Ground Source Heat Pump Sub-Slab Heat Exchange Loop Performance in a Cold Climate

Energy Technology Data Exchange (ETDEWEB)

Mittereder, N.; Poerschke, A.

2013-11-01

This report presents a cold-climate project that examines an alternative approach to ground source heat pump (GSHP) ground loop design. The innovative ground loop design is an attempt to reduce the installed cost of the ground loop heat exchange portion of the system by containing the entire ground loop within the excavated location beneath the basement slab. Prior to the installation and operation of the sub-slab heat exchanger, energy modeling using TRNSYS software and concurrent design efforts were performed to determine the size and orientation of the system. One key parameter in the design is the installation of the GSHP in a low-load home, which considerably reduces the needed capacity of the ground loop heat exchanger. This report analyzes data from two cooling seasons and one heating season. Upon completion of the monitoring phase, measurements revealed that the initial TRNSYS simulated horizontal sub-slab ground loop heat exchanger fluid temperatures and heat transfer rates differed from the measured values. To determine the cause of this discrepancy, an updated model was developed utilizing a new TRNSYS subroutine for simulating sub-slab heat exchangers. Measurements of fluid temperature, soil temperature, and heat transfer were used to validate the updated model.

ATLAS Cold Leg Top Slot Break Analysis using RELAP5

Energy Technology Data Exchange (ETDEWEB)

Sung, Haejung; Lee, Sang Ik; Park, Ju-Hyun; Choi, Tong-Soo [KEPCO NF, Daejeon (Korea, Republic of)

2016-10-15

U.S. Nuclear Regulatory Commission (US-NRC) has been reviewing the design certification application for APR1400 submitted by Korea Electric Power Corporation (KEPCO). The main concern about cold leg top slot break is that cladding temperature might be increased by core uncover due to four loop seal reformation following flooding of safety injection water. An integral effect test for cold leg top slot break was performed by KAERI (Korea Atomic Energy Research Institute) using ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation), which is a scaled down experimental facility for APR1400. In this study, RELAP5/MOD3.3/Patch04 is assessed by experimental result of ATLAS cold leg top slot break. Also, thermal hydraulic phenomena by four loop seals reformation is observed by RELAP5 result. The RELAP5/MOD3.3/Patch04 is assessed by the experimental result of ATLAS cold leg top slot break. The top slot break is described by offtake model, and the mass flow rate is fairly well estimated. The RELAP5 well predicts the correlation between general trend and four loop seal reformation. The pressure of the core region and the cladding temperature tends to increase during four loop seal reformation due to steam path blockage on four loop seals. It is presumed that the code cannot estimate two phase phenomena by loop seal clearing as same as experiments. In terms of cladding temperature, loop seal reformation due to loop seal elevation of APR1400 does not need to be the issue, since the void fraction at the active top core is maintained over 0.4.

Cold Vacuum Drying (CVD) OCRWM Loop Error Determination

International Nuclear Information System (INIS)

PHILIPP, B.L.

2000-01-01

Characterization is specifically identified by the Richland Operations Office (RL) for the Office of Civilian Radioactive Waste Management (OCRWM) of the US Department of Energy (DOE), as requiring application of the requirements in the Quality Assurance Requirements and Description (QARD) (RW-0333P DOE 1997a). Those analyses that provide information that is necessary for repository acceptance require application of the QARD. The cold vacuum drying (CVD) project identified the loops that measure, display, and record multi-canister overpack (MCO) vacuum pressure and Tempered Water (TW) temperature data as providing OCRWM data per Application of the Office of Civilian Radioactive Waste Management (OCRWM) Quality Assurance Requirements to the Hanford Spent Nuclear Fuel Project HNF-SD-SNF-RPT-007. Vacuum pressure transmitters (PT 1*08, 1*10) and TW temperature transmitters (TIT-3*05, 3*12) are used to verify drying and to determine the water content within the MCO after CVD

Thermodynamic analysis of a novel dual-loop organic Rankine cycle for engine waste heat and LNG cold

International Nuclear Information System (INIS)

Sung, Taehong; Kim, Kyung Chun

2016-01-01

Highlights: • A novel dual ORC system is designed for engine waste heat and LNG cold. • Exhaust gas and jacket cooling water are considered as heat sources. • LNG and boil-off gas are considered as heat sinks. • ORC loops are optimized to produce the maximum net work output. - Abstract: The marine sector produces a large portion of total air pollution, so the emissions of the engines used must be improved. This can be achieved using a new eco-friendly engine and waste-heat recovery system. A dual-fuel (DF) engine has been introduced for LNG carriers that is eco-friendly and has high thermal efficiency since it uses natural gas as fuel. The thermal efficiency could be further improved with the organic Rankine cycle (ORC). A novel dual-loop ORC system was designed for DF engines. The upper ORC loop recovers waste heat from the exhaust gas, and the bottom ORC loop recovers waste heat from the jacket cooling water and LNG cold. Both ORC loops were optimized to produce the maximum net work output. The optimum simple dual-loop ORC with n-pentane and R125 as working fluids produces an additional power output of 729.1 kW, which is 4.15% of the original engine output. Further system improvement studies were conducted using a recuperator and preheater, and the feasibility of using boil-off gas as a heat sink was analyzed. Optimization of the system configuration revealed that the preheater and recuperator with n-pentane and R125 as working fluids increase the maximum net work output by 906.4 kW, which is 5.17% of the original engine output.

Results of molten salt panel and component experiments for solar central receivers: Cold fill, freeze/thaw, thermal cycling and shock, and instrumentation tests

Energy Technology Data Exchange (ETDEWEB)

Pacheco, J.E.; Ralph, M.E.; Chavez, J.M.; Dunkin, S.R.; Rush, E.E.; Ghanbari, C.M.; Matthews, M.W.

1995-01-01

Experiments have been conducted with a molten salt loop at Sandia National Laboratories in Albuquerque, NM to resolve issues associated with the operation of the 10MW{sub e} Solar Two Central Receiver Power Plant located near Barstow, CA. The salt loop contained two receiver panels, components such as flanges and a check valve, vortex shedding and ultrasonic flow meters, and an impedance pressure transducer. Tests were conducted on procedures for filling and thawing a panel, and assessing components and instrumentation in a molten salt environment. Four categories of experiments were conducted: (1) cold filling procedures, (2) freeze/thaw procedures, (3) component tests, and (4) instrumentation tests. Cold-panel and -piping fill experiments are described, in which the panels and piping were preheated to temperatures below the salt freezing point prior to initiating flow, to determine the feasibility of cold filling the receiver and piping. The transient thermal response was measured, and heat transfer coefficients and transient stresses were calculated from the data. Freeze/thaw experiments were conducted with the panels, in which the salt was intentionally allowed to freeze in the receiver tubes, then thawed with heliostat beams. Slow thermal cycling tests were conducted to measure both how well various designs of flanges (e.g., tapered flanges or clamp type flanges) hold a seal under thermal conditions typical of nightly shut down, and the practicality of using these flanges on high maintenance components. In addition, the flanges were thermally shocked to simulate cold starting the system. Instrumentation such as vortex shedding and ultrasonic flow meters were tested alongside each other, and compared with flow measurements from calibration tanks in the flow loop.

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.)

UPTF loop seal tests and their RELAP simulation

International Nuclear Information System (INIS)

Tuomainen, M.; Tuunanen, J.

1997-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-11-21

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

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

International Nuclear Information System (INIS)

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

2004-10-01

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

Technical specification of HANARO fuel test loop

Energy Technology Data Exchange (ETDEWEB)

Kim, J. Y

1998-03-01

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

Analysis of severe accidents on fast reactor test loop

International Nuclear Information System (INIS)

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

1975-01-01

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

Liquid Lead-Bismuth Materials Test Loop

International Nuclear Information System (INIS)

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

2002-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2002-04-01

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

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

The Numerical Sensitivity Study of Cold Leg Top Slot Break for ATLAS

Energy Technology Data Exchange (ETDEWEB)

Sung, Hae Jung; Lee, Sang Ik; Park, Ju hyun; Choi, Tong Soo [KEPCO, Daejeon (Korea, Republic of)

2016-05-15

In this study, cold leg top slot break calculations are performed by RELAP5/MOD3.3 Patch04 for the ATLAS test which is the scaled down experimental facility for the APR1400. The test condition for base case is selected as 0.1016 m (4 in.) based on the break size of the APR1400. In addition, sensitivity studies about break size, break distance from vessel, and pressurizer location are performed until the initiation of simultaneous injection of 2.5 hours. The loop seal reformation occurs early, and the duration of final loop seal reformation is longer as the break is close to vessel. Nonetheless, PCT increased by loop seal reformation is not identified since core uncovery does not occur. In this study, it is confirmed through RELAP5 simulation of the ATLAS test that cold leg top slot break for the APR1400 is not a safety issue in perspective of the loop seal reformation.

Thermal hydraulic considerations and mock-up tests for developing two-phase thermo-siphon loop of CARR-CNS

International Nuclear Information System (INIS)

Shejiao, Du; Qincheng, Bi; Tingkuan, Chen; Quanke, Feng

2005-01-01

The main component of the China Advanced Research Reactor Cold Neutron Source (CARR-CNS), which is under design, is a two-phase thermo-siphon loop of hydrogen. It consists of a condenser, a single tube with counter current flow avoiding flooding and a cylindrical-annulus moderator cell. The mockup tests were carried out using a full-scale loop with Freon-113, to validate the self-regulating characteristics of the loop, void fraction less than 20% in the liquid of the moderator cell and the requirements for establishing the condition under which the inner shell of the moderator cell has only vapor and the outer shell liquid. In the case of these mockup tests the density ratio of liquid to vapor and the volumetric vapor evaporation rate due to heat load are kept the same as those in normal operation of the CARR-CNS. The results show that the loop has the self-regulating characteristics and the inner shell of the moderator cell contains only vapor, the outer shell liquid. The average void fraction of the moderator cell was verified less than 20% under the volumetric vapor generation of 0.65 l/s corresponding to the nuclear heating of 800 W in the case of the liquid hydrogen. The local void fraction in the liquid hydrogen increases with the increase of the loop pressure under the condition of a constant volumetric evaporation

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-07-01

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

Xenon oscillation tests in four-loop PWR cores

International Nuclear Information System (INIS)

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

1980-01-01

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

305 Building Cold Test Facility Management Plan

International Nuclear Information System (INIS)

Whitehurst, R.

1994-01-01

This document provides direction for the conduct of business in Building 305 for cold testing tools and equipment. The Cold Test Facility represents a small portion of the overall building, and as such, the work instructions already implemented in the 305 Building will be utilized. Specific to the Cold Test there are three phases for the tools and equipment as follows: 1. Development and feature tests of sludge/fuel characterization equipment, fuel containerization equipment, and sludge containerization equipment to be used in K-Basin. 2. Functional and acceptance tests of all like equipment to be installed and operated in K-Basin. 3. Training and qualification of K-Basin Operators on equipment to be installed and operated in the Basin

Conceptual Design for a High-Temperature Gas Loop Test Facility

Energy Technology Data Exchange (ETDEWEB)

James B. Kesseli

2006-08-01

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

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

International Nuclear Information System (INIS)

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

1981-12-01

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

Summary of ALSEP Test Loop Solvent Irradiation Testing

Energy Technology Data Exchange (ETDEWEB)

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

2016-08-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-12-15

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

Hardware-in-the-Loop Testing

Data.gov (United States)

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

LMFBR with booster pump in pumping loop

International Nuclear Information System (INIS)

Rubinstein, H.J.

1975-01-01

A loop coolant circulation system is described for a liquid metal fast breeder reactor (LMFBR) utilizing a low head, high specific speed booster pump in the hot leg of the coolant loop with the main pump located in the cold leg of the loop, thereby providing the advantages of operating the main pump in the hot leg with the reliability of cold leg pump operation

Accident analysis of HANARO fuel test loop

Energy Technology Data Exchange (ETDEWEB)

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

1998-03-01

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

Gas Test Loop Booster Fuel Hydraulic Testing

International Nuclear Information System (INIS)

Gas Test Loop Hydraulic Testing Staff

2006-01-01

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

Gas Test Loop Booster Fuel Hydraulic Testing

Energy Technology Data Exchange (ETDEWEB)

Gas Test Loop Hydraulic Testing Staff

2006-09-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Kim, J. Y.

1997-10-01

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

Light Duty Utility Arm system pre-operational (cold test) test plan

International Nuclear Information System (INIS)

Bennett, K.L.

1995-01-01

The Light Duty Utility (LDUA) Cold Test Facility, located in the Hanford 400 Area, will be used to support cold testing (pre- operational tests) of LDUA subsystems. Pre-operational testing is composed of subsystem development testing and rework activities, and integrated system qualification testing. Qualification testing will be conducted once development work is complete and documentation is under configuration control. Operational (hot) testing of the LDUA system will follow the testing covered in this plan and will be covered in a separate test plan

French nuclear plant safeguard pump qualification testing: EPEC test loop

International Nuclear Information System (INIS)

Guesnon, H.

1985-01-01

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

Fast Flux Test Facility primary sodium valves

International Nuclear Information System (INIS)

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

1977-01-01

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

ac power control in the Core Flow Test Loop

International Nuclear Information System (INIS)

McDonald, D.W.

1980-01-01

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

Cold moderator test facilities working group

International Nuclear Information System (INIS)

Bauer, Guenter S.; Lucas, A. T.

1997-09-01

The working group meeting was chaired by Bauer and Lucas.Testing is a vital part of any cold source development project. This applies to specific physics concept verification, benchmarking in conjunction with computer modeling and engineering testing to confirm the functional viability of a proposed system. Irradiation testing of materials will always be needed to continuously extend a comprehensive and reliable information database. An ever increasing worldwide effort to enhance the performance of reactor and accelerator based neutron sources, coupled with the complexity and rising cost of building new generation facilities, gives a new dimension to cold source development and testing programs. A stronger focus is now being placed on the fine-tuning of cold source design to maximize its effectiveness in fully exploiting the facility. In this context, pulsed spallation neutron sources pose an extra challenge due to requirements regarding pulse width and shape which result from a large variety of different instrument concepts. The working group reviewed these requirements in terms of their consequences on the needs for testing equipment and compiled a list of existing and proposed facilities suitable to carry out the necessary development work.

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

International Nuclear Information System (INIS)

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

1982-01-01

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

Conceptual Design of Forced Convection Molten Salt Heat Transfer Testing Loop

Energy Technology Data Exchange (ETDEWEB)

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

2010-09-01

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

A simple blowdown code for SUPER-SARA loop conditions

International Nuclear Information System (INIS)

Fritz, G.

1981-01-01

The Super Sara test programme (SSTP) is aimed to study in pile the fuel and cluster behaviour under two types of accident conditions: - the ''Large break loss of coolant'' condition (LB-Loca), - the ''Severe fuel damage'' (SFD) in a boildown caused by a small break. BIVOL was developed for the LB-Loca situation. This code is made for a loop where essentially two volumes define the thermohydraulics during the blowdown. In the SUPERSARA loop these two volumes are represented by the hot leg and cold leg pipings together with the respective upper and lower plenum of the test section

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

Science.gov (United States)

Balouch, Masih N.

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

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

International Nuclear Information System (INIS)

Cao Xuewu; Li Zhaohuan

1991-07-01

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

Current status for TRR-II Cold Neutron Source

International Nuclear Information System (INIS)

Lee, C.H.; Guung, T.C.; Lan, K.C.; Wang, C.H.; Chan, Y.K.; Shieh, D.J.

2001-01-01

The Taiwan Research Reactor (TRR) project (TRR-II) is carrying out at Institute of Nuclear Energy Research (INER) from October 1998 to December 2006. The purpose of Cold Neutron Source (CNS) project is to build entire CNS facility to generate cold neutrons within TRR-II reactor. The objective of CNS design is to install CNS facility with a competitive brightness of cold neutron beam to other facilities in the world. Based on the TRR-II CNS project schedule, the conceptual design for TRR-II CNS facility has been completed and the mock-up test facility for full-scale hydrogen loop has been designed. (author)

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

International Nuclear Information System (INIS)

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

2004-12-01

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

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

Grooved cold moderator tests

International Nuclear Information System (INIS)

Inoue, K.; Kiyanagi, Y.; Iwasa, H.; Watanabe, N.; Ikeda, S.; Carpenter, J.M.; Ishikawa, Y.

1983-01-01

We performed some grooved cold moderator experiments for methane at 20 K by using the Hokkaido University linac to obtain information to be used in the planning of the KENS-I' project. Cold neutron gains, spatial distribution of emitted beams and time distribution of the neutrons in the grooved cold moderator were measured. Furthermore, we assessed the effects of the grooved cold moderator on the performances of the spectrometers presently installed at the KENS-I cold source. We concluded that the grooved cold moderator benefited appreciably the performances of the spectrometers

Assessment of RELAP/MOD3 using BETHSY 6.2TC 6-inch cold leg side break comparative test

International Nuclear Information System (INIS)

Chung, Young-Jong; Jeong, Jae-Jun; Chang, Won-Pyo; Kim, Dong-Su

1996-10-01

This report presents the results of the RELAP5/MOD3 Version 7j assessment on BETHSY 6.2TC. BETHSY 6.2TC test corresponding to a six inch cold leg break LOCA of the Pressurizer Water Reactor(PWR). The primary objective of the test was to provide reference data of two facilities of different scales (BETHSY and LSTF facility). On the other hand, the present calculation aims at analysis of RELAP5/N4OD3 capability on the small break LOCA simulation, The results of calculation have shown that the RELAP5/MOD3 reasonably predicts occurrences as well as trends of the major phenomena such as primary pressure, timing of loop seal clearing, liquid hold up, etc. However, some disagreements also have been found in the predictions of loop seal clearing, collapsed core water level after loop seal clearing, and accumulator injection behaviors. For better understanding of discrepancies in same predictions, several sensitivity calculations have been performed as well. These include the changes of two-phase discharge coefficient at the break junction and some corrections of the interphase drag term. As result, change of a single parameter has not improved the overall predictions and it has been found that the interphase drag model has still large uncertainties

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

Cold Leak Tests of LHC Beam Screens

CERN Document Server

Collomb-Patton, C; Jenninger, B; Kos, N

2009-01-01

In order to guide the high energy proton beams inside its two 27 km long vacuum rings, the Large Hadron Collider (LHC) at CERN, Geneva, makes use of superconducting technology to create the required magnetic fields. More than 4000 beam screens, cooled at 7 20 K, are inserted inside the 1.9 K beam vacuum tubes to intercept beam induced heat loads and to provide dynamic vacuum stability. As extremely high helium leak tightness is required, all beam screens have been leak tested under cold conditions in a dedicated test stand prior to their installation. After describing the beam screen design and its functions, this report focuses on the cold leak test sequence and discusses the results.

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

Institute of Scientific and Technical Information of China (English)

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

2017-01-01

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

Seismic proving test of BWR primary loop recirculation system

International Nuclear Information System (INIS)

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

1987-01-01

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

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

Failure of PWR-RHRS under cold shutdown conditions: Experimental results from the PKL test facility

International Nuclear Information System (INIS)

Mandl, R.M.; Umminger, K.J.; Logt, J.V.D.

1991-01-01

The Residual Heat Removal System (RHRS) of a PWR is designed to transfer thermal energy from the core after plant shutdown and maintain the plant in cold shutdown or refuelling conditions for extended periods of time. Initial reactor cooling after shutdown is achieved by dissipating heat through the steam generators (SGs) and discharging steam to the condenser by means of the Turbine Bypass System (TBS). When the reactor coolant temperature has dropped to about 160C and pressure has been reduced to 30 bar the RHRS is placed into operation. it reduces the coolant temperature to 50C within 20 hours after shutdown. The time margin for establishing alternate methods of heat removal following a failure of the RHRS depends on the Reactor Coolant System (RCS) temperature, the decay heat rate and the amount of RCS inventory. During some shutdown operations the RCS may be partially drained (e. g. to perform SG inspections). Decreased primary system inventory can significantly reduce the time available to recover the RHRS's function prior to bulk boiling and possible core uncovery. In the PKL test facility, which simulates a 1,300 MWe 4-loop PWR on a scale 1:145, a failure of RHRS under cold shutdown conditions was performed. This presentation gives a brief description of the test facility followed by the test objectives and results of this experiment

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

International Nuclear Information System (INIS)

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

1982-01-01

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

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

International Nuclear Information System (INIS)

Oakley, D.J.

1978-01-01

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

Summary of Test Results From a 1 kWe-Class Free-Piston Stirling Power Convertor Integrated With a Pumped NaK Loop

Science.gov (United States)

Briggs, Maxwell H.; Geng, Steven M.; Pearson, J. Boise; Godfroy, Thomas J.

2010-01-01

As a step towards development of Stirling power conversion for potential use in Fission Surface Power (FSP) systems, a pair of commercially available 1 kW class free-piston Stirling convertors was modified to operate with a NaK liquid metal pumped loop for thermal energy input. This was the first-ever attempt at powering a free-piston Stirling engine with a pumped liquid metal heat source and is a major FSP project milestone towards demonstrating technical feasibility. The tests included performance mapping the convertors over various hot and cold-end temperatures, piston amplitudes and NaK flow rates; and transient test conditions to simulate various start-up and fault scenarios. Performance maps of the convertors generated using the pumped NaK loop for thermal input show increases in power output over those measured during baseline testing using electric heating. Transient testing showed that the Stirling convertors can be successfully started in a variety of different scenarios and that the convertors can recover from a variety of fault scenarios.

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

International Nuclear Information System (INIS)

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

2005-12-01

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

Status of TRR-II cold neutron source

International Nuclear Information System (INIS)

Lee, C.H.; Guung, T.C.; Lan, K.C.; Chan, Y.K.; Wang, C.H.; Chen, S.K.

2001-01-01

The Taiwan research reactor improvement and the utilization promotion project (TRR-II) with a vertical cold neutron source (CNS) is carrying out at the Institute of Nuclear Energy Research (INER). The CNS with a two-phase thermosiphon loop consists of an annular cylindrical moderator cell, a single moderator transfer tube and a condenser. A cylindrical annulus moderator cell with boiling liquid hydrogen at 1.2 bar and 20.7 K gives an optimum moderation for cold neutrons in the wavelength range between 4 A and 15 A. The moderator cell lies around 400 mm away from the core center. Its perturbed thermal flux is about 1.4 x 10 14 cm -2 s -1 . It is close to the maximum thermal neutron flux area in D 2 O tank to get the maximum possible brightness about 1 x 10 12 n cm -2 s -1 A -1 sterad -1 at 4 A. An experimental study for thermal-hydraulic characteristics of the two-phase thermosiphon loop has been performed on a full-scale mockup loop using a Freon-11 as a working fluid. The objective of the mockup testing is to validate operation and heat removal capacity in CNS hydrogen loop design. Moreover, this loop will be used to demonstrate no onset of flooding and flow oscillations in a single transfer tube under CNS normal and abnormal conditions. The flooding limitation, the liquid level, and the void fraction in the moderator cell as a function of the initial Freon-11 inventory, the heat load, and the moderator cell geometry are also reported. (orig.)

Scoping erosion flow loop test results in support of Hanford WTP

International Nuclear Information System (INIS)

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

2015-01-01

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

Bipolar plasma vaporization using plasma-cutting and plasma-loop electrodes versus cold-knife transurethral incision for the treatment of posterior urethral stricture: a prospective, randomized study.

Science.gov (United States)

Cai, Wansong; Chen, Zhiyuan; Wen, Liping; Jiang, Xiangxin; Liu, Xiuheng

2016-01-01

Evaluate the efficiency and safety of bipolar plasma vaporization using plasma-cutting and plasma-loop electrodes for the treatment of posterior urethral stricture. Compare the outcomes following bipolar plasma vaporization with conventional cold-knife urethrotomy. A randomized trial was performed to compare patient outcomes from the bipolar and cold-knife groups. All patients were assessed at 6 and 12 months postoperatively via urethrography and uroflowmetry. At the end of the first postoperative year, ureteroscopy was performed to evaluate the efficacy of the procedure. The mean follow-up time was 13.9 months (range: 12 to 21 months). If re-stenosis was not identified by both urethrography and ureteroscopy, the procedure was considered "successful". Fifty-three male patients with posterior urethral strictures were selected and randomly divided into two groups: bipolar group (n=27) or cold-knife group (n=26). Patients in the bipolar group experienced a shorter operative time compared to the cold-knife group (23.45±7.64 hours vs 33.45±5.45 hours, respectively). The 12-month postoperative Qmax was faster in the bipolar group than in the cold-knife group (15.54±2.78 ml/sec vs 18.25±2.12 ml/sec, respectively). In the bipolar group, the recurrence-free rate was 81.5% at a mean follow-up time of 13.9 months. In the cold-knife group, the recurrence-free rate was 53.8%. The application of bipolar plasma-cutting and plasma-loop electrodes for the management of urethral stricture disease is a safe and reliable method that minimizes the morbidity of urethral stricture resection. The advantages include a lower recurrence rate and shorter operative time compared to the cold-knife technique.

Evaluation of cold testing for Tokai Vitrification Facility

International Nuclear Information System (INIS)

Yoshioka, Masahiro; Inada, Eiichi

1994-01-01

The cold testing of the Tokai Vitrification Facility (TVF) was completed at the end of March, 1994 through the tests of nearly two years since May in 1992. The cold testing was carried out in order to evaluate the process equipment, product quality control, remote maintenance capability. The test results shown that TVF has enough performance with safety to treat the liquid waste in each process, and to control the product quality. For the remote maintenance of process equipment in the vitrification cell, the remote maintenance capability was confirmed for all remote equipment in the cell. The improvements were taken for some equipment with problem from the point of the operability and maintenance. It was confirmed by these test results that the TVF can go forward to the hot test operation using actual waste. (author)

Supercritical CO2 test loop operation and first test results

International Nuclear Information System (INIS)

Wright, Steven A.; Pickard, Paul S.

2009-01-01

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

Smart Home Hardware-in-the-Loop Testing

Energy Technology Data Exchange (ETDEWEB)

Pratt, Annabelle

2017-07-12

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

Experimental loop for SH2 (LECS)

International Nuclear Information System (INIS)

Strehar, N.R.; Bruzzoni, Pablo; Moras, J.J.; Cogozzo, E.O.

1981-01-01

An experimental loop is described for circulation of SH 2 that operates at 2 x 10 6 Pascal and 33 deg C. It was designed and constructed with the purpose of experimentally studying the hydraulic instability phenomenon that can be detected in cold isotopic exchange columns in the Girdler-Sulfide (GS) process of heavy water production. The main features of the different components of the loop are described, as well as the materials, the measurement and control instruments and the auxiliary equipment used, and finally the measuring methods to qualify and quantify the formation of froth. Furthermore, the loop's transportable metallic container is described, which allows to transport and connect it to CNEA's experimental heavy water plant or to any other heavy water plants using the GS method. Some tests made with inert gases that intended to verify the equipment's performance and to select the most adequate sieve trays for its operation are discussed. (M.E.L.) [es

The process and risk of the CPR1000 cold function test in the cold area

International Nuclear Information System (INIS)

Liu Tinghao; Zhang Jian; Ji Dapeng; Shi Quanjian; Tian Kuo

2014-01-01

Hong yanhe nuclear power station is the first CPR1000 reactor which is under construction in the cold area of north China. It is also the first time to carry out the cold functional test (CFT) in the winter of north China. The preparation and process of CFT are described in the paper. According to the experience feedback of CFT of Unit 1, the risk and solution which are significance for the CFT of the other NPS in the cold area are analysed. (authors)

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

International Nuclear Information System (INIS)

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

1987-06-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-07-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Test of cold asphalt storability based on alternative approaches

Science.gov (United States)

Abaffyová, Zora; Komačka, Jozef

2017-09-01

Cold asphalt products for potholes repairs should be workable (soft enough) for long time to ensure their applicability. Storability is assessed indirectly using various tests of workability. Therefore, simple test methods (self-compaction and disintegration test) was developed and verified to investigate changes of storability of this group of cold asphalts. Selfcompaction of the tested mixture in the upturned Abram’s cone for the cement concrete slump test and in the mould for the California Bearing Ratio test was assessed in first stage. After that the video record of disintegration test was taken. During this test, the mould was lifted up and the mixture fell off the mould (Abram’s cone) or disintegrate (CBR mould). The drop of surface after 10 min self-compaction and netto time related to falling out or disintegration of the mixture were used to evaluate the mixture from storability point of view. It was found out the self-compaction test has not a potential to reveal and prove changes of mixture properties. Based on the disintegration test results it can be stated this test at 5 °C using the upturned Abram’s cone could be a suitable approach to determine qualitative changes of a cold mixture from storability point of view.

A one-loop test of string duality

International Nuclear Information System (INIS)

Vafa, C.

1995-01-01

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

The High Flux Isotope Reactor (HFIR) cold source project at ORNL

International Nuclear Information System (INIS)

Selby, D.L.; Lucas, A.T.; Chang, S.J.; Freels, J.D. . E-mail-yb2@ornl.gov

1998-01-01

Following the decision to cancel the Advanced Neutron Source (ANS) Project at Oak Ridge National Laboratory (ORNL), it was determined that a hydrogen cold source should be retrofitted into an existing beam tube of the High Flux Isotope Reactor (HFIR) at ORNL. The preliminary design of this system has been completed and an 'approval in principle' of the design has been obtained from the internal ORNL safety review committees and the U.S. Department of Energy (DOE) safety review committee. The cold source concept is basically a closed loop forced flow supercritical hydrogen system. The supercritical approach was chosen because of its enhanced stability in the proposed high heat flux regions. Neutron and gamma physics of the moderator have been analyzed using the 3D Monte Carlo code MCNP 1 A D structural analysis model of the moderator vessel, vacuum tube, and beam tube was completed to evaluate stress loadings and to examine the impact of hydrogen detonations in the beam tube. A detailed ATHENA 2 system model of the hydrogen system has been developed to simulate loop performance under normal and off-normal transient conditions. Semi-prototypic hydrogen loop tests of the system have been performed at the Arnold Engineering Design Center (AEDC) located in Tullahoma, Tennessee to verify the design and benchmark the analytical system model. A 3.5 kW refrigerator system has been ordered and is expected to be delivered to ORNL by the end of this calendar year. Our present schedule shows the assembling of the cold source loop on site during the fall of 1999 for final testing before insertion of the moderator plug assembly into the reactor beam tube during the end of the year 2000. (author)

Installation Test of Cold Neutron Soruce In-pool Assembly

International Nuclear Information System (INIS)

Lee, Kye Hong; Choi, J.; Wu, S. I.; Kim, Y. K.; Cho, Y. G.; Lee, C. H.; Kim, K. R.

