Test facility for rewetting experiments at CDTN

International Nuclear Information System (INIS)

Rezende, Hugo C.; Mesquita, Amir Z.; Ladeira, Luiz C.D.; Santos, Andre A.C.

2015-01-01

One of the most important subjects in nuclear reactor safety analysis is the reactor core rewetting after a Loss-of-Coolant Accident (LOCA) in a Light Water Reactor LWR. Several codes for the prediction of the rewetting evolution are under development based on experimental results. In a Pressurized Water Reactor (PWR) the reflooding phase of a LOCA is when the fuel rods are rewetted from the bottom of the core to its top after having been totally uncovered and dried out. Out-of-pile reflooding experiments performed with electrical heated fuel rod simulators show different quench behavior depending the rods geometry. A test facility for rewetting experiments (ITR - Instalacao de Testes de Remolhamento) has been constructed at the Thermal Hydraulics Laboratory of the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), with the objective of performing investigations on basic phenomena that occur during the reflood phase of a LOCA in a PWR, using tubular and annular test sections. This paper presents the design aspects of the facility, and the current stage of the works. The mechanical aspects of the installation as its instrumentation are described. Two typical tests are presented and results compered with theoretical calculations using computer code. (author)

Test facility for rewetting experiments at CDTN

Energy Technology Data Exchange (ETDEWEB)

Rezende, Hugo C.; Mesquita, Amir Z.; Ladeira, Luiz C.D.; Santos, Andre A.C., E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (SETRE/CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores

2015-07-01

One of the most important subjects in nuclear reactor safety analysis is the reactor core rewetting after a Loss-of-Coolant Accident (LOCA) in a Light Water Reactor LWR. Several codes for the prediction of the rewetting evolution are under development based on experimental results. In a Pressurized Water Reactor (PWR) the reflooding phase of a LOCA is when the fuel rods are rewetted from the bottom of the core to its top after having been totally uncovered and dried out. Out-of-pile reflooding experiments performed with electrical heated fuel rod simulators show different quench behavior depending the rods geometry. A test facility for rewetting experiments (ITR - Instalacao de Testes de Remolhamento) has been constructed at the Thermal Hydraulics Laboratory of the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), with the objective of performing investigations on basic phenomena that occur during the reflood phase of a LOCA in a PWR, using tubular and annular test sections. This paper presents the design aspects of the facility, and the current stage of the works. The mechanical aspects of the installation as its instrumentation are described. Two typical tests are presented and results compered with theoretical calculations using computer code. (author)

Operating experience of steam generator test facility

International Nuclear Information System (INIS)

Sureshkumar, V.A.; Madhusoodhanan, G.; Noushad, I.B.; Ellappan, T.R.; Nashine, B.K.; Sylvia, J.I.; Rajan, K.K.; Kalyanasundaram, P.; Vaidyanathan, G.

2006-01-01

Steam Generator (SG) is the vital component of a Fast Reactor. It houses both water at high pressure and sodium at low pressure separated by a tube wall. Any damage to this barrier initiates sodium water reaction that could badly affect the plant availability. Steam Generator Test Facility (SGTF) has been set up in Indira Gandhi Centre for Atomic Research (IGCAR) to test sodium heated once through steam generator of 19 tubes similar to the PFBR SG dimension and operating conditions. The facility is also planned as a test bed to assess improved designs of the auxiliary equipments used in Fast Breeder Reactors (FBR). The maximum power of the facility is 5.7 MWt. This rating is arrived at based on techno economic consideration. This paper covers the performance of various equipments in the system such as Electro magnetic pumps, Centrifugal sodium pump, in-sodium hydrogen meters, immersion heaters, and instrumentation and control systems. Experience in the system operation, minor modifications, overall safety performance, and highlights of the experiments carried out etc. are also brought out. (author)

Final air test results for the 1/5-scale Mark I boiling water reactor pressure suppression experiment

International Nuclear Information System (INIS)

Collins, E.K.; Lai, W.

1977-01-01

A loss-of-coolant accident (LOCA) in a boiling-water reactor (BWR) power plant has never occurred. However, because this type of accident is particularly severe, it is used as a principal basis for design. During a hypothetical LOCA in a Mark I BWR, air followed by steam is injected from a drywell into a toroidal wetwell about half-filled with water. A series of consistent, versatile, and accurate air-water tests simulating LOCA conditions was completed in the Lawrence Livermore Laboratory 1/5-Scale Mark I BWR Pressure Suppression Experimental Facility. Results from this test series were used to quantify the vertical loading function and to study the associated fluid dynamic phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variations of hydrodynamic-generated vertical loads with changes in drywell pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the 1/5-scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings which are invariant. These groupongs show that if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor; the time when these forces occur is reduced by the square root of the scale factor

Development and Certification of Station Development Test Objective (SDTO) Experiment # 15012-U, "Near RealTime Water Quality Monitoring Demonstration for ISS Biocides Using Colorimetric Solid Phase Extraction (CSPE)"

Science.gov (United States)

Gazda, Daniel B.; Nolan, Daniel J.; Rutz, Jeffrey A.; Shcultz, John R.; Siperko, Lorraine M.; Porter, Marc D,; Lipert, Robert J.; Limardo, Jose G.; McCoy, J. Torin

2009-01-01

Scientists and engineers from the Wyle Integrated Science and Engineering Group are working with researchers at the University of Utah and Iowa State University to develop and certify an experimental water quality monitoring kit based on Colorimetric Solid Phase Extraction (CSPE). The kit will be launched as a Station Development Test Objective (SDTO) experiment and evaluated on the International Space Station (ISS) to determine the acceptability of CSPE technology for routine inflight water quality monitoring. Iodine and silver, the biocides used in the US and Russian on-orbit water systems, will serve as test analytes for the technology evaluation. This manuscript provides an overview of the CSPE SDTO experiment and details the development and certification of the experimental water quality monitoring kit. Initial results from reagent and standard solution stability testing and environmental testing performed on the kit hardware are also reported.

Experiment HFR-B1: A preliminary analysis of the water-vapor injection experiments in capsule 3

International Nuclear Information System (INIS)

Myers, B.F.

1993-01-01

A preliminary analysis of the response of uranium oxycarbide (UCO) fuel to water vapor addition in capsule 3 of experiment HFR-B1 (HFR-B1/3) has been conducted. The analysis provides an early indication of the behavior of fission gas release under a wider range of water-vapor pressures and of temperatures than heretofore studied. A preliminary analysis of selected aspects of the water-vapor injection tests in capsule 3 of experiment HFR-B1 is presented. The release of fission gas stored in bubbles and the diffusive release of fission-gas atoms are distinguished. The dependence of the release of stored fission gas ( 85m Kr) on water-vapor pressure, P(H 2 O), and temperature were established taking into account the contributing mechanisms of gaseous release, the effect of graphite hydrolysis, and the requirement of consistency with experiment HRB-17 in which similar water-vapor injection tests were conducted. The dependence on P(H 2 O) becomes weaker as temperatures increase above 770 degree C; the activation energy for release of stored-fission gas is 393 kJ/mol. Isorelease curves for the pressure-temperature plane were deduced from a derived functional relation. The stored-fission gas releases as a function of P(H 2 O) at a common temperature for experiments HFR-B1 and HRB-17 differ by a factor of 4; this discrepancy could be attributed to the differences in fission-rate density and neutron flux between the two experiments. Diffusive release of fission gas occurred during and after the release of stored gas. The ratio of diffusive release during water-vapor injection to that prior to injection varied in contrast to the results from HRB-17. The variation was attributed to the practice of injecting water vapor into HFR-B1 before sintering of the fuel, hydrolyzed in the previous test, was completed. The derived activation energy for diffusive release is 23.6 kJ/mol

Experiment HFR-B1: A preliminary analysis of the water-vapor injection experiments in capsule 3

Energy Technology Data Exchange (ETDEWEB)

Myers, B.F.

1993-08-01

A preliminary analysis of the response of uranium oxycarbide (UCO) fuel to water vapor addition in capsule 3 of experiment HFR-B1 (HFR-B1/3) has been conducted. The analysis provides an early indication of the behavior of fission gas release under a wider range of water-vapor pressures and of temperatures than heretofore studied. A preliminary analysis of selected aspects of the water-vapor injection tests in capsule 3 of experiment HFR-B1 is presented. The release of fission gas stored in bubbles and the diffusive release of fission-gas atoms are distinguished. The dependence of the release of stored fission gas ({sup 85m}Kr) on water-vapor pressure, P(H{sub 2}O), and temperature were established taking into account the contributing mechanisms of gaseous release, the effect of graphite hydrolysis, and the requirement of consistency with experiment HRB-17 in which similar water-vapor injection tests were conducted. The dependence on P(H{sub 2}O) becomes weaker as temperatures increase above 770{degree}C; the activation energy for release of stored-fission gas is 393 kJ/mol. Isorelease curves for the pressure-temperature plane were deduced from a derived functional relation. The stored-fission gas releases as a function of P(H{sub 2}O) at a common temperature for experiments HFR-B1 and HRB-17 differ by a factor of 4; this discrepancy could be attributed to the differences in fission-rate density and neutron flux between the two experiments. Diffusive release of fission gas occurred during and after the release of stored gas. The ratio of diffusive release during water-vapor injection to that prior to injection varied in contrast to the results from HRB-17. The variation was attributed to the practice of injecting water vapor into HFR-B1 before sintering of the fuel, hydrolyzed in the previous test, was completed. The derived activation energy for diffusive release is 23.6 kJ/mol.

Preliminary test of an ultrasonic liquid film sensor for high-temperature steam-water two-phase flow experiments

International Nuclear Information System (INIS)

Aoyama, Goro; Nagayoshi, Takuji; Baba, Atsushi

2014-01-01

A prototype liquid film sensor for high-temperature steam-water experiments has been developed. The sensor shape simulates a boiling water reactor (BWR) fuel rod. The pulse-echo method can be utilized to measure the thickness of the liquid film covering the sensor surface. A piezoelectric element is soldered onto the inside of the sensor casing which consists of two curved casing pieces. After the piezoelectric element is attached, the two casing pieces are laser welded together. It is confirmed that the temperature rise at the time of the laser welding does not influence soldering of the piezoelectric element. The pressure proof test shows that the sensor can be used at a high-pressure condition of 7 MPa. Simple air-water experiments are done at atmospheric pressure to confirm the liquid film thickness can be measured with the sensor. The fluctuation of the liquid film thickness is satisfactorily captured with the sensor. The minimum and maximum thicknesses are 0.084 and 0.180 mm, respectively. The amplitude of the waveform at 286°C is predicted by the calculation based on the acoustic impedance. It is expected that the sensor is able to measure the liquid film thickness even at BWR operating conditions. (author)

Lithium-lead/water interaction. Large break experiments

International Nuclear Information System (INIS)

Savatteri, C.; Gemelli, A.

1991-01-01

One current concept in fusion blanket module design is to utilize water as coolant and liquid lithium-lead as breeding/neutron-multiplier material. Considering the possibility of certain off-normal events, it is possible that water leakage into the liquid metal may occur due to a tube rupture. The lithium-lead/water contact can lead to a thermal and chemical reaction which should provoke an intolerable pressure increase in the blanket module. For realistic simulation of such in-blanket events, the Blanket Safety Test (BLAST) facility has been built. It simulates the transient event by injecting subcooled water under high pressure into a stagnant pool of about 500 kg liquid Pb-17Li. Eight fully instrumented large break tests were carried out under different conditions. The aim of the experiments is to study the chemical and thermal process and particularly: The pressurization history of the reaction vessel, the formation and deposition of the reaction products, the identification and propagation of the reaction zones and the temperature transient in the liquid metal. In this paper the results of all tests performed are presented and discussed. (orig.)

Investigation and mitigation of condensation induced water hammer by stratified flow experiments

Science.gov (United States)

Kadakia, Hiral J.

This research primarily focuses on the possibility of using stratified flow in preventing an occurrence of condensation induced water hammer (CIWH) in horizontal pipe involving steam and subcooled water. A two-phase flow loop simulating the passive safety systems of an advanced light water reactor was constructed and a series of stratified flow experiments were carried out involving a system of subcooled water, saturated water, and steam. Special instruments were designed to measure steam flow rate and subcooled liquid velocity. These experiments showed that when flow field conditions meet certain criteria CIWH does occur. Flow conditions used in experiments were typically observed in passive safety systems of an advanced light water cooled reactor. This research summarizes a) literature research and other experimental data that signify an occurrence of CIWH, b) experiments in an effort to show an occurrence of CIWH and the ability to prevent CIWH, c) qualitative and quantitative results to underline the mechanism of CIWH, d) experiments that show CIWH can be prevented under certain conditions, and e) guidelines for the safe operating conditions. Based on initial experiment results it was observed that Bernoulli's effect can play an important role in wave formation and instability. A separate effect table top experiment was constructed with plexi-glass. A series of entrance effect tests and stratified experiments were carried out with different fluids to study wave formation and wave bridging. Special test series experiments were carried out to investigate the presence of a saturated layer. The effect of subcooled water and steam flow on wedge length and depth were recorded. These experiments helped create a model which calculates wedge and depth of wedge for a given condition of steam and subcooled water. A very good comparison between the experiment results and the model was obtained. These experiments also showed that the presence of saturated layer can mitigate

Water Flow Experiments

Indian Academy of Sciences (India)

year undergraduate student at Ashoka University,. Sonipat, Haryana. This article studies how the height of water varies with time when water ... Experiment using a one-bottle system with a small bore tube at- tached to .... restricting free flow.

Life-cycle testing of receiving waters with Ceriodaphnia dubia

Energy Technology Data Exchange (ETDEWEB)

Stewart, A.J.; Konetsky, B.K.

1996-12-31

Seven-day tests with Ceriodaphnia dubia are commonly used to estimate toxicity of effluents or receiving waters but can sometimes yield {open_quotes}no toxicity{close_quotes} outcomes even if pollutants are present. We conducted two sets of full life-cycle tests with C. dubia to (1) see if tests with longer exposure periods would reveal evidence for toxicity that might not be evident from 7-day tests, and (2) determine the relative importance of water quality versus food as factors influencing C. dubia reproduction. In the first set of tests, C. dubia was reared in diluted mineral water (negative control), water from a stream impacted by coal fly-ash, or water from a retention basin containing sediments contaminated with mercury, other metals and polychlorinated biphenyls. The second set of tests used water from the retention basin only, but this water was either filtered or not filtered, and food was either added or not added, prior to testing. C. dubia survival and reproduction did not differ much among the three water types in the first set of tests, but these two parameters were strongly affected by the filtering and food-addition treatments in the second set of tests. Thus, C. dubia appeared to be relatively insensitive to general water-quality factors, but quite sensitive to food-related factors. Regression analyses showed that the predictability of life-time reproduction by C. dubia from the results of 7-day tests was very low (R{sup 2}< 0.35) in five of the six experiments. The increase in predictability as a function of test duration also differed among water types in the first set of tests, and among treatments in the second set of tests. Thus, 7-day tests with C. dubia may be used to quantify water-quality problems, but it may not be possible to reliably extrapolate the results of these tests to longer time scales.

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

International Nuclear Information System (INIS)

Feldman, E.M.; Patton, M.L. Jr.; Sackett, K.E.

1977-02-01

Recorded test data are presented for Test S-05-1 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among 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-05-1 was conducted from initial conditions of 2263 psia and 544 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the vessel lower plenum to simulate emergency core coolant injection in a PWR, with the flow rate based on system volume scaling

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.

Experiences made with tritium-containing water used as tracer in laboratory experiments with fluvioglacial gravels

International Nuclear Information System (INIS)

Klotz, D.; Rauert, W.

1982-01-01

Batch tests performed on 11 different Bavarian fluvioglacial gravels led to tritium distribution coefficients, which deviated not or only insignificantly from zero within the range of experimental accuracy applied to routine testings. The result of nine flow experiments in a gravelfilled column was a mean retardation factor of 1.01 +- 0.01. These experiments thus showed - as it had been expected - that 3 HHO is not significantly delayed with regard to the flow or movement of the water. (orig.) [de

Millisecond-period meltdown experiments on prompt-burst effects and molten-tin-water dropping experiments

International Nuclear Information System (INIS)

Wright, R.W.; Coats, R.L.; Schmidt, T.R.; Arakeri, V.H.

1976-01-01

The U.S. Nuclear Regulatory Commission has initiated a program of confirmatory research for the safety assessment of LMFBR plants. In the sodium-fuel interactions area, this research includes a series of real-time in-pile experiments on the pressure and work potential of prompt-burst excursions as well as laboratory dropping experiments with molten tin and water. The in-pile experiments are performed by Sandia Laboratories in the Annular Core Pulse Reactor (ACPR), which has a minimum period of 1.3 milliseconds. These single-pin experiments are performed in a piston-loaded, stagnent-sodium autoclave, that is conceptually similar to the one used in the S-11 TREAT test. Unlike the S-11 test, however, realistic radial temperature profiles are obtained in the fuel, the cladding, and the sodium by pre-pulsing the reactor about 1/2 second before the main pulse. A series of preparatory runs have been made with helium-filled capsules and at low energy with sodium-filled capsules. The first significant fuel-coolant interaction run is scheduled for late March 1976. This will be a double-pulsed run at 2700 j/gm UO 2 . A continuing series of experiments is planned with oxide and advanced fuels in both fresh and irradiated form. In molten-tin-water dropping experiments at UCLA, microsecond duration multi-flash photography has been used for event diagnostics. Transition or nucleate boiling was found to trigger energetic interactions or vapor explosions. Temperature stratification in the water was found to reduce the threshold tin temperature necessary to produce vapor explosions below that the predicted by the coolant homogeneous nucleation hypothesis. Interaction zone growth times of a few msec. were measured. (auth.)

Millisecond-Period Meltdown Experiments on Prompt - Burst Effects and Molten-Tin-Water Dropping Experiments

International Nuclear Information System (INIS)

Wright, R.W.; Coats, R.L.; Schmidt, T.R.; Arakeri, V.H.

1976-01-01

The U.S. Nuclear Regulatory Commission has initiated a program of confirmatory research for the safety assessment of LMFBR plants. In the sodium-fuel interactions area, this research includes a series of real-time in-pile experiments on the pressure and work potential of prompt-burst excursions as well as laboratory dropping experiments with molten tin and water. The in-pile experiments are performed by Sandia Laboratories in the Annular Core Pulse Reactor (ACPR), which has a minimum period of 1.3 milliseconds. These single-pin experiments are performed in a piston-loaded, stagnant-sodium autoclave, that is conceptually similar to the one used in the S-11 TREAT test. Unlike the S-11 test, however, realistic radial temperature profiles are obtained in the fuel, the cladding, and the sodium by pre-pulsing the reactor about 1/2 second before the main pulse. A series of preparatory runs have been made with helium-filled capsules and at low energy with sodium-filled capsules. The first significant fuel-coolant interaction run is scheduled for late March 1976. This will be a double-pulsed run at 2700 j/gm UO 2 . A continuing series of experiments is planned with oxide and advanced fuels in both fresh and irradiated form. In molten-tin-water dropping experiments at UCLA, microsecond duration multi-flash photography has been used for event diagnostics. Transition or nucleate boiling was found to trigger energetic interactions or vapor explosions. Temperature stratification in the water was found to reduce the threshold tin temperature necessary to produce vapor explosions below that the predicted by the coolant homogeneous nucleation hypothesis. Interaction zone growth times of a few msec were measured

Evaluation of water transport behavior in sodium fire experiment-II

Energy Technology Data Exchange (ETDEWEB)

Nakagiri, Toshio [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

2000-02-01

Evaluation of water transport behavior in Sodium Fire-II (Run-D4) was performed. Results of other experiments performed in Oarai-Engineering Center were considered in the evaluation, and the results of the evaluation were compared with the calculated results of ASSCOPS code. The main conclusions are described below. (1) It was estimated that aerosol hydrates were not formed in the test cell in the experiment, because of high gas temperatures (200degC - 300degC), but water vapor absorption by the formation of aerosol hydrates and water vapor condensation were occurred in humility measure line, because of low gas temperature (20degC - 40degC). Therefore, it was considered appropriate that measured water vapor concentration in the humidity measure line was different from the real concentration in the test cell. (2) Water vapor concentration in the test cell was assumed to be about 35,000 ppm during sodium leak, and reached to about 70,000 ppm because of water release from heated concrete (over 100degC) walls after 190 min from sodium leak started. The assumed value of about 35,000 ppm during sodium leak almost agree with assumed value from the quantity of aerosol in the humidity measure line, but no support for the value of about 70,000 ppm after 190 min could be found. Therefore, water release rate from heated concrete walls can change with their temperature history. (author)

Studies of water-in-oil emulsions : testing of emulsion formation in OHMSETT

International Nuclear Information System (INIS)

Fingas, M.; Fieldhouse, B.

2001-01-01

A study was conducted to determine the stability of water-in-oil emulsions in the OHMSETT tank facility. The results were then compared with previous laboratory studies which suggested that the stability of emulsions can be grouped into four categories, stable, unstable, meso-stable and entrained. It has been determined that entrained emulsions can retain oil by viscous forces long enough for interfacial agents, resins and asphaltenes to stabilize the droplets. This paper also described the difference in viscosity between the 4 categories of emulsion stability. The OHMSETT tests were conducted in two series of one week each. The first series of tests were conducted in July and involved 12 experiments on 2 different types of oils which were placed at varying thicknesses on the water. The second set of tests were conducted in November and involved 12 experiments on 6 oils. The rheological properties of the oils were measured and compared to the same oils undergoing emulsification in the laboratory. The oils and water-in-oil states produced were found to have analogous properties between the laboratory and the first set of tests at the OHMSETT facility. All the oils tested produced entrained water-in-oil states in both the laboratory and the test tank. The energy in the two test conditions was found to be similar, with the OHMSETT emulsions similar to one produced in the laboratory at high energies. The second series of tests at OHMSETT did not result in the expected water in-oil- states. This unexpected result was most likely due to the residual surfactant from an earlier dispersant experiment. The study showed that the conditions for emulsion formation are analogous in the OHMSETT tank and in the laboratory tests. The level of energy is considered to be the major variant. It was concluded that the energy levels between the laboratory mixing experiments and the OHMSETT is similar. It was shown that surfactants left over from dispersant testing inhibited the formation

Supercritical water natural circulation flow stability experiment research

Energy Technology Data Exchange (ETDEWEB)

Ma, Dongliang; Zhou, Tao; Li, Bing [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering; North China Electric Power Univ., Beijing (China). Inst. of Nuclear Thermalhydraulic Safety and Standardization; North China Electric Power Univ., Beijing (China). Beijing Key Lab. of Passive Safety Technology for Nuclear Energy; Huang, Yanping [Nuclear Power Institute of China, Chengdu (China). Science and Technology on Reactor System Design Technology Lab.

2017-12-15

The Thermal hydraulic characteristics of supercritical water natural circulation plays an important role in the safety of the Generation-IV supercritical water-cooled reactors. Hence it is crucial to conduct the natural circulation heat transfer experiment of supercritical water. The heat transfer characteristics have been studied under different system pressures in the natural circulation systems. Results show that the fluctuations in the subcritical flow rate (for natural circulation) is relatively small, as compared to the supercritical flow rate. By increasing the heating power, it is observed that the amplitude (and time period) of the fluctuation tends to become larger for the natural circulation of supercritical water. This tends to show the presence of flow instability in the supercritical water. It is possible to observe the flow instability phenomenon when the system pressure is suddenly reduced from the supercritical pressure state to the subcritical state. At the test outlet section, the temperature is prone to increase suddenly, whereas the blocking effect may be observed in the inlet section of the experiment.

Large scale FCI experiments in subassembly geometry. Test facility and model experiments

International Nuclear Information System (INIS)

Beutel, H.; Gast, K.

A program is outlined for the study of fuel/coolant interaction under SNR conditions. The program consists of a) under water explosion experiments with full size models of the SNR-core, in which the fuel/coolant system is simulated by a pyrotechnic mixture. b) large scale fuel/coolant interaction experiments with up to 5kg of molten UO 2 interacting with liquid sodium at 300 deg C to 600 deg C in a highly instrumented test facility simulating an SNR subassembly. The experimental results will be compared to theoretical models under development at Karlsruhe. Commencement of the experiments is expected for the beginning of 1975

Experiment data report for semiscale Mod-1 test S-06-4 (LOFT counterpart test)

International Nuclear Information System (INIS)

Gillins, R.L.; Sackett, K.E.; Coppin, C.E.

1977-12-01

Recorded test data are presented for Test S-06-4 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among 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-06-4 was conducted from initial conditions of 15,653 kPa and 564 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact loop to simulate emergency core coolant injection in a PWR. The heater rods in the electrically heated core were operated at an axial peak power density which was 100 percent of the maximum peak power density

Experiment data report for semiscale Mod-1 test S-06-1 (LOFT counterpart test)

International Nuclear Information System (INIS)

Collins, B.L.; Patton, M.L. Jr.; Sackett, K.E.

1977-07-01

Recorded test data are presented for Test S-06-1 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying an hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-1 was conducted from initial conditions of 15 568 kPa and 564 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact loop to simulate emergency core coolant injection in a PWR. The heater rods in the electrically heated core were operated at an axial peak power density which was 30% of the maximum peak power density

Experiment data report for semiscale Mod-1 Test S-06-2 (LOFT counterpart test)

International Nuclear Information System (INIS)

Patton, M.L. Jr.; Collins, B.L.; Sackett, K.E.

1977-08-01

Recorded test data are presented for Test S-06-2 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying an hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-2 was conducted from initial conditions of 15 513 kPa and 563 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact loop to simulate emergency core coolant injection in a PWR. The heater rods in the electrically heated core were operated at an axial peak power density which was 50% of the maximum peak power density

Steam line rupture experiments with the PPOOLEX test facility

International Nuclear Information System (INIS)

Laine, J.; Puustinen, M.

2008-07-01

The results of the steam line rupture experiment series in 2007 with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology are reported. The test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. Air was blown into the dry well compartment and from there through a DN200 blowdown pipe to the condensation pool. Altogether five experiments, each consisting of several blows (tests), were carried out. The main purpose of the experiment series was to study the initial phase of a postulated steam line break accident inside a BWR containment. Specifically, thermal stratification in the dry well compartment and ejection of water plug from the blowdown pipe were of interest. In addition, the effect of counterpressure on bubble dynamics was studied. A temperature difference of approximately 15 deg. C between the upper and lower part of the dry well was measured. In the wet well gas space, a temperature difference of more than 30 deg. C was registered. These were measured during the compression period of the tests. Towards the end of the tests the temperature differences tended to disappear. To get a more detailed picture of temperature distribution in the wet well, especially close to the water level, a dense net of measurements is required in future experiments. In longer experiments, heat conduction to structures and heat losses to surroundings should also be taken into account. Ejection of water plugs from the blowdown pipe did not cause notable loads to the structures due to the suppressing effect of the dry well compartment. The maximum measured pressure pulse at the pool bottom was only 10 kPa and the maximum strain amplitude at the pool bottom rounding was negligible both in axial and circumferential direction. As the counterpressure of the system increased, but the flow rate remained the same, the maximum size of the air bubbles at the blowdown pipe outlet got smaller and

Steam line rupture experiments with the PPOOLEX test facility

Energy Technology Data Exchange (ETDEWEB)

Laine, J.; Puustinen, M. (Lappeenranta Univ. of Technology, Nuclear Safety Research Unit (Finland))

2008-07-15

The results of the steam line rupture experiment series in 2007 with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology are reported. The test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. Air was blown into the dry well compartment and from there through a DN200 blowdown pipe to the condensation pool. Altogether five experiments, each consisting of several blows (tests), were carried out. The main purpose of the experiment series was to study the initial phase of a postulated steam line break accident inside a BWR containment. Specifically, thermal stratification in the dry well compartment and ejection of water plug from the blowdown pipe were of interest. In addition, the effect of counterpressure on bubble dynamics was studied. A temperature difference of approximately 15 deg. C between the upper and lower part of the dry well was measured. In the wet well gas space, a temperature difference of more than 30 deg. C was registered. These were measured during the compression period of the tests. Towards the end of the tests the temperature differences tended to disappear. To get a more detailed picture of temperature distribution in the wet well, especially close to the water level, a dense net of measurements is required in future experiments. In longer experiments, heat conduction to structures and heat losses to surroundings should also be taken into account. Ejection of water plugs from the blowdown pipe did not cause notable loads to the structures due to the suppressing effect of the dry well compartment. The maximum measured pressure pulse at the pool bottom was only 10 kPa and the maximum strain amplitude at the pool bottom rounding was negligible both in axial and circumferential direction. As the counterpressure of the system increased, but the flow rate remained the same, the maximum size of the air bubbles at the blowdown pipe outlet got smaller and

The water purification system for the low background counting test facility of the Borexino experiment at Gran Sasso

International Nuclear Information System (INIS)

Balata, M.; Cadonati, L.; Laubenstein, M.; Heusser, G.; Giammarchi, M.G.; Scardaoni, R.; Torri, V.; Cecchet, G.; De Bari, A.; Perotti, A.

1996-01-01

The Borexino experiment, for the study of solar neutrino physics, requires radiopurity at the level of 5 x 10 -16 g/g 238 U equivalent (or 6 x 10 -9 Bq/kg) on a detector mass of many tons of scintillator. Feasibility studies are performed in a counting test facility now operating at LNGS, which consists of 4 t of liquid scintillator viewed by 100 photomultipliers and shielded by 100 t of water. The accomplishment of this goal requires the shielding liquid, water, to be at the 10 -13 g/g contamination level (1.2 x 10 -6 Bq/kg) or better. This paper describes the water purification system; it consists of a combination of several purification processes to remove particulate, radioactive ions, dissolved gases and other impurities. Residual contaminations are measured by analytical or direct-counting techniques. For radon measurement, particularly challenging at this low activity levels, a low background counting method has been developed. (orig.)

Experiment data report for semiscale Mod-1 Test S-06-5. (LOFT counterpart test)

International Nuclear Information System (INIS)

1977-06-01

Recorded test data are presented for Test S-06-5 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among 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-06-5 was conducted from initial conditions of 2272 psia and 536 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold legs of the intact and broken loops to simulate emergency core coolant injection in a PWR. The purpose of Test S-06-5 was to assess the influence of the break nozzle geometry on core thermal and system response and on the subcooled and low quality mass flow rates at the break locations

FCI experiments in the corium/water system

Energy Technology Data Exchange (ETDEWEB)

Huhtiniemi, I.; Hohmann, H.; Magallon, D.

1995-09-01

The KROTOS fuel coolant interaction (FCI) tests aim at providing benchmark data to examine the effect of fuel/coolant initial conditions and mixing on explosion energetics. Experiments, fundamental in nature, are performed in well-controlled geometries and are complementary to the FARO large scale tests. Recently, a new test series was started using 3 kg of prototypical core material (80 w/o UO{sub 2}, 20 w/o ZrO{sub 2}) which was poured into a water column of {le} 1.25 m in height (95 mm and 200 mm in diameter) under 0.1 MPa ambient pressure. Four tests have been performed in the test section of 95 mm in diameter (ID) with different subcooling levels (10-80K) and with and without an external trigger. Additionally, one test has been performed with a test section of 200 mm in diameter (ID) and with an external trigger. No spontaneous or triggered energetic FCIs (steam explosions) have been observed in these corium tests. This is in sharp contrast with the steam explosions observed in the previously reported Al{sub 2}O{sub 3} test series which had the same initial conditions of ambient pressure and subcooling. The post-test analysis of the corium experiments indicated that strong vaporisation at the melt/water contact led to a partial expulsion of the melt from the test section into the pressure vessel. In order to avoid this and to obtain a good penetration and premixing os the corium melt, an additional test has been performed with a larger diameter test section. In all the UO{sub 2}-ZrO{sub 2} tests an efficient quenching process (0.7-1.2 MW/kg-melt) with total fuel fragmentation (mass mean diameter 1.4-2.5 mm) was observed. Results from Al{sub 2}O{sub 3} tests under the same initial conditions are also presented for further confirmation of the observed differences in behaviour between Al{sub 2}O{sub 3} and UO{sub 2}-ZrO{sub 2} melts.

PPOOLEX experiments on the dynamics of free water surface in the blowdown pipe

International Nuclear Information System (INIS)

Laine, J.; Puustinen, M.; Raesaenen, A.

2013-04-01

This report summarizes the results of the thermal stratification and mixing experiments carried out with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology. Steam was blown into the dry well compartment and from there through the vertical DN200 blowdown pipe to the condensation pool filled with sub-cooled water. The main objective of the experiments was to obtain verification data for the development of the Effective Momentum Source (EMS) and Effective Heat Source (EHS) models to be implemented in GOTHIC code by KTH. A detailed test matrix and procedure put together on the basis of pre-test calculations was provided by KTH before the experiments. Altogether six experiments were carried out. The experiments consisted of a small steam flow rate stratification period and of a higher flow rate mixing period. The dry well structures were heated up to approximately 130 deg. C before the stratification period was initiated. The initial water bulk temperature in the condensation pool was 13-16 deg. C. During the low steam flow rate (85-105 g/s) period steam condensed mainly inside the blowdown pipe. As a result temperatures remained constant below the blowdown pipe outlet while they increased towards the pool surface layers indicating strong thermal stratification of the wet well pool water. In the end of the stratification period the temperature difference between the pool bottom and surface was 15-30 deg. C depending on the test parameters and the duration of the low flow rate period. In the beginning of the mixing phase the steam flow rate was increased rapidly to 300-425 g/s to mix the pool water totally. Depending on the used steam flow rate and initial pool water temperature it took 150-500 s to achieve total mixing. If the test was continued long enough the water pool began to stratify again after the water bulk temperature had reached ∼50 deg. C despite of steam mass flux belonging to the chugging region of the

PPOOLEX experiments on the dynamics of free water surface in the blowdown pipe

Energy Technology Data Exchange (ETDEWEB)

Laine, J.; Puustinen, M.; Raesaenen, A. [Lappeenranta Univ. of Technology, Lappeenranta (Finland)

2013-04-15

This report summarizes the results of the thermal stratification and mixing experiments carried out with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology. Steam was blown into the dry well compartment and from there through the vertical DN200 blowdown pipe to the condensation pool filled with sub-cooled water. The main objective of the experiments was to obtain verification data for the development of the Effective Momentum Source (EMS) and Effective Heat Source (EHS) models to be implemented in GOTHIC code by KTH. A detailed test matrix and procedure put together on the basis of pre-test calculations was provided by KTH before the experiments. Altogether six experiments were carried out. The experiments consisted of a small steam flow rate stratification period and of a higher flow rate mixing period. The dry well structures were heated up to approximately 130 deg. C before the stratification period was initiated. The initial water bulk temperature in the condensation pool was 13-16 deg. C. During the low steam flow rate (85-105 g/s) period steam condensed mainly inside the blowdown pipe. As a result temperatures remained constant below the blowdown pipe outlet while they increased towards the pool surface layers indicating strong thermal stratification of the wet well pool water. In the end of the stratification period the temperature difference between the pool bottom and surface was 15-30 deg. C depending on the test parameters and the duration of the low flow rate period. In the beginning of the mixing phase the steam flow rate was increased rapidly to 300-425 g/s to mix the pool water totally. Depending on the used steam flow rate and initial pool water temperature it took 150-500 s to achieve total mixing. If the test was continued long enough the water pool began to stratify again after the water bulk temperature had reached {approx}50 deg. C despite of steam mass flux belonging to the chugging region

Experiment data report for Semiscale Mod-1 Test S-05-5 (alternate ECC injection test)

International Nuclear Information System (INIS)

Collins, B.L.; Patton, M.L. Jr.; Sackett, K.E.

1977-04-01

Recorded test data are presented for Test S-05-5 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among 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-05-5 was conducted from initial conditions of 2263 psia and 537 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. The upper plenum was vented through a reflood bypass line interconnecting the hot and cold legs of the broken loop

Critical heat flux experiments in a circular tube with heavy water and light water. (AWBA Development Program)

International Nuclear Information System (INIS)

Williams, C.L.; Beus, S.G.

1980-05-01

Experiments were performed to establish the critical heat flux (CHF) characteristics of heavy water and light water. Testing was performed with the up-flow of heavy and of light water within a 0.3744 inch inside diameter circular tube with 72.3 inches of heated length. Comparisons were made between heavy water and light water critical heat flux levels for the same local equilibrium quality at CHF, operating pressure, and nominal mass velocity. Results showed that heavy water CHF values were, on the average, 8 percent below the light water CHF values

Design of a water electrolysis flight experiment

Science.gov (United States)

Lee, M. Gene; Grigger, David J.; Thompson, C. Dean; Cusick, Robert J.

1993-01-01

Supply of oxygen (O2) and hydrogen (H2) by electolyzing water in space will play an important role in meeting the National Aeronautics and Space Administration's (NASA's) needs and goals for future space missios. Both O2 and H2 are envisioned to be used in a variety of processes including crew life support, spacecraft propulsion, extravehicular activity, electrical power generation/storage as well as in scientific experiment and manufacturing processes. The Electrolysis Performance Improvement Concept Study (EPICS) flight experiment described herein is sponsored by NASA Headquarters as a part of the In-Space Technology Experiment Program (IN-STEP). The objective of the EPICS is to further contribute to the improvement of the SEF technology, specifially by demonstrating and validating the SFE electromechanical process in microgravity as well as investigating perrformance improvements projected possible in a microgravity environment. This paper defines the experiment objective and presents the results of the preliminary design of the EPICS. The experiment will include testing three subscale self-contained SFE units: one containing baseline components, and two units having variations in key component materials. Tests will be conducted at varying current and thermal condition.

Radiological operating experience at FFTF [Fast Flux Test Facility

International Nuclear Information System (INIS)

Bunch, W.L.; Prevo, P.R.

1986-11-01

The Fast Flux Test Facility has been in operation for approximately five years, including about one thousand days of full power operation of the Fast Test Reactor. During that time the collective dose equivalents received by operating personnel have been about two orders of magnitude lower than those typically received at commercial light water reactors. No major contamination problems have been encountered in operating and maintaining the plant, and release of radioactive gas to the environment has been minimal and well below acceptable limits. All shields have performed satisfactorily. Experience to date indicates an apparent radiological superiority of liquid metal reactor systems over current light water plants

Experiment data report for LOFT nonnuclear test L1-3

International Nuclear Information System (INIS)

Millar, G.M.

1977-04-01

Test L1-3 was the third in a series of five nonnuclear isothermal blowdown tests conducted by the Loss of Fluid Test (LOFT) Program. For this test the LOFT Facility was configured to simulate a loss-of-coolant accident in a large pressurized water reactor resulting from a 200 percent double-ended shear break in a cold leg of the primary coolant system. A hydraulic core simulator assembly was installed in place of the nuclear core. The initial conditions in the primary coolant system intact loop were: temperature at 540 0 F, pressure at 2256 psig, and loop flow at 2.34 x 10 6 lbm/hr. During system depressurization, emergency core cooling water was specified to be injected into the lower plenum of the reactor vessel using an accumulator, a low-pressure injection system pump, and a high-pressure injection system pump to provide data on the effects of emergency core cooling on the system thermal-hydraulic response. Injection into the lower plenum was initiated from the high- and low-pressure injection systems. Injection from the accumulator, however, was not initiated because a valve was inadvertently left closed. The experiment, therefore, was not completely successful in that one of the objectives outlined in the experiment operating specification for this test was not accomplished. Test L1-3 was repeated at Test L1-3A to meet the experimental requirements. Despite these difficulties, Test L1-3 did provide very valuable data to verify experiment repeatability

Experiment data report for Semiscale Mod-1 Test S-05-4 (alternate ECC injection test)

International Nuclear Information System (INIS)

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

1977-03-01

Recorded test data are presented for Test S-05-4 of the Semiscale Mod-1 alternate emergency core coolant injection test series. These tests are among 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-05-4 was conducted from initial conditions of 2266 psia and 543 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the cold leg of each loop and into the vessel upper plenum to simulate emergency core coolant injection in a PWR. The upper plenum coolant injection was scaled according to the heat stored in the metal mass of the upper plenum

Experiment data report for Semiscale Mod-1 Test S-05-3 (alternate ECC injection test)

International Nuclear Information System (INIS)

Feldman, E.M.; Patton, M.L. Jr.; Sackett, K.E.

1977-03-01

Recorded test data are presented for Test S-05-3 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among 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-05-3 was conducted from initial conditions of 2263 psia and 545 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the cold leg sides of the intact and broken loops and into the vessel upper plenum to simulate emergency core coolant injection in a PWR. For Test S-05-3, specifically the effects of upper plenum coolant injection on core thermal and system response were being investigated

Study on core make-up water experiment of AC600 make-up water tank

International Nuclear Information System (INIS)

Ji Fuyun; Li Changlin; Zheng Hua; Liu Shaohua; Xu Xiaolan

1999-01-01

The core makeup tank (CMT) is a principal component of the passive high pressure safety injection systems for AC600 and has a function to inject cold borated water into reactor vessel during abnormal events. The purpose of this experiment is to verify the gravity drain behavior of the CMT and to provide experimental data to verify the computer codes used in the safety analyses. Five experiments with simulative small and medium break conditions are conducted at AC600 core makeup tank performance test facility of Nuclear Power Institute of China (NPIC). The author provides the results of one test. The simulated accident is a small break loss-of-coolant accident

Price vector effects in choice experiments: an empirical test

International Nuclear Information System (INIS)

Hanley, Nick; Wright, Robert E.; Adamowicz, Wiktor

2005-01-01

This paper investigates whether the preference and willingness-to-pay estimates obtained from the choice experiment method of estimating non-market values are sensitive to the vector of prices used in the experimental design. We undertake this test in the context of water quality improvements under the European Union's new Water Framework Directive. Using a mixed logit model, which allows for differing scale between the two samples, we find no significant impact of changing the price vector on estimates of preferences or willingness-to-pay. (author) (Choice modelling; Non-market valuation; Design effects; Water Framework Directive)

Experiment data report for semiscale Mod-1 test S-28-2 (steam generator tube rupture test)

International Nuclear Information System (INIS)

Patton, M.L.; Sackett, K.E.

