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)

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.

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

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.

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

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

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

Hanford Tank Farms Waste Feed Flow Loop Phase VI: PulseEcho System Performance Evaluation

Energy Technology Data Exchange (ETDEWEB)

Denslow, Kayte M.; Bontha, Jagannadha R.; Adkins, Harold E.; Jenks, Jeromy WJ; Hopkins, Derek F.

2012-11-21

This document presents the visual and ultrasonic PulseEcho critical velocity test results obtained from the System Performance test campaign that was completed in September 2012 with the Remote Sampler Demonstration (RSD)/Waste Feed Flow Loop cold-test platform located at the Monarch test facility in Pasco, Washington. This report is intended to complement and accompany the report that will be developed by WRPS on the design of the System Performance simulant matrix, the analysis of the slurry test sample concentration and particle size distribution (PSD) data, and the design and construction of the RSD/Waste Feed Flow Loop cold-test platform.

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

Uranyl Nitrate Flow Loop

International Nuclear Information System (INIS)

Ladd-Lively, Jennifer L

2008-01-01

The objectives of the work discussed in this report were to: (1) develop a flow loop that would simulate the purified uranium-bearing aqueous stream exiting the solvent extraction process in a natural uranium conversion plant (NUCP); (2) develop a test plan that would simulate normal operation and disturbances that could be anticipated in an NUCP; (3) use the flow loop to test commercially available flowmeters for use as safeguards monitors; and (4) recommend a flowmeter for production-scale testing at an NUCP. There has been interest in safeguarding conversion plants because the intermediate products [uranium dioxide (UO 2 ), uranium tetrafluoride (UF 4 ), and uranium hexafluoride (UF 6 )] are all suitable uranium feedstocks for producing special nuclear materials. Furthermore, if safeguards are not applied virtually any nuclear weapons program can obtain these feedstocks without detection by the International Atomic Energy Agency (IAEA). Historically, IAEA had not implemented safeguards until the purified UF 6 product was declared as feedstock for enrichment plants. H. A. Elayat et al. provide a basic definition of a safeguards system: 'The function of a safeguards system on a chemical conversion plant is in general terms to verify that no useful nuclear material is being diverted to use in a nuclear weapons program'. The IAEA now considers all highly purified uranium compounds as candidates for safeguarding. DOE is currently interested in 'developing instruments, tools, strategies, and methods that could be of use to the IAEA in the application of safeguards' for materials found in the front end of the nuclear fuel cycle-prior to the production of the uranium hexafluoride or oxides that have been the traditional starting point for IAEA safeguards. Several national laboratories, including Oak Ridge, Los Alamos, Lawrence Livermore, and Brookhaven, have been involved in developing tools or techniques for safeguarding conversion plants. This study was sponsored by

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.

CFD and experimental data of closed-loop wind tunnel flow

Directory of Open Access Journals (Sweden)

John Kaiser Calautit

2016-06-01

Full Text Available The data presented in this article were the basis for the study reported in the research articles entitled ‘A validated design methodology for a closed loop subsonic wind tunnel’ (Calautit et al., 2014 [1], which presented a systematic investigation into the design, simulation and analysis of flow parameters in a wind tunnel using Computational Fluid Dynamics (CFD. The authors evaluated the accuracy of replicating the flow characteristics for which the wind tunnel was designed using numerical simulation. Here, we detail the numerical and experimental set-up for the analysis of the closed-loop subsonic wind tunnel with an empty test section.

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

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.

Flow rate and temperature characteristics in steady state condition on FASSIP-01 loop during commissioning

Science.gov (United States)

Juarsa, M.; Giarno; Rohman, A. N.; Heru K., G. B.; Witoko, J. P.; Sony Tjahyani, D. T.

2018-02-01

The need for large-scale experimental facilities to investigate the phenomenon of natural circulation flow rate becomes a necessity in the development of nuclear reactor safety management. The FASSIP-01 loop has been built to determine the natural circulation flow rate performance in the large-scale media and aimed to reduce errors in the results for its application in the design of new generation reactors. The commissioning needs to be done to define the capability of the FASSIP-01 loop and to prescribe the experiment limitations. On this commissioning, two scenarios experimental method has been used. The first scenario is a static condition test which was conducted to verify measurement system response during 24 hours without electrical load in heater and cooler, there is water and no water inside the rectangular loop. Second scenario is a dynamics condition that aims to understand the flow rate, a dynamic test was conducted using heater power of 5627 watts and coolant flow rate in the HSS loop of 9.35 LPM. The result of this test shows that the temperature characterization on static test provide a recommendation, that the experiments should be done at night because has a better environmental temperature stability compared to afternoon, with stable temperature around 1°C - 3°C. While on the dynamic test, the water temperature difference between the inlet-outlets in the heater area is quite large, about 7 times the temperature difference in the cooler area. The magnitude of the natural circulation flow rate calculated is much larger at about 300 times compared to the measured flow rate with different flow rate profiles.

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)

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

CANFLEX fuel bundle cross-flow endurance test (test report)

International Nuclear Information System (INIS)

Hong, Sung Deok; Chung, C. H.; Chang, S. K.; Kim, B. D.

1997-04-01

As part of the normal refuelling sequence of CANDU nuclear reactor, both new and irradiated bundles can be parked in the cross-flow region of the liner tubes. This situation occurs normally for a few minutes. The fuel bundle which is subjected to the cross-flow should be capable of withstanding the consequences of cross flow for normal periods, and maintain its mechanical integrity. The cross-flow endurance test was conducted for CANFLEX bundle, latest developed nuclear fuel, at CANDU-Hot Test Loop. The test was carried out during 4 hours at the inlet cross-flow region. After the test, the bundle successfully met all acceptance criteria after the 4 hours cross-flow test. (author). 2 refs., 3 tabs

CANFLEX fuel bundle cross-flow endurance test (test report)

Energy Technology Data Exchange (ETDEWEB)

Hong, Sung Deok; Chung, C. H.; Chang, S. K.; Kim, B. D.

1997-04-01

As part of the normal refuelling sequence of CANDU nuclear reactor, both new and irradiated bundles can be parked in the cross-flow region of the liner tubes. This situation occurs normally for a few minutes. The fuel bundle which is subjected to the cross-flow should be capable of withstanding the consequences of cross flow for normal periods, and maintain its mechanical integrity. The cross-flow endurance test was conducted for CANFLEX bundle, latest developed nuclear fuel, at CANDU-Hot Test Loop. The test was carried out during 4 hours at the inlet cross-flow region. After the test, the bundle successfully met all acceptance criteria after the 4 hours cross-flow test. (author). 2 refs., 3 tabs.

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.

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

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

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.

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

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

Real-time display of flow-pressure-volume loops.

Science.gov (United States)

Morozoff, P E; Evans, R W

1992-01-01

Graphic display of respiratory waveforms can be valuable for monitoring the progress of ventilated patients. A system has been developed that can display flow-pressure-volume loops as derived from a patient's respiratory circuit in real time. It can also display, store, print, and retrieve ventilatory waveforms. Five loops can be displayed at once: current, previous, reference, "ideal," and previously saved. Two components, the data-display device (DDD) and the data-collection device (DCD), comprise the system. An IBM 286/386 computer with a graphics card (VGA) and bidirectional parallel port is used for the DDD; an eight-bit microprocessor card and an A/D convertor card make up the DCD. A real-time multitasking operating system was written to control the DDD, while the DCD operates from in-line assembly code. The DCD samples the pressure and flow sensors at 100 Hz and looks for a complete flow waveform pattern based on flow slope. These waveforms are then passed to the DDD via the mutual parallel port. Within the DDD a process integrates the flow to create a volume signal and performs a multilinear regression on the pressure, flow, and volume data to calculate the elastance, resistance, pressure offset, and coefficient of determination. Elastance, resistance, and offset are used to calculate Pr and Pc where: Pr[k] = P[k]-offset-(elastance.V[k]) and Pc[k] = P[k]-offset-(resistance.F[k]). Volume vs. Pc and flow vs. Pr can be displayed in real time. Patient data from previous clinical tests were loaded into the device to verify the software calculations. An analog waveform generator was used to simulate flow and pressure waveforms that validated the system.(ABSTRACT TRUNCATED AT 250 WORDS)

Flow-volume loops measured with electrical impedance tomography in pediatric patients with asthma.

Science.gov (United States)

Ngo, Chuong; Dippel, Falk; Tenbrock, Klaus; Leonhardt, Steffen; Lehmann, Sylvia

2018-05-01

Electrical impedance tomography (EIT) provides information on global and regional ventilation during tidal breathing and mechanical ventilation. During forced expiration maneuvers, the linearity of EIT and spirometric data has been documented in healthy persons. The present study investigates the potential diagnostic use of EIT in pediatric patients with asthma. EIT and spirometry were performed in 58 children with asthma (average age ± SD: 11.86 ± 3.13 years), and 58 healthy controls (average age ± SD: 12.12 ± 2.9 years). The correlation between EIT data and simultaneously acquired spirometric data were tested for FEV 1 , FEV 0.5 , MEF 75 , MEF 50 , and MEF 25 . Binary classification tests were performed for the EIT-derived Tiffeneau index FEV 1 /FVC and the bronchodilator test index ΔFEV 1 . Average flow-volume (FV) loops were generated for patients with pathologic spirometry to demonstrate the feasibility of EIT for graphic diagnosis of asthma. Spirometry and global EIT-based FV loops showed a strong correlation (P  0.9 in FEV 1 and FEV 0.5 ). In all criteria, the binary classification tests yielded high specificity (>93%), a high positive predictive value (≥75%) and a high negative predictive value (>80%), while sensitivity was higher in ΔFEV 1 (86.67%) and lower in FEV 1 /FVC (25% and 35.29%). A typical concave shape of the EIT-derived average FV loops was observed for asthmatic children with improvement after bronchospasmolysis. Global FV loops derived from EIT correlate well with spirometry. Positive bronchospasmolysis can be observed in EIT-derived FV loops. Flow-volume loops originated from EIT have a potential to visualize pulmonary function. © 2018 Wiley Periodicals, Inc.

Topologically protected loop flows in high voltage AC power grids

International Nuclear Information System (INIS)

Coletta, T; Delabays, R; Jacquod, Ph; Adagideli, I

2016-01-01

Geographical features such as mountain ranges or big lakes and inland seas often result in large closed loops in high voltage AC power grids. Sizable circulating power flows have been recorded around such loops, which take up transmission line capacity and dissipate but do not deliver electric power. Power flows in high voltage AC transmission grids are dominantly governed by voltage angle differences between connected buses, much in the same way as Josephson currents depend on phase differences between tunnel-coupled superconductors. From this previously overlooked similarity we argue here that circulating power flows in AC power grids are analogous to supercurrents flowing in superconducting rings and in rings of Josephson junctions. We investigate how circulating power flows can be created and how they behave in the presence of ohmic dissipation. We show how changing operating conditions may generate them, how significantly more power is ohmically dissipated in their presence and how they are topologically protected, even in the presence of dissipation, so that they persist when operating conditions are returned to their original values. We identify three mechanisms for creating circulating power flows, (i) by loss of stability of the equilibrium state carrying no circulating loop flow, (ii) by tripping of a line traversing a large loop in the network and (iii) by reclosing a loop that tripped or was open earlier. Because voltages are uniquely defined, circulating power flows can take on only discrete values, much in the same way as circulation around vortices is quantized in superfluids. (paper)

Joint test rig for testing and calibrating of different methods of two-phase mass flow measurement

International Nuclear Information System (INIS)

Reimann, J.; Arnold, G.; Chung, M.; Hahn, H.; John, H.; Mueller, S.; Wanner, E.

1977-01-01

The start-up of the steady-state steam-water loop has been finished. The planned maximal values of the mass flow rate as function of quality and pressure are reached. The components for the steady-state air-water loop have been ordered, the loop has been built up, first function tests have been carried out. Because of the additional work of the extension for air-water flows, the blowdown test rig was delayed. Calculations for the security of the pressure vessel have begun. During the experiments the knowledge of the flow regime and the apparent density is essential. To detect flow regime, impedance probes were developed and have been tested in steam-water flows at pressures up to 150 at. The probe signals can be adjointed to flow patterns even in those cases when high speed movies could not be interpreted definitely. To measure the apparent density a multiple γ-beam densitometer is developed. The collimator block and the mounting support for the γ-source were manufactured, the shielding and cooling of the scintillator has begun. (orig./RW) [de

Construction of the two-phase critical flow test facility

International Nuclear Information System (INIS)

Chung, C. H.; Chang, S. K.; Park, H. S.; Min, K. H.; Choi, N. H.; Kim, C. H.; Lee, S. H.; Kim, H. C.; Chang, M. H.

2002-03-01

The two-phase critical test loop facility has been constructed in the KAERI engineering laboratory for the simulation of small break loss of coolant accident entrained with non-condensible gas of SMART. The test facility can operate at 12 MPa of pressure and 0 to 60 C of sub-cooling with 0.5 kg/s of non- condensible gas injection into break flow, and simulate up to 20 mm of pipe break. Main components of the test facility were arranged such that the pressure vessel containing coolant, a test section simulating break and a suppression tank inter-connected with pipings were installed vertically. As quick opening valve opens, high pressure/temperature coolant flows through the test section forming critical two-phase flow into the suppression tank. The pressure vessel was connected to two high pressure N2 gas tanks through a control valve to control pressure in the pressure vessel. Another N2 gas tank was also connected to the test section for the non-condensible gas injection. The test facility operation was performed on computers supported with PLC systems installed in the control room, and test data such as temperature, break flow rate, pressure drop across test section, gas injection flow rate were all together gathered in the data acquisition system for further data analysis. This test facility was classified as a safety related high pressure gas facility in law. Thus the loop design documentation was reviewed, and inspected during construction of the test loop by the regulatory body. And the regulatory body issued permission for the operation of the test facility

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.

Joint test rig for testing and calibrating of different methods of two-phase mass flow measurement

International Nuclear Information System (INIS)

Reimann, J.; Demski, A.; Hahn, H.; Harten, U.; John, H.; Megerle, A.; Mueller, S.; Pawlak, L.; Wanner, E.

1977-01-01

The steam-water loop was completed by building in two throttling valves upstream of the mixing chamber. By producing steam by throttling the total mass flow may be increased up to 35% compared to the former method of operating the loop. Furthermore, throttling stabilizes the single phase mass flow measurement. The data aquisition system and computation of the reference values has been finished. The computer program contains the equations of state of steam/water and the calibration curves for all signal transducers. The 5 beam γ-densitometer has been finished mechanically and supplied with the electronics. First calibration tests are fully satisfactory. The instrumentation of the air-water loop completed. At low quality the mass fluxes are increased by a factor of 5 compared with the steam-water-loop. The regime of dispersed bubble flow is fully reached in the test section. To detect flow regimes air-water as well as in steam-water flow, a local impedance probe was used. In addition, the phase distribution across the channel could be detected by traversing the probe. The boundaries of the air-water flow regimes detected by the probe are in good correspondance with other investigations. For the first time, such experiments have been carried out in horizontal steam-water flow. The results indicate that the region of slug flow becomes smaller with increasing pressure. (orig./RW) [de

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

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

CANFLEX fuel bundle cross-flow endurance test 2 (test procedure)

Energy Technology Data Exchange (ETDEWEB)

Hong, Sung Deok; Chung, C. H.; Chang, S. K. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-04-01

This report describes test procedure of cross-flow 2 test for CANFLEX fuel. In October 1996. a cross-flow test was successfully performed in the KAERI Hot Test Loop for four hours at a water flow rate of 31kg/s, temperature of 266 deg C and inlet pressure of 11MPa, but it is requested more extended time periods to determine a realistic operational margin for the CANFLEX bundle during abnormal refuelling operations. The test shall be conducted for twenty two hours under the reactor conditions. After an initial period of ten hours, the test shall be stopped at the intervals of four hours for bundle inspection and inspect the test bundle end-plate to end-cap welds for failure or crack propagation using liquid penetrant examination. 2 refs., 1 fig. (Author)

Assessment and monitoring of flow limitation and other parameters from flow/volume loops.

Science.gov (United States)

Dueck, R

2000-01-01

Flow/volume (F/V) spirometry is routinely used for assessing the type and severity of lung disease. Forced vital capacity (FVC) and timed vital capacity (FEV1) provide the best estimates of airflow obstruction in patients with asthma, chronic obstructive pulmonary disease (COPD) and emphysema. Computerized spirometers are now available for early home recognition of asthma exacerbation in high risk patients with severe persistent disease, and for recognition of either infection or rejection in lung transplant patients. Patients with severe COPD may exhibit expiratory flow limitation (EFL) on tidal volume (VT) expiratory F/V (VTF/V) curves, either with or without applying negative expiratory pressure (NEP). EFL results in dynamic hyperinflation and persistently raised alveolar pressure or intrinsic PEEP (PEEPi). Hyperinflation and raised PEEPi greatly enhance dyspnea with exertion through the added work of the threshold load needed to overcome raised pleural pressure. Esophageal (pleural) pressure monitoring may be added to VTF/V loops for assessing the severity of PEEPi: 1) to optimize assisted ventilation by mask or via endotracheal tube with high inspiratory flow rates to lower I:E ratio, and 2) to assess the efficacy of either pressure support ventilation (PSV) or low level extrinsic PEEP in reducing the threshold load of PEEPi. Intraoperative tidal volume F/V loops can also be used to document the efficacy of emphysema lung volume reduction surgery (LVRS) via disappearance of EFL. Finally, the mechanism of ventilatory constraint can be identified with the use of exercise tidal volume F/V loops referenced to maximum F/V loops and static lung volumes. Patients with severe COPD show inspiratory F/V loops approaching 95% of total lung capacity, and flow limitation over the entire expiratory F/V curve during light levels of exercise. Surprisingly, patients with a history of congestive heart failure may lower lung volume towards residual volume during exercise

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

Study on corrosion test techniques in lead bismuth eutectic flow. Joint research report in JFY2002

International Nuclear Information System (INIS)

Takahashi, Minoru; Sekimoto, Hiroshi

2003-03-01

The evaluation of corrosion behaviors of core and structural materials in lead bismuth eutectic is one of the key issues for the utilization of lead bismuth eutectic as a coolant of the primary loops of lead bismuth cooled fast breeder reactors (FBRs) and the intermediate heat transport media of new-type steam generators of the sodium cooled FBRs. The purpose of the present study is to establish corrosion test techniques in lead bismuth eutectic flow. The techniques of steel corrosion test and oxygen control in flowing lead bismuth eutectic, and the technologies of a lead bismuth flow test at high temperature and high velocity were developed through corrosion test using a lead bismuth flow test loop of the Tokyo Institute of Technology in JFY2002. The major results are summarized as follows: (1) Techniques of fabrication, mount and rinse of corrosion specimens, measurement method of weight loss, and SEM/EDX analysis method have been established through lead bismuth corrosion test. (2) Weight losses were measured, corrosion and lead bismuth-adhered layers and eroded parts were observed in two 1000 hr-corrosion tests, and the results were compared with each other for twelve existing steels including ODS, F82H and SUH-3. (3) An oxygen sensor made of zirconia electrolyte structurally resistant to thermal stress and thermal shock was developed and tested in the lead bismuth flow loop. Good performance has been obtained. (4) An oxygen control method by injecting argon and hydrogen mixture gas containing steam into lead bismuth was applied to the lead bismuth flow loop, and technical issues for the development of the oxygen control method were extracted. (5) Technical measures for freezing and leakage of lead bismuth in the flow loop were accumulated. (6) Technical measures for flow rate decrease/blockage due to precipitation of oxide and corrosion products in a low temperature section of the lead bismuth flow loop were accumulated. (7) Electromagnetic flow meters with MI

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

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

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.)

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

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)

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)

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)

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.

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

A generalised correlation for the steady state flow in single-phase natural circulation loops

International Nuclear Information System (INIS)

Vijayan, P.K.; Bade, M.H.; Saha, D.; Sinha, R.K.; Venkat Raj, V.

2000-08-01

To establish the heat transport capability of natural circulation loops, it is essential to know the flow rate. A generalized correlation for steady state flow valid for uniform and non-uniform diameter loops has been theoretically derived

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

Hypervapotron flow testing with rapid prototype models

International Nuclear Information System (INIS)

Driemeyer, D.; Hellwig, T.; Kubik, D.; Langenderfer, E.; Mantz, H.; McSmith, M.; Jones, B.; Butler, J.

1995-01-01

A flow test model of the inlet section of a three channel hypervapotron plate that has been proposed as a heat sink in the ITER divertor was prepared using a rapid prototyping stereolithography process that is widely used for component development in US industry. An existing water flow loop at the University of Illinois is being used for isothermal flow tests to collect pressure drop data for comparison with proposed vapotron friction factor correlations. Differential pressure measurements are taken, across the test section inlet manifold, the vapotron channel (about a seven inch length), the outlet manifold and the inlet-to-outlet. The differential pressures are currently measured with manometers. Tests were conducted at flow velocities from 1--10 m/s to cover the full range of ITER interest. A tap was also added for a small hypodermic needle to inject dye into the flow channel at several positions to examine the nature of the developing flow field at the entrance to the vapotron section. Follow-on flow tests are planned using a model with adjustable flow channel dimensions to permit more extensive pressure drop data to be collected. This information will be used to update vapotron design correlations for ITER

Validation of the generalized model of two-phase thermosyphon loop based on experimental measurements of volumetric flow rate

Science.gov (United States)

Bieliński, Henryk

2016-09-01

The current paper presents the experimental validation of the generalized model of the two-phase thermosyphon loop. The generalized model is based on mass, momentum, and energy balances in the evaporators, rising tube, condensers and the falling tube. The theoretical analysis and the experimental data have been obtained for a new designed variant. The variant refers to a thermosyphon loop with both minichannels and conventional tubes. The thermosyphon loop consists of an evaporator on the lower vertical section and a condenser on the upper vertical section. The one-dimensional homogeneous and separated two-phase flow models were used in calculations. The latest minichannel heat transfer correlations available in literature were applied. A numerical analysis of the volumetric flow rate in the steady-state has been done. The experiment was conducted on a specially designed test apparatus. Ultrapure water was used as a working fluid. The results show that the theoretical predictions are in good agreement with the measured volumetric flow rate at steady-state.

Expanding and Contracting Coronal Loops as Evidence of Vortex Flows Induced by Solar Eruptions

Energy Technology Data Exchange (ETDEWEB)

Dudík, J. [Astronomical Institute of the Czech Academy of Sciences, Fričova 298, 251 65 Ondřejov (Czech Republic); Zuccarello, F. P.; Aulanier, G.; Schmieder, B.; Démoulin, P., E-mail: jaroslav.dudik@asu.cas.cz [LESIA, Observatoire de Paris, Psl Research University, CNRS, Sorbonne Universits, UPMC Univ. Paris 06, Univ. Paris Diderot, Sorbonne Paris Cit, 5 place Jules Janssen, F-92195 Meudon (France)

2017-07-20

Eruptive solar flares were predicted to generate large-scale vortex flows at both sides of the erupting magnetic flux rope. This process is analogous to a well-known hydrodynamic process creating vortex rings. The vortices lead to advection of closed coronal loops located at the peripheries of the flaring active region. Outward flows are expected in the upper part and returning flows in the lower part of the vortex. Here, we examine two eruptive solar flares, the X1.1-class flare SOL2012-03-05T03:20 and the C3.5-class SOL2013-06-19T07:29. In both flares, we find that the coronal loops observed by the Atmospheric Imaging Assembly in its 171 Å, 193 Å, or 211 Å passbands show coexistence of expanding and contracting motions, in accordance with the model prediction. In the X-class flare, multiple expanding and contracting loops coexist for more than 35 minutes, while in the C-class flare, an expanding loop in 193 Å appears to be close by and cotemporal with an apparently imploding loop arcade seen in 171 Å. Later, the 193 Å loop also switches to contraction. These observations are naturally explained by vortex flows present in a model of eruptive solar flares.

Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

International Nuclear Information System (INIS)

McCulloch, R.W.; MacPherson, R.E.

1983-03-01

The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm 2 , 1000 0 C cladding temperature, and (2) 40 h at 40 W/cm 2 , 1200 0 C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370 0 C

Development of Flow Boiling and Condensation Experiment on the International Space Station- Normal and Low Gravity Flow Boiling Experiment Development and Test Results

Science.gov (United States)

Nahra, Henry K.; Hall, Nancy R.; Hasan, Mohammad M.; Wagner, James D.; May, Rochelle L.; Mackey, Jeffrey R.; Kolacz, John S.; Butcher, Robert L.; Frankenfield, Bruce J.; Mudawar, Issam;

CANFLEX fuel bundle cross-flow endurance test 2 (Test report)

Energy Technology Data Exchange (ETDEWEB)

Hong, Sung Deok; Chung, C. H.; Chang, S. K.; Kim, B. D. [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-02-01

This report describes cross-flow endurance test 2 that was conducted at the CANDU-Hot Test Loop. The test was completed on March 30, 1999 using a new CANFLEX bundle, built by KAERI. It was carried out for a total of 22 hours. After an initial period of ten hours, the test was stopped at the intervals of four hours for bundle inspection and inter-element gap measurement[7]. The test bundle end-plate to end-cap welds were inspected carefully for failure or crack propagation using liquid penetrant examination especially at the heat-affected zones. 12 refs., 4 figs., 10 tabs. (Author)

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

Mercury flow tests (first report). Wall friction factor measurement tests and future tests plan

International Nuclear Information System (INIS)

Kaminaga, Masanori; Kinoshita, Hidetaka; Haga, Katsuhiro; Hino, Ryutaro; Sudo, Yukio

1999-07-01

In the neutron science project at JAERI, we plan to inject a pulsed proton beam of a maximum power of 5 MW from a high intense proton accelerator into a mercury target in order to produce high energy neutrons of a magnitude of ten times or more than existing facilities. The neutrons produced by the facility will be utilized for advanced field of science such as the life sciences etc. An urgent issue in order to accomplish this project is the establishment of mercury target technology. With this in mind, a mercury experimental loop with the capacity to circulate mercury up to 15 L/min was constructed to perform thermal hydraulic tests, component tests and erosion characteristic tests. A measurement of the wall friction factor was carried out as a first step of the mercury flow tests, while testing the characteristic of components installed in the mercury loop. This report presents an outline of the mercury loop and experimental results of the wall friction factor measurement. From the wall friction factor measurement, it was made clear that the wettability of the mercury was improved with an increase of the loop operation time and at the same time the wall friction factors were increased. The measured wall friction factors were much lower than the values calculated by the Blasius equation at the beginning of the loop operation because of wall slip caused by a non-wetted condition. They agreed well with the values calculated by the Blasius equation within a deviation of 10% when the sum of the operation time increased more than 11 hours. This report also introduces technical problems with a mercury circulation and future tests plan indispensable for the development of the mercury target. (author)

Performance assessment of mass flow rate measurement capability in a large scale transient two-phase flow test system

International Nuclear Information System (INIS)

Nalezny, C.L.; Chapman, R.L.; Martinell, J.S.; Riordon, R.P.; Solbrig, C.W.