2006-04-01

Before installation of the final cold neutron source in-pool assembly (IPA) in the vertical CN hole at the HANARO, the research reactor, the installation test of IPA has been conducted in the CN hole of the reactor using a full-scaled mock-up in-pool assembly. The well-known cold neutron sources, being safely operated or being now constructed, had been constructed together with each research reactor; therefore, there was little limitation to obtain the optimal cold neutron source since a cold neutron source had been decided to be installed in the reactor from the beginning of the design for the reactor construction. Unlikely, the HANARO has been operated for 10 years so that we have got lots of design limitation in terms of the decisions in the optimal shape, size, minimal light-water gap, and adhesion degree to the CN beam tube, IPA installation tools, etc. for the construction of the CNS. Accordingly, the main objective of this test is to understand any potential problem or interference happened inside the reactor by installing the mock-up IPA and installation bracket. The outcomes from this test is reflected on the finalizing process of the IPA detail design

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

Summary of Test Results From a 1 kW(sub e)-Class Free-Piston Stirling Power Convertor Integrated With a Pumped NaK Loop

Science.gov (United States)

Briggs, Maxwell H.; Geng, Steven M.; Pearson, J. Boise; Godfroy, Thomas J.

2010-01-01

As a step towards development of Stirling power conversion for potential use in Fission Surface Power (FSP) systems, a pair of commercially available 1 kW class free-piston Stirling convertors was modified to operate with a NaK liquid metal pumped loop for thermal energy input. This was the first-ever attempt at powering a free-piston Stirling engine with a pumped liquid metal heat source and is a major FSP project milestone towards demonstrating technical feasibility. The tests included performance mapping the convertors over various hot and cold-end temperatures, piston amplitudes and NaK flow rates; and transient test conditions to simulate various start-up and fault scenarios. Performance maps of the convertors generated using the pumped NaK loop for thermal input show increases in power output over those measured during baseline testing using electric heating. Transient testing showed that the Stirling convertors can be successfully started in a variety of different scenarios and that the convertors can recover from a variety of fault scenarios.

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-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

Optimal tests for electroweak loop effects

International Nuclear Information System (INIS)

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

1986-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-07-01

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

A study of sodium oxide crystallization mechanisms and kinetics in cold traps

International Nuclear Information System (INIS)

Latge, C.

1984-04-01

After showing up the present lack of data on crystallization mechanisms and kinetics, a number of tests were conducted on a sodium test loop equipped with two experimental cold traps. The effects of several geometric and thermohydraulic parameters on purification efficiency were also studied. The test results were used to develop a simulation model. An optimization code based on the model can be used to determine the nucleation and growth kinetics

An sRNA and Cold Shock Protein Homolog-Based Feedforward Loop Post-transcriptionally Controls Cell Cycle Master Regulator CtrA.

Science.gov (United States)

Robledo, Marta; Schlüter, Jan-Philip; Loehr, Lars O; Linne, Uwe; Albaum, Stefan P; Jiménez-Zurdo, José I; Becker, Anke

2018-01-01

Adjustment of cell cycle progression is crucial for bacterial survival and adaptation under adverse conditions. However, the understanding of modulation of cell cycle control in response to environmental changes is rather incomplete. In α-proteobacteria, the broadly conserved cell cycle master regulator CtrA underlies multiple levels of control, including coupling of cell cycle and cell differentiation. CtrA levels are known to be tightly controlled through diverse transcriptional and post-translational mechanisms. Here, small RNA (sRNA)-mediated post-transcriptional regulation is uncovered as an additional level of CtrA fine-tuning. Computational predictions as well as transcriptome and proteome studies consistently suggested targeting of ctrA and the putative cold shock chaperone cspA5 mRNAs by the trans- encoded sRNA ( trans- sRNA) GspR (formerly SmelC775) in several Sinorhizobium species. GspR strongly accumulated in the stationary growth phase, especially in minimal medium (MM) cultures. Lack of the gspR locus confers a fitness disadvantage in competition with the wild type, while its overproduction hampers cell growth, suggesting that this riboregulator interferes with cell cycle progression. An eGFP-based reporter in vivo assay, involving wild-type and mutant sRNA and mRNA pairs, experimentally confirmed GspR-dependent post-transcriptional down-regulation of ctrA and cspA5 expression, which most likely occurs through base-pairing to the respective mRNA. The energetically favored secondary structure of GspR is predicted to comprise three stem-loop domains, with stem-loop 1 and stem-loop 3 targeting ctrA and cspA5 mRNA, respectively. Moreover, this work reports evidence for post-transcriptional control of ctrA by CspA5. Thus, this regulation and GspR-mediated post-transcriptional repression of ctrA and cspA5 expression constitute a coherent feed-forward loop, which may enhance the negative effect of GspR on CtrA levels. This novel regulatory circuit involving

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

International Nuclear Information System (INIS)

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

2005-12-01

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

High pressure experimental water loop

International Nuclear Information System (INIS)

Grenon, M.

1958-01-01

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

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

Source localization using a non-cocentered orthogonal loop and dipole (NCOLD) array

Institute of Scientific and Technical Information of China (English)

Liu Zhaoting; Xu Tongyang

2013-01-01

A uniform array of scalar-sensors with intersensor spacings over a large aperture size generally offers enhanced resolution and source localization accuracy, but it may also lead to cyclic ambiguity. By exploiting the polarization information of impinging waves, an electromagnetic vec-tor-sensor array outperforms the unpolarized scalar-sensor array in resolving this cyclic ambiguity. However, the electromagnetic vector-sensor array usually consists of cocentered orthogonal loops and dipoles (COLD), which is easily subjected to mutual coupling across these cocentered dipoles/loops. As a result, the source localization performance of the COLD array may substantially degrade rather than being improved. This paper proposes a new source localization method with a non-cocentered orthogonal loop and dipole (NCOLD) array. The NCOLD array contains only one dipole or loop on each array grid, and the intersensor spacings are larger than a half-wave-length. Therefore, unlike the COLD array, these well separated dipoles/loops minimize the mutual coupling effects and extend the spatial aperture as well. With the NCOLD array, the proposed method can efficiently exploit the polarization information to offer high localization precision.

Plutonium Immobilization Program cold pour tests

International Nuclear Information System (INIS)

Hovis, G.L.; Stokes, M.W.; Smith, M.E.; Wong, J.W.

1999-01-01

The Plutonium Immobilization Program (PIP) is a joint venture between the Savannah River Site, Lawrence Livermore National Laboratory, Argonne National Laboratory, and Pacific Northwest National Laboratory to carry out the disposition of excess weapons-grade plutonium. This program uses the can-in-canister (CIC) approach. CIC involves encapsulating plutonium in ceramic forms (or pucks), placing the pucks in sealed stainless steel cans, placing the cans in long cylindrical magazines, latching the magazines to racks inside Defense Waste Processing Facility (DWPF) canisters, and filling the DWPF canisters with high-level waste glass. This process puts the plutonium in a stable form and makes it attractive for reuse. At present, the DWPF pours glass into empty canisters. In the CIC approach, the addition of a stainless steel rack, magazines, cans, and ceramic pucks to the canisters introduces a new set of design and operational challenges: All of the hardware installed in the canisters must maintain structural integrity at elevated (molten-glass) temperatures. This suggests that a robust design is needed. However, the amount of material added to the DWPF canister must be minimized to prevent premature glass cooling and excessive voiding caused by a large internal thermal mass. High metal temperatures, minimizing thermal mass, and glass flow paths are examples of the types of technical considerations of the equipment design process. To determine the effectiveness of the design in terms of structural integrity and glass-flow characteristics, full-scale testing will be conducted. A cold (nonradioactive) pour test program is planned to assist in the development and verification of a baseline design for the immobilization canister to be used in the PIP process. The baseline design resulting from the cold pour test program and CIC equipment development program will provide input to Title 1 design for second-stage immobilization. The cold pour tests will be conducted in two

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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

International Nuclear Information System (INIS)

Carbajo, J.J.

1984-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Kim, J. Y.

1997-07-01

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

Nondestructive characterization of recovery and recrystallization in cold rolled low carbon steel by magnetic hysteresis loops

International Nuclear Information System (INIS)

Martinez-de-Guerenu, A.; Gurruchaga, K.; Arizti, F.

2007-01-01

How structure sensitive parameters derived from hysteresis loops can provide nondestructive information about the evolution of the microstructure of cold rolled low carbon steel as a result of recovery and recrystallization processes during the annealing is shown. The coercive field, remanent induction and hysteresis losses can be used to monitor the decrease in the dislocation density during recovery. These parameters are also influenced by the average grain refinement that takes place during recrystallization, which compensates the variation produced by the annihilation of dislocations during recrystallization. The maximum of the induction and of the relative differential permeability are shown to be very sensitive to the onset and to the monitoring of the recrystallization, respectively. The correlations between coercive field and remanent induction and hysteresis losses can also be used to distinguish between recovery and recrystallization

Cold moderators at ORNL

International Nuclear Information System (INIS)

Lucas, A. T.

1997-09-01

The Advanced Neutron Source (ANS) cold moderators were not an 'Oak Ridge first', but would have been the largest both physically and in terms of cold neutron flux. Two cold moderators were planned each 410 mm in diameter and containing about 30L of liquid deuterium. They were to be completely independent of each other. A modular system design was used to provide greater reliability and serviceability. When the ANS was terminated, upgrading of the resident High Flux Isotope Reactor (HFIR) was examined and an initial study was made into the feasibility of adding a cold source. Because the ANS design was modular, it was possible to use many identical design features. Sub-cooled liquid at 4 bar abs was initially chosen for the HFIR design concept, but this was subsequently changed to 15 bar abs to operate above the critical pressure. As in the ANS, the hydrogen will operate at a constant pressure throughout the temperature range and a completely closed loop with secondary containment was adopted. The heat load of 2 kW made the heat flux comparable with that of the ANS. Subsequent studies into the construction of cryogenic moderators for the proposed new Synchrotron Neutron source indicated that again many of the same design concepts could be used. By connecting the two cold sources together in series, the total heat load of 2 kW is very close to that of the HFIR allowing a very similar supercritical hydrogen system to be configured. The two hydrogen moderators of the SNS provide a comparable heat load to the HFIR moderator. It is subsequently planned to connect the two in series and operate from a single cold loop system, once again using supercritical hydrogen. The spallation source also provided an opportunity to re-examine a cold pellet solid methane moderator operating at 20K.

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

International Nuclear Information System (INIS)

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

1985-11-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

The use of cold coagulation for the treatment of cervical intraepithelial neoplasia

LENUS (Irish Health Repository)

Wyse, A

2017-05-01

In 2015, Cold Coagulation was introduced as a treatment for cervical intraepithelial neoplasia (CIN) at our colposcopy clinic. We reviewed the 6-month follow up data of the first 200 women who underwent Cold Coagulation using cytology and HPV status as tests of cure (TOC). A random sample of 200 patients treated by Large Loop Excision of the Transformation Zone (LLETZ) during the same period was used to compare treatment outcome. Six months following treatment,173 (86.5%) of the women treated by CC and 167 (83.5%) treated by LLETZ had negative cytology. (x2= P>0.05). 148 (74%) treated by Cold Coagulation and 166 (83%) treated by LLETZ were HPV negative (x2= P<0.05). One hundred and thirty-nine (70%) women treated by Cold Coagulation and 152 (76%) treated with LLETZ had normal cytology and were HPV negative. This audit of our initial experience supports the observation that Cold Coagulation is as effective as LLETZ in the management of CIN when cervical cytology is used as a test of cure.

ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

Energy Technology Data Exchange (ETDEWEB)

Takeda, Takeshi; Ohtsu, Iwao [Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokaimura (Japan)

2017-08-15

An experiment using the Primaerkreislaeufe Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg small-break loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

Directory of Open Access Journals (Sweden)

Takeshi Takeda

2017-08-01

Full Text Available An experiment using the Primӓrkreislӓufe Versuchsanlage (PKL was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF on a cold leg small-break loss-of-coolant accident with an accident management (AM measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

Detection Test for Leakage of CO2 into Sodium Loop

International Nuclear Information System (INIS)

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

2015-01-01

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

Tests in the ATLE loop on the PIUS design

International Nuclear Information System (INIS)

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

1992-01-01

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

Mechanisms Engineering Test Loop - Phase 1 Status Report

Energy Technology Data Exchange (ETDEWEB)

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

2016-09-01

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

Engineering, Manufacture and Preliminary Testing of the ITER Toroidal Field (TF) Magnet Helium Cold Circulator

Science.gov (United States)

Rista, P. E. C.; Shull, J.; Sargent, S.

2015-12-01

The ITER cryodistribution system provides the supercritical Helium (SHe) forced flow cooling to the magnet system using cold circulators. The cold circulators are located in each of five separate auxiliary cold boxes planned for use in the facility. Barber-Nichols Inc. has been awarded a contract from ITER-India for engineering, manufacture and testing of the Toroidal Field (TF) Magnet Helium Cold Circulator. The cold circulator will be extensively tested at Barber-Nichols’ facility prior to delivery for qualification testing at the Japan Atomic Energy Agency's (JAEA) test facility at Naka, Japan. The TF Cold Circulator integrates features and technical requirements which Barber-Nichols has utilized when supplying helium cold circulators worldwide over a period of 35 years. Features include a vacuum-jacketed hermetically sealed design with a very low helium leak rate, a heat shield for use with both nitrogen & helium cold sources, a broad operating range with a guaranteed isentropic efficiency over 70%, and impeller design features for high efficiency. The cold circulator will be designed to meet MTBM of 17,500 hours and MTBF of 36,000 hours. Vibration and speed monitoring are integrated into a compact package on the rotating assembly with operation and health monitoring in a multi-drop PROFIBUS communication environment using an electrical cabinet with critical features and full local and network PLC interface and control. For the testing in Japan and eventual installation in Europe, the cold circulator must be certified to the Japanese High Pressure Gas Safety Act (JHPGSA) and CE marked in compliance with the European Pressure Equipment Directive (PED) including Essential Safety Requirements (ESR). The test methodology utilized at Barber-Nichols’ facility and the resulting test data, validating the high efficiency of the TF Cold Circulator across a broad operating range, are important features of this paper.

Engineering, Manufacture and Preliminary Testing of the ITER Toroidal Field (TF) Magnet Helium Cold Circulator

International Nuclear Information System (INIS)

C Rista, P E; Shull, J; Sargent, S

2015-01-01

The ITER cryodistribution system provides the supercritical Helium (SHe) forced flow cooling to the magnet system using cold circulators. The cold circulators are located in each of five separate auxiliary cold boxes planned for use in the facility. Barber-Nichols Inc. has been awarded a contract from ITER-India for engineering, manufacture and testing of the Toroidal Field (TF) Magnet Helium Cold Circulator. The cold circulator will be extensively tested at Barber-Nichols’ facility prior to delivery for qualification testing at the Japan Atomic Energy Agency's (JAEA) test facility at Naka, Japan. The TF Cold Circulator integrates features and technical requirements which Barber-Nichols has utilized when supplying helium cold circulators worldwide over a period of 35 years. Features include a vacuum-jacketed hermetically sealed design with a very low helium leak rate, a heat shield for use with both nitrogen and helium cold sources, a broad operating range with a guaranteed isentropic efficiency over 70%, and impeller design features for high efficiency. The cold circulator will be designed to meet MTBM of 17,500 hours and MTBF of 36,000 hours. Vibration and speed monitoring are integrated into a compact package on the rotating assembly with operation and health monitoring in a multi-drop PROFIBUS communication environment using an electrical cabinet with critical features and full local and network PLC interface and control. For the testing in Japan and eventual installation in Europe, the cold circulator must be certified to the Japanese High Pressure Gas Safety Act (JHPGSA) and CE marked in compliance with the European Pressure Equipment Directive (PED) including Essential Safety Requirements (ESR). The test methodology utilized at Barber-Nichols’ facility and the resulting test data, validating the high efficiency of the TF Cold Circulator across a broad operating range, are important features of this paper. (paper)

Hanford spent nuclear fuel cold vacuum drying proof of performance test procedure

International Nuclear Information System (INIS)

McCracken, K.J.

1998-01-01

This document provides the test procedure for cold testing of the first article skids for the Cold Vacuum Drying (CVD) process at the Facility. The primary objective of this testing is to confirm design choices and provide data for the initial start-up parameters for the process. The current scope of testing in this document includes design verification, drying cycle determination equipment performance testing of the CVD process and MCC components, heat up and cool-down cycle determination, and thermal model validation

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

International Nuclear Information System (INIS)

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

1979-01-01

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

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

International Nuclear Information System (INIS)

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

1984-01-01

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

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

International Nuclear Information System (INIS)

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

2007-12-01

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

An Investigation of Loop Seal Clearings in ATLAS SBLOCA Tests

Energy Technology Data Exchange (ETDEWEB)

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

2013-05-15

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

Microscopic observations of palladium used for cold fusion

International Nuclear Information System (INIS)

Matsumoto, T.

1991-01-01

This paper examines the microscopic structures of palladium metals used for cold fusion experiments. Tiny spot defects suggesting cold fusion have been observed in grain boundaries as the Nattoh model predicts. The relationship between these defects and a series of neutron busts and an indirect loop of hydrogen chain reactions are discussed

Start-up of a cold loop in a VVER-440, the 7th AER benchmark calculation with HEXTRAN-SMABRE-PORFLO

International Nuclear Information System (INIS)

Hovi, Ville; Taivassalo, Veikko; Haemaelaeinen, Anitta; Raety, Hanna; Syrjaelahti, Elina

2017-01-01

The 7 th dynamic AER benchmark is the first in which three-dimensional thermal hydraulics codes are supposed to be applied. The aim is to get a more precise core inlet temperature profile than the sector temperatures available typically with system codes. The benchmark consists of a start-up of the sixth, isolated loop in a VVER-440 plant. The isolated loop initially contains cold water without boric acid and the start-up leads to a somewhat asymmetrical core power increase due to feedbacks in the core. In this study, the 7 th AER benchmark is calculated with the three-dimensional nodal reactor dynamics code HEXTRAN-SMABRE coupled with the porous computational fluid dynamics code PORFLO. These three codes are developed at VTT. A novel two-way coupled simulation of the 7 th AER benchmark was performed successfully demonstrating the feasibility and advantages of the new reactor analysis framework. The modelling issues for this benchmark are reported and some evaluation against the previously reported comparisons between the system codes is provided.

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

International Nuclear Information System (INIS)

Calvaresi, C.; Moreschi, L.F.

1983-01-01

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

Comparação entre os processos de Cold Test e Hot Test em uma empresa fabricante de motores diesel

Directory of Open Access Journals (Sweden)

Pablo Fogaça

2017-10-01

Full Text Available Resumo O teste a frio (Cold Test é um processo inovador no Brasil com relação aos testes em motores diesel. Esse processo foi implementado em uma empresa fabricante de motores diesel, com o intuito de ser o principal teste de validação antes de os motores serem enviados ao cliente. Partindo-se dessa hipótese, surgiu o principal motivo desta investigação: é possível a substituição plena do teste a quente (Hot Test pelo teste a frio (Cold Test? Essa foi uma pergunta impactante, visto que o Hot Test é um processo confiável e consagrado há mais de 18 anos. O método utilizado nesta pesquisa foi o estudo de caso. A investigação detalhada identificou vários critérios de comparação dos dois processos, tais como: fluxogramas, eficácia, eficiência, complexidade, níveis de aprovação, planilhas de bordo e documentos FMEA da empresa. Os resultados mostraram que o processo Cold Test é mais vantajoso e sustentável, entretanto os dois processos são complementares. O estudo comprovou que não foi possível a plena substituição do Hot Test pelo Cold Test, principalmente pelo fato da não detecção de vazamentos no processo Cold Test.

First cold test of TESLA superconducting RF cavity in horizontal cryostat (CHECHIA)

International Nuclear Information System (INIS)

Kuzminski, J.

1996-01-01

In the framework of the TESLA project, the horizontal cryostat (CHECHIA) was built to test a superconducting RF cavity equipped with its helium vessel, magnetic shielding, cold tuner, main coupler and higher order modes couplers under realistic conditions before final assembly of eight cavities into TESLA Test Facility cryo-module. The results of the first cold tests in CHECHIA, performed at DESY with a 9-cell cavity (C19) to be used in the TTF injector are presented. (author)

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

Directory of Open Access Journals (Sweden)

Haiyu Gao

2018-02-01

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

Development of 3-Pin Fuel Test Loop and Utilization Technology

International Nuclear Information System (INIS)

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

2007-06-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-12-15

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

A new Friction and Lubrication Test for Cold Forging

DEFF Research Database (Denmark)

Bay, Niels; Wibom, Ole; Aalborg Nielsen, J

1995-01-01

This paper presents a new friction and lubrication test for cold forging. The test allows controlled variation of the surface expansion in the range 0-2000%, the tool temperature in the range 20-270°C and the sliding length between 0 and infinite. Friction is decreasing with increasing temperature...

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

Energy Technology Data Exchange (ETDEWEB)

Kim, J. Y.

1998-02-01

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

Internal friction measurement with a kilohertz range test facility of cold-worked and neutron-irradiated gold

International Nuclear Information System (INIS)

Grandchamp, Pierre-Andre

1970-02-01

The author studies properties of the Bordoni peak in 99,999 % gold. The following features are considered: - influence of type and rate of cold-work on relaxation strength. After traction cold-work, the height of the Bordoni peak is roughly proportional to the rate of plastic deformation, - relaxation features. The relaxation energies and attempt frequencies of the Bordoni peak are determined for poly - and single crystals. The broadening of experimental peaks is studied, - effect of dislocation pinning on the Bordoni peak. One shows that the experience leads to a relation: Q"-"1_m_a_x ∼ Λl"2 where Q"-"1_m_a_x is the height of the peak, Λ the dislocation density and l the mean loop length of dislocations which are implied in the phenomenon [fr

Radioactive waste incineration system cold demonstration test, (2)

Energy Technology Data Exchange (ETDEWEB)

Hozumi, Masahiro; Seike, Yasuhiko; Takaoku, Yoshinobu; Yamanaka, Yasuhiro; Asahara, Masaharu; Katagiri, Keishi; Matsumoto, Kenji; Nagae, Madoka

1985-12-01

It is urgently necessary to solve the radioactive waste problem. As an effective means for the volume reduction of low-level radioactive wastes, an improved incineration system is greatly required. SHI's Waste Incineration (WIS) licensed by Combustion Engineering, Inc., has the significant advantage of processing a variety of wastes. We started a cold demonstration test in April, 1984 to verify the excellent performance of WIS. The test was successfully completed in September, 1985 with the record of more than 1000 hours of incineration testing time. In the present paper, we describe the test results during one and half years of test period.

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

Directory of Open Access Journals (Sweden)

Gaurav Sepolia

2014-01-01

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

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

International Nuclear Information System (INIS)

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

1991-01-01

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

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

International Nuclear Information System (INIS)

Benhamou, C.

1992-01-01

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

Start-up of a cold loop in a VVER-440, the 7{sup th} AER benchmark calculation with HEXTRAN-SMABRE-PORFLO

Energy Technology Data Exchange (ETDEWEB)

Hovi, Ville; Taivassalo, Veikko; Haemaelaeinen, Anitta; Raety, Hanna; Syrjaelahti, Elina [VTT Technical Research Centre of Finland Ltd, VTT (Finland)

2017-09-15

The 7{sup th} dynamic AER benchmark is the first in which three-dimensional thermal hydraulics codes are supposed to be applied. The aim is to get a more precise core inlet temperature profile than the sector temperatures available typically with system codes. The benchmark consists of a start-up of the sixth, isolated loop in a VVER-440 plant. The isolated loop initially contains cold water without boric acid and the start-up leads to a somewhat asymmetrical core power increase due to feedbacks in the core. In this study, the 7{sup th} AER benchmark is calculated with the three-dimensional nodal reactor dynamics code HEXTRAN-SMABRE coupled with the porous computational fluid dynamics code PORFLO. These three codes are developed at VTT. A novel two-way coupled simulation of the 7{sup th} AER benchmark was performed successfully demonstrating the feasibility and advantages of the new reactor analysis framework. The modelling issues for this benchmark are reported and some evaluation against the previously reported comparisons between the system codes is provided.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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

Science.gov (United States)

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

2016-01-01

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

Loop corrections and a new test of inflation

CERN Document Server

Tasinato, Gianmassimo; Nurmi, Sami; Wands, David

2013-01-01

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

CLOSED LOOP AOCS TESTING OF AN AUTONOMOUS STAR TRACKER

DEFF Research Database (Denmark)

Jørgensen, John Leif

1999-01-01

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

Cold test plan for the Old Hydrofracture Facility tank contents removal project, Oak Ridge National Laboratory, Oak Ridge, Tennessee

International Nuclear Information System (INIS)

1997-11-01

This Old Hydrofracture Facility (OHF) Tanks Contents Removal Project Cold Test Plan describes the activities to be conducted during the cold test of the OHF sluicing and pumping system at the Tank Technology Cold Test Facility (TTCTF). The TTCTF is located at the Robotics and Process Systems Complex at the Oak Ridge National Laboratory (ORNL). The cold test will demonstrate performance of the pumping and sluicing system, fine-tune operating instructions, and train the personnel in the actual work to be performed. After completion of the cold test a Technical Memorandum will be prepared documenting completion of the cold test, and the equipment will be relocated to the OHF site

Gas Test Loop Functional and Technical Requirements

International Nuclear Information System (INIS)

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

2004-01-01

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

Gas Test Loop Functional and Technical Requirements

Energy Technology Data Exchange (ETDEWEB)

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

2004-09-01

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

Sandia Sodium Purification Loop (SNAPL) description and operations manual

International Nuclear Information System (INIS)

Acton, R.U.; Weatherbee, R.L.; Smith, L.A.; Mastin, F.L.; Nowotny, K.E.

1985-08-01

Sandia's Sodium Purification Loop was constructed to purify sodium for fast reactor safety experiments. An oxide impurity of less than 10 parts per million is required by these in-pile experiments. Commercial, reactor grade sodium is purchased in 180 kg drums. The sodium is melted and transferred into the unit. The unit is of a loop design and purification is accomplished by ''cold trapping.'' Sodium purified in this loop has been chemically analysed at one part per million oxygen by weight. 5 refs., 22 figs., 7 tabs

Space station common module thermal management: Design and construction of a test bed

Science.gov (United States)

Barile, R. G.

1986-01-01

In this project, a thermal test bed was designed, simulated, and planned for construction. The thermal system features interior and exterior thermal loads and interfacing with the central-radiator thermal bus. Components of the test bed include body mounted radiator loop with interface heat exchangers (600 Btu/hr); an internal loop with cabin air-conditioning and cold plates (3400 Btu/hr); interface heat exchangers to the central bus (13,000 Btu/hr); and provisions for new technology including advanced radiators, thermal storage, and refrigeration. The apparatus will be mounted in a chamber, heated with lamps, and tested in a vacuum chamber with LN2-cooled walls. Simulation of the test bed was accomplished using a DEC PRO 350 computer and the software package TK! olver. Key input variables were absorbed solar radiation and cold plate loads. The results indicate temperatures on the two loops will be nominal when the radiation and cold plate loads are in the range of 25% to 75% of peak loads. If all loads fall to zero, except the cabin air system which was fixed, the radiator fluid will drop below -100 F and may cause excessive pressure drop. If all loads reach 100%, the cabin air temperature could rise to 96 F.

Cold-rolled steel strip X-ray thickness gauge

International Nuclear Information System (INIS)

Tong Jianmin; Cong Peng; Li Litao

2010-01-01

This paper introduces a cold-rolled steel strip X-ray thickness gauge. This gauge uses two-detector construction including penetrating ionization chamber and measuring ionization chamber. Standard magazine box is composed of three rotating plates driving by stepper motor, including 13 pieces of standard sample to build up 154 thickness value, obtaining standard sample curve covering thickness range of 0.1 mm to 5 mm. Automation system include Siemens S7-200 PLC as key controlling unit, embedded controlling system for data acquisition and computing, and PC as man-machine interface, and employ Ethernet (TCP/IP) or RS485/232 as communication protocol. Reversing Cold mill AGC closed-loop control operation test demonstrates that the gauge can adapt to the severe production environment, operate stably and reliably, measurement precision can reached to ±0.19%, reproducibility to ±0.09%, and stability to ±0.06%, response time range from 4 ms to 200 ms and be adjustable. So it can meet the high demanding of cold-rolled plate/strip production. (authors)

High brilliant thermal and cold moderator for the HBS neutron source project Jülich

International Nuclear Information System (INIS)

Cronert, T; Zakalek, P; Rücker, U; Brückel, T; Dabruck, J P; Doege, P E; Nabbi, R; Bessler, Y; Hofmann, M; Butzek, M; Klaus, M; Lange, C; Hansen, W

2016-01-01

The proposed High Brilliance Neutron Source (HBS), recognized within the Helmholtz Association of German Research Centres, will optimize the entire chain from particle source through particle accelerator, target, moderator, reflector, shielding, beam extraction, beam transport all the way to the detector, utilizing the nuclear Be(p,n) or Be(d,n) reaction in the lower MeV energy range. A D 2 O moderating reflector prototype (MRP) and a cold source were constructed and build according to MCNP parameter studies. The MRP was tested in a feasibility study at the TREFF instrument at MLZ (Garching). Cold beam extraction from the flux maximum within the moderator based on liquid para H 2 and other cold moderators will be tested by energy spectroscopy via TOF-method. Different ratios of liquid ortho/para H 2 will be fed to the cold moderator. The ratio will be controlled by feeding from reservoires of natural liquid H 2 and a storage loop with an ortho/para converter and determined via online heat capacity measurement. (paper)

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

Energy Technology Data Exchange (ETDEWEB)

Yoder, JR.G.L.