1977-10-01

Recorded test data are presented for Test S-28-2 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among 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-28-2 was conducted from initial conditions of 15 936 kPa and 558 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. For Test S-28-2, accumulator injection into the intact loop hot leg was provided to simulate simulate the rupture of six steam generator tubes

Experiment data report for Semiscale Mod-1 test S-28-3 (steam generator tube rupture test)

International Nuclear Information System (INIS)

Gillins, R.L.; Sackett, K.E.

1977-10-01

Recorded test data are presented for Test S-28-3 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among 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-28-3 was conducted from initial conditions of 15621 kPa and 555 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Twelve steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg

Experiment data report for semiscale Mod-1 test S-28-4 (steam generator tube rupture test)

International Nuclear Information System (INIS)

Esparza, V.; Sackett, K.E.

1977-10-01

Recorded test data are presented for Test S-28-4 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among 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-28-4 was conducted from initial conditions of 15 646 kPa and 557 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Thirty steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg

Post-test analysis of the experiment 5.2C - total loss of feed water at the BETHSY test facility

Energy Technology Data Exchange (ETDEWEB)

Krepper, E; Schaefer, F

1998-10-01

The BETHSY-test facility is a 1:100 scaled thermohydraulic model of a 900 MW(el) pressurized water reactor (FRAMATOME). The test facility is mainly designed to investigate various accident scenarios and to provide an experimental data base for code validation and for the verification of accident management measures. (orig.)

Emission of biocides from treated materials: test procedures for water and air.

Science.gov (United States)

Schoknecht, Ute; Wegner, Robby; Horn, Wolfgang; Jann, Oliver

2003-01-01

Methods for the determination of biocide emissions from treated materials into water and air were developed and tested in order to support a comparative ecological assessment of biocidal products. Leaching tests, experiments with simulated rain, extraction cleaning of carpets and emission chamber tests were performed with a series of treated materials. The experiments focused on the effect of changes in the procedure as well as characteristics of the specimens and demonstrate the suitability of the proposed methods for biocides of different product types. It was demonstrated that emissions of biocides into water can be compared on the basis of leaching tests in which the emission kinetics of the active ingredients are recorded. However, the water volume per surface area and the timetable for water changes have to be defined in such tests. Functions of flux rates related to time can be well described for inorganic compounds, whereas modelling of the data is more complicated for organic substances. Emission chamber tests using 20-litre and 23-litre glass exsiccators, originally developed to study volatile organic compounds, were successfully adapted for the investigation of the emission of biocides from treated materials which are usually semi volatile organic compounds. However test parameters and the method of analysis have to be adapted to the substances to be determined. Generally, it was found that the emission curves for the semi volatile organic compounds investigated differ from those of volatile organic compounds.

Review of irradiation experiments for water reactor safety research

International Nuclear Information System (INIS)

Tobioka, Toshiaki

1977-02-01

A review is made of irradiation experiments for water reactor safety research under way in both commercial power plants and test reactors. Such experiments are grouped in two; first, LWR fuel performance under normal and abnormal operating conditions, and second, irradiation effects on fracture toughness in LWR vessels. In the former are fuel densification, swelling, and the influence of power ramp and cycling on fuel rod, and also fuel rod behavior under accident conditions in in-reactor experiment. In the latter are the effects of neutron exposure level on the ferritic steel of pressure vessels, etc.. (auth.)

Hot functional test chemistry - long term experience

International Nuclear Information System (INIS)

Vonkova, K.; Kysela, J.; Marcinsky, M.; Martykan, M.

2010-01-01

Primary circuit materials undergo general corrosion in high temperature, deoxygenated, neutral or mildly alkaline solutions to form thin oxide films. These oxide layers (films) serve as protective film and mitigate the further corrosion of primary materials. Inner chromium-rich oxide layer has low cation diffusion coefficients and thus control iron and nickel transport from the metal surface to the outer layer and their dissolution into the coolant. Much less corrosion products are generated by the compact, integral and stable oxide (passivation) layer. For the latest Czech and Slovak stations commissioned (Temelin and Mochovce) a modified Hot Functional Test (HFT) chemistry was developed in the NRI Rez. Chromium rich surface layer formatted due to modified HTF chemistry ensures lower corrosion rates and radiation field formation and thus also mitigates crud formation during operation. This procedure was also designed to prepare the commissioned unit for the further proper water chemistry practise. Mochovce 1 (SK) was the first station commissioned using these recommendations in 1998. Mochovce 2 (1999) and Temelin 1 and 2 (CZ - 2000 and 2002) were subsequently commissioned using these guidelines too. The main principles of the controlled primary water chemistry applied during the hot functional tests are reviewed and importance of the water chemistry, technological and other relevant parameters is stressed regarding to the quality of the passive layer formed on the primary system surfaces. Samples from Mochovce indicated that duplex oxide layers up to 20 μm thick were produced, which were mainly magnetite substituted with nickel and chromium (e.g. 60-65% Fe, 18-28% Cr, 9-12% Ni, <1% Mn and 1-2% Si on a stainless steel primary circuit sample). Long term operation experience from both nuclear power plants are discussed in this paper. Radiation field, occupational radiation exposure and corrosion layers evolution during the first c. ten years of operation are

VVER operational experience - effect of preconditioning and primary water chemistry on radioactivity build-up

International Nuclear Information System (INIS)

Zmitko, M.; Kysela, J.; Dudjakova, K.; Martykan, M.; Janesik, J.; Hanus, V.; Marcinsky, P.

2004-01-01

The primary coolant technology approaches currently used in VVER units are reviewed and compared with those used in PWR units. Standard and modified water chemistries differing in boron-potassium control are discussed. Preparation of the VVER Primary Water Chemistry Guidelines in the Czech Republic is noted. Operational experience of some VVER units, operated in the Czech Republic and Slovakia, in the field of the primary water chemistry, and radioactivity transport and build-up are presented. In Mochovce and Temelin units, a surface preconditioning (passivation) procedure has been applied during hot functional tests. The main principles of the controlled primary water chemistry applied during the hot functional tests are reviewed and importance of the water chemistry, technological and other relevant parameters is stressed regarding to the quality of the passive layer formed on the primary system surfaces. The first operational experience obtained in the course of beginning of these units operation is presented mainly with respect to the corrosion products coolant and surface activities. Effect of the initial passivation performed during hot functional tests and the primary water chemistry on corrosion products radioactivity level and radiation situation is discussed. (author)

Experiment data report for Semiscale Mod-1 Test S-28-1 (steam generator tube rupture test series)

International Nuclear Information System (INIS)

Collins, B.L.; Coppin, C.E.; Sackett, K.E.

1977-10-01

Recorded test data are presented for Test S-28-1 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among 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-28-1 was conducted from initial conditions of 15 767 kPa and 557 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Sixty steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg

Experience of Ko-Ri Unit 1 water chemistry

International Nuclear Information System (INIS)

Tae Il Lee

1983-01-01

The main focus is placed on operational experience in secondary system water chemistry (especially the steam generator) of the Ko-Ri nuclear power plant Unit 1, Republic of Korea, but primary side chemistry is also discussed. The major concern of secondary water chemistry in a PWR is that the condition of the steam generator be well maintained. Full flow deep bed condensate polishers have recently been installed and operation started in July 1982. Boric acid treatment of the steam generator was stopped and only the all volatile treatment method was used thereafter. A review of steam generator integrity, the chemistry control programme, secondary water quality, etc. is considered to be of great value regarding the operation of Unit 1 and future units now under startup testing or construction in the Republic of Korea. (author)

Design Development of SMART ECC Water Asymmetric Two-phase choking test facility

Energy Technology Data Exchange (ETDEWEB)

Cho, Young Il; Cho, Seok; Ko, Yung Joo; Shin, Yong Cheol; Kwon, Tae Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-05-15

SMART pressurized water reactor type is different from the existing integral NSSS commercial pressurized water reactor system which is equipped with the main features. In addition RCS piping is removed and the feature of the SBLOCA is a major design break accident. The TASS / SMR code is analyzed SMART SBLOCA. In order to verify analysis code, SMART analysis for verification of conservatism is promoting using data for experiments with Integral Effect Test and Separate Effect. In this paper, the design feature of the SWAT (SMART ECC Water Asymmetric Two-phase choking test facility) is described. SWAT is linearly reduced to a 1/5 ratio while the geometrical shape is conserved. In major shape of SMART ECC injection performance test, distortions which caused by gravitational effects are minimized. Because both the emergency core cooling water injection nozzle height and the break nozzle height match the RCP Suction Nozzle height in test section of the main forms. The main part of the test section is SG-side upper down-comer. The boundary conditions are saturated steam and water flow condition and drain flow rate to control the collapsed water level in the down-comer

Water-cooled, fire boom blanket, test and evaluation for system prototype development

International Nuclear Information System (INIS)

Stahovec, J. G.; Urban, R. W.

1999-01-01

Initial development of actively cooled fire booms indicated that water-cooled barriers could withstand direct oil fire for several hours with little damage if cooling water were continuously supplied. Despite these early promising developments, it was realized that to build reliable full-scale system for Navy host salvage booms would require several development tests and lengthy evaluations. In this experiment several types of water-cooled fire blankets were tested at the Oil and Hazardous Materials Simulated Test Tank (OHMSETT). After the burn test the blankets were inspected for damage and additional tests were conducted to determine handling characteristics for deployment, recovery, cleaning and maintenance. Test results showed that water-cooled fire boom blankets can be used on conventional offshore oil containment booms to extend their use for controlling large floating-oil marine fires. Results also demonstrated the importance of using thermoset rubber coated fabrics in the host boom to maintain sufficient reserve seam strength at elevated temperatures. The suitability of passively cooled covers should be investigated to protect equipment and boom from indirect fire exposure. 1 ref., 2 tabs., 8 figs

Simulation of water hammer experiments using RELAP5 code

International Nuclear Information System (INIS)

Kaliatka, A.; Vaisnoras, M.

2005-01-01

The rapid closing or opening of a valve causes pressure transients in pipelines. The fast deceleration of the liquid results in high pressure surges upstream the valve, thus the kinetic energy is transformed into the potential energy, which leads to the temporary pressure increases. This phenomenon is called water hammer. The intensity of water hammer effects will depend upon the rate of change in the velocity or momentum. Generally water hammer can occur in any thermal-hydraulic systems and it is extremely dangerous for the thermal-hydraulic system since, if the pressure induced exceeds the pressure range of a pipe given by the manufacturer, it can lead to the failure of the pipeline integrity. Due to its potential for damage of pipes, water hammer has been a subject of study since the middle of the nineteenth century. Many theoretical and experimental investigations were performed. The experimental investigation of the water hammer tests performed at Fraunhofer Institute for Environmental, Safety and Energy Technology (UMSICHT) [1] and Cold Water Hammer experiment performed by Forschungszentrum Rossendorf (CWHTF) [2] should be mentioned. The UMSICHT facility in Oberhausen was modified in order to simulate a piping system and associated supports that are typical for a nuclear power plant [3]. The Cold water hammer experiment is interesting and instructive because it covers a wide spectrum of particularities. One of them is sub-cooled water interaction with condensing steam at the closed end of the vertical pipe at room temperature and corresponding saturation pressure [4]. In the paper, the capabilities of RELAP5 code to correctly represent the water hammer phenomenon are presented. Paper presents the comparison of RELAP5 calculated and measured at UMSICHT and CWHTF test facilities pressure transient values after the fast closure (opening) of valves. The analyses of rarefaction wave travels inside the pipe and condensation of vapour bubbles in the liquid column

Experiment data report for LOFT nonnuclear Test L1-4

International Nuclear Information System (INIS)

Batt, D.L.

1977-07-01

Test L1-4 was the fourth in a series of five nonnuclear isothermal blowdown tests conducted by the Loss of Fluid Test (LOFT) Program. Test L1-4 was the first Nuclear Regulatory Commission standard problem (International Problem No. 5 and U.S. Problem No. 7) experiment conducted at LOFT. Data from this test will be compared with predictions generated by the standard problem participants. For this test the LOFT Facility was configured to simulate a loss-of-coolant accident in a large pressurized water reactor resulting from a 200% double-ended offset shear break in a cold leg of the primary coolant system. A hydraulic core simulator assembly was installed in place of the nuclear core. The initial conditions in the primary coolant system intact loop were temperature at 279 0 C, gauge pressure at 15.65 MPa, and intact loop flow at 268.4 kg/s. During system depressurization into a simulated containment, emergency core cooling water was injected into the primary coolant system cold leg to provide data on the effects of emergency core cooling on system thermalhydraulic response

Design of a materials testing experiment for the INTOR

International Nuclear Information System (INIS)

Vogel, M.A.; Opperman, E.K.

1981-01-01

The United States, Japan, USSR and the European community are jointly participating in the design of an International Tokamak Reactor called INTOR. In support of the US contribution to the INTOR design, the features of an experiment for bulk neutron irradiation damage studies were developed. It is anticipated that materials testing will be an important part of the programmatic mission of INTOR and consequently the requirements for materials testing in INTOR must be identified early in the reactor design to insure compatibility. The design features of the experiment, called a Channel Test, are given in this paper. The major components of the channel test are the water cooled heat sink (channel module) and the specimen capsule. The temperature within each of the 153 specimen capsules is predetermined by engineering the thermal barrier between the specimen capsule and heat sink. Individual capsules can be independently accessed and are designed to operate at a predetermined temperature within the range of 50 to 700 0 C. The total irradiation volume within a single channel test is 45 liters. Features of the channel test that result in experimental versatility and simplified remote access and handling are discussed

Start-up test of the prototype heavy water reactor 'FUGEN', (1)

International Nuclear Information System (INIS)

Ando, Hideki; Kawahara, Toshio

1982-01-01

The advanced thermal prototype reactor ''Fugen'' is a heavy water-moderated, boiling light water-cooled power reactor with electric output of 165 MW, which has been developed since 1966 as a national project. The start-up test was begun in March, 1978, being scheduled for about one year, and in March, 1979, it passed the final pre-use inspection and began the full scale operation. In this paper, the result of the start-up test of Fugen is reported. From the experience of the start-up test of Fugen, the following matters are important for the execution of start-up test. 1) Exact testing plan and work schedule, 2) the organization to perform the test, 3) the rapid evaluation of test results and the reflection to next testing plan, and 4) the reflection of test results to rated operation, regular inspection and so on. In the testing plan, the core characteristics peculiar to Fugen, and the features of heavy water-helium system, control system and other equipment were added to the contents of the start-up test of BWRs. The items of the start-up test were reactor physics test, plant equipment performance test, plant dynamic characteristic test, chemical and radiation measurement, and combined test. The organization to perform the start-up test, and the progress and the results of the test are reported. (Kako, I.)

GTS-LCS, in-situ experiment 2. Modeling of tracer test 09-03

International Nuclear Information System (INIS)

Manette, M.; Saaltink, M.W.; Soler, J.M.

2015-02-01

Within the framework of the GTS-LCS project (Grimsel Test Site - Long-Term Cement Studies), an in-situ experiment lasting about 5 years was started in 2009 to study water-cement-rock interactions in a fractured granite. Prior to the experiment, a tracer test was performed to characterize the initial flow and transport properties of the rock around the experimental boreholes. This study reports on the model interpretation of tracer test 09-03. The calculations were performed by means of a two-dimensional model (homogeneous fracture plane including 3 boreholes) using the Retraso-CodeBright software package. In the tracer test, Grimsel groundwater containing the tracer (uranine) was circulated in the emplacement borehole during 43 days (zero injection flow rate). Circulation continued without tracer afterwards. Water was extracted at the observation and extraction boreholes. Results from a model sensitivity analysis comparing model results with measured tracer concentrations showed 3 cases where the evolution of tracer concentrations in the 3 different boreholes was satisfactory. In these cases a low-permeability skin affected the emplacement and observation boreholes. No skin appeared to affect the extraction borehole. The background hydraulic gradient seems to have no effect on the results of the tracer test. These results will be applied in the calculation of the initial flow field for the reactive transport phase of in-situ experiment 2 (interaction between pre-hardened cement and fractured granite at Grimsel). (orig.)

Water chemistry experiences in a Japanese BWR during preoperational test period

International Nuclear Information System (INIS)

Nagasawa, K.; Maeda, K.; Nakamura, M.

1994-01-01

The interest in organic impurities intrusion into the BWR's coolant has been increased in recent years. Organic impurities can be measured as total organic carbons (TOC) by using commercial TOC measuring instruments. The corrosive impurities such as sulfates and chlorides, produced by decomposition of the TOC, will enhance the crack growth rate of stress corrosion cracking (SCC) in the reactor cooling system. During the preoperational test period, the reactor water quality changed remarkably by TOC intrusion. The TOC intrusion was monitored by measuring main steam conductivity. It is appeared that the main steam conductivity measurement is one of the effective and reliable techniques to detect the TOC intrusion into the reactor. (authors). 4 figs., 5 refs

Steam--water mixing in nuclear reactor safety loss-of-coolant experiments

International Nuclear Information System (INIS)

Naff, S.A.; Schwarz, W.F.

1978-01-01

Computer models used to predict the response of reactors to hypothesized accidents necessarily incorporate approximating assumptions. To verify the models by comparing predicted and measured responses in test facilities, these assumptions must be confirmed to be realistic. Recent experiments in facilities capable of repeatedly duplicating the transient behavior of a pressurized water reactor undergoing a pipe rupture show that the assumption of complete water-steam mixing during the transient results in the predicted decompression being faster than that observed. Water reactor safety studies currently in progress include programs aimed at the verification of computer models or ''codes'' used to predict reactor system responses to various hypothesized accidents. The approach is to compare code predictions of transients with the actual test transients in experimental facilities. The purpose of this paper is to explain an important instance in which predictions and data are not in complete agreement and to indicate the significance to water reactor safety studies

Experiment of the downcomer effective water head during a reflood phase of PWR LOCA

International Nuclear Information System (INIS)

Sudo, Yukio; Murao, Yoshio

1978-12-01

The results and analysis are described of a downcomer effective water head experiment. Downcomer effective water head is the driving force to feed an emergency coolant to the core during a reflood phase of PWR LOCA. The test rig has dimensions of the full-scale height and gap. Experimental conditions are: downcomer wall temperature = 250 0 -- 300 0 C, back pressure = 1 atm, coolant temperature = 98 0 -- 100 0 C, extraction water velocity = 0 -- 2 cm/s, and gap size = 200 mm. The effective water head histories obtained by experiment were compared with those predicted from the heat release from the downcomer walls. The heat release was calculated from the temperature histories indicated by thermocouples instrumented in and on the walls during experiment. The following were revealed: (1) The relation of heat flux and superheat (q vs ΔT sub(s)) obtained in the experiment is much different from that in pool boiling. (2) The predicted effective water head is in good agreement with the experimental one after 120 sec from the initiation of coolant injection. (3) The effect of extraction water velocity is negligible. (4) The effect of initial wall temperatures is evident. (author)

Optimum coagulant forecasting by modeling jar test experiments using ANNs

Science.gov (United States)

Haghiri, Sadaf; Daghighi, Amin; Moharramzadeh, Sina

2018-01-01

Currently, the proper utilization of water treatment plants and optimizing their use is of particular importance. Coagulation and flocculation in water treatment are the common ways through which the use of coagulants leads to instability of particles and the formation of larger and heavier particles, resulting in improvement of sedimentation and filtration processes. Determination of the optimum dose of such a coagulant is of particular significance. A high dose, in addition to adding costs, can cause the sediment to remain in the filtrate, a dangerous condition according to the standards, while a sub-adequate dose of coagulants can result in the reducing the required quality and acceptable performance of the coagulation process. Although jar tests are used for testing coagulants, such experiments face many constraints with respect to evaluating the results produced by sudden changes in input water because of their significant costs, long time requirements, and complex relationships among the many factors (turbidity, temperature, pH, alkalinity, etc.) that can influence the efficiency of coagulant and test results. Modeling can be used to overcome these limitations; in this research study, an artificial neural network (ANN) multi-layer perceptron (MLP) with one hidden layer has been used for modeling the jar test to determine the dosage level of used coagulant in water treatment processes. The data contained in this research have been obtained from the drinking water treatment plant located in Ardabil province in Iran. To evaluate the performance of the model, the mean squared error (MSE) and correlation coefficient (R2) parameters have been used. The obtained values are within an acceptable range that demonstrates the high accuracy of the models with respect to the estimation of water-quality characteristics and the optimal dosages of coagulants; so using these models will allow operators to not only reduce costs and time taken to perform experimental jar tests

Experiences on removal of sodium-water reaction products in SWAT-3

International Nuclear Information System (INIS)

Tanabe, H.; Hiroi, H.; Sato, M.; Otaka, J.

2002-01-01

This report summarizes experiences and information concerning the removal of sodium water reaction products (SMRP) obtained through large leak tests of the Steam Generator Safety Test Facility (SWAT-3) at PNC/OEC, which were conducted to validate the safety design of steam generators of a prototype LMFBR Monju. The following three problems are discussed here: (1) drainability of SWRP, (2) removal of SWRP by using a cold trap, and (3) steam cleaning of SWRP. (author)

Nevada test site water-supply wells

International Nuclear Information System (INIS)

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

1996-05-01

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

Experiments on interactions between zirconium-containing melt and water (ZREX). Hydrogen generation and chemical augmentation of energetics

Energy Technology Data Exchange (ETDEWEB)

Cho, D.H.; Armstrong, D.R.; Gunther, W.H. [Argonne National Lab., IL (United States); Basu, S.

1998-01-01

The results of the first data series of experiments on interactions between zirconium-containing melt and water are described. These experiments involved dropping 1-kg batches of pure zirconium or zirconium-zirconium dioxide mixture melt into a column of water. A total of nine tests were conducted, including four with pure zirconium melt and five with Zr-ZrO{sub 2} mixture melt. Explosions took place only in those tests which were externally triggered. While the extent of zirconium oxidation in the triggered experiments was quite extensive, the estimated explosion energetics were found to be very small compared to the combined thermal and chemical energy available. (author)

Experiment for water-flow measurement by pulsed-neutron activation

International Nuclear Information System (INIS)

Drozdowicz, K.

1994-08-01

An experiment is presented which constitutes a feasibility study for applying the neutron activation method for measurement of the water mass transport in pipings, e.g. in nuclear power stations. The fast neutron generator has been used as a pulsed-neutron activation source for oxygen in water which circulated in a closed system. The γ radiation of the nitrogen product isotope has been measured by the scintillation detectors placed in two positions at the piping. The two time distributions of the pulses have been recorded by a multiscaler (a software design based on CAMAC). The water flow velocity has been estimated from the peak-to-peak time distance. The tests have been performed under different experimental conditions (the neutron pulse duration, the time channel width, the water flow velocity) to define the stability, reproducibility and reliability of the measurement. The detailed results are presented in tables and in time distribution plots. The method has been found useful for the application considered. 4 refs, 17 figs, 5 tabs

Controlled laboratory experiments and modeling of vegetative filter strips with shallow water tables

Science.gov (United States)

Fox, Garey A.; Muñoz-Carpena, Rafael; Purvis, Rebecca A.

2018-01-01

Natural or planted vegetation at the edge of fields or adjacent to streams, also known as vegetative filter strips (VFS), are commonly used as an environmental mitigation practice for runoff pollution and agrochemical spray drift. The VFS position in lowlands near water bodies often implies the presence of a seasonal shallow water table (WT). In spite of its potential importance, there is limited experimental work that systematically studies the effect of shallow WTs on VFS efficacy. Previous research recently coupled a new physically based algorithm describing infiltration into soils bounded by a water table into the VFS numerical overland flow and transport model, VFSMOD, to simulate VFS dynamics under shallow WT conditions. In this study, we tested the performance of the model against laboratory mesoscale data under controlled conditions. A laboratory soil box (1.0 m wide, 2.0 m long, and 0.7 m deep) was used to simulate a VFS and quantify the influence of shallow WTs on runoff. Experiments included planted Bermuda grass on repacked silt loam and sandy loam soils. A series of experiments were performed including a free drainage case (no WT) and a static shallow water table (0.3-0.4 m below ground surface). For each soil type, this research first calibrated VFSMOD to the observed outflow hydrograph for the free drainage experiments to parameterize the soil hydraulic and vegetation parameters, and then evaluated the model based on outflow hydrographs for the shallow WT experiments. This research used several statistical metrics and a new approach based on hypothesis testing of the Nash-Sutcliffe model efficiency coefficient (NSE) to evaluate model performance. The new VFSMOD routines successfully simulated the outflow hydrographs under both free drainage and shallow WT conditions. Statistical metrics considered the model performance valid with greater than 99.5% probability across all scenarios. This research also simulated the shallow water table experiments with

Air-water tests in support of LLTR series II Test A-4

International Nuclear Information System (INIS)

Chen, K.

1980-07-01

A series of tests injecting air into a tank of stagnant water was conducted in June 1980 utilizing the GE Plenum Mixing Test Facility in San Jose, California. The test was concerned with investigating the behavior of air jets at a submerged orifice in water over a wide range of flow rates. The main objective was to improve the basic understanding of gas-liquid phenomena (e.g., leak dynamics, gas bubble agglomeration, etc.) in a simulated tube bundle through visualization. The experimental results from these air-water tests will be used as a guide to help select the leak size for LLTR Series II Test A-4 because air-water system is a good simulation of water-sodium mixture

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

International Nuclear Information System (INIS)

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

2001-01-01

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

The Functional Test for Agility Performance is a Reliable Quick Decision-Making Test for Skilled Water Polo Players

Directory of Open Access Journals (Sweden)

Tucher Guilherme

2015-06-01

Full Text Available The reliability of the Functional Test for Agility Performance has only been evaluated in water polo players in a small group of novice athletes. Thus, the aim of this study was to evaluate the reliability of the Functional Test for Agility Performance in skilled water polo players. Forty-two athletes (17.81 ± 3.24 years old with a minimum of 5 years of competitive experience (7.05 ± 2.84 years and playing at the national or international level were evaluated. The Functional Test for Agility Performance is characterized as a specific open decision-making test where a tested player moves as quickly as possible in accordance to a pass made by another player. The time spent in the test was measured by two experienced coaches. Descriptive statistics, repeated measures analysis of variance (ANOVA, 95% limit of agreement (LOA, intraclass correlation coefficient (ICC and standard error of measurements (SEM were used for data analysis. Athletes completed the Functional Test for Agility Performance in 4.15 0.47 s. The ICC value was 0.87 (95% IC = 0.80-0.92. The SEM varied between 0.24 and 0.38 s. The LOA was 1.20 s and the CV average considering each individual trial was 6%. The Functional Test for Agility Performance was shown to be a reliable quick decision-making test for skilled water polo players.

A water-cooling solution for PC-racks of the LHC experiments

CERN Document Server

Vannerem, P

2004-01-01

With ever increasing power consumption and heat dissipation of todays CPUs, cooling of rack-mounted PCs is an issue for the future online farms of the LHC experiments. In order to investigate the viability of a water-cooling solution, a prototype PC-farm rack has been equipped with a commercially available retrofitted heat exchanger. The project has been carried out as a collaboration of the four LHC experiments and the PH-ESS group . This note reports on the results of a series of cooling and power measurements of the prototype rack with configurations of 30 to 48 PCs. The cooling performance of the rack-cooler is found to be adequate; it extracts the heat dissipated by the CPUs efficiently into the cooling water. Hence, the closed PC rack transfers almost no heat into the room. The measurements and the failure tests show that the rack-cooler concept is a viable solution for the future PC farms of the LHC experiments.

Two-phase flow experiments in a model of the hot leg of a pressurised water reactor. Technical report

Energy Technology Data Exchange (ETDEWEB)

Seidel, Tobias; Vallee, Christophe; Lucas, Dirk; Beyer, Matthias; Deendarlianto

2011-09-15

In order to investigate the two-phase flow behaviour in a complex reactor-typical geometry and to supply suitable data for CFD code validation, a model of the hot leg of a pressurised water reactor was built at FZD. The hot leg model is operated in the pressure chamber of the TOPFLOW test facility, which is used to perform high-pressure experiments under pressure equilibrium with the inside atmosphere of the chamber. This technique makes it possible to visualise the two-phase flow through large windows, also at reactor-typical pressure levels. In order to optimise the optical observation possibilities, the test section was designed with a rectangular cross-section. Experiments were performed with air and water at 1.5 and 3.0 bar at room temperature as well as with steam and water at 15, 30 and 50 bar and the corresponding saturation temperature (i.e. up to 264 C). The total of 194 runs are divided into 4 types of experiments covering stationary co-current flow, counter-current flow, flow without water circulation and transient counter-current flow limitation (CCFL) experiments. This report provides a detailed documentation of the experiments including information on the experimental setup, experimental procedure, test matrix and on the calibration of the measuring devices. The available data is described and data sheets were arranged for each experiment in order to give an overview of the most important parameters. For the cocurrent flow experiments, water level histograms were arranged and used to characterise the flow in the hot leg. In fact, the form of the probability distribution was found to be sensitive to the boundary conditions and, therefore, is useful for the CFD comparison. Furthermore, the flooding characteristics of the hot leg model plotted in terms of the classical Wallis parameter or Kutateladze number were found to fail to properly correlate the data of the air/water and steam/water series. Therefore, a modified Wallis parameter is proposed, which

Two-phase flow experiments in a model of the hot leg of a pressurised water reactor. Technical report

International Nuclear Information System (INIS)

Seidel, Tobias; Vallee, Christophe; Lucas, Dirk; Beyer, Matthias; Deendarlianto

2011-09-01

In order to investigate the two-phase flow behaviour in a complex reactor-typical geometry and to supply suitable data for CFD code validation, a model of the hot leg of a pressurised water reactor was built at FZD. The hot leg model is operated in the pressure chamber of the TOPFLOW test facility, which is used to perform high-pressure experiments under pressure equilibrium with the inside atmosphere of the chamber. This technique makes it possible to visualise the two-phase flow through large windows, also at reactor-typical pressure levels. In order to optimise the optical observation possibilities, the test section was designed with a rectangular cross-section. Experiments were performed with air and water at 1.5 and 3.0 bar at room temperature as well as with steam and water at 15, 30 and 50 bar and the corresponding saturation temperature (i.e. up to 264 C). The total of 194 runs are divided into 4 types of experiments covering stationary co-current flow, counter-current flow, flow without water circulation and transient counter-current flow limitation (CCFL) experiments. This report provides a detailed documentation of the experiments including information on the experimental setup, experimental procedure, test matrix and on the calibration of the measuring devices. The available data is described and data sheets were arranged for each experiment in order to give an overview of the most important parameters. For the cocurrent flow experiments, water level histograms were arranged and used to characterise the flow in the hot leg. In fact, the form of the probability distribution was found to be sensitive to the boundary conditions and, therefore, is useful for the CFD comparison. Furthermore, the flooding characteristics of the hot leg model plotted in terms of the classical Wallis parameter or Kutateladze number were found to fail to properly correlate the data of the air/water and steam/water series. Therefore, a modified Wallis parameter is proposed, which

Water Intake by Soil, Experiments for High School Students.

Science.gov (United States)

1969

Presented are a variety of surface run-off experiments for high school students. The experiments are analogies to basic concepts about water intake, as related to water delivery, soil properties and management, floods, and conservation measures. The materials needed to perform the experiments are easily obtainable. The experiments are followed by…

Water-hammer experiences in the startup and early operation of the Shearon Harris plant

International Nuclear Information System (INIS)

Casada, D.

1987-01-01

During the startup testing and initial operation of any power plant, systems undergo planned transients to confirm their ability to respond as designed. In addition to planned testing, unplanned transients infrequently occur as equipment is initially used. Some of both the planned and unplanned transients result in water-hammer events. Several water-hammer events experienced during preoperational testing, power ascension testing, and early operation of the Shearon Harris nuclear power plant are presented. The cause, effect, and corrective actions undertaken to prevent recurrence for each event discussed are provided. In addition, examples of preventive measures taken, based on industry experience accrued at other plants, to prevent a similar occurrence at Shearon Harris are discussed

Simulation experiments for a large leak sodium-water reaction analysis. Volume 4. IHTS/relief system simulation tests

International Nuclear Information System (INIS)

Ploeger, D.W.

1978-09-01

Tests were performed in which a simplified 1/8-scale model of the intermediate heat transfer system and relief system of a LMFBR was subjected to a simulated sodium-water reaction in a steam generator. Pressures in the intermediate heat exchanger (IHX) and in the pipe were measured. The flow of water through the relief system was photographed and its velocity was measured. The forces on the relief system elbows resulting from the fluid flow were also measured. The tests were performed primarily to validate pulse propagation codes used for design and for direct use as design data

Seismic tomography Technology for the Water Infiltration Experiment

International Nuclear Information System (INIS)

Descour, J.

2001-01-01

NSA Engineering, Inc., conducted seismic tomography surveys in Niche No.3 in the Exploratory Studies Facility (ESF), Yucca Mountain, Nevada, and Alcove No.8 in the Enhanced Characterization of the Repository Block (ECRB) cross drift as part of the Infiltration Experiment being conducted in Niche No.3. NSA Engineering is a direct support contractor to the Yucca Mountain Project. This report documents the work performed from August 14 through 30, 2000, prior to the beginning of the infiltration experiment. The objective of the seismic tomography survey was to investigate the flow path of water between access drifts and more specifically to (Kramer 2000): (1) Conduct a baseline seismic tomography survey prior to the infiltration experiment; (2) Produce 2-D and 3-D tomographic images of the rock volume between Alcove No.8 and Niche No.3; (3) Correlate tomography results with published structural and lithological features, and with other geophysical data such as ground penetrating radar (GPR); and (4) Results of this survey will form a baseline with which to compare subsequent changes to the rock mass. These changes may be as a result of the water infiltration tests that could be conducted in Alcove No.8 in 2001. The scope of this reported work is to use the velocity tomograms to: (a) assess the structures and lithologic features within the surveyed area and/or volume between the two access drifts; and (b) provide information on the structural state of the rock mass as inferred by the velocity signatures of the rock prior to the beginning of the infiltration experiment

Small-Scale Water Ingression and Crust Strength Tests (SSWICS) SSWICS-6 test data report: thermal hydraulic results, Rev. 0

International Nuclear Information System (INIS)

Lomperski, S.; Farmer, M.T.; Kilsdonk, D.; Aeschlimann, B.

2011-01-01

The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core/concrete interaction (MCCI) when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. However, due to the integral nature of these tests, several questions regarding the crust freezing behavior could not be adequately resolved. These questions include: (1) To what extent does water ingression into the crust increase the melt quench rate above the conduction-limited rate and how is this affected by melt composition and system pressure? (2) What is the fracture strength of the corium crust when subjected to a thermal-mechanical load and how does it depend upon the melt composition? A series of separate-effects experiments are being conducted to address these issues. The first employs an apparatus designed to measure the quench rate of a pool of corium (∼ φ 30 cm; up to 20 cm deep). The main parameter to be varied in these quench tests is the melt composition since it is thought to have a critical influence on the crust cracking behavior which, in turn, alters quench rate. The issue of crust strength is being addressed with a second apparatus designed to mechanically load the crust produced by the quench tests. This apparatus measures the fracture strength of the crust while it is either at room temperature or above, the latter state being achieved with a heating element placed below the crust. The two apparatuses used to measure the melt quench rate and crust strength are jointly referred to as SSWICS (Small-Scale Water Ingression and Crust Strength). This report describes results of the sixth water ingression test, designated SSWICS-6. This test investigated

Disturbances to metal partitioning during toxicity testing of iron(II)-rich estuarine pore waters and whole sediments.

Science.gov (United States)

Simpson, Stuart L; Batley, Graeme E

2003-02-01

Metal partitioning is altered when suboxic estuarine sediments containing Fe(II)-rich pore waters are disturbed during collection, preparation, and toxicity testing. Experiments with model Fe(II)-rich pore waters demonstrated the rates at which adsorptive losses of Cd, Cu, Ni, Mn, Pb, and Zn occur upon exposure to air. Experiments with Zn-contaminated estuarine sediments demonstrated large and often unpredictable changes to metal partitioning during sediment storage, removal of organisms, and homogenization before testing. Small modifications to conditions, such as aeration of overlying waters, caused large changes to the metal partitioning. Disturbances caused by sediment collection required many weeks for reestablishment of equilibrium. Bioturbation by benthic organisms led to oxidation of pore-water Fe(II) and lower Zn fluxes because of the formation of Fe hydroxide precipitates that adsorb pore-water Zn. For five weeks after the addition of organisms to sediments, Zn fluxes increased slowly as the organisms established themselves in the sediments, indicating that the establishment of equilibrium was not rapid. The results are discussed in terms of the dynamic nature of suboxic, Fe(II)-rich estuarine sediments, how organisms perturb their environment, and the importance of understanding chemistry in toxicity testing with whole sediments or pore water. Recommendations are provided for the handling of sediments for toxicity testing.

Air-water mixing experiments for direct vessel injection of KNGR

International Nuclear Information System (INIS)

Hwang, Do Hyun

2000-02-01

Two air-water mixing experiments are conducted to understand the flow behavior in the downcomer for Direct Vessel Injection (DVI) of Korean Next Generation Reactor (KNGR). In the first experiment which is an air-water experiment in the rectangular channel with the gap size of 1cm, the width of water film is proportional to the water and air velocities and the inclined angle is proportional to the water velocity only, regardless of the water velocity injected in the rectangular channel. It is observed that the amount of entrained water is negligible. In the second experiment which is a full-scaled water jetting experiment without air flow, the width of water film is proportional to the flow rate injected from the pipe exit and the film thickness of water varies from 1.0mm to 5.0mm, and the maximum thickness does not exceed 5.0mm. The amount of water separated from the liquid film after striking of water jetting on the wall is measured. The amount of separation water is proportional to the flow rate, but the separation ratio in the full-scaled water jetting is not over 15%. A simplified physical model, which is designed to predict the trajectories of the width of water film, is validated through the comparison with experiment results. The 13 .deg. upward water droplet of the water injected from the pipe constitutes the outermost boundary at 1.7m below from pipe level, after the water impinges against the wall. In the model, the parameter, η which represents the relationship between the jetting velocity and the initial spreading velocity, is inversely proportional to the water velocity when it impinges against the wall. The error of the predictions by the model is decreased within 14% to the experimental data through use of exponential fitting of η for the jetting water velocity

Modernized CDTN's air-water experimental test circuit: initial results

International Nuclear Information System (INIS)

Pessoa, Mácio A.; Sobrinho, Mauricio R. da S.; Salomão, Eduardo A.; Ferreira, Arthur F.J.; Navarro, Moysés A.; Santos, André A. Campagnole dos

2017-01-01

The Counter Current Flow Limitation (CCFL) phenomenon, specifically the control that the gas exerts in a liquid flow in the opposite direction, is of real importance in the study of design and operation of various industrial sectors, particularly the nuclear industry. In nuclear engineering, such a phenomenon can occur in a loss of coolant accident (LOCA) of a Pressurized Water Reactor (PWR) when there is the need to re-flood the reactor core during an emergency cooling process. The CCFL phenomenon is being investigated at the Nuclear Technology Development Center (CDTN) thermo-hydraulics laboratory in order to better understand the flow and its limitations and thereby contribute to the improvement of its modeling for analysis of severe accidents. For this, a series of experiments were performed in CDTN in a reduced scale acrylic test section of the 'hot leg' of a PWR. The new proposed circuit is a closed loop and no water has to be discharged during the experiment. This is only possible due to the Python program, which is associated to the data acquisition system and can interface with the automated valves through the outputs of the data acquisition board to control the experiment. The trials compare the CCFL behavior for 500mm lengths of the horizontal section, for inclined duct slope 50° for a diameter of 54mm pipe's diameter. This paper describes the new tests in comparison to tests performed in the past. (author)

Modernized CDTN's air-water experimental test circuit: initial results

Energy Technology Data Exchange (ETDEWEB)

Pessoa, Mácio A.; Sobrinho, Mauricio R. da S.; Salomão, Eduardo A.; Ferreira, Arthur F.J.; Navarro, Moysés A.; Santos, André A. Campagnole dos, E-mail: marcioaraujopessoa@gmail.com, E-mail: mauricio.sobrinho223@gmail.com, E-mail: e.a.salomao@gmail.com, E-mail: arthur1303@gmail.com, E-mail: moysesnavarro@yahoo.com.br, E-mail: aacs@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2017-07-01

The Counter Current Flow Limitation (CCFL) phenomenon, specifically the control that the gas exerts in a liquid flow in the opposite direction, is of real importance in the study of design and operation of various industrial sectors, particularly the nuclear industry. In nuclear engineering, such a phenomenon can occur in a loss of coolant accident (LOCA) of a Pressurized Water Reactor (PWR) when there is the need to re-flood the reactor core during an emergency cooling process. The CCFL phenomenon is being investigated at the Nuclear Technology Development Center (CDTN) thermo-hydraulics laboratory in order to better understand the flow and its limitations and thereby contribute to the improvement of its modeling for analysis of severe accidents. For this, a series of experiments were performed in CDTN in a reduced scale acrylic test section of the 'hot leg' of a PWR. The new proposed circuit is a closed loop and no water has to be discharged during the experiment. This is only possible due to the Python program, which is associated to the data acquisition system and can interface with the automated valves through the outputs of the data acquisition board to control the experiment. The trials compare the CCFL behavior for 500mm lengths of the horizontal section, for inclined duct slope 50° for a diameter of 54mm pipe's diameter. This paper describes the new tests in comparison to tests performed in the past. (author)

Analysis of acoustic data from UK sodium/water reaction test facilities

International Nuclear Information System (INIS)

Rowley, R.; Mcknight, J.A.; Airey, J.