1979-01-01

Mass flow is an important measured variable in the Loss-of-Fluid Test (LOFT) Program. Large uncertainties in mass flow measurements in the LOFT piping during LOFT coolant experiments requires instrument testing in a transient two-phase flow loop that simulates the geometry of the LOFT piping. To satisfy this need, a transient two-phase flow loop has been designed and built. The load cell weighing system, which provides reference mass flow measurements, has been analyzed to assess its capability to provide the measurements. The analysis consisted of first performing a thermal-hydraulic analysis using RELAP4 to compute mass inventory and pressure fluctuations in the system and mass flow rate at the instrument location. RELAP4 output was used as input to a structural analysis code SAPIV which is used to determine load cell response. The computed load cell response was then smoothed and differentiated to compute mass flow rate from the system. Comparison between computed mass flow rate at the instrument location and mass flow rate from the system computed from the load cell output was used to evaluate mass flow measurement capability of the load cell weighing system. Results of the analysis indicate that the load cell weighing system will provide reference mass flows more accurately than the instruments now in LOFT

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)

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)

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

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.

Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

Energy Technology Data Exchange (ETDEWEB)

McCulloch, R.W.; MacPherson, R.E.

1983-03-01

The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

Full sized tests on a french coolant pump under two-phase flow

International Nuclear Information System (INIS)

Huchard, J.C.; Bore, C.; Dueymes, E.

1997-01-01

The French Safety Authorities required EDF to demonstrate the ability of the new N4 main coolant pump to withstand two-phase flow conditions without damage. Therefore three full sized tests, simulating a bleeding flow on the primary system, were performed on a laboratory test loop under real operating conditions (temperature = 290 deg. C, pressure = 155 b, flowrate = 7 m 3 /s; electrical power = 7 MW). The maximum value of the mean void fraction reached 75 %. The outcome of the tests is very positive: the mechanical behaviour of the main coolant pump is good, even at high void fraction. The maximum vibration levels were below the limits fixed by the manufacturer. Correlations between the mechanical behaviour of the pump and the pressure pulsation in the test loop have been found. (authors)

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.

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

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.

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...

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)

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.

Tritium Management Loop Design Status

Energy Technology Data Exchange (ETDEWEB)

Rader, Jordan D. [ORNL; Felde, David K. [ORNL; McFarlane, Joanna [ORNL; Greenwood, Michael Scott [ORNL; Qualls, A L. [ORNL; Calderoni, Pattrick [Idaho National Laboratory (INL)

2017-12-01

This report summarizes physical, chemical, and engineering analyses that have been done to support the development of a test loop to study tritium migration in 2LiF-BeF2 salts. The loop will operate under turbulent flow and a schematic of the apparatus has been used to develop a model in Mathcad to suggest flow parameters that should be targeted in loop operation. The introduction of tritium into the loop has been discussed as well as various means to capture or divert the tritium from egress through a test assembly. Permeation was calculated starting with a Modelica model for a transport through a nickel window into a vacuum, and modifying it for a FLiBe system with an argon sweep gas on the downstream side of the permeation interface. Results suggest that tritium removal with a simple tubular permeation device will occur readily. Although this system is idealized, it suggests that rapid measurement capability in the loop may be necessary to study and understand tritium removal from the system.

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.

The onset of flows and instabilities in a thermosyphon with parallel loops

International Nuclear Information System (INIS)

Zvirin, Y.

1986-01-01

A theoretical study is presented for the stability of various steady flows in a thermosyphon with multiple vertical channels. The main interest is in the onset of motion from a rest state or in a stagnant branch, therefore laminar flow is considered and a one-dimensional model is used to describe the flow and temperature fields. The steady state solutions include a state of no flow (rest) in the whole system and two basic flow configurations: a single loop between two channels while the others are stagnant and a symmetric flow. For a three-channel system the latter consists of an upward velocity in one branch and downward velocities in the other two. The mirror image of these basic flows are also steady state solutions. A critical modified number is found to be the stability margin for the onset of motion from a rest state in the entire system. This result was obtained both by a study of the steady state solution and by the stability analysis. The steady flow with a stagnant loop is always unstable while the symmetric flow solution in the system considered here is always stable. (orig./HP)

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.

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

Low-pressure dynamics of a natural-circulation two-phase flow loop

International Nuclear Information System (INIS)

Manera, A.; Kruijf, W.J.M. de; Hartmann, H.; Mudde, R.F.; Hagen, T.H.J.J. van der

2001-01-01

Flashing induced oscillations in a natural circulation loop are studied as function of heating power and inlet subcooling in symmetrical and asymmetrical power conditions. To unveil the effects of power/velocity asymmetries on the two-phase flow stability at low power and low pressure conditions different signals at several locations in the loop are recorded. In particular a Laser Doppler Anemometry set-up is used to measure the velocity simultaneously in two parallel channels and a wire-mesh sensor is used to measure the 2D void fraction distribution in a section of the ascendant part of the loop. (orig.)

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.

The development of flow test technology for PWR fuel assembly

International Nuclear Information System (INIS)

Chung, Moon Ki; Cha, Chong Hee; Chung, Chang Hwan; Chun, Se Young; Song, Chul Hwa; Chung, Heung Joon; Won, Soon Yeun; Cho, Yeong Rho; Kim, Bok Deuk

1988-05-01

KAERI has an extensive program to develope PWR fuel assembly. In relation to the program, development of flow test technology is needed to evaluate the thermal hydraulic compactibility and mechanical integrity of domestically fabricated nuclear fuels. A high-pressure and high-temperature flow test facility was designed to test domestically fabricated fuel assembly. The test section of the facility has capacity of a 6x6 full length PWR fuel assembly. A flow test rig was designed and installed at Cold Test Loop to carry out model experiments with 5x5 rod assembly under atmosphere pressure to get information about the characteristics of pressure loss of spacer grids and velocity distribution in the subchannels. LDV measuring technology was established using TSI's Laser Dopper Velocimeter 9100-3 System

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.

Mineral carbonation: energy costs of pretreatment options and insights gained from flow loop reaction studies

Energy Technology Data Exchange (ETDEWEB)

Penner, Larry R.; O' Connor, William K.; Dahlin, David C.; Gerdemann, Stephen J.; Rush, Gilbert E.

2004-01-01

Sequestration of carbon as a stable mineral carbonate has been proposed to mitigate environmental concerns that carbon dioxide may with time escape from its sequestered matrix using alternative sequestration technologies. A method has been developed to prepare stable carbonate products by reacting CO2 with magnesium silicate minerals in aqueous bicarbonate/chloride media at high temperature and pressure. Because this approach is inherently expensive due to slow reaction rates and high capital costs, studies were conducted to improve the reaction rates through mineral pretreatment steps and to cut expenses through improved reactor technology. An overview is given for the estimated cost of the process including sensitivity to grinding and heating as pretreatment options for several mineral feedstocks. The energy costs are evaluated for each pretreatment in terms of net carbon avoided. New studies with a high-temperature, high-pressure flow-loop reactor have yielded information on overcoming kinetic barriers experienced with processing in stirred autoclave reactors. Repeated tests with the flow-loop reactor have yielded insights on wear and failure of system components, on challenges to maintain and measure flow, and for better understanding of the reaction mechanism.

Mineral carbonation: energy costs of pretreatment options and insights gained from flow loop reaction studies

International Nuclear Information System (INIS)

Penner, Larry R.; O'Connor, William K.; Dahlin, David C.; Gerdemann, Stephen J.; Rush, Gilbert E.

2004-01-01

Sequestration of carbon as a stable mineral carbonate has been proposed to mitigate environmental concerns that carbon dioxide may with time escape from its sequestered matrix using alternative sequestration technologies. A method has been developed to prepare stable carbonate products by reacting CO2 with magnesium silicate minerals in aqueous bicarbonate/chloride media at high temperature and pressure. Because this approach is inherently expensive due to slow reaction rates and high capital costs, studies were conducted to improve the reaction rates through mineral pretreatment steps and to cut expenses through improved reactor technology. An overview is given for the estimated cost of the process including sensitivity to grinding and heating as pretreatment options for several mineral feedstocks. The energy costs are evaluated for each pretreatment in terms of net carbon avoided. New studies with a high-temperature, high-pressure flow-loop reactor have yielded information on overcoming kinetic barriers experienced with processing in stirred autoclave reactors. Repeated tests with the flow-loop reactor have yielded insights on wear and failure of system components, on challenges to maintain and measure flow, and for better understanding of the reaction mechanism

Experimental and computational analysis of pressure response in a multiphase flow loop

Science.gov (United States)

Morshed, Munzarin; Amin, Al; Rahman, Mohammad Azizur; Imtiaz, Syed

2016-07-01

The characteristics of multiphase fluid flow in pipes are useful to understand fluid mechanics encountered in the oil and gas industries. In the present day oil and gas exploration is successively inducing subsea operation in the deep sea and arctic condition. During the transport of petroleum products, understanding the fluid dynamics inside the pipe network is important for flow assurance. In this case the information regarding static and dynamic pressure response, pressure loss, optimum flow rate, pipe diameter etc. are the important parameter for flow assurance. The principal aim of this research is to represents computational analysis and experimental analysis of multi-phase (L/G) in a pipe network. This computational study considers a two-phase fluid flow through a horizontal flow loop with at different Reynolds number in order to determine the pressure distribution, frictional pressure loss profiles by volume of fluid (VOF) method. However, numerical simulations are validated with the experimental data. The experiment is conducted in 76.20 mm ID transparent circular pipe using water and air in the flow loop. Static pressure transducers are used to measure local pressure response in multiphase pipeline.

Measurement of flow rate in the third loop of PWR

International Nuclear Information System (INIS)

Gao Shufan.

1986-01-01

The range of flow rate was 14000-50000 m 3 /h. The diameter of main tube was 2.6 m. A special made pitot set was placed on the main tube in order to accurately measure the flow rate. A cross slideway and a guide devicc were used to prevent the pitot vibration. Method of equal annular area was used in the measurement. The error was less than 4.2%. A pitot cylinder flowmeter was set also on the main tube to supervise the total flow rate of the third loop

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

Study and development of an air conditioning system operating on a magnetic heat pump cycle (design and testing of flow directors)

Science.gov (United States)

Wang, Pao-Lien

1992-01-01

This report describes the fabrication, design of flow director, fluid flow direction analysis and testing of flow director of a magnetic heat pump. The objectives of the project are: (1) to fabricate a demonstration magnetic heat pump prototype with flow directors installed; and (2) analysis and testing of flow director and to make sure working fluid loops flow through correct directions with minor mixing. The prototype was fabricated and tested at the Development Testing Laboratory of Kennedy Space Center. The magnetic heat pump uses rear earth metal plates rotate in and out of a magnetic field in a clear plastic housing with water flowing through the rotor plates to provide temperature lift. Obtaining the proper water flow direction has been a problem. Flow directors were installed as flow barriers between separating point of two parallel loops. Function of flow directors were proven to be excellent both analytically and experimentally.

Brief description of out-of-pile test facilities for study in corrosion and fission product behaviour in flowing sodium

International Nuclear Information System (INIS)

Iizawa, K.; Sekiguchi, N.; Atsumo, H.

1976-01-01

The experimental methods to perform tests for study in corrosion and fission products behaviour in flowing sodium are outlined. Flow diagrams for the activated materials and fission products behaviour test loop are given

Distributed flow sensing for closed-loop speed control of a flexible fish robot.

Science.gov (United States)

Zhang, Feitian; Lagor, Francis D; Yeo, Derrick; Washington, Patrick; Paley, Derek A

2015-10-23

Flexibility plays an important role in fish behavior by enabling high maneuverability for predator avoidance and swimming in turbulent flow. This paper presents a novel flexible fish robot equipped with distributed pressure sensors for flow sensing. The body of the robot is molded from soft, hyperelastic material, which provides flexibility. Its Joukowski-foil shape is conducive to modeling the fluid analytically. A quasi-steady potential-flow model is adopted for real-time flow estimation, whereas a discrete-time vortex-shedding flow model is used for higher-fidelity simulation. The dynamics for the flexible fish robot yield a reduced model for one-dimensional swimming. A recursive Bayesian filter assimilates pressure measurements to estimate flow speed, angle of attack, and foil camber. The closed-loop speed-control strategy combines an inverse-mapping feedforward controller based on an average model derived for periodic actuation of angle-of-attack and a proportional-integral feedback controller utilizing the estimated flow information. Simulation and experimental results are presented to show the effectiveness of the estimation and control strategy. The paper provides a systematic approach to distributed flow sensing for closed-loop speed control of a flexible fish robot by regulating the flapping amplitude.

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)

Application of computational fluid dynamics to closed-loop bioreactors: I. Characterization and simulation of fluid-flow pattern and oxygen transfer.

Science.gov (United States)

Littleton, Helen X; Daigger, Glen T; Strom, Peter F

2007-06-01

A full-scale, closed-loop bioreactor (Orbal oxidation ditch, Envirex brand technologies, Siemens, Waukesha, Wisconsin), previously examined for simultaneous biological nutrient removal (SBNR), was further evaluated using computational fluid dynamics (CFD). A CFD model was developed first by imparting the known momentum (calculated by tank fluid velocity and mass flowrate) to the fluid at the aeration disc region. Oxygen source (aeration) and sink (consumption) terms were introduced, and statistical analysis was applied to the CFD simulation results. The CFD model was validated with field data obtained from a test tank and a full-scale tank. The results indicated that CFD could predict the mixing pattern in closed-loop bioreactors. This enables visualization of the flow pattern, both with regard to flow velocity and dissolved-oxygen-distribution profiles. The velocity and oxygen-distribution gradients suggested that the flow patterns produced by directional aeration in closed-loop bioreactors created a heterogeneous environment that can result in dissolved oxygen variations throughout the bioreactor. Distinct anaerobic zones on a macroenvironment scale were not observed, but it is clear that, when flow passed around curves, a secondary spiral flow was generated. This second current, along with the main recirculation flow, could create alternating anaerobic and aerobic conditions vertically and horizontally, which would allow SBNR to occur. Reliable SBNR performance in Orbal oxidation ditches may be a result, at least in part, of such a spatially varying environment.

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.

Description of the sodium loop ML-3

International Nuclear Information System (INIS)

Torre, de la M.; Melches, I; Lapena, J.; Martinez, T.A.; Miguel, de D.; Duran, F.

1979-01-01

The sodium loop ML-3 is described. The main objective of this facility is to obtain mechanical property data for LMFBR materials in creep and low cycle fatigue testing in flowing sodium. ML-3 includes 10 test stations for creep and two for fatigue. It is possible to operate simultaneously at three different temperature levels. The maximum operating temperature is 650 deg C at flow velocities up to 5 m/s. The ML-3 loop has been located in a manner that permits the fill/dump tank cover gas and security systems to be shared with an earlier circuit, the ML-1. (author)

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

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

Can flow-volume loops be used to diagnose exerciseinduced laryngeal obstructions?

DEFF Research Database (Denmark)

Christensen, Pernille Melia; Maltbæk, Niels; Jørgensen, Inger M

2013-01-01

BACKGROUND: Pre- and post-exercise flow-volume loops are often recommended as an easy non-invasive method for diagnosing or excluding exercise-induced laryngeal obstructions in patients with exercise-related respiratory symptoms. However, at present there is no evidence for this recommendation...

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.

MHD PbLi experiments in MaPLE loop at UCLA

International Nuclear Information System (INIS)

Courtessole, C.; Smolentsev, S.; Sketchley, T.; Abdou, M.

2016-01-01

Highlights: • The paper overviews the MaPLE facility at UCLA: one-of-a-few PbLi MHD loop in the world. • We present the progress achieved in development and testing of high-temperature PbLi flow diagnostics. • The most important MHD experiments carried out since the first loop operation in 2011 are summarized. - Abstract: Experiments on magnetohydrodynamic (MHD) flows are critical to understanding complex flow phenomena in ducts of liquid metal blankets, in particular those that utilize eutectic alloy lead–lithium as breeder/coolant, such as self-cooled, dual-coolant and helium-cooled lead–lithium blanket concepts. The primary goal of MHD experiments at UCLA using the liquid metal flow facility called MaPLE (Magnetohydrodynamic PbLi Experiment) is to address important MHD effects, heat transfer and flow materials interactions in blanket-relevant conditions. The paper overviews the one-of-a-kind MaPLE loop at UCLA and presents recent experimental activities, including the development and testing of high-temperature PbLi flow diagnostics and experiments that have been performed since the first loop operation in 2011. We also discuss MaPLE upgrades, which need to be done to substantially expand the experimental capabilities towards a new class of MHD flow phenomena that includes buoyancy effects.

MHD PbLi experiments in MaPLE loop at UCLA

Energy Technology Data Exchange (ETDEWEB)

Courtessole, C., E-mail: cyril@fusion.ucla.edu; Smolentsev, S.; Sketchley, T.; Abdou, M.

2016-11-01

Highlights: • The paper overviews the MaPLE facility at UCLA: one-of-a-few PbLi MHD loop in the world. • We present the progress achieved in development and testing of high-temperature PbLi flow diagnostics. • The most important MHD experiments carried out since the first loop operation in 2011 are summarized. - Abstract: Experiments on magnetohydrodynamic (MHD) flows are critical to understanding complex flow phenomena in ducts of liquid metal blankets, in particular those that utilize eutectic alloy lead–lithium as breeder/coolant, such as self-cooled, dual-coolant and helium-cooled lead–lithium blanket concepts. The primary goal of MHD experiments at UCLA using the liquid metal flow facility called MaPLE (Magnetohydrodynamic PbLi Experiment) is to address important MHD effects, heat transfer and flow materials interactions in blanket-relevant conditions. The paper overviews the one-of-a-kind MaPLE loop at UCLA and presents recent experimental activities, including the development and testing of high-temperature PbLi flow diagnostics and experiments that have been performed since the first loop operation in 2011. We also discuss MaPLE upgrades, which need to be done to substantially expand the experimental capabilities towards a new class of MHD flow phenomena that includes buoyancy effects.

Research and design of 3He pressure control loop

International Nuclear Information System (INIS)

Huang Xin; Zhang Peisheng; Tang Guoliang; Zhang Aimin; Zhang Yingchao

2008-01-01

In order to carry out power transient tests for PWR fuel element in China Advanced Research Reactor (CARR), the research and conceptual design of 3He pressure control loop were completed. The working principle, design parameters and technological flow of the loop were described. It is seen that the a He loop can adjust the power of the tested PWR fuel element rapidly, evenly and flexibly and it is an optimal path to realize the power transient regulation for tested PWR fuel. (authors)

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

Fluid-flow pressure measurements and thermo-fluid characterization of a single loop two-phase passive heat transfer device

Science.gov (United States)

Ilinca, A.; Mangini, D.; Mameli, M.; Fioriti, D.; Filippeschi, S.; Araneo, L.; Roth, N.; Marengo, M.

2017-11-01

A Novel Single Loop Pulsating Heat Pipe (SLPHP), with an inner diameter of 2 mm, filled up with two working fluids (Ethanol and FC-72, Filling Ratio of 60%), is tested in Bottom Heated mode varying the heating power and the orientation. The static confinement diameter for Ethanol and FC-72, respectively 3.4 mm and 1.7mm, is above and slightly under the inner diameter of the tube. This is important for a better understanding of the working principle of the device very close to the limit between the Loop Thermosyphon and Pulsating Heat Pipe working modes. With respect to previous SLPHP experiments found in the literature, such device is designed with two transparent inserts mounted between the evaporator and the condenser allowing direct fluid flow visualization. Two highly accurate pressure transducers permit local pressure measurements just at the edges of one of the transparent inserts. Additionally, three heating elements are controlled independently, so as to vary the heating distribution at the evaporator. It is found that peculiar heating distributions promote the slug/plug flow motion in a preferential direction, increasing the device overall performance. Pressure measurements point out that the pressure drop between the evaporator and the condenser are related to the flow pattern. Furthermore, at high heat inputs, the flow regimes recorded for the two fluids are very similar, stressing that, when the dynamic effects start to play a major role in the system, the device classification between Loop Thermosyphon and Pulsating Heat Pipe is not that sharp anymore.

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.

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.

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)

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

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

Improved Application of Local Models to Steel Corrosion in Lead-Bismuth Loops

International Nuclear Information System (INIS)

Zhang Jinsuo; Li Ning

2003-01-01

The corrosion of steels exposed to flowing liquid metals is influenced by local and global conditions of flow systems. The present study improves the previous local models when applied to closed loops by incorporating some global condition effects. In particular the bulk corrosion product concentration is calculated based on balancing the dissolution and precipitation in the entire closed loop. Mass transfer expressions developed in aqueous medium and an analytical expression are tested in the liquid-metal environments. The improved model is applied to a pure lead loop and produces results closer to the experimental data than the previous local models do. The model is also applied to a lead-bismuth eutectic (LBE) test loop. Systematic studies illustrate the effects of the flow rate, the oxygen concentration in LBE, and the temperature profile on the corrosion rate

Power flow control based solely on slow feedback loop for heart pump applications.

Science.gov (United States)

Wang, Bob; Hu, Aiguo Patrick; Budgett, David

2012-06-01

This paper proposes a new control method for regulating power flow via transcutaneous energy transfer (TET) for implantable heart pumps. Previous work on power flow controller requires a fast feedback loop that needs additional switching devices and resonant capacitors to be added to the primary converter. The proposed power flow controller eliminates these additional components, and it relies solely on a slow feedback loop to directly drive the primary converter to meet the heart pump power demand and ensure zero voltage switching. A controlled change in switching frequency varies the resonant tank shorting period of a current-fed push-pull resonant converter, thus changing the magnitude of the primary resonant voltage, as well as the tuning between primary and secondary resonant tanks. The proposed controller has been implemented successfully using an analogue circuit and has reached an end-to-end power efficiency of 79.6% at 10 W with a switching frequency regulation range of 149.3 kHz to 182.2 kHz.

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

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

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

Conceptual design of helium experimental loop

International Nuclear Information System (INIS)

Yu Xingfu; Feng Kaiming

2007-01-01

In a future demonstration fusion power station (DEMO), helium is envisaged as coolant for plasma facing components, such as blanket and dive,or. All these components have a very complex geometry, with many parallel cooling channels, involving a complex helium flow distribution. Test blanket modules (TBM) of this concept will under go various tests in the experimental reactor ITER. For the qualification of TBM, it is indispensable to test mock-ups in a helium loop under realistic pressure and temperature profiles, in order to validate design codes, especially regarding mass flow and heat transition processes in narrow cooling channels. Similar testing must be performed for DEMO blanket, currently under development. A Helium Experimental Loop (HELOOP) is planed to be built for TBM tests. The design parameter of temperature, pressure, flow rate is 550 degree C, 10 MPa, l kg/s respectively. In particular, HELOOP is able to: perform full-scale tests of TBM under realistic conditions; test other components of the He-cooling system in ITER; qualify the purification circuit; obtain information for the design of the ITER cooling system. The main requirements and characteristics of the HELOOP facility and a preliminary conceptual design are described in the paper. (authors)

Flow characteristics of natural circulation in a lead-bismuth eutectic loop

Institute of Scientific and Technical Information of China (English)

Chen-Chong Yue; Liu-Li Chen; Ke-Feng Lyu; Yang Li; Sheng Gao; Yue-Jing Liu; Qun-Ying Huang

2017-01-01

Lead and lead-alloys are proposed in future advanced nuclear system as coolant and spallation target.To test the natural circulation and gas-lift and obtain thermal-hydraulics data for computational fluid dynamics (CFD) and system code validation,a lead-bismuth eutectic rectangular loop,the KYLIN-Ⅱ Thermal Hydraulic natural circulation test loop,has been designed and constructed by the FDS team.In this paper,theoretical analysis on natural circulation thermal-hydraulic performance is described and the steady-state natural circulation experiment is performed.The results indicated that the natural circulation capability depends on the loop resistance and the temperature and center height differences between the hot and cold legs.The theoretical analysis results agree well with,while the CFD deviate from,the experimental results.

A detailed BWR recirculation loop model for RELAP

Energy Technology Data Exchange (ETDEWEB)

Araiza-Martínez, Enrique, E-mail: enrique.araiza@inin.gob.mx; Ortiz-Villafuerte, Javier, E-mail: javier.ortiz@inin.gob.mx; Castillo-Durán, Rogelio, E-mail: rogelio.castillo@inin.gob.mx

2017-01-15

Highlights: • A new detailed BWR recirculation loop model was developed for RELAP. • All jet pumps, risers, manifold, suction and control valves, and recirculation pump are modeled. • Model is tested against data from partial blockage of two jet pumps. • For practical applications, simulation results showed good agreement with available data. - Abstract: A new detailed geometric model of the whole recirculation loop of a BWR has been developed for the code RELAP. This detailed model includes the 10 jet pumps, 5 risers, manifold, suction and control valves, and the recirculation pump, per recirculation loop. The model is tested against data from an event of partial blockage at the entrance nozzle of one jet pump in both recirculation loops. For practical applications, simulation results showed good agreement with data. Then, values of parameters considered as figure of merit (reactor power, dome pressure, core flow, among others) for this event are compared against those from the common 1 jet pump per loop model. The results show that new detailed model led to a closer prediction of the reported power change. The detailed recirculation loop model can provide more reliable boundary condition data to a CFD models for studies of, for example, flow induced vibration, wear, and crack initiation.

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.

An experimental study of two-phase natural circulation in an adiabatic flow loop

International Nuclear Information System (INIS)

Tan, M.J.; Lambert, G.A.; Ishii, Mamoru.

1988-01-01

An experimental investigation was conducted to study the two-phase flow aspect of the phenomena of interruption and resumption of natural circulation, two-phase flow patterns and pattern transitions in the hot legs of B and W light water reactor systems. The test facility was a scaled adiabatic loop designed in accordance with the scaling criteria developed by Kocamustafaogullari and Ishii. The diameter and the height of the hot leg were 10 cm and 5.5 m, respectively; the working fluid pair was nitrogen-water. The effects of the thermal center in the steam generators, friction loss in the cold leg, and configuration of the inlet to the hot leg on the flow conditions in the hot leg were investigated by varying the water level in a gas separator, controlling the size of opening of a friction loss control valve, and using two inlet geometries. Methods for estimating the distribution parameter and the average drift velocity are proposed so that they may be used in the application of one-dimensional drift-flux model to the analysis of the interruption and resumption of natural circulation in a similar geometry. 7 refs., 17 figs., 4 tabs

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.)

Measurement technique developments for LBE flows

Energy Technology Data Exchange (ETDEWEB)

Buchenau, D., E-mail: d.buchenau@fzd.de [Forschungszentrum Dresden-Rossendorf (FZD), 01314 Dresden (Germany); Eckert, S.; Gerbeth, G. [Forschungszentrum Dresden-Rossendorf (FZD), 01314 Dresden (Germany); Stieglitz, R. [Karlsruhe Institute of Technology (KIT), 76344 Eggenstein-Leopoldshafen (Germany); Dierckx, M. [SCK-CEN, Belgian Nuclear Research Centre, 2400 Mol (Belgium)

2011-08-31

We report on the development of measurement techniques for flows in lead-bismuth eutectic alloys (LBE). This paper covers the test results of newly developed contactless flow rate sensors as well as the development and test of the LIDAR technique for operational free surface level detection. The flow rate sensors are based on the flow-induced disturbance of an externally applied AC magnetic field which manifests itself by a modified amplitude or a modified phase of the AC field. Another concept of a force-free contactless flow meter uses a single cylindrical permanent magnet. The electromagnetic torque on the magnet caused by the liquid metal flow sets the magnet into rotation. The operation of those sensors has been demonstrated at liquid metal test loops for which comparative flow rate measurements are available, as well as at the LBE loops THESYS at KIT and WEBEXPIR at SCK-CEN. For the level detection a commercial LIDAR system was successfully tested at the WEBEXPIR facility in Mol and the THEADES loop in Karlsruhe.