2006-03-08

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

Systematic Unit Testing in a Read-eval-print Loop

DEFF Research Database (Denmark)

Nørmark, Kurt

2010-01-01

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

Conceptural design of multipurpose sodium test loop

International Nuclear Information System (INIS)

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

1982-01-01

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

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

International Nuclear Information System (INIS)

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

2016-03-01

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

Development of Aerosol Scrubbing Test Loop for Containment Filtered Venting System

International Nuclear Information System (INIS)

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

2016-01-01

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

Development of Aerosol Scrubbing Test Loop for Containment Filtered Venting System

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

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

Cold-knife conisation and large loop excision of transformation zone significantly increase the risk for spontaneous preterm birth: a population-based cohort study.

Science.gov (United States)

Jančar, Nina; Mihevc Ponikvar, Barbara; Tomšič, Sonja

2016-08-01

Our aim was to explore the association between cold-knife conisation and large loop excision of transformation zone (LLETZ) with spontaneous preterm birth in a large 10-year national sample. We wanted to explore further the association of these procedures with preterm birth according to gestation. We conducted a population based retrospective cohort study, using data from national Medical Birth Registry. The study population consisted of all women giving birth to singletons in the period 2003-2012 in Slovenia, excluding all induced labors and elective cesarean sections before 37 weeks of gestation (N=192730). We compared the prevalence of spontaneous preterm births (before 28 weeks, before 32 weeks, before 34 weeks and before 37 weeks of gestation) in women with cold-knife conisation or LLETZ compared to women without history of conisation, calculating odds ratios (OR), adjusted for potential confounders. Chi-square test was used for descriptive analysis. Logistic regression analyses were performed to estimate crude odds ratio (OR) and adjusted odds ratio (aOR) and their 95% confidence intervals (95% CI) with two-sided probability (p) values. A total of 8420 (4.4%) women had a preterm birth before 37 weeks of gestation, 2250 (1.2%) before 34 weeks of gestation, 1333 (0.7%) before 32 weeks of gestation and 603 (0.3%) before 28 weeks of gestation. A total of 4580 (2.4%) women had some type of conisation in their medical history: 2083 (1.1%) had cold-knife conisation and 2498 (1.3%) had LLETZ. In women with history of cold-knife conisation, the adjusted OR for preterm birth before 37 weeks of gestation was 3.13 (95% CI; 2.74-3.57) and for preterm birth before 28 weeks of gestation 5.96 (95% CI; 4.3-8.3). In women with history of LLETZ, the adjusted OR was 1.95 (95% CI; 1.68-2.25) and 2.88 (95% CI; 1.87-4.43), respectively. Women with cervical excision procedure of any kind have significantly increased odds for preterm birth, especially for preterm birth before 28

The Phillips Laboratory capillary pumped loop test facility

Science.gov (United States)

Gluck, Donald F.; Kaylor, Marc C.

1996-03-01

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

Hypothermic general cold adaptation induced by local cold acclimation.

Science.gov (United States)

Savourey, G; Barnavol, B; Caravel, J P; Feuerstein, C; Bittel, J H

1996-01-01

To study relationships between local cold adaptation of the lower limbs and general cold adaptation, eight subjects were submitted both to a cold foot test (CFT, 5 degrees C water immersion, 5 min) and to a whole-body standard cold air test (SCAT, 1 degree C, 2 h, nude at rest) before and after a local cold acclimation (LCA) of the lower limbs effected by repeated cold water immersions. The LCA induced a local cold adaptation confirmed by higher skin temperatures of the lower limbs during CFT and a hypothermic insulative general cold adaptation (decreased rectal temperature and mean skin temperature P adaptation was related to the habituation process confirmed by decreased plasma concentrations of noradrenaline (NA) during LCA (P general cold adaptation was unrelated either to local cold adaptation or to the habituation process, because an increased NA during SCAT after LCA (P syndrome" occurring during LCA.

High-pressure test loop design and application

International Nuclear Information System (INIS)

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

1980-07-01

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

Inhibition of TRPM8 channels reduces pain in the cold pressor test in humans.

Science.gov (United States)

Winchester, Wendy J; Gore, Katrina; Glatt, Sophie; Petit, Wendy; Gardiner, Jennifer C; Conlon, Kelly; Postlethwaite, Michael; Saintot, Pierre-Philippe; Roberts, Sonia; Gosset, James R; Matsuura, Tomomi; Andrews, Mark D; Glossop, Paul A; Palmer, Michael J; Clear, Nicola; Collins, Susie; Beaumont, Kevin; Reynolds, David S

2014-11-01

The transient receptor potential (subfamily M, member 8; TRPM8) is a nonselective cation channel localized in primary sensory neurons, and is a candidate for cold thermosensing, mediation of cold pain, and bladder overactivity. Studies with TRPM8 knockout mice and selective TRPM8 channel blockers demonstrate a lack of cold sensitivity and reduced cold pain in various rodent models. Furthermore, TRPM8 blockers significantly lower body temperature. We have identified a moderately potent (IC50 = 103 nM), selective TRPM8 antagonist, PF-05105679 [(R)-3-[(1-(4-fluorophenyl)ethyl)(quinolin-3-ylcarbonyl)amino]methylbenzoic acid]. It demonstrated activity in vivo in the guinea pig bladder ice water and menthol challenge tests with an IC50 of 200 nM and reduced core body temperature in the rat (at concentrations >1219 nM). PF-05105679 was suitable for acute administration to humans and was evaluated for effects on core body temperature and experimentally induced cold pain, using the cold pressor test. Unbound plasma concentrations greater than the IC50 were achieved with 600- and 900-mg doses. The compound displayed a significant inhibition of pain in the cold pressor test, with efficacy equivalent to oxycodone (20 mg) at 1.5 hours postdose. No effect on core body temperature was observed. An unexpected adverse event (hot feeling) was reported, predominantly periorally, in 23 and 36% of volunteers (600- and 900-mg dose, respectively), which in two volunteers was nontolerable. In conclusion, this study supports a role for TRPM8 in acute cold pain signaling at doses that do not cause hypothermia. Copyright © 2014 by The American Society for Pharmacology and Experimental Therapeutics.

Disassembly and removal of sodium instrumentation test loop

International Nuclear Information System (INIS)

Ishikawa, Okinobu; Onojima, Takamitu; Nagai, Keiichi

2000-07-01

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

First cold test of TESLA superconducting RF cavity in horizontal cryostat (CHECHIA)

International Nuclear Information System (INIS)

Kuzminski, J.

1996-01-01

In the framework of the TESLA project, the horizontal cryostat (CHECHIA) was built to test a superconducting RF cavity equipped with its helium vessel, magnetic shielding, cold tuner, main coupler and higher order modes couplers under realistic conditions before final assembly of eight cavities into TESLA Test Facility cryo-module. The results of the first cold tests in CHECHIA, performed at DESY with a 9-cell cavity (C19) to be used in the TTF injector are presented. Additional measurements of mechanical stability under RF operation (frequency variation with He pressure, Lorentz detuning) and cryogenic and electric measurements of power dissipation are presented. (author)

Status of the cold test facility for the JT-60SA tokamak toroidal field coils

Energy Technology Data Exchange (ETDEWEB)

Abdel Maksoud, Walid, E-mail: walid.abdelmaksoud@cea.fr; Bargueden, Patrick; Bouty, André; Dispau, Gilles; Donati, André; Eppelle, Dominique; Genini, Laurent; Guiho, Patrice; Guihard, Quentin; Joubert, Jean-Michel; Kuster, Olivier; Médioni, Damien; Molinié, Frédéric; Sinanna, Armand; Solenne, Nicolas; Somson, Sébastien; Vieillard, Laurence

2015-10-15

Highlights: • The 5 K cryogenic loop includes a 500 W refrigerator and a She cold pump. • The coils are energized thanks to a 25.7 kA power supply and HTS current leads. • Temperature margin tests between 5 K and 7.5 K will be made on each coil. • A magnet safety system protects each double pancake of the coil in case of quench. • Instrumentation is monitored on a 1 Hz to 10 kHz fast acquisition system. - Abstract: JT-60SA is a fusion experiment which is jointly constructed by Japan and Europe and which shall contribute to the early realization of fusion energy, by providing support to the operation of ITER, and by addressing key physics issues for ITER and DEMO. In order to achieve these goals, the existing JT-60U experiment will be upgraded to JT-60SA by using superconducting coils. The 18 TF coils of the JT-60SA device will be provided by European industry and tested in a Cold Test Facility (CTF) at CEA Saclay. The coils will be tested at the nominal current of 25.7 kA and will be cooled with supercritical helium between 5 K and 7.5 K to check the temperature margin against a quench. The main objective of these tests is to check the TF coils performance and hence mitigate the fabrication risks. The most important components of the facility are: a 11.5 m × 6.5 m large cryostat in which the TF coils will be thermally insulated by vacuum; a 500 W helium refrigerator and a valve box to cool the coils down to 5 K and circulate 24 g/s of supercritical helium through the winding pack and through the casing; a power supply and HTS current leads to energize the coil; the control and instrumentation equipment (sensors, PLC's, supervision system, fast data acquisition system, etc.) and the Magnet Safety System (MSS) that protects the coils in case of quench. The paper will give an overview of the design of this large facility and the status of its realization.

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

Science.gov (United States)

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

1996-10-01

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

Assessment of the MARS-KS Code Using Atlas 6-inch cold leg Break Test

Energy Technology Data Exchange (ETDEWEB)

Kang, D. G.; Kim, J. S.; Ahn, S. H.; Seul, K. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2012-03-15

An integral effect test on the SBLOCA (Small-Break Loss of Coolant Accident) aiming at 6-inch cold leg bottom break, SB-CL-09, was conducted with the Atlas on November, 13, 2009, by KAERI. In this study, the calculations using MARS-KS Vt1.2 code were conducted for 6-inch cold leg break test of Atlas (SB-CL-09) which is the second domestic standard problem (Dsp-02) to assess MARS-KS code capability to simulate the transient thermal-hydraulic behavior for SBLOCA. The steady state was determined by conducting a null transient calculation and the errors between the calculated and measured values are acceptable for almost primary/secondary system parameters. The predicted pressurizer pressure agrees relatively well with the experimental data and the predicted break flow and mass are in good agreement with experiment. In MARS-KS calculation, the decrease of core collapsed water level is predicted well in blowdown phase, but just before LSC, water level is higher than experiment. However, the sudden decrease and increase of water level is higher than experiment. However, the sudden decrease and increase of water level at the LSC are predicted qualitatively. After LSC, there is another water level dip at Sit injection time which is not in experiment. It is considered that this phenomenon is caused by rapid depressurization of downcomer due to significant condensation rate of vapor in downcomer when Sit water flows in it. For the downcomer water level is predicted well, however, it is significantly over-predicted at SIT injection time, water level is predicted well, however, it is significantly over-predicted at SIT injection time after SIT water flows in downcomer. Predicted cladding temperature generally agrees well with the experiment, while there is peak at SIT injection time in calculation which is not in experiment. The loop seals of 1A, 2B intermediate leg are cleared around 400 seconds in experiment, while only that of 1A is cleared in MARS-KS calculation at the

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

International Nuclear Information System (INIS)

Gerstner, Douglas M.

2009-01-01

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

Test Results From a Pair of 1-kWe Dual-Opposed Free-Piston Stirling Power Convertors Integrated With a Pumped NaK Loop

Science.gov (United States)

Geng, Steven M.; Briggs, Maxwell H.; Penswick, L. Barry; Pearson, J. Boise; Godfroy, Thomas J.

2011-01-01

As a step towards development of Stirling power conversion for potential use in Fission Surface Power (FSP) systems, a pair of commercially available 1-kW-class free-piston Stirling convertors were modified to operate with a NaK (sodium (Na) and potassium (K)) liquid metal pumped loop for thermal energy input. This was the first-ever attempt at powering a free-piston Stirling engine with a pumped liquid metal heat source and is a major FSP project milestone towards demonstrating technical feasibility. The convertors were successfully tested at the Marshall Space Flight Center (MSFC) from June 6 through July 14, 2009. The convertors were operated for a total test time of 66 hr and 16 min. The tests included (a) performance mapping the convertors over various hot- and cold-end temperatures, piston amplitudes, and NaK flow rates and (b) transient test conditions to simulate various startup (i.e., low-, medium-, and high-temperature startups) and fault scenarios (i.e., loss of heat source, loss of NaK pump, convertor stall, etc.). This report documents the results of this testing

Conceptual design of HANARO cold neutron source

International Nuclear Information System (INIS)

Lee, Chang Hee; Sim, Cheul Muu; Park, K. N.; Choi, Y. H.

2002-07-01

The purpose of the cold source is to increase the available neutron flux delivered to instruments at wavelength 4 ∼ 12 A. The major engineering targets of this CNS facility is established for a reach out of very high gain factors in consideration with the cold neutron flux, moderator, circulation loop, heat load, a simplicity of the maintenance of the facility, safety in the operation of the facility against the hydrogen explosion and a layout of a minimum physical interference with the present facilities. The cold source project has been divided into 5 phases: (1) pre-conceptual (2) conceptual design (3) Testing (4) detailed design and procurement (5) installation and operation. Although there is sometime overlap between the phases, in general, they are sequential. The pre-conceptual design and concept design of KCNS has been performed on elaborations of PNPI Russia and review by Technicatome, Air Liquid, CILAS France. In the design of cold neutron source, the characteristics of cold moderators have been studied to obtain the maximum gain of cold neutron, and the analysis for radiation heat, design of hydrogen system, vacuum system and helium system have been performed. The possibility for materialization of the concept in the proposed conceptual design has been reviewed in view of securing safety and installing at HANARO. Above all, the thermosiphon system to remove heat by circulation of sub-cooled two phase hydrogen has been selected so that the whole device could be installed in the reactor pool with the reduced volume. In order to secure safety, hydrogen safety has been considered on protection to prevent from hydrogen-oxygen reaction at explosion of hydrogen-oxygen e in the containment. A lay out of the installation, a maintenance and quality assurance program and a localization are included in this report. Requirements of user, regulatory, safety, operation, maintenance should be considered to be revised for detailed design, testing, installation

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

1981-05-01

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

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

International Nuclear Information System (INIS)

1981-05-01

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

Hanford Tank Farms Waste Certification Flow Loop Test Plan

Energy Technology Data Exchange (ETDEWEB)

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

2010-01-01

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

Optimal closed-loop identification test design for internal model control

NARCIS (Netherlands)

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

2000-01-01

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

Testing cold protection according to EN ISO 20344: is there any professional footwear that does not pass?

Science.gov (United States)

Kuklane, Kalev; Ueno, Satoru; Sawada, Shin-Ichi; Holmér, Ingvar

2009-01-01

The present Comité Européen de Normalisation (CEN) and International Organization for Standardization (ISO) standards for safety, protective and occupational footwear EN ISO 20344-20347 classify footwear as cold protective by a pass/fail test where the limits are set for an allowed 10 degrees C temperature drop inside the footwear during 30 min at a temperature gradient of approximately 40 degrees C. It is questionable if a simple pass/fail test of this kind provides approved footwear that really protects the feet from cooling in exposures ranging from temperatures at +18 degrees C to as low as or even lower than -50 degrees C. This study selected for testing some professional footwear that could certainly not be considered as cold protective. Some footwear that could be used in cold was selected with as low insulation as the not cold-intended footwear. Also, a boot intended for cold was selected to be tested according to a modified standard at a temperature gradient of 70 degrees C. The footwear selection was based on insulation measurements with a thermal foot model. All footwear did pass the test. Although it is clear for the user that a sandal, a mesh shoe or a thin textile shoe is not cold protective, it is not as clear that an item of safety footwear, that has as low insulation as those mentioned above, could be classified as cold protective according to the present standards. Because of this, the user might have a deceptive feeling of safety and may be exposed to higher risks. As practically all professional footwear may pass this cold test, then the method/requirements should be radically changed or such a test should be removed from the standards.

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-07-01

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1999-10-01

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

Alternate Reductant Cold Cap Evaluation Furnace Phase II Testing

Energy Technology Data Exchange (ETDEWEB)

Johnson, F. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Stone, M. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Miller, D. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2014-09-03

Savannah River Remediation (SRR) conducted a Systems Engineering Evaluation (SEE) to determine the optimum alternate reductant flowsheet for the Defense Waste Processing Facility (DWPF). Specifically, two proposed flowsheets (nitric–formic–glycolic and nitric–formic–sugar) were evaluated based upon results from preliminary testing. Comparison of the two flowsheets among evaluation criteria indicated a preference towards the nitric–formic–glycolic flowsheet. Further research and development of this flowsheet eliminated the formic acid, and as a result, the nitric–glycolic flowsheet was recommended for further testing. Based on the development of a roadmap for the nitric–glycolic acid flowsheet, Waste Solidification Engineering (WS-E) issued a Technical Task Request (TTR) to address flammability issues that may impact the implementation of this flowsheet. Melter testing was requested in order to define the DWPF flammability envelope for the nitric-glycolic acid flowsheet. The Savannah River National Laboratory (SRNL) Cold Cap Evaluation Furnace (CEF), a 1/12^th scale DWPF melter, was selected by the SRR Alternate Reductant project team as the melter platform for this testing. The overall scope was divided into the following sub-tasks as discussed in the Task Technical and Quality Assurance Plan (TTQAP): Phase I - A nitric–formic acid flowsheet melter test (unbubbled) to baseline the CEF cold cap and vapor space data to the benchmark melter flammability models; Phase II - A nitric–glycolic acid flowsheet melter test (unbubbled and bubbled) to: Define new cold cap reactions and global kinetic parameters in support of the melter flammability model development; Quantify off-gas surging potential of the feed; Characterize off-gas condensate for complete organic and inorganic carbon species. After charging the CEF with cullet from Phase I CEF testing, the melter was slurry-fed with glycolic flowsheet based SB6-Frit 418 melter feed at 36% waste

Relationships of self-identified cold tolerance and cold-induced vasodilatation in the finger

Science.gov (United States)

Park, Joonhee; Lee, Joo-Young

2016-04-01

This study was conducted to investigate relationships of self-identified cold tolerance and cold-induced vasodilatation (CIVD) in the finger. Nine males and 34 females participated in the following 2 tests: a CIVD test and a self-reported survey. The CIVD test was conducted 30-min cold-water immersion (3.8 ± 0.3 °C) of the middle finger at an air temperature of 27.9 ± 0.1 °C. The self-reported questionnaire consisted of 28 questions about whole and local body cold and heat tolerances. By a cluster analysis on the survey results, the participants were divided into two groups: high self-identified cold tolerance (HSCT, n = 25) and low self-identified cold tolerance (LSCT, n = 18). LSCT had lower self-identified cold tolerance ( P cold or heat tolerance had relationships with cold tolerance index, T max, and amplitude ( P cold tolerance classified through a standardized survey could be a good index to predict physiological cold tolerance.

MES lead bismuth forced circulation loop and test results

International Nuclear Information System (INIS)

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

2003-01-01

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

Cold Test and Performance Evaluation of Prototype Cryoline-X

Science.gov (United States)

Shah, N.; Choukekar, K.; Kapoor, H.; Muralidhara, S.; Garg, A.; Kumar, U.; Jadon, M.; Dash, B.; Bhattachrya, R.; Badgujar, S.; Billot, V.; Bravais, P.; Cadeau, P.

2017-12-01

The multi-process pipe vacuum jacketed cryolines for the ITER project are probably world’s most complex cryolines in terms of layout, load cases, quality, safety and regulatory requirements. As a risk mitigation plan, design, manufacturing and testing of prototype cryoline (PTCL) was planned before the approval of final design of ITER cryolines. The 29 meter long PTCL consist of 6 process pipes encased by thermal shield inside Outer Vacuum Jacket of DN 600 size and carries cold helium at 4.5 K and 80 K. The global heat load limit was defined as 1.2 W/m at 4.5 K and 4.5 W/m at 80 K. The PTCL-X (PTCL for Group-X cryolines) was specified in detail by ITER-India and designed as well as manufactured by Air Liquide. PTCL-X was installed and tested at cryogenic temperature at ITER-India Cryogenic Laboratory in 2016. The heat load at 4.5 K and 80 K, estimated using enthalpy difference method, was found to be approximately 0.8 W/m at 4.5 K, 4.2 W/m at 80 K, which is well within the defined limits. Thermal shield temperature profile was also found to be satisfactory. Paper summarizes the cold test results of PTCL-X

Hybrid Combustion-Gasification Chemical Looping

Energy Technology Data Exchange (ETDEWEB)

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

2009-01-07

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

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

International Nuclear Information System (INIS)

Willms, R.S.

1989-01-01

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

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

International Nuclear Information System (INIS)

1992-01-01

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

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

ALTERNATE REDUCTANT COLD CAP EVALUATION FURNACE PHASE I TESTING

Energy Technology Data Exchange (ETDEWEB)

Johnson, F.; Miller, D.; Zamecnik, J.; Lambert, D.

2014-04-22

Savannah River Remediation (SRR) conducted a Systems Engineering Evaluation (SEE) to determine the optimum alternate reductant flowsheet for the Defense Waste Processing Facility (DWPF). Specifically, two proposed flowsheets (nitric–formic–glycolic and nitric–formic–sugar) were evaluated based upon results from preliminary testing. Comparison of the two flowsheets among evaluation criteria indicated a preference towards the nitric–formic–glycolic flowsheet. Further evaluation of this flowsheet eliminated the formic acid1, and as a result, the nitric–glycolic flowsheet was recommended for further testing. Based on the development of a roadmap for the nitric–glycolic acid flowsheet, Waste Solidification Engineering (WS-E) issued a Technical Task Request (TTR) to address flammability issues that may impact the implementation of this flowsheet. Melter testing was requested in order to define the DWPF flammability envelope for the nitric glycolic acid flowsheet. The Savannah River National Laboratory (SRNL) Cold Cap Evaluation Furnace (CEF), a 1/12th scale DWPF melter, was selected by the SRR Alternate Reductant project team as the melter platform for this testing. The overall scope was divided into the following sub-tasks as discussed in the Task Technical and Quality Assurance Plan (TTQAP): Phase I - A nitric–formic acid flowsheet melter test (unbubbled) to baseline the Cold Cap Evaluation Furnace (CEF) cold cap and vapor space data to the benchmark melter flammability models Phase II - A nitric–glycolic acid flowsheet melter test (unbubbled and bubbled) to: o Define new cold cap reactions and global kinetic parameters for the melter flammability models o Quantify off-gas surging potential of the feed o Characterize off-gas condensate for complete organic and inorganic carbon species Prior to startup, a number of improvements and modifications were made to the CEF, including addition of cameras, vessel support temperature measurement, and a heating

Construction, commissioning and initial operation of 2400W refrigerator and cold test stand for CDM testing

International Nuclear Information System (INIS)

Dubbs, J.D.; Kreinbrink, K.

1994-01-01

Air Products and CVI collaborated to design, construct and commission a refrigerator, test stands and integrated control system for the General Dynamics Collider Dipole Magnet Cold Test Facility (CTF) in Hammond, LA. The original project schedule required the cold test facility to be operational within 17 months of the notice to proceed. Midway through the project, changes in General Dynamics magnet testing requirements necessitated doubling the plant capacity, but the on stream date for the initial capacity increment could not be relaxed. The Air Products/CVI team had to adapt the project execution strategy to mitigate the schedule impact of the expansion in a cost effective manner without impacting system functionality, quality or safety. An equally challenging aspect of the job was that the (CTF) was being designed while several major systems that would interface with the CTF were being engineered. General Dynamics, Air Products and CVI had to work very closely to manage the interface issues. The teams efforts were very successful. The Hammond refrigerator/liquifier was started up on schedule. The first two test stands are currently being commissioned and will be on stream just six weeks later than the pre-expansion schedule target and all four test stands will be operational in time to support General Dynamics magnet testing requirements

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Design of closed-loop nitrogen Joule-Thomson refrigeration cycle for 67 K with sub-atmospheric device

Energy Technology Data Exchange (ETDEWEB)

Lee, C.; Lee, J.; Jeong, S. [Cryogenic Engineering Laboratory, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2013-05-15

Closed-loop J-T (Joule-Thomson) refrigeration cycle is advantageous compared to common open loop N{sub 2} decompression system in terms of nitrogen consumption. In this study, two closed-loop pure N{sub 2} J-T refrigeration systems with sub-atmospheric device for cooling High Temperature Superconductor (HTS) power cable are investigated. J-T cooling systems include 2-stage compressor, 2-stage precooling cycle, J-T valve and a cold compressor or an auxiliary vacuum pump at the room temperature. The cold compressor and the vacuum pump are installed after the J-T valve to create sub-atmospheric condition. The temperature of 67 K is possible by lowering the pressure up to 24 kPa at the cold part. The optimized hydrocarbon mixed refrigerant (MR) J-T system is applied for precooling stage. The cold head of precooling MR J-T have the temperature from 120 K to 150 K. The various characteristics of cold compressor are investigated and applied to design parameter of the cold compressor. The Carnot efficiency of cold compressor system is calculated as 16.7% and that of vacuum pump system as 16.4%. The efficiency difference between the cold compressor system and the vacuum pump system is due to difference of enthalpy change at cryogenic temperature, enthalpy change at room temperature and different work load at the pre-cooling cycle. The efficiency of neon-nitrogen MR J-T system is also presented for comparison with the sub-atmospheric devices. These systems have several pros and cons in comparison to typical MR J-T systems such as vacuum line maintainability, system's COP and etc. In this paper, the detailed design of the subcooled N{sub 2} J-T systems are examined and some practical issues of the sub-atmospheric devices are discussed.

O teste de frio sem solo em sementes de abóbora The modified cold germination test in squash seeds

Directory of Open Access Journals (Sweden)

Derblai Casaroli

2006-12-01

Full Text Available Este trabalho teve como objetivo determinar o período de exposição à baixa temperatura no teste de frio sem solo, para avaliação de diferentes potenciais fisiológicos em lotes de sementes de abóbora. Foram avaliados seis lotes de sementes de abóbora cultivar "Menina Brasileira" (Cucurbita moschata Duch. produzidas na safra 2003/2004. As sementes foram submetidas a diferentes períodos (três, cinco, sete e nove dias de exposição à baixa temperatura para condução do teste de frio sem solo, a uma temperatura constante de 10°C. Utilizou-se o delineamento inteiramente casualizado, em que os tratamentos constituíram um fatorial 6x4 (seis lotes e quatro períodos de exposição, com quatro repetições de 50 sementes. As médias foram comparadas pelo teste de Tukey em nível de 5% de probabilidade de erro. Realizou-se também o teste de correlação simples de Pearson .Os resultados permitiram concluir que o período de três dias de exposição à baixa temperatura foi o mais indicado para a condução do teste de frio sem solo em sementes de abóbora.This research was aimed at determining the exposition period at low temperature in the modified cold germination test, to evaluate physiological potential of squash seed lots. Six seed lots of squash, variety Menina Brasileira (Cucurbita moschata Duch., produced in 2003/2004, were submitted at different exposition times for the development of the cold test. The modified cold gernation test was carried out at 10°C of constant temperature, during exposition time of three, five, seven and nine days. A completely randomized desing was used to the cold test, 0 constituting a factorial 6x4, corresponding to the six seed lots and four exposition time (3, 5, 7, and 9 days, with four repetitions. The averages were compared by the Tukey test, at 5% of probability. The correlation test of Person (r was used too. For the modified cold germination test, the period of three days was the most

Hybrid Cooling Loop Technology for Robust High Heat Flux Cooling, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — Advanced Cooling Technologies, Inc. (ACT) proposes to develop a hybrid cooling loop and cold plate technology for space systems thermal management. The proposed...

Progress in construction of liquid metal LiPb experimental loops in China

International Nuclear Information System (INIS)

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

2007-01-01

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Physical separations soil washing system cold test results

Energy Technology Data Exchange (ETDEWEB)

McGuire, J.P.

1993-07-28

This test summary describes the objectives, methodology, and results of a physical separations soil-washing system setup and shakedown test using uncontaminated soil. The test is being conducted in preparation for a treatability test to be conducted in the North Pond of the 300-FF-1 Operable Unit. It will be used to assess the feasibility of using a physical separations process to reduce the volume of contaminated soils in the 300-FF-1 Operable Unit. The test is described in DOE-RL (1993). The setup test was conducted at an uncontrolled area located approximately 3.2 km northwest of the 300-FF-1 Operable Unit. The material processed was free of contamination. The physical separation equipment to be used in the test was transferred to the US Department of Energy (DOE) by the US Environmental Protection Agency (EPA) Risk Reduction Engineering Laboratory. On May 13, 1993, soil-washing equipment was moved to the cold test location. Design assistance and recommendation for operation was provided by the EPA.

Soundness confirmation through cold test of the system equipment of HTTR

International Nuclear Information System (INIS)

Ono, Masato; Shinohara, Masanori; Iigaki, Kazuhiko; Tochio, Daisuke; Nakagawa, Shigeaki; Shimazaki, Yosuke

2014-01-01

HTTR was established at the Oarai Research and Development Center of Japan Atomic Energy Agency, for the purpose of the establishment and upgrading of high-temperature gas-cooled reactor technology infrastructure. Currently, it performs a safety demonstration test in order to demonstrate the safety inherent in high-temperature gas-cooled reactor. After the Great East Japan Earthquake, it conducted confirmation test for the purpose of soundness survey of facilities and equipment, and it confirmed that the soundness of the equipment was maintained. After two years from the confirmation test, it has not been confirmed whether the function of dynamic equipment and the soundness such as the airtightness of pipes and containers are maintained after receiving the influence of damage or deterioration caused by aftershocks generated during two years or aging. To confirm the soundness of these facilities, operation under cold state was conducted, and the obtained plant data was compared with confirmation test data to evaluate the presence of abnormality. In addition, in order to confirm through cold test the damage due to aftershocks and degradation due to aging, the plant data to compare was supposed to be the confirmation test data, and the evaluation on abnormality of the plant data of machine starting time and normal operation data was performed. (A.O.)