1990-01-01

This paper describes acoustic measurements made during a number of sodium/water reaction experiments in the UK. The tests have included water and steam injections through both realistic (fatigue crack) defects and machined orifices and have covered a range of experimental conditions including those appropriate to the inlet and outlet regions of the EFR steam generators. Injection rates were typically in the range 0.1 to 30 g/s. Where possible, gas injections were also included in the test programme for comparison, since it is anticipated that a practical SGU acoustic leak detection system would include a facility for gas injections to allow system calibration, and to confirm transmission properties within the SGU. The test sections were instrumented with accelerometers on waveguides and in some cases included an under-sodium microphone situated about 300mm above the reaction zone. Tape recordings were made during the tests and used for detailed analysis off-line, although an audible output from one of the acoustic channels was used to monitor the progress of the injections and provide information for the rig operators. A comparison of the signal amplitudes measured during the experiments with typical reactor background noise was made and an estimate of the detection sensitivity of an acoustic monitoring system was deduced. 3 refs, 5 figs, 1 tab

Cold leg condensation tests. Task C. Steam--water interaction tests

International Nuclear Information System (INIS)

Brodrick, J.R.; Loiselle, V.

1974-03-01

A report is presented of tests to determine the condensation efficiency of ECC water injected into a quality fluid mixture flowing through the cold leg. In particular, a specific objective was to determine if the mixture of ECC water and quality fluid reached thermodynamic equilibrium before exiting the cold leg. Further, the stability of the ECC water/quality fluid interaction would be assessed by interpretation of thermocouple records and utilization of a section of cold leg piping with view ports to film the interaction whenever possible. The cold leg condensation tests showed complete condensation of the 5 lbm/sec steam quality mixtures in the cold leg by the ECC water flows of the test matrix. The cold leg exit fluid temperature remained below the saturation temperature and had good agreement with the predicted cold leg outlet temperature, calculated assuming total condensation. (U.S.)

Water chemistry experience of nuclear power plants in Japan

International Nuclear Information System (INIS)

Ishigure, Kenkichi; Abe, Kenji; Nakajima, Nobuo; Nagao, Hiroyuki; Uchida, Shunsuke.

1989-01-01

Japanese LWRs have experienced several troubles caused by corrosions of structural materials in the past ca. 20 years of their operational history, among which are increase in the occupational radiation exposures, intergranular stress corrosion cracking (IGSCC) of stainless steel piping in BWR, and steam generator corrosion problems in PWR. These problems arised partly from the improper operation of water chemistry control of reactor coolant systems. Consequently, it has been realized that water chemistry control is one of the most important factors to attain high availability and reliability of LWR, and extensive researches and developments have been conducted in Japan to achieve the optimum water chemistry control, which include the basic laboratory experiments, analyses of plant operational data, loop tests in operating plants and computer code developments. As a result of the continuing efforts, the Japanese LWR plants have currently attained a very high performance in their operation with high availability and low occupational radiation exposures. A brief review is given here on the R and D of water chemistry in Japan. (author)

Supercritical Water Mixture (SCWM) Experiment in the High Temperature Insert-Reflight (HTI-R)

Science.gov (United States)

Hicks, Michael C.; Hegde, Uday G.; Garrabos, Yves; Lecoutre, Carole; Zappoli, Bernard

2013-01-01

Current research on supercritical water processes on board the International Space Station (ISS) focuses on salt precipitation and transport in a test cell designed for supercritical water. This study, known as the Supercritical Water Mixture Experiment (SCWM) serves as a precursor experiment for developing a better understanding of inorganic salt precipitation and transport during supercritical water oxidation (SCWO) processes for the eventual application of this technology for waste management and resource reclamation in microgravity conditions. During typical SCWO reactions any inorganic salts present in the reactant stream will precipitate and begin to coat reactor surfaces and control mechanisms (e.g., valves) often severely impacting the systems performance. The SCWM experiment employs a Sample Cell Unit (SCU) filled with an aqueous solution of Na2SO4 0.5-w at the critical density and uses a refurbished High Temperature Insert, which was used in an earlier ISS experiment designed to study pure water at near-critical conditions. The insert, designated as the HTI-Reflight (HTI-R) will be deployed in the DECLIC (Device for the Study of Critical Liquids and Crystallization) Facility on the International Space Station (ISS). Objectives of the study include measurement of the shift in critical temperature due to the presence of the inorganic salt, assessment of the predominant mode of precipitation (i.e., heterogeneously on SCU surfaces or homogeneously in the bulk fluid), determination of the salt morphology including size and shapes of particulate clusters, and the determination of the dominant mode of transport of salt particles in the presence of an imposed temperature gradient. Initial results from the ISS experiments will be presented and compared to findings from laboratory experiments on the ground.

In-situ tuff water migration/heater experiment: experimental plan

International Nuclear Information System (INIS)

Johnstone, J.K.

1980-08-01

Tuffs on the Nevada Test Site (NTS) are currently under investigation as a potential isolation medium for heat-producing nuclear wastes. The National Academy of Sciences has concurred in our identification of the potentially large water content (less than or equal to 40 vol %) of tuffs as one of the important issues affecting their suitability for a repository. This Experimental Plan describes an in-situ experiment intended as an initial assessment of water generation/migration in response to a thermal input. The experiment will be conducted in the Grouse Canyon Welded Tuff in Tunnel U12g (G-Tunnel) located in the north-central region of the NTS. While the Grouse Canyon Welded Tuff is not a potential repository medium, it has physical, thermal, and mechanical properties very similar to those tuffs currently under consideration and is accessible at depth (400 m below the surface) in an existing facility. Other goals of the experiment are to support computer-code and instrumentation development, and to measure in-situ thermal properties. The experimental array consists of a central electrical heater, 1.2 m long x 10.2 cm diameter, surrounded by three holes for measuring water-migration behavior, two holes for measuring temperature profiles, one hole for measuring thermally induced stress in the rock, and one hole perpendicular to the heater to measure displacement with a laser. This Experimental Plan describes the experimental objectives, the technical issues, the site, the experimental array, thermal and thermomechanical modeling results, the instrumentation, the data-acquisition system, posttest characterization, and the organizational details

Testing theory in practice: a simple experiment

NARCIS (Netherlands)

Terpstra, R.; Brezocnik, Z.; Kapus, T.; Ferreira Pires, Luis; Heerink, A.W.; Tretmans, G.J.

1996-01-01

In this paper we discuss the experiences gained in conducting a simple testing experiment. The goal of this experiment is to apply the abstract, formal testing framework [8] in a practical setting, and to indicate the critical aspects in its application to realistic testing situations. For that

In-Pile Experiment of a New Hafnium Aluminide Composite Material to Enable Fast Neutron Testing in the Advanced Test Reactor

Energy Technology Data Exchange (ETDEWEB)

Donna Post Guillen; Douglas L. Porter; James R. Parry; Heng Ban

2010-06-01

A new hafnium aluminide composite material is being developed as a key component in a Boosted Fast Flux Loop (BFFL) system designed to provide fast neutron flux test capability in the Advanced Test Reactor. An absorber block comprised of hafnium aluminide (Al3Hf) particles (~23% by volume) dispersed in an aluminum matrix can absorb thermal neutrons and transfer heat from the experiment to pressurized water cooling channels. However, the thermophysical properties, such as thermal conductivity, of this material and the effect of irradiation are not known. This paper describes the design of an in-pile experiment to obtain such data to enable design and optimization of the BFFL neutron filter.

Engineering aspects of the experiment and results of animal tests. [Apollo 17 Biological Cosmic Ray Experiment

Science.gov (United States)

Look, B. C.; Tremor, J. W.; Barrows, W. F.; Zabower, H. R.; Suri, K.; Park, E. G., Jr.; Durso, J. A.; Leon, H. A.; Haymaker, W.; Lindberg, R. G.

1975-01-01

A closed passive system independent of support from the spacecraft or its crew was developed to house five pocket mice for their flight on Apollo XVII. The reaction of potassium superoxide with carbon dioxide and water vapor to produce oxygen provided a habitable atmosphere within the experiment package. The performance of the system and the ability of the mice to survive the key preflight tests gave reasonable assurance that the mice would also withstand the Apollo flight.-

Prior storm experience moderates water surge perception and risk.

Directory of Open Access Journals (Sweden)

Gregory D Webster

Full Text Available BACKGROUND: How accurately do people perceive extreme water speeds and how does their perception affect perceived risk? Prior research has focused on the characteristics of moving water that can reduce human stability or balance. The current research presents the first experiment on people's perceptions of risk and moving water at different speeds and depths. METHODS: Using a randomized within-person 2 (water depth: 0.45, 0.90 m ×3 (water speed: 0.4, 0.8, 1.2 m/s experiment, we immersed 76 people in moving water and asked them to estimate water speed and the risk they felt. RESULTS: Multilevel modeling showed that people increasingly overestimated water speeds as actual water speeds increased or as water depth increased. Water speed perceptions mediated the direct positive relationship between actual water speeds and perceptions of risk; the faster the moving water, the greater the perceived risk. Participants' prior experience with rip currents and tropical cyclones moderated the strength of the actual-perceived water speed relationship; consequently, mediation was stronger for people who had experienced no rip currents or fewer storms. CONCLUSIONS: These findings provide a clearer understanding of water speed and risk perception, which may help communicate the risks associated with anticipated floods and tropical cyclones.

Radionuclide migration experiments related to an underground nuclear test: II. modeling studies

International Nuclear Information System (INIS)

Tompson, A.; Carle, S.F.; Smith, D.K.; Hudson, G.B.; Bruton, C.J.

2001-01-01

Full text: The goal of this project is to improve our understanding of water and radionuclide migration in both saturated and unsaturated geologic media by coupling advanced simulation techniques, available characterization data, and radioanalytical measurements in the context of a remarkable field experiment. Between 1975 and 1991, groundwater was steadily pumped from a well adjacent to a 1965 underground test conducted in alluvium at the Nevada Test Site. The experiment was primarily conducted in order to elicit information on radionuclide migration through the saturated zone between the test and the well. The effluent was monitored. discharged to an unlined ditch, and allowed to infiltrate into the ground during flow towards a dry lake, about a kilometer away. The 16 years of pumping and infiltration created an unexpected second experiment in which the migration of the ditch effluent through the 200 meters of unsaturated media, back to the water table, could be studied. Pumping and effluent data are being utilized in conjunction with chemical measurements made in groundwater and a series of numerical models to better understand the movement of radionuclides in the system, both between the test and the well, and between the ditch and the water table. The release of radionuclides away from a testing area will be controlled by local groundwater flow rates, by their dissolution from solidified melt glass produced by the test, and by chemical sorption processes that retard their migration rates in chemically reactive geologic media. Only the more mobile and less reactive radionuclides (e.g.. tritium, 14 C, 36 Cl, 85 Kr, and 129 I) were measured in the well effluent. The movement of these radionuclides through the unsaturated media beneath the ditch will be affected additionally by the capillary nature of moisture movement under unsaturated conditions and by their interaction with and potential mass exchange with the gas (air) phase. Results of numerical simulations

Preparations for deuterium tritium experiments on the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.; Barnes, G.

1994-04-01

The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR). These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a Fluorinet trademark system, modification of the vacuum system to handle tritium, preparation and testing of the neutral beam system for tritium operation and a final deuterium-deuterium (D-D) run to simulate expected deuterium-tritium (D-T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D-T experiments using D-D have been performed. The physics objectives of D-T operation are production of ∼ 10 megawatts (MW) of fusion power, evaluation of confinement and heating in deuterium-tritium plasmas, evaluation of α-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined α-particles. Experimental results and theoretical modeling in support of the D-T experiments are reviewed

Preparations for deuterium--tritium experiments on the Tokamak Fusion Test Reactor*

International Nuclear Information System (INIS)

Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Aschroft, D.; Barnes, C.W.; Barnes, G.; Batchelor, D.B.; Bateman, G.; Batha, S.; Baylor, L.A.; Beer, M.; Bell, M.G.; Biglow, T.S.; Bitter, M.; Blanchard, W.; Bonoli, P.; Bretz, N.L.; Brunkhorst, C.; Budny, R.; Burgess, T.; Bush, H.; Bush, C.E.; Camp, R.; Caorlin, M.; Carnevale, H.; Chang, Z.; Chen, L.; Cheng, C.Z.; Chrzanowski, J.; Collazo, I.; Collins, J.; Coward, G.; Cowley, S.; Cropper, M.; Darrow, D.S.; Daugert, R.; DeLooper, J.; Duong, H.; Dudek, L.; Durst, R.; Efthimion, P.C.; Ernst, D.; Faunce, J.; Fonck, R.J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G.Y.; Furth, H.P.; Garzotto, V.; Gentile, C.; Gettelfinger, G.; Gilbert, J.; Gioia, J.; Goldfinger, R.C.; Golian, T.; Gorelenkov, N.; Gouge, M.J.; Grek, B.; Grisham, L.R.; Hammett, G.; Hanson, G.R.; Heidbrink, W.; Hermann, H.W.; Hill, K.W.; Hirshman, S.; Hoffman, D.J.; Hosea, J.; Hulse, R.A.; Hsuan, H.; Jaeger, E.F.; Janos, A.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Kamperschroer, J.; Kesner, J.; Kugel, H.; Kwon, S.; Labik, G.; Lam, N.T.; LaMarche, P.H.; Laughlin, M.J.; Lawson, E.; LeBlanc, B.; Leonard, M.; Levine, J.; Levinton, F.M.; Loesser, D.; Long, D.; Machuzak, J.; Mansfield, D.E.; Marchlik, M.; Marmar, E.S.; Marsala, R.; Martin, A.; Martin, G.; Mastrocola, V.; Mazzucato, E.; McCarthy, M.P.; Majeski, R.; Mauel, M.; McCormack, B.; McCune, D.C.; McGuire, K.M.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Milora, S.L.; Monticello, D.; Mueller, D.; Murakami, M.; Murphy, J.A.; Nagy, A.; Navratil, G.A.; Nazikian, R.; Newman, R.; Nishitani, T.; Norris, M.; O'Connor, T.; Oldaker, M.; Ongena, J.; Osakabe, M.; Owens, D.K.; Park, H.; Park, W.; Paul, S.F.; Pavlov, Y.I.; Pearson, G.; Perkins, F.; Perry, E.; Persing, R.; Petrov, M.; Phillips, C.K.; Pitcher, S.; Popovichev, S.; Qualls, A.L.; Raftopoulos, S.; Ramakrishnan, R.; Ramsey, A.; Rasmussen, D.A.; Redi, M.H.

1994-01-01

The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a Fluorinert TM system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium--deuterium (D--D) run to simulate expected deuterium--tritium (D--T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D--T experiments using D--D have been performed. The physics objectives of D--T operation are production of ∼10 MW of fusion power, evaluation of confinement, and heating in deuterium--tritium plasmas, evaluation of α-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined α particles. Experimental results and theoretical modeling in support of the D--T experiments are reviewed

A review of experiments and results from the transient reactor test (TREAT) facility

International Nuclear Information System (INIS)

Deitrich, L. W.

1998-01-01

The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

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)

Barrier erosion control test plan: Gravel mulch, vegetation, and soil water interactions

Energy Technology Data Exchange (ETDEWEB)

Waugh, W.J.; Link, S.O. (Pacific Northwest Lab., Richland, WA (USA))

1988-07-01

Soil erosion could reduce the water storage capacity of barriers that have been proposed for the disposal of near-surface waste at the US Department of Energy's Hanford Site. Gravel mixed into the top soil surface may create a self-healing veneer that greatly retards soil loss. However, gravel admixtures may also enhance infiltration of rainwater, suppress plant growth and water extraction, and lead to the leaching of underlying waste. This report describes plans for two experiments that were designed to test hypotheses concerning the interactive effects of surface gravel admixtures, revegetation, and enhanced precipitation on soil water balance and plant abundance. The first experiment is a factorial field plot set up on the site selected as a soil borrow area for the eventual construction of barriers. The treatments, arranged in a a split-split-plot design structure, include two densities of gravel admix, a mixture of native and introduced grasses, and irrigation to simulate a wetter climate. Changes in soil water storage and plant cover are monitored with neutron moisture probes and point intercept sampling, respectively. The second experiment consists of an array of 80 lysimeters containing several different barrier prototypes. Surface treatments are similar to the field-plot experiment. Drainage is collected from a valve at the base of each lysimeter tube, and evapotranspiration is estimated by subtraction. The lysimeters are also designed to be coupled to a whole-plant gas exchange system that will be used to conduct controlled experiments on evapotranspiration for modeling purposes. 56 refs., 6 figs., 8 tabs.

Life-cycle testing of receiving waters with Ceriodaphnia dubia

International Nuclear Information System (INIS)

Stewart, A.J.; Beane, B.K.

1995-01-01

Seven-day tests with Ceriodaphnia are commonly used to estimate the toxicity of effluents or receiving waters, but may yield no toxicity outcomes even when pollutants are present (a possible type II error). The authors conducted two sets of full life-cycle tests with C. dubia to (1) see if tests with longer exposure periods revealed evidence for toxicity that might not be evident from shorter tests, and (2) determine the relative importance of water quality versus food as factors influencing C. dubia reproduction. In the first set of tests, daphnids were reared in diluted mineral water (control), water from a stream impacted by coal fly-ash, or water from a mercury-contaminated retention basin. The second set of tests used water from the retention basin only, but this water was either filtered or not filtered, and food was either added or not added. C. dubia survival and reproduction did not differ much among the three waters in the first set of tests. However, both parameters were strongly affected by the filtering and food-addition treatments in the second set of tests. Thus, C. dubia seems to be moderately insensitive to general water-quality factors, but quite sensitive to food-related parameters. Regression analysis showed that the predictability of life-time reproduction of C. dubia from 7-day test results was low in five of six cases. The increase in predictability as a function of test duration also differed among water types (first set of tests), and among treatments (second set of tests). Thus, 7-day tests with C. dubia may be used to quantify water-quality problems, but it may not be possible to reliably extrapolate the results of such tests to longer time scales

Verification of the code ATHLET by post-test analysis of two experiments performed at the CCTF integral test facility

International Nuclear Information System (INIS)

Krepper, E.; Schaefer, F.

2001-03-01

In the framework of the external validation of the thermohydraulic code ATHLET Mod 1.2 Cycle C, which has been developed by the GRS, post test analyses of two experiments were done, which were performed at the japanese test facility CCTF. The test facility CCTF is a 1:25 volume-scaled model of a 1000 MW pressurized water reactor. The tests simulate a double end break in the cold leg of the PWR with ECC injection into the cold leg and with combined ECC injection into the hot and cold legs. The evaluation of the calculated results shows, that the main phenomena can be calculated in a good agreement with the experiment. Especially the behaviour of the quench front and the core cooling are calculated very well. Applying a two-channel representation of the reactor model the radial behaviour of the quench front could be reproduced. Deviations between calculations and experiment can be observed simulating the emergency injection in the beginning of the transient. Very high condensation rates were calculated and the pressure decrease in this phase of the transient is overestimated. Besides that, the pressurization due to evaporation in the refill phase is underestimated by ATHLET. (orig.) [de

Preoperational test report, raw water system

Energy Technology Data Exchange (ETDEWEB)

Clifton, F.T.

1997-10-29

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

Preoperational test report, raw water system

International Nuclear Information System (INIS)

Clifton, F.T.

1997-01-01

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

[Study on tests of genetics experiments in universities].

Science.gov (United States)

Jie, He; Hao, Zhang; Lili, Zhang

2015-03-01

Based on the present situation and the development of experiment tests in universities, we introduced a reform in tests of genetics experiments. According to the teaching goals and course contents of genetics experiment, the tests of genetics experiments contain four aspects on the performance of students: the adherence to the experimental procedures, the depth of participation in experiment, the quality of experiment report, and the mastery of experiment principles and skills, which account for 10 %, 20 %, 40 % and 30 % in the total scores, respectively. All four aspects were graded quantitatively. This evaluation system has been tested in our experiment teaching. The results suggest that it has an effect on the promotion of teaching in genetics experiments.

A fast alternative to core plug tests for optimising injection water salinity for EOR

DEFF Research Database (Denmark)

Hassenkam, Tue; Andersson, Martin Peter; Hilner, Emelie Kristin Margareta

2014-01-01

of the clays which would lead to permanent reservoir damage but evidence of effectiveness at moderate salinity would offer the opportunity to dispose of produced water. The goal is to define boundary conditions so injection water salinity is high enough to prevent reservoir damage and low enough to induce...... the low salinity effect while keeping costs and operational requirements at a minimum. Traditional core plug testing for optimising conditions has some limitations. Each test requires a fresh sample, core testing requires sophisticated and expensive equipment, and reliable core test data requires several...... experiments can be done relatively quickly on very little material, it gives the possibility of testing salinity response on samples from throughout a reservoir and for gathering statistics. Our approach provides a range of data that can be used to screen core plug testing conditions and to provide extra data...

Operational experience, evolution and developments in water chemistry in Indian Nuclear Power Plants - an overview

International Nuclear Information System (INIS)

Prasad, Y.S.R.

2000-01-01

Lessons learnt from the experiences at nuclear power plants have enriched the understanding of corrosion behaviour in water systems. The need for proper water chemistry control not only during operation but also during fabrication and preoperational tests is clearly seen. It should not be construed that maintenance of proper water chemistry is a panacea for all corrosion and other associated problems. Unless adequate care is taken in selection of material and sound design and fabrication practices are followed, no regime of water chemistry can help in eliminating failure due to corrosion

Alternative Water Processor Test Development

Science.gov (United States)

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

2012-01-01

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

Aspects on testing methods for acid attacks on concrete - further experiments

International Nuclear Information System (INIS)

Romben, L.

1980-01-01

The report presents a number of control experiments which have been performed to verify test procedures of the resistance of concrete and concrete products. A more detailed study of the calcium dissolution and the layer dissolution process has been performed. The mechanisms which control the rate of attack in the long term are discussed. The proposed method is intended for use in studying the resistance to attacks by acid water solutions. (G.B.)

Online monitoring of steam/water chemistry of a fast breeder test reactor

International Nuclear Information System (INIS)

Subramanian, K.G.; Suriyanarayanan, A.; Thirunavukarasu, N.; Naganathan, V.R.; Panigrahi, B.S.; Jambunathan, D.

2005-01-01

Operating experience with the once-through steam generator of a fast breeder test reactor (FBTR) has shown that an efficient water chemistry control played a major role in minimizing corrosion related failures of steam generator tubes and ensuring steam generator tube integrity. In order to meet the stringent feedwater and steam quality specifications, use of fast and sensitive online monitors to detect impurity levels is highly desirable. Online monitoring techniques have helped in achieving feedwater of an exceptional degree of purity. Experience in operating the online monitors in the steam/water system of a FBTR is discussed in detail in this paper. In addition, the effect of excess hydrazine in the feedwater on the steam generator leak detection system and the need for a hydrazine online meter are also discussed. (orig.)

[Eosin Y-water test for sperm function examination].

Science.gov (United States)

Zha, Shu-wei; Lü, Nian-qing; Xu, Hao-qin

2015-06-01

Based on the principles of the in vitro staining technique, hypotonic swelling test, and water test, the Eosin Y-water test method was developed to simultaneously detect the integrity of the sperm head and tail and sperm membrane structure and function. As a widely used method in clinical laboratories in China, the Eosin Y-water test is methodologically characterized by three advantages. Firstly, both the sperm head and tail can be detected at the same time, which allows easy and comprehensive assessment of membrane damage in different parts of sperm. Secondly, distilled water is used instead of the usual formula solution to simplify and standardize the test by eliminating any potential effects on the water molecules through the sperm membrane due to different osmotic pressure or different sugar proportions and electrolyte solutions. Thirdly, the test takes less time and thus can be repeated before and after treatment. This article focuses on the fundamental principles and modification of the Eosin Y-water test and its application in sperm function examination and routine semen analysis for male infertility, assessment of the quality of sperm retrieved by testicular fine needle aspiration, semen cryopreservation program development, and evaluation of sperm membrane integrity after microwave radiation.

Test Management Framework for the ATLAS Experiment

CERN Document Server

Kazarov, Andrei; The ATLAS collaboration; Avolio, Giuseppe

2018-01-01

Test Management Framework for the Data Acquisition of the ATLAS Experiment Data Acquisition (DAQ) of the ATLAS experiment is a large distributed and inhomogeneous system: it consists of thousands of interconnected computers and electronics devices that operate coherently to read out and select relevant physics data. Advanced diagnostics capabilities of the TDAQ control system are a crucial feature which contributes significantly to smooth operation and fast recovery in case of the problems and, finally, to the high efficiency of the whole experiment. The base layer of the verification and diagnostic functionality is a test management framework. We have developed a flexible test management system that allows the experts to define and configure tests for different components, indicate follow-up actions to test failures and describe inter-dependencies between DAQ or detector elements. This development is based on the experience gained with the previous test system that was used during the first three years of th...

Water Hammer Test

Science.gov (United States)

2008-01-01

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

[Cultivation of a permanent fish cell line in serum-free media special experiences with a cytotoxicity test for waste water samples

Science.gov (United States)

Kohlpoth, Martin; Rusche, Brigitte

1997-01-01

The use of fetal calf serum (FCS) as standard medium additive for the cell cultivation must be regarded critically from the point of view of animal welfare as well as for scientific reasons and makes it necessary to look for alternatives. In the last years an in vitro cytotoxicity assay for the testing of industrial waste waters with the permanent fish cell line RTG-2 was established and pre-validated as an alternative to the fish test with the golden orfe. The application of FCS is also a special problem with regard to the testing of waste waters in a cytotoxicity test so that FCS-alternatives were tested. The RTG-2 cells were successfully adapted to the two solvents Basal Medium Supplement (BMS) and Ultroser-G (U-G) that are used to replace serum. The characterisation of these adapted cell lines showed no significant differences in growth rate, adhesion rate, viability and sensitivity to chemicals in comparison to the original RTG-2 cells. On the determination of the cytotoxicity of industrial waste waters the RTG-2 cells adapted to the BMS medium indicated a clearly higher toxicity of the waste water samples than the original RTG-2 cells. This result confirms the thesis that serum components react with waste water elements and thus change the bio-availability of toxic compounds.

Test of prototype liquid-water-content meter for aircraft use

Science.gov (United States)

Gerber, Hermann E.

1993-01-01

This report describes the effort undertaken to meet the objectives of National Science Foundation Grant ATM-9207345 titled 'Test of Prototype Liquid-Water-Content Meter for Aircraft Use.' Three activities were proposed for testing the new aircraft instrument, PVM-100A: (1) Calibrate the PVM-100A in a facility where the liquid-water-content (LWC) channel, and the integrated surface area channel (PSA) could be compared to standard means for LWC and PSA measurements. Scaling constant for the channels were to be determined in this facility. The fog/wind tunnel at ECN, Petten, The Netherlands was judged the most suitable facility for this effort. (2) Expose the PVM-100A to high wind speeds similar to those expected on research aircraft, and test the anti-icing heaters on the PVM-100A under typical icing conditions expected in atmospheric clouds. The high-speed icing tunnel at NRC, Ottawa, Canada was to be utilized. (3) Operate the PVM-100A on an aircraft during cloud penetrations to determine its stability and practicality for such measurements. The C-131A aircraft of the University of Washington was the aircraft of opportunity for these-tests, which were to be conducted during the 4-week Atlantic Stratocumulus Transition Experiment (ASTEX) in June of 1992.

Pre- and post- test calculation for the parameter-SF1 experiment with ATHLET-CD

Energy Technology Data Exchange (ETDEWEB)

Erdmann, W.; Trambauer, K.; Stuckert, J. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koln (Germany)

2006-07-01

The main objective of the PARAMETER-SF1 experiment in the frame of the ISTC project 3194 is the experimental and analytical investigation of the Russian VVER-1000 fuel rod assemblies behavior under simulated conditions of a severe accident. The special feature is to study the effect of flooding a superheated test bundle from the top (top quenching) which has not yet been investigated at all. - Simulation of the PARAMETER test facility To calculate the special effects of the top quenching, some aspects are important: detailed simulation of the bundle top, top and bottom quench front, heat losses at top/bottom of bundle, electrical heater power. - Main initial and boundary conditions The proposed initial and boundary conditions for the double-blind pre-test calculation were quite different from the actual experimental data during the test e.g.: electric power, mass flow (water, steam, argon), temperature. - Conclusions: first experiment with top flooding proposed initial condition given in the specification could not be performed during the experiment, bundle parameters deviated from anticipated values, thus, the pre-calculations not comparable with the experiment, post-calculations with ATHLET-CD showed good agreement with experiment data, top flooding is well predicted, calculational results sensitive with respect to: boundary conditions, nodalization. (authors)

[Biological testing of water with different structural states in rats and frogs].

Science.gov (United States)

Farashchuk, N F; Mikhaylova, R I; Telenkova, O G

2014-01-01

The effect of water samples with different structural states on some physiological indices of white laboratory rats, 5 groups of 10 animals (5 females and 5 males) and frogs has been studied. The investigation was performed for 1 month. For the determination of the content of liquid crystal associates (LCA) in water samples there was used the dilatometric method, the performance of experimental animals was studied by the swimming test (up to total fatigue). The performed experiment on growing rats with the use of water with varying degrees of structuredness showed that according to the weight gain there were optimal water "Lekor" and tap water, treated with Bioptron (the content of the structured fraction is 5.06 +/- 0.09% and 6.9 +/- 0.23%, respectively). On physical performance the best indices were in animals consumed water treated with the Bioptron lamp. In performance of experiments on frogs it was found that cardiac function in animals under experimental conditions over time weakens spontaneously: heart rate and cardiac output decline. Therefore, the effect of different water samples on the cardiac function was assessed on the intensity of its decrease for 15 minutes. In the experimental study of the effect of water with different content of LCA on heart rate and cardiac output of the frog it was found that the optimum level of structuredness of water is within the range of 5.06 +/- 0.09% (in water "Lekor") - 6.9 +/- 0.23% (tap water treated with Bioptron). All the other water samples, the content of nanocrystals in which was below or above this range, has a pronounced inhibitory effect on the heart performance of the frog.

The testing effect for mediator final test cues and related final test cues in online and laboratory experiments.

Science.gov (United States)

Coppens, Leonora C; Verkoeijen, Peter P J L; Bouwmeester, Samantha; Rikers, Remy M J P

2016-05-31

The testing effect is the finding that information that is retrieved during learning is more often correctly retrieved on a final test than information that is restudied. According to the semantic mediator hypothesis the testing effect arises because retrieval practice of cue-target pairs (mother-child) activates semantically related mediators (father) more than restudying. Hence, the mediator-target (father-child) association should be stronger for retrieved than restudied pairs. Indeed, Carpenter (2011) found a larger testing effect when participants received mediators (father) than when they received target-related words (birth) as final test cues. The present study started as an attempt to test an alternative account of Carpenter's results. However, it turned into a series of conceptual (Experiment 1) and direct (Experiment 2 and 3) replications conducted with online samples. The results of these online replications were compared with those of similar existing laboratory experiments through small-scale meta-analyses. The results showed that (1) the magnitude of the raw mediator testing effect advantage is comparable for online and laboratory experiments, (2) in both online and laboratory experiments the magnitude of the raw mediator testing effect advantage is smaller than in Carpenter's original experiment, and (3) the testing effect for related cues varies considerably between online experiments. The variability in the testing effect for related cues in online experiments could point toward moderators of the related cue short-term testing effect. The raw mediator testing effect advantage is smaller than in Carpenter's original experiment.

Cytogenotoxicity screening of source water, wastewater and treated water of drinking water treatment plants using two in vivo test systems: Allium cepa root based and Nile tilapia erythrocyte based tests.

Science.gov (United States)

Hemachandra, Chamini K; Pathiratne, Asoka

2017-01-01

Biological effect directed in vivo tests with model organisms are useful in assessing potential health risks associated with chemical contaminations in surface waters. This study examined the applicability of two in vivo test systems viz. plant, Allium cepa root based tests and fish, Oreochromis niloticus erythrocyte based tests for screening cytogenotoxic potential of raw source water, water treatment waste (effluents) and treated water of drinking water treatment plants (DWTPs) using two DWTPs associated with a major river in Sri Lanka. Measured physico-chemical parameters of the raw water, effluents and treated water samples complied with the respective Sri Lankan standards. In the in vivo tests, raw water induced statistically significant root growth retardation, mitodepression and chromosomal abnormalities in the root meristem of the plant and micronuclei/nuclear buds evolution and genetic damage (as reflected by comet scores) in the erythrocytes of the fish compared to the aged tap water controls signifying greater genotoxicity of the source water especially in the dry period. The effluents provoked relatively high cytogenotoxic effects on both test systems but the toxicity in most cases was considerably reduced to the raw water level with the effluent dilution (1:8). In vivo tests indicated reduction of cytogenotoxic potential in the tested drinking water samples. The results support the potential applications of practically feasible in vivo biological test systems such as A. cepa root based tests and the fish erythrocyte based tests as complementary tools for screening cytogenotoxicity potential of the source water and water treatment waste reaching downstream of aquatic ecosystems and for evaluating cytogenotoxicity eliminating efficacy of the DWTPs in different seasons in view of human and ecological safety. Copyright © 2016 Elsevier Ltd. All rights reserved.

Experiments in a floating water bridge

Science.gov (United States)

Woisetschläger, Jakob; Gatterer, Karl; Fuchs, Elmar C.

2010-01-01

In a high-voltage direct-current experiment, a watery connection formed between two beakers filled with deionized water, giving the impression of a `floating water bridge'. Having a few millimeters diameter and up to 2.5 cm length, this watery connection reveals a number of interesting phenomena currently discussed in water science. Focusing on optical measurement techniques, the flow through the bridge was visualized and data were recorded such as flow velocity and directions, heat production, density fluctuations, pH values, drag force and mass transfer. To provide a better understanding of the basic phenomena involved the discussion references related literature.

Linear Aerospike SR-71 Experiment (LASRE) dumps water after first in-flight cold flow test

Science.gov (United States)

1998-01-01

The NASA SR-71A successfully completed its first cold flow flight as part of the NASA/Rocketdyne/Lockheed Martin Linear Aerospike SR-71 Experiment (LASRE) at NASA's Dryden Flight Research Center, Edwards, California on March 4, 1998. During a cold flow flight, gaseous helium and liquid nitrogen are cycled through the linear aerospike engine to check the engine's plumbing system for leaks and to check the engine operating characterisitics. Cold-flow tests must be accomplished successfully before firing the rocket engine experiment in flight. The SR-71 took off at 10:16 a.m. PST. The aircraft flew for one hour and fifty-seven minutes, reaching a maximum speed of Mach 1.58 before landing at Edwards at 12:13 p.m. PST. 'I think all in all we had a good mission today,' Dryden LASRE Project Manager Dave Lux said. Flight crew member Bob Meyer agreed, saying the crew 'thought it was a really good flight.' Dryden Research Pilot Ed Schneider piloted the SR-71 during the mission. Lockheed Martin LASRE Project Manager Carl Meade added, 'We are extremely pleased with today's results. This will help pave the way for the first in-flight engine data-collection flight of the LASRE.' The LASRE experiment was designed to provide in-flight data to help Lockheed Martin evaluate the aerodynamic characteristics and the handling of the SR-71 linear aerospike experiment configuration. The goal of the project was to provide in-flight data to help Lockheed Martin validate the computational predictive tools it was using to determine the aerodynamic performance of a future reusable launch vehicle. The joint NASA, Rocketdyne (now part of Boeing), and Lockheed Martin Linear Aerospike SR-71 Experiment (LASRE) completed seven initial research flights at Dryden Flight Research Center. Two initial flights were used to determine the aerodynamic characteristics of the LASRE apparatus (pod) on the back of the SR-71. Five later flights focused on the experiment itself. Two were used to cycle gaseous

Predicting the occurrence of mixed mode failure associated with hydraulic fracturing, part 2 water saturated tests

Energy Technology Data Exchange (ETDEWEB)

Bauer, Stephen J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Broome, Scott Thomas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Choens, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Barrow, Perry Carl [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-09-14

Seven water-saturated triaxial extension experiments were conducted on four sedimentary rocks. This experimental condition was hypothesized more representative of that existing for downhole hydrofracture and thus it may improve our understanding of the phenomena. In all tests the pore pressure was 10 MPa and confirming pressure was adjusted to achieve tensile and transitional failure mode conditions. Using previous work in this LDRD for comparison, the law of effective stress is demonstrated in extension using this sample geometry. In three of the four lithologies, no apparent chemo-mechanical effect of water is apparent, and in the fourth lithology test results indicate some chemo-mechanical effect of water.

Wind and water tunnel testing of a morphing aquatic micro air vehicle.

Science.gov (United States)

Siddall, Robert; Ortega Ancel, Alejandro; Kovač, Mirko

2017-02-06

Aerial robots capable of locomotion in both air and water would enable novel mission profiles in complex environments, such as water sampling after floods or underwater structural inspections. The design of such a vehicle is challenging because it implies significant propulsive and structural design trade-offs for operation in both fluids. In this paper, we present a unique Aquatic Micro Air Vehicle (AquaMAV), which uses a reconfigurable wing to dive into the water from flight, inspired by the plunge diving strategy of water diving birds in the family Sulidae . The vehicle's performance is investigated in wind and water tunnel experiments, from which we develop a planar trajectory model. This model is used to predict the dive behaviour of the AquaMAV, and investigate the efficacy of passive dives initiated by wing folding as a means of water entry. The paper also includes first field tests of the AquaMAV prototype where the folding wings are used to initiate a plunge dive.

11 CFR 100.131 - Testing the waters.

Science.gov (United States)

2010-01-01

... 11 Federal Elections 1 2010-01-01 2010-01-01 false Testing the waters. 100.131 Section 100.131 Federal Elections FEDERAL ELECTION COMMISSION GENERAL SCOPE AND DEFINITIONS (2 U.S.C. 431) Exceptions to Expenditures § 100.131 Testing the waters. (a) General exemption. Payments made solely for the purpose of...

11 CFR 100.72 - Testing the waters.

Science.gov (United States)

2010-01-01

... 11 Federal Elections 1 2010-01-01 2010-01-01 false Testing the waters. 100.72 Section 100.72 Federal Elections FEDERAL ELECTION COMMISSION GENERAL SCOPE AND DEFINITIONS (2 U.S.C. 431) Exceptions to Contributions § 100.72 Testing the waters. (a) General exemption. Funds received solely for the purpose of...

Mark I 1/5-scale boiling water reactor pressure suppresion experiment quick-look report

International Nuclear Information System (INIS)

Lai, W.; Collins, E.K.

1977-01-01

This report is intended as a ''quick-look'' report summarizing the experimental results obtained from pressure suppression experiment numbers 2.1, 2.2, and 2.3 that were performed on the Lawrence Livermore Laboratory's 1/5-scale boiling water reactor (BWR) Mark I pressure suppression experimental facility on April 26, 1977. A brief description of the general nature of the tests and a summary of the actual tests that were performed are given

Post-test analysis of ROSA-III experiment RUNs 705 and 706

International Nuclear Information System (INIS)

Koizumi, Yasuo; Soda, Kunihisa; Kikuchi, Osamu; Tasaka, Kanji; Shiba, Masayoshi

1980-07-01

The purpose of ROSA-III experiment with a scaled BWR Test facility is to examine primary coolant thermal-hydraulic behavior and performance of ECCS during a postulated loss-of-coolant accident of BWR. The results provide the information for verification and improvement of reactor safety analysis codes. RUNs 705 and 706 assumed a 200% double-ended break at the recirculation pump suction. RUN 705 was an isothermal blowdown test without initial power and initial core flow. In RUN 706 for an average core power and no ECCS, the main steam line and feed water line were isolated immediately on the break. Post-test analysis of RUNs 705 and 706 was made with computer code RELAP4J. The agreement in system pressure between calculation and experiment was satisfactory. However, the calculated heater rod surface temperature were significantly higher than the experimental ones. The calculated axial temperature profile was different in tendency from the experimental one. The calculated mixture level behavior in the core was different from the liquid void distribution observed in experiment. The rapid rise of fuel rod surface temperature was caused by the reduction of heat transfer coefficient attributed to the increase of quality. The need was indicated for improvement of analytical model of void distribution in the core, and also to performe a characteristic test of recirculation line under reverse flow and to examine the core inlet flow rate experimentally and analytically. (author)

Operating experiences on ammonia water exchange system at Heavy Water Plant, Talcher (paper No. 6.12)

International Nuclear Information System (INIS)

Venkat Ram, D.; Sharma, A.K.

1992-01-01

The Heavy Water Plant at Talcher employs bithermal ammonia hydrogen exchange process for the production of heavy water. The paper describes about the existing ammonia water exchange column, its start-up, operating experience and the problems encountered in operation of the column. The operating experiences gained and the data collected over the last few years can be utilised for design and operation of new ammonia water exchange column. (V.R). 2 figs

PROPOSAL FOR BUILD A TEST STAND TO THE STUDY WATER RAM BASED ON THE RECOMMENDATIONS CONSTRUCTION

Directory of Open Access Journals (Sweden)

Dariusz GRYGO

2016-05-01

Full Text Available The main aim of this article was developed of a test stand to a study physics phenomena accruing at work in a ram water based on the construction recommendations. The experience gained by a number of attempts allowed to the developed construction recommendations that allowed to adjust the test stand to perform various studies of physics phenomena occurring at the work of water ram and easy adaptions to the new requirements. The article includes: the review of the test stands on which performed the study in the history and in the present times, description of the construction recommendations of the universal test stand and the proposal of the test stand developed based on the construction recommendations. Motivation to take up this issue follows from the interest of devices that work in an unconventional way, devices used a renewable energy that can be used in household e.g. to improve their energy balance, delivery water to houses, drinking water for animals, fields irrigation, etc. Analysis of the literature of the water ram, test stands and their performance shown that it is small and typically old development in popular science form. In times of intensive search new sources of renewable energy reactivation of this type equipment may be highly probable.