Hummingbirds generate bilateral vortex loops during hovering: evidence from flow visualization

Science.gov (United States)

Pournazeri, Sam; Segre, Paolo S.; Princevac, Marko; Altshuler, Douglas L.

2013-01-01

Visualization of the vortex wake of a flying animal provides understanding of how wingbeat kinematics are translated into the aerodynamic forces for powering and controlling flight. Two general vortex flow patterns have been proposed for the wake of hovering hummingbirds: (1) The two wings form a single, merged vortex ring during each wing stroke; and (2) the two wings form bilateral vortex loops during each wing stroke. The second pattern was proposed after a study with particle image velocimetry that demonstrated bilateral source flows in a horizontal measurement plane underneath hovering Anna's hummingbirds ( Calypte anna). Proof of this hypothesis requires a clear perspective of bilateral pairs of vortices. Here, we used high-speed image sequences (500 frames per second) of C. anna hover feeding within a white plume to visualize the vortex wake from multiple perspectives. The films revealed two key structural features: (1) Two distinct jets of downwards airflow are present under each wing; and (2) vortex loops around each jet are shed during each upstroke and downstroke. To aid in the interpretation of the flow visualization data, we analyzed high-speed kinematic data (1,000 frames per second) of wing tips and wing roots as C. anna hovered in normal air. These data were used to refine several simplified models of vortex topology. The observed flow patterns can be explained by either a single loop model with an hourglass shape or a bilateral model, with the latter being more likely. When hovering in normal air, hummingbirds used an average stroke amplitude of 153.6° (range 148.9°-164.4°) and a wingbeat frequency of 38.5 Hz (range 38.1-39.1 Hz). When hovering in the white plume, hummingbirds used shallower stroke amplitudes ( bar{x} = 129.8°, range 116.3°-154.1°) and faster wingbeat frequencies ( bar{x} = 41.1 Hz, range 38.5-44.7 Hz), although the bilateral jets and associated vortices were observed across the full kinematic range. The plume did not

A LabVIEW model incorporating an open-loop arterial impedance and a closed-loop circulatory system.

Science.gov (United States)

Cole, R T; Lucas, C L; Cascio, W E; Johnson, T A

2005-11-01

While numerous computer models exist for the circulatory system, many are limited in scope, contain unwanted features or incorporate complex components specific to unique experimental situations. Our purpose was to develop a basic, yet multifaceted, computer model of the left heart and systemic circulation in LabVIEW having universal appeal without sacrificing crucial physiologic features. The program we developed employs Windkessel-type impedance models in several open-loop configurations and a closed-loop model coupling a lumped impedance and ventricular pressure source. The open-loop impedance models demonstrate afterload effects on arbitrary aortic pressure/flow inputs. The closed-loop model catalogs the major circulatory waveforms with changes in afterload, preload, and left heart properties. Our model provides an avenue for expanding the use of the ventricular equations through closed-loop coupling that includes a basic coronary circuit. Tested values used for the afterload components and the effects of afterload parameter changes on various waveforms are consistent with published data. We conclude that this model offers the ability to alter several circulatory factors and digitally catalog the most salient features of the pressure/flow waveforms employing a user-friendly platform. These features make the model a useful instructional tool for students as well as a simple experimental tool for cardiovascular research.

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

Simulation and analysis on fields of temperature and flow rate of liquid LIPB in DRAGON-I loop

Energy Technology Data Exchange (ETDEWEB)

Zhu, Z.; Huang, Q.; Zhang, M.; Gao, S.; Wu, Y. [Chinese Academy of Science (China). Inst. of Plasma Physics

2007-07-01

LiPb loop is the most important experimental facility used to study key issues for liquid metal LiPb blanket of fusion reactors. The first thermal convection LiPb loop DRAGON-I was built in 2005 in ASIPP (Institute of Plasma Physics, Chinese Academy of Science), China. The temperatures for the hot leg and cold leg in the loop are 480 C and 420 C, respectively. It is necessary to do research on features and distributions of the fields of temperature and flow rate for liquid metal LiPb in the loop for safe operation of loop and analysis of corrosion behavior of materials used in it. The fields of LiPb temperature and flow rate in the loop were simulated by the popular commercial CFD (Computational Fluid Dynamics) software FLUENT in two-dimensional (2D) and three-dimensional (3D) models. In the simulations and calculations, segregated solver and viscous models of k-epsilon etc. were selected, the properties of LiPb and material of loop pipe were input and the boundary conditions were setup. It was shown that the results for 2D and 3D models were comparable, the temperature field of liquid LiPb was found to be changed continuously between hot leg and cold leg of the loop because of their temperature difference, the temperature of outer-pipes are about 20 C averagely higher than that of the LiPb in the same section of the pipe, the maximum value of thermal stress of pipes was identified near to the bottom of the hot leg. So two or three heating sections in the hot leg might be needed to heat the outer-pipes of hot leg in order to keep the constant temperature of 480 C along the hot leg. The flow rate of LiPb was revealed to be about 0.2 m/s in theory, and it fluctuated little inside the pipe except for the places of upper two corners of the loop. These results will be helpful for the analysis of corrosion behavior of materials with liquid LiPb. (orig.)

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.)

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...

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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

A three-dimensional mathematical model to predict air-cooling flow and temperature distribution of wire loops in the Stelmor air-cooling system

International Nuclear Information System (INIS)

Hong, Lingxiang; Wang, Bo; Feng, Shuai; Yang, Zhiliang; Yu, Yaowei; Peng, Wangjun; Zhang, Jieyu

2017-01-01

Highlights: • A 3-dimentioanl mathematical models for complex wire loops was set up in Stelmor. • The air flow field in the cooling process was simulated. • The convective heat transfer coefficient was simulated coupled with air flow field. • The temperature distribution with distances was predicted. - Abstract: Controlling the forced air cooling conditions in the Stelmor conveyor line is important for improving the microstructure and mechanical properties of steel wire rods. A three-dimensional mathematical model incorporating the turbulent flow of the cooling air and heat transfer of the wire rods was developed to predict the cooling process in the Stelmor air-cooling line of wire rolling mills. The distribution of cooling air from the plenum chamber and the forced convective heat transfer coefficient for the wire loops were simulated at the different locations over the conveyor. The temperature profiles and cooling curves of the wire loops in Stelmor conveyor lines were also calculated by considering the convective heat transfer, radiative heat transfer as well as the latent heat during transformation. The calculated temperature results using this model agreed well with the available measured results in the industrial tests. Thus, it was demonstrated that this model can be useful for studying the air-cooling process and predicting the temperature profile and microstructure evolution of the wire rods.

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.

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.

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.

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.

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)

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

International Nuclear Information System (INIS)

Han, Lu; Bollas, George M.

2016-01-01

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

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.

Compatibility tests of steels in flowing liquid lead-bismuth

International Nuclear Information System (INIS)

Barbier, F.; Benamati, G.; Fazio, C.; Rusanov, A.

2001-01-01

The behaviour of steels exposed to flowing Pb-55Bi was evaluated. The materials tested are the two austenitic steels AISI 316L and 1.4970, and the six martensitic steels Optifer IVc, T91, Batman 27, Batman 28, EP823 and EM10 which were exposed to flowing Pb-55Bi for 1000, 2000 and 3000 h and at two temperatures (573 and 743 K). The corrosion tests were conducted in the non-isothermal loop of IPPE-Obninsk under a controlled oxygen level (10 -6 wt%). The compatibility study showed that at a lower temperature, a very thin oxide layer (<1 μm) was formed on the steels. At higher temperature, austenitic steels also exhibited a thin oxide layer sufficient to prevent their dissolution in the melt. A thicker oxide, which grew according to a parabolic law, was observed on the surface of the martensitic steels. The oxidation resistance behaviour of the martensitic steels was correlated with their alloying elements

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)

Liquid Hydrogen Recirculation System for Forced Flow Cooling Test of Superconducting Conductors

Science.gov (United States)

Shirai, Y.; Kainuma, T.; Shigeta, H.; Shiotsu, M.; Tatsumoto, H.; Naruo, Y.; Kobayashi, H.; Nonaka, S.; Inatani, Y.; Yoshinaga, S.

2017-12-01

The knowledge of forced flow heat transfer characteristics of liquid hydrogen (LH2) is important and necessary for design and cooling analysis of high critical temperature superconducting devices. However, there are few test facilities available for LH2 forced flow cooling for superconductors. A test system to provide a LH2 forced flow (∼10 m/s) of a short period (less than 100 s) has been developed. The test system was composed of two LH2 tanks connected by a transfer line with a controllable valve, in which the forced flow rate and its period were limited by the storage capacity of tanks. In this paper, a liquid hydrogen recirculation system, which was designed and fabricated in order to study characteristics of superconducting cables in a stable forced flow of liquid hydrogen for longer period, was described. This LH2 loop system consists of a centrifugal pump with dynamic gas bearings, a heat exchanger which is immersed in a liquid hydrogen tank, and a buffer tank where a test section (superconducting wires or cables) is set. The buffer tank has LHe cooled superconducting magnet which can produce an external magnetic field (up to 7T) at the test section. A performance test was conducted. The maximum flow rate was 43.7 g/s. The lowest temperature was 22.5 K. It was confirmed that the liquid hydrogen can stably circulate for 7 hours.

Construction of helium engineering demonstration loop (HENDEL M+A) for VHTR

International Nuclear Information System (INIS)

Shimomura, Saneaki; Tanaka, Toshiyuki; Nakano, Tadasuke

1983-01-01

The mother and adapter sections of the large structural component demonstration test loop, alias Helium Engineering Demonstration Loop, for the multipurpose, high temperature gas-cooled experimental reactor were completed in March, 1982. This facility was constructed by Fuji Electric Co., Ltd. and Kawasaki Heavy Industries Ltd. as the main contractors, and by the cooperation with Mitsubishi Heavy Industries Ltd. and Ishikawajima Harima Heavy Industries Co., Ltd. The HENDEL M+A is the testing facility of the largest scale in the world, which can handle 1000 deg C, 40 kgf/cm 2 G helium at a half flow rate of one core cooling loop of the experimental reactor. With the HENDEL M+A, the demonstration tests of fuel assembly stacks, in-core structures, large flow rate and high temperature equipment are planned. The HENDEL M+A comprises two mother loops, an adapter loop, and common auxiliary systems fon measurement and control (In), refining (Mp), makeup (Mu) and cooling water (Uc). The construction and function of such main equipment as a heater, circulators and internally insulated piping are described. The progress of the construction and the main experience during the construction, the process of operation and the performance are reported. (Kako, I.)

Free convection in a partially submerged fluid loop

International Nuclear Information System (INIS)

Britt, T.E.; Wood, D.C.

1982-01-01

Several natural convection loop systems are studied in order to determine the operational characteristics for a multiple loop container which is used to cool failed nuclear reactor assemblies. Both analytical and experimental studies were undertaken to examine flow in both circular and rectangular flow loops. It was found that when a circular loop is heated at the bottom and cooled at the top, recirculation cells form at all input power fluxes. At fluxes between 0.1 W/cm 2 and 0.7 W/cm 2 the cells caused flow oscillations and reversals. With the circular loop heated from the side, no recirculation cells were observed at the power fluxes up to 1.5 W/cm. Boiling did not occur in the circular loop. For a rectangular loop heated and cooled on its vertical sides, no recirculation cells or flow reversals were seen. At input power fluxes above 1.2 W/cm 2 , periodic boiling in the heated side caused flow oscillations

Analysis of data obtained in two-phase flow tests of primary heat transport pumps

International Nuclear Information System (INIS)

Currie, T.C.

1986-06-01

This report analyzes data obtained in two-phase flow tests of primary heat transport pumps performed during the period 1980-1983. Phenomena which have been known to cause pump-induced flow oscillations in pressurized piping systems under two-phase conditions are reviewed and the data analyzed to determine whether any of the identified phenomena could have been responsible for the instabilities observed in those tests. Tentative explanations for the most severe instabilities are given based on those analyses. It is shown that suction pipe geometry probably plays an important role in promoting instabilities, so additional experiments to investigate the effect of suction pipe geometry on the stability of flow in a closed pipe loop under two-phase conditions are recommended

Regulation of liquid metal coolant flow rate in experimental loops

International Nuclear Information System (INIS)

Kozlov, F.A.; Laptev, G.I.

1987-01-01

The possibility to use the VRT-2, RPA-T and R 133 analog temperature regulators for the automated regulation of liquid metal flow rate in the experimental loops for investigations on sodium and sodium-potassium alloy technology is considered. The RPA-T device is shown to be the most convenient one; it is characterized by the following parameters: measuring modulus transfer coefficient is 500; the range of regulating modulus proportionality factor variation - 0.3 - 50; the range of the regulating modulus intergrating time constant variation - 5 - 500 s

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.)

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

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.

Evaluation report on CCTF CORE-I REFLOOD TEST Cl-1, (Run 010)

International Nuclear Information System (INIS)

Sudoh, Takashi; Murao, Yoshio.

1983-09-01

This report describes the effects of the loop flow resistance on the thermohydraulic behavior in the primary system during the reflood phase. The investigation is based on the results of the test Cl-1 which was performed with increased loop flow resistance in the Cylindrical Core Test Facility (CCTF) at Japan Atomic Energy Research Institute. The loop flow resistance was about 40% higher in the present test than in the reference test Cl-5. The results of two tests were compared and the following conclusions were obtained: 1) The total loop flow rate and the core flooding rate were reduced by about 20% with the increased loop flow resistance 2) The core heat transfer was also lowered, then, the turnaround and the quench times extended at the locations above the core midplane. 3) The measured maximum temperature in the core was 50 K higher for the present test than for the reference test. (author)

Experimental and theoretical study on density wave instability in low quality natural circulation loop

Energy Technology Data Exchange (ETDEWEB)

Jia, Hai Jun [Tsinghua Univ., Beijing, BJ (China); Song, Jin Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-07-01

A series of thermal-hydraulic experimental research has been performed at INET of Beijing University in two full scale test loops, HRTL-5 and HRTL-200 simulating the HR-5 and HR-200 district heating reactors. The homogeneous equilibrium flow model and the drift flux model are employed to analyze the flow in the HRTL-200 test loop. The frequency domain linear stability analysis program LINTAB has been developed based on the homogeneous equilibrium results from HRTL-200 test facility have been used to validate the LINTAB code. Analysis results using LINTAB showed a good agreement with the test results. (author). 20 refs., 24 figs.

Experimental and theoretical study on density wave instability in low quality natural circulation loop

International Nuclear Information System (INIS)

Jia, Hai Jun; Song, Jin Ho

1997-07-01

A series of thermal-hydraulic experimental research has been performed at INET of Beijing University in two full scale test loops, HRTL-5 and HRTL-200 simulating the HR-5 and HR-200 district heating reactors. The homogeneous equilibrium flow model and the drift flux model are employed to analyze the flow in the HRTL-200 test loop. The frequency domain linear stability analysis program LINTAB has been developed based on the homogeneous equilibrium results from HRTL-200 test facility have been used to validate the LINTAB code. Analysis results using LINTAB showed a good agreement with the test results. (author). 20 refs., 24 figs

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

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.

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

Experimental data report for transient flow calibration facility tests IIB101, IIB102 and IIB201

International Nuclear Information System (INIS)

Wambach, J.L.

1980-01-01

Thermal-hydraulic response data are presented for the transient performance tests of a pitot tube rake (IIB201) and a modular drag disc-turbine transducer (DTT) rake (IIB101, IIB102). The tests were conducted in a system which provided full scale simulation of the pressure vessel and broken loop hot leg piping of the Loss of Fluid Test Facility (LOFT). A load cell system was used to provide a reference mass flow rate measurement

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)

Analytical study of flow instability behaviour in a boiling two-phase natural circulation loop under low quality conditions

International Nuclear Information System (INIS)

Nayak, A.K.; Kumar, N.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

2002-01-01

Analytical investigations have been carried out to study the flow instability behaviour in a boiling two-phase natural circulation loop under low quality conditions. For this purpose, the computer code TINFLO-S has been developed. The code solves the conservation equations of mass, momentum and energy and equation of state for homogeneous equilibrium twophase flow using linear analytical technique. The results of the code have been validated with the experimental data of the loop for both the steady state and stability. The study reveals that the stability behaviour of low quality flow oscillations is different from that of the high quality flow oscillations. The instability reduces with increase in power and throttling at the inlet of the heater. The instability first increases and then reduces with increase in pressure at any subcooling. The effects of diameter of riser pipe, heater and the height of the riser on this instability are also investigated. (orig.) [de

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.

Multiphase Venturi Dual Energy Gamma Ray combination performance in NUEX flow loop; Desempenho no flowloop do NUEX da medicao multifasica Venturi Dual Energy Gamma Ray

Energy Technology Data Exchange (ETDEWEB)

Barreiros, Claudio; Taranto, Cleber; Costa, Alcemir [PETROBRAS S.A., Rio de Janeiro, RJ (Brazil); Pinguet, Bruno; Heluey, Vitor; Bessa, Fabiano; Loicq, Olivier [Schlumberger Servicos de Petroleo Ltda., Rio de Janeiro, RJ (Brazil)

2008-07-01

The Multiphase Venturi Dual Energy Gamma Ray Combination, Vx* technology, arrived in Brazil in 2000. PETROBRAS, Brazilian Oil Company, has been putting big efforts in its production business and also has demonstrated a large interest in having a multiphase meter approved by ANP for back allocation purposes. The oil industry was looking for ways to improve the back allocation process using an approved on line multiphase flow measurement device, thus replacing punctual test done today by a permanent monitoring device. Considering this scenario, a partnership project between PETROBRAS and Schlumberger was created in Brazil. The main objective of this project, which was held in NUEX flow loop, was to demonstrate to INMETRO (Brazilian Metrology Institute) that the Multiphase Venturi Dual Energy Gamma Ray Combination meter is able to be used for back allocation purpose. PETROBRAS and Schlumberger elaborated a complete methodology in the NUEX flow loop to demonstrate the results and benefits of the Multiphase Venturi Dual Energy Gamma Ray Combination meter. The test was witnessed by INMETRO and had a very good performance at the end. The results were within what was expected by Schlumberger, PETROBRAS and INMETRO. These results has been very useful to PETROBRAS in order to start using the Venturi Dual Energy Gamma Ray technology for well allocation purposes. (author)

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

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.)

A flow test for calibrating 177 core tubes of 1/5-scale reactor flow model for Yonggwang nuclear units 3 and 4

International Nuclear Information System (INIS)

Lee, Byung Jin; Jang, Ho Cheol; Cheong, Jong Sik; Kuh, Jung Eui

1990-01-01

A flow test was performed to find out the hydraulic characteristics of every one of 177 core tubes, representing a fuel assembly respectively, as a preparatory step of 1/5 scale reactor flow model test for Yonggwang Nuclear Units (hereafter YGN) 3 and 4. The axial hydraulic resistance of the fuel assembly was simulated in the square core tube with six orifice plates positioned along the tube length; core support structure below each fuel assembly was done in the core upstream geometry section of the test loop. For each core tube the pressure differentials across the inlet, exit orifice plate and overall tube length were measured, along with the flow rates and temperatures of the test fluid. The measured pressure drops were converted to pressure loss or flow metering coefficients. The metering coefficient of the inlet orifice plate was sensitive to the configuration and location of the upstream geometry. The hydraulic resistance of the core tubes were reasonably coincided with a target value and consistent. The polynomial curve fits of the calibrated coefficients for the 177 core tubes were obtained with reasonable data scatters

Flow analysis of HANARO flow simulated test facility

International Nuclear Information System (INIS)

Park, Yong-Chul; Cho, Yeong-Garp; Wu, Jong-Sub; Jun, Byung-Jin

2002-01-01

The HANARO, a multi-purpose research reactor of 30 MWth open-tank-in-pool type, has been under normal operation since its initial critical in February, 1995. Many experiments should be safely performed to activate the utilization of the NANARO. A flow simulated test facility is being developed for the endurance test of reactivity control units for extended life times and the verification of structural integrity of those experimental facilities prior to loading in the HANARO. This test facility is composed of three major parts; a half-core structure assembly, flow circulation system and support system. The half-core structure assembly is composed of plenum, grid plate, core channel with flow tubes, chimney and dummy pool. The flow channels are to be filled with flow orifices to simulate core channels. This test facility must simulate similar flow characteristics to the HANARO. This paper, therefore, describes an analytical analysis to study the flow behavior of the test facility. The computational flow analysis has been performed for the verification of flow structure and similarity of this test facility assuming that flow rates and pressure differences of the core channel are constant. The shapes of flow orifices were determined by the trial and error method based on the design requirements of core channel. The computer analysis program with standard k - ε turbulence model was applied to three-dimensional analysis. The results of flow simulation showed a similar flow characteristic with that of the HANARO and satisfied the design requirements of this test facility. The shape of flow orifices used in this numerical simulation can be adapted for manufacturing requirements. The flow rate and the pressure difference through core channel proved by this simulation can be used as the design requirements of the flow system. The analysis results will be verified with the results of the flow test after construction of the flow system. (author)

Results of the General Atomic deposition loop program

International Nuclear Information System (INIS)

Hanson, D.L.

1976-01-01

The transport behavior of fission products in flowing helium streams has been studied to determine their deposition and re-entrainment characteristics. Such information is required for the design and safety analysis of high-temperature gas-cooled reactors (HTGRs). A small high-pressure, high-temperature loop was constructed for deposition studies at near-HTGR conditions. Five loop experiments were performed to determine the plateout distribution of iodine, strontium, and cesium. In general, the plateout activity showed an exponential decrease with distance from the source with enhanced plateout at flow disturber locations (contractions, bends, etc.) and especially in a chill section where the surface was cooled. Blowdown tests were performed on selected loop specimens to determine the amount of re-entrainment caused by abnormally high wall shear stresses. The liftoff fraction (fractional amount removed) was shown to vary approximately linearly with the shear ratio (defined as the ratio of the steady state wall shear stress under blowdown conditions to that under normal operating conditions). Blowdown results are also reported for pipe sections taken from the GAIL-IV in-pile loop. Attempts were made to correlate these plateout data with the PAD code (Plateout Activity Distribution) which was developed for prediction of plateout distribution in an HTGR primary circuit. Because of inadequate modeling of the effects of the chill section, the agreement was generally poor. Consequently, to test further the PAD code, a review of the available plateout literature was made. Plateout distributions in the Peach Bottom and Dragon HTGRs and the Battelle Memorial Institute out-of-pile loop were successfully modeled

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.

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.

Steady state flow analysis of two-phase natural circulation in multiple parallel channel loop

International Nuclear Information System (INIS)

Bhusare, V.H.; Bagul, R.K.; Joshi, J.B.; Nayak, A.K.; Kannan, Umasankari; Pilkhwal, D.S.; Vijayan, P.K.

2016-01-01

Highlights: • Liquid circulation velocity increases with increasing superficial gas velocity. • Total two-phase pressure drop decreases with increasing superficial gas velocity. • Channels with larger driving force have maximum circulation velocities. • Good agreement between experimental and model predictions. - Abstract: In this work, steady state flow analysis has been carried out experimentally in order to estimate the liquid circulation velocities and two-phase pressure drop in air–water multichannel circulating loop. Experiments were performed in 15 channel circulating loop. Single phase and two-phase pressure drops in the channels have been measured experimentally and have been compared with theoretical model of Joshi et al. (1990). Experimental measurements show good agreement with model.

Computer simulation of natural circulation in FFTF secondary loops

International Nuclear Information System (INIS)

Beaver, T.R.; Turner, D.M.; Additon, S.L.

1979-07-01

A thermal/hydraulic model of the FFTF secondary heat transport loop has been calibrated against transient natural circulation test data collected March to May 1979. The tests verified that the transition to natural convective flow could be effected from near isothermal conditions without excessive cooling at the air dump heat exchangers. Key empirical parameters of pressure drop and heat loss were found to be at 88% and 81% of pretest estimates, respectively. Pretest piping thermal transport and flow calculational models required no further revision to produce good agreement with test data

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

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.

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.

Summary and evaluation: fuel dynamics loss-of-flow experiments (tests L2, L3, and L4)

International Nuclear Information System (INIS)

Barts, E.W.; Deitrich, L.W.; Eberhart, J.G.; Fischer, A.K.; Meek, C.C.

1975-09-01

Three similar experiments conducted to support the analyses of hypothetical LMFBR unprotected-loss-of-flow accidents are summarized and evaluated. The tests, designated L2, L3, and L4, provided experimental data against which accident-analysis codes could be compared, so as to guide further analysis and modeling of the initiating phases of the hypothetical accident. The tests were conducted using seven-pin bundles of mixed-oxide fuel pins in Mark-II flowing-sodium loops in the TREAT reactor. Test L2 used fresh fuel. Tests L3 and L4 used irradiated fuel pins having, respectively, ''intermediate-power'' (no central void) and ''high-power'' (fully developed central void) microstructure. 12 references

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

UPTF-TRAM test A2. Formation of stratified flow in the hot leg

International Nuclear Information System (INIS)

Tenckhoff; Brand, B.; Weiss, P.

1992-10-01

The separate effect UPTF TRAM Test A2 consisting of six runs was designed to investigate flow regimes in the hot leg of a pressurized water reactor under two-phase natural circulation conditions. In particular, the following phenomena were investigated: - Formation of different flow regimes, e.g. stratified and slug flow in the hot leg under different boundary conditions; -Correlation between flow regime and boundary conditions of the system (mass flows, water level etc.); - Mechanism of the transport of water into the steam generator. The test runs are divided into two groups: a) Test Runs 01a, 01b and 02b with steam injection through the core simulator: In these test runs the steam injection through the core simulator was increased stepwise. In each step the steam injection was kept constant for about 100 s in order to observe steady water distribution in the hot leg and SG-simulator of broken loop. b) Test Runs 03c, 04c and 04d with steam and water injection through the core simulator: These test runs were performed at a constant steam injection rate and the water injection rate was increased stepwise. In order to verify the consistency of scaling with the pressure, the test runs were carried out at different pressures as: a) Runs 01a and 01b at 15 bar, and Run 02b at 3 bar b) Runs 03c, 04c and 04d at 15, 3 and 5 bar respectively. A preliminary evaluation of the test is presented in the Quick Look Report. (orig.) [de

METAL:LIC target failure diagnostics by means of liquid metal loop vibrations monitoring

International Nuclear Information System (INIS)

Dementjevs, S.; Barbagallo, F.; Wohlmuther, M.; Thomsen, K.; Zik, A.; Nikoluskins, R.

2014-01-01

A target mock-up, developed as an European Spallation Source comparative solution, (METAL:LIC) has been tested in a dedicated lead bismuth eutectic (LBE) loop in the Institute of Physics at the University of Latvia. In particular, the feasibility of diagnostic vibration monitoring has been investigated. The loop parameters were: operation temperature 300°C; tubing ∅100 mm, overall length 8 m; electromagnetic pump based on permanent magnets, flow rate 180 kg/s. With sufficient static pressure of a few bars, cavitation was avoided. The vibrations in the loop were measured and analyzed. Several vibrational characteristics of the set-up were derived including resonance frequencies and the dependence of excited vibrations on flow conditions and the pump rotation speed. A high sensitivity to obstructions in the loop has been confirmed, and several indicators for target failure diagnostics were tested and compared. A problem in the electromagnetic pump's gear box has been detected in a very early state long before it manifested itself in the operation of the loop. The vibration monitoring has been demonstrated as a sensitive and reliable probe for the target failure diagnostics. (author)

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.

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.

