublimation of this water translates to a cooling of the LCVG water that circulates through the Sublimator. The quality of the EMU Transport Loop water is maintained through the use of a water processing kit (ALCLR - Airlock Cooling Loop Remediation) that is used to periodically clean and disinfect the water circuit. Opportunities to reduce crew time associated with ALCLR operations include a detailed review of the historical water quality data for evidence to support an extension to the implementation cycle. Furthermore, an EMU returned after 2-years of use on the ISS (International Space Station) is being used as a test bed to evaluate the results of extended and repeated ALCLR implementation cycles. Finally, design, use and on-orbit location enhancements to the ALCLR kit components are being considered to allow the implementation cycle to occur in parallel with other EMU maintenance and check-out activities, and to extend the life of the ALCLR kit components. These efforts are undertaken to reduce the crew-time and logistics burdens for the EMU, while ensuring the long-term health of the EMU water circuits for a post- Shuttle 6-year service life.

  20. Chemical decontamination for decommissioning purposes. (Vigorous decontamination tests of steel samples in a special test loop)

    International Nuclear Information System (INIS)

    Bregani, F.; Pascali, R.; Rizzi, R.

    1984-01-01

    The aim of the research activities described was to develop vigorous decontamination techniques for decommissioning purposes, taking into account the cost of treatment of the radwaste, to achieve possibly unrestricted release of the treated components, and to obtain know-how for in situ hard decontamination. The decontamination procedures for strong decontamination have been optimized in static and dynamic tests (DECO-loop). The best values have been found for: (i) hydrochloric acid: 4 to 5% vol. at low temperature, 0.7 to 1% vol. at high temperature (80 0 C); (ii) hydrofluoric plus nitric acid: 1.5% vol. HF + 5% vol. HNO 3 at low temperature; 0.3 to 0.5% vol. HF + 2.5 to 5% vol. HNO 3 at high temperature. High flow rates are not necessary, but a good re-circulation of the solution is needed. The final contamination levels, after total oxide removal, are in accordance with limits indicated for unrestricted release of materials in some countries. The arising of the secondary waste is estimated. Decontamination of a 10 m 2 surface would typically produce 0.5 to 3.0 kg of dry waste, corresponding to 1.6 to 10 kg of concrete conditioned waste

  1. FY 1993 progress report on the ANS thermal-hydraulic test loop operation and results

    Energy Technology Data Exchange (ETDEWEB)

    Siman-Tov, M.; Felde, D.K.; Farquharson, G. [and others

    1994-07-01

    The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). Highly subcooled heavy-water coolant flows vertically upward at a very high mass flux of almost 27 MG/m{sup 2}-s. In a parallel fuel plate configuration as in the ANSR, the flow is subject to a potential excursive static-flow instability that can very rapidly lead to flow starvation and departure from nucleate boiling (DNB) in the ``hot channel``. The current correlations and experimental data bases for flow excursion (FE) and critical heat flux (CHF) seldom evaluate the specific combination of ANSR operating parameters. The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. A series of FE tests with water flowing vertically upward was completed over a nominal heat flux range of 6 to 17 MW/m{sup 2}, a mass flux range of 8 to 28 Mg/m{sup 2}-s, an exit pressure range of 1.4 to 2.1 MPa, and an inlet temperature range of 40 to 50 C. FE experiments were also conducted using as ``soft`` a system as possible to secure a true FE phenomena (actual secondary burnout). True DNB experiments under similar conditions were also conducted. To the author`s knowledge, no other FE data have been reported in the literature to date that dover such a combination of conditions of high mass flux, high heat flux, and moderately high pressure.

  2. FY 1993 progress report on the ANS thermal-hydraulic test loop operation and results

    International Nuclear Information System (INIS)

    Siman-Tov, M.; Felde, D.K.; Farquharson, G.

    1994-07-01

    The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). Highly subcooled heavy-water coolant flows vertically upward at a very high mass flux of almost 27 MG/m 2 -s. In a parallel fuel plate configuration as in the ANSR, the flow is subject to a potential excursive static-flow instability that can very rapidly lead to flow starvation and departure from nucleate boiling (DNB) in the ''hot channel''. The current correlations and experimental data bases for flow excursion (FE) and critical heat flux (CHF) seldom evaluate the specific combination of ANSR operating parameters. The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. A series of FE tests with water flowing vertically upward was completed over a nominal heat flux range of 6 to 17 MW/m 2 , a mass flux range of 8 to 28 Mg/m 2 -s, an exit pressure range of 1.4 to 2.1 MPa, and an inlet temperature range of 40 to 50 C. FE experiments were also conducted using as ''soft'' a system as possible to secure a true FE phenomena (actual secondary burnout). True DNB experiments under similar conditions were also conducted. To the author's knowledge, no other FE data have been reported in the literature to date that dover such a combination of conditions of high mass flux, high heat flux, and moderately high pressure

  3. Optimal closed-loop identification test design for internal model control

    NARCIS (Netherlands)

    Zhu, Y.; Bosch, van den P.P.J.

    2000-01-01

    In this work, optimal closed-loop test design for control is studied. Simple design formulas are derived based on the asymptotic theory of Ljung. The control scheme used is internal model control (IMC) and the design constraint is the power of the process output or that of the reference signal. The

  4. FY 1995 progress report on the ANS thermal-hydraulic test loop operation and results

    Energy Technology Data Exchange (ETDEWEB)

    Siman-Tov, M.; Felde, D.K.; Farquharson, G.; McDuffee, J.L.; McFee, M.T.; Ruggles, A.E.; Wendel, M.W.; Yoder, G.L.

    1997-07-01

    The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. Special consideration was given to allow operation of the system in a stiff mode (constant flow) and in a soft mode (constant pressure drop) for proper implementation of true FE and DNB experiments. The facility is also designed to examine other T/H phenomena, including onset of incipient boiling (IB), single-phase heat transfer coefficients and friction factors, and two-phase heat transfer and pressure drop characteristics. Tests will also be conducted that are representative of decay heat levels at both high pressure and low pressure as well as other quasi-equilibrium conditions encountered during transient scenarios. A total of 22 FE tests and 2 CHF tests were performed during FY 1994 and FY 1995 with water flowing vertically upward. Comparison of these data as well as extensive data from other investigators led to a proposed modification to the Saha and Zuber correlation for onset of significant void (OSV), applied to FE prediction. The modification takes into account a demonstrated dependence of the OSV or FE thermal limits on subcooling levels, especially in the low subcooling regime.

  5. FY 1995 progress report on the ANS thermal-hydraulic test loop operation and results

    International Nuclear Information System (INIS)

    Siman-Tov, M.; Felde, D.K.; Farquharson, G.; McDuffee, J.L.; McFee, M.T.; Ruggles, A.E.; Wendel, M.W.; Yoder, G.L.

    1997-07-01

    The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. Special consideration was given to allow operation of the system in a stiff mode (constant flow) and in a soft mode (constant pressure drop) for proper implementation of true FE and DNB experiments. The facility is also designed to examine other T/H phenomena, including onset of incipient boiling (IB), single-phase heat transfer coefficients and friction factors, and two-phase heat transfer and pressure drop characteristics. Tests will also be conducted that are representative of decay heat levels at both high pressure and low pressure as well as other quasi-equilibrium conditions encountered during transient scenarios. A total of 22 FE tests and 2 CHF tests were performed during FY 1994 and FY 1995 with water flowing vertically upward. Comparison of these data as well as extensive data from other investigators led to a proposed modification to the Saha and Zuber correlation for onset of significant void (OSV), applied to FE prediction. The modification takes into account a demonstrated dependence of the OSV or FE thermal limits on subcooling levels, especially in the low subcooling regime

  6. 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.

  7. Testing FlexRay ECUs with a hardware-in-the-loop simulator; Test von FlexRay-Steuergeraeten am Hardware-in-the-Loop Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Stroop, J.; Koehl, S. [dSPACE GmbH, Paderborn (Germany); Peller, M.; Riedesser, P. [BMW AG, Muenchen (Germany)

    2005-07-01

    To master the data communication of complex and safety relevant systems within future vehicles, the BMW Group prepares the application of FlexRay. The accompanying development process plays an important role for the quality, stability and reliability of those systems. Hardware-in-the-loop simulation and test stands are indispensable constituents and they are an integral part of the validation process. The following contribution describes the technology that is used within the BMW Group in more detail, especially in terms of communication networks with FlexRay. (orig.)

  8. Revitalisation of Control and Data Acquisition Systems for Corrosion Test Loop

    International Nuclear Information System (INIS)

    Khairul Handono; Kiswanta; Edy Sumarno

    2008-01-01

    The replacement of control and data acquisition systems for Corrosion Test Loop (CTL) has been conducted. The aim of revitalisation for CTL is to increase controller system performance Kent 4000 which is based on PLC. On the other side revitalisation of acquisition data system is done to build computer based data retrieval system for transformation gauging of parameters in thermalhydraulic experiment of CTL. Previously, data collector system used indicator recorder analog, while data recording is done manually, which caused causing very slow response and the result is less accurate. To increase the user quality of data collector system, the data acquisition system is developed with application program Visual Basic and acquisition apparatus card of data. Result of the activity of revitalisation CTL is to obtain of control systems based on PLC and data acquisition system capable to present information in the form of temperature, pressure and cooling water level interactively, namely easy to read, quickly, realtime and accurate. This results give the improvement of control systems performance and data acquisition system which data storage of acquisition into hard disk in the form of file and further processed in the form of tables or graph to facilitate the analysis. (author)

  9. Test Requirements and Conceptual Design for a Potassium Test Loop to Support an Advanced Potassium Rankine Cycle Power Conversion Systems

    Energy Technology Data Exchange (ETDEWEB)

    Yoder, JR.G.L.

    2006-03-08

    Parameters for continuing the design and specification of an experimental potassium test loop are identified in this report. Design and construction of a potassium test loop is part of the Phase II effort of the project ''Technology Development Program for an Advanced Potassium Rankine Power Conversion System''. This program is supported by the National Aeronautics and Space Administration. Design features for the potassium test loop and its instrumentation system, specific test articles, and engineered barriers for ensuring worker safety and protection of the environment are described along with safety and environmental protection requirements to be used during the design process. Information presented in the first portion of this report formed the basis to initiate the design phase of the program; however, the report is a living document that can be changed as necessary during the design process, reflecting modifications as additional design details are developed. Some portions of the report have parameters identified as ''to be determined'' (TBD), reflecting the early stage of the overall process. In cases where specific design values are presently unknown, the report attempts to document the quantities that remain to be defined in order to complete the design of the potassium test loop and supporting equipment.

  10. Release of fission products from a fuel rod with an artificial hole through cladding irradiated in an in-pile water loop, (2)

    International Nuclear Information System (INIS)

    Ishiwatari, Nasumi

    1978-11-01

    To make clear the iodine spiking phenomenon from a defective fuel rod into the primary coolant, the fuel rod (UO 2 pellets, with stainless steel sheath) with an artificial pin hole was irradiated in the inpile test section of water loop JMTR.OWL-1. Experimental conditions were depressurization and temperature drop of the primary loop coolant and diameter and position of the pin hole. Iodine 131 and cesium 137 in loop coolant were measured under various coolant conditions. The inventory and translation rate of iodine 131 in fuel rod related to irradiation histories were calculated. The levels of I-131 and Cs-137 released to loop coolant from fuel rod were compared. Comparison of the results with LWRs was made by way of the spiked amount and release rate of iodine 131. (author)

  11. Measurement data of cesium 137 yields in primary coolant of an in-pile water loop in fission products release experiment

    International Nuclear Information System (INIS)

    Ishiwatari, Nasumi; Nagai, Hitoshi; Takeda, Tsuneo

    1979-03-01

    Series of fuel rods (UO 2 pellets sheathed with stainless steel) having an artificial pinhole were irradiated in the in-pile test section of water loop JMTR OWL-1. Presented are the results of measurements of cesium 137 yields in primary coolant of OWL-1 from 1975 to 1978. (author)

  12. Vanadium—lithium in-pile loop for comprehensive tests of vanadium alloys and multipurpose coatings

    Science.gov (United States)

    Lyublinski, I. E.; Evtikhin, V. A.; Ivanov, V. B.; Kazakov, V. A.; Korjavin, V. M.; Markovchev, V. K.; Melder, R. R.; Revyakin, Y. L.; Shpolyanskiy, V. N.

    1996-10-01

    The reliable information on design and material properties of self-cooled Li sbnd Li blanket and liquid metal divertor under neutron radiation conditions can be obtained using the concept of combined technological and material in-pile tests in a vanadium—lithium loop. The method of in-pile loop tests includes studies of vanadium—base alloys resistance, weld resistance under mechanical stress, multipurpose coating formation processes and coatings' resistance under the following conditions: high temperature (600-700°C), lithium velocities up to 10 m/s, lithium with controlled concentration of impurities and technological additions, a neutron load of 0.4-0.5 MW/m 2 and level of irradiation doses up to 5 dpa. The design of such an in-pile loop is considered. The experimental data on corrosion and compatibility with lithium, mechanical properties and welding technology of the vanadium alloys, methods of coatings formation and its radiation tests in lithium environment in the BOR-60 reactor (fast neutron fluence up to 10 26 m -2, irradiation temperature range of 500-523°C) are presented and analyzed as a basis for such loop development.

  13. Analysis of ATLAS LTC-04R Test for Loop Seal Reformation Phenomena using RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Sang-Gyu; Kim, Dae-Hun; Kim, Han-Gon [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The loop seal reformation issue was selected to be the analysis topic of the DSP-04 based on the technical discussion between the participants and the operating agencies (KAERI and KINS) and domestic experts meetings. After that, KAERI performed LTC-04R test which is 4 inch top-slot cold-leg break test using ATLAS facility in December 27, 2015. KHNP CRI, as a participant of the DSP-04, performed the blind calculation and open calculation using RELAP5/Mod3.3 patch 3. This paper deals with the results of open calculation for ATLAS LTC-04R test. The results of several sensitivity analyses such as the critical flow modeling sensitivity and break flow system modeling sensitivity will be discussed. Several possible factors in the loop seal reformation behavior are examined in the sensitivity analysis. Heat loss modeling, fine break system modeling, fine loop seal nodalization and off-take modeling are not significant factor in the loop seal reformation. Still critical flow model and discharge coefficient are dominant factors. Based on the ATLAS LTC-04R, Ransom-Trapp model shows better prediction in the break flow than the Henry-Fauske model.

  14. PETER loop. Multifunctional test facility for thermal hydraulic investigations of PWR fuel elements

    International Nuclear Information System (INIS)

    Ganzmann, I.; Hille, D.; Staude, U.

    2009-01-01

    The reliable fuel element behavior during the complete fuel cycle is one of the fundamental prerequisites of a safe and efficient nuclear power plant operation. The fuel element behavior with respect to pressure drop and vibration impact cannot be simulated by means of fluid-structure interaction codes. Therefore it is necessary to perform tests using fuel element mock-ups (1:1). AREVA NP has constructed the test facility PETER (PWR fuel element tests in Erlangen) loop. The modular construction allows maximum flexibility for any type of fuel elements. Modern measuring instrumentation for flow, pressure and vibration characterization allows the analysis of cause and consequences of thermal hydraulic phenomena. PETER loop is the standard test facility for the qualification of dynamic fuel element behavior in flowing fluid and is used for failure mode analysis.

  15. Scoping erosion flow loop test results in support of Hanford WTP

    International Nuclear Information System (INIS)

    Duignan, M.; Imrich, K.; Fowley, M.; Restivo, M.; Reigel, M.

    2015-01-01

    The Waste Treatment and Immobilization Plant (WTP) will process Hanford Site tank waste by converting the waste into a stable glass form. Before the tank waste can be vitrified, the baseline plan is to process the waste through the Pretreatment (PT) Facility where it will be mixed in various process vessels using Pulse Jet Mixers (PJM) and transferred to the High Level Waste (HLW) or Low Activity Waste (LAW) vitrification facilities. The Department of Energy (DOE) and Defense Nuclear Facility Safety Board (DNFSB), as well as independent review groups, have raised concerns regarding the design basis for piping erosion in the PT Facility. Due to the complex nature of slurry erosion/corrosion wear and the unique conditions that exist within the PT Facility, additional testing has been recommended by these entities. Pipe loop testing is necessary to analyze the potential for localized wear at elbows and bends, close the outstanding PT and HLW erosion/corrosion technical issues, and underpin BNI's design basis for a 40-year operational life for black cell piping and vessels. SRNL is consulting with the DOE Office of River Protection (ORP) to resolve technical concerns related to piping erosion/corrosion (wear) design basis for PT. SRNL was tasked by ORP to start designing, building, and testing a flow loop to obtain long-term total-wear rate data using bounding simulant chemistry, operating conditions, and prototypical materials. The initial test involved a scoping paint loop to locate experimentally the potential high-wear locations. This information will provide a basis for the placement of the many sensitive wear measurement instruments in the appropriate locations so that the principal flow-loop test has the best chance to estimate long-term erosion and corrosion. It is important to note that the scoping paint loop test only utilized a bounding erosion simulant for this test. A full chemical simulant needs to be added for the complete test flow loop. The

  16. Utilization of the RELAP4 in the experimental water loop of the IPEN-CNEN/SP

    International Nuclear Information System (INIS)

    Sabundjian, G.

    1991-09-01

    For a better security of the nuclear power station and the population, thermal hydraulics codes for accident analysis were developed. Within all codes, the RELAP4, developed by Aerojet Nuclear Company of the Idaho Falls Engineering Laboratory (U.S.A.), has been the more used mostly by the fact of been available to the public. These code presents the most satisfactory results when used for various sort of problems. Basely the paper analyzes and compares the RELAP4/MOD3 and MOD5 versions, through the simulation on the Water Experimental Loop (C.E.A.) of the IPEN, in the steady state and in accidents conditions. Although the C.E.A. had been constructed to simulate PWR and BWR reactors, in these paper the tests were made only for PWR. Through the results analysis found in these paper, we concluded that the model adaptation for the simulations made with the RELAP4 code (version MOD3 and MOD5), were satisfactory. The accidents simulations indicates that the C.E.A. can support accident situation with the emergency system acting. Comparing the two versions, the MOD5 presented the better results compared with the real estimations. (author)

  17. Hardware-in-the-loop (HIL) Test of Demand as Frequency Controlled Reserve (DFR)

    DEFF Research Database (Denmark)

    Wu, Qiuwei; Zimmermann, K.; Østergaard, Jacob

    2016-01-01

    This paper presents the hardware-in-the-loop (HIL) test of the demand as frequency controlled reserve (DFR). The HIL test refers to a test in which parts of a pure simulation have been replaced by actual physical components. It is used to understand the behavior of a new device or controller....... The DFR has been tested by offline simulations to illustrate the efficacy of this technology. The DFR control logics have been implemented in the SmartBox. The HIL was conducted by having the SmartBox connected to the real time simulations and the performance of the SmartBox was tested with difference...

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

    International Nuclear Information System (INIS)

    Ding Shengyao; Xu Kun; Yu Baosheng; Ling Yude

    2006-01-01

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

  19. Thermal characteristics and performance of Ag-water nanofluid: Application to natural circulation loops

    International Nuclear Information System (INIS)

    Koca, Halil Dogacan; Doganay, Serkan; Turgut, Alpaslan

    2017-01-01

    Highlights: • Thermal conductivity and viscosity of Ag-water nanofluid were measured. • Thermal performance of Ag-water nanofluid was compared with water. • Effectiveness enhanced up to 11% with 1 wt% Ag-water nanofluid. • Effectiveness of Ag-water nanofluid samples increased with inclination angle. • Ag-water nanofluid has potential to be used in flat-plate solar collectors. - Abstract: The goal of this study is to investigate the thermal conductivity, viscosity and thermal performance in a single-phase natural circulation mini loop of Ag-water nanofluid which can be a potential working fluid for natural convective flat-plate solar collectors. The silver-water nanofluid with 5 wt% concentration, which contains also polyvinylpyrrolidone (PVP) with 1.25 wt%, was purchased. Then, the sample was diluted with de-ionized water to four different concentrations of 0.25, 0.5, 0.75 and 1 wt%. Thermal conductivity and viscosity were measured by 3ω method and Brookfield rheometer, respectively. An effectiveness factor was used to define the thermal performance of Ag-water nanofluids for different inclination angles and heating powers. The results showed that nanofluid samples are thermally less conductive than the literature, at ambient temperature (23 °C). The viscosity of nanofluid decreases significantly with increasing temperature and increases with increasing concentration. Our measurements appear to be more compatible with PVP solution results available in the literature. Effectiveness is enhanced up to 11% with 1 wt% concentrated nanofluid compared to de-ionized water and the effectiveness of the mini loop indicates an enhancement with increase in inclination angle and particle concentration at whole applied power. According to obtained results, it is concluded that Ag-water nanofluid has a promising potential to be used in natural convective flat-plate solar collector.

  20. TLTA/6431, Two-Loop-Test-Apparatus, BWR/6 Simulator, Small-Break LOCA

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: The Two-Loop-Test-Apparatus (TLTA) is a 1:624 volume scaled BWR/6 simulator. It was the predecessor of the better-scaled FIST facility. The facility is capable of full BWR system pressure and has a simulated core with a full size 8 x 8, full power single bundle of indirect electrically heated rods. All major BWR systems are simulated including lower plenum, guide tube, core region (bundle and bypass), upper plenum, steam separator, steam dome, annular downcomer, recirculation loops and ECC injection systems. The fundamental scaling consideration was to achieve real-time response. A number of the scaling compromises present in TLTA were corrected in the FIST configuration. These compromises include a number of regional volumes and component elevations. 2 - Description of test: 64.45 sqcm small break LOCA with activation of the full emergency core cooling system, but without activation of the automatic decompression system

  1. Fusion fuel purification during the Tritium Systems Test Assembly 3-week loop experiment

    International Nuclear Information System (INIS)

    Willms, R.S.

    1989-01-01

    During the time period from April 19, 1989--May 5, 1989, the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory (LANL) conducted its longest continuous integrated loop operation to date. This provided an opportunity to test some hitherto unproven capabilities of the TSTA Fuel Cleanup System (FCU). Previous FCU tests were reported. The purpose of the FCU is to remove impurities from a stream of hydrogen isotopes (Q 2 ) representative of torus exhaust gas. During this run impurities loadings ranging from 60 to 179 sccm of 90% N 2 and 10% CH 4 were fed to the FCU. Each of the two FCU main flow molecular sieve beds (MSB's) were filled to breakthrough three times. The MSB's were regenerated during loop operations. 2 refs., 6 figs., 2 tabs

  2. Analyses of the Anticipated Operational Occurrences for the HANARO Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. K.; Sim, B. S.; Chi, D. Y.; Lee, C. Y.; Ahn, S. H

    2007-12-15

    The analyses of anticipated operational occurrences of the HANARO fuel test loop have been carried out by using the MARS/FTL{sub A} code, which is a modified version of the MARS code. A critical heat flux correlation on the three rods with triangular array was implemented in the MARS/FTL{sub A} code. The correlation was obtained from the critical heat fluxes measured at a test section, which is the same geometry of the in-pile test section of the HANARO fuel test loop. The anticipated operational occurrences of the HANARO fuel test loop are the inadvertent closure of the isolation valves, the over-power transient of the HANARO, the stuck open of the safety valves, and the loss of HANARO class IV power. A minimum DNBR (Departure from Nucleate Boiling Ratio) was predicted in the inadvertent closure of the isolation valves. It is indicated that the minimum DNBR of 1.85 is greater than the design limit DNBR of 1.39. The maximum coolant pressure calculated in the anticipated operational occurrences is also less than the 110 percents of the design pressure.

  3. Analyses of the Anticipated Operational Occurrences for the HANARO Fuel Test Loop

    International Nuclear Information System (INIS)

    Park, S. K.; Sim, B. S.; Chi, D. Y.; Lee, C. Y.; Ahn, S. H.

    2007-12-01

    The analyses of anticipated operational occurrences of the HANARO fuel test loop have been carried out by using the MARS/FTL A code, which is a modified version of the MARS code. A critical heat flux correlation on the three rods with triangular array was implemented in the MARS/FTL A code. The correlation was obtained from the critical heat fluxes measured at a test section, which is the same geometry of the in-pile test section of the HANARO fuel test loop. The anticipated operational occurrences of the HANARO fuel test loop are the inadvertent closure of the isolation valves, the over-power transient of the HANARO, the stuck open of the safety valves, and the loss of HANARO class IV power. A minimum DNBR (Departure from Nucleate Boiling Ratio) was predicted in the inadvertent closure of the isolation valves. It is indicated that the minimum DNBR of 1.85 is greater than the design limit DNBR of 1.39. The maximum coolant pressure calculated in the anticipated operational occurrences is also less than the 110 percents of the design pressure

  4. Resistance temperature sensor aging degradation identification using LCSR (Loop Current Step Response) test

    International Nuclear Information System (INIS)

    Santos, Roberto Carlos dos; Goncalves, Iraci Martine Pereira

    2013-01-01

    Most critical process temperatures in nuclear power plants are measured using RTD (Resistance Temperature Detector) and thermocouples. In a PWR (Pressure Water Reactor) plant, the primary coolant temperature and feedwater temperature are measured using RTDs, and the temperature of the water that exits the reactor core is measured using thermocouples. These thermocouples are mainly used for temperature monitoring purposes and are therefore not generally subject to very stringent requirements for accuracy and response-time performance. In contrast, primary coolant RTDs typically feed the plant's control and safety systems and must, therefore, be very accurate and have good dynamic performance. The response time of RTDs and thermocouples has been characterized by a single parameter called the Plunge Time Constant. This is defined as the time it takes the sensor output to achieve 63.2 percent of its final value after a step change in temperature is impressed on its surface. This step change is typically achieved by suddenly immersing the sensor in a rotating tank of water, called Plunge Test. In nuclear reactors, however, plunge testing is inconvenient because the sensor must be removed from the reactor coolant piping and taken to a laboratory for testing. Nuclear reactor service conditions of 150 bar and 300°C are difficult to reproduce in the laboratory. Therefore, all laboratory tests are performed at much milder conditions, and the results are extrapolated to service conditions. This leads to significant errors in the measurement of sensor response times and an insitu test method called LCSR - Loop Current Step Response test was developed in the mid-1970s to measure remotely the response time of RTDs. In the LCSR method, the sensing element is heated by an electric current; the current causes Joule heating in the sensor and results in a temperature transient inside the sensor. The temperature transient in the element is recorded, and from this transient, the

  5. IR1 flow tube and In-Pile Test Section Pressure drop test for the 3-Pin Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. H.; Park, K. N.; Chi, D. Y.; Sim, B. S.; Park, S. K.; Lee, J. M.; Lee, C. Y.; Kim, H. N

    2006-02-15

    The in-pile Section (IPS) of 3-pin Fuel Test Loop(FTL) shall be installed in the vertical hole call IR1 of HANARO reactor core. In order to verify the pressure drop and flow rate both the inside region of IPS at the annular region between IPS and IR1 flow tube, a pressure drop was measured by varing the flow rate on both regions. The measured pressure drop in the annular region is 209kpa at 14.9kg/s which meets the limiting condition of operation of 200kpa. The measured pressure drop in side the IPS becomes 260.25kpa which is lower than the designed value of 306.65kpa. As the pressure drop is lower than the design value, it is quite conservative from the safety and operating point of view.

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  7. Effect of closed loop cooling water transit time on containment cooling

    International Nuclear Information System (INIS)

    Smith, R.P.; Vossahlik, J.E.; Goodwin, E.F.

    1996-01-01

    Long term containment cooling analyses in nuclear plant systems are usually conducted assuming a quasi steady-state process, that is, a steady state evaluation of the cooling system is completed for each calculational step. In reality, fluid transport in the system, and heat addition to system components may affect the heat removal rate of the system. Transient effects occurring during system startup may affect the maximum temperatures experienced in the system. It is important to ensure that such transient effects do not affect operation of the system (e.g., cause a high temperature trip). To evaluate the effect of fluid transit delays, a closed loop cooling water system model has been developed that incorporates the fluid transport times when determining the closed loop cooling system performance. This paper describes the closed loop cooling system model as implemented in the CONTEMPT-LT/028 code. The evaluation of the transient temperature response of the closed loop cooling system using the model is described. The paper also describes the effect of fluid transit time on the overall containment cooling performance

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  9. Evaluation and comparison of closed-loop wash-water system

    International Nuclear Information System (INIS)

    Whitney, P.M.; Greer, C.R.

    1991-01-01

    Effluent from vehicle and equipment cleaning is known to contain a variety of potential pollutants, the most common being hydrocarbons and suspended solids. Proper treatment and discharge of this effluent is a growing concern as environmental awareness increases. In the United States, discharge of this effluent to municipal sewage treatment systems requires a permit from local authorities, discharge to surface waters requires a federal permit and, in most cases, discharge to the ground in prohibited. Furthermore, discharge to ground and surface waters can cause soil or groundwater contamination resulting in property devaluation, adverse impact on human health, fines from regulatory agencies, expensive cleanup and negative publicity. Effluent from vehicle washing typically does not meet the minimum pollutant levels allowed by regulatory agencies for discharge to surface waters or sewage treatment plants. Because of the liability associated with discharge to ground and surface waters and the difficulty in meeting municipal sewer discharge permit requirements, closed-loop wastewater treatment is an attractive alternative to discharge. Evaluation and comparison of systems from each category constitute the basis of this paper. Factors involved in selecting a system and available water-treatment technologies are discussed. The conclusion summarizes the results of the system comparison and makes recommendations for selecting and installing closed-loop water treatment systems for vehicle and equipment cleaning

  10. Aircraft Thermal Management Using Loop Heat Pipes: Experimental Simulation of High Acceleration Environments Using the Centrifuge Table Test Bed (Postprint)

    National Research Council Canada - National Science Library

    Fleming, Andrew J; Leland, Quinn H; Yerkes, Kirk L; Elston, Levi J; Thomas, Scott K

    2006-01-01

    The objective of this paper is to describe the design of an experiment that will examine the effects of elevated acceleration environments on a high-temperature, titanium-water loop heat pipe for actuator cooling...

  11. Thermal-hydraulic calculation and water hammer analysis on CEFR loop system

    International Nuclear Information System (INIS)

    Hao Pengfei; Zhang Xiwen; Cai Weidong; Wang Xuefang

    1997-01-01

    China Experimental Fast Reactor (CEFR) is one of the '863' High-technical Projects. It is necessary to study the hydraulic and thermal Characteristic of CEFR loop system in order to guarantee the safety of operation. The results of the thermal-hydraulic calculation have been given. The main points are as follows: 1. The simplified model is built according to the loop system of CEFR, and the calculation method which is called 'NODE'-'BRANCH' is applied. This method includes two aspects, one is the theoretical analysis that is based on fluid mechanics and heat transfer theory. The other is the engineering calculation. These two aspects are connected in the computation. On the basis of the work mentioned above, the stable state computation is presented. In order to prevent serious damage caused by power failure accident, the courses of surplus reactor heat removing through two different systems have been simulated in the computation. 2. By using the fluid dynamics theory, the simplified model and the equipment boundary conditions of loop system are given. The water hammer computation is processed during the valve closing and pump stopping accidents. Some pictures of water hammer wave are presented, and the most dangerous state in the accident is also given

  12. Efficiency of the pre-heater against flow rate on primary the beta test loop

    International Nuclear Information System (INIS)

    Edy Sumarno; Kiswanta; Bambang Heru; Ainur R; Joko P

    2013-01-01

    Calculation of efficiency of the pre-heater has been carried out against the flow rate on primary the BETA Test Loop. BETA test loop (UUB) is a facilities of experiments to study the thermal hydraulic phenomenon, especially for thermal hydraulic post-LOCA (Lost of Coolant Accident). Sequences removal on the BETA Test Loop contained a pre-heater that serves as a getter heat from the primary side to the secondary side, determination of efficiency is to compare the incoming heat energy with the energy taken out by a secondary fluid. Characterization is intended to determine the performance of a pre-heater, then used as tool for analysis, and as a reference design experiments. Calculation of efficiency methods performed by operating the pre-heater with fluid flow rate variation on the primary side. Calculation of efficiency on the results obtained that the efficiency change with every change of flow rate, the flow rate is 71.26% on 163.50 ml/s and 60.65% on 850.90 ml/s. Efficiency value can be even greater if the pre-heater tank is wrapped with thermal insulation so there is no heat leakage. (author)

  13. Cutting a drop of water pinned by wire loops using a superhydrophobic surface and knife.

    Directory of Open Access Journals (Sweden)

    Ryan Yanashima

    Full Text Available A water drop on a superhydrophobic surface that is pinned by wire loops can be reproducibly cut without formation of satellite droplets. Drops placed on low-density polyethylene surfaces and Teflon-coated glass slides were cut with superhydrophobic knives of low-density polyethylene and treated copper or zinc sheets, respectively. Distortion of drop shape by the superhydrophobic knife enables a clean break. The driving force for droplet formation arises from the lower surface free energy for two separate drops, and it is modeled as a 2-D system. An estimate of the free energy change serves to guide when droplets will form based on the variation of drop volume, loop spacing and knife depth. Combining the cutting process with an electrofocusing driving force could enable a reproducible biomolecular separation without troubling satellite drop formation.

  14. Detection Test for Leakage of CO{sub 2} into Sodium Loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sun Hee; Wi, Myung-Hwan; Min, Jae Hong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This report is about the facility for the detection test for leakage of CO{sub 2} into sodium loop. The facility for the detection test for leakage of CO{sub 2} into sodium loop was introduced. The test will be carried out. Our experimental results are going to be expected to be used for approach methods to detect CO{sub 2} leaking into sodium in heat exchangers. A sodium-and-carbon dioxide (Na-CO{sub 2}) heat exchanger is one of the key components for the supercritical CO{sub 2} Brayton cycle power conversion system of sodium-cooled fast reactors (SFRs). A printed circuit heat exchanger (PCHE) is considered for the Na-CO{sub 2} heat exchanger, which is known to have potential for reducing the volume occupied by the exchangers compared to traditional shell-and-tube heat exchangers. Among various issues about the Na- CO{sub 2} exchanger, detection of CO{sub 2} leaking into sodium in the heat exchanger is most important thing for its safe operation. It is known that reaction products from sodium and CO{sub 2} such as sodium carbonate (Na{sub 2}CO{sub 3}) and amorphous carbon are hardly soluble in sodium, which cause plug sodium channels. Detection technique for Na{sub 2}CO{sub 3} in sodium loop has not been developed yet. Therefore, detection of CO{sub 2} and CO from reaction of sodium and CO{sub 2} are proper to detect CO{sub 2} leakage into sodium loop.

  15. Steady state and linear stability analysis of a supercritical water natural circulation loop

    International Nuclear Information System (INIS)

    Sharma, Manish; Pilkhwal, D.S.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2010-01-01

    Supercritical water (SCW) has excellent heat transfer characteristics as a coolant for nuclear reactors. Besides it results in high thermal efficiency of the plant. However, the flow can experience instabilities in supercritical water reactors, as the density change is very large for the supercritical fluids. A computer code SUCLIN using supercritical water properties has been developed to carry out the steady state and linear stability analysis of a SCW natural circulation loop. The conservation equations of mass, momentum and energy have been linearized by imposing small perturbation in flow rate, enthalpy, pressure and specific volume. The equations have been solved analytically to generate the characteristic equation. The roots of the equation determine the stability of the system. The code has been qualitatively assessed with published results and has been extensively used for studying the effect of diameter, height, heater inlet temperature, pressure and local loss coefficients on steady state and stability behavior of a Supercritical Water Natural Circulation Loop (SCWNCL). The present paper describes the linear stability analysis model and the results obtained in detail.

  16. Simulation of Two-Phase Natural Circulation Loop for Core Cather Cooling Using Air Water

    International Nuclear Information System (INIS)

    Revankar, S. T.; Huang, S. F.; Song, K. W.; Rhee, B. W.; Park, R. J.; Song, J. H.

    2012-01-01

    A closed loop natural circulation system employs thermally induced density gradients in single phase or two-phase liquid form to induce circulation of the working fluid thereby obviating the need for any mechanical moving parts such as pumps and pump controls. This increases the reliability and safety of the cooling system and reduces installation, operation and maintenance costs. That is the reason natural circulation cooling has been considered in advanced reactor core cooling and in engineered safety systems. Natural circulation cooling has been proposed to remove reactor decay heat by external vessel cooling for in-vessel core retention during sever accident scenario. Recently in APR1400 reactor core catcher design natural circulation cooling is proposed to stabilize and cool the corium ejected from the reactor vessel following core melt and breach of reactor vessel. The natural circulation flow is similar to external vessel cooling where water flows through an inclined narrow gap below hot surface and is heated to produce boiling. The two-phase natural circulation enables cooling of the corium pool collected on core catcher. Due to importance of this problem this paper focuses simulation of the two-phase natural circulation through inclined gap using air-water system. Scaling criteria for air-water loop are derived that enable simulation of the flow regimes and natural circulation flow rates in such systems using air-water system

  17. Alternative Water Processor Test Development

    Science.gov (United States)

    Pickering, Karen D.; Mitchell, Julie; Vega, Leticia; Adam, Niklas; Flynn, Michael; Wjee (er. Rau); Lunn, Griffin; Jackson, Andrew

    2012-01-01

    The Next Generation Life Support Project is developing an Alternative Water Processor (AWP) as a candidate water recovery system for long duration exploration missions. The AWP consists of biological water processor (BWP) integrated with a forward osmosis secondary treatment system (FOST). The basis of the BWP is a membrane aerated biological reactor (MABR), developed in concert with Texas Tech University. Bacteria located within the MABR metabolize organic material in wastewater, converting approximately 90% of the total organic carbon to carbon dioxide. In addition, bacteria convert a portion of the ammonia-nitrogen present in the wastewater to nitrogen gas, through a combination of nitrogen and denitrification. The effluent from the BWP system is low in organic contaminants, but high in total dissolved solids. The FOST system, integrated downstream of the BWP, removes dissolved solids through a combination of concentration-driven forward osmosis and pressure driven reverse osmosis. The integrated system is expected to produce water with a total organic carbon less than 50 mg/l and dissolved solids that meet potable water requirements for spaceflight. This paper describes the test definition, the design of the BWP and FOST subsystems, and plans for integrated testing.

  18. Flica: a code for the thermodynamic study of a reactor or a test loop

    International Nuclear Information System (INIS)

    Fajeau, M.

    1969-01-01

    This code handles the thermal problems of water loops or reactor cores under the following conditions: High or low pressure, steady state or transient behavior, one or two phases - Three-dimensional thermodynamic study of the flow in cylindrical geometry - Unidimensional study of heat transfer in heating elements - Neutronic studies can be coupled and a schematic representation of the safety rod behavior is given. The number of cells described in a flow cross-section is presently less than 20. This code is the logical following of FLID and CACTUS of which it constitutes a synthesis. (author) [fr

  19. Engineering design of IFMIF/EVEDA lithium test loop. Electro-magnetic pump and pressure drop

    International Nuclear Information System (INIS)

    Kondo, Hiroo; Furukawa, Tomohiro; Hirakawa, Yasushi; Iuchi, Hiroshi; Kanemura, Takuji; Ida, Mizuho; Watanabe, Kazuyoshi; Wakai, Eiichi; Nakamura, Kazuyuki; Horiike, H.; Yamaoka, N.; Matsushita, I.