In-pile cladding tests at NRI Rez and PIE capabilities and experience

International Nuclear Information System (INIS)

Zmitko, M.

2002-01-01

In-pile cladding corrosion test facilities and relevant post-irradiation capabilities at NRI Rez plc are overviewed. Basic information about the research rector LVR-15 and in-pile water loops is given. An experience in the field of Zr-alloy cladding corrosion testing and investigation of cladding corrosion behaviour is demonstrated for two experimental programmes conducted at NRI Rez in the past period. The first example describes results obtained at studying of corrosion behaviour of advanced Zr-alloys under PWR conditions with a special concern to a high lithium content and subcooled surface boiling. The second example informs about completion of the experimental programme supported by the IAEA which is focused on investigation of Zircaloy-4 cladding behaviour under VVER water chemistry, thermal-hydraulic and irradiation conditions with the main to obtain experimental data for an assessment of the Zircaloy-4 cladding compatibility with VVER conditions. (author)

Analyses for experiment on sodium-water reaction temperature by the CHAMPAGNE code

International Nuclear Information System (INIS)

Yoshioka, Naoki; Kishida, Masako; Yamada, Yumi

2000-03-01

In this work, analyses on sodium-water reaction temperature in the new SWAT-1(SWAT-1R) test were completed by the CHAMPAGNE code in order to understand void and velocity distribution in sodium system, which was difficult to be measured in experiments. The application method of the RELAP5/Mod2 code was investigated to LMFBR steam generator (SG) blow down analysis, too. The following results were obtained. (1) Analyses on sodium-water reaction temperature in the SWAT-1R test. 1) Analyses were carried out for the SWAT-1R test under the condition water leak rate 600 g/s by treating the pressure loss coefficient, the interface friction coefficient and the coefficient related to reaction rate as parameters. The effect and mechanism of each parameter on the shape of reaction zone were well understood by these analyses. 2) The void and velocity distribution in sodium system were estimated by use of the most suitable parameters. These analytical results are expected to be useful for planning of the SWAT-1R test and evaluation of test result. (2) Investigation of the RELAP5/Mod2 code. 1) The items to be improved in the RELAP5/Mod2 code were clarified to apply this code to the FBR SG blow down analysis. 2) One of these items was an addition of the shell-side (sodium-side) model. A sodium-side model was designed and added to the RELAP5/Mod2 code. Test calculations were carried out by this improved code and the basic function of this code was confirmed. (author)

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.;

Tritium-gas/water-vapor monitor. Tests and evaluation

International Nuclear Information System (INIS)

Jalbert, R.A.

1982-07-01

A tritium gas/water-vapor monitor was designed and built by the Health Physics Group at the Los Alamos National Laboratory. In its prototype configuration, the monitor took the shape of two separate instruments: a (total) tritium monitor and a water-vapor monitor. Both instruments were tested and evaluated. The tests of the (total) tritium monitor, basically an improved version of the standard flow-through ion-chamber instrument, are briefly reported here and more completely elsewhere. The tests of the water-vapor monitor indicated that the novel approach used to condense water vapor for scintillation counting has a number of serious drawbacks and that further development of the instrument is unwarranted

Orion Ground Test Article Water Impact Tests: Photogrammetric Evaluation of Impact Conditions

Science.gov (United States)

Vassilakos, Gregory J.; Mark, Stephen D.

2018-01-01

The Ground Test Article (GTA) is an early production version of the Orion Crew Module (CM). The structural design of the Orion CM is being developed based on LS-DYNA water landing simulations. As part of the process of confirming the accuracy of LS-DYNA water landing simulations, the GTA water impact test series was conducted at NASA Langley Research Center (LaRC) to gather data for comparison with simulations. The simulation of the GTA water impact tests requires the accurate determination of the impact conditions. To accomplish this, the GTA was outfitted with an array of photogrammetry targets. The photogrammetry system utilizes images from two cameras with a specialized tracking software to determine time histories for the 3-D coordinates of each target. The impact conditions can then be determined from the target location data.

Power-Cooling-Mismatch Test Series Test PCM-7. Experiment operating specifications

International Nuclear Information System (INIS)

Sparks, D.T.; Smith, R.H.; Stanley, C.J.

1979-02-01

The experiment operating specifications for the Power-Cooling-Mismatch (PCM) Test PCM-7 to be conducted in the Power Burst Facility are described. The PCM Test Series was designed on the basis of a parametric evaluation of fuel behavior response with cladding temperature, rod internal pressure, time in film boiling, and test rod power being the variable parameters. The test matrix, defined in the PCM Experiment Requirements Document (ERD), encompasses a wide range of situations extending from pre-CHF (critical heat flux) PCMs to long duration operation in stable film boiling leading to rod failure

Experimental system description for air-water CCFL tests of the 161-rod FLECHT-SEASET test vessel upper plenum

International Nuclear Information System (INIS)

Fogdall, S.P.; Anderson, J.L.

1983-01-01

A series of countercurrent flow limiting (CCFL) experiments has been performed by EG and G Idaho, Inc. in the Steam-Air-Water (SAW) test facility at the Idaho National Engineering Laboratory on behalf of the US Nuclear Regulatory Commission (NRC). Tests were performed in a mockup of the vessel for the 161-Rod Systems Effects Test (SET) facility of the FLECHT-SEASET program, conducted by the Westinghouse Electric Corporation. Westinghouse and the NRC will use the test results to provide a CCFL correlation to predict the flooding behavior in the upper plenum of the SET vessel. This paper presents a description of the experimental system and the test conduct, including data validation and uncertainty analysis. The test objectives centered on experimentally obtaining coefficients in the Wallis correlation for flooding with the specific vessel geometry. The test conditions and vessel configuration are described and the design of the test loop, instrumentation, and data acquisition are discussed. The establishment of a test point and the resultant data are described

Effects of elevated CO2 on predator avoidance behaviour by reef fishes is not altered by experimental test water

Directory of Open Access Journals (Sweden)

Philip L. Munday

2016-10-01

Full Text Available Pioneering studies into the effects of elevated CO2 on the behaviour of reef fishes often tested high-CO2 reared fish using control water in the test arena. While subsequent studies using rearing treatment water (control or high CO2 in the test arena have confirmed the effects of high CO2 on a range of reef fish behaviours, a further investigation into the use of different test water in the experimental arena is warranted. Here, we used a fully factorial design to test the effect of rearing treatment water (control or high CO2 and experimental test water (control or high CO2 on antipredator responses of larval reef fishes. We tested antipredator behaviour in larval clownfish Amphiprion percula and ambon damselfish Pomacentrus amboinensis, two species that have been used in previous high CO2 experiments. Specifically, we tested if: (1 using control or high CO2 water in a two channel flume influenced the response of larval clownfish to predator odour; and (2 using control or high CO2 water in the test arena influenced the escape response of larval damselfish to a startle stimulus. Finally, (3 because the effects of high CO2 on fish behaviour appear to be caused by altered function of the GABA-A neurotransmitter we tested if antipredator behaviours were restored in clownfish treated with a GABA antagonist (gabazine in high CO2 water. Larval clownfish reared from hatching in control water (496 µatm strongly avoided predator cue whereas larval clownfish reared from hatching in high CO2 (1,022 µatm were attracted to the predator cue, as has been reported in previous studies. There was no effect on fish responses of using either control or high CO2 water in the flume. Larval damselfish reared for four days in high CO2 (1,051 µatm exhibited a slower response to a startle stimulus and slower escape speed compared with fish reared in control conditions (464 µatm. There was no effect of test water on escape responses. Treatment of high-CO2 reared

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

International Nuclear Information System (INIS)

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

2013-05-01

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

Comparison between UMSICHT water hammer experiments and calculations using RELAP5/MOD3.3

International Nuclear Information System (INIS)

Messing, Ralf

2008-01-01

Water hammer phenomena regularly occur in piping systems of nuclear power plants, e.g. after the rapid closure of valves. As the loads are usually several times higher than during normal operation pipe rupture is possible and therefore the integrity of the nuclear power plant can be endangered. In recent years extensive studies have been performed to assess the capabilities of the widely-used transient thermohydraulic system codes like RELAP5 for the simulation of water hammers and pressure surges in piping systems. The parameters affecting the results of such simulations are on the one side related to the numerical method, in particular the grid step size Δx and time step size Δt, and on the other hand to the models for present physical effects like the fluid-structure-interaction (FSI), unsteady friction or the release of dissolved air. In many studies experimental data obtained at the Fraunhofer Institute UMSICHT at Oberhausen/Germany has been used for code validation. In a vast campaign reliable data has been measured under varying boundary conditions for pressure surges in a pipe after a fast valve closure. Details of the experimental set-up are described in /1/ and /2/. Tiselj and Petelin /3/ provided theoretical background on the properties of the numerical scheme used by RELAP5 and Tiselj and Cerne /4/ highlighted the behavior at very small time steps. Kaliatka and al. /5/ investigated the influence of the grid step size and the time step size in RELAP5 on pressure transients obtained in UMSICHT experiments. Neuhaus and Dudlik /2/,/6/ studied the effects of fluidstructure- interaction, unsteady friction and degassing of dissolved air in tape-water filled pipes on pressure surges. They compared their numerical results again with experimental data from the UMSICHT test facility and observed a significant impact of all three parameters on the pressure-surge amplitudes and frequency. Barten and al. /15/ performed calculations of experiment 329 of the UMSICHT

Post test calculation of the experiment 'small break loss-of- coolant test' SBL-22 at the Finnish integral test facility PACTEL with the thermohydraulic code ATHLET

International Nuclear Information System (INIS)

Lischke, W.; Vandreier, B.

1997-01-01

At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory

Test results on reuse of reclaimed shower water - A summary

Science.gov (United States)

Verostko, Charles E.; Garcia, Rafael; Sauer, Richard; Reysa, Richard P.; Linton, Arthur T.

1989-01-01

Results are presented from tests to evaluate a microgravity whole body shower and waste water recovery system design for possible use on the Space Station. Several water recovery methods were tested, including phase change distillation, a thermoelectric hollow fiber membrane evaporation subsystem, and a reverse osmosis dynamic membrane system. Consideration is given to the test hardware, the types of soaps evaluated, the human response to showering with reclaimed water, chemical treatment for microbial control, the procedures for providing hygienic water, and the quality of water produced by the systems. All three of the waste water recovery systems tested successfully produced reclaimed water for reuse.

International water and sanitation technology transfers, experiences from Europe

NARCIS (Netherlands)

Krozer, Yoram; Hophmayer Tokich, Sharon

2016-01-01

Possibilities of transferring cost-effective, innovative water and wastewater technologies on public water markets are discussed based on experiences of the Dutch water business cluster in the Central and Eastern European Countries. These transfers evolved under suitable conditions, among others

4D ERT Monitoring of Subsurface Water Pipe Leakage During a Controlled Field Experiment

Science.gov (United States)

Inauen, C.; Chambers, J. E.; Wilkinson, P. B.; Meldrum, P.; Swift, R. T.; Uhlemann, S.; Gunn, D.; Dashwood, B.; Taxil, J.; Curioni, G.

2016-12-01

Locating and delineating leakage from subsurface pipelines is an important task for civil engineers. 4D Electrical Resistivity Tomography (ERT) allows changes in subsurface resistivity to be imaged at a high spatial and temporal resolution in a minimally invasive manner. It is therefore a promising tool to supplement conventional point-sensing techniques to monitor subsurface flow processes. To assess the efficacy of ERT for pipe leakage monitoring several controlled leak experiments were carried out at a test site in Blagdon, Bristol, UK. To simulate the leak, a plastic pipe with a hole was buried below a flat, grassed area at a depth of 0.7 m, representing a standard UK mains water pipe installation. The water table at the site lies well below the surface meaning that the experiment took entirely place in the vadose zone, where changes in resistivity are primarily sensitive to water content variations. The ERT array covered an area of 6.5m x 6.5m around the leak location. Data acquisition was carried out with the BGS PRIME (Proactive Infrastructure Monitoring and Evaluation) system, which facilitates remote scheduling and autonomous ERT data collection and transmission. To obtain the resistivity changes of the subsurface a 4D inversion was carried out using a Gauss-Newton approach with spatial and temporal smoothness constraints. We were able to reliably observe the onset, spread and cessation of the leakage. Measurements from in-situ soil sensors at several depths above and below the leak complemented the ERT data and allowed us to assess their reliability and directly relate them to hydrogeological processes. Moreover, through experimental tests with soil samples from the test area, a Waxman-Smits relation was obtained to directly convert the changes in electrical resistivity to gravimetric soil moisture content. With future experiments on the test site more work is planned towards survey optimization, automated processing and tracking of leakage plumes.

Test results and supporting analysis of a near-surface heater experiment in the Eleana argillite

International Nuclear Information System (INIS)

McVey, D.F.; Lappin, A.R.; Thomas, R.K.

1979-01-01

A preliminary evaluation of the in-situ thermomechanical response of argillite to heating was obtained from a near-surface heater test in the Eleana Formation, at the United States Department of Energy, Nevada Test Site. The experiment consisted of a 3.8 kW, 3-m long x 0.3-m diameter electrical heater in a central hole surrounded by peripheral holes containing instrumentation to measure temperature, gas pressures, and vertical displacement. A thermal model of the experiment agreed well with experimental results; a comparison of measured and predicted temperatures indicates that some nonmodeled vertical transport of water and water vapor occurred near the heater, especially at early times. A mechanical model indicated that contraction of expandable clays in the argillite produced a region 1.5 - 2.0 m in radius, in which opening of preexisting joints occurred as a result of volumetric contraction. Results of thermal and mechanical modeling, laboratory property measurements, experimental temperature measurements, and post-test observations are all self-consistent and provide preliminary information on the in-situ response of argillaceous rocks to the emplacement of heat-producing nuclear waste

Integrated water management system - Description and test results. [for Space Station waste water processing

Science.gov (United States)

Elden, N. C.; Winkler, H. E.; Price, D. F.; Reysa, R. P.

1983-01-01

Water recovery subsystems are being tested at the NASA Lyndon B. Johnson Space Center for Space Station use to process waste water generated from urine and wash water collection facilities. These subsystems are being integrated into a water management system that will incorporate wash water and urine processing through the use of hyperfiltration and vapor compression distillation subsystems. Other hardware in the water management system includes a whole body shower, a clothes washing facility, a urine collection and pretreatment unit, a recovered water post-treatment system, and a water quality monitor. This paper describes the integrated test configuration, pertinent performance data, and feasibility and design compatibility conclusions of the integrated water management system.

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Experiment data report for Semiscale Mod-1 test S-02-5 (blowdown heat transfer test)

International Nuclear Information System (INIS)

1975-12-01

Recorded test data are presented for Test S-02-5 of the Semiscale Mod-1 blowdown heat transfer test series. Test S-02-5 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 water-cooled nuclear reactor system and to provide data for the assessment of the Loss-of-Fluid Test (LOFT) design basis. Test S-02-5 was conducted from an initial cold leg fluid temperature of 544 0 F and an initial pressure of 2,253 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization transient with full core power (1.6 MW). An electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core. System flow was set to achieve the full design core temperature differential of 66 0 F. The flow resistance of the intact loop was based on core area scaling. During system depressurization, core power was reduced from the initial level of 1.6 MW in such a manner as to simulate the surface heat flux response of the LOFT nuclear fuel rods until such time that departure from nucleate boiling occurs

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

Attachment of composite porous supra-particles to air-water and oil-water interfaces: theory and experiment.

Science.gov (United States)

Paunov, Vesselin N; Al-Shehri, Hamza; Horozov, Tommy S

2016-09-29

We developed and tested a theoretical model for the attachment of fluid-infused porous supra-particles to a fluid-liquid interface. We considered the wetting behaviour of agglomerated clusters of particles, typical of powdered materials dispersed in a liquid, as well as of the adsorption of liquid-infused colloidosomes at the liquid-fluid interface. The free energy of attachment of a composite spherical porous supra-particle made from much smaller aggregated spherical particles to the oil-water interface was calculated. Two cases were considered: (i) a water-filled porous supra-particle adsorbed at the oil-water interface from the water phase, and, (ii) an oil-filled porous supra-particle adsorbed at the oil-water interface from the oil-phase. We derived equations relating the three-phase contact angle of the smaller "building block" particles and the contact angle of the liquid-infused porous supra-particles. The theory predicts that the porous supra-particle contact angle attached at the liquid interface strongly depends on the type of fluid infused in the particle pores and the fluid phase from which it approaches the liquid interface. We tested the theory by using millimetre-sized porous supra-particles fabricated by evaporation of droplets of polystyrene latex suspension on a pre-heated super-hydrophobic surface, followed by thermal annealing at the glass transition temperature. Such porous particles were initially infused with water or oil and approached to the oil-water interface from the infusing phase. The experiment showed that when attaching at the hexadecane-water interface, the porous supra-particles behaved as hydrophilic when they were pre-filled with water and hydrophobic when they were pre-filled with hexadecane. The results agree with the theoretically predicted contact angles for the porous composite supra-particles based on the values of the contact angles of their building block latex particles measured with the Gel Trapping Technique. The

A boiling-water reactor concept for low radiation exposure based on operating experience

International Nuclear Information System (INIS)

Koine, Y.; Uchida, S.; Izumiya, M.; Miki, M.

1983-01-01

A review of boiling-water reactor (BWR) operating experience indicates the significant role of water chemistry in determining the radiation dose rate contributing to occupational exposure. The major contributor among the radioactive species involved is identified as 60 Co, produced by neutron activation of 59 Co originating from structural materials. Iron crud, a fine solid form of corrosion product in the reactor water, is also shown to enhance the radiation dose rate. A theoretical study, supported by the operating experience and an extensive confirmatory test, led to the computerized analytical model called DR CRUD which is capable of predicting long-term radiation dose buildup. It accounts for the mechanism of radiation buildup through corrosion products such as irons, cobalts and other radioactive elements; their generation, transport, activation, interaction and deposition in the reactor coolant system are simulated. A scoping analysis, using this model as a tool, establishes the base line of the BWR concept for low occupational exposure. The base line consists of a set of target values for an annual exposure of 200 man.rem in an 1100 MW(e) BWR unit. They are the parameters that will be built into the design such as iron and cobalt inputs to the reactor water, and the capability of the reactor and the condensate purification system. Applicable means of technology are identified to meet the targets, ranging from improved water chemistry to the purification technique, optimized material selection and the recommended operational procedure. Extensive test programmes provide specifications of these means for use in BWRs. Combinations of their application are reviewed to define the concept of reduced exposure. Analytical study verifies the effectiveness of the proposed BWR concept in achieving a low radiation dose rate; occupational exposure is reduced to 200 man.rem/a. (author)

Water NSTF Design, Instrumentation, and Test Planning

Energy Technology Data Exchange (ETDEWEB)

Lisowski, Darius D.; Gerardi, Craig D.; Hu, Rui; Kilsdonk, Dennis J.; Bremer, Nathan C.; Lomperski, Stephen W.; Kraus, Adam R.; Bucknor, Matthew D.; Lv, Qiuping; Farmer, Mitchell T.

2017-08-01

The following report serves as a formal introduction to the water-based Natural convection Shutdown heat removal Test Facility (NSTF) program at Argonne. Since 2005, this US Department of Energy (DOE) sponsored program has conducted large scale experimental testing to generate high-quality and traceable validation data for guiding design decisions of the Reactor Cavity Cooling System (RCCS) concept for advanced reactor designs. The most recent facility iteration, and focus of this report, is the operation of a 1/2 scale model of a water-RCCS concept. Several features of the NSTF prototype align with the conceptual design that has been publicly released for the AREVA 625 MWt SC-HTGR. The design of the NSTF also retains all aspects common to a fundamental boiling water thermosiphon, and thus is well poised to provide necessary experimental data to advance basic understanding of natural circulation phenomena and contribute to computer code validation. Overall, the NSTF program operates to support the DOE vision of aiding US vendors in design choices of future reactor concepts, advancing the maturity of codes for licensing, and ultimately developing safe and reliable reactor technologies. In this report, the top-level program objectives, testing requirements, and unique considerations for the water cooled test assembly are discussed, and presented in sufficient depth to support defining the program’s overall scope and purpose. A discussion of the proposed 6-year testing program is then introduced, which outlines the specific strategy and testing plan for facility operations. The proposed testing plan has been developed to meet the toplevel objective of conducting high-quality test operations that span across a broad range of single- and two-phase operating conditions. Details of characterization, baseline test cases, accident scenario, and parametric variations are provided, including discussions of later-stage test cases that examine the influence of geometric

Experience of water chemistry and radiation levels in Swedish BWRs

International Nuclear Information System (INIS)

Ivars, R.; Elkert, J.

1981-01-01

From the BWR operational experience in Sweden it has been found that the occupational radiation exposures have been comparatively low in an international comparison. One main reason for the favourable conditions is the good water chemistry performance. This paper deals at first with the design considerations of water chemistry and materials selection. Next, the experience of water chemistry and radiation levels are provided. Finally, some methods to further reduce the radiation sources are discussed. (author)

Pretest thermal analysis of the Tuff Water Migration/In-Situ Heater Experiment

International Nuclear Information System (INIS)

Bulmer, B.M.

1980-02-01

This report describes the pretest thermal analysis for the Tuff Water Migration/In-Situ Heater Experiment to be conducted in welded tuff in G-tunnel, Nevada Test Site. The parametric thermal modeling considers variable boiling temperature, tuff thermal conductivity, tuff emissivity, and heater operating power. For nominal tuff properties, some near field boiling is predicted for realistic operating power. However, the extent of boiling will be strongly determined by the ambient (100% water saturated) rock thermal conductivity. In addition, the thermal response of the heater and of the tuff within the dry-out zone (i.e., bounded by boiling isotherm) is dependent on the temperature variation of rock conductivity as well as the extent of induced boiling

Low cost sonoluminescence experiment in pressurized water

Science.gov (United States)

Bernal, L.; Insabella, M.; Bilbao, L.

2012-06-01

We present a low cost design for demostration and mesurements of light emmision from a sonoluminescence experiment. Using presurized water introduced in an acrylic cylinder and one piezoelectric from an ultrasonic cleaner, we are able to generate cavitacion zones with emission of light. The use of argon to pressurize the water improves the emission an the light can be seen at naked eye in a softlit ambient.

Low cost sonoluminescence experiment in pressurized water

Energy Technology Data Exchange (ETDEWEB)

Bernal, L; Insabella, M [LADOP, University of Mar del Plata (Argentina); Bilbao, L [INFIP, University of Buenos Aires and CONICET (Argentina)

2012-06-19

We present a low cost design for demostration and mesurements of light emission from a sonoluminescence experiment. Using pressurized water introduced in an acrylic cylinder and one piezoelectric from an ultrasonic cleaner, we are able to generate cavitacion zones with emission of light. The use of argon to pressurize the water improves the emission an the light can be seen at naked eye in a softlit ambient.

Low cost sonoluminescence experiment in pressurized water

International Nuclear Information System (INIS)

Bernal, L; Insabella, M; Bilbao, L

2012-01-01

We present a low cost design for demostration and mesurements of light emission from a sonoluminescence experiment. Using pressurized water introduced in an acrylic cylinder and one piezoelectric from an ultrasonic cleaner, we are able to generate cavitacion zones with emission of light. The use of argon to pressurize the water improves the emission an the light can be seen at naked eye in a softlit ambient.

Conducting Closed Habitation Experiments: Experience from the Lunar Mars Life Support Test Project

Science.gov (United States)

Barta, Daniel J.; Edeen, Marybeth A.; Henninger, Donald L.

2006-01-01

The Lunar-Mars Life Support Test Project (LMLSTP) was conducted from 1995 through 1997 at the National Aeronautics and Space Administration s (NASA) Johnson Space Center (JSC) to demonstrate increasingly longer duration operation of integrated, closed-loop life support systems that employed biological and physicochemical techniques for water recycling, waste processing, air revitalization, thermal control, and food production. An analog environment for long-duration human space travel, the conditions of isolation and confinement also enabled studies of human factors, medical sciences (both physiology and psychology) and crew training. Four tests were conducted, Phases I, II, IIa and III, with durations of 15, 30, 60 and 91 days, respectively. The first phase focused on biological air regeneration, using wheat to generate enough oxygen for one experimental subject. The systems demonstrated in the later phases were increasingly complex and interdependent, and provided life support for four crew members. The tests were conducted using two human-rated, atmospherically-closed test chambers, the Variable Pressure Growth Chamber (VPGC) and the Integrated Life Support Systems Test Facility (ILSSTF). Systems included test articles (the life support hardware under evaluation), human accommodations (living quarters, kitchen, exercise equipment, etc.) and facility systems (emergency matrix system, power, cooling, etc.). The test team was managed by a lead engineer and a test director, and included test article engineers responsible for specific systems, subsystems or test articles, test conductors, facility engineers, chamber operators and engineering technicians, medical and safety officers, and science experimenters. A crew selection committee, comprised of psychologists, engineers and managers involved in the test, evaluated male and female volunteers who applied to be test subjects. Selection was based on the skills mix anticipated for each particular test, and utilized

Recent performance experience with US light water reactor self-actuating safety and relief valves

Energy Technology Data Exchange (ETDEWEB)

Hammer, C.G.

1996-12-01

Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

Plutonium Recycle Test Reactor (PRTR). Operating Experience and Supporting R and D, Its Application to Heavy-Water Power Reactor Design and Operation

Energy Technology Data Exchange (ETDEWEB)

Harty, H. [Battelle Memorial Institute, Pacific Northwest Laboratories, Richland, WA (United States)

1968-04-15

Convincing answers to questions about heavy-water, pressure-tube, power reactors, e.g. pressure-tube serviceability, heavy-water management problems, long-term behaviour of special pressure-tube reactor components, and unique operating maintenance problems (compared to light-water reactors) must be based on actual operating experience with that type of reactor. PRTR operating experience and supporting R and D studies, although not always simple extrapolations to power reactors, can be summarized in a context applicable to future heavy-water power reactors, as follows: 1. Pressure-tube life, in a practical case, need not be limited by creep, gross hydriding, corrosion, or mechanical damage. The possibility that growth of a defect (perhaps service-induced) to a size that is critical under certain operating conditions, remains a primary unknown in pressure- tube life extrapolations. A pressure-tube failure in PRTR (combined with gross release of fuel material) proved only slightly more inconvenient, time consuming, and damaging to the reactor proper, than occurred with a gross failure of a fuel element in PRTR. 2. Routine operating losses of heavy water appear tractable in heavy-water-cooled power reactors; losses from low-pressure systems can be insignificant over the life of a plant. Non-routine losses may prove to be the largest component of loss over the life of a plant. 3. The performance of special components in PRTR, e.g. the calandria and shields, has not deteriorated despite being subjected to non-standard operating conditions. The calandria now contains a light-water reflector with single barrier separation from the heavy-water moderator. The carbon steel shields (containing carbon steel shot) show no deterioration based on pressure drop measurements and piping activation immediately outside the shields. The helium pressurization system (for primary coolant pressurization) remains a high maintenance system, and cannot be recommended for power reactors, based

U.S. experience with hydrogen water chemistry in boiling water reactors

International Nuclear Information System (INIS)

Cowan, R.L.; Head, R.A.; Indig, M.E.; Ruiz, C.P.; Simpson, J.L.

1988-01-01

Hydrogen water chemistry in boiling water reactors is currently being adopted by many utilities in the U.S., with eleven units having completed preimplementation test programs, four units operating permanently with hydrogen water chemistry, and six other units in the process of installing permanent equipment. Intergranular stress corrosion cracking protection is required for the recirculation piping system and other regions of the BWR systems. The present paper explores progress in predicting and monitoring hydrogen water chemistry response in these areas. Testing has shown that impurities can play an important role in hydrogen water chemistry. Evaluation of their effects are also performed. Both computer modeling and in plant measurements show that each plant will respond uniquely to feedwater hydrogen addition. Thus, each plant has its own unique hydrogen requirement for recirculation system protecion. Furthermore, the modeling, and plant measurements show that different regions of the BWR respond differently to hydrogen injection. Thus, to insure protection of components other than the recirculation systems may require more (or less) hydrogen demand than indicated by the recirculation system measurements. In addition, impurities such as copper can play a significant role in establishing hydrogen demand. (Nogami, K.)

Post test calculation of the experiment `small break loss-of- coolant test` SBL-22 at the Finnish integral test facility PACTEL with the thermohydraulic code ATHLET

Energy Technology Data Exchange (ETDEWEB)

Lischke, W.; Vandreier, B. [Univ. for Applied Sciences, Zittau/Goerlitz (Germany). Dept. of Nuclear Technology

1997-12-31

At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory. 5 refs.

Post test calculation of the experiment `small break loss-of- coolant test` SBL-22 at the Finnish integral test facility PACTEL with the thermohydraulic code ATHLET

Energy Technology Data Exchange (ETDEWEB)

Lischke, W; Vandreier, B [Univ. for Applied Sciences, Zittau/Goerlitz (Germany). Dept. of Nuclear Technology

1998-12-31

At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory. 5 refs.

Current experience and a new modeling on water hammer due to steam condensation in PWR secondary system

International Nuclear Information System (INIS)

Kawanishi, K.; Kasahara, J.; Ueno, T.; Suzuta, T.

1998-01-01

There have been possibilities to occur water hammer in pipelines of turbine system for nuclear or fossil fuel power plants. According to the NUREG report, approximately 150 events have been reported since 1969, we also have an experience recently. Water hammer occurs due to sudden steam condensation with pressure pulse. This kind of pressure pulses has been made by alternative producing and condensing of steam slug in the pipe and its frequency relates subcooling and pipe structures. This paper presents our current experience on water hammer with some experimental studies. The present experiment has been performed to obtain the data base for evaluating the pressure pulses. The test pipe was horizontal tubes with dead end connected to vertical tube which simulating drain line in PWR secondary system. The main results are shown as follows; Magnitude of pressure pulse depends drain velocity and initial subcooling. Pipe structure effects on the frequency and continual time of water hammer phenomenon. A new modeling for quantitative explanation of the phenomena is also presented

OECD MCCI Small-Scale Water Ingression and Crust Strength Tests (SSWICS) SSWICS-3 test data report: thermal Hydraulic results, Rev. 0 February 19, 2003

International Nuclear Information System (INIS)

Lomperski, S.; Farmer, M.T.; Kilsdonk, D.; Aeschlimann, B.

2011-01-01

The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core/concrete interaction (MCCI) when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. However, due to the integral nature of these tests, several questions regarding the crust freezing behavior could not be adequately resolved. These questions include: (1) To what extent does water ingression into the crust increase the melt quench rate above the conduction-limited rate and how is this affected by melt composition and system pressure and (2) What is the fracture strength of the corium crust when subjected to a thermal-mechanical load and how does it depend upon the melt composition? A series of separate-effects experiments are being conducted to address these issues. The first employs an apparatus designed to measure the quench rate of a pool of corium (∼φ30 cm; up to 20 cm deep). The main parameter to be varied in these quench tests is the melt composition since it is thought to have a critical influence on the crust cracking behavior which, in turn, alters quench rate. The issue of crust strength will be addressed with a second apparatus designed to mechanically load the crust produced by the quench tests. This apparatus will measure the fracture strength of the crust while under a thermal load created by a heating element beneath the crust. The two apparatuses used to measure the melt quench rate and crust strength are jointly referred to as SSWICS (Small-Scale Water Ingression and Crust Strength). This report describes results of the third water ingression test, designated SSWICS-3. This test investigated the quenching behavior of a fully oxidized PWR

OECD MCCI project Small-Scale Water Ingression and Crust Strength Tests (SSWICS) SSWICS-1 test data report: thermal hydraulic results. Rev. 0 September 20, 2002

International Nuclear Information System (INIS)

Lomperski, S.; Farmer, M.T.; Kilsdonk, D.J.; Aeschlimann, R.W.; Basu, S.

2011-01-01

The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core/concrete interaction (MCCI) when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. However, due to the integral nature of these tests, several questions regarding the crust freezing behavior could not be adequately resolved. These questions include: (1) To what extent does water ingression into the crust increase the melt quench rate above the conduction-limited rate and how is this affected by melt composition and system pressure and (2) What is the fracture strength of the corium crust when subjected to a thermal-mechanical load and how does it depend upon the melt composition? A series of separate-effects experiments are being conducted to address these issues. The first employs an apparatus designed to measure the quench rate of a pool of corium (∼φ30 cm; up to 20 cm deep). The main parameter to be varied in these quench tests is the melt composition since it is thought to have a critical influence on the crust cracking behavior which, in turn, alters quench rate. The issue of crust strength will be addressed with a second apparatus designed to mechanically load the crust produced by the quench tests. This apparatus will measure the fracture strength of the crust while under a thermal load created by a heating element beneath the crust. The two apparatuses used to measure the melt quench rate and crust strength are jointly referred to as SSWICS (Small-Scale Water Ingression and Crust Strength). This report describes results of the first water ingression test, designated SSWICS-1. The report includes a description of the test apparatus, the

[Virus adsorption from batch experiments as influenced by air-water interface].

Science.gov (United States)

Zhang, Hui; Zhao, Bing-zi; Zhang, Jia-bao; Zhang, Cong-zhi; Wang, Qiu-ying; Chen, Ji

2007-12-01

The presence of air-water interface in batch sorption experiments may result in inaccurate estimation of virus adsorption onto various soils. A batch sorption experiment was conducted to compare the adsorption results of MS2 in different soils under presence/absence of air-water interface. Soils with sterilization/nonterilization treatment were used. Virus recovery efficiency in a blank experiment (no soil) was also evaluated as affected by different amount of air-water interface. The presence of air-water interface altered the results of virus adsorption in different soils with different extent, with Sandy fluvo-aquic soil being the most considerably affected, followed by Red loam soil, and the least being Red clay soil, probably because of different soil properties associated with virus adsorption/inactivation. Soil sterilization resulted in more significant difference of virus adsorption onto the Sandy fluvo-aquic soil between the presence and absence of air-water interface, while a reduced difference was observed in the Red loam soil. The presence of air-water interface significantly decreased virus recovery efficiency, with the values being decreased with increase in the amount of air-water interface. Soil particles likely prohibit viruses from reaching the air-water interface or alter the forces at the solid-water-air interface so that the results from the blank experiment did not truly represent results from control blank, which probably resulted in adsorption difference between presence and absence of the air-water interface.

Development test procedure High Pressure Water Jet System

International Nuclear Information System (INIS)

Crystal, J.B.

1995-01-01

Development testing will be performed on the water jet cleaning fixture to determine the most effective arrangement of water jet nozzles to remove contamination from the surfaces of canisters and other debris. The following debris may be stained with dye to simulate surface contaminates: Mark O, Mark I, and Mark II Fuel Storage Canisters (both stainless steel and aluminum), pipe of various size, (steel, stainless, carbon steel and aluminum). Carbon steel and stainless steel plate, channel, angle, I-beam and other surfaces, specifically based on the Scientific Ecology Group (SEG) inventory and observations of debris within the basin. Test procedure for developmental testing of High Pressure Water Jet System

Development of a test facility for analyzing transients in supercritical water-cooled reactors by fractional scaling analysis

Energy Technology Data Exchange (ETDEWEB)

Roberto, Thiago D., E-mail: thiagodbtr@gmail.com [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil); Silva, Mário A. B. da, E-mail: mabs500@gmail.com [Departamento de Energia Nuclear (CTG/UFPE), Av. Professor Luiz Freire, 1000, Recife 50740-540, PE (Brazil); Lapa, Celso M.F., E-mail: lapa@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil)

2016-01-15

The feasibility of performing experiments using water under supercritical conditions is limited by technical and financial difficulties. These difficulties can be overcome by using model fluids that are characterized by feasible supercritical conditions, that is, lower critical pressure and critical temperature. Experimental investigations are normally used to determine the conditions under which model fluids reliably represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine the model fluids that represent supercritical fluids in a transient state. Recently, a similar technique known as fractional scaling analysis was developed to establish the conditions under which experiments can be performed using models that represent transients in prototypes. This paper presents a fractional scaling analysis application to determine parameters for a test facility in which transient conditions in supercritical water-cooled reactors are simulated by using carbon dioxide as a model fluid, whose critical point conditions are more feasible than those of water. Similarity is obtained between water (prototype) and carbon dioxide (model) by depressurization in a simple vessel. The main parameters required for the construction of a future test facility are obtained using the proposed method.

Development of a test facility for analyzing transients in supercritical water-cooled reactors by fractional scaling analysis

International Nuclear Information System (INIS)

Roberto, Thiago D.; Silva, Mário A. B. da; Lapa, Celso M.F.

2016-01-01

The feasibility of performing experiments using water under supercritical conditions is limited by technical and financial difficulties. These difficulties can be overcome by using model fluids that are characterized by feasible supercritical conditions, that is, lower critical pressure and critical temperature. Experimental investigations are normally used to determine the conditions under which model fluids reliably represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine the model fluids that represent supercritical fluids in a transient state. Recently, a similar technique known as fractional scaling analysis was developed to establish the conditions under which experiments can be performed using models that represent transients in prototypes. This paper presents a fractional scaling analysis application to determine parameters for a test facility in which transient conditions in supercritical water-cooled reactors are simulated by using carbon dioxide as a model fluid, whose critical point conditions are more feasible than those of water. Similarity is obtained between water (prototype) and carbon dioxide (model) by depressurization in a simple vessel. The main parameters required for the construction of a future test facility are obtained using the proposed method.

Full Scale Drinking Water System Decontamination at the Water Security Test Bed

Data.gov (United States)

U.S. Environmental Protection Agency — The EPA’s Water Security Test Bed (WSTB) facility is a full-scale representation of a drinking water distribution system. In collaboration with the Idaho National...

Experiment data report for semiscale Mod-1 test S-04-1 (baseline ECC test)

International Nuclear Information System (INIS)

Crapo, H.S.; Collins, B.L.; Sackett, K.E.

1976-09-01

Recorded test data are presented for Test S-04-1 of the Semiscale Mod-1 Baseline ECC Test Series. This test is among 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 system. Test S-04-1 was conducted from an initial cold leg fluid temperature of 542 0 F and an initial pressure of 2,263 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization and reflood transient using system volume scaled coolant injection parameters. System flow was set to achieve a core fluid temperature differential of 66 0 F at a full core power of 1.6 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a flat radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.6 MW in such a manner as to simulate the surface heat flux response of nuclear fuel rods until such time that departure from nucleate boiling might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core cooling injection into both the intact and broken loops. Coolant injection was continued until test termination at 200 seconds after initiation of blowdown

Water column velocimeter for NSRR experiment. Characteristics and data processing procedure

International Nuclear Information System (INIS)

Sugiyama, Tomoyuki; Fuketa, Toyoshi

2000-11-01

In order to clarify fuel behavior under reactivity initiated accident (RIA) conditions, pulse irradiation experiments on fuel rods are carried out in the Nuclear Safety Research Reactor (NSRR). One of concerns at fuel failure is mechanical energy generation in the reactor vessel. The mechanical energy is generated by a water hammer or a pressure impact occurred at fuel failure, and has possibility to damage reactor structures. Thus, the amount of generated mechanical energy is critical information for the safety evaluation of power reactor. In the NSRR experiments, the mechanical energy due to the water hammer is evaluated as the kinetic energy of the jumping water column at fuel failure, and the velocity of the water column is measured by the float type water column velocimeter. This report presents characteristics of the water column velocimeter and the procedure of data processing in the NSRR experiments. (author)

Outline of sodium-water reaction test in case of large leak with SWAT-3 testing equipments

International Nuclear Information System (INIS)

Sato, Minoru

1978-01-01

The key component in sodium-cooled fast reactors in steam generators, and the sodium-water reaction owing to the break of heating tubes may cause serious damages in equipments and pipings. The main factor controlling this phenomenon is the rate of leak of water. When the rate of water leak is small, the propagation of heating tube breaking may occur owing to ''wastage phenomenon'', on the other hand, when the rate of water leak is large, the phenomena of explosive pressure and flow occur due to the reaction heat and a large quantity of hydrogen generated by the reaction. In PNC, the testing equipments of SWAT-2 for small water leak and SWAT-1 for large leak were constructed, and the development test has been carried out to establish the method of safety design experimentally. The synthetic test equipment for the safety of steam generators, SWAT-3, was constructed to carry out the large water leak test in the scale close to actual plants. The object of the test, the outline of the test equipment, the phenomena of pressure and flow in the water injection test, the confirmation of the occurrence of secondary breaking of adjacent heating tubes, and the disposal of reaction products are described in this paper. This test is till going on, and the final conclusion will be reported later. (Kako, I.)

Protecting drinking water: water quality testing and PHAST in South Africa.

Science.gov (United States)

Breslin, E D

2000-01-01

The paper presents an innovative field-based programme that uses a simple total coliform test and the approach of PHAST (Participatory Hygiene And Sanitation Transformation) to help communities exploring possible water quality problems and actions that can be taken to address them. The Mvula Trust, a South African water and environmental sanitation NGO, has developed the programme. It is currently being tested throughout South Africa. The paper provides two case studies on its implementation in the field, and suggests ways in which the initiative can be improved in the future.

New test for oil soluble/water dispersible gas pipeline inhibitors

Energy Technology Data Exchange (ETDEWEB)

Stegmann, D.W.; Asperger, R.G.

1987-01-01

The wheel test provides good mixing of the condensate and water phases, the coupons are exposed to both phases. Therefore, the wheel test cannot distinguish between inhibitors that need continuous mixing of the these phases to maintain a water dispersion of the inhibitor and inhibitors that will self disperse into the water. This concept becomes important for pipelines in stratified flow where the water can settle out. In these cases with low turbulence, the inhibitor must self disperse into the water to be effective. The paper describes a test method to measure the effectiveness of an inhibitor and its ability to self disperse. The effectiveness of several inhibitors as predicted by the new test method is discussed relative to data from the wheel test and breaker tests. Field performance of these inhibitors in a gas gathering line, with liquids in stratified flow, are cities and compared with the results of the various laboratory tests.