Detection of cold pain, cold allodynia and cold hyperalgesia in freely behaving rats

Directory of Open Access Journals (Sweden)

Woolf Clifford J

2005-12-01

Full Text Available Abstract Background Pain is elicited by cold, and a major feature of many neuropathic pain states is that normally innocuous cool stimuli begin to produce pain (cold allodynia. To expand our understanding of cold induced pain states we have studied cold pain behaviors over a range of temperatures in several animal models of chronic pain. Results We demonstrate that a Peltier-cooled cold plate with ± 1°C sensitivity enables quantitative measurement of a detection withdrawal response to cold stimuli in unrestrained rats. In naïve rats the threshold for eliciting cold pain behavior is 5°C. The withdrawal threshold for cold allodynia is 15°C in both the spared nerve injury and spinal nerve ligation models of neuropathic pain. Cold hyperalgesia is present in the spared nerve injury model animals, manifesting as a reduced latency of withdrawal response threshold at temperatures that elicit cold pain in naïve rats. We also show that following the peripheral inflammation produced by intraplantar injection of complete Freund's adjuvant, a hypersensitivity to cold occurs. Conclusion The peltier-cooled provides an effective means of assaying cold sensitivity in unrestrained rats. Behavioral testing of cold allodynia, hyperalgesia and pain will greatly facilitate the study of the neurobiological mechanisms involved in cold/cool sensations and enable measurement of the efficacy of pharmacological treatments to reduce these symptoms.

Structure design and simulation research of active magnetic bearing for helium centrifugal cold compressor

Science.gov (United States)

Y Zhang, S.; Pan, W.; Wei, C. B.; Wu, J. H.

2017-12-01

Helium centrifugal cold compressors are utilized to pump gaseous helium from saturated liquid helium tank to obtain super-fluid helium in cryogenic refrigeration system, which is now being developed at TIPC, CAS. Active magnetic bearing (AMB) is replacing traditional oil-fed bearing as the optimal supporting assembly for cold compressor because of its many advantages: free of contact, high rotation speed, no lubrication and so on. In this paper, five degrees of freedom for AMB are developed for the helium centrifugal cold compressor application. The structure parameters of the axial and radial magnetic bearings as well as hardware and software of the electronic control system is discussed in detail. Based on modal analysis and critical speeds calculation, a control strategy combining PID arithmetic with other phase compensators is proposed. Simulation results demonstrate that the control method not only stables AMB system but also guarantees good performance of closed-loop behaviour. The prior research work offers important base and experience for test and application of AMB experimental platform for system centrifugal cold compressor.

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

Energy Technology Data Exchange (ETDEWEB)

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

2005-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-15

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

An alternative to the traditional cold pressor test: the cold pressor arm wrap.

Science.gov (United States)

Porcelli, Anthony John

2014-01-16

Recently research on the relationship between stress and cognition, emotion, and behavior has greatly increased. These advances have yielded insights into important questions ranging from the nature of stress' influence on addiction(1) to the role of stress in neural changes associated with alterations in decision-making(2,3). As topics being examined by the field evolve, however, so too must the methodologies involved. In this article a practical and effective alternative to a classic stress induction technique, the cold pressor test (CPT), is presented: the cold pressor arm wrap (CPAW). CPT typically involves immersion of a participant's dominant hand in ice-cold water for a period of time(4). The technique is associated with robust activation of the sympatho-adrenomedullary (SAM) axis (and release of catecholamines; e.g. adrenaline and noradrenaline) and mild-to-moderate activation of the hypothalamic-pituitary-adrenal (HPA) axis with associated glucocorticoid (e.g. cortisol) release. While CPT has been used in a wide range of studies, it can be impractical to apply in some research environments. For example use of water during, rather than prior to, magnetic resonance imaging (MRI) has the potential to damage sensitive and expensive equipment or interfere with acquisition of MRI signal. The CPAW is a practical and effective alternative to the traditional CPT. Composed of a versatile list of inexpensive and easily acquired components, CPAW makes use of MRI-safe gelpacs cooled to a temperature similar to CPT rather than actual water. Importantly CPAW is associated with levels of SAM and HPA activation comparable to CPT, and can easily be applied in a variety of research contexts. While it is important to maintain specific safety protocols when using the technique, these are easy to implement if planned for. Creation and use of the CPAW will be discussed.

Preliminary Design of the Liquid Lead Corrosion Test Loop

International Nuclear Information System (INIS)

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

2005-01-01

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

Large lithium loop experience

International Nuclear Information System (INIS)

Kolowith, R.; Owen, T.J.; Berg, J.D.; Atwood, J.M.

1981-10-01

An engineering design and operating experience of a large, isothermal, lithium-coolant test loop are presented. This liquid metal coolant loop is called the Experimental Lithium System (ELS) and has operated safely and reliably for over 6500 hours through September 1981. The loop is used for full-scale testing of components for the Fusion Materials Irradiation Test (FMIT) Facility. Main system parameters include coolant temperatures to 430 0 C and flow to 0.038 m 3 /s (600 gal/min). Performance of the main pump, vacuum system, and control system is discussed. Unique test capabilities of the ELS are also discussed

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

International Nuclear Information System (INIS)

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

2013-01-01

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

A COLD FLARE WITH DELAYED HEATING

International Nuclear Information System (INIS)

Fleishman, Gregory D.; Pal'shin, Valentin D.; Lysenko, Alexandra L.; Meshalkina, Natalia; Kashapova, Larisa K.; Altyntsev, Alexander T.

2016-01-01

Recently, a number of peculiar flares have been reported that demonstrate significant nonthermal particle signatures with low, if any, thermal emission, which implies a close association of the observed emission with the primary energy release/electron acceleration region. This paper presents a flare that appears “cold” at the impulsive phase, while displaying delayed heating later on. Using hard X-ray data from Konus- Wind , microwave observations by SSRT, RSTN, NoRH, and NoRP, context observations, and three-dimensional modeling, we study the energy release, particle acceleration, and transport, and the relationships between the nonthermal and thermal signatures. The flaring process is found to involve the interaction between a small loop and a big loop with the accelerated particles divided roughly equally between them. Precipitation of the electrons from the small loop produced only a weak thermal response because the loop volume was small, while the electrons trapped in the big loop lost most of their energy in the coronal part of the loop, which resulted in coronal plasma heating but no or only weak chromospheric evaporation, and thus unusually weak soft X-ray emission. The energy losses of the fast electrons in the big tenuous loop were slow, which resulted in the observed delay of the plasma heating. We determined that the impulsively accelerated electron population had a beamed angular distribution in the direction of the electric force along the magnetic field of the small loop. The accelerated particle transport in the big loop was primarily mediated by turbulent waves, which is similar to other reported cold flares.

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

Thermal performance of horizontal closed-loop oscillating heat-pipe with check valves

International Nuclear Information System (INIS)

Rittidech, S.; Pipatpaiboon, N.; Thongdaeng, S.

2010-01-01

This research investigated the thermal performance of various horizontal closed-loop oscillating heat-pipe systems with check valves (HCLOHPs/CVs). Numerous test systems were constructed using copper capillary tubes with assorted inner diameters, evaporator lengths, and check valves. The test systems were evaluated under normal operating conditions using ethanol, R123, and distilled water as working fluids. The system's evaporator sections were heated by hot water from a hot bath, and the heat was removed from the condenser sections by cold water from a cool bath. The adiabatic sections were well insulated with foam insulators. The heat-transfer performance of the various systems was evaluated in terms of the rate of heat transferred to the cold water at the condenser. The results showed that the heat-transfer performance of an HCLOHP/CV system could be improved by decreasing the evaporator length. The highest performance of all tested systems was obtained when the maximum number of system check valves was 2. The maximum heat flux occurred with a 2 mm inner diameter tube, and R123 was determined to be the most suitable working fluid

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

Energy Technology Data Exchange (ETDEWEB)

Kim, J. Y.

1998-02-01

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

Summary of TRUEX Radiolysis Testing Using the INL Radiolysis Test Loop

Energy Technology Data Exchange (ETDEWEB)

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

2012-03-01

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

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

International Nuclear Information System (INIS)

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

2007-02-01

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

Aerobic exercise and cold pressor test induce hypoalgesia in active and inactive men and women

DEFF Research Database (Denmark)

Vægter, Henrik Bjarke; Handberg, Gitte; Jørgensen, Maria N.

2015-01-01

). Conditioned pain modulation (CPM) was assessed by cold pressor testing. Exercise-induced hypoalgesia (EIH) was assessed after 15 minutes bicycling at a heart rate corresponding to 75% VO2max. A control session of 15 minutes quiet rest was also included. Pressure pain thresholds (PPTs) were recorded...... and after exercise, PPTs increased to the same degree in active and inactive subjects, and the CPM and EIH responses were correlated (P CPM response immediately after cold pressor test was maintained in women but not in men. CONCLUSIONS: Cold pressor stimulation and aerobic exercise caused...... comparable multisegmental increases in PPT in active and inactive men and women. The CPM and EIH responses were correlated, but they have different temporal manifestation of hypoalgesia....

Feasibility and repeatability of cold and mechanical quantitative sensory testing in normal dogs

Science.gov (United States)

Briley, Jessica D.; Williams, Morika D.; Freire, Mila; Griffith, Emily H.; Lascelles, B. Duncan X.

2015-01-01

Feasibility and inter-session repeatability of cold and mechanical quantitative sensory testing (QST) were assessed in 24 normal dogs. Cold thermal latencies were evaluated using a thermal probe (0 °C) applied to three pelvic limb sites. Mechanical thresholds were measured using an electronic von Frey anesthesiometer (EVF) and a blunt-probed pressure algometer (PA) applied to the dorsal aspect of the metatarsus. All QST trials were performed with dogs in lateral recumbency. Collection of cold QST data was easy (feasible) in 19/24 (79%) dogs. However, only 18.4%, 18.9% and 13.2% of cold QST trials elicited a response at the medial tibia, third digital pad and plantar metatarsal regions, respectively. Collection of mechanical QST data was easy (feasible) in 20/24 (83%) dogs for both EVF and PA. At consecutive sampling times, approximately 2 weeks apart, the average EVF sensory thresholds were 414 ± 186 g and 379 ± 166 g, respectively, and the average PA sensory thresholds were 1089 ± 414 g and 1028 ± 331 g, respectively. There was no significant difference in inter-session or inter-limb threshold values for either mechanical QST device. The cold QST protocol in this study was achievable, but did not provide consistently quantifiable results. Both mechanical QST devices tested provided repeatable, reliable sensory threshold measurements in normal, client-owned dogs. These findings contribute to the validation of the EVF and PA as tools to obtain repeated QST data over time in dogs to assess somatosensory processing changes. PMID:24268475

Does acetaminophen/hydrocodone affect cold pulpal testing in patients with symptomatic irreversible pulpitis? A prospective, randomized, double-blind, placebo-controlled study.

Science.gov (United States)

Fowler, Sara; Fullmer, Spencer; Drum, Melissa; Reader, Al

2014-12-01

The purpose of this prospective randomized, double-blind, placebo-controlled study was to determine the effects of a combination dose of 1000 mg acetaminophen/10 mg hydrocodone on cold pulpal testing in patients experiencing symptomatic irreversible pulpitis. One hundred emergency patients in moderate to severe pain diagnosed with symptomatic irreversible pulpitis of a mandibular posterior tooth randomly received, in a double-blind manner, identical capsules of either a combination of 1000 mg acetaminophen/10 hydrocodone or placebo. Cold testing with Endo-Ice (1,1,1,2 tetrafluoroethane; Hygenic Corp, Akron, OH) was performed at baseline and every 10 minutes for 60 minutes. Pain to cold testing was recorded by the patient using a Heft-Parker visual analog scale. Patients' reaction to the cold application was also rated. Cold testing at baseline and at 10 minutes resulted in severe pain for both the acetaminophen/hydrocodone and placebo groups. Although pain ratings decreased from 20-60 minutes, the ratings still resulted in moderate pain. Patient reaction to cold testing showed that 56%-62% had a severe reaction. Although the reactions decreased in severity over the 60 minutes, 20%-34% still had severe reactions at 60 minutes. Regarding pain and patients' reactions to cold testing, there were no significant differences between the combination acetaminophen/hydrocodone and placebo groups at any time period. A combination dose of 1000 mg of acetaminophen/10 mg of hydrocodone did not statistically affect cold pulpal testing in patients presenting with symptomatic irreversible pulpitis. Patients experienced moderate to severe pain and reactions to cold testing. Copyright © 2014 American Association of Endodontists. Published by Elsevier Inc. All rights reserved.

JRR-3 cold neutron source facility H2-O2 explosion safety proof testing

International Nuclear Information System (INIS)

Hibi, T.; Fuse, H.; Takahashi, H.; Akutsu, C.; Kumai, T.; Kawabata, Y.

1990-01-01

A cold Neutron Source (CNS) will be installed in Japan Research Reactor-3 (JRR-3) in Japan Atomic Energy Research Institute (JAERI) during its remodeling project. This CNS holds liquid hydrogen at a temperature of about 20 K as a cold neutron source moderator in the heavy water area of the reactor to moderate thermal neutrons from the reactor to cold neutrons of about 5 meV energy. In the hydrogen circuit of the CNS safety measures are taken to prevent oxygen/hydrogen reaction (H 2 -O 2 explosion). It is also designed in such manner that, should an H 2 -O 2 explosion take place, the soundness of all the components can be maintained so as not to harm the reactor safety. A test hydrogen circuit identical to that of the CNS (real components designed by TECHNICATOME of France) was manufactured to conduct the H 2 -O 2 explosion test. In this test, the detonation that is the severest phenomenon of the oxygen/hydrogen reaction took place in the test hydrogen circuit to measure the exerted pressure on the components and their strain, deformation, leakage, cracking, etc. Based on the results of this measurement, the structural strength of the test hydrogen circuit was analyzed. The results of this test show that the hydrogen circuit components have sufficient structural strength to withstand an oxygen/hydrogen reaction

Multiple Flow Loop SCADA System Implemented on the Production Prototype Loop

Energy Technology Data Exchange (ETDEWEB)

Baily, Scott A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dalmas, Dale Allen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wheat, Robert Mitchell [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Woloshun, Keith Albert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dale, Gregory E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-11-16

The following report covers FY 15 activities to develop supervisory control and data acquisition (SCADA) system for the Northstar Moly99 production prototype gas flow loop. The goal of this effort is to expand the existing system to include a second flow loop with a larger production-sized blower. Besides testing the larger blower, this system will demonstrate the scalability of our solution to multiple flow loops.

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

Science.gov (United States)

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

2016-01-01

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

Techno-economic evaluation for the heat integration of vaporisation cold energy in natural gas processing

International Nuclear Information System (INIS)

Koku, Oludolapo; Perry, Simon; Kim, Jin-Kuk

2014-01-01

Highlights: • Development of thermal integration modelling framework for the utilisation of LNG cold energy. • Feasibility study for various design options for the integration of low-temperature cold energy. • Provision of a design approach for achieving efficient use of cold energy in LNG terminals. • Understanding of techno-economic impacts associated with the thermal integration of LNG cold energy. - Abstract: This paper addresses a conceptual study investigating the techno-economic feasibility for the thermal Integration of LNG cold vaporisation energy in power generation applications. In conventional regasification systems, this valuable LNG cold energy is often being wasted to ambient heat sources, representing a thermodynamic inefficient process with a significant thermal impact on the local environment. A combined facility consisting of a non-integrated Combined Cycle Power Plant (CCPP) and an LNG receiving terminal employing traditional Open Rack Vaporisers (ORV) technology, has been modelled, as a base case. Retrofit strategies for the integration of LNG cold energy have been investigated, and their impacts on power production and system efficiency are systematically compared. Retrofit design options considered in this work include the use of a propane Rankine cycle coupled with the direct expansion of natural gas, the integration of a closed-loop water cycle or open-loop water circuit with a steam Rankine cycle, and the facilitation of integrated air cooling for a gas turbine

Feasibility study on the transient fuel test loop installation

International Nuclear Information System (INIS)

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

1997-02-01

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

Development of a model and test equipment for cold flow tests at 500 atm of small nuclear light bulb configurations

Science.gov (United States)

Jaminet, J. F.

1972-01-01

A model and test equipment were developed and cold-flow-tested at greater than 500 atm in preparation for future high-pressure rf plasma experiments and in-reactor tests with small nuclear light bulb configurations. With minor exceptions, the model chamber is similar in design and dimensions to a proposed in-reactor geometry for tests with fissioning uranium plasmas in the nuclear furnace. The model and the equipment were designed for use with the UARL 1.2-MW rf induction heater in tests with rf plasmas at pressures up to 500 atm. A series of cold-flow tests of the model was then conducted at pressures up to about 510 atm. At 504 atm, the flow rates of argon and cooling water were 3.35 liter/sec (STP) and 26 gal/min, respectively. It was demonstrated that the model is capable of being operated for extended periods at the 500-atm pressure level and is, therefore, ready for use in initial high-pressure rf plasma experiments.

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

International Nuclear Information System (INIS)

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

1990-11-01

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

Effect of cold working on the corrosion resistance of JPCA stainless steel in flowing Pb–Bi at 450 °C

International Nuclear Information System (INIS)

Rivai, Abu Khalid; Saito, Shigeru; Tezuka, Masao; Kato, Chiaki; Kikuchi, Kenji

2012-01-01

Development of a high performance proton beam window material is one of the critical issues for the deployment of the accelerator-driven transmutation system (ADS) with liquid Pb–Bi eutectic as a spallation target and coolant. In the present study, we applied 20% cold work treatment to JPCA austenitic stainless steel and investigated it from the corrosion behavior viewpoint. The corrosion test of 20% cold-worked JPCA SS has been carried in the JLBL-1 (JAEA Lead–Bismuth Loop-1) apparatus. The maximum temperature, the temperature difference, the flow velocity and the exposure time of the liquid Pb–Bi were 450 °C, 100 °C, 1 m/s, and 1000 h, respectively. For comparison analysis, JPCA SS without cold working was also tested in the same time and conditions with the 20% cold-worked JPCA SS. The results showed a different corrosion behavior between the JPCA SS without and with cold working. As for the JPCA SS without cold working, Pb–Bi penetrated into the matrix through a ferrite layer which was formed because of constituent metals dissolution from the matrix into Pb–Bi. As for the 20% cold-worked JPCA SS, dissolution attack occurred only partially and formed localized superficial pitting corrosion. It was found that the different corrosion behavior occurred because the cold working induced a structure transformation from γ-austenite to α′-martensite and affected the corrosion resistance of the JPCA SS in flowing Pb–Bi at 450 °C.

Corrosion and deposition behaviour of 60Co and 54Mn in the SNR mockup loop for the primary sodium system

International Nuclear Information System (INIS)

Menken, G.; Reichel, H.

1976-01-01

The SNR corrosion mockup loop, simulating the SNR primary system is described. The influence of hydraulic conditions and temperature on the deposition behaviour is studied. γ-spectroscopy measurements at the pipe-work and removable samples allowed to determine the distribution of radioactive corrosion products in the loop and by-pass system. The release rate of Mn 54 could be reduced by a factor of 3 by decreasing the cold trap temperature from 165 0 C to 105 0 C while the Co 60 release rate could be reduced by a factor of 14, respectively. High temperature loop sections (873K) representing 13% of the loop surface absorbed 50-60% of the released Co 60 and only 8 - 18% of the Mn 54. The cold trap absorbed not more than 1% of the Co 60 and 10% of the Mn 54 inventory. (author)

Teste de frio na avaliação do vigor de sementes de feijão Cold test for bean seed vigor evaluation

Directory of Open Access Journals (Sweden)

Marcelo Hissnauer Miguel

1999-01-01

Full Text Available O presente trabalho teve como objetivo estudar o comportamento de sementes de feijão submetidas a diferentes metodologias do teste de frio, comparativamente a outros testes de vigor tradicionalmente utilizados na avaliação da qualidade fisiológica dessas sementes. As metodologias do teste de frio utilizadas foram: caixa plástica com terra, rolo de papel com terra e rolo de papel sem terra, nas temperaturas de 10ºC e de 15ºC e períodos de exposição, de três, cinco e sete dias. Paralelamente foram conduzidos teste padrão de germinação, primeira contagem de germinação, envelhecimento acelerado, condutividade elétrica e emergência de plântulas em campo. As diferentes metodologias do teste de frio foram comparadas pelo teste de Tukey e, posteriormente, foram estabelecidas correlações com os demais testes de vigor. A análise dos dados e a interpretação dos resultados permitiram concluir que a metodologia do rolo de papel sem terra, nos períodos de três e cinco dias à 10ºC e à 15ºC, proporcionou melhor correlação com a primeira contagem de germinação e emergência de plântulas em campo.To assess the efficiency of cold test for bean seed vigor determination, two cultivars were evaluated by three different methodologies under two temperatures (10ºC and 15ºC and three cold period (3, 5 and 7 days regimes. The results were compared to the routine seed quality tests, such as the standard germination test, first germination count, accelerated aging, electrical conductivity and emergence in the field. Tests were conducted in two seasons, four months apart. The results obtained using the various assays were compared by Tukey's test. Correlation analysis between the cold test methods and other vigor tests were carried out. Paper rolls without soil for a period of three to five days at 10ºC and 15ºC correlated better with the standard vigor tests, especially the first germination count and field emergence. To assess the

Effect of cold work on tensile behavior of irradiated type 316 stainless steel

International Nuclear Information System (INIS)

Klueh, R.L.; Maziasz, P.J.

1986-01-01

Tensile specimens were irradiated in ORR at 250, 290, 450, and 500 0 C to produce a displacement damage of approx.5 dpa and 40 at. ppM He. Irradiation at 250 and 290 0 C caused an increase in yield stress and ultimate tensile strength and a decrease in ductility relative to unaged and thermally aged controls. The changes were greatest for the 20%-cold-worked steel and lowest for the 50%-cold-worked steel. Irradiation at 450 0 C caused a slight relative decrease in strength for all cold-worked conditions. A large decrease was observed at 500 0 C, with the largest decrease occurring for the 50%-cold-worked specimen. No bubble, void, or precipitate formation was observed for specimens examined by transmission electron microscopy (TEM). The irradiation hardening was correlated with Frank-loop and ''black-dot'' loop damage. A strength decrease at 500 0 C was correlated with dislocation network recovery. Comparison of tensile and TEM results from ORR-irradiated steel with those from steels irradiated in the High Flux Isotope Reactor and the Experimental Breeder Reactor indicated consistent strength and microstructure changes

Cold pressor test on atomic bomb survivors, Nagasaki

Energy Technology Data Exchange (ETDEWEB)

Sasaki, Tomoyoshi; Sweedler, D R; Okamoto, Akira

1964-03-12

Cold pressor test was performed on a sample of 1156 atomic bomb survivors and other persons (ages ranging between 15 to 81 years) residing in Nagasaki City. Response values differed according to such factors as age, sex, blood pressure and month of examination. The response in systolic pressure increased with age but no evidence was found to support an acceleration of aging by irradiation. The response in diastolic blood pressure showed no change with age, but differed between Comparison Groups during the summer months. However, this was apparently due to some other cause than exposure to the atomic bomb. 25 references, 8 tables.

Design and operation of a small (benchtop) pumped sodium loop

International Nuclear Information System (INIS)

Trevillion, E.A.; Rowe, D.M.J.

1975-08-01

The report outlines the design and operation of a small (benchtop) pumped sodium loop (sodium, 650g). The loop incorporates a diffusion cold trap to control the oxygen impurity level in the sodium and a sodium sampler/distillation unit to enable sodium samples to be analysed for impurities. Sodium flow rates of up to 5.5cm.s -1 (1cm 3 .s -1 ) have been achieved at temperatures up to 673.2K (400 0 C) and temperatures of up to 1023.2K (750 0 C) have been achieved under static conditions. A device for the addition and removal of metallic speciments to and from the loop sodium without contamination of either the specimens or the sodium is also described. (author)

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

International Nuclear Information System (INIS)

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

1980-08-01

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

Effect of cold work on low-temperature sensitization behaviour of austenitic stainless steels

Energy Technology Data Exchange (ETDEWEB)

Kain, V. E-mail: vivkain@apsara.barc.ernet.in; Chandra, K.; Adhe, K.N.; De, P.K

2004-09-01

The effects of cold work and low-temperature sensitization heat treatment of non-sensitized austenitic stainless steels have been investigated and related to the cracking in nuclear power reactors. Types 304, 304L and 304LN developed martensite after 15% cold working. Heat treatment of these cold worked steels at 500 deg. C led to sensitization of grain boundaries and the matrix and a desensitization effect was seen in 11 days due to fast diffusion rate of chromium in martensite. Types 316L and 316LN did not develop martensite upon cold rolling due to its chemical composition suppressing the martensite transformation (due to deformation) temperature, hence these were not sensitized at 500 deg. C. The sensitization of the martensite phase was always accompanied by a hump in the reactivation current peak in the double loop electrochemical potentiokinetic reactivation test, thus providing a test to detect such sensitization. It was shown that bending does not produce martensite and therefore, is a better method to simulate weld heat affected zone. Bending and heating at 500 deg. C for 11 days led to fresh precipitation due to increased retained strain and desensitization of 304LN due to faster diffusion rate of chromium along dislocations. The as received or solution annealed 304 and 304LN with 0.15% nitrogen showed increased sensitization after heat treatment at 500 deg. C, indicating the presence of carbides/nitrides.

Molecular Structural Basis for the Cold Adaptedness of the Psychrophilic β-Glucosidase BglU in Micrococcus antarcticus.

Science.gov (United States)

Miao, Li-Li; Hou, Yan-Jie; Fan, Hong-Xia; Qu, Jie; Qi, Chao; Liu, Ying; Li, De-Feng; Liu, Zhi-Pei

2016-01-22

Psychrophilic enzymes play crucial roles in cold adaptation of microbes and provide useful models for studies of protein evolution, folding, and dynamic properties. We examined the crystal structure (2.2-Å resolution) of the psychrophilic β-glucosidase BglU, a member of the glycosyl hydrolase 1 (GH1) enzyme family found in the cold-adapted bacterium Micrococcus antarcticus. Structural comparison and sequence alignment between BglU and its mesophilic and thermophilic counterpart enzymes (BglB and GlyTn, respectively) revealed two notable features distinct to BglU: (i) a unique long-loop L3 (35 versus 7 amino acids in others) involved in substrate binding and (ii) a unique amino acid, His299 (Tyr in others), involved in the stabilization of an ordered water molecule chain. Shortening of loop L3 to 25 amino acids reduced low-temperature catalytic activity, substrate-binding ability, the optimal temperature, and the melting temperature (Tm). Mutation of His299 to Tyr increased the optimal temperature, the Tm, and the catalytic activity. Conversely, mutation of Tyr301 to His in BglB caused a reduction in catalytic activity, thermostability, and the optimal temperature (45 to 35°C). Loop L3 shortening and H299Y substitution jointly restored enzyme activity to the level of BglU, but at moderate temperatures. Our findings indicate that loop L3 controls the level of catalytic activity at low temperatures, residue His299 is responsible for thermolability (particularly heat lability of the active center), and long-loop L3 and His299 are jointly responsible for the psychrophilic properties. The described structural basis for the cold adaptedness of BglU will be helpful for structure-based engineering of new cold-adapted enzymes and for the production of mutants useful in a variety of industrial processes at different temperatures. Copyright © 2016, American Society for Microbiology. All Rights Reserved.

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-01

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

Tests of cold helium compressors at Fermilab

International Nuclear Information System (INIS)

Peterson, T.J.; Fuerst, J.D.

1987-10-01

Fermilab has tested two cold helium compressors for possible installation in the satellite refrigerator buildings of the Tevatron cryogenic system. Operating conditions required to obtain an overall Tevatron energy upgrade from 900 to 1000 GeV are (for each of 24 machines): 52 g/s mass flow rate, 0.7 atm inlet pressure, 1.4 atm exhaust pressure. Acceptable efficiency is in the 60% range. Both Creare, Inc., and Cryogenic Consultants, Inc. (CCI), have supplied units for evaluation. The Creare machine is a high speed centrifugal pump/compressor which yielded 60% adiabatic efficiency (including an approximately 20 watt heat leak) with a 1.0 atm inlet pressure and 55 g/s flow rate. Certain mechanical difficulties were present, chiefly the device's inability to withstand two-phase flow. CCI supplied a reciprocating unit which, after initial testing and modification, achieved 59% efficiency with an approximate 35 watt heat leak at a 0.7 atm inlet pressure and 48 g/s flow rate. Although the device lacks the smooth, quiet operating characteristics of a turbomachine, it has endured mechanically throughout testing and is entirely insensitive to two-phase flow

Stability of the Horizontal Curvature of the LHC Cryodipoles During Cold Tests

CERN Document Server

Cano, E D Fernandez; García-Pérez, J; Jeanneret, Jean Bernard; Poncet, A; Seyvet, F; Tovar-Gonzalez, A; Wildner, E; IEEE Trans. Nucl. Sci.

2006-01-01

The LHC will be composed of 1232 horizontally curved, 15 meter long, superconducting dipole magnets cooled at 1.9 K. They are supported within their vacuum vessel by three Glass Fiber Reinforced Epoxy (GFRE) support posts. Each cryodipole is individually cold tested at CERN before its installation and interconnection in the LHC 27 km circumference tunnel. As the magnet geometry under cryogenic operation is extremely important for the LHC machine aperture, a new method has been developed at CERN in order to monitor the magnet curvature change between warm and cold states. It enabled us to conclude that there is no permanent horizontal curvature change of the LHC dipole magnet between warm and cold states, although a systematic horizontal transient deformation during cool-down was detected. This deformation generates loads in the dipole supporting system; further investigation permitted us to infer this behavior to the asymmetric thermal contraction of the rigid magnet thermal shield during cool-down. Controlli...