    2011-01-01

    The Engineering Validation and Engineering Design Activities (EVEDA) for the International Fusion Materials Irradiation Facility (IFMIF) is proceeding as one of the ITER Broader Approach (ITER-BA). A Li circulation loop for testing hydraulic stability of the Li target (high speed free-surface flow of liquid Li as a beam target) and Li purification traps are under construction in the Japan Atomic Energy Agency as a major Japanese activities in the EVEDA. This paper presents specification of an electro-magnetic pump (EMP) for the EVEDA Li Test Loop (ELTL) and evaluation of the pressure drop in the main loop of the ELTL. The EMP circulates the liquid Li at a large flow rate up to 0.05 m 3 /s (3000 l/min) under a vacuum cover gas (Ar) pressure of 10 -3 Pa, thus the evaluation of cavitation generation is a crucial issue. The EMP used in the ELTL consists of two EMPs aligned in series through a U-tube whose size of one EMP is 0.8 m square and 2.6 m in length. The calculation of the pressure drop in the main Li loop to the EMP is approx. 25 kPa at the design maximum flow rate of 0.05 m 3 /s. On the other hand the height from the EMP to a Li tank to supply Li to the EMP is designed to be 9.72 m, and secures a static pressure and the cavitation number of 18 kPa and 3.4 respectively at the maximum flow rate in a vacuum condition. As a result, it is confirmed to prevent cavitation at the inlet of the EMP in this design. (author)

  20. EPRI flow-loop/in situ test program for motor-operated valves

    International Nuclear Information System (INIS)

    Hosler, J.F.; Dorfman, L.S.

    1994-01-01

    The Electric Power Research Institute is undertaking a comprehensive research program to develop and validate methods for predicting the performance of common motor-operated gate, global, and butterfly valves. To assess motor-operated valve (MOV) performance characteristics and provide a basis for methods validation, full-scale testing was conducted on 62 MOVs. Tests were performed in four flow-loop facilities and in nine nuclear units. Forty-seven gate, five globe, and 10 butterfly valves were tested under a wide range of flow and differential pressure conditions. The paper describes the test program scope, test configurations, instrumentation and data acquisition, testing approach, and data analysis methods. Key results are summarized

  1. The Application of Hardware in the Loop Testing for Distributed Engine Control

    Science.gov (United States)

    Thomas, George L.; Culley, Dennis E.; Brand, Alex

    2016-01-01

    The essence of a distributed control system is the modular partitioning of control function across a hardware implementation. This type of control architecture requires embedding electronics in a multitude of control element nodes for the execution of those functions, and their integration as a unified system. As the field of distributed aeropropulsion control moves toward reality, questions about building and validating these systems remain. This paper focuses on the development of hardware-in-the-loop (HIL) test techniques for distributed aero engine control, and the application of HIL testing as it pertains to potential advanced engine control applications that may now be possible due to the intelligent capability embedded in the nodes.

  2. Analysis of the SBLOCAs in HANARO pool for the 3-pin fuel test loop

    International Nuclear Information System (INIS)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, Y. J.

    2004-09-01

    Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Small Break Loss Of Coolant Accidents (SBLOCAs) in HANARO pool for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the SBLOCAs. The location of the pipe break is assumed at the hill taps connecting the cold and hot legs in HANARO pool to the inlet and outlet nozzles of the In-Pile test Section (IPS). The break size is also assumed less than 20% of the cross section area of the pipe. The test fuels are heated up when the cold leg break occur. However, they are not heated up when the hot leg break occur. The maximum Peak Cladding Temperatures (PCT) are predicted to be about 906.9 .deg. C for the cold leg break accident in PWR fuel test mode and 971.9 .deg. C in CANDU fuel test mode respectively. The critical break size is about the 6% of the cross section area of the pipe for PWR fuel test mode and the 8% for CANDU fuel test mode. The PCTs meet the design criterion of commercial PWR fuel that the maximum PCT is lower than 1204 .deg. C

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

    International Nuclear Information System (INIS)

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

    2014-05-01

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  5. Test system design for Hardware-in-Loop evaluation of PEM fuel cells and auxiliaries

    Energy Technology Data Exchange (ETDEWEB)

    Randolf, Guenter; Moore, Robert M. [Hawaii Natural Energy Institute, University of Hawaii, Honolulu, HI (United States)

    2006-07-14

    In order to evaluate the dynamic behavior of proton exchange membrane (PEM) fuel cells and their auxiliaries, the dynamic capability of the test system must exceed the dynamics of the fastest component within the fuel cell or auxiliary component under test. This criterion is even more critical when a simulated component of the fuel cell system (e.g., the fuel cell stack) is replaced by hardware and Hardware-in-Loop (HiL) methodology is employed. This paper describes the design of a very fast dynamic test system for fuel cell transient research and HiL evaluation. The integration of the real time target (which runs the simulation), the test stand PC (that controls the operation of the test stand), and the programmable logic controller (PLC), for safety and low-level control tasks, into one single integrated unit is successfully completed. (author)

  6. Analysis of the October 5, 1979 lithium spill and fire in the Lithium Processing Test Loop

    International Nuclear Information System (INIS)

    Maroni, V.A.; Beatty, R.A.; Brown, H.L.; Coleman, L.F.; Foose, R.M.; McPheeters, C.C.; Slawecki, M.; Smith, D.L.; Van Deventer, E.H.; Weston, J.R.

    1981-12-01

    On October 5, 1979, the Lithium Processing Test Loop (LPTL) developed a lithium leak in the electromagnetic (EM) pump channel, which damaged the pump, its surrounding support structure, and the underlying floor pan. A thorough analysis of the causes and consequences of the pump failure was conducted by personnel from CEN and several other ANL divisions. Metallurgical analyses of the elliptical pump channel and adjacent piping revealed that there was a significant buildup of iron-rich crystallites and other solid material in the region of the current-carrying bus bars (region of high magnetic field), which may have resulted in a flow restriction that contributed to the deterioration of the channel walls. The location of the failure was in a region of high residual stress (due to cold work produced during channel fabrication); this failure is typical of other cold work/stress-related failures encountered in components operated in forced-circulation lithium loops. Another important result was the isolation of crystals of a compound characterized as Li/sub x/CrN/sub y/. Compounds of this type are believed to be responsible for much of the Fe, Cr, and Ni mass transfer encountered in lithium loops constructed of stainless steel. The importance of nitrogen in the mass-transfer mechanism has long been suspected, but the existence of stable ternary Li-M-N compounds (M = Fe, Cr, Ni) had not previously been verified

  7. Customer interface document for the Molten Salt Test Loop (MSTL) system.

    Energy Technology Data Exchange (ETDEWEB)

    Pettit, Kathleen; Kolb, William J.; Gill, David Dennis; Briggs, Ronald D.

    2012-03-01

    The National Solar Thermal Test Facility at Sandia National Laboratories has a unique test capability called the Molten Salt Test Loop (MSTL) system. MSTL is a test capability that allows customers and researchers to test components in flowing, molten nitrate salt. The components tested can range from materials samples, to individual components such as flex hoses, ball joints, and valves, up to full solar collecting systems such as central receiver panels, parabolic troughs, or linear Fresnel systems. MSTL provides realistic conditions similar to a portion of a concentrating solar power facility. The facility currently uses 60/40 nitrate 'solar salt' and can circulate the salt at pressure up to 600psi, temperature to 585 C, and flow rate of 400-600GPM depending on temperature. The purpose of this document is to provide a basis for customers to evaluate the applicability to their testing needs, and to provide an outline of expectations for conducting testing on MSTL. The document can serve as the basis for testing agreements including Work for Others (WFO) and Cooperative Research and Development Agreements (CRADA). While this document provides the basis for these agreements and describes some of the requirements for testing using MSTL and on the site at Sandia, the document is not sufficient by itself as a test agreement. The document, however, does provide customers with a uniform set of information to begin the test planning process.

  8. Hardware in the loop simulation test platform of fuel cell backup system

    Directory of Open Access Journals (Sweden)

    Ma Tiancai

    2015-01-01

    Full Text Available Based on an analysis of voltage mechanistic model, a real-time simulation model of the proton exchange membrane (PEM fuel cell backup system is developed, and verified by the measurable experiment data. The method of online parameters identification for the model is also improved. Based on the software LabVIEW/VeriStand real-time environment and the PXI Express hardware system, the PEM fuel cell system controller hardware in the loop (HIL simulation plat-form is established. Controller simulation test results showed the accuracy of HIL simulation platform.

  9. Accuracy of small diameter sheathed thermocouples for the core flow test loop

    International Nuclear Information System (INIS)

    Anderson, R.L.; Kollie, T.G.

    1979-04-01

    This report summarizes the research and development on 0.5-mm-diameter, compacted, metal sheathed thermocouples. The objectives of this research effort have been: to identify and analyze the sources of temperature measurement errors in the use of 0.5-mm-diameter sheathed thermocouples to measure the surface temperature of the cladding of fuel-rod simulators in the Core Flow Test Loop (CFTL) at ORNL; to devise methods for reducing or correcting for these temperature measurement errors; to estimate the overall temperature measurement uncertainties; and to recommend modifications in the manufacture, installation, or materials used to minimize temperature measurement uncertainties in the CFTL experiments

  10. Characterization of a solar photovoltaic/loop-heat-pipe heat pump water heating system

    International Nuclear Information System (INIS)

    Zhang, Xingxing; Zhao, Xudong; Xu, Jihuan; Yu, Xiaotong

    2013-01-01

    Highlights: ► Describing concept and operating principle of the PV/LHP heat pump water heating system. ► Developing a numerical model to evaluate the performance of the system. ► Experimental testing of the prototype system. ► Characterizing the system performance using parallel comparison between the modelling and experimental results. ► Investigating the impact of the operating conditions to the system’s performance. -- Abstract: This paper introduced the concept, potential application and benefits relating to a novel solar photovoltaic/loop-heat-pipe (PV/LHP) heat pump system for hot water generation. On this basis, the paper reported the process and results of characterizing the performance of such a system, which was undertaken through dedicated thermo-fluid and energy balance analyses, computer model development and operation, and experimental verification and modification. The fundamental heat transfer, fluid flow and photovoltaic governing equations were applied to characterize the energy conversion and transfer processes occurring in each part and whole system layout; while the energy balance approach was utilized to enable inter-connection and resolution of the grouped equations. As a result, a dedicated computer model was developed and used to calculate the operational parameters, optimise the geometrical configurations and sizes, and recommend the appropriate operational condition relating to the system. Further, an experimental rig was constructed and utilized to acquire the relevant measurement data that thus enabled the parallel comparison between the simulation and experiment. It is concluded that the testing and modelling results are in good agreement, indicating that the model has the reasonable accuracy in predicting the system’s performance. Under the given experimental conditions, the electrical, thermal and overall efficiency of the PV/LHP module were around 10%, 40% and 50% respectively; whilst the system’s overall performance

  11. Hardware-in-the-Loop environment for testing and commissioning of space controllers; Hardware-in-the-Loop Umgebung zum Test und zur Inbetriebnahme von Raumreglern

    Energy Technology Data Exchange (ETDEWEB)

    Adlhoch, Alexander; Becker, Martin [Hochschule Biberach (Germany). Inst. fuer Gebaeude- und Energiesysteme

    2012-07-01

    The energy-efficient and optimal functioning of room controllers in terms of indoor air climates is influenced mainly by the control algorithm and the optimal adjustment of the parameters of controllers used in terms of space requirements. In the practical operation, deficits in the function or parameters of the controller are hardly or only with great effort metrological detectable, but have a significant impact on the energy consumption and / or the indoor climate comfort. In a hardware-in-the-loop (HIL) environment, room controllers can be examined in terms of the function under defined conditions, and different controllers can be evaluated comparatively. It is also possible to adjust the parameters of the controller before the commissioning. The HiL environment presented in the contribution under consideration consists of a model of the controlled system, a hardware coupler and a real controller to be tested. Among the spatial models, it can be selected from a plurality of different types of space which in turn can be assigned by means of different spatial parameters and environmental models. These combinations enable a replication of a test scenario corresponding to the later application. The hardware coupler provides a selection of physical inputs and outputs as well as interfaces to different bus systems (for example KNX, LON, EnOcean) for connecting different types of controllers. The construction and operation of a HIL test stand for space controller is presented based on first practical control tests. At this, the focus is on the suitability of this test environment for a variety of different controllers as well as development assistance and assistance for the adjustment of parameters. The HiL environments developed in the joint research project HiL RHK1 for the testing of space controllers, controllers for HVAC systems and refrigeration technology controllers have been developed so that the HiL environments can be coupled to a multi-HIL environment. This

  12. Analysis of eventual accidents in a water experimental loop, using the Relap 4 computer code

    International Nuclear Information System (INIS)

    Fernandes Filho, T.L.

    1981-01-01

    Transients caused by accidents as (1) loss of coolant, (2) failure in the principal pump and (3) power excursions were analysed. In the accident simulation, the Relap 4/Mod 3 computer code was used. The results obtained with the steady state model showed to be consistent with the project-and operation data of the experimental loop. For all the accidents analysed that considered the performance of safety systems, the highest temperature of the heating rods in the testing section did not exceed the permissible temperature. (E.G.) [pt

  13. An equivalent ground thermal test method for single-phase fluid loop space radiator

    Directory of Open Access Journals (Sweden)

    Xianwen Ning

    2015-02-01

    Full Text Available Thermal vacuum test is widely used for the ground validation of spacecraft thermal control system. However, the conduction and convection can be simulated in normal ground pressure environment completely. By the employment of pumped fluid loops’ thermal control technology on spacecraft, conduction and convection become the main heat transfer behavior between radiator and inside cabin. As long as the heat transfer behavior between radiator and outer space can be equivalently simulated in normal pressure, the thermal vacuum test can be substituted by the normal ground pressure thermal test. In this paper, an equivalent normal pressure thermal test method for the spacecraft single-phase fluid loop radiator is proposed. The heat radiation between radiator and outer space has been equivalently simulated by combination of a group of refrigerators and thermal electrical cooler (TEC array. By adjusting the heat rejection of each device, the relationship between heat flux and surface temperature of the radiator can be maintained. To verify this method, a validating system has been built up and the experiments have been carried out. The results indicate that the proposed equivalent ground thermal test method can simulate the heat rejection performance of radiator correctly and the temperature error between in-orbit theory value and experiment result of the radiator is less than 0.5 °C, except for the equipment startup period. This provides a potential method for the thermal test of space systems especially for extra-large spacecraft which employs single-phase fluid loop radiator as thermal control approach.

  14. Natural Circulation High Pressure Loop Dynamics Around Operating Point, Tests and Modelling With Retran 02

    International Nuclear Information System (INIS)

    Masriera, N.A; Doval, A.S; Mazufri, C.M

    2000-01-01

    The Natural Circulation High Pressure Loop (CAPCN) reproduces in scale all the one-dimensional thermal-hydraulic phenomena occurring in the primary loop of CAREM-25 reactor.It plays an important role in the qualification process of calculating computer codes.This facility demanded to develop several technological solutions in order to achieve the measuring and control quality required by that process.This engineering and experimental development allowed completing the first stage of dynamic tests during 1998.The trends of recorded data were systematically evaluated in terms of the deviations of main variables in response to different perturbations.By this analysis a group of eight transients was selected, providing a Minimum Representative Set (MRS) of dynamic tests, allowing the evaluation of all dynamic phenomena.Each of these transients was simulated with RETRAN-02, using a spreadsheet to facilitate the consistent elaboration and modification of input files.Comparing measured data and computer simulations, it may be concluded that it is possible to reproduce the dynamic response of all the transients with a level of approximation quite homogeneous and generally acceptable.It is possible to identify the detailed physical models that fit better the dynamic phenomena, and which of the limitations of RETRAN code are more relevant

  15. Interoperability Testing Using the Hardware-in-the-Loop Test Tool

    National Research Council Canada - National Science Library

    Buxton, Bradley

    1999-01-01

    In the past, missile defense testing was limited to models and simulations, which provided maximum flexibility but often lacked realism, or to live fire testing, which provided realism but proved very expensive...

  16. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Baek, Won Pil; Song, C. H.; Kim, Y. S.

    2007-02-01

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform the tests for design, operation, and safety regulation of pressurized water reactors. In the first phase of this project (1997.8∼2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished. In the second phase (2002.4∼2005.2), an optimized design of the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) was established and the construction of the facility was almost completed. In the third phase (2005.3∼2007.2), the construction and commission tests of the ATLAS are to be completed and some first-phase tests are to be conducted

  17. Long Duration Testing of a Spacesuit Water Membrane Evaporator Prototype

    Science.gov (United States)

    Bue, Grant C.; Makinen, Janice; Cox, Marlon; Watts, Carly; Campbell, Colin; Vogel, Matthew; Colunga, Aaron; Conger, Bruce

    2012-01-01

    The Spacesuit Water Membrane Evaporator (SWME) is a heat-rejection device that is being developed to perform thermal control for advanced spacesuits. Cooling is achieved by circulating water from the liquid cooling garment (LCG) through hollow fibers (HoFi s), which are small hydrophobic tubes. Liquid water remains within the hydrophobic tubes, but water vapor is exhausted to space, thereby removing heat. A SWME test article was tested over the course of a year, for a total of 600 cumulative hours. In order to evaluate SWME tolerance to contamination due to constituents caused by distillation processes, these constituents were allowed to accumulate in the water as evaporation occurred. A test article was tested over the course of a year for a total of 600 cumulative hours. The heat rejection performance of the SWME degraded significantly--below 700 W, attributable to the accumulation of rust in the circulating loop and biofilm growth. Bubble elimination capability, a feature that was previously proven with SWME, was compromised during the test, most likely due to loss of hydrophobic properties of the hollow fibers. The utilization of water for heat rejection was shown not to be dependent on test article, life cycle, heat rejection rate, or freezing of the membranes.

  18. Modeling and Closed Loop Flight Testing of a Fixed Wing Micro Air Vehicle

    Directory of Open Access Journals (Sweden)

    Harikumar Kandath

    2018-03-01

    Full Text Available This paper presents the nonlinear six degrees of freedom dynamic modeling of a fixed wing micro air vehicle. The static derivatives of the micro air vehicle are obtained through the wind tunnel testing. The propeller effects on the lift, drag, pitching moment and side force are quantified through wind tunnel testing. The dynamic derivatives are obtained through empirical relations available in the literature. The trim conditions are computed for a straight and constant altitude flight condition. The linearized longitudinal and lateral state space models are obtained about trim conditions. The variations in short period mode, phugoid mode, Dutch roll mode, roll subsidence mode and spiral mode with respect to different trim operating conditions is presented. A stabilizing static output feedback controller is designed using the obtained model. Successful closed loop flight trials are conducted with the static output feedback controller.

  19. Construction and performance tests of Helium Engineering Demonstration Loop (HENDEL) for VHTR

    International Nuclear Information System (INIS)

    Hishida, M.; Tanaka, T.; Shimomura, H.; Sanokawa, K.

    1984-01-01

    A helium engineering demonstration loop (HENDEL) was constructed and operated in JAERI in order to develop the high-temperature key components of an experimental very high temperature gas cooled reactor, like fuel stack, in-core reactor structure, hot gas duct, intermediate heat exchanger. Performance tests as well as demonstration of integrity are carried out with large-size or actual-size models of key components. The key components to be tested in HENDEL are: fuel stack and control rod; core supporting structure, or bottom structure of rector core exposed to direct impingement of high temperature core outlet flow; reactor internal components and structure; high temperature components in heat removal system (primary and secondary cooling systems)

  20. Numerical comparison of thermal hydraulic aspects of supercritical carbon dioxide and subcritical water-based natural circulation loop

    Energy Technology Data Exchange (ETDEWEB)

    Sarkar, Milan Krishna Singhar; Basu, Dipankar Narayan [Dept. of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati (India)

    2017-02-15

    Application of the supercritical condition in reactor core cooling needs to be properly justified based on the extreme level of parameters involved. Therefore, a numerical study is presented to compare the thermalhydraulic performance of supercritical and single-phase natural circulation loops under low-to-intermediate power levels. Carbon dioxide and water are selected as respective working fluids, operating under an identical set of conditions. Accordingly, a three-dimensional computational model was developed, and solved with an appropriate turbulence model and equations of state. Large asymmetry in velocity and temperature profiles was observed in a single cross section due to local buoyancy effect, which is more prominent for supercritical fluids. Mass flow rate in a supercritical loop increases with power until a maximum is reached, which subsequently corresponds to a rapid deterioration in heat transfer coefficient. That can be identified as the limit of operation for such loops to avoid a high temperature, and therefore, the use of a supercritical loop is suggested only until the appearance of such maxima. Flow-induced heat transfer deterioration can be delayed by increasing system pressure or lowering sink temperature. Bulk temperature level throughout the loop with water as working fluid is higher than supercritical carbon dioxide. This is until the heat transfer deterioration, and hence the use of a single-phase loop is prescribed beyond that limit.

  1. Fabrication of the instrumented fuel rods for the 3-Pin Fuel Test Loop at HANARO

    International Nuclear Information System (INIS)

    Sohn, Jae Min; Park, Sung Jae; Shin, Yoon Tag; Lee, Jong Min; Ahn, Sung Ho; Kim, Soo Sung; Kim, Bong Goo; Kim, Young Ki; Lee, Ki Hong; Kim, Kwan Hyun

    2008-09-01

    The 3-Pin Fuel Test Loop(hereinafter referred to as the '3-Pin FTL') facility has been installed at HANARO(High-flux Advanced Neutron Application Reactor) and the 3-Pin FTL is under a test operation. The purpose of this report is to fabricate the instrumented fuel rods for the 3-Pin FTL. The fabrication of these fuel rods was based on experiences and technologies of the instrumented fuel rods for an irradiation fuel capsule. The three instrumented fuel rods of the 3-Pin FTL have been designed. The one fuel rod(180 .deg. ) was designed to measure the centerline temperature of the nuclear fuels and the internal pressure of the fuel rod, and others(60 .deg. and 300 .deg. ) were designed to measure the centerline temperature of the fuel pellets. The claddings were made of the reference material 1 and 2 and new material 1 and 2. And nuclear fuel was used UO 2 (2.0w/o) pellet type with large grain and standard grain. The major procedures of fabrication are followings: (1) the assembling and weld of fuel rods with the pellet mockups and the sensor mockups for the qualification tests, (2) the qualification tests(dimension measurements, tensile tests, metallography examinations and helium leak tests) of weld, (3) the assembling and weld of instrumented fuel rods with the nuclear pellets and the sensors for the irradiation test, and (4) the qualification tests(the helium leak test, the dimensional measurement, electric resistance measurements of sensors) of test fuel rods. Satisfactory results were obtained for all the qualification tests of the instrumented fuel rods for the 3-Pin FTL. Therefore the three instrumented fuel rods for the 3-Pin FTL have been fabricated successfully. These will be installed in the In-Pile Section of 3-Pin FTL. And the irradiation test of these fuel rods is planned from the early next year for about 3 years at HANARO

  2. Testing the master constraint programme for loop quantum gravity: V. Interacting field theories

    International Nuclear Information System (INIS)

    Dittrich, B; Thiemann, T

    2006-01-01

    This is the fifth and final paper in our series of five in which we test the master constraint programme for solving the Hamiltonian constraint in loop quantum gravity. Here we consider interacting quantum field theories, specifically we consider the non-Abelian Gauss constraints of Einstein-Yang-Mills theory and 2 + 1 gravity. Interestingly, while Yang-Mills theory in 4D is not yet rigorously defined as an ordinary (Wightman) quantum field theory on Minkowski space, in background-independent quantum field theories such as loop quantum gravity (LQG) this might become possible by working in a new, background-independent representation. While for the Gauss constraint the master constraint can be solved explicitly, for the 2 + 1 theory we are only able to rigorously define the master constraint operator. We show that the, by other methods known, physical Hilbert is contained in the kernel of the master constraint, however, to systematically derive it by only using spectral methods is as complicated as for 3 + 1 gravity and we therefore leave the complete analysis for 3 + 1 gravity

  3. Testing SUSY at the LHC: Electroweak and Dark matter fine tuning at two-loop order

    CERN Document Server

    Cassel, S; Ross, G G

    2010-01-01

    In the framework of the Constrained Minimal Supersymmetric Standard Model (CMSSM) we evaluate the electroweak fine tuning measure that provides a quantitative test of supersymmetry as a solution to the hierarchy problem. Taking account of current experimental constraints we compute the fine tuning at two-loop order and determine the limits on the CMSSM parameter space and the measurements at the LHC most relevant in covering it. Without imposing the LEPII bound on the Higgs mass, it is shown that the fine tuning computed at two-loop has a minimum $\\Delta=8.8$ corresponding to a Higgs mass $m_h=114\\pm 2$ GeV. Adding the constraint that the SUSY dark matter relic density should be within present bounds we find $\\Delta=15$ corresponding to $m_h=114.7\\pm 2$ GeV and this rises to $\\Delta=17.8$ ($m_h=115.9\\pm 2$ GeV) for SUSY dark matter abundance within 3$\\sigma$ of the WMAP constraint. We extend the analysis to include the contribution of dark matter fine tuning. In this case the overall fine tuning and Higgs mas...

  4. Operation manual for the core flow test loop zone power-supply controller

    Energy Technology Data Exchange (ETDEWEB)

    Harper, R.E.

    1981-11-01

    The core flow test loop, which is part of the Gas-Cooled Fast Breeder Reactor Program (GCFR) at ORNL, is a high-pressure, high-temperature, out-of-reactor helium circulation system that is being constructed to permit study of the performance at steady-state and transient conditions of simulated segments of core assemblies for a GCFR demonstration plant. The simulated core segments, which are divided into zones, contain electrical heating elements to simulate the heat generated by fission. To control the power which is applied to a zone, a novel multitapped transformer and zone power control system have been designed and built which satisfy stringent design criteria. The controller can match power output to demand to within better than +-1% over a 900:1 dynamic range and perform full-power transients within 1 s. The power is applied in such a way as to minimize the electromagnetic interference at the bandwidth of the loop instrumentation, and the controller incorporates several error detection techniques, making it inherently fail-safe. The operation manual describes the specifications, operating instructions, error detection capabilities, error recovery, troubleshooting, calibration and QA procedures, and maintenance requirements. Also included are sections on the theory of operation, circuitry description, and a complete set of schematics.

  5. Home-Based Risk of Falling Assessment Test Using a Closed-Loop Balance Model.

    Science.gov (United States)

    Ayena, Johannes C; Zaibi, Helmi; Otis, Martin J-D; Menelas, Bob-Antoine J

    2016-12-01

    The aim of this study is to improve and facilitate the methods used to assess risk of falling at home among older people through the computation of a risk of falling in real time in daily activities. In order to increase a real time computation of the risk of falling, a closed-loop balance model is proposed and compared with One-Leg Standing Test (OLST). This balance model allows studying the postural response of a person having an unpredictable perturbation. Twenty-nine volunteers participated in this study for evaluating the effectiveness of the proposed system which includes seventeen elder participants: ten healthy elderly ( 68.4 ±5.5 years), seven Parkinson's disease (PD) subjects ( 66.28 ±8.9 years), and twelve healthy young adults ( 28.27 ±3.74 years). Our work suggests that there is a relationship between OLST score and the risk of falling based on center of pressure measurement with four low cost force sensors located inside an instrumented insole, which could be predicted using our suggested closed-loop balance model. For long term monitoring at home, this system could be included in a medical electronic record and could be useful as a diagnostic aid tool.

  6. The Self-Calibration Test of flowmeter installed in STELLA(Sodium Integral Effect Test Loop for Safety Simulation and Assessment) facility

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Minhwan; Jeong, Ji-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The objective of this study is to describe the procedure of the self-calibration test for the flowmeters and to analyze the result of the test. In this work, the test procedure of the self-calibration of two flowmeters (FT-101, FT-102) installed in STELLA facility was described and the test result was analyzed. In regard to the long-term SFR development plan, a large-scale sodium thermal-hydraulic test project is being progressed by KAERI. This project is called STELLA (Sodium Integral Effect Test Loop for Safety Simulation and Assessment), and it is proceeding by adopting the QA (Quality Assurance) program. Due to the specificity of an experiment using sodium(Na) categorized as Class 3(pyrophoric material and water-prohibiting substance) by the Safety Control of Dangerous Substances Act, it is necessary to apply QA in consideration of the sodium experiment environment in certain parts. The one of them is about calibration of measuring instrument such as a flowmeter, thermocouple and pressure gauge. It is described in the QAP (Quality Assurance Procedures) of KAERI that calibration work should be conducted in accordance with self-calibration procedures in a special case where conventional calibration is not practicable. The calibration of two flowmeters (FT-101, FT-102) installed in STELLA facility is the typical example. As a result of test, it was confirmed that the flowmeters meet the pass criterion. Therefore, it was concluded that the flowmeters maintain instrument capacity a year ago.

  7. Automated ECU software tests on hardware-in-the-loop test benches; Automatisierte ECU-Software-Tests an Hardware-in-the-Loop-Pruefstaenden

    Energy Technology Data Exchange (ETDEWEB)

    Voegl, R. [AVL List GmbH, Graz (Austria). Abteilung Kalibrierung Ottomotoren; Duerager, Ch. [AVL List GmbH, Graz (Austria). Abteilung Kalibriermethodik; Beer, W.; Martini, E. [AVL List GmbH, Graz (Austria)

    2005-08-01

    Due to the continuous increase in complexity of engine application, AVL List decided to develop methods for automated ECU software commissioning on HiL test benches. This required the intensive co-operation of the departments for calibration methodology, calibration and electrical/electronic engineering. The result is a practical orientated collection of methods that significantly increase the test coverage for a new software version without extending the commissioning time. (orig.)

  8. Improvement of Measurement Accuracy of Coolant Flow in a Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jintae; Kim, Jong-Bum; Joung, Chang-Young; Ahn, Sung-Ho; Heo, Sung-Ho; Jang, Seoyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, to improve the measurement accuracy of coolant flow in a coolant flow simulator, elimination of external noise are enhanced by adding ground pattern in the control panel and earth around signal cables. In addition, a heating unit is added to strengthen the fluctuation signal by heating the coolant because the source of signals are heat energy. Experimental results using the improved system shows good agreement with the reference flow rate. The measurement error is reduced dramatically compared with the previous measurement accuracy and it will help to analyze the performance of nuclear fuels. For further works, out of pile test will be carried out by fabricating a test rig mockup and inspect the feasibility of the developed system. To verify the performance of a newly developed nuclear fuel, irradiation test needs to be carried out in the research reactor and measure the irradiation behavior such as fuel temperature, fission gas release, neutron dose, coolant temperature, and coolant flow rate. In particular, the heat generation rate of nuclear fuels can be measured indirectly by measuring temperature variation of coolant which passes by the fuel rod and its flow rate. However, it is very difficult to measure the flow rate of coolant at the fuel rod owing to the narrow gap between components of the test rig. In nuclear fields, noise analysis using thermocouples in the test rig has been applied to measure the flow velocity of coolant which circulates through the test loop.

  9. Water Recovery from Brines to Further Close the Water Recovery Loop in Human Spaceflight

    Science.gov (United States)

    Jackson, W. Andrew; Barta, Daniel J.; Anderson, Molly S.; Lange, Kevin E.; Hanford, Anthony J.; Shull, Sarah A.; Carter, D. Layne

    2014-01-01

    Further closure of water recovery systems will be necessary for future long duration human exploration missions. NASA's Space Technology Roadmap for Human Health, Life Support and Habitation Systems specified a milestone to advance water management technologies during the 2015 to 2019 timeframe to achieve 98% H2O recovery from a mixed wastewater stream containing condensate, urine, hygiene, laundry, and water derived from waste. This goal can only be achieved by either reducing the amount of brines produced by a water recovery system or by recovering water from wastewater brines. NASA convened a Technical Interchange Meeting (TIM) on the topic of Water Recovery from Brines (WRB) that was held on January14-15th, 2014 at Johnson Space Center. Objectives of the TIM were to review systems and architectures that are sources of brines and the composition of brines they produce, review the state of the art in NASA technology development and perspectives from other industries, capture the challenges and difficulties in developing brine processing hardware, identify key figures of merit and requirements to focus technology development and evaluate candidate technologies, and identify other critical issues including microgravity sensitivity, and concepts of operation, safety. This paper represents an initial summary of findings from the workshop.

  10. Water from abandoned mines as a heat source: practical experiences of open- and closed-loop strategies, United Kingdom

    OpenAIRE

    Banks, D; Athresh, A; Al-Habaibeh, A; Burnside, N

    2017-01-01

    Pilot heat pump systems have been installed at two former collieries in Yorkshire/Derbyshire, England, to extract heat from mine water. The installations represent three fundamental configurations of heat exchanger. At Caphouse Colliery, mine water is pumped through a heat exchanger coupled to a heat pump and then discharged to waste (an open-loop heat exchange system). The system performs with high thermal efficiency, but the drawbacks are: (1) it can only be operated when mine water is bein...

  11. Modification of heating system on HeaTiNG-02 test section of beta test loop

    International Nuclear Information System (INIS)

    Sagino; Dedy Haryanto; Riswan Djambiar; Edy Sumarno

    2013-01-01

    Modifications have been carried out on the heating test section heating-02 on the integration strand Beta Test (UUB). The activities carried out to overcome the obstacles that arise in the test section when used. Constraint that often arises is the fall of the heating source super chantal when it reaches a certain temperature. To mitigate the super chantal is initially converted into a horizontal vertical position. Change from vertical to horizontal position on super chantal aims to stabilize the position of super chantal, so it needs to be modified in the heating system. Modification activities include manufacturing, installation and testing of super chantal and refractory stone as super chantal support. Manufacturing refractory stone formation and assembly into the heater in accordance with design modifications that have been done in electromechanical workshop obtained using some machine tools. Testing results of fabrication has been done by providing voltage 110 volts until it reaches operating temperature 400°C. Test results obtained super chantal stable position when it reaches operating temperature, and heater of heating-02 test section feasible to be used for experiments. (author)

  12. Real-Time Hardware-in-the-Loop Laboratory Testing for Multisensor Sense and Avoid Systems

    Directory of Open Access Journals (Sweden)

    Giancarmine Fasano

    2013-01-01

    Full Text Available This paper focuses on a hardware-in-the-loop facility aimed at real-time testing of architectures and algorithms of multisensor sense and avoid systems. It was developed within a research project aimed at flight demonstration of autonomous non-cooperative collision avoidance for Unmanned Aircraft Systems. In this framework, an optionally piloted Very Light Aircraft was used as experimental platform. The flight system is based on multiple-sensor data integration and it includes a Ka-band radar, four electro-optical sensors, and two dedicated processing units. The laboratory test system was developed with the primary aim of prototype validation before multi-sensor tracking and collision avoidance flight tests. System concept, hardware/software components, and operating modes are described in the paper. The facility has been built with a modular approach including both flight hardware and simulated systems and can work on the basis of experimentally tested or synthetically generated scenarios. Indeed, hybrid operating modes are also foreseen which enable performance assessment also in the case of alternative sensing architectures and flight scenarios that are hardly reproducible during flight tests. Real-time multisensor tracking results based on flight data are reported, which demonstrate reliability of the laboratory simulation while also showing the effectiveness of radar/electro-optical fusion in a non-cooperative collision avoidance architecture.

  13. An automatic sodium-loop for testing the lon-term behaviour of sintered bodies flowed through by gas

    International Nuclear Information System (INIS)

    Barkleit, G.; George, G.; Haase, I.; Kiessling, W.

    1980-08-01

    An automatic sodium loop NAKOS for testing the long-term behaviour of porous stainless steel bodies which are flowed through by gas is described. The loop using a special safety protection system is capable of working without control up to 1000 h. During a 500 h-experiment the safety system and the gas permeability measuring method for testing the porous bodies were tested. Both first results of the behaviour of sintered bodies in liquid sodium of high purity and temperatures of about 850 K and some details of the production of these bodies are given. (author)

  14. A hardware-in-the-loop simulation platform for prototyping and testing of wind generator controllers

    Energy Technology Data Exchange (ETDEWEB)

    Paquin, J.N.; Dufour, C.; Belanger, J. [OPAL-RT Technologies Inc., Montreal, PQ (Canada)

    2008-07-01

    Engineers from different specialized fields need to be involved in meeting the growing demand for integrated renewable energy sources into existing power grids. The integration of distributed generation (DG) sources significantly changes the characteristics of an entire network and requires analysis of power quality, transient response to fault occurrences, protection coordination studies and controller interaction studies. Power electronic converters are a considerable challenge. Accurately simulating fast switching devices requires the use of very small time steps to solve the system's equations. Off-line simulation is often used in the field. However, it is time consuming if no precision compromise has been made on models. In addition, off-line simulation tools do not offer the wide range of possibilities available with state-of-the-art distributed real-time simulators that combine the efforts of control engineers and specialists from wind turbine manufacturers, who need to test their controllers using hardware-in-the-loop (HIL), together with those of network planning engineers from public utilities, who will conduct interconnection, interaction and protection studies. This paper focused on the prototyping and testing of DG controllers using hardware-in-the-loop simulation. The model was described and consisted of a 10-turbine wind farm connected to a single feeder, simulated using an eMEGAsim real-time simulator equipped with 8-processor cores. One of the wind turbines was controlled using an externally emulated controller. It was modeled and simulated using a dual-processor core real-time simulator, which interacted with the plant model via analog and fast digital inputs and outputs. The effectiveness of the technology was demonstrated by comparing fully numerical simulation results with an HIL-connected DFIG controller simulation. The sampling effect of the digital simulator was correctly compensated for. The simulator could be driven directly by real

  15. The temperature control and water quality regulation for steam generator secondary side hydrostatic test

    International Nuclear Information System (INIS)

    Xiao Bo; Liu Dongyong

    2014-01-01

    The secondary side hydrostatic test for the steam generator of M310 unit is to verify the pressure tightness of steam generator secondary side tube sheet and related systems. As for the importance of the steam generator, the water temperature and water quality of hydrostatic test has strict requirements. The discussion on the water temperature control and water quality regulation for the secondary loop hydrostatic test of Fuqing Unit 1 contribute greatly to the guiding work for the preparation of the steam generator pressure test for M310 unit. (authors)

  16. Integrated Human-in-the-Loop Ground Testing - Value, History, and the Future

    Science.gov (United States)

    Henninger, Donald L.

    2016-01-01

    Systems for very long-duration human missions to Mars will be designed to operate reliably for many years and many of these systems will never be returned to Earth. The need for high reliability is driven by the requirement for safe functioning of remote, long-duration crewed systems and also by unsympathetic abort scenarios. Abort from a Mars mission could be as long as 450 days to return to Earth. The key to developing a human-in-the-loop architecture is a development process that allows for a logical sequence of validating successful development in a stepwise manner, with assessment of key performance parameters (KPPs) at each step; especially important are KPPs for technologies evaluated in a full systems context with human crews on Earth and on space platforms such as the ISS. This presentation will explore the implications of such an approach to technology development and validation including the roles of ground and space-based testing necessary to develop a highly reliable system for long duration human exploration missions. Historical development and systems testing from Mercury to the International Space Station (ISS) to ground testing will be reviewed. Current work as well as recommendations for future work will be described.