The Performance test of Mechanical Sodium Pump with Water Environment

International Nuclear Information System (INIS)

Cho, Chungho; Kim, Jong-Man; Ko, Yung Joo; Jeong, Ji-Young; Kim, Jong-Bum; Ko, Bock Seong; Park, Sang Jun; Lee, Yoon Sang

2015-01-01

As contrasted with PWR(Pressurized light Water Reactor) using water as a coolant, sodium is used as a coolant in SFR because of its low melting temperature, high thermal conductivity, the high boiling temperature allowing the reactors to operate at ambient pressure, and low neutron absorption cross section which is required to achieve a high neutron flux. But, sodium is violently reactive with water or oxygen like the other alkali metal. So Very strict requirements are demanded to design and fabricate of sodium experimental facilities. Furthermore, performance testing in high temperature sodium environments is more expensive and time consuming and need an extra precautions because operating and maintaining of sodium experimental facilities are very difficult. The present paper describes performance test results of mechanical sodium pump with water which has been performed with some design changes using water test facility in SAM JIN Industrial Co. To compare the hydraulic characteristic of model pump with water and sodium, the performance test of model pump were performed using vender's experimental facility for mechanical sodium pump. To accommodate non-uniform thermal expansion and to secure the operability and the safety, the gap size of some parts of original model pump was modified. Performance tests of modified mechanical sodium pump with water were successfully performed. Water is therefore often selected as a surrogate test fluid because it is not only cheap, easily available and easy to handle but also its important hydraulic properties (density and kinematic viscosity) are very similar to that of the sodium. Normal practice to thoroughly test a design or component before applied or installed in reactor is important to ensure the safety and operability in the sodium-cooled fast reactor (SFR). So, in order to estimate the hydraulic behavior of the PHTS pump of DSFR (600 MWe Demonstraion SFR), the performance tests of the model pump such as performance

Tracer experiment by using radioisotope in surface water environment

International Nuclear Information System (INIS)

Suh, K.S.; Kim, K.C.; Chun, I.Y.; Jung, S.H.; Lee, C.W.

2007-01-01

Complete text of publication follows. 1. Objective An expansion of industrial activities and urbanization result in still increasing amount of pollutants discharged into surface water. Discharged pollutants in surface water have harmful effects on the ecology of a river system and human beings. Pollutants discharged into surface water is transported and dispersed under conditions characteristic to particular natural water receiver. Radiotracer method is a useful tool for monitoring the pollutant dispersion and description of mixing process taking place in natural streams. A tracer experiment using radioisotope was carried out to investigate the characteristics of a pollutant transport and a determination of the diffusion coefficients in a river system. 2. Methods The upper area of the Keum river was selected for the tracer experiment, which is located in a mid west of Korea. The measurements of the velocity and bathymetry before a tracer experiment were performed to select the sampling lines for a detection of the radioisotope. The radioisotope was instantaneously injected into a flow as a point source by an underwater glass-vial crusher. The detection was made with 60 2inch NaI(Tl) scintillation detectors at 3 transverse lines at a downstream position. The multi-channel data acquisition systems were used to collect and process the signals transmitted from the detectors. Two-dimensional numerical models were used to simulate the hydraulic parameters and the concentration distributions of the radioisotope injected into the river. 3. Results and Conclusion The calculated results such as velocity and concentrations were compared with the measured ones. The dispersion characteristics of the radioisotope were analyzed according to a variation of the flow rate, water level and diffusion coefficients. Also, the diffusion coefficients were calculated by using the measured concentrations and the coefficients obtained from the field experiment were compared with the ones

Simulation experiments of small break LOCA in upper plenum joint pipe for 5 MW heating test reactor

International Nuclear Information System (INIS)

Bo Jinhai; Jiang Shengyao; Zhang Youjie; Tong Yunxian; Sun Shusen; Yao Meisheng

1988-12-01

A simulation experiment of small break LOCA is introduced, which was caused by the breakdown of a small size or middle size pipe located at upper plenum, or by unexpected opening the safety valve. In the tests, the system pressure, temperature, void fraction and total loss of water were studied. The results showed that the total loss of water was nearly 20% of initial loading water. It means under this condition the 5MW low temperature heating reactor being built in Institute of Nnclear Engergy Technology of Tsinghua University is safe

High heat flux tests at divertor relevant conditions on water-cooled swirl tube targets

International Nuclear Information System (INIS)

Schlosser, J.; Boscary, J.

1994-01-01

High heat flux experiments were performed to provide a technology for heat flux removal under NET/ITER relevant conditions. The water-cooled rectangular test sections were made of hardened copper with a stainless steel twisted tape installed inside a circular channel and one-side heated. The tests aimed to investigate the heat transfer and the critical heat flux in the subcooled boiling regime. A CHF data base of 63 values was established. Test results have shown the thermalhydraulic ability of swirl tubes to sustain an incident heat flux up to a 30 MW.m -2 range. (author) 10 refs.; 7 figs

Cold vacuum drying residual free water test description

International Nuclear Information System (INIS)

Pajunen, A.L.

1997-01-01

Residual free water expected to remain in a Multi-Canister Overpack (MCO) after processing in the Cold Vacuum Drying (CVD) Facility is investigated based on three alternative models of fuel crevices. Tests and operating conditions for the CVD process are defined based on the analysis of these models. The models consider water pockets constrained by cladding defects, water constrained in a pore or crack by flow through a porous bed, and water constrained in pores by diffusion. An analysis of comparative reaction rate constraints is also presented indicating that a pressure rise test can be used to show MCO's will be thermally stable at operating temperatures up to 75 C

EBR-II blanket fuel leaching test using simulated J-13 well water

International Nuclear Information System (INIS)

Fonnesbeck, J. E.

1999-01-01

This paper discusses the results of a pulsed-flow leaching test using simulated J-13 well water leachant. This test was performed on three blanket fuel segments from the ANL-W EBR-II nuclear reactor which were originally made up of depleted uranium (DU). This experiment was designed to mimic conditions which would exist if, upon disposal of this material in a geological repository, it came in direct contact with groundwater. These segments were contained in pressure vessels and maintained at a constant temperature of 90 C. Weekly aliquots of leachate were taken from the three vessels and replaced with an equal volume of fresh leachant. These weekly aliquots were analyzed for both 90 Sr and 137 Cs. The results of the pulsed-flow leach test showed the formation of uranium oxide (UO 2 ) and uranium hydride (UH 3 ) particulate with rapid release of the 137 Cs and 90 Sr to the leachant. On the fifth week of sampling, one of the vessels became over pressurized and vented gas when opened. The most reasonable explanation for the presence of gas in this vessel is that the unoxidized uranium metal in the blanket segment could have reacted with the surrounding water leachant to form hydrogen. However, an investigation is currently being undertaken to both qualify and quantify H 2 formation during uranium spent nuclear fuel corrosion in water

Numerical simulation of air- and water-flow experiments in a block of variably saturated, fractured tuff from Yucca Mountain, Nevada

International Nuclear Information System (INIS)

Kwicklis, E.M.; Healy, R.W.; Thamir, F.; Hampson, D.

1998-01-01

Numerical models of water movement through variably saturated, fractured tuff have undergone little testing against experimental data collected from relatively well-controlled and characterized experiments. This report used the results of a multistage experiment on a block of variably saturated, fractured, welded tuff and associated core samples to investigate if those results could be explained using models and concepts currently used to simulate water movement in variably saturated, fractured tuff at Yucca Mountain, Nevada, the potential location of a high-level nuclear-waste repository. Aspects of the experiment were modeled with varying degrees of success. Imbibition experiments performed on cores of various lengths and diameters were adequately described by models using independently measured permeabilities and moisture-characteristic curves, provided that permeability reductions resulting from the presence of entrapped air were considered. Entrapped gas limited maximum water saturations during imbibition to approximately 0.70 to 0,80 of the fillable porosity values determined by vacuum saturation. A numerical simulator developed for application to fluid flow problems in fracture networks was used to analyze the results of air-injection tests conducted within the tuff block through 1.25-cm-diameter boreholes. These analyses produced estimates of transmissivity for selected fractures within the block. Transmissivities of other fractures were assigned on the basis of visual similarity to one of the tested fractures. The calibrated model explained 53% of the observed pressure variance at the monitoring boreholes (with the results for six outliers omitted) and 97% of the overall pressure variance (including monitoring and injection boreholes) in the subset of air-injection tests examined

Water electrolysis system refurbishment and testing

Science.gov (United States)

Greenough, B. M.

1972-01-01

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

Field testing at the Climax Stock on the Nevada Test Site: spent fuel test and radionuclide migration experiments

International Nuclear Information System (INIS)

Ballou, L.B.; Isherwood, D.J.; Patrick, W.C.

1982-01-01

Two field tests in the Climax Stock are being conducted. The Climax Stock, a granitic instrusive, has been administratively excluded from consideration as a full-scale repository site. However, it provides a readily available facility for field testing with high-level radioactive materials at a depth (420 m) approaching that of a repository. The major test activity in the 1980 fiscal year has been initiation of the Spent Fuel Test-Climax (SFT-C). This test, which was authorized in June 1978, is designed to evaluate the generic feasibility of geologic storage and retrievability of commercial power reactor spent fuel assemblies in a granitic medium. In addition, the test is configured and instrumented to provide thermal and thermomechanical response data that will be relevant to the design of a repository in hard crystalline rock. The other field activity in the Climax Stock is a radionuclide migration test. It combines a series of field and laboratory migration experiments with the use of existing hydrologic models for pretest predictions and data interpretation. Goals of this project are to develop: (1) field measurement techniques for radionuclide migration studies in a hydrologic regime where the controlling mechanism is fracture permeability; (2) field test data on radionuclide migration; and (3) a comparison of laboratory- and field-measured retardation factors. This radionuclide migration test, which was authorized in the middle of the 1980 fiscal year, is in the preliminary design phase. The detailed program plan was prepared and subjected to formal peer review in August. In September/October researchers conducted preliminary flow tests with water in selected near-vertical fractures intersected by small horizontal boreholes. These tests were needed to establish the range of pressures, flow rates, and other operating parameters to be used in conducting the nuclide migration tests. 21 references, 14 figures, 1 table

Sodium-water reaction test facility (SWAT-3)

International Nuclear Information System (INIS)

Shimazu, Hisashi; Ukechi, Kazutoshi; Sasakura, Kazutake; Kusunoki, Junichi

1976-01-01

In the development of the liquid metal cooled fast breeder reactor (LMFBR), the steam generator (SG) is considered one of the most important components. The Power Reactor and Nuclear Fuel Development Corporation (PNC) is now promoting the research and development of the SG system used with the prototype fast breeder reactor ''Monju''. In this research, the phenomena of the sodium-water reaction in the SG are the key which must be investigated for the solution of problems. The test facility (SWAT-3) simulating Monju's SG on the scale of 1/2.5 was designed, fabricated and installed by IHI at Oarai Engineering Center of PNC, its pre-operation being accomplished in February 1975. The purpose of SWAT-3 is summarized as follows: (1) To perform an overall test on the safety of Monju's SG and intermediate heat transport system under the design condition against sodium-water reaction accidents. (2) To investigate the damage of the SG structure caused by the sodium-water reaction, and the possibility of repair and recovery operations. The first test was accomplished successfully on June 9, 1975. As a result of the test, the fundamental function of this test facility was proven to be satisfactory as expected. (auth.)

Loss-of-Fluid Test findings in pressurized water reactor core's thermal-hydraulic behavior

International Nuclear Information System (INIS)

Russell, M.

1983-01-01

This paper summarizes the pressurized water reactor (PWR) core's thermal-hydraulic behavior findings from experiments performed at the Loss-of-Fluid Test (LOFT) Facility at the Idaho National Engineering Laboratory. The potential impact of these findings on the safety and economics of PWR's generation of electricity is also discussed. Reviews of eight important findings in the core's physical behavior and in experimental methods are presented with supporting evidence

The microelectronics and photonics test bed (MPTB) space, ground test and modeling experiments

International Nuclear Information System (INIS)

Campbell, A.

1999-01-01

This paper is an overview of the MPTB (microelectronics and photonics test bed) experiment, a combination of a space experiment, ground test and modeling programs looking at the response of advanced electronic and photonic technologies to the natural radiation environment of space. (author)

The effect of isolation on reproduction and growth of Pseudosuccinea columella (Pulmonata: Lymnaeidae: a snail-conditioned water experiment

Directory of Open Access Journals (Sweden)

Gutiérrez Alfredo

2002-01-01

Full Text Available A snail-conditioned water experiment was conducted in Pseudosuccinea columella to test the possible role of a chemical interaction between snails on the diminished growth and fecundity rates found for snails raised in pairs compared to those raised in complete isolation. The results permit to discard the hypothesis of an inhibition of growth and reproduction between snails due to factors released into the water.

Why a regional approach to postgraduate water education makes sense - the WaterNet experience in Southern Africa

Science.gov (United States)

Jonker, L.; van der Zaag, P.; Gumbo, B.; Rockström, J.; Love, D.; Savenije, H. H. G.

2012-03-01

This paper reports the experience of a regional network of academic departments involved in water education that started as a project and evolved, over a period of 12 yr, into an independent network organisation. The paper pursues three objectives. First, it argues that it makes good sense to organise postgraduate education and research on water resources on a regional scale. This is because water has a transboundary dimension that poses delicate sharing questions, an approach that promotes a common understanding of what the real water-related issues are, results in future water specialists speaking a common (water) language, enhances mutual respect, and can thus be considered an investment in future peace. Second, it presents the WaterNet experience as an example that a regional approach can work and has an impact. Third, it draws three generalised lessons from the WaterNet experience. Lesson 1: For a regional capacity building network to be effective, it must have a legitimate ownership structure and a clear mandate. Lesson 2: Organising water-related training opportunities at a regional and transboundary scale makes sense - not only because knowledge resources are scattered, but also because the topic - water - has a regional and transboundary scope. Lesson 3: Jointly developing educational programmes by sharing expertise and resources requires intense intellectual management and sufficient financial means.

Testing Preference Axioms in Discrete Choice experiments

DEFF Research Database (Denmark)

Hougaard, Jens Leth; Østerdal, Lars Peter; Tjur, Tue

Recent studies have tested the preference axioms of completeness and transitivity, and have detected other preference phenomena such as unstability, learning- and tiredness effects, ordering effects and dominance, in stated preference discrete choice experiments. However, it has not been explicitly...... of the preference axioms and other preference phenomena in the context of stated preference discrete choice experiments, and examine whether or how these can be subject to meaningful (statistical) tests...

The minimal ice water caloric test compared with established vestibular caloric test procedures.

Science.gov (United States)

Schmäl, Frank; Lübben, Björn; Weiberg, Kerstin; Stoll, Wolfgang

2005-01-01

Caloric testing of the vestibular labyrinth is usually performed by classical caloric test procedures (CCTP) using water warmed to 30 degrees C and 44 degrees C. Ice water irrigation (4 degrees C) is usually not performed, although it might be useful as a bedside test. To verify the validity of the Minimal Ice Water Caloric Test (MIWCT), comparative video-oculographic investigations were performed in 22 healthy subjects using ice water (0.5 ml, 1.0 ml, 2 ml), CCTP, and cold air (27 degrees C). Frequency, amplitude, slow phase velocity (SPV), the onset, and the duration of nystagmus were documented. After addition of three ice cubes, the temperature of conventional tap water (16 degrees C) fell within 13 min to 4 degrees C. In pessimum position the subjects demonstrated no nystagmus response. Compared to CCTP, MIWCT was associated with a significantly later onset of nystagmus and a significant prolongation of the nystagmus reaction. In contrast to air stimulation (27 degrees C), a significant Spearman's correlation was noted between MIWCT (1 and 2 ml) and established CCTP in respect of essential nystagmus parameters (frequency, amplitude and SPV). Furthermore, MIWCT (0.5 and 1 ml) showed a higher sensitivity and specificity with regard to the detection of canal paresis based on Jongkees' formula compared to stimulation with air 27 degrees C. Thus, MIWCT appears to be a suitable procedure for bedside investigation of vestibular function outside the vestibular laboratory, e.g. in a hospital ward, where bedridden patients with vertigo occasionally require vestibular testing.

Experimental evaluation of the drag coefficient of water rockets by a simple free-fall test

Energy Technology Data Exchange (ETDEWEB)

Barrio-Perotti, R; Blanco-Marigorta, E; Argueelles-Diaz, K; Fernandez-Oro, J [Departamento de Energia, Universidad de Oviedo, Campus de Viesques, 33271 Gijon, Asturias (Spain)], E-mail: barrioraul@uniovi.es

2009-09-15

The flight trajectory of a water rocket can be reasonably calculated if the magnitude of the drag coefficient is known. The experimental determination of this coefficient with enough precision is usually quite difficult, but in this paper we propose a simple free-fall experiment for undergraduate students to reasonably estimate the drag coefficient of water rockets made from plastic soft drink bottles. The experiment is performed using relatively small fall distances (only about 14 m) in addition with a simple digital-sound-recording device. The fall time is inferred from the recorded signal with quite good precision, and it is subsequently introduced as an input of a Matlab (registered) program that estimates the magnitude of the drag coefficient. This procedure was tested first with a toy ball, obtaining a result with a deviation from the typical sphere value of only about 3%. For the particular water rocket used in the present investigation, a drag coefficient of 0.345 was estimated.

[Reduction of animal experiments in experimental drug testing].

Science.gov (United States)

Behrensdorf-Nicol, H; Krämer, B

2014-10-01

In order to ensure the quality of biomedical products, an experimental test for every single manufactured batch is required for many products. Especially in vaccine testing, animal experiments are traditionally used for this purpose. For example, efficacy is often determined via challenge experiments in laboratory animals. Safety tests of vaccine batches are also mostly performed using laboratory animals. However, many animal experiments have clear inherent disadvantages (low accuracy, questionable transferability to humans, unclear significance). Furthermore, for ethical reasons and animal welfare aspects animal experiments are also seen very critical by the public. Therefore, there is a strong trend towards replacing animal experiments with methods in which no animals are used ("replacement"). If a replacement is not possible, the required animal experiments should be improved in order to minimize the number of animals necessary ("reduction") and to reduce pain and suffering caused by the experiment to a minimum ("refinement"). This "3R concept" is meanwhile firmly established in legislature. In recent years many mandatory animal experiments have been replaced by alternative in vitro methods or improved according to the 3R principles; numerous alternative methods are currently under development. Nevertheless, the process from the development of a new method to its legal implementation takes a long time. Therefore, supplementary regulatory measures to facilitate validation and acceptance of new alternative methods could contribute to a faster and more consequent implementation of the 3R concept in the testing of biomedical products.

Impacts of Personal Experience: Informing Water Conservation Extension Education

Science.gov (United States)

Huang, Pei-wen; Lamm, Alexa J.

2017-01-01

Extension educators have diligently educated the general public about water conservation. Incorporating audiences' personal experience into educational programming is recommended as an approach to effectively enhance audiences' adoption of water conservation practices. To ensure the impact on the audiences and environment, understanding the…

Experience on KKNPP VVER 1000 MWe water chemistry

International Nuclear Information System (INIS)

Ganesh, S.; Selvaraj, S.; Balasubramanian, M.R.; Selvavinayagam, P.; Pillai, Suresh Kumar

2015-01-01

Kudankulam Nuclear Power Project consists of pressurized water reactor (VVER) 2 x 1000 MWe constructed in collaboration with Russian Federation at Kudankulam in Tirunelveli District, Tamilnadu. Unit - 1 attained criticality on July 13 th 2013 and the unit was synchronized to grid on 22 nd October 2013. This paper highlights experience gained on water chemistry regime for primary and secondary circuit. (author)

Dielectric-breakdown tests of water at 6 MV

Directory of Open Access Journals (Sweden)

W. A. Stygar

2009-01-01

Full Text Available We have conducted dielectric-breakdown tests on water subject to a single unipolar pulse. The peak voltages used for the tests range from 5.8 to 6.8 MV; the effective pulse widths range from 0.60 to 1.1  μs; and the effective areas tested range from 1.8×10^{5} to 3.6×10^{6}  cm^{2}. The tests were conducted on water-insulated coaxial capacitors. The two electrodes of each capacitor have outer and inner radii of 99 and 56 cm, respectively. Results of the tests are consistent with predictions of the water-dielectric-breakdown relation developed in [Phys. Rev. ST Accel. Beams 9, 070401 (2006PRABFM1098-440210.1103/PhysRevSTAB.9.070401].

Channel type reactors with supercritical water coolant. Russian experience

International Nuclear Information System (INIS)

Kuznetsov, Y.N.; Gabaraev, B.A.

2003-01-01

Transition to coolant of supercritical parameters allows for principle engineering-andeconomic characteristics of light-water nuclear power reactors to be substantially enhanced. Russian experience in development of channel-type reactors with supercritical water coolant has demonstrated advantages and practical feasibility of such reactors. (author)

Results of an experiment in a Zion-like geometry to investigate the effect of water on the containment basement floor on direct containment heating (DCH) in the Surtsey Test Facility: The IET-4 test

International Nuclear Information System (INIS)

Allen, M.D.; Blanchat, T.K.; Pilch, M.; Nichols, R.T.

1992-09-01

This document discusses the fourth experiment of the Integral Effects Test (IET-4) series which was conducted to investigate the effects of high pressure melt ejection on direct containment heating. Scale models (1:10) of the Zion reactor pressure vessel (RPV), cavity, instrument tunnel, and subcompartment structures were constructed in the Surtsey Test Facility at Sandia National Laboratories. ne RPV was modeled with a melt generator that consisted of a steel pressure barrier, a cast MgO crucible, and a thin steel inner liner. The melt generator/crucible had a hemispherical bottom head containing a graphite limitor plate with a 3.5-cm exit hole to simulate the ablated hole in the RPV bottom head that would be tonned by tube ejection in a severe nuclear power plant accident. The reactor cavity model contained 3.48 kg of water with a depth of 0.9 cm that corresponded to condensate levels in the Zion plant. A 43-kg initial charge of iron oxide/aluminum/chromium thermite was used to simulate corium debris on the bottom head of the RPV. Molten thermite was ejected into the scaled reactor cavity by 6.7 MPa steam. IET-4 replicated the third experiment in the IET series (IET-3), except the Surtsey vessel contained slightly more preexisting oxygen (9.6 mol.% vs. 9.0 mol.%), and water was placed on the basement floor inside the crane wall. The cavity pressure measurements showed that a small steam explosion occurred in the cavity at about the same time as the steam explosion in IET-1. The oxygen in the Surtsey vessel in IET-4 resulted in a vigorous hydrogen bum, which caused a significant increase in the peak pressure, 262 kPa compared to 98 kPa in the IET-1 test. EET-3, with similar pre-existing oxygen concentrations, also had a large peak pressure of 246 kPa

Irradiation experiments and materials testing capabilities in High Flux Reactor in Petten

International Nuclear Information System (INIS)

Luzginova, N.; Blagoeva, D.; Hegeman, H.; Van der Laan, J.

2011-01-01

The text of publication follows: The High Flux Reactor (HFR) in Petten is a powerful multi-purpose research and materials testing reactor operating for about 280 Full Power Days per year. In combination with hot cells facilities, HFR provides irradiation and post-irradiation examination services requested by nuclear energy research and development programs, as well as by industry and research organizations. Using a variety of the custom developed irradiation devices and a large experience in executing irradiation experiments, the HFR is suitable for fuel, materials and components testing for different reactor types. Irradiation experiments carried out at the HFR are mainly focused on the understanding of the irradiation effects on materials; and providing databases for irradiation behavior of materials to feed into safety cases. The irradiation experiments and materials testing at the HFR include the following issues. First, materials irradiation to support the nuclear plant life extensions, for instance, characterization of the reactor pressure vessel stainless steel claddings to insure structural integrity of the vessel, as well as irradiation of the weld material coupons to neutron fluence levels that are representative for Light Water Reactors (LWR) internals applications. Secondly, development and qualification of the structural materials for next generation nuclear fission reactors as well as thermo-nuclear fusion machines. The main areas of interest are in both conventional stainless steel and advanced reduced activation steels and special alloys such as Ni-base alloys. For instance safety-relevant aspects of High Temperature Reactors (HTR) such as the integrity of fuel and structural materials with increasing neutron fluence at typical HTR operating conditions has been recently assessed. Thirdly, support of the fuel safety through several fuel irradiation experiments including testing of pre-irradiated LWR fuel rods containing UO 2 or MOX fuel. Fourthly

Endurance Test and Evaluation of Alkaline Water Electrolysis Cells

Science.gov (United States)

Kovach, Andrew J.; Schubert, Franz H.; Chang, B. J.; Larkins, Jim T.

1985-01-01

The overall objective of this program is to assess the state of alkaline water electrolysis cell technology and its potential as part of a Regenerative Fuel Cell System (RFCS) of a multikilowatt orbiting powerplant. The program evaluates the endurance capabilities of alkaline electrolyte water electrolysis cells under various operating conditions, including constant condition testing, cyclic testing and high pressure testing. The RFCS demanded the scale-up of existing cell hardware from 0.1 sq ft active electrode area to 1.0 sq ft active electrode area. A single water electrolysis cell and two six-cell modules of 1.0 sq ft active electrode area were designed and fabricated. The two six-cell 1.0 sq ft modules incorporate 1.0 sq ft utilized cores, which allow for minimization of module assembly complexity and increased tolerance to pressure differential. A water electrolysis subsystem was designed and fabricated to allow testing of the six-cell modules. After completing checkout, shakedown, design verification and parametric testing, a module was incorporated into the Regenerative Fuel Cell System Breadboard (RFCSB) for testing at Life Systems, Inc., and at NASA JSC.

Laboratory effectiveness testing of water-in-oil emulsion breakers

International Nuclear Information System (INIS)

Fingas, M.F.; Fieldhouse, B.; Bier, I.; Conrod, D.; Tennyson, E.

1995-01-01

The physics and chemistry of water-in-oil emulsions dominate the development of effectiveness tests. Emulsions are variable in stability--this variability is largely dependent on oil type and degree of weathering. These factors complicate the development of a test. Emulsions which have low stability will apparently break easily with chemical emulsion breakers. Broken emulsions will form a foam-like material, called rag, which retains water which is not part of the stable emulsions. Analytical methods used to determine the final stability of the broken or unbroken emulsion were evaluated. Measurements of water content and viscosity measurements show correlation to emulsion stability. Viscosity provides a more reliable measure of emulsion stability but water content measurements are more convenient and are largely used in this study. Twelve tests were developed in the past. Two testing methods have been developed to a usable stage. These tests are described and data using them provided. The effects of mixing time, agent amount, settling time and mixing energy on effectiveness results are presented

Contact sponge water absorption test implemented for in situ measures

Science.gov (United States)

Gaggero, Laura; Scrivano, Simona

2016-04-01

The contact sponge method is a non-destructive in-situ methodology used to estimate a water uptake coefficient. The procedure, unlike other in-situ measurement was proven to be directly comparable to the water uptake laboratory measurements, and was registered as UNI 11432:2011. The UNI Normal procedure requires to use a sponge with known density, soaked in water, weighed, placed on the material for 1 minute (UNI 11432, 2011; Pardini & Tiano, 2004), then weighed again. Difficulties arise in operating on test samples or on materials with porosity varied for decay. While carrying on the test, fluctuations in the bearing of the environmental parameters were negligible, but not the pressure applied to the surface, that induced the release of different water amounts towards the material. For this reason we designed a metal piece of the same diameter of the plate carrying the sponge, to be screwed at the tip of a pocket penetrometer. With this instrument the sponge was kept in contact with the surface for 1 minute applying two different loads, at first pushed with 0.3 kg/cm2 in order to press the sponge, but not its holder, against the surface. Then, a load of 1.1 kg/ cm2 was applied, still avoiding deviating the load to the sponge holder. We applied both the current and our implemented method to determine the water absorption by contact sponge on 5 fresh rock types (4 limestones: Fine - and Coarse grained Pietra di Vicenza, Rosso Verona, Breccia Aurora, and the silicoclastic Macigno sandstone). The results show that 1) the current methodology imply manual skill and experience to produce a coherent set of data; the variable involved are in fact not only the imposed pressure but also the compression mechanics. 2) The control on the applied pressure allowed reproducible measurements. Moreover, 3) the use of a thicker sponge enabled to apply the method even on rougher surfaces, as the device holding the sponge is not in contact with the tested object. Finally, 4) the

Condensation driven water hammer studies for feed water distribution pipe

International Nuclear Information System (INIS)

Savolainen, S.; Katajala, S.; Elsing, B.; Nurkkala, P.; Longvinov, S.A.; Trunov, N.B.; Sitnik, Yu.K.

1997-01-01

Special T-shaped feedwater distribution pipes were installed in steam generators at the Loviisa (Finland) and Rovno (Russia) nuclear power plants. The new shape was tested in an extensive testing programme. Since the tubes frequently suffer from corrosion damage, large-scale water hammer experiments were performed on a model facility in 1996. The main objectives of the water hammer experiments were to find out the prevailing parameters leading to water hammers, as well as the sensitivity of hammering to boundary conditions. A water hammer may occur when the mass flow rate into the steam generator exceeds 6 kg/s and the temperature difference between steam generator and feedwater exceeds 100 degC. Visual experiments and stress analyses of the pipe were also carried out. The weakest part, the T-joint, may hold against such water hammers only for a limited time of the order of few minutes. (M.D.)

The Reliability of a Functional Agility Test for Water Polo

Directory of Open Access Journals (Sweden)

Tucher Guilherme

2014-07-01

Full Text Available Few functional agility tests for water polo take into consideration its specific characteristics. The preliminary objective of this study was to evaluate the reliability of an agility test for water polo players. Fifteen players (16.3 ± 1.8 years old with a minimum of two years of competitive experience were evaluated. A Functional Test for Agility Performance (FTAP was designed to represent the context of this sport. Several trials were performed to familiarize the athlete with the movement. Two experienced coaches measured three repetitions of the FTAP. Descriptive statistics, repeated measures analysis of variance (ANOVA, 95% limit of agreement (LOA, intraclass correlation coefficient (ICC and standard error of measurements (SEM were used for data analysis. It was considered that certain criteria of reliability measures were met. There was no significant difference between the repetitions, which may be explained by an effect of the evaluator, the ability of the players or fatigue (p > 0.05. The ICC average from evaluators was high (0.88. The SEM varied between 0.13 s and 0.49 s. The CV average considering each individual was near 6-7%. These values depended on the condition of measurement. As the FTAP contains some characteristics that create a degree of unpredictability, the same athlete may reach different performance results, increasing variability. An adjustment in the sample, familiarization and careful selection of subjects help to improve this situation and enhance the reliability of the indicators.

Cellular changes in the skin of the rainbow trout (Oncorhynchus Mykiss) after short-term exposure to increased water temperature: Part 2: Experiments with Rhine water

International Nuclear Information System (INIS)

Iger, I.; Jenner, H.A.

1994-01-01

The aim of the study on the title subject is to show that salmonides in a simulated cooling water outlet area of an electric power plant along the river Rhine do not experience long-term or permanent negative consequences of their presence in the heated water. Laboratory experiments were carried out with 'test trouts'. The water in which the trouts were kept was heated up from 9C to 16C in 30 minutes. The temperature of 16C was maintained for 3 hours. After the heating up skin samples were taken at different periods for 35 days. Histochemical and electron microscopic investigations were carried out. The short term increase of the water temperature clearly effected the ultrastructure of the fish. The changes in the epidermis did not recover within 35 days. It is suggested that a number of the found cellular reactions can be used as indicators for environmental stress. These reactions concern the production of lysosomes and the migration of macrophages in the epidermis. No serious effects as skin decomposition and an excess of mucous secretion were found

Long Duration Testing of a Spacesuit Water Membrane Evaporator Prototype

Science.gov (United States)

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

2012-01-01

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

Water purification by corona-above-water treatment

NARCIS (Netherlands)

Pemen, A.J.M.; Heesch, van E.J.M.; Hoeben, W.F.L.M.

2012-01-01

Advanced oxidation technologies (AOT), such as non-thermal plasmas, are considered to be very promising for the purpose of water treatment. The goal of this study is to test the feasibility of "Corona-above-water" technology for the treatment of drinking water. Experiments have been performed on the

Experience feedback of an operation event during the experiment of feed-water pump switch

International Nuclear Information System (INIS)

Sun Shuhai; Li Huasheng; Zhang Hao

2012-01-01

In this paper an event is summarized and analyzed, which caused the quit of the high-pressure heaters and the nuclear power rising, during the experiment of the driven feed-water pump switch. The good experience feedback on this event is brought out through gathering related information of domestic nuclear plants. (authors)

Characterizing experiments of the PPOOLEX test facility

Energy Technology Data Exchange (ETDEWEB)

Puustinen, M.; Laine, J. (Lappeenranta Univ. of Technology, Nuclear Safety Research Unit (Finland))

2008-07-15

This report summarizes the results of the characterizing test series in 2007 with the scaled down PPOOLEX facility designed and constructed at Lappeenranta University of Technology. Air and steam/air mixture was blown into the dry well compartment and from there through a DN200 blowdown pipe to the condensation pool (wet well). Altogether eight air and four steam/air mixture experiments, each consisting of several blows (tests), were carried out. The main purpose of the experiment series was to study the general behavior of the facility and the performance of basic instrumentation. Proper operation of automation, control and safety systems was also tested. The test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. The facility is equipped with high frequency measurements for capturing different aspects of the investigated phenomena. The general behavior of the PPOOLEX facility differs significantly from that of the previous POOLEX facility because of the closed two-compartment structure of the test vessel. Heat-up by several tens of degrees due to compression in both compartments was the most obvious evidence of this. Temperatures also stratified. Condensation oscillations and chugging phenomenon were encountered in those tests where the fraction of non-condensables had time to decrease significantly. A radical change from smooth condensation behavior to oscillating one occurred quite abruptly when the air fraction of the blowdown pipe flow dropped close to zero. The experiments again demonstrated the strong diminishing effect that noncondensable gases have on dynamic unsteady loadings experienced by submerged pool structures. BWR containment like behavior related to the beginning of a postulated steam line break accident was observed in the PPOOLEX test facility during the steam/air mixture experiments. The most important task of the research project, to produce experimental data for code simulation purposes, can be

Experiment data report for semiscale Mod-1 test S-04-2 (baseline ECC test)

International Nuclear Information System (INIS)

Crapo, H.S.; Collins, B.L.; Sackett, K.E.

1976-09-01

Recorded test data are presented for Test S-04-2 of the Semiscale Mod-1 Baseline ECC test series. This test is among Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor system. Test S-04-2 was conducted from an initial cold leg fluid temperature of 542 0 F and an initial pressure of 2,263 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization and reflood transient using emergency core coolant injection parameters based on downcomer volume scaling. System flow was set to achieve a core fluid temperature differential of 66 0 F at a full core power of 1.6 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a flat radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.6 MW to simulate the surface heat flux response of nuclear fuel rods until such sime that departure from nucleate boiling might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core coolant injection into both the intact and broken loops. Coolant injection was continued until test termination at 200 seconds after initiation of blowdown. The purpose of the report is to make available the uninterpreted data from Test S-04-2 for future data analysis and test results reporting activities. The data, presented in the form of graphs in engineering units, have been analyzed only to the extent necessary to assure that they are reasonable and consistent

EBO feed water distribution system, experience gained from operation

Energy Technology Data Exchange (ETDEWEB)

Matal, O. [Energovyzkum, Brno (Switzerland); Schmidt, S.; Mihalik, M. [Atomove Elektrarne Bohunice, Jaslovske Bohunice (Switzerland)

1997-12-31

Advanced feed water distribution systems of the EBO design have been installed into steam generators at Units 3 and 4 of the NPP Jaslovske Bohunice (VVER 440). Experiences gained from the operation of steam generators with the advanced feed water distribution systems are discussed in the paper. (orig.). 4 refs.

EBO feed water distribution system, experience gained from operation

Energy Technology Data Exchange (ETDEWEB)

Matal, O [Energovyzkum, Brno (Switzerland); Schmidt, S; Mihalik, M [Atomove Elektrarne Bohunice, Jaslovske Bohunice (Switzerland)

1998-12-31

Advanced feed water distribution systems of the EBO design have been installed into steam generators at Units 3 and 4 of the NPP Jaslovske Bohunice (VVER 440). Experiences gained from the operation of steam generators with the advanced feed water distribution systems are discussed in the paper. (orig.). 4 refs.

Facility for generating crew waste water product for ECLSS testing

Science.gov (United States)

Buitekant, Alan; Roberts, Barry C.

1990-01-01

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

An apparatus to estimate the hydrodynamic coefficients of autonomous underwater vehicles using water tunnel testing.

Science.gov (United States)

Nouri, N M; Mostafapour, K; Bahadori, R

2016-06-01

Hydrodynamic coefficients or hydrodynamic derivatives of autonomous underwater vehicles (AUVs) play an important role in their development and maneuverability. The most popular way of estimating their coefficients is to implement captive model tests such as straight line tests and planar motion mechanism (PMM) tests in the towing tanks. This paper aims to develop an apparatus based on planar experiments of water tunnel in order to estimate hydrodynamic derivatives due to AUVs' acceleration and velocity. The capability of implementing straight line tests and PMM ones using mechanical oscillators located in the downstream flow of the model is considered in the design procedure of the system. The hydrodynamic derivatives that resulted from the acceleration and velocity of the AUV model were estimated using the apparatus that we developed. Static and dynamics test results were compared for the similar derivatives. The findings showed that the system provided the basis for conducting static tests, i.e., straight-line and dynamic tests that included pure pitch and pure heave. By conducting such tests in a water tunnel, we were able to eliminate errors related to the time limitation of the tests and the effects of surface waves in the towing tank on AUVs with applications in the deep sea.

Technical report on operating experience with boiling water reactor offgas systems

International Nuclear Information System (INIS)

Lo, R.; Barrett, L.; Grimes, B.; Eisenhut, D.

1978-03-01

Over 100 reactor years of Boiling Water Reactor (BWR) operating experience have been accumulated since the first commercial operation of BWRs. A number of incidents have occurred involving the ''offgas'' of these Boiling Water Reactors. This report describes the generation and processing of ''offgas'' in Boiling Water Reactors, the safety considerations regarding systems processing the ''offgas'', operating experience involving ignitions or explosions of ''offgas'' and possible measures to reduce the likelihood of future ignitions or explosions and to mitigate the consequences of such incidents should they occur

The Quality Testing of Water from Microbiology and Radioactivity

International Nuclear Information System (INIS)

Zainul Kamal; Yazid, M.; Mulyaningsih; Iim lmroatin

2002-01-01

The quality testing of well water from microbiologic and radioactivity has been done. The samples were taken from Degolan and Lodadi village, Ngemplak, Sleman. The quality testing based from standard procedure of microbiologic and environmental radioactivity. From the experimentally results showed that E. Coli in well water = 5 - 920 JPT / 100 ml, Streptococcus in well water 0 - 4 JPT /100 ml, E. Coli and Streptococcus in PAM water 0 JPT / 100 ml, radioactivity β totally in well water 0.08-0.34 Bq/l and in PAM water 0.08 - 0.31 Bq/l. From the dates required could be concluded that in microbiologically aspects the value of E. Coli and Streptococcus in well water higher than the threshold value from Health Department Rl 416/Menkes/PER/IX/1990, in radioactivity aspect lower than the threshold value from Health Department RI 416/Menkes/PER/IX/1990. (author)

Testing and use of radar water level sensors by the U.S. Geological Survey

Science.gov (United States)

Fulford, Janice M.

2016-01-01

The United States Geological Survey uses water-level (or stage) measurements to compute streamflow at over 8000 stream gaging stations located throughout the United States (waterwatch.usgs.gov, 2016). Streamflow (or discharge) is computed at five minute to hourly intervals from a relationship between water level and discharge that is uniquely determined for each station. The discharges are posted hourly to WaterWatch (waterwatch. usgs.gov) and are used by water managers to issue flood warnings and manage water supply and by other users of water information to make decisions. The accuracy of the water-level measurement is vital to the accuracy of the computed discharge. Because of the importance of water-level measurements, USGS has an accuracy policy of 0.02 ft or 0.2 percent of reading (whichever is larger) (Sauer and Turnipseed, 2010). Older technologies, such as float and shaft-encoder systems, bubbler systems and submersible pressure sensors, provide the needed accuracy but often require extensive construction to install and are prone to malfunctioning and damage from floating debris and sediment. No stilling wells or orifice lines need to be constructed for radar installations. During the last decade testing by the USGS Hydrologic Instrumentation Facility(HIF) found that radar water-level sensors can provide the needed accuracy for water-level measurements and because the sensor can be easily attached to bridges, reduce the construction required for installation. Additionally, the non-contact sensing of water level minimizes or eliminates damage and fouling from floating debris and sediment. This article is a brief summary of the testing efforts by the USGS HIF and field experiences with models of radar water-level sensors in streamflow measurement applications. Any use of trade names in this article is for descriptive purposes only and does not imply endorsement by the U.S. Government.

