WorldWideScience

Sample records for heat shield testing

  1. Orion Heat Shield Manufacturing Producibility Improvements for the EM-1 Flight Test Program

    Science.gov (United States)

    Koenig, William J.; Stewart, Michael; Harris, Richard F.

    2018-01-01

    This paper describes how the ORION program is incorporating improvements in the heat shield design and manufacturing processes reducing programmatic risk and ensuring crew safety in support of NASA's Exploration missions. The approach for the EFT-1 heat shield utilized a low risk Apollo heritage design and manufacturing process using an Avcoat TPS ablator with a honeycomb substrate to provide a one piece heat shield to meet the mission re-entry heating environments. The EM-1 mission will have additional flight systems installed to fly to the moon and return to Earth. Heat shield design and producibility improvements have been incorporated in the EM-1 vehicle to meet deep space mission requirements. The design continues to use the Avcoat material, but in a block configuration to enable improvements in consistant and repeatable application processes using tile bonding experience developed on the Space Shuttle Transportation System Program.

  2. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    International Nuclear Information System (INIS)

    McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

    1986-02-01

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior

  3. Multifunctional Hot Structure Heat Shield

    Data.gov (United States)

    National Aeronautics and Space Administration — This project is performing preliminary development of a Multifunctional Hot Structure (HOST) heat shield for planetary entry. Results of this development will...

  4. BWR spent fuel storage cask performance test. Volume 2. Pre- and post-test decay heat, heat transfer, and shielding analyses

    International Nuclear Information System (INIS)

    Wiles, L.E.; Lombardo, N.J.; Heeb, C.M.; Jenquin, U.P.; Michener, T.E.; Wheeler, C.L.; Creer, J.M.; McCann, R.A.

    1986-06-01

    This report describes the decay heat, heat transfer, and shielding analyses conducted in support of performance testing of a Ridhihalgh, Eggers and Associates REA 2033 boiling water reactor (BWR) spent fuel storage cask. The cask testing program was conducted for the US Department of Energy (DOE) Commercial Spent Fuel Management Program by the Pacific Northwest Laboratory (PNL) and by General Electric at the latters' Morris Operation (GE-MO) as reported in Volume I. The analyses effort consisted of performing pretest calculations to (1) select spent fuel for the test; (2) symmetrically load the spent fuel assemblies in the cask to ensure lateral symmetry of decay heat generation rates; (3) optimally locate temperature and dose rate instrumentation in the cask and spent fuel assemblies; and (4) evaluate the ORIGEN2 (decay heat), HYDRA and COBRA-SFS (heat transfer), and QAD and DOT (shielding) computer codes. The emphasis of this second volume is on the comparison of code predictions to experimental test data in support of the code evaluation process. Code evaluations were accomplished by comparing pretest (actually pre-look, since some predictions were not completed until testing was in progress) predictions with experimental cask testing data reported in Volume I. No attempt was made in this study to compare the two heat transfer codes because results of other evaluations have not been completed, and a comparison based on one data set may lead to erroneous conclusions

  5. Thermophysical Properties of Heat Resistant Shielding Material

    International Nuclear Information System (INIS)

    Porter, W.D.

    2004-01-01

    This project was aimed at determining thermal conductivity, specific heat and thermal expansion of a heat resistant shielding material for neutron absorption applications. These data are critical in predicting the structural integrity of the shielding under thermal cycling and mechanical load. The measurements of thermal conductivity and specific heat were conducted in air at five different temperatures (-31 F, 73.4 F, 140 F, 212 F and 302 F). The transient plane source (TPS) method was used in the tests. Thermal expansion tests were conducted using push rod dilatometry over the continuous range from -40 F (-40 C) to 302 F (150 C)

  6. Heating profiles on ICRF antenna Faraday shields

    International Nuclear Information System (INIS)

    Taylor, D.J.; Baity, F.W.; Hahs, C.L. Riemer, B.W.; Ryan, D.M.; Williamson, D.E.

    1992-01-01

    Poor definition of the heating profiles that occur during normal operation of Faraday shields for ion cyclotron resonant frequency (ICRF) antennas has complicated the mechanical design of ICRF system components. This paper reports that at Oak Ridge National Laboratory (ORNL), Faraday shield analysis is being used in defining rf heating profiles. In recent numerical analyses of proposed hardware for the Burning Plasma Experiment (BPX) and DIII-D, rf magnetic fields at Faraday shield surfaces were calculated, providing realistic predictions of the induced skin currents flowing on the shield elements and the resulting dissipated power profile. Detailed measurements on mock-ups of the Faraday shields for DIII-D and the Tokamak Fusion Test Reactor (TFTR) confirmed the predicted magnetic field distributions. A conceptual design for an uncooled Faraday shield for the BPX ion cyclotron resonance heating (ICRH) antenna, which should withstand the proposed long-pulse operation, has been completed. The analytical effort is described in detail, with emphasis on the design work for the BPX ICRH antenna conceptual design and for the replacement Faraday shield for the DIII-D FWCD antenna. Results of analyses are shown, and configuration issues involved in component modeling are discussed

  7. Shielding benchmark test

    International Nuclear Information System (INIS)

    Kawai, Masayoshi

    1984-01-01

    Iron data in JENDL-2 have been tested by analyzing shielding benchmark experiments for neutron transmission through iron block performed at KFK using CF-252 neutron source and at ORNL using collimated neutron beam from reactor. The analyses are made by a shielding analysis code system RADHEAT-V4 developed at JAERI. The calculated results are compared with the measured data. As for the KFK experiments, the C/E values are about 1.1. For the ORNL experiments, the calculated values agree with the measured data within an accuracy of 33% for the off-center geometry. The d-t neutron transmission measurements through carbon sphere made at LLNL are also analyzed preliminarily by using the revised JENDL data for fusion neutronics calculation. (author)

  8. Core test reactor shield cooling system analysis

    International Nuclear Information System (INIS)

    Larson, E.M.; Elliott, R.D.

    1971-01-01

    System requirements for cooling the shield within the vacuum vessel for the core test reactor are analyzed. The total heat to be removed by the coolant system is less than 22,700 Btu/hr, with an additional 4600 Btu/hr to be removed by the 2-inch thick steel plate below the shield. The maximum temperature of the concrete in the shield can be kept below 200 0 F if the shield plug walls are kept below 160 0 F. The walls of the two ''donut'' shaped shield segments, which are cooled by the water from the shield and vessel cooling system, should operate below 95 0 F. The walls of the center plug, which are cooled with nitrogen, should operate below 100 0 F. (U.S.)

  9. Heat transfer pipe shielding device for heat exchanger

    International Nuclear Information System (INIS)

    Hanawa, Jun.

    1991-01-01

    The front face and the rear face of a frame that surrounds the circumference of the water chamber body of a multi-tube heat exchanger are covered by a rotational shielding plate. A slit is radially formed to the shielding plate for the insertion of a probe or cleaner to the heat transfer pipe and a deflector is disposed on the side opposite to the slit. The end of the heat transfer pipe to be inspected is exposed to the outer side by way of the slit by the rotation of the shielding plate, and the probe or cleaner is inserted in the heat transfer pipe to conduct an eddy current injury monitoring test or cleaning. The inside of the water chamber and the heat transfer pipe is exhausted by a ventilation nozzle disposed to the frame. Accordingly, a shielding effect upon inspection and cleaning can be obtained and, in addition, inspection and exhaustion at the cleaning position can be conducted easily. Since the operation for attachment and detachment is easy, the effect of reducing radiation dose per unit can be obtained by the shortening of the operation time. (N.H.)

  10. Compression Pad Cavity Heating Augmentation on Orion Heat Shield

    Science.gov (United States)

    Hollis, Brian R.

    2011-01-01

    An experimental study has been conducted to assess the effects of compression pad cavities on the aeroheating environment of the Project Orion Crew Exploration Vehicle heat shield. Testing was conducted in Mach 6 and 10 perfect-gas wind tunnels to obtain heating measurements in and around the compression pads cavities using global phosphor thermography. Data were obtained over a wide range of Reynolds numbers that produced laminar, transitional, and turbulent flow within and downstream of the cavities. The effects of cavity dimensions on boundary-layer transition and heating augmentation levels were studied. Correlations were developed for transition onset and for the average cavity-heating augmentation.

  11. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.H.

    1990-03-01

    The GA-4 and GA-9 spent fuel shipping casks employ a solid neutron shielding material. During a hypothetical thermal accident, any combustion of the neutron shield must not compromise the ability of the cask to contain the radioactive contents. A two-phase thermal testing program was carried out to assist in selecting satisfactory shielding materials. In the first phase, small-scale screening tests were performed on nine candidate materials using ASTM procedures. From these initial results, three of the nine candidates were chosen for inclusion in the second phase of testing, These materials were Bisco Products NS-4-FR, Reactor Experiments 201-1, and Reactor Experiments 207. In the second phase, each selected material was fabricated into a test article which simulated a full-scale of neutron shield from the cask. The test article was heated in an environmental prescribed by NRC regulations. Results of this second testing phase showed that all three materials are thermally acceptable

  12. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.N.

    1990-01-01

    The GA-4 and GA-9 spent fuel shipping casks employ a solid neutron shielding material. During a hypothetical thermal accident, any combustion of the neutron shield must not compromise the ability of the cask to contain the radioactive contents. A two-phase thermal testing program was carried out to assist in selecting satisfactory shielding materials. In the first phase, small-scale screening tests were performed on nine candidate materials using ASTM procedures. From these initial results, three of the nine candidates were chosen for inclusion in the second phase of testing. These materials were Bisco Products NS-4-FR, Reactor Experiments 201-1, and Reactor Experiments 207. In the second phase, each selected material was fabricated into a test article which simulated a full-scale section of neutron shield from the cask. The test article was heated in an environment prescribed by NRC regulations. Results of this second testing phase show that all three materials are thermally acceptable

  13. Mars Exploration Rover Heat Shield Recontact Analysis

    Science.gov (United States)

    Raiszadeh, Behzad; Desai, Prasun N.; Michelltree, Robert

    2011-01-01

    The twin Mars Exploration Rover missions landed successfully on Mars surface in January of 2004. Both missions used a parachute system to slow the rover s descent rate from supersonic to subsonic speeds. Shortly after parachute deployment, the heat shield, which protected the rover during the hypersonic entry phase of the mission, was jettisoned using push-off springs. Mission designers were concerned about the heat shield recontacting the lander after separation, so a separation analysis was conducted to quantify risks. This analysis was used to choose a proper heat shield ballast mass to ensure successful separation with low probability of recontact. This paper presents the details of such an analysis, its assumptions, and the results. During both landings, the radar was able to lock on to the heat shield, measuring its distance, as it descended away from the lander. This data is presented and is used to validate the heat shield separation/recontact analysis.

  14. Highly heat removing radiation shielding material

    International Nuclear Information System (INIS)

    Asano, Norio; Hozumi, Masahiro.

    1990-01-01

    Organic materials, inorganic materials or metals having excellent radiation shielding performance are impregnated into expanded metal materials, such as Al, Cu or Mg, having high heat conductivity. Further, the porosity of the expanded metals and combination of the expanded metals and the materials to be impregnated are changed depending on the purpose. Further, a plurality of shielding materials are impregnated into the expanded metal of the same kind, to constitute shielding materials. In such shielding materials, impregnated materials provide shielding performance against radiation rays such as neutrons and gamma rays, the expanded metals provide heat removing performance respectively and they act as shielding materials having heat removing performance as a whole. Accordingly, problems of non-informity and discontinuity in the prior art can be dissolved be provide materials having flexibility in view of fabrication work. (T.M.)

  15. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.H.

    1993-01-01

    In May-June 1989 the first series of full-scale thermal tests was performed on three shielding materials: Bisco Products NS-4-FR, and Reactor Experiments RX-201 and RX-207. The tests are described in Thermal Testing of Solid Neutron Shielding Materials, GA-A19897, R.H. Boonstra, General Atomics (1990), and demonstrated the acceptability of these materials in a thermal accident. Subsequent design changes to the cask rendered these materials unattractive in terms of weight or adequate service temperature margin. For the second test series a material specification was developed for a polypropylene based neutron shield with a softening point of at least 280degF. Table 1 lists the neutron shield materials tested. The Envirotech and Bisco materials are not polypropylene, but were tested as potential backup materials in the event that a satisfactory polypropylene could not be found. The Bisco modified NS-4 and Reactor Experiments HMPP are both acceptable materials from a thermal accident standpoint for use in the shipping cask. Tests of the Kobe PP-R01 and Envirotech HDPE were stopped for safety reasons, due to inability to deal with the heavy smoke, before completion of the 30-minute heating phase. However these materials may prove satisfactory if they could undergo the complete heating. (J.P.N.)

  16. Heating profiles on ICRF antenna Faraday shields

    International Nuclear Information System (INIS)

    Taylor, D.J.; Baity, F.W.; Hahs, C.L.; Riemer, B.W.; Ryan, P.M.; Williamson, D.E.

    1991-01-01

    A conceptual design for an uncooled Faraday shield for the BPX ion cyclotron resonance heating (ICRH) antenna, which should withstand the proposed long-pulse operation, has been completed. A high-heat-flux, uncooled Faraday shield has also been designed for the fast-wave current drive (FWCD) antenna on D3-D. For both components, the improved understanding of the heating profiles made it possible to design for heat fluxes that would otherwise have been too close to mechanically established limits. The analytical effort is described in detail, with emphasis on the design work for the BPX ICRH antenna conceptual design and for the replacement Faraday shield for the D3-D FWCD antenna. Results of analyses are shown, and configuration issues involved in component modeling are discussed. 3 refs., 6 figs., 2 tabs

  17. Space Shuttle Orbiter AFT heat shield seal

    Science.gov (United States)

    Walkover, L. J.

    1979-01-01

    The evolution of the orbiter aft heat shield seal (AHSS) design, which involved advancing mechanical seal technology in severe thermal environment is discussed. The baseline design, various improvements for engine access, and technical problem solution are presented. It is a structure and mechanism at the three main propulsion system (MPS) engine interfaces to the aft compartment structure. Access to each MPS engine requires disassembly and removal of the AHSS. Each AHSS accommodates the engine movement, is exposed to an extremely high temperature environment, and is part of the venting control of the aft compartment.

  18. Castor-1C spent fuel storage cask decay heat, heat transfer, and shielding analyses

    International Nuclear Information System (INIS)

    Rector, D.R.; McCann, R.A.; Jenquin, U.P.; Heeb, C.M.; Creer, J.M.; Wheeler, C.L.

    1986-12-01

    This report documents the decay heat, heat transfer, and shielding analyses of the Gesellschaft fuer Nuklear Services (GNS) CASTOR-1C cask used in a spent fuel storage demonstration performed at Preussen Elektra's Wurgassen nuclear power plant. The demonstration was performed between March 1982 and January 1984, and resulted in cask and fuel temperature data and cask exterior surface gamma-ray and neutron radiation dose rate measurements. The purpose of the analyses reported here was to evaluate decay heat, heat transfer, and shielding computer codes. The analyses consisted of (1) performing pre-look predictions (predictions performed before the analysts were provided the test data), (2) comparing ORIGEN2 (decay heat), COBRA-SFS and HYDRA (heat transfer), and QAD and DOT (shielding) results to data, and (3) performing post-test analyses if appropriate. Even though two heat transfer codes were used to predict CASTOR-1C cask test data, no attempt was made to compare the two codes. The codes are being evaluated with other test data (single-assembly data and other cask data), and to compare the codes based on one set of data may be premature and lead to erroneous conclusions

  19. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.H.

    1992-09-01

    Two legal-weight truck casks the GA-4 and GA-9, will carry four PWR and nine BWR spent fuel assemblies, respectively. Each cask has a solid neutron shielding material separating the steel body and the outer steel skin. In the thermal accident specified by NRC regulations in 10CFR Part 71, the cask is subjected to an 800 degree C environment for 30 minutes. The neutron shield need not perform any shielding function during or after the thermal accident, but its behavior must not compromise the ability of the cask to contain the radioactive contents. In May-June 1989 the first series of full-scale thermal tests was performed on three shielding materials: Bisco Products NS-4-FR, and Reactor Experiments RX-201 and RX-207. The tests are described in Thermal Testing of Solid Neutron Shielding Materials, GA-AL 9897, R. H. Boonstra, General Atomics (1990), and demonstrated the acceptability of these materials in a thermal accident. Subsequent design changes to the cask rendered these materials unattractive in terms of weight or adequate service temperature margin. For the second test series, a material specification was developed for a polypropylene based neutron shield with a softening point of at least 280 degree F. The neutron shield materials tested were boronated (0.8--4.5%) polymers (polypropylene, HDPE, NS-4). The Envirotech and Bisco materials are not polypropylene, but were tested as potential backup materials in the event that a satisfactory polypropylene could not be found

  20. Heating Augmentation Due to Compression Pad Cavities on the Project Orion CEV Heat Shield

    Science.gov (United States)

    Hollis, Brian R.

    2009-01-01

    An experimental study has been conducted to assess the effects of compression pad cavities on the aeroheating environment of the Project Orion CEV heat-shield. Testing was conducted in Mach 6 and Mach 10 perfect-gas wind tunnels to obtain heating measurements in and around the compression pads cavities using global phosphor thermography. Data were obtained over a wide range of Reynolds numbers that produced laminar, transitional, and turbulent flow within and downstream of the cavities. The effects of cavity dimensions on boundary-layer transition and heating augmentation levels were studied. Correlations were developed for transition onset and for the average cavity-heating augmentation.

  1. Seismic proof test of shielding block walls

    International Nuclear Information System (INIS)

    Ohte, Yukio; Watanabe, Takahide; Watanabe, Hiroyuki; Maruyama, Kazuhide

    1989-01-01

    Most of the shielding block walls used for building nuclear facilities are built by dry process. When a nuclear facility is designed, seismic waves specific at each site are set as input seismic motions and they are adopted in the design. Therefore, it is necessary to assure safety of the shielding block walls for earthquake by performing anti-seismic experiments under the conditions at each site. In order to establish the normal form that can be applied to various seismic conditions in various areas, Shimizu Corp. made an actual-size test samples for the shielding block wall and confirmed the safety for earthquake and validity of normalization. (author)

  2. Bremsstrahlung converter debris shields: test and analysis

    International Nuclear Information System (INIS)

    Reedy, E.D. Jr.; Perry, F.C.

    1983-10-01

    Electron beam accelerators are commonly used to create bremsstrahlung x-rays for effects testing. Typically, the incident electron beam strikes a sandwich of three materials: (1) a conversion foil, (2) an electron scavenger, and (3) a debris shield. Several laboratories, including Sandia National Laboratories, are developing bremsstrahlung x-ray sources with much larger test areas (approx. 200 to 500 cm 2 ) than ever used before. Accordingly, the debris shield will be much larger than before and subject to loads which could cause shield failure. To prepare for this eventuality, a series of tests were run on the Naval Surface Weapons Center's Casino electron beam accelerator (approx. 1 MeV electrons, 100 ns FWHM pulse, 45 kJ beam energy). The primary goal of these tests was to measure the stress pulse which loads a debris shield. These measurements were made with carbon gages mounted on the back of the converter sandwich. At an electron beam fluence of about 1 kJ/cm 2 , the measured peak compressive stress was typically in the 1 to 2 kbar range. Measured peak compressive stress scaled in a roughly linear manner with fluence level as the fluence level was increased to 10 kJ/cm 2 . The duration of the compressive pulse was on the order of microseconds. In addition to the stress wave measurements, a limited number of tests were made to investigate the type of damage generated in several potential shield materials

  3. Evaluation of Heat Shields from RTS Wright Industries Magnesium and Uranium Beds

    CERN Document Server

    Korinko, P S

    2002-01-01

    Heat shields from a factory test of the furnaces that will be used to heat the magnesium and uranium beds for the tritium extraction facility (TEF) were examined to determine the cause of discoloration. The samples were examined using visual, optical microscopy, electron microscopy, x-ray spectroscopy, and Auger electron spectroscopy.

  4. Regolith Derived Heat Shield for Planetary Body Entry and Descent System with In Situ Fabrication

    Science.gov (United States)

    Hogue, Michael D.; Meuller, Robert P.; Sibille, Laurent; Hintze, Paul E.; Rasky, Daniel J.

    2012-01-01

    . Platinum, titanium, helium 3, and other metals, elements and minerals are all high-value commodities in limited supply on Earth, and it may be profitable to mine these substances throughout the Solar System and return them to Earth, if an economical method can be found. To date, several private corporations have been launched to pursue these goals. Because the heat shield is the last element to be used in an Earth-return mission, a high penalty is paid in the propellant mass required to carry the heat shield to the destination and back. If the heat shield could be manufactured in space, and then outfitted on the spacecraft prior to the reentry at Earth, then significant propellant and mass savings could be achieved during launch and space operations. Preliminary mission architecture scenarios are described, which explain the potential benefits that may be derived from using an in-situ fabricated regolith heat shield. In order to prove that this is a feasible technology concept, this project successfully fabricated heat shield materials from mineral simulant materials of lunar and Martian regolith by two methods: 1) Sintering and 2) Binding the simulant with a "room-temperature vulcanizing" (RTV) silicone formulated to withstand high temperatures. Initially a third type of fabrication was planned using the hot waste stream from regolith ISRU processes. This fabrication method was discarded since the resulting samples would be too dense and brittle for heat shields. High temperature flame tests at KSC and subsequent arc jet tests at Ames Research Center (ARC) have proved promising. These coupon tests show favorable materials properties and have the potential to be a new way of fabricating heat shields for space entry into planetary atmospheres.

  5. Superconductor shields test chamber from ambient magnetic fields

    Science.gov (United States)

    Hildebrandt, A. F.

    1965-01-01

    Shielding a test chamber for magnetic components enables it to maintain a constant, low magnetic field. The chamber is shielded from ambient magnetic fields by a lead foil cylinder maintained in a superconducting state by liquid helium.

  6. The clinical testing of male gonad shields. Technical report

    International Nuclear Information System (INIS)

    Church, W.W.; Burnett, B.M.

    1975-11-01

    Two types of male gonad shields, designed for use with support garments, were tested in a number of hospitals and clinics throughout the United States. The clinical evaluation consisted of: (1) measuring dose reduction with thermoluminescent dosimeters; and (2) determining acceptability of the shields for routine use in x-ray facilities, through the use of survey forms completed by patients, technologists, and facilities. The shields proved to provide a basis for a very satisfactory male gonad shield program

  7. The clinical testing of male gonad shields. Technical report

    Energy Technology Data Exchange (ETDEWEB)

    Church, W.W.; Burnett, B.M.

    1975-11-01

    Two types of male gonad shields, designed for use with support garments, were tested in a number of hospitals and clinics throughout the United States. The clinical evaluation consisted of: (1) measuring dose reduction with thermoluminescent dosimeters; and (2) determining acceptability of the shields for routine use in x-ray facilities, through the use of survey forms completed by patients, technologists, and facilities. The shields proved to provide a basis for a very satisfactory male gonad shield program. (GRA)

  8. Shielding benchmark tests of JENDL-3

    International Nuclear Information System (INIS)

    Kawai, Masayoshi; Hasegawa, Akira; Ueki, Kohtaro; Yamano, Naoki; Sasaki, Kenji; Matsumoto, Yoshihiro; Takemura, Morio; Ohtani, Nobuo; Sakurai, Kiyoshi.

    1994-03-01

    The integral test of neutron cross sections for major shielding materials in JENDL-3 has been performed by analyzing various shielding benchmark experiments. For the fission-like neutron source problem, the following experiments are analyzed: (1) ORNL Broomstick experiments for oxygen, iron and sodium, (2) ASPIS deep penetration experiments for iron, (3) ORNL neutron transmission experiments for iron, stainless steel, sodium and graphite, (4) KfK leakage spectrum measurements from iron spheres, (5) RPI angular neutron spectrum measurements in a graphite block. For D-T neutron source problem, the following two experiments are analyzed: (6) LLNL leakage spectrum measurements from spheres of iron and graphite, and (7) JAERI-FNS angular neutron spectrum measurements on beryllium and graphite slabs. Analyses have been performed using the radiation transport codes: ANISN(1D Sn), DIAC(1D Sn), DOT3.5(2D Sn) and MCNP(3D point Monte Carlo). The group cross sections for Sn transport calculations are generated with the code systems PROF-GROUCH-G/B and RADHEAT-V4. The point-wise cross sections for MCNP are produced with NJOY. For comparison, the analyses with JENDL-2 and ENDF/B-IV have been also carried out. The calculations using JENDL-3 show overall agreement with the experimental data as well as those with ENDF/B-IV. Particularly, JENDL-3 gives better results than JENDL-2 and ENDF/B-IV for sodium. It has been concluded that JENDL-3 is very applicable for fission and fusion reactor shielding analyses. (author)

  9. Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

    International Nuclear Information System (INIS)

    Ware, A.G.; Longhurst, G.R.

    1981-12-01

    This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available

  10. Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

    Energy Technology Data Exchange (ETDEWEB)

    Ware, A.G.; Longhurst, G.R.

    1981-12-01

    This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available.

  11. PEP radiation shielding tests in SLAC A Beam

    International Nuclear Information System (INIS)

    Ash, W.; DeStaebler, H.; Harris, J.; Jenkins, T.; Murray, J.

    1977-09-01

    Radiation shielding tests designed to simulate possible conditions in and around the PEP experimental halls were conducted. The SLAC A Beam was targeted in the block tunnel at a point about midway between End Station A and Beam Dump East. At that site it was relatively easy to rearrange the concrete block structure to simulate the various shielding configurations under consideration for PEP. Extensive surveys of neutron and ionizing radiation were made. Complete results of the shielding tests are given

  12. Effects of polarization-charge shielding in microwave heating

    Energy Technology Data Exchange (ETDEWEB)

    Lin, M. S.; Lin, S. M.; Chiang, W. Y.; Barnett, L. R.; Chu, K. R., E-mail: krchu@yahoo.com.tw [Department of Physics, National Taiwan University, Taipei 106, Taiwan (China)

    2015-08-15

    Heating of dielectric objects by radio frequency (RF) and microwaves has long been a method widely employed in scientific research and industrial applications. However, RF and microwave heating are often susceptible to an excessive temperature spread due to uneven energy deposition. The current study elucidates an important physical reason for this difficulty and proposes an effective remedy. Non-spherical samples are placed in an anechoic chamber, where it is irradiated by a traveling microwave wave with 99% intensity uniformity. Polarization charges induced on the samples tend to partially cancel the incident electric field and hence reduce the heating rate. The polarization-charge shielded heating rate is shown to be highly dependent on the sample's shape and its orientation relative to the wave electric field. For samples with a relatively high permittivity, the resultant uneven heating can become a major cause for the excessive temperature spread. It is also demonstrated that a circularly polarized wave, with its rapidly rotating electric field, can effectively even out the heating rate and hence the temperature spread.

  13. Manufacture and testing of the CTB&SBB thermal shield for the ITER magnet feeder system

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Kun; Liu, Chen; Song, Yuntao; Feng, Hansheng; Ding, Kaizhong, E-mail: kzding@ipp.ac.cn; Wang, Tanbin; Ji, Hui

    2015-10-15

    The system of International Thermonuclear Experimental Reactor (ITER) feeders is responsible for the power, helium cooling, and instrumentation of the magnets of the coil terminal box and S-bend box (CTB&SBB) thermal shield outside the cryostat. An 80-K rectangular Al thermal shield is hung inside the CTB&SBB to reduce the thermal radiation heat loads of 4.5-K helium. The American Society of Interventional Pain Physicians (ASIPP) will supply all the 31 sets of feeders for ITER. A manufactured prototype of CTB&SBB thermal shield is first quality-tested before the commencement of the series production. First, a detailed configuration of the rectangular Al thermal shield is presented in this article. The paper also presents more information on the manufacturing process of the thermal shield, especially the welding process, the procedure for ensuring good weld quality, and the use of a specially designed tool to ensure <5-mm deformation on such a 7.3-m-long thermal shield during welding. In addition, the cold test and results, including the cooling process with 13-bar and 17.5-g/s 80-K He gas, and the temperature distribution on different panels of the thermal shield are presented. The whole process of manufacture and testing lays a good foundation for the series production of the thermal shield.

  14. Thermal stress evaluation of the Viking RTG heat shield

    International Nuclear Information System (INIS)

    Stadter, J.T.; Weiss, R.O.

    1976-03-01

    Thermal stress analyses of the Viking RTG heat shield are presented. The primary purpose of the analyses was to determine the effects of the end cap and the finite length of the heat shield on the peak tensile stress in the barrel wall. The SAAS III computer code was used to calculate the thermal stresses; axisymmetric and plane section analyses were performed for a variety of temperature distributions. The study consisted of three parts. In the first phase, the influence of the end cap on the barrel wall stresses was examined by parametrically varying the modulus of elasticity of the contact zone between the end cap and the barrel. The second phase was concerned with stresses occurring as a result of an orbital decay reentry trajectory, and the effects of the magnitude and shape of the axial temperature gradient. The final part of the study was concerned with the circumferentially nonuniform temperature distribution which develops during a side-on stable reentry. The last part includes a comparison of stresses generated for a hexagonal cross section with those generated for a circular cross section

  15. Evaluation of alternative methods of simulating asymmetric bulk heating in fusion reactor blanket/shield components

    International Nuclear Information System (INIS)

    Deis, G.A.; Longhurst, G.R.; Miller, L.G.; Wadkins, R.P.; Wessol, D.E.

    1981-10-01

    As a part of Phase O, Test Program Element-II of the Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program, a study was conducted by EG and G Idaho, Inc., to identify, characterize, and recommend alternative approaches for simulating fusion bulk heating in blanket/shield components. This is the report on that effort. Since the usefulness of any simulation approach depends upon the particular experiment considered, classes of problem types (thermal-hydraulic, thermomechanical, etc.) and material types (structure, solid breeder, etc.) are developed. The evaluation of the various simulation approaches is performed for the various significant combinations of problem class and material class. The simulation approaches considered are discrete-source heating, direct resistance, electromagnetic induction, microwave heating, and nuclear heating. From the evaluations performed for each experiment type, discrete - source heating emerges as a good approach for bulk heating simulation in thermal - hydraulics experiments, and nuclear heating appears to be a good approach in experiments addressing thermomechanics and combined thermal-hydraulic/thermomechanics

  16. Heat shield manifold system for a midframe case of a gas turbine engine

    Energy Technology Data Exchange (ETDEWEB)

    Mayer, Clinton A.; Eng, Jesse; Schopf, Cheryl A.

    2017-07-25

    A heat shield manifold system for an inner casing between a compressor and turbine assembly is disclosed. The heat shield manifold system protects the outer case from high temperature compressor discharge air, thereby enabling the outer case extending between a compressor and a turbine assembly to be formed from less expensive materials than otherwise would be required. In addition, the heat shield manifold system may be configured such that compressor bleed air is passed from the compressor into the heat shield manifold system without passing through a conventional flange to flange joint that is susceptible to leakage.

  17. Magnetic shielding tests for MFTF-B neutral beamlines

    International Nuclear Information System (INIS)

    Kerns, J.; Fabyan, J.; Wood, R.; Koger, P.

    1983-01-01

    A test program to determine the effectiveness of various magnetic shielding designs for MFTF-B beamlines was established at Lawrence Livermore National Laboratory (LLNL). The proposed one-tenth-scale shielding-design models were tested in a uniform field produced by a Helmholtz coil pair. A similar technique was used for the MFTF source-injector assemblies, and the model test results were confirmed during the Technology Demonstration in 1982. The results of these tests on shielding designs for MFTF-B had an impact on the beamline design for MFTF-B. The iron-core magnet and finger assembly originally proposed were replaced by a simple, air-core, race-track-coil, bending magnet. Only the source injector needs to be magnetically shielded from the fields of approximately 400 gauss

  18. Gamma-ray shielding design and performance test of WASTEF

    International Nuclear Information System (INIS)

    Matsumoto, Seiichiro; Aoyama, Saburo; Tashiro, Shingo; Nagai, Shiro

    1984-06-01

    The Waste Safety Testing Facility (WASTEF) was planned in 1978 to test the safety performance of HLW vitrified forms under the simulated conditions of long term storage and disposal, and completed in August 1981. The designed feature of the facility is to treat the vitrified forms contain actual high-level wastes of 5 x 10 4 Ci in maximum with 5 units of concrete shilded hot cells (3 units : Bate-Gamma cells, 2 units : Alpha-Gamma cells) and one units of Alpha-Gamma lead shielded cell, and to store radioactivity of 10 6 Ci in maximum. The safety performance of this facility is fundamentally maintained with confinement of radioactivity and shielding of the radiation. This report describes the method of gamma-ray shielding design, evaluation of the shielding test performed by using sealded gamma-ray sources(Co-60). (author)

  19. Fabrication and testing of small scale mock-ups of ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Hatano, Toshihisa; Sato, Satoshi; Suzuki, Satoshi; Yokoyama, Kenji; Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Takatsu, Hideyuki; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1998-12-01

    Small scale mock-ups of the primary first wall, the baffle first wall, the shield block and a partial model for the edge of the primary first wall module were designed and fabricated incorporating most of the key design features of the ITER shielding blanket. All mock-ups featured the DSCu heat sink, the built-in SS coolant tubes within the heat sink and the SS shield block. CFC tiles was used as the protection armor for the baffle first wall mock-up. The small scale shield block mock-up, integrated with the first wall, was designed to have a poloidal curvature specified in the ITER design. Fabrication routes of mock-ups were decided based on the single step solid HIP of DSCu/DSCu, DSCu/SS and SS/SS reflecting the results of previous joining techniques development and testing. For attaching the CFC tiles onto DSCu heat sink in the fabrication of the baffle first wall mock-up, a two-step brazing was tried. All mock-ups and the partial model were successfully fabricated with a satisfactory dimensional accuracy. The small scale primary first wall mock-up was thermo-mechanically tested under high heat fluxes of 5-7 MW/m{sup 2} for 2500 cycles in total. Satisfactory heat removal performance and integrity of the mock-up against cyclic high heat flux loads were confirmed by measurement during the tests and destructive examination after the tests. Similar high heat flux tests were also performed with the small scale baffle first wall mock-up under 5-10 MW/m{sup 2} for 4500 cycles in total resulting in sufficient heat removal capability and integrity confirmed by measurements during the tests. (author)

  20. Fabrication and testing of small scale mock-ups of ITER shielding blanket

    International Nuclear Information System (INIS)

    Hatano, Toshihisa; Sato, Satoshi; Suzuki, Satoshi; Yokoyama, Kenji; Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Takatsu, Hideyuki; Ohara, Yoshihiro

    1998-12-01

    Small scale mock-ups of the primary first wall, the baffle first wall, the shield block and a partial model for the edge of the primary first wall module were designed and fabricated incorporating most of the key design features of the ITER shielding blanket. All mock-ups featured the DSCu heat sink, the built-in SS coolant tubes within the heat sink and the SS shield block. CFC tiles was used as the protection armor for the baffle first wall mock-up. The small scale shield block mock-up, integrated with the first wall, was designed to have a poloidal curvature specified in the ITER design. Fabrication routes of mock-ups were decided based on the single step solid HIP of DSCu/DSCu, DSCu/SS and SS/SS reflecting the results of previous joining techniques development and testing. For attaching the CFC tiles onto DSCu heat sink in the fabrication of the baffle first wall mock-up, a two-step brazing was tried. All mock-ups and the partial model were successfully fabricated with a satisfactory dimensional accuracy. The small scale primary first wall mock-up was thermo-mechanically tested under high heat fluxes of 5-7 MW/m 2 for 2500 cycles in total. Satisfactory heat removal performance and integrity of the mock-up against cyclic high heat flux loads were confirmed by measurement during the tests and destructive examination after the tests. Similar high heat flux tests were also performed with the small scale baffle first wall mock-up under 5-10 MW/m 2 for 4500 cycles in total resulting in sufficient heat removal capability and integrity confirmed by measurements during the tests. (author)

  1. Testing of the PELSHIE shielding code using Benchmark problems and other special shielding models

    International Nuclear Information System (INIS)

    Language, A.E.; Sartori, D.E.; De Beer, G.P.

    1981-08-01

    The PELSHIE shielding code for gamma rays from point and extended sources was written in 1971 and a revised version was published in October 1979. At Pelindaba the program is used extensively due to its flexibility and ease of use for a wide range of problems. The testing of PELSHIE results with the results of a range of models and so-called Benchmark problems is desirable to determine possible weaknesses in PELSHIE. Benchmark problems, experimental data, and shielding models, some of which were resolved by the discrete-ordinates method with the ANISN and DOT 3.5 codes, were used for the efficiency test. The description of the models followed the pattern of a classical shielding problem. After the intercomparison with six different models, the usefulness of the PELSHIE code was quantitatively determined [af

  2. Temperature distribution due to the heat generation in nuclear reactor shielding

    International Nuclear Information System (INIS)

    Torres, L.M.R.

    1985-01-01

    A study is performed for calculating nuclear heating due to the interaction of neutrons and gamma-rays with matter. Modifications were implemented in the ANISN and DOT 3.5 codes, that solve the transport equation using the discrete ordinate method, in one two-dimensions respectively, to include nuclear heating calculations in these codes. In order to determine the temperature distribution, using the finite difference method, a numerical model was developed for solving the heat conduction equation in one-dimension, in plane, cylindrical and spherical geometries, and in two-dimensions, X-Y and R-Z geometries. Based on these models, computer programs were developed for calculating the temperature distribution. Tests and applications of the implemented modifications were performed in problems of nuclear heating and temperature distribution due to radiation energy deposition in fission and fusion reactor shields. (Author) [pt

  3. Materials development for ITER shielding and test blanket in China

    Energy Technology Data Exchange (ETDEWEB)

    Chen, J.M., E-mail: Chenjm@swip.ac.cn [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wu, J.H.; Liu, X.; Wang, P.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wang, Z.H.; Li, Z.N. [Ningxia Orient Non-ferrous Metals Group Co. Ltd., P.O. Box 105, Shizuishan (China); Wang, X.S.; Zhang, P.C. [China Academy of Engineering Physics, P.O. Box 919-71, Mianyang 621900 (China); Zhang, N.M.; Fu, H.Y.; Liu, D.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China)

    2011-10-01

    China is a member of the ITER program and is developing her own materials for its shielding and test blanket modules. The materials include vacuum-hot-pressing (VHP) Be, CuCrZr alloy, 316L(N) and China low activation ferritic/martensitic (CLF-1) steels. Joining technologies including Be/Cu hot isostatic pressing (HIP) and electron beam (EB) weldability of 316L(N) were investigated. Chinese VHP-Be showed good properties, with BeO content and ductility that satisfy the ITER requirements. Be/Cu mock-ups were fabricated for Be qualification tests at simulated ITER vertical displacement event (VDE) and heat flux cycling conditions. Fine microstructure and good mechanical strength of the CuCrZr alloy were achieved by a pre-forging treatment, while the weldability of 316L(N) by EB was demonstrated for welding depths varying from 5 to 80 mm. Fine microstructure, high strength, and good ductility were achieved in CLF-1 steel by an optimized normalizing, tempering and aging procedure.

  4. Evaporation and vapor shielding of CFC targets exposed to plasma heat fluxes relevant to ITER ELMs

    International Nuclear Information System (INIS)

    Safronov, V.M.; Arkhipov, N.I.; Landman, I.S.; Pestchanyi, S.E.; Toporkov, D.A.; Zhitlukhin, A.M.

    2009-01-01

    Carbon fibre composite NB31 was tested at plasma gun facility MK-200UG by plasma heat fluxes relevant to Edge Localised Modes in ITER. The paper reports the results obtained on the evaporation threshold of carbon fibre composite, the velocity of carbon vapor motion along and across the magnetic field lines, and the parameters of carbon plasma such as temperature, density and ionization state. First experimental results on investigation of the vapor shield onset conditions are presented also. The obtained experimental data are compared with the results of numerical modeling.

  5. Experimental assessment on the thermal effects of the neutron shielding and heat-transfer fin of dual purpose casks on open pool fire

    International Nuclear Information System (INIS)

    Bang, Kyoung-Sik; Yu, Seung-Hwan; Lee, Ju-Chan; Seo, Ki-Seog; Choi, Woo-Seok

    2016-01-01

    Highlights: • An open pool fire test was performed to estimate not only the combustion effect of the neutron shielding but also the effect of the heat transfer fin of the dual purpose cask. • The heat transfer to the inside of the dual purpose cask was reduced, when the neutron shielding burns. • The surface temperatures are lower in the present of the heat transfer fins. • If inflammable material is used as the components of the cask, evaluating thermal integrity using the thermal test would be desirable. - Abstract: Dual purpose casks are used for storage and transport of spent nuclear fuel assemblies. They must therefore satisfy the requirements prescribed in the Korea Nuclear Safety Security Commission Act 2014-50, the IAEA Safety Standard Series No. SSR-6, and US 10 CFR Part 71. These regulatory guidelines classify the dual purpose cask as a Type B package and state that a Type B package must be able to withstand a temperature of 800 °C for a period of 30 min. NS-4-FR is used as neutron shielding of the dual purpose cask. Heat transfer fins are embedded to enhance heat transfer from the cask body to the outer-shell because the thermal conductivity of NS-4-FR is not good. However, accurately simulating not only the combustion effect of the neutron shielding but also the effect of the heat transfer fin in the thermal analysis is not easy. Therefore, an open pool fire test was conducted using a one-sixth slice of a real cask to estimate these effects at a temperature of 800 °C for a period of 30 min. The temperature at the central portion of the neutron shielding was lower when the neutron shielding in contact with the outer cask burned because the neutron shielding absorbed the surrounding latent heat as the neutron shielding burned. Therefore, the heat transfer to the inside of the dual purpose cask was reduced. The surface temperature was lower when a heat transfer fin was installed because the high heat generated by the flame was transferred to the

  6. Experimental assessment on the thermal effects of the neutron shielding and heat-transfer fin of dual purpose casks on open pool fire

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Kyoung-Sik, E-mail: nksbang@kaeri.re.kr; Yu, Seung-Hwan; Lee, Ju-Chan; Seo, Ki-Seog; Choi, Woo-Seok

    2016-08-01

    Highlights: • An open pool fire test was performed to estimate not only the combustion effect of the neutron shielding but also the effect of the heat transfer fin of the dual purpose cask. • The heat transfer to the inside of the dual purpose cask was reduced, when the neutron shielding burns. • The surface temperatures are lower in the present of the heat transfer fins. • If inflammable material is used as the components of the cask, evaluating thermal integrity using the thermal test would be desirable. - Abstract: Dual purpose casks are used for storage and transport of spent nuclear fuel assemblies. They must therefore satisfy the requirements prescribed in the Korea Nuclear Safety Security Commission Act 2014-50, the IAEA Safety Standard Series No. SSR-6, and US 10 CFR Part 71. These regulatory guidelines classify the dual purpose cask as a Type B package and state that a Type B package must be able to withstand a temperature of 800 °C for a period of 30 min. NS-4-FR is used as neutron shielding of the dual purpose cask. Heat transfer fins are embedded to enhance heat transfer from the cask body to the outer-shell because the thermal conductivity of NS-4-FR is not good. However, accurately simulating not only the combustion effect of the neutron shielding but also the effect of the heat transfer fin in the thermal analysis is not easy. Therefore, an open pool fire test was conducted using a one-sixth slice of a real cask to estimate these effects at a temperature of 800 °C for a period of 30 min. The temperature at the central portion of the neutron shielding was lower when the neutron shielding in contact with the outer cask burned because the neutron shielding absorbed the surrounding latent heat as the neutron shielding burned. Therefore, the heat transfer to the inside of the dual purpose cask was reduced. The surface temperature was lower when a heat transfer fin was installed because the high heat generated by the flame was transferred to the

  7. Thermal Analysis of a SHIELD Electromigration Test Structure

    Energy Technology Data Exchange (ETDEWEB)

    Benson, David A.; Bowman, Duane J.; Mitchell, Robert T.

    1999-05-01

    The steady state and transient thermal behavior of an electromigration test structure was analyzed. The test structure was a Sandia SHIELD (Self-stressing HIgh fregquency rELiability Device) electromigration test device manufactured by an outside vendor. This device has a high frequency oscillator circuit, a buffer circuit to isolate and drive the metal line to the tested (DUT), the DUT to be electromigrated itself, a metal resistance thermometry monitor, and a heater elment to temperature accelerate the electromigration effect.

  8. Dynamic Open-Rotor Composite Shield Impact Test Report

    Science.gov (United States)

    Seng, Silvia; Frankenberger, Charles; Ruggeri, Charles R.; Revilock, Duane M.; Pereira, J. Michael; Carney, Kelly S.; Emmerling, William C.

    2015-01-01

    The Federal Aviation Administration (FAA) is working with the European Aviation Safety Agency to determine the certification base for proposed new engines that would not have a containment structure on large commercial aircraft. Equivalent safety to the current fleet is desired by the regulators, which means that loss of a single fan blade will not cause hazard to the aircraft. NASA Glenn and Naval Air Warfare Center (NAWC) China Lake collaborated with the FAA Aircraft Catastrophic Failure Prevention Program to design and test a shield that would protect the aircraft passengers and critical systems from a released blade that could impact the fuselage. This report documents the live-fire test from a full-scale rig at NAWC China Lake. NASA provided manpower and photogrammetry expertise to document the impact and damage to the shields. The test was successful: the blade was stopped from penetrating the shield, which validates the design analysis method and the parameters used in the analysis. Additional work is required to implement the shielding into the aircraft.

  9. Tests of a thermal acoustic shield with a supersonic jet

    Science.gov (United States)

    Pickup, N.; Mangiarotty, R. A.; Okeefe, J. V.

    1981-10-01

    Fuel economy is a key element in the design of a future supersonic transport (SST). Variable cycle engines are being developed to provide the most economic combination of characteristics for a range of cruise speeds extending from subsonic speeds for overland flights to the supersonic cruise speeds. For one of these engines, the VCE-702, some form of noise suppression is needed for takeoff/sideline thrusts. The considered investigation is primarily concerned with scale model static tests of one particular concept for achieving that reduction, the thermal acoustic shield (TAS), which could also benefit other candidate SST engines. Other noise suppression devices being considered for SST application are the coannular nozzle, an internally ventilated nozzle, and mechanical suppressors. A test description is provided, taking into account the model configurations, the instrumentation, the test jet conditions, and aspects of screech noise control. Attention is given to shield thickness effects, a spectrum analysis, suppression and performance loss, and installed performance.

  10. Shielded regeneration heating element for a particulate filter

    Science.gov (United States)

    Gonze, Eugene V [Pinckney, MI; Ament, Frank [Troy, MI

    2011-01-04

    An exhaust system includes a particulate filter (PF) that is disposed downstream from an engine. The PF filters particulates within an exhaust from the engine. A heating element heats particulate matter in the PF. A catalyst substrate or a flow converter is disposed upstream from said heating element. The catalyst substrate oxidizes the exhaust prior to reception by the heating element. The flow converter converts turbulent exhaust flow to laminar exhaust flow prior to reception by the heating element.

  11. Phase 2 testing of ENDF/B-VI shielding data

    International Nuclear Information System (INIS)

    Ingersoll, D.T.; Wright, R.Q.; Slater, C.O.

    1992-01-01

    Version 6 of the US Evaluated Nuclear Data File (ENDF/B-VI) was released in early 1990 and is currently undergoing phase 2 testing. In Phase 2 testing, the evaluated data are approximately processed and used in an integral manner to predict the solution of previously specified benchmark experiments. Results are presented for the initial testing of several light elements and structural materials which are important for shielding applications. These initial tests indicate that the relatively subtle changes made to the iron data and the major modernization of the boron-11 data in Version 6 both represent significant and positive advancements in the quality of the evaluated data files

  12. New possibility of magnetic ripple shielding for specific heat measurements in hybrid magnets

    NARCIS (Netherlands)

    Tarnawski, Z.; Meulen, der, H. van; Franse, J.J.M.; Kadowaki, K.; Veenhuizen, P.A.; Klaasse, J.

    1988-01-01

    A test of the new high Tc superconducting materials for magnetic ripple shielding has been carried out. It was found that magnetic ripples of 0.0009 T (peak-to-peak) in the frequency range below 20 kHz can be completely shielded in high static fields by a 2 mm thick Y-Ba-Cu-O screen.

  13. Design and Testing of Improved Spacesuit Shielding Components

    International Nuclear Information System (INIS)

    Ware, J.; Ferl, J.; Wilson, J.W.; Clowdsley, M.S.; DeAngelis, G.; Tweed, J.; Zeitlin, C.J.

    2002-01-01

    In prior studies of the current Shuttle Spacesuit (SSA), where basic fabric lay-ups were tested for shielding capabilities, it was found that the fabric portions of the suit give far less protection than previously estimated due to porosity and non-uniformity of fabric and LCVG components. In addition, overall material transmission properties were less than optimum. A number of alternate approaches are being tested to provide more uniform coverage and to use more efficient materials. We will discuss in this paper, recent testing of new material lay-ups/configurations for possible use in future spacesuit designs

  14. Rotating shield ceiling for the compact ignition tokamak test cell

    International Nuclear Information System (INIS)

    Commander, J.C.

    1986-01-01

    For the next phase of the United States fusion program, a compact, high-field, toroidal ignition machine with liquid nitrogen cooled copper coils, designated the Compact Ignition Tokamak (CIT), is proposed. The CIT machine will be housed in a test cell with design features developed during preconceptual design. Configured as a right cylinder, the selected test cell design features: a test cell and basement with thick concrete shielding walls, and floor; leak tight tritium seals; and operational characteristics well suited to the circular CIT machine configuration and radially oriented ancillary equipment and systems

  15. Heat-shield for Extreme Entry Environment Technology (HEEET) Development Status

    Science.gov (United States)

    Venkatapathy, Ethiraj; Ellerby, Don; Gage, Peter

    2016-01-01

    The Heat shield for Extreme Entry Environment Technology (HEEET) Project is a NASA STMD and SMD co-funded effort. The goal is to develop and mission infuse a new ablative Thermal Protection System that can withstand extreme entry. It is targeted to support NASA's high priority missions, as defined in the latest decadal survey, to destinations such as Venus and Saturn in-situ robotic science missions. Entry into these planetary atmospheres results in extreme heating. The entry peak heat-flux and associated pressure are estimated to be between one and two orders of magnitude higher than those experienced by Mars Science Laboratory or Lunar return missions. In the recent New Frontiers community announcement NASA has indicated that it is considering providing an increase to the PI managed mission cost (PIMMC) for investigations utilizing the Heat Shield for Extreme Entry Environment Technology (HEEET) and in addition, NASA is considering limiting the risk assessment to only their accommodation on the spacecraft and the mission environment. The HEEET ablative TPS utilizes 3D weaving technology to manufacture a dual layer material architecture. The 3-D weaving allows for flat panels to be woven. The dual layer consists of a top layer designed to withstand the extreme external environment while the inner or insulating layer by design, is designed to achieve low thermal conductivity, and it keeps the heat from conducting towards the structure underneath. Both arc jet testing combined with material properties have been used to develop thermal response models that allows for comparison of performance with heritage carbon phenolic. A 50% mass efficiency is achieved by the dual layer construct compared to carbon phenolic for a broad range of missions both to Saturn and Venus. The 3-D woven flat preforms are molded to achieve the shape as they are compliant and then resin infusion with curing forms a rigid panels. These panels are then bonded on to the aeroshell structure. Gaps

  16. Heating facility for blanket and performance test

    Energy Technology Data Exchange (ETDEWEB)

    Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Sato, Satoshi; Hatano, Toshihisa; Takatsu, Hideyuki; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Hara, Shigemitsu

    1999-03-01

    A design and a fabrication of heating test facility for a mock-up of the blanket module to be installed in International Thermonuclear Experimental Reactor (ITER) have been conducted to evaluate/demonstrate its heat removal performance and structural soundness under cyclic heat loads. To simulate surface heat flux to the blanket module, infrared heating method is adopted so as to heat large surface area uniformly. The infrared heater is used in vacuum environment (10{sup -4} Torr{approx}), and the lamps are cooled by air flowing through an annulus between the lamp and a cover tube made of quartz glass. Elastomer O rings (available to be used up to {approx}300degC) and used for vacuum seal at outer surface of the cover tube. To prevent excessive heating of the O ring, the end part of the cover tube is specially designed including the tube shape, flow path of air and gold coating on the surface of the cover tube to protect the O ring against thermal radiation from glowing tungsten filament. To examine the performance of the facility, steady state and cyclic operation of the infrared heater were conducted using a small-scaled shielding blanket mock-up as a test specimen. The important results are as follows: (1) Heat flux at the surface of the small-scaled mock-up measured by a calorimeter was {approx}0.2 MW/m{sup 2}. (2) A comparison of thermal analysis results and measured temperature responses showed that the small-scaled mock-up had good heat removal performance. (3) Steady state operation and cyclic operation with step response between the rated and zero powers of the infrared heater were successfully performed, and it was confirmed that this heating facility was well-prepared and available for the thermal cyclic test of a blanket module. (author)

  17. Arc Heated Scramjet Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Arc Heated Scramjet Test Facility is an arc heated facility which simulates the true enthalpy of flight over the Mach number range of about 4.7 to 8 for free-jet...

  18. Space Environmental Testing of the Electrodynamic Dust Shield Technology

    Science.gov (United States)

    Calle, Carlos I.; Mackey, P. J.; Hogue, M. D.; Johansen, M .R.; Yim, H.; Delaune, P. B.; Clements, J. S.

    2013-01-01

    NASA's exploration missions to Mars and the moon may be jeopardized by dust that will adhere to surfaces of (a) Optical systems, viewports and solar panels, (b) Thermal radiators, (c) Instrumentation, and (d) Spacesuits. We have developed an active dust mitigation technology, the Electrodynamic Dust Shield, a multilayer coating that can remove dust and also prevents its accumulation Extensive testing in simulated laboratory environments and on a reduced gravity flight shows that high dust removal performance can be achieved Long duration exposure to the space environment as part of the MISSE-X payload will validate the technology for lunar missions.

  19. Seismic analysis of the mirror fusion test facility shielding vault

    International Nuclear Information System (INIS)

    Gabrielsen, B.L.; Tsai, K.

    1981-04-01

    This report presents a seismic analysis of the vault in Building 431 at Lawrence Livermore National Laboratory which houses the mirror Fusion Test Facility. The shielding vault structure is approximately 120 ft long by 80 ft wide and is constructed of concrete blocks approximately 7 x 7 x 7 ft. The north and south walls are approximately 53 ft high and the east wall is approximately 29 ft high. These walls are supported on a monolithic concrete foundation that surrounds a 21-ft deep open pit. Since the 53-ft walls appeared to present the greatest seismic problem they were the first investigated

  20. Shield design for the Fusion Materials Irradiation Test facility

    International Nuclear Information System (INIS)

    Carter, L.L.; Mann, F.M.; Morford, R.J.; Wilcox, A.D.; Johnson, D.L.; Huang, S.T.

    1983-03-01

    The shield design for the Fusion Materials Irradiation Test facility is based upon one-, two- and three-dimensional transport calculations with experimental measurements utilized to refine the nuclear data including the neutron cross sections from 20 to 50 MeV and the gamma ray and neutron source terms. The high energy neutrons and deuterons produce activation products from the numerous reactions that are kinematically allowed. The analyses for both beam-on and beam-off (from the activation products) conditions have required extensive nuclear data libraries and the utilization of Monte Carlo, discrete ordinates, point kernel and auxiliary computer codes

  1. SP-100 GES/NAT radiation shielding systems design and development testing

    International Nuclear Information System (INIS)

    Disney, R.K.; Kulikowski, H.D.; McGinnis, C.A.; Reese, J.C.; Thomas, K.; Wiltshire, F.

    1991-01-01

    Advanced Energy Systems (AES) of Westinghouse Electric Corporation is under subcontract to the General Electric Company to supply nuclear radiation shielding components for the SP-100 Ground Engineering System (GES) Nuclear Assembly Test to be conducted at Westinghouse Hanford Company at Richland, Washington. The radiation shielding components are integral to the Nuclear Assembly Test (NAT) assembly and include prototypic and non-prototypic radiation shielding components which provide prototypic test conditions for the SP-100 reactor subsystem and reactor control subsystem components during the GES/NAT operations. W-AES is designing three radiation shield components for the NAT assembly; a prototypic Generic Flight System (GFS) shield, the Lower Internal Facility Shield (LIFS), and the Upper Internal Facility Shield (UIFS). This paper describes the design approach and development testing to support the design, fabrication, and assembly of these three shield components for use within the vacuum vessel of the GES/NAT. The GES/NAT shields must be designed to operate in a high vacuum which simulates space operations. The GFS shield and LIFS must provide prototypic radiation/thermal environments and mechanical interfaces for reactor system components. The NAT shields, in combination with the test facility shielding, must provide adequate radiation attenuation for overall test operations. Special design considerations account for the ground test facility effects on the prototypic GFS shield. Validation of the GFS shield design and performance will be based on detailed Monte Carlo analyses and developmental testing of design features. Full scale prototype testing of the shield subsystems is not planned

  2. Early test facilities and analytic methods for radiation shielding: Proceedings

    International Nuclear Information System (INIS)

    Ingersoll, D.T.; Ingersoll, J.K.

    1992-11-01

    This report represents a compilation of eight papers presented at the 1992 American Nuclear Society/European Nuclear Society International Meeting. The meeting is of special significance since it commemorates the fiftieth anniversary of the first controlled nuclear chain reaction. The papers contained in this report were presented in a special session organized by the Radiation Protection and Shielding Division in keeping with the historical theme of the meeting. The paper titles are good indicators of their content and are: (1) The origin of radiation shielding research: The Oak Ridge experience, (2) Shielding research at the hanford site, (3) Aircraft shielding experiments at General Dynamics Fort Worth, 1950-1962, (4) Where have the neutrons gone?, a history of the tower shielding facility, (5) History and evolution of buildup factors, (6) Early shielding research at Bettis atomic power laboratory, (7) UK reactor shielding: then and now, (8) A very personal view of the development of radiation shielding theory

  3. Shielding design for testing room of large container scanner

    International Nuclear Information System (INIS)

    Liu Yisi; Miao Qitian; Zhou Liye

    1997-01-01

    Testing facility for large container scanner is a most advanced anti-smuggle tool. The X-ray scanning principle is adopted in this system. The X-ray was collimated a ted as a fan-shape beam. The accelerator only supplies the ray beam when the container is scanned. The irradiation time is less than one minute per test. The X-ray burst irradiation and highly collimated a ted scanning beam of this system is different from the common industrial irradiation accelerator. The shielding design of the 1:1 large container scanner introduced has better collimation level because of tri-collimation. The irradiation dose is less than 150 μGy per test, which is obviously lower than importations

  4. Method of measurement on materials shielding effectiveness test in time domain

    International Nuclear Information System (INIS)

    Liu Shunkun; Han Jun; Chen Xiangyue

    2009-01-01

    Windows method is a measurement of slot coupling effect in nature when it is used to measure material's shielding effectiveness. The error of measurement will become serious when it is used to measure material's shielding effectiveness in low frequency band. It is difficult to measure material's shielding effectiveness of electromagnetic pulse with Windows method. Device under test method (DUT method) was presented in this paper to overcome the limitations of Windows method in material's shielding effectiveness test. The method can be used to measure any material's shielding Effectiveness effectively through the design and the test of the DUT.The method was used to measure shielding effectiveness of special cement .Compared with theoretical analysis,the measurement result prove the DUT method to be very efficient in material's shielding effectiveness test. (authors)

  5. Shielding tests for a new ship for the transport of spent nuclear fuels

    International Nuclear Information System (INIS)

    Ito, D.; Kitano, T.; Akiyama, H.; Ueki, K.; Sanui, T.

    1998-01-01

    a new ship for the transport of spent nuclear fuels which uses serpentine concrete as its major shielding material has been constructed. The shielding calculations are based on DOT3.5 code (CCC-276) and the DLC23). Experiments with Cf-252 and Co-60 sources were carried out to confirm the validity of this method of calculating the shielding effectiveness of serpentine concrete. In these experiments, neutron and gamma-ray dose equivalent rates were measured in various arrangements to simulate the shielding structures of the ship, the calculations for each arrangement were performed by this shielding calculation method. For both neutron and gamma-rays, the calculation results agreed with the experiments very well, confirming that this calculation method used in the ship's shielding design is valid. Two kinds of on-board gamma-ray shielding tests were performed to confirm the ship's actual shielding effectiveness. In one kind of test, gamma-ray dose equivalent rates were measured for each shielding wall using Co-60 sources. In the other kind of test, gamma-ray dose equivalent rates in the ship's accommodation area were measured when a strong Co-60 source was placed in a loaded shipping cask's position. In both gamma-ray shielding tests all measured dose equivalent rates were less than the calculated values, confirming that the ship's actual shielding is sufficient to meet safety requirements. (authors)

  6. Antithermal shield for rockets with heat evacuation by infrared radiation reflection

    Directory of Open Access Journals (Sweden)

    Ioan RUSU

    2010-12-01

    Full Text Available At high speed, the friction between the air mass and the rocket surface causes a localheating of over 1000 Celsius degrees. For the heat protection of the rocket, on its outside surfacethermal shields are installed.Studying the Coanda effect, the fluid flow on solids surface, respectively, the author Ioan Rusuhas discovered by simply researches that the Coanda effect could be /extended also to the fluid flowon discontinuous solids, namely, on solids provided with orifices. This phenomenon was named by theauthor, the expanded Coanda effect. Starting with this discovery, the author has invented a thermalshield, registered at The State Office for inventions and Trademarks OSIM, deposit F 2010 0153This thermal shield:- is built as a covering rocket sheet with many orifices installed with a minimum space fromthe rocket body- takes over the heat fluid generated by the frontal part of the rocket and avoids the directcontact between the heat fluid and the rocket body- ensures the evacuation of the infrared radiation, generated by the heat fluid flowing overthe shield because of the extended Coanda effect by reflection from the rocket bodysurface.

  7. Final Test Report: Hexavalent Chrome Free Coatings for Electronics Electromagnetic Interference (EMI) Shielding Effectiveness (SE)

    Science.gov (United States)

    Kessel, Kurt R.