  17. Diagnostics of high-speed liquid lithium jet for IFMIF/EVEDA lithium test loop

    International Nuclear Information System (INIS)

    Kanemura, Takuji; Kondo, Hiroo; Furukawa, Tomohiro; Sugiura, Hirokazu; Horiike, Hiroshi; Yamaoka, Nobuo; Ida, Mizuho; Nakamura, Kazuyuki; Matsushita, Izuru

    2011-01-01

    Regarding R and Ds on the International Fusion Materials Irradiation Facility (IFMIF), hydraulic stability of the liquid Li jet simulating the IFMIF Li target is planned to be validated using EVEDA Li Test Loop (ELTL). IFMIF is an accelerator-based deuteron-lithium (Li) neutron source for research and development of fusion reactor materials. The stable Li target is required in IFMIF to maintain the quality of the neutron fluence and integrity of the Li target itself. This paper presents diagnostics of the Li jet to be implemented in validation tests of the jet stability in ELTL, and those specifications and methodologies are introduced. In the tests, the following physical parameters need to be measured; thickness of the jet; surface structure (height, length/width and frequency of free-surface waves); local flow velocity at the free surface; and Li evaporation rate. With regard to measurement of jet thickness and the surface wave height, a contact-type liquid level sensor is to be used. As for measurement of wave velocity and visual understanding of detailed free-surface structure, a high-speed video camera is to be leveraged. With respect to Li evaporation measurement, weight change of specimens installed near the free surface and frequency change of a crystal quartz are utilized. (author)

  18. Pre-test analysis of a LBLOCA using the design data of the ATLAS facility, a reduced-height integral effect test loop for PWRs

    International Nuclear Information System (INIS)

    Hyun-Sik Park; Ki-Yong Choi; Dong-Jin Euh; Tae-Soon Kwon; Won-Pil Baek

    2005-01-01

    Full text of publication follows: The simulation capability of the KAERI integral effect test facility, ATLAS (Advanced Thermalhydraulic Test Loop for Accident Simulation), has been assessed for a large-break loss-of-coolant accident (LBLOCA) transient. The ATLAS facility is a 1/2 height-scaled, 1/144 area-scaled (1/288 in volume scale), and full-pressure test loop based on the design features of the APR1400, an evolutionary pressurized water reactor that has been developed by Korean industry. The APR1400 has four mechanically separated hydraulic trains for the emergency core cooling system (ECCS) with direct vessel injection (DVI). The APR1400 design features have brought about several new safety issues related to the LBLOCA including the steam-water interaction, ECC bypass, and boiling in the reactor vessel downcomer. The ATLAS facility will be used to investigate the multiple responses between the systems or between the components during various anticipated transients. The ATLAS facility has been designed according to a scaling method that is mainly based on the model suggested by Ishii and Kataoka. The ATLAS facility is being evaluated against the prototype plant APR1400 with the same control logics and accident scenarios using the best-estimated code, MARS. This paper briefly introduces the basic design features of the ATLAS facility and presents the results of pre-test analysis for a postulated LBLOCA of a cold leg. The LBLOCA analyses has been conducted to assess the validity of the applied scaling law and the similarity between the ATLAS facility and the APR1400. As the core simulator of the ATLAS facility has the 10% capability of the scaled full power, the blowdown phase can not be simulated, and the starting point of the accident scenario is around the end of blowdown. So it is an important problem to find the correct initial conditions. For the analyzed LBLOCA scenario, the ATLAS facility showed very similar thermal-hydraulic characteristics to the APR

  19. The Impact of Curriculum Looping on Standardized Literacy and Mathematics Test Scores

    Science.gov (United States)

    Nessler, Ralph D.

    2010-01-01

    There is a lack of research on the practice of curriculum looping and student achievement. The purpose of this study was to examine academic achievement between students in looping classes and those in nonlooping classes. The theoretical model of this study was based on the social cognitive theory of Bandura and Maslow's hierarchy of needs theory.…

  20. Excitation power quantities in phase resonance testing of nonlinear systems with phase-locked-loop excitation

    Science.gov (United States)

    Peter, Simon; Leine, Remco I.

    2017-11-01

    Phase resonance testing is one method for the experimental extraction of nonlinear normal modes. This paper proposes a novel method for nonlinear phase resonance testing. Firstly, the issue of appropriate excitation is approached on the basis of excitation power considerations. Therefore, power quantities known from nonlinear systems theory in electrical engineering are transferred to nonlinear structural dynamics applications. A new power-based nonlinear mode indicator function is derived, which is generally applicable, reliable and easy to implement in experiments. Secondly, the tuning of the excitation phase is automated by the use of a Phase-Locked-Loop controller. This method provides a very user-friendly and fast way for obtaining the backbone curve. Furthermore, the method allows to exploit specific advantages of phase control such as the robustness for lightly damped systems and the stabilization of unstable branches of the frequency response. The reduced tuning time for the excitation makes the commonly used free-decay measurements for the extraction of backbone curves unnecessary. Instead, steady-state measurements for every point of the curve are obtained. In conjunction with the new mode indicator function, the correlation of every measured point with the associated nonlinear normal mode of the underlying conservative system can be evaluated. Moreover, it is shown that the analysis of the excitation power helps to locate sources of inaccuracies in the force appropriation process. The method is illustrated by a numerical example and its functionality in experiments is demonstrated on a benchmark beam structure.

  1. Irradiation of Superheater Test Fuel Elements in the Steam Loop of the R2 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ravndal, F

    1967-12-15

    The design, fabrication, irradiation results, and post-irradiation examination for three superheater test fuel elements are described. During the spring of 1966 these clusters, each consisting of six fuel rods, were successfully exposed in the superheater loop No. 5 in the R2 reactor for a maximum of 24 days at a maximum outer cladding surface temperature of {approx} 650 deg C. During irradiation the linear heat rating of the rods was in the range 400-535 W/cm. The diameter of the UO{sub 2} pellets was 11.5 and 13.0 mm; the wall thickness of the 20/25 Nb and 20/35 cladding was in every case 0.4 mm. The diametrical gap between fuel and cladding was one of the main parameters and was chosen to be 0.05, 0.07 and 0.10 mm. These experiments, to be followed by one high cladding temperature irradiation ({approx} 750 deg C) and one long time irradiation ({approx} 6000 MWd/tU), were carried out to demonstrate the operational capability of short superheater test fuel rods at steady and transient operational environments for the Marviken superheater fuel elements and also to provide confirmation of design criteria for the same fuel elements.

  2. Thermal performance test of hot gas ducts of helium engineering demonstration loop (HENDEL)

    International Nuclear Information System (INIS)

    Hishida, Makoto; Kunitomi, Kazuhiko; Ioka, Ikuo; Umenishi, Koji; Kondo, Yasuo; Tanaka, Toshiyuki; Shimomura, Hiroaki

    1984-01-01

    A hot gas duct provided with internal thermal insulation is supposed to be used for an experimental very high-temperature gas-cooled reactor (VHTR) which has been developed by the Japan Atomic Energy Research Institute (JAERI). This type of hot gas duct has not been used so far in industrial facilities, and only a couple of tests on such a large-scale model of hot gas duct have been conducted. The present test was to investigate the thermal performance of the hot gas ducts which are installed as parts of a helium engineering demonstration loop (HENDEL) of JAERI. Uniform temperature and heat flux distributions at the surface of the duct were observed, the experimental correlation being obtained for the effective thermal conductivity of the internal thermal insulation layer. The measured temperature distribution of the pressure tube was in good agreement with the calculation by a TRUMP heat transfer computer code. The temperature distribution of the inner tube of VHTR hot gas duct was evaluated, and no hot spot was detected. These results would be very valuable for the design and development of VHTR. (author)

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

    International Nuclear Information System (INIS)

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

    2010-11-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  5. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  6. Numerical Simulation of a Single-Phase Closed-Loop Thermo-Siphon in LORELEI Test Device

    International Nuclear Information System (INIS)

    Gitelman, D.; Shenha, H.; Gonnier, Ch.; Tarabelli, D.; Sasson, A.; Weiss, Y.; Katz, M.

    2014-01-01

    The LORELEI experimental setup in the Jules Horowitz Reactor (JHR) is dedicated for the study of fuel during a Loss of Coolant Accident (LOCA). The main objective of the LORELEI(2) (Light-Water One-Rod Equipment for LOCA Experimental Investigation) is to study the thermal-mechanical behavior of fuel during such an accident and to produce a short half-life fission products source term. In order to study those phenomena, the fuel sample will experience a transient neutron flux field, which in turn will generate a Linear Heat Generation Rate (LHGR) and determine the temperature of the fuel and its cladding, simulating the behavior of the fuel and the cladding during a LOCA accident. In order to reproduce a LOCA-type transient sequence, the experimental test device will be located on a displacement device. The displacement device moves the test device in the flux field in order to generate a representing LHGR in the fuel or temperature of its cladding. The LOCA-type transient sequence has four major features: „h An adiabatic heating of the fuel up to the ballooning and burst occurrence. „h High temperature plateau which will promote clad oxidation. „h Passive precooling by thermal inertia. „h Water re-flooding and quenching. The challenge in the thermo-hydraulic design of the LORELEI test section is in defining a one closed water capsule design that can operate as a thermo-siphon at re-irradiation phase and also can reproduce all LOCA-type transient sequence phases. This design should be validated and verified to fill all safety and regulation requirements. This work aims to investigate fluid flow behavior of a single-phase thermo-siphon in the LORELEI test device, as part of the conceptual design and optimization study. The complexity of the flow field in the LORELEI test device, as a closed-loop thermo-siphon, is due to the opposing forces in the device - buoyancy forces and natural convection flow generated (mainly) by the fuel power in the hot channel

  7. UAS-NAS Integrated Human in the Loop: Test Environment Report

    Science.gov (United States)

    Murphy, Jim; Otto, Neil; Jovic, Srba

    2015-01-01

    The desire and ability to fly Unmanned Aircraft Systems (UAS) in the National Airspace System (NAS) is of increasing urgency. The application of unmanned aircraft to perform national security, defense, scientific, and emergency management are driving the critical need for less restrictive access by UAS to the NAS. UAS represent a new capability that will provide a variety of services in the government (public) and commercial (civil) aviation sectors. The growth of this potential industry has not yet been realized due to the lack of a common understanding of what is required to safely operate UAS in the NAS. NASA's UAS Integration in the NAS Project is conducting research in the areas of Separation Assurance/Sense and Avoid Interoperability (SSI), Human Systems Integration (HSI), and Communication to support reducing the barriers of UAS access to the NAS. This research was broken into two research themes namely, UAS Integration and Test Infrastructure. UAS Integration focuses on airspace integration procedures and performance standards to enable UAS integration in the air transportation system, covering Sense and Avoid (SAA) performance standards, command and control performance standards, and human systems integration. The focus of the Test Infrastructure theme was to enable development and validation of airspace integration procedures and performance standards, including the execution of integrated test and evaluation. In support of the integrated test and evaluation efforts, the Project developed an adaptable, scalable, and schedulable relevant test environment incorporating live, virtual, and constructive elements capable of validating concepts and technologies for unmanned aircraft systems to safely operate in the NAS. To accomplish this task, the Project planned to conduct three integrated events: a Human-in-the-Loop simulation and two Flight Test series that integrated key concepts, technologies and/or procedures in a relevant air traffic environment. Each of

  8. An Installation of IPS Bypass Line at the Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Ho Young; Ahn, G. H.; Lee, M.; Kim, M. S.; Cho, S. H.; Han, J. S.; Hur, S. O. [KAERI, Daejeon (Korea, Republic of)

    2010-12-15

    The Fuel Test Loop(FTL) was installed for the national goal of self-supporting technology in the field of design and construction of nuclear power plant. The FTL with the fuel irradiation equipment is essential in developing, improving and inspecting the fuel of CANDU type or PWR type nuclear power plant. The FTL should be operated at the same conditions of commercial nuclear power plant such as temperature, pressure, flow rate, neutron flux and so on. Starting designing in December 2001, the FTL was installed from March 2007 to August 2008. Especially the In Pile Section(IPS) was installed at IR1 hole in August 2008. Until September 2009 after loading the test fuel, a series of power escalation tests (LSD, CSB1, CSB2, HSB, HOP) were conducted. And it was operated at the condition of CSB2 for the 8 cycles from October 2009 to July 2010. But it could not be normally operated in early 2010, because the high radiation released from irradiated materials due to the worn down bearing of main cooling pump. So, we removed the IPS and installed a newly designed IPS bypass line to prevent increasing high radiation. In this report we will present preliminary works, main works processes, devices of making work environments, a designing and manufacturing of IPS bypass line and a rack of IPS, installing know-hows, problems and solutions broke out during the work etc. We believe that our efforts to complete successful installing and operating of the FTL system will contribute for the efficient utilization of HANARO

  9. An Experimental Study of Natural Circulation in a Loop with Parallel Flow Test Sections

    International Nuclear Information System (INIS)

    Mathisen, R.P.; Eklind, O.

    1965-10-01

    The dynamic behaviour of a natural circulation loop parallel round duct channels has been studied. The test sections were both electrically heated and the power distribution was uniform along the 4300 mm heated length of the 20 mm dia. channels. The inter channel interference and the threshold of flow instability were obtained by using a dynamically calibrated flowmeter in each channel. The pressure was 50 bars and the sub-cooling 6 deg C. The main parameters varied, were the flow restrictions in the one-phase and two-phase sections. The instability data were correlated to the resistance coefficients due to these restrictions. Theoretical calculations for parallel channels in natural circulation have been compared with the experimental results. For the conditions determined by the above mentioned magnitudes, the steady-state computations are in excellent agreement with experiment. The transients are also nearly similar, except for the resonance frequency which for the theoretical case is higher by an amount between 0.3 and 0.5 c.p.s

  10. Battery algorithm verification and development using hardware-in-the-loop testing

    Energy Technology Data Exchange (ETDEWEB)

    He, Yongsheng [General Motors Global Research and Development, 30500 Mound Road, MC 480-106-252, Warren, MI 48090 (United States); Liu, Wei; Koch, Brain J. [General Motors Global Vehicle Engineering, Warren, MI 48090 (United States)

    2010-05-01

    Battery algorithms play a vital role in hybrid electric vehicles (HEVs), plug-in hybrid electric vehicles (PHEVs), extended-range electric vehicles (EREVs), and electric vehicles (EVs). The energy management of hybrid and electric propulsion systems needs to rely on accurate information on the state of the battery in order to determine the optimal electric drive without abusing the battery. In this study, a cell-level hardware-in-the-loop (HIL) system is used to verify and develop state of charge (SOC) and power capability predictions of embedded battery algorithms for various vehicle applications. Two different batteries were selected as representative examples to illustrate the battery algorithm verification and development procedure. One is a lithium-ion battery with a conventional metal oxide cathode, which is a power battery for HEV applications. The other is a lithium-ion battery with an iron phosphate (LiFePO{sub 4}) cathode, which is an energy battery for applications in PHEVs, EREVs, and EVs. The battery cell HIL testing provided valuable data and critical guidance to evaluate the accuracy of the developed battery algorithms, to accelerate battery algorithm future development and improvement, and to reduce hybrid/electric vehicle system development time and costs. (author)

  11. Experimental tests for the Babu-Zee two-loop model of Majorana neutrino masses

    International Nuclear Information System (INIS)

    Sierra, Diego Aristizabal; Hirsch, Martin

    2006-01-01

    The smallness of the observed neutrino masses might have a radiative origin. Here we revisit a specific two-loop model of neutrino mass, independently proposed by Babu and Zee. We point out that current constraints from neutrino data can be used to derive strict lower limits on the branching ratio of flavour changing charged lepton decays, such as μ→eγ. Non-observation of Br(μ→eγ) at the level of 10 -13 would rule out singly charged scalar masses smaller than 590 GeV (5.04 TeV) in case of normal (inverse) neutrino mass hierarchy. Conversely, decay branching ratios of the non-standard scalars of the model can be fixed by the measured neutrino angles (and mass scale). Thus, if the scalars of the model are light enough to be produced at the LHC or ILC, measuring their decay properties would serve as a direct test of the model as the origin of neutrino masses

  12. Experimental tests for the Babu-Zee two-loop model of Majorana neutrino masses

    International Nuclear Information System (INIS)

    Aristizabal, D.

    2006-01-01

    Abstract: The smallness of the observed neutrino masses might have a radiative origin. Here we revisit a specific two-loop model of neutrino mass, independently proposed by Babu and Zee. We point out that current constraints from neutrino data can be used to derive strict lower limits on the branching ratio of flavour changing charged lepton decays, such as μ → e γ. Non-observation of Br(μ → e γ) at the level of 10 -13 would rule out singly charged scalar masses smaller than 590 GeV (5.04 TeV) in case of normal (inverse) neutrino mass hierarchy. Conversely, decay branching ratios of the non-standard scalars of the model can be fixed by the measured neutrino angles (and mass scale). Thus, if the scalars of the model are light enough to be produced at the LHC or ILC, measuring their decay properties would serve as a direct test of the model as the origin of neutrino masses. (author)

  13. Battery algorithm verification and development using hardware-in-the-loop testing

    Science.gov (United States)

    He, Yongsheng; Liu, Wei; Koch, Brain J.

    Battery algorithms play a vital role in hybrid electric vehicles (HEVs), plug-in hybrid electric vehicles (PHEVs), extended-range electric vehicles (EREVs), and electric vehicles (EVs). The energy management of hybrid and electric propulsion systems needs to rely on accurate information on the state of the battery in order to determine the optimal electric drive without abusing the battery. In this study, a cell-level hardware-in-the-loop (HIL) system is used to verify and develop state of charge (SOC) and power capability predictions of embedded battery algorithms for various vehicle applications. Two different batteries were selected as representative examples to illustrate the battery algorithm verification and development procedure. One is a lithium-ion battery with a conventional metal oxide cathode, which is a power battery for HEV applications. The other is a lithium-ion battery with an iron phosphate (LiFePO 4) cathode, which is an energy battery for applications in PHEVs, EREVs, and EVs. The battery cell HIL testing provided valuable data and critical guidance to evaluate the accuracy of the developed battery algorithms, to accelerate battery algorithm future development and improvement, and to reduce hybrid/electric vehicle system development time and costs.

  14. An Experimental Study of Natural Circulation in a Loop with Parallel Flow Test Sections

    Energy Technology Data Exchange (ETDEWEB)

    Mathisen, R P; Eklind, O

    1965-10-15

    The dynamic behaviour of a natural circulation loop parallel round duct channels has been studied. The test sections were both electrically heated and the power distribution was uniform along the 4300 mm heated length of the 20 mm dia. channels. The inter channel interference and the threshold of flow instability were obtained by using a dynamically calibrated flowmeter in each channel. The pressure was 50 bars and the sub-cooling 6 deg C. The main parameters varied, were the flow restrictions in the one-phase and two-phase sections. The instability data were correlated to the resistance coefficients due to these restrictions. Theoretical calculations for parallel channels in natural circulation have been compared with the experimental results. For the conditions determined by the above mentioned magnitudes, the steady-state computations are in excellent agreement with experiment. The transients are also nearly similar, except for the resonance frequency which for the theoretical case is higher by an amount between 0.3 and 0.5 c.p.s.

  15. 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)

  16. Very High Temperature Test of Alloy617 Compact Heat Exchanger in Helium Experimental Loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chan Soo; Park, Byung-Ha; Kim, Eung-Seon [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The Intermediate Heat eXchanger (IHX) is a key-challenged high temperature component which determines the efficiency and the economy of VHTR system. Heat generated in the VHTR fuel block is transferred from the VHTR to the intermediate loop through IHX. In the present, the shell-helical tube heat exchanger is generally used as IHX of the helium cooled reactor. Recently, a Printed Circuit Heat Exchanger (PCHE) is one of the candidates for the IHX in a VHTR because its operation temperature and pressure are larger than any other compact heat exchanger types. These test results show that there is no problem in operation of HELP at the very high temperature experimental condition and the alloy617 compact heat exchanger can be operated in the very high temperature condition above 850℃. In the future, the high temperature structural analysis will be studied to estimate the thermal stress during transient and thermal shock condition. The conditions and evaluation standard for the alloy 617 diffusion bonding will be minutely studied to fabricate the large-scale PCHE for the high temperature condition.

  17. Nevada test site water-supply wells

    International Nuclear Information System (INIS)

    Gillespie, D.; Donithan, D.; Seaber, P.

    1996-05-01

    A total of 15 water-supply wells are currently being used at the Nevada Test Site (NTS). The purpose of this report is to bring together the information gleaned from investigations of these water-supply wells. This report should serve as a reference on well construction and completion, static water levels, lithologic and hydrologic characteristics of aquifers penetrated, and general water quality of water-supply wells at the NTS. Possible sources for contamination of the water-supply wells are also evaluated. Existing wells and underground nuclear tests conducted near (within 25 meters (m)) or below the water table within 2 kilometers (km) of a water-supply were located and their hydrogeologic relationship to the water-supply well determined

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

    International Nuclear Information System (INIS)

    Covelli, B.; Alder, H.P.

    1988-03-01

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

  19. Design and Testing of Bi-Functional, P-Loop-Targeted MDM2 Inhibitors

    National Research Council Canada - National Science Library

    Prives, Carol L; Stockwell, Brent R

    2007-01-01

    Our proposal is to design and evaluate a novel class of bifunctional MDM2 inhibitors, based on the discovery that nucleotides can bind to the P-loop of MDM2 and cause its relocalization to the nucleolus...

  20. Design and Testing of Bi-Functional, P-Loop-Targeted MDM2 Inhibitors

    National Research Council Canada - National Science Library

    Prives, Carol L

    2006-01-01

    This proposal is to design and evaluate a novel class of bifunctional MDM2 inhibitors, based on the discovery that nucleotides can bind to the P-loop of MDM2 and cause its relocalization to the nucleolus...

  1. Ballast Water Treatment Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: Provides functionality for the full-scale testing and controlled simulation of ship ballasting operations for assessment of aquatic nuisance species (ANS)...

  2. Heat transfer in the in-pile test section and penetration region of 3-pin fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Dae Young; Lee, Chung Young; Sim, Bong Shick; Park, Kook Nam; Park, Su Ki; Lee, Jong Min; Kim, Young Jin

    2003-12-01

    This report studies two types of normal heat transfer. One is the heat loss from the pressure vessel of In-Pile Test Section to HANARO pool water via IPS insulation gas gap. The other is the heat transfer of the Penetration Cooling Water System including the effect of the Foamglas insulator at the penetration region. The heat transfer from IPS insulation gas gap has been performed according to the detail design results from NUKEM. The heat loss also occurs at the concrete penetration region between the HANARO pool water and the FTL pipe gallery. The Foamglas insulator has been already installed at the MCW piping of the penetration region. This insulation effect has been reviewed. The Penetration Cooling Water System has been designed to fulfill the design requirement not to exceed the allowable temperature at the penetration concrete wall. The cooling ability and heat loss of PCW system has been reviewed with the insulation effect.

  3. Power-Hardware-In-the-Loop (PHIL) Test of VSC-based HVDC connection for Offshore Wind Power Plants (WPPs)

    DEFF Research Database (Denmark)

    Sharma, Ranjan; Cha, Seung-Tae; Wu, Qiuwei

    2011-01-01

    This paper presents a power-hardware-in-the-loop (PHIL) test for an offshore wind power plant (WPP) interconnected to the onshore grid by a VSC-based HVDC connection. The intention of the PHIL test is to verify the control coordination between the plant side converter of the HVDC connection...... the successful control coordination between the WPP and the plant side VSC converter of the HVDC connection of the WPP....

  4. Hardware-in-the-loop environment for the design and test of regulators in the refrigeration technology; Hardware-in-the-Loop Umgebung fuer den Entwurf und Test von kaeltetechnischen Reglern

    Energy Technology Data Exchange (ETDEWEB)

    Koeberle, Thomas; Becker, Martin [Hochschule Biberach (Germany). Inst. fuer Gebaeude- und Energiesysteme

    2012-07-01

    In its Directive on energy efficiency, The European Commission has specified a target value for saving primary energy by 20 % up the year 2020. According to current projections, the target will be failed by half. This is why significantly more efforts are needed. The energy consumption of the technical supply of coldness amounts nearly 14 % of the total consumption of electrical power in Germany. The Research Council Coldness specifies that the potential for energy conservation in the refrigeration engineering amounts up to 40 %. Thus, more effort must be made in the refrigeration engineering in order to reach the energy saving targets of the European Commission. Thereby, the major portion of the potential of energy saving consists of the establishment of the requirements in regard to system concepts and components as well as the control. The control of refrigeration systems provides a simple possibility of intervention to optimize the energy efficiency. The optimization of control parameters is usually achieved only with great experience and knowledge. There are any tools which facilitate an objective comparison of optimization measures of control concepts, strategies and settings. In order to facilitate the evaluation and comparison of refrigeration controls, a hardware-in-the-loop test environment was set up at the University of Applied Sciences in Biberach (Federal Republic of Germany). The test environment facilitates an implementation of a controller in a simulation environment so that the controller drives the simulation model of the chiller. Due to this procedure, tests are possible under standardized and reproducible conditions. The impact of modified control parameters, disturbances and modifications in the regulatory approach can be investigated by means of the possibility of a targeted impacting of individual disturbances. The test rig was designed, built and tested at the University of Applied Sciences in Biberach. Simulation models were adapted to the

  5. A New Built-in Self Test Scheme for Phase-Locked Loops Using Internal Digital Signals

    Science.gov (United States)

    Kim, Youbean; Kim, Kicheol; Kim, Incheol; Kang, Sungho

    Testing PLLs (phase-locked loops) is becoming an important issue that affects both time-to-market and production cost of electronic systems. Though a PLL is the most common mixed-signal building block, it is very difficult to test due to internal analog blocks and signals. In this paper, we propose a new PLL BIST (built-in self test) using the distorted frequency detector that uses only internal digital signals. The proposed BIST does not need to load any analog nodes of the PLL. Therefore, it provides an efficient defect-oriented structural test scheme, reduced area overhead, and improved test quality compared with previous approaches.

  6. What we've learned about the installation and operation of water source, closed loop systems in Iowa

    Energy Technology Data Exchange (ETDEWEB)

    Bierbaum, A.C.

    1985-08-01

    Studies showing that groundwater heat pumps compare favorably with other heating systems convinced the Iowa Association of Electric Co-ops that there was a good market potential and an opportunity to forestall the need for additional generating capacity. The author describes the marketing effort and its experiences as well as workshops to discuss the results of a closed loop performance questionnaire and the performance of water source heat pumps at varying water temperatures. Results of the study emphasize the importance of selecting the proper heat pump and size for closed loops and the importance of manufacturers listing performances down to 25 to 30 deg. F. Oversizing as much as 200% seems to work well for air conditioning, while undersizing causes inefficient operation. 6 tables.

  7. Safety analyses for an in-pile SCWR fuel qualification test loop

    Energy Technology Data Exchange (ETDEWEB)

    Schulenberg, T.; Raque, M. [Karlsruhe Inst. of Tech., Karlsruhe (Germany)

    2014-07-01

    As a nuclear facility cooled with supercritical water has never been built nor operated in the past, the planned SCWR fuel qualification test will give the first experience with supercritical water-cooled nuclear systems in general. With a fuel inventory of almost 1 kg of UO{sub 2} with almost 20% enrichment, the supercritical pressure test section inside a low pressure, pool type research reactor needs to be cooled properly even in case of a number of postulated design basis accidents. Depressurization systems and emergency cooling systems will need to be designed with similar reliability as for a prototype reactor to ensure the integrity of barriers retaining the radioactive material. The paper reports about the safety concept and summarizes the safety analyses which have been performed in this context. (author)

  8. Using Causal Loop Diagramming to Explore the Drivers of the Sustained Functionality of Rural Water Services in Timor-Leste

    Directory of Open Access Journals (Sweden)

    Kate Neely

    2016-01-01

    Full Text Available It is recognized that international water sector development work has issues with a lack of sustained positive outcomes. A large driver of this outcome is how NGOs work with communities to implement and then manage water services. Many NGOs tend to focus their efforts on improving their reach and organisational growth by continually engaging in new projects. This behaviour is largely driven by short-term donor funding models that reward extended coverage, leaving little focus on sustained outcomes. Similarly, community-based management (CBM schemes often impede sustained services as a result of the community’s limited capacity to operate and maintain the technology. To explore these complicated drivers on water service sustainability, we used causal loop diagramming to analyse the key aspect influencing the combined dynamics between NGOs, donors and CBM. We demonstrate this methodology through a study in Timor-Leste, where we gathered data necessary to develop and apply causal loop diagrams to analyse rural water supply program outcomes. The analysis of these diagrams allowed identification of leverage points used to suggest structural changes for sustained benefits of water services. These structural changes emphasize the importance of increased robustness and reliability of water technology and the associated impact this has on community satisfaction and, conjointly, on water service sustainability.

  9. Development of treatment technology of radio-contaminated coolant in fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1997-10-01

    In 1995, the installation of KMRR located in KAERI provided a milestone in independence of nuclear technologies in Korea. The independence of technologies is only possible through the enormous investment for research and through the active approaches for various experiments. The performance of various experiments enhanced the risk of environmental pollution and the nuclear fuel irradiation test is one of those experiments. The damage of fuel which might happen any time in irradiation test, will discharge high level radioactive materials from the inside of failed fuel and will gradually contaminate the cooling water in near vicinity. Accordingly, the proper management of coolant having high temperature and high level . (author). refs., tabs., figs.

  10. Development of treatment technology of radio-contaminated coolant in fuel test loop

    International Nuclear Information System (INIS)

    Kim, J. Y.

    1997-10-01

    In 1995, the installation of KMRR located in KAERI provided a milestone in independence of nuclear technologies in Korea. The independence of technologies is only possible through the enormous investment for research and through the active approaches for various experiments. The performance of various experiments enhanced the risk of environmental pollution and the nuclear fuel irradiation test is one of those experiments. The damage of fuel which might happen any time in irradiation test, will discharge high level radioactive materials from the inside of failed fuel and will gradually contaminate the cooling water in near vicinity. Accordingly, the proper management of coolant having high temperature and high level . (author). refs., tabs., figs

  11. Accident analysis in the water loop of the nuclear engineering department of IPEN using the RELAP4 code

    International Nuclear Information System (INIS)

    Fernandes Filho, T.L.

    1980-06-01

    A thermal-hydraulic analysis to describe the transient behavior in the water loop of the Nuclear Engineering Department of the Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo, Brazil, was performed. Postulated accidents such as those resulting from (1) loss of coolant, (2) main pump failure and (3) power excursions, were studied. The computer code RELAP4/Mod.3 was employed as the principal tool of analysis. (Author) [pt

  12. Comparison of thermo-hydraulic analysis with measurements for HELIOS. The scaled integral test loop for PEACER

    International Nuclear Information System (INIS)

    Cho, Jae Hyun; Lim, Jun; Kim, Ji Hak; Hwang, Il Soon

    2009-01-01

    A scaled-down Lead-Bismuth Eutectic circulating integral test loop named as HELIOS (Heavy Eutectic liquid metal Loop for Integral test of Operability and Safety of PEACER) has been employed to characterize steady-state isothermal forced circulation behavior and non-isothermal natural circulation capability of the lead and lead-alloy cooled advanced nuclear energy systems (LACANES). In this time, thermal-hydraulic experiments have been carried out using HELIOS following rigorous calibration campaigns on sensors for temperature and pressure, especially isothermal steady-state forced convection using by the pump. The isothermal steady-state forced convection test was performed to obtain the pressure loss information including friction loss coefficients and form loss coefficients. Then its data were compared with multi-approaching analysis including hand calculation results and computer simulation code results. (MARS-LBE, CFX). We report the results of comparisons between the analysis and measurements together. (author)

  13. PDCI Wide-Area Damping Control: PSLF Simulations of the 2016 Open and Closed Loop Test Plan

    Energy Technology Data Exchange (ETDEWEB)

    Wilches Bernal, Felipe [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Pierre, Brian Joseph [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Elliott, Ryan Thomas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Schoenwald, David A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Byrne, Raymond H. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Neely, Jason C. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Trudnowski, Daniel J. [Montana Tech of the Univ. of Montana, Butte, MT (United States); Donnelly, Matthew K. [Montana Tech of the Univ. of Montana, Butte, MT (United States)

    2017-03-01

    To demonstrate and validate the performance of the wide-are a damping control system, the project plans to conduct closed-loop tests on the PDCI in summer/fall 2016. A test plan details the open and closed loop tests to be conducted on the P DCI using the wide-area damping control system. To ensure the appropriate level of preparedness, simulations were performed in order to predict and evaluate any possible unsafe operations before hardware experiments are attempted. This report contains the result s from these simulations using the power system dynamics software PSLF (Power System Load Flow, trademark of GE). The simulations use the WECC (Western Electricity Coordinating Council) 2016 light summer and heavy summer base cases.

  14. Some loop experiments in the NRX reactor to study the corrosion of mild steel by flowing water at 90oF

    International Nuclear Information System (INIS)

    Allison, G.M.

    1956-11-01

    This work was undertaken to find the water conditions necessary for minimum corrosion in the mild steel thermal shield recirculating systems in NRX and NRU. This report contains the chemical and corrosion results obtained by operating three mild steel loops in which water at 85-95 o F was recirculated through test sections located in J-rod positions in the NRX reactor. Lowest corrosion rates were found when the water was maintained at pH 10.5 with or without oxygen being present. In both cases the corrosion was general in nature and no pitting occurred. At pH 7 with oxygen present in the water severe pitting took place and the corrosion rate was several times higher than similar conditions without oxygen in the water. Under oxygen-free conditions the corrosion product was Fe 3 O 4 . At pH 7 and with 3-5 ppm of O 2 in the water the corrosion product was a mixture of Fe 3 O 4 and γ-Fe 2 O 3 . At high pH with oxygen present Fe 3 O 4 predominated with some traces of Fe 2 O 3 . The systems tested may he listed in order of increasing corrosiveness: High pH with or without O 2 in the water 2 present and continual purification 2 present and no purification or pH control 2 present. (author)

  15. The model of the thermal and hydraulic behaviour of a out-of-pile test loop; Model thermohidraulickog ponasanja vanreaktorskog exksperimentalnog cirkulacionog kola

    Energy Technology Data Exchange (ETDEWEB)

    Vehauc, A; Stosic, Z [Institut za nuklearne nauke Boris Kidric, Voinca, Belgrade (Yugoslavia)

    1988-07-01

    A complex circulation loop was modeled and a simulation program developed for the determination of the pressure, temperature, velocity and flow rate distribution in legs of the loop. The model was used to study the thermal and hydraulic behaviour of an out-of-pile test loop at IBK-ITE. For a given set of conditions in the test section, the model yields data on all the operating modes possible with the existing control system and in consequence on the optimum operating conditions for the loop as a whole. (author)

  16. Evaluation of T-111 forced-convection loop tested with lithium at 13700C

    International Nuclear Information System (INIS)

    DeVan, J.H.; Long, E.L. Jr.

    1975-04-01

    A T-111 alloy (Ta--8 percent W--2 percent Hf) forced-convection loop containing molten lithium was operated 3000 h at a maximum temperature of 1370 0 C. Flow velocities up to 6.3 m/s were used. The results obtained in this forced-convection loop are very similar to those observed in lower velocity thermal-convection loops of T-111 containing lithium. Weight changes were determined at 93 positions around the loop. The maximum dissolution rate occurred at the maximum wall temperature of the loop and was less than 1.3 μ m/year. Mass transfer of hafnium, nitrogen, and, to a lesser extent, carbon occurred from the hotter to cooler regions. Exposed surfaces in the highest temperature region were found to be depleted in hafnium to a depth of 60 μ m with no detectable change in tungsten content. There was some loss in room-temperature tensile strength for specimens exposed to lithium at 1370 0 C, attributable to depletion of hafnium and nitrogen and to attendant grain growth. (U.S.)

  17. Measurement of Li target thickness in the EVEDA Li Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Kanemura, Takuji, E-mail: kanemura.takuji@jaea.go.jp [Japan Atomic Energy Agency, 4002 Narita, O-arai, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan); Kondo, Hiroo; Furukawa, Tomohiro; Hirakawa, Yasushi [Japan Atomic Energy Agency, 4002 Narita, O-arai, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan); Hoashi, Eiji; Yoshihashi, Sachiko; Horiike, Hiroshi [Osaka University, 2-1 Yamada-oka, Suita, Osaka 565-0871 (Japan); Wakai, Eiichi [Japan Atomic Energy Agency, 4002 Narita, O-arai, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan)

    2015-10-15

    Highlights: • The objective is to validate stability of the IFMIF liquid Li target flowing at 15 m/s. • Design requirement of target thickness fluctuation is ±1 mm. • Mean and maximum wave amplitude are 0.26 and 1.46 mm, respectively. • Average thickness can be well predicted with developed analytical model. • Li target was adequately stable and satisfied design requirement. - Abstract: A high-speed (nominal: 15 m/s, range: 10–16 m/s) liquid lithium wall jet is planned to serve as the target for two 40 MeV and 125 mA deuteron beams in the International Fusion Materials Irradiation Facility (IFMIF). The design requirement of target thickness stability is 25 ± 1 mm under a vacuum of 10{sup −3} Pa. This paper presents the results of the target thickness measurement conducted in the EVEDA Li Test Loop under a wide range of conditions including the IFMIF condition (target speed of 10, 15, and 20 m/s; vacuum pressure of 10{sup −3} Pa; and Li temperature of 250 °C). For measurement, we use a laser probe method that we developed in advance; this method generates statistical measurements method using a laser distance meter. The measurement results obtained under the IFMIF nominal condition (15 m/s, 10{sup −3} Pa, 250 °C) at the IFMIF beam center are as follows: average target thickness = 26.08 ± 0.09 mm (2σ), mean wave amplitude = 0.26 ± 0.01 mm (2σ), and maximum wave amplitude = 1.46 ± 0.25 mm (2σ). Of the total wave components, 99.7% are within the design requirement. The analytically predicted target thickness is in excellent agreement with the experimental data, resulting in successful characterization of the Li target thickness.

  18. Measurement of lithium target surface velocity in the IFMIF/EVEDA lithium test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kanemura, Takuji, E-mail: kanemura.takuji@jaea.go.jp [Japan Atomic Energy Agency, 4002 Narita, O-arai, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan); Kondo, Hiroo; Furukawa, Tomohiro; Hirakawa, Yasushi [Japan Atomic Energy Agency, 4002 Narita, O-arai, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan); Hoashi, Eiji [Osaka University, 2-1 Yamada-oka, Suita, Osaka 565-0871 (Japan); Yoshihashi, Sachiko; Horiike, Hiroshi [Fukui University of Technology, Gakuen 3-6-1, Fukui-shi, Fukui 910-8505 (Japan); Wakai, Eiichi [Japan Atomic Energy Agency, 4002 Narita, O-arai, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan)

    2016-11-01

    Highlights: • The objective is to measure the free-surface velocity field of the IFMIF Li target. • The Li target has an important role to remove 10 MW heat input from a deuteron beam. • The free-surface of the Li target is under the most severe heat load condition. • Measured surface velocities are almost equal to cross-sectional average velocities. • It was confirmed that the IFMIF Li target has adequate heat removal performance. - Abstract: In the framework of the Engineering Validation and Engineering Design Activities (EVEDA) project of the International Fusion Materials Irradiation Facility (IFMIF), we measured surface velocity fields of a lithium (Li) target at the EVEDA Li test loop under specifically-designated IFMIF conditions (target speeds of 10, 15, and 20 m/s, vacuum pressure of 10{sup −3} Pa, and Li temperature of 250 °C). In the current design of the IFMIF, the free surface of the Li target is under a most severe heat load condition with respect to Li boiling. The objective of this study is to measure the actual free-surface velocity under these IFMIF conditions to evaluate the heat removal performance of the Li target. The measured results (using the surface-wave tracking method that our team developed) showed two-dimensional time-averaged velocity distributions around the IFMIF beam footprint being virtually uniform, and close to the cross-sectional average velocity. The uniformity of the velocity distributions was less than 1 m/s. The comparison between the measured and analyzed surface velocity at the beam center showed that the analysis accurately predicts the measurement results within a margin of 3%. Finally, it was confirmed that the Li target delivers adequate heat removal performance in the IFMIF as designed.

  19. Manufacturing of In-Pile Test Section(IPS) Mock-up for the 3-Pin Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. M.; Park, K. N.; Chi, D. Y. (and others)

    2005-10-15

    Manufacturing process of IPS Mock-up was initiated in late of 2003 with DAEWOO Precision industries Company. Manufacturing drawings due to detail drawings are composed of Outer assembly and Inner assembly. Welding of IPS Mock-up was performed by the GMAW(Gas Metal Arc Welding) process. After the welding process, non-destructive examination was conducted. Leak test was performed to the Main cooling water part and Neon gas inter-space gap part by the He gas injection with the pressure of 6.0 kg{sub f}/cm{sup 2} and 30 minutes holding time. the result was shown that there was no leak at the Neon gas inter-space gap part but leak was occurred at Main cooling water part according to imperfect screw of purge plug. so, it was re-finished and test was performed to certify the leak tightness. To satisfy the HANARO Limiting Operation Condition, IPS should be tested ahead of installation at the HANARO reactor by the use of test facilities. IPS Mock-up and its test facilities will be designed and used for the test of 'HANARO flow tube pressure drop', 'IPS inner pressure drop' and 'IPS inner vibration'.