Understanding the self-sustained oscillating two-phase flow motion in a closed loop pulsating heat pipe

International Nuclear Information System (INIS)

Spinato, Giulia; Borhani, Navid; Thome, John R.

2015-01-01

In the framework of efficient thermal management schemes, pulsating heat pipes (PHPs) represent a breakthrough solution for passive on-chip two-phase flow cooling of micro-electronics. Unfortunately, the unique coupling of thermodynamics, hydrodynamics and heat transfer, responsible for the self-sustained pulsating two-phase flow in such devices, presents many challenges to the understanding of the underlying physical phenomena which have so far eluded accurate prediction. In this experimental study, the novel time-strip image processing technique was used to investigate the thermo-flow dynamics of a single-turn channel CLPHP (closed loop pulsating heat pipe) charged with R245fa and tested under different operating conditions. The resulting frequency data confirmed the effect of flow pattern, and thus operating conditions, on the oscillating behavior. Dominant frequencies from 1.2 Hz for the oscillating regime to 0.6 Hz for the unidirectional flow circulation regime were measured, whilst wide spectral bands were observed for the unstable circulation regime. In order to analytically assess the observed trends in the spectral behavior, a spring-mass-damper system model was developed for the two-phase flow motion. As well as showing that system stiffness and mass have an effect on the two-phase flow dynamics, further insights into the flow pattern transition mechanism were also gained. - Highlights: • A novel synchronized thermal and visual investigation technique was applied to a CLPHP. • Thermal and hydrodynamic behaviors were analyzed by means of spectral analysis. • 3D frequency spectra for temperature and flow data show significant trends. • A spring-mass-damper system model was developed for the two-phase flow motion. • System stiffness and mass have an effect on the two-phase flow dynamics.

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

Assessment of RELAP5-3D copyright using data from two-dimensional RPI flow tests

International Nuclear Information System (INIS)

Davis, C.B.

1998-01-01

The capability of the RELAP5-3D copyright computer code to perform multi-dimensional thermal-hydraulic analysis was assessed using data from steady-state flow tests conducted at Rensselaer Polytechnic Institute (RPI). The RPI data were taken in a two-dimensional test section in a low-pressure air/water loop. The test section consisted of a thin vertical channel that simulated a two-dimensional slice through the core of a pressurized water reactor. Single-phase and two-phase flows were supplied to the test section in an asymmetric manner to generate a two-dimensional flow field. A traversing gamma densitometer was used to measure void fraction at many locations in the test section. High speed photographs provided information on the flow patterns and flow regimes. The RPI test section was modeled using the multi-dimensional component in RELAP5-3D Version BF06. Calculations of three RPI experiments were performed. The flow regimes predicted by the base code were in poor agreement with those observed in the tests. The two-phase regions were observed to be in the bubbly and slug flow regimes in the test. However, nearly all of the junctions in the horizontal direction were calculated to be in the stratified flow regime because of the relatively low velocities in that direction. As a result, the void fraction predictions were also in poor agreement with the measured values. Significantly improved results were obtained in sensitivity calculations with a modified version of the code that prevented the horizontal junctions from entering the stratified flow regime. These results indicate that the code's logic in the determination of flow regimes in a multi-dimensional component must be improved. The results of the sensitivity calculations also indicate that RELAP5-3D will provide a significant multi-dimensional hydraulic analysis capability once the flow regime prediction is improved

Advanced multi-evaporator loop thermosyphon

International Nuclear Information System (INIS)

Mameli, M.; Mangini, D.; Vanoli, G.F.T.; Araneo, L.; Filippeschi, S.; Marengo, M.

2016-01-01

A novel prototype of multi-evaporator closed loop thermosyphon is designed and tested at different heaters position, inclinations and heat input levels, in order to prove that a peculiar arrangement of multiple heaters may be used in order to enhance the flow motion and consequently the thermal performance. The device consists in an aluminum tube (Inner/Outer tube diameter 3.0 mm/5.0 mm), bent into a planar serpentine with five U-turns and partially filled with FC-72, 50% vol. The evaporator zone is equipped with five heated patches (one for each U-turn) in series with respect to the flow path. In the first arrangement, heaters are wrapped on each bend symmetrically, while in the second layout heaters are located on the branch just above the U-turn, non-symmetrical with respect to the gravity direction, in order to promote the fluid circulation in a preferential direction. The condenser zone is cooled by forced air and equipped with a 50 mm transparent section for the flow pattern visualization. The non-symmetrical heater arrangement effectively promotes a stable fluid circulation and a reliable operation for a wider range of heat input levels and orientations with respect to the symmetrical case. In vertical position, the heat flux dissipation exceeds the pool boiling heat transfer limit for FC-72 by 75% and the tube wall temperatures in the evaporator zone are kept lower than 80 °C. Furthermore, the heat flux capability is up to five times larger with respect to the other existing wickless heat pipe technologies demonstrating the attractiveness of the new concept for electronic cooling thermal management. - Highlights: • A novel passive heat transfer device named Multi-Evaporator Loop Thermosyphon is tested. • The loop is investigated at different heating patterns, inclinations and heat power levels. • The non-symmetrical heating configuration promotes the fluid circulation within the loop. • The performance in terms of maximum heat flux exceeds the

Fast Flux Test Facility primary sodium valves

International Nuclear Information System (INIS)

Rabe, G.B.; Ezra, B.C.

1977-01-01

The design and development of the valves used in the primary sodium coolant loop of the Fast Flux Test Facility is described. One tilting-disk check valve is used in the cold leg of the coolant loop. It is designed to limit flow reversal in the loop while maintaining a low pressure drop during forward flow. Two isolation valves are used in each coolant loop--one in the cold leg and one in the hot leg. They are of the motor-operated swinging-gate type. The design, analysis, and testing programs undertaken to develop and qualify these valves are described

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.

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.

Performance Evaluation of the International Space Station Flow Boiling and Condensation Experiment (FBCE) Test Facility

Science.gov (United States)

Hasan, Mohammad; Balasubramaniam, R.; Nahra, Henry; Mackey, Jeff; Hall, Nancy; Frankenfield, Bruce; Harpster, George; May, Rochelle; Mudawar, Issam; Kharangate, Chirag R.;

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.

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...

PHOTOSPHERIC PROPERTIES OF WARM EUV LOOPS AND HOT X-RAY LOOPS

Energy Technology Data Exchange (ETDEWEB)

Kano, R. [National Astronomical Observatory of Japan, 2-21-1 Osawa, Mitaka, Tokyo 181-8588 (Japan); Ueda, K. [Department of Astronomy, Graduate School of Science, University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-0033 (Japan); Tsuneta, S., E-mail: ryouhei.kano@nao.ac.jp [Institute of Space and Astronautical Science, Japan Aerospace Exploration Agency, 3-1-1 Yoshinodai, Chuo, Sagamihara, Kanagawa 252-5210 (Japan)

2014-02-20

We investigate the photospheric properties (vector magnetic fields and horizontal velocity) of a well-developed active region, NOAA AR 10978, using the Hinode Solar Optical Telescope specifically to determine what gives rise to the temperature difference between ''warm loops'' (1-2 MK), which are coronal loops observed in EUV wavelengths, and ''hot loops'' (>3 MK), coronal loops observed in X-rays. We found that outside sunspots, the magnetic filling factor in the solar network varies with location and is anti-correlated with the horizontal random velocity. If we accept that the observed magnetic features consist of unresolved magnetic flux tubes, this anti-correlation can be explained by the ensemble average of flux-tube motion driven by small-scale random flows. The observed data are consistent with a flux tube width of ∼77 km and horizontal flow at ∼2.6 km s{sup –1} with a spatial scale of ∼120 km. We also found that outside sunspots, there is no significant difference between warm and hot loops either in the magnetic properties (except for the inclination) or in the horizontal random velocity at their footpoints, which are identified with the Hinode X-Ray Telescope and the Transition Region and Coronal Explorer. The energy flux injected into the coronal loops by the observed photospheric motion of the magnetic fields is estimated to be 2 × 10{sup 6} erg s{sup –1} cm{sup –2}, which is the same for both warm and hot loops. This suggests that coronal properties (e.g., loop length) play a more important role in giving rise to temperature differences of active-region coronal loops than photospheric parameters.

High temperature storage loop :

Energy Technology Data Exchange (ETDEWEB)

Gill, David Dennis; Kolb, William J.

2013-07-01

A three year plan for thermal energy storage (TES) research was created at Sandia National Laboratories in the spring of 2012. This plan included a strategic goal of providing test capability for Sandia and for the nation in which to evaluate high temperature storage (>650ÀC) technology. The plan was to scope, design, and build a flow loop that would be compatible with a multitude of high temperature heat transfer/storage fluids. The High Temperature Storage Loop (HTSL) would be reconfigurable so that it was useful for not only storage testing, but also for high temperature receiver testing and high efficiency power cycle testing as well. In that way, HTSL was part of a much larger strategy for Sandia to provide a research and testing platform that would be integral for the evaluation of individual technologies funded under the SunShot program. DOEs SunShot program seeks to reduce the price of solar technologies to 6/kWhr to be cost competitive with carbon-based fuels. The HTSL project sought to provide evaluation capability for these SunShot supported technologies. This report includes the scoping, design, and budgetary costing aspects of this effort

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

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.

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.

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.

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...

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)

A review of investigations on flow instabilities in natural circulation boiling loops

International Nuclear Information System (INIS)

Gonella V Durga Prasad; Manmohan Pandey; Manjeet S Kalra

2005-01-01

Full text of publication follows: Steam generation systems are subjected to flow instabilities due to parametric fluctuations, inlet conditions etc., which may result in mechanical vibrations of components (called flow induced vibrations) and system control problems. Analysis of these instabilities in natural circulation boiling loops is very important for the safety of nuclear reactors and other boiling systems. This paper presents the state of the art in this area by reviewing over 100 contributions made in the past 30 years. A large number of experimental and numerical investigations have been conducted to study and understand the conditions for inception of flow instabilities, parametric effects of instabilities, and the system behavior under such conditions. Work done on instabilities due to channel thermal-hydraulics as well as neutronics-thermohydraulics coupling has been reviewed. Different methods of analysis used by researchers and results obtained by them have been discussed. Various numerical techniques adopted and computer codes developed have also been discussed. The knowledge obtained from the investigations made in the past three decades has been summarized to present the state of the art of the understanding of flow instabilities. (authors)

Concepts of self-acting circulation loops for downward heat transfer (reverse thermosiphons)

International Nuclear Information System (INIS)

Dobriansky, Y.

2011-01-01

This paper reviews the scientific and technical knowledge related to general self-acting flow loops (thermosiphons and heat pipes) that transmit heat upwards and self-acting reverse flow loops that transmit heat downwards. This paper classifies the heat and mass transfer processes that take place in general flow loops and analyses the nomenclature applied in the literature. It also presents the principles of operation of sixteen reverse flow loops; four of the loops are powered by an external source of energy, while the remaining loops are self-acting. Of the self-acting loops, vapor was used for heat transfer in seven of them and liquid was used in the remaining ones. Based on the available research results, a list of the advantages and disadvantages of both types of loops is presented.

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

International Nuclear Information System (INIS)

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

1991-01-01

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

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)

Technical description of the test section SUCOT to investigate a water/steam two-phase flow

International Nuclear Information System (INIS)

Daubner, M.; Janssens-Maenhout, G.; Knebel, J.U.

2002-02-01

Within the POOLTHY Project of the European Union (Euratom Fourth Framework Programme contract FJ4J-CT95-0003) an active/passive concept (SUCO-Programme) was investigated which controls the heat removal after a potential core melt-down accident in an evolutionary light water reactor by spreading and stabilising the core melt in the reactor sump and flooding the melt with sump water from above. The experiments were performed in the test facility SUCOT which was designed and erected at the Institute for Nuclear and Energy Technologies (IKET). The report gives an overview on the SUCO-Programme and the scaling analysis, which was applied to design the test facility SUCOT. A detailed technical description of the test facility SUCOT is given, in which the natural circulation driven two-phase flow within the reactor sump and relevant phenomena such as flow boiling, disperse bubbly flow with and without mass transfer, and geysering are investigated. The major components of the test facility, the three-loop system and the instrumentation are described. Finally, a perspective for future application of the gained knowledge is given. (orig.) [de

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)

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.

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.

Comparison of numerical results with experimental data for single-phase natural convection in an experimental sodium loop

International Nuclear Information System (INIS)

Ribando, R.J.

1979-01-01

A comparison is made between computed results and experimental data for single-phase natural convection in an experimental sodium loop. The tests were conducted in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) Facility, an engineering-scale high temperature sodium facility at the Oak Ridge National Laboratory used for thermal-hydraulic testing of simulated LMFBR subassemblies at normal and off-normal operating conditions. Heat generation in the 19 pin assembly during these tests was typical of decay heat levels. Tests were conducted both with zero initial forced flow and with a small initial forced flow. The bypass line was closed in most tests, but open in one. The computer code used to analyze these tests [LONAC (LOw flow and NAtural Convection)] is an ORNL-developed, fast running, one-dimensional, single-phase finite difference model for simulating forced and free convection transients in the THORS loop

Assessment of the adequacy of bronchial stenting by flow-volume loops

Energy Technology Data Exchange (ETDEWEB)

McLaren, Clare A.; Roebuck, Derek J. [Great Ormond Street Hospital for Children, Department of Radiology, London (United Kingdom); Pigott, Nick; Elliott, Martin J. [Great Ormond Street Hospital for Children, Cardiothoracic Unit, London (United Kingdom); Dunne, Catherine [Great Ormond Street Hospital for Children, Department of Physiotherapy, London (United Kingdom)

2006-08-15

Airway compression is a common problem in children with certain forms of congenital heart disease. Although various surgical approaches are available to overcome this form of airway obstruction, internal stenting is necessary in a minority of patients. It can be difficult to assess the success of stenting at the time of the procedure, and the interval to successful extubation is usually used as an outcome measure. Measurement of relevant parameters of respiratory physiology with flow-volume and volume-pressure loops permits immediate quantitative assessment of the adequacy of stenting. A 3-month-old infant who underwent bronchial stenting and physiological assessment at the time of the procedure is described. (orig.)

Assessment of the adequacy of bronchial stenting by flow-volume loops

International Nuclear Information System (INIS)

McLaren, Clare A.; Roebuck, Derek J.; Pigott, Nick; Elliott, Martin J.; Dunne, Catherine

2006-01-01

Airway compression is a common problem in children with certain forms of congenital heart disease. Although various surgical approaches are available to overcome this form of airway obstruction, internal stenting is necessary in a minority of patients. It can be difficult to assess the success of stenting at the time of the procedure, and the interval to successful extubation is usually used as an outcome measure. Measurement of relevant parameters of respiratory physiology with flow-volume and volume-pressure loops permits immediate quantitative assessment of the adequacy of stenting. A 3-month-old infant who underwent bronchial stenting and physiological assessment at the time of the procedure is described. (orig.)

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.

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

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)

Operating problem of low specific speed pumps operating in closed hydraulic loop

International Nuclear Information System (INIS)

Rajput, A.K.

1979-01-01

Results of the studies of pressure pulsations caused by the centrifugal pump driving a typical sodium test loop are presented. The method of characteristics has been used for solving the equations of unsteady fluid flow in closed hydraulic loops with various boundary points, important of which are pump, control valve and heater tank (acting hydraulically as surge tank). Mathematical and computational models used for calculations are described. (M.G.B.)

Performance Tests for Bubble Blockage Device

International Nuclear Information System (INIS)

Ha, Kwang Soon; Wi, Kyung Jin; Park, Rae Joon; Wan, Han Seong

2014-01-01

Postulated severe core damage accidents have a high threat risk for the safety of human health and jeopardize the environment. Versatile measures have been suggested and applied to mitigate severe accidents in nuclear power plants. To improve the thermal margin for the severe accident measures in high-power reactors, engineered corium cooling systems involving boiling-induced two-phase natural circulation have been proposed for decay heat removal. A boiling-induced natural circulation flow is generated in a coolant path between a hot vessel wall and cold coolant reservoir. In general, it is possible for some bubbles to be entrained in the natural circulation loop. If some bubbles entrain in the liquid phase flow passage, flow instability may occur, that is, the natural circulation mass flow rate may be oscillated. A new device to block the entraining bubbles is proposed and verified using air-water test loop. To avoid bubbles entrained in the natural circulation flow loop, a new device was proposed and verified using an air-water test loop. The air injection and liquid circulation loop was prepared, and the tests for the bubble blockage devices were performed by varying the geometry and shape of the devices. The performance of the bubble blockage device was more effective as the area ratio of the inlet to the down-comer increased, and the device height decreased. If the device has a rim to generate a vortex zone, the bubbles will be most effectively blocked

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.

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

Operating experience with gas-bearing circulators in a high-pressure helium loop

International Nuclear Information System (INIS)

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

1988-01-01

A high-pressure engineering test loop has been designed and constructed at the Oak Ridge National Laboratory for circulating helium through a test chamber at temperatures to 1,000 deg. C. The purpose of this loop is to determine the thermal and structural performance of proposed components for the primary loops of gas-cooled nuclear reactors. Three gas-bearing circulators, mounted in series, provide a maximum volumetric flow of 0.47 m 3 /s and a maximum head of 78 kJ/kg at operating pressures from 0.1 to 10.7 MPa. Control of gaseous impurities in the circulating gas was the significant operating requirement that dictated the choice of a circulator that is lubricated by the circulating gas. The motor for each circulator is contained within the pressure boundary, and it is cooled by circulating the gas in the motor cavity over water-cooled coils. Each motor is rated at 200 kW at a speed of 23,500 rpm. The circulators have been operated in the loop for more than 5,000 h. The flow of the gas in the loop is controlled by varying the speed of the circulators through the use of individual 250-kVA, solid state power supplies that can be continuously varied in frequency from 50 to 400 Hz. To prevent excessive wear on the gas bearings during startup, the circulator motor accelerates the rotor to 3,000 rpm in less than one second. During operation, no problems associated with the gas bearings, per se, were encountered; however, related problems pointed to design considerations that should be included in future applications of circulators of this type. The primary test that has been conducted in this loop required sustained operation for several weeks without interruption. After a number of unscheduled interruptions, the operating goals were attained. During part of this period, the loop was operated with only two circulators installed in the pressure vessels with a guard installed in the third vessel to protect the closure flange from the gas temperatures. Unattended

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.

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

Comparison of numerical results with experimental data for single-phase natural convection in an experimental sodium loop

International Nuclear Information System (INIS)

Ribando, R.J.

1979-01-01

A comparison is made between computed results and experimental data for a single-phase natural convection test in an experimental sodium loop. The test was conducted in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility, an engineering-scale high temperature sodium loop at the Oak Ridge National Laboratory (ORNL) used for thermal-hydraulic testing of simulated Liquid Metal Fast Breeder Reactor (LMFBR) subassemblies at normal and off-normal operating conditions. Heat generation in the 19 pin assembly during the test was typical of decay heat levels. The test chosen for analysis in this paper was one of seven natural convection runs conducted in the facility using a variety of initial conditions and testing parameters. Specifically, in this test the bypass line was open to simulate a parallel heated assembly and the test was begun with a pump coastdown from a small initial forced flow. The computer program used to analyze the test, LONAC (LOw flow and NAtural Convection) is an ORNL-developed, fast-running, one-dimensional, single-phase, finite-difference model used for simulating forced and free convection transients in the THORS loop

Dynamical behaviour of natural convection in closed loops

International Nuclear Information System (INIS)

Ehrhard, P.

1988-04-01

A one dimensional model is presented together with experiments, which describe the natural convective flow in closed loops heated at the bottom and cooled in the upper semicircle. Starting from a single loop, mechanical and thermal coupling with a second loop is discussed. The experiments and the theoretical model both concurrently demonstrate that the investigated natural convection is clearly influenced by non-linear effects. Beside the variety of stable steady flows there are extensive subcritical ranges of convective flow. In these parameter ranges subcritical instabilities of the steady state flow could occur in the presence of finite amplitude disturbances. However, the supercritical, global unstable range is characterized by chaotic histories of the variables of state. Non-symmetric heating generates an imperfect bifurcation out of the steady solution with zero velocity in the loop. This effect stabilizes the flow in the preferred direction. The flow in the opposite direction only remains stable in a small isolated interval of the heating parameter. Furthermore the calculations with the model equations demonstrate that a stable periodic behaviour of the flow is possible in a small parameter window. However, it has not been possible to verify this particular effect in the experiments conducted to date. (orig./GL) [de

Experimental investigations and modeling of a loop thermosyphon for cooling with zero electrical consumption

International Nuclear Information System (INIS)

Chehade, Ali; Louahlia-Gualous, Hasna; Le Masson, Stéphane; Lépinasse, Eric

2015-01-01

This paper presents an analytical model for a thermosyphon loop developed for cooling air inside a telecommunication cabinet. The proposed model is based on the combination of thermal and hydraulic management of two-phase flow in the loop. Experimental tests on a closed thermosyphon loop are conducted with different working fluids that could be used for electronic cooling. Correlations for condensation and evaporation heat transfer in the thermosyphon loop are proposed. They are used in the model to calculate condenser and evaporator thermal resistances in order to predict the cabinet operating temperature, the loop's mass flow rate and pressure drops. Furthermore, various figures of merit proposed in the previous works are evaluated in order to be used for selection of the best loop's working fluid. The comparative studies show that the present model well predicts the experimental data. The mean deviation between the predictions of the theoretical model with the measurements for operating temperature is about 6%. Besides, the model is used to define an optimal liquid and vapor lines diameters and the effect of the ambient temperature on the fluid's mass flow rate and pressure drop. - Highlights: • Modeling of thermosyphon loop for cooling telecommunication cabinet. • The cooling system operates with zero electrical consumption. • The new correlations are proposed for condensation and evaporation heat transfer. • FOM equation is defined for selecting the best working fluid. • The proposed model well predicts the experimental data and operating temperature

A statistical learning strategy for closed-loop control of fluid flows

Science.gov (United States)

Guéniat, Florimond; Mathelin, Lionel; Hussaini, M. Yousuff

2016-12-01

This work discusses a closed-loop control strategy for complex systems utilizing scarce and streaming data. A discrete embedding space is first built using hash functions applied to the sensor measurements from which a Markov process model is derived, approximating the complex system's dynamics. A control strategy is then learned using reinforcement learning once rewards relevant with respect to the control objective are identified. This method is designed for experimental configurations, requiring no computations nor prior knowledge of the system, and enjoys intrinsic robustness. It is illustrated on two systems: the control of the transitions of a Lorenz'63 dynamical system, and the control of the drag of a cylinder flow. The method is shown to perform well.

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

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

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.

Experimental data report for transient flow calibration facility tests IIIA101, IIIA102, IIIA201, and IIIA202

International Nuclear Information System (INIS)

Wambach, J.L.

1980-01-01

Thermal-hydraulic response data are presented for the transient performance tests of an ECC pitot tube rake (IIIA201, IIIA202) and both an ECC pitot tube rake and modular drag disc-turbine transducer (DTT) rake (IIIA101, IIIA102). The tests were conducted in a system which provided full scale simulation of the pressure vessel and intact loop cold leg piping of the Loss of Fluid Test Facility (LOFT). A load cell system was used to provide a reference mass flow rate measurement

The Test for Flow Characteristics of Tubular Fuel Assembly(II) - Experimental results and CFD analysis

International Nuclear Information System (INIS)

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

2006-12-01

A test facility had been established for the experiment of velocity distribution and pressure drop in a tubular fuel. A basic test had been conducted to examine the performance of the test loop and to verify the accuracy of measurement by pitot-tube. In this report, test results and CFD analysis for the hydraulic characteristics of a tubular fuel, following the previous tests, are described. Coolant velocities in all channels were measured using pitot-tube and the effect of flow rate change on the velocity distribution was also examined. The pressure drop through the tubular fuel was measured for various flow rates in range of 1 kg/s to 21 kg/s to obtain a correlation of pressure drop with variation of flow rate. In addition, a CFD(Computational Fluid Dynamics) analysis was also done to find out the hydraulic characteristics of tubular fuel such as velocity distribution and pressure drop. As the results of CFD analysis can give us a detail insight on coolant flow in the tubular fuel, the CFD method is a very useful tool to understand the flow structure and phenomena induced by fluid flow. The CFX-10, a commercial CFD code, was used in this study. The two results by the experiment and the CFD analysis were investigated and compared with each other. Overall trend of velocity distribution by CFD analysis was somewhat different from that of experiment, but it would be reasonable considering measurement uncertainties. The CFD prediction for pressure drop of a tubular fuel shows a tolerably good agreement with experiment within 8% difference

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.

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)

Initial liquid metal magnetohydrodynamic thin film flow experiments in the MeGA-loop facility at UCLA

International Nuclear Information System (INIS)

Morley, N.B.; Gaizer, A.A.; Tillack, M.S.; Abdou, M.A.

1995-01-01

Free surface thin film flows of liquid metal were investigated experimentally in the presence of a coplanar magnetic field. This investigation was performed in a new magnetohydrodynamic (MHD) flow facility, the MeGA-loop, utilizing a low melting temperature lead-bismuth alloy as the working metal. Owing to the relatively low magnetic field produced by the present field coil system, the ordinary hydrodynamic and low MHD interaction regimes only were investigated. At the high flow speeds necessary for self cooling, the importance of a well designed and constructed channel becomes obvious. Partial MHD drag, increasing the film height, is observed as Haβ 2 becomes greater than unity. MHD laminarization of the turbulent film flows is observed when Haβ/Re>0.002, but fully laminar flow was not reached. Suggestions for facility upgrades to achieve greater MHD interaction are presented in the context of these initial results. (orig.)

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

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)

Design validation and performance of closed loop gas recirculation system

International Nuclear Information System (INIS)

Kalmani, S.D.; Majumder, G.; Mondal, N.K.; Shinde, R.R.; Joshi, A.V.

2016-01-01

A pilot experimental set up of the India Based Neutrino Observatory's ICAL detector has been operational for the last 4 years at TIFR, Mumbai. Twelve glass RPC detectors of size 2 × 2 m 2 , with a gas gap of 2 mm are under test in a closed loop gas recirculation system. These RPCs are continuously purged individually, with a gas mixture of R134a (C 2 H 2 F 4 ), isobutane (iC 4 H 10 ) and sulphur hexafluoride (SF 6 ) at a steady rate of 360 ml/h to maintain about one volume change a day. To economize gas mixture consumption and to reduce the effluents from being released into the atmosphere, a closed loop system has been designed, fabricated and installed at TIFR. The pressure and flow rate in the loop is controlled by mass flow controllers and pressure transmitters. The performance and integrity of RPCs in the pilot experimental set up is being monitored to assess the effect of periodic fluctuation and transients in atmospheric pressure and temperature, room pressure variation, flow pulsations, uniformity of gas distribution and power failures. The capability of closed loop gas recirculation system to respond to these changes is also studied. The conclusions from the above experiment are presented. The validations of the first design considerations and subsequent modifications have provided improved guidelines for the future design of the engineering module gas system.

Comparison of numerical results with experimental data for single-phase natural convection in an experimental sodium loop. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Ribando, R.J.