SRF cavity testing using a FPGA Self Excited Loop

CERN Document Server

Ben-Zvi, Ilan

2018-01-01

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

Long-Duration Testing of a Temperature-Swing Adsorption Compressor for Carbon Dioxide for Closed-Loop Air Revitalization Systems

Science.gov (United States)

Rosen, Micha; Mulloth, Lila; Varghese, Mini

2005-01-01

This paper describes the results of long-duration testing of a temperature-swing adsorption compressor that has application in the International Space Station (ISS) and future spacecraft for closing the air revitalization loop. The air revitalization system of the ISS operates in an open loop mode and relies on the resupply of oxygen and other consumables from Earth for the life support of astronauts. A compressor is required for delivering the carbon dioxide from a removal assembly to a reduction unit to recover oxygen and thereby closing the air-loop. The TSAC is a solid-state compressor that has the capability to remove CO2 from a low-pressure source, and subsequently store, compress, and deliver at a higher pressure as required by a processor. The TSAC is an ideal interface device for CO2 removal and reduction units in the air revitalization loop of a spacecraft for oxygen recovery. The TSAC was developed and its operation was successfully verified in integration tests with the flight-like Carbon Dioxide Removal Assembly (CDRA) at Marshall Space Flight Center prior to the long-duration tests. Long-duration tests reveal the impacts of repeated thermal cycling on the compressor components and the adsorbent material.

Regeneration of the cold trap of the PEC mechanism testing plant

International Nuclear Information System (INIS)

Caponetti, R.; Petrazzuolo, F.

1984-01-01

Experimentation on prototypes of PEC reactor blocking mechanisms is presently in course at Casaccia Cre in the experimental engineering division of the fast reactor department. After a brief description of the experimental cycle of the components, this repor shows the design criteria of a selected method for the regeneration of mechenism testing plant cold trap

Compression test of cold-formedsteel perforated profile with steel sheathing

Directory of Open Access Journals (Sweden)

Shamanin Aleksandr Yur’evich

2015-05-01

Full Text Available The subject of this paper is the stability and strength of cold-formed and perforated steel sigma-section columns with steel sheathing of different thickness. Ceilings with and without steel sheathing of different thickness are tested to failure in compression on a laboratory machine, which was based on a manual hydraulic jack. Series of 4 experiments with full-scale walls (2.5 m height were carried out. Also, for examination of the role of boundary conditions, the sheet in a ceiling is either left free or connected to base with screws.In civil engineering there are many experiments and methodologies for calculating the strength and buckling of ceiling with the sheathing of various materials, such as oriented strand board and gypsum board. However, for producing superstructures of ships the materials with high plastic properties and strength characteristics are required. For example steel possesses such properties. It was the main reason for conducting a series of experiments and studying the behavior of cold-formed steel columns with steel sheathing. During the experiments the deformation of the cross-section of three equally spaced cross sections was determined, as well as the axial deformation of the central column in the ceiling with steel sheathing.The test results showed the influence of the thickness of sheathing and boundary condition of a sheet on the strength and buckling of ceiling. According to the results of the tests it is necessary to evaluate the impact of the sheathing made of different materials and if necessary to carry out further tests.

Testing of Rice Stocks for Their Survival of Winter Cold

Directory of Open Access Journals (Sweden)

Hiroshi Ikehashi

2018-03-01

Full Text Available Rice cultivation is considered to be initiated by vegetative propagation of sprout from wild perennial stocks. To test whether any presently cultivated rice cultivar can survive the winter cold or not, rice stocks of several cultivars including indica and japonica types were placed in a shallow pool from October to April in 2015–2016 and 2016–2017. During the coldest period of the winter, the bases of the stocks were placed 5–6 cm below the surface of water, where temperatures ranged from 3 °C to 5 °C, while the surface was frozen for two or three times and covered with snow for a day. Only one cultivar, Nipponbare, a japonica type, survived the winter cold and regenerated sprouts in the end of April or early May. A possibility to develop perennial cultivation of rice or perennial hybrid rice is discussed.

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

International Nuclear Information System (INIS)

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

2007-02-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2006-07-03

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1988-07-01

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

Cold Test Results of the ISAC-II Medium Beta High Gradient Cryomodule

CERN Document Server

Laxdal, R E; Clark, G S; Fong, K; Mitra, A K; Poirier, R L; Rawnsley, B; Ries, T; Sekatchev, I; Stanford, G; Zvyagintsev, V

2004-01-01

Many proposals (RIA, Eurisol, ISAC-II) are emerging for a new generation of high gradient heavy ion accelerators. The ISAC-II medium beta cryomodule represents the first realized application that encorporates many new techniques to improve the performance over machines presently being used for beam delivery. The machine lattice, compatible with multi-charge acceleration, uses high field (9T) superconducting solenoids with bucking coils for active fringe field compensation. The bulk niobium quarter wave medium beta cavity produces 6 MV/m over an effective length of 18cm with a peak surface field of ~30 MV/m. TRIUMF has developed a mechanical tuner capable of both coarse (kHz) and fine (Hz) frequency adjustments of the cavity. The demonstrated tuner resolution is better than 0.1 μm (0.6 Hz). A new rf coupling loop has been developed that operates at 200 Watts forward power with less than 0.5 Watt of power being added to the helium load. Cold alignment in ISAC-II has been done with rf pick-ups using a ...

Alternative procedure for the cold test for soybean seeds Procedimento alternativo para o teste de frio em semente de soja

Directory of Open Access Journals (Sweden)

Bruno Guilherme Torres Licursi Vieira

2010-10-01

Full Text Available The growing demand for high quality soybean [Glycine max (L. Merrill] seeds requires a precise seed quality control system from the seed industry. One way to accomplish this is by improving vigor testing. Cold test has been traditionally employed for corn seeds. However, it has also been used for other seed crops such as cotton (Gossypium spp., soybean (Glycine Max, dry bean (Phaseolus vulgaris and pea (Pisum sativum. This study was carried out with the objective of adjusting an alternative procedure for the cold test to determine soybean seed vigor. Six commercial soybean seed lots of the cultivar BRS 133 were used. The physiological potential of the seed lots was evaluated by germination on paper towel and sand box, seedling field emergence, tetrazolium, accelerated aging and electrical conductivity tests. Seed moisture content was also determined. The temperature used for the cold test procedures was 10ºC during five days. Four cold test procedures were evaluated: i plastic boxes with soil; ii rolled paper towel with soil; iii rolled paper towel without soil, and iv an alternative procedure, using rolled paper towel without soil under cold water. A completely randomized experimental design with eight replications was used and the means were compared by the Tukey test (p = 0.05. To verify the dependence between the alternative test and others single linear correlation was used. All cold test procedures had similar coefficients of variation (CV, highlighting that rolled paper towel with soil and the alternative procedure had the best performance, with an average of 94% and 93% normal seedlings and CV of 3.2% and 3.6%, respectively. The alternative procedure has satisfactory results for estimating soybean seed vigor, yielding consistent results compared to the traditional procedure.A crescente demanda por semente de soja [Glycine max (L. Merrill] de alta qualidade tem requerido da indústria de sementes um controle de qualidade mais preciso

Numerical simulation of losses along a natural circulation helium loop

Energy Technology Data Exchange (ETDEWEB)

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

2016-06-30

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

Reactor loops at Chalk River

International Nuclear Information System (INIS)

Sochaski, R.O.

1962-07-01

This report describes broadly the nine in-reactor loops, and their components, located in and around the NRX and NRU reactors at Chalk River. First an introduction and general description is given of the loops and their function, supplemented with a table outlining some loop specifications and nine simplified flow sheets, one for each individual loop. The report then proceeds to classify each loop into two categories, the 'main loop circuit' and the 'auxiliary circuit', and descriptions are given of each circuit's components in turn. These components, in part, are comprised of the main loop pumps, the test section, loop heaters, loop coolers, delayed-neutron monitors, surge tank, Dowtherm coolers, loop piping. Here again photographs, drawings and tables are included to provide a clearer understanding of the descriptive literature and to include, in tables, some specifications of the more important components in each loop. (author)

Plutonium immobilization program - Cold pour Phase 1 test results

International Nuclear Information System (INIS)

Hamilton, L.

2000-01-01

The Plutonium Immobilization Project will disposition excess weapons grade plutonium. It uses the can-in-canister approach that involves placing plutonium-ceramic pucks in sealed cans that are then placed into Defense Waste Processing Facility canisters. These canisters are subsequently filled with high-level radioactive waste glass. This process puts the plutonium in a stable form and makes it unattractive for reuse. A cold (non-radioactive) glass pour program was performed to develop and verify the baseline design for the canister and internal hardware. This paper describes the Phase 1 scoping test results

Plutonium Immobilization Program - Cold pour Phase 1 test results

International Nuclear Information System (INIS)

Hamilton, L.

2000-01-01

The Plutonium Immobilization Project will disposition excess weapons grade plutonium. It uses the can-in-canister approach that involves placing plutonium-ceramic pucks in sealed cans that are then placed into Defense Waste Processing Facility canisters. These canisters are subsequently filled with high-level radioactive waste glass. This process puts the plutonium in a stable form and makes it unattractive for reuse. A cold (non-radioactive) glass pour program was performed to develop and verify the baseline design for the canister and internal hardware. This paper describes the Phase 1 scoping test results

The hydrogen influenced cold cracking tendency of two high strength low alloy steels - evaluated by the implant-test

International Nuclear Information System (INIS)

Neumann, V.; Schoenherr, W.

1978-01-01

A possible way of evaluating the hydrogen influenced cold cracking tendency of constructional steels is the implant test. Using this testing method, it is possible to adjust extensively independently of one other the three influencing parameters - hydrogen content of the welding deposit and the heat-affected zone, hardness structure and stresses - and to examine their effect on the crack behaviour. Due to the same microstructure formation in the heat affected zone of the implant samples and in the non-heat affected regions from the consequent position of the heat affected zone of component seams, welding conditions can be determined with suitable changing of the sample whose application to the real component practically excludes the danger of cold cracking. The broken surfaces in cold cracking are partly ductile and poor in deformation. The deformation-poor fracturing can possibly take an intercrystalline or transcrystalline course according to the chemical composition of the steel. The investigation confirm the theories and test results of other authors: The formation of deformation-poor, typical fracture sections for cold cracking was only obtained when there was a clear delay between putting on the test load and fracture of the sample. (orig./RW) [de

New Technique for Cryogenically Cooling Small Test Articles

Science.gov (United States)

Rodriquez, Karen M.; Henderson, Donald J.

2011-01-01

Convective heat removal techniques to rapidly cool small test articles to Earth-Moon L2 temperatures of 77 K were accomplished through the use of liquid nitrogen (LN2). By maintaining a selected pressure range on the saturation curve, test articles were cooled below the LN2 boiling point at ambient pressure in less than 30 min. Difficulties in achieving test pressures while maintaining the temperature tolerance necessitated a modification to the original system to include a closed loop conductive cold plate and cryogenic shroud

Performance Tests for Bubble Blockage Device

International Nuclear Information System (INIS)

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

2014-01-01

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

High flux isotope reactor cold source preconceptual design study report

International Nuclear Information System (INIS)

Selby, D.L.; Bucholz, J.A.; Burnette, S.E.

1995-12-01

In February 1995, the deputy director of Oak Ridge National Laboratory (ORNL) formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced Neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. The anticipated cold source will consist of a cryogenic LH 2 moderator plug, a cryogenic pump system, a refrigerator that uses helium gas as a refrigerant, a heat exchanger to interface the refrigerant with the hydrogen loop, liquid hydrogen transfer lines, a gas handling system that includes vacuum lines, and an instrumentation and control system to provide constant system status monitoring and to maintain system stability. The scope of this project includes the development, design, safety analysis, procurement/fabrication, testing, and installation of all of the components necessary to produce a working cold source within an existing HFIR beam tube. This project will also include those activities necessary to transport the cold neutron beam to the front face of the present HFIR beam room. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and research and development (R and D), (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the preconceptual phase and establishes the concept feasibility. The information presented includes the project scope, the preliminary design requirements, the preliminary cost and schedule, the preliminary performance data, and an outline of the various plans for completing the project

Operational techniques employed for the liquid sodium source term control loops

International Nuclear Information System (INIS)

Chulos, L.E.

1976-01-01

Four Source Term Control Loops (STCLs) have been designed, constructed, and placed into operation at the Hanford Engineering Development Laboratory (HEDL) as part of the Radioactivity Control Technology program. The data obtained are used to determine the corrosion and deposition of LMFBR materials, including corrosion product radionuclides, in a non-isothermal flowing sodium system. The paper discusses operation of the STCL Facilities and, in particular, the methods used for controlling the oxygen content of the liquid sodium. These methods include cold trapping techniques, hot trapping, seeding the cold traps with sodium oxide, and precipitating the oxygen in the cold trap in a controlled manner. Operational problems encountered with the STCL Facilities and the techniques for correcting these problems are also discussed

Summary of ROSA-4 LSTF first phase test program and station blackout (TMLB) test results

International Nuclear Information System (INIS)

Tasaka, K.; Kukita, Y.; Anoda, Y.

1990-01-01

This paper summarizes major test results obtained at the ROSA-4 Large Scale Test Facility (LSTF) during the first phase of the test program. The results from a station blackout (TMLB) test conducted at the end of the first-phase program are described in some detail. The LSTF is an integral test facility being operated by the Japan Atomic Energy Research Institute for simulation of pressurized water reactor (PWR) thermal-hydraulic responses during small-break loss-of-coolant accidents (SBLOCAs) and operational/abnormal transients. It is a 1/48 volumetrically scaled, full-height, full-pressure simulator of a Westinghouse-type 4-loop PWR. The facility includes two symmetric primary loops each one containing an active inverted-U tube steam generator and an active reactor coolant pump. The loop horizontal legs are sized to conserve the scaled (1/24) volumes as well as the length to the square root of the diameter ratio in order to simulate the two-phase flow regime transitions. The primary objective of the LSTF first-phase program was to define the fundamental PWR thermal-hydraulic responses during SBLOCAs and transients. Most of the tests were conducted with simulated component/operator failures, including unavailability of the high pressure injection system and auxiliary feedwater system, as well as operator failure to take corrective actions. The forty-two first phase tests included twenty-nine SBLOCA tests conducted mainly for cold leg breaks, three abnormal transient tests and ten natural circulation tests. Attempts were made in several of the SBLOCA tests to simulate the plant recovery procedures as well as candidate accident management measures for prevention of high-pressure core melt situation. The natural circulation tests simulated the single-phase and two-phase natural circulation as well as reflux condensation behavior in the primary loops in steady or quasi-steady states

Experience on the removal of impurities from liquid metal systems by cold-trapping

Energy Technology Data Exchange (ETDEWEB)

Bray, J. A.

1963-10-15

Experience in impurity removal by cold-trapping, which was obtained on DFR and its associated liquid metal rigs, is reviewed. The development of the present DFR cold-trapping system is outlined, and the operation of the additional pumped loops, which were required in order to control the reactor impurity levels, are described. Operation of the liquid metal rigs ancillary to the reactor project is discussed with particular reference to the control of impurity levels. (auth)

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-10-15

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

Impairment of exercise performance following cold water immersion is not attenuated after 7 days of cold acclimation.

Science.gov (United States)

Jones, Douglas M; Roelands, Bart; Bailey, Stephen P; Buono, Michael J; Meeusen, Romain

2018-03-19

It is well-documented that severe cold stress impairs exercise performance. Repeated immersion in cold water induces an insulative type of cold acclimation, wherein enhanced vasoconstriction leads to greater body heat retention, which may attenuate cold-induced exercise impairments. The purpose of this study, therefore, was to investigate changes in exercise performance during a 7-day insulative type of cold acclimation. Twelve healthy participants consisting of eight males and four females (mean ± SD age: 25.6 ± 5.2 years, height: 174.0 ± 8.9 cm, weight: 75.6 ± 13.1 kg) performed a 20 min self-paced cycling test in 23 °C, 40% humidity without prior cold exposure. Twenty-four hours later they began a 7-day cold acclimation protocol (daily 90 min immersion in 10 °C water). On days one, four, and seven of cold acclimation, participants completed the same cycling test. Measurements of work completed, core and skin temperatures, heart rate, skin blood flow, perceived exertion, and thermal sensation were measured during each cycling test. Successful insulative cold acclimation was observed. Work produced during the baseline cycling test (220 ± 70 kJ) was greater (p immersions (195 ± 58, 197 ± 60, and 194 ± 62 kJ) despite similar ratings of perceived exertion during each test, suggesting that cold exposure impaired cycling performance. This impairment, however, was not attenuated over the cold acclimation period. Results suggest that insulative cold acclimation does not attenuate impairments in exercise performance that were observed following acute cold water immersion.

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

DEFF Research Database (Denmark)

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

2016-01-01

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

Proposals for cold testing of the ITER TF coils

International Nuclear Information System (INIS)

Libeyre, P.; Ciazynski, D.; Dolgetta, N.; Duchateau, J.L.; Lyraud, C.; Kircher, F.; Schild, T.; Fietz, W.H.; Zahn, G.

2005-01-01

The ITER Toroidal Field (TF) magnet system will be made of 18 coils using Nb 3 Sn as superconducting material. These coils will operate at a maximum field of 11.8 T for a nominal current of 68 kA carried by a dual channel cable-in-conduit conductor cooled by a forced flow of supercritical helium at 4.5 K. In each coil, seven 760 m conductor lengths wound in double pancakes will be connected to each other by low resistance joints. As a final step of the reception tests, it is proposed to perform cold tests of these coils at liquid helium temperature after completion of their manufacture. The testing shall include high voltage tests to check the quality of the insulation, leak tests and pressure drop measurements of the hydraulic circuits as well as measurement of the joint resistances. Testing the coils up to nominal current is a discussed option, addressing on one hand measurement of the electrical performances in self field and on the other hand the mechanical behaviour of the coils. To perform these tests, a dedicated test facility has to be built, allowing possible simultaneous testing of two coils, assembled together in a twin coil configuration, similarly to their assembly in the torus. (authors)

Proposals for cold testing of the ITER TF coils

Energy Technology Data Exchange (ETDEWEB)

Libeyre, P.; Ciazynski, D.; Dolgetta, N.; Duchateau, J.L.; Lyraud, C. [Association Euratom/CEA Cadarache, Dept. de Recherches sur la Fusion Controlee (DRFC), 13 - Saint-Paul-lez-Durance (France); Kircher, F.; Schild, T. [CEA Saclay, Dept. d' Astrophysique, de Physique des Particules, de Physique Nucleaire et de l' Instrumentation Associee, 91- Gif sur Yvette (France); Fietz, W.H.; Zahn, G. [Association Euratom-Forschungszentrum Karlsruhe, Karlsruhe (Germany)

2005-07-01

The ITER Toroidal Field (TF) magnet system will be made of 18 coils using Nb{sub 3}Sn as superconducting material. These coils will operate at a maximum field of 11.8 T for a nominal current of 68 kA carried by a dual channel cable-in-conduit conductor cooled by a forced flow of supercritical helium at 4.5 K. In each coil, seven 760 m conductor lengths wound in double pancakes will be connected to each other by low resistance joints. As a final step of the reception tests, it is proposed to perform cold tests of these coils at liquid helium temperature after completion of their manufacture. The testing shall include high voltage tests to check the quality of the insulation, leak tests and pressure drop measurements of the hydraulic circuits as well as measurement of the joint resistances. Testing the coils up to nominal current is a discussed option, addressing on one hand measurement of the electrical performances in self field and on the other hand the mechanical behaviour of the coils. To perform these tests, a dedicated test facility has to be built, allowing possible simultaneous testing of two coils, assembled together in a twin coil configuration, similarly to their assembly in the torus. (authors)

Quantitative and Qualitative Responses to Topical Cold in Healthy Caucasians Show Variance between Individuals but High Test-Retest Reliability.

Directory of Open Access Journals (Sweden)

Penny Moss

Full Text Available Increased sensitivity to cold may be a predictor of persistent pain, but cold pain threshold is often viewed as unreliable. This study aimed to determine the within-subject reliability and between-subject variance of cold response, measured comprehensively as cold pain threshold plus pain intensity and sensation quality at threshold. A test-retest design was used over three sessions, one day apart. Response to cold was assessed at four sites (thenar eminence, volar forearm, tibialis anterior, plantar foot. Cold pain threshold was measured using a Medoc thermode and standard method of limits. Intensity of pain at threshold was rated using a 10cm visual analogue scale. Quality of sensation at threshold was quantified with indices calculated from subjects' selection of descriptors from a standard McGill Pain Questionnaire. Within-subject reliability for each measure was calculated with intra-class correlation coefficients and between-subject variance was evaluated as group coefficient of variation percentage (CV%. Gender and site comparisons were also made. Forty-five healthy adults participated: 20 male, 25 female; mean age 29 (range 18-56 years. All measures at all four test sites showed high within-subject reliability: cold pain thresholds r = 0.92-0.95; pain rating r = 0.93-0.97; McGill pain quality indices r = 0.87-0.85. In contrast, all measures showed wide between-subject variance (CV% between 51.4% and 92.5%. Upper limb sites were consistently more sensitive than lower limb sites, but equally reliable. Females showed elevated cold pain thresholds, although similar pain intensity and quality to males. Females were also more reliable and showed lower variance for all measures. Thus, although there was clear population variation, response to cold for healthy individuals was found to be highly reliable, whether measured as pain threshold, pain intensity or sensation quality. A comprehensive approach to cold response testing therefore may add

Quantitative and Qualitative Responses to Topical Cold in Healthy Caucasians Show Variance between Individuals but High Test-Retest Reliability.

Science.gov (United States)

Moss, Penny; Whitnell, Jasmine; Wright, Anthony

2016-01-01

Increased sensitivity to cold may be a predictor of persistent pain, but cold pain threshold is often viewed as unreliable. This study aimed to determine the within-subject reliability and between-subject variance of cold response, measured comprehensively as cold pain threshold plus pain intensity and sensation quality at threshold. A test-retest design was used over three sessions, one day apart. Response to cold was assessed at four sites (thenar eminence, volar forearm, tibialis anterior, plantar foot). Cold pain threshold was measured using a Medoc thermode and standard method of limits. Intensity of pain at threshold was rated using a 10cm visual analogue scale. Quality of sensation at threshold was quantified with indices calculated from subjects' selection of descriptors from a standard McGill Pain Questionnaire. Within-subject reliability for each measure was calculated with intra-class correlation coefficients and between-subject variance was evaluated as group coefficient of variation percentage (CV%). Gender and site comparisons were also made. Forty-five healthy adults participated: 20 male, 25 female; mean age 29 (range 18-56) years. All measures at all four test sites showed high within-subject reliability: cold pain thresholds r = 0.92-0.95; pain rating r = 0.93-0.97; McGill pain quality indices r = 0.87-0.85. In contrast, all measures showed wide between-subject variance (CV% between 51.4% and 92.5%). Upper limb sites were consistently more sensitive than lower limb sites, but equally reliable. Females showed elevated cold pain thresholds, although similar pain intensity and quality to males. Females were also more reliable and showed lower variance for all measures. Thus, although there was clear population variation, response to cold for healthy individuals was found to be highly reliable, whether measured as pain threshold, pain intensity or sensation quality. A comprehensive approach to cold response testing therefore may add validity and

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-09-01

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

BWR recirculation loop discharge line break LOCA tests with break areas of 50 and 100% assuming HPCS failure at ROSA-III test facility

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Tasaka, Kanji; Yonomoto, Taisuke; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Murata, Hideo; Shiba, Masayoshi; Iriko, Masanori.

1985-03-01

This report presents the experimental results of RUN 962 and RUN 963 in ROSA-III program, which are 50 and 100 % break LOCA tests at the BWR recirculation pump discharge line, respectively. The ROSA-III test facility simulates a volumetrically scaled (1/424) BWR system and has four half-length electrically heated fuel bundles, two active recirculation loops, three types of ECCSs and steam and feedwater systems. The experimental data of RUN 962 and RUN 963 were compared with those of RUN 961, a 200 % discharge line break test to study the break area effects on the transient thermal hydraulic phenomena. The least flow areas at the jet pump drive nozzles and recirculation pump discharge nozzle in the broken recirculation loop limitted the discharge flows from the pressure vessel and the depressurization rate in the 100 and 200 % break tests, whereas the least flow area at break nozzle limitted the depressurization rate in the 50 % break test. The highest PCT was observed in the 50 % break test among the three tests. (author)

Advancing viral RNA structure prediction: measuring the thermodynamics of pyrimidine-rich internal loops.

Science.gov (United States)

Phan, Andy; Mailey, Katherine; Saeki, Jessica; Gu, Xiaobo; Schroeder, Susan J

2017-05-01

Accurate thermodynamic parameters improve RNA structure predictions and thus accelerate understanding of RNA function and the identification of RNA drug binding sites. Many viral RNA structures, such as internal ribosome entry sites, have internal loops and bulges that are potential drug target sites. Current models used to predict internal loops are biased toward small, symmetric purine loops, and thus poorly predict asymmetric, pyrimidine-rich loops with >6 nucleotides (nt) that occur frequently in viral RNA. This article presents new thermodynamic data for 40 pyrimidine loops, many of which can form UU or protonated CC base pairs. Uracil and protonated cytosine base pairs stabilize asymmetric internal loops. Accurate prediction rules are presented that account for all thermodynamic measurements of RNA asymmetric internal loops. New loop initiation terms for loops with >6 nt are presented that do not follow previous assumptions that increasing asymmetry destabilizes loops. Since the last 2004 update, 126 new loops with asymmetry or sizes greater than 2 × 2 have been measured. These new measurements significantly deepen and diversify the thermodynamic database for RNA. These results will help better predict internal loops that are larger, pyrimidine-rich, and occur within viral structures such as internal ribosome entry sites. © 2017 Phan et al.; Published by Cold Spring Harbor Laboratory Press for the RNA Society.

Solar thermoelectric cooling using closed loop heat exchangers with macro channels

Science.gov (United States)

Atta, Raghied M.

2017-07-01

In this paper we describe the design, analysis and experimental study of an advanced coolant air conditioning system which cools or warms airflow using thermoelectric (TE) devices powered by solar cells. Both faces of the TE devices are directly connected to closed-loop highly efficient channels plates with macro scale channels and liquid-to-air heat exchangers. The hot side of the system consists of a pump that moves a coolant through the hot face of the TE modules, a radiator that drives heat away into the air, and a fan that transfer the heat over the radiator by forced convection. The cold side of the system consists also of a pump that moves coolant through the cold face of the TE modules, a radiator that drives cold away into the air, and a fan that blows cold air off the radiator. The system was integrated with solar panels, tested and its thermal performance was assessed. The experimental results verify the possibility of heating or cooling air using TE modules with a relatively high coefficient of performance (COP). The system was able to cool a closed space of 30 m3 by 14 °C below ambient within 90 min. The maximum COP of the whole system was 0.72 when the TE modules were running at 11.2 Å and 12 V. This improvement in the system COP over the air cooled heat sink is due to the improvement of the system heat exchange by means of channels plates.

Final report on the Controlled Cold Helium Spill Test in the LHC tunnel at CERN

CERN Document Server

Dufay-Chanat, L; Casas-Cubillos, J; Chorowski, M; Grabowski, M; Jedrusyna, A; Lindell, G; Nonis, M; Koettig, T; Vauthier, N; van Weelderen, R; Winkler, T

2015-01-01

The 27 km circumference LHC underground tunnel is a space in which the helium cooled LHC magnets are installed. The vacuum enclosures of the superconducting magnets are protected by over-pressure safety relief devices that open whenever cold helium escapes either from the magnet cold enclosure or from the helium supply headers, into this vacuum enclosure. A 3-m long no stay zone around these devices is defined based on scale model studies, protecting the personnel against cold burns or asphyxia caused by such a helium release event. Recently, several simulation studies have been carried out modelling the propagation of the helium/air mixture, resulting from the opening of such a safety device, along the tunnel. The released helium flows vary in the range between 1 kg/s and 0.1 kg/s. To validate these different simulation studies, real life mock-up tests have been performed inside the LHC tunnel, releasing helium flow rates of 1 kg/s, 0.3 kg/s and 0.1 kg/s. For each test, up to 1000 liters of liquid helium wer...

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

Directory of Open Access Journals (Sweden)

Xianwen Ning

2015-02-01

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

High-temperature helium-loop facility

International Nuclear Information System (INIS)

Tokarz, R.D.

1981-09-01

The high-temperature helium loop is a facility for materials testing in ultrapure helium gas at high temperatures. The closed loop system is capable of recirculating high-purity helium or helium with controlled impurities. The gas loop maximum operating conditions are as follows: 300 psi pressure, 500 lb/h flow rate, and 2100 0 F temperature. The two test sections can accept samples up to 3.5 in. diameter and 5 ft long. The gas loop is fully instrumented to continuously monitor all parameters of loop operation as well as helium impurities. The loop is fully automated to operate continuously and requires only a daily servicing by a qualified operator to replenish recorder charts and helium makeup gas. Because of its versatility and high degree of parameter control, the helium loop is applicable to many types of materials research. This report describes the test apparatus, operating parameters, peripheral systems, and instrumentation system. The experimental capabilities and test conand presents the results that have been obtained. The study has been conducted using a four-phase approach. The first phase develops the solution to the steady-state radon-diffusion equation in one-dimensieered barriers; disposal charge analysis; analysis of spent fuel policy implementation; spent f water. Field measurements and observations are reported for each site. Analytical data and field measurements are presented in tables and maps. Uranium concentrations in the sediments which were above detection limits ranged from 0.10 t 51.2 ppM. The mean of the logarithms of the uranium concentrations was 0.53. A group of high uranium concentrations occurs near the junctions of quadrangles AB, AC, BB, a 200 mK. In case 2), x-ray studies of isotopic phase separation in 3 He-- 4 He bcc solids were carried out by B. A. Fraass

Cold water recovery reduces anaerobic performance.