Code of practice for the release of hydrostatic test water from hydrostatic testing of petroleum liquid and gas pipelines

International Nuclear Information System (INIS)

1999-01-01

This booklet describes a series of administrative procedures regarding the code of practice in Alberta for the release of hydrostatic test water from hydrostatic testing of petroleum liquid and gas pipelines. The topics covered include the registration process, the type and quality of water to use during the test, and the analytical methods to be used. Reporting schedule and record keeping information are also covered. Schedule 1 discusses the requirements for the release of hydrostatic test water to land, while Schedule 2 describes the requirements for the release of hydrostatic test water to receiving water. 3 tabs

Operating experiences since rise-to-power test in high temperature engineering test reactor (HTTR)

International Nuclear Information System (INIS)

Tochio, Daisuke; Watanabe, Shuji; Motegi, Toshihiro; Kawano, Shuichi; Kameyama, Yasuhiko; Sekita, Kenji; Kawasaki, Kozo

2007-03-01

The rise-to-power test of the High Temperature Engineering Test Reactor (HTTR) was actually started in April 2000. The rated thermal power of 30MW and the rated reactor outlet coolant temperature of 850degC were achieved in the middle of Dec. 2001. After that, the reactor thermal power of 30MW and the reactor outlet coolant temperature of 950degC were achieved in the final rise-to-power test in April 2004. After receiving the operation licensing at 850degC, the safety demonstration tests have conducted to demonstrate inherent safety features of the HTGRs as well as to obtain the core and plant transient data for validation of safety analysis codes and for establishment of safety design and evaluation technologies. This paper summarizes the HTTR operating experiences for six years from start of the rise-to-power test that are categorized into (1) Operating experiences related to advanced gas-cooled reactor design, (2) Operating experiences for improvement of the performance, (3) Operating experiences due to fail of system and components. (author)

To What Extent is Drinking Water Tested in Sub-Saharan Africa? A Comparative Analysis of Regulated Water Quality Monitoring.

Science.gov (United States)

Peletz, Rachel; Kumpel, Emily; Bonham, Mateyo; Rahman, Zarah; Khush, Ranjiv

2016-03-02

Water quality information is important for guiding water safety management and preventing water-related diseases. To assess the current status of regulated water quality monitoring in sub-Saharan Africa, we evaluated testing programs for fecal contamination in 72 institutions (water suppliers and public health agencies) across 10 countries. Data were collected through written surveys, in-person interviews, and analysis of microbial water quality testing levels. Though most institutions did not achieve the testing levels specified by applicable standards or World Health Organization (WHO) Guidelines, 85% of institutions had conducted some microbial water testing in the previous year. Institutions were more likely to meet testing targets if they were suppliers (as compared to surveillance agencies), served larger populations, operated in urban settings, and had higher water quality budgets (all p water providers and rural public health offices will require greater attention and additional resources to achieve regulatory compliance for water quality monitoring in sub-Saharan Africa. The cost-effectiveness of water quality monitoring should be improved by the application of risk-based water management approaches. Efforts to strengthen monitoring capacity should pay greater attention to program sustainability and institutional commitment to water safety.

Liquid nitrogen - water interaction experiments for fusion reactor accident scenarios

International Nuclear Information System (INIS)

Duckworth, R.; Murphy, J.; Pfotenhauer, J.; Corradini, M.

2001-01-01

With the implementation of superconducting magnets in fusion reactors, the possibility exists for the interaction between water and cryogenic systems. The interaction between liquid nitrogen and water was investigated experimentally and numerically. The rate of pressurization and peak pressure were found to be driven thermodynamically by the expansion of the water and the boil-off of the liquid nitrogen and did not have a vapor explosion nature. Since the peak pressure was small in comparison to previous work with stratified geometries, the role of the geometry of the interacting fluids has been shown to be significant. Comparisons of the peak pressure and the rate of pressurization with respect to the ratio of the liquid nitrogen mass to water mass reveal no functional dependence as was observed in the liquid helium-water experiments. A simple thermodynamic model provides a fairly good description of the pressure rise data. From the data, the model will allow one to extract the interaction area of the water. As with previous liquid helium-water interaction experiments, more extensive investigation of the mass ratio and interaction geometry is needed to define boundaries between explosive and non-explosive conditions. (authors)

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

Artificial enzyme-powered microfish for water-quality testing.

Science.gov (United States)

Orozco, Jahir; García-Gradilla, Victor; D'Agostino, Mattia; Gao, Wei; Cortés, Allan; Wang, Joseph

2013-01-22

We present a novel micromotor-based strategy for water-quality testing based on changes in the propulsion behavior of artificial biocatalytic microswimmers in the presence of aquatic pollutants. The new micromotor toxicity testing concept mimics live-fish water testing and relies on the toxin-induced inhibition of the enzyme catalase, responsible for the biocatalytic bubble propulsion of tubular microengines. The locomotion and survival of the artificial microfish are thus impaired by exposure to a broad range of contaminants, that lead to distinct time-dependent irreversible losses in the catalase activity, and hence of the propulsion behavior. Such use of enzyme-powered biocompatible polymeric (PEDOT)/Au-catalase tubular microengine offers highly sensitive direct optical visualization of changes in the swimming behavior in the presence of common contaminants and hence to a direct real-time assessment of the water quality. Quantitative data on the adverse effects of the various toxins upon the swimming behavior of the enzyme-powered artificial swimmer are obtained by estimating common ecotoxicological parameters, including the EC(50) (exposure concentration causing 50% attenuation of the microfish locomotion) and the swimmer survival time (lifetime expectancy). Such novel use of artificial microfish addresses major standardization and reproducibility problems as well as ethical concerns associated with live-fish toxicity assays and hence offers an attractive alternative to the common use of aquatic organisms for water-quality testing.

Site characterization and validation - equipment design and techniques used in single borehole hydraulic testing, simulated drift experiment and crosshole testing

International Nuclear Information System (INIS)

Holmes, D.C.; Sehlstedt, M.

1991-10-01

This report describes the equipment and techniques used to investigate the variation of hydrogeological parameters within a fractured crystalline rock mass. The testing program was performed during stage 3 of the site characterization and validation programme at the Stripa mine in Sweden. This programme used a multidisciplinary approach, combining geophysical, geological and hydrogeological methods, to determine how groundwater moved through the rock mass. The hydrogeological work package involved three components. Firstly, novel single borehole techniques (focused packer testing) were used to determine the distribution of hydraulic conductivity and head along individual boreholes. Secondly, water was abstracted from boreholes which were drilled to simulate a tunnel (simulated drift experiment). Locations and magnitudes of flows were measured together with pressure responses at various points in the SCV rock mass. Thirdly, small scale crosshole tests, involving detailed interference testing, were used to determine the variability of hydrogeological parameters within previously identified, significant flow zones. (au)

Thermal effect on water retention curve of bentonite: experiment and thermodynamic modeling

International Nuclear Information System (INIS)

Qin Bing; Chen Zhenghai; Sun Faxin; Liu Yuemiao; Wang Ju

2012-01-01

The thermal effects on water retention curve of GMZ bentonite were investigated experimentally and theoretically. Water retention tests were conducted on GMZ bentonite at five temperatures ranging from 20℃ to 100℃. Test results showed that the water retention capacity and the hysteresis of the water retention curve decreased with increasing temperature, and that the water retention curves at different temperatures were almost parallel to each other. Based on the thermodynamics of sorption, a model was established to describe the temperature influence on the water retention curve. The model was validated by comparing the model predictions and the test results. (authors)

Performance test for a solar water heater

Science.gov (United States)

1979-01-01

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

Some tests of wet tropospheric calibration for the CASA Uno Global Positioning System experiment

Science.gov (United States)

Dixon, T. H.; Wolf, S. Kornreich

1990-01-01

Wet tropospheric path delay can be a major error source for Global Positioning System (GPS) geodetic experiments. Strategies for minimizing this error are investigted using data from CASA Uno, the first major GPS experiment in Central and South America, where wet path delays may be both high and variable. Wet path delay calibration using water vapor radiometers (WVRs) and residual delay estimation is compared with strategies where the entire wet path delay is estimated stochastically without prior calibration, using data from a 270-km test baseline in Costa Rica. Both approaches yield centimeter-level baseline repeatability and similar tropospheric estimates, suggesting that WVR calibration is not critical for obtaining high precision results with GPS in the CASA region.

Testing biological effects of hand-washing grey water for reuse in irrigation on an urban farm: a case study.

Science.gov (United States)

Khan, Mohammad Zain; Sim, Yei Lin; Lin, Yang Jian; Lai, Ka Man

2013-01-01

The feasibility of reusing hand-washing grey water contaminated with antibacterial hand-washing liquid for irrigation purposes in an urban farm is explored in this case study. Experiments are carried out to investigate if the quality of this grey water allows for its reuse in agriculture as per the guidelines established by the World Health Organization (WHO). However, there is no guideline to test the biological effect of grey water prior to agricultural use. It is plausible that the antibacterial property of the grey water can harm the soil microbial system and plants when applied to land, even if all other water quality parameters satisfy the WHO limit. We use algae (Chlorella vulgaris) and indigenous soil bacteria as initial plant and soil bacteria indicators, respectively, to test the potential inhibition of the water on plants and soil bacteria. Results show that the chemical oxygen demand (COD) of the grey water is 10% higher than the WHO permissible level, while all other water quality parameters are within the limits after four days of our experimental period. An inhibitory effect is observed in all of the biological tests. However, the inhibitory effect on algae and soil bacteria is not observed after the four-day period. The case study demonstrates a new approach for testing the biological effect of grey water, which can be used in conjunction with the WHO guideline, and provides data for this urban farm to set up a future water treatment system for grey-water reuse in irrigation.

To What Extent is Drinking Water Tested in Sub-Saharan Africa? A Comparative Analysis of Regulated Water Quality Monitoring

Directory of Open Access Journals (Sweden)

Rachel Peletz

2016-03-01

Full Text Available Water quality information is important for guiding water safety management and preventing water-related diseases. To assess the current status of regulated water quality monitoring in sub-Saharan Africa, we evaluated testing programs for fecal contamination in 72 institutions (water suppliers and public health agencies across 10 countries. Data were collected through written surveys, in-person interviews, and analysis of microbial water quality testing levels. Though most institutions did not achieve the testing levels specified by applicable standards or World Health Organization (WHO Guidelines, 85% of institutions had conducted some microbial water testing in the previous year. Institutions were more likely to meet testing targets if they were suppliers (as compared to surveillance agencies, served larger populations, operated in urban settings, and had higher water quality budgets (all p < 0.05. Our results indicate that smaller water providers and rural public health offices will require greater attention and additional resources to achieve regulatory compliance for water quality monitoring in sub-Saharan Africa. The cost-effectiveness of water quality monitoring should be improved by the application of risk-based water management approaches. Efforts to strengthen monitoring capacity should pay greater attention to program sustainability and institutional commitment to water safety.

Study on elastic-plastic fracture toughness test in high temperature water

International Nuclear Information System (INIS)

Miura, Yasufumi

2016-01-01

Structural integrity of internal components in light water reactors is important for the safety of operation and service lifetime. Fracture toughness is important parameter for structural integrity assessment of nuclear power plant. In general, fracture toughness of materials which compose the components in light water reactor is obtained with fracture toughness tests in air although some components are subjected to high temperature water because of the difficulty of fracture toughness test in high temperature water. However, the effects of high temperature water and hydrogen on fracture behavior of the structural materials in nuclear power plant such as low alloy steel, cast austenitic stainless steel, and Ni base alloy are concerned recently. In this study, elastic-plastic fracture toughness test of low alloy steel in simulated BWR water environment was studied. Fracture toughness test in high temperature water with original clip gage and normalization data reduction technique was established. The difference of fracture toughness J_Q tested in air between using elastic unload compliance method and normalization data reduction technique was also discussed. As a result, obtained value with normalization data reduction technique tended to be higher than the value with elastic unload compliance. (author)

Experiment data report for Semiscale Mod-1 Test S-29-1 (integral test with asymmetrical break)

International Nuclear Information System (INIS)

Crapo, H.S.; Jensen, M.F.; Sackett, K.E.

1976-07-01

Recorded test data are presented for Test S-29-1 of the Semiscale Mod-1 special heat transfer test series. This test is among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident (LOCA) in a pressurized-water reactor system. Test S-29-1 was conducted from an initial cold leg fluid temperature of 544 0 F and an initial pressure of 2,260 psia. An asymmetrical offset shear cold leg break was used to investigate the system response to a depressurization transient with a flow distribution different from that associated with a symmetrical cold leg break. System flow was set to achieve a core fluid temperature differential of 66 0 F at full core power of 1.6 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a flat radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.6 MW to simulate the surface heat flux response of nuclear fuel rods until such time that departure from nucleate boiling (DNB) might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core cooling injection into both the intact and broken loops. Coolant injection was continued until test termination at 200 seconds after initiation of blowdown

Testing of unlubricated water-immersed carbon-steel ball bearings

International Nuclear Information System (INIS)

Wensel, R.G.

1977-07-01

Wear data for untreated, chemically oxidized, autoclave oxidized, and hard chromium plated ball bearings operating in water at different radial loads, speeds and water temperatures are presented. Results were highly variable, even for replicated tests, making wear results difficult to interpret. Within the ranges used in these tests, load, speed and water temperature had no consistent effect on wear rate or operating torque. One treatment, the Du-Lite process, resulted in a significant reduction in wear rate. All oxide-forming treatments studied, resulted in improved corrosion resistance. (author)

Cooperation on impingement wastage experiment of Mod. 9Cr-1Mo steel using SWAT-1R sodium-water reaction test facility

International Nuclear Information System (INIS)

Beauchamp, F.; Allou, A.; Nishimura, M.; Umeda, R.

2013-01-01

Conclusion: • 6 experiments were carried out in the SWAT-1R facility of JAEA Oarai R&D Center to study the wastage resistance of the Mod. 9Cr-1Mo steel (T91) straight tubes. • These experiments were performed under the cooperation between CEA and JAEA. • The experiments were conducted successfully: - all the tubes were punctured by the reaction jet, - wastage and steam/water leak rates were obtained, - experimental results brought some new determining sets of wastage data on T91. • This fruitful cooperation has contributed to: - expanding the wastage database on T91, - upgrading wastage rates prediction from modelling, - the safety demonstration of future steam generators units

A Field Test of Electromigration as a Method for Remediating Sulfate from Shallow Ground Water

Science.gov (United States)

Patterson, C.G.; Runnells, D.D.

1996-01-01

Electromigration offers a potential tool for remediating ground water contaminated with highly soluble components, such as Na+, Cl-, NO3-, and SO4-. A field experiment was designed to test the efficacy of electromigration for preconcentrating dissolved SO42- in ground water associated with a fossil-fuel power plant. Two shallow wells, 25 feet apart (one 25 feet deep, the other 47 feet deep), were constructed in the upper portion of an unconfined alluvial aquifer. The wells were constructed with a double-wall design, with an outer casing of 4-inch PVC and an inner tube of 2-inch PVC; both were fully slotted (0.01 inch). Electrodes were constructed by wrapping the inner tubing with a 100-foot length of rare-earth metal oxide/copper wire. An electrical potential of 10.65 volts DC was applied, and tests were run for periods of 12, 44, and 216 hours. Results showed large changes in the pH from the initial pH of ground water of about 7.5 to values of approximately 2 and 12 at the anode and cathode, respectively. Despite the fact that the test conditions were far from ideal, dissolved SO42- was significantly concentrated at the anode. Over a period of approximately nine days, the concentration of SO42- at the anode reached what appeared to be a steady-state value of 2200 mg/L, compared to the initial value in ground water of approximately 1150 mg/L. The results of this field test should encourage further investigation of electromigration as a tool in the remediation of contaminated ground water.

Ground water pollution by roof runoff infiltration evidenced with multi-tracer experiments.

Science.gov (United States)

Ammann, Adrian A; Hoehn, Eduard; Koch, Sabine

2003-03-01

The infiltration of urban roof runoff into well permeable subsurface material may have adverse effects on the ground water quality and endanger drinking water resources. Precipitation water from three different roofs of an industrial complex was channelled to a pit and infiltrated into a perialpine glaciofluvial gravel-and-sand aquifer. A shaft was constructed at the bottom of the pit and equipped with an array of TDR probes, lysimeters and suction cups that allowed measuring and sampling soil water at different depths. A fast infiltration flow was observed during natural rainfall events and during artificial infiltration experiments. For a better understanding of the behaviour of contaminants, experiments were conducted with cocktails of compounds of different reactivity (ammonium, strontium, atratone) and of non-reactive tracers (uranine, bromide, naphthionate), which represent different classes of pollutants. The experiment identified cation exchange reactions influencing the composition of the infiltrating water. These processes occurred under preferential flow conditions in macropores of the material. Measuring concentration changes under the controlled inflow of tracer experiments, the pollution potential was found to be high. Non-reactive tracers exhibited fast breakthrough and little sorption.

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

International Nuclear Information System (INIS)

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

1999-01-01

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

LOFT/LP-LB-1, Loss of Fluid Test, Large-Break LOCA Experiment

International Nuclear Information System (INIS)

1989-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, Thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCE is expected to closely model a LPWR LOCA. 2 - Description of test: Experiment LP-LB-1 was conducted on 3 February 1984 in the Loss-of-Fluid Test (LOFT) facility at the Idaho National Engineering Laboratory under the auspices of the Organization for Economic Cooperation and Development. The primary objectives of Experiment LP-LB-1 were to determine system transient characteristics and to assess code predictive capabilities for design basis large-break loss-of-coolant accidents in pressurized water reactors (PWRs). This experiment simulated a double-ended offset shear of one inlet pipe in a four-loop PWR and was initiated from conditions representative of licensing limits in a PWR. Other boundary conditions for the simulation were loss of offsite power, rapid primary coolant pump coast down, and United Kingdom minimum safeguard emergency core coolant injection rates. The nuclear fuel rods were not pressurized. The transient was initiated by opening the quick-opening blowdown valves in the broken loop hot and cold legs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

LOFT/LP-FP-1B, Loss of Fluid Test, Fission Product Release Experiment

International Nuclear Information System (INIS)

1989-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The seventh OECD LOFT experiment was conducted on 19 December 1984. It was the first of the two experiments to be performed in the LOFT facility with intentional release of fission products. Its objectives were to obtain data on fission product release from the fuel-cladding gap into vapor and reflood water and to collect data on transport of these fission products through and out of the reactor coolant system. The experiment was initiated by a reactor scram with one second delayed opening of the quick-opening blowdown valves. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

Results of water corrosion in static cell tests representing multi-metal assemblies in the hydraulic circuits of Tore supra

Energy Technology Data Exchange (ETDEWEB)

Lipa, M. [CEA/DSM/DRFC Centre de Cadarache, 13 - Saint-Paul lez Durance (France); Blanchet, J.; Cellier, F. [Framatome, 71 - Saint Marcel (France). Centre Technique

2007-07-01

Following experiences obtained with steam generator tubes of nuclear power plants, a cooling water quality of AVT (all volatile treatment) has been defined based on demineralised water with adjustment of the pH value to about 9.0/7.0 (25 C/200 C) by addiction of ammoniac, and hydrazine in order to absorb oxygen dissolved in water. At that time, a simplified water corrosion test program has been performed using static (no circulation) test cell samples made of above mentioned TS metal combinations. All test cell samples, prepared and filled with AVT water, were performed at 280 C and 65 bars in an autoclave during 3000 hours. The test cell water temperature has been chosen to be sufficient above the TS component working temperature, in order to accelerate an eventual corrosion process. Generally all above mentioned metal combinations survived the test campaign without stress corrosion cracking, with the exception of the memory metal junction (creep in Cu) and the bellows made of St-St 316L and Inconel 625 while 316 Ti bellows survived. In contrary to the vacuum brazed Cu-LSTP to St-St samples, some of flame brazed Cu to St-St samples failed either in the braze joint or in the copper structure itself. For comparison, a spot weld of an inflated 316L panel sample, filled voluntary with a caustic solution of pH 11.5 (25 C), failed after 90 h of testing (intergranular cracking at the spot weld), while an identical sample containing AVT water of pH 9.0 (25 C) survived without damage. The results of these tests, performed during 1986 and 1997, have never been published and therefore are presented more in detail in this paper since corrosion in hydraulic circuits is also an issue of ITER. Up to day, the TS cooling water plant operates with an above mentioned water treatment and no water leaks have been detected on in-vessel components originating from water corrosion at high temperature and high pressure. (orig.)

Results of water corrosion in static cell tests representing multi-metal assemblies in the hydraulic circuits of Tore supra

International Nuclear Information System (INIS)

Lipa, M.; Blanchet, J.; Cellier, F.

2007-01-01

Following experiences obtained with steam generator tubes of nuclear power plants, a cooling water quality of AVT (all volatile treatment) has been defined based on demineralised water with adjustment of the pH value to about 9.0/7.0 (25 C/200 C) by addiction of ammoniac, and hydrazine in order to absorb oxygen dissolved in water. At that time, a simplified water corrosion test program has been performed using static (no circulation) test cell samples made of above mentioned TS metal combinations. All test cell samples, prepared and filled with AVT water, were performed at 280 C and 65 bars in an autoclave during 3000 hours. The test cell water temperature has been chosen to be sufficient above the TS component working temperature, in order to accelerate an eventual corrosion process. Generally all above mentioned metal combinations survived the test campaign without stress corrosion cracking, with the exception of the memory metal junction (creep in Cu) and the bellows made of St-St 316L and Inconel 625 while 316 Ti bellows survived. In contrary to the vacuum brazed Cu-LSTP to St-St samples, some of flame brazed Cu to St-St samples failed either in the braze joint or in the copper structure itself. For comparison, a spot weld of an inflated 316L panel sample, filled voluntary with a caustic solution of pH 11.5 (25 C), failed after 90 h of testing (intergranular cracking at the spot weld), while an identical sample containing AVT water of pH 9.0 (25 C) survived without damage. The results of these tests, performed during 1986 and 1997, have never been published and therefore are presented more in detail in this paper since corrosion in hydraulic circuits is also an issue of ITER. Up to day, the TS cooling water plant operates with an above mentioned water treatment and no water leaks have been detected on in-vessel components originating from water corrosion at high temperature and high pressure. (orig.)

Operating Experience Review of Tritium-in-Water Monitors

Energy Technology Data Exchange (ETDEWEB)

S. A. Bruyere; L. C. Cadwallader

2011-09-01

Monitoring tritium facility and fusion experiment effluent streams is an environmental safety requirement. This paper presents data on the operating experience of a solid scintillant monitor for tritium in effluent water. Operating experiences were used to calculate an average monitor failure rate of 4E-05/hour for failure to function. Maintenance experiences were examined to find the active repair time for this type of monitor, which varied from 22 minutes for filter replacement to 11 days of downtime while waiting for spare parts to arrive on site. These data support planning for monitor use; the number of monitors needed, allocating technician time for maintenance, inventories of spare parts, and other issues.

Seabed gallery intakes: Investigation of the water pretreatment effectiveness of the active layer using a long-term column experiment

KAUST Repository

Dehwah, Abdullah; Missimer, Thomas M.

2017-01-01

) in improving raw water quality was conducted by using a long-term bench-scale columns experiment. Two different media types, silica and carbonate sand, were tested in 1 m columns to evaluate the effectiveness of media type in terms of algae, bacteria, Natural

Code of practice for the temporary diversion of water from hydrostatic testing of pipelines: Water Act - Water (Ministerial) Regulation No. 205

International Nuclear Information System (INIS)

1999-01-01

A series of administrative procedures are described in this regulation regarding the temporary diversion of water for hydrostatic testing of pipelines in Alberta. These include the notification and timing of the release of water, where the test will occur and the possible effect of water release on the environment. The requirements of a hydrostatic test plan are described in point form. In accordance with the regulation, a person responsible for the testing must ensure that the testing does not cause a significant adverse effect on the aquatic environment or on any household user. Should this occur, all adverse effects on the aquatic environment should be fully mitigated. The reporting schedule and record keeping information are also covered. 2 tabs., 1 figs

Computer-Adaptive Testing: Implications for Students' Achievement, Motivation, Engagement, and Subjective Test Experience

Science.gov (United States)

Martin, Andrew J.; Lazendic, Goran

2018-01-01

The present study investigated the implications of computer-adaptive testing (operationalized by way of multistage adaptive testing; MAT) and "conventional" fixed order computer testing for various test-relevant outcomes in numeracy, including achievement, test-relevant motivation and engagement, and subjective test experience. It did so…

Immunotoxicity and genotoxicity testing for in-flight experiments under microgravity

Science.gov (United States)

Hansen, Peter-Diedrich; Hansen, Peter-Diedrich; Unruh, Eckehardt

choice of phagocytes from invertebrates is justified to study the universal validity of innate immune responses. The TRIPLELUX-B Experiment contributes to risk assessment concerning immunotoxicity under space flight conditions. The components of the phagocytosis test system for the BIOLAB are now established and the technical realization of the TRIPLELUX- B Experiment is in final progess. The components of the fully automated AEC (Advanced Experimental Containment) will be demonstrated in the poster. There will be two AECs for reference measurements at 1xg and 0xg. The AEC of the TRIPLELUX-B experiment will contribute to a real time operational monitoring for immunotoxicity testing on earth. The AEC will allow "real time monitoring" providing automated observations of immunotoxicity in coastal and inland waters.

Simulation of power maneuvering experiment of MASLWR test facility by MARS-KS code

Energy Technology Data Exchange (ETDEWEB)

Park, Ju Yeop [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-10-15

In this ICSP, experimental data obtained from MASLWR (Mulit-Application Small Light Water Reactor) test facility located at Oregon state university in the US have been simulated by various thermal-hydraulic codes of each participant of the ICSP and compared among others. MASLWR test facility is a mock-up of a passive integral type reactor equipped with helical coil steam generator. Since SMART reactor which is currently being developed in Korea also adopts a helical coil steam generator, Korea Institute of Nuclear Safety (KINS) has joined this ICSP to assess the applicability of a domestic regulatory audit thermal-hydraulic code (i. e. MARS-KS code) for the SMART reactor including wall-to-fluid heat transfer model modification based on independent international experiment data. In the ICSP, two types of transient experiments have been focused and they are 1) loss of feedwater transient with subsequent ADS operation and long term cooling (SP-2) and normal operating conditions at different power levels. In the present study, KINS simulation result by the MARS-KS code (KS-002 version) for the SP-3 experiment is presented in detail and conclusion on MARS-KS code performance drawn through this simulation is described. Performance of the MARS-KS code is evaluated through the simulation of the power maneuvering experiment of the MASLWR test facility. Steady run shows the helical coil specific heat transfer model of the code is reasonable. However, identified discrepancy of the primary mass flowrate at transient run shows code performance for pressure drop needs to be improved considering sensitivity of the flowrate to the pressure drop at natural circulation.

Large-scale experiments on aerosol behavior in light water reactor containments

International Nuclear Information System (INIS)

Schock, W.; Bunz, H.; Adams, R.E.; Tobias, M.L.; Rahn, F.J.

1988-01-01

Recently, three large-scale experimental programs were carried out dealing with the behavior of aerosols during core-melt accidents in light water reactors (LWRs). In the Nuclear Safety Pilot Plant (NSPP) program, the principal behaviors of different insoluble aerosols and of mixed aerosols were measured in dry air atmospheres and in condensing steam-air atmospheres contained in a 38-m/sup 3/ steel vessel. The Demonstration of Nuclear Aerosol Behavior (DEMONA) program used a 640-m/sup 3/ concrete containment model to simulate typical accident sequence conditions, and measured the behavior of different insoluble aerosols and mixed aerosols in condensing and transient atmospheric conditions. Part of the LWR Aerosol Containment Experiments (LACE) program was also devoted to aerosol behavior in containment; and 852-m/sup 3/ steel vessel was used, and the aerosols were composed of mixtures of insoluble and soluble species. The results of these experiments provide a suitable data base for validation of aerosol behavior codes. Fundamental insight into details of aerosol behavior in condensing environments has been gained through the results of the NSPP tests. Code comparisons have been and are being performed in the DEMONA and LACE experiments

Benchmark experiments to test plutonium and stainless steel cross sections. Topical report

International Nuclear Information System (INIS)

Jenquin, U.P.; Bierman, S.R.

1978-06-01

The Nuclear Regulatory Commission (NRC) commissioned Battelle, Pacific Northwest Laboratory (PNL) to ascertain the accuracy of the neutron cross sections for the isotopes of plutonium and the constituents of stainless steel and determine if improvements can be made in application to criticality safety analysis. NRC's particular area of interest is in the transportation of light-water reactor spent fuel assemblies. The project was divided into two tasks. The first task was to define a set of integral experimental measurements (benchmarks). The second task is to use these benchmarks in neutronics calculations such that the accuracy of ENDF/B-IV plutonium and stainless steel cross sections can be assessed. The results of the first task are given in this report. A set of integral experiments most pertinent to testing the cross sections has been identified and the code input data for calculating each experiment has been developed

The Sequential Probability Ratio Test: An efficient alternative to exact binomial testing for Clean Water Act 303(d) evaluation.

Science.gov (United States)

Chen, Connie; Gribble, Matthew O; Bartroff, Jay; Bay, Steven M; Goldstein, Larry

2017-05-01

The United States's Clean Water Act stipulates in section 303(d) that states must identify impaired water bodies for which total maximum daily loads (TMDLs) of pollution inputs into water bodies are developed. Decision-making procedures about how to list, or delist, water bodies as impaired, or not, per Clean Water Act 303(d) differ across states. In states such as California, whether or not a particular monitoring sample suggests that water quality is impaired can be regarded as a binary outcome variable, and California's current regulatory framework invokes a version of the exact binomial test to consolidate evidence across samples and assess whether the overall water body complies with the Clean Water Act. Here, we contrast the performance of California's exact binomial test with one potential alternative, the Sequential Probability Ratio Test (SPRT). The SPRT uses a sequential testing framework, testing samples as they become available and evaluating evidence as it emerges, rather than measuring all the samples and calculating a test statistic at the end of the data collection process. Through simulations and theoretical derivations, we demonstrate that the SPRT on average requires fewer samples to be measured to have comparable Type I and Type II error rates as the current fixed-sample binomial test. Policymakers might consider efficient alternatives such as SPRT to current procedure. Copyright © 2017 Elsevier Ltd. All rights reserved.

Water chemistry management during hot functional test

International Nuclear Information System (INIS)

Yokoyama, Jiro; Kanda, Tomio; Kagawa, Masaru

1988-01-01

To reduce radiation exposure in light water reactor, it is important decrease radioactive corrosion product which is a radiation source. One of the countermeasures is to improve water quality during plant trial operation to form a stable oxide film and to minimize metal release to the coolant at the beginning of commercial operation. This study reviews the optimum water quality conditions to form a chromium rich oxide film during hot functional test (HFT) that is thought to be stable under the PWR condition and reduce the release of Ni that is the source of Co-58, the main radiation source of exposure. (author)

Automated corrosion fatigue crack growth testing in pressurized water environments

International Nuclear Information System (INIS)

Ceschini, L.J.; Liaw, P.K.; Rudd, G.E.; Logsdon, W.A.

1984-01-01

This paper describes in detail a novel approach to construct a test facility for developing corrosion fatigue crack growth rate (FCGR) properties in aggressive environments. The environment studied is that of a pressurized water reactor (PWR) at 288 0 C (550 0 F) and 13.8 MPa (200 psig). To expedite data generation, each chamber was designed to accommodate two test specimens. A common water recirculation and pressurization system was employed to service two test chambers. Thus, four fatigue crack propagation rate tests could be conducted simultaneously in the pressurized water environment. The data analysis was automated to minimize the typically high labor costs associated with corrosion fatigue crack propagation testing. Verification FCGR tests conducted on an ASTM A469 rotor steel in a room temperature air environment as well as actual PWR environment FCGR tests performed on an ASTM A533 Grade B Class 2 pressure vessel steel demonstrated that the dual specimen test facility is an excellent system for developing the FCGR properties of materials in adverse environments

Critical heat flux experiments for high conversion light water reactor, (3)

International Nuclear Information System (INIS)

Iwamura, Takamichi; Okubo, Tsutomu; Suemura, Takayuki; Hiraga, Fujio; Murao, Yoshio

1990-03-01

As a part of the thermal-hydraulic feasibility study of a high conversion light water reactor (HCLWR), critical heat flux (CHF) experiments were performed using triangular array rod bundles under steady-state and flow reduction transient conditions. The geometries of test sections were: rod outer diameter 9.5 mm, number of rods 4∼7, heated length 0.5∼1.0 m, and pitch to diameter ratio (P/D) 1.126∼1.2. The simulated fuel rod was a stainless steel tube and uniformly heated electrically with direct current. In the steady-state tests, pressures ranged: 1.0∼3.9 Mpa, mass velocities: 460∼4270 kg/s·m 2 , and exit qualities: 0.02∼0.35. In the transient tests, the times to CHF detection ranged from 0.5 to 25.4 s. The steady-state CHF's for the 4-rod test sections were higher than those for the 7-rod test sections with respect to the bundle averaged flow conditions. The measured CHF's increased with decreasing the heated length and decreased with decreasing the P/D. Based on the local flow conditions obtained with the subchannel analysis code COBRA-IV-I, KfK correlation agreed with the CHF data within 20 %, while WSC-2, EPRI-B and W, EPRI-Columbia and Kattor correlations failed to give satisfactory agreements. Under flow reduction rates less than 6 %/s, no significant difference in the onset conditions of DNB (departure from nucleate boiling) was recognized between the steady-state and transient conditions. At flow reduction rates higher than 6 %/s, on the other hand, the DNB occurred earlier than the DNB time predicted with the steady-state experiments. (author)

HACCP and water safety plans in Icelandic water supply: preliminary evaluation of experience.

Science.gov (United States)

Gunnarsdóttir, María J; Gissurarson, Loftur R

2008-09-01

Icelandic waterworks first began implementing hazard analysis and critical control points (HACCP) as a preventive approach for water safety management in 1997. Since then implementation has been ongoing and currently about 68% of the Icelandic population enjoy drinking water from waterworks with a water safety plan based on HACCP. Preliminary evaluation of the success of HACCP implementation was undertaken in association with some of the waterworks that had implemented HACCP. The evaluation revealed that compliance with drinking water quality standards improved considerably following the implementation of HACCP. In response to their findings, waterworks implemented a large number of corrective actions to improve water safety. The study revealed some limitations for some, but not all, waterworks in relation to inadequate external and internal auditing and a lack of oversight by health authorities. Future studies should entail a more comprehensive study of the experience with the use of HACCP with the purpose of developing tools to promote continuing success.

Plant experience with temporary reverse osmosis makeup water systems

International Nuclear Information System (INIS)

Polidoroff, C.

1986-01-01

Pacific Gas and Electric (PG and E) Company's Diablo Canyon Power Plant (DCPP), which is located on California's central coast, has access to three sources of raw water: creek water, well water, and seawater. Creek and well water are DCPP's primary sources of raw water; however, because their supply is limited, these sources are supplemented with seawater. The purpose of this paper is to discuss the temporary, rental, reverse osmosis systems used by PG and E to process DCPP's raw water into water suitable for plant makeup. This paper addresses the following issues: the selection of reverse osmosis over alternative water processing technologies; the decision to use vendor-operated temporary, rental, reverse osmosis equipment versus permanent PG and E-owned and -operated equipment; the performance of DCPP's rental reverse osmosis systems; and, the lessons learned from DCPP's reverse osmosis system rental experience that might be useful to other plants considering renting similar equipment

Development and Testing of Infrared Water Current Meter | Ezenne ...

African Journals Online (AJOL)

Continuous monitoring of the river flow is essential for assessing water availability. River flow velocity is crucial to simulate discharge hydrographs of water in the hydrological system.This study developed a digital water current meter with infrared. The infrared current meter was tested using Ebonyi River at Obollo-Etiti and ...

Preliminary results of ecotoxicological assessment of an Acid Mine Drainage (AMD) passive treatment system testing water quality of depurated lixiviates

OpenAIRE

Miguel Sarmiento, Aguasanta; Bonnail, Estefanía; Nieto Liñán, José Miguel; Valls Casillas, Tomás Ángel del

2017-01-01

The current work reports on the preliminary results of a toxicity test using screening experiments to check the efficiency of an innovative passive treatment plant designed for acid mine drainage purification. Bioassays took place with water samples before and after the treatment system and in the river, once treated water is discharged. Due to the high toxicity of the water collected at the mouth of the mine (before the treatment plant), the bioassay was designed and developed with respect t...

Investigation of water content in primary upper shield of high temperature engineering test reactor (HTTR)

International Nuclear Information System (INIS)

Sumita, Junya; Sawa, Kazuhiro; Mogi, Haruyoshi; Itahashi, Shuuji; Kitami, Toshiyuki; Akutu, Youichi; Fuchita, Yasuhiro; Kawaguchi, Toru; Moriya, Masahiro

1999-09-01

A primary upper shield of the High Temperature Engineering Test Reactor (HTTR) is composed of concrete (grout) which is packed into iron frames. The main function of the primary upper shield is to attenuate neutron and gamma ray from the core, that leads to satisfy dose equivalent rate limit of operating floor and stand-pipe room. Water content in the concrete is one of the most important things because it strongly affects neutron-shielding ability. Then, we carried out out-of-pile experiments to investigate relationship between temperature and water content in the concrete. Based on the experimental results, a hydrolysis-diffusion model was developed to investigate water release behavior from the concrete. The model showed that water content used for shielding design in the primary upper shield of the HTTR will be maintained if temperature during operating life is under 110degC. (author)

Development flight tests of JetStar LFC leading-edge flight test experiment

Science.gov (United States)

Fisher, David F.; Fischer, Michael C.

1987-01-01

The overall objective of the flight tests on the JetStar aircraft was to demonstrate the effectiveness and reliability of laminar flow control under representative flight conditions. One specific objective was to obtain laminar flow on the JetStar leading-edge test articles for the design and off-design conditions. Another specific objective was to obtain operational experience on a Laminar Flow Control (LFC) leading-edge system in a simulated airline service. This included operational experience with cleaning requirements, the effect of clogging, possible foreign object damage, erosion, and the effects of ice particle and cloud encounters. Results are summarized.

The testing of a steam-water separating device used for vertical steam generators

International Nuclear Information System (INIS)

Ding Xunshen; Cui Baoyuan; Xue Yunkui; Liu Shixun

1989-01-01

The air-water screening tests of a steam-water separating device used for vertical steam generators at low pressure are introduced. The article puts emphasis on the qualification test of the steam-water separating device at hot conditions in a high temperature and pressure water test rig. The performance of the comprehensive test of the steam-water separating device indicates that the humidity of the steam at the drier exit is much less than the specified amount of 0.25%

Core-concrete interactions with overlying water pools. The WETCOR-1 test

Energy Technology Data Exchange (ETDEWEB)

Blose, R.E. [Ktech Corp., Albuquerque, NM (United States); Powers, D.A.; Copus, E.R.; Brockmann, J.E.; Simpson, R.B.; Lucero, D.A. [Sandia National Labs., Albuquerque, NM (United States)

1993-11-01

The WETCOR-1 test of simultaneous interactions of a high-temperature melt with water and a limestone/common-sand concrete is described. The test used a 34.1-kg melt of 76.8 w/o Al{sub 2}O{sub 3}, 16.9 w/o CaO, and 4.0 w/o SiO{sub 2} heated by induction using tungsten susceptors. Once quasi-steady attack on concrete by the melt was established, an attempt was made to quench the melt at 1850 K with 295 K water flowing at 57 liters per minute. Net power into the melt at the time of water addition was 0.61 {plus_minus} 0.19 W/cm{sup 3}. The test configuration used in the WETCOR-1 test was designed to delay melt freezing to the walls of the test fixture. This was done to test hypotheses concerning the inherent stability of crust formation when high-temperature melts are exposed to water. No instability in crust formation was observed. The flux of heat through the crust to the water pool maintained over the melt in the test was found to be 0.52 {plus_minus} 0.13 MW/m{sup 2}. Solidified crusts were found to attenuate aerosol emissions during the melt concrete interactions by factors of 1.3 to 3.5. The combination of a solidified crust and a 30-cm deep subcooled water pool was found to attenuate aerosol emissions by factors of 3 to 15.

Modeling and experiments on the drive characteristics of high-strength water hydraulic artificial muscles

Science.gov (United States)

Zhang, Zengmeng; Hou, Jiaoyi; Ning, Dayong; Gong, Xiaofeng; Gong, Yongjun

2017-05-01

Fluidic artificial muscles are popular in robotics and function as biomimetic actuators. Their pneumatic version has been widely investigated. A novel water hydraulic artificial muscle (WHAM) with high strength is developed in this study. WHAMs can be applied to underwater manipulators widely used in ocean development because of their environment-friendly characteristics, high force-to-weight ratio, and good bio-imitability. Therefore, the strength of WHAMs has been improved to fit the requirements of underwater environments and the work pressure of water hydraulic components. However, understanding the mechanical behaviors of WHAMs is necessary because WHAMs use work media and pressure control that are different from those used by pneumatic artificial muscles. This paper presents the static and dynamic characteristics of the WHAM system, including the water hydraulic pressure control circuit. A test system is designed and built to analyze the drive characteristics of the developed WHAM. The theoretical relationships among the amount of contraction, pressure, and output drawing force of the WHAM are tested and verified. A linearized transfer function is proposed, and the dynamic characteristics of the WHAM are investigated through simulation and inertia load experiments. Simulation results agree with the experimental results and show that the proposed model can be applied to the control of WHAM actuators.

To What Extent is Drinking Water Tested in Sub-Saharan Africa? A Comparative Analysis of Regulated Water Quality Monitoring

OpenAIRE

Rachel Peletz; Emily Kumpel; Mateyo Bonham; Zarah Rahman; Ranjiv Khush

2016-01-01

Water quality information is important for guiding water safety management and preventing water-related diseases. To assess the current status of regulated water quality monitoring in sub-Saharan Africa, we evaluated testing programs for fecal contamination in 72 institutions (water suppliers and public health agencies) across 10 countries. Data were collected through written surveys, in-person interviews, and analysis of microbial water quality testing levels. Though most institutions did no...

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)

Water, Forests, People: The Swedish Experience in Building Resilient Landscapes.