1979-01-01

A comparison is made between computed results and experimental data for a single-phase natural convection test in an experimental sodium loop. The test was conducted in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility, an engineering-scale high temperature sodium loop at the Oak Ridge National Laboratory (ORNL) used for thermal-hydraulic testing of simulated Liquid Metal Fast Breeder Reactor (LMFBR) subassemblies at normal and off-normal operating conditions. Heat generation in the 19 pin assembly during the test was typical of decay heat levels. The test chosen for analysis in this paper was one of seven natural convection runs conducted in the facility using a variety of initial conditions and testing parameters. Specifically, in this test the bypass line was open to simulate a parallel heated assembly and the test was begun with a pump coastdown from a small initial forced flow. The computer program used to analyze the test, LONAC (LOw flow and NAtural Convection) is an ORNL-developed, fast-running, one-dimensional, single-phase, finite-difference model used for simulating forced and free convection transients in the THORS loop.

Mixed convection in a two-phase flow cooling loop

International Nuclear Information System (INIS)

Janssens-Maenhout, G.; Daubner, M.; Knebel, J.U.

2002-03-01

This report summarizes the numerical simulations using the CFD code CFX4.1 which has additional models for subcooled flow boiling phenomena and the interfacial forces. The improved CFX4.1 code can be applied to the design of boiling induced mixed convection cooling loops in a defined parameter range. The experimental part describes the geysering experiments and the instability effects on the two-phase natural circulation flow. An experimentally validated flow pattern map in the Phase Change Number - Subcooling Number (N PCh - N Sub ) diagram defines the operational range in which flow instabilities such as geysering can be expected. One important perspective of this combined experimental/numerical work, which is in the field of two-phase flow, is its application to the development of accelerator driven systems (ADS). The main objective on an ADS is its potential to transmute minor actinides and long-lived fission products, thus participating in closing the fuel cycle. The development of an ADS is an important issue within the Euratom Fifth FP on Partitioning and Transmutation. One concept of an ADS, which is investigated in more detail within the ''preliminary design study of an experimental ADS'' Project (PDS-XADS) of the Euratom Fifth FP, is the XADS lead-bismuth cooled Experimental ADS of ANSALDO. An essential feature of this concept is the natural circulation of the primary coolant within the reactor pool. The natural circulation, which is driven by the density differences between the blanket and the heat exchanger, is enhanced by the injection of the nitrogen cover gas through spargers located in a riser part just above the blanket. This so-called gas-lift pump system has not been investigated in more detail nor has this gas-lift pump system been numerically/experimentally confirmed. The knowledge gained within the SUCO Programe, i.e. the modelling of the interfacial forces, the experimental work on flow instabilities and the modelling of the interfacial area

Mixed convection in a two-phase flow cooling loop

Energy Technology Data Exchange (ETDEWEB)

Janssens-Maenhout, G.; Daubner, M.; Knebel, J.U.

2002-03-01

This report summarizes the numerical simulations using the CFD code CFX4.1 which has additional models for subcooled flow boiling phenomena and the interfacial forces. The improved CFX4.1 code can be applied to the design of boiling induced mixed convection cooling loops in a defined parameter range. The experimental part describes the geysering experiments and the instability effects on the two-phase natural circulation flow. An experimentally validated flow pattern map in the Phase Change Number - Subcooling Number (N{sub PCh} - N{sub Sub}) diagram defines the operational range in which flow instabilities such as geysering can be expected. One important perspective of this combined experimental/numerical work, which is in the field of two-phase flow, is its application to the development of accelerator driven systems (ADS). The main objective on an ADS is its potential to transmute minor actinides and long-lived fission products, thus participating in closing the fuel cycle. The development of an ADS is an important issue within the Euratom Fifth FP on Partitioning and Transmutation. One concept of an ADS, which is investigated in more detail within the ''preliminary design study of an experimental ADS'' Project (PDS-XADS) of the Euratom Fifth FP, is the XADS lead-bismuth cooled Experimental ADS of ANSALDO. An essential feature of this concept is the natural circulation of the primary coolant within the reactor pool. The natural circulation, which is driven by the density differences between the blanket and the heat exchanger, is enhanced by the injection of the nitrogen cover gas through spargers located in a riser part just above the blanket. This so-called gas-lift pump system has not been investigated in more detail nor has this gas-lift pump system been numerically/experimentally confirmed. The knowledge gained within the SUCO Programe, i.e. the modelling of the interfacial forces, the experimental work on flow instabilities and the

TSTA loop operation with 100 grams-level of tritium

International Nuclear Information System (INIS)

Yoshida, Hiroshi; Fukui, Hiroshi; Hirata, Shingo

1988-12-01

A fully integrated loop operation test of Tritium systems Test Assembly (TSTA) with 107 grams of tritium was completed at Los Alamos National Laboratory (LANL) in June, 1988. In this test, a compound cryopump with a charcoal panel was incorporated into the main process loop for the first time. The objectives were (i) to demonstrate the compound cryopump system with different flow rates and impurities, (ii) to demonstrate the regeneration of the compound cryopump system, (iii) to accumulate operating experience with other process systems such as the fuel cleanup system, the isotope separation system, the tritium supply and recovery system, etc. and (iv) to improve the data-base on TSTA safety systems such as the secondary containment system, tritium waste treatment system and tritium monitoring system. This report briefly describes characteristics of the main subsystems observed during the milestone run. (author)

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

Lessons from wet gas flow metering systems using differential measurements devices: Testing and flow modelling results

Energy Technology Data Exchange (ETDEWEB)

Cazin, J.; Couput, J.P.; Dudezert, C. et al

2005-07-01

A significant number of wet gas meters used for high GVF and very high GVF are based on differential pressure measurements. Recent high pressure tests performed on a variety of different DP devices on different flow loops are presented. Application of existing correlations is discussed for several DP devices including Venturi meters. For Venturi meters, deviations vary from 9% when using the Murdock correlation to less than 3 % with physical based models. The use of DP system in a large domain of conditions (Water Liquid Ratio) especially for liquid estimation will require information on the WLR This obviously raises the question of the gas and liquid flow metering accuracy in wet gas meters and highlight needs to understand AP systems behaviour in wet gas flows (annular / mist / annular mist). As an example, experimental results obtained on the influence of liquid film characteristics on a Venturi meter are presented. Visualizations of the film upstream and inside the Venturi meter are shown. They are completed by film characterization. The AP measurements indicate that for a same Lockhart Martinelli parameter, the characteristics of the two phase flow have a major influence on the correlation coefficient. A 1D model is defined and the results are compared with the experiments. These results indicate that the flow regime influences the AP measurements and that a better modelling of the flow phenomena is needed even for allocation purposes. Based on that, lessons and way forward in wet gas metering systems improvement for allocation and well metering are discussed and proposed. (author) (tk)

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

Performance Tests of Three Flow Distributors Using SMART-ITL with 1-Train CMT

Energy Technology Data Exchange (ETDEWEB)

Bae, Hwang; Ryu, Sung Uk; Shin, Yong-Cheol; Ko, Yung-Joo; Min, Kyoung-Ho; Ryu, Hyo Bong; Park, Jong-Kuk; Bang, Yun-Gon; Chae, Young-Jong; Yi, Sung-Jae; Park, Hyun-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

Passive safety systems (PSSs) are key tools to remove the heat from the core or containment. Safety improvements for SMART have been studied since the Standard Design Approval (SDA) for SMART was certificated in 2012. Active safety systems such as safety injection pumps are replaced by a passive system, which is a kind of the gravity injection system with core makeup tanks (CMT) and safety injection tanks (SIT). All tanks for the passive safety systems are located higher than a pressurized reactor vessel, whose injection nozzles are located around the reactor coolant pumps (RCP). An Integral Test Loop for the SMART design (SMART-ITL) has been constructed and its commissioning tests finished in 2012. SMART-ITL is scaled down by the volume scaling methodology. Its height is conserved and its volume scale ratio is 1/49. The SMART-ITL has all fluid systems of SMART together with a break system and instruments. Recently, a test program to validate the performance of SMART Passive Safety System (PSS) was launched. A scaled-down test facility for SMART PSS was additionally installed at the existing SMART-ITL facility and a set of validation tests were performed. In this paper, the performance tests of the flow distributors using SMART-ITL with 1-train CMT will be discussed. A 1-train passive safety system including a CMT and SIT, which is operated only by gravity force, was additionally installed in the SMART-ITL to replace the active safety system for the SMART design. Several performance tests for the flow distributors were carried out to estimate a designed flow rate. 1. The peak flow rate in a hot test does not reach the value in a cold test, and the approaching time to peak is also delayed during the early stage of gravity injection. 2.. It is verified that the flow rate from a gravity injection depends on the differential pressure in the injection pipe line including a friction and form drag, which can be adjusted by controlling the resistance coefficient.

Effect of Loop Diameter on the Steady State and Stability Behaviour of Single-Phase and Two-Phase Natural Circulation Loops

Directory of Open Access Journals (Sweden)

P. K. Vijayan

2008-01-01

Full Text Available In natural circulation loops, the driving force is usually low as it depends on the riser height which is generally of the order of a few meters. The heat transport capability of natural circulation loops (NCLs is directly proportional to the flow rate it can generate. With low driving force, the straightforward way to enhance the flow is to reduce the frictional losses. A simple way to do this is to increase the loop diameter which can be easily adopted in pressure tube designs such as the AHWR and the natural circulation boilers employed in fossil-fuelled power plants. Further, the loop diameter also plays an important role on the stability behavior. An extensive experimental and theoretical investigation of the effect of loop diameter on the steady state and stability behavior of single- and two-phase natural circulation loops have been carried out and the results of this study are presented in this paper.

Testing the role of meander cutoff in promoting gene flow across a riverine barrier in ground skinks (Scincella lateralis.

Directory of Open Access Journals (Sweden)

Nathan D Jackson

Full Text Available Despite considerable attention, the long-term impact of rivers on species diversification remains uncertain. Meander loop cutoff (MLC is one river phenomenon that may compromise a river's diversifying effects by passively transferring organisms from one side of the river to the other. However, the ability of MLC to promote gene flow across rivers has not been demonstrated empirically. Here, we test several predictions of MLC-mediated gene flow in populations of North American ground skinks (Scincella lateralis separated by a well-established riverine barrier, the Mississippi River: 1 individuals collected from within meander cutoffs should be more closely related to individuals across the river than on the same side, 2 individuals within meander cutoffs should contain more immigrants than individuals away from meander cutoffs, 3 immigration rates estimated across the river should be highest in the direction of the cutoff event, and 4 the distribution of alleles native to one side of the river should be better predicted by the historical rather than current path of the river. To test these predictions we sampled 13 microsatellite loci and mitochondrial DNA from ground skinks collected near three ancient meander loops. These predictions were generally supported by genetic data, although support was stronger for mtDNA than for microsatellite data. Partial support for genetic divergence of samples within ancient meander loops also provides evidence for the MLC hypothesis. Although a role for MLC-mediated gene flow was supported here for ground skinks, the transient nature of river channels and morphologies may limit the long-term importance of MLC in stemming population divergence across major rivers.

Study on the stability of a single-phase natural circulation flow in a closed loop. Demonstrative experiments on the higher-mode density wave oscillation

International Nuclear Information System (INIS)

Nishihara, Takashi

1997-01-01

Single-phase natural circulation loops are very important systems driven by the density variation generated thermally and have various applications in energy systems. Many theoretical and experimental works have been carried out on them and it has been known that the oscillatory instability can occur under some conditions. Most of the works on the oscillatory instability have been limited to specific geometry of the loops and they have paid attention only to the instability of fundamental mode, which has the period approximately equal to the item that the fluid goes round the loop, hereinafter referred to as the typical period. The author had applied the linear stability analysis to the simplified rectangular loop to investigate the basic stability characteristics of a natural circulation flow in a closed loop. The results indicate that various higher-mode oscillatory instabilities can be caused with a period approximately equal to one nth of the typical period according to parameters such as the pressure loss coefficient, the locations of a heat source and a heat sink, and so on. In this report, experimental tests were carried out and it was demonstrated that the higher-mode oscillatory instability can be caused with features as predicted in the analysis. The stability analysis was applied to the geometry of the experimental apparatus. The analytical results and those of experiments were compared with regard to the mode and the region of the parameters to be unstable and they have a good agreement qualitatively. (author)

Corrosion behaviour of martensitic and austenitic steels in flowing lead-bismuth eutectic

International Nuclear Information System (INIS)

Martin-Munoz, F.J.; Soler-Crespo, L.; Gomez-Briceno, D.

2011-01-01

The LINCE loop is a forced convection loop designed for long-term corrosion tests in lead-bismuth eutectic (LBE) at CIEMAT. The LBE volume of in the loop is 250 l and the maximum flow velocity in the region of specimens is approximately 1 m s -1 . An oxygen control system has been implemented in the loop. The corrosion behaviour of AISI 316L and T91 steels was investigated in flowing LBE at temperatures of 575 and 725 K for exposure times of 2000, 5000 and 10,000 h. At 575 K, the results showed a good response, with no weight loss detected in any of the materials after exposure to the flowing LBE up to 10,000 h. A similar behaviour was observed for the specimens tested at 725 K during 2000 and 10,000 h. Specimens extracted at intermediate time (5000 h) showed an anomalous behaviour with important weight loss. These specimens were placed at the bottom of the hot test section, and this position probably made them to suffer an accused process of cavitation-erosion.

Computational stability appraisal of rectangular natural circulation loop: Effect of loop inclination

International Nuclear Information System (INIS)

Krishnani, Mayur; Basu, Dipankar N.

2017-01-01

Highlights: • Computational model developed for single-phase rectangular natural circulation loop. • Role of loop inclination to vertical on thermalhydraulic stability is explored. • Inclination has strong stabilizing effect due to lower effective gravitation force. • Increase in tilt angle reduces settling time and highest amplitude of oscillation. • An angle of 15° is suggested for the selected loop geometry. - Abstract: Controlling stability behavior of single-phase natural circulation loops, without significantly affecting its steady-state characteristics, is a topic of wide research interest. Present study explores the role of loop inclination on a particular loop geometry. Accordingly a 3D computational model of a rectangular loop is developed and transient conservation equations are solved to obtain the temporal variation in flow parameters. Starting from the quiescent state, simulations are performed for selected sets of operating conditions and also with a few selected inclination angles. System experiences instability at higher heater powers and also with higher sink temperatures. Inclination is found to have a strong stabilizing influence owing to the reduction in the effective gravitational acceleration and subsequent decline in local buoyancy effects. The settling time and highest amplitude of oscillations substantially reduces for a stable system with a small inclination. Typically-unstable systems can also suppress the oscillations, when subjected to tilting, within a reasonable period of time. It is possible to stabilize the loop within shorter time span by increasing the tilt angle, but at the expense of reduction in steady-state flow rate. Overall a tilt angle of 15° is suggested for the selected geometry. Results from the 3D model is compared with the predictions from an indigenous 1D code. While similar qualitative influence of inclination is observed, the 1D model predicts early appearance of the stability threshold and hence hints

Numerical simulation of losses along a natural circulation helium loop

Energy Technology Data Exchange (ETDEWEB)

Knížat, Branislav, E-mail: branislav.knizat@stuba.sk; Urban, František, E-mail: frantisek.urban@stuba.sk; Mlkvik, Marek, E-mail: marek.mlkvik@stuba.sk; Ridzoň, František, E-mail: frantisek.ridzon@stuba.sk; Olšiak, Róbert, E-mail: robert.olsiak@stuba.sk [Slovak University of Technology in Bratislava, Faculty of Mechanical Engineering, Nám. slobody 17, 812 31 Bratislava, Slovak Republik (Slovakia)

2016-06-30

A natural circulation helium loop appears to be a perspective passive method of a nuclear reactor cooling. When designing this device, it is important to analyze the mechanism of an internal flow. The flow of helium in the loop is set in motion due to a difference of hydrostatic pressures between cold and hot branch. Steady flow at a requested flow rate occurs when the buoyancy force is adjusted to resistances against the flow. Considering the fact that the buoyancy force is proportional to a difference of temperatures in both branches, it is important to estimate the losses correctly in the process of design. The paper deals with the calculation of losses in branches of the natural circulation helium loop by methods of CFD. The results of calculations are an important basis for the hydraulic design of both exchangers (heater and cooler). The analysis was carried out for the existing model of a helium loop of the height 10 m and nominal heat power 250 kW.

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

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

Optic flow estimation on trajectories generated by bio-inspired closed-loop flight.

Science.gov (United States)

Shoemaker, Patrick A; Hyslop, Andrew M; Humbert, J Sean

2011-05-01

We generated panoramic imagery by simulating a fly-like robot carrying an imaging sensor, moving in free flight through a virtual arena bounded by walls, and containing obstructions. Flight was conducted under closed-loop control by a bio-inspired algorithm for visual guidance with feedback signals corresponding to the true optic flow that would be induced on an imager (computed by known kinematics and position of the robot relative to the environment). The robot had dynamics representative of a housefly-sized organism, although simplified to two-degree-of-freedom flight to generate uniaxial (azimuthal) optic flow on the retina in the plane of travel. Surfaces in the environment contained images of natural and man-made scenes that were captured by the moving sensor. Two bio-inspired motion detection algorithms and two computational optic flow estimation algorithms were applied to sequences of image data, and their performance as optic flow estimators was evaluated by estimating the mutual information between outputs and true optic flow in an equatorial section of the visual field. Mutual information for individual estimators at particular locations within the visual field was surprisingly low (less than 1 bit in all cases) and considerably poorer for the bio-inspired algorithms that the man-made computational algorithms. However, mutual information between weighted sums of these signals and comparable sums of the true optic flow showed significant increases for the bio-inspired algorithms, whereas such improvement did not occur for the computational algorithms. Such summation is representative of the spatial integration performed by wide-field motion-sensitive neurons in the third optic ganglia of flies.

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

TREAT MK III Loop Thermoelastoplastic Stress Analysis for the L03 Experiment

Energy Technology Data Exchange (ETDEWEB)

Kennedy, James M.

1981-03-01

The STRAW code was used to analyze the static response of a TREAT MK III loop subjected to thermal and mechanical loadings arising from an accident situation for the purpose of determining the defiections and stresses. This analysis provides safety support for the L03 reactivity accident study. The analysis was subdivided into two tasks: (1) an analysis of a flow blockage accident (Cases A and B), where all the energy is assumed deposited in the test leg, resulting in a temperature increase from 530°F to 1720°F, with a small internal pressure throughout the loop and (2) an analysis of a second flow blockage accident (Cases C and D), where again, all the energy is assumed to he deposited in the test leg, resulting in a temperature rise from 530°F to 1845°F, with a small internal pressure throughout the loop. The purpose of these two tasks was to determine if loop failure can occur with the thermal differential across the pump and test legs. Also of interest is whether an undesirable amount of loop lateral deflection will be caused by the thermal differential. A two dimensional analysis of the TREAT MK III loop was performed. The analysis accounted for material nonlinearities, both as a function of temperature and stress, and geometric nonlinearities arising from large deflections. Straight beam elements with annular cross sections were used to model the loop. The analyses show that the maximum strains are less than 21% of their failure strains for all subcases of Cases A and B. For all subcases of cases C and D, the maximum strains are less than 53% of their failure strains. The failure strain is 27.9% for the material at 530°F, 38.1% at 1720°F and 17.8% at 1845°F. Large lateral deflections are observed when the loop is not constrained except at its clamped support--as much as 8.6 inches. However, by accounting for the constraint of the concrete biological shield, the maximum lateral deflection was reduced to less than 0.05 inches at the points of concern.

Investigation on premature occurrence of critical heat flux under oscillatory flow and power conditions

International Nuclear Information System (INIS)

Vishnoi, A.K.; Dasgupta, A.; Chandraker, D.K.; Nayak, A.K.; Vijayan, P.K.

2015-01-01

Two-phase natural circulation loops have extensive applications in nuclear and process industries. One of the major concerns with natural circulation is the occurrence of the various types of flow instabilities, which can cause premature boiling crisis due to flow and power oscillations. In this work a transient computer code COPCOS (Code for Prediction of CHF under Oscillating flow and power condition) has been developed to predict the premature occurrence of CHF (critical heat flux) under oscillating flow and power. The code incorporates conduction equation of the fuel and coolant energy equation. For CHF prediction, CHF look-up table developed by Groeneveld is used. A facility named CHF and Instability Loop (CHIL) has been set up to study the effect of oscillatory flow on CHF. CHF and Instability Loop (CHIL) is a simple rectangular loop having a 10.5 mm ID and 1.2 m long test section. The flow through the test section is controlled by a canned motor pump using a Variable Frequency Drive (VFD). This leads to the ability of having a very precise control over flow oscillations which can be induced in the test section. The effect of frequency and amplitude of flow oscillation on occurrence of premature CHF has been investigated in this facility using COPCOS. Full paper covers details of COPCOS code, description of the facility and effect of frequency and the effect of oscillatory flow on CHF in the facility. (author)

Valveless pumping mechanics of the embryonic heart during cardiac looping: Pressure and flow through micro-PIV.

Science.gov (United States)

Bark, D L; Johnson, B; Garrity, D; Dasi, L P

2017-01-04

Cardiovascular development is influenced by the flow-induced stress environment originating from cardiac biomechanics. To characterize the stress environment, it is necessary to quantify flow and pressure. Here, we quantify the flow field in a developing zebrafish heart during the looping stage through micro-particle imaging velocimetry and by analyzing spatiotemporal plots. We further build upon previous methods to noninvasively quantify the pressure field at a low Reynolds number using flow field data for the first time, while also comparing the impact of viscosity models. Through this method, we show that the atrium builds up pressure to ~0.25mmHg relative to the ventricle during atrial systole and that atrial expansion creates a pressure difference of ~0.15mmHg across the atrium, resulting in efficient cardiac pumping. With these techniques, it is possible to noninvasively fully characterize hemodynamics during heart development. Copyright © 2016 Elsevier Ltd. All rights reserved.

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

Progress in construction of liquid metal LiPb experimental loops in China

International Nuclear Information System (INIS)

Zhang, M.; Zhu, Z.; Gao, S.; Song, Y.; Li, C.; Huang, Q.; Wu, Y.

2007-01-01

The activities of FDS series fusion reactors design with liquid tritium breeder blankets have been performed at ASIPP (Institute of Plasma Physics, Chinese Academy of Sciences) for years. In the designs, CLAM (China Low Activation Martensitic steel), is considered as the primary candidate structural material and LiPb eutectic as both tritium breeder and coolant of the blankets. Therefore, researches on LiPb experimental loop and construction of LiPb loop are severely needed in order to carry out experimental study on the compatibility of candidate structural materials for fusion reactors such as CLAM etc., flowing characteristics of LiPb and Magnetohydrodynamic (MHD) effect and so on, which is essential to researches of China liquid LiPb blankets. A lot of work has been done at ASIPP on design, manufacture and experiments for the series LiPb experimental loops i.e. Dragon-I, Dragon-II, Dragon-III and Dragon-IV. Dragon-I is a thermal convection LiPb loop made of SS316L steel and operating at 500 degree C. The first 3000 hour loop operation at 480 degree C for compatibility test on CLAM was done. Dragon-II and Dragon-III are also thermal convection LiPb loops, made of Inconel 600 and SiCr/SiC, and operating at 700 degree C and 1000 degree C, respectively, to obtain corrosion results of materials such as SiCr/SiC composite. Dragon-II has already been built up and under testing. Dragon-III is under construction. Base on requirement for experiments on characteristics of LiPb on its flow, MHD effect and corrosion to materials, Dragon-IV forced convection loop is being designed. The operation temperature ranges from 480 degree C at the cold leg to 700 degree C at the hot leg, the magnetic field is about 2-5T. Experiments and related studies in those loops are underway. (authors)

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)

Experimental investigation on premature occurrence of critical heat flux under oscillatory flow

International Nuclear Information System (INIS)

Vishnoi, A.K.; Dasgupta, A.; Chandraker, D.K.; Nayak, A.K.; Rama Rao, A.; Hegde, Nandan D.

2016-01-01

Two-phase natural circulation loops have extensive applications in nuclear and process industries. One of the major concerns with natural circulation is the occurrence of the various types of flow instabilities, which can cause premature boiling crisis due to flow and power oscillations. In this work, experimental investigation on CHF under oscillatory flow was carried out in a facility named CHF and Instability Loop (CHIL). CHIL is a simple rectangular loop having a 10.5 mm ID and 1.1 m long test section. The flow through the test section is controlled by a canned motor pump using a Variable Frequency Drive (VFD). The effect of frequency and amplitude of flow oscillation on occurrence of premature CHF has been investigated for this facility using a transient computer code COPCOS (Code for Prediction of CHF under Oscillating flow and power condition). The code incorporates conduction equation of the fuel and coolant energy equation. For CHF prediction, CHF look-up table developed by Groeneveld is used. Full paper covers description of the facility, experimental procedure, experimental results and data analysis using COPCOS. (author)

Experimental investigations in high-pressure natural circulation loop: progress report for the period January-June, 1999

International Nuclear Information System (INIS)

Naveen Kumar; Rajalakshmi, R.; Kulkarni, R.D.; Sagar, T.V.; Vijayan, P.K.; Saha, D.

2000-02-01

The Advanced Heavy Water Reactor employs natural circulation as the normal mode of coolant circulation. This is expected to enhance safety and reliability as it eliminates all safety issues associated with the pump failure. Two-phase natural circulation, however, is susceptible to several types of instabilities. In addition, the flow rate in a natural circulation loop is a dependent quantity and is not known a priori. Reliable calculations of the flow rate and stability behaviour are essential to ensure the success of AHWR design. Hence computer codes developed to predict the steady state flow rate and stability behaviour require validation against test data under natural circulation. For this purpose a high-pressure natural circulation loop has been designed, constructed and commissioned. Steady state experiments have been carried out in this loop to study the effect of pressure on natural circulation flow rate. The experimental results for this case are presented in this report. More experiments are planned in future to study the various aspects of two-phase natural circulation. (author)

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.)

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

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)

Analysis of flow-induced vibrations in the PEC design

International Nuclear Information System (INIS)

Cornaggia, L.; Reale, M.; Martelli, A.; Zambelli, M.

1986-01-01

This paper summarizes the studies performed for the Italian PEC fast reactor test facility with regard to flow-induced vibration problems. Reference is made to the reactor-block, the primary and secondary coolant loops and the emergency loops. Studies in progress and future developments foreseen are also mentioned. (author)

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.

Evaluation report on CCTF core-II reflood test C2 - 8 (Run 67)

International Nuclear Information System (INIS)

Akimoto, Hajime; Iguchi, Tadashi; Okubo, Tsutomu; Murao, Yoshio; Okabe, Kazuharu; Sugimoto, Jun.

1987-01-01

In order to study the system pressure effect of the core cooling and flow behavior during the reflood phase of a PWR LOCA, a test was performed with CCTF under the system pressure pf 0.15 MPa as a counterpart test of the CCTF test C2-1 (system pressure 0.42 MPa) and the CCTF test C2-4 (system pressure 0.20 MPa). Through the comparisons of results from these three tests, the following conclusions were obtained: (1) The higher system pressure resulted in the lower temperature rise, the shorter turnaround time and the shorter quench time as observed in the CCTF Core-I system pressure effect tests. (2) The higher system pressure resulted in higher core water head, higher upper plenum water head, higher mass flow rate through the primary loops. On the other hand, the higher system pressure resulted in lower downcomer water head and lower pressure drop through the primary loops and the broken cold leg. These system pressure effects on the flow behavior in the primary system are almost the same as observed in the system pressure effect tests in the CCTF Core-I test series. (3) Before the mixture level in the upper plenum reached the level of the hot leg nozzle, the loop flow resistance coefficient of the intact loops was nearly constant regardless of the system pressure. After the mixture level reached the level of the hot leg nozzle, the loop flow resistance coefficient was increased due to the water accumulation in the hot leg piping and the inlet plenum of the steam generator in these tests. (J.P.N.)