Science.gov (United States)

Crowe, M J; O'Connor, D; Rudd, D

2007-12-01

This study investigated the effects of cold water immersion on recovery from anaerobic cycling. Seventeen (13 male, 4 female) active subjects underwent a crossover, randomised design involving two testing sessions 2 - 6 d apart. Testing involved two 30-s maximal cycling efforts separated by a one-hour recovery period of 10-min cycling warm-down followed by either passive rest or 15-min cold water immersion (13 - 14 degrees C) with passive rest. Peak power, total work and postexercise blood lactate were significantly reduced following cold water immersion compared to the first exercise test and the control condition. These variables did not differ significantly between the control tests. Peak exercise heart rate was significantly lower after cold water immersion compared to the control. Time to peak power, rating of perceived exertion, and blood pH were not affected by cold water immersion compared to the control. Core temperature rose significantly (0.3 degrees C) during ice bath immersion but a similar increase also occurred in the control condition. Therefore, cold water immersion caused a significant decrease in sprint cycling performance with one-hour recovery between tests.

Testing experience with fast flux test facility

International Nuclear Information System (INIS)

Noordhoff, B.H.; McGough, C.B.; Nolan, J.E.

1975-01-01

Early FFTF project planning emphasized partial and full-scale testing of major reactor and plant prototype components under expected environmental conditions, excluding radiation fields. Confirmation of component performance during FFTF service was considered essential before actual FFTF startup, to provide increased assurance against FFTF startup delays or operational difficulties and downtime. Several new sodium facilities were constructed, and confirmation tests on the prototype components are now in progress. Test conditions and results to date are reported for the primary pump, intermediate heat exchanger, sodium-to-air dump heat exchanger, large and small sodium valves, purification cold trap, in-vessel handling machine, instrument tree, core restraint, control rod system, low-level flux monitor, closed loop ex-vessel machine, refueling equipment, and selected maintenance equipment. The significance and contribution of these tests to the FFTF and Liquid Metal Fast Breeder Reactor (LMFBR) program are summarized. (U.S.)

High-Temperature Structural Analysis of a Small-Scale PHE Prototype under the Test Condition of a Small-Scale Gas Loop

International Nuclear Information System (INIS)

Song, K.; Hong, S.; Park, H.

2012-01-01

A process heat exchanger (PHE) is a key component for transferring the high-temperature heat generated from a very high-temperature reactor (VHTR) to a chemical reaction for the massive production of hydrogen. The Korea Atomic Energy Research Institute has designed and assembled a small-scale nitrogen gas loop for a performance test on VHTR components and has manufactured a small-scale PHE prototype made of Hastelloy-X alloy. A performance test on the PHE prototype is underway in the gas loop, where different kinds of pipelines connecting to the PHE prototype are tested for reducing the thermal stress under the expansion of the PHE prototype. In this study, to evaluate the high-temperature structural integrity of the PHE prototype under the test condition of the gas loop, a realistic and effective boundary condition imposing the stiffness of the pipelines connected to the PHE prototype was suggested. An equivalent spring stiffness to reduce the thermal stress under the expansion of the PHE prototype was computed from the bending deformation and expansion of the pipelines connected to the PHE. A structural analysis on the PHE prototype was also carried out by imposing the suggested boundary condition. As a result of the analysis, the structural integrity of the PHE prototype seems to be maintained under the test condition of the gas loop.

Controlled Cold Helium Spill Test in the LHC Tunnel at CERN

Science.gov (United States)

Koettig, T.; Casas-Cubillos, J.; Chorowski, M.; Dufay-Chanat, L.; Grabowski, M.; Jedrusyna, A.; Lindell, G.; Nonis, M.; Vauthier, N.; van Weelderen, R.; Winkler, T.; Bremer, J.

The helium cooled magnets of the LHC particle accelerator are installed in a confined space, formed by a 27 km circumference 3.8 m diameter underground tunnel. The vacuum enclosures of the superconducting LHC magnets are protected by a lift plate against excessive overpressure created by eventual leaks from the magnet helium bath, or from the helium supply headers. A three-meter long no stay zone has been defined centered to these plates, based on earlier scale model studies, to protect the personnel against the consequences of an eventual opening of such a lift plate. More recently several simulation studies have been carried out modelling the propagation of the resulting helium/air mixture along the tunnel in case of such a cold helium release at a rate in the range of 1 kg/s. To validate the different scale models and simulation studies, real life mock-up tests have been performed in the LHC, releasing about 1000 liter of liquid helium under standard operational tunnel conditions. Data recorded during these tests include oxygen level, temperature and flow speed as well as video recordings, taken up- and downstream of the spill point (-100 m to +200 m) with respect to the ventilation direction in the LHC tunnel. The experimental set-up and measurement results are presented. Generic effects found during the tests will be discussed to allow the transposal to possible cold helium release cases in similar facilities.

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

Thermal hydraulic tests of a liquid hydrogen cold neutron source. NISTIR 5026

International Nuclear Information System (INIS)

Siegwarth, J.D.; Olson, D.A.; Lewis, M.A.; Rowe, J.M.; Williams, R.E.; Kopetka, P.

1995-01-01

Liquid hydrogen cold neutron source designed at NBSR contains neutron moderator chamber. The NIST-B electrically heated glass moderator chamber used to test the NBSR chamber testing showed the following results: Stable operation possible up to at least 2200 watts with two-phase flow; LH 2 mass quickly reaches new, stable value after heat load change; Void fraction well below 20 at anticipated power and pressure; Restart of H 2 flow verified after extending supply line; Visual inspection showed no dryout or unexpected voids

Annealing dislocation loops in OKh16N15M3T steel implanted by helium

International Nuclear Information System (INIS)

Utkelbaev, B.D.; Reutov, V.F.; Zhdan, G.T.

1993-01-01

With the use of electron microscopy a study was made into the influence of preliminary thermomechanical treatment on the process of dislocation loop development in austenitic stainless steel type OKh16N15M3T with helium on annealing. Preliminary treatment was shown to prevent dislocation loop formation to a greater or lesser extent. Preliminary 'cold' working and thermal ageing of the material are the most effective ways to suppress radiation defect formation when annealing helium implanted steel

Liquid lithium loop system to solve challenging technology issues for fusion power plant

Science.gov (United States)

Ono, M.; Majeski, R.; Jaworski, M. A.; Hirooka, Y.; Kaita, R.; Gray, T. K.; Maingi, R.; Skinner, C. H.; Christenson, M.; Ruzic, D. N.

2017-11-01

Steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor concept and its variant, the active liquid lithium divertor concept, taking advantage of the enhanced or non-coronal Li radiation in relatively poorly confined divertor plasmas. To maintain the LL purity in a 1 GW-electric class fusion power plant, a closed LL loop system with a modest circulating capacity of ~1 l s-1 is envisioned. We examined two key technology issues: (1) dust or solid particle removal and (2) real time recovery of tritium from LL while keeping the tritium inventory level to an acceptable level. By running the LL-loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to the outside where the dust/impurities can be removed by relatively simple dust filter, cold trap and/or centrifugal separation systems. With ~1 l s-1 LL flow, even a small 0.1% dust content by weight (or 0.5 g s-1) suggests that the LL-loop could carry away nearly 16 tons of dust per year. In a 1 GW-electric (or ~3 GW fusion power) fusion power plant, about 0.5 g s-1 of tritium is needed to maintain the fusion fuel cycle

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Handling Ibuprofen increases pain tolerance and decreases perceived pain intensity in a cold pressor test.

Directory of Open Access Journals (Sweden)

Abraham M Rutchick

Full Text Available Pain contributes to health care costs, missed work and school, and lower quality of life. Extant research on psychological interventions for pain has focused primarily on developing skills that individuals can apply to manage their pain. Rather than examining internal factors that influence pain tolerance (e.g., pain management skills, the current work examines factors external to an individual that can increase pain tolerance. Specifically, the current study examined the nonconscious influence of exposure to meaningful objects on the perception of pain. Participants (N = 54 completed a cold pressor test, examined either ibuprofen or a control object, then completed another cold pressor test. In the second test, participants who previously examined ibuprofen reported experiencing less intense pain and tolerated immersion longer (relative to baseline than those who examined the control object. Theoretical and applied implications of these findings are discussed.

Experiment data report for semiscale Mod-1 Tests S-01-4 and S-01-4A (isothermal blowdown with core resistance simulator)

International Nuclear Information System (INIS)

Zender, S.N.; Jensen, M.F.; Sackett, K.E.

1975-03-01

Recorded test data are presented for Tests S-01-4 and S-01-4A of the semiscale Mod-1 isothermal blowdown test series. These tests are among several semiscale Mod-1 experiments which are counterparts of the planned Loss-of-Fluid Test (LOFT) nonnuclear experiments. System hardware is representative of LOFT design based on volumetric scaling methods, and initial conditions duplicate those identified for the LOFT nonnuclear tests. Tests S-01-4 and S-01-4A employed an intact loop resistance that was similar to that of Test S-01-3 and low relative to that of Test S-01-2. An orificed structure was used in the pressure vessel to simulate the LOFT core simulator. The tests were initiated at initial isothermal conditions of about 2250 psig and 540 0 F by a simulated offset shear of the cold-leg broken-loop piping. During system depressurization, coolant was injected into the cold leg of the intact loop to provide data on the effects of emergency core cooling on system response. Following the blowdown portion of Test S-01-4, coolant spray was introduced into the pressure suppression tank to determine the response of the pressure suppression system. The uninterpreted data are presented. 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. (U.S.)

DETECTING NANOFLARE HEATING EVENTS IN SUBARCSECOND INTER-MOSS LOOPS USING Hi-C

International Nuclear Information System (INIS)

Winebarger, Amy R.; Moore, Ronald; Cirtain, Jonathan; Walsh, Robert W.; De Pontieu, Bart; Title, Alan; Hansteen, Viggo; Golub, Leon; Korreck, Kelly; Weber, Mark; Kobayashi, Ken; DeForest, Craig; Kuzin, Sergey

2013-01-01

The High-resolution Coronal Imager (Hi-C) flew aboard a NASA sounding rocket on 2012 July 11 and captured roughly 345 s of high-spatial and temporal resolution images of the solar corona in a narrowband 193 Å channel. In this paper, we analyze a set of rapidly evolving loops that appear in an inter-moss region. We select six loops that both appear in and fade out of the Hi-C images during the short flight. From the Hi-C data, we determine the size and lifetimes of the loops and characterize whether these loops appear simultaneously along their length or first appear at one footpoint before appearing at the other. Using co-aligned, co-temporal data from multiple channels of the Atmospheric Imaging Assembly on the Solar Dynamics Observatory, we determine the temperature and density of the loops. We find the loops consist of cool (∼10 5 K), dense (∼10 10 cm –3 ) plasma. Their required thermal energy and their observed evolution suggest they result from impulsive heating similar in magnitude to nanoflares. Comparisons with advanced numerical simulations indicate that such dense, cold and short-lived loops are a natural consequence of impulsive magnetic energy release by reconnection of braided magnetic field at low heights in the solar atmosphere.

Design, fabrication, commissioning, and testing of a 250 g/s, 2-K helium cold compressor system

International Nuclear Information System (INIS)

V. Ganni; D. M. Arenius; B. S. Bevins; W. C. Chronis; J. D. Creel; J. D. Wilson Jr.

2002-01-01

In June 1999 the Thomas Jefferson National Accelerator Facility (TJNAF) Cryogenic Systems Group had completed the design, fabrication, and commissioning of a cold compressor system capable of pumping 250 g/s of 2-K helium vapor to a pressure above 1 bar. The 2-K cold box consists of five stages of centrifugal variable speed compressors with LN2 cooled drive motors and magnetic bearings, a plate fin heat exchanger, and an LN2 shield system. The new 2-K cold box (referred to as the SCN) was built as a redundant system to an existing four stage cold compressor SCM cold box that was commissioned in May 1994. The SCN has been in continuous service supporting the facility experiments since commissioning. This system has achieved a significant improvement in the total 2-K refrigeration system capacity and stability and has substantially increased the operating envelope both in cold compressor flow and operating pressure range. This paper describes the cold box configuration and the experience s in the design, fabrication, commissioning and performance evaluation. The capacity of the system for various operating pressures (0.040 to 0.025 bar at the load corresponding to a total compressor pressure ratio of 28 to 54) is presented. An effort is made to characterize the components and their operating data over the tested range. This includes the return side pressure drop in the distribution system, the heat exchanger, and the cold compressor characteristics. The system design parameters and their effects on performance are outlined

Tritium Management Loop Design Status

Energy Technology Data Exchange (ETDEWEB)

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

2017-12-01

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

A two-loop test of M(atrix) theory

International Nuclear Information System (INIS)

Becker, K.

1997-01-01

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

Commissioning of the Opal reactor cold neutron source

International Nuclear Information System (INIS)

Thiering, R.; Lu, W.; Ullah, R.

2006-01-01

Full text: At OPAL, Australia's first cold neutron facility will form an essential part of the reactor's research programs. Fast neutrons, born in the core of a reactor, interact with a cryogenic material, in this case liquid deuterium, to give them very low energies ( 1 0 m eV). A cold neutron flux of 1.4 1 0 E 1 4 n /cm 2/ s is expected, with a peak in the energy spectrum at 4.2m eV. The cold neutron source reached cryogenic conditions for the first time in late 2005. The cold neutron source operates with a sub-cooled liquid Deuterium moderator at 24 K. The moderator chamber, which contains the deuterium, has been constructed from AlMg 5. The thermosiphon and moderator chamber are cooled by helium gas, in a natural convection thermosiphon loop. The helium refrigeration system utilises the Brayton cycle, and is fully insulated within a high vacuum environment. Despite the proximity of the cold neutron source to the reactor core, it has been considered as effectively separate to the reactor system, due to the design of its special vacuum containment vessel. As OPAL is a multipurpose research reactor, used for beam research as well as radiopharmaceutical production and industrial irradiations, the cold neutron source has been designed with a stand-by mode, to maximise production. The stand-by mode is a warm operating mode using only gaseous deuterium at ambient temperatures (∼ 3 00 K ), allowing for continued reactor operations whilst parts of the cold source are unavailable or in maintenance. This is the first time such a stand-by feature has been incorporated into a cold source facility

Temporary core liquid level depression during cold-leg small-break LOCA effect of break size and power level

International Nuclear Information System (INIS)

Koizumi, Y.; Kumamaru, H.; Mimura, Y.; Kukita, Y.; Tasaka, K.

1989-01-01

Cold-leg small break LOCA experiments (0.5-10% break) were conducted at the large scale test facility (LSTF), a volumetrically-scaled (1/48) simulator of a PWR, of the ROSA-IV Program. When a break area was less than 2.5% of the scaled cold-leg flow area, the core liquid level was temporarily further depressed to the bottom elevation of the crossover leg during the loop seal clearing early in the transient only by the manometric pressure balance since no coolant remained in the upper portion of the primary system. When the break size was larger than 5%, the core liquid level was temporarily further depressed lower than the bottom elevation of the crossover leg during the loop seal clearing since coolant remained at the upper portion of the primary system; the steam generator (SG) U-tube upflow side and the SG inlet plenum, due to counter current flow limiting by updrafting steam while the coolant drained. The amount of coolant trapped there was dependent on the vapor velocity (core power); the larger the core power, the lower the minimum core liquid level. The RELAP5/MOD2 code reasonable predicted phenomena observed in the experiments. (orig./DG)

Adaptable Single Active Loop Thermal Control System (TCS) for Future Space Missions

Science.gov (United States)

Mudawar, Issam; Lee, Seunghyun; Hasan, Mohammad

2015-01-01

This presentation will examine the development of a thermal control system (TCS) for future space missions utilizing a single active cooling loop. The system architecture enables the TCS to be reconfigured during the various mission phases to respond, not only to varying heat load, but to heat rejection temperature as well. The system will consist of an accumulator, pump, cold plates (evaporators), condenser radiator, and compressor, in addition to control, bypass and throttling valves. For cold environments, the heat will be rejected by radiation, during which the compressor will be bypassed, reducing the system to a simple pumped loop that, depending on heat load, can operate in either a single-phase liquid mode or two-phase mode. For warmer environments, the pump will be bypassed, enabling the TCS to operate as a heat pump. This presentation will focus on recent findings concerning two-phase flow regimes, pressure drop, and heat transfer coefficient trends in the cabin and avionics micro-channel heat exchangers when using the heat pump mode. Also discussed will be practical implications of using micro-channel evaporators for the heat pump.

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

Energy Technology Data Exchange (ETDEWEB)

Bischi, Aldo

2012-05-15

can be used for the hot rig design debugging and it is at the same time the hydrodynamic small scale model of a ten times larger industrial application. The adopted scaling strategy and design brought to the construction of one of the world biggest and more complex fluidized bed cold flow model reactor systems. The air and fuel reactor have a height of 5 m and a diameter of respectively 0.230 and 0.144 m. The selected particles are fine and heavy being classifiable as high density Geldart A; there is almost no published literature regarding those particles utilization in circulating fluidized beds. Extensive test campaigns have been performed to hydrodynamically validate the proposed designs. It was possible to understand and evaluate the operational window, the sensitivity to the input parameters and the key design details performance. Control strategies were qualitatively developed. The presented double loop architecture design showed good stability and flexibility at the same time, so that can also suit the requirements of other chemical processes based on two complementary reactions taking place simultaneously and continuously.(Author)

An empirical model for friction in cold forging

DEFF Research Database (Denmark)

Bay, Niels; Eriksen, Morten; Tan, Xincai

2002-01-01

With a system of simulative tribology tests for cold forging the friction stress for aluminum, steel and stainless steel provided with typical lubricants for cold forging has been determined for varying normal pressure, surface expansion, sliding length and tool/work piece interface temperature...... of normal pressure and tool/work piece interface temperature. The model is verified by process testing measuring friction at varying reductions in cold forward rod extrusion. KEY WORDS: empirical friction model, cold forging, simulative friction tests....

Heat Load Sharing in a Capillary Pumped Loop with Multiple Evaporators and Multiple Condensers

Science.gov (United States)

Ku, Jentung

2005-01-01

This paper describes the heat load sharing function among multiple parallel evaporators in a capillary pumped loop (CPL). In the normal mode of operation, the evaporators cool the instruments by absorbing the waste heat. When an instruments is turned off, the attached evaporator can keep it warm by receiving heat from other evaporators serving the operating instruments. This is referred to as heat load sharing. A theoretical basis of heat load sharing is given first. The fact that the wicks in the powered evaporators will develop capillary pressure to force the generated vapor to flow to cold locations where the pressure is lower leads to the conclusion that heat load sharing is an inherent function of a CPL with multiple evaporators. Heat load sharing has been verified with many CPLs in ground tests. Experimental results of the Capillary Pumped Loop 3 (CAPL 3) Flight Experiment are presented in this paper. Factors that affect the amount of heat being shared are discussed. Some constraints of heat load sharing are also addressed.

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

Science.gov (United States)

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

2016-01-01

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

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

International Nuclear Information System (INIS)

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

1984-01-01

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

Experiment data report for semiscale Mod-1 test S-02-3 (blowdown heat transfer test)

International Nuclear Information System (INIS)

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

1975-09-01

Recorded test data are presented for Test S-02-3 of the Semiscale Mod-1 blowdown heat transfer test series. Test S-02-3 was conducted from an initial cold leg fluid temperature of 544 0 F and an initial pressure of 2,263 psig. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization transient with a moderately heated core (75 percent design power level). An electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core. System flow was also set at the 75 percent design level to achieve full core temperature differential. 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 1.2 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 (DNB) occurs. Blowdown to the pressure suppression system was accomplished without simulated emergency core coolant injection or pressure suppression system coolant spray

Pumped Fluid Loop Heat Rejection and Recovery Systems for Thermal Control of the Mars Science Laboratory

Science.gov (United States)

Bhandari, Pradeep; Birur, Gajanana; Prina, Mauro; Ramirez, Brenda; Paris, Anthony; Novak, Keith; Pauken, Michael

2006-01-01

This viewgraph presentation reviews the heat rejection and heat recovery system for thermal control of the Mars Science Laboratory (MSL). The MSL mission will use mechanically pumped fluid loop based architecture for thermal control of the spacecraft and rover. The architecture is designed to harness waste heat from an Multi Mission Radioisotope Thermo-electric Generator (MMRTG) during Mars surface operations for thermal control during cold conditions and also reject heat during the cruise aspect of the mission. There are several test that are being conducted that will insure the safety of this concept. This architecture can be used during any future interplanetary missions utilizing radioisotope power systems for power generation.

DETECTING NANOFLARE HEATING EVENTS IN SUBARCSECOND INTER-MOSS LOOPS USING Hi-C

Energy Technology Data Exchange (ETDEWEB)

Winebarger, Amy R.; Moore, Ronald; Cirtain, Jonathan [NASA Marshall Space Flight Center, ZP 13, Huntsville, AL 35812 (United States); Walsh, Robert W. [University of Central Lancashire, Preston, Lancashire PR1 2HE (United Kingdom); De Pontieu, Bart; Title, Alan [Lockheed Martin Solar and Astrophysics Lab, 3251 Hanover St., Org. A0215, Bldg. 252, Palo Alto, CA 94304 (United States); Hansteen, Viggo [Institute of Theoretical Astrophysics, University of Oslo, P.O. Box 1029 Blindern, N-0315 Oslo (Norway); Golub, Leon; Korreck, Kelly; Weber, Mark [Harvard-Smithsonian Center for Astrophysics, 60 Garden St., Cambridge, MA 02138 (United States); Kobayashi, Ken [Center for Space Plasma and Aeronomic Research, University of Alabama in Huntsville, 320 Sparkman Dr, Huntsville, AL 35805 (United States); DeForest, Craig [Southwest Research Institute, 1050 Walnut Street, Suite 300, Boulder, CO 80302 (United States); Kuzin, Sergey, E-mail: amy.r.winebarger@nasa.gov [P.N. Lebedev Physical institute of the Russian Academy of Sciences, Leninskii prospekt 53 119991, Moscow (Russian Federation)

2013-07-01

The High-resolution Coronal Imager (Hi-C) flew aboard a NASA sounding rocket on 2012 July 11 and captured roughly 345 s of high-spatial and temporal resolution images of the solar corona in a narrowband 193 A channel. In this paper, we analyze a set of rapidly evolving loops that appear in an inter-moss region. We select six loops that both appear in and fade out of the Hi-C images during the short flight. From the Hi-C data, we determine the size and lifetimes of the loops and characterize whether these loops appear simultaneously along their length or first appear at one footpoint before appearing at the other. Using co-aligned, co-temporal data from multiple channels of the Atmospheric Imaging Assembly on the Solar Dynamics Observatory, we determine the temperature and density of the loops. We find the loops consist of cool ({approx}10{sup 5} K), dense ({approx}10{sup 10} cm{sup -3}) plasma. Their required thermal energy and their observed evolution suggest they result from impulsive heating similar in magnitude to nanoflares. Comparisons with advanced numerical simulations indicate that such dense, cold and short-lived loops are a natural consequence of impulsive magnetic energy release by reconnection of braided magnetic field at low heights in the solar atmosphere.

Long-life fatigue test results for two nickel-base structural alloys

International Nuclear Information System (INIS)

Mowbray, D.F.; Giaquinto, E.V.; Mehringer, F.J.

1978-11-01

The results are reported of fatigue tests on two nickel--base alloys, hot-cold-worked and stress-relieved nickel--chrome--iron Alloy 600 and mill-annealed nickel--chrome--moly--iron Alloy 625 in which S-N data were obtained in the life range of 10 6 to 10 10 cycles. The tests were conducted in air at 600 0 F, in the reversed membrane loading mode, at a frequency of approx. 1850 Hz. An electromagnetic, closed loop servo-controlled machine was built to perform the tests. A description of the machine is given

LoopIng: a template-based tool for predicting the structure of protein loops.

KAUST Repository

Messih, Mario Abdel

2015-08-06

Predicting the structure of protein loops is very challenging, mainly because they are not necessarily subject to strong evolutionary pressure. This implies that, unlike the rest of the protein, standard homology modeling techniques are not very effective in modeling their structure. However, loops are often involved in protein function, hence inferring their structure is important for predicting protein structure as well as function.We describe a method, LoopIng, based on the Random Forest automated learning technique, which, given a target loop, selects a structural template for it from a database of loop candidates. Compared to the most recently available methods, LoopIng is able to achieve similar accuracy for short loops (4-10 residues) and significant enhancements for long loops (11-20 residues). The quality of the predictions is robust to errors that unavoidably affect the stem regions when these are modeled. The method returns a confidence score for the predicted template loops and has the advantage of being very fast (on average: 1 min/loop).www.biocomputing.it/loopinganna.tramontano@uniroma1.itSupplementary data are available at Bioinformatics online.

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

International Nuclear Information System (INIS)

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

2004-12-01

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

Backward can extrusion with conical,rotating punch as a cold forging tribology test

DEFF Research Database (Denmark)

Ceron, Ermanno; Bay, Niels; Tetsuo, A.

2011-01-01

with an analysis of the sliding velocity distribution along the punch nose. The latter is determined by FE analysis of the test. Results show friction stress for unalloyed low C-steel provided with different types of lubricants, e.g. phosphate coating plus soap, phosphate coating plus MoS2 and single bathe...... lubrication with PULS. The new test is so severe, that it is possible to break down the best lubrication systems for cold forging, such as phosphate coating plus soap and MoS2....

Cold flow model study of an oxyfuel combustion pilot plant

Energy Technology Data Exchange (ETDEWEB)

Guio-Perez, D.C.; Tondl, G.; Hoeltl, W.; Proell, T.; Hofbauer, H. [Vienna University of Technology, Institute of Chemical Engineering, Vienna (Austria)

2011-12-15

The fluid-dynamic behavior of a circulating fluidized bed pilot plant for oxyfuel combustion was studied in a cold flow model, down-scaled using Glicksman's criteria. Pressures along the unit and the global circulation rate were used for characterization. The analysis of five operating parameters and their influence on the system was carried out; namely, total solids inventory and the air velocity of primary, secondary, loop seal and support fluidizations. The cold flow model study shows that the reactor design allows stable operation at a wide range of fluidization rates, with results that agree well with previous observations described in the literature. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Preparation of HEC Serial Modules for Beam and Cold Tests at CERN in 2000

CERN Multimedia

Fischer, A.

2001-01-01

Photo 1 - Three mated HEC-1 modules in the clean room, prepared for beam tests. Photo 2 - Three mated HEC-1 modules in the clean room, prepared for beam tests. Photo 3 - Three mated HEC-2 modules in the clean room, prepared for beam tests. Photo 4 - Three mated HEC-2 modules in the clean room, prepared for beam tests. Photo 5 - Close-up of the inner-connecting bars of three mated HEC-2 modules, prepared for beam tests. Photo 6 - Mechanical acceptance tests for HEC serial modules at CERN. Photo 7 - Mechanical acceptance tests for HEC serial modules at CERN. Photo 8 - Four HEC-2 modules in the cryostat, prepared for cold tests.

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

Energy Technology Data Exchange (ETDEWEB)

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

2005-12-15

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

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

Medium-term experiences with in-situ gamma-spectrometry of the primary loop transport processes at Paks NPP

International Nuclear Information System (INIS)

Raics, P.; Sztaricskai, T.; Szabo, J.; Szegedi, S.

2001-01-01

Surface activity of 15 corrosion/erosion and fission products was determined by in-situ gamma-spectrometry for 2-2 locations on the hot and cold legs of each loop, respectively. Gamma-dosimetry in the assay points was performed. Activity profiles of ion exchange columns were analyzed. Combined measurements along the steam generators completed the characterization of the primary circuits. Most of this technique was regularly included into all maintenance periods. Data evaluation was performed for the surface contaminations as well as coolant activities and reactor operation features for years 1985-2001. Trends and tendencies were investigated in the time behavior of the specific activities. Asymmetry in the surface contamination at the primary loop points, cold-leg activity inversion, water chemistry effects, isotope selectivity were observed. Correlations in different parameters have been calculated and analyzed. (R.P.)

Structural, magnetic and electrical transport properties in cold-drawn thin Fe-rich wires

International Nuclear Information System (INIS)

Garcia, C.; Chizhik, A.; Val, J.J. del; Zhukov, A.; Blanco, J.M.; Gonzalez, J.

2005-01-01

Microstructural (X-ray diffraction), magnetic properties (hysteresis loop), electrical resistivity, magneto-impedance and stress impedance effects have been investigated in cold-drawn Fe 77.5 B 15 Si 7.5 amorphous wire. Initial amorphous wire (obtained by the in-rotating-water technique) with diameter of 125 μm was submitted to cold-drawn process decreasing the diameter to 50 μm. Such cold-drawn wire was treated by current annealing (currents of 190, 210, 220 and 230 mA during times between 1 and 45 min) for tailoring the magnetic and electrical transport properties. A qualitative analysis of the magnetoimpedance and stress impedance effects is given by considering the influence of the magnetoelastic anisotropy and frequency of the AC driving electrical current on the circular permeability

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

Science.gov (United States)

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

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

Mass transfer of steels for FBR in sodium loop

International Nuclear Information System (INIS)

Susukida, Hiroshi; Yonezawa, Toshio; Ueda, Mitsuo; Imazu, Takayuki; Kiyokawa, Teruyuki.

1976-06-01

In order to grasp quantitatively the corrosion and mass transfer of steels for FBR in sodium loop and to establish their allowable stress value and corrosion rate, a special sodium loop for material testing was designed and fabricated and the steels were given 3010 hours exposing test in the sodium loop. This paper gives the outline of the sodium loop and the results of the test. (1) Carburization and a slight increase in weight were observed in the specimens of type 304 stainless steel exposed in the sodium loop for 3010 hours, while decarburization was observed in the specimens of 2 1/4 Cr-1 Mo steel. It is considered that these phenomena were caused by the downstream factor of the sodium loop. (2) A remarkable decrease of Charpy absorbed energy was observed in the specimens of type 304 stainless steel exposed in the sodium loop. It is considered that this resulted from the weakening of the grain boundary due to heat history and mass transfer. (3) The specimens exposed in the sodium loop must be washed by ultrasonic waves in a water bath after washing in alcohol. (auth.)