Science.gov (United States)

Eriksson, Mats; Samuelson, Lotta; Jägrud, Linnéa; Mattsson, Eskil; Celander, Thorsten; Malmer, Anders; Bengtsson, Klas; Johansson, Olof; Schaaf, Nicolai; Svending, Ola; Tengberg, Anna

2018-05-21

A growing world population and rapid expansion of cities increase the pressure on basic resources such as water, food and energy. To safeguard the provision of these resources, restoration and sustainable management of landscapes is pivotal, including sustainable forest and water management. Sustainable forest management includes forest conservation, restoration, forestry and agroforestry practices. Interlinkages between forests and water are fundamental to moderate water budgets, stabilize runoff, reduce erosion and improve biodiversity and water quality. Sweden has gained substantial experience in sustainable forest management in the past century. Through significant restoration efforts, a largely depleted Swedish forest has transformed into a well-managed production forest within a century, leading to sustainable economic growth through the provision of forest products. More recently, ecosystem services are also included in management decisions. Such a transformation depends on broad stakeholder dialog, combined with an enabling institutional and policy environment. Based on seminars and workshops with a wide range of key stakeholders managing Sweden's forests and waters, this article draws lessons from the history of forest management in Sweden. These lessons are particularly relevant for countries in the Global South that currently experience similar challenges in forest and landscape management. The authors argue that an integrated landscape approach involving a broad array of sectors and stakeholders is needed to achieve sustainable forest and water management. Sustainable landscape management-integrating water, agriculture and forests-is imperative to achieving resilient socio-economic systems and landscapes.

Operating experience of natural circulation core cooling in boiling water reactors

International Nuclear Information System (INIS)

Kullberg, C.; Jones, K.; Heath, C.

1993-01-01

General Electric (GE) has proposed an advanced boiling water reactor, the Simplified Boiling Water Reactor (SBWR), which will utilize passive, gravity-driven safety systems for emergency core coolant injection. The SBWR design includes no recirculation loops or recirculation pumps. Therefore the SBWR will operate in a natural circulation (NC) mode at full power conditions. This design poses some concerns relative to stability during startup, shutdown, and at power conditions. As a consequence, the NRC has directed personnel at several national labs to help investigate SBWR stability issues. This paper will focus on some of the preliminary findings made at the INEL. Because of the broad range of stability issues this paper will mainly focus on potential geysering instabilities during startup. The two NC designs examined in detail are the US Humboldt Bay Unit 3 BWR-1 plant and Dodewaard plant in the Netherlands. The objective of this paper will be to review operating experience of these two plants and evaluate their relevance to planned SBWR operational procedures. For completeness, experimental work with early natural circulation GE test facilities will also be briefly discussed

The application and testing of diatom-based indices of stream water ...

African Journals Online (AJOL)

The application and testing of diatom-based indices of stream water quality in Chinhoyi Town, Zimbabwe. ... PROMOTING ACCESS TO AFRICAN RESEARCH ... test the applicability of foreign diatom-based water quality assessment indices to ...

Heavy water critical experiments on plutonium lattice

International Nuclear Information System (INIS)

Miyawaki, Yoshio; Shiba, Kiminori

1975-06-01

This report is the summary of physics study on plutonium lattice made in Heavy Water Critical Experiment Section of PNC. By using Deuterium Critical Assembly, physics study on plutonium lattice has been carried out since 1972. Experiments on following items were performed in a core having 22.5 cm square lattice pitch. (1) Material buckling (2) Lattice parameters (3) Local power distribution factor (4) Gross flux distribution in two region core (5) Control rod worth. Experimental results were compared with theoretical ones calculated by METHUSELAH II code. It is concluded from this study that calculation by METHUSELAH II code has acceptable accuracy in the prediction on plutonium lattice. (author)

A comparison between burn-out data for 19-rod cluster test-sections cooled by Freon-12 at 155 lb/in2 (abs), and by water at 1000 lb/in2 in vertical upflow

International Nuclear Information System (INIS)

Stevens, G.F.; Wood, R.W.

1966-01-01

Previous experiments on the Winfrith Freon Rig have produced scaling factors which relate these Freon experiments to the corresponding experiments in water with an accuracy of about 10%. It has also been found that the Freon rig is accurate, economical and easy to use. The scaling factors so obtained have now been tested against data for 19-rod clusters which had previously been tested at Columbia University. This report presents the results of the rod cluster tests in which comparison is made between Freon-12 and water for three test-sections which differ in the means of spacing the individual rods. All the test-sections were heated uniformly with respect to length, but had a radial flux depression of nominally 0.70/1.0. The results provide strong evidence that the scaling factor method using Freon-12 at 155 lb/in 2 (abs) is a useful technique for predicting the behaviour at burn-out of complicated test-sections cooled by boiling water at 1000 lb/in 2 with only one-eighteenth of the power required for the water experiment. In particular, the Freon tests reproduce closely the relative burn-out powers previously measured in water. It has also been found that repeated rebuilding of a nominally unchanged cluster from the same components can produce burn-out powers differing by ± 6%. This new result illustrates the power and value of the Freon technique. (author)

Acceptance Test Report for the high pressure water jet system canister cleaning fixture

Energy Technology Data Exchange (ETDEWEB)

Burdin, J.R.

1995-10-25

This Acceptance Test confirmed the test results and recommendations, documented in WHC-SD-SNF-DTR-001, Rev. 0 Development Test Report for the High Pressure Water Jet System Nozzles, for decontaminating empty fuel canisters in KE-Basin. Optimum water pressure, water flow rate, nozzle size and overall configuration were tested

Acceptance Test Report for the high pressure water jet system canister cleaning fixture

International Nuclear Information System (INIS)

Burdin, J.R.

1995-01-01

This Acceptance Test confirmed the test results and recommendations, documented in WHC-SD-SNF-DTR-001, Rev. 0 Development Test Report for the High Pressure Water Jet System Nozzles, for decontaminating empty fuel canisters in KE-Basin. Optimum water pressure, water flow rate, nozzle size and overall configuration were tested

Manufacturing and testing experience for FFTF major safety related components

International Nuclear Information System (INIS)

Peckinpaugh, C.L.

1976-01-01

Experience with FFTF Heat Transport System components during design, manufacturing, and prototype testing is dscussed. Specifically the special design features and the results of the testing performed to assure that the designs provide for safe operation are outlined. Particular emphasis is placed on the full size prototype testing programs and the valuable experience gained

HIV testing and counseling: test providers' experiences of best practices.

Science.gov (United States)

Myers, Ted; Worthington, Catherine; Haubrich, Dennis J; Ryder, Karen; Calzavara, Liviana

2003-08-01

Although education is central to HIV testing and counseling, little is known about the educational processes within the testing experience. This study investigated test providers' understandings of testing and counseling best practices. Interviews with a purposive sample of 24 test providers were thematically analyzed. Analysis revealed five best practices specific to HIV education and public health--ensuring information and education for HIV risk reduction, individualization of risk assessment, ensuring test results are given in person, providing information and referrals, and facilitating partner notification--and six practices not specific to HIV counseling relationship building. The latter were building trust and rapport; maintaining professional boundaries; ensuring a comfortable, safe environment; ensuring confidentiality; imparting nonjudgmntal attitude; and self-determination. The identified best practices demonstrated remarkable consistency across respondent subgroups. Although counseling was seen as largely educational and with a preventive focus, it included individualized messages based on assessments of risk, knowledge, and social and cultural characteristics.

Investigation of CTBT OSI Radionuclide Techniques at the DILUTED WATERS Nuclear Test Site

Energy Technology Data Exchange (ETDEWEB)

Baciak, James E.; Milbrath, Brian D.; Detwiler, Rebecca S.; Kirkham, Randy R.; Keillor, Martin E.; Lepel, Elwood A.; Seifert, Allen; Emer, Dudley; Floyd, Michael

2012-11-01

Under the Comprehensive Nuclear-Test-Ban Treaty (CTBT), a verification regime that includes the ability to conduct an On-Site Inspection (OSI) will be established. The Treaty allows for an OSI to include many techniques, including the radionuclide techniques of gamma radiation surveying and spectrometry and environmental sampling and analysis. Such radioactivity detection techniques can provide the “smoking gun” evidence that a nuclear test has occurred through the detection and quantification of indicative recent fission products. An OSI faces restrictions in time and manpower, as dictated by the Treaty; not to mention possible logistics difficulties due to the location and climate of the suspected explosion site. It is thus necessary to have a good understanding of the possible source term an OSI will encounter and the proper techniques that will be necessary for an effective OSI regime. One of the challenges during an OSI is to locate radioactive debris that has escaped an underground nuclear explosion (UNE) and settled on the surface near and downwind of ground zero. To support the understanding and selection of sampling and survey techniques for use in an OSI, we are currently designing an experiment, the Particulate Release Experiment (PRex), to simulate a small-scale vent from an underground nuclear explosion. PRex will occur at the Nevada National Security Site (NNSS). The project is conducted under the National Center for Nuclear Security (NCNS) funded by the National Nuclear Security Agency (NNSA). Prior to the release experiment, scheduled for Spring of 2013, the project scheduled a number of activities at the NNSS to prepare for the release experiment as well as to utilize the nuclear testing past of the NNSS for the development of OSI techniques for CTBT. One such activity—the focus of this report—was a survey and sampling campaign at the site of an old UNE that vented: DILUTED WATERS. Activities at DILUTED WATERS included vehicle-based survey

Evaluation of annual corrosion tests for aggressive water

Science.gov (United States)

Dubová, V.; Ilavský, J.; Barloková, D.

2011-12-01

Internal corrosion has a significant effect on the useful life of pipes, the hydraulic conditions of a distribution system and the quality of the water transported. All water is corrosive under some conditions, and the level of this corrosion depends on the physical and chemical properties of the water and properties of the pipe material. Galvanic treatment is an innovation for protecting against corrosion, and this method is also suitable for removal of water stone too. This method consists of the electrogalvanic principle, which is generated by the flowing of water between a zinc anode and the cupro-alloy cover of a column. This article presents experimental corrosion tests at water resource Pernek (This water resource-well marked as HL-1 is close to the Pernek of village), where the device is operating based on this principle.

'Not the swab!' Young men's experiences with STI testing.

Science.gov (United States)

Shoveller, Jean A; Knight, Rod; Johnson, Joy; Oliffe, John L; Goldenberg, Shira

2010-01-01

In Canada, STI rates are high and rising, especially amongst young men. Meanwhile, the needs of young men regarding STI testing services are poorly understood, as are the socio-cultural and structural factors that influence young men's sexual health-seeking behaviours. To better understand this phenomenon, we draw on interviews with 45 men (ages 15-25) from British Columbia, Canada. Our research reveals how structural forces (e.g. STI testing procedures) interact with socio-cultural factors (e.g. perceptions of masculinities and feminities) to shape young men's experiences with STI testing. STI testing was characterised as both a potentially sexualised experience (e.g. fears of getting an erection during genital examinations), and as a process where young men experience multiple vulnerabilities associated with exposing the male body in clinical service sites. In response, participants drew on dominant ideals of masculinity to reaffirm their predominately hetero-normative gender identities. Despite growing up in an era where sexual health promotion efforts have been undertaken, participants did not feel they had permission to engage in discussions with other men about sexual health issues. Attending to young men's perspectives on STI testing represents a starting point in reforming our approaches to addressing how socio-cultural and structural factors shape these experiences.

Simulation of power maneuvering experiment of MASLWR test facility by MARS-KS code

Energy Technology Data Exchange (ETDEWEB)

Park, Ju Yeop [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-10-15

In the present study, KINS simulation result by the MARS-KS code (KS-002 version) for the SP-3 experiment is presented in detail and conclusion on MARS-KS code performance drawn through this simulation is described. Performance of the MARS-KS code is evaluated through the simulation of the power maneuvering experiment of the MASLWR test facility. Steady run shows the helical coil specific heat transfer model of the code is reasonable. However, identified discrepancy of the primary mass flowrate at transient run shows code performance for pressure drop needs to be improved considering sensitivity of the flowrate to the pressure drop at natural circulation. Since 2009, IAEA has conducted a research program entitled as ICSP (International Collaborative Standard Problem) on integral PWR design to evaluate current the state of the art of thermal-hydraulic code in simulating natural circulation flow within integral type reactor. In this ICSP, experimental data obtained from MASLWR (Multi-Application Small Light Water Reactor) test facility located at Oregon state university in the US have been simulated by various thermal-hydraulic codes of each participant of the ICSP and compared among others. MASLWR test facility is a mock-up of a passive integral type reactor equipped with helical coil steam generator. Since SMART reactor which is currently being developed in Korea also adopts a helical coil steam generator, Korea Institute of Nuclear Safety (KINS) has joined this ICSP to assess the applicability of a domestic regulatory audit thermal-hydraulic code (i. e. MARS-KS code) for the SMART reactor including wall-to-fluid heat transfer model modification based on independent international experiment data. In the ICSP, two types of transient experiments have been focused and they are loss of feedwater transient with subsequent ADS operation and long term cooling (SP-2) and normal operating conditions at different power levels (SP-3)

Space Shuttle Boundary Layer Transition Flight Experiment Ground Testing Overview

Science.gov (United States)

Berger, Karen T.; Anderson, Brian P.; Campbell, Charles H.

2014-01-01

In support of the Boundary Layer Transition (BLT) Flight Experiment (FE) Project in which a manufactured protuberance tile was installed on the port wing of Space Shuttle Orbiter Discovery for STS-119, STS- 128, STS-131 and STS-133 as well as Space Shuttle Orbiter Endeavour for STS-134, a significant ground test campaign was completed. The primary goals of the test campaign were to provide ground test data to support the planning and safety certification efforts required to fly the flight experiment as well as validation for the collected flight data. These test included Arcjet testing of the tile protuberance, aerothermal testing to determine the boundary layer transition behavior and resultant surface heating and planar laser induced fluorescence (PLIF) testing in order to gain a better understanding of the flow field characteristics associated with the flight experiment. This paper provides an overview of the BLT FE Project ground testing. High-level overviews of the facilities, models, test techniques and data are presented, along with a summary of the insights gained from each test.

Assessment of soil phosphorus tests for situations in Australia where reactive phosphate rock and water-soluble fertilizers are used

International Nuclear Information System (INIS)

McLaughlin, M.J.

2002-01-01

A selection of commonly used soil P tests, which included anion and cation exchange resin membranes, were compared in a glasshouse experiment using subterranean clover, and evaluated in the field at 19 sites from the National Reactive Phosphate Rock Project in 1993 and at 6 sites in 1995. The ability of the soil P tests to predict plant response was used to evaluate the tests. In the glasshouse experiment, the resin test was less effective than the Bray and Colwell tests in its ability to assess the level of plant available P from the different fertilizer treatments. Seventy one percent of the variation in total P content of the subterranean clover tops was explained by resin extractable P values, whereas the Colwell procedure accounted for 81% and the Bray 1 procedure accounted for 78%. Water and CaCl 2 extracts were poor predictors of P content. In the field experiments, all tests evaluated performed poorly in describing the relationship between soil test P and the level of P applied and relative yield and soil test P over a wide range of soil types and environments. The best performing test was the Bray 1 test, though the relationship was poor. (author)

Emergence of a signal from background noise in the "memory of water" experiments: how to explain it?

Science.gov (United States)

Beauvais, Francis

2012-01-01

After more than 20 years, the case of the "memory of water" still has not been resolved satisfactorily. After the affair with the journal Nature, Benveniste extended his results on high dilutions to an "electromagnetic biology" and then to a "digital biology," where electromagnetic signals supposed to be emitted from biologically active solutions were said to be stored on magnetic memories. Although the results obtained by Benveniste and coworkers were obvious, the difficulties in reproducibility by other teams created doubt of the reality of the alleged phenomenon. In a first step, we analyzed a set of experiments obtained by Benveniste's team in the 1990s. We quantified the relationship between "expected" effects (ie, labels of the tested samples) and apparatus outcomes, and we defined the experimental conditions to observe significant correlations. We concluded that the results of these experiments were related to experimenter-dependent correlations, which did not support the initial "memory of water" hypothesis. The fact that a signal emerged from background noise, however, remained puzzling. Therefore, in a second step, we described Benveniste's experiments according to the relational interpretation of quantum physics of C. Rovelli. In this interpretation, the state of a system is observer-dependent and the collapse of the wave function appears only in the states relative to a given observer. This interpretation allowed us to elaborate a model describing Benveniste's experiments in which the emergence of a signal from background noise was described by the entanglement of the experimenter with the observed system. In conclusion, the pursuit of the experimental "proof" to support the "memory of water" hypothesis has prevented other interpretations. Although our hypothesis does not definitely dismiss the possibility of "memory of water," the experimenter-dependent entanglement could be an attractive alternative interpretation of Benveniste's experiments

Acceptance test procedure for High Pressure Water Jet System

International Nuclear Information System (INIS)

Crystal, J.B.

1995-01-01

The overall objective of the acceptance test is to demonstrate a combined system. This includes associated tools and equipment necessary to perform cleaning in the 105 K East Basin (KE) for achieving optimum reduction in the level of contamination/dose rate on canisters prior to removal from the KE Basin and subsequent packaging for disposal. Acceptance tests shall include necessary hardware to achieve acceptance of the cleaning phase of canisters. This acceptance test procedure will define the acceptance testing criteria of the high pressure water jet cleaning fixture. The focus of this procedure will be to provide guidelines and instructions to control, evaluate and document the acceptance testing for cleaning effectiveness and method(s) of removing the contaminated surface layer from the canister presently identified in KE Basin. Additionally, the desired result of the acceptance test will be to deliver to K Basins a thoroughly tested and proven system for underwater decontamination and dose reduction. This report discusses the acceptance test procedure for the High Pressure Water Jet

Experiment data report for Test RIA 1-2 (Reactivity Initiated Accident Test Series)

International Nuclear Information System (INIS)

Zimmermann, C.L.; White, C.E.; Evans, R.P.

1979-06-01

Recorded test data are presented for the second of six planned tests in the Reactivity Initiated Accident (RIA) Test Series I, Test RIA 1-2. This test, conducted at the Power Burst Facility, had the following objectives: (1) characterize the response of preirradiated fuel rods during an RIA event conducted at boiling water reactor hot-startup conditions; and (2) evaluate the effect of rod internal pressure on preirradiated fuel rod response during an RIA event. The data from Test RIA 1-2 are graphed in engineering units and have been appraised for quality and validity. These uninterpreted data are presented for use in the nuclear fuel behavior research field before detailed analysis and interpretation have been completed

10 CFR Appendix T to Subpart B of... - Uniform Test Method for Measuring the Water Consumption of Water Closets and Urinals

Science.gov (United States)

2010-01-01

... 10 Energy 3 2010-01-01 2010-01-01 false Uniform Test Method for Measuring the Water Consumption of... Appendix T to Subpart B of Part 430—Uniform Test Method for Measuring the Water Consumption of Water... previous step. The final water consumption value shall be rounded to one decimal place. b. The test...

Temelin 1000 MW Units active testing stage - commissioning experience

International Nuclear Information System (INIS)

Ubra, O.; Riha, V.

2003-01-01

There were three basic stages of the NPP Temelin Units active testing stage- The zero and low power testing, The power ascension testing and Plant Performance Test. The main objective of the start- up process stages and the testing procedures including some operational experience are described in the paper. (author)

Critical and Exponential Experiments on 19-Rod Clusters (R3 Fuel) in Heavy Water

Energy Technology Data Exchange (ETDEWEB)

Persson, R; Wikdahl, C E; Zadworski, Z

1962-03-15

Buckling measurements on clusters of 19 UO{sub 2} rods in heavy water have been performed in an exponential assembly and by means of substitution measurements in a critical facility. The material buckling was determined as a function of lattice pitch (range of V{sub mod} /V{sub fuel}: 7-22), internal spacing, void, and temperature (20 < T < 90 deg C). The change of diffusion coefficients (about 6-8 per cent) caused by voids was studied with single test fuel assemblies. The progressive substitution measurements have been analysed by means of a modified one-group perturbation theory in combination with an unconventional cell definition. The buckling differences between test and reference lattices are of the order of -1.0 to -3.5/m{sup 2}, The results of the exponential and the critical experiments are compared with similar measurements on the same kind of fuel at the Savannah River Laboratory. This comparison shows that the results of the various experiments agree quite well, whereas theoretical predictions fail in the extreme ranges.

Tests to determine water uptake behaviour of tunnel backfill

Energy Technology Data Exchange (ETDEWEB)

Dixon, David (Atomic Energy of Canada Limited (AECL) (Canada)); Anttila, S.; Viitanen, M. (Poeyry InfRa Oy (Finland)); Keto, Paula (Saanio and Riekkola Oy, Helsinki (Finland))

2008-12-15

A series of 27 large-scale tests have been completed at the 420 level of SKB's Aespoe Hard Rock Laboratory. These tests have examined the influence of natural Aespoe fracture zone water on the movement of water into and through assemblies of Friedland clay blocks and bentonite pellets/ granules. These tests have established the manner in which groundwater may influence backfill and backfilling operations at the repository-scale. Tests have established that it is critical to provide a clay block backfilling system with lateral support and confinement as quickly as possible following block installation. Exposure of the blocks to even low rates of water ingress can result in rapid loss of block cohesion and subsequent slumping of the block materials into the spaces between the blocks and the tunnel walls. Installation of granular or pelletized bentonite clay between the blocks and the walls resulted in a system that was generally stable and not prone to unacceptable short-term strains as water entered. Inflow of water into a backfilled volume does not result in uniform wetting of the pellet/granule filled volume and as a result there is the potential for rapid movement of water from the point(s) of ingress to the downstream face of the backfill. Depending on the inflow rate and flow path(s) developed this flow can be via discrete flow channels that are essentially non-erosive or else they can develop highly erosive flow paths through the clay block materials. Erosion generally tends to be highest in the period immediately following first water exit from the backfill and then decreases as preferential flow paths develop to channel the water directly through the backfill, bypassing large volumes of unsaturated backfill. At the scale examined in this study inflow rates of 0.1 l/min or less do not tend to be immediately problematic when the source is 0.6 m distant from the downstream face of the backfill. At larger scales or longer distances from the working face, it

Tests to determine water uptake behaviour of tunnel backfill

International Nuclear Information System (INIS)

Dixon, David; Anttila, S.; Viitanen, M.; Keto, Paula

2008-12-01

A series of 27 large-scale tests have been completed at the 420 level of SKB's Aespoe Hard Rock Laboratory. These tests have examined the influence of natural Aespoe fracture zone water on the movement of water into and through assemblies of Friedland clay blocks and bentonite pellets/ granules. These tests have established the manner in which groundwater may influence backfill and backfilling operations at the repository-scale. Tests have established that it is critical to provide a clay block backfilling system with lateral support and confinement as quickly as possible following block installation. Exposure of the blocks to even low rates of water ingress can result in rapid loss of block cohesion and subsequent slumping of the block materials into the spaces between the blocks and the tunnel walls. Installation of granular or pelletized bentonite clay between the blocks and the walls resulted in a system that was generally stable and not prone to unacceptable short-term strains as water entered. Inflow of water into a backfilled volume does not result in uniform wetting of the pellet/granule filled volume and as a result there is the potential for rapid movement of water from the point(s) of ingress to the downstream face of the backfill. Depending on the inflow rate and flow path(s) developed this flow can be via discrete flow channels that are essentially non-erosive or else they can develop highly erosive flow paths through the clay block materials. Erosion generally tends to be highest in the period immediately following first water exit from the backfill and then decreases as preferential flow paths develop to channel the water directly through the backfill, bypassing large volumes of unsaturated backfill. At the scale examined in this study inflow rates of 0.1 l/min or less do not tend to be immediately problematic when the source is 0.6 m distant from the downstream face of the backfill. At larger scales or longer distances from the working face, it is

A PMT mass testing facility for the JUNO experiment

Energy Technology Data Exchange (ETDEWEB)

Tietzsch, Alexander; Alsheimer, Isabell; Blum, David; Lachenmaier, Tobias; Sterr, Tobias [Physikalisches Institut, Universitaet Tuebingen (Germany); Bein, Bosse; Bick, Daniel; Ebert, Joachim; Hagner, Caren; Rebber, Henning; Steppat, Lisa; Wonsak, Bjoern [Institut fuer Experimentalphysik, Universitaet Hamburg (Germany)

2016-07-01

The JUNO (Jiangmen Underground Neutrino Observatory) experiment will be one of the big neutrino oscillation experiments starting in the next years. The main goal of JUNO is the determination of the neutrino mass hierarchy. To detect the sub-dominant effects in the oscillation pattern which depend on the mass hierarchy, the JUNO detector is planned with almost 20 kt fiducial volume, high light yield and energy resolution of better than 3%. In order to reach this, roughly 17000 newly developed high QE PMTs for the central detector, and additionally 2000 for the veto will be used. Each PMT has to be tested and characterized before it will be mounted in the experiment. This talk gives an overview on our plans and strategy for the mass test of all PMTs, and on the current status of the experimental test setup and next steps. The testing facility is developed in a cooperation between the Physical Institutes in Tuebingen and Hamburg within the JUNO collaboration.

Effect of Fuel Wobbe Number on Pollutant Emissions from Advanced Technology Residential Water Heaters: Results of Controlled Experiments

Energy Technology Data Exchange (ETDEWEB)

Rapp, Vi H.; Singer, Brett C.

2014-03-01

The research summarized in this report is part of a larger effort to evaluate the potential air quality impacts of using liquefied natural gas in California. A difference of potential importance between many liquefied natural gas blends and the natural gas blends that have been distributed in California in recent years is the higher Wobbe number of liquefied natural gas. Wobbe number is a measure of the energy delivery rate for appliances that use orifice- or pressure-based fuel metering. The effect of Wobbe number on pollutant emissions from residential water heaters was evaluated in controlled experiments. Experiments were conducted on eight storage water heaters, including five with “ultra low-NO{sub X}” burners, and four on-demand (tankless) water heaters, all of which featured ultra low-NO{sub X} burners. Pollutant emissions were quantified as air-free concentrations in the appliance flue and fuel-based emission factors in units of nanogram of pollutant emitter per joule of fuel energy consumed. Emissions were measured for carbon monoxide (CO), nitrogen oxides (NO{sub X}), nitrogen oxide (NO), formaldehyde and acetaldehyde as the water heaters were operated through defined operating cycles using fuels with varying Wobbe number. The reference fuel was Northern California line gas with Wobbe number ranging from 1344 to 1365. Test fuels had Wobbe numbers of 1360, 1390 and 1420. The most prominent finding was an increase in NO{sub X} emissions with increasing Wobbe number: all five of the ultra low-NO{sub X} storage water heaters and two of the four ultra low-NO{sub X} on-demand water heaters had statistically discernible (p<0.10) increases in NO{sub X} with fuel Wobbe number. The largest percentage increases occurred for the ultra low-NO{sub X} water heaters. There was a discernible change in CO emissions with Wobbe number for all four of the on-demand devices tested. The on-demand water heater with the highest CO emissions also had the largest CO increase

Heat tracer test in an alluvial aquifer: Field experiment and inverse modelling

Science.gov (United States)

Klepikova, Maria; Wildemeersch, Samuel; Hermans, Thomas; Jamin, Pierre; Orban, Philippe; Nguyen, Frédéric; Brouyère, Serge; Dassargues, Alain

2016-09-01

Using heat as an active tracer for aquifer characterization is a topic of increasing interest. In this study, we investigate the potential of using heat tracer tests for characterization of a shallow alluvial aquifer. A thermal tracer test was conducted in the alluvial aquifer of the Meuse River, Belgium. The tracing experiment consisted in simultaneously injecting heated water and a dye tracer in an injection well and monitoring the evolution of groundwater temperature and tracer concentration in the pumping well and in measurement intervals. To get insights in the 3D characteristics of the heat transport mechanisms, temperature data from a large number of observation wells closely spaced along three transects were used. Temperature breakthrough curves in observation wells are contrasted with what would be expected in an ideal layered aquifer. They reveal strongly unequal lateral and vertical components of the transport mechanisms. The observed complex behavior of the heat plume is explained by the groundwater flow gradient on the site and heterogeneities in the hydraulic conductivity field. Moreover, due to high injection temperatures during the field experiment a temperature-induced fluid density effect on heat transport occurred. By using a flow and heat transport numerical model with variable density coupled with a pilot point approach for inversion of the hydraulic conductivity field, the main preferential flow paths were delineated. The successful application of a field heat tracer test at this site suggests that heat tracer tests is a promising approach to image hydraulic conductivity field. This methodology could be applied in aquifer thermal energy storage (ATES) projects for assessing future efficiency that is strongly linked to the hydraulic conductivity variability in the considered aquifer.

Design requirements for the supercritical water oxidation test bed

International Nuclear Information System (INIS)

Svoboda, J.M.; Valentich, D.J.

1994-05-01

This report describes the design requirements for the supercritical water oxidation (SCWO) test bed that will be located at the Idaho National Engineering Laboratory (INEL). The test bed will process a maximum of 50 gph of waste plus the required volume of cooling water. The test bed will evaluate the performance of a number of SCWO reactor designs. The goal of the project is to select a reactor that can be scaled up for use in a full-size waste treatment facility to process US Department of Energy mixed wastes. EG ampersand G Idaho, Inc. will design and construct the SCWO test bed at the Water Reactor Research Test Facility (WRRTF), located in the northern region of the INEL. Private industry partners will develop and provide SCWO reactors to interface with the test bed. A number of reactor designs will be tested, including a transpiring wall, tube, and vessel-type reactor. The initial SCWO reactor evaluated will be a transpiring wall design. This design requirements report identifies parameters needed to proceed with preliminary and final design work for the SCWO test bed. A flow sheet and Process and Instrumentation Diagrams define the overall process and conditions of service and delineate equipment, piping, and instrumentation sizes and configuration Codes and standards that govern the safe engineering and design of systems and guidance that locates and interfaces test bed hardware are provided. Detailed technical requirements are addressed for design of piping, valves, instrumentation and control, vessels, tanks, pumps, electrical systems, and structural steel. The approach for conducting the preliminary and final designs and environmental and quality issues influencing the design are provided

Human posture experiments under water: ways of applying the findings to microgravity

Science.gov (United States)

Dirlich, Thomas

For the design and layout human spacecraft interiors the Neutral Body Posture (NBP) in micro-gravity is of great importance. The NBP has been defined as the stable, replicable and nearly constant posture the body "automatically" assumes when a human relaxes in microgravity. Furthermore the NBP, as published, suggests that there is one standard neutral posture for all individuals. Published experiments from space, parabolic flights and under water on the other hand show strong inter-individual variations of neutral (relaxed) postures. This might originate from the quite small sample sizes of subjects analyzed or the different experiment conditions, e. g. space and under water. Since 2008 a collaborative research project focussing on human postures and motions in microgravity has been ongoing at the Technische Univer-sitüt München (TUM). This collaborative effort is undertaken by the Institute of Astronautics a (LRT) and the Institute of Ergonomics (LfE). Several test campaigns have been conducted in simulated microgravity under water using a specially designed standardized experiment setup. Stereo-metric HD video footage and anthropometric data from over 50 subjects (female and male) has been gathered in over 80 experiments. The video data is analyzed using PCMAN software, developed by the LfE, resulting in a 3D volumetric CAD-based model of each subject and posture. Preliminary and ongoing analysis of the data offer evidence for the existence of intra-individually constant neutral postures, as well as continuously recurring relaxation strate-gies. But as with the data published prior the TUM experiments show quite a large variation of inter-individual postures. These variation might be induced or influenced by the special environmental conditions in the underwater experiment. Thus in present paper ways of stan-dardizing data and applying the findings gathered under water to real microgravity are being discussed. The following influences stemming from the

The preference for water nipples vs. water bowls in dairy goats

Directory of Open Access Journals (Sweden)

Andersen Inger L

2011-09-01

Full Text Available Abstract Background Previous studies have reported that the design of the water dispensers can influence the water intake in farm animals. Horses and dairy cows seem to prefer to drink from an open surface whereas sheep and pigs apparently prefer water nipples, probably because of the worse water quality in water bowls. The aim of the present study was to examine the preference of dairy goats for water nipples or water bowls. Methods In each of the two experiments (exp. 1, dry goats, exp. 2 lactating goats, 42 dairy goats were allotted into 6 groups of 7 goats. In period 1, the goats had access to a water nipple. In period 2, they had access to a water bowl and in period 3 (preference test they had access to both a water nipple and a water bowl. Water usage and wastage was recorded and water intake (water usage - water wastage was calculated for each group for the two last days of each period. In experiment 2, water samples from each dispenser were analyzed for heterotrophy germs at 22°C, Escherichia coli and turbidity. Results Water usage was higher from water nipples than from water bowls both in experiment 1 (dry goats and experiment 2 (lactating goats. There was however, no difference in water intake from water nipples and water bowls. In the preference test (period 3, the water intake tended to be higher from the water nipple than from the water bowl both for the dry goats (exp. 1 and lactating goats (exp. 2. Especially for the dry goats, the differences between groups were large. Turbidity and heterotrophy germs were much higher in the samples from the water bowls than from the water nipples. Water wastage from the water bowls was negligible compared to the water nipples. From the water nipples the water wastage was 30% and 23% of water usage for the dry and lactating goats respectively. Conclusions We conclude that type of water dispenser (nipple or bowl was probably of minor importance for water intake in goats, but water bowls had a

Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

Energy Technology Data Exchange (ETDEWEB)

T. A. Tomberlin; S. B. Grover

2004-11-01

The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

International Nuclear Information System (INIS)

T. A. Tomberlin; S. B. Grover

2004-01-01

The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment

Test results of sodium-water reaction testing in near prototypical LMR steam generator

International Nuclear Information System (INIS)

Boardman, C.E.; Hui, M.; Neely, H.H.

1990-01-01

An extensive test program has been performed in the United States to investigate the effects of large sodium-water reaction events in LMFBR steam generators. Tests were conducted in the Large Leak Test Rig (LLTR) located at the Energy Technology Engineering Center (ETEC). The program was divided into two phases, Series I and Series II, for the purpose of satisfying near-term and long-term needs. Series II was further subdivided into large and intermediate leak tests. This paper will emphasize the Series II intermediate leak tests and resulting conclusions for steam generator design and operation. 11 figs, 2 tabs

HDR flood-water storage-tank modal vibration tests

International Nuclear Information System (INIS)

Gorman, V.W.; Thinnes, G.L.

1983-01-01

Modal vibration tests were conducted by EG and G Idaho on two vessels located at West Germany's Heissdampfreaktor (HDR) facility which is 25 kilometers east of Frankfurt. The tests were performed during May and June 1982 for the US Nuclear Regulatory Commission (NRC) as part of their cooperative effort with Kernforschungszentrum Karlsruhe (KfK) of West Germany. The primary purpose for performing this task was to determine modal properties (frequencies, mode shapes and associated damping ratios) in order to eventually provide guidelines for standards development by the NRC in modeling similar vessels. One of the vessels tested was a flood water storage tank (FWST) for empty, half full and full water conditions. The FWST was excited randomly with an electromagnetic shaker and by impulsive hammer blows. Excitation or input forces together with measured vessel responses were processed by a digital modal analyzer and stored on magnetic disks for subsequent evaluation

Experience in maintenance of pumps in test facilities at R.E.D., B.A.R.C. [Paper No.: II-5

International Nuclear Information System (INIS)

Nangia, H.V.; Soni, K.L.; Kamath, K.V.; Mahajan, S.C.

1981-01-01

Centrifugal pumps are used for circulating water in various test facilities which are meant for testing reactor components. Operating pressures are about 85 bars at 250 degC. Reciprocating pumps are used for make up service at pressures upto 160 bars. A primary heat transport pump for 200 MWe Candu type reactor, is being tested for checking its performance and for collection of base data. The following are discussed: (i) failure of hydrostatic seal and hydrostatic bearing and its causes, (ii) subsequent repairs and reconditioning, (iii) steps taken to avoid repetition of such failures, and (iv) problems in assembly. For a pump used in the fuelling machine test facility, following maintenance problems are discussed: (1) failure of various components like bearings, mechanical, seals, wear rings, etc., (2) causes of failure and steps taken to remedy the deficiencies noted, (3) experience with the indigenous antifriction bearings, and (4) experience with indigenous spiral wound gaskets. For the reciprocating pumps, leakage through gland packing is a problem. Experience with various types of packing and other parameters, affecting leakage are discussed. (author)

Creating a testing field where delta technology and water innovations are tested and demonstrated with the help of citizen science methods

Science.gov (United States)

de Vries, Sandra; Rutten, Martine; de Vries, Liselotte; Anema, Kim; Klop, Tanja; Kaspersma, Judith

2017-04-01

In highly populated deltas, much work is to be done. Complex problems ask for new and knowledge driven solutions. Innovations in delta technology and water can bring relief to managing the water rich urban areas. Testing fields form a fundamental part of the knowledge valorisation for such innovations. In such testing fields, product development by start-ups is coupled with researchers, thus supplying new scientific insights. With the help of tests, demonstrations and large-scale applications by the end-users, these innovations find their way to the daily practices of delta management. More and more cities embrace the concept of Smart Cities to tackle the ongoing complexity of urban problems and to manage the city's assets - such as its water supply networks and other water management infrastructure. Through the use of new technologies and innovative systems, data are collected from and with citizens and devices - then processed and analysed. The information and knowledge gathered are keys to enabling a better quality of life. By testing water innovations together with citizens in order to find solutions for water management problems, not only highly spatial amounts of data are provided by and/or about these innovations, they are also improved and demonstrated to the public. A consortium consisting of a water authority, a science centre, a valorisation program and two universities have joined forces to create a testing field for delta technology and water innovations using citizen science methods. In this testing field, the use of citizen science for water technologies is researched and validated by facilitating pilot projects. In these projects, researchers, start-ups and citizens work together to find the answer to present-day water management problems. The above mentioned testing field tests the use of crowd-sourcing data as for example hydrological model inputs, or to validate remote sensing applications, or improve water management decisions. Currently the

Comparison of water absorption methods: testing the water absorption of recently quarried and weathered porous limestone on site and under laboratory conditions

Science.gov (United States)

Rozgonyi-Boissinot, Nikoletta; Agárdi, Tamás; Karolina Cebula, Ágnes; Török, Ákos

2017-04-01

The water absorption of weathering sensitive stones is a critical parameter that influences durability. The current paper compares different methods of water absorption tests by using on site and laboratory tests. The aims of the tests were to assess the water absorption of un-weathered quarry stones and various weathering forms occurring on porous limestone monuments. For the tests a Miocene porous limestone was used that occurs in Central and Western Hungary and especially near and in Budapest. Besides the Hungarian occurrences the same or very similar porous limestones are found in Austria, Slovakia and in the Czech Republic. Several quarries were operating in these countries. Due to the high workability the stone have been intensively used as construction material from the Roman period onward. The most prominent monuments made of this stone were built in Vienna and in Budapest during the 18th -19th century and in the early 20th century. The high porosity and the micro-fabric of the stone make it prone to frost- and salt weathering. Three different limestone types were tested representing coarse-, medium- and fine grained lithologies. The test methods included Rilem tube (Karsten tube) tests and capillary water absorption tests. The latter methodology has been described in detail in EN 1925:2000. The test results of on-site tests of weathered porous limestone clearly show that the water absorption of dissolved limestone surfaces and crumbling or micro-cracked limestone is similar. The water absorption curves have similar inclinations marking high amount of absorbed water. To the contrary, the white weathering crusts covered stone blocks and black crusts have significantly lower water absorptions and many of these crusts are considered as very tight almost impermeable surfaces. Capillary water absorption tests in the laboratory allowed the determination of maximum water absorption of quarried porous limestone. Specimens were placed in 3 mm of water column and the

Estimation of Water Quality

International Nuclear Information System (INIS)

Vetrinskaya, N.I.; Manasbayeva, A.B.

1998-01-01

Water has a particular ecological function and it is an indicator of the general state of the biosphere. In relation with this summary, the toxicological evaluation of water by biologic testing methods is very actual. The peculiarity of biologic testing information is an integral reflection of all totality properties of examination of the environment in position of its perception by living objects. Rapid integral evaluation of anthropological situation is a base aim of biologic testing. If this evaluation has deviations from normal state, detailed analysis and revelation of dangerous components could be conducted later. The quality of water from the Degelen gallery, where nuclear explosions were conducted, was investigated by bio-testing methods. The micro-organisms (Micrococcus Luteus, Candida crusei, Pseudomonas algaligenes) and water plant elodea (Elodea canadensis Rich) were used as test-objects. It is known that the transporting functions of cell membranes of living organisms are violated the first time in extreme conditions by difference influences. Therefore, ion penetration of elodeas and micro-organisms cells, which contained in the examination water with toxicants, were used as test-function. Alteration of membrane penetration was estimated by measurement of electrolytes electrical conductivity, which gets out from living objects cells to distillate water. Index of water toxic is ratio of electrical conductivity in experience to electrical conductivity in control. Also, observations from common state of plant, which was incubated in toxic water, were made. (Chronic experience conducted for 60 days.) The plants were incubated in water samples, which were picked out from gallery in the years 1996 and 1997. The time of incubation is 1-10 days. The results of investigation showed that ion penetration of elodeas and micro-organisms cells changed very much with influence of radionuclides, which were contained in testing water. Changes are taking place even in

The use of peracetic acid in drinking water systems: flow tests

International Nuclear Information System (INIS)

Ragazzo, P.; Navazio, G.