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

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.

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

Two-phase flow model with nonequilibrium and critical flow

International Nuclear Information System (INIS)

Sureau, H.; Houdayer, G.

1976-01-01

The model proposed includes the three conservation equations (mass, momentum, energy) applied to the two phase flows and a fourth partial derivative equation which takes into account the nonequilibriums and describes the mass transfer process. With this model, the two phase critical flow tests performed on the Moby-Dick loop (CENG) with several geometries, are interpreted by a unique law. Extrapolations to industrial dimension problems show that geometry and size effects are different from those obtained with earlier models (Zaloudek, Moody, Fauske) [fr

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

On vortex loops and filaments: three examples of numerical predictions of flows containing vortices.

Science.gov (United States)

Krause, Egon

2003-01-01

Vortex motion plays a dominant role in many flow problems. This article aims at demonstrating some of the characteristic features of vortices with the aid of numerical solutions of the governing equations of fluid mechanics, the Navier-Stokes equations. Their discretized forms will first be reviewed briefly. Thereafter three problems of fluid flow involving vortex loops and filaments are discussed. In the first, the time-dependent motion and the mutual interaction of two colliding vortex rings are discussed, predicted in good agreement with experimental observations. The second example shows how vortex rings are generated, move, and interact with each other during the suction stroke in the cylinder of an automotive engine. The numerical results, validated with experimental data, suggest that vortex rings can be used to influence the spreading of the fuel droplets prior to ignition and reduce the fuel consumption. In the third example, it is shown that vortices can also occur in aerodynamic flows over delta wings at angle of attack as well as pipe flows: of particular interest for technical applications of these flows is the situation in which the vortex cores are destroyed, usually referred to as vortex breakdown or bursting. Although reliable breakdown criteria could not be established as yet, the numerical predictions obtained so far are found to agree well with the few experimental data available in the recent literature.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-15

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

The influence of core bypass flow during SBLOCA

International Nuclear Information System (INIS)

Maselj, A.; Jurkovic, M.

1996-01-01

Many parameters affect the behaviour of a NPP during a Small Break Loss of Coolant Accident (SBLOCA). The bypass flow between the core side and the downcomer is one of them. Different PWRs have different values of core bypass flow. In spite of the complexity of the real situation in the primary system during SBLOCA, some fundamental details of the phenomena can be explained with simplified mathematical models, which relate on basic parameters of the primary coolant. These models define the conditions for loop seal clearance and final results are confirmed with measured values. The analysis presented in the paper refers to Bethsy Test 9.1.b SB LOCA scenario, with variation of core bypass flow. Basic RELAP5 input model calculation results show very good agreement with the experimental data. The core liquid level depression before loop seal clearance is lower in case of smaller core bypass flow. This affects the fuel clad temperature because of different heat transfer mechanisms. Time of loop seal clearance is delayed with larger core bypass flow and consequently lower differential pressure between downcomer and core. (author)

Design Principles for Closed Loop Supply Chains

NARCIS (Netherlands)

H.R. Krikke (Harold); C.P. Pappis (Costas); G.T. Tsoulfas; J.M. Bloemhof-Ruwaard (Jacqueline)

2001-01-01

textabstractIn this paper we study design principles for closed loop supply chains. Closed loop supply chains aim at closing material flows thereby limiting emission and residual waste, but also providing customer service at low cost. We study 'traditional' and 'new' design principles known in the

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

Evaluation of piping heat transfer, piping flow regimes, and steam generator heat transfer for the Semiscale Mod-1 isothermal tests

International Nuclear Information System (INIS)

French, R.T.

1975-08-01

Selected experimental data pertinent to piping heat transfer, transient fluid flow regimes, and steam generator heat transfer obtained during the Semiscale Mod-1 isothermal blowdown test series (Test Series 1) are analyzed. The tests in this first test series were designed to provide counterparts to the LOFT nonnuclear experiments. The data from the Semiscale Mod-1 intact and broken loop piping are evaluated to determine the surface heat flux and average heat transfer coefficients effective during the blowdown transient and compared with well known heat transfer correlations used in the RELAP4 computer program. Flow regimes in horizontal pipe sections are calculated and compared with data obtained from horizontal and vertical densitometers and with an existing steady state flow map. Effects of steam generator heat transfer are evaluated quantitatively and qualitatively. The Semiscale Mod-1 data and the analysis presented in this report are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict system response to piping heat transfer, piping flow regimes, and steam generator heat transfer during a postulated loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). 16 references. (auth)

Formation of Al{sub 2}O{sub 3}/FeAl coatings on a 9Cr-1Mo steel, and corrosion evaluation in flowing Pb-17Li loop

Energy Technology Data Exchange (ETDEWEB)

Majumdar, Sanjib, E-mail: sanjib@barc.gov.in [High Temperature Materials Development Section, Materials Processing & Corrosion Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai (India); Paul, Bhaskar [High Temperature Materials Development Section, Materials Processing & Corrosion Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai (India); Chakraborty, Poulami [Materials Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai (India); Kishor, Jugal; Kain, Vivekanand [High Temperature Materials Development Section, Materials Processing & Corrosion Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai (India); Dey, Gautam Kumar [Materials Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai (India); Materials Group, Bhabha Atomic Research Centre, Trombay, Mumbai (India)

2017-04-01

Iron aluminide coating layers were formed on a ferritic martensitic grade 9Cr-1Mo (P 91) steel using pack aluminizing process. The formation of different aluminide compositions such as orthorhombic-Fe{sub 2}Al{sub 5}, B2-FeAl and A2-Fe(Al) on the pack chemistry and heat treatment conditions have been established. About 4–6 μm thick Al{sub 2}O{sub 3} scale was formed on the FeAl phase by controlled heat treatment. The corrosion tests were conducted using both the FeAl and Al{sub 2}O{sub 3}/FeAl coated specimens in an electro-magnetic pump driven Pb-17Li Loop at 500 °C for 5000 h maintaining a flow velocity of 1.5 m/s. The detailed characterization studies using scanning electron microscopy, back-scattered electron imaging and energy dispersive spectrometry revealed no deterioration of the coating layers after the corrosion tests. Self-healing oxides were formed at the cracks generated in the aluminide layers during thermal cycling and protected the base alloy (steel) from any kind of elemental dissolution or microstructural degradation. - Highlights: •Al{sub 2}O{sub 3}/FeAl coating produced on P91 steel by pack aluminizing and heat treatment. •Corrosion tests of coated steel conducted in flowing Pb-17Li loop at 500 °C for 5000 h. •Coating was protective against molten metal corrosion during prolonged exposure. •Self-healing protective oxides formed in the cracks generated in aluminide layers.

Thermohaline loops, Stommel box models, and the Sandström theorem

OpenAIRE

Wunsch, Carl

2005-01-01

The Stommel two-box, two flow-regime box model is kinematically and dynamically equivalent to the flow in a onedimensional fluid loop, although one having awkward and extreme mixing coefficients. More generally, such a loop, when heated and cooled at the same geopotential, provides a simple example of the working of the Sandström theorem, with flow intensity capable of increasing or decreasing with growing diffusion. Stress dominates real oceanic flows, and its introduction into the purely th...

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.

Mechanism of formation of loop-type prominences

International Nuclear Information System (INIS)

Uralov, A.M.; Fedorov, L.V.

1978-01-01

Chromospheric gas heated to high temperatures flows out to the corona, filling and carrying up arches of the coronal magnetic field. Under the action of the magnetic tension and of the gravitation, a part of matter contained in the field tubes begins to fall back. The magnetic pressure of the magnetic loop reduced to its original size prevents the vertical fall of gas. At the loop top, braking of gas is most significant, due to field quasi-transversality. Here, in the first place gas compression and cooling by emission of radiation occurs, the already visible matter thereafter flowing away from the condensation point, thus marking the loop contours. A continuous return to the state of equilibrium of new field tubes with matter leads to an apparent ascent of the arch structure into the corona

Mercury Thermal Hydraulic Loop (MTHL) Summary Report

Energy Technology Data Exchange (ETDEWEB)

Felde, David K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Crye, Jason Michael [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wendel, Mark W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yoder, Jr, Graydon L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Farquharson, George [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jallouk, Philip A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McFee, Marshall T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ruggles, Art E. [Univ. of Tennessee, Knoxville, TN (United States); Carbajo, Juan J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-03-01

The Spallation Neutron Source (SNS) is a high-power linear accelerator built at Oak Ridge National Laboratory (ORNL) which incorporates the use of a flowing liquid mercury target. The Mercury Thermal Hydraulic Loop (MTHL) was constructed to investigate and verify the heat transfer characteristics of liquid mercury in a rectangular channel. This report provides a compilation of previously reported results from the water-cooled and electrically heated straight and curved test sections that simulate the geometry of the window cooling channel in the target nose region.

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

Supersonic flow over a pitching delta wing using surface pressure measurements and numerical simulations

Directory of Open Access Journals (Sweden)

Mostafa HADIDOOLABI

2018-01-01

Full Text Available Experimental and numerical methods were applied to investigating high subsonic and supersonic flows over a 60° swept delta wing in fixed state and pitching oscillation. Static pressure coefficient distributions over the wing leeward surface and the hysteresis loops of pressure coefficient versus angle of attack at the sensor locations were obtained by wind tunnel tests. Similar results were obtained by numerical simulations which agreed well with the experiments. Flow structure around the wing was also demonstrated by the numerical simulation. Effects of Mach number and angle of attack on pressure distribution curves in static tests were investigated. Effects of various oscillation parameters including Mach number, mean angle of attack, pitching amplitude and frequency on hysteresis loops were investigated in dynamic tests and the associated physical mechanisms were discussed. Vortex breakdown phenomenon over the wing was identified at high angles of attack using the pressure coefficient curves and hysteresis loops, and its effects on the flow features were discussed.

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)

Peach Bottom Cycle 2 Low Flow Stability Tests analysis using RELAP5/PARCS

International Nuclear Information System (INIS)

Costa, A.L.; Salah, A.B.; D'Auria, F.

2004-01-01

Nowadays, the coupled codes technique, which consists in incorporating threedimensional (3D) neutron modeling of the reactor core into system codes, is extensively used for simulating transients that involve core spatial asymmetric phenomena and strong feedback effects between core neutronics and reactor loop thermal-hydraulics. So, in this work, the coupled codes technique using RELAP5/3.3-PARCS is applied to simulate stability transients in a BWR (Boiling Water Reactor). Validation has been performed against Peach Bottom-2 Low-Flow Stability Tests. In these transients dynamically complex neutron kinetics coupling with thermal-hydraulics events take place in response to a core pressure perturbation. The calculated coupled code results are herein compared against the available experimental data. (author)

Computational simulation of flow and heat transfer in single-phase natural circulation loops

International Nuclear Information System (INIS)

Pinheiro, Larissa Cunha

2017-01-01

Passive decay heat removal systems based on natural circulation are essential assets for the new Gen III+ nuclear power reactors and nuclear spent fuel pools. The aim of the present work is to study both laminar and turbulent flow and heat transfer in single-phase natural circulation systems through computational fluid dynamics simulations. The working fluid is considered to be incompressible with constant properties. In the way, the Boussinesq Natural Convection Hypothesis was applied. The model chosen for the turbulence closure problem was the k -- εThe commercial computational fluid dynamics code ANSYS CFX 15.0 was used to obtain the numerical solution of the governing equations. Two single-phase natural circulation circuits were studied, a 2D toroidal loop and a 3D rectangular loop, both with the same boundary conditions of: prescribed heat flux at the heater and fixed wall temperature at the cooler. The validation and verification was performed with the numerical data provided by DESRAYAUD et al. [1] and the experimental data provided by MISALE et al. [2] and KUMAR et al. [3]. An excellent agreement between the Reynolds number (Re) and the modified Grashof number (Gr_m), independently of Prandtl Pr number was observed. However, the convergence interval was observed to be variable with Pr, thus indicating that Pr is a stability governing parameter for natural circulation. Multiple steady states was obtained for Pr = 0,7. Finally, the effect of inclination was studied for the 3D circuit, both in-plane and out-of-plane inclinations were verified for the steady state laminar regime. As a conclusion, the Re for the out-of-plane inclination was in perfect agreement with the correlation found for the zero inclination system, while for the in-plane inclined system the results differ from that of the corresponding vertical loop. (author)

Simulation experiments for hot-leg U-bend two-phase flow phenomena

International Nuclear Information System (INIS)

Ishii, M.; Hsu, J.T.; Tucholke, D.; Lambert, G.; Kataoka, I.

1986-01-01

In order to study the two-phase natural circulation and flow termination during a small break loss of coolant accident in LWR, simulation experiments have been performed. Based on the two-phase flow scaling criteria developed under this program, an adiabatic hot leg U-bend simulation loop using nitrogen gas and water and a Freon 113 boiling and condensation loop were built. The nitrogen-water system has been used to isolate key hydrodynamic phenomena from heat transfer problems, whereas the Freon loop has been used to study the effect of phase changes and fluid properties. Various tests were carried out to establish the basic mechanism of the flow termination and reestablishment as well as to obtain essential information on scale effects of parameters such as the loop frictional resistance, thermal center, U-bend curvature and inlet geometry. In addition to the above experimental study, a preliminary modeling study has been carried out for two-phase flow in a large vertical pipe at relatively low gas fluxes typical of natural circulation conditions

SOLA-LOOP analysis of a back pressure check valve

International Nuclear Information System (INIS)

Travis, J.R.

1984-01-01

The SOLA-LOOP computer code for transient, nonequilibrium, two-phase flows in networks has been coupled with a simple valve model to analyze a feedwater pipe breakage with a back-pressure check valve. Three tests from the Superheated Steam Reactor Safety Program Project (PHDR) at Kahl, West Germany, are analyzed, and the calculated transient back-pressure check valve behavior and fluid dynamics effects are found to be in excellent agreement with the experimentally measured data

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.

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.

Apparatus for measuring fluid flow

Science.gov (United States)

Smith, J.E.; Thomas, D.G.

Flow measuring apparatus includes a support loop having strain gages mounted thereon and a drag means which is attached to one end of the support loop and which bends the sides of the support loop and induces strains in the strain gages when a flow stream impacts thereon.

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

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)

PONDEROMOTIVE ACCELERATION IN CORONAL LOOPS

International Nuclear Information System (INIS)

Dahlburg, R. B.; Obenschain, K.; Laming, J. M.; Taylor, B. D.

2016-01-01

Ponderomotive acceleration has been asserted to be a cause of the first ionization potential (FIP) effect, the well-known enhancement in abundance by a factor of 3–4 over photospheric values of elements in the solar corona with FIP less than about 10 eV. It is shown here by means of numerical simulations that ponderomotive acceleration occurs in solar coronal loops, with the appropriate magnitude and direction, as a “by-product” of coronal heating. The numerical simulations are performed with the HYPERION code, which solves the fully compressible three-dimensional magnetohydrodynamic equations including nonlinear thermal conduction and optically thin radiation. Numerical simulations of coronal loops with an axial magnetic field from 0.005 to 0.02 T and lengths from 25,000 to 75,000 km are presented. In the simulations the footpoints of the axial loop magnetic field are convected by random, large-scale motions. There is a continuous formation and dissipation of field-aligned current sheets, which act to heat the loop. As a consequence of coronal magnetic reconnection, small-scale, high-speed jets form. The familiar vortex quadrupoles form at reconnection sites. Between the magnetic footpoints and the corona the reconnection flow merges with the boundary flow. It is in this region that the ponderomotive acceleration occurs. Mirroring the character of the coronal reconnection, the ponderomotive acceleration is also found to be intermittent.

PONDEROMOTIVE ACCELERATION IN CORONAL LOOPS

Energy Technology Data Exchange (ETDEWEB)

Dahlburg, R. B.; Obenschain, K. [LCP and FD, Naval Research Laboratory, Washington, DC 20375 (United States); Laming, J. M. [Space Science Division, Naval Research Laboratory, Washington, DC 20375 (United States); Taylor, B. D. [AFRL Eglin AFB, Pensacola, FL 32542 (United States)

2016-11-10

Ponderomotive acceleration has been asserted to be a cause of the first ionization potential (FIP) effect, the well-known enhancement in abundance by a factor of 3–4 over photospheric values of elements in the solar corona with FIP less than about 10 eV. It is shown here by means of numerical simulations that ponderomotive acceleration occurs in solar coronal loops, with the appropriate magnitude and direction, as a “by-product” of coronal heating. The numerical simulations are performed with the HYPERION code, which solves the fully compressible three-dimensional magnetohydrodynamic equations including nonlinear thermal conduction and optically thin radiation. Numerical simulations of coronal loops with an axial magnetic field from 0.005 to 0.02 T and lengths from 25,000 to 75,000 km are presented. In the simulations the footpoints of the axial loop magnetic field are convected by random, large-scale motions. There is a continuous formation and dissipation of field-aligned current sheets, which act to heat the loop. As a consequence of coronal magnetic reconnection, small-scale, high-speed jets form. The familiar vortex quadrupoles form at reconnection sites. Between the magnetic footpoints and the corona the reconnection flow merges with the boundary flow. It is in this region that the ponderomotive acceleration occurs. Mirroring the character of the coronal reconnection, the ponderomotive acceleration is also found to be intermittent.

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.

Experiment data report for Semiscale Mod-1 Test S-05-2 (alternate ECC injection test)

International Nuclear Information System (INIS)

Feldman, E.M.; Collins, B.L.; Sackett, K.E.

1977-02-01

Recorded test data are presented for Test S-05-2 of the Semiscale Mod-1 alternate emergency core coolant (ECC) injection test series. This test is one of several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-2 was conducted from an initial cold leg fluid temperature of 545 0 F and an initial pressure of 2263 psia. A simulated double-ended offset shear cold leg break was used to investigate core and system response to a depressurization and reflood transient with ECC injection at the intact loop pump suction and broken loop cold leg. A reduced lower plenum volume was used for this test to more accurately represent the lower plenum of a PWR, based on system volume scaling. System flow was set to achieve a core fluid temperature differential of 65 0 F at a core power level of 1.44 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a slightly peaked radial power profile was used in the pressure vessel to simulate the predicted surface heat flux of nuclear fuel rods during a loss-of-coolant accident

Heat pipes et two-phase loops for spacecraft applications. ESA programmes

Energy Technology Data Exchange (ETDEWEB)

Supper, W [European Space Agency / ESTEC. Thermal control and life support division (France)

1997-12-31

This document is a series of transparencies presenting the current and future applications of heat pipes in spacecraft and the activities in the field of capillary pumped two-phase loops: thermal tests, high-efficiency low pressure drop condensers, theoretical understanding of evaporator function, optimization of liquid and vapor flows, trade-off between low and high conductivity wicks, development of high capillary capacity wicks etc.. (J.S.)

Heat pipes et two-phase loops for spacecraft applications. ESA programmes

Energy Technology Data Exchange (ETDEWEB)

Supper, W. [European Space Agency / ESTEC. Thermal control and life support division (France)

1996-12-31

This document is a series of transparencies presenting the current and future applications of heat pipes in spacecraft and the activities in the field of capillary pumped two-phase loops: thermal tests, high-efficiency low pressure drop condensers, theoretical understanding of evaporator function, optimization of liquid and vapor flows, trade-off between low and high conductivity wicks, development of high capillary capacity wicks etc.. (J.S.)

Preliminary test results and CFD analysis for moderator circulation test at Korea

Energy Technology Data Exchange (ETDEWEB)

Kim, H.T. [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of); Im, S.H.; Sung, H.J. [Korea Advanced Inst. of Science and Tech., Daejeon (Korea, Republic of); Seo, H.; Bang, I.C. [Ulsan National Inst. of Science and Tech., Ulsan (Korea, Republic of)

2014-07-01

Korea Atomic Energy Research Institute (KAERI) is carrying out a scaled-down moderator test program to simulate the CANDU6 moderator circulation phenomena during steady state operation and accident conditions. This research program includes the construction of the Moderator Circulation Test (MCT) facility, production of the validation data for self-reliant CFD tools, and development of optical measurement system using the Particle Image Velocimetry (PIV). The MCT facility includes a primary circulation loop (pipe lines, a primary side pump, a heat exchanger, valves, flow meters) and a secondary side loop (pipe lines, a secondary side pump, and an external cooling tower). The loop leakage test and non-heating test are performed in the present work. In the present work the PIV technique is used to measure the velocity distributions in the scaled moderator tank of MCT under iso-thermal test conditions. The preliminary PIV measurement data are obtained and compared with CFX code predictions. (author)

Preliminary test results and CFD analysis for moderator circulation test at Korea

International Nuclear Information System (INIS)

Kim, H.T.; Im, S.H.; Sung, H.J.; Seo, H.; Bang, I.C.

2014-01-01

Korea Atomic Energy Research Institute (KAERI) is carrying out a scaled-down moderator test program to simulate the CANDU6 moderator circulation phenomena during steady state operation and accident conditions. This research program includes the construction of the Moderator Circulation Test (MCT) facility, production of the validation data for self-reliant CFD tools, and development of optical measurement system using the Particle Image Velocimetry (PIV). The MCT facility includes a primary circulation loop (pipe lines, a primary side pump, a heat exchanger, valves, flow meters) and a secondary side loop (pipe lines, a secondary side pump, and an external cooling tower). The loop leakage test and non-heating test are performed in the present work. In the present work the PIV technique is used to measure the velocity distributions in the scaled moderator tank of MCT under iso-thermal test conditions. The preliminary PIV measurement data are obtained and compared with CFX code predictions. (author)

Material analyses of foam-based SiC FCI after dynamic testing in PbLi in MaPLE loop at UCLA

Energy Technology Data Exchange (ETDEWEB)

Gonzalez, Maria, E-mail: maria.gonzalez@ciemat.es [LNF-CIEMAT, Avda Complutense, 40, 28040 Madrid (Spain); Rapisarda, David; Ibarra, Angel [LNF-CIEMAT, Avda Complutense, 40, 28040 Madrid (Spain); Courtessole, Cyril; Smolentsev, Sergey; Abdou, Mohamed [Fusion Science and Technology Center, UCLA (United States)

2016-11-01

Highlights: • Samples from foam-based SiC FCI were analyzed by looking at their SEM microstructure and elemental composition. • After finishing dynamic experiments in the flowing hot PbLi, the liquid metal ingress has been confirmed due to infiltration through local defects in the protective inner CVD layer. • No direct evidences of corrosion/erosion were observed; these defects could be related to the manufacturing process. - Abstract: Foam-based SiC flow channel inserts (FCIs) developed and manufactured by Ultramet, USA are currently under testing in the flowing hot lead-lithium (PbLi) alloy in the MaPLE loop at UCLA to address chemical/physical compatibility and to access the MHD pressure drop reduction. UCLA has finished the first experimental series, where a single uninterrupted long-term (∼6500 h) test was performed on a 30-cm FCI segment in a magnetic field up to 1.8 T at the temperature of 300 °C and maximum flow velocities of ∼ 15 cm/s. After finishing the experiments, the FCI sample was extracted from the host stainless steel duct and cut into slices. Few of them have been analyzed at CIEMAT as a part of the joint collaborative effort on the development of the DCLL blanket concept in the EU and the US. The initial inspection of the slices using optical microscopic analysis at UCLA showed significant PbLi ingress into the bulk FCI material that resulted in degradation of insulating properties of the FCI. Current material analyses at CIEMAT are based on advanced techniques, including characterization of FCI samples by FESEM to study PbLi ingress, imaging of cross sections, composition analysis by EDX and crack inspection. These analyses suggest that the ingress was caused by local defects in the protective inner CVD layer that might be originally present in the FCI or occurred during testing.

Hybrid Combustion-Gasification Chemical Looping

Energy Technology Data Exchange (ETDEWEB)

Herbert Andrus; Gregory Burns; John Chiu; Gregory Lijedahl; Peter Stromberg; Paul Thibeault

2009-01-07

} separation, and also syngas production from coal with the calcium sulfide (CaS)/calcium sulfate (CaSO{sub 4}) loop utilizing the PDU facility. The results of Phase I were reported in Reference 1, 'Hybrid Combustion-Gasification Chemical Looping Coal Power Development Technology Development Phase I Report' The objective for Phase II was to develop the carbonate loop--lime (CaO)/calcium carbonate (CaCO{sub 3}) loop, integrate it with the gasification loop from Phase I, and ultimately demonstrate the feasibility of hydrogen production from the combined loops. The results of this program were reported in Reference 3, 'Hybrid Combustion-Gasification Chemical Looping Coal Power Development Technology Development Phase II Report'. The objective of Phase III is to operate the pilot plant to obtain enough engineering information to design a prototype of the commercial Chemical Looping concept. The activities include modifications to the Phase II Chemical Looping PDU, solids transportation studies, control and instrumentation studies and additional cold flow modeling. The deliverable is a report making recommendations for preliminary design guidelines for the prototype plant, results from the pilot plant testing and an update of the commercial plant economic estimates.

Development of Capillary Loop Convective Polymerase Chain Reaction Platform with Real-Time Fluorescence Detection

Directory of Open Access Journals (Sweden)

Wen-Pin Chou

2017-02-01

Full Text Available Polymerase chain reaction (PCR has been one of the principal techniques of molecular biology and diagnosis for decades. Conventional PCR platforms, which work by rapidly heating and cooling the whole vessel, need complicated hardware designs, and cause energy waste and high cost. On the other hand, partial heating on the various locations of vessels to induce convective solution flows by buoyancy have been used for DNA amplification in recent years. In this research, we develop a new convective PCR platform, capillary loop convective polymerase chain reaction (clcPCR, which can generate one direction flow and make the PCR reaction more stable. The U-shaped loop capillaries with 1.6 mm inner diameter are designed as PCR reagent containers. The clcPCR platform utilizes one isothermal heater for heating the bottom of the loop capillary and a CCD device for detecting real-time amplifying fluorescence signals. The stable flow was generated in the U-shaped container and the amplification process could be finished in 25 min. Our experiments with different initial concentrations of DNA templates demonstrate that clcPCR can be applied for precise quantification. Multiple sample testing and real-time quantification will be achieved in future studies.

Determination of Heritage SSME Pogo Suppressor Resistance and Inertance from Waterflow Pulse Testing

Science.gov (United States)

McDougal, Chris; Eberhart, Chad; Lee, Erik

2016-01-01

Waterflow tests of a heritage Space Shuttle Main Engine pogo suppressor were performed to experimentally quantify the resistance and inertance provided by the suppressor. Measurements of dynamic pressure and flow rate in response to pulsing flow were made throughout the test loop. A unique system identification methodology combined all sensor measurements with a one-dimensional perturbational flow model of the complete water flow loop to spatially translate physical measurements to the device under test. Multiple techniques were then employed to extract the effective resistance and inertance for the pogo suppressor. Parameters such as steady flow rate, perturbational flow rate magnitude, and pulse frequency were investigated to assess their influence on the behavior of the pogo suppressor dynamic response. These results support validation of the RS-25 pogo suppressor performance for use on the Space Launch System Core Stage.

Density wave oscillations of a boiling natural circulation loop induced by flashing

Energy Technology Data Exchange (ETDEWEB)

Furuya, Masahiro; Inada, Fumio; Yasuo, Akira [Central Research Institute of Electric Power Industry, Tokyo (Japan)

1995-09-01

Experiments are conducted to investigate two-phase flow instabilities in a boiling natural circulation loop with a chimney due to flashing in the chimney at lower pressure. The test facility used in this experiment is designed to have non-dimensional values which are nearly equal to those of natural circulation BWR. Stability maps in reference to the heat flux, the inlet subcooling, the system pressure are presented. This instability is suggested to be density wave oscillations due to flashing in the chimney, and the differences from other phenomena such as flow pattern oscillations and geysering phenomena are discussed by investigating the dynamic characteristics, the oscillation period, and the transient flow pattern.