  20. Loop kinematics

    International Nuclear Information System (INIS)

    Migdal, A.A.

    1982-01-01

    Basic operators acting in the loop space are introduced. The topology of this space and properties of the Stokes type loop functionals are discussed. The parametrically invariant loop calculus developed here is used in the loop dynamics

  1. Analysis of natural circulation stability in a low pressure thermohydraulic test loop

    International Nuclear Information System (INIS)

    Jafari, J.; D'Auria, F.; Kazeminejad, H.; Davilu, H.

    2002-01-01

    This paper discusses an instability study of a natural circulation (NC) loop performed with the aid of Relap5 thermal-hydraulic system code. This loop has been designed and constructed for the analysis of relevant thermohydraulic parameters of a nuclear reactor. In this study, the main parameters for the stability of NC are identified and characterized through the execution of proper code runs. The obtained stability boundary (SB) in the dimensionless Zuber- Sub-cooling plane is compared with the SB reported in referenced literature. The agreement of predicted NC stability boundaries with the results of independent studies demonstrates both the capability of the mentioned code in assessing NC loop stability and the quality of the performed calculations.(author)

  2. Detection of Toxoplasma gondii oocysts in different water resources by Loop Mediated Isothermal Amplification (LAMP).

    Science.gov (United States)

    Gallas-Lindemann, Carmen; Sotiriadou, Isaia; Mahmoodi, Mohammad Reza; Karanis, Panagiotis

    2013-02-01

    Human toxoplasmosis is potentially contracted due to consumption of contaminated drinking water and represents an increasing public health risk worldwide. Toxoplasma gondii oocysts can be resistant to standard disinfection processes, including UV radiation. Increased awareness of the risk of waterborne toxoplasmosis outbreaks has led to an increase in research interest in the detection of oocysts in environmental water systems. Ninety-five environmental water samples from the Lower Rhine area in Germany have been included in the study and examined for the presence of Toxoplasma. Water samples were filtered or flocculated by aluminum sulfate and purified by sucrose density gradient. DNA was then extracted, and the DNA samples were then examined by LAMP analysis. T. gondii DNA was detected in eight out of 83 (9.6%) influent and effluent samples obtained from wastewater treatment plants. All samples (n=12) from the surface, ground, raw and tap waters tested negative. The purpose of this work was to investigate the occurrence and distribution of Toxoplasma oocysts on the Lower Rhine in Germany. Our study provides evidence that the assay is a sensitive, specific, rapid and cost effective method for the detection of T. gondii and is useful for both the investigations of cases of waterborne outbreaks and for identifying the source of contamination. Copyright © 2012 Elsevier B.V. All rights reserved.

  3. The regeneration test of the secondary loop condensate polishing mixed bed resin in Qinshan NPP

    International Nuclear Information System (INIS)

    Xu Meijing; Dong Liming

    1995-12-01

    There are four condensate polishing mixed beds in the water chemical treatment plant of Qinshan NPP. 2125 kg of D001-TR type cation exchange resin, 2000 kg of D201-TR type anion exchange resin, and 375 kg of S-TR type inert resin are filled into each mixed bed. The bed height of resin is 1.2 m and the volume is about 2.7 m 3 . In order to regenerate the exhausted resin out of the bed, the pre-designed condensate polishing mixed bed regeneration process was used to regenerate the first exhausted resin. After the resin was scrubbed and separated, cation resin and anion resin were respectively regenerated, rinsed to resume the exchange capability of the resin. The regenerated mixed bed is able to keep higher efficiency for condensate polishing. The outlet water quality and the resin service-life are able to meet the design requirements or more favorable than that. During the test, some main cations and anions in the blow-off water at each procedure were analyzed. The analyzed results were used to make pre-designed regeneration process better. The test results proved that pre-designed process is reasonable and effective. (6 refs., 6 figs., 7 tabs.)

  4. Pressure and Temperature of the Room 1 for the Pipe Break Accidents of the 3-Pin Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, H. R

    2005-08-15

    This report deals with the prediction of the pressure and temperature of the room 1 for the pipe break accidents of the 3-pin fuel test loop. The 3-pin fuel test loop is an experimental facility for nuclear fuel tests at the operation conditions similar to those of PWR and CANDU power plants. Because the most processing systems of the 3-pin fuel test loop are placed in the room 1. The structural integrity of the room 1 should be evaluated for the postulated accident conditions. Therefore the pressures and temperatures of the room 1 needed for the structural integrity evaluation have been calculated by using MARS code. The pressures and temperatures of the room 1 have been calculated in various conditions such as the thermal hydraulic operation parameters, the locations of pipe break, and the thermal properties of the room 1 wall. It is assumed that the pipe break accident occurs in the letdown operation without regeneration, because the mass and energy release to the room 1 is expected to be the largest. As a result of the calculations the maximum pressure and temperature are predicted to be 208kPa and 369.2K(96.0 .deg. C) in case the heat transfer is considered in the room 1 wall. However the pressure and temperature are asymptotically 243kPa and 378.1K(104.9 .deg. C) assuming that the heat transfer does not occur in the room 1 wall.

  5. Note: A dual temperature closed loop batch reactor for determining the partitioning of trace gases within CO2-water systems.

    Science.gov (United States)

    Warr, Oliver; Rochelle, Christopher A; Masters, Andrew J; Ballentine, Christopher J

    2016-01-01

    An experimental approach is presented which can be used to determine partitioning of trace gases within CO2-water systems. The key advantages of this system are (1) The system can be isolated with no external exchange, making it ideal for experiments with conservative tracers. (2) Both phases can be sampled concurrently to give an accurate composition at each phase at any given time. (3) Use of a lower temperature flow loop outside of the reactor removes contamination and facilitates sampling. (4) Rapid equilibration at given pressure/temperature conditions is significantly aided by stirring and circulating the water phase using a magnetic stirrer and high-pressure liquid chromatography pump, respectively.

  6. Experimental study of a water thermo-capillary loop; Etude experimentale d`une boucle thermocapillaire a eau

    Energy Technology Data Exchange (ETDEWEB)

    Lefriec, C; Alexandre, A [Ecole Nationale Superieure de Mecanique et d` Aerotechnique (ENSMA), 86 - Poitiers (France)

    1997-12-31

    This paper presents a bench scale experiment of a water thermo-capillary loop which allows to improve the understanding of its functioning mechanisms thanks to the internal visualisation of each component using transparent walls. The advantages of water are its non-toxicity, its high chemical compatibility with several materials and its low functioning pressure. The experimental device is presented and the functioning regimes of each component is analyzed: condenser (flow visualization, influence of tilt), evaporator (quality of heat exchange between teeth and porous medium, bubbles, heat exchange coefficient). (J.S.)

  7. Experimental study of a water thermo-capillary loop; Etude experimentale d`une boucle thermocapillaire a eau

    Energy Technology Data Exchange (ETDEWEB)

    Lefriec, C.; Alexandre, A. [Ecole Nationale Superieure de Mecanique et d`Aerotechnique (ENSMA), 86 - Poitiers (France)

    1996-12-31

    This paper presents a bench scale experiment of a water thermo-capillary loop which allows to improve the understanding of its functioning mechanisms thanks to the internal visualisation of each component using transparent walls. The advantages of water are its non-toxicity, its high chemical compatibility with several materials and its low functioning pressure. The experimental device is presented and the functioning regimes of each component is analyzed: condenser (flow visualization, influence of tilt), evaporator (quality of heat exchange between teeth and porous medium, bubbles, heat exchange coefficient). (J.S.)

  8. Development and Testing of a Temperature-swing Adsorption Compressor for Carbon Dioxide in Closed-loop Air Revitalization Systems

    Science.gov (United States)

    Mulloth, Lila M.; Rosen, Micha; Affleck, David; LeVan, M. Douglas; Wang, Yuan

    2005-01-01

    The air revitalization system of the International Space Station (ISS) operates in an open loop mode and relies on the resupply of oxygen and other consumables from earth for the life support of astronauts. A compressor is required for delivering the carbon dioxide from a removal assembly to a reduction unit to recover oxygen and thereby dosing the air-loop. We have developed a temperature-swing adsorption compressor (TSAC) that is energy efficient, quiet, and has no rapidly moving parts for performing these tasks. The TSAC is a solid-state compressor that has the capability to remove CO2 from a low- pressure source, and subsequently store, compress, and deliver at a higher pressure as required by a processor. The TSAC is an ideal interface device for CO2 removal and reduction units in the air revitalization loop of a spacecraft for oxygen recovery. This paper discusses the design and testing of a TSAC for carbon dioxide that has application in the ISS and future spacecraft for closing the air revitalization loop.

  9. Contribution to perfecting eddy current testing of steam generator tubes of sodium cooled breeders: description of the Monacault loop for the study of sodium deposit influence

    International Nuclear Information System (INIS)

    Lapicore, A.; Lemarquis, J.C.; Oberlin, C.; Pigeon, M.

    1981-12-01

    In the event of sodium-water reaction in the steam generator of a sodium cooled breeder reactor, it is essential to be able to monitor the local loss of thickness of the tubes located in the reaction area. A method for monitoring the tubes by an eddy current probe is being developed for Super Phenix. The sodium deposits on the outer wall of the tubes, as well as their prolonged contact with high temperature sodium are likely to bring about a change in the signals picked up. A test loop, Monacault, has been built in order to clarify the importance of these parameters (effect of sodium deposits, reproducibility of the wetting at different temperatures). It includes three test cells containing the sample tubes having a total of 61 standard defects to be tested. The first results on the wetting of tubes are given and discussed [fr

  10. Water soluble contrast enema examination of the integrity of the rectal anastomosis prior to loop ileostomy reversal may be superfluous.

    Science.gov (United States)

    Larsson, Anna; Lindmark, Gudrun; Syk, Ingvar; Buchwald, Pamela

    2015-03-01

    Defunctioning loop ileostomy in low anterior resection (LAR) is routinely used to reduce consequences of anastomotic leakage. The purpose of this study was to analyze which examination technique is optimal for evaluating the integrity of the anastomosis prior to loop ileostomy reversal. Retrospective analysis of 95 patients who had been subjected to LAR at Helsingborg Hospital and Skåne University Hospital, Sweden, was undertaken between January 2007 and June 2009. The examination techniques of the rectal anastomosis prior to reversal and the clinical outcome after reversal were studied. Radiologic anastomosis control using water soluble contrast enema, digital rectal examination (DRE), and rectoscopy were performed in 53 % (50/95), 98 % (93/95), and 69 % (66/95), respectively. In two patients, no control of the anastomosis was performed before reversal. Fifty-two percent (49/95) of the patients were examined using all techniques. Six patients demonstrated leakage detected before reversal of which two were only radiological leakages. These two patients underwent loop ileostomy reversal after delay without complications. They were the only ones where the three examination techniques did not prove coherence. Four patients had symptomatic leakage; these were detected with rectoscopy and DRE and verified with enema. Three patients developed anastomotic leakage after loop ileostomy reversal despite normal preoperative examinations. Two of these patients had rectovaginal fistulas (AVFs). This retrospective study indicates that contrast enema does not provide additional information if rectoscopy and DRE are normal. Despite negative examinations, three of nine leakages were diagnosed after loop ileostomy reversal. Especially, AVFs seem difficult to diagnose.

  11. Reflooding Experimental On Beta Test Loop : The Characterisation And Preliminary Experiment

    International Nuclear Information System (INIS)

    Khairul, H.; Antariksawan, Anhar R.; Sumamo, Edy; Kiswanta; Giarno; Joko, P.; H, Ismu

    2001-01-01

    The characterisation and preliminary experiment of reflooding had been conducted. The characteristics of main system and component had been identified completely. From these characteristics the experiment condition was set up : heated rod voltage was 20 volt, frequency,of pump was 19 Hz, flow rate was 1 m3/h. The first of experiment did not show the phenomena of rewetting. Possibly because the heated rod temperature was too low. For the second experiment, the voltage of heated rod was increased to 22 Volt and the flow rate was decreased. The result was that the nucleation boiling on the surfaced of heated rod, was observed during the water re flooded the test section

  12. Dynamic Modeling of Process Technologies for Closed-Loop Water Recovery Systems

    Science.gov (United States)

    Allada, Rama Kumar; Lange, Kevin E.; Anderson, Molly S.

    2012-01-01

    Detailed chemical process simulations are a useful tool in designing and optimizing complex systems and architectures for human life support. Dynamic and steady-state models of these systems help contrast the interactions of various operating parameters and hardware designs, which become extremely useful in trade-study analyses. NASA s Exploration Life Support technology development project recently made use of such models to compliment a series of tests on different waste water distillation systems. This paper presents dynamic simulations of chemical process for primary processor technologies including: the Cascade Distillation System (CDS), the Vapor Compression Distillation (VCD) system, the Wiped-Film Rotating Disk (WFRD), and post-distillation water polishing processes such as the Volatiles Removal Assembly (VRA). These dynamic models were developed using the Aspen Custom Modeler (Registered TradeMark) and Aspen Plus(Registered TradeMark) process simulation tools. The results expand upon previous work for water recovery technology models and emphasize dynamic process modeling and results. The paper discusses system design, modeling details, and model results for each technology and presents some comparisons between the model results and available test data. Following these initial comparisons, some general conclusions and forward work are discussed.

  13. Dynamic Modeling of Process Technologies for Closed-Loop Water Recovery Systems

    Science.gov (United States)

    Allada, Rama Kumar; Lange, Kevin; Anderson, Molly

    2011-01-01

    Detailed chemical process simulations are a useful tool in designing and optimizing complex systems and architectures for human life support. Dynamic and steady-state models of these systems help contrast the interactions of various operating parameters and hardware designs, which become extremely useful in trade-study analyses. NASA s Exploration Life Support technology development project recently made use of such models to compliment a series of tests on different waste water distillation systems. This paper presents dynamic simulations of chemical process for primary processor technologies including: the Cascade Distillation System (CDS), the Vapor Compression Distillation (VCD) system, the Wiped-Film Rotating Disk (WFRD), and post-distillation water polishing processes such as the Volatiles Removal Assembly (VRA) that were developed using the Aspen Custom Modeler and Aspen Plus process simulation tools. The results expand upon previous work for water recovery technology models and emphasize dynamic process modeling and results. The paper discusses system design, modeling details, and model results for each technology and presents some comparisons between the model results and available test data. Following these initial comparisons, some general conclusions and forward work are discussed.

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

    Directory of Open Access Journals (Sweden)

    Li Gang

    2016-01-01

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

  15. COBALT: A GN&C Payload for Testing ALHAT Capabilities in Closed-Loop Terrestrial Rocket Flights

    Science.gov (United States)

    Carson, John M., III; Amzajerdian, Farzin; Hines, Glenn D.; O'Neal, Travis V.; Robertson, Edward A.; Seubert, Carl; Trawny, Nikolas

    2016-01-01

    The COBALT (CoOperative Blending of Autonomous Landing Technology) payload is being developed within NASA as a risk reduction activity to mature, integrate and test ALHAT (Autonomous precision Landing and Hazard Avoidance Technology) systems targeted for infusion into near-term robotic and future human space flight missions. The initial COBALT payload instantiation is integrating the third-generation ALHAT Navigation Doppler Lidar (NDL) sensor, for ultra high-precision velocity plus range measurements, with the passive-optical Lander Vision System (LVS) that provides Terrain Relative Navigation (TRN) global-position estimates. The COBALT payload will be integrated onboard a rocket-propulsive terrestrial testbed and will provide precise navigation estimates and guidance planning during two flight test campaigns in 2017 (one open-loop and closed- loop). The NDL is targeting performance capabilities desired for future Mars and Moon Entry, Descent and Landing (EDL). The LVS is already baselined for TRN on the Mars 2020 robotic lander mission. The COBALT platform will provide NASA with a new risk-reduction capability to test integrated EDL Guidance, Navigation and Control (GN&C) components in closed-loop flight demonstrations prior to the actual mission EDL.

  16. Some loop experiments in the NRX reactor to study the corrosion of mild steel by flowing water at 90{sup o}F

    Energy Technology Data Exchange (ETDEWEB)

    Allison, G. M.

    1956-11-15

    This work was undertaken to find the water conditions necessary for minimum corrosion in the mild steel thermal shield recirculating systems in NRX and NRU. This report contains the chemical and corrosion results obtained by operating three mild steel loops in which water at 85-95{sup o}F was recirculated through test sections located in J-rod positions in the NRX reactor. Lowest corrosion rates were found when the water was maintained at pH 10.5 with or without oxygen being present. In both cases the corrosion was general in nature and no pitting occurred. At pH 7 with oxygen present in the water severe pitting took place and the corrosion rate was several times higher than similar conditions without oxygen in the water. Under oxygen-free conditions the corrosion product was Fe{sub 3}O{sub 4}. At pH 7 and with 3-5 ppm of O{sub 2} in the water the corrosion product was a mixture of Fe{sub 3}O{sub 4} and {gamma}-Fe{sub 2}O{sub 3}. At high pH with oxygen present Fe{sub 3}O{sub 4} predominated with some traces of Fe{sub 2}O{sub 3}. The systems tested may he listed in order of increasing corrosiveness: High pH with or without O{sub 2} in the water < water at pH 7 with no O{sub 2} present and continual purification < water with no O{sub 2} present and no purification or pH control < water at pH 7 with 3-5 ppm of O{sub 2} present. (author)

  17. Preoperational test report, raw water system

    Energy Technology Data Exchange (ETDEWEB)

    Clifton, F.T.

    1997-10-29

    This represents the preoperational test report for the Raw Water System, Project W-030. Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks. The system supplies makeup water to the W-030 recirculation evaporative cooling towers for tanks AY1O1, AY102, AZ1O1, AZ102. The Raw Water pipe riser and associated strainer and valving is located in the W-030 diesel generator building. The tests verify correct system operation and correct indications displayed by the central Monitor and Control System.

  18. Preoperational test report, raw water system

    International Nuclear Information System (INIS)

    Clifton, F.T.

    1997-01-01

    This represents the preoperational test report for the Raw Water System, Project W-030. Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks. The system supplies makeup water to the W-030 recirculation evaporative cooling towers for tanks AY1O1, AY102, AZ1O1, AZ102. The Raw Water pipe riser and associated strainer and valving is located in the W-030 diesel generator building. The tests verify correct system operation and correct indications displayed by the central Monitor and Control System

  19. Risk analysis for a radiolysis gas detonation in an in-pile loop with supercritical water

    International Nuclear Information System (INIS)

    Zeiger, T.; Raque, M.; Kuznetsov, M.; Redlinger, R.; Schulenberg, T.

    2012-01-01

    The SCWR (supercritical water reactor) -FQT project is a cooperation between European and Chinese partners aimed to test the fuel SCWR elements under reactor conditions. In the frame of this work the risk of radiolysis gas production in the active range of the test track was assessed. The radiolysis gas could accumulate in an emergency cooling system with stagnating coolant. The ignition of this radiolysis gas could cause pressure peaks that are able to damage the primary coolant circuit. Pressure increase and deformations in case of ignition of accumulated gas were investigated. As piping material the Ti stabilized austenitic steel 08Ch18N10T was assumed, the simulation was performed using the ANSYS code. The results show that pipes without significant wall thickness enhancement cannot withstand the radiolysis gas detonation.

  20. Detailed Modeling of Distillation Technologies for Closed-Loop Water Recovery Systems

    Science.gov (United States)

    Allada, Rama Kumar; Lange, Kevin E.; Anderson, Molly S.

    2011-01-01

    Detailed chemical process simulations are a useful tool in designing and optimizing complex systems and architectures for human life support. Dynamic and steady-state models of these systems help contrast the interactions of various operating parameters and hardware designs, which become extremely useful in trade-study analyses. NASA?s Exploration Life Support technology development project recently made use of such models to compliment a series of tests on different waste water distillation systems. This paper presents efforts to develop chemical process simulations for three technologies: the Cascade Distillation System (CDS), the Vapor Compression Distillation (VCD) system and the Wiped-Film Rotating Disk (WFRD) using the Aspen Custom Modeler and Aspen Plus process simulation tools. The paper discusses system design, modeling details, and modeling results for each technology and presents some comparisons between the model results and recent test data. Following these initial comparisons, some general conclusions and forward work are discussed.

  1. Performance test for a solar water heater

    Science.gov (United States)

    1979-01-01

    Two reports describe procedures and results of performance tests on domestic solar powered hot water system. Performance tests determine amount of energy collected by system, amount of energy delivered to solar source, power required to operate system and maintain proper tank temperature, overall system efficiency, and temperature distribution in tank.

  2. Sensitivity analysis on the interfacial drag in SPACE code to simulate UPTF separate effect test about loop seal clearance phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sukho; Lim, Sanggyu; You, Gukjong; Park, Youngsheop [Korea Hydro and Nuclear Power Company, Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    The nuclear thermal hydraulic system code known as SPACE (Safety and Performance Analysis CodE) was developed and its V and V (Verification and Validation) have been conducted using well-known SETs (Separate Effect Tests) and IETs (Integral Effect Tests). At the same time, the SBLOCA (Small Break Loss of Coolant Accident) methodology in accordance with Appendix K of 10CFR50 for the APR1400 (Advanced Power Reactor 1400) was developed and applied to regulatory body for licensing in 2013. Especially, the SBLOCA methodology developed using SPACE v2.14 code adopts inherent test matrix independent of V and V test to show its conservatism for important phenomena. In this paper, the predictability of SPACE code for UPTF (Upper Plenum Test Facility) test simulating loop seal clearance of SBLOCA important phenomena and the related sensitivity analysis are introduced.

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

    International Nuclear Information System (INIS)

    Matijević, Mario; Pevec, Dubravko; Trontl, Krešimir

    2015-01-01

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

  4. Water electrolysis system refurbishment and testing

    Science.gov (United States)

    Greenough, B. M.

    1972-01-01

    The electrolytic oxygen generator for the back-up water electrolysis system in a 90-day manned test was refurbished, improved and subjected to a 182-day bench test. The performance of the system during the test demonstrated the soundness of the basic electrolysis concept, the high development status of the automatic controls which allowed completely hands-off operation, and the capability for orbital operation. Some design improvements are indicated.

  5. Research of the heat exchanging processes running in the heating and hot water supply loops of the coil heat exchangers

    Directory of Open Access Journals (Sweden)

    Ірина Геннадіївна Шитікова

    2016-11-01

    Full Text Available The fuel-energy complex research has made it possible to disclose a huge power-saving potential in the municipal heat-and-power engineering. Power-and-resource-saving units and systems are becoming extremely urgent because of the power engineering crisis expansion. The self-adjusting heat supply system from the individual heating points with the heat-accumulating units and coil heat exchangers for independent heating and water supply systems has been examined. Coil heat exchangers are used in municipal heating for heat transfer (e.g. geothermal waters for the independent mains of the heating and hot water supply systems. The heat engineering calculation of the heating and accumulating unit with the coil heat exchanger for independent heat supply systems from individual heater was performed and experimental data were received at the experimental industrial unit under the laboratory conditions. The peculiarities of the flows in the intertubular space, their influence on the heat exchange and temperatures of the first and intermediate mains have been shown. It is important to know the processes running inside the apparatus to be able to improve the technical characteristics of the three-loop coil heat exchanger. The task solution will make it possible to save the materials consumption for the three-loop coil heat exchangers in the future

  6. Nonlinear dynamic response analysis in piping system for a loss of coolant accident in primary loop of pressurized water reactor

    International Nuclear Information System (INIS)

    Zhang Xiwen; He Feng; Hao Pengfei; Wang Xuefang

    2000-01-01

    Based on the elaborate force and moment analysis with characteristics method and control-volume integrating method for the piping system of primary loop under pressurized water reactor' loss of coolant accident (LOCA) conditions, the nonlinear dynamic response of this system is calculated by the updated Lagrangian formulation (ADINA code). The piping system and virtual underpinning are specially processed, the move displacement of the broken pipe with time is accurately acquired, which is very important and useful for the design of piping system and virtual underpinning

  7. Modulation of friction dynamics in water by changing the combination of the loop- and graft-type poly(ethylene glycol) surfaces.

    Science.gov (United States)

    Seo, Ji-Hun; Tsutsumi, Yusuke; Kobari, Akinori; Shimojo, Masayuki; Hanawa, Takao; Yui, Nobuhiko

    2015-02-07

    A Velcro-like poly(ethylene glycol) (PEG) interface was prepared in order to control the friction dynamics of material surfaces. Graft- and loop-type PEGs were formed on mirror-polished Ti surfaces using an electrodeposition method with mono- and di-amine functionalized PEGs. The friction dynamics of various combinations of PEG surfaces (i.e., graft-on-graft, loop-on-loop, graft-on-loop, and loop-on-graft) were investigated by friction testing. Here, only the Velcro-like combinations (graft-on-loop and loop-on-graft) exhibited a reversible friction behavior (i.e., resetting the kinetic friction coefficient and the reappearance of the maximum static friction coefficient) during the friction tests. The same tendency was observed when the molecular weights of loop- and graft-type PEGs were tested at 1 k and 10 k, respectively. This indicates that a Velcro-like friction behavior could be induced by simply changing the conformation of PEGs, which suggests a novel concept of altering polymer surfaces for the effective control of friction dynamics.

  8. Breakdown voltage at the electric terminals of GCFR-core flow test loop fuel rod simulators in helium and air

    International Nuclear Information System (INIS)

    Huntley, W.R.; Conley, T.B.

    1979-12-01

    Tests were performed to determine the ac and dc breakdown voltage at the terminal ends of a fuel rod simulator (FRS) in helium and air atmospheres. The tests were performed at low pressures (1 to 2 atm) and at temperatures from 20 to 350 0 C (68 to 660 0 F). The area of concern was the 0.64-mm (0.025-in.) gap between the coaxial conductor of the FRS and the sheaths of the four internal thermocouples as they exit the FRS. The tests were prformed to ensure a sufficient safety margin during Core Flow Test Loop (CFTL) operations that require potentials up to 350 V ac at the FRS terminals. The primary conclusion from the test results is that the CFTL cannot be operated safely if the terminal ends of the FRSs are surrounded by a helium atmosphere but can be operated safely in air

  9. Experimental test of an eco-evolutionary dynamic feedback loop between evolution and population density in the green peach aphid.

    Science.gov (United States)

    Turcotte, Martin M; Reznick, David N; Daniel Hare, J

    2013-05-01

    An eco-evolutionary feedback loop is defined as the reciprocal impacts of ecology on evolutionary dynamics and evolution on ecological dynamics on contemporary timescales. We experimentally tested for an eco-evolutionary feedback loop in the green peach aphid, Myzus persicae, by manipulating initial densities and evolution. We found strong evidence that initial aphid density alters the rate and direction of evolution, as measured by changes in genotype frequencies through time. We also found that evolution of aphids within only 16 days, or approximately three generations, alters the rate of population growth and predicts density compared to nonevolving controls. The impact of evolution on population dynamics also depended on density. In one evolution treatment, evolution accelerated population growth by up to 10.3% at high initial density or reduced it by up to 6.4% at low initial density. The impact of evolution on population growth was as strong as or stronger than that caused by a threefold change in intraspecific density. We found that, taken together, ecological condition, here intraspecific density, alters evolutionary dynamics, which in turn alter concurrent population growth rate (ecological dynamics) in an eco-evolutionary feedback loop. Our results suggest that ignoring evolution in studies predicting population dynamics might lead us to over- or underestimate population density and that we cannot predict the evolutionary outcome within aphid populations without considering population size.

  10. Hardware in the Loop Testing of an Iodine-Fed Hall Thruster

    Science.gov (United States)

    Polzin, Kurt A.; Peeples, Steven R.; Cecil, Jim; Lewis, Brandon L.; Molina Fraticelli, Jose C.; Clark, James P.

    2015-01-01

    chamber (it is under 10(exp -6) torr at -75 C), making it possible to 'cryopump' the propellant with lower-cost recirculating refrigerant-based systems as opposed to using liquid nitrogen or low temperature gaseous helium cryopanels. In the present paper, we describe testing performed using an iodine-fed 200 W Hall thruster mounted to a thrust stand and operated in conjunction with MSFCs Small Projects Rapid Integration and Test Environment (SPRITE) Portable Hardware In the Loop (PHIL) hardware. This work is performed in support of the iodine satellite (iSAT) project, which aims to fly a 200-W iodine-fed thruster on a 12-U CubeSat. The SPRITE PHIL hardware allows a given vehicle to do a checkout of its avionics algorithm by allowing it to monitor and feed data to simulated sensors and effectors in a digital environment. These data are then used to determine the attitude of the vehicle and a separate computer is used to interpret the data set and visualize it using a 3D graphical interface. The PHIL hardware allows the testing of the vehicles bus by providing 'real' hardware interfaces (in the case of this test a real RS422 bus) and specific components can be modeled to show their interactions with the avionics algorithm (e.g. a thruster model). For the iSAT project the PHIL is used to visualize the operating cycle of the thruster and the subsequent effect this thrusting has on the attitude of the satellite over a given period of time. The test is controlled using software running on an Andrews Space Cortex 160 flight computer. This computer is the current baseline for a full iSAT mission. While the test could be conducted with a lab computer and software, the team chose to exercise the propulsion system with a representative CubeSat-class computer. For purposes of this test, the "flight" software monitored the propulsion and PPU systems, controlled operation of the thruster, and provided thruster state data to the PHIL simulation. Commands to operate the thruster were

  11. Water chemistry management during hot functional test

    International Nuclear Information System (INIS)

    Yokoyama, Jiro; Kanda, Tomio; Kagawa, Masaru

    1988-01-01

    To reduce radiation exposure in light water reactor, it is important decrease radioactive corrosion product which is a radiation source. One of the countermeasures is to improve water quality during plant trial operation to form a stable oxide film and to minimize metal release to the coolant at the beginning of commercial operation. This study reviews the optimum water quality conditions to form a chromium rich oxide film during hot functional test (HFT) that is thought to be stable under the PWR condition and reduce the release of Ni that is the source of Co-58, the main radiation source of exposure. (author)

  12. Automation of secondary loop operation in Indus-2 LCW plant

    International Nuclear Information System (INIS)

    Srinivas, L.; Pandey, R.M.; Yadav, R.P.; Gupta, S.; Gandhi, M.L.; Thakurta, A.C.

    2013-01-01

    Indus-2 Low Conductivity Water (LCW) plant has two loops, primary loop and secondary loop. The primary loop mainly supplies LCW to magnets, power supplies and RF systems at constant flow rate. The secondary loop extracts heat from the primary loop through heat exchangers to maintain the supply water temperature of the primary loop around a set value. The supply water temperature of the primary loop is maintained by operating the pumps and cooling towers in the secondary loop. The desired water flow rate in the secondary loop is met by the manual operation of the required number of the pumps. The automatic operation of the pumps and the cooling towers is proposed to replace the existing inefficient manual operation. It improves the operational reliability and ensures the optimum utilization of the pumps and the cooling towers. An algorithm has been developed using LabView programming to achieve optimized operation of the pumps and the cooling towers by incorporating First-In-First-Out (FIFO) logic. It also takes care of safety interlocks, and generates alarms. The program exchanges input and output signals of the plant using existing SCADA system. In this paper, the development of algorithm, its design and testing are elaborated. In the end, the results obtained thereof are discussed. (author)

  13. Results of water chemistry control in the in-pile ''Callisto'' loop (an experimental PWR rig installed in the BR2 reactor)

    International Nuclear Information System (INIS)

    Weber, M.; Benoit, P.; Dekeyser, J.; Verwimp, A.

    1994-01-01

    Since June 1992, a new experimental facility, called CALLISTO, is being irradiated in the BR2 materials testing reactor at Mol, Belgium. The main objective of the present test campaign is to study the behaviour of advanced fuel to high burn-up rates in a realistic PWR environment. Three in-pile sections, containing each 9 fuel rods, are loaded inside the reactor vessel and are connected to a common out-of-pile pressurized water circulation loop (ref.1). The later is branched-off into a purification circuit (feed-bleed concept) and further equipped with safety and auxiliary systems. To cope with the test programme, the equipments are designed so that the guidelines of a PWR primary water chemistry can be followed (ref.2). Real steady-state conditions cannot be observed because the typical BR2 cycle (3 weeks running/3 weeks shut-down) is much shorter and because the rig is cooled down during each reactor shut-down. The purpose of this poster is to provide results of chemical parameters recorded during the cycling behaviour of the CALLISTO primary water. (authors). 4 figs., 1 tab., 2 refs

  14. Special power supply and control system for the gas-cooled fast reactor-core flow test loop

    International Nuclear Information System (INIS)

    Hudson, T.L.

    1981-09-01

    The test bundle in the Gas-Cooled Fast Reactor-Core Flow Test Loop (GCFR-CFTL) requires a source of electrical power that can be controlled accurately and reliably over a wide range of steady-state and transient power levels and skewed power distributions to simulate GCFR operating conditions. Both ac and dc power systems were studied, and only those employing silicon-controlled rectifiers (SCRs) could meet the requirements. This report summarizes the studies, tests, evaluations, and development work leading to the selection. it also presents the design, procurement, testing, and evaluation of the first 500-kVa LMPL supply. The results show that the LMPL can control 60-Hz sine wave power from 200 W to 500 kVA

  15. Spatial-temporal evolution of self-organized loop-patterns on a water surface and a diffuse discharge in the gap

    Science.gov (United States)

    Li, Xuechen; Geng, Jinling; Jia, Pengying; Zhang, Panpan; Zhang, Qi; Li, Yaru

    2017-11-01

    Excited by an alternating current voltage, a patterned discharge and a diffuse discharge are generated in a needle to liquid configuration. Using an intensified charge-coupled device (ICCD), temporal evolution of the discharge between the two electrodes is investigated for the diffuse mode and the patterned mode, respectively. For the diffuse mode, the positive discharge is in a glow regime, and the negative discharge is in a Townsend discharge regime. For the patterned mode, the discharge always belongs to the Townsend discharge regime. Moreover, in the patterned mode, various patterns including the single loop, single loop with the surrounding corona, triple loops, and concentric loops with a central spot are observed on the water surface with the increasing positive peak-value of the applied voltage (Upp). Temporally resolved images of the loop-patterns are captured on the water surface. From the electrical measurements and the ICCD imaging, it is found that the loop pattern emerges after the discharge bridges the two electrodes. Then, it begins to evolve and finally degenerates with the decrease in the discharge current. The pattern does not disappear until the discharge quenches. Formation of the loop-patterns is attributed to the role of negative ions.

  16. Heat Integration of the Water-Gas Shift Reaction System for Carbon Sequestration Ready IGCC Process with Chemical Looping

    Energy Technology Data Exchange (ETDEWEB)

    Juan M. Salazara; Stephen E. Zitney; Urmila M. Diwekara

    2010-01-01

    Integrated gasification combined cycle (IGCC) technology has been considered as an important alternative for efficient power systems that can reduce fuel consumption and CO2 emissions. One of the technological schemes combines water-gas shift reaction and chemical-looping combustion as post gasification techniques in order to produce sequestration-ready CO2 and potentially reduce the size of the gas turbine. However, these schemes have not been energetically integrated and process synthesis techniques can be applied to obtain an optimal flowsheet. This work studies the heat exchange network synthesis (HENS) for the water-gas shift reaction train employing a set of alternative designs provided by Aspen energy analyzer (AEA) and combined in a process superstructure that was simulated in Aspen Plus (AP). This approach allows a rigorous evaluation of the alternative designs and their combinations avoiding all the AEA simplifications (linearized models of heat exchangers). A CAPE-OPEN compliant capability which makes use of a MINLP algorithm for sequential modular simulators was employed to obtain a heat exchange network that provided a cost of energy that was 27% lower than the base case. Highly influential parameters for the pos gasification technologies (i.e. CO/steam ratio, gasifier temperature and pressure) were calculated to obtain the minimum cost of energy while chemical looping parameters (oxidation and reduction temperature) were ensured to be satisfied.

  17. Water NSTF Design, Instrumentation, and Test Planning

    Energy Technology Data Exchange (ETDEWEB)

    Lisowski, Darius D.; Gerardi, Craig D.; Hu, Rui; Kilsdonk, Dennis J.; Bremer, Nathan C.; Lomperski, Stephen W.; Kraus, Adam R.; Bucknor, Matthew D.; Lv, Qiuping; Farmer, Mitchell T.

    2017-08-01

    The following report serves as a formal introduction to the water-based Natural convection Shutdown heat removal Test Facility (NSTF) program at Argonne. Since 2005, this US Department of Energy (DOE) sponsored program has conducted large scale experimental testing to generate high-quality and traceable validation data for guiding design decisions of the Reactor Cavity Cooling System (RCCS) concept for advanced reactor designs. The most recent facility iteration, and focus of this report, is the operation of a 1/2 scale model of a water-RCCS concept. Several features of the NSTF prototype align with the conceptual design that has been publicly released for the AREVA 625 MWt SC-HTGR. The design of the NSTF also retains all aspects common to a fundamental boiling water thermosiphon, and thus is well poised to provide necessary experimental data to advance basic understanding of natural circulation phenomena and contribute to computer code validation. Overall, the NSTF program operates to support the DOE vision of aiding US vendors in design choices of future reactor concepts, advancing the maturity of codes for licensing, and ultimately developing safe and reliable reactor technologies. In this report, the top-level program objectives, testing requirements, and unique considerations for the water cooled test assembly are discussed, and presented in sufficient depth to support defining the program’s overall scope and purpose. A discussion of the proposed 6-year testing program is then introduced, which outlines the specific strategy and testing plan for facility operations. The proposed testing plan has been developed to meet the toplevel objective of conducting high-quality test operations that span across a broad range of single- and two-phase operating conditions. Details of characterization, baseline test cases, accident scenario, and parametric variations are provided, including discussions of later-stage test cases that examine the influence of geometric

  18. Analysis of water hammer-structure interaction in piping system for a loss of coolant accident in primary loop of pressurized water reactor

    International Nuclear Information System (INIS)

    Zhang Xiwen; Yang Jinglong; He Feng; Wang Xuefang

    2000-01-01

    The conventional analysis of water hammer and dynamics response of structure in piping system is divided into two parts, and the interaction between them is neglected. The mechanism of fluid-structure interaction under the double-end break pipe in piping system is analyzed. Using the characteristics method, the numerical simulation of water hammer-structure interaction in piping system is completed based on 14 parameters and 14 partial differential equations of fluid-piping cell. The calculated results for a loss of coolant accident (LOCA) in primary loop of pressurized water reactor show that the waveform and values of pressure and force with time in piping system are different from that of non-interaction between water hammer and structure in piping system, and the former is less than the later

  19. BWR recirculation loop discharge line break LOCA tests with break areas of 50 and 100% assuming HPCS failure at ROSA-III test facility

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Tasaka, Kanji; Yonomoto, Taisuke; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Murata, Hideo; Shiba, Masayoshi; Iriko, Masanori.