    2016-01-01

    The test results for Salt Spray Resistance, Static Heat and Humidity and Marine Environment can be found in Sections 3.1.3.3, 3.1.4.3 and 3.1.5.3 respectively. In summary, both the Metalast TCP and SurTec 650 Type 2 conversion coatings perform very similar to the incumbent Type 1 conversion coating against both 6061 and 5052 aluminum under all three test conditions. Significant prior work was performed to select the aluminum and conversion coating included within this test cycle; Reference - NASA GSDO Program Hexavalent Chrome Alternatives Final Pretreatments Test Report Task Order: NNH12AA45D September 01, 2013. As illustrated in the data, the 6061 aluminum panels SLIGHTLY out-performed the 5052 aluminum panels. Individual shielding effectiveness graphs for each panel are included within Appendix C and D. One other notable effect found during review of the data is that the Test Panels exposed to B117 Salt Fog reduced in shielding effectiveness significantly more than the Marine Environment Test Panels. The shielding effectiveness of the Marine Test Panels was approximately 20dB higher than the Test Panels that underwent B117 Salt Fog Exposure. The intent of this evaluation was not to maximize shielding effectiveness values. The same Parker Chomerics Cho-Seal 6503 gasket material was used for all panels with aluminum and conversion coating variants. A typical EMI gasket design for corrosive environments would be done quite differently. The intent was to execute a test that would provide the best possible evaluation of different aluminum materials and conversion coatings in corrosive environments. The test program achieved this intent. The fact that the two aluminums and two Type II conversion coatings performed similar to the incumbent Type 1 conversion coating is a positive outcome. It was desired to have an outcome that further differentiation the performance of two aluminum types and two conversion coating types but this could not be extracted by the test

  8. Evaporation and Vapor Shielding of CFC Targets Exposed to Plasma Heat Fluxes Relevant to ITER ELMs

    International Nuclear Information System (INIS)

    Safronov, V.; Arkhipov, N.I.; Toporkov, D.A.; Zhitlukhin, A.M.; Landman, I.

    2007-01-01

    Full text of publication follows: Carbon-fibre composite (CFC) is foreseen presently as armour material for the divertor target in ITER. During the transient processes such as instabilities of Edge Localized Modes (ELMs) the target as anticipated will be exposed to the plasma heat loads of a few MJ/m 2 on the time scale of a fraction of ms, which causes an intense evaporation at the target surface and contaminates tokamak plasma by evaporated carbon. The ITER transient loads are not achievable at existing tokamaks therefore for testing divertor armour materials other facilities, in particular plasma guns are employed. In the present work the CFC targets have been tested for ITER at the plasma gun facility MK- 200 UG in Troitsk by ELM relevant heat fluxes. The targets in the applied magnetic field up to 2 T were irradiated by hydrogen plasma streams of diameter 6 - 8 cm, impact ion energy 2 - 3 keV, pulse duration 0.05 ms and energy density varying in the range 0.05 - 1 MJ/m 2 . Primary attention has been focused on the measurement of evaporation threshold and investigation of carbon vapor properties. Fast infrared pyrometer, optical and VUV spectrometers, framing cameras and plasma calorimeters were applied as diagnostics. The paper reports the results obtained on the evaporation threshold of CFC, the evaporation rate of the carbon fibers oriented parallel and perpendicular to the exposed target surface, the velocity of carbon vapor motion along and across the magnetic field lines, and the parameters of carbon plasma such as temperature, density and ionization state measured up to the distance 15 cm at varying plasma load. First experimental results on investigation of the vapor shield onset conditions are presented also. (authors)

  9. X-Ray Micro-Tomography Applied to Nasa's Materials Research: Heat Shields, Parachutes and Asteroids

    Science.gov (United States)

    Panerai, Francesco; Borner, Arnaud; Ferguson, Joseph C.; Mansour, Nagi N.; Stern, Eric C.; Barnard, Harold S.; Macdowell, Alastair A.; Parkinson, Dilworth Y.

    2017-01-01

    X-ray micro-tomography is used to support the research on materials carried out at NASA Ames Research Center. The technique is applied to a variety of applications, including the ability to characterize heat shield materials for planetary entry, to study the Earth- impacting asteroids, and to improve broadcloths of spacecraft parachutes. From micro-tomography images, relevant morphological and transport properties are determined and validated against experimental data.

  10. Radiative Heating in MSL Entry: Comparison of Flight Heating Discrepancy to Ground Test and Predictive Models

    Science.gov (United States)

    Cruden, Brett A.; Brandis, Aaron M.; White, Todd R.; Mahzari, Milad; Bose, Deepak

    2014-01-01

    During the recent entry of the Mars Science Laboratory (MSL), the heat shield was equipped with thermocouple stacks to measure in-depth heating of the thermal protection system (TPS). When only convective heating was considered, the derived heat flux from gauges in the stagnation region was found to be underpredicted by as much as 17 W/sq cm, which is significant compared to the peak heating of 32 W/sq cm. In order to quantify the contribution of radiative heating phenomena to the discrepancy, ground tests and predictive simulations that replicated the MSL entry trajectory were performed. An analysis is carried through to assess the quality of the radiation model and the impact to stagnation line heating. The impact is shown to be significant, but does not fully explain the heating discrepancy.

  11. Helium leak testing of superconducting magnets, thermal shields and cryogenic lines of SST -1

    International Nuclear Information System (INIS)

    Thankey, P.L.; Joshi, K.S.; Semwal, P.; Pathan, F.S.; Raval, D.C.; Khan, Z.; Patel, R.J.; Pathak, H.A.

    2005-01-01

    Tokamak SST - 1 is under commissioning at Institute for Plasma Research. It comprises of a toroidal doughnut shaped plasma chamber, surrounded by liquid helium cooled superconducting magnets, housed in a cryostat chamber. The cryostat has two cooling circuits, (1) liquid nitrogen cooling circuit operating at 80 K to minimize the radiation heat load on the magnets, and (2) liquid helium cooling circuit to cool magnets and cold mass support structure to 4.5 K. In this paper we describe (a) the leak testing of copper - SS joints, brazing joints, interconnecting joints of the superconducting magnets, and (b) the leak testing of the liquid nitrogen cooling circuit, comprising of the main supply header, the thermal shields, interconnecting pipes, main return header and electrical isolators. All these tests were carried out using both vacuum and sniffer methods. (author)

  12. Grimsel Test Site: heat test, final report

    International Nuclear Information System (INIS)

    Schneefuss, J.; Glaess, F.; Gommlich, G.; Schmidt, M.

    1989-05-01

    The Swiss concept for the storage of radioactive waste consists in placing it in compact, dense rock formations. An experiment 'Heat Test' carried out by the 'Gesellschaft fuer Strahlen- und Umweltforschung' in Nagra's Grimsel rock laboratory simulated the heat production of stored radioactive waste. The aim was to evaluate processes for the demonstration of the suitability of a final repository for heat-producing radioactive waste in cristalline rock, to investigate the thermic, mechanic and hydraulic reactions to an artificial heat source, and to develop corresponding calculating models. The duration of the tests was about 3 years. In this report the measured thermic, mechanic and hydraulic reactions are documented and discussed in detail. A simple, rotation symmetrical FEM-model was used for the preparatory and experiment-accompanying modelling of the thermomechanical conditions in the heat test. The test showed that suitable measuring methods for the surveillance of the geomechanics of a final repository are available and that the reactions of the crystalline host rock to the heat source remain locally limited and can be modelled with relatively small effort. 29 refs., 33 figs., 10 tabs

  13. Test of thermal shields for early warning station detectors

    DEFF Research Database (Denmark)

    Petersen, Jesper

    1997-01-01

    The properties of thermal shields around NaI crystal scintillators for early warning stations have been checked in order to assure that external temperature variations cannot influence the stability of the measurements....

  14. Thermal, epithermal and thermalized neutron attenuation properties of ilmenite-serpentine heat resistant concrete shield

    International Nuclear Information System (INIS)

    Kany, A.M.I.; El-Gohary, M.I.; Kamal, S.M.

    1994-01-01

    Experimental measurements were carried out to study the attenuation properties of low-energy neutrons transmitted through unheated and preheated barriers of heavy-weight, highly hydrated and heat-resistant concrete shields. The concrete shields under investigation have been prepared from naturally occurring ilmenite and serpentine Egyptian ores. A collimated beam obtained from an Am-Be source was used as a source of neutrons, while the measurements of total thermal, epithermal, and thermalized neutron fluxes were performed using a BF-3 detector, multichannel analyzer and Cd filter. Results show that the ilmenite-serpentine concrete proved to be a better thermal, epithermal and thermalized neutron attenuator than the ordinary concrete especially at a high temperature of concrete exposure. (Author)

  15. Test and performance of a BGO Compton-suppression shield for GAMMASPHERE

    International Nuclear Information System (INIS)

    Carpenter, M.P.; Khoo, T.L.; Ahmad, I.

    1994-01-01

    Bismuth germanate (BGO) compton-suppression shields have been constructed to surround the Ge detectors of the GAMMASPHERE array. A shield consists of six hexagonal tapered BGO elements, each coupled to two 1-inch x 1-inch photomultiplier tubes. In addition, a cylindrical BGO detector is placed behind the Ge detector to intercept the forward scattered gamma rays. One hundred ten such shields are planned for the GAMMASPHERE array. Procedures for measuring the performance of these shields have been developed. Large (70 %) Ge detectors when used with these shields give a peak-to-total ratio of better tan 0.60. To date more than 85 shield have been tested and approved for use in GAMMASPHERE

  16. Joule loss on a Faraday shield of JT-60 ICRF test antenna

    International Nuclear Information System (INIS)

    Fujii, Tsuneyuki; Saigusa, Mikio; Ikeda, Yoshitaka; Kimura, Haruyuki; Hirashima, Teruhisa; Uehara, Munenori.

    1988-01-01

    Joule loss on a Faraday shield of JT-60 ICRF test antenna with a conductive casing is investigated at the frequency range of 120 MHz. The magnetic field radiated from the antenna is measured by three-dimensionally scanning an rf probe both inside and outside the antenna casing. The magnetic field perpendicular to the Faraday shield, B x , is found to be the largest component near the Faraday shield. It consequently gives the major part of the joule loss on the Faraday shield. The temperature distribution of the Faraday shield due to joule loss is measured directly with a thermocamera. It is confirmed that the area of the high temperature rise is consistent with the peak positions of the B x field. Faraday shield resistance which is estimated from power measurements agrees with the theoretical value. (author)

  17. Development of a pencil-type single shield graphite quasi-adiabatic calorimeter and comparison of its performance with a double-shield graphite calorimeter for the measurement of nuclear heat deposition rate in a fusion environment

    International Nuclear Information System (INIS)

    Joneja, O.P.; Rosselet, M.; Ligou, J.; Gardel, P.

    1995-01-01

    Recently, heat deposition rate measurements were reported that used a quasi-adiabatic double-shield graphite calorimeter. It was found that for a better understanding of nuclear heating due to incident radiation, having a calorimeter that could be conveniently moved axially and radially inside large material blocks would be advisable. Here, a simpler design, based on three elements, i.e., core, jacket, and shield is conceived. The fabrication and testing details are presented, and the performance of the current calorimeter is compared with a double-shield calorimeter under similar conditions. Such a system is found to be extremely sensitive and can be employed successfully at the LOTUS facility for future nuclear heat deposition rate measurements in large blocks of materials. The current design paves the way for the convenient testing of a large amount of kerma factor data required for constructing future fusion machines. The same configuration with minor changes can be extended to most of the fusion materials of interest. The core of the new calorimeter measures 11 mm in diameter and height and has overall dimensions of 24 mm in diameter and 180 mm in height. The response of the calorimeter is measured by placing it in front of the Haefely neutron generator. 12 refs., 16 figs., 9 tabs

  18. Long term testing of materials for tube shielding, stage 2; Laangtidsprovning av tubskyddsmaterial, etapp 2

    Energy Technology Data Exchange (ETDEWEB)

    Norling, Rikard; Hjoernhede, Anders; Mattsson, Mattias

    2012-02-15

    Circulating Fluidized Bed (CFB) boilers are commonly used for combustion of biomass and are used to some extent for Waste-to-Energy (WtE) plants as well. The superheaters of the latter are for obvious reasons more prone to suffer from high temperature corrosion caused by the corrosive species in the fuel, mainly chlorides. Frequently the final (hottest) superheater is positioned in the loop seal, where the circulating bed material is returned to the furnace after being separated from the flue gas by a cyclone. The environment in the loop seal is relatively free of chlorides, since these primarily follow the flue gas into the convection pass. Hence, higher steam temperature can be allowed without excessive damage to the final superheater. On the other hand the superheaters, which are located in the convection pass, are more exposed to the corrosive species of the flue gas. Further, they are eroded by particles entrained in the gas flow. Particles and condensing gaseous species are to a large extent deposited on the superheaters, which limits the heat transfer and promotes corrosion. The deposits are regularly removed e.g. by soot blowers. The pressurized steam from soot blowers causes additional erosion damage to that caused by entrained particles. It shall be noted that the actual damage is caused by a combined mechanism of erosion and corrosion denoted erosion-corrosion, which usually results in dramatically accelerated wear. To avoid excessive erosion damage on the superheater tubes the first tube row of each bundle is protected by tube shielding. In its simplest form the shields are made from a steel sheet that has been bent into a semi-circular half-cylinder shell. These shields are attached onto the wind-side of the tubes by hangers. A typical material for tube shielding is the austenitic high temperature resistant stainless steel 253MA. Life of tube shielding depends on numerous factors such as boiler design, superheater location, fuel and operating

  19. Astronaut Ronald Sega with Wake Shield Facility on test stand at JSC

    Science.gov (United States)

    1991-01-01

    The Wake Shield Facility is displayed on a test stand at JSC. Astronaut Ronald M. Sega, mission specialist for STS-60, is seen with the facility during a break in testing in the acoustic and vibration facility at JSC.

  20. Developmental testing of partially volatile neutron shields for high-performance shipping casks

    International Nuclear Information System (INIS)

    Pope, R.B.; Allen, G.C.; Rack, H.J.; Joseph, B.J.; Dupree, S.A.

    1980-01-01

    Results of the phase one tests have demonstrated that the neutron-shielding concept described in this paper is a viable design option for spent fuel shipping casks. The tests have shown that the Boro-silicone 236 shield is superior to the other shield materials considered. Repeated TGA, aging and fire tests demonstrated the reliability of the data. A second phase of the test program is now being pursued where the Boro-silicone 236 is injected into all-steel slab sections, and cured in place. 5 tables

  1. Materials tests and analyses of Faraday shield tubes for ICRF [ion cyclotron resonant frequency] antennas

    International Nuclear Information System (INIS)

    King, J.F.; Baity, F.W.; Hoffman, D.J.; Walls, J.C.; Taylor, D.J.

    1988-01-01

    The ion cyclotron resonant frequency (ICRF) antennas for heating fusion plasmas require careful analysis of the materials selected for the design and the successful fabrication of high integrity braze bonds. Graphite tiles are brazed to Inconel 625 Faraday shield tubes to protect the antenna from the plasma. The bond between the graphite and Inconel tube is difficult to achieve due to the different coefficients of thermal expansion. A 2-D stress analysis showed the graphite could be bonded to Inconel with a Ag-Cu-Ti braze alloy without cracking the graphite. Brazing procedures and nondestructive examination methods have been developed for these joints. This paper presents the results of our joining development and proof testing. 2 refs., 3 figs

  2. Safety verification of radiation shielding and heat transfer for a model for dry

    International Nuclear Information System (INIS)

    Yu, Haiyan; Tang, Xiaobin; Wang, Peng; Chen, Feida; Chai, Hao; Chen, Da

    2015-01-01

    Highlights: • New type of dry spent fuel storage was designed. • MC method and FEM were used to verify the reliability of new storage. • Radiation shield and heat transfer both meet IAEA standards: 2 mSv/h, 0.1 mSv/h and 190 °C, 85 °C. • Provided possibilities for future implementation of this type of dry storage. - Abstract: The goal of this research is to develop a type of dry spent fuel storage called CHN-24 container, which could contain an equivalent load of 45 GWD/MTU of spent fuel after 10 years cooling. Basically, radiation shielding performance and safe removal of decay heat, which play important roles in the safety performance, were checked and validated using the Monte Carlo method and finite element analysis to establish the radiation dose rate calculation model and three-dimensional heat transfer model for the CHN-24 container. The dose rates at the surface of the container and at a distance of 1 m from the surface were 0.42 mSv/h and 0.06 mSv/h, respectively. These conform to the International Atomic Energy Agency (IAEA) radioactive material transportation safety standards 2 mSv/h and 0.1 mSv/h. The results shows that the CHN-24 container maintains its structural and material integrity under the condition of normal thermal steady-state heat transfer as well as in case of extreme fire as evinced by transient-state analysis. The temperature inside and on the surface of the container were 150.91 °C and 80 °C under normal storage conditions, which indicated that the design also conform to IAEA heat transfer safety standards of 190 °C and 85 °C

  3. Shielding designs and tests of a new exclusive ship for transporting spent nuclear fuels

    International Nuclear Information System (INIS)

    Ono, M.; Ito, D.; Kitano, T.; Ueki, K.; Akiyama, H.; Obara, I.; Sanui, T.

    2000-01-01

    The Rokuei-Maru, a ship built specially for the transport of spent nuclear fuels in casks, was launched April in 1996. She is the first ship to comply with special Japanese regulations, KAISA 520, based on the INF code. DOT3.5 and MCNP-4A were used for the evaluation of dose equivalent rates of her shielding structures. On-board gamma-ray shielding tests were executed to confirm the effectiveness of the ship's shielding performance. The tests confirmed that effective shielding has been achieved and the dose equivalent rate in the accommodation and other inhabited spaces is sufficiently lower than the regulated limitations. This was achieved by employing the appropriate calculation methods and shielding materials. (author)

  4. High performance inboard shield design for the compact TIBER-II test reactor: Appendix A-2

    International Nuclear Information System (INIS)

    El-Guebaly, L.A.; Sviatoslavsky, I.N.

    1987-01-01

    The compactness of the TIBER-II reactor has placed a premium on the design of a high performance inboard shield to protect the inner legs of the toroidal field (TF) coils. The available space for shield is constrained to 48 cm and the use of tungsten is mandatory to protect the magnet against the 1.53 MW/m 2 neutron wall loading. The primary requirement for the shield is to limit the fast neutron fluence to 10 19 n/cm 2 . In an optimization study, the performance of various candidate materials for protecting the magnet was examined. The optimum shield consists of a 40 cm thick W layer, followed by an 8 cm thick H 2 O/LiNO 3 layer. The mechanical design of the shield calls for tungsten blocks within SS stiffened panels. All the coolant channels are vertical with more of them in the front where there is a high heat load. The coolant pressure is 0.2 MPa and the maximum structural surface temperature is 0 C. The effects of the detailed mechanical design of the shield and the assembly gaps between the shield sectors on the damage in the magnet were analyzed and peaking factors of ∼2 were found at the hot spots. 2 refs., 6 figs., 2 tabs

  5. Shielding plugs

    International Nuclear Information System (INIS)

    Makishima, Kenji.

    1986-01-01

    Purpose: In shielding plugs of an LMFBR type reactor, to restrain natural convection of heat in an annular space between a thermal shield layer and a shield shell, to prevent the lowering of heat-insulation performance, and to alleviate a thermal stress in a reactor container and the shield shell. Constitution: A ring-like leaf spring split in the direction of height is disposed in an annular space between a thermal shield layer and a shield shell. In consequence, the space is partitioned in the direction of height and, therefore, if axial temperature conditions and space width are the same and the space is low, the natural convection is hard to occur. Thus the rise of upper surface temperature of the shielding plugs can prevent the lowering of the heat insulation performance which will result in the increment of shielding plug cooling capacity, thereby improving reliability. In the meantime, since there is mounted an earthquake-resisting support, the thermal shield layer will move for a slight gap in case of an earthquake, being supported by the earthquake-resisting support, and the movement of the thermal shield layer is restricted, thereby maintaining integrity without increasing the stroke of the ring-like spring. (Kawakami, Y.)

  6. Analysis and Testing of a Composite Fuselage Shield for Open Rotor Engine Blade-Out Protection

    Science.gov (United States)

    Pereira, J. Michael; Emmerling, William; Seng, Silvia; Frankenberger, Charles; Ruggeri, Charles R.; Revilock, Duane M.; Carney, Kelly S.

    2016-01-01

    The Federal Aviation Administration is working with the European Aviation Safety Agency to determine the certification base for proposed new engines that would not have a containment structure on large commercial aircraft. Equivalent safety to the current fleet is desired by the regulators, which means that loss of a single fan blade will not cause hazard to the Aircraft. The NASA Glenn Research Center and The Naval Air Warfare Center (NAWC), China Lake, collaborated with the FAA Aircraft Catastrophic Failure Prevention Program to design and test lightweight composite shields for protection of the aircraft passengers and critical systems from a released blade that could impact the fuselage. LS-DYNA® was used to predict the thickness of the composite shield required to prevent blade penetration. In the test, two composite blades were pyrotechnically released from a running engine, each impacting a composite shield with a different thickness. The thinner shield was penetrated by the blade and the thicker shield prevented penetration. This was consistent with pre-test LS-DYNA predictions. This paper documents the analysis conducted to predict the required thickness of a composite shield, the live fire test from the full scale rig at NAWC China Lake and describes the damage to the shields as well as instrumentation results.

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

    International Nuclear Information System (INIS)

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

    1999-09-01

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

  8. Development of high-performance shielding material by heat curing method

    Energy Technology Data Exchange (ETDEWEB)

    Miura, Toshimasa; Hirao, Yoshihiro; Hayashi, Takayuki; Okuno, Koichi; Sato, Osamu [National Maritime Research Institute, Ibaraki (Japan)

    2002-07-01

    A high-performance shielding material is developed by a heat curing method. It is mainly made of a thermosetting resin, lead powder, and a boron compound. To make the resin, a single functional monomer stearyl methacrylate (SMA) is used. To get good dispersion of lead and the boron compound in the resin, the viscosity of the SMA is increased by adding a small amount of a peroxide into the liquid monomer and heating up to the temperature of 100 .deg. C. Next, a peroxide, lead powder, a boron compound, a three functional monomer, and a curing accelerator are mixed into the viscous SMA. The mixture is cured in an atmosphere of nitrogen after removing bubbles using a vacuum pump. Measured properties of the cured material are as follows. The curing rate of SMA is 97 %. The density is kept 2.35 g/cm{sub 3} in the range from room temperature to 150 .deg. C. The weight-change measured by a thermogravimetry is 0.16 % in the range from room temperature to 200 .deg. C. Details of fragments in the gas released from the material is analyzed by a gas chromatography and a mass spectrometry. The hydrogen content of the material is 6.04x10 {sub 22} /cm{sub 3} . The shielding effect is calculated for a fission source by an Sn code ANISN. The shielding effect of the curing material is excellent. For example, concrete shield of a certain thickness can be replaced by the material having a thickness less than a half of concrete. Several samples of the material are irradiated at an irradiation equipment of the research reactor JRR-4 installed at Japan Atomic Energy Research Institute. At the 14{sub th} day after irradiating with the thermal neutron fluence of 6.6x10{sub 15} /cm{sub 2} , the radioactivity is less than one tenth of 75 Bq/g above which materials are regulated as the radioactive substance in Japan.

  9. Normalizing effect on fatigue crack propagation at the heat-affected zone of AISI 4140 steel shielded metal arc weldings

    Directory of Open Access Journals (Sweden)

    B. Vargas-Arista

    2013-01-01

    Full Text Available The fractography and mechanical behaviour of fatigue crack propagation in the heat-affected zone (HAZ of AISI 4140 steel welded using the shielded metal arc process was analysed. Different austenitic grain size was obtained by normalizing performed at 1200 °C for 5 and 10 hours after welding. Three point bending fatigue tests on pre-cracked specimens along the HAZ revealed that coarse grains promoted an increase in fatigue crack growth rate, hence causing a reduction in both fracture toughness and critical crack length, and a transgranular brittle final fracture with an area fraction of dimple zones connecting cleavage facets. A fractographic analysis proved that as the normalizing time increased the crack length decreased. The increase in the river patterns on the fatigue crack propagation in zone II was also evidenced and final brittle fracture because of transgranular quasicleavage was observed. Larger grains induced a deterioration of the fatigue resistance of the HAZ.

  10. COMPUTATIONAL FLUID DYNAMICS INVESTIGATION ON THE USE OF HEAT SHIELDS FOR THERMAL MANAGEMENT IN A CAR UNDERHOOD

    Directory of Open Access Journals (Sweden)

    S.Y. Lam

    2012-12-01

    Full Text Available Temperature variations inside a car underhood are largely controlled by the heat originating from the engine block and the exhaust manifold. Excessive temperatures in the underhood can lead to the faster deterioration of engine components and may affect the thermal comfort level inside the passenger cabin. This paper presents computational fluid dynamics investigations to assess the performance of a heat shield in lowering the peak temperature of the engine components and firewall in the underhood region of a typical passenger car. The simulation used the finite volume method with the standard k-ε turbulence model and an isothermal model for the heat transfer calculations. The results show that the heat shield managed to reduce the peak temperature of the engine components and firewall by insulating the intense heat from the engine block and exhaust and regulating the airflow inside the underhood region.

  11. Hydramite II screening tests of potential bremsstrahlung converter debris shield materials

    International Nuclear Information System (INIS)

    Reedy, E.D. Jr.; Hedemann, M.A.; Stark, M.A.

    1986-03-01

    Results of a brief test series aimed at screening a number of potential bremsstrahlung converter debris shield materials are reported. These tests were run on Sandia National Laboratories' Hydramite II accelerator using a diode configuration which produces a pinched electron beam. The materials tested include: (1) laminated Kevlar 49/polyester and E-glass/polyester composites, (2) a low density laminated Kevlar 49 composite, and (3) two types of through-the-thickness reinforced Kevlar 49 composites. As expected, tests using laminated Kevlar 49/polyester shields showed that shield permanent set (i.e., permanent deflection) increased with increasing tantalum conversion foil thickness and decreased with increasing shield thickness. The through-the-thickness reinforced composites developed localized, but severe, back surface damage. The laminated composites displayed little back surface damage, although extensive internal matrix cracking and ply delaminations were generated. Roughly the same degree of permanent set was produced in shields made from the low density Kevlar 49 composite and the Kevlar 49/polyester. The E-glass reinforced shields exhibited relatively low levels of permanent set

  12. Erosion and migration of tungsten employed at the central column heat shield of ASDEX Upgrade

    International Nuclear Information System (INIS)

    Krieger, K.; Gong, X.; Balden, M.; Hildebrandt, D.; Maier, H.; Rohde, V.; Roth, J.; Schneider, W.

    2002-01-01

    In ASDEX Upgrade, tungsten was employed as plasma facing material at the central column heat shield in the plasma main chamber. The campaign averaged tungsten erosion flux was determined by measuring the difference of the W-layer thickness before and after the experimental campaign using ion beam analysis methods. The observed lateral variation and the total amount of eroded tungsten are attributed to erosion by impact of ions from the scrape-off layer plasma. Migration and redeposition of eroded tungsten were investigated by quantitative analysis of deposited tungsten on collector probes and wall samples. The obtained results, as well as the spectroscopically observed low tungsten plasma penetration probability, indicate that a major fraction of the eroded tungsten migrates predominantly through direct transport channels in the outer plasma scrape-off layer without entering the confined plasma

  13. Thermal fatigue cracks in gas turbine heat shield plates; Thermoermuedungsrisse in Hitzeschildplatten von Gasturbinen

    Energy Technology Data Exchange (ETDEWEB)

    Neidel, Andreas; Riesenbeck, Susanne [Siemens AG, Berlin (Germany). Gas Turbine Plant Berlin Labs.

    2012-07-01

    There are numerous possible designations for the damage mechanism described in this case study. As a consequence, the terminology is far from being consistent. In this context, the Anglo-Saxon language area has to be taken into consideration. On the one hand many failure analysis reports have to be written in English, on the other hand it is meanwhile expected to use English terms in reports written in German, the latter in an effort to standardize the internal nomenclature. Therefore, it is advisable for damage analysts to know technical terms in both languages, at least for the most important damage mechanisms occurring in their respective fields of activity. In the present case, individual ceramic coated metal heat shield plates have been replaced after several ten thousand operating hours and several hundred start-up and shut-down procedures, i.e. machine start-ups, due to cracks in the central locating hole.

  14. Efficient hydrogen production using heat in neutron shield of fusion reactor

    International Nuclear Information System (INIS)

    Okano, Kunihiko; Asaoka, Yoshiyuki; Hiwatari, Ryouji; Yoshida, Tomoaki

    2001-01-01

    In future perspective of energy supply, a hydrogen energy cycle is expected to play an important role as a CO 2 free fuel for mobile or co-generation systems. Fusion power plants should offer advantages, compatibilities and/or synergistic effects with or in such future energy systems. In this paper, a comprehensive power station, in which a fusion plant is integrated with a hydrogen production plant, is proposed. A tenuous heat source in the outboard shield, which is unsuitable to produce high-pressure and high-temperature steam for efficient electric power generation, is used for the hydrogen production. This integrated system provides some synergistic effects and it would be advantageous over any independent use of each plant. (author)

  15. Study on Load-displacement Test of Rubber Pads of Coal Mine Roadway Constructed by Shield

    Science.gov (United States)

    Yang, Yue; Chen, Xiaoguo; Yang, Liyun

    2017-12-01

    Shield method construction of coal mine roadway is the future trend of the development of deep coal mining. The main shaft supporting is the segment. There is rubber pads between segment and segment. The performance of compression deformation of rubber pad is essential for the overall stability of lining. Through load test, displacement of the rubber pad under load, the thrust force law of the rubber pad deformation, and provide a theoretical basis for the stability analysis of coal mine tunnel shield construction.

  16. Pre-installation empirical testing of room shielding for high dose rate remote afterloaders

    International Nuclear Information System (INIS)

    Klein, E.E.; Grigsby, P.W.; Williamson, J.F.; Meigooni, A.S.

    1993-01-01

    PURPOSE: Many facilities are acquiring high dose rate remote afterloading units. It is economical that these units be placed in existing shielded teletherapy rooms. Scatter-radiation barriers marginally protect uncontrolled areas from a high dose rate source especially in a room that houses a non-dynamic Cobalt-60 unit. In addition the exact thickness and material composition of the barriers are unknown and therefore, a calculation technique may give misleading results. Also, it would be impossible to evaluate an entire wall barrier by taking isolated core samples in order to assist in the calculations. A quick and inexpensive measurement of dose equivalent using a rented high activity 192Ir source evaluates the barriers and locates shielding deficiencies. METHODS AND MATERIALS: We performed transmission calculations for primary and scattered radiation based on National Council on Radiation Protection and Measurements Reports 49 and 51, respectively. We then rented a high activity 21.7 Ci (8.03 x 10(11) Bq) Ir-192 source to assess our existing teletherapy room shielding for adequacy and voids. This source was placed at the proposed location for clinical high dose rate treatment and measurements were performed. RESULTS: No deficiencies were found in controlled areas surrounding the room, but large differences were found between the calculated and measured values. Our survey located a region in the uncontrolled area above the room requiring augmented shielding which was not predicted by the calculations. A canopy shield was designed to potentially augment the shielding in the ceiling direction. CONCLUSION: Pre-installation testing by measurement is an invaluable method for locating shielding deficiencies and avoiding unnecessary enhancement of shielding particularly when there is lack of information of the inherent shielding

  17. Pulse-echo ultrasonic inspection system for in-situ nondestructive inspection of Space Shuttle RCC heat shields.

    Energy Technology Data Exchange (ETDEWEB)

    Roach, Dennis Patrick; Walkington, Phillip D.; Rackow, Kirk A.

    2005-06-01

    The reinforced carbon-carbon (RCC) heat shield components on the Space Shuttle's wings must withstand harsh atmospheric reentry environments where the wing leading edge can reach temperatures of 3,000 F. Potential damage includes impact damage, micro cracks, oxidation in the silicon carbide-to-carbon-carbon layers, and interlaminar disbonds. Since accumulated damage in the thick, carbon-carbon and silicon-carbide layers of the heat shields can lead to catastrophic failure of the Shuttle's heat protection system, it was essential for NASA to institute an accurate health monitoring program. NASA's goal was to obtain turnkey inspection systems that could certify the integrity of the Shuttle heat shields prior to each mission. Because of the possibility of damaging the heat shields during removal, the NDI devices must be deployed without removing the leading edge panels from the wing. Recently, NASA selected a multi-method approach for inspecting the wing leading edge which includes eddy current, thermography, and ultrasonics. The complementary superposition of these three inspection techniques produces a rigorous Orbiter certification process that can reliably detect the array of flaws expected in the Shuttle's heat shields. Sandia Labs produced an in-situ ultrasonic inspection method while NASA Langley developed the eddy current and thermographic techniques. An extensive validation process, including blind inspections monitored by NASA officials, demonstrated the ability of these inspection systems to meet the accuracy, sensitivity, and reliability requirements. This report presents the ultrasonic NDI development process and the final hardware configuration. The work included the use of flight hardware and scrap heat shield panels to discover and overcome the obstacles associated with damage detection in the RCC material. Optimum combinations of custom ultrasonic probes and data analyses were merged with the inspection procedures needed to

  18. BUGLE-93 (ENDF/B-VI) cross-section library data testing using shielding benchmarks

    International Nuclear Information System (INIS)

    Hunter, H.T.; Slater, C.O.; White, J.E.

    1994-01-01

    Several integral shielding benchmarks were selected to perform data testing for new multigroup cross-section libraries compiled from the ENDF/B-VI data for light water reactor (LWR) shielding and dosimetry. The new multigroup libraries, BUGLE-93 and VITAMIN-B6, were studied to establish their reliability and response to the benchmark measurements by use of radiation transport codes, ANISN and DORT. Also, direct comparisons of BUGLE-93 and VITAMIN-B6 to BUGLE-80 (ENDF/B-IV) and VITAMIN-E (ENDF/B-V) were performed. Some benchmarks involved the nuclides used in LWR shielding and dosimetry applications, and some were sensitive specific nuclear data, i.e. iron due to its dominant use in nuclear reactor systems and complex set of cross-section resonances. Five shielding benchmarks (four experimental and one calculational) are described and results are presented

  19. Heat leak testing of a superconducting RHIC dipole magnet at Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    DeLalio, J.T.; Brown, D.P.; Sondericker, J.H.

    1993-01-01

    Brookhaven National Laboratory is currently performing heat load tests on a superconducting dipole magnet. The magnet is a prototype of the 360, 8 cm bore, arc dipole magnets that will be used in the Relativistic Heavy Ion Collider (RMC). An accurate measurement of the heat load is needed to eliminate cumulative errors when determining the REUC cryogenic system load requirements. The test setup consists of a dipole positioned between two quadrupoles in a common vacuum tank and heat shield. Piping and instrumentation are arranged to facilitate measurement of the heat load on the primary 4.6 K magnet load and the secondary 55 K heat shield load. Initial results suggest that the primary heat load is well below design allowances. The secondary load was found to be higher than estimated, but remained close to the budgeted amount. Overall, the dipole performed to specifications

  20. Shutdown dose rate analysis of European test blanket modules shields in ITER Equatorial Port #16

    Energy Technology Data Exchange (ETDEWEB)

    Juárez, Rafael, E-mail: rjuarez@ind.uned.es [Departamento de Ingeniería Energética, ETSII-UNED, Calle Juan del Rosal 12, Madrid 28040 (Spain); Sauvan, Patrick; Perez, Lucia [Departamento de Ingeniería Energética, ETSII-UNED, Calle Juan del Rosal 12, Madrid 28040 (Spain); Panayotov, Dobromir; Vallory, Joelle; Zmitko, Milan; Poitevin, Yves [Fusion for Energy (F4E), Torres Diagonal Litoral B3, Josep Pla 2, Barcelona 08019 (Spain); Sanz, Javier [Departamento de Ingeniería Energética, ETSII-UNED, Calle Juan del Rosal 12, Madrid 28040 (Spain)

    2016-11-01

    Highlights: • Nuclear analysis for European TBMs and shields, in ITER Equatorial Port #16, has been conducted in support of the ‘Concept Design Review’ from ITER. • The objective of the work is the characterization of the Shutdown Dose Rates at Equatorial Port #16 interspace. • The role played by the TBM and TBM shields, the equatorial port gaps and the vacuum vessel permeation, in terms of neutron flux transmission is assessed. • The role played by the TBM, TBM shields, Port Plug Frame, Pipe Forest and the machine in terms of activation is also investigated. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). An essential element of the Conceptual Design Review (CDR) of these TBSs is the demonstration of capability of Test Blanket Modules (TBM) and their shields to fulfil their function and comply with the design requirements. One of the TBM shields highly relevant design aspects is the project target for shutdown dose rates (SDDR) in the interspace. We investigated two functions of the TBMs and TBM shields—the neutron flux attenuation along the shields, and the reduction of the activation of the components contributing to SDDR. It is shown that TBMs and TBM shields reduce significantly the neutron flux in the port plug (PP). In terms of neutron flux attenuation, the TBM shield provides sufficient neutron flux reduction, being responsible for 5 × 10{sup 6} n/cm{sup 2} s at port interspace, while the EPP gaps and BSM gaps are responsible for 5 × 10{sup 7} n/cm{sup 2} s each. When considering closed upper, lower and lateral neighbour equatorial ports (thus, excluding the cross-talk between ports), a SDDR of 121 μSv/h averaged near the port closure flange was obtained, out of which, only 4 μSv/h are due to the activation of TBMs and TBM shields. Maximum SDDR in the range

  1. Optimization of the Mu2e Production Solenoid Heat and Radiation Shield

    Science.gov (United States)

    Pronskikh, V. S.; Coleman, R.; Glenzinski, D.; Kashikhin, V. V.; Mokhov, N. V.

    2014-03-01

    The Mu2e experiment at Fermilab is designed to study the conversion of a negative muon to electron in the field of a nucleus without emission of neutrinos. Observation of this process would provide unambiguous evidence for physics beyond the Standard Model, and can point to new physics beyond the reach of the LHC. The main parts of the Mu2e apparatus are its superconducting solenoids: Production Solenoid (PS), Transport Solenoid (TS), and Detector Solenoid (DS). Being in the vicinity of the beam, PS magnets are most subjected to the radiation damage. In order for the PS superconducting magnet to operate reliably, the peak neutron flux in the PS coils must be reduced by 3 orders of magnitude by means of sophisticatedly designed massive Heat and Radiation Shield (HRS), optimized for the performance and cost. An issue with radiation damage is related to large residual electrical resistivity degradation in the superconducting coils, especially its Al stabilizer. A detailed MARS15 analysis and optimization of the HRS has been carried out both to satisfy the Mu2e requirements to the radiation quantities (such as displacements per atom, peak temperature and power density in the coils, absorbed dose in the insulation, and dynamic heat load) and cost. Results of MARS15 simulations of these radiation quantities are reported and optimized HRS models are presented; it is shown that design levels satisfy all requirements.

  2. Auxiliary Heat Exchanger Flow Distribution Test

    International Nuclear Information System (INIS)

    Kaufman, J.S.; Bressler, M.M.

    1983-01-01

    The Auxiliary Heat Exchanger Flow Distribution Test was the first part of a test program to develop a water-cooled (tube-side), compact heat exchanger for removing heat from the circulating gas in a high-temperature gas-cooled reactor (HTGR). Measurements of velocity and pressure were made with various shell side inlet and outlet configurations. A flow configuration was developed which provides acceptable velocity distribution throughout the heat exchanger without adding excessive pressure drop

  3. Effects of drop testing on scale model shipping containers shielded with depleted uranium

    International Nuclear Information System (INIS)

    Butler, T.A.

    1980-02-01

    Three scale model shipping containers shielded with depleted uranium were dropped onto an essentially unyielding surface from various heights to determine their margins to failure. This report presents the results of a thorough posttest examination of the models to check for basic structural integrity, shielding integrity, and deformations. Because of unexpected behavior exhibited by the depleted uranium shielding, several tests were performed to further characterize its mechanical properties. Based on results of the investigations, recommendations are made for improved container design and for applying the results to full-scale containers. Even though the specimens incorporated specific design features, the results of this study are generally applicable to any container design using depleted uranium

  4. Radiation safety aspects during nondestructive testing of reactor shielding components by gamma radiometry

    International Nuclear Information System (INIS)

    Viswanathan, S.; Jose, M.T.; Venkatraman, B.

    2016-01-01

    In nuclear facilities, effective shielding of radioactive components and structures are essential to ensure radiation protection to operating personnel. The shield structures are made of lead, steel and concrete with varying thickness of up to 1200 mm. It needs to be verified for shielding integrity, presence of voids, blowholes and defects to avoid exposure to workers and to public at large. Radiometry using gamma source serves as excellent tool for non-destructive examination of such structures and components. Gamma sources of high activity up to 50 Curies (gamma camera type) depending on the thickness of component have to be used. During the testing exposure to the operating personnel needs to be minimized, this requires certain safety procedures to be followed. This paper focuses the methodology to be adapted by means of selection of source, effective training of personnel, compliance with safety requirements and maintenance of source devices

  5. Initial progress in the first wall, blanket, and shield Engineering Test Program for magnetically confined fusion-power reactors

    International Nuclear Information System (INIS)

    Herman, H.; Baker, C.C.; Maroni, V.A.

    1981-10-01

    The first wall/blanket/shield (FW/B/S) Engineering Test Program (ETP) progressed from the planning stage into implementation during July, 1981. The program, generic in nature, comprises four Test Program Elements (TPE's), the emphasis of which is on defining the performance parameters for the Fusion Engineering Device (FED) and the major fusion device to follow FED. These elements are: (1) nonnuclear thermal-hydraulic and thermomechanical testing of first wall and component facsimiles with emphasis on surface heat loads and heat transient (i.e., plasma disruption) effects; (2) nonnuclear and nuclear testing of FW/B/S components and assemblies with emphasis on bulk (nuclear) heating effects, integrated FW/B/S hydraulics and mechanics, blanket coolant system transients, and nuclear benchmarks; (3) FW/B/S electromagnetic and eddy current effects testing, including pulsed field penetration, torque and force restraint, electromagnetic materials, liquid metal MHD effects and the like; and (4) FW/B/S Assembly, Maintenance and Repair (AMR) studies focusing on generic AMR criteria, with the objective of preparing an AMR designers guidebook; also, development of rapid remote assembly/disassembly joint system technology, leak detection and remote handling methods

  6. Testing of ENDF/B-VI data for shielding applications

    International Nuclear Information System (INIS)

    Ingersoll, D.T.

    1995-01-01

    Version VI of the U.S. Evaluated Nuclear Data File (END/B-VI) was released for open distribution in 1990 after an extensive multiyear, multilaboratory evaluation effort coordinated by the Cross Section Evaluation Working Group (CSEWG). More than 75 of the 320 evaluations contained in the library are new for Version VI, including many relatively important nuclides and many with substantial changes to the cross-section data. Also, several important changes were made to the basic data formats for Version VI to permit better representation of the data and to allow additional types of data. Although these format changes yielded improved evaluations, they hindered processing of the data for use in applications codes and delayed the timely integral testing of the data. It has been only during the past year that significant integral testing of END/B-VI data has been achieved. The results and conclusions from some of these efforts are summarized in the following paragraphs

  7. Material and electromagnetic properties of Faraday shields for ion cyclotron heating antennas

    International Nuclear Information System (INIS)

    Hoffman, D.J.; Baity, F.W.; Becraft, W.R.; Caughman, J.B.O.; Tsai, C.C.

    1985-01-01

    The Faraday shields for ion cyclotron antennas must transmit magnetic waves and absorb little RF power. To investigate these properties, we have constructed 27 Faraday shields in many configurations, including chevrons, tubes, straps, concentric rings, various layered shields, conventionally leafed straps, and replicas of the Faraday shields for ASDEX, the Joint European Torus (JET), TEXTOR, and Alcator-C. We have measured the magnetic flux and observed loading at various operating resistances by using dielectric sheets or magnetic-coupled loads. Each Faraday shield effects a net change in the characteristic inductance of the antenna, resulting in a reduction of wave coupling. However, the load experienced by the antenna is not always reduced because the Faraday shield itself acts as a load. We differentiate between these effects experimentally. The net result of the study is that the Faraday shields now in use cost up to a factor of 50% of coupling. This, of course, reduces the power handling capability by 50% as well. However, configurations exist that are easily cooled and result in a reduction of less than 5% in loading

  8. Material and electromagnetic properties of Faraday shields for ion cyclotron heating antennas

    International Nuclear Information System (INIS)

    Hoffman, D.J.; Becraft, W.R.; Baity, F.W.; Caughman, J.B.O.; Tsai, C.C.

    1985-01-01

    The Faraday shields for ion cyclotron antennas must transmit magnetic waves and adsorb little rf power. To investigate these properties, we have constructed 27 Faraday shields in many configurations, including chevrons, tubes, straps, concentric rings, various layered shields, conventionally leafed straps, and replicas of the Faraday shields for ASDEX, the Joint European Torus (JET), TEXTOR, and Alcator-C. We have measured the magnetic flux and observed loading at various operating resistances by using dielectric sheets or magnetic-coupled loads. Each Faraday shield effects a net change in the characteristic inductance of the antenna, resulting in a reduction of wave coupling. However, the load experienced by the antenna is not always reduced because the Faraday shield itself acts as a load. We differentiate between these effects experimentally. The net result of the study is that the Faraday shields now in use cost up to a factor of 50% of coupling. This, of course, reduces the power handling capability by 50% as well. However, configurations exist that are easily cooled and result in a reduction of less than 5% in loading

  9. Bulk-shield design for the Fusion Materials Irradiation Test facility

    International Nuclear Information System (INIS)

    Carter, L.L.; Mann, F.M.; Morford, R.J.; Johnson, D.L.; Huang, S.T.

    1982-07-01

    The accelerator-based Fusion Materials Irradiation Test (FMIT) facility will provide a high-fluence, fusion-like radiation environment for the testing of materials. While the neutron spectrum produced in the forward direction by the 35 MeV deuterons incident upon a flowing lithium target is characterized by a broad peak around 14 MeV, a high energy tail extends up to about 50 MeV. Some shield design considerations are reviewed

  10. Layer-splitting technique for testing the recursive scheme for multilayer shields gamma ray buildup factors

    International Nuclear Information System (INIS)

    Alkhatib, Sari F.; Park, Chang Je; Jeong, Hae Yong; Lee, Yongdeok

    2016-01-01

    Highlights: • A simple formalism is suggested for the recursive approach and then it is used to produce buildup factors for certain multilayer shields. • The newly layer-splitting technique is implemented on the studied cases for testing the suggested formalism performance. • The buildup factors are generated using cubic polynomial fitting functions that are produced based on previous well-acknowledge data. - Abstract: This study illustrates the implementation of the newly suggested layer-splitting testing technique. This technique is introduced in order to be implemented in examining suggested formalisms for the recursive scheme (or iterative scheme). The recursive scheme is a concept used in treating and producing the gamma ray buildup factors in the case of multilayer shields. The layer-splitting technique simply enforces the scheme to treat a single layer of one material as two separated layers with similar characteristics. Thus it subjects the scheme to an abnormal definition of the multilayer shield that will test its performance in treating the successive layers. Thus, it will act as a method of verification for the approximations and assumptions taken in consideration. A simple formalism was suggested for the recursive scheme then the splitting technique was implemented on it. The results of implementing both the suggested formalism and the splitting technique are then illustrated and discussed. Throughout this study, cubic polynomial fitting functions were used to generate the data of buildup factors for the basic single-media that constitute the multilayer shields understudy. This study is limited to the cases of multiple shields consisting of repeated consecutive thin layers of lead–water and iron–water shields for 1 MeV gamma rays. The produced results of the buildup factor values through the implementation of the suggested formalism showed good consistency with the Monte Carlo simulation results of Lin and Jiang work. In the implementation of

  11. Peripheral dose to the testes: the design and clinical use of a practical and effective gonadal shield

    International Nuclear Information System (INIS)

    Fraass, B.A.; Kinsella, T.J.; Harrington, F.S.; Galtstein, E.

    1985-01-01

    A simple and practical gonadal shield has been developed for use near megavoltage radiation fields. The lead shield encloses only the testes, allowing its use with nearly any radiation field that does not include the testes. The dose to the testes with and without the shield has been measured extensively both in phantoms and on patients. The gonadal shield allows a 3 to 10-fold reduction in dose to the testes depending primarily on the distance from the field edge to the gonads. When the shield is used, the gonadal dose is always less than 1% of the patient's prescription dose. Based on our patient studies of testicular injury following conventionally-fractionated irradiation, a dose of less than 50 cGy should preserve normal testicular function

  12. Peripheral dose to the testes: the design and clinical use of a practical and effective gonadal shield.

    Science.gov (United States)

    Fraass, B A; Kinsella, T J; Harrington, F S; Glatstein, E

    1985-03-01

    A simple and practical gonadal shield has been developed for use near megavoltage radiation fields. The lead shield encloses only the testes, allowing its use with nearly any radiation field that does not include the testes. The dose to the testes with and without the shield has been measured extensively both in phantoms and on patients. The gonadal shield allows a 3 to 10-fold reduction in dose to the testes depending primarily on the distance from the field edge to the gonads. When the shield is used, the gonadal dose is always less than 1% of the patient's prescription dose. Based on our patient studies of testicular injury following conventionally-fractionated irradiation, a dose of less than 50 cGy (1% of a typical 5000 cGy treatment regimen) should preserve normal testicular function.

  13. Nuclear Rocket Test Facility Decommissioning Including Controlled Explosive Demolition of a Neutron-Activated Shield Wall

    International Nuclear Information System (INIS)

    Michael Kruzic

    2007-01-01

    Located in Area 25 of the Nevada Test Site, the Test Cell A Facility was used in the 1960s for the testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program. The facility was decontaminated and decommissioned (D and D) in 2005 using the Streamlined Approach For Environmental Restoration (SAFER) process, under the Federal Facilities Agreement and Consent Order (FFACO). Utilities and process piping were verified void of contents, hazardous materials were removed, concrete with removable contamination decontaminated, large sections mechanically demolished, and the remaining five-foot, five-inch thick radiologically-activated reinforced concrete shield wall demolished using open-air controlled explosive demolition (CED). CED of the shield wall was closely monitored and resulted in no radiological exposure or atmospheric release

  14. The ParaShield Entry Vehicle Concept: Basic Theory and Flight Test Development

    OpenAIRE

    Akin, David

    1990-01-01

    With the emergence of microsatellite launch vehicle technology and the development of interest in space commercialization, there is a renewed need for entry vehicle technology to return mass from low earth orbit. This paper documents the ParaShield concept of the Space Systems Laboratory, which is an ultra-low ballistic coefficient (ULβ) entry vehicle. Trajectory simulations show that as the ballistic coefficient is lowered into the range of 100-150 Pa (2-3lb/ft2) the total heat load and peak...

  15. Generation and testing of the shielding data library EURLIB for fission and fusion technology

    International Nuclear Information System (INIS)

    Caglioti, E.; Hehn, G.; Herrnberger, V.; Mattes, M.; Nicks, R.; Penkuhn, H.

    1977-01-01

    For the common field of core physics and shielding, the CSEWG group structure of 239 fast neutron groups had been proposed, of which the 100 neutron groups of the EURLIB Library is a sub-set for shielding. This standard group Library EURLIB had been initiated by the NEA-specialist group on shielding benchmarks in 1974. The wide acceptance of the Library for interpretation of benchmarks in the NEA program represents an important step forward in the standardization of group data which is the basic requirement for a useful collaboration. On the other side the interpretation of a series of different benchmark experiments with the EURLIB Library provides the best check of the cross section data for neutron and gamma-rays showing the needs for further improvements. The paper describes the joint work of IKE, Stuttgart and EURATOM, Ispra in generating multigroup libraries for neutron and gamma-rays. Special effort has been devoted to improve the flux weighting for both types of radiation and proper treatment of thermal neutrons. The coupled multigroup Library of 100 neutron and 20 gamma groups is collapsed into few group structures for typical designs of LWR, LMFBR, gas cooled and thermonuclear reactors. The work for optimal few group representation is done in cooperation with EIR, Wurenlingen. The testing of the EURLIB Library is a common effort of several institutions participating in the NEA shielding benchmark program

  16. Neutron shieldings

    International Nuclear Information System (INIS)

    Tarutani, Kohei

    1979-01-01

    Purpose: To decrease the stresses resulted by the core bendings to the base of an entrance nozzle. Constitution: Three types of round shielding rods of different diameter are arranged in a hexagonal tube. The hexagonal tube is provided with several spacer pads receiving the loads from the core constrain mechanism at its outer circumference, a handling head for a fuel exchanger at its top and an entrance nozzle for self-holding the neutron shieldings and flowing heat-removing coolants at its bottom. The diameters for R 1 , R 2 and R 3 for the round shielding rods are designed as: 0.1 R 1 2 1 and 0.2 R 1 2 1 . Since a plurality of shielding rods of small diameter are provided, soft structure are obtained and a plurality of coolant paths are formed. (Furukawa, Y.)

  17. Experimental Studies of the Aerothermal Characteristics of the Project Orion CEV heat Shield in High Speed Transitional and Turbulent Flows

    Science.gov (United States)

    Wadhams, T.P.; MacLean, M.; Holden, M.S.; Cassady, A.M.

    2009-01-01

    An experimental program has been completed by CUBRC exploring laminar, transitional, and turbulent flows over a 7.0% scale model of the Project ORION CEV geometry. This program was executed primarily to answer questions concerning the increase in heat transfer on the windward, or "hot shoulder" of the CEV heat shield from laminar to turbulent flow. To answer these questions CUBRC constructed and instrumented a 14.0 inch diameter Project ORION CEV model and ran a range of Reynolds numbers based on diameter from 1.0 to over 40 million at a Mach number of 8.0. These Reynolds numbers were selected to cover laminar to turbulent heating data on the "hot shoulder". Data obtained during these runs will be used to guide design decisions as they apply to heat shield thickness and extent. Several experiments at higher enthalpies were achieved to obtain data for code validation with real gas effects and transition. CUBRC also performed computation studies of these experiments to aid in the data reduction process and study turbulence modeling.

  18. Design heating test section HeaTiNG-02

    International Nuclear Information System (INIS)

    Riswan Djambiar; Sagino; Dedy Haryanto; Joko Prasetio Witoko

    2013-01-01

    HeaTiNG-02 is a component test loop BETA which serves as a heater in conducting experimental heat transfer processes in two-phase flow in narrow slit-shaped plate, considering this phenomenon is one of the conditions postulated accident scenarios a NPP type PWR. To produce heat for the heating component takes the AC power the source voltage can be set from 0 Volts to 220 Volts with no more than a maximum power of 25 KVA. To obtain the thermal conditions on HeaTiNG-02 heating wire dimensions need to be determined and the corresponding voltage so that it will an expected power. Determination of the dimensions of the heater wire through calculations using electricity formulations. Retrieved draft heater test BETA (UUB) HeaTiNG-02 use material super kanthal (FeCuAl) in diameter (Ø) = 2 mm and wire length 3770 mm. Voltage regulators with a maximum power of 25 kVA with a minimum voltage of 0 volts and a maximum of 220 volts. Heater is used as the base to form refractory stone trench. (author)

  19. Integral data testing of JENDL-3.2 for fusion reactor and shielding applications

    International Nuclear Information System (INIS)

    Oyama, Yukio

    1995-01-01

    Integral data testing of JENDL-3.2 is being performed in the activities of two working groups of the Japanese Nuclear Data Committee. The continuous and group-wise libraries prepared from JENDL-3.2 are planned to be tested by the working groups. In this paper, the continuous library FSXLIB-J3R2 processed from JENDL-3.2 for MCNP was tested for fission and fusion neutrons using data of integral experiments and compared to the results of JENDL-3.1. The results of integral data testing of JENDL-3.2 for fusion and shielding application are reviewed. (author)

  20. Studies of the ultrasonic testing scheme on bonding quality in shield blanket of ITER

    International Nuclear Information System (INIS)

    Shi Sichao; Shen Jingling; He Fengqi; Jin Wanping

    2007-01-01

    International Thermonuclear Experimental Reactor (ITER) is an international cooperative item. One of its components, the First Wall (FW) functioning as neutron shielding and cooling, is an important part. According to the component materials, structural features, testing requirements of the FW, and the ultrasonic propagation characteristics, it is suggested that Broad-band ultrasonic can be used to test the bonding quality of the FW. According to the case mentioned above, the Broad-band Ultrasonic Testing scheme was presented, and the ultrasonic testing feasibility was analyzed theoretically in this paper. (authors)

  1. Performance testing of a hydrogen heat pipe

    International Nuclear Information System (INIS)

    Alario, J.; Kosson, R.

    1980-01-01

    Test results are presented for a reentrant groove heat pipe with hydrogen working fluid. The heat pipe became operational between 20 and 30 K after a cooldown from 77 K without any difficulty. Steady-state performance data taken over a 19 to 23 K temperature range indicated the following: (1) maximum heat transport capacity 5.4 W-m (2) static wicking height 1.42 cm and (3) overall heat pipe conductance 1.7 W/C. These data agreed remarkably well with extrapolations made from comparable ammonia test results. The maximum heat transport capacity is 9.5% larger than the extrapolated value, but the static wicking height is the same. The overall conductance is 29% of the ammonia value, which is close to the ratio of liquid thermal conductivities (24%). Also, recovery from a completely frozen condition was accomplished within 5 min by simply applying an evaporater heat load of 1.8 W

  2. Comparison of heat-testing methodology.

    Science.gov (United States)

    Bierma, Mark M; McClanahan, Scott; Baisden, Michael K; Bowles, Walter R

    2012-08-01

    Patients with irreversible pulpitis occasionally present with a chief complaint of sensitivity to heat. To appropriately diagnose the offending tooth, a variety of techniques have been developed to reproduce this chief complaint. Such techniques cause temperature increases that are potentially damaging to the pulp. Newer electronic instruments control the temperature of a heat-testing tip that is placed directly against a tooth. The aim of this study was to determine which method produced the most consistent and safe temperature increase within the pulp. This consistency facilitates the clinician's ability to differentiate between a normal pulp and irreversible pulpitis. Four operators applied the following methods to each of 4 extracted maxillary premolars (for a total of 16 trials per method): heated gutta-percha, heated ball burnisher, hot water, and a System B unit or Elements unit with a heat-testing tip. Each test was performed for 60 seconds, and the temperatures were recorded via a thermocouple in the pulp chamber. Analysis of the data was performed by using the intraclass correlation coefficient. The least consistent warming was found with hot water. The heat-testing tip also demonstrated greater consistency between operators compared with the other methods. Hot water and the heated ball burnisher caused temperature increases high enough to damage pulp tissue. The Elements unit with a heat-testing tip provides the most consistent warming of the dental pulp. Copyright © 2012 American Association of Endodontists. Published by Elsevier Inc. All rights reserved.

  3. Tests of shielding effectiveness of Kevlar and Nextel onboard the International Space Station and the Foton-M3 capsule.

    Science.gov (United States)

    Pugliese, M; Bengin, V; Casolino, M; Roca, V; Zanini, A; Durante, M

    2010-08-01

    Radiation assessment and protection in space is the first step in planning future missions to the Moon and Mars, where mission and number of space travelers will increase and the protection of the geomagnetic shielding against the cosmic radiation will be absent. In this framework, the shielding effectiveness of two flexible materials, Kevlar and Nextel, were tested, which are largely used in the construction of spacecrafts. Accelerator-based tests clearly demonstrated that Kevlar is an excellent shield for heavy ions, close to polyethylene, whereas Nextel shows poor shielding characteristics. Measurements on flight performed onboard of the International Space Station and of the Foton-M3 capsule have been carried out with special attention to the neutron component; shielded and unshielded detectors (thermoluminescence dosemeters, bubble detectors) were exposed to a real radiation environment to test the shielding properties of the materials under study. The results indicate no significant effects of shielding, suggesting that thin shields in low-Earth Orbit have little effect on absorbed dose.

  4. Structural analysis of the Passive Magnetic Shield for the ITER Heating Neutral Beam Injector system

    Energy Technology Data Exchange (ETDEWEB)

    Cabrera, Santiago, E-mail: santiago.cabrera@ciemat.es [CIEMAT Laboratorio Nacional de Fusión, Avda. Complutense 40, 28040 Madrid (Spain); Rincón, Esther; Ahedo, Begoña; Alonso, Javier; Barrera, Germán; Ramos, Francisco; Ríos, Luis [CIEMAT Laboratorio Nacional de Fusión, Avda. Complutense 40, 28040 Madrid (Spain); El-Ouazzani, Anass; García, Pablo [ITER Organization, Route de Vinon-sur-Verdon – CS 90 046, 13067 St Paul Lez Durance Cedex (France); Agarici, Gilbert [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3 – 07/08, 08019 Barcelona (Spain)

    2015-10-15

    The ITER Passive Magnetic Shield (PMS) main function is to protect the Neutral Beam Injector (NBI) from the external magnetic field coming from the tokamak, and to shield the NB cell from the radiation coming from all activated components. The shielding from the external magnetic field is performed in association with the Active Compensation Cooled Correction Coils (ACCC). The Bushing and Transmission Line (TL) PMS also provides structural support for HV bushing, allowing its maintenance and providing air sealing function between NBI cell and High Voltage deck room. The paper summarizes the structural analyses performed in order to evaluate the mechanical behaviour of the HNB PMS under operation combined with seismic event. The RCC-MR Code is used to validate the design, assuming creep is negligible, since the structure is expected to be at room temperature. P-type damage is assessed.

  5. Thermal design of top shield for PFBR

    International Nuclear Information System (INIS)

    Gajapathy, R.; Jalaludeen, S.; Selvaraj, A.; Bhoje, S.B.

    1988-01-01

    India's Liquid Metal Cooled Fast Breeder Reactor programme started with the construction of loop type 13MW(e) Fast Breeder Test Reactor (FBTR) which attained criticality in October 1985. With the experience of FBTR, the design work on pool type 500 MW(e) Prototype Fast Breeder Reactor (PFBR) which will be a forerunner for future commercial fast breeder reactors, has been started. The Top Shield forms the cover for the main vessel which contains the primary circuit. Argon cover gas separates the Top Shield from the free level of hot sodium pool (803K). The Top Shield which is of box type construction consists of control plug, two rotatable plugs and roof slab, assembled together, which provide biological shielding, thermal shielding and leak tight containment at the top of the main vessel. Heat is transferred from the sodium pool to the Top Shield through argon cover gas and through components supported by it and dipped in the sodium pool. The Top Shield should be maintained at the desired operating temperature by incorporating a cooling system inside it. Insulation may be provided below the bottom plate to reduce the heat load to the cooling system, if required. The thermal design of Top Shield consists of estimation of heat transfer to the Top Shield, selection of operating temperature, assessment of insulation requirement, design of cooling system and evaluation of transient temperature changes

  6. Innovative technologies for Faraday shield cooling

    International Nuclear Information System (INIS)

    Rosenfeld, J.H.; Lindemuth, J.E.; North, M.T.; Goulding, R.H.