The impact of interpreted flow regimes during constant head injection tests on the estimated transmissivity from injection tests and difference flow logging

Energy Technology Data Exchange (ETDEWEB)

Hjerne, Calle; Ludvigsson, Jan-Erik; Harrstroem, Johan [Geosigma AB, Uppsala (Sweden)

2013-04-15

A large number of constant head injection tests were carried out in the site investigation at Forsmark using the Pipe String System, PSS3. During the original evaluation of the tests the dominating transient flow regimes during both the injection and recovery period were interpreted together with estimation of hydraulic parameters. The flow regimes represent different flow and boundary conditions during the tests. Different boreholes or borehole intervals may display different distributions of flow regimes. In some boreholes good agreement was obtained between the results of the injection tests and difference flow logging with Posiva flow log (PFL) but in other boreholes significant discrepancies were found. The main objective of this project is to study the correlation between transient flow regimes from the injection tests and other borehole features such as transmissivity, depth, geology, fracturing etc. Another subject studied is whether observed discrepancies between estimated transmissivity from difference flow logging and injection tests can be correlated to interpreted flow regimes. Finally, a detailed comparison between transient and stationary evaluation of transmissivity from the injection tests in relation to estimated transmissivity from PFL tests in corresponding sections is made. Results from previous injection tests in 5 m sections in boreholes KFM04, KFM08A and KFM10A were used. Only injection tests above the (test-specific) measurement limit regarding flow rate are included in the analyses. For all of these tests transient flow regimes were interpreted. In addition, results from difference flow logging in the corresponding 5 m test sections were used. Finally, geological data of fractures together with rock and fracture zone properties have been used in the correlations. Flow regimes interpreted from the injection period of the tests are generally used in the correlations but deviations between the interpreted flow regimes from the injection and

16 x 16 Vantage+ Fuel Assembly Flow Vibrational Testing

International Nuclear Information System (INIS)

Chambers, Martin; Kurincic, Bojan

2014-01-01

Nuklearna Elektrarna Krsko (NEK) has experienced leaking fuel after increasing the cycle duration to 18 months. The leaking fuel mechanism has predominantly been consistent over multiple cycles and is typically observed in highly irradiated Fuel Assemblies (FA) after around 4 years of continuous operation that were located at the core periphery (baffle). The cause of the leaking fuel is due to Grid-To-Rod-Fretting (GRTF) and occasional debris fretting. NEK utilises a 16x16 Vantage+ FA design with all Inconel structural mixing vane grids (8 in total), Zirlo thimbles, Integral Fuel Burnable Absorber (IFBA) rods with enriched ZrB2, enriched Annular Blanket, Debris Filter Bottom Nozzle (DFBN), Removable Top Nozzle (RTN) and Zirlo fuel cladding material with a high burnup capability of 60 GWD/MTU. Numerous design and operational changes are thought to have reduced the original 16x16 FA design margin to fretting resistance of either vibration or its wear work rate, such as significant power uprate (spring force loss, rod creep down...), operational cycle duration increase from 12 to 18 months (increasing residence time as well as lead FA and fuel rod burnup values), Reactor Coolant System flow increase (increased vibration), removal of Thimble Plugs (increased bypass flow, increased vibration) and Zirc-4 to Zirlo cladding change (decreasing wear work rate). The fuel rod to grid spring as well as dimple contact areas are relatively smaller than other FA designs that exhibit good in-reactor fretting performance. A FA design change project to address the small rod to dimple / spring contact area and utilise fuel cladding oxide coating is currently being pursued with the fuel supplier. The FA vibrational properties are very important to the in-reactor FA performance and reliability. The 16x16 Vantage+ vibrational testing was performed with a full size FA in the Fuel Assembly Compatibility Testing (FACTS) loop that is able to provide full flow rates at elevated temperature

RELAPS choked flow model and application to a large scale flow test

International Nuclear Information System (INIS)

Ransom, V.H.; Trapp, J.A.

1980-01-01

The RELAP5 code was used to simulate a large scale choked flow test. The fluid system used in the test was modeled in RELAP5 using a uniform, but coarse, nodalization. The choked mass discharge rate was calculated using the RELAP5 choked flow model. The calulations were in good agreement with the test data, and the flow was calculated to be near thermal equilibrium

Micro-Columnated Loop Heat Pipe: The Future of Electronic Substrates

Science.gov (United States)

Dhillon, Navdeep Singh

. To predict the overall heat carrying capacity of the muCLHP in the capillary pumping limit, an analytical model was developed to account for a steady state pressure balance in the device flow loop. Based on this model, a design optimization study, employing monotonicity analysis and numerical optimization techniques, was undertaken. It was found that an optimized muCLHP device can absorb heat fluxes as large as 1293 W/cm2 when water is used as a working fluid. A finite volume method-based numerical model was also developed to compute the rates of thin-film evaporation from the patterned surface of the secondary wick. The numerical results indicated that, by properly optimizing the dual-scale wick topology, allowable evaporative heat fluxes can be made commensurate with the heat flux performance predicted by the capillary pumping limit. The latter part of the dissertation deals with the fabrication, packaging, and experimental testing of several in-plane-wicking micro loop heat pipe (muLHP) prototypes. These devices were fabricated on silicon and Pyrex substrates and closely resemble the muCLHP design philosophy, with the exception that the CPS wick is substituted with an easier to fabricate in-plane wick. A novel thermal-flux method was developed for the degassing and fluid charging of the muLHP prototypes. Experiments were conducted to study the process of evaporation and dynamics of the liquid and vapor phases in the device flow loop. Using these results, the overall device and individual component topologies critical to the operation of the two-phase flow loop were identified. A continuous two-phase device flow loop was demonstrated for applied evaporator heat fluxes as high as 41 W/cm2. The performance of these devices, currently found to be limited by the motive temperature head requirement, can be significantly improved by implementing the parasitic heat flow-reduction strategies developed in this work. The 3-D thin-film evaporation model, when integrated into

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.

Development of multiplex loop mediated isothermal amplification (m-LAMP) label-based gold nanoparticles lateral flow dipstick biosensor for detection of pathogenic Leptospira.

Science.gov (United States)

Nurul Najian, A B; Engku Nur Syafirah, E A R; Ismail, Nabilah; Mohamed, Maizan; Yean, Chan Yean

2016-01-15

In recent years extensive numbers of molecular diagnostic methods have been developed to meet the need of point-of-care devices. Efforts have been made towards producing rapid, simple and inexpensive DNA tests, especially in the diagnostics field. We report on the development of a label-based lateral flow dipstick for the rapid and simple detection of multiplex loop-mediated isothermal amplification (m-LAMP) amplicons. A label-based m-LAMP lateral flow dipstick assay was developed for the simultaneous detection of target DNA template and a LAMP internal control. This biosensor operates through a label based system, in which probe-hybridization and the additional incubation step are eliminated. We demonstrated this m-LAMP assay by detecting pathogenic Leptospira, which causes the re-emerging disease Leptospirosis. The lateral flow dipstick was developed to detect of three targets, the LAMP target amplicon, the LAMP internal control amplicon and a chromatography control. Three lines appeared on the dipstick, indicating positive results for all representative pathogenic Leptospira species, whereas two lines appeared, indicating negative results, for other bacterial species. The specificity of this biosensor assay was 100% when it was tested with 13 representative pathogenic Leptospira species, 2 intermediate Leptospira species, 1 non-pathogenic Leptospira species and 28 other bacteria species. This study found that this DNA biosensor was able to detect DNA at concentrations as low as 3.95 × 10(-1) genomic equivalent ml(-1). An integrated m-LAMP and label-based lateral flow dipstick was successfully developed, promising simple and rapid visual detection in clinical diagnostics and serving as a point-of-care device. Copyright © 2015 Elsevier B.V. All rights reserved.

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...

Ultrasonic flow-meter test in sodium

International Nuclear Information System (INIS)

Ishii, Y.; Uno, O.; Kamei, M.

1978-01-01

As a part of the R and D programme for the prototype fast breeder reactor MONJU, an ultrasonic flow-meter (USFM) test is being carried out in sodium in the O-Arai Engineering Center of PNC. Prior to the present test, an in-water test was done at the manufacturer's as a preliminary investigation. The results reported here are the results up to the present. Calibration tests using the actual fluid were conducted on a 12-inch ultrasonic flow-meter with guide rods fabricated for sodium flow measurement. The test conditions in sodium were a temperature of 200 approximately 400 0 C and flow-rates of 0 approximately 6m/s. The main results are: (1) The linearity of output signal was good and accuracy was within 1%; (2) The alternating type of the USFM was much better than the fixed type in temperature change; (3) 2MHz of transducer frequency was better than 3MHz in sodium; (4) The S/N ratio of the ultrasonic signal and the length/diameter effect in a wide range in sodium surpassed the in-water test. (author)

Flooding in a loop with a vertical and a horizontal tube connected by an elbow

International Nuclear Information System (INIS)

Yan Changqi

1994-01-01

The experimental research of flooding and flow-reverse in a test loop which a vertical and a horizontal tube connected by an elbow is introduced. According to the experimental results, the effects of the elbow on flooding and flow-reverse is analyzed. The experimental results is compared with the results obtained in vertical tubes. The effect of horizontal tube length and hysteresis in de-flooding are analyzed. Dimensionless parameters was used in data process. The correlations for predicting the flooding point, de-flooding point, completed carry up and flow reverse points are given

Fission Surface Power Technology Demonstration Unit Test Results

Science.gov (United States)

Briggs, Maxwell H.; Gibson, Marc A.; Geng, Steven M.; Sanzi, James L.

2016-01-01

The Fission Surface Power (FSP) Technology Demonstration Unit (TDU) is a system-level demonstration of fission power technology intended for use on manned missions to Mars. The Baseline FSP systems consists of a 190 kWt UO2 fast-spectrum reactor cooled by a primary pumped liquid metal loop. This liquid metal loop transfers heat to two intermediate liquid metal loops designed to isolate fission products in the primary loop from the balance of plant. The intermediate liquid metal loops transfer heat to four Stirling Power Conversion Units (PCU), each of which produce 12 kWe (48 kW total) and reject waste heat to two pumped water loops, which transfer the waste heat to titanium-water heat pipe radiators. The FSP TDU simulates a single leg of the baseline FSP system using an electrically heater core simulator, a single liquid metal loop, a single PCU, and a pumped water loop which rejects the waste heat to a Facility Cooling System (FCS). When operated at the nominal operating conditions (modified for low liquid metal flow) during TDU testing the PCU produced 8.9 kW of power at an efficiency of 21.7 percent resulting in a net system power of 8.1 kW and a system level efficiency of 17.2 percent. The reduction in PCU power from levels seen during electrically heated testing is the result of insufficient heat transfer from the NaK heater head to the Stirling acceptor, which could not be tested at Sunpower prior to delivery to the NASA Glenn Research Center (GRC). The maximum PCU power of 10.4 kW was achieved at the maximum liquid metal temperature of 875 K, minimum water temperature of 350 K, 1.1 kg/s liquid metal flow, 0.39 kg/s water flow, and 15.0 mm amplitude at an efficiency of 23.3 percent. This resulted in a system net power of 9.7 kW and a system efficiency of 18.7 percent.

Thermal and hydrodynamic characteristics of supercritical CO2 natural circulation in closed loops

International Nuclear Information System (INIS)

Chen, Lin; Deng, Bi-Li; Jiang, Bin; Zhang, Xin-Rong

2013-01-01

Highlights: ► We model thermosyphon heat transfer and stability with super-/trans-critical turbulence model incorporated. ► Potentials of super-/trans-critical CO 2 thermosyphon are confirmed. ► Three characteristics found: flow instability; high flow rate with density wave; heat transfer discrepancies. ► Major laws of system stability factors are different compared with traditional fluids. ► Traditional thermosyphon flow correlation has its limitations and deserves further development. -- Abstract: Natural convective flow of supercritical fluids has become a hot topic in engineering applications. Natural circulation thermosyphon using supercritical/trans-critical CO 2 can be a potential choice for effectively transportation of heat and mass without pumping devices. This paper presents a series of numerical investigations into the fundamental features in a supercritical/trans-critical CO 2 based natural circulation loop. New heat transport model aiming at trans-critical thermosyphon heat transfer and stability is proposed with supercritical/trans-critical turbulence model incorporated. In this study, the fundamentals include the basic flow and heat transfer behavior of the above loop, the effect of heat source temperature on system stability, the effect of loop diameter on natural convection supercritical CO 2 loop and its coupling effect with heat source temperature and the effect of constant changing heat input condition and system behavior evolution during unsteady input or failure conditions. The fundamental potentials of supercritical/trans-critical CO 2 based natural convection system are confirmed. Basic supercritical CO 2 closed loop flow and heat transfer behaviors are clarified. During this study, the CO 2 loop stability map are also put forward and introduced as an important feature of supercritical CO 2 system. Stability factors of natural convective trans-critical CO 2 flow and its implications on real system control are also discussed in

Evaluation report on CCTF Core-I reflood tests Cl-2 (Run 11) and Cl-3 (Run 12)

International Nuclear Information System (INIS)

Akimoto, Hajime; Murao, Yoshio

1983-06-01

In order to clarify the effect of the initial superheat of the downcomer wall on the system and the core cooling behaviors during the reflood phase of a PWR-LOCA, two tests were performed with the Cylindrical Core Test Facility (CCTF). One is the superheated wall test (test Cl-2) with the initial superheat of 79K, as in the actual PWR, and the other is the saturated wall test (test Cl-3) without any initial superheat. Through the comparisons of the test results from these two tests, the following conclusions were obtained. (1) The initial superheat of the downcomer wall resulted in the lower downcomer water head as observed in our separate-effect tests for the downcomer water head. (2) The superheat also caused the core inlet subcooling to be decreased, and led to the lower core water head. (3) The mass flow rate through the intact loop was reduced only by 4% by the initial superheat of the downcomer wall because the core water head was reduced as well as the downcomer water head. Whereas the mass flow rate through the broken loop was increased because of the increased pressure drop through the broken cold leg. (4) The difference of the core inlet mass flow rate was small between the superheated and the saturated wall tests. It can be considered that small difference of the core inlet mass flow rate results from the compensation of the decreased mass flow rate through the intact loops by the increased mass flow rate through the broken loop. (5) The main discrepancies of the core cooling and the carry-over behaviors between two CCTF tests, were consistent with those observed in the parametric tests for the core inlet subcooling of the FLECHT LOW FLOODING TEST series. (author)

Investigation of reflood models by coupling REFLA-1D and multi-loop system model

International Nuclear Information System (INIS)

Sugimoto, Jun; Murao, Yoshio

1983-09-01

A system analysis code REFLA-1DS was developed by coupling reflood analysis code REFLA-1D and a multi-loop primary system model. The reflood models in the code were investigated for the development of the integral system analysis code. The REFLA-1D, which was developed with the small scale reflood experiment at JAERI, consists of one-dimensional core model and a primary system model with a constant loop resistance. The multi-loop primary system model was developed with the Cylindrical Core Test Facility of JAERI's large scale reflood tests. The components modeled in the code are the upper plenum, the steam generator, the coolant pump, the ECC injection port, the downcomer and the broken cold leg nozzle. The coupling between the two models in REFLA-1DS is accomplished by applying the equivalent flow resistance calculated with the multiloop model to the REFLA-1D. The characteristics of the code is its simplicity of the system model and the solution method which enables the fast running and the easy reflood analysis for the further model development. A fairly good agreement was obtained with the results of the Cylindrical Core Test Facility for the calculated water levels in the downcomer, the core and the upper plenum. A qualitatively good agreement was obtained concerning the parametric effects of the system pressure, the ECC flow rate and the initial clad temperature. Needs for further code improvements of the models, however, were pointed out. These include the problem concerning the generation rate of the steam and water droplets in the core in an early period, the effect of the flow oscillation on the core cooling, the heat release from the downcomer wall, and the stable system calculation. (author)

Integrated, digital experiment transient control and safety protection of an in-pile test

International Nuclear Information System (INIS)

Thomas, R.W.; Whitacre, R.F.; Klingler, W.B.

1982-01-01

The Sodium Loop Safety Facility experimental program has demonstrated that in-pile loop fuel failure transient tests can be digitally controlled and protected with reliability and precision. This was done in four nuclear experiments conducted in the Engineering Test Reactor operated by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory. Loop sodium flow and reactor power transients can be programmed to sponsor requirements and verified prior to the test. Each controller has redundancy, which reduces the effect of single failures occurring during test transients. Feedback and reject criteria are included in the reactor power control. Timed sequencing integrates the initiation of the controllers, programmed safety set-points, and other experiment actions (e.g., planned scram). Off-line and on-line testing is included. Loss-of-flow, loss-of-piping-integrity, boiling-window, transient-overpower, and local fault tests have been successfully run using this system

Influence of test tube material on subcooled flow boiling critical heat flux in short vertical tube

International Nuclear Information System (INIS)

Hata, Koichi; Shiotsu, Masahiro; Noda, Nobuaki

2007-01-01

The steady state subcooled flow boiling critical heat flux (CHF) for the flow velocities (u=4.0 to 13.3 m/s), the inlet subcoolings (ΔT sub,in =48.6 to 154.7 K), the inlet pressure (P in =735.2 to 969.0 kPa) and the increasing heat input (Q 0 exp(t/τ), τ=10, 20 and 33.3 s) are systematically measured with the experimental water loop. The 304 Stainless Steel (SUS304) test tube of inner diameter (d=6 mm), heated length (L=66 mm) and L/d=11 with the inner surface of rough finished (Surface roughness, Ra=3.18 μm), the Cupro Nickel (Cu-Ni 30%) test tube of d=6 mm, L=60 mm and L/d=10 with Ra=0.18 μm and the Platinum (Pt) test tubes of d=3 and 6 mm, L=66.5 and 69.6 mm, and L/d=22.2 and 11.6 respectively with Ra=0.45 μm are used in this work. The CHF data for the SUS304, Cu-Ni 30% and Pt test tubes were compared with SUS304 ones for the wide ranges of d and L/d previously obtained and the values calculated by the authors' published steady state CHF correlations against outlet and inlet subcoolings. The influence of the test tube material on CHF is investigated into details and the dominant mechanism of subcooled flow boiling critical heat flux is discussed. (author)

Influence of Test Tube Material on Subcooled Flow Boiling Critical Heat Flux in Short Vertical Tube

International Nuclear Information System (INIS)

Koichi Hata; Masahiro Shiotsu; Nobuaki Noda

2006-01-01

The steady state subcooled flow boiling critical heat flux (CHF) for the flow velocities (u = 4.0 to 13.3 m/s), the inlet subcooling (ΔT sub,in = 48.6 to 154.7 K), the inlet pressure (P in = 735.2 to 969.0 kPa) and the increasing heat input (Q 0 exp(t/t), t = 10, 20 and 33.3 s) are systematically measured with the experimental water loop. The 304 Stainless Steel (SUS304) test tubes of inner diameters (d = 6 mm), heated lengths (L = 66 mm) and L/d = 11 with the inner surface of rough finished (Surface roughness, R a = 3.18 μm), the Cupro Nickel (Cu-Ni 30%) test tubes of d = 6 mm, L = 60 mm and L/d = 10 with R a = 0.18 μm and the Platinum (Pt) test tubes of d = 3 and 6 mm, L = 66.5 and 69.6 mm, and L/d 22.2 and 11.6 respectively with R a = 0.45 μm are used in this work. The CHF data for the SUS304, Cu-Ni 30% and Pt test tubes were compared with SUS304 ones for the wide ranges of d and L/d previously obtained and the values calculated by the authors' published steady state CHF correlations against outlet and inlet subcooling. The influence of the test tube material on CHF is investigated into details and the dominant mechanism of subcooled flow boiling critical heat flux is discussed. (authors)

A Role for the Cytoskeleton in Heart Looping

Directory of Open Access Journals (Sweden)

Kersti K. Linask

2007-01-01

Full Text Available Over the past 10 years, key genes involved in specification of left-right laterality pathways in the embryo have been defined. The read-out for misexpression of laterality genes is usually the direction of heart looping. The question of how dextral looping direction occurred mechanistically and how the heart tube bends remains unknown. It is becoming clear from our experiments and those of others that left-right differences in cell proliferation in the second heart field (anterior heart field drives the dextral direction. Evidence is accumulating that the cytoskeleton is at the center of laterality, and the bending and rotational forces associated with heart looping. If laterality pathways are modulated upstream, the cytoskeleton, including nonmuscle myosin II (NMHC-II, is altered downstream within the cardiomyocytes, leading to looping abnormalities. The cytoskeleton is associated with important mechanosensing and signaling pathways in cell biology and development. The initiation of blood flow during the looping period and the inherent stresses associated with increasing volumes of blood flowing into the heart may help to potentiate the process. In recent years, the steps involved in this central and complex process of heart development that is the basis of numerous congenital heart defects are being unraveled.

Natural Circulation Characteristics of a Symmetric Loop under Inclined Conditions

Directory of Open Access Journals (Sweden)

Xingtuan Yang

2014-01-01

Full Text Available Natural circulation is an important process for primary loops of some marine integrated reactors. The reactor works under inclined conditions when severe accidents happen to the ship. In this paper, to investigate the characteristics of natural circulation, experiments were conducted in a symmetric loop under the inclined angle of 0~45°. A CFD model was also set up to predict the behaviors of the loop beyond the experimental scope. Total circulation flow rate decreases with the increase of inclined angle. Meanwhile one circulation is depressed while the other is enhanced, and accordingly the disparity between the branch circulations arises and increases with the increase of inclined angle. Circulation only takes place in one branch circuit at large inclined angle. Also based on the CFD model, the influences of flow resistance distribution and loop configuration on natural circulation are predicted. The numerical results show that to design the loop with the configuration of big altitude difference and small width, it is favorable to reduce the influence of inclination; however too small loop width will cause severe reduction of circulation ability at large angle inclination.

Open-loop heat-recovery dryer

Science.gov (United States)

TeGrotenhuis, Ward Evan

2013-11-05

A drying apparatus is disclosed that includes a drum and an open-loop airflow pathway originating at an ambient air inlet, passing through the drum, and terminating at an exhaust outlet. A passive heat exchanger is included for passively transferring heat from air flowing from the drum toward the exhaust outlet to air flowing from the ambient air inlet toward the drum. A heat pump is also included for actively transferring heat from air flowing from the passive heat exchanger toward the exhaust outlet to air flowing from the passive heat exchanger toward the drum. A heating element is also included for further heating air flowing from the heat pump toward the drum.

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.

Experimental Study of Single Phase Flow in a Closed-Loop Cooling System with Integrated Mini-Channel Heat Sink

Directory of Open Access Journals (Sweden)

Lei Ma

2016-06-01

Full Text Available The flow and heat transfer characteristics of a closed-loop cooling system with a mini-channel heat sink for thermal management of electronics is studied experimentally. The heat sink is designed with corrugated fins to improve its heat dissipation capability. The experiments are performed using variable coolant volumetric flow rates and input heating powers. The experimental results show a high and reliable thermal performance using the heat sink with corrugated fins. The heat transfer capability is improved up to 30 W/cm2 when the base temperature is kept at a stable and acceptable level. Besides the heat transfer capability enhancement, the capability of the system to transfer heat for a long distance is also studied and a fast thermal response time to reach steady state is observed once the input heating power or the volume flow rate are varied. Under different input heat source powers and volumetric flow rates, our results suggest potential applications of the designed mini-channel heat sink in cooling microelectronics.

Study and analysis on the flow induced vibration of the core barrel of PWR

International Nuclear Information System (INIS)

Yao Weida; Shi Guolin; Jiang Nanyan; Peng YongYong; Zhang Huijun; Wang Yufen; Xie Yongcheng; Guo Chunhua; Shen Qinping

1989-01-01

The deduction of the resemblance criterion and the design of the test model by applying flow-solid coupling theory are described. The model analysis of a core barrel both in the air and stationary water were performed in a 1:10 model, thus obtaining the dynamic characteristic. In a 1:5 reactor model with a hydraulic closed loop, the inner structure and support were modeled for performing hydraulic closed loop, the inner structure and support were modeled for performing hydraulic vibration test of the core barrel. The flow induced pulse pressure of the core barrel and corresponding response were obtained by using miniature pressure capsule, strain gauge and accelerometer. Power spectrum, correlation functions, transfer function and amplitudes under different flow velocities were calculated. The hydraulic vibration test shows that the core barrel will be in safety during its 30-year life time

Inspection of piping wall loss with flow accelerated corrosion accelerated simulation test

International Nuclear Information System (INIS)

Ryu, Kyung Ha; Kim, Ji Hak; Hwang, Il Soon; Lee, Na Young; Kim, Ji Hyun

2009-01-01

Flow Accelerated Corrosion (FAC) has become a hot issue for aging of passive components. Ultrasonic Technique (UT) has been adopted to inspect the secondary piping of Nuclear Power Plants (NPPs). UT, however, uses point detection method, which results in numerous detecting points and thus takes time. We developed an Equipotential Switching Direct Current Potential Drop (ES-DCPD) method to monitor the thickness of piping that covers wide range of piping at once time. Since the ES-DCPD method covers area, not a point, it needs less monitoring time. This can be a good approach to broad carbon steel piping system such as secondary piping of NPPs. In this paper, FAC accelerated simulation test results is described. We realized accelerated FAC phenomenon by 2 times test: 23.7% thinning in 216.7 hours and 51% thinning in 795 hours. These were monitored by ES-DCPD and traditional UT. Some parameters of water chemistry are monitored and controlled to accelerate FAC process. As sensitive factors on FAC, temperature and pH was changed during the test. The wall loss monitored results reflected these changes of water chemistry successfully. Developed electrodes are also applied to simulation loop to monitor water chemistry. (author)

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.