2000-01-01

In a previous research, a preliminary study was carried out on the disinfection efficiency of peracetic acid (PAA), comparing it to that of other disinfectants that are typically used, in batch tests with dosage values ranging from 0.5 to 5 ppm. The study was carried out on samples of water collected from several significant points of the treatment process at the main water treatment plant in Jesolo (Venice, Italy). On the basis of results (basically positive at that time) obtained from these tests, a 400 litre/hour pilot plant was built, as a lower scale reproduction of the drink water treatment system mentioned earlier, in order to study the characteristics of PAA even in tests that could more realistically simulate the flow of water along the process. These tests essentially confirmed the kinetics of the spontaneous hydrolysis to CH 3 COOH+H 2 O 2 and those of dismutation to CH 3 COOH+O 2 , with half-life time values ranging from 3 to 12 hours, depending on the characteristics of the water (especially the pH factor) and the PAA concentration values [it

Validation of the ATHLET-code 2.1A by calculation of the ECTHOR experiment; Validierung des ATHLET-Codes 2.1A anhand des Einzeleffekt-Tests ECTHOR

Energy Technology Data Exchange (ETDEWEB)

Jung, Andreas; Sarkadi, Peter; Schaffrath, Andreas [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany)

2010-05-15

Before a numerical code (e.g. ATHLET) is used for simulation of physical phenomena being new or unknown for the code and/or the user, the user ensures the applicability of the code and his own experience of handling with it by means of a so-called validation. Parametric studies with the code are executed for that matter and the results have to be compared with verified experimental data. Corresponding reference values are available in terms of so-called single-effect-tests (e.g. ECTHOR). In this work the system-code ATHLET Mod. 2.1 Cycle A is validated by post test calculation of the ECTHOR experiment due to the above named aspects. With the ECTHOR-tests the clearing of a water-filled model of a loop seal by means of an air-stream was investigated including momentum exchange at the phase interface under adiabatic and atmospheric conditions. The post test calculations show that the analytical results meet the experimental data within the reproducibility of the experiments. Further findings of the parametric studies are: - The experimental results obtained with the system water-air (ECTHOR) can be assigned to a water-steam-system, if the densities of the phases are equal in both cases. - The initial water level in the loop seal has no influence on the results as long as the gas mass flow is increased moderately. - The loop seal is appropriately nodalized if the mean length of the control volumes accords approx. 1.5 tim es the hydraulic pipe diameter. (orig.)

Validation of the ATHLET-code 2.1A by calculation of the ECTHOR experiment; Validierung des ATHLET-Codes 2.1A anhand des Einzeleffekt-Tests ECTHOR

Energy Technology Data Exchange (ETDEWEB)

Jung, Andreas; Sarkadi, Peter; Schaffrath, Andreas [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany)

2010-06-15

Before a numerical code (e.g. ATHLET) is used for simulation of physical phenomena being new or unknown for the code and/or the user, the user ensures the applicability of the code and his own experience of handling with it by means of a so-called validation. Parametric studies with the code are executed for that matter und the results have to be compared with verified experimental data. Corresponding reference values are available in terms of so-called single-effect-tests (e.g. ECTHOR). In this work the system-code ATHLET Mod. 2.1 Cycle A is validated by post test calculation of the ECTHOR experiment due to the above named aspects. With the ECTHOR-tests the clearing of a water-filled model of a loop seal by means of an air-stream was investigated including momentum exchange at the phase interface under adiabatic and atmospheric conditions. The post test calculations show that the analytical results meet the experimental data within the reproducibility of the experiments. Further findings of the parametric studies are: - The experimental results obtained with the system water-air (ECTHOR) can be assigned to a water-steam-system, if the densities of the phases are equal in both cases. - The initial water level in the loop seal has no influence on the results as long as the gas mass flow is increased moderately. - The loop seal is appropriately nodalized if the mean length of the control volumes accords approx. 1.5 times the hydraulic pipe diameter. (orig.)

Does the Fizeau Experiment Really Test Special Relativity?

Science.gov (United States)

Clement, Gerard

1980-01-01

The motivation and interpretation of the Fizeau experiment are reviewed, and its status as a test of special relativity is discussed. It is shown, with the aid of a simplified, purely mechanical model of the propagation of light in matter, that the experiment actually cannot discriminate between Galilean and relativistic kinematics. (Author/SK)

Automatically generated acceptance test: A software reliability experiment

Science.gov (United States)

Protzel, Peter W.

1988-01-01

This study presents results of a software reliability experiment investigating the feasibility of a new error detection method. The method can be used as an acceptance test and is solely based on empirical data about the behavior of internal states of a program. The experimental design uses the existing environment of a multi-version experiment previously conducted at the NASA Langley Research Center, in which the launch interceptor problem is used as a model. This allows the controlled experimental investigation of versions with well-known single and multiple faults, and the availability of an oracle permits the determination of the error detection performance of the test. Fault interaction phenomena are observed that have an amplifying effect on the number of error occurrences. Preliminary results indicate that all faults examined so far are detected by the acceptance test. This shows promise for further investigations, and for the employment of this test method on other applications.

Elemental Water Impact Test: Phase 2 36-Inch Aluminum Tank Head

Science.gov (United States)

Vassilakos, Gregory J.

2014-01-01

Spacecraft are being designed based on LS-DYNA simulations of water landing impacts. The Elemental Water Impact Test (EWIT) series was undertaken to assess the accuracy of LS-DYNA water impact simulations. EWIT Phase 2 featured a 36-inch aluminum tank head. The tank head was outfitted with one accelerometer, twelve pressure transducers, three string potentiometers, and four strain gages. The tank head was dropped from heights of 1 foot and 2 feet. The focus of this report is the correlation of analytical models against test data. As a measure of prediction accuracy, peak responses from the baseline LS-DYNA model were compared to peak responses from the tests.

78 FR 2340 - Energy Conservation Program: Test Procedures for Residential Water Heaters and Commercial Water...

Science.gov (United States)

2013-01-11

... corresponded to the levels in the American Society of Heating, Refrigerating and Air Conditioning Engineers... provides a brief history of DOE's more recent test procedure rulemakings related to residential water... performance (e.g., such as ambient air temperature, ambient relative humidity, and inlet water temperature...

2D/3D program. Upper plenum test facility - UPTF. Test No. 1

International Nuclear Information System (INIS)

1987-01-01

Test No.1 was a quasi-steady state, separate effect test involving the UPTF-System with blocked break valves and blocked pump simulators. Initially the test vessel, the cold and hot leg nozzels as well as the pump seals were completely filled witht hot water in this test. This test was designed to investigate the fluid-fluid mixing phenomena and the development of the fluid and wall temperature fields in the cold leg and downcomer region of a PWR. The experiment was performed by injecting a cold water stream into one cold leg of UPTF while the system was initially filled with stagnant hot water. (orig.)

Testing of the method for water microleakage detection from OH hydroxyl spectral lines at the L-2M stellarator

International Nuclear Information System (INIS)

Voronov, G. S.; Berezhetskii, M. S.; Bondar’, Yu. F.; Vafin, I. Yu.; Vasil’kov, D. G.; Voronova, E. V.; Grebenshchikov, S. E.; Grishina, I. A.; Larionova, N. F.; Letunov, A. A.; Logvinenko, V. P.; Meshcheryakov, A. I.; Pleshkov, E. I.; Khol’nov, Yu. V.; Fedyanin, O. I.; Tsygankov, V. A.; Shchepetov, S. V.; Kurnaev, V. A.; Vizgalov, I. V.; Urusov, V. A.

2013-01-01

Results are presented from L-2M stellarator experiments on testing a possible method for detection of water microleakages in the cooling system of the first wall and vacuum chamber of ITER. The method consists in the spectroscopic detection of spectral lines of the OH hydroxyl, which forms via the dissociation of water molecules in plasma. Emission in the spectral band of 305–310 nm can be detected even at water leakage rates less than 10 −4 Pa m 3 /s. Chemical reactions between water and boron compounds on the vacuum chamber wall delay the detection of leakages up to ∼2000 s. A similar phenomenon can be expected when a leakage will occur in ITER, where the materials suggested for the first wall (Be, Li) can also chemically react with water.

Endurance test and evaluation of alkaline water electrolysis cells

Science.gov (United States)

Burke, K. A.; Schubert, F. H.

1981-01-01

Utilization in the development of multi-kW low orbit power systems is discussed. The following technological developments of alkaline water electrolysis cells for space power application were demonstrated: (1) four 92.9 cm2 single water electrolysis cells, two using LST's advanced anodes and two using LST's super anodes; (2) four single cell endurance test stands for life testing of alkaline water electrolyte cells; (3) the solid performance of the advanced electrode and 355 K; (4) the breakthrough performance of the super electrode; (5) the four single cells for over 5,000 hours each significant cell deterioration or cell failure. It is concluded that the static feed water electrolysis concept is reliable and due to the inherent simplicity of the passive water feed mechanism coupled with the use of alkaline electrolyte has greater potential for regenerative fuel cell system applications than alternative electrolyzers. A rise in cell voltage occur after 2,000-3,000 hours which was attributed to deflection of the polysulfone end plates due to creepage of the thermoplastic. More end plate support was added, and the performance of the cells was restored to the initial performance level.

Light water reactor mixed-oxide fuel irradiation experiment

International Nuclear Information System (INIS)

Hodge, S.A.; Cowell, B.S.; Chang, G.S.; Ryskamp, J.M.

1998-01-01

The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding

Experiment and mathematical model for the heat transfer in water around 4 °C

Science.gov (United States)

Ogawa, Naohisa; Kaneko, Fumitoshi

2017-03-01

Water, which is the habitat for a variety of living creatures, has a maximum density at 4.0 °C. This crucial property is considered to play a very important role in the biology of a lake and also has a close relationship with the areas of environmentology and geoscience. It would be desirable for students to confirm this important property of water themselves by carrying out simple experiments. However, it is not easy to detect the maximum density at 4.0 °C because the temperature dependence of the water density is very small close to its freezing point. For example, the density of water is 0.999 975 g cm-3 at 4.0 °C and 0.999 850 g cm-3 at 0.1 °C. The aim in this manuscript is to demonstrate a simple experiment to detect 4.0 °C as the temperature of maximum density, in which the time dependence of the water temperature is measured at several different depths by chilling the water surface. This is a simple experiment that can also be performed by high school students. We also present a mathematical model that can explain the results of this experiment.

Laboratory and field tests of the Sutron RLR-0003-1 water level sensor

Science.gov (United States)

Fulford, Janice M.; Bryars, R. Scott

2015-01-01

Three Sutron RLR-0003-1 water level sensors were tested in laboratory conditions to evaluate the accuracy of the sensor over the manufacturer’s specified operating temperature and distance-to-water ranges. The sensor was also tested for compliance to SDI-12 communication protocol and in field conditions at a U.S. Geological Survey (USGS) streamgaging site. Laboratory results were compared to the manufacturer’s accuracy specification for water level and to the USGS Office of Surface Water (OSW) policy requirement that water level sensors have a measurement uncertainty of no more than 0.01 foot or 0.20 percent of the indicated reading. Except for one sensor, the differences for the temperature testing were within 0.05 foot and the average measurements for the sensors were within the manufacturer’s accuracy specification. Two of the three sensors were within the manufacturer’s specified accuracy and met the USGS accuracy requirements for the laboratory distance to water testing. Three units passed a basic SDI-12 communication compliance test. Water level measurements made by the Sutron RLR-0003-1 during field testing agreed well with those made by the bubbler system and a Design Analysis Associates (DAA) H3613 radar, and they met the USGS accuracy requirements when compared to the wire-weight gage readings.

History of the water chemistry for the few tube test model

International Nuclear Information System (INIS)

Moss, S.A.; Simpson, J.L.

1979-09-01

The water chemistry activities carried out in support of the Few Tube Test are described. This test was conducted to provide design confirmation data for the Clinch River Breeder Reactor Project (CRBRP) steam generators. Proposed CRBRP chemistry was followed; all volatile treatment (AVT) of water was carried out with on-line monitoring capability

Chemical Explosion Experiments to Improve Nuclear Test Monitoring - Developing a New Paradigm for Nuclear Test Monitoring with the Source Physics Experiments (SPE)

International Nuclear Information System (INIS)

Snelson, Catherine M.; Abbott, Robert E.; Broome, Scott T.; Mellors, Robert J.; Patton, Howard J.; Sussman, Aviva J.; Townsend, Margaret J.; Walter, William R.

2013-01-01

A series of chemical explosions, called the Source Physics Experiments (SPE), is being conducted under the auspices of the U.S. Department of Energy's National Nuclear Security Administration (NNSA) to develop a new more physics-based paradigm for nuclear test monitoring. Currently, monitoring relies on semi-empirical models to discriminate explosions from earthquakes and to estimate key parameters such as yield. While these models have been highly successful monitoring established test sites, there is concern that future tests could occur in media and at scale depths of burial outside of our empirical experience. This is highlighted by North Korean tests, which exhibit poor performance of a reliable discriminant, mb:Ms (Selby et al., 2012), possibly due to source emplacement and differences in seismic responses for nascent and established test sites. The goal of SPE is to replace these semi-empirical relationships with numerical techniques grounded in a physical basis and thus applicable to any geologic setting or depth

Some new fatigue tests in high temperature water and liquid sodium environment

International Nuclear Information System (INIS)

Hattori, Takahiro; Yamauchi, Takayoshi; Kanasaki, Hiroshi; Kondo, Yoshiyuki; Endo, Tadayoshi.

1987-01-01

To evaluate the fatigue strength of structural materials for PWR or FBR plants, fatigue test data must be obtained in an environment of simulated primary and secondary water for PWR or of high temperature liquid sodium for FBR. Generally, such tests make it necessary to prepare expensive facilities, so when large amount of fatigue data are required, it is necessary to rationalize and simplify the fatigue tests while maintaining high accuracy. At the Takasago Research Development Center, efforts to rationalize facilities and maintain accuracy in fatigue tests have been made by developing new test methods and improving conventional techniques. This paper introduces a new method of low cycle fatigue test in high temperature water, techniques for automatic measurement of crack initiation and propagation in high temperature water environment and a multiple type fatigue testing machine for high temperature liquid sodium. (author)

Correction of Pressure Drop in Steam and Water System in Performance Test of Boiler

Science.gov (United States)

Liu, Jinglong; Zhao, Xianqiao; Hou, Fanjun; Wu, Xiaowu; Wang, Feng; Hu, Zhihong; Yang, Xinsen

2018-01-01

Steam and water pressure drop is one of the most important characteristics in the boiler performance test. As the measuring points are not in the guaranteed position and the test condition fluctuation exsits, the pressure drop test of steam and water system has the deviation of measuring point position and the deviation of test running parameter. In order to get accurate pressure drop of steam and water system, the corresponding correction should be carried out. This paper introduces the correction method of steam and water pressure drop in boiler performance test.

Field Validation of Toxicity Tests to Evaluate the Potential for Beneficial Use of Produced Water

Energy Technology Data Exchange (ETDEWEB)

Joseph Bidwell; Jonathan Fisher; Naomi Cooper

2008-03-31

limited to sediment depths of 10 cm or greater, which is outside of the primary zone of biological activity. Further, exposure to site sediments did not have any effects on test organisms, and macroinvertebrate communities did not indicate impairment at the oil production site as compared to a reference site. In situ experiments with H. azteca and C. fluminea, indicated a sublethal site effect (on growth of both species), but these could not be definitively linked with produced water infiltration. Severe weather conditions (drought followed by flooding) negatively influenced the intensity of lake sampling aimed at delineating produced water infiltration. Due to the lack of clear evidence of produced water infiltration into the sub-littoral zone of the lake, it was not possible to assess whether the laboratory bioassays of produced water effectively indicate risk in the receiving system. However, the acutely toxic nature of the produced water and general lack of biological effects in the lake at the oil production site suggest minimal to no produced water infiltration into surficial lake sediments and the near-shore water column. This study was able to demonstrate the utility of ion toxicity modeling to support data from toxicity identification evaluations aimed at identifying key toxic constituents in produced water. This information could be used to prioritize options for treating produced water in order to reduce toxic constituents and enhance options for reuse. The study also demonstrated how geographic information systems, toxicity modeling, and toxicity assessment could be used to facilitate future site assessments.

Final report on 3-D experiment project air-water upper plenum experiments

International Nuclear Information System (INIS)

Jacoby, J.K.; Mohr, C.M.

1978-11-01

The results are presented from upper plenum air-water reflood behavior testing performed as part of the program to investigate three-dimensional aspects of PWR LOCA research. Tests described were performed at near ambient temperature and pressure in a plexiglass vessel which included the important features of the upper core and upper plenum regions corresponding to a single fuel bundle in both Westinghouse Electric Corporation (Trojan) and Kraftwerk Union (KKU) PWR designs. The data included observed two-phase flow characteristics, particularly with regard to countercurrent flow, and cinematography of the characteristic upper plenum flow patterns

Treatability tests on water from a low-level waste burial ground

International Nuclear Information System (INIS)

Taylor, P.A.

1990-01-01

Lab-scale treatability tests on trench water from a low-level waste burial ground have shown that the water can be successfully treated by existing wastewater treatment plants at Oak Ridge National Laboratory. Water from the four most highly contaminated trenches that had been identified to date was used in the treatability tests. The softening and ion exchange processes used in the Process Wastewater Treatment Plant removed Sr-90 from the trench water, which was the only radionuclide present at above the discharge limits. The air stripping and activated carbon adsorption processes used in the Nonradiological Wastewater Treatment Plant removed volatile and semi-volatile organics, which were the main contaminants in the trench water, to below detection limits. 6 refs., 2 figs., 7 tabs

OECD MCCI Small-Scale Water Ingression and Crust Strength tests (SSWICS) design report, Rev. 2 October 31, 2002

International Nuclear Information System (INIS)

Farmer, M.; Lomperski, S.; Kilsdonk, D.; Aeschlimann, B.; Pfeiffer, P.

2011-01-01

The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core/concrete interaction (MCCI) when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. However, due to the integral nature of these tests, several questions regarding the crust freezing behavior could not be adequately resolved. These questions include: (1) To what extent does water ingression into the crust increase the melt quench rate above the conduction-limited rate and how is this affected by melt composition and system pressure and (2) What is the fracture strength of the corium crust when subjected to a thermal-mechanical load and how does it depend upon the melt composition? A series of separate-effects experiments are planned to address these issues. The first employs an apparatus designed to measure the quench rate of a pool of corium (∼φ30 cm; up to 20 cm deep). The main parameter to be varied in these quench tests is the melt composition since it is thought to have a critical influence on the crust cracking behavior which, in turn, alters quench rate. A description of the test apparatus, instrumentation, data reduction, and test matrix are the subject of the first portion of this report. The issue of crust strength will be addressed with a second apparatus designed to mechanically load the crust produced by the quench tests. This apparatus will measure the fracture strength of the crust while under a thermal load created by a heating element beneath the crust. The introduction of a thermal gradient across the crust is thought to be important for these tests because of uncertainty in the magnitude of the thermal stresses and thus their relative

TESTING OF CARBONACEOUS ADSORBENTS FOR REMOVAL OF POLLUTANTS FROM WATER

Directory of Open Access Journals (Sweden)

RAISA NASTAS

2012-03-01

Full Text Available Testing of carbonaceous adsorbents for removal of pollutants from water. Relevant direction for improving of quality of potable water is application of active carbons at various stages of water treatments. This work includes complex research dealing with testing of a broad spectrum of carbonaceous adsorbents for removal of hydrogen sulfide and nitrite ions from water. The role of the surface functional groups of carbonaceous adsorbents, their acid-basic properties, and the influence of the type of impregnated heteroatom (N, O, or metals (Fe, Cu, Ni, on removal of hydrogen sulfide species and nitrite ions have been researched. The efficiency of the catalyst obtained from peach stones by impregnation with Cu2+ ions of oxidized active carbon was established, being recommended for practical purposes to remove the hydrogen sulfide species from the sulfurous ground waters. Comparative analysis of carbonaceous adsorbents reveals the importance of surface chemistry for oxidation of nitrite ions.

Can a Century Old Experiment Reveal Hidden Properties of Water?

Directory of Open Access Journals (Sweden)

Elmar C. Fuchs

2010-08-01

Full Text Available In 1893 Sir William Armstrong placed a cotton thread between two wine glasses filled with chemically pure water. After applying a high voltage, a watery connection formed, and after some time, the cotton thread was pulled into one of the glasses, leaving a rope of water suspended between the two glasses. Although being a very simple experiment, it is of special interest since it comprises a number of phenomena currently tackled in modern water science like electrolysis-less charge transport and nanobubbles. This work gives some background information about water research in general and describes the water bridge phenomenon from the viewpoint of different fields such as electrohydrodynamics and quantum field theory. It is shown that the investigation of the floating water bridge led to new discoveries about water, both in the macroscopic and microscopic realm – but these were merely “hidden” in that sense that they only become evident upon application of electric fields.

U.S. EPA Water Technology Innovation Cluster Leaders Meeting - "Successfully Supporting Early-Stage Companies: The Role of Technology Testing" Meeting Summary Report

Science.gov (United States)

The goals of this workshop were to: (1) increase the cluster leaders’ level of knowledge regarding past and current water technology testing programs, facilities and requirements; (2) learn from the experiences of technology vendors in getting innovative, commercial-ready product...

Operating experiences of gas purification system of Heavy Water Plant Talcher (Paper No. 1.11)

International Nuclear Information System (INIS)

Bhattacharya, R.; Mohanty, P.R.; Pandey, B.L.

1992-01-01

The operating experiences with the purification system installed at Heavy Water Plant, Talcher for purification of feed synthesis gas from fertilizer plant is described. The purification system has performed satisfactorily even with levels of impurities as much as 15 to 20 ppm of oxygen and carbon monoxide. The system could not however be tested at designed gas throughput and on a sustained basis. However, increase in gas throughput upto the design value is not expected to pose any problem on the performance of the purification system. (author). 5 figs

Validation of the TRACR3D code for soil water flow under saturated/unsaturated conditions in three experiments

International Nuclear Information System (INIS)

Perkins, B.; Travis, B.; DePoorter, G.

1985-01-01

Validation of the TRACR3D code in a one-dimensional form was obtained for flow of soil water in three experiments. In the first experiment, a pulse of water entered a crushed-tuff soil and initially moved under conditions of saturated flow, quickly followed by unsaturated flow. In the second experiment, steady-state unsaturated flow took place. In the final experiment, two slugs of water entered crushed tuff under field conditions. In all three experiments, experimentally measured data for volumetric water content agreed, within experimental errors, with the volumetric water content predicted by the code simulations. The experiments and simulations indicated the need for accurate knowledge of boundary and initial conditions, amount and duration of moisture input, and relevant material properties as input into the computer code. During the validation experiments, limitations on monitoring of water movement in waste burial sites were also noted. 5 references, 34 figures, 9 tables

Limitations of the inspection and testing concepts for pressurised components from the viewpoint of operating experience

International Nuclear Information System (INIS)

Schulz, H.

2001-01-01

The role of in-service inspection and testing is to contribute to a safe and reliable operation of systems, structures and components. It is therefore the objective of inspections and tests to identify malfunctions and degradations at a stage early enough to avoid detrimental impacts on safety as well as on the reliability of the plant. Taking mostly the pressure boundary of German light-water reactors as an example, it is the intention of this paper to analyse how successful present inspection and testing requirements are and to discuss limitations. Based on a review of the world-wide operating experience the following questions of a more generic nature are addressed: - Are the relevant damage mechanisms being addressed in our codes and standards? - What are the criteria to develop a representative scope of inspection? - How to maintain a sufficient level of information for a decreasing number of nuclear power plants in operation? It can be concluded that the revision of codes and standards according to lessons learned from operating experience remains as an ongoing process. Furthermore, the criteria applied to derive a representative scope of inspection need to be addressed in more detail, specifically with respect to corrosion. The continuous evaluation of operating experience of a large number of plants is the most valuable source to identify beginning degradations. (author)

Large block migration experiments: INTRAVAL phase 1, Test Case 9

Energy Technology Data Exchange (ETDEWEB)

Gureghian, A.B.; Noronha, C.J. (Battelle, Willowbrook, IL (USA). Office of Waste Technology Development); Vandergraaf, T.T. (Atomic Energy of Canada Ltd., Ottawa, ON (Canada))

1990-08-01

The development of INTRAVAL Test Case 9, as presented in this report, was made possible by a past subsidiary agreement to the bilateral cooperative agreement between the US Department of Energy (DOE) and Atomic Energy of Canada Limited (AECL) encompassing various aspects of nuclear waste disposal research. The experimental aspect of this test case, which included a series of laboratory experiments designed to quantify the migration of tracers in a single, natural fracture, was undertaken by AECL. The numerical simulation of the results of these experiments was performed by the Battelle Office of Waste Technology Development (OWTD) by calibrating an in-house analytical code, FRACFLO, which is capable of predicting radionuclide transport in an idealized fractured rock. Three tracer migration experiments were performed, using nonsorbing uranine dye for two of them and sorbing Cs-137 for the third. In addition, separate batch experiments were performed to determine the fracture surface and rock matrix sorption coefficients for Cs-137. The two uranine tracer migration experiment were used to calculate the average fracture aperture and to calibrate the model for the fracture dispersivity and matrix diffusion coefficient. The predictive capability of the model was then tested by simulating the third, Cs-137, tracer test without changing the parameter values determined from the other experiments. Breakthrough curves of both the experimental and numerical results obtained at the outlet face of the fracture are presented for each experiment. The reported spatial concentration profiles for the rock matrix are based solely on numerical predictions. 22 refs., 12 figs., 8 tabs.

Water table tests of proposed heat transfer tunnels for small turbine vanes

Science.gov (United States)

Meitner, P. L.

1974-01-01

Water-table flow tests were conducted for proposed heat-transfer tunnels which were designed to provide uniform flow into their respective test sections of a single core engine turbine vane and a full annular ring of helicopter turbine vanes. Water-table tests were also performed for the single-vane test section of the core engine tunnel. The flow in the heat-transfer tunnels was shown to be acceptable.

Synthetic sea water - An improved stress corrosion test medium for aluminum alloys

Science.gov (United States)

Humphries, T. S.; Nelson, E. E.

1973-01-01

A major problem in evaluating the stress corrosion cracking resistance of aluminum alloys by alternate immersion in 3.5 percent salt (NaCl) water is excessive pitting corrosion. Several methods were examined to eliminate this problem and to find an improved accelerated test medium. These included the addition of chromate inhibitors, surface treatment of specimens, and immersion in synthetic sea water. The results indicate that alternate immersion in synthetic sea water is a very promising stress corrosion test medium. Neither chromate inhibitors nor surface treatment (anodize and alodine) of the aluminum specimens improved the performance of alternate immersion in 3.5 percent salt water sufficiently to be classified as an effective stress corrosion test method.

Artificial-Crack-Behavior Test Evaluation of the Water-Leakage Repair Materials Used for the Repair of Water-Leakage Cracks in Concrete Structures

Directory of Open Access Journals (Sweden)

Soo-Yeon Kim

2016-09-01

Full Text Available There are no existing standard test methods at home and abroad that can verify the performance of water leakage repair materials, and it is thus very difficult to perform quality control checks in the field of water leakage repair. This study determined that the key factors that have the greatest impact on the water leakage repair materials are the micro-behaviors of cracks, and proposed an artificial-crack-behavior test method for the performance verification of the repair materials. The performance of the 15 kinds of repair materials that are currently being used in the field of water leakage repair was evaluated by applying the proposed test method. The main aim of such a test method is to determine if there is water leakage by injecting water leakage repair materials into a crack behavior test specimen with an artificial 5-mm crack width, applying a 2.5 mm vertical behavior load at 100 cycles, and applying 0.3 N/mm2 constant water pressure. The test results showed that of the 15 kinds of repair materials, only two effectively sealed the crack and thus stopped the water leakage. The findings of this study confirmed the effectiveness of the proposed artificial-crack-behavior test method and suggest that it can be used as a performance verification method for checking the responsiveness of the repair materials being used in the field of water leakage repair to the repetitive water leakage behaviors that occur in concrete structures. The study findings further suggest that the use of the proposed test method makes it possible to quantify the water leakage repair quality control in the field.

Performance study of protective clothing against hot water splashes: from bench scale test to instrumented manikin test.

Science.gov (United States)

Lu, Yehu; Song, Guowen; Wang, Faming

2015-03-01

Hot liquid hazards existing in work environments are shown to be a considerable risk for industrial workers. In this study, the predicted protection from fabric was assessed by a modified hot liquid splash tester. In these tests, conditions with and without an air spacer were applied. The protective performance of a garment exposed to hot water spray was investigated by a spray manikin evaluation system. Three-dimensional body scanning technique was used to characterize the air gap size between the protective clothing and the manikin skin. The relationship between bench scale test and manikin test was discussed and the regression model was established to predict the overall percentage of skin burn while wearing protective clothing. The results demonstrated strong correlations between bench scale test and manikin test. Based on these studies, the overall performance of protective clothing against hot water spray can be estimated on the basis of the results of the bench scale hot water splashes test and the information of air gap size entrapped in clothing. The findings provide effective guides for the design and material selection while developing high performance protective clothing. Published by Oxford University Press on behalf of the British Occupational Hygiene Society 2014.

Radiation education using local environment. Educational experiment using Misasa spring water

International Nuclear Information System (INIS)

Nakamura, Mariko; Esaka, Takao; Kamata, Masahiro

2005-01-01

Hoping that use of natural radioactivity as teaching materials helps learners to understand the existence of radiation in nature, the authors developed several kinds of safe and inexpensive experiments for elementary and junior high school education using hot spring water taken from Misasa, situated in Tottori prefecture, Japan. Here, they report the details of experimental procedure to observe the radioactive equilibrium between Rn 222 released from the hot spring water and its daughters as well as the decay after isolation from Rn 222. The experiment needs no hazardous chemicals nor Bunsen burners, and can be carried out in normal classrooms without any special apparatus. (S. Ohno)

Experience of 12 kA / 16 V SMPS during the HTS Current Leads Test

Science.gov (United States)

Panchal, P.; Christian, D.; Panchal, R.; Sonara, D.; Purwar, G.; Garg, A.; Nimavat, H.; Singh, G.; Patel, J.; Tanna, V.; Pradhan, S.

2017-04-01

As a part of up gradation plans in SST-1 Tokamak, one pair of 3.3 kA rated prototype hybrid current leads were developed using Di-BSCCO as High Temperature Superconductors (HTS) and the copper heat exchanger. In order to validate the manufacturing procedure prior to go for series production of such current leads, it was recommended to test these current leads using dedicated and very reliable DC switch mode power supply (SMPS). As part of test facility, 12 kA, 16 VDC programmable SMPS was successfully installed, commissioned and tested. This power supply has special features such as modularity, N+1 redundancy, very low ripple voltage, precise current measurements with Direct Current Current Transformer, CC/CV modes with auto-crossover and auto-sequence programming. As a part of acceptance of this converter, A 5.8 mΩ water-cooled resistive dummy load and PLC based SCADA system is designed, developed for commissioning of power supply. The same power supply was used for the testing of the prototype HTS current leads. The paper describes the salient features and experience of state-of-art of power supply and results obtained from this converter during the HTS current leads test.

Lixiviation of polymer matrix parcels of nuclear wastes in an environment with a low water content with respect to the standard characterisation test

International Nuclear Information System (INIS)

Reynaud, Vincent

1996-01-01

It is generally admitted that, in a nuclear waste storage site, a possible return of radionuclides towards the biosphere would mainly occur by leaching of coated items and their transport by natural waters. Therefore, lixiviation properties of coated nuclear wastes are among the most important. The objective of this research thesis is therefore to compare the activity release of samples of ion exchange polymer coated by a polymer (epoxy or polyester) matrix. Two types of tests have been performed: a standard test (sample immersion in water) and a lysimeter test (simulation of the geological environment by means of glass balls). The lixiviation of tritium-containing water is studied after a 300 day long experiment. The modelling of the release of tritium-containing water by using Fick equations gives good results. Factors influencing the lixiviation of cobalt ions and caesium ions are studied, and the lixiviation of these both ions is then modelled [fr

Post-test analysis of ROSA-III experiment Run 702

International Nuclear Information System (INIS)

Koizumi, Yasuo; Kikuchi, Osamu; Soda, Kunihisa

1980-01-01

The purpose of the ROSA-III experiment with a scaled BWR test facility is to examine primary coolant thermal-hydraulic behavior and performance of ECCS during a posturated loss-of-coolant accident of BWR. The results provide information for verification and improvement of reactor safety analysis codes. Run 702 assumed a 200% split break at the recirculation pump suction line under an average core power without ECCS activation. Post - test analysis of the Run 702 experiment was made with computer code RELAP4J. Agreement of the calculated system pressure and the experiment one was good. However, the calculated heater surface temperatures were higher than the measured ones. Also, the axial temperature distribution was different in tendency from the experimental one. From these results, the necessity was indicated of improving the analytical model of void distribution in the core and the nodalization in the pressure vassel, in order to make the analysis more realistic. And also, the need of characteristic test was indicated for ROSA-III test facility components, such as jet pump and piping form loss coefficient; likewise, flow rate measurements must be increased and refined. (author)

The research of materials and water chemistry for supercritical water-cooled reactors in Research Centre Rez

International Nuclear Information System (INIS)

Zychova, Marketa; Fukac, Rostislav; Vsolak, Rudolf; Vojacek, Ales; Ruzickova, Mariana; Vonkova, Katerina

2012-09-01

Research Centre Rez (CVR) is R and D company based in the Czech Republic. It was established as the subsidiary of the Nuclear Research Institute Rez plc. One of the main activities of CVR is the research of materials and chemistry for the generation IV reactor systems - especially the supercritical water-cooled one. For these experiments is CVR equipped by a supercritical water loop (SCWL) and a supercritical water autoclave (SCWA) serving for research of material and Supercritical Water-cooled Reactor (SCWR) environment compatibility experiments. SCWL is a research facility designed to material, water chemistry, radiolysis and other testing in SCWR environment, SCWA serves for complementary and supporting experiments. SCWL consists of auxiliary circuits (ensuring the required parameters as temperature, pressure and chemical conditions in the irradiation channel, purification and measurements) and irradiation channel (where specimens are exposed to the SCWR environment). The design of the loop is based on many years of experience with loop design for various types of corrosion/water chemistry experiments. Designed conditions in the test area of SCWL are 600 deg. C and 25 MPa. SCWL was designed in 2008 within the High Performance Light Water Reactor Phase 2 project and built during 2008 and 2009. The trial operations were performed in 2010 and 2011 and were divided into three phases - the first phase to verify the functionality of auxiliary circuits of the loop, the second phase to verify the complete facility (auxiliary circuits and functional irradiation channel internals) and the third phase to verify the feasibility of corrosion tests with the complete equipment and specimens. All three trial operations were very successful - designed conditions and parameters were reached. (authors)

Oscillatory water sorption test for determining water uptake behavior in bread crust

NARCIS (Netherlands)

Nieuwenhuijzen, N.H. van; Tromp, R.H.; Hamer, R.J.; Vliet, T. van

2007-01-01

In this work, water sorption kinetics of bread crust are described using an oscillatory sorption test in combination with a Langmuir type equation. Both kinetic and thermodynamic information could be obtained at the same time. An advantage of applying a Langmuir type equation for a quantitative

Test results of six-month test of two water electrolysis systems

Science.gov (United States)

Mills, E. S.; Wells, G. W.

1972-01-01

The two water electrolysis systems used in the NASA space station simulation 90-day manned test of a regenerative life support system were refurbished as required and subjected to 26-weeks of testing. The two electrolysis units are both promising systems for oxygen and hydrogen generation and both needed extensive long-term testing to evaluate the performance of the respective cell design and provide guidance for further development. Testing was conducted to evaluate performance in terms of current, pressure, variable oxygen demands, and orbital simulation. An automatic monitoring system was used to record, monitor and printout performance data at one minute, ten minute or one-hour intervals. Performance data is presented for each day of system operation for each module used during the day. Failures are analyzed, remedial action taken to eliminate problems is discussed and recommendations for redesign for future space applications are stated.

Sediment filtration can reduce the N load of the waste water discharge - a full-scale lake experiment

Science.gov (United States)

Aalto, Sanni L.; Saarenheimo, Jatta; Karvinen, Anu; Rissanen, Antti J.; Ropponen, Janne; Juntunen, Janne; Tiirola, Marja

2016-04-01

European commission has obliged Baltic states to reduce nitrate load, which requires high investments on the nitrate removal processes and may increase emissions of greenhouse gases, e.g. N2O, in the waste water treatment plants. We used ecosystem-scale experimental approach to test a novel sediment filtration method for economical waste water N removal in Lake Keurusselkä, Finland between 2014 and 2015. By spatially optimizing the waste water discharge, the contact area and time of nitrified waste water with the reducing microbes of the sediment was increased. This was expected to enhance microbial-driven N transformation and to alter microbial community composition. We utilized 15N isotope pairing technique to follow changes in the actual and potential denitrification rates, nitrous oxide formation and dissimilatory nitrate reduction to ammonium (DNRA) in the lake sediments receiving nitrate-rich waste water input and in the control site. In addition, we investigated the connections between observed process rates and microbial community composition and functioning by using next generation sequencing and quantitative PCR. Furthermore, we estimated the effect of sediment filtration method on waste water contact time with sediment using the 3D hydrodynamic model. We sampled one year before the full-scale experiment and observed strong seasonal patterns in the process rates, which reflects the seasonal variation in the temperature-related mixing patterns of the waste water within the lake. During the experiment, we found that spatial optimization enhanced both actual and potential denitrification rates of the sediment. Furthermore, it did not significantly promote N2O emissions, or N retention through DNRA. Overall, our results indicate that sediment filtration can be utilized as a supplemental or even alternative method for the waste water N removal.

Early effects of water inflow into a deposition hole. Laboratory tests results

International Nuclear Information System (INIS)

Sanden, Torbjoern; Boergesson, Lennart

2010-12-01

During the installation of buffer and canister in a deposition hole a number of different problems can arise. The problems are mainly connected to water flow from fractures in the rock into the deposition hole. At some conditions it probably will be necessary to protect the installed buffer blocks with a special sheet made of rubber or plastic. This report deals with two processes that can occur and are possible to strongly influence the buffer during installation: 1. Erosion. Erosion of bentonite from the deposition hole up to the voids in the tunnel backfill. This process will continue until a tunnel plug have been built and the voids in the backfill are filled with water. 2. Heave. Early wetting of the pellets filling may cause a heave of the buffer blocks that will decrease the density of the buffer. An erosion model has been suggested /Sanden et al. 2008b/ which makes it possible to estimate the amount of eroded material for a certain water flow rate during a certain time. In order to verify the model and investigate how the buffer in a deposition hole behaves when exposed to a water flow, a number of different tests have been performed: - Test type 1. Simulation of water flow out from a deposition hole. The deposition hole was made of steel and had a radial scale of 1:4 and a height of 0.6 meter. The pellets slot was scaled 1:1. After filling the deposition hole with buffer blocks and pellets, a constant water flow was applied in a point at the bottom. The discharged water at the top was collected and the amount of eroded material determined. The displacement of the blocks and pellets surfaces was also measured during the test. - Test type 2. The influence of test length on the erosion rate was investigated by performing tests with Plexiglas tubes of different lengths (0.4 and 4 meter). The tubes were positioned vertically, filled with pellets and a point inflow was applied in the bottom. The discharged water was collected at the top and the amount of eroded

Early effects of water inflow into a deposition hole. Laboratory tests results

Energy Technology Data Exchange (ETDEWEB)

Sanden, Torbjoern; Boergesson, Lennart (Clay Technology AB, Lund (Sweden))

2010-12-15

During the installation of buffer and canister in a deposition hole a number of different problems can arise. The problems are mainly connected to water flow from fractures in the rock into the deposition hole. At some conditions it probably will be necessary to protect the installed buffer blocks with a special sheet made of rubber or plastic. This report deals with two processes that can occur and are possible to strongly influence the buffer during installation: 1. Erosion. Erosion of bentonite from the deposition hole up to the voids in the tunnel backfill. This process will continue until a tunnel plug have been built and the voids in the backfill are filled with water. 2. Heave. Early wetting of the pellets filling may cause a heave of the buffer blocks that will decrease the density of the buffer. An erosion model has been suggested /Sanden et al. 2008b/ which makes it possible to estimate the amount of eroded material for a certain water flow rate during a certain time. In order to verify the model and investigate how the buffer in a deposition hole behaves when exposed to a water flow, a number of different tests have been performed: - Test type 1. Simulation of water flow out from a deposition hole. The deposition hole was made of steel and had a radial scale of 1:4 and a height of 0.6 meter. The pellets slot was scaled 1:1. After filling the deposition hole with buffer blocks and pellets, a constant water flow was applied in a point at the bottom. The discharged water at the top was collected and the amount of eroded material determined. The displacement of the blocks and pellets surfaces was also measured during the test. - Test type 2. The influence of test length on the erosion rate was investigated by performing tests with Plexiglas tubes of different lengths (0.4 and 4 meter). The tubes were positioned vertically, filled with pellets and a point inflow was applied in the bottom. The discharged water was collected at the top and the amount of eroded

The water run of the Lavia test borehole - an interpritation of the test data

International Nuclear Information System (INIS)

Ylinen, Arto

1986-04-01

In 1984 a 1001 metres deep testhole was drilled in Lavia, Western Finland, for developing and testing equipments and methods to investigate the properties of rock for its suitability as a site for a nuclear waste repository. During the same year, constant head injection tests were performed in the hole. The method was two-packer test with 30 metres long test interval. Totally 30, mainly 2 hours' 2000kPa, tests were made in the subsequent depths. The data, pressure, flow rate and temperature, during the injection and fall-off phases, were stored in the numerical form in micro-computer diskettes. The aim of this study is to make an interpretation of the test data for getting knowledge about the rock ability of conducting water. The traditional homogeneous Horner and reciprocal flow interpretations are made and the water conducting parametres are estimated. Also a new method of taking into account the differences in fracture types (fillings, roughness, shape, etc.) and the causing different transmissivity, is derived and used for the fracture interpretation. As a result, a dense spectrum for flow parametres is obtained from the same data. Finally, the relevance of different parametres and their suitability for testing and modelling the flow in the rock is evaluated, and new schemes for further investigation are proposed. (author)