    1995-01-01

    Alternative advanced technologies are being evaluated for use in cooling the Faraday shields used for protection of ion cyclotron range of frequencies (ICR) antennae in Tokamaks. Two approaches currently under evaluation include heat pipe cooling and gas cooling. A Monel/water heat pipe cooled Faraday shield has been successfully demonstrated. Heat pipe cooling offers the advantage of reducing the amount of water discharged into the Tokamak in the event of a tube weld failure. The device was recently tested on an antenna at Oak Ridge National Laboratory. The heat pipe design uses inclined water heat pipes with warm water condensers located outside of the plasma chamber. This approach can passively remove absorbed heat fluxes in excess of 200 W/cm 2 ;. Helium-cooled Faraday shields are also being evaluated. This approach offers the advantage of no liquid discharge into the Tokamak in the event of a tube failure. Innovative internal cooling structures based on porous metal cooling are being used to develop a helium-cooled Faraday shield structure. This approach can dissipate the high heat fluxes typical of Faraday shield applications while minimizing the required helium blower power. Preliminary analysis shows that nominal helium flow and pressure drop can sufficiently cool a Faraday shield in typical applications. Plans are in progress to fabricate and test prototype hardware based on this approach

  7. Model-based analysis of nonstationary thermal mode in premises with an insolation passive heating system with a three-layer translucent shield

    International Nuclear Information System (INIS)

    Avezova, N.R.; Avezov, R.R.; Rashidov, Y.K. et al.

    2014-01-01

    The results of the model-based study of nonstationary thermal mode in premises with an insolation passive heating system with a three-layer translucent shield are presented. The article is aimed at determining daily variations in the air temperature of the heated premise on typical heating season days and analyzing the optimization of the thermal capacity of the short-term (daily) thermal battery of the heating system on this basis. (author)

  8. Radiation Resistance Test of Wireless Sensor Node and the Radiation Shielding Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Li, Liqan; Sur, Bhaskar [Atomic Energy of Canada Limited, Ontario (Canada); Wang, Quan [University of Western Ontario, Ontario (Canada); Deng, Changjian [The University of Electronic Science and Technology, Chengdu (China); Chen, Dongyi; Jiang, Jin [Applied Physics Branch, Ontario (Korea, Republic of)

    2014-08-15

    A wireless sensor network (WSN) is being developed for nuclear power plants. Amongst others, ionizing radiation resistance is one essential requirement for WSN to be successful. This paper documents the work done in Chalk River Laboratories of Atomic Energy of Canada Limited (AECL) to test the resistance to neutron and gamma radiation of some WSN nodes. The recorded dose limit that the nodes can withstand before being damaged by the radiation is compared with the radiation environment inside a typical CANDU (CANada Deuterium Uranium) power plant reactor building. Shielding effects of polyethylene, cadmium and lead to neutron and gamma radiations are also analyzed using MCNP simulation. The shielding calculation can be a reference for the node case design when high dose rate or accidental condition (like Fukushima) is to be considered.

  9. TIBER II/ETR [Engineering Test Reactor] nuclear shielding and optional tritium breeding system: An overview

    International Nuclear Information System (INIS)

    Lee, J.D.; Sawan, M.

    1987-01-01

    TIBER II, the Tokamak Ignition/Burn Experimental Reactor II, is a design concept developed as the US candidate for an International Engineering Test Reactor (ETR). An important objective of this design is to minimize cost by minimizing major radius while providing a wall loading greater than 1.0 MW/m2 and a total fluence greater than 3.0 MWY/m2 needed for blanket module testing. The shielding required for the superconducting TF coils is an important element in setting TIBER II's 3.0m major radius. 6 refs., 1 fig., 1 tab

  10. Shielding Gas and Heat Input Effects on the Mechanical and Metallurgical Characterization of Gas Metal Arc Welding of Super Martensitic Stainless Steel (12Cr5Ni2Mo) Joints

    Science.gov (United States)

    Prabakaran, T.; Prabhakar, M.; Sathiya, P.

    This paper deals with the effects of shielding gas mixtures (100% CO2, 100% Ar and 80 % Ar + 20% CO2) and heat input (3.00, 3.65 and 4.33kJ/mm) on the mechanical and metallurgical characteristics of AISI 410S (American Iron and Steel Institute) super martensitic stainless steel (SMSS) by gas metal arc welding (GMAW) process. AISI 410S SMSS with 1.2mm diameter of a 410 filler wire was used in this study. A detailed microstructural analysis of the weld region as well as the mechanical properties (impact, microhardness and tensile tests at room temperature and 800∘C) was carried out. The tensile and impact fracture surfaces were further analyzed through scanning electron microscope (SEM). 100% Ar shielded welds have a higher amount of δ ferrite content and due to this fact the tensile strength of the joints is superior to the other two shielded welds.

  11. Experimental study of carbon materials behavior under high temperature and VUV radiation: Application to Solar Probe+ heat shield

    International Nuclear Information System (INIS)

    Eck, J.; Sans, J.-L.; Balat-Pichelin, M.

    2011-01-01

    The aim of the Solar Probe Plus (SP+) mission is to understand how the solar corona is heated and how the solar wind is accelerated. To achieve these goals, in situ measurements are necessary and the spacecraft has to approach the Sun as close as 9.5 solar radii. This trajectory induces extreme environmental conditions such as high temperatures and intense Vacuum Ultraviolet radiation (VUV). To protect the measurement and communication instruments, a heat shield constituted of a carbon material is placed on the top of the probe. In this study, the physical and chemical behavior of carbon materials is experimentally investigated under high temperatures (1600-2100 K), high vacuum (10 -4 Pa) and VUV radiation in conditions near those at perihelion for SP+. Thanks to several in situ and ex situ characterizations, it was found that VUV radiation induced modification of outgassing and of mass loss rate together with alteration of microstructure and morphology.

  12. Investigation of effect of post weld heat treatment conditions on residual stress for ITER blanket shield blocks

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hun-Chea, E-mail: hcjung@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, Sa-Woong [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Yun-Hee [Division of Convergence Technology, Korea Research Institute of Standard and Science (KRISS), Daejeon (Korea, Republic of); Baek, Seung-Wook [Division of Industrial Metrology, Korea Research Institute of Standard and Science (KRISS), Daejeon (Korea, Republic of); Ha, Min-Su; Shim, Hee-Jin [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    Highlights: • PWHT for ITER blanket shield block should be performed for dimensional stability. • Investigation of the effect of PWHT conditions on properties was performed. • Instrumented indentation method for evaluation of properties was used. • Residual stress and hardness decreased with increasing PWHT temperature. • Optimization of PWHT conditions would be needed for satisfaction of requirement. - Abstract: The blanket shield block (SB) shall be required the tight tolerance because SB interfaces with many components, such as flexible support keypads, First Wall (FW) support contact surfaces, FW central bolt, electrical strap contact surfaces and attachment inserts for both FW and Vacuum Vessel (VV). In order to fulfil the tight tolerance requirement, stress relieving shall be performed for dimensional stability after cover welding operation. In this paper, effect of Post Weld Heat Treatment (PWHT) conditions, temperature and holding time, was investigated on the residual stress and hardness. The 316L Stainless Steel (SS) was prepared and welded by manual TIG welding by using filler material with 2.4 mm of diameter. Welded 316L SS plate was machined to prepare the specimen for PWHT. PWHT was implemented at 250, 300, 400 °C for 2 and 3 h (400 °C only) and residual stress after relaxation were determined. The evaluation of residual stress and hardness for each specimen was carried out by instrumented indentation technique. The residual stress and hardness were decreased with increasing the heat treatment temperature and holding time.

  13. Heat shields for aircraft - A new concept to save lives in crash fires.

    Science.gov (United States)

    Neel, C. B.; Parker, J. A.; Fish, R. H.; Henshaw, J.; Newland, J. H.; Tempesta, F. L.

    1971-01-01

    A passenger compartment surrounded by a fire-retardant shell, to protect the occupants long enough for the fire to burn out or for fire-fighting equipment to reach the aircraft and extinguish it, is proposed as a new concept for saving lives in crash fires. This concept is made possible by the recent development of two new fire-retardant materials: a very lightweight foam plastic, called polyisocyanurate foam, and an intumescent paint. Exposed to heat, the intumescent paint expands to many times its original thickness and insulates the surface underneath it. Demonstration tests are illustrated, described and discussed. However, some problems, such as preventing fuselage rupture and protecting windows, must be solved before such a system can be used.

  14. Radiation shielding lead shield

    International Nuclear Information System (INIS)

    Dei, Shoichi.

    1991-01-01

    The present invention concerns lead shields for radiation shielding. Shield boxes are disposed so as to surround a pipeline through which radioactive liquids, mists or like other objects are passed. Flanges are formed to each of the end edges of the shield boxes and the shield boxes are connected to each other by the flanges. Upon installation, empty shield boxes not charged with lead particles and iron plate shields are secured at first at the periphery of the pipeline. Then, lead particles are charged into the shield boxes. This attains a state as if lead plate corresponding to the depth of the box is disposed. Accordingly, operations for installation, dismantling and restoration can be conducted in an empty state with reduced weight to facilitate the operations. (I.S.)

  15. Monte Carlo based demonstration of sufficiently dimensioned shielding for a Co-60 testing facility

    International Nuclear Information System (INIS)

    Wind, Michael; Beck, Peter; Latocha, Marcin

    2015-01-01

    The electrical properties of electronic equipment can be changed in an ionized radiation field. The knowledge of these changes is necessary for applications in space, in air traffic and nuclear medicine. Experimental tests will be performed in Co-60 radiation fields in the irradiation facility (TEC facility) of the Seibersdorf Labor GmbH that is in construction. The contribution deals with a simulation that is aimed to calculate the local dose rate within and outside the building for demonstration of sufficient dimensioning of the shielding in compliance with the legal dose rate limits.

  16. Wake Shield Target Protection

    International Nuclear Information System (INIS)

    Valmianski, Emanuil I.; Petzoldt, Ronald W.; Alexander, Neil B.

    2003-01-01

    The heat flux from both gas convection and chamber radiation on a direct drive target must be limited to avoid target damage from excessive D-T temperature increase. One of the possibilities of protecting the target is a wake shield flying in front of the target. A shield will also reduce drag force on the target, thereby facilitating target tracking and position prediction. A Direct Simulation Monte Carlo (DSMC) code was used to calculate convection heat loads as boundary conditions input into ANSYS thermal calculations. These were used for studying the quality of target protection depending on various shapes of shields, target-shield distance, and protective properties of the shield moving relative to the target. The results show that the shield can reduce the convective heat flux by a factor of 2 to 5 depending on pressure, temperature, and velocity. The protective effect of a shield moving relative to the target is greater than the protective properties of a fixed shield. However, the protective effect of a shield moving under the drag force is not sufficient for bringing the heat load on the target down to the necessary limit. Some other ways of diminishing heat flux using a protective shield are discussed

  17. Development of neutron shielding material for cask

    International Nuclear Information System (INIS)

    Najima, K.; Ohta, H.; Ishihara, N.; Matsuoka, T.; Kuri, S.; Ohsono, K.; Hode, S.

    2001-01-01

    Since 1980's Mitsubishi Heavy Industries, Ltd (MHI) has established transport and storage cask design 'MSF series' which makes higher payload and reliability for long term storage. MSF series transport and storage cask uses new-developed neutron shielding material. This neutron shielding material has been developed for improving durability under high condition for long term. Since epoxy resin contains a lot of hydrogen and is comparatively resistant to heat, many casks employ epoxy base neutron shielding material. However, if the epoxy base neutron shielding material is used under high temperature condition for a long time, the material deteriorates and the moisture contained in it is released. The loss of moisture is in the range of several percents under more than 150 C. For this reason, our purpose was to develop a high durability epoxy base neutron shielding material which has the same self-fire-extinction property, high hydrogen content and so on as conventional. According to the long-time heating test, the weight loss of this new neutron shielding material after 5000 hours heating has been lower than 0.04% at 150 C and 0.35% at 170 C. A thermal test was also performed: a specimen of neutron shielding material covered with stainless steel was inserted in a furnace under condition of 800 C temperature for 30 minutes then was left to cool down in ambient conditions. The external view of the test piece shows that only a thin layer was carbonized

  18. Study of temperature effect on the physical properties of ilmenite-serpentine heat resistant concrete radiation shields

    International Nuclear Information System (INIS)

    Kany, A.M.I.; EL-Fouly, M.M.; EL-Gohary, M.I.; Makatious, A.S.; Kamal, S.M.

    1990-01-01

    A series of experimental studies have been carried out to determine the change in unit weigh, compressive strength, water content and neutron macroscopic cross section of a new type of concrete shields made from egyptian ilmenite and serpentine ores when heated for long period at temperatures up to 600 degree C. Results show that the unit weight of the cure concrete has a value of 2.98 Ton/M 3 and decreases with increasing temperature, while the compressive strength reaches a maximum value of 19 Ton/M 2 at 100 degree C. The differential thermal analysis (D.T.A.) of this concrete shows three endothermic peaks at 100 degree C, 48 degree C and 740 degree C. Also, the thermogravimetry analysis (T.G.A.) shows that the cure concrete retains about 11% water content of the total sample weigh and still retains 4.5% of its initial value when heated for long period at 600 degree C. Results also show that the neutron macroscopic cross section (for neutrons of energies < 1 MeV) of the ilmenite-serpentine heat resistant concrete decreases to 18.6% of its initial value after heating to 600 degree C

  19. Test of magnetic shielding cases for a 3'' phototube attached to a lead glass counter

    International Nuclear Information System (INIS)

    Ogawa, K.; Sumiyoshi, T.; Takasaki, F.

    1985-09-01

    Effect of a magnetic shielding for a phototube of 3'' diameter attached to a lead glass counter has been studied using permalloy shielding cases with two kinds of shapes. Both cases show sufficient shielding effect with magnetic field up to around 30 gauss. (author)

  20. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

  1. I2S-LWR Activation Analysis of Heat Exchangers Using Hybrid Shielding Methodology with SCALE6.1

    International Nuclear Information System (INIS)

    Matijevic, M.; Pevec, D.; Jecmenica, R.

    2016-01-01

    The Integral Inherently Safe Light Water Reactor (I2S-LWR) concept developed by Georgia Tech is a novel PWR reactor delivering electric power of 1000 MWe while implementing inherent safety features typical for Generation III+ small modular reactors. The main safety feature is based on integral primary circuit configuration, bringing together compact design of the reactor core with 121 fuel assembly (FA), control rod drive mechanism (CRDM), 8 primary heat exchangers (PHE), 4 passive decay heat removal systems (DHRS), 8 pumps, and other integral components. A high power density core based on silicide fuel is selected to achieve a high thermal power which is extracted with PHEs placed in the annual region between the barrel and the vessel. The complex and integrated design of I2S-LWR leads to activation of integral components, mainly made from stainless steel, so accurate and precise Monte Carlo (MC) simulations are needed to quantify potential dose rates to personnel during routine maintenance operation. This shielding problem is therefore very challenging one, posing a non-trivial neutron flux solution in a phase space. This paper presents the performance of the hybrid shielding methodologies CADIS/FW-CADIS implemented in the MAVRIC sequence of the SCALE6.1 code package. The main objective was to develop a detailed MC shielding model of the I2S-LWR reactor along with effective variance reduction (VR) parameters and to calculate neutron fluence rates inside PHEs. Such results are then utilized to find neutron activation rate distribution via 60Co generation inside of a stack of microchannel heat exchangers (MCHX), which will be periodically withdrawn for the maintenance. 59Co impurities are the main cause of (n,gamma) radiative gamma dose to personnel via neutron activation since 60Co has half-life of 5.27 years and is emitting high energy gamma rays (1.17 MeV and 1.33 MeV). The developed MC model was successfully used to find converged fluxes inside all 8 stacks of

  2. A new practical model of testes shield: the effectiveness during abdominopelvic computed tomography.

    Science.gov (United States)

    Sancaktutar, Ahmet Ali; Bozkurt, Yaşar; Önder, Hakan; Söylemez, Haluk; Atar, Murat; Penbegül, Necmettin; Ziypak, Tevfik; Tekbaş, Güven; Tepeler, Abdülkadir

    2012-01-01

    The goal of our prospective study was to measure the effect of a new standard model male gonad shield on the testicular radiation exposure during routine abdominopelvic computed tomography (CT). Two hundred male patients who underwent upper abdominal and pelvic CT examinations were included in our study. To prepare the testes shield (TS), 2 No. 8 fluoroscopy radiation-protection gloves made of bismuth (0.35 mm lead equivalent) were used. These gloves were invaginated into one another and their fingers were turned inside out. Scrotums of all patients were pushed into these lead-containing gloves. Upper abdominal CT (n = 6), pelvic CT (n = 9), and abdominopelvic scanning (n = 185) were performed. Immediately after the CT examinations and at postprocedural day 1, the scrotal examinations were repeated. None of the patients exhibited scrotal laceration, edema, eruption, erythema, tenderness, or pain. During the CT examinations, 22 patients (11%) felt unrest because of their exposed genital regions, without any adverse effect on the procedure. Dosimetric measurements of radioactivity inside the TS (dosimeter I) and outside it (dosimeter II) were 6.8 and 69.00 mSv, respectively. Accordingly, the TS we used in our study reduced the radiation exposure of the testes by 90.2% (10.1 times). We think that the use of this radioprotective TS during radiological diagnostic and therapeutic procedures is an appropriate approach from both a medical and legal perspective. Therefore, we recommend this userfriendly, practical, low-cost, and effective TS for all radiologic procedures.

  3. Development and testing of multigroup library with correction of self-shielding effects in fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Zou Jun; He Zhaozhong; Zeng Qin; Qiu Yuefeng; Wang Minghuang

    2010-01-01

    A multigroup library HENDL2.1/SS (Hybrid Evaluated Nuclear Data Library/Self-Shielding) based on ENDF/B-VII.0 evaluate data has been generated using Bondarenko and flux calculator method for the correction of self-shielding effect of neutronics analyses. To validate the reliability of the multigroup library HENDL2.1/SS, transport calculations for fusion-fission hybrid system FDS-I were performed in this paper. It was verified that the calculations with the HENDL2.1/SS gave almost the same results with MCNP calculations and were better than calculations with the HENDL2.0/MG which is another multigroup library without self-shielding correction. The test results also showed that neglecting resonance self-shielding caused underestimation of the K eff , neutron fluxes and waste transmutation ratios in the multigroup calculations of FDS-I.

  4. Radiation shielding

    International Nuclear Information System (INIS)

    Yue, D.D.

    1979-01-01

    Details are given of a cylindrical electric penetration assembly for carrying instrumentation leads, used in monitoring the performance of a nuclear reactor, through the containment wall of the reactor. Effective yet economical shielding protection against both fast neutron and high-energy gamma radiation is provided. Adequate spacing within the assembly allows excessive heat to be efficiently dissipated and means of monitoring all potential radiation and gas leakage paths are provided. (UK)

  5. Phase Change Material Heat Exchanger Life Test

    Science.gov (United States)

    Lillibridge, Sean; Stephan, Ryan

    2009-01-01

    Low Lunar Orbit (LLO) poses unique thermal challenges for the orbiting space craft, particularly regarding the performance of the radiators. The IR environment of the space craft varies drastically from the light side to the dark side of the moon. The result is a situation where a radiator sized for the maximal heat load in the most adverse situation is subject to freezing on the dark side of the orbit. One solution to this problem is to implement Phase Change Material (PCM) Heat Exchangers. PCM Heat Exchangers act as a "thermal capacitor," storing thermal energy when there is too much being produced by the space craft to reject to space, and then feeding that energy back into the thermal loop when conditions are more favorable. Because they do not use an expendable resource, such as the feed water used by sublimators and evaporators, PCM Heat Exchangers are ideal for long duration LLO missions. In order to validate the performance of PCM Heat Exchangers, a life test is being conducted on four n-Pentadecane, carbon filament heat exchangers. Fluid loop performance, repeatability, and measurement of performance degradation over 2500 melt-freeze cycles will be performed.

  6. Eddy current testing of heat exchangers tubes

    International Nuclear Information System (INIS)

    Gouez, J.F.; Rieusset, A.; Groix, F.

    An automatic system for Eddy Current testing of heat exchangers tubes of warships was developed. The advantages are an exposure of the controller limited at the time required to put in place the system and a reduced time of control [fr

  7. Transmission test of the polyethylene shield against 40 and 65 MeV quasi monochrome neutron

    International Nuclear Information System (INIS)

    Nakao, Makoto; Nakamura, Takashi; Sakuya, Yoshimasa; Nauchi, Yasushi; Nakao, Noriaki; Tanaka, Susumu; Sakamoto, Yukio; Nakajima, Hiroshi; Nakane, Yoshihiro.

    1996-01-01

    Using 40 and 65 MeV quasi monochrome neutron of the AVF cyclotron installed at Takasaki Laboratory, Japan Atomic Energy Research Institute, the neutron energy spectra were measured after transmitting the polyethylene shield. Results of the shielding experiments using concrete and iron recognized as main shielding material were proposed previously. As data obtained in the experiments were useful for a bench-mark experiment to investigate for shielding calculation and sectional data set, a shielding calculation simulated with new experiment to compare with and investigate for the previous experimental data. As a result, it was found that calculation result of neutron flux transmitting through the polyethylene shield showed difference with increase of the shield thickness. And, reducing distance of the peak neutron was also found to be over-estimated in its calculation value, such as three and five times on 43 MeV at 120 and 180 cm thick, respectively. (G.K.)

  8. A Reliability Comparison of Classical and Stochastic Thickness Margin Approaches to Address Material Property Uncertainties for the Orion Heat Shield

    Science.gov (United States)

    Sepka, Steve; Vander Kam, Jeremy; McGuire, Kathy

    2018-01-01

    The Orion Thermal Protection System (TPS) margin process uses a root-sum-square approach with branches addressing trajectory, aerothermodynamics, and material response uncertainties in ablator thickness design. The material response branch applies a bond line temperature reduction between the Avcoat ablator and EA9394 adhesive by 60 C (108 F) from its peak allowed value of 260 C (500 F). This process is known as the Bond Line Temperature Material Margin (BTMM) and is intended to cover material property and performance uncertainties. The value of 60 C (108 F) is a constant, applied at any spacecraft body location and for any trajectory. By varying only material properties in a random (monte carlo) manner, the perl-based script mcCHAR is used to investigate the confidence interval provided by the BTMM. In particular, this study will look at various locations on the Orion heat shield forebody for a guided and an abort (ballistic) trajectory.

  9. Shielding performances analysis for the IFMIF test facility based on high-fidelity Monte Carlo neutronic calculations

    Energy Technology Data Exchange (ETDEWEB)

    Kondo, Keitaro, E-mail: kondo.keitaro@jaea.go.jp; Arbeiter, Frederik; Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng; Tian, Kuo

    2015-10-15

    Highlights: • A detailed geometry model with pipe penetrations and gaps was prepared for the IFMIF test cell. • The neutron streaming effect due to gaps and pipes with shielding plugs was investigated. • The present analysis revealed that the streaming effect can be mitigated by some counter measures. • Occupational workers can access to the room above the test cell during operation. - Abstract: The IFMIF Test Cell (TC) design was developed and optimized in the EVEDA phase, and finally the reference TC design was proposed. The present study is devoted to further investigations of open issues on the reference TC design. In order to examine the neutron streaming effect caused by pipe penetrations and gaps around removable shielding plugs, a new geometry model for neutronic analyses has been prepared directly from engineering CAD data by utilizing the McCad conversion software. All removable shielding plugs are separately described in the model and a detailed description of pipes was incorporated into the model. The calculation result suggests that the streaming effect is mitigated if the pipe penetration is designed appropriately, while the gaps around the shielding plugs above the TC have large impact on the radiation dose in the access cell. The concept of the reference TC design has been basically validated from the neutronics point of view, although the streaming effect should be compensated by the shielding capability of the test cell cover plate so that occupational workers can access to the access cell during operation.

  10. Shielding performances analysis for the IFMIF test facility based on high-fidelity Monte Carlo neutronic calculations

    International Nuclear Information System (INIS)

    Kondo, Keitaro; Arbeiter, Frederik; Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng; Tian, Kuo

    2015-01-01

    Highlights: • A detailed geometry model with pipe penetrations and gaps was prepared for the IFMIF test cell. • The neutron streaming effect due to gaps and pipes with shielding plugs was investigated. • The present analysis revealed that the streaming effect can be mitigated by some counter measures. • Occupational workers can access to the room above the test cell during operation. - Abstract: The IFMIF Test Cell (TC) design was developed and optimized in the EVEDA phase, and finally the reference TC design was proposed. The present study is devoted to further investigations of open issues on the reference TC design. In order to examine the neutron streaming effect caused by pipe penetrations and gaps around removable shielding plugs, a new geometry model for neutronic analyses has been prepared directly from engineering CAD data by utilizing the McCad conversion software. All removable shielding plugs are separately described in the model and a detailed description of pipes was incorporated into the model. The calculation result suggests that the streaming effect is mitigated if the pipe penetration is designed appropriately, while the gaps around the shielding plugs above the TC have large impact on the radiation dose in the access cell. The concept of the reference TC design has been basically validated from the neutronics point of view, although the streaming effect should be compensated by the shielding capability of the test cell cover plate so that occupational workers can access to the access cell during operation.

  11. Test results from a helium gas-cooled porous metal heat exchanger

    International Nuclear Information System (INIS)

    North, M.T.; Rosenfeld, J.H.; Youchison, D.L.

    1996-01-01

    A helium-cooled porous metal heat exchanger was built and tested, which successfully absorbed heat fluxes exceeding all previously tested gas-cooled designs. Helium-cooled plasma-facing components are being evaluated for fusion applications. Helium is a favorable coolant for fusion devices because it is not a plasma contaminant, it is not easily activated, and it is easily removed from the device in the event of a leak. The main drawback of gas coolants is their relatively poor thermal transport properties. This limitation can be removed through use of a highly efficient heat exchanger design. A low flow resistance porous metal heat exchanger design was developed, based on the requirements for the Faraday shield for the International Thermonuclear Experimental Reactor (ITER) device. High heat flux tests were conducted on two representative test articles at the Plasma Materials Test Facility (PMTF) at Sandia National Laboratories. Absorbed heat fluxes as high as 40 MW/m 2 were successfully removed during these tests without failure of the devices. Commercial applications for electronics cooling and other high heat flux applications are being identified

  12. The other side of the coin: urban heat islands as shields from extreme cold

    Science.gov (United States)

    Yang, J.; Bou-Zeid, E.

    2017-12-01

    Extensive studies focusing on urban heat islands (UHIs) during hot periods create a perception that UHIs are invariably hazardous to human health and the sustainability of cities. Consequently, cities have invested substantial resources to try to mitigate UHIs. These urban policies can have serious repercussions since the health risks associated with cold weather are in fact higher than for heat episodes, yet wintertime UHIs have hardly been explored. We combine ground observations from 12 U.S. cities and high-resolution simulations to show that UHIs not only warm urban areas in the winter, but also further intensify during cold waves by up to 1.32 ± 0.78 oC (mean ± standard deviation) at night. Urban heat islands serve as shelters against extreme colds and provide invaluable benefits of reducing health risks and heating demand. More importantly, our simulations indicate that standard UHI mitigation measures such as green or cool roofs reduce these cold time amenities to different extents. Cities, particularly in cool and cold temperate climates, should hence revisit policies and efforts that are only desgined for hot periods. A paradigm shift is urgently needed to give an equal weight to the wintertime benefits of UHIs in the sustainability and resilience blueprints of cities.

  13. Demonstration test on manufacturing 200 l drum inner shielding material for recycling of reactor operating metal scrap

    International Nuclear Information System (INIS)

    Umemura, A.; Kimura, K.; Ueno, H.

    1993-01-01

    Low-level reactor wastes should be safely recycled considering those resource values, the reduction of waste disposal volume and environmental effects. The reasonable recycling system of reactor operating metal scrap has been studied and it was concluded that the 200 liter drum inner shielding material is a very promising product for recycling within the nuclear industry. The drum inner shielding material does not require high quality and so it is expected to be easily manufactured by melting and casting from roughly sorted scrap metals. This means that the economical scrap metal recycling system can be achieved by introducing it. Furthermore its use will ensure safety because of being contained in a drum. In order to realize this recycling system with the drum inner shielding material, the demonstration test program is being conducted. The construction of the test facility, which consists of a melting and refining furnace, a casting apparatus, a machining apparatus etc., was finishing in September, 1992

  14. Development of a bellows assembly with RF-shield for KEKB II: abrasion and pumping down tests

    International Nuclear Information System (INIS)

    Suetsugu, Yusuke; Kanazawa, Ken-ichi; Kawahara, Masaharu; Harada, Yosuke; Kaneko, Motosada

    1997-01-01

    A bellows assembly with RF-shield as been designed and developed for the KEK B-factory (KEKB). The RF-shield is a usual finger-type but has special spring-fingers to press contact-fingers (shield-fingers) surely onto inner tube (beam tube). In a chain of design studies an abrasion test of the contact-fingers was performed in vacuum. A quantity of generated metal particles was estimated and expected to have little harm on the beam lifetime if the inner tube is coated with silver. The gas desorption rate and the residual gas components of the bellows assembly were also measured as a final bench test. The gas desorption rate of 1 - 1.5x10 -10 Pa·l/s/cm 2 was obtained after a bake at 150degC for 24 hours. (author)

  15. Test of a new gonad shield in radiographic hip joint examinations of sucklings and infants

    International Nuclear Information System (INIS)

    Bronsch, T.

    1977-01-01

    Preparation and application of a shield consisting of lead rubber are described. Using the shield, a considerable decrease of radiation exposure to male and female infants could be achieved. Therefore it is recommended for application in mass examinations of hip joints. (author)

  16. Shielding modefication and safety review on Mutsu

    International Nuclear Information System (INIS)

    Osanai, Masao

    1978-01-01

    The Japan Atomic Energy Commission requests strongly to repair the shielding and make general safety inspection on Mutsu after an accident of radiation leakage from the reactor. The content and procedure of this repair of shielding and general safety inspection are outlined. The neutron leakage location in the reactor proper, technical shielding investigation, conceptual design of relating shielding repair, the mock up test of the shielding on the neutron streaming, the final conceptual design of repair, the relating research and development experiment and the detailed basic design of repair are explained, comparing the original design and the modified one. The modified design depends on the experimental results of neutron streaming test between the reactor vessel and the primary shield. As for the general safety inspection, the functional test of control rod driving mechanism and other main components, the flaw detection for heat transfer tubes of the steam generator and primary cooling pipings are carried out in hardwares, and the integrity analysis of fuel assemblies, stress corrosion cracking of fuel claddings and primary cooling pipings, the natural circulation analysis of primary cooling system, and integrity check of the heat transfer tubes of steam generator are carried out in softwares. The burst test and the strength test after high temperature oxidation for fuel claddings made of stainless steel were carried out. (Nakai, Y.)

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

    International Nuclear Information System (INIS)

    Sagino; Dedy Haryanto; Riswan Djambiar; Edy Sumarno

    2013-01-01

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

  18. Investigation the effect of outdoor air infiltration on the heat-shielding characteristics the outer walls of high-rise buildings

    Science.gov (United States)

    Vytchikov, Yu. S.; Kostuganov, A. B.; Saparev, M. E.; Belyakov, I. G.

    2018-03-01

    The presented article considers the influence of infiltrated outdoor air on the heat-shielding characteristics of the exterior walls of modern residential and public buildings. A review of the sources devoted to this problem confirmed its relevance at the present time, especially for high-rise buildings. The authors of the article analyzed the effect of longitudinal and transverse air infiltration on the heat-shielding characteristics of the outer wall of a 25-story building that was built in Samara. The results showed a significant reduction of the reduced resistance to the heat transfer of the outer wall when air is infiltrated through it. There are the results of full-scale examination of external walls to confirm the calculated data. Based on the results of the study carried out by the authors of the article, general recommendations on the internal finishing of the outer walls of high-rise buildings are given.

  19. SU-E-T-243: Design of a Novel Testing Port for Radiation Protection and Shielding Measurements

    International Nuclear Information System (INIS)

    Tanny, S; Parsai, E; Harrell, D; Noller, J; Chopra, M

    2015-01-01

    Purpose: The majority of radiation shielding research utilizes Monte Carlo simulation because of the difficulty in eliminating secondary radiations from measurements. We have designed a test port into a primary barrier of our newest vault to allow for shielding measurements while ensuring adequate protection to the public and staff during normal machine operation. This port allows for measurement of attenuation values of shielding materials, differential dose albedos, and radiation scatter fractions. Methods: The vault design utilized the maze as part of a compound primary barrier. The test port is contained within the maze and is centered along isocenter. The inner 30 cm has a 20×20 cm 2 opening, while the remaining length has a 30×30 cm 2 opening. The block that contains the port has a density of 200 pcf to minimize internal scatter. The 30×30 cm 2 opening is occupied by removable 215 pcf concrete blocks. The innermost and outermost blocks activate an interlock wired into the beam-enable loop. This disallows beam-on in treatment mode if the interlock isn’t closed. The interlock can be overridden in service mode, or by-passed via an override switch in case of circuit failure. Results: The test port was installed in August. The beam is disabled when the interlock is tripped. Measurements taken when the primary beam is not incident on the port are indistinguishable from background. Ambient dose levels surrounding the vault with the designed shielding blocks in place are all within allowable limits for occupational workers. Conclusions: We have designed and installed a unique testing port for radiation protection and shielding measurements. This port is appropriately interlocked and designed to mitigate any risks of incidental exposure to staff or members of the public. The test port design allows measurements with “good geometry” and efficient removal of contaminating sources of radiation present in many shielding measurements. Daniel Harrell and Jim Noller

  20. Analysis of a shield design for a DT neutron generator test facility.

    Science.gov (United States)

    Chichester, D L; Pierce, G D

    2007-10-01

    Independent numerical simulations have been performed using the MCNP5 and SCALE5 radiation transport codes to evaluate the effectiveness of a concrete facility designed to shield personnel from neutron radiation emitted from DT neutron generators. The analysis considered radiation source terms of 14.1 MeV monoenergetic neutrons located at three discrete locations within the two test vaults in the facility, calculating neutron and photon dose rates at 44 locations around the facility using both codes. In addition, dose rate contours were established throughout the facility using the MCNP5 mesh tally feature. Neutron dose rates calculated outside of the facility are predicted to be below 0.01 mrem/h at all locations when all neutron generator source terms are operating within the facility. Similarly, the neutron dose rate in one empty test vault when the adjacent test vault is being utilized is also less then 0.01 mrem/h. For most calculation locations outside the facility the photon dose rates were less then the neutron dose rates by a factor of 10 or more.

  1. Manufacturing and testing of full scale prototype for ITER blanket shield block

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sa-Woong, E-mail: swkim12@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, Duck-Hoi; Jung, Hun-Chea [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Sung-Ki [WONIL Co., Ltd., Haman (Korea, Republic of); Kang, Sung-Chan [POSCO Specialty Steel Co., Ltd., Changwon (Korea, Republic of); Zhang, Fu; Kim, Byoung-Yoon [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ahn, Hee-Jae; Lee, Hyeon-Gon; Jung, Ki-Jung [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-04-15

    Highlights: • 316L(N)-IG forged steel was successfully fabricated and qualified. • Related R&D activities were implemented to resolve the fabrication issues. • SB #8 FSP was successfully manufactured with conventional fabrication techniques. • All of the validation tests were carried out and met the acceptance criteria. - Abstract: Based on the preliminary design of the ITER blanket shield block (SB) #8, the full scale prototype (FSP) has been manufactured and tested in accordance with pre-qualification program, and related R&D was performed to resolve the technical issues of fabrication. The objective of the SB pre-qualification program is to demonstrate the acceptable manufacturing quality by successfully passing the formal test program. 316L(N)-IG stainless steel forging blocks with 1.80L × 1.12W × 0.43t (m) were developed by using an electric arc furnace, and as a result, the material properties were satisfied with technical specification. In the course of applying conventional fabrication techniques such as cutting, milling, drilling and welding of the forged stainless steel block for the manufacturing of the SB #8 FSP, several technical problems have been addressed. And also, the hydraulic connector of cross-forged material re-melted by electro slag or vacuum arc requires the application of advanced joining techniques such as automatic bore TIG and friction welding. Many technical issues – drilling, welding, slitting, non-destructive test and so on – have been raised during manufacturing. Associated R&D including the computational simulation and coupon testing has been done in collaboration with relevant industries in order to resolve these engineering issues. This paper provides technical key issues and their possible resolutions addressed during the manufacture and formal test of the SB #8 FSP, and related R&D.

  2. An experimental study of Alfven wave heating using electrostatically shielded antennas in TCA

    International Nuclear Information System (INIS)

    Borg, G.G.; Joye, B.

    1990-01-01

    Despite the wide acceptance of electrostatic screens in ICRH for the protection of the plasma from the near fields of rf antennas, it has always been considered that low voltages at low frequency have made such screens unnecessary in Alfven wave heating (AWH). Despite this, AWH performs rather poorly as a heating method; the results being confused by a density rise up to 300 % of the target density. It is known that the density increase arises neither from impurity injection nor from a change in recycling. In addition, an extensive range of phenomena have been observed in the plasma scrape-off layer (SOL). During AWH, the SOL density is observed to decrease, the SOL floating potential is perturbed in a way that reflects the Alfven wave spectrum, the antennas charge negatively and draw a large current from the plasma and harmonics have been observed on the edge wave fields. The cause and correlation of these effects with each other and their impression on the bulk plasma response was not known. Experimental results from the TORTUS tokamak have indicated that the density increase might be eliminated by electrostatic screens. In their case, two AWH experiments were performed. In the first, an unshielded OFHC copper loop antenna was excited at a given power and, in the second, the same antenna was excited at the same power after installation of an aluminium, TiN coated, slotted screen. The density increase in the first case was shown to be completely eliminated in the second, although spectroscopic measurements revealed a difference in the plasma O(II) and Cu(I) content for each case. (author) 2 refs., 3 figs

  3. On the hydraulic behaviour of ITER Shield Blocks #14 and #08. Computational analysis and comparison with experimental tests

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128, Palermo (Italy); Merola, M.; Mitteau, R.; Raffray, R. [ITER Organization, Route de Vinon sur Verdon, 13067 Saint Paul, Lez Durance (France); Vallone, E., E-mail: eug.vallone@gmail.com [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128, Palermo (Italy)

    2016-11-01

    Highlights: • A benchmarking activity has been carried out focusing the attention on the cooling circuits of ITER Shield Blocks #08 and #14. • A theoretical-computational fluid-dynamic approach based on the Finite Volume Method has been followed, adopting a commercial code. • Hydraulic characteristic functions and spatial distributions of coolant mass flow rate, velocity and pressure drop have been assessed. • Results obtained have allowed code benchmarking for Blanket modules and the numerical predictions have been found to be generally lower than but quite close to the experimental results (lower than 10%). - Abstract: As a consequence of its position and functions, the ITER blanket system will be subjected to significant heat loads under nominal reference conditions. Therefore, the design of its cooling system is particularly demanding. Coolant water is distributed individually to the 440 blanket modules (BMs) through manifold piping, which makes it a highly parallelized system. The mass flow rate distribution is finely tuned to meet all operation constraints: adequate margin to burn out in the plasma facing components, even distribution of water flow among the so-called plasma-facing “fingers” of the Blanket First Wall panels, high enough water flow rate to avoid excessive water temperature in the outlet pipes, maximum allowable water velocity lower than 7 m/s in manifold pipes. Furthermore the overall pressure drop and flow rate in each BM shall be within the fixed specified design limit to avoid an unduly unbalance of cooling among the 440 modules. Analyses have to be carried out following a computational fluid-dynamic (CFD) approach based on the finite volume method and adopting a CFD commercial code to assess the thermal-hydraulic behaviour of each single circuit of the ITER blanket cooling system. This paper describes the code benchmarking needed to determine the best method to get reliable and timely results. Since experimental tests are

  4. Laboratory tests on neutron shields for gamma-ray detectors in space

    CERN Document Server

    Hong, J; Hailey, C J

    2000-01-01

    Shields capable of suppressing neutron-induced background in new classes of gamma-ray detectors such as CdZnTe are becoming important for a variety of reasons. These include a high cross section for neutron interactions in new classes of detector materials as well as the inefficient vetoing of neutron-induced background in conventional active shields. We have previously demonstrated through Monte-Carlo simulations how our new approach, supershields, is superior to the monolithic, bi-atomic neutron shields which have been developed in the past. We report here on the first prototype models for supershields based on boron and hydrogen. We verify the performance of these supershields through laboratory experiments. These experimental results, as well as measurements of conventional monolithic neutron shields, are shown to be consistent with Monte-Carlo simulations. We discuss the implications of this experiment for designs of supershields in general and their application to future hard X-ray/gamma-ray experiments...

  5. Modeling, Testing and Deploying a Multifunctional Radiation Shielding / Hydrogen Storage Unit, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — This project addresses two vital problems for long-term space travel activities: radiation shielding and hydrogen storage for power and propulsion. While both...

  6. High temperature glass thermal control structure and coating. [for application to spacecraft reusable heat shielding

    Science.gov (United States)

    Stewart, D. A.; Goldstein, H. E.; Leiser, D. B. (Inventor)

    1983-01-01

    A high temperature stable and solar radiation stable thermal control coating is described which is useful either as such, applied directly to a member to be protected, or applied as a coating on a re-usable surface insulation (RSI). It has a base coat layer and an overlay glass layer. The base coat layer has a high emittance, and the overlay layer is formed from discrete, but sintered together glass particles to give the overlay layer a high scattering coefficient. The resulting two-layer space and thermal control coating has an absorptivity-to-emissivity ratio of less than or equal to 0.4 at room temperature, with an emittance of 0.8 at 1200 F. It is capable of exposure to either solar radiation or temperatures as high as 2000 F without significant degradation. When used as a coating on a silica substrate to give an RSI structure, the coatings of this invention show significantly less reduction in emittance after long term convective heating and less residual strain than prior art coatings for RSI structures.

  7. Material shielding of power frequency magnetic fields: Research and testing results from the EPRI Power Delivery Center-Lenox. Final report

    International Nuclear Information System (INIS)

    Anderson, C.B.

    1998-06-01

    Extensive investigations of a variety of material shielding methods have been performed at the EPRI Power Delivery Center--Lenox, Massachusetts. This work is part of a larger shielding investigation being done for EPRI by Electric Research and Management, Inc. (ERM) as part of the Magnetic Field Management Target in the EPRI Environment Group. Part of this work, involving cylinders of material, is to be included in a shielding handbook being prepared by ERM. Material shielding tests, not included in the handbook, as well as additional material shielding research, including testing, analyses, and computer simulations performed at the EPRI Power Delivery Center--Lenox are documented here. One of the major complications of using materials to shield magnetic fields is the mathematical complexity of the phenomenon involved. The result is that analytical solutions exist only for a very small number of simple geometries such as spheres, infinitely long cylinders, and infinite sheets. In practice, the materials typically come in the form of sheets. At present, there are no analytical methods for directly determining the shielding effectiveness of finite sheets of material, however, EPRI is sponsoring work in this area. There are some methods based on conformal mapping which can provide a solution for simple two-dimensional sheets. While such methods are useful in gaining insight into the mechanisms of shielding, they are not realistic enough to provide accurate shielding estimates. Empirical techniques are still required to determine the shielding effectiveness of material sheets. The material shielding tests and computer simulations are described in the report. The results of these tests and simulations have been used to develop a number of material shielding design rules for use in practical applications

  8. Model surface conductivity effect for the electromagnetic heat shield in re-entry flight

    International Nuclear Information System (INIS)

    Matsuda, Atsushi; Otsu, Hirotaka; Kawamura, Masaaki; Konigorski, Detlev; Takizawa, Yuji; Abe, Takashi

    2008-01-01

    Effects of model surface conductivity on shock layer enhancement by an applied magnetic field in weakly ionized supersonic plasma flow with a large Hall parameter (β∼300) was investigated experimentally. The shock layer structures of test models of two kinds were measured using laser absorption spectroscopy, in the large Hall parameter situation. One was an insulated model; the other was a conductive spherical blunt model. The shock layer enhancement phenomenon by the applied magnetic field was more pronounced for the insulated model than for the conductive model. This tendency agrees with the computational fluid dynamics result, at least qualitatively

  9. Development of heat resistant concrete and its application to concrete casks. Improvement of neutron shielding performance of concrete in high temperature environment

    International Nuclear Information System (INIS)

    Owaki, Eiji; Hata, Akihito; Sugihara, Yutaka; Shimojo, Jun; Taniuchi, Hiroaki; Mantani, Kenichi

    2003-01-01

    Heat resistant concrete with hydrogen, which is able to shield neutron at more than 100degC, was developed. Using this new type concrete, a safety concrete cask having the same concept of metal casks was designed and produced. The new type cask omitted the inhalation and exhaust vent of the conventional type concrete casks. The new concrete consists of Portland cement added calcium hydroxide, iron powder and iron fiber. It showed 2.17 g/cm 3 density, 10.8 mass% water content, 1.4 W/(m·K) thermal conductivity at 150degC. Increasing of heat resistance made possible to produce the perfect sealing type structure, which had high shielding performance of radiation no consideration for streaming of radiation. Moreover, a monitor of sealing can be set. General view of concrete casks, outer view of 1/3 scaled model, cask storage system in the world, properties of new developed heat resistant concrete, results of shielding calculation are contained. (S.Y.)

  10. Pilot tests for dismantling by blasting of the biological shield of a shut down nuclear power station

    International Nuclear Information System (INIS)

    Freund, H.U.

    1995-01-01

    Following free-field tests on concrete blocks the feasibility of explosive dismantling of the biological shield of nuclear power stations has been succesfully tested at the former hotsteam reaction in Karlstein/Main Germany. For this purpose a model shield of scale 1:2 was embedded into the reactor structure at which bore-hole blasting tests employing up to about 15 kg of explosive were performed. An elaborate measurement system allowed to receive detailed information on the blast side-effects: Special emphasis was focussed on the quantitative registration of the dynamic blast loads; data for the transfer of the dismantling method to the removal of real ractor structures were obtained. (orig.) [de

  11. Estimating Orion Heat Shield Failure Due To Ablator Cracking During The EFT-1 Mission

    Science.gov (United States)

    Vander Kam, Jeremy C.; Gage, Peter

    2016-01-01

    The Orion EFT-1 heatshield suffered from two major certification challenges: First, the mechanical properties used in design were not evident in the flight hardware and second, the flight article itself cracked during fabrication. The combination of these events motivated the Orion Program to pursue an engineering-level Probabilistic Risk Assessment (PRA) as part of heatshield certification rationale. The PRA provided loss of Mission (LOM) likelihoods considering the probability of a crack occurring during the mission and the likelihood of subsequent structure over-temperature. The methods and input data for the PRA are presented along with a discussion of the test data used to anchor the results. The Orion program accepted an EFT-1 Loss of Vehicle (LOV) risk of 1-in-160,000 due to in-mission Avcoat cracking based on the results of this analysis. Conservatisms in the result, along with future considerations for Exploration Missions (EM) are also addressed.

  12. Ammunition Peculiar Equipment (APE) 1995, NIR Propellant Analyzer, to MIL-STD-398, Military Standard Shields, Operational for Ammunition Operations, Criteria for Design of and Tests for Acceptance

    National Research Council Canada - National Science Library

    2003-01-01

    ... (SJMAC-DEM) to test the Ammunition Peculiar Equipment (APE) 1995 NIR Propellant Analyzer, to MIL-STD-398, "Military Standard Shields, Operational for Ammunition Operations, Criteria for Design of and Tests for Acceptance...

  13. Validation test for carbon-14 migration and accumulation in a Canadian shield lake

    International Nuclear Information System (INIS)

    Bird, G.A.

    1996-09-01

    This particular BIOMOVS II Technical Report is concerned with modelling the transfer of C-14 through the aquatic food chain following release to a Canadian shield lake. Model performance has been tested against field data supplied by Atomic Energy of Canada Limited. Carbon-14 was added in 1978 to the epilimnion of a small Canadian Shield lake to investigate primary production and carbon dynamics. Data from this experiment were used within BIOMOVS II to provide a validation test, which involved modelling the fate of the C-14 added to the lake. The nature of the spike and the subsequent monitoring allowed the investigation of both short-term processes relevant to evaluation of the impacts of accidental releases as well as longer-term processes relevant to routine release and to solid waste disposal. Four models participated in the scenario: 1) a simple mass balance model of a lake (AECL, Whiteshell Laboratories, Canada); 2) a relatively complex deterministic dynamic compartment model (QuantiSci Ltd.,UK); 3) a complex deterministic model (Studsvik Model A) and a more complex probabilistic model (Studsvik Model B; Studsvik Eco and Safety AB, Sweden). Endpoints were C-14 concentrations in water, sediment and whitefish over a thirteen year period. Each model produced reasonable predictions when compared to the observed data and when uncertainty is taken into consideration. About 0.2 to 0.4% of the initial C-14 inventory to the lakes remained in the water at the end of the study, because of internal recycling of C-14 from sediments. The simple AECL model did not account for this internal recycling of C-14 and, in this respect, its predictions were not as realistic as those of the QuantiSci and Studsvik models for concentrations in water. However, the AECL model predictions for the C-14 inventory remaining in lake sediment were closest to the observed values. Overall, Studsvik Model B was the most accurate in simulating C-14 concentrations in water and in whitefish, but

  14. Validation test for carbon-14 migration and accumulation in a Canadian shield lake

    Energy Technology Data Exchange (ETDEWEB)

    Bird, G.A. [AECL, Pinawa, MB (Canada). Whiteshell Labs.] [and others

    1996-09-01

    This particular BIOMOVS II Technical Report is concerned with modelling the transfer of C-14 through the aquatic food chain following release to a Canadian shield lake. Model performance has been tested against field data supplied by Atomic Energy of Canada Limited. Carbon-14 was added in 1978 to the epilimnion of a small Canadian Shield lake to investigate primary production and carbon dynamics. Data from this experiment were used within BIOMOVS II to provide a validation test, which involved modelling the fate of the C-14 added to the lake. The nature of the spike and the subsequent monitoring allowed the investigation of both short-term processes relevant to evaluation of the impacts of accidental releases as well as longer-term processes relevant to routine release and to solid waste disposal. Four models participated in the scenario: (1) a simple mass balance model of a lake (AECL, Whiteshell Laboratories, Canada); (2) a relatively complex deterministic dynamic compartment model (QuantiSci Ltd.,UK); (3) a complex deterministic model (Studsvik Model A) and a more complex probabilistic model (Studsvik Model B; Studsvik Eco and Safety AB, Sweden). Endpoints were C-14 concentrations in water, sediment and whitefish over a thirteen year period. Each model produced reasonable predictions when compared to the observed data and when uncertainty is taken into consideration. About 0.2 to 0.4% of the initial C-14 inventory to the lakes remained in the water at the end of the study, because of internal recycling of C-14 from sediments. The simple AECL model did not account for this internal recycling of C-14 and, in this respect, its predictions were not as realistic as those of the QuantiSci and Studsvik models for concentrations in water. However, the AECL model predictions for the C-14 inventory remaining in lake sediment were closest to the observed values. Overall, Studsvik Model B was the most accurate in simulating C-14 concentrations in water and in whitefish, but

  15. Tower Shielding Reactor II design and operation report. Vol. 3. Assembling and testing of the control mechanism assembly

    International Nuclear Information System (INIS)

    Ward, D.R.; Holland, L.B.

    1979-09-01

    The mechanisms that are operated to control the reactivity of the Tower Shielding Reactor II(TSR-II) are mounted on a Control Mechanism Housing (CMH) that is centered inside the reactor core. The information required to procure, fabricate, inspect, and assemble a CMH is contained in the ORNL engineering drawings listed in the appropriate sections. The components are fabricated and inspected from these drawings in accordance with a Quality Assurance Plan and a Manufacturing Plan. The material in this report describes the acceptance and performance tests of CMH subassemblies used ty the Tower Shielding Facility (TSF) staff but it can also be used by personnel fabricating the components. This information which was developed and used before the advent of the formalized QA Program and Manufacturing Plans evolved during the fabrication and testing of the first five CMHs

  16. Development and testing of aluminum micro channel heat sink

    Science.gov (United States)

    Kumaraguruparan, G.; Sornakumar, T.

    2010-06-01

    Microchannel heat sinks constitute an innovative cooling technology for the removal of a large amount of heat from a small area and are suitable for electronics cooling. In the present work, Tool Steel D2 grade milling slitting saw type plain milling cutter is fabricated The microchannels are machined in aluminum work pieces to form the microchannel heat sink using the fabricated milling cutter in an horizontal milling machine. A new experimental set-up is fabricated to conduct the tests on the microchannel heat sink. The heat carried by the water increases with mass flow rate and heat input. The heat transfer coefficient and Nusselt number increases with mass flow rate and increased heat input. The pressure drop increases with Reynolds number and decreases with input heat. The friction factor decreases with Reynolds number and decreases with input heat. The thermal resistance decreases with pumping power and decreases with input heat.

  17. Structural design of shield-integrated thin-wall vacuum vessel and manufacturing qualification tests for International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Shimizu, Katsusuke; Shibui, Masanao; Koizumi, Koichi; Kanamori, Naokazu; Nishio, Satoshi; Sasaki, Takashi; Tada, Eisuke

    1992-09-01

    Conceptual design of shield-integrated thin-wall vacuum vessel has been done for ITER (International Thermonuclear Experimental Reactor). The vacuum vessel concept is based on a thin-double-wall structure, which consists of inner and outer plates and rib stiffeners. Internal shielding structures, which provide neutron irradiation shielding to protect TF coils, are set up between the inner plate and the outer plate of the vessel to avoid complexity of machine systems such as supporting systems of blanket modules. The vacuum vessel is assembled/disassembled by remote handling, so that welding joints are chosen as on-site joint method from reliability of mechanical strength. From a view point of assembling TF coils, the vacuum vessel is separated at the side of port, and is divided into 32 segments similar to the ITER-CDA reference design. Separatrix sweeping coils are located in the vacuum vessel to reduce heat fluxes onto divertor plates. Here, the coil structure and attachment to the vacuum vessel have been investigated. A sectorized saddle-loop coil is available for assembling and disassembling the coil. To support electromagnetic loads on the coils, they are attached to the groove in the vacuum vessel by welding. Flexible multi-plate supporting structure (compression-type gravity support), which was designed during CDA, is optimized by investigating buckling and frequency response properties, and concept on manufacturing and fabrication of the gravity support are proposed. Partial model of the vacuum vessel is manufactured for trial, so that fundamental data on welding and fabrication are obtained. From mechanical property tests of weldment and partial models, mechanical intensity and behaviors of the weldment are obtained. Informations on FEM-modeling are obtained by comparing analysis results with experimental results. (author)

  18. Shielding practice

    International Nuclear Information System (INIS)

    Sauermann, P.F.

    1985-08-01

    The basis of shielding practice against external irradiation is shown in a simple way. For most sources of radiation (point sources) occurring in shielding practice, the basic data are given, mainly in the form of tables, which are required to solve the shielding problems. The application of these data is explained and discussed using practical examples. Thickness of shielding panes of glove boxes for α and β radiation; shielding of sealed γ-radiography sources; shielding of a Co-60 radiation source, and of the manipulator panels for hot cells; damping factors for γ radiation and neutrons; shielding of fast and thermal neutrons, and of bremsstrahlung (X-ray tubes, Kr-85 pressure gas cylinders, 42 MeV betatrons, 20 MeV linacs); two-fold shielding (lead glass windows for hot cells, 14 MeV neutron generators); shielding against scattered radiation. (orig./HP) [de

  19. Use of a radio frequency shield during 1.5 and 3.0 Tesla magnetic resonance imaging: experimental evaluation.

    Science.gov (United States)

    Favazza, Christopher P; King, Deirdre M; Edmonson, Heidi A; Felmlee, Joel P; Rossman, Phillip J; Hangiandreou, Nicholas J; Watson, Robert E; Gorny, Krzysztof R

    2014-01-01

    Radiofrequency (RF) shields have been recently developed for the purpose of shielding portions of the patient's body during magnetic resonance imaging (MRI) examinations. We present an experimental evaluation of a commercially available RF shield in the MRI environment. All tests were performed on 1.5 T and 3.0 T clinical MRI scanners. The tests were repeated with and without the RF shield present in the bore, for comparison. Effects of the shield, placed within the scanner bore, on the RF fields generated by the scanner were measured directly using tuned pick-up coils. Attenuation, by as much as 35 dB, of RF field power was found inside the RF shield. These results were supported by temperature measurements of metallic leads placed inside the shield, in which no measurable RF heating was found. In addition, there was a small, simultaneous detectable increase (∼1 dB) of RF power just outside the edges of the shield. For these particular scanners, the autocalibrated RF power levels were reduced for scan locations prescribed just outside the edges of the shield, which corresponded with estimations based on the pick-up coil measurements. Additionally, no significant heating during MRI scanning was observed on the shield surface. The impact of the RF shield on the RF fields inside the magnet bore is likely to be dependent on the particular model of the RF shield or the MRI scanner. These results suggest that the RF shield could be a valuable tool for clinical MRI practices.

  20. First Wall, Blanket, Shield Engineering Technology Program

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1982-01-01

    The First Wall/Blanket/Shield Engineering Technology Program sponsored by the Office of Fusion Energy of DOE has the overall objective of providing engineering data that will define performance parameters for nuclear systems in advanced fusion reactors. The program comprises testing and the development of computational tools in four areas: (1) thermomechanical and thermal-hydraulic performance of first-wall component facsimiles with emphasis on surface heat loads; (2) thermomechanical and thermal-hydraulic performance of blanket and shield component facsimiles with emphasis on bulk heating; (3) electromagnetic effects in first wall, blanket, and shield component facsimiles with emphasis on transient field penetration and eddy-current effects; (4) assembly, maintenance and repair with emphasis on remote-handling techniques. This paper will focus on elements 2 and 4 above and, in keeping with the conference participation from both fusion and fission programs, will emphasize potential interfaces between fusion technology and experience in the fission industry

  1. Heat pumps in field test; Waermepumpen im Feldtest

    Energy Technology Data Exchange (ETDEWEB)

    Becker, R. [Fraunhofer-Institut fuer Solare Energiesysteme (ISE), Freiburg im Breisgau (Germany); Miara, M.; Russ, C.

    2007-09-15

    The Fraunhofer ISE has launched two field tests of newly installed heat pumps in 2006. Both deal with the measurement of a high number of heat pump units under real conditions in small houses. Values of volume flows, temperatures, heat quantity and electricity consumption are collected and daily saved and analysed at the Fraunhofer ISE. (orig.)

  2. Structural shielding at irradiation tests with x-rays up to 400 kV

    International Nuclear Information System (INIS)

    Rabitsch, H.; Schachinger, E.

    1979-12-01

    Tables worked out in accordance to the dimensioning method following DIN 54113 are given for the determination protective barriers thicknesses against effective and stray radiation in practical operation conditions. The tables comprise irradiation operation with directional and non-directional radiation sources radiators beams collimated and are restricted to steel as dispersive material for the barrier thicknesses against stray radiation. The calculations of protective barriers against stray radiation show that operation with properly chosen primary beam limiting orifice yields in considerable savings in protective shield thickness when compared to operation without or with mismatched orifice to film size. It further shows a relative indifference against changes of the stray area (film size) or the focus - film sheet distance. The protective shield thickness for the most common exposure conditions calculated according to the Austrian Regulations for Radiation Protection are on the safe side when compared to protective barriere thicknesses calculated according to DIN 54113. (V.M.)

  3. Radiation shielding tests in the Meson beamline in the master substation area

    International Nuclear Information System (INIS)

    Coleman, R.; Kissel, W.; Leveling, A.; Moore, C.D.; Vylet, V.

    1991-04-01

    A review of shielding uncovered a weak region in a portion of the proton beam transport to the Meson Area. Preliminary CASIM Monte Carlo studies indicated dose rates at the surface under abnormal operating conditions would be above the Fermilab Radiation Guide limits. Measurements made on December 15 and 16 confirmed this concern. Further comparisons of data with CASIM predictions are discussed. 5 refs., 22 figs., 8 tabs

  4. Pre-evaluation of fusion shielding benchmark experiment

    International Nuclear Information System (INIS)

    Hayashi, K.; Handa, H.; Konno, C.

    1994-01-01

    Shielding benchmark experiment is very useful to test the design code and nuclear data for fusion devices. There are many types of benchmark experiments that should be done in fusion shielding problems, but time and budget are limited. Therefore it will be important to select and determine the effective experimental configurations by precalculation before the experiment. The authors did three types of pre-evaluation to determine the experimental assembly configurations of shielding benchmark experiments planned in FNS, JAERI. (1) Void Effect Experiment - The purpose of this experiment is to measure the local increase of dose and nuclear heating behind small void(s) in shield material. Dimension of the voids and its arrangements were decided as follows. Dose and nuclear heating were calculated both for with and without void(s). Minimum size of the void was determined so that the ratio of these two results may be larger than error of the measurement system. (2) Auxiliary Shield Experiment - The purpose of this experiment is to measure shielding properties of B 4 C, Pb, W, and dose around superconducting magnet (SCM). Thickness of B 4 C, Pb, W and their arrangement including multilayer configuration were determined. (3) SCM Nuclear Heating Experiment - The purpose of this experiment is to measure nuclear heating and dose distribution in SCM material. Because it is difficult to use liquid helium as a part of SCM mock up material, material composition of SCM mock up are surveyed to have similar nuclear heating property of real SCM composition

  5. Electromagnetic shielding

    International Nuclear Information System (INIS)

    Tzeng, Wen-Shian V.