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)

In-situ Condition Monitoring of Components in Small Modular Reactors Using Process and Electrical Signature Analysis. Final report, volume 1. Development of experimental flow control loop, data analysis and plant monitoring

Energy Technology Data Exchange (ETDEWEB)

Upadhyaya, Belle [Univ. of Tennessee, Knoxville, TN (United States); Hines, J. Wesley [Univ. of Tennessee, Knoxville, TN (United States); Damiano, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mehta, Chaitanya [Univ. of Tennessee, Knoxville, TN (United States); Collins, Price [Univ. of Tennessee, Knoxville, TN (United States); Lish, Matthew [Univ. of Tennessee, Knoxville, TN (United States); Cady, Brian [Univ. of Tennessee, Knoxville, TN (United States); Lollar, Victor [Univ. of Tennessee, Knoxville, TN (United States); de Wet, Dane [Univ. of Tennessee, Knoxville, TN (United States); Bayram, Duygu [Univ. of Tennessee, Knoxville, TN (United States)

2015-12-15

The research and development under this project was focused on the following three major objectives: Objective 1: Identification of critical in-vessel SMR components for remote monitoring and development of their low-order dynamic models, along with a simulation model of an integral pressurized water reactor (iPWR). Objective 2: Development of an experimental flow control loop with motor-driven valves and pumps, incorporating data acquisition and on-line monitoring interface. Objective 3: Development of stationary and transient signal processing methods for electrical signatures, machinery vibration, and for characterizing process variables for equipment monitoring. This objective includes the development of a data analysis toolbox. The following is a summary of the technical accomplishments under this project: - A detailed literature review of various SMR types and electrical signature analysis of motor-driven systems was completed. A bibliography of literature is provided at the end of this report. Assistance was provided by ORNL in identifying some key references. - A review of literature on pump-motor modeling and digital signal processing methods was performed. - An existing flow control loop was upgraded with new instrumentation, data acquisition hardware and software. The upgrading of the experimental loop included the installation of a new submersible pump driven by a three-phase induction motor. All the sensors were calibrated before full-scale experimental runs were performed. - MATLAB-Simulink model of a three-phase induction motor and pump system was completed. The model was used to simulate normal operation and fault conditions in the motor-pump system, and to identify changes in the electrical signatures. - A simulation model of an integral PWR (iPWR) was updated and the MATLAB-Simulink model was validated for known transients. The pump-motor model was interfaced with the iPWR model for testing the impact of primary flow perturbations (upsets) on

In-situ Condition Monitoring of Components in Small Modular Reactors Using Process and Electrical Signature Analysis. Final report, volume 1. Development of experimental flow control loop, data analysis and plant monitoring

International Nuclear Information System (INIS)

Upadhyaya, Belle; Hines, J. Wesley; Damiano, Brian; Mehta, Chaitanya; Collins, Price; Lish, Matthew; Cady, Brian; Lollar, Victor; De Wet, Dane; Bayram, Duygu

2015-01-01

The research and development under this project was focused on the following three major objectives: Objective 1: Identification of critical in-vessel SMR components for remote monitoring and development of their low-order dynamic models, along with a simulation model of an integral pressurized water reactor (iPWR). Objective 2: Development of an experimental flow control loop with motor-driven valves and pumps, incorporating data acquisition and on-line monitoring interface. Objective 3: Development of stationary and transient signal processing methods for electrical signatures, machinery vibration, and for characterizing process variables for equipment monitoring. This objective includes the development of a data analysis toolbox. The following is a summary of the technical accomplishments under this project: - A detailed literature review of various SMR types and electrical signature analysis of motor-driven systems was completed. A bibliography of literature is provided at the end of this report. Assistance was provided by ORNL in identifying some key references. - A review of literature on pump-motor modeling and digital signal processing methods was performed. - An existing flow control loop was upgraded with new instrumentation, data acquisition hardware and software. The upgrading of the experimental loop included the installation of a new submersible pump driven by a three-phase induction motor. All the sensors were calibrated before full-scale experimental runs were performed. - MATLAB-Simulink model of a three-phase induction motor and pump system was completed. The model was used to simulate normal operation and fault conditions in the motor-pump system, and to identify changes in the electrical signatures. - A simulation model of an integral PWR (iPWR) was updated and the MATLAB-Simulink model was validated for known transients. The pump-motor model was interfaced with the iPWR model for testing the impact of primary flow perturbations (upsets) on

Project description: ORNL PWR blowdown heat transfer separate-effects program, Thermal-Hydraulic Test Facility (THTF)

International Nuclear Information System (INIS)

1976-02-01

The ORNL Pressurized-Water Reactor Blowdown Heat Transfer (PWR-BDHT) Program is an experimental separate-effects study of the relations among the principal variables that can alter the rate of blowdown, the presence of flow reversal and rereversal, time delay to critical heat flux, the rate at which dryout progresses, and similar time-related functions that are important to LOCA analysis. Primary test results will be obtained from the Thermal-Hydraulic Test Facility (THTF), a large nonnuclear pressurized-water loop that incorporates a 49-rod electrically heated bundle. Supporting experiments will be carried out in two additional test loops - the Forced Convection Test Facility (FCTF), a small high-pressure facility in which single heater rods can be tested in annular geometry; and an air-water loop which is used to evaluate two-phase flow-measuring instrumentation

An efficiency booster for energy conversion in natural circulation loops

Energy Technology Data Exchange (ETDEWEB)

Wang, Dongqing, E-mail: wangdongqing@stu.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China); Beijing Computational Science Research Center, Beijing 100084 (China); Jiang, Jin, E-mail: jjiang@eng.uwo.ca [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario N6A 5B9 (Canada); Beijing Computational Science Research Center, Beijing 100084 (China)

2016-08-01

Highlights: • Low driving power conversion efficiency of natural circulation loops is proved. • The low conversion efficiency leads to low heat transfer capacity of such loops. • An efficiency booster is designed with turbine to increase the efficiency. • Performance of the proposed booster has been numerically simulated. • The booster drastically enhances heat transfer capacity of such loops. - Abstract: In this paper, the capacity of a natural circulation loop for transferring heat from a heat source to a heat sink has been analyzed. It is concluded that the capacity of the natural circulation loop depends on the conversion efficiency of the thermal energy from the heat source to the driving force for the circulation of the flow. The low conversion efficiency leading to weak driving force in such loops has been demonstrated analytically and validated through simulation results. This issue has resulted in a low heat transfer capacity in the circulation loop. To increase the heat transfer capacity, one has to improve this efficiency. To meet such a need, a novel efficiency booster has been developed in this paper. The booster essentially increases the flow driving force and hence significantly improves the overall heat transfer capacity. Design and analysis of this booster have been performed in detail. The performance has been examined through extensive computer simulations. It is concluded that the booster can indeed drastically improve the heat transfer capacity of the natural circulation loop.

An efficiency booster for energy conversion in natural circulation loops

International Nuclear Information System (INIS)

Wang, Dongqing; Jiang, Jin

2016-01-01

Highlights: • Low driving power conversion efficiency of natural circulation loops is proved. • The low conversion efficiency leads to low heat transfer capacity of such loops. • An efficiency booster is designed with turbine to increase the efficiency. • Performance of the proposed booster has been numerically simulated. • The booster drastically enhances heat transfer capacity of such loops. - Abstract: In this paper, the capacity of a natural circulation loop for transferring heat from a heat source to a heat sink has been analyzed. It is concluded that the capacity of the natural circulation loop depends on the conversion efficiency of the thermal energy from the heat source to the driving force for the circulation of the flow. The low conversion efficiency leading to weak driving force in such loops has been demonstrated analytically and validated through simulation results. This issue has resulted in a low heat transfer capacity in the circulation loop. To increase the heat transfer capacity, one has to improve this efficiency. To meet such a need, a novel efficiency booster has been developed in this paper. The booster essentially increases the flow driving force and hence significantly improves the overall heat transfer capacity. Design and analysis of this booster have been performed in detail. The performance has been examined through extensive computer simulations. It is concluded that the booster can indeed drastically improve the heat transfer capacity of the natural circulation loop.

A closed-loop analysis of the tubuloglomerular feedback mechanism

DEFF Research Database (Denmark)

Holstein-Rathlou, N H

1991-01-01

The tubuloglomerular feedback (TGF) mechanism is of importance in the regulation of glomerular filtration rate (GFR). A second mechanism of potential importance is the change in proximal pressure caused by a change, for example, in the rate of proximal fluid reabsorption. The quantitative contrib...... and the late proximal flow rate, with changes in the proximal pressure of lesser importance. Furthermore, under closed-loop conditions the operating point for the TGF mechanism is at or close to the point of maximal sensitivity....... nl/min in steps of 5 nl/min. The open-loop gain (OLG) was 3.1 (range 1.5-9.9, n = 13) at the unperturbed tubular flow rate, and decreased as the tubular flow rate was either increased or decreased. The proximal pressure increased by 0.21 +/- 0.03 mmHg per unit increase in late proximal flow rate (nl...

Remote Sampler Demonstration Isolok Configuration Test

International Nuclear Information System (INIS)

Kelly, Steve E.

2016-01-01

The accuracy and precision of a new Isolok sampler configuration was evaluated using a recirculation flow loop. The evaluation was performed using two slurry simulants of Hanford high-level tank waste. Through testing, the capability of the Isolok sampler was evaluated. Sample concentrations were compared to reference samples that were simultaneously collected by a two-stage Vezin sampler. The capability of the Isolok sampler to collect samples that accurately reflect the contents in the test loop improved – biases between the Isolok and Vezin samples were greatly reduce for fast settling particles.

Remote Sampler Demonstration Isolok Configuration Test

Energy Technology Data Exchange (ETDEWEB)

Kelly, Steve E. [Washington River Protection Solutions, LLC, Richland, WA (United States)

2016-06-08

The accuracy and precision of a new Isolok sampler configuration was evaluated using a recirculation flow loop. The evaluation was performed using two slurry simulants of Hanford high-level tank waste. Through testing, the capability of the Isolok sampler was evaluated. Sample concentrations were compared to reference samples that were simultaneously collected by a two-stage Vezin sampler. The capability of the Isolok sampler to collect samples that accurately reflect the contents in the test loop improved – biases between the Isolok and Vezin samples were greatly reduce for fast settling particles.

Test Methodologies for Hydrogen Sensor Performance Assessment: Chamber vs. Flow Through Test Apparatus: Preprint

Energy Technology Data Exchange (ETDEWEB)

Buttner, William J [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Hartmann, Kevin S [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Schmidt, Kara [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Cebolla, Rafeal O [Joint Research Centre, Petten, the Netherlands; Weidner, Eveline [Joint Research Centre, Petten, the Netherlands; Bonato, Christian [Joint Research Centre, Petten, the Netherlands

2017-11-06

Certification of hydrogen sensors to standards often prescribes using large-volume test chambers [1, 2]. However, feedback from stakeholders such as sensor manufacturers and end-users indicate that chamber test methods are often viewed as too slow and expensive for routine assessment. Flow through test methods potentially are an efficient, cost-effective alternative for sensor performance assessment. A large number of sensors can be simultaneously tested, in series or in parallel, with an appropriate flow through test fixture. The recent development of sensors with response times of less than 1s mandates improvements in equipment and methodology to properly capture the performance of this new generation of fast sensors; flow methods are a viable approach for accurate response and recovery time determinations, but there are potential drawbacks. According to ISO 26142 [1], flow through test methods may not properly simulate ambient applications. In chamber test methods, gas transport to the sensor can be dominated by diffusion which is viewed by some users as mimicking deployment in rooms and other confined spaces. Alternatively, in flow through methods, forced flow transports the gas to the sensing element. The advective flow dynamics may induce changes in the sensor behaviour relative to the quasi-quiescent condition that may prevail in chamber test methods. One goal of the current activity in the JRC and NREL sensor laboratories [3, 4] is to develop a validated flow through apparatus and methods for hydrogen sensor performance testing. In addition to minimizing the impact on sensor behaviour induced by differences in flow dynamics, challenges associated with flow through methods include the ability to control environmental parameters (humidity, pressure and temperature) during the test and changes in the test gas composition induced by chemical reactions with upstream sensors. Guidelines on flow through test apparatus design and protocols for the evaluation of

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)

Vortex Diode Analysis and Testing for Fluoride Salt-Cooled High-Temperature Reactors

International Nuclear Information System (INIS)

Yoder, Graydon L. Jr.; Elkassabgi, Yousri M.; De Leon, Gerardo I.; Fetterly, Caitlin N.; Ramos, Jorge A.; Cunningham, Richard Burns

2012-01-01

Fluidic diodes are presently being considered for use in several fluoride salt-cooled high-temperature reactor designs. A fluidic diode is a passive device that acts as a leaky check valve. These devices are installed in emergency heat removal systems that are designed to passively remove reactor decay heat using natural circulation. The direct reactor auxiliary cooling system (DRACS) uses DRACS salt-to-salt heat exchangers (DHXs) that operate in a path parallel to the core flow. Because of this geometry, under normal operating conditions some flow bypasses the core and flows through the DHX. A flow diode, operating in reverse direction, is-used to minimize this flow when the primary coolant pumps are in operation, while allowing forward flow through the DHX under natural circulation conditions. The DRACSs reject the core decay heat to the environment under loss-of-flow accident conditions and as such are a reactor safety feature. Fluidic diodes have not previously been used in an operating reactor system, and therefore their characteristics must be quantified to ensure successful operation. This report parametrically examines multiple design parameters of a vortex-type fluidic diode to determine the size of diode needed to reject a particular amount of decay heat. Additional calculations were performed to size a scaled diode that could be tested in the Oak Ridge National Laboratory Liquid Salt Flow Loop. These parametric studies have shown that a 152.4 mm diode could be used as a test article in that facility. A design for this diode is developed, and changes to the loop that will be necessary to test the diode are discussed. Initial testing of a scaled flow diode has been carried out in a water loop. The 150 mm diode design discussed above was modified to improve performance, and the final design tested was a 171.45 mm diameter vortex diode. The results of this testing indicate that diodicities of about 20 can be obtained for diodes of this size. Experimental

FLOW TESTING AND ANALYSIS OF THE FSP-1 EXPERIMENT

Energy Technology Data Exchange (ETDEWEB)

Hawkes, Grant L.; Jones, Warren F.; Marcum, Wade; Weiss, Aaron; Howard, Trevor

2017-06-01

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A flow versus delta P curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported herein. Fuel plate deformation tests were also performed and reported.

A comparative approach to closed-loop computation.

Science.gov (United States)

Roth, E; Sponberg, S; Cowan, N J

2014-04-01

Neural computation is inescapably closed-loop: the nervous system processes sensory signals to shape motor output, and motor output consequently shapes sensory input. Technological advances have enabled neuroscientists to close, open, and alter feedback loops in a wide range of experimental preparations. The experimental capability of manipulating the topology-that is, how information can flow between subsystems-provides new opportunities to understand the mechanisms and computations underlying behavior. These experiments encompass a spectrum of approaches from fully open-loop, restrained preparations to the fully closed-loop character of free behavior. Control theory and system identification provide a clear computational framework for relating these experimental approaches. We describe recent progress and new directions for translating experiments at one level in this spectrum to predictions at another level. Operating across this spectrum can reveal new understanding of how low-level neural mechanisms relate to high-level function during closed-loop behavior. Copyright © 2013 Elsevier Ltd. All rights reserved.

Engineering design and development of lead lithium loop for thermo-fluid MHD studies

International Nuclear Information System (INIS)

Kumar, M.; Patel, Anita; Jaiswal, A.; Ranjan, A.; Mohanta, D.; Sahu, S.; Saraswat, A.; Rao, T.S.; Mehta, V.; Bhattacharyay, R.; Rajendra Kumar, E.

2017-01-01

In the frame of the design and development of LLCB TBM, number of R and D activities is in progress in the area of Pb-Li technology development. Molten Pb-Li is used as a tritium breeder and also as a coolant for the internals of the TBM structure. In presence of strong plasma confining toroidal magnetic field, motion of electrically conducting Pb-Li leads to Magneto Hydro Dynamic (MHD) phenomena, as a consequence of which the flow profile of Pb-Li is significantly modified inside the Pb-Li channels of TBM. This causes additional pressure drop inside TBM and affects the heat transfer from internal structure. The detail studies of these MHD effects are of prime importance for successful design of LLCB TBM and its performance evaluation. Although, various numerical MHD codes have been developed, validated in simple flow configuration and are being used to study MHD phenomena in LLCB TBM, experimental validation of these codes in TBM relevant complex flow geometry is yet to be performed. A Pb-Li MHD experimental loop is, therefore, being developed at IPR to perform thermo-fluid MHD experiments in various LLCB TBM relevant flow configuration. MHD experiments are planned with different test sections instrumented with potential pins, thermo couples, etc. under a uniform magnetic field of ∼1.4 T. The obtained experimental data will be analyzed to understand the MHD phenomena in TBM like flow configuration and also for validation of MHD codes. This paper describes the detailed process as well as engineering design of the Pb-Li MHD loop and its major components along with the plan of MHD experiments in various test mock ups. (author)

Numerical modeling of supercritical CO{sub 2} natural circulation loop

Energy Technology Data Exchange (ETDEWEB)

Archana, V., E-mail: archanav@barc.gov.in [Homi Bhabha National Institute, Mumbai, Maharashtra 400 094 (India); Vaidya, A.M., E-mail: avaidya@barc.gov.in [Bhabha Atomic Research Centre, Mumbai, Maharashtra 400 085 (India); Vijayan, P.K., E-mail: vijayanp@barc.gov.in [Bhabha Atomic Research Centre, Mumbai, Maharashtra 400 085 (India)

2015-11-15

Highlights: • Supercritical CO{sub 2} natural circulation loop is modeled by in-house developed 1D and 2D axi-symmetric CFD codes. • Steady state characteristics of VHVC configuration of supercritical CO{sub 2} natural circulation loop are studied over a range of power. • Improved accuracy of predictions by 2D axi-symmetric formulation over 1D formulation is demonstrated. • The validity of correlations used in 1D model such as friction factor and heat transfer correlations is analyzed. • Simulation results shows normal, enhanced and deteriorated heat transfer regimes in supercritical CO{sub 2} natural circulation loop. - Abstract: The objective of this research project is to estimate steady state characteristics of supercritical natural circulation loop (SCNCL) using computational methodology and to compliment on-going experimental investigation of the same at the authors’ organization. For computational investigation, a couple of in-house codes are developed. At first, formulation and a corresponding computer program for the SCNCL based on conservation equations written in 1D framework is developed. Comparison of 1D code results with experimental data showed that, under some operating conditions, there is deviation between computed results and experimental data. To improve predictive capability, it was thought to model the SCNCL using conservation equations in 2D axi-symmetric framework. An existing 2D axi-symmetric code (named NAFA), which was developed and validated for supercritical fluid flow in pipes, is modified for natural circulation loop (NCL) geometry. The modified code, named NAFA-Loop, is subsequently used to compute the steady state characteristics of the SCNCL. These results are compared with experimental data. The steady state characteristics such as the variation of mass flow rate with power, velocity and temperature profiles in heater and cooler are studied using NAFA-Loop. The computed velocity and temperature fields show that the

CFD simulation of a four-loop PWR at asymmetric operation conditions

Energy Technology Data Exchange (ETDEWEB)

Cheng, Jian-Ping; Yan, Li-Ming; Li, Feng-Chen, E-mail: lifch@hit.edu.cn

2016-04-15

Highlights: • A CFD numerical simulation procedure was established for simulating RPV of VVER-1000. • The established CFD approach was validated by comparing with available data. • Thermal hydraulic characteristics under asymmetric operation condition were investigated. • Apparent influences of the shutdown loop on its neighboring loops were obtained. - Abstract: The pressurized water reactor (PWR) with multiple loops may have abnormal working conditions with coolant pumps out of running in some loops. In this paper, a computational fluid dynamics (CFD) numerical study of the four-loop VVER-1000 PWR pressure vessel model was presented. Numerical simulations of the thermohydrodynamic characteristics in the pressure vessel were carried out at different inlet conditions with four and three loops running, respectively. At normal stead-state condition (four-loop running), different parameters were obtained for the full fluid domain, including pressure losses across different parts, pressure, velocity and temperature distributions in the reactor pressure vessel (RPV) and mass flow distribution of the coolant at the inlet of reactor core. The obtained results for pressure losses matched with the experimental reference values of the VVER-1000 PWR at Tianwan nuclear power plant (NPP). For most fuel assemblies (FAs), the inlet flow rates presented a symmetrical distribution about the center under full-loop operation conditions, which accorded with the practical distribution. These results indicate that it is now possible to study the dynamic transition process between different asymmetric operation conditions in a multi-loop PWR using the established CFD method.

Development of multiplex loop mediated isothermal amplification (m-LAMP) label-based gold nanoparticles lateral flow dipstick biosensor for detection of pathogenic Leptospira

International Nuclear Information System (INIS)

Nurul Najian, A.B.; Engku Nur Syafirah, E.A.R.; Ismail, Nabilah; Mohamed, Maizan; Yean, Chan Yean

2016-01-01

In recent years extensive numbers of molecular diagnostic methods have been developed to meet the need of point-of-care devices. Efforts have been made towards producing rapid, simple and inexpensive DNA tests, especially in the diagnostics field. We report on the development of a label-based lateral flow dipstick for the rapid and simple detection of multiplex loop-mediated isothermal amplification (m-LAMP) amplicons. A label-based m-LAMP lateral flow dipstick assay was developed for the simultaneous detection of target DNA template and a LAMP internal control. This biosensor operates through a label based system, in which probe-hybridization and the additional incubation step are eliminated. We demonstrated this m-LAMP assay by detecting pathogenic Leptospira, which causes the re-emerging disease Leptospirosis. The lateral flow dipstick was developed to detect of three targets, the LAMP target amplicon, the LAMP internal control amplicon and a chromatography control. Three lines appeared on the dipstick, indicating positive results for all representative pathogenic Leptospira species, whereas two lines appeared, indicating negative results, for other bacterial species. The specificity of this biosensor assay was 100% when it was tested with 13 representative pathogenic Leptospira species, 2 intermediate Leptospira species, 1 non-pathogenic Leptospira species and 28 other bacteria species. This study found that this DNA biosensor was able to detect DNA at concentrations as low as 3.95 × 10 −1 genomic equivalent ml −1 . An integrated m-LAMP and label-based lateral flow dipstick was successfully developed, promising simple and rapid visual detection in clinical diagnostics and serving as a point-of-care device. - Highlights: • We develop multiplex LAMP label-based lateral flow dipstick biosensor for detection of pathogenic Leptospira. • We design primers for multiplex LAMP targeting the conserved LipL32 gene of pathogenic Leptospira and LAMP internal

Development of multiplex loop mediated isothermal amplification (m-LAMP) label-based gold nanoparticles lateral flow dipstick biosensor for detection of pathogenic Leptospira

Energy Technology Data Exchange (ETDEWEB)

Nurul Najian, A.B.; Engku Nur Syafirah, E.A.R.; Ismail, Nabilah [Department of Medical Microbiology & Parasitology, School of Medical Sciences, Health Campus, Universiti Sains Malaysia, 16150 Kubang Kerian, Kelantan (Malaysia); Mohamed, Maizan [Faculty of Veterinary Medicine, Universiti Malaysia Kelantan, City Campus, Pengkalan Chepa, Locked Bag 36, 16100 Kota Bharu, Kelantan (Malaysia); Yean, Chan Yean, E-mail: yeancyn@yahoo.com [Department of Medical Microbiology & Parasitology, School of Medical Sciences, Health Campus, Universiti Sains Malaysia, 16150 Kubang Kerian, Kelantan (Malaysia); Institute for Research in Molecular Medicine (INFORMM), Health Campus, Universiti Sains Malaysia, 16150 Kubang Kerian, Kelantan (Malaysia)

2016-01-15

In recent years extensive numbers of molecular diagnostic methods have been developed to meet the need of point-of-care devices. Efforts have been made towards producing rapid, simple and inexpensive DNA tests, especially in the diagnostics field. We report on the development of a label-based lateral flow dipstick for the rapid and simple detection of multiplex loop-mediated isothermal amplification (m-LAMP) amplicons. A label-based m-LAMP lateral flow dipstick assay was developed for the simultaneous detection of target DNA template and a LAMP internal control. This biosensor operates through a label based system, in which probe-hybridization and the additional incubation step are eliminated. We demonstrated this m-LAMP assay by detecting pathogenic Leptospira, which causes the re-emerging disease Leptospirosis. The lateral flow dipstick was developed to detect of three targets, the LAMP target amplicon, the LAMP internal control amplicon and a chromatography control. Three lines appeared on the dipstick, indicating positive results for all representative pathogenic Leptospira species, whereas two lines appeared, indicating negative results, for other bacterial species. The specificity of this biosensor assay was 100% when it was tested with 13 representative pathogenic Leptospira species, 2 intermediate Leptospira species, 1 non-pathogenic Leptospira species and 28 other bacteria species. This study found that this DNA biosensor was able to detect DNA at concentrations as low as 3.95 × 10{sup −1} genomic equivalent ml{sup −1}. An integrated m-LAMP and label-based lateral flow dipstick was successfully developed, promising simple and rapid visual detection in clinical diagnostics and serving as a point-of-care device. - Highlights: • We develop multiplex LAMP label-based lateral flow dipstick biosensor for detection of pathogenic Leptospira. • We design primers for multiplex LAMP targeting the conserved LipL32 gene of pathogenic Leptospira and LAMP

Response of the primary piping loop to an HCDA

International Nuclear Information System (INIS)

Chang, Y.W.; Moneim, M.T.A.; Wang, C.Y.; Gvildys, J.

1975-01-01

The paper describes a method for analyzing the response of the primary piping loop that consists of straight pipes, elbows, and other components connected in series and subject to hypothetical core disruptive accident (HCDA) loads at both ends of the loop. The complete hydrodynamic equations in two-dimensions, that include both the nonlinear convective and viscous dissipation terms are used for the fluid dynamics together with the implicit ICE technique. The external walls of the pipes and components are treated as thin shells in which the analysis accounts for the membrane and bending strength of the wall, elastic-plastic material behavior, as well as large deformation under the effect of transient loading conditions. In the straight pipes, the flow is assumed to be axisymmetric; in the elbow regions, the two dimensions considered are the r and theta directions. The flow in the other components is also assumed to be axisymmetric; the components are modeled as a circular cylinder, in which the radius of the cylinder can be varied to conform with the outside shape of the component and the flow area inside can be changed independently from the outside shape. However, they must remain axially symmetric. The method is applied to a piping loop which consists of six elastic-plastic pipes and five rigid elbows connected in series and subjected to pressure pulses at both ends of the loop

Characterization of natural circulation looping of emergency cooling systems in naval and advanced reactors

International Nuclear Information System (INIS)

Macedo, Luiz Alberto; Baptista Filho, Benedito Dias

2000-01-01

This paper describes the natural circuit looping, resumes the main project characteristics, presents results of the hydraulic characterization, consisting of pressure loss measurements, and presents results from calibration tests of the power and flow measurements and the first experiments in natural circulation. Those experiments comprised transients in natural circulation with application of application of power steps. The results shown a non linear behaviour of the magnetic flow meter and a dependence on the fluid temperature as well. The assembly circuit/instrumentation/data acquisition system is suitable for the research on emergency cooling passive systems

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.)

Bistable flow spectral analysis. Repercussions on jet pumps

International Nuclear Information System (INIS)

Gavilan Moreno, C.J.

2011-01-01

Highlights: → The most important thing in this paper, is the spectral characterization of the bistable flow in a Nuclear Power Plant. → This paper goes deeper in the effect of the bistable flow over the jet pump and the induced vibrations. → The jet pump frequencies are very close to natural jet pump frequencies, in the 3rd and 6th mode. - Abstract: There have been many attempts at characterizing and predicting bistable flow in boiling water reactors (BWRs). Nevertheless, in most cases the results have only managed to develop models that analytically reproduce the phenomenon (). Modeling has been forensic in all cases, while the capacity of the model focus on determining the exclusion areas on the recirculation flow map. The bistability process is known by its effects given there is no clear definition of its causal process. In the 1980s, Hitachi technicians () managed to reproduce bistable flow in the laboratory by means of pipe geometry, similar to that which is found in recirculation loops. The result was that the low flow pattern is formed by the appearance of a quasi stationary, helicoidal vortex in the recirculation collector's branches. This vortex creates greater frictional losses than regions without vortices, at the same discharge pressure. Neither the behavior nor the dynamics of these vortices were characterized in this paper. The aim of this paper is to characterize these vortices in such a way as to enable them to provide their own frequencies and their later effect on the jet pumps. The methodology used in this study is similar to the one used previously when analyzing the bistable flow in tube arrays with cross flow (). The method employed makes use of the power spectral density function. What differs is the field of application. We will analyze a Loop B with a bistable flow and compare the high and low flow situations. The same analysis will also be carried out on the loop that has not developed the bistable flow (Loop A) at the same moments