    1985-03-01

    This report presents the experimental results of RUN 962 and RUN 963 in ROSA-III program, which are 50 and 100 % break LOCA tests at the BWR recirculation pump discharge line, respectively. The ROSA-III test facility simulates a volumetrically scaled (1/424) BWR system and has four half-length electrically heated fuel bundles, two active recirculation loops, three types of ECCSs and steam and feedwater systems. The experimental data of RUN 962 and RUN 963 were compared with those of RUN 961, a 200 % discharge line break test to study the break area effects on the transient thermal hydraulic phenomena. The least flow areas at the jet pump drive nozzles and recirculation pump discharge nozzle in the broken recirculation loop limitted the discharge flows from the pressure vessel and the depressurization rate in the 100 and 200 % break tests, whereas the least flow area at break nozzle limitted the depressurization rate in the 50 % break test. The highest PCT was observed in the 50 % break test among the three tests. (author)

  20. Large scale sodium-water reaction tests for Monju steam generators

    International Nuclear Information System (INIS)

    Sato, M.; Hiroi, H.; Hori, M.

    1976-01-01

    To demonstrate the safe design of the steam generator system of the prototype fast reactor Monju against the postulated large leak sodium-water reaction, a large scale test facility SWAT-3 was constructed. SWAT-3 is a 1/2.5 scale model of the Monju secondary loop on the basis of the iso-velocity modeling. Two tests have been conducted in SWAT-3 since its construction. The test items using SWAT-3 are discussed, and the description of the facility and the test results are presented

  1. Dynamic Characterization of a Low Cost Microwave Water-Cut Sensor in a Flow Loop

    KAUST Repository

    Karimi, Muhammad Akram; Arsalan, Muhammad; Shamim, Atif

    2017-01-01

    Inline precise measurement of water fraction in oil (i.e. water-cut [WC]) finds numerous applications in oil and gas industry. This paper presents the characterization of an extremely low cost, completely non-intrusive and full range microwave water

  2. PaTAVTT: A Hardware-in-the-Loop Scaled Platform for Testing Autonomous Vehicle Trajectory Tracking

    Directory of Open Access Journals (Sweden)

    Zhigang Xu

    2017-01-01

    Full Text Available With the advent of autonomous vehicles, in particular its adaptability to harsh conditions, the research and development of autonomous vehicles attract significant attention by not only academia but also practitioners. Due to the high risk, high cost, and difficulty to test autonomous vehicles under harsh conditions, the hardware-in-the-loop (HIL scaled platform has been proposed as it is a safe, inexpensive, and effective test method. This platform system consists of scaled autonomous vehicle, scaled roadway, monitoring center, transmission device, positioning device, and computers. This paper uses a case of the development process of tracking control for high-speed U-turn to build the tracking control function. Further, a simplified vehicle dynamics model and a trajectory tracking algorithm have been considered to build the simulation test. The experiment results demonstrate the effectiveness of the HIL scaled platform.

  3. Closing the loop: integrating human impacts on water resources to advanced land surface models

    Science.gov (United States)

    Zaitchik, B. F.; Nie, W.; Rodell, M.; Kumar, S.; Li, B.

    2016-12-01

    Advanced Land Surface Models (LSMs), including those used in the North American Land Data Assimilation System (NLDAS), offer a physically consistent and spatially and temporally complete analysis of the distributed water balance. These models are constrained both by physically-based process representation and by observations ingested as meteorological forcing or as data assimilation updates. As such, they have become important tools for hydrological monitoring and long-term climate analysis. The representation of water management, however, is extremely limited in these models. Recent advances have brought prognostic irrigation routines into models used in NLDAS, while assimilation of Gravity Recovery and Climate Experiment (GRACE) derived estimates of terrestrial water storage anomaly has made it possible to nudge models towards observed states in water storage below the root zone. But with few exceptions these LSMs do not account for the source of irrigation water, leading to a disconnect between the simulated water balance and the observed human impact on water resources. This inconsistency is unacceptable for long-term studies of climate change and human impact on water resources in North America. Here we define the modeling challenge, review instances of models that have begun to account for water withdrawals (e.g., CLM), and present ongoing efforts to improve representation of human impacts on water storage across models through integration of irrigation routines, water withdrawal information, and GRACE Data Assimilation in NLDAS LSMs.

  4. Morphing Wing-Tip Open Loop Controller and its Validation During Wind Tunnel Tests at the IAR-NRC

    Directory of Open Access Journals (Sweden)

    Mohamed Sadok GUEZGUEZ

    2016-09-01

    Full Text Available In this project, a wing tip of a real aircraft was designed and manufactured. This wing tip was composed of a wing and an aileron. The wing was equipped with a composite skin on its upper surface. This skin changed its shape (morphed by use of 4 electrical in-house developed actuators and 32 pressure sensors. These pressure sensors measure the pressures, and further the loads on the wing upper surface. Thus, the upper surface of the wing was morphed using these actuators with the aim to improve the aerodynamic performances of the wing-tip. Two types of ailerons were designed and manufactured: one aileron is rigid (non-morphed and one morphing aileron. This morphing aileron can change its shape also for the aerodynamic performances improvement. The morphing wing-tip internal structure is designed and manufactured, and is presented firstly in the paper. Then, the modern communication and control hardware are presented for the entire morphing wing tip equipped with actuators and sensors having the aim to morph the wing. The calibration procedure of the wing tip is further presented, followed by the open loop controller results obtained during wind tunnel tests. Various methodologies of open loop control are presented in this paper, and results obtained were obtained and validated experimentally through wind tunnel tests.

  5. Reactor loops at Chalk River

    International Nuclear Information System (INIS)

    Sochaski, R.O.

    1962-07-01

    This report describes broadly the nine in-reactor loops, and their components, located in and around the NRX and NRU reactors at Chalk River. First an introduction and general description is given of the loops and their function, supplemented with a table outlining some loop specifications and nine simplified flow sheets, one for each individual loop. The report then proceeds to classify each loop into two categories, the 'main loop circuit' and the 'auxiliary circuit', and descriptions are given of each circuit's components in turn. These components, in part, are comprised of the main loop pumps, the test section, loop heaters, loop coolers, delayed-neutron monitors, surge tank, Dowtherm coolers, loop piping. Here again photographs, drawings and tables are included to provide a clearer understanding of the descriptive literature and to include, in tables, some specifications of the more important components in each loop. (author)

  6. Development of metal fuel and study of construction materials (I-IV), Part V, Vol. II, Project of the device for irradiation of metal uranium in the reactor; 2. Construction of the loop for uranium radiation creep testing

    International Nuclear Information System (INIS)

    Mihajlovic, A.; Pavlovic, A.

    1965-11-01

    This volume includes the design description for construction of the loop for testing uranium radiation creep. It covers the following: construction of the loop head, protection closure; system for pressure regulation and uranium temperature regulation; system for recording samples dilatation and temperature. Testing of components and the loop on the whole is described as well as the safety reports

  7. Rise time reduction of thermal actuators operated in air and water through optimized pre-shaped open-loop driving

    International Nuclear Information System (INIS)

    Larsen, T; Doll, J C; Loizeau, F; Pruitt, B L; Hosseini, N; Fantner, G E; Peng, A W; Ricci, A J

    2017-01-01

    Electrothermal actuators have many advantages compared to other actuators used in micro-electro-mechanical systems (MEMS). They are simple to design, easy to fabricate and provide large displacements at low voltages. Low voltages enable less stringent passivation requirements for operation in liquid. Despite these advantages, thermal actuation is typically limited to a few kHz bandwidth when using step inputs due to its intrinsic thermal time constant. However, the use of pre-shaped input signals offers a route for reducing the rise time of these actuators by orders of magnitude. We started with an electrothermally actuated cantilever having an initial 10–90% rise time of 85 μ s in air and 234 μ s in water for a standard open-loop step input. We experimentally characterized the linearity and frequency response of the cantilever when operated in air and water, allowing us to obtain transfer functions for the two cases. We used these transfer functions, along with functions describing desired reduced rise-time system responses, to numerically simulate the required input signals. Using these pre-shaped input signals, we improved the open-loop 10–90% rise time from 85 μ s to 3 μ s in air and from 234 μ s to 5 μ s in water, an improvement by a factor of 28 and 47, respectively. Using this simple control strategy for MEMS electrothermal actuators makes them an attractive alternative to other high speed micromechanical actuators such as piezoelectric stacks or electrostatic comb structures which are more complex to design, fabricate, or operate. (paper)

  8. Rise Time Reduction of Thermal Actuators Operated in Air and Water through Optimized Pre-Shaped Open-Loop Driving.

    Science.gov (United States)

    Larsen, T; Doll, J C; Loizeau, F; Hosseini, N; Peng, A W; Fantner, G; Ricci, A J; Pruitt, B L

    2017-01-01

    Electrothermal actuators have many advantages compared to other actuators used in Micro-Electro-Mechanical Systems (MEMS). They are simple to design, easy to fabricate and provide large displacements at low voltages. Low voltages enable less stringent passivation requirements for operation in liquid. Despite these advantages, thermal actuation is typically limited to a few kHz bandwidth when using step inputs due to its intrinsic thermal time constant. However, the use of pre-shaped input signals offers a route for reducing the rise time of these actuators by orders of magnitude. We started with an electrothermally actuated cantilever having an initial 10-90% rise time of 85 μs in air and 234 μs in water for a standard open-loop step input. We experimentally characterized the linearity and frequency response of the cantilever when operated in air and water, allowing us to obtain transfer functions for the two cases. We used these transfer functions, along with functions describing desired reduced rise-time system responses, to numerically simulate the required input signals. Using these pre-shaped input signals, we improved the open-loop 10-90% rise time from 85 μs to 3 μs in air and from 234 μs to 5 μs in water, an improvement by a factor of 28 and 47, respectively. Using this simple control strategy for MEMS electrothermal actuators makes them an attractive alternative to other high speed micromechanical actuators such as piezoelectric stacks or electrostatic comb structures which are more complex to design, fabricate, or operate.

  9. The ξ/ξ2nd ratio as a test for Effective Polyakov Loop Actions

    Science.gov (United States)

    Caselle, Michele; Nada, Alessandro

    2018-03-01

    Effective Polyakov line actions are a powerful tool to study the finite temperature behaviour of lattice gauge theories. They are much simpler to simulate than the original (3+1) dimensional LGTs and are affected by a milder sign problem. However it is not clear to which extent they really capture the rich spectrum of the original theories, a feature which is instead of great importance if one aims to address the sign problem. We propose here a simple way to address this issue based on the so called second moment correlation length ξ2nd. The ratio ξ/ξ2nd between the exponential correlation length and the second moment one is equal to 1 if only a single mass is present in the spectrum, and becomes larger and larger as the complexity of the spectrum increases. Since both ξexp and ξ2nd are easy to measure on the lattice, this is an economic and effective way to keep track of the spectrum of the theory. In this respect we show using both numerical simulation and effective string calculations that this ratio increases dramatically as the temperature decreases. This non-trivial behaviour should be reproduced by the Polyakov loop effective action.

  10. Coral mucus fuels the sponge loop in warm- and cold-water coral reef ecosystems

    NARCIS (Netherlands)

    Rix, L.; de Goeij, J.M.; Mueller, C.E.; Struck, U.; Middelburg, J.J.; van Duyl, F.C.; Al-Horani, F.A.; Wild, C.; Naumann, M.S.; Van Oevelen, D.

    2016-01-01

    Shallow warm-water and deep-sea cold-water corals engineer the coral reef framework and fertilize reef communities by releasing coral mucus, a source of reef dissolved organic matter (DOM). By transforming DOM into particulate detritus, sponges play a key role in transferring the energy and

  11. Assurance of Learning, "Closing the Loop": Utilizing a Pre and Post Test for Principles of Finance

    Science.gov (United States)

    Flanegin, Frank; Letterman, Denise; Racic, Stanko; Schimmel, Kurt

    2010-01-01

    Since there is no standard national Pre and Post Test for Principles of Finance, akin to the one for Economics, by authors created one by selecting questions from previously administered examinations. The Cronbach's Alpha of 0.851, exceeding the minimum of 0.70 for reliable pen and paper test, indicates that our Test can detect differences in…

  12. Safety Technology Research Program in the field of pressurized water reactors. 1. Technical report on advancement project RS 36/2. Emergency cooling program service life experiments: reflooding experiments involving the primary loop systems

    International Nuclear Information System (INIS)

    Schweickert, H.; Kremin, H.; Mandl, R.; Riedle, V.; Ruthrof, K.; Sarkar, J.; Schmidt, H.

    The reflooding of the hot reactor core is to be examined for a pressurized water reactor (PWR), using a model of the entire primary loop system. The scale of the model is to be 1:340 in cross-section, with the heights represented full-scale. In addition to the goals of the project, a description of the test facility, including data collection and control equipment is presented. The instrumentation, the planned test program and the test procedure are briefly set forth

  13. Development of criteria for early ascertainment of damage by monitoring of vibration in primary loops of pressurized water reactors

    International Nuclear Information System (INIS)

    Neumeyer, F.; Schramm, K.; Wehling, H.J.; Bauernfeind, V.; Sunder, R.

    1988-11-01

    Nuclear Safety Standards KTA 3201.4 and KTA 3204 prescribe for all light water reactors installed in the Federal Republic of Germany that their vibration behaviour must be measurable in service. For this reason, most PWR plants in the Federal Republic of Germany are equipped with vibration monitoring systems of the KWU-SUeS type or at least with SUeS subsystems. The said safety standards do not prescribe how vibration should be measured. Over the past 20 years, multifarious efforts have been made in the FR Germany to arrive at objective assessment criteria for use in vibration monitoring. Under the terms of reference of this project, conditions were created for the efficient use of the measurement procedures, and general criteria for the evaluation of the results of vibration monitoring of PWR primary loop components were defined under consideration of all experience to date. (orig.) With 20 refs., 6 tabs., 145 figs [de

  14. Use of loop-seals for the control of the overpressures in hydraulic transients evolving in a sea service water system

    Energy Technology Data Exchange (ETDEWEB)

    Canetta, D.; Capozza, A.; Iovino, G.

    1985-01-01

    The transient response following pump trip-offs and start-ups was investigated in the sea water system of a nuclear power plant. Specific care was devoted to water column separation and cavity collapse phenomena. A computer program designed for analysis of complex hydraulic networks was used. It is found that dangerous overpressures can be avoided by the use of loop seals. The design of the vacuum breaker valves of the loop seals and the optimization of overall transient behavior is discussed. 1 reference.

  15. Power ramping test in the JMTR for PCI study of water reactor fuel

    International Nuclear Information System (INIS)

    Nakata, H.; Kanbara, M.; Ichikawa, M.

    1984-01-01

    Power ramping test is essential for PCI study of water reactor fuel. Boiling water capsules have been used for the tests in the JMTR. Heat generation of fuel rod in the capsule can be changed by the He-3 power control facility during reactor operation. Four specially designed fuel rods have been ramped to about 41-43 kW/m; two of them have small gaps filled with iodine, the other two are equipped with centerline temperature thermocouple. Fuel rod elongation detector is equipped to each capsule. For the fuel rods with small gap, unique contraction followed by ordinary fuel relaxation behaviour was observed right after the fast ramping. None of them failed. Future programme includes a series of tests of fuel rods irradiated in the high-pressure water loop at the JMTR and a verification test of remedy fuel which allows daily-load-following operation of BWRs. (author)

  16. Design, Development, and Testing of a UAV Hardware-in-the-Loop Testbed for Aviation and Airspace Prognostics Research

    Science.gov (United States)

    Kulkarni, Chetan; Teubert, Chris; Gorospe, George; Burgett, Drew; Quach, Cuong C.; Hogge, Edward

    2016-01-01

    The airspace is becoming more and more complicated, and will continue to do so in the future with the integration of Unmanned Aerial Vehicles (UAVs), autonomy, spacecraft, other forms of aviation technology into the airspace. The new technology and complexity increases the importance and difficulty of safety assurance. Additionally, testing new technologies on complex aviation systems & systems of systems can be very difficult, expensive, and sometimes unsafe in real life scenarios. Prognostic methodology provides an estimate of the health and risks of a component, vehicle, or airspace and knowledge of how that will change over time. That measure is especially useful in safety determination, mission planning, and maintenance scheduling. The developed testbed will be used to validate prediction algorithms for the real-time safety monitoring of the National Airspace System (NAS) and the prediction of unsafe events. The framework injects flight related anomalies related to ground systems, routing, airport congestion, etc. to test and verify algorithms for NAS safety. In our research work, we develop a live, distributed, hardware-in-the-loop testbed for aviation and airspace prognostics along with exploring further research possibilities to verify and validate future algorithms for NAS safety. The testbed integrates virtual aircraft using the X-Plane simulator and X-PlaneConnect toolbox, UAVs using onboard sensors and cellular communications, and hardware in the loop components. In addition, the testbed includes an additional research framework to support and simplify future research activities. It enables safe, accurate, and inexpensive experimentation and research into airspace and vehicle prognosis that would not have been possible otherwise. This paper describes the design, development, and testing of this system. Software reliability, safety and latency are some of the critical design considerations in development of the testbed. Integration of HITL elements in

  17. System-Level Heat Transfer Analysis, Thermal- Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor. A Preliminary Study

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish [Argonne National Lab. (ANL), Argonne, IL (United States); Soppet, William [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, Saurin [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, Ken [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-01-03

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in April 2015 under the work package for environmentally assisted fatigue under DOE's Light Water Reactor Sustainability program. In this report, updates are discussed related to a system level preliminary finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent thermal-mechanical stress analysis were performed for typical design-basis thermal-mechanical fatigue cycles. The in-air fatigue lives of components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in NUREG-6909. The discussed models and results are very preliminary. Further advancement of the discussed model is required for more accurate life prediction of reactor components. This report only presents the work related to finite element modelling activities. However, in between multiple tensile and fatigue tests were conducted. The related experimental results will be presented in the year-end report.

  18. Advanced Photovoltaic Inverter Control Development and Validation in a Controller-Hardware-in-the-Loop Test Bed

    Energy Technology Data Exchange (ETDEWEB)

    Prabakar, Kumaraguru [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Shirazi, Mariko [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Singh, Akanksha [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Chakraborty, Sudipta [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

    2017-11-07

    Penetration levels of solar photovoltaic (PV) generation on the electric grid have increased in recent years. In the past, most PV installations have not included grid-support functionalities. But today, standards such as the upcoming revisions to IEEE 1547 recommend grid support and anti-islanding functions-including volt-var, frequency-watt, volt-watt, frequency/voltage ride-through, and other inverter functions. These functions allow for the standardized interconnection of distributed energy resources into the grid. This paper develops and tests low-level inverter current control and high-level grid support functions. The controller was developed to integrate advanced inverter functions in a systematic approach, thus avoiding conflict among the different control objectives. The algorithms were then programmed on an off-the-shelf, embedded controller with a dual-core computer processing unit and field-programmable gate array (FPGA). This programmed controller was tested using a controller-hardware-in-the-loop (CHIL) test bed setup using an FPGA-based real-time simulator. The CHIL was run at a time step of 500 ns to accommodate the 20-kHz switching frequency of the developed controller. The details of the advanced control function and CHIL test bed provided here will aide future researchers when designing, implementing, and testing advanced functions of PV inverters.

  19. Ergonomic Investigation On The Layout And Design Of The Main Control Room Of (MCR)Reactor Thermalhydraulic Testing Loop

    International Nuclear Information System (INIS)

    DARLlS; WIDAGDO, SUHARYO

    2000-01-01

    Ergonomics investigation on the layout and design of the reactor thermalhydraulic testing loop main control room has been done. This reason is needed to be done as the primary step for evaluating of operator workload. The operator work load be influence on the operator performance, and finally would influencing the installation operation safety. Generally, the factors that is influencing on operator performance are the layout and design of MCR and its supporting physical environments factors for instance lighting, noising and climatic condition respectively. From investigation had been done, cod be identified that ergonomics point of view not implemented yet on the main control console design, especially on the alarm panel, and also found a little bit brightness problem. Otherwise the temperature and noise room are still in the tolerance boundary

  20. Hardware in the loop testing and evaluation of seaborne search radars

    CSIR Research Space (South Africa)

    Strydom, JJ

    2012-09-01

    Full Text Available for independent testing and evaluation of radar systems. The CSIR digital radio frequency memory (DRFM) hardware technology is used as the basis of these test systems. DRFM's are traditionally used for EW applications, but processing power of field programmable... environment simulation (RES) on digital radio frequency memory (DRFM) platforms can be utilised to test the performance of a search radar in a sea clutter Y ra n ge X r a n g e S h a p e p a r a m e t e r 0 1 2 3 4 x 1 0 4 - 3 - 2 - 1...

  1. Non-Parametric, Closed-Loop Testing of Autonomy in Unmanned Aircraft Systems, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — The proposed Phase I program aims to develop new methods to support safety testing for integration of Unmanned Aircraft Systems into the National Airspace (NAS) with...

  2. Design of in-pile section monitoring system in fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Sim, B. S.; Park, K. N.; Park, S. K.; Chi, D. Y.; Lee, J. M.; Ahn, S. H.; Lee, C. Y.; Kim, Y. J. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    In this paper a brief summary of the monitoring system of In-Pile Section(IPS) is described. To meet the user requirements on the test fuel and irradiation conditions, various instruments are installed on the test fuel pin itself and the appropriate locations in IPS. The requirements and descriptions for instruments, gas supply system, and the data acquisition system to sample and record parameters are described.

  3. The UK MT-3 ballooning test in the Battelle NRU loop

    International Nuclear Information System (INIS)

    Gibson, I.H.; Coddington, P.; Healey, T.; Mann, C.A.

    1982-04-01

    The ballooning response of PWR fuel rods under postulated LOCA conditions has been the subject of intensive investigations. The UK MT-3 ballooning test was a co-operative international exercise which was technically very successful. This report describes the pre-test planning and calculations carried out in the UK and then describes in some detail the thermal hydraulic and clad deformation results. (U.K.)

  4. Elaboration of a guideline for counterpart testing of integral loop systems

    International Nuclear Information System (INIS)

    Karwat, H.

    1988-01-01

    A previous study on the problems of scaling and extrapolation of experimental results indicated the necessity to overcome some uncertainty in the achievable code accuracy when applied to full size reactor systems. The empirical approach in simulating interfacial relationships within systems behaviour codes in combination with the liberty of nodalisation requires a careful assessment of the overall validity of the code verification concept. Several new test facilities are presently under construction to study fluid-dynamic effects of interest for the behaviour of PWRs under small break LOCA conditions in a larger dimension and with somewhat modified scaling considerations. The test results expected from these new facilities may be useful to assess the overall validity of the analytical simulation concept of the applied computer codes. Test facilities of interest within the European Community are the - LOBI Test Facility (vol. scaling 1/700) at JRC-Ispra - SPES Test Facility (vol. scaling 1/425) in Italy - BETHSY Facility (vol scaling 1/100) in France and outside Europe the - ROSA IV Facility (vol. scaling 1/50) in Japan. Main objective of the study was the elaboration of a guideline for the particular requirements of meaningful counterpart testing with the final aim to qualify the overall validity of code simulation concepts and the expected predictive accuracy of code applications for full size prototype PWRs

  5. Large lithium loop experience

    International Nuclear Information System (INIS)

    Kolowith, R.; Owen, T.J.; Berg, J.D.; Atwood, J.M.

    1981-10-01

    An engineering design and operating experience of a large, isothermal, lithium-coolant test loop are presented. This liquid metal coolant loop is called the Experimental Lithium System (ELS) and has operated safely and reliably for over 6500 hours through September 1981. The loop is used for full-scale testing of components for the Fusion Materials Irradiation Test (FMIT) Facility. Main system parameters include coolant temperatures to 430 0 C and flow to 0.038 m 3 /s (600 gal/min). Performance of the main pump, vacuum system, and control system is discussed. Unique test capabilities of the ELS are also discussed

  6. Behavior of copper corrosion products in water loops of heat-exchange units

    International Nuclear Information System (INIS)

    Zarembo, V.I.; Kritskii, V.G.; Slobodov, A.A.; Puchkov, L.V.

    1989-01-01

    This communication is dedicated to an examination of copper corrosion products (CP) in the conditions of real aqueous-chemical regime (ACR) parameters. The deposition of these CP in steam-generating zones (up to 85% of their total amount) stimulate local types of corrosion. The solubility in Cu CP (Cu 2 O, CuO, Cu(OH) 2 )-water (H 2 O)-gas (H 2 , O 2 )-conditioning additives (HCl, KOH) systems was determined by computer modeling according to the minimum Gibbs energy criterion on the basis of selected and matched thermodynamic constants for various chemical forms of copper under standard conditions. As a result of the authors' calculations they obtained the solubilities in water of CuO, Cu 2 O and Cu(OH) 2 when changing the dosage of active gases from 0 to 10 -2 mole/kg of water, of acid or equal to that of saturated vapor of pure water. Thus, they were able to monitor the behavior of copper CP in conditions modeling those of real ACR in operating heat exchange units, including in conditions deviating from the standard

  7. Power-hardware-in-the-loop test of VSC-HVDC connection for off-shore wind power plants

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, Ranjan [Siemens Wind Power A/S, Brande (Denmark); Technical Univ. of Denmark (Denmark). Center for Electric Technology; Cha, Seung T.; Wu, Qiuwei; Rasmussen, Tonny W.; Oestergaard, Jacob [Technical Univ. of Denmark (Denmark). Center for Electric Technology; Jensen, Kim H. [Siemens Wind Power A/S, Brande (Denmark)

    2011-07-01

    This paper present a power-hardware-in-the-loop (PHIL) test of an off-shore wind power plant (WPP) interconnected to the on-shore grid via a VSC-HVDC connection. The intention of the PHIL test is to verify the hardware interaction and the control co-ordination between the plant side VSC of the HVDC system and the wind turbines within the WPP in order to ensure smooth operation of the WPP under both normal and fault operating condition. The PHIL test platform is comprised of a real time digital simulator (RTDS), a Spitzenberger Spies three phase 7,5 kW power amplifier, a purpose built VSC and a DC chopper. The WPP is simulated in the RTDS as a single full-scale wind turbine. The simulated WPP interacts with the WPP side VSC through the power amplifier. The interface between the RTDS and the power amplifier is done via an analogue GTAO I/O card of the RTDS and the input channel of the amplifier. The amplifier scales up the voltages at the point of connection of the WPP in the RTDS to the voltage level for the WPP side VSC. The WPP side VSC converter is equipped with a DC chopper. The test results show the successful control coordination between the WPP and the plant side VSC converter of the HVDC connection of the WPP. (orig.)

  8. Water deprivation test in children with polyuria.

    Science.gov (United States)

    Wong, Lap Ming; Man, Sze Shun

    2012-01-01

    Polyuria is an uncommon clinical presentation in paediatric practice. When diabetes mellitus has been excluded by history taking and preliminary investigations and impaired renal concentrating ability is confirmed, water deprivation test (WDT) is necessary to differentiate among central diabetes insipidus (CDI), nephrogenic diabetes insipidus, or primary polydipsia. Traditionally, responsiveness to desmopressin injection is defined as urine osmolality >750 mOsm/kg. However, that level could not be reached in the review of our patients. We discuss how to arrive at the diagnosis of CDI in WDT. An approach to polyuria and WDT will also be discussed.

  9. Application of reliability techniques to prioritize BWR [boiling water reactor] recirculation loop welds for in-service inspection

    International Nuclear Information System (INIS)

    Holman, G.S.

    1989-12-01

    In January 1988 the US Nuclear Regulatory Commission issued Generic Letter 88-01 together with NUREG-0313, Revision 2, ''Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping,'' to implement NRC long-range plans for addressing the problem of stress corrosion cracking in boiling water reactor piping. NUREG-0313 presents guidelines for categorizing BWR pipe welds according to their SCC condition (e.g., presence of known cracks, implementation of measures for mitigating SCC) as well as recommended inspection schedules (e.g., percentage of welds inspected, inspection frequency) for each weld category. NUREG-0313 does not, however, specify individual welds to be inspected. To address this issue, the Lawrence Livermore National Laboratory developed two recommended inspection samples for welds in a typical BWR recirculation loop. Using a probabilistic fracture mechanics model, LLNL prioritized loop welds on the basis of estimated leak probabilities. The results of this evaluation indicate that riser welds and bypass welds should be given priority attention over other welds. Larger-diameter welds as a group can be considered of secondary importance compared to riser and bypass welds. A ''blind'' comparison between the probability-based inspection samples and data from actual field inspections indicated that the probabilistic analysis generally captured the welds which the field inspections identified as warranting repair or replacement. Discrepancies between the field data and the analytic results can likely be attributed to simplifying assumptions made in the analysis. The overall agreement between analysis and field experience suggests that reliability techniques -- when combined with historical experience -- represent a sound technical basis on which to define meaningful weld inspection programs. 13 refs., 8 figs., 5 tabs

  10. Water testing of the FMIT lithium target

    International Nuclear Information System (INIS)

    Hassberger, J.A.; Ingham, J.G.

    1981-11-01

    Results of water tests of the lithium target design for the Fusion Materials Irradiation Test (FMIT) Facility demonstrate hydraulic features essential for acceptable target operation and confirm predictions of the target performance. This high speed, free surface, curved-wall jet has been shown to generate a stable surface shape and to provide the high velocities and pressures within the fluid needed to remove the 3.5 MW of power generated within the jet during FMIT operation. Measurements of the jet performance are found to fall within limits bounded by one- and two-dimensional potential flow predictions. This agreement between measured and predicted performance provides for a significant level of confidence in the ability of the FMIT lithium target to meet its design and functional objectives

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

    International Nuclear Information System (INIS)

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

    2012-09-01

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

  12. Multi-loop social learning for sustainable land and water governance: Towards a research agenda on the potential of virtual learning platforms

    NARCIS (Netherlands)

    Medema, W.J.; Wals, A.E.J.; Adamowski, J.

    2014-01-01

    Managing social-ecological systems and human well being in a sustainable way requires knowledge of these systems in their full complexity. Multi-loop social learning is recognized as a crucial element to sustainable decision-making for land and water resources management involving a process of

  13. Parametric tests of the effects of water chemistry impurities on corrosion of Zr-alloys under simulated BWR condition

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, S; Ito, K [Nippon Nuclear Fuel Development Co. Ltd., Oarai, Ibaraki (Japan); Lin, C C [GE Nucklear Energy (United States); Cheng, B [Electric Power Research Inst. (United States); Ikeda, T [Toshiba Corp. (Japan); Oguma, M [Hitachi, Ltd (Japan); Takei, T [Tokyo Electric Power Co., Inc. (Japan); Vitanza, C; Karlsen, T M [Institutt for Energiteknikk, Halden (Norway). OECD Halden Reaktor Projekt

    1997-02-01

    The Halden BWR corrosion test loop was constructed to evaluate the impact of water chemistry variables, heat flux and boiling condition on corrosion performance of Zr-alloys in a simulated BWR environment. The loop consists of two in-core rigs, one for testing fuel rod segments and the other for evaluating water chemistry variables utilizing four miniautoclaves. Ten coupon specimens are enclosed in each miniautoclave. The Zr-alloys for the test include Zircaloy-2 having different nodular corrosion resistance and five new alloys. The first and second of the six irradiation tests planned in this program were completed. Post-irradiation examination of those test specimens have shown that the test loop is capable of producing nodular corrosion on the fuel rod cladding tested under the reference chemistry condition. The miniautoclave tests showed that nodular corrosion could be formed without flux and boiling under some water chemistry conditions and the new alloys, generally, had higher corrosion resistance than the Zircaloy in high oxygen environments. (author). 5 refs, 4 figs, 5 tabs.

  14. Independent modification on water lubrication loop of radial-axial bearing of Russian reactor coolant pump

    International Nuclear Information System (INIS)

    Gu Yingbin

    2012-01-01

    Water lubrication was used for radial-axial bearings of 1391M reactor coolant pumps at both units of Tianwan Nuclear Power Plant Phase I Project, which was the first trial on large commercial pressurized water reactors in the world. As a prototype, there were inherent deficiencies leading to a series of operational events. Jiangsu Nuclear Power Corporation conducted the independent innovative technical modification to cope with the defects, and succeeded in reducing heat removal rate of the radial-axial bearings of the reactor coolant pumps, mitigating or preventing the cavitation abrasion of the bearings and improving the cooling effects. This paper illustrates the reasons of the innovative modification, the design and implementation preparation of modification program, the implementation process and evaluation of modification effect, including detailed follow-up work program. (author)

  15. The CABRI facility: Implementation of a pressurized water loop and related safety review

    International Nuclear Information System (INIS)

    Cabrillat, J.C.; Maegey, M.; Bourguignon, D.; Miachon, G.; Forestier, F.; Coulon, J.P.; Faury, M.

    2003-01-01

    The CABRI reactor operated by CEA at the CADARACHE Nuclear Center in France provided the condition for safety studies on nuclear fuel. Initially designed to support investigations on Fast Reactor fuel, large modifications are underway to provide representative conditions for studies on Light Water Fuel types. A general overview of these modifications, the related safety review and supporting studies are described in the paper. (author)

  16. Thermodynamic and experimental study on heat transfer mechanism of miniature loop heat pipe with water-copper nanofluid

    Science.gov (United States)

    Wang, Xiao-wu; Wan, Zhen-ping; Tang, Yong

    2018-02-01

    A miniature loop heat pipe (mLHP) is a promising device for heat dissipation of electronic products. Experimental study of heat transfer performance of an mLHP employing Cu-water nanofluid as working fluid was conducted. It is found that, when input power is above 25 W, the temperature differences between the evaporator wall and vapor of nanofluid, Te - Tv, and the total heat resistance of mLHP using nanofluid are always lower than those of mLHP using de-ionized water. The values of Te - Tv and total heat resistance of mLHP using nanofluid with concentration 1.5 wt. % are the lowest, while when the input power is 25 W, the values of Te - Tv and total heat resistance of mLHP using de-ionized water are even lower than those of mLHP using nanofluid with concentration 2.0 wt. %. At larger input power, the dominant interaction is collision between small bubbles and nanoparticles which can facilitate heat transfer. While at lower input power, nanoparticles adhere to the surface of large bubble. This does not benefit boiling heat transfer. For mLHP using nanofluid with larger concentration, for example 2.0%, the heat transfer may even be worse compared with using de-ionized water at lower input power. The special structure of the mLHP in this study, two separated chambers in the evaporator, produces an extra pressure difference and contributes to the heat transfer performance of the mLHP.

  17. Hardware-in-the-Loop Test for Automatic Voltage Regulator of Synchronous Condenser

    DEFF Research Database (Denmark)

    Nguyen, Ha Thi; Yang, Guangya; Nielsen, Arne Hejde

    2018-01-01

    Automatic voltage regulator (AVR) plays an important role in volt/var control of synchronous condenser (SC) in power systems. Test AVR performance in steady-state and dynamic conditions in real grid is expensive, low efficiency, and hard to achieve. To address this issue, we implement hardware...

  18. Real-Time Hardware-in-the-Loop Testing for Digital Controllers

    DEFF Research Database (Denmark)

    Cha, Seung-Tae; Kwon, Park In; Wu, Qiuwei

    2012-01-01

    of the power electronics hardware are not included in the RTDS. Instead, the control algorithms are coded using the native C code and downloaded to the dedicated digital signal processor (DSP)/microcontrollers. The two experimental applications illustrate the effectiveness of the HIL controller testing...

  19. Experimental loop for SH2 (LECS)

    International Nuclear Information System (INIS)

    Strehar, N.R.; Bruzzoni, Pablo; Moras, J.J.; Cogozzo, E.O.

    1981-01-01

    An experimental loop is described for circulation of SH 2 that operates at 2 x 10 6 Pascal and 33 deg C. It was designed and constructed with the purpose of experimentally studying the hydraulic instability phenomenon that can be detected in cold isotopic exchange columns in the Girdler-Sulfide (GS) process of heavy water production. The main features of the different components of the loop are described, as well as the materials, the measurement and control instruments and the auxiliary equipment used, and finally the measuring methods to qualify and quantify the formation of froth. Furthermore, the loop's transportable metallic container is described, which allows to transport and connect it to CNEA's experimental heavy water plant or to any other heavy water plants using the GS method. Some tests made with inert gases that intended to verify the equipment's performance and to select the most adequate sieve trays for its operation are discussed. (M.E.L.) [es

  20. Qualification program for JHR fuel elements: Irradiation of the first JHR test assembly in the BR2-Evita loop

    International Nuclear Information System (INIS)

    Anselmet, M.-C.; Lemoine, P.; Koonen, E.; Benoit, P.; Gouat, P.; Claes, W.; Geens, F.; Miras, G.; Brisson, S.

    2010-01-01

    An experimental program has been designed by CEA to qualify the behaviour of the JHR fuel under conditions representative of the reactor operating ones. This program uses the SCK.CEN facilities, irradiating JHR lead test elements in the BR2 reactor, inside its central channel which has been particularly arranged for this objective (Evita loop). As a first step in the program, a two cycle irradiation (4 weeks by cycle) started mid-July 2009 and ended mid-November (EVITA-1). After a cooling phase, this first JHR lead test element will be submitted to post-irradiation examination. The second JHR test element began its irradiation in the first quarter of 2010; its unloading is planned before the end of 2010, after 5 cycles in the BR2 reactor. The results of these two experiments are expected as input information for the Safety Authority Report. This paper presents the qualification program with the objectives assigned to each phase (irradiation, examination). A first interpretation of the irradiation data for the first element is presented, so as the information available on the progress of the following phases of the programme. (author)

  1. The EL-4 reactor. Changing of a pressure tube on a test loop

    International Nuclear Information System (INIS)

    Foulquier, H.; Clara, P.

    1964-01-01

    Right from the beginning of the EL-4 project, the research convected with the overall design of the reactor was guided by the various technical specifications resulting from a justifiable concern about the reliability. The external and internal tubes of each layer situated in the reactor block had in particular to be interchangeable. The research alone into the dismantling of the external tube, i.e in fact the pressure tube, justified a certain number of full-scale tests on a model. The tests carried out under relevant conditions on a non-irradiated structure made it possible to define a complete ranger of of positioning and un-positioning sequences at a distance for such a pressure tube. (authors) [fr

  2. Studies of the steam generator degraded tubes behavior on BRUTUS test loop

    Energy Technology Data Exchange (ETDEWEB)

    Chedeau, C.; Rassineux, B. [EDF/DER/MTC, Moret Sur Loing (France); Flesch, B. [EDF/EPN/DMAINT, Paris (France)] [and others

    1997-04-01

    Studies for the evaluation of steam generator tube bundle cracks in PWR power plants are described. Global tests of crack leak rates and numerical calculations of crack opening area are discussed in some detail. A brief overview of thermohydraulic studies and the development of a mechanical probabilistic design code is also given. The COMPROMIS computer code was used in the studies to quantify the influence of in-service inspections and maintenance work on the risk of a steam generator tube rupture.

  3. Functional Assessment of Battery Management System Tested on Hardware-in-the-Loop Simulator

    DEFF Research Database (Denmark)

    Kalogiannis, Theodoros; Stroe, Daniel-Ioan; Swierczynski, Maciej Jozef

    2017-01-01

    . The BMS under test is interacting in real-time with the emulated battery pack and several of its functions such as current, voltage and State of Charge (SOC) estimation are evaluated. Also, passive balancing experiments are conducted during charging in order to assess different balancing settings...... of the BMS as far as their respective balancing times and deviations in SOC are concerned. Finally, the energy efficiency of the BMS is accurately measured which, depending on the requirements, can achieve more than 99% of efficiency. In general, it is demonstrated that the HIL setup can provide...... representative outcomes at minimum resources and time requirements....