Evaluation report on CCTF CORE-I REFLOOD TEST Cl-4 (Run 13) and Cl-15 (Run 24)

International Nuclear Information System (INIS)

Sudoh, Takashi; Murao, Yoshio.

1983-08-01

The tests Cl-4 and Cl-15 were performed with the Cylindrical Core Test Facility (CCTF) to investigate the effects of the depressurization process to simulate the refill phase, and the effects of the nitrogen to be injected after the end of the accumulator injection on the thermo-hydraulic behavior in the core and primary loop system during refill and reflood phases. In these tests, after the lower plenum was filled to 0.9m level with saturated water at 0.6 MPa, the accumulator water was injected into three intact cold legs in the depressurization period from 0.6 MPa to 0.2 MPa. The water in the lower plenum voided during the depressurization and the significant steam condensation occurred in or near the intact cold legs. The condensation caused high steam flow rate in the intact loops and the lower plenum flashing resulted in suppressed core water accumulation. The slightly lower core heat transfer coefficient due to the less core water caused the higher turnaround temperature and the longer quench time than those of the normal reflood test without the depressurization process. The nitrogen injection followed the accumulator injection was allowed in the test Cl-15. However, significant effects of the nitrogen injection was not observed. (author)

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

Directory of Open Access Journals (Sweden)

Ma Tiancai

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-03-01

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

Acid Pit Stabilization Project (Volume 1 - Cold Testing) and (Volume 2 - Hot Testing)

International Nuclear Information System (INIS)

Loomis, G. G.; Zdinak, A. P.; Ewanic, M. A.; Jessmore, J. J.

1998-01-01

During the summer and fall of Fiscal Year 1997, a Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) Treatability Study was performed at the Idaho National Engineering and Environmental Laboratory. The study involved subsurface stabilization of a mixed waste contaminated soil site called the Acid Pit. This study represents the culmination of a successful technology development effort that spanned Fiscal Years 1994-1996. Research and development of the in situ grout stabilization technique was conducted. Hardware and implementation techniques are currently documented in a patent pending with the United States Patent and Trademark Office. The stabilization technique involved using jet grouting of an innovative grouting material to form a monolith out of the contamination zone. The monolith simultaneously provides a barrier to further contaminant migration and closes voids in the soil structure against further subsidence. This is accomplished by chemical incorporation of contaminants into less soluble species and achieving a general reduction in hydraulic conductivity within the monolith. The grout used for this study was TECT-HG, a relatively dense iron oxide-based cementitious grout. The treatability study involved cold testing followed by in situ stabilization of the Acid Pit. Volume 1 of this report discusses cold testing, performed as part of a ''Management Readiness Assessment'' in preparation for going hot. Volume 2 discusses the results of the hot Acid Pit Stabilization phase of this project. Drilling equipment was specifically rigged to reduce the spread of contamination, and all grouting was performed under a concrete block containing void space to absorb any grout returns. Data evaluation included examination of implementability of the grouting process and an evaluation of the contaminant spread during grouting. Following curing of the stabilized pit, cores were obtained and evaluated for toxicity characteristic leach ing

Cold-Leg Small Break LOCA Analysis of APR1400 Plant Using a SPACE/sEM Code

International Nuclear Information System (INIS)

Lim, Sang Gyu; Lee, Suk Ho; Yu, Keuk Jong; Kim, Han Gon; Lee, Jae Yong

2013-01-01

The Small Break Loss-of-Coolant Accident (SBLOCA) evaluation methodology (EM) for APR1400, called sEM, is now being developed using SPACE code. SPACE/sEM is to set up a conservative evaluation methodology in accordance with appendix K of 10 CFR 50. Major required and acceptable features of the evaluation models are described as below. - Fission product decay : 1.2 times of ANS97 decay curve - Critical flow model : Henry-Fauske Moody two phase critical flow model - Metal-Water reaction model : Baker-Just equation - Critical Heat Flux (CHF) : B and W, Barnett and Modified Barnett correlation - Post-CHF : Groeneveld 5.7 film boiling correlation A series of test matrix is established to validate SPACE/sEM code in terms of major SBLOCA phenomena, e.g. core level swelling and boiling, core heat transfer, critical flow, loop seal clearance and their integrated effects. The separated effect tests (SETs) and integrated effect tests (IETs) are successfully performed and these results shows that SPACE/sEM code has a conservatism comparing with experimental data. Finally, plant calculations of SBLOCA for APR1400 are conducted as described below. - Break location sensitivity : DVI line, hot-leg, cold-leg, pump suction leg. - Break size spectrum : 0.4ft 2 ∼0.02ft 2 (DVI) 0.5ft 2 ∼0.02ft 2 (hot-leg, cold-leg, pump suction leg) This paper deals with break size spectrum analysis of cold-leg break accidents. Based on the calculation results, emergency core cooling system (ECCS) performances of APR1400 and typical SBLOCA phenomena can be evaluated. Cold-leg SBLOCA analysis for APR1400 is performed using SPACE/sEM code under harsh environment condition. SPACE/sEM code shows the typical SBLOCA behaviors and it is reasonably predicted. Although SPACE/sEM code has conservative models and correlations based on appendix K of 10 CFR 50, PCT does not exceed the requirement (1477 K). It is concluded that ECCS in APR1400 has a sufficient performance in cold-leg SBLOCA

Cold-Leg Small Break LOCA Analysis of APR1400 Plant Using a SPACE/sEM Code

Energy Technology Data Exchange (ETDEWEB)

Lim, Sang Gyu; Lee, Suk Ho; Yu, Keuk Jong; Kim, Han Gon; Lee, Jae Yong [Central Research Institute, KHNP, Ltd., Daejeon (Korea, Republic of)

2013-10-15

The Small Break Loss-of-Coolant Accident (SBLOCA) evaluation methodology (EM) for APR1400, called sEM, is now being developed using SPACE code. SPACE/sEM is to set up a conservative evaluation methodology in accordance with appendix K of 10 CFR 50. Major required and acceptable features of the evaluation models are described as below. - Fission product decay : 1.2 times of ANS97 decay curve - Critical flow model : Henry-Fauske Moody two phase critical flow model - Metal-Water reaction model : Baker-Just equation - Critical Heat Flux (CHF) : B and W, Barnett and Modified Barnett correlation - Post-CHF : Groeneveld 5.7 film boiling correlation A series of test matrix is established to validate SPACE/sEM code in terms of major SBLOCA phenomena, e.g. core level swelling and boiling, core heat transfer, critical flow, loop seal clearance and their integrated effects. The separated effect tests (SETs) and integrated effect tests (IETs) are successfully performed and these results shows that SPACE/sEM code has a conservatism comparing with experimental data. Finally, plant calculations of SBLOCA for APR1400 are conducted as described below. - Break location sensitivity : DVI line, hot-leg, cold-leg, pump suction leg. - Break size spectrum : 0.4ft{sup 2}∼0.02ft{sup 2}(DVI) 0.5ft{sup 2}∼0.02ft{sup 2}(hot-leg, cold-leg, pump suction leg) This paper deals with break size spectrum analysis of cold-leg break accidents. Based on the calculation results, emergency core cooling system (ECCS) performances of APR1400 and typical SBLOCA phenomena can be evaluated. Cold-leg SBLOCA analysis for APR1400 is performed using SPACE/sEM code under harsh environment condition. SPACE/sEM code shows the typical SBLOCA behaviors and it is reasonably predicted. Although SPACE/sEM code has conservative models and correlations based on appendix K of 10 CFR 50, PCT does not exceed the requirement (1477 K). It is concluded that ECCS in APR1400 has a sufficient performance in cold-leg SBLOCA.

Hot or cold? Is monothermal caloric testing useful and cost-effective?

Science.gov (United States)

Bush, Matthew L; Bingcang, Christopher M; Chang, Edward T; Fornwalt, Brandon; Rayle, Christopher; Gal, Thomas J; Jones, Raleigh O; Shinn, Jennifer B

2013-06-01

Videonystagmography (VNG) is used widely in the assessment of balance dysfunction. The full test battery can be time-consuming and can induce patient discomfort. The purpose of this study was to examine the value of monothermal caloric testing in predicting unilateral caloric weakness, as well as abnormal VNG vestibular and nonvestibular eye movement, while considering the time and reimbursement associated with these tests. In a retrospective review of 645 patients who completed a comprehensive VNG test battery with bithermal caloric testing, we calculated the specificity, sensitivity, and predictive values of monothermal caloric testing in relation to bithermal caloric results and noncaloric VNG results. With unilateral vestibular weakness (UVW) defined as a 25% interear difference, warm-air monothermal caloric testing yielded a sensitivity of 87% and a negative predictive value of 90% for predicting UVW. With a 10% UVW definition, the warm-air caloric testing sensitivity increased to 95% and the negative predictive value to 92%. Warm-air monothermal caloric testing had a positive predictive value of 85% and a negative predictive value of 18% for predicting noncaloric VNG findings; cold-air monothermal and bithermal testing displayed similar results. Isolated monothermal testing is a sensitive screening tool for detecting UVW, but is not adequate for predicting noncaloric VNG results.

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

International Nuclear Information System (INIS)

Sugimoto, Jun; Murao, Yoshio

1983-09-01

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

Countermeasure against thermal fatigue crack of primary loop recirculation pump in BWR

International Nuclear Information System (INIS)

Noda, Hiroshi; Narabayashi, Tadashi; Takahashi, Yuuji

2008-01-01

The reactor water was fed to the purge water of the mechanical seal on the original design of the primary loop recirculation pump. Because the mechanical seal had a short life due to the cruds in the reactor water, the clean purge water was adopted instead of the reactor water. After this modification, the shallow cracks were found on the surface of the pump shaft and casing cover due to the temperature fluctuation between the cold purge water and the hot pump discharge water. The fundamental mechanism and countermeasure were investigated by scale test, mock-up test and so on. The flow barrier with a heater was contrived through these tests. It has been introduced gradually in operating and constructing PLR pumps after its completion in 1995. The PLR pumps are overhauled around every 10 years in Japan. The first overhaul of the PLR pumps showed no cracks around the pump shaft and casing over after 10 years' operation. This paper presents both its development process and inspection results. (author)

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Experimental Investigation of Tensile Test on Connection of Cold-formed Cut-curved Steel Section

Science.gov (United States)

Sani, Mohd Syahrul Hisyam Mohd; Muftah, Fadhluhartini; Rahman, Nurul Farraheeda Abdul; Fakri Muda, Mohd

2017-08-01

Cold-formed steel (CFS) is widely used as structural and non-structural components such as roof trusses and purlin. A CFS channel section with double intermediate web stiffener and lipped is chosen based on the broader usage in roof truss construction. CFS section is cut to form cold-formed pre-cut-curved steel section and lastly strengthened by several types of method or likely known as connection to establish the cold-formed cut-curved steel (CFCCS) section. CFCCS is proposed to be used as a top chord section in the roof truss system. The CFCCS is to resist the buckling phenomena of the roof truss structure and reduced the compression effect on the top chord. The tensile test connection of CFCCS section, especially at the flange element with eight types of connection by welding, plate with self-drilling screw and combination is investigated. The flange element is the weakest part that must be solved first other than the web element because they are being cut totally, 100% of their length for curving process. The testing is done using a universal testing machine for a tensile load. From the experiment, specimen with full welding has shown as a good result with an ultimate load of 13.37 kN and reported having 35.41% when compared with normal specimen without any of connection methods. Furthermore, the experimental result is distinguished by using Eurocode 3. The failure of a full welding specimen is due to breaking at the welding location. Additionally, all specimens with either full weld or spot weld or combination failed due to breaking on weld connection, but specimen with flange plate and self-drilling screw failed due to tilting and bearing. Finally, the full welding specimen is chosen as a good connection to perform the strengthening method of CFCCS section.

Frequent extreme cold exposure and brown fat and cold-induced thermogenesis: A study in a monozygotic twin

NARCIS (Netherlands)

M.J. Vosselman (Maarten J.); G.H.E.J. Vijgen (Guy H. E. J.); B.R.M. Kingma (Boris R. M.); B. Brans (Boudewijn); W.D. Van Marken Lichtenbelt (Wouter D.)

2014-01-01

textabstractIntroduction: Mild cold acclimation is known to increase brown adipose tissue (BAT) activity and cold-induced thermogenesis (CIT) in humans. We here tested the effect of a lifestyle with frequent exposure to extreme cold on BAT and CIT in a Dutch man known as 'the Iceman', who has

Spectroscopy with cold and ultra-cold neutrons

Directory of Open Access Journals (Sweden)

Abele Hartmut

2015-01-01

Full Text Available We present two new types of spectroscopy methods for cold and ultra-cold neutrons. The first method, which uses the R×B drift effect to disperse charged particles in a uniformly curved magnetic field, allows to study neutron β-decay. We aim for a precision on the 10−4 level. The second method that we refer to as gravity resonance spectroscopy (GRS allows to test Newton’s gravity law at short distances. At the level of precision we are able to provide constraints on any possible gravity-like interaction. In particular, limits on dark energy chameleon fields are improved by several orders of magnitude.

Final report on the Controlled Cold Helium Spill Test in the LHC tunnel at CERN

International Nuclear Information System (INIS)

Dufay-Chanat, L; Bremer, J; Casas-Cubillos, J; Koettig, T; Vauthier, N; Van Weelderen, R; Winkler, T; Chorowski, M; Grabowski, M; Jedrusyna, A; Lindell, G; Nonis, M

2015-01-01

The 27 km circumference LHC underground tunnel is a space in which the helium cooled LHC magnets are installed. The vacuum enclosures of the superconducting magnets are protected by over-pressure safety relief devices that open whenever cold helium escapes either from the magnet cold enclosure or from the helium supply headers, into this vacuum enclosure. A 3-m long no stay zone around these devices is defined based on scale model studies, protecting the personnel against cold burns or asphyxia caused by such a helium release event. Recently, several simulation studies have been carried out modelling the propagation of the helium/air mixture, resulting from the opening of such a safety device, along the tunnel. The released helium flows vary in the range between 1 kg/s and 0.1 kg/s. To validate these different simulation studies, real life mock-up tests have been performed inside the LHC tunnel, releasing helium flow rates of 1 kg/s, 0.3 kg/s and 0.1 kg/s. For each test, up to 1000 liters of liquid helium were released under standard operational tunnel conditions. The data recorded include oxygen concentration, temperature and flow speed measurements, and video footage used to assess qualitatively the visibility. These measurements have been made in the up- and downstream directions, with respect to the air ventilation flow, of the spill point.This paper presents the experimental set-up under which these release tests were made, the effects of these releases on the atmospheric tunnel condition as a function of the release flow rate. We discuss the modification to the personnel access conditions to the LHC tunnel that are presently implemented as a result of these tests. (paper)

Final report on the Controlled Cold Helium Spill Test in the LHC tunnel at CERN

Science.gov (United States)

Dufay-Chanat, L.; Bremer, J.; Casas-Cubillos, J.; Chorowski, M.; Grabowski, M.; Jedrusyna, A.; Lindell, G.; Nonis, M.; Koettig, T.; Vauthier, N.; van Weelderen, R.; Winkler, T.

2015-12-01

The 27 km circumference LHC underground tunnel is a space in which the helium cooled LHC magnets are installed. The vacuum enclosures of the superconducting magnets are protected by over-pressure safety relief devices that open whenever cold helium escapes either from the magnet cold enclosure or from the helium supply headers, into this vacuum enclosure. A 3-m long no stay zone around these devices is defined based on scale model studies, protecting the personnel against cold burns or asphyxia caused by such a helium release event. Recently, several simulation studies have been carried out modelling the propagation of the helium/air mixture, resulting from the opening of such a safety device, along the tunnel. The released helium flows vary in the range between 1 kg/s and 0.1 kg/s. To validate these different simulation studies, real life mock-up tests have been performed inside the LHC tunnel, releasing helium flow rates of 1 kg/s, 0.3 kg/s and 0.1 kg/s. For each test, up to 1000 liters of liquid helium were released under standard operational tunnel conditions. The data recorded include oxygen concentration, temperature and flow speed measurements, and video footage used to assess qualitatively the visibility. These measurements have been made in the up- and downstream directions, with respect to the air ventilation flow, of the spill point. This paper presents the experimental set-up under which these release tests were made, the effects of these releases on the atmospheric tunnel condition as a function of the release flow rate. We discuss the modification to the personnel access conditions to the LHC tunnel that are presently implemented as a result of these tests.

Characterization of the deformation texture after tensile test and cold rolling of a Ti-6Al-4V sheet alloy

International Nuclear Information System (INIS)

Mehdi, B; Badji, R; Azzeddine, H; Alili, B; Bradai, D; Ji, V

2015-01-01

The deformation texture after cold rolling and tensile test of an industrial Ti-6Al-4V sheet alloy was studied using X-ray diffraction. The alloy was subjected to a cold rolling to different thickness reductions (from 20% to 60%) and then tensile tests have been carried out along three directions relatively to the rolling direction (0°, 45° and 90°). The experimental results were compared to the existing literature and discussed in terms of active plastic deformation mechanisms. (paper)

Studies of the remagnetization process in cold drawn Fe-rich thin amorphous wires

International Nuclear Information System (INIS)

Gawronski, P.; Zhukov, A.P.; Blanco, J.M.; Gonzalez, J.; Zhukova, V.; Kulakowski, K.

2007-01-01

Results on local hysteresis loops for Fe-rich amorphous cold-drawn wires with the sample lengths of 30 mm and 70 mm under applied stress, σ, up to 1300 MPa are reported. For both lengths, the magnetization, M, increases with σ and the coercive field, H C , decreases. Near the wire end, M results to be larger than at the wire center, and H C sharply decreases. This reduction is eliminated if the applied stress is large enough. It is assumed, that the magnetic configuration at the wire end for the cold-drawn wires is different from the other wires families

Cold acclimation increases cold tolerance independently of diapause programing in the bean bug, Riptortus pedestris.

Science.gov (United States)

Rozsypal, J; Moos, M; Goto, S G

2017-10-17

The bean bug (Riptortus pedestris) is a pest of soybeans and other legumes in Japan and other Asian countries. It enters a facultative adult diapause on exposure to short days. While photoperiodism and diapause are well understood in R. pedestris, knowledge of cold tolerance is very limited, as is information on the effect of diapause on cold tolerance. We examined the effect of photoperiod, cold acclimation, and feeding status on cold tolerance in R. pedestris. We found that cold acclimation significantly increased survival at -10°C in both long- and short-day adult R. pedestris. Since the difference in cold survival between long- and short-day cold-acclimated groups was only marginal, we conclude that entering diapause is not crucial for R. pedestris to successfully pass through cold acclimation and become cold tolerant. We observed similar effects in 5th instar nymphs, with both long- and short-day cold-acclimated groups surviving longer cold exposures compared with non-acclimated groups. Starvation, which was tested only in adult bugs, had only a negligible and negative impact on cold survival. Although cold tolerance significantly increased with cold acclimation in adult bugs, supercooling capacity unexpectedly decreased. Our results suggest that changes in supercooling capacity as well as in water content are unrelated to cold tolerance in R. pedestris. An analysis of metabolites revealed differences between the treatments, and while several metabolites markedly increased with cold acclimation, their concentrations were too low to have a significant effect on cold tolerance.

Sensing loop performance monitoring in the safety systems of nuclear power stations

International Nuclear Information System (INIS)

Colley, R.C.; Widmeyer, M.; Weiss, J.H.; Wiegle, H.R.

1991-01-01

This paper reports on plant technical specifications and NRC regulatory guides which require testing of sensing loops to detect degradation and failure. Industry efforts have focused on specific manual testing to detect individual failure modes such as increased response time and calibration drift. Recent work performed by EPRI and by others using instrument loop data, failure modes, and effects analyses (FMEAs), and experience with utility on-line sensor health monitoring programs has established qualitative physical models of the sensing loop. This methodology has demonstrated that sensing loop cross comparison techniques can provide equivalent indication of sensing loop performance. It also provides more frequent sensing loop health indication than manual testing and reduces the requirement for manual testing

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

Science.gov (United States)

Plebani, Mario; Lippi, Giuseppe

2011-07-01

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

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

Science.gov (United States)

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

2005-01-01

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

Research and design of 3He pressure control loop

International Nuclear Information System (INIS)

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

2008-01-01

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

The effect of cold work on grain boundary precipitation and sensitization in nitrogen added type 316L stainless steels

International Nuclear Information System (INIS)

Seo, Moo Hong; Chun, Byong Sun; Oh, Yong Jun; Ryu, Woo Seog; Hong, Jun Hwa

1998-01-01

The precipitation and sensitization behavior of nitrogen added type 316L Stainless Steels (SS) were investigated by using specimens cold worked for 0∼40%. The alloys had a variation in nitrogen content from 0.04 to 0.15%. To quantify the degree of sensitization, Double-Loop Electrochemical Potentiokinetic Reactivation (DL-EPR) test was performed in a 0.1M H 2 SO 4 + 0.01M KSCN solution at 30 .deg. C. The addition of nitrogen increased sensitization resistance by retarding the onset of M 23 C 6 precipitation and shifted Time-Temperature-Sensitization(TTS) curve to higher temperature and longer time range. Cold work accelerated the M 23 C 6 precipitation and sensitization kinetic due to the increase in dislocation density. However, the acceleration of sensitization was found to depend on the added nitrogen content in the alloys. The alloys with high nitrogen(>0.1%N) content exhibited higher acceleration of the sensitization as a function of the cold work than that with low nitrogen content. From the microstructural analysis, this was found to be attributed to the development of intensive slip bands during cold work and retardation of dislocation annihilation during subsequent aging in the alloys with high nitrogen content

2010 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond

International Nuclear Information System (INIS)

Lewis, Mike

2011-01-01

This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond from November 1, 2009 through October 31, 2010. The report contains the following information: (1) Facility and system description; (2) Permit required effluent monitoring data and loading rates; (3) Groundwater monitoring data; (4) Status of compliance activities; and (5) Discussion of the facility's environmental impacts. During the 2010 permit year, approximately 164 million gallons of wastewater were discharged to the Cold Waste Pond. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

Facts about the Common Cold

Science.gov (United States)

... different viruses. Rhinovirus is the most common cause, accounting for 10 to 40 percent of colds. Other common cold viruses include coronavirus and ... RSS | Terms Of Use | Privacy | Sitemap Our Family Of Sites ... Introduction Risk Factors Screening Symptoms Tumor Testing Summary '; var ...

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

International Nuclear Information System (INIS)

Dittrich, B; Thiemann, T

2006-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-07-01

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

Multi-Megawatt-Scale Power-Hardware-in-the-Loop Interface for Testing Ancillary Grid Services by Converter-Coupled Generation: Preprint

Energy Technology Data Exchange (ETDEWEB)

Koralewicz, Przemyslaw J [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Gevorgian, Vahan [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Wallen, Robert B [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

2017-07-26

Power-hardware-in-the-loop (PHIL) is a simulation tool that can support electrical systems engineers in the development and experimental validation of novel, advanced control schemes that ensure the robustness and resiliency of electrical grids that have high penetrations of low-inertia variable renewable resources. With PHIL, the impact of the device under test on a generation or distribution system can be analyzed using a real-time simulator (RTS). PHIL allows for the interconnection of the RTS with a 7 megavolt ampere (MVA) power amplifier to test multi-megawatt renewable assets available at the National Wind Technology Center (NWTC). This paper addresses issues related to the development of a PHIL interface that allows testing hardware devices at actual scale. In particular, the novel PHIL interface algorithm and high-speed digital interface, which minimize the critical loop delay, are discussed.

Finalize field testing of cold climate heat pump (CCHP) based on tandem vapor injection compressors

Energy Technology Data Exchange (ETDEWEB)

Shen, Bo [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Baxter, Van D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Abdelaziz, Omar [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rice, C. Keith [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-03-01

This report describes the system diagram and control algorithm of a prototype air-source cold climate heat pump (CCHP) using tandem vapor injection (VI) compressors. The prototype was installed in Fairbanks, Alaska and underwent field testing starting in 09/2016. The field testing results of the past six months, including compressor run time fractions, measured COPs and heating capacities, etc., are presented as a function of the ambient temperature. Two lessons learned are also reported.

Cold/menthol TRPM8 receptors initiate the cold-shock response and protect germ cells from cold-shock-induced oxidation.

Science.gov (United States)

Borowiec, Anne-Sophie; Sion, Benoit; Chalmel, Frédéric; D Rolland, Antoine; Lemonnier, Loïc; De Clerck, Tatiana; Bokhobza, Alexandre; Derouiche, Sandra; Dewailly, Etienne; Slomianny, Christian; Mauduit, Claire; Benahmed, Mohamed; Roudbaraki, Morad; Jégou, Bernard; Prevarskaya, Natalia; Bidaux, Gabriel

2016-09-01

Testes of most male mammals present the particularity of being externalized from the body and are consequently slightly cooler than core body temperature (4-8°C below). Although, hypothermia of the testis is known to increase germ cells apoptosis, little is known about the underlying molecular mechanisms, including cold sensors, transduction pathways, and apoptosis triggers. In this study, using a functional knockout mouse model of the cold and menthol receptors, dubbed transient receptor potential melastatine 8 (TRPM8) channels, we found that TRPM8 initiated the cold-shock response by differentially modulating cold- and heat-shock proteins. Besides, apoptosis of germ cells increased in proportion to the cooling level in control mice but was independent of temperature in knockout mice. We also observed that the rate of germ cell death correlated positively with the reactive oxygen species level and negatively with the expression of the detoxifying enzymes. This result suggests that the TRPM8 sensor is a key determinant of germ cell fate under hypothermic stimulation.-Borowiec, A.-S., Sion, B., Chalmel, F., Rolland, A. D., Lemonnier, L., De Clerck, T., Bokhobza, A., Derouiche, S., Dewailly, E., Slomianny, C., Mauduit, C., Benahmed, M., Roudbaraki, M., Jégou, B., Prevarskaya, N., Bidaux, G. Cold/menthol TRPM8 receptors initiate the cold-shock response and protect germ cells from cold-shock-induced oxidation. © The Author(s).

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

Energy Technology Data Exchange (ETDEWEB)

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

2006-07-14

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

The Combined Effect of Cold and Moisture on Manual Performance.

Science.gov (United States)

Ray, Matthew; Sanli, Elizabeth; Brown, Robert; Ennis, Kerri Ann; Carnahan, Heather

2018-02-01

Objective The aim of this study was to investigate the combined effect of cold and moisture on manual performance and tactile sensitivity. Background People working in the ocean environment often perform manual work in cold and wet conditions. Although the independent effects of cold and moisture on hand function are known, their combined effect has not been investigated. Method Participants completed sensory (Touch-Test, two-point discrimination) and motor (Purdue Pegboard, Grooved Pegboard, reef knot untying) tests in the following conditions: dry hand, wet hand, cold hand, and cold and wet hand. Results For the Purdue Pegboard and knot untying tasks, the greatest decrement in performance was observed in the cold-and-wet-hand condition, whereas the decrements seen in the cold-hand and wet-hand conditions were similar. In the Grooved Pegboard task, the performance decrements exhibited in the cold-and-wet-hand condition and the cold-hand condition were similar, whereas no decrement was observed in the wet-hand condition. Tactile sensitivity was reduced in the cold conditions for the Touch-Test but not the two-point discrimination test. The combined effect of cold and moisture led to the largest performance decrements except when intrinsic object properties helped with grasp maintenance. The independent effects of cold and moisture on manual performance were comparable. Application Tools and equipment for use in the cold ocean environment should be designed to minimize the effects of cold and moisture on manual performance by including object properties that enhance grasp maintenance and minimize the fine-dexterity requirements.

Contribution of laser Doppler flowmetry with venoarteriolar reflex, cold, and rewarming testing, and intravital capillaroscopy to diagnose Raynaud's phenomenon

Directory of Open Access Journals (Sweden)

Zeman J

2014-05-01

Full Text Available Jan Zeman,1 Oksana Turyanytsya,1 Vojtĕch Kapsa,2 Mojmír Eliáš3 1Department of Clinical Cardiology and Angiology, Hospital Bulovka, 2Charles University in Prague, Faculty of Mathematics and Physics, 3Kooperativa a.s., Pobrezni, Prague, Czech Republic Background: The early differential diagnosis of Raynaud’s phenomenon (RP is crucial for the prognosis and therapy of these patients. In our microcirculatory laboratory, we use intravital capillaroscopy (IC, plethysmography (P, and laser Doppler flowmetry (LDF for examining acrosyndromes. We combine LDF with venoarteriolar reflex test, cold test, and rewarming test to achieve more reliable diagnoses of acrosyndromes. Patients and methods: We examined LDF and IC according to a strict protocol using a battery of tests (venoarteriolar reflex test, cold test, rewarming test applied to five different groups of people and compared their results: healthy controls, primary Raynaud’s phenomenon (PRP, systemic scleroderma, vibration white finger, and peripheral artery occlusive disease. Our tests included 340 individuals (72 patients plus 268 controls. Results: Although all tests provided some differences between controls and patients, only the rewarming test offered significant results for differential diagnoses. Conclusion: IC and LDF combined with the battery of tests (venoarteriolar reflex test, cold test, rewarming test under standard conditions can be used as reliable tools to distinguish between PRP and some types of secondary RP (especially in the case of systemic scleroderma, vibration white fingers, or peripheral artery occlusive disease; RPs with organic occlusions of the small arteries causing the diseases. Our methodology can help to distinguish between other types of RP, as well. Keywords: Raynaud’s phenomenon, acrosyndrome, laser Doppler flowmetry, intravital capillaroscopy, scleroderma, vibration white finger, peripheral artery occlusive disease

Rogowski Loop design for NSTX

International Nuclear Information System (INIS)

McCormack, B.; Kaita, R.; Kugel, H.; Hatcher, R.