    1991-01-01

    Electromagnetic interference (EMI) shielding materials are well known in the art in forms such as gaskets, caulking compounds, adhesives, coatings and the like for a variety of EMI shielding purposes. In the past, where high shielding performance is necessary, EMI shielding has tended to use silver particles or silver coated copper particles dispersed in a resin binder. More recently, aluminum core silver coated particles have been used to reduce costs while maintaining good electrical and physical properties. (author). 8 figs

  6. High heat flux testing of ITER ICH&CD antenna beryllium faraday screen bars mock-ups

    International Nuclear Information System (INIS)

    Courtois, X.; Meunier, L.; Kuznetsov, V.; Beaumont, B.; Lamalle, P.; Conchon, D.; Languille, P.

    2016-01-01

    Highlights: • ITER ICH&CD antenna beryllium faraday screen bars mock-ups were manufactured. • The mock-ups are submitted to high heat loads to test their heat exhaust capabilities. • The mock-ups withstand without damage the design limit load. • Lifetime is gradually reduced when the heat load is augmented beyond the design limit. • Thermal and mechanical behavior are reproducible, and coherent with the calculation. - Abstract: The Faraday Screen (FS) is the plasma facing component of ITER ion cyclotron heating antennas shielding. The requirement for the high heat exhaust, and the limitation of the temperatures to minimize strain and thus offer sufficient resistance to fatigue, imply the need for high conductivity materials and a high cooling flow rate. The FS bars are constructed by a hipping process involving beryllium tiles, a pure copper layer, a copper chrome zirconium alloy for the cooling channel and a stainless steel backing strip. Two FS bars small scale mock-ups were manufactured and tested under high heat flux. They endured 15,000 heating cycles without degradation under nominal heat flux, and revealed growing flaws when the heat flux was progressively augmented beyond. In this case, the ultrasonic test confirms a strong delamination of the Be tiles.

  7. Secondary gamma-ray data for shielding calculation

    International Nuclear Information System (INIS)

    Miyasaka, Sunichi

    1979-01-01

    In deep penetration transport calculations, the integral design parameters is determined mainly by secondary particles which are produced by interactions of the primary radiation with materials. The shield thickness and the biological dose rate at a given point of a bulk shield are determined from the contribution from secondary gamma rays. The heat generation and the radiation damage in the structural and shield materials depend strongly on the secondary gamma rays. In this paper, the status of the secondary gamma ray data and its further problems are described from the viewpoint of shield design. The secondary gamma-ray data in ENDF/B-IV and POPOP4 are also discussed based on the test calculations made for several shield assemblies. (author)

  8. The potential of high heat generating granites as EGS source to generate power and reduce CO2 emissions, western Arabian shield, Saudi Arabia

    Science.gov (United States)

    Chandrasekharam, D.; Lashin, A.; Al Arifi, N.; Al Bassam, A.; El Alfy, M.; Ranjith, P. G.; Varun, C.; Singh, H. K.

    2015-12-01

    Saudi Arabia's dependence on oil and gas to generate electricity and to desalinate sea water is widely perceived to be economically and politically unsustainable. A recent business as usual simulation concluded that the Kingdom would become an oil importer by 2038. There is an opportunity for the country to over come this problem by using its geothermal energy resources. The heat flow and heat generation values of the granites spread over a cumulative area of 161,467 sq. km and the regional stress regime over the western Saudi Arabian shield strongly suggest that this entire area is potential source of energy to support 1) electricity generation, 2) fresh water generation through desalination and 3) extensive agricultural activity for the next two decades. The country can adopt a policy to harness this vast untapped enhanced geothermal systems (EGS) to mitigate climate and fresh water related issues and increase the quantity of oil for export. The country has inherent expertise to develop this resource.

  9. Flow and pressure profiles for the primary heat transport system of Rajasthan Atomic Power Station for the operation with few isolated reactor channels near the end shield cracks

    Energy Technology Data Exchange (ETDEWEB)

    Gaikwad, A J; Chaki, S K; Sehgal, R L; Venkat Raj, V [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    The RAPS (Rajasthan Atomic Power Station) unit-1 is now operating at reduced power due to the removal of fifteen fuel channels for repair of south end shield cracks. The power level is restricted to 50% of the full power capacity as a precautionary measure. The relative difference that operation at 50% power and higher power would make to the end shield structure is being currently analysed with a view to operate this reactor at higher power levels. As a prerequisite, a detailed thermal hydraulic analysis is essential to assess the effect of reactor operation with isolated channels on the primary heat transport (PHT) system pressure, flow, temperature. The adequacy of the existing trip set points for the plant operation under this mode is also required to be assessed. In the present study, analysis of the PHT system has been carried out to determine the flow and pressure profiles for the RAPS heat transport system for operation of the reactor with isolated channels. (author). 5 refs., 1 fig., 1 tab.

  10. Density-tunable lightweight polymer composites with dual-functional ability of efficient EMI shielding and heat dissipation.

    Science.gov (United States)

    Lee, Seung Hwan; Yu, Seunggun; Shahzad, Faisal; Kim, Woo Nyon; Park, Cheolmin; Hong, Soon Man; Koo, Chong Min

    2017-09-21

    Lightweight dual-functional materials with high EMI shielding performance and thermal conductivity are of great importance in modern cutting-edge applications, such as mobile electronics, automotive, aerospace, and military. Unfortunately, a clear material solution has not emerged yet. Herein, we demonstrate a simple and effective way to fabricate lightweight metal-based polymer composites with dual-functional ability of excellent EMI shielding effectiveness and thermal conductivity using expandable polymer bead-templated Cu hollow beads. The low-density Cu hollow beads (ρ ∼ 0.44 g cm -3 ) were fabricated through electroless plating of Cu on the expanded polymer beads with ultralow density (ρ ∼ 0.02 g cm -3 ). The resulting composites that formed a continuous 3D Cu network with a very small Cu content (∼9.8 vol%) exhibited excellent EMI shielding (110.7 dB at 7 GHz) and thermal conductivity (7.0 W m -1 K -1 ) with isotropic features. Moreover, the densities of the composites are tunable from 1.28 to 0.59 g cm -3 in accordance with the purpose of their applications. To the best of our knowledge, the resulting composites are the best lightweight dual-functional materials with exceptionally high EMI SE and thermal conductivity performance among synthetic polymer composites.

  11. Nano lead oxide and epdm composite for development of polymer based radiation shielding material: Gamma irradiation and attenuation tests

    Science.gov (United States)

    Özdemir, T.; Güngör, A.; Akbay, I. K.; Uzun, H.; Babucçuoglu, Y.

    2018-03-01

    It is important to have a shielding material that is not easily breaking in order to have a robust product that guarantee the radiation protection of the patients and radiation workers especially during the medical exposure. In this study, nano sized lead oxide (PbO) particles were used, for the first time, to obtain an elastomeric composite material in which lead oxide nanoparticles, after the surface modification with silane binding agent, was used as functional material for radiation shielding. In addition, the composite material including 1%, 5%, 10%, 15% and 20% weight percent nano sized lead oxide was irradiated with doses of 81, 100 and 120 kGy up to an irradiation period of 248 days in a gamma ray source with an initial dose rate of 21.1 Gy/h. Mechanical, thermal properties of the irradiated materials were investigated using DSC, DMA, TGA and tensile testing and modifications in thermal and mechanical properties of the nano lead oxide containing composite material via gamma irradiation were reported. Moreover, effect of bismuth-III oxide addition on radiation attenuation of the composite material was investigated. Nano lead oxide and bismuth-III oxide particles were mixed with different weight ratios. Attenuation tests have been conducted to determine lead equivalent values for the developed composite material. Lead equivalent thickness values from 0.07 to 0.65 (2-6 mm sample thickness) were obtained.

  12. The performance test of anti-scattering x-ray grid with inclined shielding material by MCNP code simulation

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Jun Woo; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2016-06-15

    The scattered photons cause reduction of the contrast of radiographic image and it results in the degradation of the quality of the image. In order to acquire better quality image, an anti-scattering x-ray gird should be equipped in radiography system. The X-ray anti-scattering grid of the inclined type based on the hybrid concept for that of parallel and focused type was tested by MCNP code. The MCNPX 2.7.0 was used for the simulation based test. The geometry for the test was based on the IEC 60627 which was an international standard for diagnostic X-ray imaging equipment-Characteristics of general purpose and mammographic anti-scatter grids. The performance of grids with four inclined shielding material types was compared with that of the parallel type. The grid with completely tapered type the best performance where there were little performance difference according to the degree of inclination.

  13. Model techniques for testing heated concrete structures

    International Nuclear Information System (INIS)

    Stefanou, G.D.

    1983-01-01

    Experimental techniques are described which may be used in the laboratory to measure strains of model concrete structures representing to scale actual structures of any shape or geometry, operating at elevated temperatures, for which time-dependent creep and shrinkage strains are dominant. These strains could be used to assess the distribution of stress in the scaled structure and hence to predict the actual behaviour of concrete structures used in nuclear power stations. Similar techniques have been employed in an investigation to measure elastic, thermal, creep and shrinkage strains in heated concrete models representing to scale parts of prestressed concrete pressure vessels for nuclear reactors. (author)

  14. Technical management on commissioning test of nuclear heating reactor

    International Nuclear Information System (INIS)

    Zhang Yajun; Su Qingshan

    1999-01-01

    The commissioning is the last construction stage of a nuclear heating project. The commissioning quality will directly affect on the safe operation and availability of the heating reactor. The author presents the whole test process until the completion of the test report from the point of test documents, including the preparation and execution of the test, the management of the various unexpected events during the test. And it will be emphatically discussed that the managing procedures of the various unexpected events during the test, including temporary control change, setpoint change, unexpected events and design change

  15. Radiant heating tests of several liquid metal heat-pipe sandwich panels

    International Nuclear Information System (INIS)

    Camarda, C.J.; Basiulis, A.

    1983-08-01

    Integral heat pipe sandwich panels, which synergistically combine the thermal efficiency of heat pipes and the structural efficiency of honeycomb sandwich construction, were conceived as a means of alleviating thermal stress problems in the Langley Scramjet Engine. Test panels which utilized two different wickable honeycomb cores, facesheets with screen mesh sintered to the internal surfaces, and a liquid metal working fluid (either sodium or potassium) were tested by radiant heating at various heat load levels. The heat pipe panels reduced maximum temperature differences by 31 percent with sodium working fluid and 45 percent with potassium working fluid. Results indicate that a heat pipe sandwich panel is a potential, simple solution to the engine thermal stress problem. Other interesting applications of the concept include: cold plates for electronic component and circuit card cooling, radiators for large space platforms, low distortion large area structures (e.g., space antennas) and laser mirrors

  16. Bayesian analysis of heat pipe life test data for reliability demonstration testing

    International Nuclear Information System (INIS)

    Bartholomew, R.J.; Martz, H.F.

    1985-01-01

    The demonstration testing duration requirements to establish a quantitative measure of assurance of expected lifetime for heat pipes was determined. The heat pipes are candidate devices for transporting heat generated in a nuclear reactor core to thermoelectric converters for use as a space-based electric power plant. A Bayesian analysis technique is employed, utilizing a limited Delphi survey, and a geometric mean accelerated test criterion involving heat pipe power (P) and temperature (T). Resulting calculations indicate considerable test savings can be achieved by employing the method, but development testing to determine heat pipe failure mechanisms should not be circumvented

  17. Methodology for Life Testing of Refractory Metal / Sodium Heat Pipes

    International Nuclear Information System (INIS)

    Martin, James J.; Reid, Robert S.

    2006-01-01

    This work establishes an approach to generate carefully controlled data to find heat pipe operating life with material-fluid combinations capable of extended operation. To accomplish this goal acceleration is required to compress 10 years of operational life into 3 years of laboratory testing through a combination of increased temperature and mass fluence. Specific test series have been identified, based on American Society for Testing and Materials (ASTM) specifications, to investigate long-term corrosion rates. The refractory metal selected for demonstration purposes is a molybdenum-44.5% rhenium alloy formed by powder metallurgy. The heat pipes each have an annular crescent wick formed by hot isostatic pressing of molybdenum-rhenium wire mesh. The heat pipes are filled by vacuum distillation with purity sampling of the completed assembly. Round-the-clock heat pipe tests with 6-month destructive and non-destructive inspection intervals are conducted to identify the onset and level of corrosion. Non-contact techniques are employed to provide power to the evaporator (radio frequency induction heating at 1 to 5 kW per heat pipe) and calorimetry at the condenser (static gas gap coupled water cooled calorimeter). The planned operating temperature range extends from 1123 to 1323 K. Accomplishments before project cancellation included successful development of the heat pipe wick fabrication technique, establishment of all engineering designs, baseline operational test requirements, and procurement/assembly of supporting test hardware systems. (authors)

  18. Summary of Blast Shield and Material Testing for Development of Solid Debris Collection at the National Ignition Facility (NIF)

    International Nuclear Information System (INIS)

    Shaughnessy, D.A.; Gostic, J.M.; Moody, K.J.; Grant, P.M.; Lewis, L.A.; Hutcheon, I.D.

    2011-01-01

    The ability to collect solid debris from the target chamber following a NIF shot has application for both capsule diagnostics, particularly for fuel-ablator mix, and measuring cross sections relevant to the Stockpile Stewardship program and nuclear astrophysics. Simulations have shown that doping the capsule with up to 10 15 atoms of an impurity not otherwise found in the capsule does not affect its performance. The dopant is an element that will undergo nuclear activations during the NIF implosion, forming radioactive species that can be collected and measured after extraction from the target chamber. For diagnostics, deuteron or alpha induced reactions can be used to probe the fuel-ablator mix. For measuring neutron cross sections, the dopant should be something that is sensitive to the 14 MeV neutrons produced through the fusion of deuterium and tritium. Developing the collector is a challenge due to the extreme environment of the NIF chamber. The collector surface is exposed to a large photon flux from x-rays and unconverted laser light before it is exposed to a debris wind that is formed from vaporized material from the target chamber center. The photons will ablate the collector surface to some extent, possibly impeding the debris from reaching the collector and sticking. In addition, the collector itself must be mechanically strong enough to withstand the large amount of energy it will be exposed to, and it should be something that will be easy to count and chemically process. In order to select the best material for the collector, a variety of different metals have been tested in the NIF chamber. They were exposed to high-energy laser shots in order to evaluate their postshot surface characterization, morphology, degree of melt, and their ability to retain debris from the chamber center. The first set of samples consisted of 1 mm thick pieces of aluminum that had been fielded in the chamber as blast shields protecting the neutron activation diagnostic. Ten

  19. Thermodynamic and heat transfer analysis of heat recovery from engine test cell by Organic Rankine Cycle

    Science.gov (United States)

    Shokati, Naser; Mohammadkhani, Farzad; Farrokhi, Navid; Ranjbar, Faramarz

    2014-12-01

    During manufacture of engines, evaluation of engine performance is essential. This is accomplished in test cells. During the test, a significant portion of heat energy released by the fuel is wasted. In this study, in order to recover these heat losses, Organic Rankine Cycle (ORC) is recommended. The study has been conducted assuming the diesel oil to be composed of a single hydrocarbon such as C12H26. The composition of exhaust gases (products of combustion) have been computed (and not determined experimentally) from the stoichiometric equation representing the combustion reaction. The test cell heat losses are recovered in three separate heat exchangers (preheater, evaporator and superheater). These heat exchangers are separately designed, and the whole system is analyzed from energy and exergy viewpoints. Finally, a parametric study is performed to investigate the effect of different variables on the system performance characteristics such as the ORC net power, heat exchangers effectiveness, the first law efficiency, exergy destruction and heat transfer surfaces. The results of the study show that by utilizing ORC, heat recovery equivalent to 8.85 % of the engine power is possible. The evaporator has the highest exergy destruction rate, while the pump has the lowest among the system components. Heat transfer surfaces are calculated to be 173.6, 58.7, and 11.87 m2 for the preheater, evaporator and superheater, respectively.

  20. PENGARUH VARIASI SUHU POST WELD HEAT TREATMENT ANNEALING TERHADAP SIFAT MEKANIS MATERIAL BAJA EMS-45 DENGAN METODE PENGELASAN SHIELDED METAL ARC WELDING (SMAW

    Directory of Open Access Journals (Sweden)

    Rusiyanto Rusiyanto

    2012-02-01

    Full Text Available Penelitian ini bertujuan Untuk mengetahui nilai kekerasan Vickers material Baja EMS-45 sebelum proses pengelasan dan setelah dilakukan proses pengelasan tanpa post weld heat treatment annealing, Untuk mengetahui berapakah suhu optimal post weld heat treatment annealing untuk material baja EMS-45 dengan variasi suhu yang digunakan 350 o C, 550 o C, dan 750 C. Untuk mengetahui struktur mikro dari material baja EMS-45 akibat variasi suhu post weld heat treatment annealing pada proses pengelasan dengan menggunakan metode pengelasan shielded metal arc welding. Bahan atau material dasar yang digunakan pada penelitian ini adalah Baja EMS-45 dengan ketebalan pelat 10 mm, lebar pelat 20 mm dan panjang 100 mm. Berdasarkan hasil pengujian nilai kekerasan tertinggi setelah proses pengelasan terletak pada daerah Logam Las. Pengelasan non PWHT memiliki nilai kekerasan paling tinggi setelah proses pengelasan yaitu sebesar 183,2 VHN. Suhu optimal Post Weld Heat Treatment Annealing untuk material baja EMS-45 adalah pada suhu 750 C. Karena pada PWHT pada suhu tersebut mengalami penurunan kekerasan yang besar yaitu sebesar 127,2 VHN, sehingga material baja EMS-45 dapat memperbaiki sifat mampu mesinnya. Struktur mikro dari material baja EMS-45 sebelum proses pengelasan berupa grafit serpih, perlit dan ferit, setelah dilakukan proses pengelasan mempunyai struktur mikro berupa matrik ferit dan grafit pada daerah logam las, matrik perlit kasar dan grafit serpih pada daerah HAZ dan struktur perlit, grafit serpih dan ferit pada daerah logam induk o o

  1. Retrofitting Combined Space and Water Heating Systems: Laboratory Tests

    Energy Technology Data Exchange (ETDEWEB)

    Schoenbauer, B.; Bohac, D.; Huelman, P.; Olson, R.; Hewitt, M.

    2012-10-01

    Better insulated and tighter homes can often use a single heating plant for both space and domestic water heating. These systems, called dual integrated appliances (DIA) or combination systems, can operate at high efficiency and eliminate combustion safety issues associated by using a condensing, sealed combustion heating plant. Funds were received to install 400 DIAs in Minnesota low-income homes. The NorthernSTAR DIA laboratory was created to identify proper system components, designs, operating parameters, and installation procedures to assure high efficiency of field installed systems. Tests verified that heating loads up to 57,000 Btu/hr can be achieved with acceptable return water temperatures and supply air temperatures.

  2. Retrofitting Combined Space and Water Heating Systems. Laboratory Tests

    Energy Technology Data Exchange (ETDEWEB)

    Schoenbauer, B. [NorthernStar Building America Partnership, St. Paul, MN (United States); Bohac, D. [NorthernStar Building America Partnership, St. Paul, MN (United States); Huelman, P. [NorthernStar Building America Partnership, St. Paul, MN (United States); Olsen, R. [NorthernStar Building America Partnership, St. Paul, MN (United States); Hewett, M. [NorthernStar Building America Partnership, St. Paul, MN (United States)

    2012-10-01

    Better insulated and tighter homes can often use a single heating plant for both space and domestic water heating. These systems, called dual integrated appliances (DIA) or combination systems, can operate at high efficiency and eliminate combustion safety issues associated by using a condensing, sealed combustion heating plant. Funds were received to install 400 DIAs in Minnesota low-income homes. The NorthernSTAR DIA laboratory was created to identify proper system components, designs, operating parameters, and installation procedures to assure high efficiency of field installed systems. Tests verified that heating loads up to 57,000 Btu/hr can be achieved with acceptable return water temperatures and supply air temperatures.

  3. Design and preliminary test results at Mach 5 of an axisymmetric slotted sound shield. [for supersonic wind tunnels (noise reduction in wind tunnel nozzles)

    Science.gov (United States)

    Beckwith, I. E.; Spokowski, A. J.; Harvey, W. D.; Stainback, P. C.

    1975-01-01

    The basic theory and sound attenuation mechanisms, the design procedures, and preliminary experimental results are presented for a small axisymmetric sound shield for supersonic wind tunnels. The shield consists of an array of small diameter rods aligned nearly parallel to the entrance flow with small gaps between the rods for boundary layer suction. Results show that at the lowest test Reynolds number (based on rod diameter) of 52,000 the noise shield reduced the test section noise by about 60 percent ( or 8 db attenuation) but no attenuation was measured for the higher range of test reynolds numbers from 73,000 to 190,000. These results are below expectations based on data reported elsewhere on a flat sound shield model. The smaller attenuation from the present tests is attributed to insufficient suction at the gaps to prevent feedback of vacuum manifold noise into the shielded test flow and to insufficient suction to prevent transition of the rod boundary layers to turbulent flow at the higher Reynolds numbers. Schlieren photographs of the flow are shown.

  4. Comparing heat flow models for interpretation of precast quadratic pile heat exchanger thermal response tests

    DEFF Research Database (Denmark)

    Alberdi Pagola, Maria; Poulsen, Søren Erbs; Loveridge, Fleur

    2018-01-01

    This paper investigates the applicability of currently available analytical, empirical and numerical heat flow models for interpreting thermal response tests (TRT) of quadratic cross section precast pile heat exchangers. A 3D finite element model (FEM) is utilised for interpreting five TRTs by in...

  5. Micro-Combined Heat and Power Device Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — NIST has developed a test facility for micro-combined heat and power (micro-CHP) devices to measure their performance over a range of different operating strategies...

  6. Heat transfer performance test of PDHRS heat exchangers of PGSFR using STELLA-1 facility

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jonggan, E-mail: hong@kaeri.re.kr; Yeom, Sujin; Eoh, Jae-Hyuk; Lee, Tae-Ho; Jeong, Ji-Young

    2017-03-15

    Highlights: • Heat transfer performance test of heat exchangers of PGSFR PDHRS is conducted using STELLA-1 facility. • Steady-state test results of DHX and AHX show good agreement with theoretical results of design codes. • Design codes for DHX and AHX are validated by STELLA-1 experimental results. • Heat transport capability of DHX and AHX is turned out to be satisfactory for reliable plant operation. - Abstract: The STELLA-1 facility was designed and constructed to carry out separate effect tests of the decay heat exchanger (DHX) and natural draft sodium-to-air heat exchanger (AHX), which are key components of the safety-grade decay heat removal system in PGSFR. The DHX is a sodium-to-sodium heat exchanger with a straight tube arrangement, and the AHX is a sodium-to-air heat exchanger with a helically coiled tube arrangement. The model heat exchangers in STELLA-1 have been designed to meet their own similitude conditions from the prototype ones, of which scale ratios were set to be unity in height (or length) and 1/2.5 in heat transfer rate. Consequently, the overall heat transfer coefficients and log-mean temperature differences of the prototypes have been preserved as well. The steady-state test results for each model heat exchanger obtained from STELLA-1 showed good agreement with the theoretical results of the computer design codes for thermal-sizing and a performance analysis of the DHX and AHX. In the DHX result comparison, the discrepancies in the heat transfer rate ranged from −4.4% to 2.0%, and in the AHX result comparison, they ranged from −11.1% to 12.6%. Therefore, the first step in thermal design codes validation for sodium heat exchangers, e.g., DHX and AHX, has been successfully completed with the experimental database obtained from STELLA-1. In addition, the heat transfer performance of the DHX and AHX was found to be satisfactory enough to secure a reliable decay heat removal performance.

  7. Shielding container

    International Nuclear Information System (INIS)

    Darling, K.A.M.

    1981-01-01

    A shielding container incorporates a dense shield, for example of depleted uranium, cast around a tubular member of curvilinear configuration for accommodating a radiation source capsule. A lining for the tubular member, in the form of a close-coiled flexible guide, provides easy replaceability to counter wear while the container is in service. Container life is extended, and maintenance costs are reduced. (author)

  8. Nuclear test watchers feel political heat

    International Nuclear Information System (INIS)

    Marshall, E.

    1987-01-01

    One year after US citizen diplomats signed a remarkable pact with the Soviet Union to monitor nuclear bomb tests, they are running into some of the obstacles that regular diplomats encounter - political flak from the Pentagon and harassment by the Soviet military. But they have devised some technical solutions that they hope will get them around the roadblocks. These solutions are discussed

  9. High-heat-flux testing of helium-cooled heat exchangers for fusion applications

    International Nuclear Information System (INIS)

    Youchison, D.L.; Izenson, M.G.; Baxi, C.B.; Rosenfeld, J.H.

    1996-01-01

    High-heat-flux experiments on three types of helium-cooled divertor mock-ups were performed on the 30-kW electron beam test system and its associated helium flow loop at Sandia National Laboratories. A dispersion-strengthened copper alloy (DSCu) was used in the manufacture of all the mock-ups. The first heat exchanger provides for enhanced heat transfer at relatively low flow rates and much reduced pumping requirements. The Creare sample was tested to a maximum absorbed heat flux of 5.8 MW/m 2 . The second used low pressure drops and high mass flow rates to achieve good heat removal. The GA specimen was tested to a maximum absorbed heat flux of 9 MW/m 2 while maintaining a surface temperature below 400 degree C. A second experiment resulted in a maximum absorbed heat flux of 34 MW/m 2 and surface temperatures near 533 degree C. The third specimen was a DSCu, axial flow, helium-cooled divertor mock-up filled with a porous metal wick which effectively increases the available heat transfer area. Low mass flow and high pressure drop operation at 4.0 MPa were characteristic of this divertor module. It survived a maximum absorbed heat flux of 16 MW/m 2 and reached a surface temperature of 740 degree C. Thermacore also manufactured a follow-on, dual channel porous metal-type heat exchanger, which survived a maximum absorbed heat flux of 14 MW/m 2 and reached a maximum surface temperature of 690 degree C. 11refs., 20 figs., 3 tabs

  10. Test and evaluation report for Westinghouse Hanford Company's Hedgehog Shielded Container, Docket 94-39-7A, Type A container

    International Nuclear Information System (INIS)

    Kelly, D.L.

    1995-01-01

    This report documents the US Department of Transportation Specification 7A Type A (DOT-7A) compliance test results of the Westinghouse Hanford Company Hedgehog Shielded Container. The Hedgehog packaging configurations provide primary and secondary containment. The packaging configurations tested consisted of an internal bottle, varying in size. Testing showed that the bottles are not required for the packaging to pass Type A requirements, with the exception of the 1-liter version, in which the polyvinyl chloride (PVC)-coated glass bottle used in testing is considered a part of the containment system. The packaging configurations were evaluated and tested in February 1995. The packaging configurations described in this report are designed to ship Type A quantities of radioactive materials, normal form. Contents may be in solid or liquid form. Liquids may have a specific gravity ≤2. The solid versions would allow the shipment of normal or special form solids. The solid materials would be limited in weight--to include packaging--to the gross weight of the as-tested liquids and bottles. The packaging configurations described in this document may be transported by air, and they meet the applicable International Air Transport Association/International Civil Aviation Organization (IATA/ICAO) Dangerous Goods Regulations in addition to the DOT-7A requirements

  11. Estimation of the heat generation in vitrified waste product and shield thickness of the cask for the transportation of vitrified waste product using Monte Carlo technique

    International Nuclear Information System (INIS)

    Deepa, A.K.; Jakhete, A.P.; Mehta, D.; Kaushik, C.P.

    2011-01-01

    High Level Liquid waste (HLW) generated during reprocessing of spent fuel contains most of the radioactivity present in the spent fuel resulting in the need for isolation and surveillance for extended period of time. Major components in HLW are the corrosion products, fission products such as 137 Cs, 90 Sr, 106 Ru, 144 Ce, 125 Sb etc, actinides and various chemicals used during reprocessing of spent fuel. Fresh HLW having an activity concentration of around 100Ci/l is to be vitrified into borosilicate glass and packed in canisters which are placed in S.S overpacks for better confinement. These overpacks contain around 0.7 Million Curies of activity. Characterisation of activity in HLW and activity profile of radionuclides for various cooling periods sets the base for the study. For transporting the vitrified waste product (VWP), two most important parameters is the shield thickness of the transportation cask and the heat generation in the waste product. This paper describes the methodology used in the estimation of lead thickness for the transportation cask using the Monte Carlo Technique. Heat generation due to decay of fission products results in the increase in temperature of the vitrified waste product during interim storage and disposal. Glass being the material, not having very high thermal conductivity, temperature difference between the canister and surrounding bears significance in view of the possibility of temperature based devitrification of VWP. The heat generation in the canister and the overpack containing vitrified glass is also estimated using MCNP. (author)

  12. Hexavalent Chrome Free Coatings for Electronics Electromagnetic Interference (EMI) Shielding Effectiveness (SE) Interim Test Report

    Science.gov (United States)

    Kessel, Kurt R.

    2015-01-01

    Test specimen configuration was provided by Parker Chomerics. The EMI gasket used in this project was Cho-Seal 6503E. Black oxide alloy steel socket head bolts were used to hold the plates together. Non-conductive spacers were used to control the amount of compression on the gaskets. The following test fixture specifications were provided by Parker Chomerics. The CHO-TP09 test plate sets selected for this project consist of two aluminum plates manufactured to the specifications detailed in CHO­-TP09. The first plate, referred to as the test frame, is illustrated in Figure 1. The test frame is designed with a cutout in the center and two alternating bolt patterns. One pattern is used to bolt the test frame to the corresponding test cover plate (Figure 2), forming a test plate set. The second pattern accepts the hardware used to mount the fully assembled test plate set to the main adapter plate (Figure 3).

  13. Dynamic Test Method Based on Strong Electromagnetic Pulse for Electromagnetic Shielding Materials with Field-Induced Insulator-Conductor Phase Transition

    Science.gov (United States)

    Wang, Yun; Zhao, Min; Wang, Qingguo

    2018-01-01

    In order to measure the pulse shielding performance of materials with the characteristic of field-induced insulator-conductor phase transition when materials are used for electromagnetic shielding, a dynamic test method was proposed based on a coaxial fixture. Experiment system was built by square pulse source, coaxial cable, coaxial fixture, attenuator, and oscilloscope and insulating components. S11 parameter of the test system was obtained, which suggested that the working frequency ranges from 300 KHz to 7.36 GHz. Insulating performance is good enough to avoid discharge between conductors when material samples is exposed in the strong electromagnetic pulse field up to 831 kV/m. This method is suitable for materials with annular shape, certain thickness and the characteristic of field-induced insulator-conductor phase transition to get their shielding performances of strong electromagnetic pulse.

  14. Heat resistance study of basalt fiber material via mechanical tests

    Science.gov (United States)

    Gao, Y. Q.; Jia, C.; Meng, L.; Li, X. H.

    2017-12-01

    This paper focuses on the study of the relationship between the fracture strength of basalt rovings and temperature. Strong stretching performance of the rovings has been tested after the treatment at fixed temperatures but different heating time and then the fracture strength of the rovings exposed to the heating at different temperatures and cooled in different modes investigated. Finally, the fracture strength of the basalt material after the heat treatment was studied. The results showed that the room-temperature strength tends to decrease with an increase of the heat treatment time at 250 °C, but it has the local maximum after 2h heating. And the basalt rovings strength increased after the heat treatment up to 200 °C. It was 16.7 percent higher than the original strength. The strength depends not only on the temperature and duration of the heating, but also on the cooling mode. The value of the strength measured after cold water cooling was less by 6.3% compared with an ambient air cooling mode. The room-temperature breaking strength of the rovings heated at 200 °C and 100 °C for 2 hours each increased by about 14.6% with respect to unpretreated basalt rovings.

  15. Thermal response testing of precast pile heat exchangers

    DEFF Research Database (Denmark)

    Pagola, Maria Alberdi; Poulsen, Søren Erbs; Jensen, Rasmus Lund

    The report is organized as follows: first, the concept of TRT is explained. Second, the test sites are described. Third, the field work is presented and a summary of the future work regarding the methodology to treat the data from the tests is provided. Finally, further documentation...... of the fieldwork, the pile heat exchangers and the TRT equipment is extended in diverse appendices....

  16. Analysis of high heat flux testing of mock-ups

    International Nuclear Information System (INIS)

    Salavy, J.-F.; Giancarli, L.; Merola, M.; Picard, F.; Roedig, M.

    2003-01-01

    ITER EU Home Team is performing a large R and D effort in support of the development of high heat flux components for ITER. In this framework, this paper describes the thermal analyses, the fatigue lifetime evaluation and the transient VDE with material melting related to the high heat flux thermo-mechanical tests performed in the JUDITH facility. It reports on several mock-ups representative of different proposed component designs based on Be, W and CFC as armour materials

  17. Space Launch System Base Heating Test: Experimental Operations & Results

    Science.gov (United States)

    Dufrene, Aaron; Mehta, Manish; MacLean, Matthew; Seaford, Mark; Holden, Michael

    2016-01-01

    NASA's Space Launch System (SLS) uses four clustered liquid rocket engines along with two solid rocket boosters. The interaction between all six rocket exhaust plumes will produce a complex and severe thermal environment in the base of the vehicle. This work focuses on a recent 2% scale, hot-fire SLS base heating test. These base heating tests are short-duration tests executed with chamber pressures near the full-scale values with gaseous hydrogen/oxygen engines and RSRMV analogous solid propellant motors. The LENS II shock tunnel/Ludwieg tube tunnel was used at or near flight duplicated conditions up to Mach 5. Model development was based on the Space Shuttle base heating tests with several improvements including doubling of the maximum chamber pressures and duplication of freestream conditions. Test methodology and conditions are presented, and base heating results from 76 runs are reported in non-dimensional form. Regions of high heating are identified and comparisons of various configuration and conditions are highlighted. Base pressure and radiometer results are also reported.

  18. Radiation Shielding Materials and Containers Incorporating Same

    Energy Technology Data Exchange (ETDEWEB)

    Mirsky, Steven M.; Krill, Stephen J.; and Murray, Alexander P.

    2005-11-01

    An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound (''PYRUC'') shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

  19. MMW [multimegawatt] shielding design and analysis

    International Nuclear Information System (INIS)

    Olson, A.P.

    1988-01-01

    Reactor shielding for multimegawatt (MMW) space power must satisfy a mass constraint as well as performance specifications for neutron fluence and gamma dose. A minimum mass shield is helpful in attaining the launch mass goal for the entire vehicle, because the shield comprises about 1% to 2% of the total vehicle mass. In addition, the shield internal heating must produce tolerable temperatures. The analysis of shield performance for neutrons and gamma rays is emphasized. Topics addressed include cross section preparation for multigroup 2D S/sub n/-transport analyses, and the results of parametric design studies on shadow shield performance and mass versus key shield design variables such as cone angle, number, placement, and thickness of layers of tungsten, and shield top radius. Finally, adjoint methods are applied to the shield in order to spatially map its relative contribution to dose reduction, and to provide insight into further design optimization. 7 refs., 2 figs., 3 tabs

  20. Primary shield displacement and bowing

    International Nuclear Information System (INIS)

    Scott, K.V.

    1978-01-01

    The reactor primary shield is constructed of high density concrete and surrounds the reactor core. The inlet, outlet and side primary shields were constructed in-place using 2.54 cm (1 in) thick steel plates as the forms. The plates remained as an integral part of the shields. The elongation of the pressure tubes due to thermal expansion and pressurization is not moving through the inlet nozzle hardware as designed but is accommodated by outward displacement and bowing of the inlet and outlet shields. Excessive distortion of the shields may result in gas seal failures, intolerable helium gas leaks, increased argon-41 emissions, and shield cooling tube failures. The shield surveillance and testing results are presented

  1. Electromagnetic shield

    International Nuclear Information System (INIS)

    Miller, J.S.

    1987-01-01

    An electromagnetic shield is described comprising: closed, electrically-conductive rings, each having an open center; and binder means for arranging the rings in a predetermined, fixed relationship relative to each other, the so-arranged rings and binder means defining an outer surface; wherein electromagnetic energy received by the shield from a source adjacent its outer surface induces an electrical current to flow in a predetermined direction adjacent and parallel to the outer surface, through the rings; and wherein each ring is configured to cause source-induced alternating current flowing through the portion of the ring closest to the outer surface to electromagnetically induce an oppositely-directed current in the portion of the ring furthest from the surface, such oppositely-directed current bucking any source-induced current in the latter ring portion and thus reducing the magnitude of current flowing through it, whereby the electromagnetic shielding effected by the shield is enhanced

  2. Nuclear shields

    International Nuclear Information System (INIS)

    Linares, R.C.; Nienart, L.F.; Toelcke, G.A.

    1976-01-01

    A process is described for preparing melt-processable nuclear shielding compositions from chloro-fluoro substituted ethylene polymers, particularly PCTFE and E-CTFE, containing 1 to 75 percent by weight of a gadolinium compound. 13 claims, no drawings

  3. REACTOR SHIELD

    Science.gov (United States)

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  4. Lithospheric structure of southern Indian shield and adjoining oceans: integrated modelling of topography, gravity, geoid and heat flow data

    Science.gov (United States)

    Kumar, Niraj; Zeyen, H.; Singh, A. P.; Singh, B.

    2013-07-01

    For the present 2-D lithospheric density modelling, we selected three geotransects of more than 1000 km in length each crossing the southern Indian shield, south of 16°N, in N-S and E-W directions. The model is based on the assumption of local isostatic equilibrium and is constrained by the topography, gravity and geoid anomalies, by geothermal data, and where available by seismic data. Our integrated modelling approach reveals a crustal configuration with the Moho depth varying from ˜40 km beneath the Dharwar Craton, and ˜39 km beneath the Southern Granulite Terrane to about 15-20 km beneath the adjoining oceans. The lithospheric thickness varies significantly along the three profiles from ˜70-100 km under the adjoining oceans to ˜130-135 km under the southern block of Southern Granulite Terrane including Sri Lanka and increasing gradually to ˜165-180 km beneath the northern block of Southern Granulite Terrane and the Dharwar Craton. This step-like lithosphere-asthenosphere boundary (LAB) structure indicates a normal lithospheric thickness beneath the adjoining oceans, the northern block of Southern Granulite Terrane and the Dharwar Craton. The thin lithosphere below the southern block of Southern Granulite Terrane including Sri Lanka is, however, atypical considering its age. Our results suggest that the southern Indian shield as a whole cannot be supported isostatically only by thickened crust; a thin and hot lithosphere beneath the southern block of Southern Granulite Terrane including Sri Lanka is required to explain the high topography, gravity, geoid and crustal temperatures. The widespread thermal perturbation during Pan-African (550 Ma) metamorphism and the breakup of Gondwana during late Cretaceous are proposed as twin cause mechanism for the stretching and/or convective removal of the lower part of lithospheric mantle and its replacement by hotter and lighter asthenosphere in the southern block of Southern Granulite Terrane including Sri Lanka

  5. Performance and Facility Background Pressure Characterization Tests of NASAs 12.5-kW Hall Effect Rocket with Magnetic Shielding Thruster

    Science.gov (United States)

    Kamhawi, Hani; Huang, Wensheng; Haag, Thomas; Shastry, Rohit; Thomas, Robert; Yim, John; Herman, Daniel; Williams, George; Myers, James; Hofer, Richard; hide

    2015-01-01

    NASA's Space Technology Mission Directorate (STMD) Solar Electric Propulsion Technology Demonstration Mission (SEP/TDM) project is funding the development of a 12.5-kW Hall thruster system to support future NASA missions. The thruster designated Hall Effect Rocket with Magnetic Shielding (HERMeS) is a 12.5-kW Hall thruster with magnetic shielding incorporating a centrally mounted cathode. HERMeS was designed and modeled by a NASA GRC and JPL team and was fabricated and tested in vacuum facility 5 (VF5) at NASA GRC. Tests at NASA GRC were performed with the Technology Development Unit 1 (TDU1) thruster. TDU1's magnetic shielding topology was confirmed by measurement of anode potential and low electron temperature along the discharge chamber walls. Thermal characterization tests indicated that during full power thruster operation at peak magnetic field strength, the various thruster component temperatures were below prescribed maximum allowable limits. Performance characterization tests demonstrated the thruster's wide throttling range and found that the thruster can achieve a peak thruster efficiency of 63% at 12.5 kW 500 V and can attain a specific impulse of 3,000 s at 12.5 kW and a discharge voltage of 800 V. Facility background pressure variation tests revealed that the performance, operational characteristics, and magnetic shielding effectiveness of the TDU1 design were mostly insensitive to increases in background pressure.

  6. Testing of PCM Heat Storage Modules with Solar Collectors as Heat Source

    DEFF Research Database (Denmark)

    Englmair, Gerald; Dannemand, Mark; Johansen, Jakob Berg

    2016-01-01

    A latent heat storage based on the phase change material Sodium Acetate Trihydrate (SAT) has been tested as part of a demonstration system. The full heat storage consisted of 4 individual modules each containing about 200 kg of sodium acetate trihydrate with different additives. The aim...... was to actively utilize the ability of the material to supercool to obtain long storage periods. The modules were charged with solar heat supplied by 22.4 m2 evacuated tubular collectors. The investigation showed that it was possible to fully charge one module within a period of 270 minutes with clear skies...

  7. Optimal design of tests for heat exchanger fouling identification

    International Nuclear Information System (INIS)

    Palmer, Kyle A.; Hale, William T.; Such, Kyle D.; Shea, Brian R.; Bollas, George M.

    2016-01-01

    Highlights: • Built-in test design that optimizes the information extractable from the said test. • Method minimizes the covariance of a fault with system uncertainty. • Method applied for the identification and quantification of heat exchanger fouling. • Heat exchanger fouling is identifiable despite the uncertainty in inputs and states. - Graphical Abstract: - Abstract: Particulate fouling in plate fin heat exchangers of aircraft environmental control systems is a recurring issue in environments rich in foreign object debris. Heat exchanger fouling detection, in terms of quantification of its severity, is critical for aircraft maintenance scheduling and safe operation. In this work, we focus on methods for offline fouling detection during aircraft ground handling, where the allowable variability range of admissible inputs is wider. We explore methods of optimal experimental design to estimate heat exchanger inputs and input trajectories that maximize the identifiability of fouling. In particular, we present a methodology in which D-optimality is used as a criterion for statistically significant inference of heat exchanger fouling in uncertain environments. The optimal tests are designed on the basis of a heat exchanger model of the inherent mass, energy and momentum balances, validated against literature data. The model is then used to infer sensitivities of the heat exchanger outputs with respect to fouling metrics and maximize them by manipulating input trajectories; thus enhancing the accuracy in quantifying the fouling extent. The proposed methodology is evaluated with statistical indices of the confidence in estimating thermal fouling resistance at uncertain operating conditions, explored in a series of case studies.

  8. Introduction to the PBMR heat transfer test facility

    International Nuclear Information System (INIS)

    Rousseau, P.G.; Staden, M. van

    2008-01-01

    This paper provides an introduction to the Heat Transfer Test Facility (HTTF) that is currently being developed for PBMR (Pty.) Ltd. by M-Tech Industrial (Pty.) Ltd. in association with North-West University in South Africa. The paper provides an overview of the phenomena that will be studied and the envisaged test configurations for each of these phenomena. It also shows the layouts of the different test units namely the High Pressure Test Unit (HPTU) and the High Temperature Test Unit (HTTU) and provides an overview of the planned test schedule

  9. Foam property tests to evaluate the potential for longwall shield dust control.

    Science.gov (United States)

    Reed, W R; Beck, T W; Zheng, Y; Klima, S; Driscoll, J

    2018-01-01

    Tests were conducted to determine properties of four foam agents for their potential use in longwall mining dust control. Foam has been tried in underground mining in the past for dust control and is currently being reconsidered for use in underground coal longwall operations in order to help those operations comply with the Mine Safety and Health Administration's lower coal mine respirable dust standard of 1.5 mg/m 3 . Foams were generated using two different methods. One method used compressed air and water pressure to generate foam, while the other method used low-pressure air generated by a blower and water pressure using a foam generator developed by the U.S. National Institute for Occupational Safety and Health. Foam property tests, consisting of a foam expansion ratio test and a water drainage test, were conducted to classify foams. Compressed-air-generated foams tended to have low expansion ratios, from 10 to 19, with high water drainage. Blower-air-generated foams had higher foam expansion ratios, from 30 to 60, with lower water drainage. Foams produced within these ranges of expansion ratios are stable and potentially suitable for dust control. The test results eliminated two foam agents for future testing because they had poor expansion ratios. The remaining two foam agents seem to have properties adequate for dust control. These material property tests can be used to classify foams for their potential use in longwall mining dust control.

  10. EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

    International Nuclear Information System (INIS)

    Reid, Robert S.; Pearson, J. Bosie; Stewart, Eric T.

    2007-01-01

    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

  11. EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    REID, ROBERT S. [Los Alamos National Laboratory; PEARSON, J. BOSIE [Los Alamos National Laboratory; STEWART, ERIC T. [Los Alamos National Laboratory

    2007-01-16

    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

  12. Experimental Evaluation of the Thermal Performance of a Water Shield for a Surface Power Reactor

    International Nuclear Information System (INIS)

    Pearson, J. Boise; Stewart, Eric T.; Reid, Robert S.

    2007-01-01

    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 deg. C. The CFD model with 1/6-g predicts a maximum water temperature of 88 deg. C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield

  13. Induction Heating Model of Cermet Fuel Element Environmental Test (CFEET)

    Science.gov (United States)

    Gomez, Carlos F.; Bradley, D. E.; Cavender, D. P.; Mireles, O. R.; Hickman, R. R.; Trent, D.; Stewart, E.

    2013-01-01

    Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames. Nuclear Thermal Rockets (NTR) are capable of producing a high specific impulse by employing heat produced by a fission reactor to heat and therefore accelerate hydrogen through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements due to large thermal gradients; therefore, high-melting-point ceramics-metallic matrix composites (cermets) are one of the fuels under consideration as part of the Nuclear Cryogenic Propulsion Stage (NCPS) Advance Exploration System (AES) technology project at the Marshall Space Flight Center. The purpose of testing and analytical modeling is to determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures and obtain data to assess the properties of the non-nuclear support materials. The fission process and the resulting heating performance are well known and do not require that active fissile material to be integrated in this testing. A small-scale test bed; Compact Fuel Element Environmental Tester (CFEET), designed to heat fuel element samples via induction heating and expose samples to hydrogen is being developed at MSFC to assist in optimal material and manufacturing process selection without utilizing fissile material. This paper details the analytical approach to help design and optimize the test bed using COMSOL Multiphysics for predicting thermal gradients induced by electromagnetic heating (Induction heating) and Thermal Desktop for radiation calculations.

  14. First-wall, blanket, and shield engineering test program for magnetically confined fusion power reactors

    International Nuclear Information System (INIS)

    Maroni, V.A.

    1980-01-01

    The key engineering areas identified for early study relate to FW/B/S system thermal-hydraulics, thermomechnics, nucleonics, electromagnetics, assembly, maintenance, and repair. Programmatic guidance derived frm planning exercises involving over thirty organizations (laboratories, industries, and universities) has indicated (1) that meaningful near term engineering testing should be feasible within the bounds of a modest funding base, (2) that there are existing facilities and expertise which can be profitably utilized in this testing, and (3) that near term efforts should focus on the measurement of engineering data and the verification/calibration of predictive methods for anticipated normal operational and transient FW/B/S conditions. The remainder of this paper discusses in more detail the planning strategies, proposed approach to near term testing, and longer range needs for integrated FW/B/S test facilities

  15. Demonstration testing and evaluation of in situ heating of soil

    International Nuclear Information System (INIS)

    1995-01-01

    This document describes the Quality Assurance Project Plan (QAPP) for IITRI Project C06787 entitled open-quotes Demonstration Testing and Evaluation of In Situ Heating of Soilclose quotes. A work plan for the above mentioned work was previously submitted. This QAPP describes the sampling and analysis of soil core-samples obtained from the K-25 Site (Oak Ridge Gaseous Diffusion Plant) where an in-situ heating and soil decontamination demonstration experiment will be performed. Soil samples taken before and after the experiment will be analyzed for selected volatile organic compounds. The Work Plan mentioned above provides a complete description of the demonstration site, the soil sampling plan, test plan, etc

  16. Chinese nuclear heating test reactor and demonstration plant

    International Nuclear Information System (INIS)

    Wang Dazhong; Ma Changwen; Dong Duo; Lin Jiagui

    1992-01-01

    In this report the importance of nuclear district heating is discussed. From the viewpoint of environmental protection, uses of energy resources and transport, the development of nuclear heating in China is necessary. The development program of district nuclear heating in China is given in the report. At the time being, commissioning of the 5 MW Test Heating Reactor is going on. A 200 MWt Demonstration Plant will be built. In this report, the main characteristics of these reactors are given. It shows this type of reactor has a high inherent safety. Further the report points out that for this type of reactor the stability is very important. Some experimental results of the driving facility are included in the report. (orig.)

  17. A Liquid-Liquid Thermoelectric Heat Exchanger as a Heat Pump for Testing Phase Change Material Heat Exchangers

    Science.gov (United States)

    Sheth, Rubik B.; Makinen, Janice; Le, Hung V.

    2016-01-01

    The primary objective of the Phase Change HX payload on the International Space Station (ISS) is to test and demonstrate the viability and performance of Phase Change Material Heat Exchangers (PCM HX). The system was required to pump a working fluid through a PCM HX to promote the phase change material to freeze and thaw as expected on Orion's Multipurpose Crew Vehicle. Due to limitations on ISS's Internal Thermal Control System, a heat pump was needed on the Phase Change HX payload to help with reducing the working fluid's temperature to below 0degC (32degF). This paper will review the design and development of a TEC based liquid-liquid heat exchanger as a way to vary to fluid temperature for the freeze and thaw phase of the PCM HX. Specifically, the paper will review the design of custom coldplates and sizing for the required heat removal of the HX.

  18. Use of a radio frequency shield during 1.5 and 3.0 Tesla magnetic resonance imaging: experimental evaluation

    Directory of Open Access Journals (Sweden)

    Favazza CP

    2014-10-01

    Full Text Available Christopher P Favazza, Deirdre M King, Heidi A Edmonson, Joel P Felmlee, Phillip J Rossman, Nicholas J Hangiandreou, Robert E Watson, Krzysztof R Gorny Department of Radiology, Mayo Clinic, Rochester, MN, USA Abstract: Radiofrequency (RF shields have been recently developed for the purpose of shielding portions of the patient's body during magnetic resonance imaging (MRI examinations. We present an experimental evaluation of a commercially available RF shield in the MRI environment. All tests were performed on 1.5 T and 3.0 T clinical MRI scanners. The tests were repeated with and without the RF shield present in the bore, for comparison. Effects of the shield, placed within the scanner bore, on the RF fields generated by the scanner were measured directly using tuned pick-up coils. Attenuation, by as much as 35 dB, of RF field power was found inside the RF shield. These results were supported by temperature measurements of metallic leads placed inside the shield, in which no measurable RF heating was found. In addition, there was a small, simultaneous detectable increase (~1 dB of RF power just outside the edges of the shield. For these particular scanners, the autocalibrated RF power levels were reduced for scan locations prescribed just outside the edges of the shield, which corresponded with estimations based on the pick-up coil measurements. Additionally, no significant heating during MRI scanning was observed on the shield surface. The impact of the RF shield on the RF fields inside the magnet bore is likely to be dependent on the particular model of the RF shield or the MRI scanner. These results suggest that the RF shield could be a valuable tool for clinical MRI practices. Keywords: radiofrequency shield, magnetic resonance imaging, radiofrequency attenuation

  19. Design, fabrication, and properties of a continuous carbon-fiber reinforced Sm_2O_3/polyimide gamma ray/neutron shielding material

    International Nuclear Information System (INIS)

    Wang, Peng; Tang, Xiaobin; Chai, Hao; Chen, Da; Qiu, Yunlong

    2015-01-01

    Highlights: • Sm_2O_3 is used for neutron absorber instead of B_4C, and Sm_2O_3 has a good photon-shielding effect. • Carbon-fiber cloth and polyimide were used to enhance shielding materials’ mechanical behavior and thermal behavior. • Both Monte Carlo method and shielding test were used to evaluate shielding performance of the novel shielding material. - Abstract: The design and fabrication of shielding materials with good heat-resistance and mechanical properties is a major problem in the radiation shielding field. In this paper, based on gamma ray and neutron shielding theory, a continuous carbon-fiber reinforced Sm_2O_3/polyimide gamma ray/neutron shielding material was fabricated by hot-pressing method. The material's application behavior was subsequently evaluated using neutron shielding, photon shielding, mechanical tensile, and thermogravimetric analysis–differential scanning calorimetry tests. The results show that the tensile strength of the novel shielding material exceeds 200 MPa, which makes it of similar strength to aluminum alloy. The material does not undergo crosslinking and decomposition reactions at 300 °C and it can be used in such environments for long periods of time. The continuous carbon-fiber reinforced Sm_2O_3/polyimide material has a good shielding performance with respect to gamma rays and neutrons. The material thus has good prospects for use in fusion reactor system and nuclear waste disposal applications.

  20. Neutron shielding material

    International Nuclear Information System (INIS)

    Nodaka, M.; Iida, T.; Taniuchi, H.; Yosimura, K.; Nagahama, H.

    1993-01-01

    From among the neutron shielding materials of the 'kobesh' series developed by Kobe Steel, Ltd. for transport and storage packagings, silicon rubber base type material has been tested for several items with a view to practical application and official authorization, and in order to determine its adaptability to actual vessels. Silicon rubber base type 'kobesh SR-T01' is a material in which, from among the silicone rubber based neutron shielding materials, the hydrogen content is highest and the boron content is most optimized. Its neutron shielding capability has been already described in the previous report (Taniuchi, 1986). The following tests were carried out to determine suitability for practical application; 1) Long-term thermal stability test 2) Pouring test on an actual-scale model 3) Fire test The experimental results showed that the silicone rubber based neutron shielding material has good neutron shielding capability and high long-term fire resistance, and that it can be applied to the advanced transport packaging. (author)

  1. Image dissector photocathode solar damage test program. [solar radiation shielding using a fast optical lens

    Science.gov (United States)

    Smith, R. A.

    1977-01-01

    Image dissector sensors of the same type which will be used in the NASA shuttle star tracker were used in a series of tests directed towards obtaining solar radiation/time damage criteria. Data were evaluated to determine the predicted level of operability of the star tracker if tube damage became a reality. During the test series a technique for reducing the solar damage effect was conceived and verified. The damage concepts are outlined and the test methods and data obtained which were used for verification of the technique's feasibility are presented. The ability to operate an image dissector sensor with the solar image focussed on the photocathode by a fast optical lens under certain conditions is feasible and the elimination of a mechanical protection device is possible.

  2. Induction Heating on Dynamic Tensile Tests in CEA Saclay

    International Nuclear Information System (INIS)

    Averty, X.; Yvon, P.; Duguay, C.; Pizzanelli, J. P.; Basini, V.

    2001-01-01

    The LCMI (Laboratory for characterization of irradiated materials), located in CEA from Saclay, is in charge of the mechanical tests on irradiated materials. The dynamic tensile testing machine, in a hot cell equipped with two remote handling, has been first improved in 1995, to fulfill the French safety programs on Reactivity Initiated Accident (RIA). One objective of this machine is to obtain mechanical property data on current Zircaloy cladding types needed to quality the cladding's response under RIA or LOCA transient loading and thermal conditions. For the RIA, this means testing at strain rates up to 5 s' and heating rates up to 200 degree centigree-s''-1, while for Loss of Coolant Accidents (LOCA) testing at strain rates of 10''-3 s''-1 and heating rates of 20 degree centigree s''-1 would be appropriate. The tensile samples are machined with a spark erosion machine, directly from pieces of cladding previously de fueled. Two kinds of samples can be machined in the cladding. Axial samples in order to test axial mechanical characteristics Ring samples in order to test transverse mechanical characteristics, more representative of RIA conditions. On one hand, the axial tensile tests were performed using the Joule effect, and heating rates up to about 500 degree centigree .s''-1 were obtained. This enabled us to perform the axial tests in a satisfactory manner. On the other hand, the tensile ring were first performed in a vertical furnace with a heating rate about 0.2 degree centigree.s''-1 and a thermal stability about 1 degree centigree. For temperatures above 480 degree centigree, the mechanical characteristics showed a sharp drop which could be attributed to irradiation defect annealing. Therefore we have recently developed an induction heating system to reach heating rates high enough (200 degree centigree.s''-1) to prevent any significant annealing before performing the ring tensile tests. To apply a uniaxial tangential tension, two matching half

  3. Cooling Performance of TBM-shield Designed for Manufacturability

    International Nuclear Information System (INIS)

    Park, Seong Dae; Lee, Dong Won; Kim, Dong Jun; Yoon, Jae Sung; Ahn, Mu Young

    2016-01-01

    Helium cooled ceramic reflector (HCCR) test blanket module (TBM) is composed of four sub-modules and a common back manifold (BM). The associated shield is a water-cooled 316L(N)-IG block with internal cooling channels. The purpose of the TBM-shield is to make the condition with the allowable neutron flux and dose rate level. The radially continuous layers of water and structure were configured. The main purpose of the shield is to reduce the neutron flux by absorbing the neutron in the structure. The water could act as the moderator and cool down the structure which is heated due to the reaction with the neutrons. The moderated neutrons are easily absorbed by the structure. It could meet the criteria for the minimum neutron flux by increasing the thickness of structure. The formation of inside cooling channel in the TBM-shield should be considered while maintaining the allowable temperature range. In this work, a manufacturing process including the formation of inside cooling channel was presented. Current design and thermal analysis results for the TBM-shield were presented. The geometry of the shield blocks was considerably changed. The coolant channel was exposed to the outer surface of the TBM-shield. The overall manufacturing process is simplified compared with the previous process of CD model

  4. Cooling Performance of TBM-shield Designed for Manufacturability

    Energy Technology Data Exchange (ETDEWEB)

    Park, Seong Dae; Lee, Dong Won; Kim, Dong Jun; Yoon, Jae Sung [KAERI, Daejeon (Korea, Republic of); Ahn, Mu Young [NFRI, Daejeon (Korea, Republic of)

    2016-05-15

    Helium cooled ceramic reflector (HCCR) test blanket module (TBM) is composed of four sub-modules and a common back manifold (BM). The associated shield is a water-cooled 316L(N)-IG block with internal cooling channels. The purpose of the TBM-shield is to make the condition with the allowable neutron flux and dose rate level. The radially continuous layers of water and structure were configured. The main purpose of the shield is to reduce the neutron flux by absorbing the neutron in the structure. The water could act as the moderator and cool down the structure which is heated due to the reaction with the neutrons. The moderated neutrons are easily absorbed by the structure. It could meet the criteria for the minimum neutron flux by increasing the thickness of structure. The formation of inside cooling channel in the TBM-shield should be considered while maintaining the allowable temperature range. In this work, a manufacturing process including the formation of inside cooling channel was presented. Current design and thermal analysis results for the TBM-shield were presented. The geometry of the shield blocks was considerably changed. The coolant channel was exposed to the outer surface of the TBM-shield. The overall manufacturing process is simplified compared with the previous process of CD model.

  5. The heating operational summarization in three winters of a 5 MW test heating reactor

    International Nuclear Information System (INIS)

    Wang Dazhong; Dong Duo; Su Qingshan; Zhang Yajun

    1992-09-01

    The 5 MW THR (5 MW test heating reactor) is a new type reactor with inherent safety developed by INET (Institute of Nuclear Energy Technology). It is the first 'pressure vessel type' heating reactor in operation in the world. It was put into operation in November, 1989. Since then it has operated for three winter seasons. The total operation time has reached to 8174 hours and its availability of heating has reached to 99%. The advanced technology of this reactor has been proved in the past three years operation. The characteristics of power regulating, load following, reactivity disturbance and the variation of parameters under the condition of ATWS (anticipated transients without scram) were studied with experiments in 5 MW THR. The 5 MW THR is an ideal heating reactor and has outstanding performances

  6. Simple exercise test for the prediction of relative heat tolerance

    International Nuclear Information System (INIS)

    Kenney, W.L.; Lewis, D.A.; Anderson, R.K.; Kamon, E.

    1986-01-01

    A medical screening exercise test is presented which accurately predicts relative heat tolerance during work in very hot environments. The test consisted of 15-20 min of exercise at a standard absolute intensity of about 600 kcal/hr (140W) with the subject wearing a vapor-barrier suit. Five minutes after the subject exercised, recovery heart rate was measured. When this heart rate is used, a physiological limit (+/- approximately 5 min) can be predicted with 95% confidence for the most intense work-heat conditions found in nuclear power stations. In addition, site health and safety personnel can establish qualification criteria for work on hot jobs, based on the test results. The test as developed can be performed in an office environment with the use of a minimum of equipment by personnel with minimal expertise and training. Total maximal test duration is about 20-25 min per person and only heart rate need be monitored (simple pulse palpation will suffice). Test modality is adaptable to any ergometer, the most readily available and least expensive of which is bench-stepping. It is recommended that this test be available for use for those persons who, based upon routine medical examination or past history, are suspected of being relatively heat intolerant

  7. 2003 International High-Level Radioactive Waste Management Conference Breached Drip Shield Test and Validation of a TSPA Sub-Model

    International Nuclear Information System (INIS)

    Walton, Z.P.; Kam, J.T.

    2002-01-01

    The Engineered Barrier System (EBS) represents the system of human engineered barriers in the isolation of high-level radioactive waste in the proposed repository at Yucca Mountain. It is designed to complement and enhance the natural barriers to isolate and prevent the transport of radionuclides into the surrounding environment. The transport mechanism most frequently postulated for radionuclides is liquid water flux that has penetrated the EBS through corrosion breaches in the drip shield and waste packages (WP). A water flux-splitting model is used to predict flow through WP and drip shield breaches and is documented in the ''EBS Radionuclide Transport Abstraction''. A future revision of the ''EBS Radionuclide Transport Abstraction'' will be one component of the total system performance assessment--license application (TSPA-LA) for the Yucca Mountain repository. The flux-splitting model is conservative based on the following assumptions: (1) Drip impact occurs without a loss of water mass. (2) Dripping flux falls exactly at the crown of the drip shield as opposed to different locations on the curved surface, which will effect splashing and flow patterns. (3) The flux passing through a drip shield patch is proportional to the ratio of the length of the penetration in the axial direction to the total axial length of the drip shield. In this assumption all fluid that drips and flows from the drip shield crown toward a penetration will be collected if the axial locations of the source and patch coincide. (4) The potential for evaporation is ignored. Because of these conservatisms, the current version of the flux-splitting model is incapable of accounting for water that has been splashed from the impact location, the deviation of water paths (rivulets) from the axis of impact, and water loss due to evaporation. This paper will present the results of a series of breached drip shield tests used to collect empirical data for the initial validation and further

  8. Demonstration testing and evaluation of in situ soil heating

    International Nuclear Information System (INIS)

    Sresty, G.C.