  4. Loop-mediated isothermal amplification (LAMP) test for specific and rapid detection of Brucella abortus in cattle.

    Science.gov (United States)

    Karthik, K; Rathore, Rajesh; Thomas, Prasad; Arun, T R; Viswas, K N; Agarwal, R K; Manjunathachar, H V; Dhama, Kuldeep

    2014-01-01

    Brucella abortus, the major causative agent of abortion in cattle and a zoonotic pathogen, needs to be diagnosed at an early stage. Loop-mediated isothermal amplification (LAMP) test is easy to perform and also promising to be adapted at field level. To develop a LAMP assay for specific and rapid detection of B. abortus from clinical samples of cattle. LAMP primers were designed targeting BruAb2_0168 region using specific software tool and LAMP was optimized. The developed LAMP was tested for its specificity with 3 Brucella spp. and 11 other non-Brucella spp. Sensitivity of the developed LAMP was also carried out with known quantity of DNA. Cattle whole blood samples and aborted fetal stomach contents were collected and used for testing with developed LAMP assay and results were compared with polymerase chain reaction (PCR). The developed LAMP assay works at 61 °C for 60 min and the detection limit was observed to be 100-fold more than the conventional PCR that is commonly used for diagnosis of B. abortus. Clinical sensitivity and specificity of the developed LAMP assay was 100% when compared with Rose Bengal plate test and standard tube agglutination test. SYB® green dye I was used to visualize the result with naked eye. The novelty of the developed LAMP assay for specifically detecting B. abortus infection in cattle along with its inherent rapidness and high sensitivity can be employed for detecting this economically important pathogen of cattle at field level as well be exploited for screening of human infections.

  5. An original valveless artificial heart providing pulsatile flow tested in mock circulatory loops.

    Science.gov (United States)

    Tozzi, Piergiorgio; Maertens, Audrey; Emery, Jonathan; Joseph, Samuel; Kirsch, Matthias; Avellan, François

    2017-11-24

    We present the test bench results of a valveless total artificial heart that is potentially compatible with the pediatric population. The RollingHeart is a valveless volumetric pump generating pulsatile flow. It consists of a single spherical cavity divided into 4 chambers by 2 rotating disks. The combined rotations of both disks produce changes in the volumes of the 4 cavities (suction and ejection). The blood enters/exits the spherical cavity through 4 openings that are symmetrical to the fixed rotation axis of the first disk.Mock circulatory system: The device pumps a 37% glycerin solution through 2 parallel circuits, simulating the pulmonary and systemic circulations. Flow rates are acquired with a magnetic inductive flowmeter, while pressure sensors collect pressure in the left and right outflow and inflow tracts.In vitro test protocol: The pump is run at speeds ranging from 20 to 180 ejections per minute. The waveform of the pressure generated at the inflow and outflow of the 4 chambers and the flow rate in the systemic circulation are measured. At an ejection rate of 178 min-1, the RollingHeart pumps 5.3 L/min for a systemic maximal pressure gradient of 174 mmHg and a pulmonary maximal pressure gradient of 75 mmHg. The power input was 14 W, corresponding to an efficiency of 21%. The RollingHeart represents a new approach in the domain of total artificial heart. This preliminary study endorses the feasibility of a single valveless device acting as a total artificial heart.

  6. Test results on reuse of reclaimed shower water - A summary

    Science.gov (United States)

    Verostko, Charles E.; Garcia, Rafael; Sauer, Richard; Reysa, Richard P.; Linton, Arthur T.

    1989-01-01

    Results are presented from tests to evaluate a microgravity whole body shower and waste water recovery system design for possible use on the Space Station. Several water recovery methods were tested, including phase change distillation, a thermoelectric hollow fiber membrane evaporation subsystem, and a reverse osmosis dynamic membrane system. Consideration is given to the test hardware, the types of soaps evaluated, the human response to showering with reclaimed water, chemical treatment for microbial control, the procedures for providing hygienic water, and the quality of water produced by the systems. All three of the waste water recovery systems tested successfully produced reclaimed water for reuse.

  7. Thermal hydraulic considerations and mock-up tests for developing two-phase thermo-siphon loop of CARR-CNS

    International Nuclear Information System (INIS)

    Shejiao, Du; Qincheng, Bi; Tingkuan, Chen; Quanke, Feng

    2005-01-01

    The main component of the China Advanced Research Reactor Cold Neutron Source (CARR-CNS), which is under design, is a two-phase thermo-siphon loop of hydrogen. It consists of a condenser, a single tube with counter current flow avoiding flooding and a cylindrical-annulus moderator cell. The mockup tests were carried out using a full-scale loop with Freon-113, to validate the self-regulating characteristics of the loop, void fraction less than 20% in the liquid of the moderator cell and the requirements for establishing the condition under which the inner shell of the moderator cell has only vapor and the outer shell liquid. In the case of these mockup tests the density ratio of liquid to vapor and the volumetric vapor evaporation rate due to heat load are kept the same as those in normal operation of the CARR-CNS. The results show that the loop has the self-regulating characteristics and the inner shell of the moderator cell contains only vapor, the outer shell liquid. The average void fraction of the moderator cell was verified less than 20% under the volumetric vapor generation of 0.65 l/s corresponding to the nuclear heating of 800 W in the case of the liquid hydrogen. The local void fraction in the liquid hydrogen increases with the increase of the loop pressure under the condition of a constant volumetric evaporation

  8. High-Temperature Structural Analysis of a Small-Scale PHE Prototype under the Test Condition of a Small-Scale Gas Loop

    International Nuclear Information System (INIS)

    Song, K.; Hong, S.; Park, H.

    2012-01-01

    A process heat exchanger (PHE) is a key component for transferring the high-temperature heat generated from a very high-temperature reactor (VHTR) to a chemical reaction for the massive production of hydrogen. The Korea Atomic Energy Research Institute has designed and assembled a small-scale nitrogen gas loop for a performance test on VHTR components and has manufactured a small-scale PHE prototype made of Hastelloy-X alloy. A performance test on the PHE prototype is underway in the gas loop, where different kinds of pipelines connecting to the PHE prototype are tested for reducing the thermal stress under the expansion of the PHE prototype. In this study, to evaluate the high-temperature structural integrity of the PHE prototype under the test condition of the gas loop, a realistic and effective boundary condition imposing the stiffness of the pipelines connected to the PHE prototype was suggested. An equivalent spring stiffness to reduce the thermal stress under the expansion of the PHE prototype was computed from the bending deformation and expansion of the pipelines connected to the PHE. A structural analysis on the PHE prototype was also carried out by imposing the suggested boundary condition. As a result of the analysis, the structural integrity of the PHE prototype seems to be maintained under the test condition of the gas loop.

  9. Remote mechanized equipment for the repair and replacement of boiling water reactor recirculation loop piping

    International Nuclear Information System (INIS)

    Mauser, D.; Busch, D.F.

    1983-01-01

    Equipment has been assembled for the remote repair or replacement of boiling water reactor nuclear plant piping in the diameter range of 4 to 28 inches (10-71 cm). The objectives of this program were to produce high-quality pipe welds, reduce plant downtime, and reduce man-rem exposure. The repair strategy was to permit repair personnel to install and check out the repair subsystems and then leave the radiation zone allowing the operations to be conducted at a distance of up to 300 feet (91 m) from the operator. The complete repair system comprises subsystems for pipe severing, dimensional gaging, joint preparation, counterboring, welding, postweld nondestructive inspection (conceptual design), and audio, electronic, and visual monitoring of all operations. Components for all subsystems, excluding those for postweld nondestructive inspection, were purchased and modified as needed for integration into the repair system. Subsystems were designed for two sizes of Type 304 stainless steelpipe. For smaller, 12-inch-diameter (30.5 cm) pipe, severing is accomplished by a power hack saw and joint preparation and counterboring by an internally mounted lathe. The 22-inch-diameter (56 cm) pipe is severed, prepared, and counterbored using an externally mounted, single-point machining device. Dimensional gaging is performed to characterize the pipe geometry relative to a fixed external reference surface, allowing the placement of the joint preparation and the counterbore to be optimized. For both pipe sizes, a track-mounted gas tungsten-arc welding head with filler wire feed is used

  10. Putting humans in the loop: Using crowdsourced snow information to inform water management

    Science.gov (United States)

    Fedorov, Roman; Giuliani, Matteo; Castelletti, Andrea; Fraternali, Piero

    2016-04-01

    The unprecedented availability of user generated data on the Web due to the advent of online services, social networks, and crowdsourcing, is opening new opportunities for enhancing real-time monitoring and modeling of environmental systems based on data that are public, low-cost, and spatio-temporally dense, possibly contributing to our ability of making better decisions. In this work, we contribute a novel crowdsourcing procedure for computing virtual snow indexes from public web images, either produced by users or generated by touristic webcams, which is based on a complex architecture designed for automatically crawling content from multiple web data sources. The procedure retains only geo-tagged images containing a mountain skyline, identifies the visible peaks in each image using a public online digital terrain model, and classifies the mountain image pixels as snow or no-snow. This operation yields a snow mask per image, from which it is possible to extract time series of virtual snow indexes representing a proxy of the snow covered area. The value of the obtained virtual snow indexes is estimated in a real world water management problem. We consider the snow-dominated catchment of Lake Como, a regulated lake in Northern Italy, where snowmelt represents the most important contribution to seasonal lake storage, and we used the virtual snow indexes for informing the daily operation of the lake's dam. Numerical results show that such information is effective in extending the anticipation capacity of the lake operations, ultimately improving the system performance.

  11. A Novel 100 kW Power Hardware-in-the-Loop Emulation Test Bench for Permanent Magnet Synchronous Machines with Nonlinear Magnetics

    OpenAIRE

    Schmitt, Alexander; Richter, Jan; Gommeringer, Mario; Wersal, Thomas; Braun, Michael

    2016-01-01

    This paper presents a high dynamic power hardware-inthe-loop (PHIL) emulation test bench to mimic arbitrary permanent magnet synchronous machines with nonlinear magnetics. The proposed PHIL test bench is composed of a high performance real-time simulation system to calculate the machine behaviour and a seven level modular multiphase multilevel converter to emulate the power flow of the virtual machine. The PHIL test bench is parametrized for an automotive synchronous machine and controlled by...

  12. Simulation of fuel rod irradiation capsules in water loops by electric heater rods

    International Nuclear Information System (INIS)

    Lopez, J.; Montes, M.; Serrano, J.; Haefner, H.E.

    1984-01-01

    The out of pile simulation of irradiation devices was carried out by J.E.N. in the frame of the KfK-JEN joint experiment for irradiation of fast reactor fuel rods (IVO-FR2-Vg7). A typical single-wall-Nak (22% Na, 78% K) electrical heated capsule was fabricated and hydraulical tests were done. The capsule was instrumented with 10 thermocouples in order to obtain the radial temperature profile into the capsule in function of the electrical rod power (max. 215 w/cm), flow rate (max. 2,4 m 3 /h) and coolant temperature (max. 60degC). The experimental values are compared to the Tecap-Code results. (author)

  13. Long-Duration Testing of a Temperature-Swing Adsorption Compressor for Carbon Dioxide for Closed-Loop Air Revitalization Systems

    Science.gov (United States)

    Rosen, Micha; Mulloth, Lila; Varghese, Mini

    2005-01-01

    This paper describes the results of long-duration testing of a temperature-swing adsorption compressor that has application in the International Space Station (ISS) and future spacecraft for closing the air revitalization loop. The air revitalization system of the ISS operates in an open loop mode and relies on the resupply of oxygen and other consumables from Earth for the life support of astronauts. A compressor is required for delivering the carbon dioxide from a removal assembly to a reduction unit to recover oxygen and thereby closing the air-loop. The TSAC is a solid-state compressor that has the capability to remove CO2 from a low-pressure source, and subsequently store, compress, and deliver at a higher pressure as required by a processor. The TSAC is an ideal interface device for CO2 removal and reduction units in the air revitalization loop of a spacecraft for oxygen recovery. The TSAC was developed and its operation was successfully verified in integration tests with the flight-like Carbon Dioxide Removal Assembly (CDRA) at Marshall Space Flight Center prior to the long-duration tests. Long-duration tests reveal the impacts of repeated thermal cycling on the compressor components and the adsorbent material.

  14. Analysis, by RELAP5 code, of boron dilution phenomena in a mid-loop operation transient, performed in PKL III F2.1 RUN 1 test

    International Nuclear Information System (INIS)

    Mascari, F.; Vella, G.; Del Nevo, A.; D'Auria, F.

    2007-01-01

    The present paper deals with the post test analysis and accuracy quantification of the test PKL III F2.1 RUN 1 by RELAP5/Mod3.3 code performed in the framework of the international OECD/SETH PKL III Project. The PKL III is a full-height integral test facility (ITF) that models the entire primary system and most of the secondary system (except for turbine and condenser) of pressurized water reactor of KWU design of the 1300-MW (electric) class on a scale of 1:145. Detailed design was based to the largest possible extent on the specific data of Philippsburg nuclear power plant, unit 2. As for the test facilities of this size, the scaling concept aims to simulate overall thermal hydraulic behavior of the full-scale power plant [1]. The main purpose of the project is to investigate PWR safety issues related to boron dilution and in particular this experiment investigates (a) the boron dilution issue during mid-loop operation and shutdown conditions, and (b) assessing primary circuit accident management operations to prevent boron dilution as a consequence of loss of heat removal [2]. In this work the authors deal with a systematic procedure (developed at the university of Pisa) for code assessment and uncertainty qualification and its application to RELAP5 system code. It is used to evaluate the capability of RELAP5 to reproduce the thermal hydraulics of an inadvertent boron dilution event in a PWR. The quantitative analysis has been performed adopting the Fast Fourier Transform Based Method (FFTBM), which has the capability to quantify the errors in code predictions as compared to the measured experimental signal. (author)

  15. Integrated Testing of a Carbon Dioxide Removal Assembly and a Temperature-Swing Adsorption Compressor for Closed-Loop Air Revitalization

    Science.gov (United States)

    Knox, J. C.; Mulloth, Lila; Frederick, Kenneth; Affleck, Dave

    2003-01-01

    Accumulation and subsequent compression of carbon dioxide that is removed from space cabin are two important processes involved in a closed-loop air revitalization scheme of the International Space Station (ISS). The carbon dioxide removal assembly (CDRA) of ISS currently operates in an open loop mode without a compressor. This paper describes the integrated test results of a flight-like CDRA and a temperature-swing adsorption compressor (TSAC) for carbon dioxide removal and compression. The paper provides details of the TSAC operation at various CO2 loadings and corresponding performance of CDRA.

  16. Evaluation by the Double Loop Electrochemical Potentiokinetic Reactivation Test of Aged Ferritic Stainless Steel Intergranular Corrosion Susceptibility

    Science.gov (United States)

    Sidhom, H.; Amadou, T.; Braham, C.

    2010-12-01

    An experimental design method was used to determine the effect of factors that significantly affect the response of the double loop-electrochemical potentiokinetic reactivation (DL-EPR) test in controlling the susceptibility to intergranular corrosion (IGC) of UNS S43000 (AISI 430) ferritic stainless steel. The test response is expressed in terms of the reactivation/activation current ratio ( I r / I a pct). Test results analysed by the analysis of variance (ANOVA) method show that the molarity of the H2SO4 electrolyte and the potential scanning rate have a more significant effect on the DL-EPR test response than the temperature and the depassivator agent concentration. On the basis of these results, a study was conducted in order to determine the optimal operating conditions of the test as a nondestructive technique for evaluating IGC resistance of ferritic stainless steel components. Three different heat treatments are considered in this study: solution annealing (nonsensitized), aging during 3 hours at 773 K (500 °C) (slightly sensitized), and aging during 2 hours at 873 K (600 °C) (highly sensitized). The aim is to find the operating conditions that simultaneously ensure the selectivity of the attack (intergranular and chromium depleted zone) and are able to detect the effect of low dechromization. It is found that a potential scanning rate of 2.5 mV/s in an electrolyte composed of H2SO4 3 M solution without depassivator, at a temperature around 293 K (20 °C), is the optimal operating condition for the DL-EPR test. Using this condition, it is possible to assess the degree of sensitization (DOS) to the IGC of products manufactured in ferritic stainless steels rapidly, reliably, and quantitatively. A time-temperature-start of sensitization (TTS) diagram for the UNS S43000 (France Inox, Villepinte, France) stainless steel was obtained with acceptable accuracy by this method when the IGC sensitization criterion was set to I r / I a > 1 pct. This diagram is in

  17. Experimental investigation on thermal performance of a closed loop pulsating heat pipe (CLPHP) using methanol and distilled water at different filling ratios

    Science.gov (United States)

    Rahman, Md. Lutfor; Swarna, Anindita Dhar; Ahmed, Syed Nasif Uddin; Perven, Sanjida; Ali, Mohammad

    2016-07-01

    Pulsating Heat Pipes, the new two-phase heat transfer devices, with no counter current flow between liquid and vapor have become a modern topic for research in the field of thermal management. This paper focuses on the performance of methanol and distilled water as working fluid in a closed loop pulsating heat pipe (CLPHP). This performances are compared in terms of thermal resistance, heat transfer co-efficient, and evaporator and condenser wall temperature with variable heat inputs. Methanol and Distilled water are selected for their lower surface tension, dynamic viscosity and sensible heat. A closed loop PHP made of copper with 2mm ID and 2.5mm OD having total 8 loops are supplied with power input varied from 10W to 60W. During the experiment the PHP is kept vertical, while the filling ratio (FR) is increased gradually from 40% to 70% with 10% increment. The optimum filling ratio for a minimum thermal resistance is found to be 60% and 40% for distilled water and methanol respectively and methanol is found to be the better working fluid compared to distilled water in terms of its lower thermal resistance and higher heat transfer coefficient.

  18. Pre-Test Analysis of the MEGAPIE Spallation Source Target Cooling Loop Using the TRAC/AAA Code

    International Nuclear Information System (INIS)

    Bubelis, Evaldas; Coddington, Paul; Leung, Waihung

    2006-01-01

    A pilot project is being undertaken at the Paul Scherrer Institute in Switzerland to test the feasibility of installing a Lead-Bismuth Eutectic (LBE) spallation target in the SINQ facility. Efforts are coordinated under the MEGAPIE project, the main objectives of which are to design, build, operate and decommission a 1 MW spallation neutron source. The technology and experience of building and operating a high power spallation target are of general interest in the design of an Accelerator Driven System (ADS) and in this context MEGAPIE is one of the key experiments. The target cooling is one of the important aspects of the target system design that needs to be studied in detail. Calculations were performed previously using the RELAP5/Mod 3.2.2 and ATHLET codes, but in order to verify the previous code results and to provide another capability to model LBE systems, a similar study of the MEGAPIE target cooling system has been conducted with the TRAC/AAA code. In this paper a comparison is presented for the steady-state results obtained using the above codes. Analysis of transients, such as unregulated cooling of the target, loss of heat sink, the main electro-magnetic pump trip of the LBE loop and unprotected proton beam trip, were studied with TRAC/AAA and compared to those obtained earlier using RELAP5/Mod 3.2.2. This work extends the existing validation data-base of TRAC/AAA to heavy liquid metal systems and comprises the first part of the TRAC/AAA code validation study for LBE systems based on data from the MEGAPIE test facility and corresponding inter-code comparisons. (authors)

  19. Experimental Study on Hydrate Induction Time of Gas-Saturated Water-in-Oil Emulsion using a High-Pressure Flow Loop

    Directory of Open Access Journals (Sweden)

    Lv X.F.

    2015-11-01

    Full Text Available Hydrate is one of the critical precipitates which have to be controlled for subsea flow assurance. The induction time of hydrate is therefore a significant parameter. However, there have been few studies on the induction time of the natural gas hydrate formation in a flow loop system. Consequently, a series of experiments were firstly performed, including water, natural gas and Diesel oil, on the hydrate induction time under various conditions such as the supercooling and supersaturation degree, water cut, anti-agglomerant dosage, etc. The experiments were conducted in a high-pressure hydrate flow loop newly constructed in the China University of Petroleum (Beijing, and dedicated to flow assurance studies. Then, based on previous research, this study puts forward a method for induction time, which is characterized by clear definition, convenient measurement and good generality. Furthermore, we investigated the influences of the experimental parameters and analyzed the experimental phenomena for the hydrate induction time in a flowing system.

  20. High-temperature helium-loop facility

    International Nuclear Information System (INIS)

    Tokarz, R.D.

    1981-09-01

    The high-temperature helium loop is a facility for materials testing in ultrapure helium gas at high temperatures. The closed loop system is capable of recirculating high-purity helium or helium with controlled impurities. The gas loop maximum operating conditions are as follows: 300 psi pressure, 500 lb/h flow rate, and 2100 0 F temperature. The two test sections can accept samples up to 3.5 in. diameter and 5 ft long. The gas loop is fully instrumented to continuously monitor all parameters of loop operation as well as helium impurities. The loop is fully automated to operate continuously and requires only a daily servicing by a qualified operator to replenish recorder charts and helium makeup gas. Because of its versatility and high degree of parameter control, the helium loop is applicable to many types of materials research. This report describes the test apparatus, operating parameters, peripheral systems, and instrumentation system. The experimental capabilities and test conand presents the results that have been obtained. The study has been conducted using a four-phase approach. The first phase develops the solution to the steady-state radon-diffusion equation in one-dimensieered barriers; disposal charge analysis; analysis of spent fuel policy implementation; spent f water. Field measurements and observations are reported for each site. Analytical data and field measurements are presented in tables and maps. Uranium concentrations in the sediments which were above detection limits ranged from 0.10 t 51.2 ppM. The mean of the logarithms of the uranium concentrations was 0.53. A group of high uranium concentrations occurs near the junctions of quadrangles AB, AC, BB, a 200 mK. In case 2), x-ray studies of isotopic phase separation in 3 He-- 4 He bcc solids were carried out by B. A. Fraass

  1. Experimental system description for air-water CCFL tests of the 161-rod FLECHT-SEASET test vessel upper plenum

    International Nuclear Information System (INIS)

    Fogdall, S.P.; Anderson, J.L.

    1983-01-01

    A series of countercurrent flow limiting (CCFL) experiments has been performed by EG and G Idaho, Inc. in the Steam-Air-Water (SAW) test facility at the Idaho National Engineering Laboratory on behalf of the US Nuclear Regulatory Commission (NRC). Tests were performed in a mockup of the vessel for the 161-Rod Systems Effects Test (SET) facility of the FLECHT-SEASET program, conducted by the Westinghouse Electric Corporation. Westinghouse and the NRC will use the test results to provide a CCFL correlation to predict the flooding behavior in the upper plenum of the SET vessel. This paper presents a description of the experimental system and the test conduct, including data validation and uncertainty analysis. The test objectives centered on experimentally obtaining coefficients in the Wallis correlation for flooding with the specific vessel geometry. The test conditions and vessel configuration are described and the design of the test loop, instrumentation, and data acquisition are discussed. The establishment of a test point and the resultant data are described

  2. Mass transfer of steels for FBR in sodium loop

    International Nuclear Information System (INIS)

    Susukida, Hiroshi; Yonezawa, Toshio; Ueda, Mitsuo; Imazu, Takayuki; Kiyokawa, Teruyuki.

    1976-06-01

    In order to grasp quantitatively the corrosion and mass transfer of steels for FBR in sodium loop and to establish their allowable stress value and corrosion rate, a special sodium loop for material testing was designed and fabricated and the steels were given 3010 hours exposing test in the sodium loop. This paper gives the outline of the sodium loop and the results of the test. (1) Carburization and a slight increase in weight were observed in the specimens of type 304 stainless steel exposed in the sodium loop for 3010 hours, while decarburization was observed in the specimens of 2 1/4 Cr-1 Mo steel. It is considered that these phenomena were caused by the downstream factor of the sodium loop. (2) A remarkable decrease of Charpy absorbed energy was observed in the specimens of type 304 stainless steel exposed in the sodium loop. It is considered that this resulted from the weakening of the grain boundary due to heat history and mass transfer. (3) The specimens exposed in the sodium loop must be washed by ultrasonic waves in a water bath after washing in alcohol. (auth.)

  3. Multi-Megawatt-Scale Power-Hardware-in-the-Loop Interface for Testing Ancillary Grid Services by Converter-Coupled Generation: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Koralewicz, Przemyslaw J [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Gevorgian, Vahan [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Wallen, Robert B [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

    2017-07-26

    Power-hardware-in-the-loop (PHIL) is a simulation tool that can support electrical systems engineers in the development and experimental validation of novel, advanced control schemes that ensure the robustness and resiliency of electrical grids that have high penetrations of low-inertia variable renewable resources. With PHIL, the impact of the device under test on a generation or distribution system can be analyzed using a real-time simulator (RTS). PHIL allows for the interconnection of the RTS with a 7 megavolt ampere (MVA) power amplifier to test multi-megawatt renewable assets available at the National Wind Technology Center (NWTC). This paper addresses issues related to the development of a PHIL interface that allows testing hardware devices at actual scale. In particular, the novel PHIL interface algorithm and high-speed digital interface, which minimize the critical loop delay, are discussed.

  4. Mockup tests of void fraction in moderator cell and two-phase thermosiphon loop of cold neutron source in China Advanced Research Reactor

    International Nuclear Information System (INIS)

    Du Shejiao; Bi Qincheng; Chen Tingkuan; Feng Quanke; Li Xiaoming

    2004-01-01

    Full-scale mockup tests were carried out using freon-113 as a working fluid to verify the design of China Advanced Research Reactor (CARR) Cold neutron Source (CNS), which is a two-phase hydrogen thermosiphon loop consisting of an annular cylindrical moderator cell, two separated hydrogen transfer tubes and a condenser. The circulation characteristics, liquid level and void fraction in the moderator cell against the variation of the heat load were studied. The density ratio and the volumetric evaporating rate of the mockup test are kept the same as those of CARR CNS. The test results show that the mockup loop can establish stable circulation and has a self-regulating characteristic. Within the moderator cell, the inner shell contains only vapor and the outer shell contains the mixture of vapor-liquid with void fraction in a certain range. (authors)

  5. Life-cycle testing of receiving waters with Ceriodaphnia dubia

    Energy Technology Data Exchange (ETDEWEB)

    Stewart, A.J.; Konetsky, B.K.

    1996-12-31

    Seven-day tests with Ceriodaphnia dubia are commonly used to estimate toxicity of effluents or receiving waters but can sometimes yield {open_quotes}no toxicity{close_quotes} outcomes even if pollutants are present. We conducted two sets of full life-cycle tests with C. dubia to (1) see if tests with longer exposure periods would reveal evidence for toxicity that might not be evident from 7-day tests, and (2) determine the relative importance of water quality versus food as factors influencing C. dubia reproduction. In the first set of tests, C. dubia was reared in diluted mineral water (negative control), water from a stream impacted by coal fly-ash, or water from a retention basin containing sediments contaminated with mercury, other metals and polychlorinated biphenyls. The second set of tests used water from the retention basin only, but this water was either filtered or not filtered, and food was either added or not added, prior to testing. C. dubia survival and reproduction did not differ much among the three water types in the first set of tests, but these two parameters were strongly affected by the filtering and food-addition treatments in the second set of tests. Thus, C. dubia appeared to be relatively insensitive to general water-quality factors, but quite sensitive to food-related factors. Regression analyses showed that the predictability of life-time reproduction by C. dubia from the results of 7-day tests was very low (R{sup 2}< 0.35) in five of the six experiments. The increase in predictability as a function of test duration also differed among water types in the first set of tests, and among treatments in the second set of tests. Thus, 7-day tests with C. dubia may be used to quantify water-quality problems, but it may not be possible to reliably extrapolate the results of these tests to longer time scales.

  6. Life-cycle testing of receiving waters with Ceriodaphnia dubia

    International Nuclear Information System (INIS)

    Stewart, A.J.; Beane, B.K.

    1995-01-01

    Seven-day tests with Ceriodaphnia are commonly used to estimate the toxicity of effluents or receiving waters, but may yield no toxicity outcomes even when pollutants are present (a possible type II error). The authors conducted two sets of full life-cycle tests with C. dubia to (1) see if tests with longer exposure periods revealed evidence for toxicity that might not be evident from shorter tests, and (2) determine the relative importance of water quality versus food as factors influencing C. dubia reproduction. In the first set of tests, daphnids were reared in diluted mineral water (control), water from a stream impacted by coal fly-ash, or water from a mercury-contaminated retention basin. The second set of tests used water from the retention basin only, but this water was either filtered or not filtered, and food was either added or not added. C. dubia survival and reproduction did not differ much among the three waters in the first set of tests. However, both parameters were strongly affected by the filtering and food-addition treatments in the second set of tests. Thus, C. dubia seems to be moderately insensitive to general water-quality factors, but quite sensitive to food-related parameters. Regression analysis showed that the predictability of life-time reproduction of C. dubia from 7-day test results was low in five of six cases. The increase in predictability as a function of test duration also differed among water types (first set of tests), and among treatments (second set of tests). Thus, 7-day tests with C. dubia may be used to quantify water-quality problems, but it may not be possible to reliably extrapolate the results of such tests to longer time scales

  7. Storm Water BMP Tool Implementation Testing

    Science.gov (United States)

    2017-12-01

    Under project 2015-ORIL 7, a screening tool was developed to assist Local communities with selecting post-construction storm water best management practices (BMPs) to comply with the Ohio Environmental Protection Agencys (Ohio EPA) statewide Const...

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

    International Nuclear Information System (INIS)

    1983-03-01

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

  9. Integrated Testing of a 4-Bed Molecular Sieve and a Temperature-Swing Adsorption Compressor for Closed-Loop Air Revitalization

    Science.gov (United States)

    Knox, James C.; Mulloth, Lila M.; Affleck, David L.

    2004-01-01

    Accumulation and subsequent compression of carbon dioxide that is removed from space cabin are two important processes involved in a closed-loop air revitalization scheme of the International Space Station (ISS). The 4-Bed Molecular Sieve (4BMS) of ISS currently operates in an open loop mode without a compressor. This paper reports the integrated 4BMS and liquid-cooled TSAC testing conducted during the period of March 3 to April 18, 2003. The TSAC prototype was developed at NASA Ames Research Center (ARC). The 4BMS was modified to a functionally flight-like condition at NASA Marshall Space Flight Center (MSFC). Testing was conducted at MSFC. The paper provides details of the TSAC operation at various CO2 loadings and corresponding performance of CDRA.

  10. Double-winding Wilson loops in SU(N) Yang-Mills theory - A criterion for testing the confinement models -

    Science.gov (United States)

    Matsudo, Ryutaro; Kondo, Kei-Ichi; Shibata, Akihiro

    2018-03-01

    We examine how the average of double-winding Wilson loops depends on the number of color N in the SU(N) Yang-Mills theory. In the case where the two loops C1 and C2 are identical, we derive the exact operator relation which relates the doublewinding Wilson loop operator in the fundamental representation to that in the higher dimensional representations depending on N. By taking the average of the relation, we find that the difference-of-areas law for the area law falloff recently claimed for N = 2 is excluded for N ⩾ 3, provided that the string tension obeys the Casimir scaling for the higher representations. In the case where the two loops are distinct, we argue that the area law follows a novel law (N - 3)A1/(N - 1) + A2 with A1 and A2(A1 law when (N ⩾ 3). Indeed, this behavior can be confirmed in the two-dimensional SU(N) Yang-Mills theory exactly.

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

    International Nuclear Information System (INIS)

    Novakovic, M.

    1961-12-01

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

  12. Development and Testing of Infrared Water Current Meter | Ezenne ...

    African Journals Online (AJOL)

    Continuous monitoring of the river flow is essential for assessing water availability. River flow velocity is crucial to simulate discharge hydrographs of water in the hydrological system.This study developed a digital water current meter with infrared. The infrared current meter was tested using Ebonyi River at Obollo-Etiti and ...

  13. Modeling of cavitation in sodium flow by water flow test in prototypical LMFBR components

    International Nuclear Information System (INIS)

    Soehendro, B.; Trejo, F.; Bonilla, C.F.

    1976-02-01

    Cavitation tests of water recirculating through a venturi or a rounded inlet nozzle were carried out under steady conditions. Water temperature was varied from 100 to 195 0 F. Argon was used as cover gas and to regulate pressure in the loop. Cavitation was detected by the voltage output of a piezo-electric lead-zirconate-titanate ceramic on a titanium alloy horn facing the cavitator outlet. Three different incipient and desinent cavitation modes, designated gaseous, gaseous-vaporous, and vaporous were observed, and no significant difference was found between the conditions for incipient and desinent cavitation. Local cavitation number and fraction of equilibrium argon pressure in the gas phase at incipient and desinent cavitation are almost constant in the ranges of temperature and argon content studied. Injection of gas into the stream does not change the conditions for vaporous cavitation, but affects the gaseous and gaseous-vaporous cavitation considerably

  14. Effect of using acetone and distilled water on the performance of open loop pulsating heat pipe (OLPHP) with different filling ratios

    Science.gov (United States)

    Rahman, Md. Lutfor; Afrose, Tonima; Tahmina, Halima Khatun; Rinky, Rumana Parvin; Ali, Mohammad

    2016-07-01

    Pulsating heat pipe (PHP) is a new innovation in the modern era of miniaturizes thermal management system for its higher heating and cooling capacity. The objective of this experiment is to observe the performance of open loop pulsating heat pipe using two fluids at different filling ratios. This OLPHP is a copper capillary tube of 2.5mm outer diameter and 2mm inner diameter. It consists of 8 loops where the evaporative section is 50mm, adiabatic section is 120mm and condensation section is 80mm. The experiment is conducted with distilled water and acetone at 40%, 50%, 60%, and 70% filling ratios where 0° (vertical) is considered as definite angle of inclination. Distilled water and acetone are selected as working fluids considering their different latent heat of vaporization and surface tension. It is found that acetone shows lower thermal resistance than water at all heat inputs. Best performance of acetone is attained at 70% filling ratio. Water displays better heat transfer capability at 50% filling ratio.

  15. 11 CFR 100.131 - Testing the waters.

    Science.gov (United States)

    2010-01-01

    ... 11 Federal Elections 1 2010-01-01 2010-01-01 false Testing the waters. 100.131 Section 100.131 Federal Elections FEDERAL ELECTION COMMISSION GENERAL SCOPE AND DEFINITIONS (2 U.S.C. 431) Exceptions to Expenditures § 100.131 Testing the waters. (a) General exemption. Payments made solely for the purpose of...

  16. 11 CFR 100.72 - Testing the waters.

    Science.gov (United States)

    2010-01-01

    ... 11 Federal Elections 1 2010-01-01 2010-01-01 false Testing the waters. 100.72 Section 100.72 Federal Elections FEDERAL ELECTION COMMISSION GENERAL SCOPE AND DEFINITIONS (2 U.S.C. 431) Exceptions to Contributions § 100.72 Testing the waters. (a) General exemption. Funds received solely for the purpose of...

  17. Design study on large-scale mercury loop for engineering test of target of high-intensity proton accelerator

    International Nuclear Information System (INIS)

    Hino, Ryutaro; Haga, Katsuhiro; Aita, Hideki; Sekita, Kenji; Sudo, Yukio; Koiso, Kohji; Kaminaga, Masanori; Takahashi, Hiromichi.

    1997-03-01

    A heavy liquid-metal target has been proposed as a representative target of a 5MW-scale neutron source for a neutron scattering facility coupled with a high-intensity proton accelerator. In the report, about mercury considered to be the best material of the heavy liquid-metal target, its properties needed for the design were formulated, and results of research on mercury treatment and of evaluation of heat removal performance on the basis of generating heat obtained by a numerical calculation of a spallation reaction were presented. From these results, a 1.5MW-scale mercury loop which equals to that for the first stage operation of the neutron science program of JAERI was designed conceptually for obtaining design data of the mercury target, and basic flow diagram of the loop and specifications of components were decided: diameter of pipelines flowing mercury at the velocity below 1m/s, power of an electro-magnet pump and structure of a cooler. Through the design, engineering problems were made clear such as selection and development of mercury-resistant materials and optimization of the loop and components for decreasing mercury inventory. (author)

  18. Irradiation capability of Japanese materials test reactor for water chemistry experiments

    International Nuclear Information System (INIS)

    Hanawa, Satoshi; Hata, Kuniki; Chimi, Yasuhiro; Nishiyama, Yutaka; Nakamura, Takehiko

    2012-09-01

    Appropriate understanding of water chemistry in the core of LWRs is essential as chemical species generated due to water radiolysis by neutron and gamma-ray irradiation govern corrosive environment of structural materials in the core and its periphery, causing material degradation such as stress corrosion cracking. Theoretical model calculation such as water radiolysis calculation gives comprehensive understanding of water chemistry at irradiation field where we cannot directly monitor. For enhancement of the technology, accuracy verification of theoretical models under wide range of irradiation conditions, i.e. dose rate, temperature etc., with well quantified in-pile measurement data is essential. Japan Atomic Energy Agency (JAEA) has decided to launch water chemistry experiments for obtaining data that applicable to model verification as well as model benchmarking, by using an in-pile loop which will be installed in the Japan Materials Testing Reactor (JMTR). In order to clarify the irradiation capability of the JMTR for water chemistry experiments, preliminary investigations by water radiolysis / ECP model calculations were performed. One of the important irradiation conditions for the experiments, i.e. dose rate by neutron and gamma-ray, can be controlled by selecting irradiation position in the core. In this preliminary study, several representative irradiation positions that cover from highest to low absorption dose rate were chosen and absorption dose rate at the irradiation positions were evaluated by MCNP calculations. As a result of the calculations, it became clear that the JMTR could provide the irradiation conditions close to the BWR. The calculated absorption dose rate at each irradiation position was provided to water radiolysis calculations. The radiolysis calculations were performed under various conditions by changing absorption dose rate, water chemistry of feeding water etc. parametrically. Qualitatively, the concentration of H 2 O 2 , O 2 and

  19. High Temperature Operational Experiences of Helium Experimental Loop

    International Nuclear Information System (INIS)

    Kim, Chan Soo; Hong, Sung-Deok; Kim, Eung-Seon; Kim, Min Hwan

    2015-01-01

    The development of high temperature components of VHTR is very important because of its higher operation temperature than that of a common light water reactor and high pressure industrial process. The development of high temperature components requires the large helium loop. Many countries have high temperature helium loops or a plan for its construction. Table 1 shows various international state-of-the-art of high temperature and high pressure gas loops. HELP performance test results show that there is no problem in operation of HELP at the very high temperature experimental condition. These experimental results also provide the basic information for very high temperature operation with bench-scale intermediate heat exchanger prototype in HELP. In the future, various heat exchanger tests will give us the experimental data for GAMMA+ validation about transient T/H behavior of the IHX prototype and the optimization of the working fluid in the intermediate loop

  20. Full Scale Drinking Water System Decontamination at the Water Security Test Bed

    Data.gov (United States)

    U.S. Environmental Protection Agency — The EPA’s Water Security Test Bed (WSTB) facility is a full-scale representation of a drinking water distribution system. In collaboration with the Idaho National...

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

    Directory of Open Access Journals (Sweden)

    Mehdi Dehjourian

    2016-08-01

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

  2. Analysis of a hot-leg small break loss-of-coolant accident in a three-loop westinghouse pressurized water reactor plant

    International Nuclear Information System (INIS)

    Peterson, C.E.; Chexal, V.K.; Clements, T.B.

    1985-01-01

    The RETRAN-02 computer code was used to perform a best-estimate analysis of a 7.52-cm-diam hotleg break in a three-loop Westinghouse pressurized water reactor. This break size produced a net primary coolant mass depletion through the early portion of the transient. The primary system started to refill only after the accumulator valves opened. As the primary system refilled, there were extreme temperature differentials around the system with cold, denser fluid collecting at the lower elevations and two-phase fluid at higher elevations

  3. Study of the effect of slight variants to a 3-loop pressurized water nuclear reactor design in order to improve the reactor safety

    International Nuclear Information System (INIS)

    Castiglia, F.; Oliveri, E.; Taibi, S.; Vella, G.