2000-01-01

The Rogowski Loop is one of the most basic diagnostics for tokamak operations. On the National Spherical Torus Experiment (NSTX), the plasma current Rogowski Loop had the constraints of the very limited space available on the center stack, 5,000 volt isolation, flexibility requirements as it remained a part of the Center Stack assembly after the first phase of operation, and a +120 C temperature requirement. For the second phase of operation, four Halo Current Rogowski Loops under the Center Stack tiles will be installed having +600 C and limited space requirements. Also as part of the second operational phase, up to ten Rogowski Loops will installed to measure eddy currents in the Passive Plate support structures with +350 C, restricted space, and flexibility requirements. This presentation will provide the details of the material selection, fabrication techniques, testing, and installation results of the Rogowski Loops that were fabricated for the high temperature operational and bakeout requirements, high voltage isolation requirements, and the space and flexibility requirements imposed upon the Rogowski Loops. In the future operational phases of NSTX, additional Rogowski Loops could be anticipated that will measure toroidal plasma currents in the vacuum vessel and in the Passive Plate assemblies

Development of cold neutron spectrometers

International Nuclear Information System (INIS)

Lee, Changhee; Lee, C. H.; So, J. Y.; Park, S.; Han, Y. S.; Cho, S. J.; Moon, M. K.; Choi, Y. H.; Sun, G. M.

2012-03-01

□ Cold Neutron Triple Axsis Spectrometer (Cold-TAS) Development Ο Fabrication and Installation of the Major Cold-TAS Components Ο Performance Test of the Cold-TAS □ Cold Neutron Time-of-Flight Spectrometer(DC-TOF) Development Ο Fabrication of the Major DC-TOF Components Ο Development DC-TOF Data Reduction Software □ Expected Contribution The two world-class inelastic neutron scattering instruments measure atomic or molecular scale dynamics of meV energy range. This unprecedented measurement capability in the country will enable domestic and international scientists to observe new phenomena in their materials research to obtain world class results. Especially those who work in the fields of magnetic properties of superconductors and multiferroics, molecular dynamics, etc. will get more benefit from these two instruments

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

Energy Technology Data Exchange (ETDEWEB)

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

2005-08-15

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

Integrated Testing of a 4-Bed Molecular Sieve and a Temperature-Swing Adsorption Compressor for Closed-Loop Air Revitalization

Science.gov (United States)

Knox, James C.; Mulloth, Lila M.; Affleck, David L.

2004-01-01

Accumulation and subsequent compression of carbon dioxide that is removed from space cabin are two important processes involved in a closed-loop air revitalization scheme of the International Space Station (ISS). The 4-Bed Molecular Sieve (4BMS) of ISS currently operates in an open loop mode without a compressor. This paper reports the integrated 4BMS and liquid-cooled TSAC testing conducted during the period of March 3 to April 18, 2003. The TSAC prototype was developed at NASA Ames Research Center (ARC). The 4BMS was modified to a functionally flight-like condition at NASA Marshall Space Flight Center (MSFC). Testing was conducted at MSFC. The paper provides details of the TSAC operation at various CO2 loadings and corresponding performance of CDRA.

Loop Transfer Matrix and Loop Quantum Mechanics

International Nuclear Information System (INIS)

Savvidy, George K.

2000-01-01

The gonihedric model of random surfaces on a 3d Euclidean lattice has equivalent representation in terms of transfer matrix K(Q i ,Q f ), which describes the propagation of loops Q. We extend the previous construction of the loop transfer matrix to the case of nonzero self-intersection coupling constant κ. We introduce the loop generalization of Fourier transformation which allows to diagonalize transfer matrices, that depend on symmetric difference of loops only and express all eigenvalues of 3d loop transfer matrix through the correlation functions of the corresponding 2d statistical system. The loop Fourier transformation allows to carry out the analogy with quantum mechanics of point particles, to introduce conjugate loop momentum P and to define loop quantum mechanics. We also consider transfer matrix on 4d lattice which describes propagation of memebranes. This transfer matrix can also be diagonalized by using the generalized Fourier transformation, and all its eigenvalues are equal to the correlation functions of the corresponding 3d statistical system. In particular the free energy of the 4d membrane system is equal to the free energy of 3d gonihedric system of loops and is equal to the free energy of 2d Ising model. (author)

Commissioning of an Integral Effect Test Loop for SMART

Energy Technology Data Exchange (ETDEWEB)

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

2013-05-15

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

On the loop-loop scattering amplitudes in Abelian and non-Abelian gauge theories

International Nuclear Information System (INIS)

Meggiolaro, Enrico

2005-01-01

The high-energy elastic scattering amplitude of two colour-singlet qq-bar pairs is governed by the correlation function of two Wilson loops, which follow the classical straight lines for quark (antiquark) trajectories. This quantity is expected to be free of IR divergences, differently from what happens for the parton-parton elastic scattering amplitude, described, in the high-energy limit, by the expectation value of two Wilson lines. We shall explicitly test this IR finiteness by a direct non-perturbative computation of the loop-loop scattering amplitudes in the (pedagogic, but surely physically interesting) case of quenched QED. The results obtained for the Abelian case will be generalized to the case of a non-Abelian gauge theory with Nc colours, but stopping to the order O(g4) in perturbation theory. In connection with the above-mentioned IR finiteness, we shall also discuss some analytic properties of the loop-loop scattering amplitudes in both Abelian and non-Abelian gauge theories, when going from Minkowskian to Euclidean theory, which can be relevant to the still unsolved problem of the s-dependence of hadron-hadron total cross-sections

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

NARCIS (Netherlands)

Rix, L.; de Goeij, J.M.; Mueller, C.E.; Struck, U.; Middelburg, J.J.; van Duyl, F.C.; Al-Horani, F.A.; Wild, C.; Naumann, M.S.; Van Oevelen, D.

2016-01-01

Shallow warm-water and deep-sea cold-water corals engineer the coral reef framework and fertilize reef communities by releasing coral mucus, a source of reef dissolved organic matter (DOM). By transforming DOM into particulate detritus, sponges play a key role in transferring the energy and

Crack growth testing of cold worked stainless steel in a simulated PWR primary water environment to assess susceptibility to stress corrosion cracking

International Nuclear Information System (INIS)

Tice, D.R.; Stairmand, J.W.; Fairbrother, H.J.; Stock, A.

2007-01-01

Although austenitic stainless steels do not show a high degree of susceptibility to stress corrosion cracking (SCC) in PWR primary environments, there is limited evidence from laboratory testing that crack propagation may occur under some conditions for materials in a cold-worked condition. A test program is therefore underway to examine the factors influencing SCC propagation in good quality PWR primary coolant. Type 304 stainless steel was subjected to cold working by either rolling (at ambient or elevated temperature) or fatigue cycling, to produce a range of yield strengths. Compact tension specimens were fabricated from these materials and tested in simulated high temperature (250-300 o C) PWR primary coolant. It was observed that the degree of crack propagation was influenced by the degree of cold work, the crack growth orientation relative to the rolling direction and the method of working. (author)

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

Science.gov (United States)

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

2018-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Harper, R.E.

1981-11-01

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

2011 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond

International Nuclear Information System (INIS)

Lewis, Mike

2012-01-01

This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond from November 1, 2010 through October 31, 2011. The report contains the following information: Facility and system description Permit required effluent monitoring data and loading rates Groundwater monitoring data Status of compliance activities Noncompliance and other issues Discussion of the facility's environmental impacts During the 2011 permit year, approximately 166 million gallons of wastewater were discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing

International Nuclear Information System (INIS)

D.R. Jackson; G.R. Kiebel

1999-01-01

Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training

Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing

Energy Technology Data Exchange (ETDEWEB)

D.R. Jackson; G.R. Kiebel

1999-08-24

Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training.

Summary revaluation of cold testing of the first block of nuclear power plant Mochovce

International Nuclear Information System (INIS)

Miskolci, M.; Sarvaic, I.

1998-01-01

The document contents summary revaluation of the stage of cold testing of the first unit of nuclear power plant Mochovce. The valuation is processed in individual systems with safety significance. The process and results of system testing and their conclusions for the block readiness for active testing are summarized in the document. The valuation has been elaborated by a scientific management for start-up of nuclear power plant Mochovce as an independent conductor assistance for activation check from the nuclear safety point of view. The valuation of the activation results of systems in the first unit of nuclear power plant was processed as of 15.3.1998

Alstom's Chemical Looping Combustion Prototype for CO₂ Capture from Existing Pulverized Coal-Fired Power Plants

Energy Technology Data Exchange (ETDEWEB)

Andrus, Jr., Herbert E. [Alstom Power Inc., Windsor, CT (United States); Chiu, John H. [Alstom Power Inc., Windsor, CT (United States); Edberg, Carl D. [Alstom Power Inc., Windsor, CT (United States); Thibeault, Paul R. [Alstom Power Inc., Windsor, CT (United States); Turek, David G. [Alstom Power Inc., Windsor, CT (United States)

2012-09-30

Alstom’s Limestone Chemical Looping (LCL™) process has the potential to capture CO₂ from new and existing coal-fired power plants while maintaining high plant power generation efficiency. This new power plant concept is based on a hybrid combustion- gasification process utilizing high temperature chemical and thermal looping technology. This process could also be potentially configured as a hybrid combustion-gasification process producing a syngas or hydrogen for various applications while also producing a separate stream of CO₂ for use or sequestration. The targets set for this technology is to capture over 90% of the total carbon in the coal at cost of electricity which is less than 20% greater than Conventional PC or CFB units. Previous work with bench scale test and a 65 kWt Process Development Unit Development (PDU) has validated the chemistry required for the chemical looping process and provided for the investigation of the solids transport mechanisms and design requirements. The objective of this project is to continue development of the combustion option of chemical looping (LCL-C™) by designing, building and testing a 3 MWt prototype facility. The prototype includes all of the equipment that is required to operate the chemical looping plant in a fully integrated manner with all major systems in service. Data from the design, construction, and testing will be used to characterize environmental performance, identify and address technical risks, reassess commercial plant economics, and develop design information for a demonstration plant planned to follow the proposed Prototype. A cold flow model of the prototype will be used to predict operating conditions for the prototype and help in operator training. Operation of the prototype will provide operator experience with this new technology and performance data of the LCL-C™ process, which will be applied to the commercial design and economics and plan for a future demonstration

Short-term test-retest-reliability of conditioned pain modulation using the cold-heat-pain method in healthy subjects and its correlation to parameters of standardized quantitative sensory testing.

Science.gov (United States)

Gehling, Julia; Mainka, Tina; Vollert, Jan; Pogatzki-Zahn, Esther M; Maier, Christoph; Enax-Krumova, Elena K

2016-08-05

Conditioned Pain Modulation (CPM) is often used to assess human descending pain inhibition. Nine different studies on the test-retest-reliability of different CPM paradigms have been published, but none of them has investigated the commonly used heat-cold-pain method. The results vary widely and therefore, reliability measures cannot be extrapolated from one CPM paradigm to another. Aim of the present study was to analyse the test-retest-reliability of the common heat-cold-pain method and its correlation to pain thresholds. We tested the short-term test-retest-reliability within 40 ± 19.9 h using a cold-water immersion (10 °C, left hand) as conditioning stimulus (CS) and heat pain (43-49 °C, pain intensity 60 ± 5 on the 101-point numeric rating scale, right forearm) as test stimulus (TS) in 25 healthy right-handed subjects (12females, 31.6 ± 14.1 years). The TS was applied 30s before (TSbefore), during (TSduring) and after (TSafter) the 60s CS. The difference between the pain ratings for TSbefore and TSduring represents the early CPM-effect, between TSbefore and TSafter the late CPM-effect. Quantitative sensory testing (QST, DFNS protocol) was performed on both sessions before the CPM assessment. paired t-tests, Intraclass correlation coefficient (ICC), standard error of measurement (SEM), smallest real difference (SRD), Pearson's correlation, Bland-Altman analysis, significance level p Pain ratings during CPM correlated significantly (ICC: 0.411…0.962) between both days, though ratings for TSafter were lower on day 2 (p pain thresholds. The short-term test-retest-reliability of the early CPM-effect using the heat-cold-pain method in healthy subjects achieved satisfying results in terms of the ICC. The SRD of the early CPM effect showed that an individual change of > 20 NRS can be attributed to a real change rather than chance. The late CPM-effect was weaker and not reliable.

IP and resistivity survey at the INEL cold test pit

International Nuclear Information System (INIS)

Frangos, W.

1997-01-01

Induced polarization (IP) is a technique for detection of diffuse occurrences of metallic material. The Idaho National Laboratory (INEL) Cold Test Pit (CTP) has been carefully constructed to simulate stored hazardous waste occurrences. IP and resistivity surveys of the CTP show a very strong IP response and a modest resistivity response associated with the simulated waste. Several false positive resistivity anomalies are noted. 2-D inversion of the data accurately determine the lateral limits of the waste zone. The depth of burial indicated is deeper than reported by the construction engineers. Limited depth extent is recognized, in distinction to results from other, resistivity-based, methods, but it also appears too deep. IP holds promise as a method for locating buried waste

Simulated HLLW compositions for cold test of waste management development

International Nuclear Information System (INIS)

Banba, Tsunetaka; Kimura, Hideo; Kamizono, Hiroshi; Tashiro, Shingo

1982-07-01

Three grades of simulated high-level liquid waste (HLLW)-JW-A, JW-B, and JW-C - were proposed to be used respectively according to stages of various cold tests for safety assessment of HLW management. The composition of HLLW was estimated taking into account the spectrum of fission products and actinides, waste volume, corrosion products, and chemical additives. One of conditions, the spectrum of fission products and actinides of LWR spent fuels, was calculated by DCHAIN-code. Fuel burn-up of 28,000 MWD/tUO 2 and 33,000 MWD/tUO 2 were adopted as normal and maximum values of Japanese LWR power plants. The other conditions were estimated using the data obtained at Marcoule plant in France. (author)

Integrated Testing of a Carbon Dioxide Removal Assembly and a Temperature-Swing Adsorption Compressor for Closed-Loop Air Revitalization

Science.gov (United States)

Knox, J. C.; Mulloth, Lila; Frederick, Kenneth; Affleck, Dave

2003-01-01

Accumulation and subsequent compression of carbon dioxide that is removed from space cabin are two important processes involved in a closed-loop air revitalization scheme of the International Space Station (ISS). The carbon dioxide removal assembly (CDRA) of ISS currently operates in an open loop mode without a compressor. This paper describes the integrated test results of a flight-like CDRA and a temperature-swing adsorption compressor (TSAC) for carbon dioxide removal and compression. The paper provides details of the TSAC operation at various CO2 loadings and corresponding performance of CDRA.

Summary Of Cold Crucible Vitrification Tests Results With Savannah River Site High Level Waste Surrogates

Energy Technology Data Exchange (ETDEWEB)

Stefanovsky, Sergey; Marra, James; Lebedev, Vladimir

2014-01-13

The cold crucible inductive melting (CCIM) technology successfully applied for vitrification of low- and intermediate-level waste (LILW) at SIA Radon, Russia, was tested to be implemented for vitrification of high-level waste (HLW) stored at Savannah River Site, USA. Mixtures of Sludge Batch 2 (SB2) and 4 (SB4) waste surrogates and borosilicate frits as slurries were vitrified in bench- (236 mm inner diameter) and full-scale (418 mm inner diameter) cold crucibles. Various process conditions were tested and major process variables were determined. Melts were poured into 10L canisters and cooled to room temperature in air or in heat-insulated boxes by a regime similar to Canister Centerline Cooling (CCC) used at DWPF. The products with waste loading from ~40 to ~65 wt.% were investigated in details. The products contained 40 to 55 wt.% waste oxides were predominantly amorphous; at higher waste loadings (WL) spinel structure phases and nepheline were present. Normalized release values for Li, B, Na, and Si determined by PCT procedure remain lower than those from EA glass at waste loadings of up to 60 wt.%.

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

International Nuclear Information System (INIS)

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

1975-04-01

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

MAP Kinase Cascades Regulate the Cold Response by Modulating ICE1 Protein Stability.

Science.gov (United States)

Zhao, Chunzhao; Wang, Pengcheng; Si, Tong; Hsu, Chuan-Chih; Wang, Lu; Zayed, Omar; Yu, Zheping; Zhu, Yingfang; Dong, Juan; Tao, W Andy; Zhu, Jian-Kang

2017-12-04

Mitogen-activated protein kinase cascades are important signaling modules that convert environmental stimuli into cellular responses. We show that MPK3, MPK4, and MPK6 are rapidly activated after cold treatment. The mpk3 and mpk6 mutants display increased expression of CBF genes and enhanced freezing tolerance, whereas constitutive activation of the MKK4/5-MPK3/6 cascade in plants causes reduced expression of CBF genes and hypersensitivity to freezing, suggesting that the MKK4/5-MPK3/6 cascade negatively regulates the cold response. MPK3 and MPK6 can phosphorylate ICE1, a basic-helix-loop-helix transcription factor that regulates the expression of CBF genes, and the phosphorylation promotes the degradation of ICE1. Interestingly, the MEKK1-MKK2-MPK4 pathway constitutively suppresses MPK3 and MPK6 activities and has a positive role in the cold response. Furthermore, the MAPKKK YDA and two calcium/calmodulin-regulated receptor-like kinases, CRLK1 and CRLK2, negatively modulate the cold activation of MPK3/6. Our results uncover important roles of MAPK cascades in the regulation of plant cold response. Copyright © 2017 Elsevier Inc. All rights reserved.

Analysis of ATLAS 6-inch cold leg break simulation with MARS-KS code

Energy Technology Data Exchange (ETDEWEB)

Kim, Se Yun; Jun, Hwang Yong; Ha, Sang Jun [Korea Electric Power Company, Daejeon (Korea, Republic of)

2011-05-15

A Domestic Standard Problem (DSP) exercise using ATLAS facility has been organized by KAERI. As the second DSP exercise, the 6-inch cold leg bottom break was determined. This experiment is the counterpart test to the DVI line break to verify the safety performance of the DVI method over the traditional CLI method. Compared with the large break LOCA, the phases of the small break LOCA prior to core recovery occur over a long period. The blowdown, natural circulation, loop seal clearance, boil-off, and core recovery phase should be investigated minutely with relevant models of safety analysis codes in order to predict these thermal hydraulic phenomena correctly. To investigate the ECC bypass phenomena, a finer study on the thermalhydraulic behavior in upper annulus downcomer was carried out

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

International Nuclear Information System (INIS)

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

1979-04-01

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

Improvement of Measurement Accuracy of Coolant Flow in a Test Loop

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Best-estimate analysis of a loss-of-coolant accident in a four-loop US PWR using TRAC-PD2

International Nuclear Information System (INIS)

Ireland, J.R.

1982-01-01

A 200-percent double-ended cold-leg break loss-of-coolant accident (LOCA) in a typical US pressurized water reactor (PWR) was simulated using the Transient Reactor Analysis Code (TRAC-PD2). The reactor system modeled represented a typical US PWR with four loops (three intact, one broken) and cold-leg emergency-core-cooling systems (ECCS). The finely noded TRAC model employed 440 three dimensional (r, THETA, z) vessel cells along with approximately 300 one-dimensional cells that modeled the primary system loops. The calculated peak-clad temperature of 950 0 K occurred during blowdown and the clad temperature excursion was terminated at 175 s, when complete core quenching occurred. Accumulator flows were initiated at 10 s, when the system pressure reached 4.08 MPa, and the refill phase ended at 36 s when the lower plenum refilled. During reflood, both bottom and falling film quench fronts were calculated

Cold-water acclimation does not modify whole-body fluid regulation during subsequent cold-water immersion.

Science.gov (United States)

Stocks, J M; Patterson, M J; Hyde, D E; Jenkins, A B; Mittleman, K D; Taylor, N A S

2004-06-01

We investigated the impact of cold-water acclimation on whole-body fluid regulation using tracer-dilution methods to differentiate between the intracellular and extracellular fluid compartments. Seven euhydrated males [age 24.7 (8.7) years, mass 74.4 (6.4) kg, height 176.8 (7.8) cm, sum of eight skinfolds 107.4 (20.4) mm; mean (SD)] participated in a 14-day cold-water acclimation protocol, with 60-min resting cold-water stress tests [CWST; 18.1 (0.1) degrees C] on days 1, 8 and 15, and 90-min resting cold-water immersions [18.4 (0.4) degrees C] on intervening days. Subjects were immersed to the 4th intercostal space. Intracellular and extracellular fluid compartments, and plasma protein, electrolyte and hormone concentrations were investigated. During the first CWST, the intracellular fluid (5.5%) and plasma volumes were reduced (6.1%), while the interstitial fluid volume was simultaneously expanded (5.4%). This pattern was replicated on days 8 and 15, but did not differ significantly among test days. Acclimation did not produce significant changes in the pre-immersion distribution of total body water, or changes in plasma osmolality, total protein, electrolyte, atrial natriuretic peptide or aldosterone concentrations. Furthermore, a 14-day cold-water acclimation regimen did not elicit significant changes in body-fluid distribution, urine production, or the concentrations of plasma protein, electrolytes or the fluid-regulatory hormones. While acclimation trends were not evident, we have confirmed that fluid from extravascular cells is displaced into the interstitium during acute cold-water immersion, both before and after cold acclimation.

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

International Nuclear Information System (INIS)

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

2002-01-01

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

The 400W at 1.8K Test Facility at CEA-Grenoble

Science.gov (United States)

Roussel, P.; Girard, A.; Jager, B.; Rousset, B.; Bonnay, P.; Millet, F.; Gully, P.

2006-04-01

A new test facility with a cooling capacity respectively of 400W at 1.8K or 800W at 4.5K, is now under nominal operation in SBT (Low Temperature Department) at CEA Grenoble. It has been recently used for thermohydraulic studies of two phase superfluid helium in autumn 2004. In the near future, this test bench will allow: - to test industrial components at 1.8K (magnets, cavities of accelerators) - to continue the present studies on thermohydraulics of two phase superfluid helium - to develop and simulate new cooling loops for ITER Cryogenics, and other applications such as high Reynolds number flows This new facility consists of a cold box connected to a warm compressor station (one subatmospheric oil ring pump in series with two screw compressors). The cold box, designed by AIR LIQUIDE, comprises two centrifugal cold compressors, a cold turbine, a wet piston expander, counter flow heat exchangers and two phase separators at 4.5K and 1.8K. The new facility uses a Programmable Logic Controller (PLC) connected to a bus for the measurements. The design is modular and will allow the use of saturated fluid flow (two phase flow at 1.8K or 4.5K) or single phase fluid forced flow. Experimental results and cooling capacity in different operation modes are detailed.

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

Science.gov (United States)

Nessler, Ralph D.

2010-01-01

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

Expert Cold Structure Development

Science.gov (United States)

Atkins, T.; Demuysere, P.

2011-05-01

The EXPERT Program is funded by ESA. The objective of the EXPERT mission is to perform a sub-orbital flight during which measurements of critical aero- thermodynamic phenomena will be obtained by using state-of-the-art instrumentation. As part of the EXPERT Flight Segment, the responsibility of the Cold Structure Development Design, Manufacturing and Validation was committed to the Belgian industrial team SONACA/SABCA. The EXPERT Cold Structure includes the Launcher Adapter, the Bottom Panel, the Upper Panel, two Cross Panels and the Parachute Bay. An additional Launcher Adapter was manufactured for the separation tests. The selected assembly definition and manufacturing technologies ( machined parts and sandwich panels) were dictated classically by the mass and stiffness, but also by the CoG location and the sensitive separation interface. Used as support for the various on-board equipment, the Cold Structure is fixed to but thermally uncoupled from the PM 1000 thermal shield. It is protect on its bottom panel by a thermal blanket. As it is a protoflight, analysis was the main tool for the verification. Low level stiffness and modal analysis tests have also been performed on the Cold Structure equipped with its ballast. It allowed to complete its qualification and to prepare SONACA/SABCA support for the system dynamic tests foreseen in 2011. The structure was finally coated with a thermal control black painting and delivered on time to Thales Alenia Space-Italy end of March 201.

Natively unstructured loops differ from other loops.

Directory of Open Access Journals (Sweden)

Avner Schlessinger

2007-07-01

Full Text Available Natively unstructured or disordered protein regions may increase the functional complexity of an organism; they are particularly abundant in eukaryotes and often evade structure determination. Many computational methods predict unstructured regions by training on outliers in otherwise well-ordered structures. Here, we introduce an approach that uses a neural network in a very different and novel way. We hypothesize that very long contiguous segments with nonregular secondary structure (NORS regions differ significantly from regular, well-structured loops, and that a method detecting such features could predict natively unstructured regions. Training our new method, NORSnet, on predicted information rather than on experimental data yielded three major advantages: it removed the overlap between testing and training, it systematically covered entire proteomes, and it explicitly focused on one particular aspect of unstructured regions with a simple structural interpretation, namely that they are loops. Our hypothesis was correct: well-structured and unstructured loops differ so substantially that NORSnet succeeded in their distinction. Benchmarks on previously used and new experimental data of unstructured regions revealed that NORSnet performed very well. Although it was not the best single prediction method, NORSnet was sufficiently accurate to flag unstructured regions in proteins that were previously not annotated. In one application, NORSnet revealed previously undetected unstructured regions in putative targets for structural genomics and may thereby contribute to increasing structural coverage of large eukaryotic families. NORSnet found unstructured regions more often in domain boundaries than expected at random. In another application, we estimated that 50%-70% of all worm proteins observed to have more than seven protein-protein interaction partners have unstructured regions. The comparative analysis between NORSnet and DISOPRED2 suggested

DNA looping by FokI: the impact of twisting and bending rigidity on protein-induced looping dynamics

Science.gov (United States)

Laurens, Niels; Rusling, David A.; Pernstich, Christian; Brouwer, Ineke; Halford, Stephen E.; Wuite, Gijs J. L.

2012-01-01

Protein-induced DNA looping is crucial for many genetic processes such as transcription, gene regulation and DNA replication. Here, we use tethered-particle motion to examine the impact of DNA bending and twisting rigidity on loop capture and release, using the restriction endonuclease FokI as a test system. To cleave DNA efficiently, FokI bridges two copies of an asymmetric sequence, invariably aligning the sites in parallel. On account of the fixed alignment, the topology of the DNA loop is set by the orientation of the sites along the DNA. We show that both the separation of the FokI sites and their orientation, altering, respectively, the twisting and the bending of the DNA needed to juxtapose the sites, have profound effects on the dynamics of the looping interaction. Surprisingly, the presence of a nick within the loop does not affect the observed rigidity of the DNA. In contrast, the introduction of a 4-nt gap fully relaxes all of the torque present in the system but does not necessarily enhance loop stability. FokI therefore employs torque to stabilise its DNA-looping interaction by acting as a ‘torsional’ catch bond. PMID:22373924

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

Directory of Open Access Journals (Sweden)

Geslot Benoit

2018-01-01

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

Sodium hydride precipitation in sodium cold traps

International Nuclear Information System (INIS)

McPheeters, C.C.; Raue, D.J.

1979-10-01

A series of experiments have been performed to test a calculational model for precipitation of NaH in sodium cold traps. The calculational model, called ACTMODEL, is a computer simulation that uses the system geometry and operating conditions as input to calculate a mass transfer coefficient and the distribution of NaH in a cold trap. The ACTMODEL was tested using an analytical cold trap (ACT) that is simple and essentially one-dimensional. The ACT flow and temperature profile may be controlled at any desired condition. The ACT was analyzed destructively after each test to measure the actual NaH distribution. Excellent agreement was obtained between the ACTMODEL simulations and the experiments

UPTF test instrumentation. Measurement system identification, engineering units and computed parameters

International Nuclear Information System (INIS)

Sarkar, J.; Liebert, J.; Laeufer, R.

1992-11-01

This updated version of the previous report /1/ contains, besides additional instrumentation needed for 2D/3D Programme, the supplementary instrumentation in the inlet plenum of SG simulator and hot and cold leg of broken loop, the cold leg of intact loops and the upper plenum to meet the requirements (Test Phase A) of the UPTF Programme, TRAM, sponsored by the Federal Minister of Research and Technology (BMFT) of the Federal Republic of Germany. For understanding, the derivation and the description of the identification codes for the entire conventional and advanced measurement systems classifying the function, and the equipment unit, key, as adopted in the conventional power plants, have been included. Amendments have also been made to the appendices. In particular, the list of measurement systems covering the measurement identification code, instrument, measured quantity, measuring range, band width, uncertainty and sensor location has been updated and extended to include the supplementary instrumentation. Beyond these amendments, the uncertainties of measurements have been precisely specified. The measurement identification codes which also stand for the identification of the corresponding measured quantities in engineering units and the identification codes derived therefrom for the computed parameters have been adequately detailed. (orig.)

Loop kinematics

International Nuclear Information System (INIS)

Migdal, A.A.

1982-01-01

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Pressure field study of the Tevatron cold compressors

International Nuclear Information System (INIS)

Klebaner, A.L.; Martinez, A.; Soyars, W.M.; Theilacker, J.C.; Fermilab

2003-01-01

The Fermilab Tevatron cryogenic system utilizes high-speed centrifugal cold compressors, manufactured by Ishikawajima-Harima Heavy Industries Co. Ltd. (IHI), for high-energy operations [1]. The compressor is designed to pump 60 g/sec of 3.6 K saturated helium vapor at a pressure ratio of 2.8, with an off-design range of 40 to 70 g/sec. Operating speeds are between 40 and 95 krpm, with a speed of 80 krpm at the design point. Different heat loads and magnet quench performance of each of the twenty-four satellite refrigerators dictates different process pressure and flow rates of the cold compressors. Reducing the process flow rate can cause the centrifugal cold compressor to stop pumping and subsequently surge. Tests have been conducted at the Cryogenic Test Facility at Fermilab to map the pressure field and appropriate efficiency of the IHI hydrodynamic cold compressor. The information allows tuning of each of the twenty-four Tevatron satellite refrigerators to avoid cold compressor operation near the surge and choke lines. A new impeller has also been tested. The Tevatron cold compressor pressure field and efficiency data with the new impeller are presented in this paper