    1994-01-01

    A Treatability Study planned for the demonstration of the in situ electromagnetic (EM) heating process to remove organic solvents is described in this Work Plan. The treatability study will be conducted by heating subsurface vadose-zone soils in an organic plume adjacent to the Classified Burial Ground K-1070-D located at K-25 Site, Oak Ridge. The test is scheduled to start during the fourth quarter of FY94 and will be completed during the first quarter of FY95. The EM heating process for soil decontamination is based on volumetric heating technologies developed during the '70s for the recovery of fuels from shale and tar sands by IIT Research Institute (IITRI) under a co-operative program with the US Department of Energy (DOE). Additional modifications of the technology developed during the mid '80s are currently used for the production of heavy oil and waste treatment. Over the last nine years, a number of Government agencies (EPA, Army, AF, and DOE) and industries sponsored further development and testing of the in situ heating and soil decontamination process for the remediation of soils containing hazardous organic contaminants. In this process the soil is heated in situ using electrical energy. The contaminants are removed from the soil due to enhanced vaporization, steam distillation and stripping. IITRI will demonstrate the EM Process for in situ soil decontamination at K-25 Site under the proposed treatability study. Most of the contaminants of concern are volatile organics which can be removed by heating the soil to a temperature range of 85 to 95 C. The efficiency of the treatment will be determined by comparing the concentration of contaminants in soil samples. Samples will be obtained before and after the demonstration for a measurement of the concentration of contaminants of concern

  9. Neutron shielding material

    International Nuclear Information System (INIS)

    Suzuki, Shigenori; Iimori, Hiroshi; Kobori, Junzo.

    1980-01-01

    Purpose: To provide a neutron shielding material which incorporates preferable shielding capacity, heat resistance, fire resistance and workability by employing a mixture of thermosetting resin, polyethylene and aluminium hydroxide in special range ratio and curing it. Constitution: A mixture containing 20 to 60% by weight of thermosetting resin having preferable heat resistance, 10 to 40% by weight of polyethylene powder having high hydrogen atom density and 1000 to 60000 of molecular weight, and 15 to 55% by weight of Al(OH) 3 for imparting fire resistance and self-fire extinguishing property thereto is cured. At this time approx. 0.5 to 5% of curing catalyst of the thermosetting resin is contained in 100 parts by weight of the mixture. (Sekiya, K.)

  10. Current Status and Performance Tests of Korea Heat Load Test Facility KoHLT-EB

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sukkwon; Jin, Hyunggon; Shin, Kyuin; Choi, Boguen; Lee, Eohwak; Yoon, Jaesung; Lee, Dongwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Duckhoi; Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    A commissioning test has been scheduled to establish the installation and preliminary performance experiments of the copper hypervapotron mockups. And a qualification test will be performed to evaluate the CuCrZr duct liner in the ITER neutral beam injection facility and the ITER first wall small-scale mockups of the semi-prototype, at up to 1.5 and 5 MW/m{sup 2} high heat flux. Also, this system will be used to test other PFCs for ITER and materials for tokamak reactors. Korean high heat flux test facility(KoHLT-EB; Korea Heat Load Test facility - Electron Beam) by using an electron beam system has been constructed in KAERI to perform the qualification test for ITER blanket FW semi-prototype mockups, hypervapotron cooling devices in fusion devices, and other ITER plasma facing components. The commissioning and performance tests with the supplier of e-gun system have been performed on November 2012. The high heat flux test for hypervapotron cooling device and calorimetry were performed to measure the surface heat flux, the temperature profile and cooling performance. Korean high heat flux test facility for the plasma facing components of nuclear fusion machines will be constructed to evaluate the performance of each component. This facility for the plasma facing materials will be equipped with an electron beam system with a 60 kV acceleration gun.

  11. Radiation shielding

    International Nuclear Information System (INIS)

    Aitken, D.

    1979-01-01

    Shields for equipment in which ionising radiation is associated with high electrical gradients, for example X-ray tubes and particle accelerators, incorporate a radiation-absorbing metal, as such or as a compound, and are electrically non-conducting and can be placed in the high electrical gradient region of the equipment. Substances disclosed include dispersions of lead, tungsten, uranium or oxides of these in acrylics polyesters, PVC, ABS, polyamides, PTFE, epoxy resins, glass or ceramics. The material used may constitute an evacuable enclosure of the equipment or may be an external shield thereof. (U.K.)

  12. Qualification test for ITER HCCR-TBS mockups with high heat flux test facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Suk-Kwon, E-mail: skkim93@kaeri.re.kr [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Seong Dae; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    Highlights: • The test mockups for ITER HCCR (Helium Cooled Ceramic Reflector) TBS (Test Blanket System) in Korea were designed and fabricated. • A thermo-hydraulic analysis was performed using a high heat flux test facility by using electron beam. • The plan for qualification tests was developed to evaluate the thermo-hydraulic efficiency in accordance with the requirements of the ITER Organization. - Abstract: The test mockups for ITER HCCR (Helium Cooled Ceramic Reflector) TBS (Test Blanket System) in Korea were designed and fabricated, and an integrity and thermo-hydraulic performance test should be completed under the same or similar operation conditions of ITER. The test plan for a thermo-hydraulic analysis was developed by using a high heat flux test facility, called the Korean heat load test facility by using electron beam (KoHLT-EB). This facility is utilized for a qualification test of the plasma facing component (PFC) for the ITER first wall and DEMO divertor, and for the thermo-hydraulic experiments. In this work, KoHLT-EB will be used for the plan of the performance qualification test of the ITER HCCR-TBS mockups. This qualification tests should be performed to evaluate the thermo-hydraulic efficiency in accordance with the requirements of the ITER Organization (IO), which describe the specifications and qualifications of the heat flux test facility and test procedure for ITER PFC.

  13. Design experience: CRBRP radiation shielding

    International Nuclear Information System (INIS)

    Disney, R.K.; Chan, T.C.; Gallo, F.G.; Hedgecock, L.R.; McGinnis, C.A.; Wrights, G.N.

    1978-11-01

    The Clinch River Breeder Reactor Plant (CRBRP) is being designed as a fast breeder demonstration project in the U.S. Liquid Metal Fast Breeder Reactor (LMFBR) program. Radiation shielding design of the facility consists of a comprehensive design approach to assure compliance with design and government regulatory requirements. Studies conducted during the CRBRP design process involved the aspects of radiation shielding dealing with protection of components, systems, and personnel from radiation exposure. Achievement of feasible designs, while considering the mechanical, structural, nuclear, and thermal performance of the component or system, has required judicious trade-offs in radiation shielding performance. Specific design problems which have been addressed are in-vessel radial shielding to protect permanent core support structures, flux monitor system shielding to isolate flux monitoring systems for extraneous background sources, reactor vessel support shielding to allow personnel access to the closure head during full power operation, and primary heat transport system pipe chaseway shielding to limit intermediate heat transport system sodium system coolant activation. The shielding design solutions to these problems defined a need for prototypic or benchmark experiments to provide assurance of the predicted shielding performance of selected design solutions and the verification of design methodology. Design activities of CRBRP plant components an systems, which have the potential for radiation exposure of plant personnel during operation or maintenance, are controlled by a design review process related to radiation shielding. The program implements design objectives, design requirements, and cost/benefit guidelines to assure that radiation exposures will be ''as low as reasonably achievable''

  14. Hydrodynamical tests with an original PWR heat removal pump

    International Nuclear Information System (INIS)

    Wietstock, P.

    1984-01-01

    GKSS-Forschungszentrum performes hydrodynamical tests with an original PWR heat removal pump to analyse the influences of fluid parameters on the capacity and cavitation behavior of the pump in order to get further improvements of the quantification of the reached safety-level. It can be concluded, that in case of the tested heat removal pump the additional loads during transition from cavitation free operation into fully cavitation for the investigated operation point with 980 m 3 /h will be smaller than the alteration of loads during passing through the total characteristic. The results from cavitation tests for other operation points indicate, that this very important consequence especially for accident operation will be valid for the total specified pump flow area. (orig.)

  15. Heat transfer and friction on smooth and rough test rods

    International Nuclear Information System (INIS)

    Franken, W.M.P.; Hoogland, H.; Deijman, P.

    1977-06-01

    Results are reported on heat transfer and pressure drop tests on one smooth and nine rough test rods in an annular geometry. The wall roughness consisted of transversal ribs with various roughness pitches, rib heights and rib widths. The tests were performed with air as coolant under a wide range of experimental conditions: 10 5 5 , 1.1 2. Special attention has been given to the effect of variation of the physical coolant properties over the flow cross section. This effect could be described by the power function (Tsub(w)/Tsub(b))sup(-0.3l) in additional systematic variation of the heat transfer could be recognized, dependent on the coolant temperature level. The experimental results were correlated by the equation St = C(Tsub(in)) Resup(-0.2) Prsup(-0.6) (Tsub(w)/Tsub(b)sup(-0.31). Values of C(Tsub(in)) are given in tabular form. The thermal entrance effect has been measured on various test rods. A substantial reduction of the heat transfer coefficient was almost constant along the rough test rods

  16. Gonadal shield.

    Science.gov (United States)

    Purdy, J A; Stiteler, R D; Glasgow, G P; Mill, W B

    1975-10-01

    A secondary gonadal shield for use in the pelvic irradiation of males was designed and built using material and apparatus available with the Cerrobend blocking system. The gonadal dose was reduced to approximately 1.5 to 2.5% of the given dose.

  17. SHIELD verification and validation report

    International Nuclear Information System (INIS)

    Boman, C.

    1992-02-01

    This document outlines the verification and validation effort for the SHIELD, SHLDED, GEDIT, GENPRT, FIPROD, FPCALC, and PROCES modules of the SHIELD system code. Along with its predecessors, SHIELD has been in use at the Savannah River Site (SRS) for more than ten years. During this time the code has been extensively tested and a variety of validation documents have been issued. The primary function of this report is to specify the features and capabilities for which SHIELD is to be considered validated, and to reference the documents that establish the validation

  18. F-doped VO2 nanoparticles for thermochromic energy-saving foils with modified color and enhanced solar-heat shielding ability.

    Science.gov (United States)

    Dai, Lei; Chen, Shi; Liu, Jianjun; Gao, Yanfeng; Zhou, Jiadong; Chen, Zhang; Cao, Chuanxiang; Luo, Hongjie; Kanehira, Minoru

    2013-07-28

    F-doped VO2 (M1) nanoparticles were prepared via one-pot hydrothermal synthesis. The F-doping can minimise the size of the VO2 (M1) nanoparticles, induce a homogeneous size distribution and effectively decrease the phase transition temperature to 35 °C at 2.93% F in VO2. VO2 smart glass foils obtained by casting these nanoparticles exhibit excellent thermochromism in the near-infrared region, which suggests that these foils can be used for energy-efficient glass. Compared to a pure VO2 foil, the 2.93% F-doped VO2 foil exhibits an increased solar-heat shielding ability (35.1%) and a modified comfortable colour, while still retaining an excellent solar modulation ability (10.7%) and an appropriate visible transmittance (48.7%). The F-doped VO2 foils are the first to simultaneously meet the requirements of a reduced phase transition temperature, diluted colour and excellent thermochromic properties, and these properties make the further improved F-doped VO2 foils suitable for commercial applications in energy efficient glass.

  19. Gonadal Shielding in Radiography: A Best Practice?

    Science.gov (United States)

    Fauber, Terri L

    2016-11-01

    To investigate radiation dose to phantom testes with and without shielding. A male anthropomorphic pelvis phantom was imaged with thermoluminescent dosimeters (TLDs) placed in the right and left detector holes corresponding to the testes. Ten exposures were made of the pelvis with and without shielding. The exposed TLDs were packaged securely and mailed to the University of Wisconsin Calibration Laboratory for reading and analysis. A t test was calculated for the 2 exposure groups (no shield and shielded) and found to be significant, F = 8.306, P shield was used during pelvic imaging. Using a flat contact shield during imaging of the adult male pelvis significantly reduces radiation dose to the testes. Regardless of the contradictions in the literature on gonadal shielding, the routine practice of shielding adult male gonads during radiographic imaging of the pelvis is a best practice. © 2016 American Society of Radiologic Technologists.

  20. Measuring space radiation shielding effectiveness

    Directory of Open Access Journals (Sweden)

    Bahadori Amir

    2017-01-01

    Full Text Available Passive radiation shielding is one strategy to mitigate the problem of space radiation exposure. While space vehicles are constructed largely of aluminum, polyethylene has been demonstrated to have superior shielding characteristics for both galactic cosmic rays and solar particle events due to the high hydrogen content. A method to calculate the shielding effectiveness of a material relative to reference material from Bragg peak measurements performed using energetic heavy charged particles is described. Using accelerated alpha particles at the National Aeronautics and Space Administration Space Radiation Laboratory at Brookhaven National Laboratory, the method is applied to sample tiles from the Heat Melt Compactor, which were created by melting material from a simulated astronaut waste stream, consisting of materials such as trash and unconsumed food. The shielding effectiveness calculated from measurements of the Heat Melt Compactor sample tiles is about 10% less than the shielding effectiveness of polyethylene. Shielding material produced from the astronaut waste stream in the form of Heat Melt Compactor tiles is therefore found to be an attractive solution for protection against space radiation.

  1. Measuring space radiation shielding effectiveness

    Science.gov (United States)

    Bahadori, Amir; Semones, Edward; Ewert, Michael; Broyan, James; Walker, Steven

    2017-09-01

    Passive radiation shielding is one strategy to mitigate the problem of space radiation exposure. While space vehicles are constructed largely of aluminum, polyethylene has been demonstrated to have superior shielding characteristics for both galactic cosmic rays and solar particle events due to the high hydrogen content. A method to calculate the shielding effectiveness of a material relative to reference material from Bragg peak measurements performed using energetic heavy charged particles is described. Using accelerated alpha particles at the National Aeronautics and Space Administration Space Radiation Laboratory at Brookhaven National Laboratory, the method is applied to sample tiles from the Heat Melt Compactor, which were created by melting material from a simulated astronaut waste stream, consisting of materials such as trash and unconsumed food. The shielding effectiveness calculated from measurements of the Heat Melt Compactor sample tiles is about 10% less than the shielding effectiveness of polyethylene. Shielding material produced from the astronaut waste stream in the form of Heat Melt Compactor tiles is therefore found to be an attractive solution for protection against space radiation.

  2. Niobium 1 percent zirconium/potassium and titanium/potassium life-test heat pipe design and testing

    Science.gov (United States)

    Sena, J. Tom; Merrigan, Michael A.

    Experimental lifetime performance studies currently in progress use Niobium 1 percent Zirconium (Nb-1Zr) and Titanium (Ti) heat pipes with potassium (K) as the working fluid. A heat pipe life test matrix was developed for testing the heat pipes. Because the corrosion rates in alkali metal heat pipes are affected by temperature and working fluid evaporation flux, the variable parameters of the experimental matrix are established as steady operating temperature and input heat flux density. Total impurity inventory is a factor in corrosion rate so impurity levels are being evaluated in the heat pipe materials before and after testing. Eight Nb-1Zr/K heat pipes were designed, fabricated, and tested. Two of the heat pipes have completed testing whereas the other six are currently in test. These are gravity assist heat pipes operating in a reflux mode. The heat pipes are tested by sets, one set of two and two sets of three heat pipes. Three Ti/K heat pipes are also in life test. These heat pipes are tested as a set in a horizontal position in a capillary pumped annular flow mode. Each of the heat pipes is encapsulated in a quartz vacuum container with a water calorimeter over the vacuum container for power throughput measurements. Thermocouples are attached to the heat pipes for measuring temperature. Heat input to the heat pipes is via an RF coil. The heat pipes are operating at between 800 and 900 K, with heat input fluxes of 13.8 to 30 W/sq cm. Of the Nb-1Zr/K heat pipes, two of the heat pipes have been in operation for 14,000 hours, three over 10,000 hours, and three over 7,000 hours. The Ti/K heat pipes have been in operation for 1,266 hours.

  3. HEAT ENGINEERING TESTING OF AIR COOLING UNIT OF HORIZONTAL TYPE

    OpenAIRE

    Rohachov, Valerii Andriiovych; Semeniako, Oleksandr Volodymyrovych; Лазоренко, Р. О.; Середа, Р. М.; Parafeinyk, Volodymyr Petrovych

    2018-01-01

    The results of the thermal tests of the section of air cooler, the heat-exchange surface of which is made up of chess package of bimetal finned tubes are presented. The methods of research are presented, the experimental stand is described, the measurement errors are given. The efficiency of the experimental stand and the accuracy of the experimental data on it are confirmed. Proposed to use the stand for researches of air cooling units with other types and sections of finned tubes.

  4. Tube vibration in industrial size test heat exchanger

    International Nuclear Information System (INIS)

    Halle, H.; Wambsganss, M.W.

    1980-03-01

    Tube vibration data from tests of a specially built and instrumented, industrial-type, shell-and-tube heat exchanger are reported. The heat exchanger is nominally 0.6 m (2 ft) in dia and 3.7 m (12 ft) long. Both full tube and no-tubes-in-window bundles were tested for inlet/outlet nozzles of different sizes and with the tubes supported by seven, equally-spaced, single-segmental baffles. Prior to water flow testing, natural frequencies and damping of representative tubes were measured in air and water. Flow testing was accomplished by increasing the flow rates in stepwise fashion and also by sweeping through a selected range of flow rates. The primary variables measured and reported are tube accelerations and/or displacements and pressure drop through the bundle. Tests of the full tube bundle configuration revealed tube rattling to occur at intermediate flow rates, and fluidelastic instability, with resultant tube impacting, to occur when the flow rate exceeded a threshold level; principally, the four-span tubes were involved in the regions immediately adjacent to the baffle cut. For the range of flow rates tested, fluidelastic instability was not achieved in the no-tubes-in-window bundle; in this configuration the tubes are supported by all seven baffles and are, therefore, stiffer

  5. PANDA passive decay heat removal transient test results

    International Nuclear Information System (INIS)

    Bandurski, Th.; Dreier, J.; Huggenberger, M.

    1997-01-01

    PANDA is a large scale facility for investigating the long-term decay heat removal from the containment of a next generation of 'passive' Advanced Light Water Reactors (ALWR). PANDA was used to examine the long-term LOCA response of the Passive Containment Cooling System (PCCS) for the General Electric (GE) Simplified Boiling Water Reactor (SBWR). The first PANDA test series had the dual objectives of demonstrating the performance of the SBWR PCCS and extending the data base available for containment analysis code qualification. The test objectives also include the study of the effects of mixing and stratification of steam and noncondensible gases in the drywell (DW) and in the suppression chamber or wetwell (WW). Ten tests were conducted in the course of the PANDA SBWR Program. The tests demonstrated a favorable and robust overall PCCS performance under different conditions. The present paper focuses on the main phenomena observed during the tests with respect to PCCS operation and DW gas mixing. (author)

  6. Experience with a servo-hydraulic mechanical testing machine installed in a new shielded active facility at Windscale Nuclear Power Development Laboratories

    International Nuclear Information System (INIS)

    Garlick, A.; Hindmarch, P.; Gravenor, J.G.; Rhodes, D.

    1982-03-01

    An Instron model 1273 servo-hydraulic machine has been installed within a lead-shielded cell at Windscale in order to provide a facility capable of performing a wide range of mechanical tests on nuclear reactor structural materials and fuel assembly components. This particular type of machine was chosen because it has design features associated with the load frame, location of the actuator and adjustment and clamping of the cross-head that are especially well suited to remote operation within a shielded cell. The design of the testing facility is described and the programmes of work that have been completed over the past 11/2 years of operation are reviewed. (author)

  7. Performance of the JT-60SA cryogenic system under pulsed heat loads during acceptance tests

    Science.gov (United States)

    Hoa, C.; Bonne, F.; Roussel, P.; Lamaison, V.; Girard, S.; Fejoz, P.; Goncalves, R.; Vallet, J. C.; Legrand, J.; Fabre, Y.; Pudys, V.; Wanner, M.; Cardella, A.; Di Pietro, E.; Kamiya, K.; Natsume, K.; Ohtsu, K.; Oishi, M.; Honda, A.; Kashiwa, Y.; Kizu, K.

    2017-12-01

    The JT-60SA cryogenic system a superconducting tokamak currently under assembly at Naka, Japan. After one year of commissioning, the acceptance tests were successfully completed in October 2016 in close collaboration with Air Liquide Advanced Technologies (ALaT), the French atomic and alternative energies commission (CEA), Fusion for Energy (F4E) and the Quantum Radiological Science and Technology (QST). The cryogenic system has several cryogenic users at various temperatures: the superconducting magnets at 4.4 K, the current leads at 50 K, the thermal shields at 80 K and the divertor cryo-pumps at 3.7 K. The cryogenic system has an equivalent refrigeration power of about 9.5 kW at 4.5 K, with peak loads caused by the nuclear heating, the eddy currents in the structures and the AC losses in the magnets during cyclic plasma operation. The main results of the acceptance tests will be reported, with emphasis on the management of the challenging pulsed load operation using a liquid helium volume of 7 m3 as a thermal damper.

  8. Stress relaxation in tempered glass caused by heat soak testing

    DEFF Research Database (Denmark)

    Schneider, Jens; Hilcken, Jonas; Aronen, Antti

    2016-01-01

    Heat soak testing of tempered glass is a thermal process required after the tempering process itself to bring glasses of commercial soda-lime-silica-glass to failure that are contaminated with nickel sulphide inclusions, diameter 50 mm to 500 mm typically. Thus, the tests avoid a so-called "spont...... of commercial soda-lime-silica glass, it causes stress relaxation in tempered glass and the fracture pattern of the glass changes accordingly, especially thin glasses are affected. Based on the Tool-Narayanaswamy-Model, this paper comprises the theoretical background of the stress...

  9. Electrostatic shielding of transformers

    Energy Technology Data Exchange (ETDEWEB)

    De Leon, Francisco

    2017-11-28

    Toroidal transformers are currently used only in low-voltage applications. There is no published experience for toroidal transformer design at distribution-level voltages. Toroidal transformers are provided with electrostatic shielding to make possible high voltage applications and withstand the impulse test.

  10. Shielded container

    International Nuclear Information System (INIS)

    Fries, B.A.

    1978-01-01

    A shielded container for transportation of radioactive materials is disclosed in which leakage from the container is minimized due to constructional features including, inter alia, forming the container of a series of telescoping members having sliding fits between adjacent side walls and having at least two of the members including machine sealed lids and at least two of the elements including hand-tightenable caps

  11. Results of high heat flux tests and structural analysis of the new solid tungsten divertor tile for ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Jaksic, Nikola, E-mail: nikola.jaksic@ipp.mpg.de; Greuner, Henri; Herrmann, Albrecht; Böswirth, Bernd; Vorbrugg, Stefan

    2015-10-15

    Highlights: • The main motivation for the HHF investigation of tungsten tiles was an untypical deformation of some specimens under thermal loading, observed during the previous tests in GLADIS test facility. • A nonlinear finite element (FE) model for simulations of the GLADIS tests has been built. • The unexpected plastic deformations are mainly caused by internal stresses due to the manufacturing process. The small discrepancies among the FEA investigated and measured plastic deformations are most likely caused, beside of the practical difficulties by measuring of low items, also by tile internal stresses. • The influences of the residual stresses caused by special production processes have to be taken into account by design of the structural part made of solid tungsten. - Abstract: Tungsten as plasma-facing material for fusion devices is currently the most favorable candidate. In general solid tungsten is used for shielding the plasma chamber interior against the high heat generated from the plasma. For the purposes of implementation at ASDEX Upgrade and as a contribution to ITER the thermal performance of tungsten tiles has been extensively tested in the high heat flux test facility GLADIS during the development phase and beyond. These tests have been performed on full scale tungsten tile prototypes including their clamping and cooling structure. Simulating the adiabatically thermal loading due to plasma operation in ASDEX Upgrade, the tungsten tiles have been subjected to a thermal load with central heat flux of 10–24 MW/m{sup 2} and absorbed energy between 370 and 680 kJ. This loading results in maximum surface temperatures between 1300 °C and 2800 °C. The tests in GLADIS have been accompanied by intensive numerical investigations using FEA methods. For this purpose a multiple nonlinear finite element model has been set up. This paper discusses the main results of the high heat flux final tests and their numerical simulation. Moreover, first

  12. Results of high heat flux tests and structural analysis of the new solid tungsten divertor tile for ASDEX Upgrade

    International Nuclear Information System (INIS)

    Jaksic, Nikola; Greuner, Henri; Herrmann, Albrecht; Böswirth, Bernd; Vorbrugg, Stefan

    2015-01-01

    Highlights: • The main motivation for the HHF investigation of tungsten tiles was an untypical deformation of some specimens under thermal loading, observed during the previous tests in GLADIS test facility. • A nonlinear finite element (FE) model for simulations of the GLADIS tests has been built. • The unexpected plastic deformations are mainly caused by internal stresses due to the manufacturing process. The small discrepancies among the FEA investigated and measured plastic deformations are most likely caused, beside of the practical difficulties by measuring of low items, also by tile internal stresses. • The influences of the residual stresses caused by special production processes have to be taken into account by design of the structural part made of solid tungsten. - Abstract: Tungsten as plasma-facing material for fusion devices is currently the most favorable candidate. In general solid tungsten is used for shielding the plasma chamber interior against the high heat generated from the plasma. For the purposes of implementation at ASDEX Upgrade and as a contribution to ITER the thermal performance of tungsten tiles has been extensively tested in the high heat flux test facility GLADIS during the development phase and beyond. These tests have been performed on full scale tungsten tile prototypes including their clamping and cooling structure. Simulating the adiabatically thermal loading due to plasma operation in ASDEX Upgrade, the tungsten tiles have been subjected to a thermal load with central heat flux of 10–24 MW/m"2 and absorbed energy between 370 and 680 kJ. This loading results in maximum surface temperatures between 1300 °C and 2800 °C. The tests in GLADIS have been accompanied by intensive numerical investigations using FEA methods. For this purpose a multiple nonlinear finite element model has been set up. This paper discusses the main results of the high heat flux final tests and their numerical simulation. Moreover, first results

  13. ATLAS IV in situ heating test in Boom Clay

    International Nuclear Information System (INIS)

    Chen, Guangjing; Li, Xiangling; Verstricht, Jan; Sillen, Xavier

    2012-01-01

    Document available in extended abstract form only. The small scale in-situ ATLAS (Admissible Thermal Loading for Argillaceous Storage) tests are performed to assess the hydro-mechanical effects of a thermal transient on the host Boom clay at the HADES underground research facility in Mol, Belgium. The initial test set-up, consisting of a heater borehole and two observation boreholes, was installed in 1991-1992. The first test (later named 'ATLAS I') was then performed from July 1993 to June 1996; during this time, the heater dissipated a constant power of 900 W. During the second phase ('ATLAS II'), the heating power was doubled (1800 W) and maintained constant from June 1996 to May 1997. This was followed by shutdown and natural cooling starting from June 1997 on. To broaden the THM characterization of the Boom clay at a larger scale and at different temperature levels, the test set-up was extended in 2006 by drilling two additional instrumented boreholes (AT97E and AT98E). The heater was switched on again from April 2007 to April 2008 with a stepwise power increase, followed by an instantaneous shutdown. This phase is called 'ATLAS III'. The above tests have provided a large set of good quality and well documented data on temperature, pore water pressure and total stress; these data allowed to make several interesting observations regarding the thermal anisotropy and THM coupling in the Boom clay. The straightforward geometry and well defined boundary conditions of the tests facilitate the comparison between measurement and numerical modeling studies. Based on the three dimensional coupled THM modeling of the ATLAS III test, the good agreement between measurement and numerical modeling of temperature and pore water pressure yields a set of THM parameters and confirms the thermo-mechanical anisotropy of the Boom clay. To get a better insight in the anisotropic THM behavior of the Boom clay, a new upward instrumented borehole was drilled above the ATLAS heater at

  14. Setup for EMI Shielding Effectiveness Tests of Electrically Conductive Polymer Composites at Frequencies up to 3.0 GHz

    NARCIS (Netherlands)

    Vieira Valente, R.D.; De Ruijter, Chris; Vlasveld, Daniel; van der Zwaag, S.; Groen, W.A.

    2017-01-01

    Conductive polymer composites have been receiving increased interest both from the scientific community and industry with a special focus on electromagnetic interference (EMI) shielding applications. In this paper, we present the design, EM wave simulation, and validation through S-parameters

  15. Setup for EMI Shielding Effectiveness Tests of Electrically Conductive Polymer Composites at Frequencies up to 3.0 GHz

    NARCIS (Netherlands)

    Valente, R.; Ruijter, C. de; Vlasveld, D.; Zwaag, S. van der; Groen, P.

    2017-01-01

    Conductive polymer composites have been receiving increased interest both from the scientific community and industry with a special focus on electromagnetic interference (EMI) shielding applications. In this paper, we present the design, EM wave simulation, and validation through S-parameters

  16. Testing and Performance Validation of a Shielded Waste Segregation and Clearance Monitor Designed for the Measurement of Low Level Waste-13043

    International Nuclear Information System (INIS)

    Mason, John A.; Burke, Kevin J.; Towner, Antony C.N.; Beaven, Graham; Spence, Robert

    2013-01-01

    This paper describes the development, testing and validation of a shielded waste segregation and clearance monitor designed for the measurement of low-density low-level waste (LLW). The monitor is made of a measurement chamber surrounded by detectors and a shielded outer frame. The shielded chamber consists of a steel frame, which contains typically 1.5 inches (3.81 cm) of lead and 0.5 inches (1.27 cm) of steel shielding. Inside the shielding are plastic scintillator panels, which serve as gross gamma ray detectors. The detector panels, with embedded photomultipliers, completely surround the internal measurement chamber on all 6 sides. Care has been taken to distribute the plastic scintillator detectors in order to optimise both the efficiency for gamma ray detection and at the same time achieve a volumetric sensitivity, which is as uniform as possible. A common high voltage power supply provides the bias voltage for each of the six photomultipliers. The voltage signals arising from the detectors and photomultipliers are amplified by six sensitive amplifiers. Each amplifier incorporates a single channel analyser with both upper and lower thresholds and the digitised counts from each detector are recorded on six scalars. Operation of the device is by means of a microprocessor from which the scalars are controlled. An internal load cell linked to the microprocessor determines the weight of the waste object, and this information is used to calculate the specific activity of the waste. The monitor makes background measurements when the shielded door is closed and a sample, usually a bag of low-density waste, is not present in the measurement chamber. Measurements of the minimum detectable activity (MDA) of an earlier large volume prototype instrument are reported as part of the development of the Waste Segregation and Clearance Monitor (WSCM) described in the paper. For the optimised WSCM a detection efficiency of greater than 32% was measured using a small Cs-137

  17. Performance, Facility Pressure Effects, and Stability Characterization Tests of NASA's Hall Effect Rocket with Magnetic Shielding Thruster

    Science.gov (United States)

    Kamhawi, Hani; Huang, Wensheng; Haag, Thomas; Yim, John; Herman, Daniel; Williams, George; Gilland, James; Peterson, Peter; Hofer, Richard; Mikellides, Ioannis

    2016-01-01

    NASAs Hall Effect Rocket with Magnetic Shielding (HERMeS) 12.5 kW Technology Demonstration Unit-1 (TDU-1) Hall thruster has been the subject of extensive technology maturation in preparation for flight system development. Part of the technology maturation effort included experimental evaluation of the TDU-1 thruster with conducting and dielectric front pole cover materials in two different electrical configurations. A graphite front pole cover thruster configuration with the thruster body electrically tied to cathode and an alumina front pole cover thruster configuration with the thruster body floating were evaluated. Both configurations were also evaluated at different facility background pressure conditions to evaluate background pressure effects on thruster operation. Performance characterization tests found that higher thruster performance was attained with the graphite front pole cover configuration with the thruster electrically tied to cathode. A total thrust efficiency of 68 and a total specific impulse of 2,820 s was demonstrated at a discharge voltage of 600 V and a discharge power of 12.5 kW. Thruster stability regimes were characterized with respect to the thruster discharge current oscillations and with maps of the current-voltage-magnetic field (IVB). Analysis of TDU-1 discharge current waveforms found that lower normalized discharge current peak-to-peak and root mean square magnitudes were attained when the thruster was electrically floated with alumina front pole covers. Background pressure effects characterization tests indicated that the thruster performance and stability was mostly invariant to changes in the facility background pressure for vacuum chamber pressure below 110-5 Torr-Xe (for thruster flow rate above 8 mgs). Power spectral density analysis of the discharge current waveform showed that increasing the vacuum chamber background pressure resulted in a higher discharge current dominant frequency. Finally the IVB maps of the TDU-1

  18. Eddy current test of fin tubes for a heat exchanger

    International Nuclear Information System (INIS)

    KIm, Young Joo; Lee, Se Kyung; Chung, Min Hwa

    1992-01-01

    Eddy current probes were designed for the test of fin tubes. Fin tubes, often used for heat exchangers, have uneven outer and inner surfaces to enhance the heat emission. The surface roughness make it difficult to detect flaws employing eddy current test(ECT). In order to overcome the difficulties we performed two types of works, one is the delopment of ECT probes, and the other is the signal processing including fast Fourier transform and digital filtering. In the development of ECT probes, we adopted empirical design method. Our ECT probes for fin tubes are inside diameter type. And we are specially concerned about geometric features such as the widths of the coils composing an ECT probe. We fabricated four probes with various coil widths. Eddy current test was performed using those ECT probes on specimens with artificial flaws. After analyzing the output signals, we found that, in order for the effective testing, the width of a coil should be determined considering the pitch of the fins of a tube. And we also learned that the frequency filtering could improve the s/n ratio.

  19. Manufacturing and thermomechanical testing of actively cooled all beryllium high heat flux test pieces

    International Nuclear Information System (INIS)

    Vasiliev, N.N.; Sokolov, Yu.A.; Shatalov, G.E.

    1995-01-01

    One of the problems affiliated to ITER high heat flux elements development is a problem of interface of beryllium protection with heat sink routinely made of copper alloys. To get rid of this problem all beryllium elements could be used as heat receivers in places of enhanced thermal loads. In accordance with this objectives four beryllium test pieces of two types have been manufactured in open-quotes Institute of Berylliumclose quotes for succeeding thermomechanical testing. Two of them were manufactured in accordance with JET team design; they are round open-quotes hypervapotron typeclose quotes test pieces. Another two ones are rectangular test sections with a twisted tape installed inside of the circular channel. Preliminary stress-strain analysis have been performed for both type of the test pieces. Hypervapotrons have been shipped to JET where they were tested on JET test bed. Thermomechanical testing of pieces of the type of open-quotes swirl tape inside of tubeclose quotes have been performed on Kurchatov Institute test bed. Chosen beryllium grade properties, some details of manufacturing, results of preliminary stress-strain analysis and thermomechanical testing of the test pieces open-quotes swirl tape inside of tubeclose quotes type are given in this report

  20. HTTR demonstration test plan for industrial utilization of nuclear heat

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Ohashi, Hirofumi; Yan, Xing L.; Kubo, Shinji; Nishihara, Tetsuo; Tachibana, Yukio; Inagaki, Yoshiyuki

    2014-09-01

    Japan Atomic Energy Agency has been conducting research and development with a central focus on the utilization of High Temperature engineering Test Reactor (HTTR), the first High Temperature Gas-cooled Reactor (HTGR) in Japan, towards the realization of industrial use of nuclear heat. Several studies have made on the integration of the HTTR with thermochemical iodine-sulfur process and steam methane reforming hydrogen production plant (H 2 plant) as well as helium gas turbine power conversion system. In addition, safety standards for coupling a H 2 plant to a nuclear facility has been investigated. Based on the past design information, the present study identified test items to be validated in the HTTR demonstration test to accomplish a formulation of safety requirement and design consideration for coupling a H 2 plant to a nuclear facility as well as confirmation of overall performance of helium gas turbine system. In addition, plant concepts for the heat utilization system to be connected with the HTTR are investigated. (author)

  1. Concrete shielding for nuclear ship 'Mutsu'

    International Nuclear Information System (INIS)

    Nagase, Tetsuo; Saito, Tetsuo

    1983-01-01

    The repair works of the shielding for the nuclear ship ''Mutsu'' were completed in August, 1982. For the primary shielding, serpentine concrete was adopted as it contains a large quantity of water required for neutron shielding, and in the secondary shielding at the upper part of the reactor containment vessel, the original shielding was abolished, and the heavy concrete (high water content, high density concrete) which is effective for neutron and gamma-ray shielding was newly adopted. In this report, the design and construction using these shielding concrete are outlined. In September, 1974, Mutsu caused radiation leak during the test, and the cause was found to be the fast neutrons streaming through a gap between the reactor pressure vessel and the primary shielding. The repair works were carried out in the Sasebo Shipyard. The outline of the repair works of the shielding is described. The design condition for the shielding, the design standard for the radiation dose outside and inside the ship, the method of shielding analysis and the performance required for shielding concrete are reported. The selection of materials, the method of construction and mixing ratio, the evaluation of the soundness and properties of concrete, and the works of placing the shielding concrete are outlined. (Kako, I.)

  2. Safety evaluation for packaging (onsite) for the concrete-shielded RH TRU drum for the 327 Postirradiation Testing Laboratory

    International Nuclear Information System (INIS)

    Smith, R.J.

    1998-01-01

    This safety evaluation for packaging authorizes onsite transport of Type B quantities of radioactive material in the Concrete Shielded Remote-Handled Transuranic Waste (RH TRU) Drum per HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments. The drum will be used for transport of 327 Building legacy waste from the 300 Area to a solid waste storage facility on the Hanford Site

  3. Safety evaluation for packaging (onsite) for concrete-shielded RHTRU waste drum for the 327 postirradiation testing laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Adkins, H.E.

    1996-10-29

    This safety evaluation for packaging authorizes onsite transport of Type B quantities of radioactive material in the Concrete- Shielded Remote-Handled Transuranic Waste (RH TRU) Drum per WHC-CM-2-14, Hazardous Material Packaging and Shipping. The drum will be used for transport of 327 Building legacy waste from the 300 Area to the Transuranic Waste Storage and Assay Facility in the 200 West Area and on to a Solid Waste Storage Facility, also in the 200 Area.

  4. Simulated Lunar Testing of Metabolic Heat Regenerated Temperature Swing Adsorption

    Science.gov (United States)

    Padilla, Sebastian A.; Bower, Chad E.; Iacomini, Christie S.; Paul, Heather L.

    2012-01-01

    Metabolic heat regenerated Temperature Swing Adsorption (MTSA) technology is being developed for thermal and carbon dioxide (CO2) control for a Portable Life Support System (PLSS), as well as water recycling. An Engineering Development Unit (EDU) of the MTSA Subassembly (MTSAS) was designed and assembled for optimized Martian operations, but also meets system requirements for lunar operations. For lunar operations the MTSA sorption cycle is driven via a vacuum swing between suit ventilation loop pressure and lunar vacuum. The focus of this effort was testing in a simulated lunar environment. This environment was simulated in Paragon's EHF vacuum chamber. The objective of the testing was to evaluate the full cycle performance of the MTSA Subassembly EDU, and to assess CO2 loading and pressure drop of the wash coated aluminum reticulated foam sorbent bed. Lunar environment testing proved out the feasibility of pure vacuum swing operation, making MTSA a technology that can be tested and used on the Moon prior to going to Mars. Testing demonstrated better than expected CO2 Nomenclature loading on the sorbent and nearly replicates the equilibrium data from the sorbent manufacturer. This exceeded any of the previous sorbent loading tests performed by Paragon. Subsequently, the increased performance of the sorbent bed design indicates future designs will require less mass and volume than the current EDU rendering MTSA as very competitive for Martian PLSS applications.

  5. Ambient temperature testing of the G-tunnel heated block

    International Nuclear Information System (INIS)

    Zimmerman, R.M.; Board, M.P.; Hardin, E.L.; Voegele, M.D.

    1984-01-01

    The G-Tunnel heated block experiment is being conducted on the Nevada Test Site (NTS) as part of the Nevada Nuclear Waste Storage Investigations project (NNWSI). The purpose of the ambient temperature testing phase is to evaluate rock-mass mechanical properties of a block (≅8 m/sup 3/) under biaxial stress changes up to 7.5 MPa above an initialization in situ value of 3.1 MPa. Results indicate that the modulus of deformation ranges from 9.7 to 17.0 GPa and Poisson's ratio ranges from 0.21 to 0.33. In general, the higher values of the modulus and Poisson's ratio were influenced by fracture propagations parallel to the compressive stress field. Other measurements indicated that cross-hole compression (p) wave velocities and single fracture permeability values were relatively insensitive to stress changes above the in situ value

  6. Vibrations in water-gas heat exchangers. Design and tests

    International Nuclear Information System (INIS)

    Alexandre, M.; Allard, G.; Vangedhen, A.

    1981-01-01

    It is shown on an example how to make a complete list of the possible vibrations and how to use the data of tests and technical literature to predict damaging vibrations. The water-heavy gas tubular heat-exchanger in case is briefly described. The sources of mechanical excitations are a compressor and earthquake loadings. The various eigenmodes are described and it is shown that no resonance is possible with the compressor and that the effect of the earthquake is negligible. The excitation of the tubes by the gas flow is examined by means of Connors stability criterion; and there is no resonance with the Benard-von Karman vortices. The magnification of this latter excitation by acoustical waves is not to be feared. Satisfactory tests have been carried successively on tubes, on the casing, on the casing plus part of the tubes, on a complete prototype in workshop and in operation on site [fr

  7. Thermal design of top shield

    International Nuclear Information System (INIS)

    Raghupathy, S.; Velusamy, K.; Parthasarathy, U.; Ghosh, D.; Selvaraj, P.; Chellapandi, P.; Chetal, S.C.

    2005-01-01

    insulation to achieve a temperature of 423 K at the MV-RS junction and an average heat flux of 200 W/m 2 to the reactor vault. The heat transfer through cover gas has been verified experimentally in a test rig. The heat transfer coefficient for jet cooling has also been verified experimentally. Experimental verification of the flow distribution within TS is planned. In this paper, the detailed thermal design and analysis carried out for top shield covering all the aspects indicated above is discussed. (authors)

  8. The Effects of Shielded Metal Arc Welding (Smaw) Welding On The Mechanical Characteristics With Heating Treatment inn S45c Steel

    Science.gov (United States)

    Munawar; Abbas, Hammada; Yusran Aminy, Ahmad

    2018-02-01

    Steel material has been used mainly for making tooling, automotive components, other household needs, power generators to frame buildings and bridges. This study aimed (1) to analyze the mechanical Characteristics of S45C steel with and without heating treatments, and (2) to analyze the temperature of heating treatment which could result in the maximal strength of S45C steel after the welding process. The research was conducted in the laboratory of mechanical engineering study program, Departement of mechanical Engineering, Christian university of indonesia paulus, makassar. The method used materials, instruments, and the dimensions determination of specimen based on the proposed testing standard, Next, was to determine the mechanical caracteristics of the S45C steel wich had been welded and heated.The tensile specimens, the hardness specimen, the impact specimen, and microstructures of which,each of the 3 specimens was the specimens was the specimen without treatment, the spesimen with the welding wthout heating and the specimen of 150°C, 250° C, 300° C. The research results indicated that the treatment process of 150°C, 250°C and 300°C produced the changes of mechanic charateristics with the tensile strength of 42 kgf/mm2 when the temperature had reached 300°C, but at the temperature 300°C, the its toughness would decrease to Hi = 0.836 j/m2 and its hardness would increase to 40.83 at the temperature of 300°C. The value of the maximum strengs was reached at the heating temperature of 300°C for the tensile strength and the hardness, while at the temperature of 300°C its impact value would decrease.

  9. Testing of Solar Heated Domestic Hot Water System for Solahart Scandinavia ApS

    DEFF Research Database (Denmark)

    Andersen, Elsa

    1997-01-01

    The solar heating system marketed by Solahart Scandinavia ApS was tested in the Institutes test facility for SDHWsystems. The test results are described in the report.......The solar heating system marketed by Solahart Scandinavia ApS was tested in the Institutes test facility for SDHWsystems. The test results are described in the report....

  10. 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

  11. HELCZA-High heat flux test facility for testing ITER EU first wall components.

    Czech Academy of Sciences Publication Activity Database

    Prokůpek, J.; Samec, K.; Jílek, R.; Gavila, P.; Neufuss, S.; Entler, Slavomír

    2017-01-01

    Roč. 124, November (2017), s. 187-190 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] Institutional support: RVO:61389021 Keywords : HELCZA * High heat flux * Electron beam testing * Test facility * Plasma facing components * First wall * Divertora Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 www.sciencedirect.com/science/article/pii/S0920379617302818

  12. LMFBR post accident heat removal testing needs and conceptual design of a test facility

    International Nuclear Information System (INIS)

    Kleefeldt, K.; Kuechle, M.; Royl, P.; Werle, H.; Boenisch, G.; Heinzel, V.; Mueller, R.A.; Schramm, K.; Smidt, D.

    1977-03-01

    A study has been carried out in which the needs and requirements for a test facility were derived, enabling detailed investigation of key phenomena anticipated during the post accident heat removal (PAHR) phase as a consequence of a postulated LMFBR whole core accident. Part I of the study concentrates on demonstrating the PAHR phenomena and related testing needs. Three types of experiments were identified which require in-pile testing, ranging from 10 to 70 cm test bed diameter and correspondingly, 30 to 5 W/g minimum power density in the test fuel. In part II a conceptual design for a test facility is presented, emphasizing the capability for accomodating large test beds. This is achieved by a below-reactor-vessel testing device, neutronically coupled to a 100 MWt sodium cooled fast reactor. (orig.) [de

  13. 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

  14. General-Purpose Heat Source Development: Safety Test Program. Postimpact evaluation, Design Iteration Test 3

    International Nuclear Information System (INIS)

    Schonfeld, F.W.; George, T.G.

    1984-07-01

    The General-Purpose Heat Source(GPHS) provides power for space missions by transmitting the heat of 238 PuO 2 decay to thermoelectric elements. Because of the inevitable return of certain aborted missions, the heat source must be designed and constructed to survive both re-entry and Earth impact. The Design Iteration Test (DIT) series is part of an ongoing test program. In the third test (DIT-3), a full GPHS module was impacted at 58 m/s and 930 0 C. The module impacted the target at an angle of 30 0 to the pole of the large faces. The four capsules used in DIT-3 survived impact with minimal deformation; no internal cracks other than in the regions indicated by Savannah River Plant (SRP) preimpact nondestructive testing were observed in any of the capsules. The 30 0 impact orientation used in DIT-3 was considerably less severe than the flat-on impact utilized in DIT-1 and DIT-2. The four capsules used in DIT-1 survived, while two of the capsules used in DIT-2 breached; a small quantity (approx. = 50 μg) of 238 PuO 2 was released from the capsules breached in the DIT-2 impact. All of the capsules used in DIT-1 and DIT-2 were severely deformed and contained large internal cracks. Postimpact analyses of the DIT-3 test components are described, with emphasis on weld structure and the behavior of defects identified by SRP nondestructive testing

  15. A study of gamma shielding

    International Nuclear Information System (INIS)

    Roogtanakait, N.

    1981-01-01

    Gamma rays have high penetration power and its attenuation depends upon the thickness and the attenuation coefficient of the shield, so it is necessary to use the high density shield to attenuate the gamma rays. Heavy concrete is considered to be used for high radiation laboratory and the testing of the shielding ability and compressibility of various types of heavy concrete composed of baryte, hematite, ilmenite and galena is carried out. The results of this study show that baryte-ilmenite concrete is the most suitable for high radiation laboratory in Thailand

  16. The Role of Heat Tolerance Testing in Recovery and Return to Duty

    Science.gov (United States)

    2008-10-01

    CV diseases Hyperthyroidism Pheochromocytoma Infectious diseases Diabetes mellitus Psychiatric illness Parkinsonism Congenital abnormalities: CF...environments. To assess the heat tolerance status of prior heat stroke patient. Heat tolerance test (HTT) “HTT was effective in evaluating the heat tolerance

  17. An assessment of RELAP5 MOD3.1.1 condensation heat transfer modeling with GIRAFFE heat transfer tests

    International Nuclear Information System (INIS)

    Boyer, B.D.; Parlatan, Y.; Slovik, G.C.

    1995-01-01

    RELAP5 MOD3.1.1 is being used to simulate Loss of Coolant Accidents (LOCA) for the Simplified Boiling Water Reactor (SBWR) being proposed by General Electric (GE). One of the major components associated with the SBWR is the Passive Containment Cooling System (PCCS) which provides the long-term heat sink to reject decay heat. The RELAP5 MOD3.1.1 code is being assessed for its ability to represent accurately the PCCS. Data from the Phase 1, Step 1 Heat Transfer Tests performed at Toshiba's Gravity-Driven Integral Full-Height Test for Passive Heat Removal (GIRAFFE) facility will be used for assessing the ability of RELAP5 to model condensation in the presence of noncondensables. The RELAP5 MOD3.1.1 condensation model uses the University of California at Berkeley (UCB) correlation developed by Vierow and Schrock. The RELAP5 code uses this heat transfer coefficient with the gas velocity effect multiplier being limited to 2. This heat transfer option was used to analyze the condensation heat transfer in the GIRAFFE PCCS heat exchanger tubes in the Phase 1, Step 1 Heat Transfer Tests which were at a pressure of 3 bar and had a range of nitrogen partial pressure fractions from 0.0 to 0.10. The results of a set of RELAP5 calculations at these conditions were compared with the GIRAFFE data. The effects of PCCS cell noding on the heat transfer process were also studied. The UCB correlation, as implemented in RELAP5, predicted the heat transfer to ±5% of the data with a three--node model. The three-node model has a large cell in the entrance region which smeared out the entrance effects on the heat transfer, which tend to overpredict the condensation. Hence, the UCB correlation predicts condensation heat transfer correlation implemented in the code must be removed to allow for accurate calculations with smaller cell sizes

  18. An assessment of RELAP5 MOD3.1.1 condensation heat transfer modeling with GIRAFFE heat transfer tests

    Energy Technology Data Exchange (ETDEWEB)

    Boyer, B.D.; Parlatan, Y.; Slovik, G.C. [and others

    1995-09-01

    RELAP5 MOD3.1.1 is being used to simulate Loss of Coolant Accidents (LOCA) for the Simplified Boiling Water Reactor (SBWR) being proposed by General Electric (GE). One of the major components associated with the SBWR is the Passive Containment Cooling System (PCCS) which provides the long-term heat sink to reject decay heat. The RELAP5 MOD3.1.1 code is being assessed for its ability to represent accurately the PCCS. Data from the Phase 1, Step 1 Heat Transfer Tests performed at Toshiba`s Gravity-Driven Integral Full-Height Test for Passive Heat Removal (GIRAFFE) facility will be used for assessing the ability of RELAP5 to model condensation in the presence of noncondensables. The RELAP5 MOD3.1.1 condensation model uses the University of California at Berkeley (UCB) correlation developed by Vierow and Schrock. The RELAP5 code uses this heat transfer coefficient with the gas velocity effect multiplier being limited to 2. This heat transfer option was used to analyze the condensation heat transfer in the GIRAFFE PCCS heat exchanger tubes in the Phase 1, Step 1 Heat Transfer Tests which were at a pressure of 3 bar and had a range of nitrogen partial pressure fractions from 0.0 to 0.10. The results of a set of RELAP5 calculations at these conditions were compared with the GIRAFFE data. The effects of PCCS cell noding on the heat transfer process were also studied. The UCB correlation, as implemented in RELAP5, predicted the heat transfer to {plus_minus}5% of the data with a three--node model. The three-node model has a large cell in the entrance region which smeared out the entrance effects on the heat transfer, which tend to overpredict the condensation. Hence, the UCB correlation predicts condensation heat transfer correlation implemented in the code must be removed to allow for accurate calculations with smaller cell sizes.

  19. Performance test of miniature heat exchangers with microchannels

    International Nuclear Information System (INIS)

    Hong, Yong Ju; Koh, Deuk Yong

    2005-01-01

    Etched microchannel heat exchanger, a subfield within MEMS, has high heat flux capability. This capability makes microchannels well-suited for a wide variety of application of cooling and chemical reaction. In this study, counter flow type miniature heat exchangers, which have flat metal plates with chemically etched microchannels, were manufactured by brazing method. Four type of the heat exchangers, which have straight microchannels, wavy shape microchannels, pin-fin channels and serpentine shape microchannels, were investigated to compare their thermal and hydraulic performance. Gas to gas heat exchange experiments were performed to measure the pressure drop and effectiveness of the heat exchangers at given gas flow rates and temperature difference

  20. Problems of the power plant shield optimization

    International Nuclear Information System (INIS)

    Abagyan, A.A.; Dubinin, A.A.; Zhuravlev, V.I.; Kurachenko, Yu.A.; Petrov, Eh.E.

    1981-01-01

    General approaches to the solution of problems on the nuclear power plant radiation shield optimization are considered. The requirements to the shield parameters are formulated in a form of restrictions on a number of functionals, determined by the solution of γ quantum and neutron transport equations or dimensional and weight characteristics of shield components. Functional determined by weight-dimensional parameters (shield cost, mass and thickness) and functionals, determined by radiation fields (equivalent dose rate, produced by neutrons and γ quanta, activation functional, radiation functional, heat flux, integral heat flux in a particular part of the shield volume, total energy flux through a particular shield surface are considered. The following methods of numerical solution of simplified optimization problems are discussed: semiempirical methods using radiation transport physical leaks, numerical solution of approximate transport equations, numerical solution of transport equations for the simplest configurations making possible to decrease essentially a number of variables in the problem. The conclusion is drawn that the attained level of investigations on the problem of nuclear power plant shield optimization gives the possibility to pass on at present to the solution of problems with a more detailed account of the real shield operating conditions (shield temperature field account, its strength and other characteristics) [ru

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

    International Nuclear Information System (INIS)

    Schlosser, J.; Boscary, J.

    1994-01-01

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

  2. Climate Adaptivity and Field Test of the Space Heating Used Air-Source Transcritical CO2 Heat Pump

    Science.gov (United States)

    Song, Yulong; Ye, Zuliang; Cao, Feng

    2017-08-01

    In this study, an innovation of air-sourced transcritical CO2 heat pump which was employed in the space heating application was presented and discussed in order to solve the problem that the heating performances of the transcritical CO2 heat pump water heater deteriorated sharply with the augment in water feed temperature. An R134a cycle was adopted as a subcooling device in the proposed system. The prototype of the presented system was installed and supplied hot water for three places in northern China in winter. The field test results showed that the acceptable return water temperature can be increased up to 55°C, while the supply water temperature was raised rapidly by the presented prototype to up to 70°C directly, which was obviously appropriate to the various conditions of heating radiator in space heating application. Additionally, though the heating capacity and power dissipation decreased with the decline in ambient temperature or the augment in water temperature, the presented heat pump system performed efficiently whatever the climate and water feed temperature were. The real time COP of the presented system was generally more than 1.8 in the whole heating season, while the seasonal performance coefficient (SPC) was also appreciable, which signified that the economic efficiency of the presented system was more excellent than other space heating approaches such as fuel, gas, coal or electric boiler. As a result, the novel system will be a promising project to solve the energy issues in future space heating application.

  3. Performance of overlapped shield tunneling through an integrated physical model tests, numerical simulations and real-time field monitoring

    Directory of Open Access Journals (Sweden)

    Junlong Yang

    2017-03-01

    Full Text Available In this work, deformations and internal forces of an existing tunnel subjected to a closely overlapped shield tunneling are monitored and analyzed using a series of physical model experiments and numerical simulations. Effects of different excavation sequences and speeds are explicitly considered in the analysis. The results of the physical model experiments show that the bottom-up tunneling procedure is better than the top-down tunneling procedure. The incurred deformations and internal forces of the existing tunnel increase with the excavation speed and the range of influence areas also increase accordingly. For construction process control, real-time monitoring of the power tunnel is used. The monitoring processes feature full automation, adjustable frequency, real-time monitor and dynamic feedback, which are used to guide the construction to achieve micro-disturbance control. In accordance with the situation of crossing construction, a numerical study on the performance of power tunnel is carried out. Construction control measures are given for the undercrossing construction, which helps to accomplish the desired result and meet protection requirements of the existing tunnel structure. Finally, monitoring data and numerical results are compared, and the displacement and joint fracture change models in the power tunnel subject to the overlapped shield tunnel construction are analyzed. Keywords: Overlapped tunnel, Automatic monitoring, Micro-disturbance control

  4. Validation of intermediate heat and decay heat exchanger model in MARS-LMR with STELLA-1 and JOYO tests

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chiwoong; Ha, Kwiseok; Hong, Jonggan; Yeom, Sujin; Eoh, Jaehyuk [Sodium-cooled Fast Reactor Design Division, Korea Atomic Energy Research Institute (KAERI), 989-111, Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Jeong, Hae-yong, E-mail: hyjeong@sejong.ac.kr [Department of Nuclear Engineering, Sejong University, 209 Neungdong-ro, Gwangjin-gu, Seoul 143-747 (Korea, Republic of)

    2016-11-15

    Highlights: • The capability of the MARS-LMR for heat transfer through IHX and DHX is evaluated. • Prediction of heat transfer through IHXs and DHXs is essential in the SFR analysis. • Data obtained from the STELLA-1 and the JOYO test are analyzed with the MARS-LMR. • MARS-LMR adopts the Aoki’s correlation for tube side and Graber-Rieger’s for shell. • The performance of the basic models and other available correlations is evaluated. • The current models in MARS-LMR show best prediction for JOYO and STELLA-1 data. - Abstract: The MARS-LMR code has been developed by the Korea Atomic Energy Research Institute (KAERI) to analyze transients in a pool-type sodium-cooled fast reactor (SFR). Currently, KAERI is developing a prototype Gen-IV SFR (PGSFR) with metallic fuel. The decay heat exchangers (DHXs) and the intermediate heat exchangers (IHXs) were designed as a sodium-sodium counter-flow tube bundle type for decay heat removal system (DHRS) and intermediate heat transport system (IHTS), respectively. The IHX and DHX are important components for a heat removal function under normal and accident conditions, respectively. Therefore, sodium heat transfer models for the DHX and IHX heat exchangers were added in MARS-LMR. In order to validate the newly added heat transfer model, experimental data were obtained from the JOYO and STELLA-1 facilities were analyzed. JOYO has two different types of IHXs: type-A (co-axial circular arrangement) and type-B (triangular arrangement). For the code validation, 38 and 39 data points for type A and type B were selected, respectively. A DHX performance test was conducted in STELLA-1, which is the test facility for heat exchangers and primary pump in the PGSFR. The DHX test in STELLA-1 provided eight data points for a code validation. Ten nodes are used in the heat transfer region is used, based on the verification test for the heat transfer models. RMS errors for JOYO IHX type A and type B of 19.1% and 4.3% are obtained

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

    International Nuclear Information System (INIS)

    French, R.T.

    1975-08-01

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

  6. Water and Regolith Shielding for Surface Reactor Missions

    Science.gov (United States)

    Poston, David I.; Ade, Brian J.; Sadasivan, Pratap; Leichliter, Katrina J.; Dixon, David D.

    2006-01-01

    This paper investigates potential shielding options for surface power fission reactors. The majority of work is focused on a lunar shield that uses a combination of water in stainless-steel cans and lunar regolith. The major advantage of a water-based shield is that development, testing, and deployment should be relatively inexpensive. This shielding approach is used for three surface reactor concepts: (1) a moderated spectrum, NaK cooled, Hastalloy/UZrH reactor, (2) a fast-spectrum, NaK-cooled, SS/UO2 reactor, and (3) a fast-spectrum, K-heat-pipe-cooled, SS/UO2 reactor. For this study, each of these reactors is coupled to a 25-kWt Stirling power system, designed for 5 year life. The shields are designed to limit the dose both to the Stirling alternators and potential astronauts on the surface. The general configuration used is to bury the reactor, but several other options exist as well. Dose calculations are presented as a function of distance from reactor, depth of buried hole, water boron concentration (if any), and regolith repacked density.

  7. Water and Regolith Shielding for Surface Reactor Missions

    International Nuclear Information System (INIS)

    Poston, David I.; Sadasivan, Pratap; Dixon, David D.; Ade, Brian J.; Leichliter, Katrina J.

    2006-01-01

    This paper investigates potential shielding options for surface power fission reactors. The majority of work is focused on a lunar shield that uses a combination of water in stainless-steel cans and lunar regolith. The major advantage of a water-based shield is that development, testing, and deployment should be relatively inexpensive. This shielding approach is used for three surface reactor concepts: (1) a moderated spectrum, NaK cooled, Hastalloy/UZrH reactor, (2) a fast-spectrum, NaK-cooled, SS/UO2 reactor, and (3) a fast-spectrum, K-heat-pipe-cooled, SS/UO2 reactor. For this study, each of these reactors is coupled to a 25-kWt Stirling power system, designed for 5 year life. The shields are designed to limit the dose both to the Stirling alternators and potential astronauts on the surface. The general configuration used is to bury the reactor, but several other options exist as well. Dose calculations are presented as a function of distance from reactor, depth of buried hole, water boron concentration (if any), and regolith repacked density

  8. Jet Noise Shielding Provided by a Hybrid Wing Body Aircraft

    Science.gov (United States)

    Doty, Michael J.; Brooks, Thomas F.; Burley, Casey L.; Bahr, Christopher J.; Pope, Dennis S.