    1992-01-01

    In order to improve the safety features of a 3-loop pressurized water nuclear reactor we propose a slight design variant consisting in the introduction of a bypass hole in the divider plate of the coolant chambers of the steam generators. The aim is to reduce both the extent and the duration of the core exposure and thus the maximum value of the peak cladding temperature, in case of a hypothetical cold leg small break loss of coolant accident. The proposal, as attested by a preliminary RELAP5/MOD3 analysis, seems to deserve some attention. (6 figures) (Author)

  4. The Quality Testing of Water from Microbiology and Radioactivity

    International Nuclear Information System (INIS)

    Zainul Kamal; Yazid, M.; Mulyaningsih; Iim lmroatin

    2002-01-01

    The quality testing of well water from microbiologic and radioactivity has been done. The samples were taken from Degolan and Lodadi village, Ngemplak, Sleman. The quality testing based from standard procedure of microbiologic and environmental radioactivity. From the experimentally results showed that E. Coli in well water = 5 - 920 JPT / 100 ml, Streptococcus in well water 0 - 4 JPT /100 ml, E. Coli and Streptococcus in PAM water 0 JPT / 100 ml, radioactivity β totally in well water 0.08-0.34 Bq/l and in PAM water 0.08 - 0.31 Bq/l. From the dates required could be concluded that in microbiologically aspects the value of E. Coli and Streptococcus in well water higher than the threshold value from Health Department Rl 416/Menkes/PER/IX/1990, in radioactivity aspect lower than the threshold value from Health Department RI 416/Menkes/PER/IX/1990. (author)

  5. Cold vacuum drying residual free water test description

    International Nuclear Information System (INIS)

    Pajunen, A.L.

    1997-01-01

    Residual free water expected to remain in a Multi-Canister Overpack (MCO) after processing in the Cold Vacuum Drying (CVD) Facility is investigated based on three alternative models of fuel crevices. Tests and operating conditions for the CVD process are defined based on the analysis of these models. The models consider water pockets constrained by cladding defects, water constrained in a pore or crack by flow through a porous bed, and water constrained in pores by diffusion. An analysis of comparative reaction rate constraints is also presented indicating that a pressure rise test can be used to show MCO's will be thermally stable at operating temperatures up to 75 C

  6. Development test procedure High Pressure Water Jet System

    International Nuclear Information System (INIS)

    Crystal, J.B.

    1995-01-01

    Development testing will be performed on the water jet cleaning fixture to determine the most effective arrangement of water jet nozzles to remove contamination from the surfaces of canisters and other debris. The following debris may be stained with dye to simulate surface contaminates: Mark O, Mark I, and Mark II Fuel Storage Canisters (both stainless steel and aluminum), pipe of various size, (steel, stainless, carbon steel and aluminum). Carbon steel and stainless steel plate, channel, angle, I-beam and other surfaces, specifically based on the Scientific Ecology Group (SEG) inventory and observations of debris within the basin. Test procedure for developmental testing of High Pressure Water Jet System

  7. The Performance test of Mechanical Sodium Pump with Water Environment

    International Nuclear Information System (INIS)

    Cho, Chungho; Kim, Jong-Man; Ko, Yung Joo; Jeong, Ji-Young; Kim, Jong-Bum; Ko, Bock Seong; Park, Sang Jun; Lee, Yoon Sang

    2015-01-01

    As contrasted with PWR(Pressurized light Water Reactor) using water as a coolant, sodium is used as a coolant in SFR because of its low melting temperature, high thermal conductivity, the high boiling temperature allowing the reactors to operate at ambient pressure, and low neutron absorption cross section which is required to achieve a high neutron flux. But, sodium is violently reactive with water or oxygen like the other alkali metal. So Very strict requirements are demanded to design and fabricate of sodium experimental facilities. Furthermore, performance testing in high temperature sodium environments is more expensive and time consuming and need an extra precautions because operating and maintaining of sodium experimental facilities are very difficult. The present paper describes performance test results of mechanical sodium pump with water which has been performed with some design changes using water test facility in SAM JIN Industrial Co. To compare the hydraulic characteristic of model pump with water and sodium, the performance test of model pump were performed using vender's experimental facility for mechanical sodium pump. To accommodate non-uniform thermal expansion and to secure the operability and the safety, the gap size of some parts of original model pump was modified. Performance tests of modified mechanical sodium pump with water were successfully performed. Water is therefore often selected as a surrogate test fluid because it is not only cheap, easily available and easy to handle but also its important hydraulic properties (density and kinematic viscosity) are very similar to that of the sodium. Normal practice to thoroughly test a design or component before applied or installed in reactor is important to ensure the safety and operability in the sodium-cooled fast reactor (SFR). So, in order to estimate the hydraulic behavior of the PHTS pump of DSFR (600 MWe Demonstraion SFR), the performance tests of the model pump such as performance

  8. Summary of Test Results From a 1 kW(sub e)-Class Free-Piston Stirling Power Convertor Integrated With a Pumped NaK Loop

    Science.gov (United States)

    Briggs, Maxwell H.; Geng, Steven M.; Pearson, J. Boise; Godfroy, Thomas J.

    2010-01-01

    As a step towards development of Stirling power conversion for potential use in Fission Surface Power (FSP) systems, a pair of commercially available 1 kW class free-piston Stirling convertors was modified to operate with a NaK liquid metal pumped loop for thermal energy input. This was the first-ever attempt at powering a free-piston Stirling engine with a pumped liquid metal heat source and is a major FSP project milestone towards demonstrating technical feasibility. The tests included performance mapping the convertors over various hot and cold-end temperatures, piston amplitudes and NaK flow rates; and transient test conditions to simulate various start-up and fault scenarios. Performance maps of the convertors generated using the pumped NaK loop for thermal input show increases in power output over those measured during baseline testing using electric heating. Transient testing showed that the Stirling convertors can be successfully started in a variety of different scenarios and that the convertors can recover from a variety of fault scenarios.

  9. Summary of Test Results From a 1 kWe-Class Free-Piston Stirling Power Convertor Integrated With a Pumped NaK Loop

    Science.gov (United States)

    Briggs, Maxwell H.; Geng, Steven M.; Pearson, J. Boise; Godfroy, Thomas J.

    2010-01-01

    As a step towards development of Stirling power conversion for potential use in Fission Surface Power (FSP) systems, a pair of commercially available 1 kW class free-piston Stirling convertors was modified to operate with a NaK liquid metal pumped loop for thermal energy input. This was the first-ever attempt at powering a free-piston Stirling engine with a pumped liquid metal heat source and is a major FSP project milestone towards demonstrating technical feasibility. The tests included performance mapping the convertors over various hot and cold-end temperatures, piston amplitudes and NaK flow rates; and transient test conditions to simulate various start-up and fault scenarios. Performance maps of the convertors generated using the pumped NaK loop for thermal input show increases in power output over those measured during baseline testing using electric heating. Transient testing showed that the Stirling convertors can be successfully started in a variety of different scenarios and that the convertors can recover from a variety of fault scenarios.

  10. Coronal Heating: Testing Models of Coronal Heating by Forward-Modeling the AIA Emission of the Ansample of Coronal Loops

    Science.gov (United States)

    Malanushenko, A. V.

    2015-12-01

    We present a systemic exploration of the properties of coronal heating, by forward-modeling the emission of the ensemble of 1D quasi-steady loops. This approximations were used in many theoretical models of the coronal heating. The latter is described in many such models in the form of power laws, relating heat flux through the photosphere or volumetric heating to the strength of the magnetic field and length of a given field line. We perform a large search in the parameter space of these power laws, amongst other variables, and compare the resulting emission of the active region to that observed by AIA. We use a recently developed magnetic field model which uses shapes of coronal loops to guide the magnetic model; the result closely resembles observed structures by design. We take advantage of this, by comparing, in individual sub-regions of the active region, the emission of the active region and its synthetic model. This study allows us to rule out many theoretical models and formulate predictions for the heating models to come.

  11. [Eosin Y-water test for sperm function examination].

    Science.gov (United States)

    Zha, Shu-wei; Lü, Nian-qing; Xu, Hao-qin

    2015-06-01

    Based on the principles of the in vitro staining technique, hypotonic swelling test, and water test, the Eosin Y-water test method was developed to simultaneously detect the integrity of the sperm head and tail and sperm membrane structure and function. As a widely used method in clinical laboratories in China, the Eosin Y-water test is methodologically characterized by three advantages. Firstly, both the sperm head and tail can be detected at the same time, which allows easy and comprehensive assessment of membrane damage in different parts of sperm. Secondly, distilled water is used instead of the usual formula solution to simplify and standardize the test by eliminating any potential effects on the water molecules through the sperm membrane due to different osmotic pressure or different sugar proportions and electrolyte solutions. Thirdly, the test takes less time and thus can be repeated before and after treatment. This article focuses on the fundamental principles and modification of the Eosin Y-water test and its application in sperm function examination and routine semen analysis for male infertility, assessment of the quality of sperm retrieved by testicular fine needle aspiration, semen cryopreservation program development, and evaluation of sperm membrane integrity after microwave radiation.

  12. Experimental studies and computational benchmark on heavy liquid metal natural circulation in a full height-scale test loop for small modular reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Yong-Hoon, E-mail: chaotics@snu.ac.kr [Department of Energy Systems Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of); Cho, Jaehyun [Korea Atomic Energy Research Institute, 111 Daedeok-daero, 989 Beon-gil, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Lee, Jueun; Ju, Heejae; Sohn, Sungjune; Kim, Yeji; Noh, Hyunyub; Hwang, Il Soon [Department of Energy Systems Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of)

    2017-05-15

    Highlights: • Experimental studies on natural circulation for lead-bismuth eutectic were conducted. • Adiabatic wall boundaries conditions were established by compensating heat loss. • Computational benchmark with a system thermal-hydraulics code was performed. • Numerical simulation and experiment showed good agreement in mass flow rate. • An empirical relation was formulated for mass flow rate with experimental data. - Abstract: In order to test the enhanced safety of small lead-cooled fast reactors, lead-bismuth eutectic (LBE) natural circulation characteristics have been studied. We present results of experiments with LBE non-isothermal natural circulation in a full-height scale test loop, HELIOS (heavy eutectic liquid metal loop for integral test of operability and safety of PEACER), and the validation of a system thermal-hydraulics code. The experimental studies on LBE were conducted under steady state as a function of core power conditions from 9.8 kW to 33.6 kW. Local surface heaters on the main loop were activated and finely tuned by trial-and-error approach to make adiabatic wall boundary conditions. A thermal-hydraulic system code MARS-LBE was validated by using the well-defined benchmark data. It was found that the predictions were mostly in good agreement with the experimental data in terms of mass flow rate and temperature difference that were both within 7%, respectively. With experiment results, an empirical relation predicting mass flow rate at a non-isothermal, adiabatic condition in HELIOS was derived.

  13. Integrated water management system - Description and test results. [for Space Station waste water processing

    Science.gov (United States)

    Elden, N. C.; Winkler, H. E.; Price, D. F.; Reysa, R. P.

    1983-01-01

    Water recovery subsystems are being tested at the NASA Lyndon B. Johnson Space Center for Space Station use to process waste water generated from urine and wash water collection facilities. These subsystems are being integrated into a water management system that will incorporate wash water and urine processing through the use of hyperfiltration and vapor compression distillation subsystems. Other hardware in the water management system includes a whole body shower, a clothes washing facility, a urine collection and pretreatment unit, a recovered water post-treatment system, and a water quality monitor. This paper describes the integrated test configuration, pertinent performance data, and feasibility and design compatibility conclusions of the integrated water management system.

  14. Testing flight software on the ground: Introducing the hardware-in-the-loop simulation method to the Alpha Magnetic Spectrometer on the International Space Station

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Wenhao, E-mail: wenhao_sun@126.com [Southeast University, Nanjing 210096 (China); Cai, Xudong [Massachusetts Institute of Technology, MA 02139-4307 (United States); Meng, Qiao [Southeast University, Nanjing 210096 (China)

    2016-04-11

    Complex automatic protection functions are being added to the onboard software of the Alpha Magnetic Spectrometer. A hardware-in-the-loop simulation method has been introduced to overcome the difficulties of ground testing that are brought by hardware and environmental limitations. We invented a time-saving approach by reusing the flight data as the data source of the simulation system instead of mathematical models. This is easy to implement and it works efficiently. This paper presents the system framework, implementation details and some application examples.

  15. Evaluation of annual corrosion tests for aggressive water

    Science.gov (United States)

    Dubová, V.; Ilavský, J.; Barloková, D.

    2011-12-01

    Internal corrosion has a significant effect on the useful life of pipes, the hydraulic conditions of a distribution system and the quality of the water transported. All water is corrosive under some conditions, and the level of this corrosion depends on the physical and chemical properties of the water and properties of the pipe material. Galvanic treatment is an innovation for protecting against corrosion, and this method is also suitable for removal of water stone too. This method consists of the electrogalvanic principle, which is generated by the flowing of water between a zinc anode and the cupro-alloy cover of a column. This article presents experimental corrosion tests at water resource Pernek (This water resource-well marked as HL-1 is close to the Pernek of village), where the device is operating based on this principle.

  16. Computational fluid dynamic simulation of pressurizer safety valve loop seal purge phenomena in nuclear power plants

    International Nuclear Information System (INIS)

    Park, Jong Woon

    2012-01-01

    In Korean 3 Loop plants a water loop seal pipe is installed containing condensed water upstream of a pressurizer safety valve to protect the valve disk from the hot steam environment. The loop seal water purge time is a key parameter in safety analyses for overpressure transients, because it delays valve opening. The loop seal purge time is uncertain to measure by test and thus 3-dimensional realistic computational fluid dynamics (CFD) model is developed in this paper to predict the seal water purge time before full opening of the valve which is driven by steam after water purge. The CFD model for a typical pressurizer safety valve with a loop seal pipe is developed using the computer code of ANSYS CFX 11. Steady-state simulations are performed for full discharge of steam at the valve full opening. Transient simulations are performed for the loop seal dynamics and to estimate the loop seal purge time. A sudden pressure drop higher than 2,000 psia at the tip of the upper nozzle ring is expected from the steady-state calculation. Through the transient simulation, almost loop seal water is discharged within 1.2 second through the narrow opening between the disk and the nozzle of the valve. It can be expected that the valve fully opens at least before 1.2 second because constant valve opening is assumed in this CFX simulation, which is conservative because the valve opens fully before the loop seal water is completely discharged. The predicted loop seal purge time is compared with previous correlation. (orig.)

  17. Computational fluid dynamic simulation of pressurizer safety valve loop seal purge phenomena in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Woon [Dongguk Univ., Gyeongju (Korea, Republic of). Nuclear and Energy Engineering Dept.

    2012-11-15

    In Korean 3 Loop plants a water loop seal pipe is installed containing condensed water upstream of a pressurizer safety valve to protect the valve disk from the hot steam environment. The loop seal water purge time is a key parameter in safety analyses for overpressure transients, because it delays valve opening. The loop seal purge time is uncertain to measure by test and thus 3-dimensional realistic computational fluid dynamics (CFD) model is developed in this paper to predict the seal water purge time before full opening of the valve which is driven by steam after water purge. The CFD model for a typical pressurizer safety valve with a loop seal pipe is developed using the computer code of ANSYS CFX 11. Steady-state simulations are performed for full discharge of steam at the valve full opening. Transient simulations are performed for the loop seal dynamics and to estimate the loop seal purge time. A sudden pressure drop higher than 2,000 psia at the tip of the upper nozzle ring is expected from the steady-state calculation. Through the transient simulation, almost loop seal water is discharged within 1.2 second through the narrow opening between the disk and the nozzle of the valve. It can be expected that the valve fully opens at least before 1.2 second because constant valve opening is assumed in this CFX simulation, which is conservative because the valve opens fully before the loop seal water is completely discharged. The predicted loop seal purge time is compared with previous correlation. (orig.)

  18. Air-water tests in support of LLTR series II Test A-4

    International Nuclear Information System (INIS)

    Chen, K.

    1980-07-01

    A series of tests injecting air into a tank of stagnant water was conducted in June 1980 utilizing the GE Plenum Mixing Test Facility in San Jose, California. The test was concerned with investigating the behavior of air jets at a submerged orifice in water over a wide range of flow rates. The main objective was to improve the basic understanding of gas-liquid phenomena (e.g., leak dynamics, gas bubble agglomeration, etc.) in a simulated tube bundle through visualization. The experimental results from these air-water tests will be used as a guide to help select the leak size for LLTR Series II Test A-4 because air-water system is a good simulation of water-sodium mixture

  19. UPTF/TEST10B/RUN081, Steam/Water Flow Phenomena Reflood PWR Cold Leg Break LOCA

    International Nuclear Information System (INIS)

    1998-01-01

    1 - Description of test facility: The Upper Plenum Test Facility (UPTF) is a geometrical full-scale simulation of the primary system of the four-loop 1300 MWe Siemens/KWU pressurized water reactor (PWR) at Grafenrheinfeld. The test vessel, upper plenum and its internals, downcomer, primary loops, pressurizer and surge line are replicas of the reference plant. The core, coolant pumps, steam generators and containment of a PWR are replaced by simulators which simulate the boundary and initial conditions during end-of-blowdown, refill and reflood phase following a loss-of-coolant accident (LOCA) with a hot or cold leg break. The break size and location can be simulated in the broken loop. The emergency core coolant (ECC) injection systems at the UPTF are designed to simulate the various ECC injection modes, such as hot leg, upper plenum, cold leg, downcomer or combined hot and cold leg injection of different ECC systems of German and US/Japan PWRs. Moreover, eight vent valves are mounted in the core barrel above the hot leg nozzle elevation for simulation of ABB and B and W PWRs. The UPTF primary system is divided into the investigation and simulation areas. The investigation areas, which are the exact replicas of a GPWR, consist of the upper plenum with internals, hot legs, cold legs and downcomer. The realistic thermal-hydraulic behavior in the investigation areas is assured by appropriate initial and boundary conditions of the area interface. The boundary conditions are realized by above mentioned simulators, the setup and the operation of which are based on small-scale data and mathematical models. The simulation areas include core simulator, steam generator simulators, pump simulators and containment simulator. The steam production and entrainment in a real core during a LOCA are simulated by steam and water injection through the core simulator. 2 - Description of test: Investigation of steam/water flow phenomena at the upper tie plate and in the upper plenum and

  20. Water Pollution Detection Based on Hypothesis Testing in Sensor Networks

    Directory of Open Access Journals (Sweden)

    Xu Luo

    2017-01-01

    Full Text Available Water pollution detection is of great importance in water conservation. In this paper, the water pollution detection problems of the network and of the node in sensor networks are discussed. The detection problems in both cases of the distribution of the monitoring noise being normal and nonnormal are considered. The pollution detection problems are analyzed based on hypothesis testing theory firstly; then, the specific detection algorithms are given. Finally, two implementation examples are given to illustrate how the proposed detection methods are used in the water pollution detection in sensor networks and prove the effectiveness of the proposed detection methods.

  1. Thermo-hydraulic Analysis of a Water-cooled Printed Circuit Heat Exchanger in a Small-scale Nitrogen Loop

    International Nuclear Information System (INIS)

    Kim, Chan Soo; Hong, Sung Deok; Kim, Min Hwan; Shim, Jaesool; Lee, Gyung Dong

    2013-01-01

    The development of high-temperature heat exchangers is very important because of its higher operation temperature and pressure than those of common light water reactors and industrial process plants. In particular, the intermediate heat exchanger is a key-challenged high temperature component in a Very High Temperature gas-cooled Reactor (VHTR). A printed circuit heat exchanger is one of the candidates for an intermediate heat exchanger in a VHTR. The printed circuit heat exchanger (PCHE) was developed and commercialized by HEATRIC. The compactness is better than any other heat exchanger types, because its core matrices are fabricated by diffusion bonding with photo-chemically etched micro-channels. Various tests and analysis have been performed to verify the performance of PCHE. The thermal stress analysis of the high temperature PCHE is necessary to endure the extremely operation condition of IHX. In this study, the thermo-hydraulic analysis for the laboratory-scale PCHE is performed to provide the input data for the boundary conditions of a structural analysis. The results from the first-principal calculation are compared with those from computational fluid dynamics code analysis. COMSOL 4.3a analysis is successfully performed at the uniform pressure drop condition in a set of flow channel stacks. The heat-exchanged region concentrated to the nitrogen inlet cause the uniform mass velocity distribution in the channels, therefore there is little difference between two analytical results

  2. TESTING OF CARBONACEOUS ADSORBENTS FOR REMOVAL OF POLLUTANTS FROM WATER

    Directory of Open Access Journals (Sweden)

    RAISA NASTAS

    2012-03-01

    Full Text Available Testing of carbonaceous adsorbents for removal of pollutants from water. Relevant direction for improving of quality of potable water is application of active carbons at various stages of water treatments. This work includes complex research dealing with testing of a broad spectrum of carbonaceous adsorbents for removal of hydrogen sulfide and nitrite ions from water. The role of the surface functional groups of carbonaceous adsorbents, their acid-basic properties, and the influence of the type of impregnated heteroatom (N, O, or metals (Fe, Cu, Ni, on removal of hydrogen sulfide species and nitrite ions have been researched. The efficiency of the catalyst obtained from peach stones by impregnation with Cu2+ ions of oxidized active carbon was established, being recommended for practical purposes to remove the hydrogen sulfide species from the sulfurous ground waters. Comparative analysis of carbonaceous adsorbents reveals the importance of surface chemistry for oxidation of nitrite ions.

  3. Tritium-gas/water-vapor monitor. Tests and evaluation

    International Nuclear Information System (INIS)

    Jalbert, R.A.

    1982-07-01

    A tritium gas/water-vapor monitor was designed and built by the Health Physics Group at the Los Alamos National Laboratory. In its prototype configuration, the monitor took the shape of two separate instruments: a (total) tritium monitor and a water-vapor monitor. Both instruments were tested and evaluated. The tests of the (total) tritium monitor, basically an improved version of the standard flow-through ion-chamber instrument, are briefly reported here and more completely elsewhere. The tests of the water-vapor monitor indicated that the novel approach used to condense water vapor for scintillation counting has a number of serious drawbacks and that further development of the instrument is unwarranted

  4. Design and Testing of an Educational Water Tunnel

    Science.gov (United States)

    Kosaraju, Srinivas

    2017-11-01

    A new water tunnel is designed and tested for educational and research purposes at Northern Arizona University. The university currently owns an educational wind tunnel with a test section of 12in X 12in X 24in. However, due to limited size of test section and range of Reynolds numbers, its application is currently limited to very few experiments. In an effort to expand the educational and research capabilities, a student team is tasked to design, build and test a water tunnel as a Capstone Senior Design project. The water tunnel is designed to have a test section of 8in X 8in X 36in. and be able to test up to Re = 50E3. Multiple numerical models are used to optimize the flow field inside the test section before building the physical apparatus. The water tunnel is designed to accommodate multiple experiments for drag and lift studies. The built-in die system can deliver up to three different colors to study the streamlines and vortex shedding from the surfaces. During the first phase, a low discharge pump is used to achieve Re = 4E3 to test laminar flows. In the second phase, a high discharge pump will be used to achieve targeted Re = 50E3 to study turbulent flows.

  5. Supercritical water oxidation data acquisition testing. Final report, Volume I

    International Nuclear Information System (INIS)

    1996-11-01

    This report discusses the phase one testing of a data acquisition system for a supercritical water waste oxidation system. The system is designed to destroy a wide range of organic materials in mixed wastes. The design and testing of the MODAR Oxidizer is discussed. An analysis of the optimized runs is included

  6. Tetraethylammonium block of water flux in Aquaporin-1 channels expressed in kidney thin limbs of Henle's loop and a kidney-derived cell line.

    Directory of Open Access Journals (Sweden)

    Pannabecker Thomas L

    2002-03-01

    Full Text Available Abstract Background Aquaporin-1 (AQP1 channels are constitutively active water channels that allow rapid transmembrane osmotic water flux, and also serve as cyclic-GMP-gated ion channels. Tetraethylammonium chloride (TEA; 0.05 to 10 mM was shown previously to inhibit the osmotic water permeability of human AQP1 channels expressed in Xenopus oocytes. The purpose of the present study was to determine if TEA blocks osmotic water flux of native AQP1 channels in kidney, and recombinant AQP1 channels expressed in a kidney derived MDCK cell line. We also demonstrate that TEA does not inhibit the cGMP-dependent ionic conductance of AQP1 expressed in oocytes, supporting the idea that water and ion fluxes involve pharmacologically distinct pathways in the AQP1 tetrameric complex. Results TEA blocked water permeability of AQP1 channels in kidney and kidney-derived cells, demonstrating this effect is not limited to the oocyte expression system. Equivalent inhibition is seen in MDCK cells with viral-mediated AQP1 expression, and in rat renal descending thin limbs of Henle's loops which abundantly express native AQP1, but not in ascending thin limbs which do not express AQP1. External TEA (10 mM does not block the cGMP-dependent AQP1 ionic conductance, measured by two-electrode voltage clamp after pre-incubation of oocytes in 8Br-cGMP (10–50 mM or during application of the nitric oxide donor, sodium nitroprusside (2–4 mM. Conclusions TEA selectively inhibits osmotic water permeability through native and heterologously expressed AQP1 channels. The pathways for water and ions in AQP1 differ in pharmacological sensitivity to TEA, and are consistent with the idea of independent solute pathways within the channel structure. The results confirm the usefulness of TEA as a pharmacological tool for the analysis of AQP1 function.

  7. Natively unstructured loops differ from other loops.

    Directory of Open Access Journals (Sweden)

    Avner Schlessinger

    2007-07-01

    Full Text Available Natively unstructured or disordered protein regions may increase the functional complexity of an organism; they are particularly abundant in eukaryotes and often evade structure determination. Many computational methods predict unstructured regions by training on outliers in otherwise well-ordered structures. Here, we introduce an approach that uses a neural network in a very different and novel way. We hypothesize that very long contiguous segments with nonregular secondary structure (NORS regions differ significantly from regular, well-structured loops, and that a method detecting such features could predict natively unstructured regions. Training our new method, NORSnet, on predicted information rather than on experimental data yielded three major advantages: it removed the overlap between testing and training, it systematically covered entire proteomes, and it explicitly focused on one particular aspect of unstructured regions with a simple structural interpretation, namely that they are loops. Our hypothesis was correct: well-structured and unstructured loops differ so substantially that NORSnet succeeded in their distinction. Benchmarks on previously used and new experimental data of unstructured regions revealed that NORSnet performed very well. Although it was not the best single prediction method, NORSnet was sufficiently accurate to flag unstructured regions in proteins that were previously not annotated. In one application, NORSnet revealed previously undetected unstructured regions in putative targets for structural genomics and may thereby contribute to increasing structural coverage of large eukaryotic families. NORSnet found unstructured regions more often in domain boundaries than expected at random. In another application, we estimated that 50%-70% of all worm proteins observed to have more than seven protein-protein interaction partners have unstructured regions. The comparative analysis between NORSnet and DISOPRED2 suggested

  8. Modernized CDTN's air-water experimental test circuit: initial results

    Energy Technology Data Exchange (ETDEWEB)

    Pessoa, Mácio A.; Sobrinho, Mauricio R. da S.; Salomão, Eduardo A.; Ferreira, Arthur F.J.; Navarro, Moysés A.; Santos, André A. Campagnole dos, E-mail: marcioaraujopessoa@gmail.com, E-mail: mauricio.sobrinho223@gmail.com, E-mail: e.a.salomao@gmail.com, E-mail: arthur1303@gmail.com, E-mail: moysesnavarro@yahoo.com.br, E-mail: aacs@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    The Counter Current Flow Limitation (CCFL) phenomenon, specifically the control that the gas exerts in a liquid flow in the opposite direction, is of real importance in the study of design and operation of various industrial sectors, particularly the nuclear industry. In nuclear engineering, such a phenomenon can occur in a loss of coolant accident (LOCA) of a Pressurized Water Reactor (PWR) when there is the need to re-flood the reactor core during an emergency cooling process. The CCFL phenomenon is being investigated at the Nuclear Technology Development Center (CDTN) thermo-hydraulics laboratory in order to better understand the flow and its limitations and thereby contribute to the improvement of its modeling for analysis of severe accidents. For this, a series of experiments were performed in CDTN in a reduced scale acrylic test section of the 'hot leg' of a PWR. The new proposed circuit is a closed loop and no water has to be discharged during the experiment. This is only possible due to the Python program, which is associated to the data acquisition system and can interface with the automated valves through the outputs of the data acquisition board to control the experiment. The trials compare the CCFL behavior for 500mm lengths of the horizontal section, for inclined duct slope 50° for a diameter of 54mm pipe's diameter. This paper describes the new tests in comparison to tests performed in the past. (author)

  9. Modernized CDTN's air-water experimental test circuit: initial results

    International Nuclear Information System (INIS)

    Pessoa, Mácio A.; Sobrinho, Mauricio R. da S.; Salomão, Eduardo A.; Ferreira, Arthur F.J.; Navarro, Moysés A.; Santos, André A. Campagnole dos

    2017-01-01

    The Counter Current Flow Limitation (CCFL) phenomenon, specifically the control that the gas exerts in a liquid flow in the opposite direction, is of real importance in the study of design and operation of various industrial sectors, particularly the nuclear industry. In nuclear engineering, such a phenomenon can occur in a loss of coolant accident (LOCA) of a Pressurized Water Reactor (PWR) when there is the need to re-flood the reactor core during an emergency cooling process. The CCFL phenomenon is being investigated at the Nuclear Technology Development Center (CDTN) thermo-hydraulics laboratory in order to better understand the flow and its limitations and thereby contribute to the improvement of its modeling for analysis of severe accidents. For this, a series of experiments were performed in CDTN in a reduced scale acrylic test section of the 'hot leg' of a PWR. The new proposed circuit is a closed loop and no water has to be discharged during the experiment. This is only possible due to the Python program, which is associated to the data acquisition system and can interface with the automated valves through the outputs of the data acquisition board to control the experiment. The trials compare the CCFL behavior for 500mm lengths of the horizontal section, for inclined duct slope 50° for a diameter of 54mm pipe's diameter. This paper describes the new tests in comparison to tests performed in the past. (author)

  10. Verification of RELAP5-3D code in natural circulation loop as function of the initial water inventory

    Science.gov (United States)

    Bertani, C.; Falcone, N.; Bersano, A.; Caramello, M.; Matsushita, T.; De Salve, M.; Panella, B.

    2017-11-01

    High safety and reliability of advanced nuclear reactors, Generation IV and Small Modular Reactors (SMR), have a crucial role in the acceptance of these new plants design. Among all the possible safety systems, particular efforts are dedicated to the study of passive systems because they rely on simple physical principles like natural circulation, without the need of external energy source to operate. Taking inspiration from the second Decay Heat Removal system (DHR2) of ALFRED, the European Generation IV demonstrator of the fast lead cooled reactor, an experimental facility has been built at the Energy Department of Politecnico di Torino (PROPHET facility) to study single and two-phase flow natural circulation. The facility behavior is simulated using the thermal-hydraulic system code RELAP5-3D, which is widely used in nuclear applications. In this paper, the effect of the initial water inventory on natural circulation is analyzed. The experimental time behaviors of temperatures and pressures are analyzed. The experimental matrix ranges between 69 % and 93%; the influence of the opposite effects related to the increase of the volume available for the expansion and the pressure raise due to phase change is discussed. Simulations of the experimental tests are carried out by using a 1D model at constant heat power and fixed liquid and air mass; the code predictions are compared with experimental results. Two typical responses are observed: subcooled or two phase saturated circulation. The steady state pressure is a strong function of liquid and air mass inventory. The numerical results show that, at low initial liquid mass inventory, the natural circulation is not stable but pulsated.

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

    International Nuclear Information System (INIS)

    1989-02-01

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

  12. Thermal hydraulic behavior of a PWR under beyond-design-basis accident conditions: Conclusions from an experimental program in a 4-loop test facility (PKL)

    International Nuclear Information System (INIS)

    Umminger, K.J.; Kastner, W.; Mandl, R.M.; Weber, P.

    1993-01-01

    Within the scope of German reactor safety research, extensive experiments covering the behavior of nuclear power plants under accident conditions have been carried out in the PKL test facility which simulates a 4-loop, 1,300 MWe KWU-designed PWR. While the investigations dealing with design-basis accidents and with the efficiency of the emergency core cooling systems have been largely completed, the main interest nowadays concentrates on the investigation of beyond-design-basis accidents to demonstrate the safety margins of nuclear power plants and to investigate the contribution of the built-in safety features for a further reduction of the residual risk. The thermal hydraulic behavior of a PWR under these extreme accident conditions was experimentally investigated within the PKL III B test program. This paper presents the fundamental findings with some of the most important results being discussed in detail. Future plans are also outlined

  13. Power Hardware-in-the-Loop Testing of Multiple Photovoltaic Inverters' Volt-Var Control with Real-Time Grid Model

    Energy Technology Data Exchange (ETDEWEB)

    Chakraborty, Sudipta; Nelson, Austin; Hoke, Anderson

    2016-12-12

    Traditional testing methods fall short in evaluating interactions between multiple smart inverters providing advanced grid support functions due to the fact that such interactions largely depend on their placements on the electric distribution systems with impedances between them. Even though significant concerns have been raised by the utilities on the effects of such interactions, little effort has been made to evaluate them. In this paper, power hardware-in-the-loop (PHIL) based testing was utilized to evaluate autonomous volt-var operations of multiple smart photovoltaic (PV) inverters connected to a simple distribution feeder model. The results provided in this paper show that depending on volt-var control (VVC) parameters and grid parameters, interaction between inverters and between the inverter and the grid is possible in some extreme cases with very high VVC slopes, fast response times and large VVC response delays.

  14. Autonomous Aerial Refueling Ground Test Demonstration—A Sensor-in-the-Loop, Non-Tracking Method

    Directory of Open Access Journals (Sweden)

    Chao-I Chen

    2015-05-01

    Full Text Available An essential capability for an unmanned aerial vehicle (UAV to extend its airborne duration without increasing the size of the aircraft is called the autonomous aerial refueling (AAR. This paper proposes a sensor-in-the-loop, non-tracking method for probe-and-drogue style autonomous aerial refueling tasks by combining sensitivity adjustments of a 3D Flash LIDAR camera with computer vision based image-processing techniques. The method overcomes the inherit ambiguity issues when reconstructing 3D information from traditional 2D images by taking advantage of ready to use 3D point cloud data from the camera, followed by well-established computer vision techniques. These techniques include curve fitting algorithms and outlier removal with the random sample consensus (RANSAC algorithm to reliably estimate the drogue center in 3D space, as well as to establish the relative position between the probe and the drogue. To demonstrate the feasibility of the proposed method on a real system, a ground navigation robot was designed and fabricated. Results presented in the paper show that using images acquired from a 3D Flash LIDAR camera as real time visual feedback, the ground robot is able to track a moving simulated drogue and continuously narrow the gap between the robot and the target autonomously.

  15. Outdoor test method to determine the thermal behavior of solar domestic water heating systems

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Valladares, O.; Pilatowsky, I. [Centro de Investigacion en Energia, Universidad Nacional Autonoma de Mexico, Privada Xochicalco, s/n, Colonia Centro, 62580 Temixco, Morelos (Mexico); Ruiz, V. [Escuela Tecnica Superior de Ingenieros, Universidad de Sevilla, Camino de los Descubrimientos, s/n, Isla de la Cartuja, 41092 Sevilla, Espana (Spain)

    2008-07-15

    The dynamics of the market, the generation of new promotion programs, fiscal incentives and many other factors are to be considered for the massive application of solar domestic water heating systems (SDWHS) mainly of the compact thermosiphon type, makes it necessary to choose simple and inexpensive procedure tests that permit to know their characteristic thermal behaviors without an official standard being necessary. Moreover, it allows the comparison among systems and offers enough and reliable information to consumers and manufacturers. In most developing countries, an official national standard for SDWHS is not available, therefore it is necessary to adopt an international test procedure in which the cost and time of implementation is very important. In this work, a simple and inexpensive test method to determine the thermal behavior of SDWHS is proposed. Even though these procedure tests do not have an official standard structure they permit, by comparing different solar systems under identical solar, ambient, and initial conditions, the experimental determination of: (a) the maximum available volume of water for solar heating; (b) water temperature increment and available thermal energy at the end of the day; (c) temperature profiles (stratification) and the average temperature in the storage tank after it is homogenized; (d) the average global thermal efficiency; (e) water temperature decrement and energy lost overnight; and (f) the relationship between hot water volume and solar collector area as function of the average heating temperature. An additional proposed test permits to know the heat losses caused by the reverse flow in the collector loop. These tests will be carried out independently of the configuration between the solar collector and the storage tank, the way the fluid circulates and the type of thermal exchange. The results of this procedure test can be very useful, firstly, for the local solar manufacturers' equipment in order to design

  16. Outdoor test method to determine the thermal behavior of solar domestic water heating systems

    International Nuclear Information System (INIS)

    Garcia-Valladares, O.; Pilatowsky, I.; Ruiz, V.

    2008-01-01

    The dynamics of the market, the generation of new promotion programs, fiscal incentives and many other factors are to be considered for the massive application of solar domestic water heating systems (SDWHS) mainly of the compact thermosiphon type, makes it necessary to choose simple and inexpensive procedure tests that permit to know their characteristic thermal behaviors without an official standard being necessary. Moreover, it allows the comparison among systems and offers enough and reliable information to consumers and manufacturers. In most developing countries, an official national standard for SDWHS is not available, therefore it is necessary to adopt an international test procedure in which the cost and time of implementation is very important. In this work, a simple and inexpensive test method to determine the thermal behavior of SDWHS is proposed. Even though these procedure tests do not have an official standard structure they permit, by comparing different solar systems under identical solar, ambient, and initial conditions, the experimental determination of: (a) the maximum available volume of water for solar heating; (b) water temperature increment and available thermal energy at the end of the day; (c) temperature profiles (stratification) and the average temperature in the storage tank after it is homogenized; (d) the average global thermal efficiency; (e) water temperature decrement and energy lost overnight; and (f) the relationship between hot water volume and solar collector area as function of the average heating temperature. An additional proposed test permits to know the heat losses caused by the reverse flow in the collector loop. These tests will be carried out independently of the configuration between the solar collector and the storage tank, the way the fluid circulates and the type of thermal exchange. The results of this procedure test can be very useful, firstly, for the local solar manufacturers' equipment in order to design and

  17. Endurance Test and Evaluation of Alkaline Water Electrolysis Cells

    Science.gov (United States)

    Kovach, Andrew J.; Schubert, Franz H.; Chang, B. J.; Larkins, Jim T.

    1985-01-01

    The overall objective of this program is to assess the state of alkaline water electrolysis cell technology and its potential as part of a Regenerative Fuel Cell System (RFCS) of a multikilowatt orbiting powerplant. The program evaluates the endurance capabilities of alkaline electrolyte water electrolysis cells under various operating conditions, including constant condition testing, cyclic testing and high pressure testing. The RFCS demanded the scale-up of existing cell hardware from 0.1 sq ft active electrode area to 1.0 sq ft active electrode area. A single water electrolysis cell and two six-cell modules of 1.0 sq ft active electrode area were designed and fabricated. The two six-cell 1.0 sq ft modules incorporate 1.0 sq ft utilized cores, which allow for minimization of module assembly complexity and increased tolerance to pressure differential. A water electrolysis subsystem was designed and fabricated to allow testing of the six-cell modules. After completing checkout, shakedown, design verification and parametric testing, a module was incorporated into the Regenerative Fuel Cell System Breadboard (RFCSB) for testing at Life Systems, Inc., and at NASA JSC.

  18. Automated corrosion fatigue crack growth testing in pressurized water environments

    International Nuclear Information System (INIS)

    Ceschini, L.J.; Liaw, P.K.; Rudd, G.E.; Logsdon, W.A.