    2014-01-01

    One approach toward achieving NASA's aggressive N+2 noise goal of 42 EPNdB cumulative margin below Stage 4 is through the use of novel vehicle configurations like the Hybrid Wing Body (HWB). Jet noise measurements from an HWB acoustic test in NASA Langley's 14- by 22-Foot Subsonic Tunnel are described. Two dual-stream, heated Compact Jet Engine Simulator (CJES) units are mounted underneath the inverted HWB model on a traversable support to permit measurement of varying levels of shielding provided by the fuselage. Both an axisymmetric and low noise chevron nozzle set are investigated in the context of shielding. The unshielded chevron nozzle set shows 1 to 2 dB of source noise reduction (relative to the unshielded axisymmetric nozzle set) with some penalties at higher frequencies. Shielding of the axisymmetric nozzles shows up to 6.5 dB of reduction at high frequency. The combination of shielding and low noise chevrons shows benefits beyond the expected additive benefits of the two, up to 10 dB, due to the effective migration of the jet source peak noise location upstream for increased shielding effectiveness. Jet noise source maps from phased array results processed with the Deconvolution Approach for the Mapping of Acoustic Sources (DAMAS) algorithm reinforce these observations.

  9. Shielding modification design of the N.S. Mutsu

    International Nuclear Information System (INIS)

    Yamaji, A.; Miyakoshi, J.; Kageyama, T.; Futamura, Y.

    1983-01-01

    Shielding modification design of the N.S. Mutsu was performed for reducing the radiation doses outside the primary and the secondary shields by providing shields for neutrons streaming through the air gap between the pressure vessel and the primary shield. This was accomplished by replacing parts of the shields and adding new shields in the upper and lower sections of both primary and secondary shields, and also replacing the thermal insulator in the gap. The shielding design calculations were made using one- and two-dimensional discrete ordinates codes and also a point kernel code. Special attention was paid to the calculations of, (1) the neutrons streaming through the gap between the pressure vessel and the primary shield, (2) the radiations transmitted through the radial shield of the core in the primary shield, (3) the radiations transmitted through the upper and lower sections of the secondary shield, and (4) the dose rate equivalent in the accommodation area. Their calculational accuracies were estimated by analyzing various experiments. To support the modification, a variety of experiments and tests were carried out, which were material tests, cooling test of the primary shield, mechanical strength test of the double bottom, trial fabrication tests of new shields, performance degradation test of heavy concrete and duct streaming experiment in the secondary shield. (author)

  10. 77 FR 8178 - Test Procedures for Central Air Conditioners and Heat Pumps: Public Meeting

    Science.gov (United States)

    2012-02-14

    .... EERE-2010-BT-TP-0038] Test Procedures for Central Air Conditioners and Heat Pumps: Public Meeting... methodologies and gather comments on testing residential central air conditioners and heat pumps designed to use... residential central air conditioners and heat pumps that are single phase with rated cooling capacities less...

  11. Modular Micromachined Si Heat Removal (MOMS Heat Removal): Electronic Integration and System Test

    National Research Council Canada - National Science Library

    Brown, Elliott

    2003-01-01

    ...: (1) insulated-gated bipolar transistors (IGBTs), and (2) laterally-diffused (LD) MOSFETs. Heat pipes were found to provide little or no advantage over conventional copper-based heat spreaders in both device applications...

  12. Experiment data report for semiscale Mod-1, test S-02-7. Blowndown heat transfer test

    International Nuclear Information System (INIS)

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

    1975-11-01

    Recorded test data are presented for Test S-02-7 of the Semiscale Mod-1 blowdown heat transfer test series conducted to investigate the thermal and hydraulic phenomena accompanying an hypothesized loss-of-coolant accident (LOCA) 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-02-7 was conducted from an initial cold leg fluid temperature of 543 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 transient with full design core power (1.6 MW). An electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core with power set to provide a flat radial power profile. System flow was set to achieve the full design core temperature differential of 66 0 F. Blowdown to the pressure suppression system was accomplished without simulated emergency core cooling injection or pressure suppression system coolant spray. The uninterpreted data from Test S-02-7 are presented 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. Also included as an appendix are selected data from a test identified as Test S-02-7C. This test was an initial attempt at Test S-02-7 in which an inadvertent power trip occurred at 2.3 seconds after rupture. Selected data comparisons of the results from Test S-02-7 and S-02-7C are presented to indicate the repeatability of system behavior

  13. Tracer Testing for Estimating Heat Transfer Area in Fractured Reservoirs

    Energy Technology Data Exchange (ETDEWEB)

    Pruess, Karsten; van Heel, Ton; Shan, Chao

    2004-05-12

    A key parameter governing the performance and life-time of a Hot Fractured Rock (HFR) reservoir is the effective heat transfer area between the fracture network and the matrix rock. We report on numerical modeling studies into the feasibility of using tracer tests for estimating heat transfer area. More specifically, we discuss simulation results of a new HFR characterization method which uses surface-sorbing tracers for which the adsorbed tracer mass is proportional to the fracture surface area per unit volume. Sorption in the rock matrix is treated with the conventional formulation in which tracer adsorption is volume-based. A slug of solute tracer migrating along a fracture is subject to diffusion across the fracture walls into the adjacent rock matrix. Such diffusion removes some of the tracer from the fluid in the fractures, reducing and retarding the peak in the breakthrough curve (BTC) of the tracer. After the slug has passed the concentration gradient reverses, causing back-diffusion from the rock matrix into the fracture, and giving rise to a long tail in the BTC of the solute. These effects become stronger for larger fracture-matrix interface area, potentially providing a means for estimating this area. Previous field tests and modeling studies have demonstrated characteristic tailing in BTCs for volatile tracers in vapor-dominated reservoirs. Simulated BTCs for solute tracers in single-phase liquid systems show much weaker tails, as would be expected because diffusivities are much smaller in the aqueous than in the gas phase, by a factor of order 1000. A much stronger signal of fracture-matrix interaction can be obtained when sorbing tracers are used. We have performed simulation studies of surface-sorbing tracers by implementing a model in which the adsorbed tracer mass is assumed proportional to the fracture-matrix surface area per unit volume. The results show that sorbing tracers generate stronger tails in BTCs, corresponding to an effective

  14. Heat Transfer in Metal Foam Heat Exchangers at High Temperature

    Science.gov (United States)

    Hafeez, Pakeeza

    Heat transfer though open-cell metal foam is experimentally studied for heat exchanger and heat shield applications at high temperatures (˜750°C). Nickel foam sheets with pore densities of 10 and 40 pores per linear inch (PPI), have been used to make the heat exchangers and heat shields by using thermal spray coating to deposit an Inconel skin on a foam core. Heat transfer measurements were performed on a test rig capable of generating hot gas up to 1000°C. The heat exchangers were tested by exposing their outer surface to combustion gases at a temperature of 550°C and 750°C while being cooled by air flowing through them at room temperature at velocities up to 5 m/s. The temperature rise of the air, the surface temperature of the heat exchangers and the air temperature inside the heat exchanger were measured. The volumetric heat transfer coefficient and Nusselt number were calculated for different velocities. The heat transfer performance of the 40PPI sample brazed with the foil is found to be the most efficient. Pressure drop measurements were also performed for 10 and 40PPI metal foam. Thermographic measurements were done on 40PPI foam heat exchangers using a high temperature infrared camera. A high power electric heater was used to produce hot air at 300°C that passed over the foam heat exchanger while the cooling air was blown through it. Heat shields were made by depositing porous skins on metal foam and it was observed that a small amount of coolant leaking through the pores notably reduces the heat transfer from the hot gases. An analytical model was developed based assuming local thermal non-equilibrium that accounts for the temperature difference between solid and fluid phase. The experimental results are found to be in good agreement with the predicted values of the model.

  15. About the Scythian Shields

    Directory of Open Access Journals (Sweden)

    About the Scythian Shields

    2017-10-01

    Full Text Available Shields played major role in the armament system of the Scythians. Made from organic materials, they are poorly traced on the materials of archaeological excavations. Besides, scaly surface of shields was often perceived in practice as the remnants of the scaly armor. E. V. Chernenko was able to discern the difference between shields’ scaly plates and armor scales. The top edge of the scales was bent inwards, and shield plates had a wire fixation. These observations let significantly increase the number of shields, found in the burial complexes of the Scythians. The comparison of archaeological materials and the images of Scythian warriors allow distinguishing the main forms of Scythian shields. All shields are divided into fencing shields and cover shields. The fencing shields include round wooden shields, reinforced with bronze sheet, and round moon-shaped shields with a notch at the top, with a metal scaly surface. They came to the Scythians under the Greek influence and are known in the monuments of the 4th century BC. Oval shields with scaly surface (back cover shields were used by the Scythian cavalry. They protected the rider in case of frontal attack, and moved back in case of maneuver or closein fighting. Scythian battle tactics were based on rapid approaching the enemy and throwing spears and further rapid withdrawal. Spears stuck in the shields of enemies, forcing them to drop the shields, uncover, and in this stage of the battle the archers attacked the disorganized ranks of the enemy. That was followed by the stage of close fight. Oval form of a wooden shield with leather covering was used by the Scythian infantry and spearmen. Rectangular shields, including wooden shields and the shields pleached from rods, represented a special category. The top of such shield was made of wood, and a pleached pad on leather basis was attached to it. This shield could be a reliable protection from arrows, but it could not protect against javelins

  16. Performance Test of Korea Heat Load Test Facility (KoHLT-EB) for the Plasma Facing Components of Fusion Reactor

    International Nuclear Information System (INIS)

    Kim, Suk-Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won; Cho, Seungyon

    2014-01-01

    The main components of the plasma facing components (PFCs) in the tokamak are the blanket first wall and divertor, which include the armour materials, the heat sink with the cooling mechanism, and the diagnostics devices for the temperature measurement. The Korea Heat Load Test facility by using electron beam (KoHLT-EB) has been operating for the plasma facing components to develop fusion engineering. This electron beam facility was constructed using a 300 kW electron gun and a cylindrical vacuum chamber. Performance tests were carried out for the calorimetric calibrations with Cu dummy mockup and for the heat load test of large Cu module. For the simulation of the heat load test of each mockup, the preliminary thermal-hydraulic analyses with ANSYS-CFX were performed. For the development of the plasma facing components in the fusion reactors, test mockups were fabricated and tested in the high heat flux test facility. To perform a beam profile test, an assessment of the possibility of electron beam Gaussian power density profile and the results of the absorbed power for that profile before the test starts are needed. To assess the possibility of a Gaussian profile, for the qualification test of the Gaussian heat load profile, a calorimeter mockup and large Cu module were manufactured to simulate real heat. For this high-heat flux test, the Korean high-heat flux test facility using an electron beam system was constructed. In this facility, a cyclic heat flux test will be performed to measure the surface heat flux, surface temperature profile, and cooling capacity

  17. Performance Test of Korea Heat Load Test Facility (KoHLT-EB) for the Plasma Facing Components of Fusion Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Suk-Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The main components of the plasma facing components (PFCs) in the tokamak are the blanket first wall and divertor, which include the armour materials, the heat sink with the cooling mechanism, and the diagnostics devices for the temperature measurement. The Korea Heat Load Test facility by using electron beam (KoHLT-EB) has been operating for the plasma facing components to develop fusion engineering. This electron beam facility was constructed using a 300 kW electron gun and a cylindrical vacuum chamber. Performance tests were carried out for the calorimetric calibrations with Cu dummy mockup and for the heat load test of large Cu module. For the simulation of the heat load test of each mockup, the preliminary thermal-hydraulic analyses with ANSYS-CFX were performed. For the development of the plasma facing components in the fusion reactors, test mockups were fabricated and tested in the high heat flux test facility. To perform a beam profile test, an assessment of the possibility of electron beam Gaussian power density profile and the results of the absorbed power for that profile before the test starts are needed. To assess the possibility of a Gaussian profile, for the qualification test of the Gaussian heat load profile, a calorimeter mockup and large Cu module were manufactured to simulate real heat. For this high-heat flux test, the Korean high-heat flux test facility using an electron beam system was constructed. In this facility, a cyclic heat flux test will be performed to measure the surface heat flux, surface temperature profile, and cooling capacity.

  18. An exploration of unsaturated zone during in-situ heating test in sedimentary soft rocks

    International Nuclear Information System (INIS)

    Kubota, Kenji; Suzuki, Koichi; Ikenoya, Takafumi; Takakura, Nozomu; Tani, Kazuo

    2011-01-01

    In-situ heating test has been conducted to evaluate the influence of high temperature in an underground facility at a depth of 50 m. Resistivity monitoring is thought to be effective to map the extent of the high temperature and unsaturated zone. So we have conducted resistivity tomography during the heating test. As a result, the resistivity of the rock mass around the heating well was decreased and this area was gradually expanded from the heated area during the heating. This suggests that high temperature zone is detected by resistivity tomography. The results also suggested that resistivity was increased by unsaturation of rock mass around the heating well. (author)

  19. Thermal response test data of five quadratic cross section precast pile heat exchangers.

    Science.gov (United States)

    Alberdi-Pagola, Maria

    2018-06-01

    This data article comprises records from five Thermal Response Tests (TRT) of quadratic cross section pile heat exchangers. Pile heat exchangers, typically referred to as energy piles, consist of traditional foundation piles with embedded heat exchanger pipes. The data presented in this article are related to the research article entitled "Comparing heat flow models for interpretation of precast quadratic pile heat exchanger thermal response tests" (Alberdi-Pagola et al., 2018) [1]. The TRT data consists of measured inlet and outlet temperatures, fluid flow and injected heat rate recorded every 10 min. The field dataset is made available to enable model verification studies.

  20. Testing for cross-subsidisation in the combined heat and power generation sector

    DEFF Research Database (Denmark)

    Amundsen, Eirik S; Andersen, Per; Jensen, Frank

    2011-01-01

    In this paper we examine cross-subsidisation among combined heat and power producers in Denmark.Information on stand-alone costs for heat generation allows us to empirically compare the Faulhaber tests,tests with an upper bound on stand-alone costs (the Palmer tests) and the fully distributed cos...... test (FDC). All tests indicate a substantial amount of cross-subsidisation from heat generation to power generation. It is shown that the FDC test is closer to that of the Faulhaber tests in its results than the Palmer tests. Thus as the Faulhaber tests are considered in the literature...

  1. Testing, development and demonstration of large scale solar district heating systems

    DEFF Research Database (Denmark)

    Furbo, Simon; Fan, Jianhua; Perers, Bengt

    2015-01-01

    In 2013-2014 the project “Testing, development and demonstration of large scale solar district heating systems” was carried out within the Sino-Danish Renewable Energy Development Programme, the so called RED programme jointly developed by the Chinese and Danish governments. In the project Danish...... know how on solar heating plants and solar heating test technology have been transferred from Denmark to China, large solar heating systems have been promoted in China, test capabilities on solar collectors and large scale solar heating systems have been improved in China and Danish-Chinese cooperation...

  2. Shielding benchmark problems

    International Nuclear Information System (INIS)

    Tanaka, Shun-ichi; Sasamoto, Nobuo; Oka, Yoshiaki; Kawai, Masayoshi; Nakazawa, Masaharu.

    1978-09-01

    Shielding benchmark problems were prepared by the Working Group of Assessment of Shielding Experiments in the Research Comittee on Shielding Design of the Atomic Energy Society of Japan, and compiled by the Shielding Laboratory of Japan Atomic Energy Research Institute. Twenty-one kinds of shielding benchmark problems are presented for evaluating the calculational algorithm and the accuracy of computer codes based on the discrete ordinates method and the Monte Carlo method and for evaluating the nuclear data used in the codes. (author)

  3. Shielding Benchmark Computational Analysis

    International Nuclear Information System (INIS)

    Hunter, H.T.; Slater, C.O.; Holland, L.B.; Tracz, G.; Marshall, W.J.; Parsons, J.L.

    2000-01-01

    Over the past several decades, nuclear science has relied on experimental research to verify and validate information about shielding nuclear radiation for a variety of applications. These benchmarks are compared with results from computer code models and are useful for the development of more accurate cross-section libraries, computer code development of radiation transport modeling, and building accurate tests for miniature shielding mockups of new nuclear facilities. When documenting measurements, one must describe many parts of the experimental results to allow a complete computational analysis. Both old and new benchmark experiments, by any definition, must provide a sound basis for modeling more complex geometries required for quality assurance and cost savings in nuclear project development. Benchmarks may involve one or many materials and thicknesses, types of sources, and measurement techniques. In this paper the benchmark experiments of varying complexity are chosen to study the transport properties of some popular materials and thicknesses. These were analyzed using three-dimensional (3-D) models and continuous energy libraries of MCNP4B2, a Monte Carlo code developed at Los Alamos National Laboratory, New Mexico. A shielding benchmark library provided the experimental data and allowed a wide range of choices for source, geometry, and measurement data. The experimental data had often been used in previous analyses by reputable groups such as the Cross Section Evaluation Working Group (CSEWG) and the Organization for Economic Cooperation and Development/Nuclear Energy Agency Nuclear Science Committee (OECD/NEANSC)

  4. Residential gas-fired sorption heat pumps. Test and technology evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Naeslund, M.

    2008-12-15

    Heat pumps may be the next step in gas-fired residential space heating. Together with solar energy it is an option to combine natural gas and renewable energy. Heat pumps for residential space heating are likely to be based on the absorption or adsorption process, i.e. sorption heat pumps. Manufacturers claim that the efficiency could reach 140-160%. The annual efficiency will be lower but it is clear that gas-fired heat pumps can involve an efficiency and technology step equal to the transition from non-condensing gas boilers with atmospheric burners to condensing boilers. This report contains a review of the current sorption gas-fired heat pumps for residential space heating and also the visible development trends. A prototype heat pump has been laboratory tested. Field test results from Germany and the Netherlands are also used for a technology evaluation. The tested heat pump unit combines a small heat pump and a supplementary condensing gas boiler. Field tests show an average annual efficiency of 120% for this prototype design. The manufacturer abandoned the tested design during the project period and the current development concentrates on a heat pump design only comprising the heat pump, although larger. The heat pump development at three manufacturers in Germany indicates a commercial stage around 2010-2011. A fairly high electricity consumption compared to traditional condensing boilers was observed in the tested heat pump. Based on current prices for natural gas and electricity the cost savings were estimated to 12% and 27% for heat pumps with 120% and 150% annual efficiency respectively. There is currently no widespread performance testing procedure useful for annual efficiency calculations of gas-fired heat pumps. The situation seems to be clearer for electric compression heat pumps regarding proposed testing and calculation procedures. A German environmental label exists and gasfired sorption heat pumps are also slightly treated in the Eco-design work

  5. Neutron integral test of graphite cross sections in MeV energy region for the JENDL-3T through an analysis of WINFRITH shielding experiment

    International Nuclear Information System (INIS)

    Ueki, Kohtaro; Sakurai, Kiyoshi.

    1988-01-01

    The neutron integral tests of graphite cross sections in MeV neutron energy region for the ENDF/B-IV, JENDL-2, JENDL-3PR1 and -3T were performed through the Monte Carlo analysis of the graphite shielding experiment at the WINFRITH. The measured values were on the reaction rates of 115 In(n,n') 115m In, 27 Al(n,α) 24 Na, 32 S(n,p) 32 P, and 103 Rh(n,n') 103m Rh threshold detectors located in the graphite slabs, so that the experiment on the graphite was good at the integral test of neutron cross sections in MeV energy resion. (author)

  6. Radiation shielding device

    International Nuclear Information System (INIS)

    Nakagawa, Takahiro; Yamagami, Makoto.

    1996-01-01

    A fixed shielding member made of a radiation shielding material is constituted in perpendicular to an opening formed on radiation shielding walls. The fixed shielding member has one side opened and has other side, the upper portion and the lower portion disposed in close contact with the radiation shielding walls. Movable shielding members made of a radiation shielding material are each disposed openably on both side of the fixed shielding member. The movable shielding member has a shaft as a fulcrum on one side thereof for connecting it to the radiation shielding walls. The other side has a handle attached for opening/closing the movable shielding member. Upon access of an operator, when each one of the movable shielding members is opened/closed on every time, leakage of linear or scattered radiation can be prevented. Even when both of the movable shielding members are opened simultaneously, the fixed shielding member and the movable shielding members form labyrinth to prevent leakage of linear radioactivity. (I.N.)

  7. An assessment of RELAP5 MOD3.1.1 condensation heat transfer modeling with GIRAFFE heat transfer tests

    International Nuclear Information System (INIS)

    Boyer, B.D.; Parlatan, Y.; Slovik, G.C.; Rohatgi, U.S.

    1995-01-01

    RELAP5 MOD3.1.1 is being used to simulate Loss of Coolant Accidents (LOCA) for the Simplified Boiling Water Reactor (SBWR) being proposed by General Electric (GE). One of the major components associated with the SBWR is the Passive Containment Cooling System (PCCS) which provides the long-term heat sink to reject decay heat. The RELAP5 MOD3.1.1 code is being assessed for its ability to represent accurately the PCCS. Data from the Phase 1, Step 1 Heat Transfer Tests performed at Toshiba's Gravity-Driven Integral Full-Height Test for Passive Heat Removal (GIRAFFE) facility will be used for assessing the ability of RELAP5 to model condensation in the presence of noncondensables. The RELAP5 MOD3.1.1 condensation model uses the University of California at Berkeley (UCB) correlation developed by Vierow and Schrock. The RELAP5 code uses this heat transfer coefficient with the gas velocity effect multiplier being limited to 2. This heat transfer option was used to analyze the condensation heat transfer in the GIRAFFE PCCS heat exchanger tubes in the Phase 1, Step 1 Heat Transfer Tests which were at a pressure of 3 bar and had a range of nitrogen partial pressure fractions from 0.0 to 0.10. The results of a set of RELAP5 calculations al these conditions were compared with the GIRAFFE data. The effects of PCCS cell nodings on the heat transfer process were also studied. The UCB correlation, as implemented in RELAP5, predicted the heat transfer to ±5% of the data with a three-node model. The three-node model has a large cell in the entrance region which smeared out the entrance effects on the heat transfer, which tend to overpredict the condensation. Hence, the UCB correlation predicts condensation heat transfer in the presence of noncondensable gases with only a coarse mesh. The cell length term in the condensation heat transfer correlation implemented in the code must be removed to allow for accurate calculations with smaller cell sizes

  8. Heat transfer coefficient testing in nuclear fuel rod bundles with mixing vane grids

    International Nuclear Information System (INIS)

    Conner, Michael E.; Smith, L. David III; Holloway, Mary V.; Beasley, Donald E.

    2005-01-01

    An air heat transfer test facility was developed to test the heat transfer downstream of support grids in simulated PWR nuclear fuel rod bundles. The goal of this testing is to study the single-phase heat transfer coefficients downstream of grids with mixing vanes in a square-pitch rod bundle. The technique developed utilizes fully-heated grid spans and a specially designed thermocouple holder that can be moved axially down the rod bundle and aximuthally within a test rod. From this testing, the axial and aximuthally varying heat transfer coefficient can be determined. Different grid designs are tested and compared to determine the heat transfer enhancement associated with key grid features such as mixing vanes. (author)

  9. Thermal response test data of five quadratic cross section precast pile heat exchangers

    Directory of Open Access Journals (Sweden)

    Maria Alberdi-Pagola

    2018-06-01

    Full Text Available This data article comprises records from five Thermal Response Tests (TRT of quadratic cross section pile heat exchangers. Pile heat exchangers, typically referred to as energy piles, consist of traditional foundation piles with embedded heat exchanger pipes. The data presented in this article are related to the research article entitled “Comparing heat flow models for interpretation of precast quadratic pile heat exchanger thermal response tests” (Alberdi-Pagola et al., 2018 [1]. The TRT data consists of measured inlet and outlet temperatures, fluid flow and injected heat rate recorded every 10 min. The field dataset is made available to enable model verification studies.

  10. ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Strebkov, Yu [ENTEK, Moscow (Russian Federation); Avsjannikov, A [ENTEK, Moscow (Russian Federation); Baryshev, M [NIAT, Moscow (Russian Federation); Blinov, Yu [ENTEK, Moscow (Russian Federation); Shatalov, G [KIAE, Moscow (Russian Federation); Vasiliev, N [KIAE, Moscow (Russian Federation); Vinnikov, A [ENTEK, Moscow (Russian Federation); Chernjagin, A [DYNAMICA, Moscow (Russian Federation)

    1995-03-01

    A reference non-breeding blanket is under development now for the ITER Basic Performance Phase for the purpose of high reliability during the first stage of ITER operation. More severe operation modes are expected in this stage with first wall (FW) local heat loads up to 100-300Wcm{sup -2}. Integration of a blanket design with protective and start limiters requires new solutions to achieve high reliability, and possible use of beryllium as a protective material leads to technologies. The rigid shielding blanket concept was developed in Russia to satisfy the above-mentioned requirements. The concept is based on a copper alloy FW, austenitic stainless steel blanket structure, water cooling. Beryllium protection is integrated in the FW design. Fabrication technology and assembly procedure are described in parallel with the equipment used. (orig.).

  11. Transient shielded liquid hydrogen containers

    International Nuclear Information System (INIS)

    Varghese, A.P.; Herring, R.H.

    1990-01-01

    The storage of hydrogen in the liquid phase has been limited in duration due to the thermal performance constraints of conventional Liquid Hydrogen containers available. Conventional Liquid Hydrogen containers lose hydrogen because of their relatively high heat leak and variations in usage pattern of hydrogen due to shutdowns. Local regulations also discourage venting of hydrogen. Long term storage of Liquid Hydrogen without product loss was usually accomplished using Liquid Nitrogen sacrificial shields. This paper reports on a new low heat leak container developed and patented that will extend the storage time of liquid hydrogen by five hundred percent. The principle of operation of the Transient Shields which makes the extraordinary performance of this container feasible is described in this paper. Also covered are the impact of this new container on present applications of hydrogen and the new opportunities afforded to Liquid hydrogen in the world hydrogen market

  12. Experimental result of BWR post-CHF tests. Critical heat flux and post-CHF heat transfer coefficient. Contract research

    International Nuclear Information System (INIS)

    Iguchi, Tadashi; Anoda, Yoshinari

    2002-02-01

    Authors performed post-CHF experiments under wider pressure ranges of 2 MPa - 18 MPa, wider mass flux ranges of 33 kg/m 2 s - 1651 kg/m 2 s and wider superheat of heaters up to 500 K in comparison to experimental ranges at previous post-CHF experiments. Data on boiling transition, critical heat flux and post-CHF heat transfer coefficient were obtained. Used test section was 4x4-rod bundle with heaters, which diameter and length were the same as those of BWR nuclear fuels. As the result of the experiments, it was found that the boiling transition occurred just below several grid spacers, and that the fronts of the boiling transition region proceeded lower with increase of heated power. Heat transfer was due to nucleate boiling above grid spacers, while it was due to film boiling below grid spacers. Consequently, critical heat flux is affected on the distance from the grid spacers. Critical heat flux above the grid spacers was about 15% higher than that below the grid spacers, by comparing them under the same local condition. Heat transfer by steam turbulent flow was dominant to post-CHF heat transfer, when superheat of heaters was sufficiently high. Then, post-CHF heat transfer coefficient was predicted with heat transfer correlations for single-phase flow. On the other hand, when superhead of heaters was not sufficiently high, post-CHF heat transfer coefficient was higher than the prediction with heat transfer correlations for single-phase flow. Mass flux effect on post-CHF heat transfer coefficient was described by standardization of post-CHF heat transfer coefficient with the prediction for single-phase flow. However, pressure effect, superheat effect and effect of position were not described. Authors clarified that those effects could be described with functions of heater temperature and position. Post-CHF heat transfer coefficient was lowest just blow the grid spacers, and it increased with the lower positions. It increased by about 30% in one span of the grid

  13. Experimental result of BWR post-CHF tests. Critical heat flux and post-CHF heat transfer coefficient. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi; Anoda, Yoshinari [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwaki, Chikako [Toshiba Corp., Tokyo (Japan)

    2002-02-01

    Authors performed post-CHF experiments under wider pressure ranges of 2 MPa - 18 MPa, wider mass flux ranges of 33 kg/m{sup 2}s - 1651 kg/m{sup 2}s and wider superheat of heaters up to 500 K in comparison to experimental ranges at previous post-CHF experiments. Data on boiling transition, critical heat flux and post-CHF heat transfer coefficient were obtained. Used test section was 4x4-rod bundle with heaters, which diameter and length were the same as those of BWR nuclear fuels. As the result of the experiments, it was found that the boiling transition occurred just below several grid spacers, and that the fronts of the boiling transition region proceeded lower with increase of heated power. Heat transfer was due to nucleate boiling above grid spacers, while it was due to film boiling below grid spacers. Consequently, critical heat flux is affected on the distance from the grid spacers. Critical heat flux above the grid spacers was about 15% higher than that below the grid spacers, by comparing them under the same local condition. Heat transfer by steam turbulent flow was dominant to post-CHF heat transfer, when superheat of heaters was sufficiently high. Then, post-CHF heat transfer coefficient was predicted with heat transfer correlations for single-phase flow. On the other hand, when superhead of heaters was not sufficiently high, post-CHF heat transfer coefficient was higher than the prediction with heat transfer correlations for single-phase flow. Mass flux effect on post-CHF heat transfer coefficient was described by standardization of post-CHF heat transfer coefficient with the prediction for single-phase flow. However, pressure effect, superheat effect and effect of position were not described. Authors clarified that those effects could be described with functions of heater temperature and position. Post-CHF heat transfer coefficient was lowest just blow the grid spacers, and it increased with the lower positions. It increased by about 30% in one span of

  14. Comparison of the CATHENA model of Gentilly-2 end shield cooling system predictions to station data

    Energy Technology Data Exchange (ETDEWEB)

    Zagre, G.; Sabourin, G. [Candu Energy Inc., Montreal, Quebec (Canada); Chapados, S. [Hydro-Quebec, Montreal, Quebec (Canada)

    2012-07-01

    As part of the Gentilly-2 Refurbishment Project, Hydro-Quebec has elected to perform the End Shield Cooling Safety Analysis. A CATHENA model of Gentilly-2 End Shield Cooling System was developed for this purpose. This model includes new elements compared to other CANDU6 End Shield Cooling models such as a detailed heat exchanger and control logic model. In order to test the model robustness and accuracy, the model predictions were compared with plant measurements.This paper summarizes this comparison between the model predictions and the station measurements. It is shown that the CATHENA model is flexible and accurate enough to predict station measurements for critical parameters, and the detailed heat exchanger model allows reproducing station transients. (author)

  15. General-Purpose Heat Source Safety Verification Test program: Edge-on flyer plate tests

    International Nuclear Information System (INIS)

    George, T.G.

    1987-03-01

    The radioisotope thermoelectric generator (RTG) that will supply power for the Galileo and Ulysses space missions contains 18 General-Purpose Heat Source (GPHS) modules. The GPHS modules provide power by transmitting the heat of 238 Pu α-decay to an array of thermoelectric elements. Each module contains four 238 PuO 2 -fueled clads and generates 250 W(t). Because the possibility of a launch vehicle explosion always exists, and because such an explosion could generate a field of high-energy fragments, the fueled clads within each GPHS module must survive fragment impact. The edge-on flyer plate tests were included in the Safety Verification Test series to provide information on the module/clad response to the impact of high-energy plate fragments. The test results indicate that the edge-on impact of a 3.2-mm-thick, aluminum-alloy (2219-T87) plate traveling at 915 m/s causes the complete release of fuel from capsules contained within a bare GPHS module, and that the threshold velocity sufficient to cause the breach of a bare, simulant-fueled clad impacted by a 3.5-mm-thick, aluminum-alloy (5052-T0) plate is approximately 140 m/s

  16. Explosion overpressure test series: General-Purpose Heat Source development: Safety Verification Test program

    International Nuclear Information System (INIS)

    Cull, T.A.; George, T.G.; Pavone, D.

    1986-09-01

    The General-Purpose Heat Source (GPHS) is a modular, radioisotope heat source that will be used in radioisotope thermoelectric generators (RTGs) to supply electric power for space missions. The first two uses will be the NASA Galileo and the ESA Ulysses missions. The RTG for these missions will contain 18 GPHS modules, each of which contains four 238 PuO 2 -fueled clads and generates 250 W/sub (t)/. A series of Safety Verification Tests (SVTs) was conducted to assess the ability of the GPHS modules to contain the plutonia in accident environments. Because a launch pad or postlaunch explosion of the Space Transportation System vehicle (space shuttle) is a conceivable accident, the SVT plan included a series of tests that simulated the overpressure exposure the RTG and GPHS modules could experience in such an event. Results of these tests, in which we used depleted UO 2 as a fuel simulant, suggest that exposure to overpressures as high as 15.2 MPa (2200 psi), without subsequent impact, does not result in a release of fuel

  17. Field Test of a Steam Condenser Heat Sink Concept

    Science.gov (United States)

    1974-01-01

    stored underground for a specified time. A functional and economical heat rejection system is an important design consideration for such...per- mits the use of tunnels for other than just heat sink purposes. If existing tunnels can be used, the concept becomes economically attractive...that the water meter readings aie a valid indication of the mpu ! and that condensate was lost bv seepage thionuli the lock and or ballast into the

  18. NEPTUN/5052, PWR LOCA Cooling Heat Transfer Tests for Loft, Reflood Test

    International Nuclear Information System (INIS)

    Richner, M.; Analytis, G.Th.; Aksan, S.N.

    1993-01-01

    1 - Description of test facility: NEPTUN is designed to perform PWR LOCA simulation experiments, which provide the full length emergency cooling heat transfer tests for LOFT. Therefore the NEPTUN heater bundle with 33 electrical heater elements and 4 guide tubes simulates a section of the LOFT nuclear core. The main test loop also contains measuring systems for the carry-over rate and for the steam expelled, and a back-pressure control system. A water loop brings the water to the initial reflooding conditions. In addition, auxiliary systems maintain normal operating conditions. 2 - Description of test: Test 5052 is one of a series of 40 reflood tests performed in NEPTUN. Before the start of the test, the flooding water in its circuit is brought to the following conditions: pressure = 4.1 bar; velocity = 2.5 cm/sec; subcooling temperature = 78 C; single rod power = 2.45 kW; maximal initial cladding temperature = 867 C. 3 - Status: CSNI1013/01, 21-Jul-1993 Arrived at NEADB

  19. Development, manufacturing and testing of a gas-loaded variable conductance methanol heat pipe

    Science.gov (United States)

    Vanbuggenum, R. I. J.; Daniels, D. H. W.

    1987-02-01

    The experimental technology required to measure the performance of moderate temperature heat pipes is presented. The heat pipe manufacturing process is described. The hydrodynamic characteristics of the porous structure inside the heat pipe envelope were examined using a specially developed test rig, based upon a steady-state evaporation test. A fully automated test facility was developed and validated by testing constant conductance and variable conductance heat pipes (VCHP). Theoretical performance predictions are illustrated in terms of pressure, depicted in 3D-plots, and compared with the test results of the heat pipe performance tests. The design of the VCHP was directed towards the verification of the VCHP mathematical model. The VCHP design is validated and ready for the final testing and model verification.

  20. Concrete shielding for nuclear ship 'Mutsu'

    International Nuclear Information System (INIS)

    Nagase, Tetsuo; Nakajima, Tadao; Okumura, Tadahiko; Saito, Tetsuo

    1983-01-01

    The nuclear ship ''Mutsu'' was constructed in 1970 as the fourth in the world. On September 1, 1974, during the power raising test in the Pacific Ocean, radiation leak was detected. As the result of investigation, it was found that the cause was the fast neutrons streaming through the gap between the reactor pressure vessel and the primary shield. In order to repair the shielding facility, the Japan Nuclear Ship Research Development Agency carried out research and development and shielding design. It was decided to adopt serpentine concrete for the primary shield, which is the excellent moderator of fast neutrons even at high temperature, and heavy concrete for the secondary shield, which is effective for shielding both gamma ray and neutron beam. The repair of shielding was carried out in the Sasebo Shipyard, and completed in August, 1982. The outline of the repair work is reported. The weight increase was about 300 t. The conditions of the shielding design, the method of shielding analysis, the performance required for the shielding concrete, the preliminary experiment on heavy concrete and the construction works of serpentine concrete and heavy concrete are described. (Kako, I.)

  1. A shield against distraction

    OpenAIRE

    Halin, N.; Marsh, J.E.; Hellman, A.; Hellstrom, I.; Sörqvist, Patrik

    2014-01-01

    In this paper, we apply the basic idea of a trade-off between the level of concentration and distractibility to test whether a manipulation of task difficulty can shield against distraction. Participants read, either in quiet or with a speech noise background, texts that were displayed either in an easy-to-read or a hard-to-read font. Background speech impaired prose recall, but only when the text was displayed in the easy-to-read font. Most importantly, recall was better in the background sp...

  2. Gallium-cladding compatibility testing plan: Phase 3: Test plan for centrally heated surrogate rodlet test. Revision 2

    International Nuclear Information System (INIS)

    Morris, R.N.; Baldwin, C.A.; Wilson, D.F.

    1998-07-01

    The Fissile Materials Disposition Program (FMDP) is investigating the use of weapons grade plutonium in mixed oxide (MOX) fuel for light-water reactors (LWR). Commercial MOX fuel has been successfully used in overseas reactors for many years; however, weapons derived fuel may differ from the previous commercial fuels because of small amounts of gallium impurities. A concern presently exists that the gallium may migrate out of the fuel, react with and weaken the clad, and thereby promote loss of fuel pin integrity. Phases 1 and 2 of the gallium task are presently underway to investigate the types of reactions that occur between gallium and clad materials. This is a Level-2 document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-Oxide Fuel Irradiation Test Project Plan. This Plan summarizes the projected Phase 3 Gallium-Cladding compatibility heating test and the follow-on post test examination (PTE). This work will be performed using centrally-heated surrogate pellets, to avoid unnecessary complexities and costs associated with working with plutonium and an irradiation environment. Two sets of rodlets containing pellets prepared by two different methods will be heated. Both sets will have an initial bulk gallium content of approximately 10 ppm. The major emphasis of the PTE task will be to examine the material interactions, particularly indications of gallium transport from the pellets to the clad

  3. An assessment of the lifetime of Faraday shield elements

    International Nuclear Information System (INIS)

    Caughman, J.B.O. II; Ruzic, D.N.; Hoffman, D.J.; Langley, R.A.; Lewis, M.B.; Ryan, P.M.

    1989-01-01

    The interaction of plasma with rf fields from an ion cyclotron range of frequencies (ICRF) antenna has been studied to estimate the amount of Faraday shield erosion expected in normal ICRF heating (ICRH) operation. Plasma parameters and ion energies have been measured in the near field of an antenna and used in a model to estimate the erosion rate of the Faraday shield surface. Experiments were conducted on the RF Test Facility (RFTF), a magnetic mirror device at Oak Ridge National Laboratory (ORNL), using a single-strap resonant loop antenna with a two-tier Faraday shield. The outer tier, facing the plasma, was layered with graphite tiles. The antenna was operated at currents and voltages (∼500 A, ∼20 kV at 25 kW) within 50% of those expected in tokamaks. The time varying floating potential was measured with a capacitively coupled probe, and the time-averaged floating potential, electron temperature, and electron density were measured with a Langmuir probe. Both probes were scanned in front of the antenna. Ion energies were measured with a gridded energy analyzer located below the antenna, and samples of silicon were placed on the Faraday shield surface to estimate the incident ion energy. The capacitive probe measurements show that the rf floating potential follows the magnetic field pattern of the antenna, indicating that the electromagnetic fields are responsible for the potential formation. Plasma parameters and ion energies have been correlated with the antenna current and used in s computational model of the plasma sheath to predict the amount of erosion expected from the Faraday shield elements exposed to plasma. Predictions of light ion sputtering of candidate Faraday shield materials are presented. 19 refs., 6 figs., 1 tab

  4. An assessment of the lifetime of Faraday shield elements

    International Nuclear Information System (INIS)

    Caughman, J.B.O. II; Ruzic, D.N.; Hoffman, D.J.; Langley, R.A.; Lewis, M.B.; Ryan, P.M.

    1990-01-01

    The interaction of plasma with rf fields from an ion cyclotron range of frequencies (ICRF) antenna has been studied to estimate the amount of Faraday shield erosion expected in normal ICRF heating operation. Plasma parameters and ion energies have been measured in the near field of an antenna and used in a model to estimate the erosion rate of the Faraday shield surface. Experiments were conducted on the RF Test Facility, a magnetic mirror device at Oak Ridge National Laboratory, using a single-strap resonant loop antenna with a two-tier Faraday shield. The outer tier, facing the plasma, was layered with graphite tiles. The antenna was operated at currents and voltages within 50% of those expected in tokamaks. The time-varying floating potential was measured with a capacitively coupled probe, and the time-averaged floating potential, electron temperature, and electron density were measured with a Langmuir probe. Ion energies were measured with a gridded energy analyser located below the antenna, and samples of silicon were placed on the Faraday shield surface to estimate the incident ion energy. The capacitive probe measurements show that the rf floating potential follows the magnetic field pattern of the antenna, indicating that the electromagnetic fields are responsible for the potential formation. Plasma parameters and ion energies have been correlated with the antenna current and used in a computational model of the plasma sheath to predict the amount of erosion expected from the Faraday shield elements exposed to plasma. Predictions of light ion sputtering of candidate Faraday shield materials are presented

  5. Handout on shielding calculation

    International Nuclear Information System (INIS)

    Heilbron Filho, P.F.L.

    1991-01-01

    In order to avoid the difficulties of the radioprotection supervisors in the tasks related to shielding calculations, is presented in this paper the basic concepts of shielding theory. It also includes exercises and examples. (author)

  6. EMI Shielding Performance For Varies Frequency by Metal Plating on Mold Compound

    Directory of Open Access Journals (Sweden)

    Min Fee Tai

    2017-07-01

    Full Text Available Conformal metalization on mold compound offers new possibility for IC package design to improve features such as rigidization of the flexible core, heat sink capability, 3D-circuit patterning and the electromagnetic interference (EMI shielding. With the unique processes, the fabrication technology had enabled to achieve the high reliable performance and had passed the electrical test. Following research after the reliability concern, this paper further study the shielding effectiveness of varying coating thickness with respect to laboratory simulated EMI condition, using radio frequency from 10MHz to 5.8 GHz. Different metal namely pure nickel, nickel-phosphorous and pure plated copper are studied for their effectiveness of EMI sheilding. Our first result showed over 35-40dB of shielding effectiveness is achievable on high frequency 868-5800MHz. Nevertheless on low frequency of 10MHz, the shielding effectiveness achievement is below than 25dB. To overcome the shielding need for lower frequency, we further expanded our test by choosing ferromagentic material Nicke/Ironl-alloy in combination with thick copper plating. With this new metal combination, EMI shielding effectiveness for lower frequency is improved to 40dB.

  7. Gravity Scaling of a Power Reactor Water Shield

    International Nuclear Information System (INIS)

    Reid, Robert S.; Pearson, J. Boise

    2008-01-01

    Water based reactor shielding is being considered as an affordable option for potential use on initial lunar surface reactor power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxillary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2006). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa n . These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined

  8. Testing and Measurement Techniques in Heat Transfer and Combustion.

    Science.gov (United States)

    1980-09-01

    fine brass gauze, which have initially been heated or cooled by passing air through copper spiral coils immersed in a bath of heated oil or alcohol and...par l’analyseur peut Ctre interpr~t~e soit cornie 6tant celle r~ellement presente dans la flamme, sojt canine r~sultant d’une oxydation de NO syant...pr~l ,vement et analyse provient d’one oxydation de NO en proportion variltle suivant le type de sonde utilis6. De plus ii apparalt que cette

  9. Tests for removal of decay heat by natural convection

    International Nuclear Information System (INIS)

    Kashiwagi, E.; Wataru, M.; Gomi, Y.; Hattori, Y.; Ozaki, S.

    1993-01-01

    Interim storage technology for spent fuel by dry storage casks have been investigated. The casks are vertically placed in a storage building. The decay heat is removed from the outer cask surface by natural convection of air entering from the building wall to the roof. The air flow pattern in the storage building was governed by the natural driving pressure difference and circulating flow. The purpose of this study is to understand the mechanism of the removal of decay heat from casks by natural convection. The simulated flow conditions in the building were assumed as a natural and forced combined convection and were investigated by the turbulent quantities near wall. (author)

  10. Hot Cell Window Shielding Analysis Using MCNP

    International Nuclear Information System (INIS)

    Pope, Chad L.; Scates, Wade W.; Taylor, J. Todd

    2009-01-01

    The Idaho National Laboratory Materials and Fuels Complex nuclear facilities are undergoing a documented safety analysis upgrade. In conjunction with the upgrade effort, shielding analysis of the Fuel Conditioning Facility (FCF) hot cell windows has been conducted. This paper describes the shielding analysis methodology. Each 4-ft thick window uses nine glass slabs, an oil film between the slabs, numerous steel plates, and packed lead wool. Operations in the hot cell center on used nuclear fuel (UNF) processing. Prior to the shielding analysis, shield testing with a gamma ray source was conducted, and the windows were found to be very effective gamma shields. Despite these results, because the glass contained significant amounts of lead and little neutron absorbing material, some doubt lingered regarding the effectiveness of the windows in neutron shielding situations, such as during an accidental criticality. MCNP was selected as an analysis tool because it could model complicated geometry, and it could track gamma and neutron radiation. A bounding criticality source was developed based on the composition of the UNF. Additionally, a bounding gamma source was developed based on the fission product content of the UNF. Modeling the windows required field inspections and detailed examination of drawings and material specifications. Consistent with the shield testing results, MCNP results demonstrated that the shielding was very effective with respect to gamma radiation, and in addition, the analysis demonstrated that the shielding was also very effective during an accidental criticality.

  11. Assessment of Feasibility of the Beneficial Use of Waste Heat from the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Donna P. Guillen

    2012-07-01

    This report investigates the feasibility of using waste heat from the Advanced Test Reactor (ATR). A proposed glycol waste heat recovery system was assessed for technical and economic feasibility. The system under consideration would use waste heat from the ATR secondary coolant system to preheat air for space heating of TRA-670. A tertiary coolant stream would be extracted from the secondary coolant system loop and pumped to a new plate and frame heat exchanger, where heat would be transferred to a glycol loop for preheating outdoor air in the heating and ventilation system. Historical data from Advanced Test Reactor operations over the past 10 years indicates that heat from the reactor coolant was available (when needed for heating) for 43.5% of the year on average. Potential energy cost savings by using the waste heat to preheat intake air is $242K/yr. Technical, safety, and logistics considerations of the glycol waste heat recovery system are outlined. Other opportunities for using waste heat and reducing water usage at ATR are considered.

  12. Heat transfer in a seven-rod test bundle with supercritical pressure water (1). Experiments

    International Nuclear Information System (INIS)

    Ezato, Koichiro; Seki, Yohji; Dairaku, Masayuki; Suzuki, Satoshi; Enoeda, Mikio; Akiba, Masato; Mori, H.; Oka, Y.

    2009-01-01

    Heat transfer experiments in a seven-rod test bundle with supercritical pressure water has been carried out. The pressure drop and heat transfer coefficients (HTCs) in the test section are evaluated. In the present limited conditions, difference between HTCs at the surface facing the sub-channel center and those at the surface in the narrowest region between rods is not observed. (author)

  13. Design of emergency shield

    International Nuclear Information System (INIS)

    Soliman, S.E.

    1993-01-01

    Manufacturing of an emergency movable shield in the hot laboratories center is urgently needed for the safety of personnel in case of accidents or spilling of radioactive materials. In this report, a full design for an emergency shield is presented and the corresponding dose rates behind the shield for different activities (from 1 mCi to 5 Ci) was calculated by using micro shield computer code. 4 figs., 1 tab

  14. Shielding requirements for particle bed propulsion systems

    Science.gov (United States)

    Gruneisen, S. J.

    1991-06-01

    Nuclear Thermal Propulsion systems present unique challenges in reliability and safety. Due to the radiation incident upon all components of the propulsion system, shielding must be used to keep nuclear heating in the materials within limits; in addition, electronic control systems must be protected. This report analyzes the nuclear heating due to the radiation and the shielding required to meet the established criteria while also minimizing the shield mass. Heating rates were determined in a 2000 MWt Particle Bed Reactor (PBR) system for all materials in the interstage region, between the reactor vessel and the propellant tank, with special emphasis on meeting the silicon dose criteria. Using a Lithium Hydride/Tungsten shield, the optimum shield design was found to be: 50 cm LiH/2 cm W on the axial reflector in the reactor vessel and 50 cm LiH/2 cm W in a collar extension of the inside shield outside of the pressure vessel. Within these parameters, the radiation doses in all of the components in the interstage and lower tank regions would be within acceptable limits for mission requirements.

  15. Electromagnetically shielded building

    International Nuclear Information System (INIS)

    Takahashi, T.; Nakamura, M.; Yabana, Y.; Ishikawa, T.; Nagata, K.

    1992-01-01

    This invention relates to a building having an electromagnetic shield structure well-suited for application to an information network system utilizing electromagnetic waves, and more particularly to an electromagnetically shielded building for enhancing the electromagnetic shielding performance of an external wall. 6 figs

  16. Electromagnetically shielded building

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, T; Nakamura, M; Yabana, Y; Ishikawa, T; Nagata, K

    1992-04-21

    This invention relates to a building having an electromagnetic shield structure well-suited for application to an information network system utilizing electromagnetic waves, and more particularly to an electromagnetically shielded building for enhancing the electromagnetic shielding performance of an external wall. 6 figs.

  17. Simulation Tests in Whole Building Heat and Moisture Transfer

    DEFF Research Database (Denmark)

    Rode, Carsten; Peuhkuri, Ruut Hannele; Woloszyn, Monika

    2006-01-01

    An important part of the International Energy Agency project, ECBCS, Annex 41 is about modelling the integral heat, air and moisture transfer processes that take place in “whole buildings”. Such modelling deals with all most relevant elements of buildings: The indoor air, the building envelope...

  18. TESTING OF REFRIGERANT MIXTURES IN RESIDENTIAL HEAT PUMPS

    Science.gov (United States)

    The report gives results of an investigation of four possibilities for replacing Hydrochlorofluorocarbon-22 (HCFC-22) with the non-ozone-depleting new refrigerants R-407D and R-407C in residential heat pumps. The first and simplest scenario was a retrofit with no hardware modific...

  19. Heat Pipe Powered Stirling Conversion for the Demonstration Using Flattop Fission (DUFF) Test

    Science.gov (United States)

    Gibson, Marc A.; Briggs, Maxwell H.; Sanzi, James L.; Brace, Michael H.

    2013-01-01

    Design concepts for small Fission Power Systems (FPS) have shown that heat pipe cooled reactors provide a passive, redundant, and lower mass option to transfer heat from the fuel to the power conversion system, as opposed to pumped loop designs typically associated with larger FPS. Although many systems have been conceptually designed and a few making it to electrically heated testing, none have been coupled to a real nuclear reactor. A demonstration test named DUFF Demonstration Using Flattop Fission, was planned by the Los Alamos National Lab (LANL) to use an existing criticality experiment named Flattop to provide the nuclear heat source. A team from the NASA Glenn Research Center designed, built, and tested a heat pipe and power conversion system to couple to Flattop with the end goal of making electrical power. This paper will focus on the design and testing performed in preparation for the DUFF test.

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

    Science.gov (United States)

    Meitner, P. L.

    1974-01-01

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

  1. High temperature superconducting current lead test facility with heat pipe intercepts

    International Nuclear Information System (INIS)

    Blumenfeld, P.E.; Prenger, C.; Roth, E.W.; Stewart, J.A.

    1998-01-01

    A high temperature superconducting (HTS) current lead test facility using heat pipe thermal intercepts is under development at the Superconducting Technology Center at Los Alamos National Laboratory. The facility can be configured for tests at currents up to 1,000 A. Mechanical cryocoolers provide refrigeration to the leads. Electrical isolation is maintained by intercepting thermal energy from the leads through cryogenic heat pipes. HST lead warm end temperature is variable from 65 K to over 90 K by controlling heat pipe evaporator temperature. Cold end temperature is variable up to 30 K. Performance predictions in terms of heat pipe evaporator temperature as a function of lead current are presented for the initial facility configuration, which supports testing up to 200 A. Measurements are to include temperature and voltage gradient in the conventional and HTS lead sections, temperature and heat transfer rate in the heat pipes. as well as optimum and off-optimum performance of the conventional lead sections

  2. Actively shielded low level gamma - spectrometric system

    International Nuclear Information System (INIS)

    Mrdja, D.; Bikit, I.; Forkapic, S.; Slivka, J.; Veskovic, M.

    2005-01-01

    The results of the adjusting and testing of the actively shielded low level gamma-spectrometry system are presented. The veto action of the shield reduces the background in the energy region of 50 keV to the 2800 keV for about 3 times. (author) [sr

  3. Radiation shielding in dental radiography

    International Nuclear Information System (INIS)

    Stenstroem, B.; Rehnmark-Larsson, S.; Julin, P.; Richter, S.

    1983-01-01

    The protective effect in the thyroid region from different types of radiation shieldings at intraoral radiography has been studied as well as the reduction of the absorbed dose to the sternal and the gonadal regions. The shieldings tested were five different types of leaded aprons, of which three had an attached leaded collar and the other two were used in combination with separate soft leaded collars. Furthermore one of the soft leaded collars and an unflexible horizontal leaded shield were tested separately. Two dental x-ray machines of 60 and 65 kVp with rectangular and circular tube collimators were used. The exposure time corresponded to speed group E film. The absorbed doses were measured with two ionization chambers. No significant difference in the protective effect in the thyroid gland could be found between the different types of radiation shieldings. There was a dose reduction by approximately a factor of 2 to the thyroid region down to 0.08 mGy per full survey using parallelling technique, and below 0.001 mGy per single bitewing exposure. The shieldings reduced the thyroid dose using bisecting-angle technique by a factor of 5 down to 0.15 mGy per full survey (20 exposures). In the sternal region the combinations of apron and collar reduced the absorbed dose from a full survey to below 2 μGy compared with 18 μGy (parallelling) and 31 μGy (biscting-angle) without any shielding. With the horizontal leaded shield a reduction by a factor of 6 was obtained but no significant sternal dose reduction could be detected from the soft collar alone. The gonadal dose could be reduced by a factor of 10 with the horizontal leaded shield, parallelling technique and circular collimator. Using leaded aprons the gonadal dose was approximately one per cent of the dose without any shielding, i.e. below 0.01 μGy per single intraoral exposure. (Authors)

  4. Assessment of the Performance of a Ventilated Window Coupled with a Heat Recovery Unit through the Co-Heating Test

    Directory of Open Access Journals (Sweden)

    Ludovico Danza

    2016-01-01

    Full Text Available The aim of the article is to describe the results of an experimental campaign based on the assessment of a heat recovery unit coupled with a dynamic window. Two fully monitored and calibrated outdoor test cells are used, in order to evaluate the energy performance and the related thermal comfort. The former presents a traditional window with double-glazing, aluminum frame and indoor blind and a centrifugal extractor for the air circulation. The latter is equipped with a dynamic window with ventilated and blinded double-glazing provided with a heat exchanger. The connection of the dynamic window and heat recovery unit provides different actions: heat recovery; heat transfer reduction; pre-heating before the exchanger. Different operating configurations allowed the trends of the dynamic system to be assessed in different seasons in terms of energy saving, thermal comfort behavior and energy efficiency. The results showed an overall lower consumption of the innovative system, both in winter and summer, with 20% and 15% energy saving, respectively. In general, the dynamic system provided the best comfort conditions, even if it involves a worse behavior than expected, in the summer season.

  5. Test bench HEATREC for heat loss measurement on solar receiver tubes

    Science.gov (United States)

    Márquez, José M.; López-Martín, Rafael; Valenzuela, Loreto; Zarza, Eduardo

    2016-05-01

    In Solar Thermal Electricity (STE) plants the thermal energy of solar radiation is absorbed by solar receiver tubes (HCEs) and it is transferred to a heat transfer fluid. Therefore, heat losses of receiver tubes have a direct influence on STE plants efficiency. A new test bench called HEATREC has been developed by Plataforma Solar de Almería (PSA) in order to determinate the heat losses of receiver tubes under laboratory conditions. The innovation of this test bench consists in the possibility to determine heat losses under controlled vacuum.

  6. Electromagnetic shielding formulae

    International Nuclear Information System (INIS)

    Dahlberg, E.

    1979-02-01

    This addendum to an earlier collection of electromagnetic shielding formulae (TRITA-EPP-75-27) contains simple transfer matrices suitable for calculating the quasistatic shielding efficiency for multiple transverse-field and axial-field cylindrical and spherical shields, as well as for estimating leakage fields from long coaxial cables and the normal-incidence transmission of a plane wave through a multiple plane shield. The differences and similarities between these cases are illustrated by means of equivalent circuits and transmission line analogies. The addendum also includes a discussion of a possible heuristic improvement of some shielding formulae. (author)

  7. Shielding benchmark problems, (2)

    International Nuclear Information System (INIS)

    Tanaka, Shun-ichi; Sasamoto, Nobuo; Oka, Yoshiaki; Shin, Kazuo; Tada, Keiko.

    1980-02-01

    Shielding benchmark problems prepared by Working Group of Assessment of Shielding Experiments in the Research Committee on Shielding Design in the Atomic Energy Society of Japan were compiled by Shielding Laboratory in Japan Atomic Energy Research Institute. Fourteen shielding benchmark problems are presented newly in addition to twenty-one problems proposed already, for evaluating the calculational algorithm and accuracy of computer codes based on discrete ordinates method and Monte Carlo method and for evaluating the nuclear data used in codes. The present benchmark problems are principally for investigating the backscattering and the streaming of neutrons and gamma rays in two- and three-dimensional configurations. (author)

  8. The Active Muon Shield

    CERN Document Server

    Bezshyiko, Iaroslava

    2016-01-01

    In the SHiP beam-dump of the order of 1011 muons will be produced per second. An active muon-shield is used to magnetically deflect these muons out of the acceptance of the spectrom- eter. This note describes how this shield is modelled and optimized. The SHiP spectrometer is being re-optimized using a conical decay-vessel, and utilizing the possibility to magnetize part of the beam-dump shielding iron. A shield adapted to these new conditions is presented which is significantly shorter and lighter than the shield used in the Technical Proposal (TP), while showing a similar performance.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-12-01

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

  10. ANALISIS KESELAMATAN TERMOHIDROLIK BULK SHIELDING REAKTOR KARTINI

    Directory of Open Access Journals (Sweden)

    Azizul Khakim

    2015-10-01

    Full Text Available ABSTRAK ANALISIS KESELAMATAN TERMOHIDROLIK BULK SHIELDING REAKTOR KARTINI. Bulk shielding merupakan fasilitas yang terintegrasi dengan reaktor Kartini yang berfungsi sebagai penyimpanan sementara bahan bakar bekas. Fasilitas ini merupakan fasilitas yang termasuk dalam struktur, sistem dan komponen (SSK yang penting bagi keselamatan. Salah satu fungsi keselamatan dari sistem penanganan dan penyimpanan bahan bakar adalah mencegah kecelakaan kekritisan yang tak terkendali dan membatasi naiknya temperatur bahan bakar. Analisis keselamatan paling kurang harus mencakup analisis keselamatan dari sisi neutronik dan termo hidrolik Bulk shielding. Analisis termo hidrolik ditujukan untuk memastikan perpindahan panas dan proses pendinginan bahan bakar bekas berjalan baik dan tidak terjadi akumulasi panas yang mengancam integritas bahan bakar. Code tervalidasi PARET/ANL digunakan untuk analisis pendinginan dengan mode konveksi alam. Hasil perhitungan menunjukkan bahwa mode pendinginan konvekasi alam cukup memadai dalam mendinginkan panas sisa tanpa mengakibatkan kenaikan temperatur bahan bakar yang signifikan. Kata kunci: Bulk shielding, bahan bakar bekas, konveksi alam, PARET.   ABSTRACT THERMAL HYDRAULIC SAFETY ANALYSIS OF BULK SHIELDING KARTINI REACTOR. Bulk shielding is an integrated facility to Kartini reactor which is used for temporary spent fuels storage. The facility is one of the structures, systems and components (SSCs important to safety. Among the safety functions of fuel handling and storage are to prevent any uncontrolable criticality accidents and to limit the fuel temperature increase. Safety analyses should, at least, cover neutronic and thermal hydraulic calculations of the bulk shielding. Thermal hydraulic analyses were intended to ensure that heat removal and the process of the spent fuels cooling takes place adequately and no heat accumulation that challenges the fuel integrity. Validated code, PARET/ANL was used for analysing the

  11. Radiation shielding for fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Laboratory, Tokyo (Japan)

    2000-03-01

    Radiation shielding aspects relating fission reactors have been reviewed. Domestic activities in the past five years have been mainly described concerning nuclear data, calculation methods, shielding and skyshine experiments, Advanced Boiling Water Reactor (ABWR), Advanced Pressurized Water Reactor (APWR), High Temperature Engineering Test Reactor (HTTR), Experimental and Prototype Fast Reactors (JOYO, MONJU), Demonstration FBR, core shroud replacement of BWR, and spent fuel transportation cask and vessel. These studies have valuable information in safety and cost reduction issues of fission reactor design for not only existing reactors but also new reactor concepts in the next century. It has been concluded that we should maintain existing shielding technologies and improve these data and methods for coming generations in the next millennium. (author)

  12. Shielding effectiveness of superconductive particles in plastics

    International Nuclear Information System (INIS)

    Pienkowski, T.; Kincaid, J.; Lanagan, M.T.; Poeppel, R.B.; Dusek, J.T.; Shi, D.; Goretta, K.C.

    1988-09-01

    The ability to cool superconductors with liquid nitrogen instead of liquid helium has opened the door to a wide range of research. The well known Meissner effect, which states superconductors are perfectly diamagnetic, suggests shielding applications. One of the drawbacks to the new ceramic superconductors is the brittleness of the finished material. Because of this drawback, any application which required flexibility (e.g., wire and cable) would be impractical. Therefore, this paper presents the results of a preliminary investigation into the shielding effectiveness of YBa 2 Cu 3 O/sub 7-x/ both as a composite and as a monolithic material. Shielding effectiveness was measured using two separate test methods. One tested the magnetic (near field) shielding, and the other tested the electromagnetic (far field) shielding. No shielding was seen in the near field measurements on the composite samples, and only one heavily loaded sample showed some shielding in the far field. The monolithic samples showed a large amount of magnetic shielding. 5 refs., 5 figs

  13. Shielding modification and safety review on the nuclear ship 'Mutsu'

    International Nuclear Information System (INIS)

    Osanai, Masao

    1978-01-01

    The Japan Atomic Energy Commission (JAEC) called on the Japan Nuclear Ship Development Agency (JNSDA) for shielding modification and safety review on the nuclear ship ''Mutsu'', and JNSDA has conducted the research and development (R and D) to meet the request of JAEC for the above two items. Concerning the shield modification, the following matters are described: the study on the cause of radiation leakage which was concluded to the fast neutron streaming, the conceptual design for this modification, the mock up experiment for shielding utilizing JRR-4, the basic design following on the conceptual design, including the detailed drawings of the modified construction and the shielding analysis using RADHEAT-V3 code, and the relating experiments such as the heat transfer test of the primary shielding structure and the test of strength in stranding. As for the safety review, the survey of the troubles and the technical problems having been experienced in the light water reactor plants of land use, for example, fuel integrity, stress corrosion cracking and the leakage of steam generator tubes, the revision of the design so as to adapt to current safety standards and regulations, for example, in-service inspection, the setting of additional leak detectors in the primary cooling system, the modification of emergeney filters, etc., and the review of the design and construction corresponding to recent R and D works, such as re-evaluation of the core design, cooling capability of natural circulation, thermal stress analysis of main pipings, and the evaluation of ECCS performance are presented . (Nakai, Y.)

  14. Shielding wall for thermonuclear device

    International Nuclear Information System (INIS)

    Uchida, Takaho.

    1989-01-01

    This invention concerns shielding walls opposing to plasmas of a thermonuclear device and it is an object thereof to conduct reactor operation with no troubles even if a portion of shielding wall tiles should be damaged. That is, the shielding wall tiles are constituted as a dual layer structure in which the lower base tiles are connected by means of bolts to first walls. Further, the upper surface tiles are bolt-connected to the layer base tiles. In this structure, the plasma thermal loads are directly received by the surface layer tiles and heat is conducted by means of conduction and radiation to the underlying base tiles and the first walls. Even upon occurrence of destruction accidents to the surface layer tiles caused by incident heat or electromagnetic force upon elimination of plasmas, since the underlying base tiles remain as they are, the first walls constituted with stainless steels, etc. are not directly exposed to the plasmas. Accordingly, the integrity of the first walls having cooling channels can be maintained and sputtering intrusion of atoms of high atom number into the plasmas can be prevented. (I.S.)

  15. General-Purpose Heat Source development: Safety Verification Test Program. Bullet/fragment test series

    Energy Technology Data Exchange (ETDEWEB)

    George, T.G.; Tate, R.E.; Axler, K.M.

    1985-05-01

    The radioisotope thermoelectric generator (RTG) that will provide power for space missions contains 18 General-Purpose Heat Source (GPHS) modules. Each module contains four /sup 238/PuO/sub 2/-fueled clads and generates 250 W/sub (t)/. Because a launch-pad or post-launch explosion is always possible, we need to determine the ability of GPHS fueled clads within a module to survive fragment impact. The bullet/fragment test series, part of the Safety Verification Test Plan, was designed to provide information on clad response to impact by a compact, high-energy, aluminum-alloy fragment and to establish a threshold value of fragment energy required to breach the iridium cladding. Test results show that a velocity of 555 m/s (1820 ft/s) with an 18-g bullet is at or near the threshold value of fragment velocity that will cause a clad breach. Results also show that an exothermic Ir/Al reaction occurs if aluminum and hot iridium are in contact, a contact that is possible and most damaging to the clad within a narrow velocity range. The observed reactions between the iridium and the aluminum were studied in the laboratory and are reported in the Appendix.

  16. Testing the responses of four wheat crop models to heat stress at anthesis and grain filling.

    Science.gov (United States)

    Liu, Bing; Asseng, Senthold; Liu, Leilei; Tang, Liang; Cao, Weixing; Zhu, Yan

    2016-05-01

    Higher temperatures caused by future climate change will bring more frequent heat stress events and pose an increasing risk to global wheat production. Crop models have been widely used to simulate future crop productivity but are rarely tested with observed heat stress experimental datasets. Four wheat models (DSSAT-CERES-Wheat, DSSAT-Nwheat, APSIM-Wheat, and WheatGrow) were evaluated with 4 years of environment-controlled phytotron experimental datasets with two wheat cultivars under heat stress at anthesis and grain filling stages. Heat stress at anthesis reduced observed grain numbers per unit area and individual grain size, while heat stress during grain filling mainly decreased the size of the individual grains. The observed impact of heat stress on grain filling duration, total aboveground biomass, grain yield, and grain protein concentration (GPC) varied depending on cultivar and accumulated heat stress. For every unit increase of heat degree days (HDD, degree days over 30 °C), grain filling duration was reduced by 0.30-0.60%, total aboveground biomass was reduced by 0.37-0.43%, and grain yield was reduced by 1.0-1.6%, but GPC was increased by 0.50% for cv Yangmai16 and 0.80% for cv Xumai30. The tested crop simulation models could reproduce some of the observed reductions in grain filling duration, final total aboveground biomass, and grain yield, as well as the observed increase in GPC due to heat stress. Most of the crop models tended to reproduce heat stress impacts better during grain filling than at anthesis. Some of the tested models require improvements in the response to heat stress during grain filling, but all models need improvements in simulating heat stress effects on grain set during anthesis. The observed significant genetic variability in the response of wheat to heat stress needs to be considered through cultivar parameters in future simulation studies. © 2016 John Wiley & Sons Ltd.

  17. Heat loss and fluid leakage tests of the ROSA-III facility

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Tasaka, Kanji; Shiba, Masayoshi

    1981-12-01

    The report presents characteristic test results about the steady state heat loss, one of the inherent characteristics of the ROSA-III test facility. The steady state heat loss tests were conducted at five different temperature conditions between 111 0 C and 290 0 C . Net heat loss rates were obtained by estimating the electric power supplied to the core, heat input from the recirculation pumps and steam leakage rate. The heat loss characteristics have important contribution to analyses of the ROSA-III small break tests. A following simple relation was obtained between the net heat loss rate Q*sub(HL) (kJ/s) (*: radical) of the ROSA-III facility and the temperature difference ΔT ( 0 C) between the fluid temperature of the system and the room temperature, Q*sub(HL) = 0.56 x ΔT. (*: radical) And the steam leak flow at normal operating condition of the ROSA-III test, (P = 7.2 MPa) was obtained as 8.9 x 10 -3 kg/s and corresponding steam leakage energy as 10.5 kJ/s. The heat input from the recirculation pumps was indirectly estimated under a constant speed by assuming the heat input was equal to the brake horce power of the pumps. (author)

  18. Radiation shielding calculations for the vista spacecraft

    International Nuclear Information System (INIS)

    Sahin, Suemer; Sahin, Haci Mehmet; Acir, Adem

    2005-01-01

    The VISTA spacecraft design concept has been proposed for manned or heavy cargo deep space missions beyond earth orbit with inertial fusion energy propulsion. Rocket propulsion is provided by fusion power deposited in the inertial confined fuel pellet debris and with the help of a magnetic nozzle. The calculations for the radiation shielding have been revised under the fact that the highest jet efficiency of the vehicle could be attained only if the propelling plasma would have a narrow temperature distribution. The shield mass could be reduced from 600 tons in the original design to 62 tons. Natural and enriched lithium were the principle shielding materials. The allowable nuclear heating in the superconducting magnet coils (up to 5 mW/cm 3 ) is taken as the crucial criterion for dimensioning the radiation shielding structure of the spacecraft. The space craft mass is 6000 tons. Total peak nuclear power density in the coils is calculated as ∼5.0 mW/cm 3 for a fusion power output of 17 500 MW. The peak neutron heating density is ∼2.0 mW/cm 3 , and the peak γ-ray heating density is ∼3.0 mW/cm 3 (on different points) using natural lithium in the shielding. However, the volume averaged heat generation in the coils is much lower, namely 0.21, 0.71 and 0.92 mW/cm 3 for the neutron, γ-ray and total nuclear heating, respectively. The coil heating will be slightly lower if highly enriched 6 Li (90%) is used instead of natural lithium. Peak values are then calculated as 2.05, 2.15 and 4.2 mW/cm 3 for the neutron, γ-ray and total nuclear heating, respectively. The corresponding volume averaged heat generation in the coils became 0.19, 0.58 and 0.77 mW/cm 3

  19. 3-D analysis on arbitrarily-shaped ICRF antennas and Faraday shields

    International Nuclear Information System (INIS)

    Chen, G.L.; Whealton, J.H.; Baity, F.W.; Hoffman, D.J.; Owens, T.L.

    1986-01-01

    Cavity antennas with Faraday shields are proposed to couple ion cyclotron radio frequency power for heating fusion plasmas. This application requires small, high-power, low-frequency antennas. The results are presented of a theoretical study of the ICRF antennas being developed for this purpose at the Radio Frequency Test Facility (RFTF). The objectives of this work are to optimize experimental designs and to confirm test results

  20. MFTF-α + T shield design

    International Nuclear Information System (INIS)

    Gohar, Y.

    1985-01-01

    MFTF-α+T is a DT upgrade option of the Tandem Mirror Fusion Test Facility (MFTF-B) to study better plasma performance, and test tritium breeding blankets in an actual fusion reactor environment. The central cell insert, designated DT axicell, has a 2-MW/m 2 neutron wall loading at the first wall for blanket testing. This upgrade is completely shielded to protect the reactor components, the workers, and the general public from the radiation environment during operation and after shutdown. The shield design for this upgrade is the subject of this paper including the design criteria and the tradeoff studies to reduce the shield cost

  1. ICRF antenna Faraday shield plasma sheath model

    International Nuclear Information System (INIS)

    Whealton, J.H.; Ryan, P.M.; Raridon, R.J.

    1990-01-01

    A two-dimensional nonlinear formulation that explicitly considers the plasma edge near a Faraday shield in a self-consistent manner is used in the modeling of the ion motion for a Faraday shield concept and model suggested by Perkins. Two models are considered that may provide significant insight into the generation of impurities for ion cyclotron resonance heating (ICRH) antennas. In one of these models a significant sheath periodically forms next to the Faraday screen, with ion acoustic waves heating the ions in the plasma. (orig.)

  2. Estimation of temperature distribution in a reactor shield

    International Nuclear Information System (INIS)

    Agarwal, R.A.; Goverdhan, P.; Gupta, S.K.

    1989-01-01

    Shielding is provided in a nuclear reactor to absorb the radiations emanating from the core. The energy of these radiations appear in the form of heat. Concrete which is commonly used as a shielding material in nuclear power plants must be able to withstand the temperatures and temperature gradients appearing in the shield due to this heat. High temperatures lead to dehydration of the concrete and in turn reduce the shielding effectiveness of the material. Adequate cooling needs to be provided in these shields in order to limit the maximum temperature. This paper describes a method to estimate steady state and transient temperature distribution in reactor shields. The results due to loss of coolant in the coolant tubes have been studied and presented in the paper. (author). 5 figs

  3. Technical products for radiation shielding. Shield assembled from lead blocks for radiation protection. General technical requirements

    International Nuclear Information System (INIS)

    1981-01-01

    The object of this standard description is the general technological requirements of 50 and 100 mm thick radiation protection shields assembled from lead blocks. The standard contains the definitions, types, parameters and dimensions of shields, their technical and acceptance criteria with testing methods, tagging, packaging, transportation and storage requirements, producer's liability. Some illustrated assembling examples, preferred parameters and dosimetry methods for shield inspection are given. (R.P.)

  4. High heat flux test of tungsten brazed mock-ups developed for KSTAR divertor

    Energy Technology Data Exchange (ETDEWEB)

    Song, J.H. [National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, K.M., E-mail: kyungmin@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Hong, S.H.; Kim, H.T.; Park, S.H.; Park, H.K.; Ahn, H.J. [National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, S.K.; Lee, D.W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    The tungsten (W) brazed flat type mock-up which consists of W, OFHC-Cu (oxygen-free high conductive copper) and CuCrZr alloy has been designed for KSTAR divertor in preparation for KSTAR upgrade with 17 MW heating power. For verification of the W brazed mock-up, the high heat flux test is performed at KoHLT-EB (Korea High Heat Load Test Facility-Electron Beam) in KAERI (Korea Atomic Energy Research Institute). Three mock-ups are tested for several thousand thermal cycles with absorbed heat flux up to 5 MW/m{sup 2} for 20 s duration. There is no evidence of the failure at the bonding joints of all mock-ups after HHF test. Finite element analysis (FEA) is performed to interpret the result of the test. As a result, it is considered that the local area in the water is in the subcooled boiling regime.

  5. Development of two tier test to assess conceptual understanding in heat and temperature

    Science.gov (United States)

    Winarti; Cari; Suparmi; Sunarno, Widha; Istiyono, Edi

    2017-01-01

    Heat and temperature is a concept that has been learnt from primary school to undergraduate levels. One problem about heat and temperature is that they are presented abstractly, theoretical concept. A student conceptual frameworks develop from their daily experiences. The purpose of this research was to develop a two-tier test of heat and temperature concept and measure conceptual understanding of heat and temperature of the student. This study consist of two method is qualitative and quantitative method. The two-tier test was developed using procedures defined by Borg and Gall. The two-tier test consisted of 20 question and was tested for 137 students for collecting data. The result of the study showed that the two-tier test was effective in determining the students’ conceptual understanding and also it might be used as an alternative for assessment and evaluation of students’ achievement

  6. FELIX: construction and testing of a facility to study electromagnetic effects for first wall, blanket, and shield systems

    International Nuclear Information System (INIS)

    Praeg, W.F.; Turner, L.R.; Biggs, J.A.; Knott, M.J.; Lari, R.J.; McGhee, D.G.; Wehrle, R.B.

    1983-01-01

    An experimental test facility for the study of electromagnetic effects in the FWBS systems of fusion reactors has been constructed over the past 1-1/2 years at Argonne National Laboratory (ANL). In a test volume of 0.76 m 3 a vertical pulsed 0.5 T dipole field (B < 50 T/s) is perpendicular to a 1 T solenoid field. Power supplies of 2.75 MW and 5.5 MW and a solid state switch rated 13 kV, 13.1 kA (170 MW) control the pulsed magnetic fields. The total stored energy in the coils is 2.13 MJ. The coils are designed for a future upgrade to 4 T or the solenoid and 1 T for the dipole field (a total of 23.7 MJ). This paper describes the design and construction features of the facility. These include the power supplies, the solid state switches, winding and impregnation of large dipole saddle coils, control of the magnetic forces, computer control of FELIX and of experimental data acquisition and analysis, and an initial experimental test setup to analyze the eddy current distribution in a flat disk

  7. FELIX: Construction and testing of a facility to study electromagnetic effects for First Wall, Blanket, and Shield systems

    International Nuclear Information System (INIS)

    Praeg, W.F.; Biggs, J.; Knott, M.J.; Lari, R.J.; McGhee, D.G.; Turner, L.R.; Wehrle, R.

    1983-01-01

    An experimental test facility for the study of electromagnetic effects in the FWBS systems of fusion reactors has been constructed over the past 2-1/2 years at Argonne National Laboratory (ANL). In a test volume of 0.76 m 3 a vertical pulsed 0.5 T dipole field (B < 50 T/s) is perpendicular to a 1 T solenoid field. Power supplies of 2.75 MW and 5.5 MW and a solid state switch rated 13 kV, 13.1 kA (170 MW) control the pulsed magnetic fields. The total stored energy in the coils is 2.13 MJ. The coils are designed for a future upgrade to 4 T for the solenoid and 1 T for the dipole field (a total of 23.7 MJ). This paper describes the design and construction features of the facility. These include the power supplies, the solid state switches, winding and impregnation of large dipole saddle coils, control of the magnetic forces, computer control of FELIX and of experimental data acquisition and analysis, and an initial experimental test setup to analyze the eddy current distribution in a flat disk

  8. Results from evaporation tests to support the MWTF heat removal system design

    International Nuclear Information System (INIS)

    Crea, B.A.

    1994-01-01

    An experimental tests program was conducted to measure the evaporative heat removal from the surface of a tank of simulated waste. The results contained in this report constitute definition design data for the latest heat removal function of the MWTF primary ventilation system

  9. 78 FR 63410 - Energy Conservation Program for Consumer Products: Test Procedures for Direct Heating Equipment...

    Science.gov (United States)

    2013-10-24

    ... test procedures for direct heating equipment and pool heaters established under the Energy Policy and... U.S.C. 6293(e)(2)) The current energy conservation standards for direct heating equipment and pool... DEPARTMENT OF ENERGY 10 CFR Part 430 [Docket Number EERE-2013-BT-TP-0004] RIN 1904-AC94 Energy...

  10. Calorimeter measures high nuclear heating rates and their gradients across a reactor test hole

    Science.gov (United States)

    Burwell, D.; Coombe, J. R.; Mc Bride, J.

    1970-01-01

    Pedestal-type calorimeter measures gamma-ray heating rates from 0.5 to 7.0 watts per gram of aluminum. Nuclear heating rate is a function of cylinder temperature change, measured by four chromel-alumel thermocouples attached to the calorimeter, and known thermoconductivity of the tested material.

  11. Development of radiation shielding standards in the American Nuclear Society

    International Nuclear Information System (INIS)

    Trubey, D.K.

    1975-11-01

    The American Nuclear Society (ANS) is a standards-writing organization-member of the American National Standards Institute (ANSI). The ANS Standards Committee has a subcommittee denoted ANS-6, Shielding, whose charge is to establish standards in connection with radiation protection and shielding, to provide shielding information to other standards writing groups, and to prepare recommended sets of shielding data and test problems. This paper is a progress report of this subcommittee

  12. An assessment of the lifetime of Faraday shield elements

    International Nuclear Information System (INIS)

    Caughman, J.B.O. II; Ruzic, D.N.; Hoffman, D.J.; Langley, R.A.; Lewis, M.B.; Ryan, P.M.

    1989-01-01

    The interaction of plasma with rf fields from an ion cyclotron range of frequencies (ICRF) antenna has been studied to estimate the amount of Faraday shield erosion expected in normal ICRF heating operation. Plasma parameters and ion energies have been measured in the near field of an antenna and used in a model to estimate the erosion rate of the Faraday shield surface. Experiments were conducted on the RF Test Facility, a magnetic mirror device at Oak Ridge National Laboratory, using a single-strap resonant loop antenna with a two-tier Faraday shield. The outer tier, facing the plasma, was layered with graphite tiles. The antenna was operated at currents and voltages within 50% of those expected in tokamaks. The time-varying floating potential was measured with a capacitively coupled probe, and the time-averaged floating potential, electron temperature, and electron density were measured with Langmuir probe. Both probes were scanned in front of the antenna. Ion energies were measured with a gridded energy analyzer located below the antenna, and samples of silicon were placed on the Faraday shield surface to estimate the incident ion energy. The capacitive probe measurement show that the rf floating potential follows the magnetic field pattern of the antenna, indicating that the electromagnetic fields are responsible for the potential formation. Electron temperatures increase with rf power and can reach values ≥60 eV for an rf power of ∼25 kW. Incident ion energies ≥300 eV have been measured for the same power level. Predictions of light ion sputtering of candidate Faraday shield materials are presented. 19 refs., 6 figs., 1 tab

  13. Heat and mass release for some transient fuel source fires: A test report

    International Nuclear Information System (INIS)

    Nowlen, S.P.

    1986-10-01

    Nine fire tests using five different trash fuel source packages were conducted by Sandia National Laboratories. This report presents the findings of these tests. Data reported includes heat and mass release rates, total heat and mass release, plume temperatures, and average fuel heat of combustion. These tests were conducted as a part of the US Nuclear Regulatory Commission sponsored fire safety research program. Data from these tests were intended for use in nuclear power plant probabilistic risk assessment fire analyses. The results were also used as input to a fire test program at Sandia investigating the vulnerability of electrical control cabinets to fire. The fuel packages tested were chosen to be representative of small to moderately sized transient trash fuel sources of the type that would be found in a nuclear power plant. The highest fire intensity encountered during these tests was 145 kW. Plume temperatures did not exceed 820 0 C

  14. Radiation shielding concrete

    International Nuclear Information System (INIS)

    Kunishima, Shigeru.

    1990-01-01

    The radiation shielding concretes comprise water, cement, fine aggregates consisting of serpentines and blown mist slags, coarse aggregates consisting of serpentines and kneading materials. Since serpentines containing a relatively great amount of water of crystallization in rocks as coarse aggregates and fine aggregates, the hydrogen content in the radiation shielding concretes is increased and the neutron shielding effect is improved. In addition, since serpentines are added as the fine aggregates and blown mists slags of a great specific gravity are used, the specific gravity of the shielding concretes is increased to improve the γ-ray shielding effect. Further, by the use of the kneading material having a water reducing effect and fluidizing effect, and by the bearing effect of the spherical blown mist slags used as the fine aggregates, concrete fluidity can be increased. Accordingly, workability of the radiation shielding concretes can be improved. (T.M.)

  15. Design, fabrication, installation and shielding integrity testing of source storage container for automatic source movement system used in TLD calibration facility

    International Nuclear Information System (INIS)

    Subramanian, V.; Baskar, S.; Annalakshmi, O.; Jose, M.T.; Jayshree, C.P.; Choudry, Shreelatha

    2012-01-01

    A state-of-art TLD laboratory has been commissioned in January 2000 at Radiological Safety Division of Indira Gandhi Centre for Atomic Research (IGCAR). The laboratory provides personnel monitoring service to 2000 occupational workers from Indira Gandhi Centre for Atomic Research and Bhabha Atomic Research Centre facilities. The laboratory has been accredited by the Radiation Safety Systems Division (RSSD), Bhabha Atomic Research Centre (BARC) since year 2002. The laboratory has exclusive facility for the calibration of the TLD cards. As apart of accreditation procedure and taking into account of geometry effect, the dose rate at the card position is determined by the accreditation authorities by using graphite chamber (secondary or national standard instrument) and often re estimated by a condenser R meter (M/s Victoreen, Germany) by our laboratory. As per the regulatory requirement, the exposure protocols should be automated. Towards this an automatic source movement system has been augmented in the calibration facility. By using the system, the source will be brought to the irradiation position by pneumatically and exposures will be terminated by counter, timer and triggering system. To accomplish this task a lead container has been designed, fabricated and mounted at the beneath of the calibration table for the storage of source. As per the automation process, a lead container for the source storage has been designed and installed beneath to the Calibration Table. The container was designed to hold a 3Ci 137 Cs source, but present activity of the source is 1.2Ci. Hence, the shielding integrity was tested with higher active source (1.7Ci 60 Co). The dose rate measured outside on the circumference of the container at the middle of the source is found to be the same as calculated using QAD CGGP calculations. The top plug is so designed to avoid inadvertent upward movement of the source. Though, the shielding was not adequate on top of the top plug, however it does

  16. Testing a Quantum Heat Pump with a Two-Level Spin

    Directory of Open Access Journals (Sweden)

    Luis A. Correa

    2016-04-01

    Full Text Available Once in its non-equilibrium steady state, a nanoscale system coupled to several heat baths may be thought of as a “quantum heat pump”. Depending on the direction of its stationary heat flows, it may function as, e.g., a refrigerator or a heat transformer. These continuous heat devices can be arbitrarily complex multipartite systems, and yet, their working principle is always the same: they are made up of several elementary three-level stages operating in parallel. As a result, it is possible to devise external “black-box” testing strategies to learn about their functionality and performance regardless of any internal details. In particular, one such heat pump can be tested by coupling a two-level spin to one of its “contact transitions”. The steady state of this external probe contains information about the presence of heat leaks and internal dissipation in the device and, also, about the direction of its steady-state heat currents. Provided that the irreversibility of the heat pump is low, one can further estimate its coefficient of performance. These techniques may find applications in the emerging field of quantum thermal engineering, as they facilitate the diagnosis and design optimization of complex thermodynamic cycles.

  17. He II Heat Exchanger Test Unit for the LHC Inner Triplet

    CERN Document Server

    Blanco-Viñuela, E; Huang, Y; Nicol, T H; Peterson, T; Van Weelderen, R

    2002-01-01

    The Inner Triplet Heat Exchanger Test Unit (IT-HXTU) is a 30-m long thermal model designed at Fermilab, built in US industry, fully automated and tested at CERN as part of the US LHC program to develop the LHC Interaction Region quadrupole system. The cooling scheme of the IT-HXTU is based on heat exchange between stagnant pressurized He II in the magnet cold mass and saturated He II (two-phase) flowing in a heat exchanger located outside of and parallel to the cold mass. The purposes of this test are, among others, to validate the proposed cooling scheme and to define an optimal control strategy to be implemented in the future LHC accelerator. This paper discusses the results for the heat exchanger test runs and emphasizes the thermal and hydraulic behavior of He II for the inner triplet cooling scheme.

  18. Accelerator shielding benchmark problems

    International Nuclear Information System (INIS)

    Hirayama, H.; Ban, S.; Nakamura, T.

    1993-01-01

    Accelerator shielding benchmark problems prepared by Working Group of Accelerator Shielding in the Research Committee on Radiation Behavior in the Atomic Energy Society of Japan were compiled by Radiation Safety Control Center of National Laboratory for High Energy Physics. Twenty-five accelerator shielding benchmark problems are presented for evaluating the calculational algorithm, the accuracy of computer codes and the nuclear data used in codes. (author)

  19. Vapor compression heat pump system field tests at the tech complex

    Science.gov (United States)

    Baxter, Van D.

    1985-11-01

    The Tennessee Energy Conservation In Housing (TECH) complex has been utilized since 1977 as a field test site for several novel and conventional heat pump systems for space conditioning and water heating. Systems tested include the Annual Cycle Energy System (ACES), solar assisted heat pumps (SAHP) both parallel and series, two conventional air-to-air heat pumps, an air-to-air heat pump with desuperheater water heater, and horizontal coil and multiple shallow vertical coil ground-coupled heat pumps (GCHP). A direct comparison of the measured annual performance of the test systems was not possible. However, a cursory examination revealed that the ACES had the best performance, however, its high cost makes it unlikely that it will achieve wide-spread use. Costs for the SAHP systems are similar to those of the ACES but their performance is not as good. Integration of water heating and space conditioning functions with a desuperheater yielded significant efficiency improvement at modest cost. The GCHP systems performed much better for heating than for cooling and may well be the most efficient alternative for residences in cold climates.

  20. Avery Island heater tests: measured data for 1000 days of heating

    International Nuclear Information System (INIS)

    Van Sambeek, L.L.; Stickney, R.G.; DeJong, K.B.

    1983-10-01

    Three heater tests were conducted in the Avery Island salt mine. The measurements of temperature and displacement, and the calculation of stress in the vicinity of each heater are of primary importance in the understanding of the thermal and thermomechanical response of the salt to an emplaced heat source. This report presents the temperature, displacement, and calculated stress data gathered during the heating phase of the three heater tests. The data presented have application in the ongoing studies of the response of geologicic media to an emplaced heat source. Specifically, electric heaters, which simulate canisters of heat-generating nuclear waste, were placed in the floor of the Avery Island salt mine, and measurements were made of the response of the salt caused by the heating. The purpose of this report is to transmit the data to the scientific community; rigorous analysis and interpretation of the data are considered beyond the scope of this data report. 11 references, 46 figures

  1. 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.

  2. Solar-energy heats a transportation test center--Pueblo, Colorado

    Science.gov (United States)

    1981-01-01

    Petroleum-base, thermal energy transport fluid circulating through 583 square feet of flat-plate solar collectors accumulates majority of energy for space heating and domestic hot-water of large Test Center. Report describes operation, maintenance, and performance of system which is suitable for warehouses and similar buildings. For test period from February 1979 to January 1980, solar-heating fraction was 31 percent, solar hot-water fraction 79 percent.

  3. Mechanical design of the TIBER breeding shield

    Energy Technology Data Exchange (ETDEWEB)

    Rathke, J.; Deutsch, L. (Grumman Corp., Bethpage, NY (USA). Space Systems Div.)

    1989-04-01

    TIBER features a segmented shield assembly that provides the nuclear shielding for the superconducting toroidal field coils. In addition to its primary function, the shield also provides tritium breeding through the use of water coolant that contains 16 wt% dissolved lithium nitrate. Because the TIBER reactor need not provide electrical power, the coolant is maintained at low pressure (0.2 MPa) and low temperature (75/sup 0/C). The shield is made in several segments to facilitate assembly and allow for replacement of high heat flux components (divertor blades). The segments are designated as inboard, outboard, upper, lower, and divertor modules. In total, there are 96 separate modules in the machine, consisting of six different types. The design features of the different modules vary primarily depending on the thickness of the shield in a given location. The very thick outboard shield has a breeding zone in the inboard portion of the module, with a shielding zone behind it. The breeding zone consists of a stainless steel casing filled with beryllium spheres. The shielding zone consists of the same casing filled with steel spheres. Both of these zones have lithiated water circulated throughout to provide cooling and breeding. In zones with minimal thickness, tungsten alloys are used to achieve the required shielding. These alloys are incoprorated in subassemblies utilizing stainless steel casings surrounding blocks of tungsten heavy metal alloy. These are infiltrated with lead on final assembly to form a thermally continuous panel. Several of these panels are then assembled into an outer stainless steel case to form an inboard module. These modules also use the lithiated coolant. The details of the design are presented and discussed. (orig.).

  4. Results of heating mode performance tests of a solar-assisted heat pump

    Science.gov (United States)

    Jones, C. B.; Smetana, F. O.

    1979-01-01

    The performance of a heat pump, utilizing 8.16 square meters of low-cost solar collectors as the evaporator in a Freon-114 refrigeration cycle, was determined under actual insolation conditions during the summer and fall of 1976. C.O.P.'s (coefficient of performance) greater than 3 were obtained with condensing temperatures around 78 C and evaporating temperatures around 27 C. Ambient temperatures were about 3 C above evaporating temperatures. Similar performance levels were obtained at other insolation and temperature conditions. Experience with the system has identified some component and system changes which should increase the obtainable C.O.P. to about 4.0. These are described along with the system's design rationale. The accumulated data are presented as an appendix.

  5. Intrinsically disordered proteins as molecular shields.

    Science.gov (United States)

    Chakrabortee, Sohini; Tripathi, Rashmi; Watson, Matthew; Schierle, Gabriele S Kaminski; Kurniawan, Davy P; Kaminski, Clemens F; Wise, Michael J; Tunnacliffe, Alan

    2012-01-01

    The broad family of LEA proteins are intrinsically disordered proteins (IDPs) with several potential roles in desiccation tolerance, or anhydrobiosis, one of which is to limit desiccation-induced aggregation of cellular proteins. We show here that this activity, termed molecular shield function, is distinct from that of a classical molecular chaperone, such as HSP70 - while HSP70 reduces aggregation of citrate synthase (CS) on heating, two LEA proteins, a nematode group 3 protein, AavLEA1, and a plant group 1 protein, Em, do not; conversely, the LEA proteins reduce CS aggregation on desiccation, while HSP70 lacks this ability. There are also differences in interaction with client proteins - HSP70 can be co-immunoprecipitated with a polyglutamine-containing client, consistent with tight complex formation, whereas the LEA proteins can not, although a loose interaction is observed by Förster resonance energy transfer. In a further exploration of molecular shield function, we demonstrate that synthetic polysaccharides, like LEA proteins, are able to reduce desiccation-induced aggregation of a water-soluble proteome, consistent with a steric interference model of anti-aggregation activity. If molecular shields operate by reducing intermolecular cohesion rates, they should not protect against intramolecular protein damage. This was tested using the monomeric red fluorescent protein, mCherry, which does not undergo aggregation on drying, but the absorbance and emission spectra of its intrinsic fluorophore are dramatically reduced, indicative of intramolecular conformational changes. As expected, these changes are not prevented by AavLEA1, except for a slight protection at high molar ratios, and an AavLEA1-mCherry fusion protein is damaged to the same extent as mCherry alone. A recent hypothesis proposed that proteomes from desiccation-tolerant species contain a higher degree of disorder than intolerant examples, and that this might provide greater intrinsic stability

  6. Accelerated life tests of specimen heat pipe from Communication Technology Satellite (CTS) project

    Science.gov (United States)

    Tower, L. K.; Kaufman, W. B.

    1977-01-01

    A gas-loaded variable conductance heat pipe of stainless steel with methanol working fluid identical to one now on the CTS satellite was life tested in the laboratory at accelerated conditions for 14 200 hours, equivalent to about 70 000 hours at flight conditions. The noncondensible gas inventory increased about 20 percent over the original charge. The observed gas increase is estimated to increase operating temperature by about 2.2 C, insufficient to harm the electronic gear cooled by the heat pipes in the satellite. Tests of maximum heat input against evaporator elevation agree well with the manufacturer's predictions.

  7. Recent High Heat Flux Tests on W-Rod-Armored Mockups

    International Nuclear Information System (INIS)

    Nygren, Richard E.; Youchison, Dennis L.; McDonald, Jimmie M.; Lutz, Thomas J.; Miszkiel, Mark E.

    2000-01-01

    In the authors initial high heat flux tests on small mockups armored with W rods, done in the small electron beam facility (EBTS) at Sandia National Laboratories, the mockups exhibited excellent thermal performance. However, to reach high heat fluxes, they reduced the heated area to only a portion (approximately25%) of the sample. They have now begun tests in their larger electron beam facility, EB 1200, where the available power (1.2 MW) is more than enough to heat the entire surface area of the small mockups. The initial results indicate that, at a given power, the surface temperatures of rods in the EB 1200 tests is somewhat higher than was observed in the EBTS tests. Also, it appears that one mockup (PW-10) has higher surface temperatures than other mockups with similar height (10mm) W rods, and that the previously reported values of absorbed heat flux on this mockup were too high. In the tests in EB 1200 of a second mockup, PW-4, absorbed heat fluxes of approximately22MW/m 2 were reached but the corresponding surface temperatures were somewhat higher than in EBTS. A further conclusion is that the simple 1-D model initially used in evaluating some of the results from the EBTS testing was not adequate, and 3-D thermal modeling will be needed to interpret the results

  8. Time-varying Entry Heating Profile Replication with a Rotating Arc Jet Test Article

    Science.gov (United States)

    Grinstead, Jay Henderson; Venkatapathy, Ethiraj; Noyes, Eric A.; Mach, Jeffrey J.; Empey, Daniel M.; White, Todd R.

    2014-01-01

    A new approach for arc jet testing of thermal protection materials at conditions approximating the time-varying conditions of atmospheric entry was developed and demonstrated. The approach relies upon the spatial variation of heat flux and pressure over a cylindrical test model. By slowly rotating a cylindrical arc jet test model during exposure to an arc jet stream, each point on the test model will experience constantly changing applied heat flux. The predicted temporal profile of heat flux at a point on a vehicle can be replicated by rotating the cylinder at a prescribed speed and direction. An electromechanical test model mechanism was designed, built, and operated during an arc jet test to demonstrate the technique.

  9. A short-term rating method for heat pump heating systems; phase 5: test of the fault diagnosis systems; Kurztestmethode fuer Waermepumpenanlagen; Phase 5: Test der Fehlerdiagnosesysteme

    Energy Technology Data Exchange (ETDEWEB)

    Zogg, D.; Esfandiar, S.

    2001-07-01

    This final report for the Swiss Federal Office of Energy (SFOE) describes the testing phase of a project that developed systems for the operational monitoring and optimisation of heat pump installations along with a diagnosis system for faults. The heat pump is considered as a sub-system. The report describes two monitoring systems and a simulation model that are used to monitor the state of the heat pump both during commissioning as well as during operation. The aim is also to detect faults as early as possible during the whole of the operational life of the installation. A state-orientated approach is propagated as being cheaper than fixed service intervals or repairing after breakdown and standstill. The development of the two monitoring systems called 'HeatWatch' and 'FuzzyWatch' is described. The effort needed for the parametrisation and training of these systems is discussed. The testing of the systems on two test beds using real-life measured values for a single-family home and further simulation data is described and the results listed. The authors state that the monitoring systems can also be used for refrigeration and air-conditioning systems.

  10. Operational tests of a device with an elevator for automated outlet of heat for heating

    Energy Technology Data Exchange (ETDEWEB)

    Zinger, N.M.; Belevich, A.I.; Burd, A.L.; Krivitskiy, V.I.; Lokshin, L.S.

    1983-01-01

    The results are presented of operational tests of series produced jet pumps. The field of use of an automated elevator is determined and recommendations for increasing its operational qualities are given.

  11. Proposed heat transfer model for the gas-liquid heat transfer effects observed in the Stanford Research Institute scaled tests

    International Nuclear Information System (INIS)

    Corradini, M.; Sonin, A.A.; Todreas, N.

    1976-12-01

    In 1971-72, the Stanford Research Institute conducted a series of scaled experiments which simulated a sodium-vapor expansion in a hypothetical core disruptive accident (HCDA) for the Fast Flux Test Facility. A non-condensible explosive source was used to model the pressure-volume expansion characteristics of sodium vapor as predicted by computer code calculations. Rigid piston-cylinder experiments ( 1 / 10 and 1 / 30 scale) were undertaken to determine these expansion characteristics. The results showed that the pressure-volume characteristics depend significantly on the presence of water in the cylinder reducing the work output by about 50 percent when a sufficient water depth was present. The study presented proposes that the mechanism of heat transfer between the water and high temperature gas was due to area enhancement by Taylor instabilities at the gas-liquid interface. A simple heat transfer model is proposed which describes this energy transport process and agrees well with the experimental data from both scaled experiments. The consequences of this analysis suggest that an estimate of the heat transfer to the cold slug during a full-scale HCDA due to sodium vapor expansion and the accompanying reduction in mechanical work energy warrants further investigation. The implication of this analysis is that for either sodium or fuel vapor expansion in an HCDA, there is an inherent heat transfer mechanism which significantly reduces the work output of the expanding bubble

  12. ESBWR related passive decay heat removal tests in PANDA

    International Nuclear Information System (INIS)

    Huggenberger, M.; Aubert, C.; Bandurski, T.; Dreier, J.; Fischer, O.; Strassberger, H.J.; Yadigaroglu, G.

    1999-01-01

    A number of test series to investigate passive safety systems for the next generation of Light Water Reactors have been performed in the PANDA multi-purpose facility at the Paul Scherrer Institut (PSI). The large scale thermal-hydraulic test facility allows to investigate LWR containment phenomena and system behaviour. PANDA was first used to examine the Passive Containment Cooling System (PCCS) for the Simplified Boiling Water Reactor (SBWR). In 1996 new test series were initiated; all related to projects of the EC Fourth Framework Programme on Nuclear Fission Safety. One of these projects (TEPSS) is focused on the European Simplified Boiling Water Reactor (ESBWR). The ESBWR containment features and PCCS long-term post LOCA response were investigated in PANDA. The PCCS start-up was demonstrated, the effect of nitrogen hidden somewhere in the drywell and released later in the transient was simulated and the effect of light gases (helium) on the PCCS performance was investigated. Finally, the influence of low PCC pool levels on PCCS and containment performance was examined. The main findings were that the PCCS works as intended and shows generally a favorable and robust long-term post LOCA behaviour. The system starts working even under extreme conditions and trapped air released from the drywell later in the transient does only temporarily reduce the PCCS performance. The new PANDA test series provided an extensive data base which will contribute to further improve containment design of passive plants and allow for system code assessment in a wide parameter range. (author)

  13. INTOR radiation shielding for personnel access

    International Nuclear Information System (INIS)

    Gohar, Y.; Abdou, M.

    1981-01-01

    The INTOR reactor shield system consists of the blanket, bulk shield, penetration shield, component shield, and biological shield. The bulk shield consists of two parts: (a) the inboard shield; and (b) the outboard shield. The distinction between the different components of the shield system is essential to satisfy the different design constraints and achieve various objectives

  14. Sodium heat pipe module test for the SAFE-30 reactor prototype

    International Nuclear Information System (INIS)

    Reid, Robert S.; Sena, J. Tom; Martinez, Adam L.

    2001-01-01

    Reliable, long-life, low-cost heat pipes can enable safe, affordable space fission power and propulsion systems. Advanced versions of these systems can in turn allow rapid access to any point in the solar system. Twelve stainless steel-sodium heat pipe modules were built and tested at Los Alamos for use in a non-nuclear thermohydraulic simulation of the SAFE-30 reactor (Poston et al., 2000). SAFE-30 is a near-term, low-cost space fission system demonstration. The heat pipes were designed to remove thermal power from the SAFE-30 core, and transfer this power to an electrical power conversion system. These heat pipe modules were delivered to NASA Marshall Space Flight Center in August 2000 and were assembled and tested in a prototypical configuration during September and October 2000. The construction and test of one of the SAFE-30 modules is described

  15. Transient testing of the FFTF for decay-heat removal by natural convection

    International Nuclear Information System (INIS)

    Beaver, T.R.; Johnson, H.G.; Stover, R.L.

    1982-06-01

    This paper reports on the series of transient tests performed in the FFTF as a major part of the pre-operations testing program. The structure of the transient test program was designed to verify the capability of the FFTF to safely remove decay heat by natural convection. The series culminated in a scram from full power to complete natural convection in the plant, simulating a loss of all electrical power. Test results and acceptance criteria related to the verification of safe decay heat removal are presented

  16. Testing of a cryogenic recooler heat exchanger at Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    Nicoletti, A.; Wu, K.C.

    1993-01-01

    Brookhaven National Laboratory has tested a recooler heat exchanger intended to be used in the cryogenic system of the Relativistic Heavy Ion Collider. The unit is required to transfer 225 Watts from a supercritical helium stream flowing at 100 g/s to a helium bath boiling at 4.25 K. Measurements made with heat loads of 50 to over 450 Watts on the unit indicate its cooling capacity is greater than 400 Watts, as expected, and it will be suitable for use in the RHIC ring. Presented are the modifications made to BNL's MAGCOOL test facility that were necessary for testing, test procedure, and recooler performance

  17. BUGJEFF311.BOLIB (JEFF-3.1.1) and BUGENDF70.BOLIB (ENDF/B-VII.0) - Generation Methodology and Preliminary Testing of two ENEA-Bologna Group Cross Section Libraries for LWR Shielding and Pressure Vessel Dosimetry

    Science.gov (United States)

    Pescarini, Massimo; Sinitsa, Valentin; Orsi, Roberto; Frisoni, Manuela

    2016-02-01

    Two broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format, dedicated to LWR shielding and pressure vessel dosimetry applications, were generated following the methodology recommended by the US ANSI/ANS-6.1.2-1999 (R2009) standard. These libraries, named BUGJEFF311.BOLIB and BUGENDF70.BOLIB, are respectively based on JEFF-3.1.1 and ENDF/B-VII.0 nuclear data and adopt the same broad-group energy structure (47 n + 20 γ) of the ORNL BUGLE-96 similar library. They were respectively obtained from the ENEA-Bologna VITJEFF311.BOLIB and VITENDF70.BOLIB libraries in AMPX format for nuclear fission applications through problem-dependent cross section collapsing with the ENEA-Bologna 2007 revision of the ORNL SCAMPI nuclear data processing system. Both previous libraries are based on the Bondarenko self-shielding factor method and have the same AMPX format and fine-group energy structure (199 n + 42 γ) as the ORNL VITAMIN-B6 similar library from which BUGLE-96 was obtained at ORNL. A synthesis of a preliminary validation of the cited BUGLE-type libraries, performed through 3D fixed source transport calculations with the ORNL TORT-3.2 SN code, is included. The calculations were dedicated to the PCA-Replica 12/13 and VENUS-3 engineering neutron shielding benchmark experiments, specifically conceived to test the accuracy of nuclear data and transport codes in LWR shielding and radiation damage analyses.

  18. Nuclear code case development of printed-circuit heat exchangers with thermal and mechanical performance testing

    Energy Technology Data Exchange (ETDEWEB)

    Aakre, Shaun R. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Mechanical Engineering; Jentz, Ian W. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Mechanical Engineering; Anderson, Mark H. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Mechanical Engineering

    2018-03-27

    The U.S. Department of Energy has agreed to fund a three-year integrated research project to close technical gaps involved with compact heat exchangers to be used in nuclear applications. This paper introduces the goals of the project, the research institutions, and industrial partners working in collaboration to develop a draft Boiler and Pressure Vessel Code Case for this technology. Heat exchanger testing, as well as non-destructive and destructive evaluation, will be performed by researchers across the country to understand the performance of compact heat exchangers. Testing will be performed using coolants and conditions proposed for Gen IV Reactor designs. Preliminary observations of the mechanical failure mechanisms of the heat exchangers using destructive and non-destructive methods is presented. Unit-cell finite element models assembled to help predict the mechanical behavior of these high-temperature components are discussed as well. Performance testing methodology is laid out in this paper along with preliminary modeling results, an introduction to x-ray and neutron inspection techniques, and results from a recent pressurization test of a printed-circuit heat exchanger. The operational and quality assurance knowledge gained from these models and validation tests will be useful to developers of supercritical CO2 systems, which commonly employ printed-circuit heat exchangers.

  19. Thermal performance tests on a sodium-to-sodium heat exchanger

    International Nuclear Information System (INIS)

    Prahlad, B.; Kale, R.D.; Rajan, K.K.

    1990-01-01

    Thermal performance of a 3 MW sodium-to-sodium intermediate heat exchanger (IHX) was evaluated under temperature conditions typical of a Fast Breeder Reactor IHX. A regenerative figure of eight loop was used with the heat exchanger at the cross over point, and a 500 kW heat source and an air cooled sink to maintain the desired test conditions. The overall heat transfer coefficient was found to vary from 4.02 to 4.87 kW/m 2 ·K for Peclet numbers varying from 37 to 112.5 on the shell side and 44.4 to 133.5 on the tube side respectively. The Peclet numbers were representative of low turbulent regime in this case. While the overall heat transfer coefficient was found close to predictions using Lubarsky's correlation, it was somewhat lower than that predicted by later correlations of Spukunsky and Borishansky. The reasons for the lower overall heat transfer coefficient have been explained in terms of possible maldistribution of shell side flow in low turbulent regime reducing the effective heat transfer area and increased thermal contact resistance. Based on their findings the authors feel that heat transfer in a sodium-to-sodium heat exchanger at low Peclet numbers is expected to differ from that obtained with large Peclet numbers. (author)

  20. Thermal response testing of precast pile heat exchangers: fieldwork report

    DEFF Research Database (Denmark)

    Alberdi Pagola, Maria; Poulsen, Søren Erbs; Jensen, Rasmus Lund

    Centrum Pæle A/S, Aalborg University, VIA University College and INSERO Horsens are partners in the industrial PhD project: “Experimental and numerical characterisation of the thermo-mechanical behaviour of quadratic cross section energy piles”. This document aims to present the fieldwork...... undertaken in the project at two test sites in Denmark: one in Horsens and one in Vejle. The tasks have been carried out between January 2014 and February 2016....

  1. Thermal response testing of precast pile heat exchangers

    DEFF Research Database (Denmark)

    Pagola, Maria Alberdi; Poulsen, Søren Erbs; Jensen, Rasmus Lund

    Centrum Pæle A/S, Aalborg University, VIA University College and INSERO Horsens are partners in the industrial PhD project: “Experimental and numerical characterisation of the thermo-mechanical behaviour of quadratic cross section energy piles”. This document aims to present the fieldwork...... undertaken in the project at two test sites in Denmark: one in Horsens and one in Vejle. The tasks have been carried out between January 2014 and February 2016....

  2. Design and qualification testing of a strontium-90 fluoride heat source

    International Nuclear Information System (INIS)

    Fullam, H.T.

    1981-12-01

    The Strontium Heat Source Development Program began at the Pacific Northwest Laboratory (PNL) in 1972 and is scheduled to be completed by the end of FY-1981. The program is currently funded by the US Department of Energy (DOE) By-Product Utilization Program. The primary objective of the program has been to develop the data and technology required to permit the licensing of power systems for terrestrial applications that utilize 90 SrF 2 -fueled radioisotope heat sources. A secondary objective of the program has been to design and qualification-test a general purpose 90 SrF 2 -fueled heat source. The effort expended in the design and testing of the heat source is described. Detailed information is included on: heat source design, licensing requirements, and qualification test requirements; the qualification test procedures; and the fabrication and testing of capsules of various materials. The results obtained in the qualification tests show that the outer capsule design proposed for the 90 SrF 2 heat source is capable of meeting current licensing requirements when Hastelloy S is used as the outer capsule material. The data also indicate that an outer capsule of Hastelloy C-4 would probably also meet licensing requirements, although Hastelloy S is the preferred material. Therefore, based on the results of this study, the general purpose 90 SrF 2 heat source will consist of a standard WESF Hastelloy C-276 inner capsule filled with 90 SrF 2 and a Hastelloy S outer capsule having a 2.375-in. inner diameter and 0.500-in. wall thickness. The end closures for this study, the general purpose 90 SrF 2 heat a Hastelloy S outer capsule having a 2.375-in. inner diameter and 0.500-in. wall thickness. The end closures for the outer capsule will utilize an interlocking joint design requiring a 0.1-in. penetration closure weld

  3. A new test procedure to evaluate the performance of substations for collective heating systems

    Science.gov (United States)

    Baetens, Robin; Verhaert, Ivan

    2017-11-01

    The overall heat demand of a single dwelling, existing out of space heating and domestic hot water production, decreases due to higher insulation rates. Because of this, investing in efficient and renewable heat generation becomes less interesting. Therefore, to incorporate renewables or residual heat on a larger scale, district heating or collective heating systems grow in importance. Within this set-up, the substation is responsible for the interaction between local demand for comfort and overall energy performance of the collective heating system. Many different configurations of substations exist, which influence both local comfort and central system performance. Next to that, also hybrids exist with additional local energy input. To evaluate performance of such substations, a new experimental-based test procedure is developed in order to evaluate these different aspects, characterized by the two roles a substation has, namely as heat generator and as heat consumer. The advantage of this approach is that an objective comparison between individual and central systems regarding performance on delivering local comfort can be executed experimentally. The lab set-up consists out of three different subsystems, namely the central system, the domestic hot water consumption and the local space heating. The central system can work with different temperature regimes and control strategies, as these aspects have proven to have the largest influence on actual performance. The domestic hot water system is able to generate similar tap profiles according to eco-design regulation for domestic hot water generation. The space heating system is able to demand a modular heat load.

  4. A new test procedure to evaluate the performance of substations for collective heating systems

    Directory of Open Access Journals (Sweden)

    Baetens Robin

    2017-01-01

    Full Text Available The overall heat demand of a single dwelling, existing out of space heating and domestic hot water production, decreases due to higher insulation rates. Because of this, investing in efficient and renewable heat generation becomes less interesting. Therefore, to incorporate renewables or residual heat on a larger scale, district heating or collective heating systems grow in importance. Within this set-up, the substation is responsible for the interaction between local demand for comfort and overall energy performance of the collective heating system. Many different configurations of substations exist, which influence both local comfort and central system performance. Next to that, also hybrids exist with additional local energy input. To evaluate performance of such substations, a new experimental-based test procedure is developed in order to evaluate these different aspects, characterized by the two roles a substation has, namely as heat generator and as heat consumer. The advantage of this approach is that an objective comparison between individual and central systems regarding performance on delivering local comfort can be executed experimentally. The lab set-up consists out of three different subsystems, namely the central system, the domestic hot water consumption and the local space heating. The central system can work with different temperature regimes and control strategies, as these aspects have proven to have the largest influence on actual performance. The domestic hot water system is able to generate similar tap profiles according to eco-design regulation for domestic hot water generation. The space heating system is able to demand a modular heat load.

  5. Standard Test Method for Measuring Heat Flux Using Flush-Mounted Insert Temperature-Gradient Gages

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This test method describes the measurement of the net heat flux normal to a surface using gages inserted flush with the surface. The geometry is the same as heat-flux gages covered by Test Method E 511, but the measurement principle is different. The gages covered by this standard all use a measurement of the temperature gradient normal to the surface to determine the heat that is exchanged to or from the surface. Although in a majority of cases the net heat flux is to the surface, the gages operate by the same principles for heat transfer in either direction. 1.2 This general test method is quite broad in its field of application, size and construction. Two different gage types that are commercially available are described in detail in later sections as examples. A summary of common heat-flux gages is given by Diller (1). Applications include both radiation and convection heat transfer. The gages used for aerospace applications are generally small (0.155 to 1.27 cm diameter), have a fast time response ...

  6. A multi-functional testing instrument for heat assisted magnetic recording media

    International Nuclear Information System (INIS)

    Yang, H. Z.; Chen, Y. J.; Leong, S. H.; An, C. W.; Ye, K. D.; Hu, J. F.; Yin, M. J.

    2014-01-01

    With recent developments in heat assisted magnetic recording (HAMR), characterization of HAMR media is becoming very important. We present a multi-functional instrument for testing HAMR media, which integrates HAMR writing, reading, and a micro-magneto-optic Kerr effect (μ-MOKE) testing function. A potential application of the present instrument is to make temperature dependent magnetic property measurement using a pump-probe configuration. In the measurement, the media is heated up by a heating (intense) beam while a testing (weak) beam is overlapped with the heating beam for MOKE measurement. By heating the media with different heating beam power, magnetic measurements by MOKE at different temperatures can be performed. Compared to traditional existing tools such as the vibrating sample magnetometer, the present instrument provides localized and efficient heating at the measurement spot. The integration of HAMR writing and μ-MOKE system can also facilitate a localized full investigation of the magnetic media by potential correlation of HAMR head independent write/read performance to localized magnetic properties

  7. APL/JHU free flight tests of the General Purpose Heat Source module. Testing: 5-7 March 1984

    International Nuclear Information System (INIS)

    Baker, W.M. II.

    1984-01-01

    Purpose of the test was to obtain statistical information on the dynamics of the General Purpose Heat Source (GPHS) module at terminal speeds. Models were designed to aerodynamically and dynamically represent the GPHS module. Normal and high speed photographic coverage documented the motion of the models. This report documents test parameters and techniques for the free-spin tests. It does not include data analysis

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

    Science.gov (United States)

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

    2016-09-01

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

  9. Shielded regenerative neutron detector

    International Nuclear Information System (INIS)

    Terhune, J.H.; Neissel, J.P.

    1978-01-01

    An ion chamber type neutron detector is disclosed which has a greatly extended lifespan. The detector includes a fission chamber containing a mixture of active and breeding material and a neutron shielding material. The breeding and shielding materials are selected to have similar or substantially matching neutron capture cross-sections so that their individual effects on increased detector life are mutually enhanced

  10. Improving the fidelity of electrically heated nuclear systems testing using simulated neutronic feedback

    International Nuclear Information System (INIS)

    Bragg-Sitton, Shannon M.; Godfroy, Thomas J.; Webster, Kenny

    2010-01-01

    Nonnuclear test platforms and methodologies can be employed to reduce the overall cost, risk and complexity of testing nuclear systems while allowing one to evaluate the operation of an integrated nuclear system within a reasonable timeframe, providing valuable input to the overall system design. In a nonnuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard electric test techniques allow one to fully assess thermal, heat transfer, and stress related attributes of a given system, but these approaches fail to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and testing with nuclear fuel elements installed. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. This paper summarizes the results of initial system dynamic response testing for two electrically heated reactor concepts: a heat pipe-cooled reactor simulator with integrated heat exchanger and a gas-cooled reactor simulator with integrated Brayton power conversion system. Initial applications apply a simplified reactor kinetics model with either a single or an averaged measured state point. Preliminary results demonstrate the applicability of the dynamic test methodology to any reactor type, elucidating the variation in system response characteristics in different reactor concepts. These results suggest a need to further enhance the dynamic test approach by incorporating a more accurate model of the reactor dynamics and improved hardware instrumentation for better state estimation in application of the

  11. Radiation shielding plate

    International Nuclear Information System (INIS)

    Kobayashi, Torakichi; Sugawara, Takeo.

    1983-01-01

    Purpose: To reduce the weight and stabilize the configuration of a radiation shielding plate which is used in close contact with an object to be irradiated with radiation rays. Constitution: The radiation shielding plate comprises a substrate made of lead glass and a metallic lead coating on the surface of the substrate by means of plating, vapor deposition or the like. Apertures for permeating radiation rays are formed to the radiation shielding plate. Since the shielding plate is based on a lead glass plate, a sufficient mechanical strength can be obtained with a thinner structure as compared with the conventional plate made of metallic lead. Accordingly, if the shielding plate is disposed on a soft object to be irradiated with radiation rays, the object and the plate itself less deform to obtain a radiation irradiation pattern with distinct edges. (Moriyama, K.)

  12. Critical heat flux tests for a 12 finned-element assembly

    Energy Technology Data Exchange (ETDEWEB)

    Yang, J., E-mail: Jun.Yang@cnl.ca; Groeneveld, D.C.; Yuan, L.Q.

    2017-03-15

    Highlights: • CHF tests for a 12 finned-fuel-element assembly at highly subcooled conditions. • Test approach to maximize experimental information and minimize heater failures. • Three series of tests were completed in vertical upward light water flow. • Bundle simulators of two axial power profiles and three heated lengths were tested. • Results confirm that the prediction method predicts lower CHF values than measured. - Abstract: An experimental study was undertaken to provide relevant data to validate the current critical heat flux (CHF) prediction method of the NRU driver fuel for safety analysis, i.e., to confirm no CHF occurrence below the predicted values. The NRU driver fuel assembly consists of twelve finned fuel elements arranged in two rings – three in the inner ring and nine in the outer ring. To satisfy the experimental objective tests at very high heat fluxes, very high mass velocities, and high subcoolings were conducted where the CHF mechanism is the departure from nucleate boiling (DNB). Such a CHF experiment can be very difficult, costly and time consuming since failure of the heating surface due to rupture or melting (physical burnout) is expected when the DNB type of CHF is reached. A novel experimental approach has been developed to maximize the amount of relevant experimental information on safe operating conditions in the tests, and to minimize any possible heater failures that inherently accompany the CHF occurrence at these conditions. Three series of tests using electrically heated NRU driver fuel simulators with three heated lengths and two axial power profiles (or axial heat flux distribution (AFD)) were completed in vertical upward light water flow. Each series of tests covered two mass flow rates, several heat flux levels, and local subcoolings that bound the ranges of interest for the analysis of postulated slow loss-of-regulation accident (LORA) and loss-of-flow accident (LOFA) scenarios. Tests for each mass flow rate of

  13. Selenide isotope generator for the Galileo Mission. Axially-grooved heat pipe: accelerated life test results

    International Nuclear Information System (INIS)

    1979-08-01

    The results through SIG/Galileo contract close-out of accelerated life testing performed from June 1978 to June 1979 on axially-grooved, copper/water heat pipes are presented. The primary objective of the test was to determine the expected lifetime of axially-grooved copper/water heat pipes. The heat pipe failure rate, due to either a leak or a build-up of non-condensible gas, was determined. The secondary objective of the test was to determine the effects of time and temperature on the thermal performance parameters relevant to long-term (> 50,000 h) operation on a space power generator. The results showed that the gas generation rate appears to be constant with time after an initial sharp rise although there are indications that it drops to approximately zero beyond approx. 2000 h. During the life test, the following pipe-hours were accumulated: 159,000 at 125 0 C, 54,000 at 165 0 C, 48,000 at 185 0 C, and 8500 at 225 0 C. Heated hours per pipe ranged from 1000 to 7500 with an average of 4720. Applying calculated acceleration factors yields the equivalent of 930,000 pipe-h at 125 0 C. Including the accelerated hours on vendor tested pipes raises this number to 1,430,000 pipe-hours at 125 0 C. It was concluded that, for a heat pipe temperature of 125 0 C and a mission time of 50,000 h, the demonstrated heat pipe reliability is between 80% (based on 159,000 actual pipe-h at 125 0 C) and 98% (based on 1,430,000 accelerated pipe-h at 125 0 C). Measurements indicate some degradation of heat transfer with time, but no detectable degradation of heat transport

  14. Numerical analysis of thermal response tests with a groundwater flow and heat transfer model

    Energy Technology Data Exchange (ETDEWEB)

    Raymond, J.; Therrien, R. [Departement de Geologie et de Genie Ggeologique, Universite Laval, 1065 avenue de la medecine, Quebec (Qc) G1V 0A6 (Canada); Gosselin, L. [Departement de Genie Mecanique, Universite Laval, 1065 avenue de la medecine, Quebec (Qc) G1V 0A6 (Canada); Lefebvre, R. [Institut National de la Recherche Scientifique, Centre Eau Terre Environnement, 490 de la Couronne, Quebec (Qc) G1K 9A9 (Canada)

    2011-01-15

    The Kelvin line-source equation, used to analyze thermal response tests, describes conductive heat transfer in a homogeneous medium with a constant temperature at infinite boundaries. The equation is based on assumptions that are valid for most ground-coupled heat pump environments with the exception of geological settings where there is significant groundwater flow, heterogeneous distribution of subsurface properties, a high geothermal gradient or significant atmospheric temperature variations. To address these specific cases, an alternative method to analyze thermal response tests was developed. The method consists in estimating parameters by reproducing the output temperature signal recorded during a test with a numerical groundwater flow and heat transfer model. The input temperature signal is specified at the entrance of the ground heat exchanger, where flow and heat transfer are computed in 2D planes representing piping and whose contributions are added to the 3D porous medium. Results obtained with this method are compared to those of the line-source model for a test performed under standard conditions. A second test conducted in waste rock at the South Dump of the Doyon Mine, where conditions deviate from the line-source assumptions, is analyzed with the numerical model. The numerical model improves the representation of the physical processes involved during a thermal response test compared to the line-source equation, without a significant increase in computational time. (author)

  15. Ohmically heated toroidal experiment (OHTE) mobile ignition test reactor facility concept study

    International Nuclear Information System (INIS)

    Masson, L.S.; Watts, K.D.; Piscitella, R.R.; Sekot, J.P.; Drexler, R.L.

    1983-02-01

    This report presents the results of a study to evaluate the use of an existing nuclear test complex at the Idaho National Engineering Laboratory (INEL) for the assembly, testing, and remote maintenance of the ohmically heated toroidal experiment (OHTE) compact reactor. The portable reactor concept is described and its application to OHTE testing and maintenance requirements is developed. Pertinent INEL facilities are described and several test system configurations that apply to these facilities are developed and evaluated

  16. Validation of a new 39 neutron group self-shielded library based on the nucleonics analysis of the Lotus fusion-fission hybrid test facility performed with the Monte Carlo code

    International Nuclear Information System (INIS)

    Pelloni, S.; Cheng, E.T.