    1984-01-01

    This paper describes in detail a novel approach to construct a test facility for developing corrosion fatigue crack growth rate (FCGR) properties in aggressive environments. The environment studied is that of a pressurized water reactor (PWR) at 288 0 C (550 0 F) and 13.8 MPa (200 psig). To expedite data generation, each chamber was designed to accommodate two test specimens. A common water recirculation and pressurization system was employed to service two test chambers. Thus, four fatigue crack propagation rate tests could be conducted simultaneously in the pressurized water environment. The data analysis was automated to minimize the typically high labor costs associated with corrosion fatigue crack propagation testing. Verification FCGR tests conducted on an ASTM A469 rotor steel in a room temperature air environment as well as actual PWR environment FCGR tests performed on an ASTM A533 Grade B Class 2 pressure vessel steel demonstrated that the dual specimen test facility is an excellent system for developing the FCGR properties of materials in adverse environments

  19. Evaluation of precipitates used in strainer head loss testing: Part III. Long-term aluminum hydroxide precipitation tests in borated water

    International Nuclear Information System (INIS)

    Bahn, Chi Bum; Kasza, Ken E.; Shack, William J.; Natesan, Ken; Klein, Paul

    2011-01-01

    Research highlights: → Aluminum hydroxide precipitation boundary is similar to that for amorphous phase. → Various precipitation tests are combined into one map in temperature-'pH + p[Al] T '. → Flocculation tendency of precipitates depend on pH and total Al concentration. → DLVO theory explains qualitatively the dependency of flocculation tendency on pH. - Abstract: Long-term aluminum (Al) hydroxide precipitation tests were conducted in slightly alkaline solutions containing 2500 ppm boron. The solution temperature was cycled to obtain a temperature history more representative of emergency core cooling system temperatures after a loss-of-coolant accident. The observed Al precipitation boundary was close to predicted results for amorphous precipitates, which are higher than the solubility expected for crystalline forms. Bench-scale and loop head loss test results under various conditions were successfully combined into single map in a temperature - 'pH + p[Al] T ' domain, which yielded two bounding lines for Al hydroxide solubility in borated alkaline water that depend on whether or not loop head loss tests with Al alloy coupons are included. Precipitates were observed to form either as fine, cloudy suspensions, which showed very little tendency to settle, or as flocculated precipitates. The flocculation tendency of the precipitates can be qualitatively explained by a colloid stability theory or a phase diagram for protein solutions.

  20. Direct detection of Mycobacterium avium in environmental water and scale samples by loop-mediated isothermal amplification.

    Science.gov (United States)

    Nishiuchi, Yukiko; Tamaru, Aki; Suzuki, Yasuhiko; Kitada, Seigo; Maekura, Ryoji; Tateishi, Yoshitaka; Niki, Mamiko; Ogura, Hisashi; Matsumoto, Sohkichi

    2014-06-01

    We previously demonstrated the colonization of Mycobacterium avium complex in bathrooms by the conventional culture method. In the present study, we aimed to directly detect M. avium organisms in the environment using loop-mediated isothermal amplification (LAMP), and to demonstrate the efficacy of LAMP by comparing the results with those obtained by culture. Our data showed that LAMP analysis has detection limits of 100 fg DNA/reaction for M. avium. Using an FTA(®) elute card, DNA templates were extracted from environmental samples from bathrooms in the residences of 29 patients with pulmonary M. avium disease. Of the 162 environmental samples examined, 143 (88%) showed identical results by both methods; 20 (12%) and 123 (76%) samples were positive and negative, respectively, for M. avium. Of the remaining 19 samples (12%), seven (5%) and 12 (7%) samples were positive by the LAMP and culture methods, respectively. All samples that contained over 20 colony forming units/primary isolation plate, as measured by the culture method, were also positive by the LAMP method. Our data demonstrate that the combination of the FTA elute card and LAMP can facilitate prompt detection of M. avium in the environment.

  1. Field Evaluation of a High Throughput Loop Mediated Isothermal Amplification Test for the Detection of Asymptomatic Plasmodium Infections in Zanzibar.

    Directory of Open Access Journals (Sweden)

    Berit Aydin-Schmidt

    Full Text Available New field applicable diagnostic tools are needed for highly sensitive detection of residual malaria infections in pre-elimination settings. Field performance of a high throughput DNA extraction system for loop mediated isothermal amplification (HTP-LAMP was therefore evaluated for detecting malaria parasites among asymptomatic individuals in Zanzibar.HTP-LAMP performance was evaluated against real-time PCR on 3008 paired blood samples collected on filter papers in a community-based survey in 2015.The PCR and HTP-LAMP determined malaria prevalences were 1.6% (95%CI 1.3-2.4 and 0.7% (95%CI 0.4-1.1, respectively. The sensitivity of HTP-LAMP compared to PCR was 40.8% (CI95% 27.0-55.8 and the specificity was 99.9% (CI95% 99.8-100. For the PCR positive samples, there was no statistically significant difference between the geometric mean parasite densities among the HTP-LAMP positive (2.5 p/μL, range 0.2-770 and HTP-LAMP negative (1.4 p/μL, range 0.1-7 samples (p = 0.088. Two lab technicians analysed up to 282 samples per day and the HTP-LAMP method was experienced as user friendly.Although field applicable, this high throughput format of LAMP as used here was not sensitive enough to be recommended for detection of asymptomatic low-density infections in areas like Zanzibar, approaching malaria elimination.

  2. Long-term pumping test to study the impact of an open-loop geothermal system on seawater intrusion in a coastal aquifer: the case study of Bari (Southern Italy)

    Science.gov (United States)

    Clementina Caputo, Maria; Masciale, Rita; Masciopinto, Costantino; De Carlo, Lorenzo

    2016-04-01

    The high cost and scarcity of fossil fuels have promoted the increased use of natural heat for a number of direct applications. Just as for fossil fuels, the exploitation of geothermal energy should consider its environmental impact and sustainability. Particular attention deserves the so-called open loop geothermal groundwater heat pump (GWHP) system, which uses groundwater as geothermal fluid. From an economic point of view, the implementation of this kind of geothermal system is particularly attractive in coastal areas, which have generally shallow aquifers. Anyway the potential problem of seawater intrusion has led to laws that restrict the use of groundwater. The scarcity of freshwater could be a major impediment for the utilization of geothermal resources. In this study a new methodology has been proposed. It was based on an experimental approach to characterize a coastal area in order to exploit the low-enthalpy geothermal resource. The coastal karst and fractured aquifer near Bari, in Southern Italy, was selected for this purpose. For the purpose of investigating the influence of an open-loop GWHP system on the seawater intrusion, a long-term pumping test was performed. The test simulated the effects of a prolonged withdrawal on the chemical-physical groundwater characteristics of the studied aquifer portion. The duration of the test was programmed in 16 days, and it was performed with a constant pumping flowrate of 50 m3/h. The extracted water was outflowed into an adjacent artificial channel, by means of a piping system. Water depth, temperature and electrical conductivity of the pumped water were monitored for 37 days, including also some days before and after the pumping duration. The monitored parameters, collected in the pumping and in five observation wells placed 160 m down-gradient with respect to the groundwater flow direction, have been used to estimate different scenarios of the impact of the GWHP system on the seawater intrusion by mean of a

  3. Influence of the loop design of the feedwater- and steam quality in a power plant with pressurized water reactor

    International Nuclear Information System (INIS)

    Bennert, J.; Becher, L.

    1977-01-01

    At nuclear power plants with pressurized water reactors, condensate occurs on the high pressure part of the water-steam circuit, caused by the operation with low steam parameters. The behaviour of the electrolytes which entered into the circuit (solubility, distribution in water and/or steam) shows that these electrolytes (salts) are to be found mainly in the condensate. The insinuated electrolytes are reconcentrated during the common arrangements with 'Small Circuit' - consisting of steam generator, high pressure turbine, water separator, feedwater vessel, and have a negative influence on the feedwater - boiler water - and the steam quality. Remedy is possible by modified arrangements, during which these electrolyte-containing condensates will be treated and traced back into the main circuit. Nevertheless that the efficiency decrease is insignificant and additional efforts are necessary, a change over to these arrangements is recommendable, due to the fact that the feedwater quality, the boiler water quality, the steam quality in front of the turbine, and finally also the operational safety, as well as the availability will be improved. (orig.) [de

  4. 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

    OpenAIRE

    Hwang Bae; Dong Eok Kim; Sung-Uk Ryu; Sung-Jae Yi; Hyun-Sik Park

    2017-01-01

    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 s...

  5. Laboratory effectiveness testing of water-in-oil emulsion breakers

    International Nuclear Information System (INIS)

    Fingas, M.F.; Fieldhouse, B.; Bier, I.; Conrod, D.; Tennyson, E.

    1995-01-01

    The physics and chemistry of water-in-oil emulsions dominate the development of effectiveness tests. Emulsions are variable in stability--this variability is largely dependent on oil type and degree of weathering. These factors complicate the development of a test. Emulsions which have low stability will apparently break easily with chemical emulsion breakers. Broken emulsions will form a foam-like material, called rag, which retains water which is not part of the stable emulsions. Analytical methods used to determine the final stability of the broken or unbroken emulsion were evaluated. Measurements of water content and viscosity measurements show correlation to emulsion stability. Viscosity provides a more reliable measure of emulsion stability but water content measurements are more convenient and are largely used in this study. Twelve tests were developed in the past. Two testing methods have been developed to a usable stage. These tests are described and data using them provided. The effects of mixing time, agent amount, settling time and mixing energy on effectiveness results are presented

  6. The Space Operations Simulation Center (SOSC) and Closed-Loop Hardware Testing for Orion Rendezvous System Design

    Science.gov (United States)

    Milenkovic, Zoran; DSouza, Christopher; Huish, David; Bendle, John; Kibler, Angela

    2012-01-01

    The exploration goals of Orion / MPCV Project will require a mature Rendezvous, Proximity Operations and Docking (RPOD) capability. Ground testing autonomous docking with a next-generation sensor such as the Vision Navigation Sensor (VNS) is a critical step along the path of ensuring successful execution of autonomous RPOD for Orion. This paper will discuss the testing rationale, the test configuration, the test limitations and the results obtained from tests that have been performed at the Lockheed Martin Space Operations Simulation Center (SOSC) to evaluate and mature the Orion RPOD system. We will show that these tests have greatly increased the confidence in the maturity of the Orion RPOD design, reduced some of the latent risks and in doing so validated the design philosophy of the Orion RPOD system. This paper is organized as follows: first, the objectives of the test are given. Descriptions of the SOSC facility, and the Orion RPOD system and associated components follow. The details of the test configuration of the components in question are presented prior to discussing preliminary results of the tests. The paper concludes with closing comments.

  7. 78 FR 2340 - Energy Conservation Program: Test Procedures for Residential Water Heaters and Commercial Water...

    Science.gov (United States)

    2013-01-11

    ... corresponded to the levels in the American Society of Heating, Refrigerating and Air Conditioning Engineers... provides a brief history of DOE's more recent test procedure rulemakings related to residential water... performance (e.g., such as ambient air temperature, ambient relative humidity, and inlet water temperature...

  8. An Approach for Systematic In-the-Loop Simulations for Development and Test of a Complex Mechatronic Embedded System

    NARCIS (Netherlands)

    Soltani Nehzad, Amir; Lukkien, Johan J.; Mak, Rudolf H.; Verhoeven, Richard; van den Heuvel, Martijn M.H.P.; Skavhaug, A.; Guiochet, J.; Schoitsch, E.; Bitsch, F.

    2016-01-01

    Simulations are widely used in the engineering workflow of complex mechatronic embedded systems in various domains, such as healthcare, railway, automotive and aerospace, for analyzing, testing and validating purposes. This paper focuses on the development and test of the control software of complex

  9. Test Results From a Pair of 1-kWe Dual-Opposed Free-Piston Stirling Power Convertors Integrated With a Pumped NaK Loop

    Science.gov (United States)

    Geng, Steven M.; Briggs, Maxwell H.; Penswick, L. Barry; Pearson, J. Boise; Godfroy, Thomas J.

    2011-01-01

    As a step towards development of Stirling power conversion for potential use in Fission Surface Power (FSP) systems, a pair of commercially available 1-kW-class free-piston Stirling convertors were modified to operate with a NaK (sodium (Na) and potassium (K)) liquid metal pumped loop for thermal energy input. This was the first-ever attempt at powering a free-piston Stirling engine with a pumped liquid metal heat source and is a major FSP project milestone towards demonstrating technical feasibility. The convertors were successfully tested at the Marshall Space Flight Center (MSFC) from June 6 through July 14, 2009. The convertors were operated for a total test time of 66 hr and 16 min. The tests included (a) performance mapping the convertors over various hot- and cold-end temperatures, piston amplitudes, and NaK flow rates and (b) transient test conditions to simulate various startup (i.e., low-, medium-, and high-temperature startups) and fault scenarios (i.e., loss of heat source, loss of NaK pump, convertor stall, etc.). This report documents the results of this testing

  10. A method of calculating fission gas diffusion from UO{sub 2} fuel and its application to the X-2-f loop test

    Energy Technology Data Exchange (ETDEWEB)

    Booth, A H

    1957-09-15

    A method for calculating the fraction of the rare gas fission products that diffuses out of a UO{sub 2} fuel element under conditions In a reactor is outlined, The method is based on the values of the diffusion constant found in laboratory experiments, as described In CRDC-718, and assumes that these remain unaltered during the period that the fuel is in the reactor, The method has been applied to two types of oxide in the X-2-f loop test of 1956 and the results compared with the amounts of fission gas found by analysis of the gases collected in sheath puncture experiments, as described in CRDC-719. The calculated values depend heavily on the estimated temperatures In the fuel. They are in close agreement with the experimental values provided that, in calculating the temperature, certain assumptions are made regarding the thermal expansion of the oxide cylinder. (author)

  11. Escherichia coli. A sanitary methodology for faecal water pollution tests

    International Nuclear Information System (INIS)

    Bonadonna, L.

    2001-01-01

    Among the traditional indictors of faecal water pollution, Escherichia coli has shown to fit better with the definition of indicator organism. Till now its recovery has been time-consuming and needs confirmation tests. In this report more rapid and direct methods, based on enzymatic reactions, are presented [it

  12. Artificial enzyme-powered microfish for water-quality testing.

    Science.gov (United States)

    Orozco, Jahir; García-Gradilla, Victor; D'Agostino, Mattia; Gao, Wei; Cortés, Allan; Wang, Joseph

    2013-01-22

    We present a novel micromotor-based strategy for water-quality testing based on changes in the propulsion behavior of artificial biocatalytic microswimmers in the presence of aquatic pollutants. The new micromotor toxicity testing concept mimics live-fish water testing and relies on the toxin-induced inhibition of the enzyme catalase, responsible for the biocatalytic bubble propulsion of tubular microengines. The locomotion and survival of the artificial microfish are thus impaired by exposure to a broad range of contaminants, that lead to distinct time-dependent irreversible losses in the catalase activity, and hence of the propulsion behavior. Such use of enzyme-powered biocompatible polymeric (PEDOT)/Au-catalase tubular microengine offers highly sensitive direct optical visualization of changes in the swimming behavior in the presence of common contaminants and hence to a direct real-time assessment of the water quality. Quantitative data on the adverse effects of the various toxins upon the swimming behavior of the enzyme-powered artificial swimmer are obtained by estimating common ecotoxicological parameters, including the EC(50) (exposure concentration causing 50% attenuation of the microfish locomotion) and the swimmer survival time (lifetime expectancy). Such novel use of artificial microfish addresses major standardization and reproducibility problems as well as ethical concerns associated with live-fish toxicity assays and hence offers an attractive alternative to the common use of aquatic organisms for water-quality testing.

  13. Acceptance test procedure for High Pressure Water Jet System

    International Nuclear Information System (INIS)

    Crystal, J.B.

    1995-01-01

    The overall objective of the acceptance test is to demonstrate a combined system. This includes associated tools and equipment necessary to perform cleaning in the 105 K East Basin (KE) for achieving optimum reduction in the level of contamination/dose rate on canisters prior to removal from the KE Basin and subsequent packaging for disposal. Acceptance tests shall include necessary hardware to achieve acceptance of the cleaning phase of canisters. This acceptance test procedure will define the acceptance testing criteria of the high pressure water jet cleaning fixture. The focus of this procedure will be to provide guidelines and instructions to control, evaluate and document the acceptance testing for cleaning effectiveness and method(s) of removing the contaminated surface layer from the canister presently identified in KE Basin. Additionally, the desired result of the acceptance test will be to deliver to K Basins a thoroughly tested and proven system for underwater decontamination and dose reduction. This report discusses the acceptance test procedure for the High Pressure Water Jet

  14. Cold leg condensation tests. Task C. Steam--water interaction tests

    International Nuclear Information System (INIS)

    Brodrick, J.R.; Loiselle, V.

    1974-03-01

    A report is presented of tests to determine the condensation efficiency of ECC water injected into a quality fluid mixture flowing through the cold leg. In particular, a specific objective was to determine if the mixture of ECC water and quality fluid reached thermodynamic equilibrium before exiting the cold leg. Further, the stability of the ECC water/quality fluid interaction would be assessed by interpretation of thermocouple records and utilization of a section of cold leg piping with view ports to film the interaction whenever possible. The cold leg condensation tests showed complete condensation of the 5 lbm/sec steam quality mixtures in the cold leg by the ECC water flows of the test matrix. The cold leg exit fluid temperature remained below the saturation temperature and had good agreement with the predicted cold leg outlet temperature, calculated assuming total condensation. (U.S.)

  15. Orion Ground Test Article Water Impact Tests: Photogrammetric Evaluation of Impact Conditions

    Science.gov (United States)

    Vassilakos, Gregory J.; Mark, Stephen D.

    2018-01-01

    The Ground Test Article (GTA) is an early production version of the Orion Crew Module (CM). The structural design of the Orion CM is being developed based on LS-DYNA water landing simulations. As part of the process of confirming the accuracy of LS-DYNA water landing simulations, the GTA water impact test series was conducted at NASA Langley Research Center (LaRC) to gather data for comparison with simulations. The simulation of the GTA water impact tests requires the accurate determination of the impact conditions. To accomplish this, the GTA was outfitted with an array of photogrammetry targets. The photogrammetry system utilizes images from two cameras with a specialized tracking software to determine time histories for the 3-D coordinates of each target. The impact conditions can then be determined from the target location data.

  16. High power testing of the multiple-loop radio-frequency drive concept for the FMIT accelerator

    International Nuclear Information System (INIS)

    Fazio, M.V.; Patton, R.D.

    1984-04-01

    The Fusion Materials Irradiation Test (FMIT) accelerator requires several 600-kW rf systems to simultaneously supply rf power to a single accelerator tank. Each rf-system output must be carefully phase and amplitude controlled to achieve the proper system performance. Two 80-MHz, 600-kW rf amplifiers with phase- and amplitude-control systems have been tested into a single, high-Q resonant cavity. Experimental results are presented

  17. Water corrosion test of oxide dispersion strengthened (ODS) steel claddings

    International Nuclear Information System (INIS)

    Narita, Takeshi; Ukai, Shigeharu; Kaito, Takeji; Ohtsuka, Satoshi; Matsuda, Yasushi

    2006-07-01

    As a part of feasibility study of ODS steel cladding, its water corrosion resistance was examined under water pool condition. Although addition of Cr is effective for preventing water corrosion, excessive Cr addition leads to embrittlement due to the Cr-rich α' precipitate formation. In the ODS steel developed by the Japan Atomic Energy Agency (JAEA), the Cr content is controlled in 9Cr-ODS martensite and 12Cr-ODS ferrite. In this study, water corrosion test was conducted for these ODS steels, and their results were compared with that of conventional austenitic stainless steel and ferritic-martensitic stainless steel. Following results were obtained in this study. (1) Corrosion rate of 9Cr-ODS martensitic and 12Cr-ODS ferritic steel are significantly small and no pitting was observed. Thus, these ODS steels have superior resistance for water corrosion under the condition of 60degC and pH8-12. (2) It was showed that 9Cr-ODS martensitic steel and 12Cr-ODS ferritic steel have comparable water corrosion resistance to that of PNC316 and PNC-FMS at 60degC for 1,000h under varying pH of 8, 10. Water corrosion resistance of these alloys is slightly larger than that of PNC316 and PNC-FMS at pH12 without significant difference of appearance and uneven condition. (author)

  18. Design requirements for the supercritical water oxidation test bed

    International Nuclear Information System (INIS)

    Svoboda, J.M.; Valentich, D.J.

    1994-05-01

    This report describes the design requirements for the supercritical water oxidation (SCWO) test bed that will be located at the Idaho National Engineering Laboratory (INEL). The test bed will process a maximum of 50 gph of waste plus the required volume of cooling water. The test bed will evaluate the performance of a number of SCWO reactor designs. The goal of the project is to select a reactor that can be scaled up for use in a full-size waste treatment facility to process US Department of Energy mixed wastes. EG ampersand G Idaho, Inc. will design and construct the SCWO test bed at the Water Reactor Research Test Facility (WRRTF), located in the northern region of the INEL. Private industry partners will develop and provide SCWO reactors to interface with the test bed. A number of reactor designs will be tested, including a transpiring wall, tube, and vessel-type reactor. The initial SCWO reactor evaluated will be a transpiring wall design. This design requirements report identifies parameters needed to proceed with preliminary and final design work for the SCWO test bed. A flow sheet and Process and Instrumentation Diagrams define the overall process and conditions of service and delineate equipment, piping, and instrumentation sizes and configuration Codes and standards that govern the safe engineering and design of systems and guidance that locates and interfaces test bed hardware are provided. Detailed technical requirements are addressed for design of piping, valves, instrumentation and control, vessels, tanks, pumps, electrical systems, and structural steel. The approach for conducting the preliminary and final designs and environmental and quality issues influencing the design are provided

  19. Sodium-water reaction test facility (SWAT-3)

    International Nuclear Information System (INIS)

    Shimazu, Hisashi; Ukechi, Kazutoshi; Sasakura, Kazutake; Kusunoki, Junichi

    1976-01-01

    In the development of the liquid metal cooled fast breeder reactor (LMFBR), the steam generator (SG) is considered one of the most important components. The Power Reactor and Nuclear Fuel Development Corporation (PNC) is now promoting the research and development of the SG system used with the prototype fast breeder reactor ''Monju''. In this research, the phenomena of the sodium-water reaction in the SG are the key which must be investigated for the solution of problems. The test facility (SWAT-3) simulating Monju's SG on the scale of 1/2.5 was designed, fabricated and installed by IHI at Oarai Engineering Center of PNC, its pre-operation being accomplished in February 1975. The purpose of SWAT-3 is summarized as follows: (1) To perform an overall test on the safety of Monju's SG and intermediate heat transport system under the design condition against sodium-water reaction accidents. (2) To investigate the damage of the SG structure caused by the sodium-water reaction, and the possibility of repair and recovery operations. The first test was accomplished successfully on June 9, 1975. As a result of the test, the fundamental function of this test facility was proven to be satisfactory as expected. (auth.)

  20. HDR flood-water storage-tank modal vibration tests

    International Nuclear Information System (INIS)

    Gorman, V.W.; Thinnes, G.L.

    1983-01-01

    Modal vibration tests were conducted by EG and G Idaho on two vessels located at West Germany's Heissdampfreaktor (HDR) facility which is 25 kilometers east of Frankfurt. The tests were performed during May and June 1982 for the US Nuclear Regulatory Commission (NRC) as part of their cooperative effort with Kernforschungszentrum Karlsruhe (KfK) of West Germany. The primary purpose for performing this task was to determine modal properties (frequencies, mode shapes and associated damping ratios) in order to eventually provide guidelines for standards development by the NRC in modeling similar vessels. One of the vessels tested was a flood water storage tank (FWST) for empty, half full and full water conditions. The FWST was excited randomly with an electromagnetic shaker and by impulsive hammer blows. Excitation or input forces together with measured vessel responses were processed by a digital modal analyzer and stored on magnetic disks for subsequent evaluation

  1. Endurance test and evaluation of alkaline water electrolysis cells

    Science.gov (United States)

    Burke, K. A.; Schubert, F. H.

    1981-01-01

    Utilization in the development of multi-kW low orbit power systems is discussed. The following technological developments of alkaline water electrolysis cells for space power application were demonstrated: (1) four 92.9 cm2 single water electrolysis cells, two using LST's advanced anodes and two using LST's super anodes; (2) four single cell endurance test stands for life testing of alkaline water electrolyte cells; (3) the solid performance of the advanced electrode and 355 K; (4) the breakthrough performance of the super electrode; (5) the four single cells for over 5,000 hours each significant cell deterioration or cell failure. It is concluded that the static feed water electrolysis concept is reliable and due to the inherent simplicity of the passive water feed mechanism coupled with the use of alkaline electrolyte has greater potential for regenerative fuel cell system applications than alternative electrolyzers. A rise in cell voltage occur after 2,000-3,000 hours which was attributed to deflection of the polysulfone end plates due to creepage of the thermoplastic. More end plate support was added, and the performance of the cells was restored to the initial performance level.

  2. Test results of sodium-water reaction testing in near prototypical LMR steam generator

    International Nuclear Information System (INIS)

    Boardman, C.E.; Hui, M.; Neely, H.H.

    1990-01-01

    An extensive test program has been performed in the United States to investigate the effects of large sodium-water reaction events in LMFBR steam generators. Tests were conducted in the Large Leak Test Rig (LLTR) located at the Energy Technology Engineering Center (ETEC). The program was divided into two phases, Series I and Series II, for the purpose of satisfying near-term and long-term needs. Series II was further subdivided into large and intermediate leak tests. This paper will emphasize the Series II intermediate leak tests and resulting conclusions for steam generator design and operation. 11 figs, 2 tabs

  3. Application of a Virtual Reactivity Feedback Control Loop in Non-Nuclear Testing of a Fast Spectrum Reactor

    International Nuclear Information System (INIS)

    Bragg-Sitton, Shannon M.; Forsbacka, Matthew

    2004-01-01

    For a compact, fast-spectrum reactor, reactivity feedback is dominated by core deformation at elevated temperature. Given the use of accurate deformation measurement techniques, it is possible to simulate nuclear feedback in non-nuclear electrically heated reactor tests. Implementation of simulated reactivity feedback in response to measured deflection is being tested at the Nasa Marshall Space Flight Center Early Flight Fission Test Facility (EFF-TF). During tests of the SAFE-100 reactor prototype, core deflection was monitored using a high resolution camera. 'Virtual' reactivity feedback was accomplished by applying the results of Monte Carlo calculations (MCNPX) to core deflection measurements; the computational analysis was used to establish the reactivity worth of various core deformations. The power delivered to the SAFE-100 prototype was then adjusted accordingly via kinetics calculations. The work presented in this paper will demonstrate virtual reactivity feedback as core power was increased from 1 kWt to 10 kWt, held approximately constant at 10 kWt, and then allowed to decrease based on the negative thermal reactivity coefficient. (authors)

  4. Supercritical water oxidation benchscale testing metallurgical analysis report

    International Nuclear Information System (INIS)

    Norby, B.C.

    1993-02-01

    This report describes metallurgical evaluation of witness wires from a series of tests using supercritical water oxidation (SCWO) to process cutting oil containing a simulated radionuclide. The goal of the tests was to evaluate the technology's ability to process a highly chlorinated waste representative of many mixed waste streams generated in the DOE complex. The testing was conducted with a bench-scale SCWO system developed by the Modell Development Corporation. Significant test objectives included process optimization for adequate destruction efficiency, tracking the radionuclide simulant and certain metals in the effluent streams, and assessment of reactor material degradation resulting from processing a highly chlorinated waste. The metallurgical evaluation described herein includes results of metallographic analysis and Scanning Electron Microscopy analysis of witness wires exposed to the SCWO environment for one test series

  5. Alternative loop rings

    CERN Document Server

    Goodaire, EG; Polcino Milies, C

    1996-01-01

    For the past ten years, alternative loop rings have intrigued mathematicians from a wide cross-section of modern algebra. As a consequence, the theory of alternative loop rings has grown tremendously. One of the main developments is the complete characterization of loops which have an alternative but not associative, loop ring. Furthermore, there is a very close relationship between the algebraic structures of loop rings and of group rings over 2-groups. Another major topic of research is the study of the unit loop of the integral loop ring. Here the interaction between loop rings and group ri

  6. Maintenance of immersion ultrasonic testing on the water tube boiler

    International Nuclear Information System (INIS)

    Ishiyama, Toru; Kawasaki, Ichio; Miura, Hirohito

    2014-01-01

    There are 4-boiler in nuclear fuel cycle engineering laboratories (NCL). These boilers have been operated in the long term over 20 years. One of them, the leakage of boiler water was found at one of the generating tubes, and 2 adjoining generating tubes were corroded in Dec, 2011. These generating tubes were investigated by immersion ultrasonic testing (UT) for measure thickness of the tube. As a result, thinner tube was found in a part of a bend and near the water drum. These parts are covered with sulfide deposit, it seems that the generating tubes were corroded by sulfide. (author)

  7. Retrofitting Combined Space and Water Heating Systems: Laboratory Tests

    Energy Technology Data Exchange (ETDEWEB)

    Schoenbauer, B.; Bohac, D.; Huelman, P.; Olson, R.; Hewitt, M.

    2012-10-01

    Better insulated and tighter homes can often use a single heating plant for both space and domestic water heating. These systems, called dual integrated appliances (DIA) or combination systems, can operate at high efficiency and eliminate combustion safety issues associated by using a condensing, sealed combustion heating plant. Funds were received to install 400 DIAs in Minnesota low-income homes. The NorthernSTAR DIA laboratory was created to identify proper system components, designs, operating parameters, and installation procedures to assure high efficiency of field installed systems. Tests verified that heating loads up to 57,000 Btu/hr can be achieved with acceptable return water temperatures and supply air temperatures.

  8. Retrofitting Combined Space and Water Heating Systems. Laboratory Tests

    Energy Technology Data Exchange (ETDEWEB)

    Schoenbauer, B. [NorthernStar Building America Partnership, St. Paul, MN (United States); Bohac, D. [NorthernStar Building America Partnership, St. Paul, MN (United States); Huelman, P. [NorthernStar Building America Partnership, St. Paul, MN (United States); Olsen, R. [NorthernStar Building America Partnership, St. Paul, MN (United States); Hewett, M. [NorthernStar Building America Partnership, St. Paul, MN (United States)

    2012-10-01

    Better insulated and tighter homes can often use a single heating plant for both space and domestic water heating. These systems, called dual integrated appliances (DIA) or combination systems, can operate at high efficiency and eliminate combustion safety issues associated by using a condensing, sealed combustion heating plant. Funds were received to install 400 DIAs in Minnesota low-income homes. The NorthernSTAR DIA laboratory was created to identify proper system components, designs, operating parameters, and installation procedures to assure high efficiency of field installed systems. Tests verified that heating loads up to 57,000 Btu/hr can be achieved with acceptable return water temperatures and supply air temperatures.

  9. Exceptional F(4) higher-spin theory in AdS{sub 6} at one-loop and other tests of duality

    Energy Technology Data Exchange (ETDEWEB)

    Günaydin, Murat [Institute for Gravitation and the Cosmos Physics Department, Pennsylvania State University, University Park, PA 16802 (United States); Skvortsov, Evgeny [Arnold Sommerfeld Center for Theoretical Physics, Ludwig-Maximilians University Munich, Theresienstr. 37, D-80333 Munich (Germany); Lebedev Institute of Physics, Leninsky ave. 53, 119991 Moscow (Russian Federation); Tran, Tung [Department of Physics, Brown University, Providence, Rhode Island 02912 (United States)

    2016-11-28

    We study the higher-spin gauge theory in six-dimensional anti-de Sitter space AdS{sub 6} that is based on the exceptional Lie superalgebra F(4). The relevant higher-spin algebra was constructed in http://arxiv.org/abs/1409.2185. We determine the spectrum of the theory and show that it contains the physical fields of the Romans F(4) gauged supergravity. The full spectrum consists of an infinite tower of unitary supermultiplets of F(4) which extend the Romans multiplet to higher spins plus a single short supermultiplet. Motivated by applications to this novel supersymmetric higher-spin theory as well as to other theories, we extend the known one-loop tests of AdS/CFT duality in various directions. The spectral zeta-function is derived for the most general case of fermionic and mixed-symmetry fields, which allows one to test the Type-A and B theories and supersymmetric extensions thereof in any dimension. We also study higher-spin doubletons and partially-massless fields. While most of the tests are successfully passed, the Type-B theory in all even dimensional anti-de Sitter spacetimes presents an interesting puzzle: the free energy as computed from the bulk is not equal to that of the free fermion on the CFT side, though there is some systematics to the discrepancy.

  10. Statistical modeling of dental unit water bacterial test kit performance.

    Science.gov (United States)

    Cohen, Mark E; Harte, Jennifer A; Stone, Mark E; O'Connor, Karen H; Coen, Michael L; Cullum, Malford E

    2007-01-01

    While it is important to monitor dental water quality, it is unclear whether in-office test kits provide bacterial counts comparable to the gold standard method (R2A). Studies were conducted on specimens with known bacterial concentrations, and from dental units, to evaluate test kit accuracy across a range of bacterial types and loads. Colony forming units (CFU) were counted for samples from each source, using R2A and two types of test kits, and conformity to Poisson distribution expectations was evaluated. Poisson regression was used to test for effects of source and device, and to estimate rate ratios for kits relative to R2A. For all devices, distributions were Poisson for low CFU/mL when only beige-pigmented bacteria were considered. For higher counts, R2A remained Poisson, but kits exhibited over-dispersion. Both kits undercounted relative to R2A, but the degree of undercounting was reasonably stable. Kits did not grow pink-pigmented bacteria from dental-unit water identified as Methylobacterium rhodesianum. Only one of the test kits provided results with adequate reliability at higher bacterial concentrations. Undercount bias could be estimated for this device and used to adjust test kit results. Insensitivity to methylobacteria spp. is problematic.

  11. Review of design analysis and site installation method of ASME vessel of fuel test loop of Hanaro

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Sung Won; Kim, Jun Yeon

    1997-02-01

    The major goal of this report is to take the advantages under control, and provide basic solutions for field installation. In order to access to technical availability, the scope, limitation, and possibilities of design requirements are carefully considered in this. The chapter 3 deals with seismic stress analysis of vessels, using manufacturers` finite element analysis data. The test of manufactured equipment is scheduled to hold near future. The evaluation criteria and inspection specifications for the quality release are well illustrated in the chapter IV to efficiently check out the items. The full considerations for field installation are also reviewed with that in terms of space limit. Following the inspection and test of manufacturer`s shop, the results will be promptly reported. Based on installation principles and the analysis in the report, updated procedure and methodology for the installation will be applied to the field in more details and better breakdown precision in the coming years. (author). 14 refs., 23 tabs., 26 figs.

  12. Review of design analysis and site installation method of ASME vessel of fuel test loop of Hanaro

    International Nuclear Information System (INIS)

    Cho, Sung Won; Kim, Jun Yeon.

    1997-02-01

    The major goal of this report is to take the advantages under control, and provide basic solutions for field installation. In order to access to technical availability, the scope, limitation, and possibilities of design requirements are carefully considered in this. The chapter 3 deals with seismic stress analysis of vessels, using manufacturers' finite element analysis data. The test of manufactured equipment is scheduled to hold near future. The evaluation criteria and inspection specifications for the quality release are well illustrated in the chapter IV to efficiently check out the items. The full considerations for field installation are also reviewed with that in terms of space limit. Following the inspection and test of manufacturer's shop, the results will be promptly reported. Based on installation principles and the analysis in the report, updated procedure and methodology for the installation will be applied to the field in more details and better breakdown precision in the coming years. (author). 14 refs., 23 tabs., 26 figs

  13. Closing the feedback loop: engaging students in large first-year mathematics test revision sessions using pen-enabled screens

    Science.gov (United States)

    Donovan, Diane; Loch, Birgit

    2013-01-01

    How can active learning, peer learning and prompt feedback be achieved in large first-year mathematics classes? Further, what technologies may support these aims? In this article, we assert that test revision sessions in first-year mathematics held in a technology-enhanced lecture theatre can be highly interactive with students solving problems, learning from each other and receiving immediate feedback. This is facilitated by pen-enabled screens and synchronization software. We argue that the educational benefits achievable through the technology do outweigh the technological distractions, and that these benefits can be achieved by focused, targeted one-off sessions and not only by a semester-long, regular approach. Repeat mid-semester test revision sessions were offered on a non-compulsory basis using pen-enabled screens for all students. Students worked practice test questions and marked solutions to mathematical problems on the screens. Students' work was then displayed anonymously for their peers to see. Answers were discussed with the whole class. We discuss outcomes from two offerings of these sessions using student feedback and lecturer reflections and show the impact of participation on self-reported student confidence. Pedagogical approaches that the technology allowed for the first time in a large class are highlighted. Students responded uniformly positively.

  14. The adsorber loop concept for the contact between seawater and adsorber granulate

    International Nuclear Information System (INIS)

    Koske, P.H.; Ohlrogge, K.

    1984-01-01

    The present paper deals with the so-called ''adsorber loop concept'' in which the adsorber granulate is carried along with the seawater to be processed in a loop-like configuration and is separated again from the depleted water before this is leaving the adsorption unit. This concept enables high seawater velocities thus reducing the required bed area. Theoretical considerations are presented together with experimental results from field tests. (orig.) [de

  15. A pilot solar water disinfecting system: performance analysis and testing

    Energy Technology Data Exchange (ETDEWEB)

    Saitoh, T.S.; El-Ghetany, H.H. [Tohoku University, Sendai (Japan). Dept. of Aeronautics and Space Engineering

    2002-07-01

    In most countries, contaminated water is the major cause of most water-borne diseases. Disinfection of water may be accomplished by a number of different physical-chemical treatments including direct application of thermal energy, chemical and filtration techniques. Solar energy also can be used effectively in this field because inactivation of microorganisms is done either by heating water to a disinfecting temperature or by exposing it to ultraviolet solar radiation. A pilot solar system for disinfecting contaminated water is designed, constructed and tested. Investigations are carried out to evaluate the performance of a wooden hot box solar facility as a solar disinfectant. Experimental data show that solar energy is viable for the disinfection process. A solar radiation model is presented and compared with the experimental data. A mathematical model of the solar disinfectant is also presented. The governing equations are solved numerically via the fourth-order Runge-Kutta method. The effects of environmental conditions (ambient temperature, wind speed, solar radiation, etc.) on the performance of the solar disinfectant are examined. Results showed that the system is affected by ambient temperature, wind speed, ultraviolet solar radiation intensity, the turbidity of the water, the quantity of water exposed, the contact area between the transparent water container in the solar disinfectant and the absorber plate as well as the geometrical parameters of the system. It is pointed out that for partially cloudy conditions with a low ambient temperature and high wind speeds, the thermal efficiency of the solar disinfectant is at a minimum. The use of solar energy for the disinfection process will increase the productivity of the system while completely eliminating the coliform group bacteria at the same time. (author)

  16. Rogowski Loop design for NSTX

    International Nuclear Information System (INIS)

    McCormack, B.; Kaita, R.; Kugel, H.; Hatcher, R.

    2000-01-01

    The Rogowski Loop is one of the most basic diagnostics for tokamak operations. On the National Spherical Torus Experiment (NSTX), the plasma current Rogowski Loop had the constraints of the very limited space available on the center stack, 5,000 volt isolation, flexibility requirements as it remained a part of the Center Stack assembly after the first phase of operation, and a +120 C temperature requirement. For the second phase of operation, four Halo Current Rogowski Loops under the Center Stack tiles will be installed having +600 C and limited space requirements. Also as part of the second operational phase, up to ten Rogowski Loops will installed to measure eddy currents in the Passive Plate support structures with +350 C, restricted space, and flexibility requirements. This presentation will provide the details of the material selection, fabrication techniques, testing, and installation results of the Rogowski Loops that were fabricated for the high temperature operational and bakeout requirements, high voltage isolation requirements, and the space and flexibility requirements imposed upon the Rogowski