    1985-02-01

    The Swiss LOTUS fusion-fission hybrid test facility was used to investigate the influence of the self-shielding of resonance cross sections on the tritium breeding and on the thorium ratios. Nucleonic analyses were performed using the discrete-ordinates transport codes ANISN and ONEDANT, the surface-flux code SURCU, and the version 3 of the MCNP code for the Li 2 CO 3 and the Li 2 O blanket designs with lead, thorium and beryllium multipliers. Except for the MCNP calculation which bases on the ENDF/B-V files, all nuclear data are generated from the ENDF/B-IV basic library. For the deterministic methods three NJOY group libraries were considered. The first, a 39 neutron group self-shielded library, was generated at EIR. The second bases on the same group structure as the first does and consists of infinitely diluted cross sections. Finally the third library was processed at LANL and consists of coupled 30+12 neutron and gamma groups; these cross sections are not self-shielded. The Monte Carlo analysis bases on a continuous and on a discrete 262 group library from the ENDF/B-V evaluation. It is shown that the results agree well within 3% between the unshielded libraries and between the different transport codes and theories. The self-shielding of resonance cross sections results in a decrease of the thorium capture rate and in an increase of the tritium breeding of about 6%. The remaining computed ratios are not affected by the self-shielding of cross sections. (Auth.)

  17. Experiences with tungsten coatings in high heat flux tests and under plasma load in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Herrmann, A; Greuner, H; Fuchs, J C; Marne, P de; Neu, R

    2009-01-01

    ASDEX Upgrade was operated with about 6400 s plasma discharge during the scientific program in 2007/2008 exploring tungsten as a first wall material in tokamaks. In the first phase, the heating power was restricted to 10 MW. It was increased to 15 MW in the second phase. During this operational period, a delamination of the 200 μm W-VPS coating happened at 2 out of 128 tiles of the outer divertor and an unscheduled opening was required. In the third phase, ASDEX Upgrade was operated with partly predamaged tiles and up to 15 MW heating power. The target load was actively controlled by N 2 -seeding. This paper presents the screening test of target tiles in the high heat flux test facility GLADIS, experiences with operation and detected damages of the outer divertor as well as the heat load to the outer divertor and the reasons for the toroidal asymmetry of the divertor load.

  18. Models of the heat dynamics of solar collectors for performance testing

    DEFF Research Database (Denmark)

    Bacher, Peder; Madsen, Henrik; Perers, Bengt

    2011-01-01

    accurate estimates of parameters in physical models. The applied method is described by Kristensen et al. (2004) and implemented in the software CTSM1. Examples of successful applications of the method includes modelling the of the heat dynamics of integrated photo-voltaic modules (Friling et al., 2009......) and modelling of the heat dynamics of buildings (Madsen and Holst, 1995). Measurements obtained at a test site in Denmark during the spring 2010 are used for the modelling. The tested collector is a single glazed large area flat plate collector with selective absorber and Teflon anti convection layer. The test...

  19. ANS shielding standards for light-water reactors

    International Nuclear Information System (INIS)

    Trubey, D.K.

    1982-01-01

    The purpose of the American Nuclear Society Standards Subcommittee, ANS-6, Radiation Protection and Shielding, is to develop standards for radiation protection and shield design, to provide shielding information to other standards-writing groups, and to develop standard reference shielding data and test problems. A total of seven published ANS-6 standards are now current. Additional projects of the subcommittee, now composed of nine working groups, include: standard reference data for multigroup cross sections, gamma-ray absorption coefficients and buildup factors, additional benchwork problems for shielding problems and energy spectrum unfolding, power plant zoning design for normal and accident conditions, process radiation monitors, and design for postaccident radiological conditions

  20. Laboratory test of a prototype heat storage module based on stable supercooling of sodium acetate trihydrate

    DEFF Research Database (Denmark)

    Dannemand, Mark; Kong, Weiqiang; Fan, Jianhua

    2015-01-01

    Laboratory test of a long term heat storage module utilizing the principle of stable supercooling of 199.5 kg of sodium acetate water mixture has been carried out. Avoiding phase separation of the incongruently melting salt hydrate by using the extra water principle increased the heat storage...... capacity. An external expansion vessel minimized the pressure built up in the module while heating and reduced the risk of instable supercooling. The module was stable supercooled at indoor ambient temperature for up to two months after which it was discharged. The energy discharged after activating...

  1. Testing of a Stirling engine for heat + power cogeneration; Test eines Stirlingmotors zur Kraft-Waerme-Kopplung

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, M.; Heinen, J. [RWE Energy AG, Essen (Germany)

    2007-01-15

    As part of a technology evaluation of distributed energy generators, RWE Energy AG extensively tested a micro combined heat and power appliance, powered by a Stirling engine developed by the British firm Microgen Energy Limited. Microgen Energy Limited is a specialist in micro combined heat and power (microCHP) based on unique Free-Piston Stirling generator technology Microgen is working with leading appliance manufacturers to integrate its core technology into a range of innovative microCHP products. The investigations concentrated on the determination of capacity, efficiency and emissions, the grid connection and behaviour at start-up and under varying loads. This article summarises the results of the tests and gives an overview of micro-CHP technologies (CHP=combined heat and power) and their possible significance to the market in the future. (orig.)

  2. Results of high heat flux tests of tungsten divertor targets under plasma heat loads expected in ITER and tokamaks (review)

    Energy Technology Data Exchange (ETDEWEB)

    Budaev, V. P., E-mail: budaev@mail.ru [National Research Centre Kurchatov Institute (Russian Federation)

    2016-12-15

    Heat loads on the tungsten divertor targets in the ITER and the tokamak power reactors reach ~10MW m{sup −2} in the steady state of DT discharges, increasing to ~0.6–3.5 GW m{sup −2} under disruptions and ELMs. The results of high heat flux tests (HHFTs) of tungsten under such transient plasma heat loads are reviewed in the paper. The main attention is paid to description of the surface microstructure, recrystallization, and the morphology of the cracks on the target. Effects of melting, cracking of tungsten, drop erosion of the surface, and formation of corrugated and porous layers are observed. Production of submicron-sized tungsten dust and the effects of the inhomogeneous surface of tungsten on the plasma–wall interaction are discussed. In conclusion, the necessity of further HHFTs and investigations of the durability of tungsten under high pulsed plasma loads on the ITER divertor plates, including disruptions and ELMs, is stressed.

  3. Testing and Failure Mechanisms of Ice Phase Change Material Heat Exchangers

    Science.gov (United States)

    Leimkuehler, Thomas O.; Stephan, Ryan A.; Hawkins-Reynolds, Ebony

    2011-01-01

    Phase change materials (PCM) may be useful for thermal control systems that involve cyclical heat loads or cyclical thermal environments such as specific spacecraft orientations in Low Earth Orbit (LEO) and low beta angle Low Lunar Orbit (LLO). Thermal energy can be stored in the PCM during peak heat loads or in adverse thermal environments. The stored thermal energy can then be released later during minimum heat loads or in more favorable thermal environments. One advantage that PCM s have over evaporators in this scenario is that they do not use a consumable. The use of water as a PCM rather than the more traditional paraffin wax has the potential for significant mass reduction since the latent heat of formation of water is approximately 70% greater than that of wax. One of the potential drawbacks of using ice as a PCM is its potential to rupture its container as water expands upon freezing. In order to develop a space qualified ice PCM heat exchanger, failure mechanisms must first be understood. Therefore, a methodical experimental investigation has been undertaken to demonstrate and document specific failure mechanisms due to ice expansion in the PCM. A number of ice PCM heat exchangers were fabricated and tested. Additionally, methods for controlling void location in order to reduce the risk of damage due to ice expansion were investigated. This paper presents the results of testing that occurred from March through September of 2010 and builds on testing that occurred during the previous year.

  4. Separate effects tests for GOTHIC condensation and evaporative heat transfer models

    International Nuclear Information System (INIS)

    George, T.L.; Singh, A.

    1994-01-01

    The GOTHIC computer program, under development at EPRI/NAI, is a general purpose thermal hydraulics computer program for design, licensing, safety and operating analysis of nuclear containments and other confinement buildings. The code solves a nine equation model for three dimensional multiphase flow with separate mass, momentum and energy equations for vapor, liquid and drop phases. The vapor phase can be a gas mixture of steam and non-condensing gases. The phase balance equations are coupled by mechanistic and empirical models for interface mass, energy and momentum transfer that cover the entire flow regime from bubbly flow to film/drop flow. A variety of heat transfer correlations are available to model the fluid coupling to active and passive solid conductors. This paper focuses on the application of GOTHIC to two separate effects tests; condensation heat transfer on a vertical flat plate with varying bulk velocity, steam concentration and temperature, and evaporative heat transfer from a hot pool to a dry (superheated) atmosphere. Comparisons with experimental data is included for both tests. Results show the validity of two condensation heat transfer correlations as incorporated into GOTHIC and the interfacial heat and mass transfer models for the range of the experimental test conditions. Comparisons are also made for lumped versus multidimensional modeling for buoyancy controlled flow with evaporative heat transfer. (author). 13 refs., 1 tab., 10 figs

  5. Orion Exploration Flight Test Reaction Control System Jet Interaction Heating Environment from Flight Data

    Science.gov (United States)

    White, Molly E.; Hyatt, Andrew J.

    2016-01-01

    The Orion Multi-Purpose Crew Vehicle (MPCV) Reaction Control System (RCS) is critical to guide the vehicle along the desired trajectory during re-­-entry. However, this system has a significant impact on the convective heating environment to the spacecraft. Heating augmentation from the jet interaction (JI) drives thermal protection system (TPS) material selection and thickness requirements for the spacecraft. This paper describes the heating environment from the RCS on the afterbody of the Orion MPCV during Orion's first flight test, Exploration Flight Test 1 (EFT-1). These jet plumes interact with the wake of the crew capsule and cause an increase in the convective heating environment. Not only is there widespread influence from the jet banks, there may also be very localized effects. The firing history during EFT-1 will be summarized to assess which jet bank interaction was measured during flight. Heating augmentation factors derived from the reconstructed flight data will be presented. Furthermore, flight instrumentation across the afterbody provides the highest spatial resolution of the region of influence of the individual jet banks of any spacecraft yet flown. This distribution of heating augmentation across the afterbody will be derived from the flight data. Additionally, trends with possible correlating parameters will be investigated to assist future designs and ground testing programs. Finally, the challenges of measuring JI, applying this data to future flights and lessons learned will be discussed.

  6. Separate effects tests for GOTHIC condensation and evaporative heat transfer models

    International Nuclear Information System (INIS)

    George, T.L.; Singh, A.

    1996-01-01

    The GOTHIC computer program, under development at NAI for EPRI, is a general purpose thermal hydraulics computer program for design, licensing, safety and operating analysis of nuclear containments and other confinement buildings. The code solves a nine-equation model for three-dimensional multiphase flow with separate mass, momentum and energy equations for vapor, liquid and drop phases. The vapor phase can be a gas mixture of steam and non-condensing gases. The phase balance equations are coupled by mechanistic and empirical models for interface mass, energy and momentum transfer that cover the entire flow regime from bubbly flow to film-drop flow. A variety of heat transfer correlations are available to model the fluid coupling to active and passive solid conductors. This paper focuses on the application of GOTHIC to two separate effects tests: condensation heat transfer on a vertical flat plate with varying bulk velocity, steam concentration and temperature, and evaporative heat transfer from a hot pool to a dry (superheated) atmosphere. Comparisons with experimental data are included for both tests. Results show the validity of two condensation heat transfer correlations as incorporated into GOTHIC and the interfacial heat and mass transfer models for the range of the experimental test conditions. Comparisons are also made for lumped vs. multidimensional modeling for buoyancy-controlled flow with evaporative heat transfer. (orig.)

  7. HPGe detector shielding adjustment

    International Nuclear Information System (INIS)

    Trnkova, L.; Rulik, P.

    2008-01-01

    Low-level background shielding of HPGe detectors is used mainly for environmental samples with very low content of radionuclides. National Radiation Protection Institute (SURO) in Prague is equipped with 14 HPGe detectors with relative efficiency up to 150%. The detectors are placed in a room built from materials with low content of natural radionuclides and equipped with a double isolation of the floor against radon. Detectors themselves are placed in lead or steel shielding. Steel shielding with one of these detectors with relative efficiency of 100% was chosen to be rebuilt to achieve lower minimum detectable activity (MDA). Additional lead and copper shielding was built up inside the original steel shielding to reduce the volume of the inner space and filled with nitrogen by means of evaporating liquid nitrogen. The additional lead and copper shielding, consequent reduction of the inner volume and supply of evaporated nitrogen, caused a decrease of the background count and accordingly MDA values as well. The effect of nitrogen evaporation on the net areas of peaks belonging to radon daughters is significant. The enhanced shielding adjustment has the biggest influence in low energy range, what can be seen in collected data. MDA values in energy range from 30 keV to 400 keV decreased to 0.65-0.85 of original value, in energy range from 400 keV to 2 MeV they fell to 0.70-0.97 of original value. (authors)

  8. Shielding in experimental areas

    International Nuclear Information System (INIS)

    Stevens, A.; Tarnopolsky, G.; Thorndike, A.; White, S.

    1979-01-01

    The amount of shielding necessary to protect experimental detectors from various sources of background radiation is discussed. As illustrated an experiment has line of sight to sources extending approx. 90 m upstream from the intersection point. Packing a significant fraction of this space with shielding blocks would in general be unacceptable because primary access to the ring tunnel is from the experimental halls. (1) From basic machine design considerations and the inherent necessity to protect superconducting magnets it is expected that experimental areas in general will be cleaner than at any existing accelerator. (2) Even so, it will likely be necessary to have some shielding blocks available to protect experimental apparatus, and it may well be necessary to have a large amount of shielding available in the WAH. (3) Scraping will likely have some influence on all halls, and retractable apparatus may sometimes be necessary. (4) If access to any tunnel is needed to replace a magnet, one has 96 h (4 days) available to move shielding away to permit access without additional downtime. This (the amount of shielding one can shuffle about in 96 h) is a reasonable upper limit to shielding necessary in a hall

  9. High heat flux tests of mock-ups for ITER divertor application

    International Nuclear Information System (INIS)

    Giniatulin, R.; Gervash, A.; Komarov, V.L.; Makhankov, A.; Mazul, I.; Litunovsky, N.; Yablokov, N.

    1998-01-01

    One of the most difficult tasks in fusion reactor development is the designing, fabrication and high heat flux testing of actively cooled plasma facing components (PFCs). At present, for the ITER divertor project it is necessary to design and test components by using mock-ups which reflect the real design and fabrication technology. The cause of failure of the PFCs is likely to be through thermo-cycling of the surface with heat loads in the range 1-15 MW m -2 . Beryllium, tungsten and graphite are considered as the most suitable armour materials for the ITER divertor application. This work presents the results of the tests carried out with divertor mock-ups clad with beryllium and tungsten armour materials. The tests were carried out in an electron beam facility. The results of high heat flux screening tests and thermo-cycling tests in the heat load range 1-9 MW m -2 are presented along with the results of metallographic analysis carried out after the tests. (orig.)

  10. The application of SHIELD-HIT12A computer code to calculate of absorption dose for in vitro and in vivo test in BNCT

    International Nuclear Information System (INIS)

    Yohannes Sardjono; Hamidatul Faqqiyyah; Niels Bassler

    2014-01-01

    The projection of world population growth and increased longevity are leading to a rapid increase in the total number of middle-aged and older adults, with a corresponding increase in the number of deaths caused by non communicable diseases. It is projected that the annual number of deaths due to cardiovascular disease will increase from 17 million in 2008 to 25 million in 2030 with annual cancer deaths increasing from 7.6 million to 13 million. Boron Neutron Capture Therapy is a therapy that utilizes the absorption interaction of Boron-10 with thermal neutron and become He-4 particle and located in cell target and very short half life gamma emission. Studies were carried out to dose distribution in HER-2+ breast cancer therapy by Boron Neutron Capture Therapy (BNCT) using SHIELD Heavy Ion Therapy (HIT12A) T program. The Monte Carlo particle transport code SHIELD-HIT1 is designed to precisely simulate therapeutic beams of protons and ions in biological tissue relevant for ion beam cancer therapy. SHIELD-HIT (Heavy Ion Therapy) evolved from the common SHIELD code that models interactions of hadrons and atomic nuclei in complex extended targets in the energy range up to 1 TeV/nucleon. Through this computer code can be applied to calculate of absorption dose in cell target. (author)

  11. The ITER EC H&CD upper launcher: Design, analysis and testing of a bolted joint for the Blanket Shield Module

    NARCIS (Netherlands)

    Gessner, R.; Aiello, G.; Grossetti, G.; Meier, A.; Ronden, D.; Spaeh, P.; Scherer, T.; Schreck, S.; Strauss, D.; Vaccaro, A.

    2013-01-01

    The final design of the structural system for the ITER EC H&CD upper launcher is in progress. Many design features of the preliminary design are under revision with the aim to achieve the built-to-print-status. This paper deals with design and analysis of a bolted joint for the Blanket Shield

  12. Design and testing of a magnetic shield for the Thomson scattering photomultiplier tubes in the stray fields of the ERASMUS tokamak

    International Nuclear Information System (INIS)

    Desoppere, E.; Van Oost, G.

    1983-01-01

    A multiple coaxial shield system has been designed for the photomultiplier tubes of the ERASMUS tokamak Thomson scattering diagnostic. A stray field of 75 x 10 -4 T was reduced to 0.01 x 10 -4 T for a field parallel to the tube axis, and to 0.03 x 10 -4 T for a perpendicular field

  13. Examination of W7-X target elements after high heat flux testing

    International Nuclear Information System (INIS)

    Missirlian, M.; Durocher, A.; Schlosser, J.; Greuner, H.; Schedler, B.

    2007-01-01

    Full text of publication follows: The target elements of Wendelstein 7-X (W7-X) divertor are designed to sustain a stationary heat flux of 10 MW/m 2 and to remove a maximum power load up to 100 kW. The plasma-facing material is made of CFC NB31 flat tiles bonded to a CuCrZr copper alloy water-cooled heat sink. Before launching the serial fabrication, pre-series activities aimed at qualifying the design, the manufacturing route, the relevant non-destructive examination (NDE) methods, and at defining the acceptance criteria for the serial production. High heat flux (HHF) testing is the central activity of this qualification phase and represents a fundamental tool to predict 'critical' defects assembling. Within the framework of this qualification activity, the reception tests performed in the transient infrared thermography test bed SATIR at CEA-Cadarache and HHF testing carried out in the ion beam facility GLADIS at IPP-Garching, exhibited some tiles with thermal inhomogeneities, which initiated and developed during high heat flux testing. Hence, studies were launched in order to better understand this behaviour during cyclic heat loading. This post testing examination was mainly focused on the interface between CFC flat tiles and CuCrZr heat sink to improve if necessary the current design. HHF thermal cycling tests at ∼10 MW/m 2 for 10 s pulse duration each, allowed to assess the performances of target elements and showed some tiles with hot spots close to the edge (stable or progressing). Finally, after the HHF experimental campaign, a comprehensive analysis of some tested elements was carried out by means of infrared thermography inspection SATIR and metallographic examinations. Afterwards correlations between the non destructive SATIR inspection, HHF testing GLADIS and metallographic observation were investigated to assess damage detection, to analyse defect propagation, and to adjust the acceptance criteria valuable for the serial production. This paper will

  14. Heat transfer test in a tube using CO2 at supercritical pressures

    International Nuclear Information System (INIS)

    Kim, Hwan Yeol; Kim, Hyungrae; Song, Jin Ho; Cho, Bong Hyun; Bae, Yoon Yeong

    2005-01-01

    Heat transfer test facility, which is named as SPHINX (Supercritical Pressure Heat Transfer Investigation for NeXt Generation), has been constructed in KAERI for the study of heat transfer and pressure drop characteristics in a single tube, single rod and rod bundle at supercritical CO 2 conditions. The tests with supercritical water are difficult it terms of cost and effort, since the critical pressure and temperature of water are as high as 22.12 MPa and 374.14degC. As a substitute for water, CO 2 is selected for the test since the critical pressure and temperature of CO 2 are 7.38 MPa and 31.05degC that are much lower than those of water. This paper describes the design characteristics of the SPHINX and the experimental investigations on the heat transfer and pressure drop of a vertical single tube with an inside diameter of 4.4 mm with upward flow of supercritical CO 2 . The geometry of the single tube is the same as that of Kyushu University test performed with Freon (R22) for the direct comparison of a medium effect. The tests were performed with various heat and mass fluxes at a given pressure. The range of mass flux is 400∼1200 kg/m 2 s and the heat flux is chosen up to 150 kW/m 2 . The selected pressure are 7.75, 8.12, and 8.85 MPa. The test results are investigated and compared with the previous tests. (author)

  15. Analysis of panthers full-scale heat transfer tests with RELAP5

    International Nuclear Information System (INIS)

    Parlatan, Y.; Boyer, B.D.; Jo, J.; Rohatgi, S.

    1996-01-01

    The RELAP5 code is being assessed on the full-scale Passive Containment Cooling System (PCCS) in the Performance ANalysis and Testing of HEat Removal Systems (PANTHERS) facility at Societa Informazioni Termoidrauliche (SIET) in Italy. PANTHERS is a test facility with fall-size prototype beat exchangers for the PCCS in support of the General Electric's (GE) Simplified Boiling Water Reactor (SBWR) program. PANTHERS tests with a low noncondensable gas concentration and with a high noncondensable gas concentration were analyzed with RELAP5. The results showed that beat transfer rate decreases significantly along the PCCS tubes. In the test case with a higher inlet noncondensable gas fraction, the PCCS removed 35% less heat than in the test case with the lower noncondensable gas fraction. The dominant resistance to the overall heat transfer is the condensation beat transfer resistance inside the tubes. This resistance increased by about 5-fold between the inlet and exit of the tube due to the build up of noncondensable gases along the tube. The RELAP5 calculations also predicted that 4% to 5% of the heat removed to the PCCS pool occurs in the inlet steam piping and PCCS upper and lower headers. These piping needs to be modeled for other tests systems. The full-scale PANTHERS predictions are also compared against 1/400 scale GIRAFFE tests. GIRAFFE has 33% larger heat surface area, but its efficiency is only 15% and 23% higher than PANTHERS for the two cases analyzed This was explained by the high heat transfer resistance inside the tubes near the exit

  16. Field testing of a ceramic heat exchanger for heat recovery application

    Science.gov (United States)

    Sohal, M. S.

    1988-06-01

    AiResearch Company, Torrance, California, developed a 5 MMBtu/hr ceramic-metallic hybrid High Temperature Burner-Duct-Recuperator (HTBDR) system to recover energy from hot (up to 2500 F), dirty, and corrosive glue gas streams and preheat combustion air up to 2000 F. To reduce the cost and size of the ceramic recuperator, ceramic tubes with internal cruciform baffles were developed. The HTBDR system was tested on a 20 MMBtu/hr rotary forging furnace for about 2000 hours. To facilitate tube replacements, final design configuration uses horizontally mounted tubes. A maximum air preheat temperature of about 1916 F was achieved with a flue gas temperature of 2122 F. This represents fuel savings of about 30 to 50 percent (depending upon the amount of excess air) compared with an unrecuperated furnace. The overall design and operation of the recuperator proved to be successful up to the time of material failure. X ray diffraction of some failed components indicated that there was some residual Silicon in the interior regions and complete nitriding did not occur during the fabrication process. Degradation of failed components was probably caused by oxidation of residual silicon and by the stresses caused due to different coefficient of thermal expansion of various compounds during thermal cycling. A combination of severe and numerous thermal cycling coupled with incomplete nitriding was the most likely cause of material failure.

  17. Detailed technical plan for Test Program Element-III (TPE-III) of the first wall/blanket shield engineering test program

    International Nuclear Information System (INIS)

    Turner, L.R.; Praeg, W.F.

    1982-03-01

    The experimental requirements, test-bed design, and computational requirements are reviewed and updated. Next, in Sections 3, 4 and 5, the experimental plan, instrumentation, and computer plan, respectively, are described. Finally, Section 6 treats other considerations, such as personnel, outside participation, and distribution of results

  18. Detailed technical plan for Test Program Element-III (TPE-III) of the first wall/blanket shield engineering test program

    Energy Technology Data Exchange (ETDEWEB)

    Turner, L.R.; Praeg, W.F.

    1982-03-01

    The experimental requirements, test-bed design, and computational requirements are reviewed and updated. Next, in Sections 3, 4 and 5, the experimental plan, instrumentation, and computer plan, respectively, are described. Finally, Section 6 treats other considerations, such as personnel, outside participation, and distribution of results.

  19. PWR blowdown heat transfer separate-effects program: thermal-hydraulic test facility experimental data report for test 104

    International Nuclear Information System (INIS)

    Leon, D.M.; White, M.D.; Moore, P.A.; Hedrick, R.A.

    1978-01-01

    Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 104, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in the PWR system. Test 104 was conducted to obtain CHF in bundle 1 under blowdown conditions. The primary purpose of this report is to make the reduced instrument responses during test 104 available

  20. Heat transfer test in a vertical tube using CO2 at supercritical pressures

    International Nuclear Information System (INIS)

    Kim, Hwan Yeol; Kim, Hyungrae; Song, Jin Ho; Cho, Bong Hyun; Bae, Yoon Yeong

    2007-01-01

    Heat transfer test facility, SPHINX (Supercritical Pressure Heat Transfer Investigation for NeXt Generation), was constructed at KAERI (Korea Atomic Energy Research Institute) for an investigation of the thermal-hydraulic behaviors of supercritical CO 2 at the various geometries of the test section. The test data will be used for the reactor core design of the SCWR (SuperCritical Water-cooled Reactor). As a working fluid, CO 2 was selected to make use of the low critical pressure and temperature of CO 2 compared with water. An experimental study was carried out in the SPHINX to investigate the characteristics of heat transfer and pressure drop at a vertical single tube with an inside diameter of 4.4 mm in case of an upward flow of supercritical CO 2 . The heat and mass fluxes were varied at a given pressure. The mass flux was in the range of 400-1,200 kg/m 2 s and the heat flux was chosen up to 150 kW/m 2 . The selected pressures were 7.75, 8.12, and 8.85 MPa. A heat transfer deterioration occurred at the lower mass fluxes. The experimental heat transfer coefficients were compared with the ones predicted by several existing correlations. The standard deviation was about 20% for each correlation and an apparent discrepancy was not found among the correlations. The major components of the pressure drop were a gravitational pressure drop and a frictional pressure drop. The frictional pressure drop increases as the mass flux and heat flux increase. (author)

  1. Shielding member for thermonuclear device

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, Masanori

    1997-06-30

    In a thermonuclear device for shielding fast neutrons by shielding members disposed in a shielding vessel (vacuum vessel and structures such as a blanket disposed in the vacuum vessel), the shielding member comprises a large number of shielding wires formed fine and short so as to have elasticity. The shielding wires are sealed in a shielding vessel together with water, and when the width of the shielding vessel is changed, the shielding wires follow after the change of the width while elastically deforming in the shielding vessel, so that great stress and deformation are not formed thereby enabling to improve reliability. In addition, the length, the diameter and the shape of each of the shielding wires can be selected in accordance with the shielding space of the shielding vessel. Even if the shape of the shielding vessel is complicated, the shielding wires can be inserted easily. Accordingly, the filling rate of the shielding members can be changed easily. It can be produced more easily compared with a conventional spherical pebbles. It can be produced more easily than existent spherical shielding pebbles thereby enabling to reduce the production cost. (N.H.)

  2. 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

  3. Effects of particle size and heating time on thiobarbituric acid (TBA) test of soybean powder.

    Science.gov (United States)

    Lee, Youn-Ju; Yoon, Won-Byong

    2013-06-01

    Effects of particle size and heating time during TBA test on the thiobarbituric acid reactive substance (TBARS) of soybean (Glycine Max) powder were studied. Effects of processing variables involved in the pulverization of soybean, such as the temperature of soybean powder, the oxygen level in the vessel, and the pulverisation time, were investigated. The temperature of the soybean powder and the oxygen level had no significant influence on the TBARS (pTBA test significantly affected the TBARS. Change of TBARS during heating was well described by the fractional conversion first order kinetics model. A diffusion model was introduced to quantify the effect of particle size on TBARS. The major finding of this study was that the TBA test to estimate the level of the lipid oxidation directly from powders should consider the heating time and the mean particle sizes of the sample. Copyright © 2012 Elsevier Ltd. All rights reserved.

  4. Testing of a 4 K to 2 K heat exchanger with an intermediate pressure drop

    Energy Technology Data Exchange (ETDEWEB)

    Knudsen, Peter N. [Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States); Ganni, Venkatarao [Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States)

    2015-12-01

    Most large sub-atmospheric helium refrigeration systems incorporate a heat exchanger at the load, or in the distribution system, to counter-flow the sub-atmospheric return with the super-critical or liquid supply. A significant process improvement is theoretically obtainable by handling the exergy loss across the Joule-Thompson throttling valve supplying the flow to the load in a simple but different manner. As briefly outlined in previous publications, the exergy loss can be minimized by allowing the supply flow pressure to decrease to a sub-atmospheric pressure concurrent with heat exchange flow from the load. One practical implementation is to sub-divide the supply flow pressure drop between two heat exchanger sections, incorporating an intermediate pressure drop. Such a test is being performed at Jefferson Lab's Cryogenic Test Facility (CTF). This paper will briefly discuss the theory, practical implementation and test results and analysis obtained to date.

  5. A heated large block test for high level nuclear waste management

    International Nuclear Information System (INIS)

    Lin, W.; Wilder, D.G.; Blink, J.A.; Blair, S.C.; Buscheck, T.A.; Glassley, W.E.; Lee, K.; Owens, M.W.; Roberts, J.J.

    1995-01-01

    The radioactive decay heat from high-level nuclear waste may, depending on the thermal load, create coupled thermal-mechanical-hydrological-chemical (TMHC) processes in the host rock of a repository. A heated large block test (LBT) is designed to understand some of the TNMC processes. A block of Topopah Spring tuff of about 3 x 3 x 4.5 m was isolated at Fran Ridge, Nevada Test Site. Small blocks of the rock adjacent to the large block were collected for laboratory testing of some individual thermal-mechanical, thermal-hydrological, and thermal-chemical processes. The large block will be heated by heaters within so that a dryout zone and a condensate zone will exist simultaneously. Guard heaters on the block sides will be used to minimize horizontal heat losses. A constant load of about 4 MPa will be applied to the top and sides of the large block. The sides will be sealed with moisture and thermal barriers. Temperature, moisture content, pore pressure, chemical composition, stress, displacement, electrical resistivity, acoustic emissions, and acoustic velocities will be measured throughout the block during the heating and cool-down phases. The results from the experiments on small blocks and the tests on the large block will provide a better understanding of some concepts of the coupled TMHC processes. The progress of the project is presented in this paper

  6. An Operators View of Reliability Testing and Decay Heat Rejection Systems

    International Nuclear Information System (INIS)

    Henderson, J.D.C.

    1975-01-01

    The object of this paper is to review the in-situ testing of DHR systems, and to convey policy rather than to indicate a definitive test programme. The test policy is aimed primarily at commissioning the plant and secondly at providing such support for reliability predictions as is practical. Provisions for removal of decay heat from the core and from the reactor tank are described in papers by Broadley and Davies

  7. PWR upper/lower internals shield

    Energy Technology Data Exchange (ETDEWEB)

    Homyk, W.A. [Indian Point Station, Buchanan, NY (United States)

    1995-03-01

    During refueling of a nuclear power plant, the reactor upper internals must be removed from the reactor vessel to permit transfer of the fuel. The upper internals are stored in the flooded reactor cavity. Refueling personnel working in containment at a number of nuclear stations typically receive radiation exposure from a portion of the highly contaminated upper intervals package which extends above the normal water level of the refueling pool. This same issue exists with reactor lower internals withdrawn for inservice inspection activities. One solution to this problem is to provide adequate shielding of the unimmersed portion. The use of lead sheets or blankets for shielding of the protruding components would be time consuming and require more effort for installation since the shielding mass would need to be transported to a support structure over the refueling pool. A preferable approach is to use the existing shielding mass of the refueling pool water. A method of shielding was devised which would use a vacuum pump to draw refueling pool water into an inverted canister suspended over the upper internals to provide shielding from the normally exposed components. During the Spring 1993 refueling of Indian Point 2 (IP2), a prototype shield device was demonstrated. This shield consists of a cylindrical tank open at the bottom that is suspended over the refueling pool with I-beams. The lower lip of the tank is two feet below normal pool level. After installation, the air width of the natural shielding provided by the existing pool water. This paper describes the design, development, testing and demonstration of the prototype device.

  8. Study on concrete cask for practical use. Heat removal test under normal condition

    International Nuclear Information System (INIS)

    Takeda, Hirofumi; Wataru, Masumi; Shirai, Koji; Saegusa, Toshiari

    2005-01-01

    In Japan, it is planed to construct interim storage facilities taking account of dry storage away form reactor in 2010. Recently, a concrete cask is noticed from the economical point of view. But data for its safety analysis have not been sufficient yet. Heat removal tests using to types of full-scale concrete casks were conducted. This paper describes the results under normal condition of spent fuel storage. In the tests, data on heat removal performance and integrity of cask components were obtained for different storage periods. The change of decay heat of spent fuel was simulated using electric heaters. Reinforced Concrete cask (RC cask) and Concrete Filled Steel cask (CFS cask) were the specimen casks. The levels of decay heat at the initial period of 60 years of storage, the intermediate period (20 years of storage), and the final period (40 years of storage) correspond to 22.6 kW, 16 kW and 10 kW, respectively. Quantitative temperature data of the cask components were obtained as compared with their limit temperature. In addition, heat balance data required for heat removal analyses were obtained. (author)

  9. First in situ operation performance test of ground source heat pump in Tunisia

    International Nuclear Information System (INIS)

    Naili, Nabiha; Attar, Issam; Hazami, Majdi; Farhat, Abdelhamid

    2013-01-01

    Highlights: • Evaluate the geothermal energy in Tunisia. • Study of the performance of GSHP system for cooling space. • GSHP is a promising alternative for building cooling in Tunisia. - Abstract: The main purpose of this paper is to study the energetic potential of the deployment in Tunisia of the Ground Source Heat Pump (GSHP) system for cooling mode application. Therefore, a pilot GSHP system using horizontal Ground Heat Exchanger (GHE) was installed and experimented in the Research and Technology Center of Energy (CRTEn), Borj Cédria. The experiment is conducted in a test room with a floor area of about 12 m 2 . In the floor of the tested room is integrated a polyethylene exchanger (PEX) used as a radiant floor cooling (RFC) system. The experimental setup mainly includes the ground temperature, the temperature and flow rate of water circulating in the heat pump and the GHE, as well as the power consumption of the heat pump and circulating pumps. These experimental data are essentially used to evaluate the coefficient of performance of the heat pump (COP hp ) and the overall system (COP sys ) for continuous operation mode. The COP hp and the COP sys were found to be 4.25 and 2.88, respectively. These results reveal that the use of the ground source heat pump is very appropriate for Tunisian building cooling

  10. Standard Test Method for Measuring Heat Transfer Rate Using a Thin-Skin Calorimeter

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2005-01-01

    1.1 This test method covers the design and use of a thin metallic calorimeter for measuring heat transfer rate (also called heat flux). Thermocouples are attached to the unexposed surface of the calorimeter. A one-dimensional heat flow analysis is used for calculating the heat transfer rate from the temperature measurements. Applications include aerodynamic heating, laser and radiation power measurements, and fire safety testing. 1.2 Advantages 1.2.1 Simplicity of ConstructionThe calorimeter may be constructed from a number of materials. The size and shape can often be made to match the actual application. Thermocouples may be attached to the metal by spot, electron beam, or laser welding. 1.2.2 Heat transfer rate distributions may be obtained if metals with low thermal conductivity, such as some stainless steels, are used. 1.2.3 The calorimeters can be fabricated with smooth surfaces, without insulators or plugs and the attendant temperature discontinuities, to provide more realistic flow conditions for ...

  11. Heat removal tests on dry storage facilities for nuclear spent fuels

    International Nuclear Information System (INIS)

    Wataru, M.; Saegusa, T.; Koga, T.; Sakamoto, K.; Hattori, Y.

    1999-01-01

    In Japan, spent fuel generated in NPP is controlled and stored in dry storage facility away-from reactor. Natural convection cooling system of the storage facility is considered advantageous from both safety and economic point of view. In order to realize this type of facility it is necessary to develop an evaluation method for natural convection characteristics and to make a rational design taking account safety and economic factors. Heat removal tests with the reduces scale models of storage facilities (cask, vault and silo) identified the the flow pattern in the test modules. The temperature and velocity distributions were obtained and the heat transfer characteristics were evaluated

  12. Design and testing of a combustion-heated nineteen-converter SAVTEC array

    International Nuclear Information System (INIS)

    Nyren, T.; Fitzpatrick, G.O.; Korringa, M.; McVey, J.; Sahines, T.

    1984-01-01

    The SAVTEC (Self-Adjusting Versatile Thermionic Energy Converter) is a new design approach for achieving very close (<12μ) interelectrode spacing in a thermionic converter. Techniques were developed for fabricating an array of nineteen SAVTEC converters. The array was incorporated in an SiC protective ''hot shell'' which also served as a radiant heat source for the emitter of each converter. The completed assembly was tested with a specially constructed combustion heat source. Electric output was generated by sixteen of the nineteen converters, despite poor thermal contact in a cooling block, which resulted in high collector temperatures. Details of the array design and test results are described

  13. Heat Flux Tests of the ITER FWQMs at KoHLT-1

    International Nuclear Information System (INIS)

    Bae, Young Dug; Kim, Suk Kwon; Shin, Hee Yun; Lee, Dong Won; Hong, Bong Guen

    2009-05-01

    As a party of the ITER, especially as a procurement party of the ITER blanket, we have designed the First Wall Qualification Mockup (FWQM) and fabricated five FWQMs. Two of them have been tested up to 12,690/12,020 cycles at a heat flux higher than 0.625 MW/m 2 at the KoHLT-1 facility established in the Korea Atomic Energy Research Institute (KAERI). Two KO FWQMs successfully passed the normal heat flux tests, and there was no indication of defect in the Be-to-CuCrZr joints

  14. Estimating ISABELLE shielding requirements

    International Nuclear Information System (INIS)

    Stevens, A.J.; Thorndike, A.M.

    1976-01-01

    Estimates were made of the shielding thicknesses required at various points around the ISABELLE ring. Both hadron and muon requirements are considered. Radiation levels at the outside of the shield and at the BNL site boundary are kept at or below 1000 mrem per year and 5 mrem/year respectively. Muon requirements are based on the Wang formula for pion spectra, and the hadron requirements on the hadron cascade program CYLKAZ of Ranft. A muon shield thickness of 77 meters of sand is indicated outside the ring in one area, and hadron shields equivalent to from 2.7 to 5.6 meters in thickness of sand above the ring. The suggested safety allowance would increase these values to 86 meters and 4.0 to 7.2 meters respectively. There are many uncertainties in such estimates, but these last figures are considered to be rather conservative

  15. Scintillation counter, segmented shield

    International Nuclear Information System (INIS)

    Olson, R.E.; Thumim, A.D.

    1975-01-01

    A scintillation counter, particularly for counting gamma ray photons, includes a massive lead radiation shield surrounding a sample-receiving zone. The shield is disassembleable into a plurality of segments to allow facile installation and removal of a photomultiplier tube assembly, the segments being so constructed as to prevent straight-line access of external radiation through the shield into radiation-responsive areas. Provisions are made for accurately aligning the photomultiplier tube with respect to one or more sample-transmitting bores extending through the shield to the sample receiving zone. A sample elevator, used in transporting samples into the zone, is designed to provide a maximum gamma-receiving aspect to maximize the gamma detecting efficiency. (U.S.)

  16. 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

  17. Shields for nuclear reactors

    International Nuclear Information System (INIS)

    Aspden, G.J.

    1984-01-01

    The patent concerns shields for nuclear reactors. The roof shield comprises a normally fixed radial outer portion, a radial inner portion rotatable about a vertical axis, and a connection between the inner and outer portions. In the event of hypothecal core disruption conditions, a cantilever system on the inner wall allows the upward movement of the inner wall, in order to prevent loss of containment. (UK)

  18. Radiation shielding curtain

    International Nuclear Information System (INIS)

    Winkler, N.T.

    1976-01-01

    A radiation shield is described in the form of a stranded curtain made up of bead-chains whose material and geometry are selected to produce a cross-sectional density that is the equivalent of 0.25 mm or more of lead and which curtain may be mounted on various radiological devices to shield against scattered radiation while offering a minimum of obstruction to the radiologist

  19. Shielded cells transfer automation

    International Nuclear Information System (INIS)

    Fisher, J.J.

    1984-01-01

    Nuclear waste from shielded cells is removed, packaged, and transferred manually in many nuclear facilities. Radiation exposure is absorbed by operators during these operations and limited only through procedural controls. Technological advances in automation using robotics have allowed a production waste removal operation to be automated to reduce radiation exposure. The robotic system bags waste containers out of glove box and transfers them to a shielded container. Operators control the system outside the system work area via television cameras. 9 figures

  20. Development and testing of passive autocatalytic recombiners cooled by heat pipes

    International Nuclear Information System (INIS)

    Granzow, Christoph

    2012-01-01

    A severe accident in a nuclear power plant (NPP) can lead to core damage in conjunction with the release of large amounts of hydrogen. As hydrogen mitigation measure, passive autocatalytic recombiners (PARs) are used in today's pressurized water reactors. PARs recombine hydrogen and oxygen contained in the air to steam. The heat from this exothermic reaction causes the catalyst and its surroundings to heat up. If parts of the PAR heat up above the ignition temperature of the gas mixture, a spontaneous deflagration or detonation can occur. The aim of this work is the prevention of such high temperatures by means of passive cooling of the catalyst with heat pipes. Heat pipes are completely passive heat exchanger with a very high effective thermal conductivity. For a deeper understanding of the reaction kinetics at lower temperatures, single catalytic coated heat pipes are studied in a flow reactor. The development of a modular small-scale PAR model is then based on a test series with cooled catalyst sheets. Finally, the PAR model is tested inside a pressure vessel under boundary conditions similar to a real NPP. The experiments show, that the temperatures of the cooled catalytic sheets stay significantly below the temperature of the uncooled sheets and below the ignition temperature of the gas mixture under any set boundary conditions, although no significant reduction of the conversion efficiency can be observed. As a last point, a mathematical model of the reaction kinetics of the recombination process as well as a model of the fluid dynamic and thermohydraulic processes in a heat pipe are developed with the data obtained from the experiments.

  1. Evaluation of the shielding integrity of end-shields in PHWR type NPPs

    International Nuclear Information System (INIS)

    Sah, B.M.L.; Ramamirtham, B.; Kutty, B.S.

    1996-01-01

    In the new plants (Narora Atomic Power Plants (NAPP) onwards) relatively higher radiation fields exist on the north and south fuelling machine (FM) vault walls of the E1 100m accessible area passages. These fields were first noticed at NAPS-1 and subsequently at NAPS-2 and KAPS-1. Such surveys done at RAPS have indicated that the fields on these walls would come out to be quite low (only 1-2 mR/h) from sources other than that arising from 41 Ar contamination. RAPS/MAPS experience pointed to adequacy of shielding of the FM vault walls and sufficient overall shielding thickness of the end-shields. Further, radiometry tests of end-shields carried out at Kaiga and RAPP 3 and 4 indicated fairly satisfactory and uniform filling of balls. Hence, incomplete filling of water column of the end-shields due to any venting problem was suspected to be one possible reason for the observed high fields in NAPS and Kakrapar Atomic Power Station (KAPS). Since the presence of high radiation fields, both neutron and gamma, is of long-term concern, a special study/measurement of radiation levels on reactor face during high power operation was undertaken. In order to compare the shielding integrity of the older (RAPS/MAPS solid plate type shielding) and newer (NAPS/KAPS steel ball-filled type) end shields, these experiments were done at MAPS-2 and NAPS-2. (author). 2 refs., 2 tabs

  2. Evaluation of mechanical design fire brick at test section on the HeaTiNG-02

    International Nuclear Information System (INIS)

    Dedy Haryanto; Riswan Djambiar; Sagino; Edy Sumarno

    2013-01-01

    The activity was carried out due to the modification of the heating in the HeaTiNG-02 test section. Modification of the heater needs to be done to overcome the obstacles that arise as part of the test section is used. Constraint that often arises is the fall of the heating source with super khantal material when it reaches a certain temperature. To mitigate the super khantal position is initially converted into a vertical position horizontal. The change from vertical to horizontal position on super khantal cause any deformities in refractory fire brick which serves as a support super khantal. Manufacture of refractory design fire brick formation and mechanical strength evaluation performed using CATIA V5 R20 software. Evaluation of fireproof rock mechanics to be based on the mechanical properties of alumina as a refractory material of fire brick. The results of the analysis have a fire brick design stress greater than the bend strength alumina materials, so that the necessary checks before and after the experiment as well as the replacement of refractory fire brick if something is broken. Translational greatest displacement 0.453 mm at a temperature of 1575 K did not give any meaningful form. Thus the refractory fire brick design can be used as heating source support in the HeaTiNG-02 test section with checks before and after the operation. (author)

  3. The ITER EC H and CD upper launcher: Design, analysis and testing of a bolted joint for the Blanket Shield Module

    Energy Technology Data Exchange (ETDEWEB)

    Gessner, Robby, E-mail: robby.gessner@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Aiello, Gaetano; Grossetti, Giovanni; Meier, Andreas [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Ronden, Dennis [DIFFER – Dutch Institute for Fundamental Energy Physics, P.O. Box 1207, NL-3430 BE Nieuwegein (Netherlands); Spaeh, Peter; Scherer, Theo; Schreck, Sabine; Strauss, Dirk; Vaccaro, Alessandro [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2013-10-15

    Highlights: ► The BSM of the ECH Launcher is attached to the Launcher Main Frame by a bolted joint. ► The bolts were designed as “captive” in order to avoid their accidental removal from the joint. ► The bolted flange connection using two sets of 15 captive bolts (M22 × 2) placed along the sides. ► The captive bolt design is based on a concept that uses a dedicated spring ring, a standard spiral spring and a tensioning screw with two threads to secure the bolts in a form-locking stop. -- Abstract: The final design of the structural system for the ITER EC H and CD upper launcher is in progress. Many design features of the preliminary design are under revision with the aim to achieve the built-to-print-status. This paper deals with design and analysis of a bolted joint for the Blanket Shield Module with special perspective on Remote Handling capability. The BSM of the ECH Launcher is attached to the Launcher Main Frame by a bolted joint conceived so that in the Hot Cell Facility, RH maintenance can be performed on internal components. The joint must be capable to resist very high Electro-Magnetic loads from disruptions, while it has to sustain substantial thermal cycling during operation. Thus the need for a rigid and reliable design is essential. Beside the set of pre-stressed bolts the flanges were therefore equipped with additional shear keys to divert radial moments away from the bolts. Main focus of the work performed was the mechanical design of the joint and the assessment of the structural integrity with respect to the loads applied and its capability for maintenance by RH procedures. To fulfill a major aspect of the RH requirements, the bolts were designed as “captive” in order to avoid their accidental removal from the joint. The captive bolt design is based on a concept that uses a dedicated spring ring, a standard spiral spring and a tensioning screw with two threads to secure the bolts in a form-locking stop. The final approval phase of

  4. The ITER EC H and CD upper launcher: Design, analysis and testing of a bolted joint for the Blanket Shield Module

    International Nuclear Information System (INIS)

    Gessner, Robby; Aiello, Gaetano; Grossetti, Giovanni; Meier, Andreas; Ronden, Dennis; Spaeh, Peter; Scherer, Theo; Schreck, Sabine; Strauss, Dirk; Vaccaro, Alessandro

    2013-01-01

    Highlights: ► The BSM of the ECH Launcher is attached to the Launcher Main Frame by a bolted joint. ► The bolts were designed as “captive” in order to avoid their accidental removal from the joint. ► The bolted flange connection using two sets of 15 captive bolts (M22 × 2) placed along the sides. ► The captive bolt design is based on a concept that uses a dedicated spring ring, a standard spiral spring and a tensioning screw with two threads to secure the bolts in a form-locking stop. -- Abstract: The final design of the structural system for the ITER EC H and CD upper launcher is in progress. Many design features of the preliminary design are under revision with the aim to achieve the built-to-print-status. This paper deals with design and analysis of a bolted joint for the Blanket Shield Module with special perspective on Remote Handling capability. The BSM of the ECH Launcher is attached to the Launcher Main Frame by a bolted joint conceived so that in the Hot Cell Facility, RH maintenance can be performed on internal components. The joint must be capable to resist very high Electro-Magnetic loads from disruptions, while it has to sustain substantial thermal cycling during operation. Thus the need for a rigid and reliable design is essential. Beside the set of pre-stressed bolts the flanges were therefore equipped with additional shear keys to divert radial moments away from the bolts. Main focus of the work performed was the mechanical design of the joint and the assessment of the structural integrity with respect to the loads applied and its capability for maintenance by RH procedures. To fulfill a major aspect of the RH requirements, the bolts were designed as “captive” in order to avoid their accidental removal from the joint. The captive bolt design is based on a concept that uses a dedicated spring ring, a standard spiral spring and a tensioning screw with two threads to secure the bolts in a form-locking stop. The final approval phase of

  5. Tests on fast heating for the regeneration process of ITER cryopumps

    International Nuclear Information System (INIS)

    Day, C.; Kammerer, B.; Mack, A.

    1996-10-01

    Within the framework of the European Fusion Technology Programme, a primary vacuum pump for the ITER reactor is being developed. As the tritium accumulated by the pumps must be limited, short pumping cycles are necessary and as a consequence to that regeneration times of about 4 min only are required by the intermittently working cryopumps; approximately 60 s are available for the heating process from LHe temperature (4.2 K) to LN 2 temperature (77 K). Methods for fast heating were tested in component tests. The heating tests were performed at the TITAN test facility. According to the basic planning, the LHe-cooled panel consisted of seven flow channels in quilted design (500 x 350 mm 2 ); the detailed planning meanwhile showed that a smaller number of channels per panel will be sufficient. The panel was mounted in a LN 2 -cooled rig, which worked as first pumping stage. After having worked out a screening study comprehensive test series with three different heating methods were performed. (orig.) [de

  6. Study of Incoloy 800HT alloy tested by heat-cycling

    International Nuclear Information System (INIS)

    Velciu, L.; Meleg, T.; Pantiru, M.; Petrescu, D.; Voicu, F.

    2016-01-01

    This paper investigated Incoloy 800HT (UNS N08811) alloy after some heat-cycling tests. The study continues prior tests realized in INR Pitesti concerning utilization of some nickel-based alloys in the heat exchangers and steam generators construction. The thermal-cycling consist in a successive series of heating and cooling with some rates in a range temperature. Technical parameters of thermal cycling: 50 & 200 cycles, 25 °C/minute heating-cooling rate, temperature range 450-1000°C, and argon working medium. The analysis consisted in metallographic examination (microstructure), Vickers microhardness, and traction tests. The average grain size was determined by linear interception method (ASTM E-112). The micro hardness was calculated by the relationship of the device technical book. On the Strength-Deformation diagrams were obtained: tensile strength and elongation. The tested samples were compared with the ''as received'' material. The results showed a good metallographic and mechanical behaviour of Incoloy 800HT at these thermal-cycling tests. (authors)

  7. Ground test program for a full-size solar dynamic heat receiver

    Science.gov (United States)

    Sedgwick, L. M.; Kaufmann, K. J.; Mclallin, K. L.; Kerslake, T. W.

    1991-01-01

    Test hardware, facilities, and procedures were developed to conduct ground testing of a full-size, solar dynamic heat receiver in a partially simulated, low earth orbit environment. The heat receiver was designed to supply 102 kW of thermal energy to a helium and xenon gas mixture continuously over a 94 minute orbit, including up to 36 minutes of eclipse. The purpose of the test program was to quantify the receiver thermodynamic performance, its operating temperatures, and thermal response to changes in environmental and power module interface boundary conditions. The heat receiver was tested in a vacuum chamber using liquid nitrogen cold shrouds and an aperture cold plate. Special test equipment was designed to provide the required ranges in interface boundary conditions that typify those expected or required for operation as part of the solar dynamic power module on the Space Station Freedom. The support hardware includes an infrared quartz lamp heater with 30 independently controllable zones and a closed-Brayton cycle engine simulator to circulate and condition the helium-xenon gas mixture. The test article, test support hardware, facilities, and instrumentation developed to conduct the ground test program are all described.

  8. Identification student’s misconception of heat and temperature using three-tier diagnostic test

    Science.gov (United States)

    Suliyanah; Putri, H. N. P. A.; Rohmawati, L.

    2018-03-01

    The objective of this research is to develop a Three-Tier Diagnostic Test (TTDT) to identify the student's misconception of heat and temperature. Stages of development include: analysis, planning, design, development, evaluation and revise. The results of this study show that (1) the quality of the three-tier type diagnostic test instrument developed has been expressed well with the following details: (a) Internal validity of 88.19% belonging to the valid category. (b) External validity of empirical construct validity test using Pearson Product Moment obtained 0.43 is classified and result of empirical construct validity test obtained false positives 6.1% and false negatives 5.9% then the instrument was valid. (c) Test reliability by using Cronbach’s Alpha of 0.98 which means acceptable. (d) The 80% difficulty level test is quite difficult. (2) Student misconceptions on the temperature of heat and displacement materials based on the II test the highest (84%), the lowest (21%), and the non-misconceptions (7%). (3) The highest cause of misconception among students is associative thinking (22%) and the lowest is caused by incomplete or incomplete reasoning (11%). Three-Tier Diagnostic Test (TTDT) could identify the student's misconception of heat and temperature.

  9. Heat transfer characteristics evaluation of heat exchangers of mock-up test facility with full-scale reaction tube for HTTR hydrogen production system (Contract research)

    International Nuclear Information System (INIS)

    Shimizu, Akira; Ohashi, Hirofumi; Kato, Michio; Hayashi, Koji; Aita, Hideki; Nishihara, Tetsuo; Inaba, Yoshitomo; Takada, Shoji; Morisaki, Norihiro; Sakaki, Akihiro; Maeda, Yukimasa; Sato, Hiroyuki; Inagaki, Yoshiyuki; Hanawa, Hiromi; Fujisaki, Katsuo; Yonekawa, Hideo

    2005-06-01

    Connection of hydrogen production system by steam reforming of methane to the High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Research Institute (JAERI) has been surveyed until now. Mock-up test facility of this steam reforming system with full-scale reaction tube was constructed in FY 2001 and hydrogen of 120 Nm 3 /h was successfully produced in overall performance test. Totally 7 times operational tests were performed from March 2002 to December 2004. A lot of operational test data on heat exchanges were obtained in these tests. In this report specifications and structures of steam reformer, steam superheater, steam generator, condenser, helium gas cooler, feed gas heater and feed gas superheater were described. Heat transfer correlation equations for inside and outside tube were chosen from references. Spreadsheet programs were newly made to evaluate heat transfer characteristics from measured test data such as inlet and outlet temperature pressure and flow-rate. Overall heat-transfer coefficients obtained from the experimental data were compared and evaluated with the calculated values with heat transfer correlation equation. As a result, actual measurement values of all heat exchangers gave close agreement with the calculated values with correlation equations. Thermal efficiencies of the heat exchangers were adequate as they were well accorded with design value. (author)

  10. 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

  11. Rotating Arc Jet Test Model: Time-Accurate Trajectory Heat Flux Replication in a Ground Test Environment

    Science.gov (United States)

    Laub, Bernard; Grinstead, Jay; Dyakonov, Artem; Venkatapathy, Ethiraj

    2011-01-01

    Though arc jet testing has been the proven method employed for development testing and certification of TPS and TPS instrumentation, the operational aspects of arc jets limit testing to selected, but constant, conditions. Flight, on the other hand, produces timevarying entry conditions in which the heat flux increases, peaks, and recedes as a vehicle descends through an atmosphere. As a result, we are unable to "test as we fly." Attempts to replicate the time-dependent aerothermal environment of atmospheric entry by varying the arc jet facility operating conditions during a test have proven to be difficult, expensive, and only partially successful. A promising alternative is to rotate the test model exposed to a constant-condition arc jet flow to yield a time-varying test condition at a point on a test article (Fig. 1). The model shape and rotation rate can be engineered so that the heat flux at a point on the model replicates the predicted profile for a particular point on a flight vehicle. This simple concept will enable, for example, calibration of the TPS sensors on the Mars Science Laboratory (MSL) aeroshell for anticipated flight environments.

  12. Failure analysis of radioisotopic heat source capsules tested under multi-axial conditions

    International Nuclear Information System (INIS)

    Zielinski, R.E.; Stacy, E.; Burgan, C.E.

    In order to qualify small radioisotopic heat sources for a 25-yr design life, multi-axial mechanical tests were performed on the structural components of the heat source. The results of these tests indicated that failure predominantly occurred in the middle of the weld ramp-down zone. Examination of the failure zone by standard metallographic techniques failed to indicate the true cause of failure. A modified technique utilizing chemical etching, scanning electron microscopy, and energy dispersive x-ray analysis was employed and dramatically indicated the true cause of failure, impurity concentration in the ramp-down zone. As a result of the initial investigation, weld parameters for the heat sources were altered. Example welds made with a pulse arc technique did not have this impurity buildup in the ramp-down zone

  13. Heat resistant/radiation resistant cable and incore structure test device for FBR type reactor

    International Nuclear Information System (INIS)

    Tanimoto, Hajime; Shiono, Takeo; Sato, Yoshimi; Ito, Kazumi; Sudo, Shigeaki; Saito, Shin-ichi; Mitsui, Hisayasu.

    1995-01-01

    A heat resistant/radiation resistant coaxial cable of the present invention comprises an insulation layer, an outer conductor and a protection cover in this order on an inner conductor, in which the insulation layer comprises thermoplastic polyimide. In the same manner, a heat resistant/radiation resistant power cable has an insulation layer comprising thermoplastic polyimide on a conductor, and is provided with a protection cover comprising braid of alamide fibers at the outer circumference of the insulation layer. An incore structure test device for an FBR type reactor comprises the heat resistant/radiation resistant coaxial cable and/or the power cable. The thermoplastic polyimide can be extrusion molded, and has excellent radiation resistant by the extrusion, as well as has high dielectric withstand voltage, good flexibility and electric characteristics at high temperature. The incore structure test device for the FBR type reactor of the present invention comprising such a cable has excellent reliability and durability. (T.M.)

  14. Overview of the EU small scale mock-up tests for ITER high heat flux components

    International Nuclear Information System (INIS)

    Vieider, G.; Barabash, V.; Cardella, A.

    1998-01-01

    This task within the EU R and D for ITER was aimed at the development of basic manufacturing solutions for the high heat flux plasma facing components such as the divertor targets, the baffles and limiters. More than 50 representative small-scale mock-ups have been manufactured with beryllium, carbon and tungsten armour using various joining technologies. High heat flux testing of 20 of these mock-ups showed the carbon mono-blocks to be the most robust solution, surviving 2000 cycles at absorbed heat fluxes of up to 24 MW m -2 . With flat armour tiles rapid joint failures occurred at 5-16 MW m -2 depending on joining technology and armour material. These test results serve as a basis for the selection of manufacturing options and materials for the prototypes now being ordered. (orig.)

  15. Thermal Protection Test Bed Pathfinder Development Project

    Science.gov (United States)

    Snapp, Cooper

    2015-01-01

    In order to increase thermal protection capabilities for future reentry vehicles, a method to obtain relevant test data is required. Although arc jet testing can be used to obtain some data on materials, the best method to obtain these data is to actually expose them to an atmospheric reentry. The overprediction of the Orion EFT-1 flight data is an example of how the ground test to flight traceability is not fully understood. The RED-Data small reentry capsule developed by Terminal Velocity Aerospace is critical to understanding this traceability. In order to begin to utilize this technology, ES3 needs to be ready to build and integrate heat shields onto the RED-Data vehicle. Using a heritage Shuttle tile material for the heat shield will both allow valuable insight into the environment that the RED-Data vehicle can provide and give ES3 the knowledge and capability to build and integrate future heat shields for this vehicle.

  16. EBT-P gamma-ray-shielding analysis

    International Nuclear Information System (INIS)

    Gohar, Y.

    1983-01-01

    First, a one-dimensional scoping study was performed for the gamma-ray shield of the ELMO Bumpy Torus proof-of-principle device to define appropriate shielding material and determine the required shielding thickness. The dose-equivalent results are analyzed as a function of the radiation-shield thickness for different shielding options. A sensitivity analysis for the pessimistic case is given. The recommended shielding option based on the performance and cost is discussed. Next, a three-dimensional scoping study for the coil shield was performed for four different shielding options to define the heat load for each component and check the compliance with the design criterion of 10 watts maximum heat load per coil from the gamma-ray sources. Also, a detailed biological-dose survey was performed which included: (a) the dose equivalent inside and outside the building, (b) the dose equivalent from the two mazes of the building, and (c) the skyshine contribution to the dose equivalent

  17. Development, Testing, and Failure Mechanisms of a Replicative Ice Phase Change Material Heat Exchanger

    Science.gov (United States)

    Leimkuehler, Thomas O.; Hansen, Scott; Stephan, Ryan A.

    2010-01-01

    Phase change materials (PCM) may be useful for thermal control systems that involve cyclical heat loads or cyclical thermal environments such as Low Earth Orbit (LEO) and Low Lunar Orbit (LLO). Thermal energy can be stored in the PCM during peak heat loads or in adverse thermal environments. The stored thermal energy can then be released later during minimum heat loads or in more favorable thermal environments. One advantage that PCM's have over evaporators in this scenario is that they do not use a consumable. Wax PCM units have been baselined for the Orion thermal control system and also provide risk mitigation for the Altair Lander. However, the use of water as a PCM has the potential for significant mass reduction since the latent heat of formation of water is approximately 70% greater than that of wax. One of the potential drawbacks of using ice as a PCM is its potential to rupture its container as water expands upon freezing. In order to develop a space qualified ice PCM heat exchanger, failure mechanisms must first be understood. Therefore, a methodical experimental investigation has been undertaken to demonstrate and document specific failure mechanisms due to ice expansion in the PCM. An ice PCM heat exchanger that replicates the thermal energy storage capacity of an existing wax PCM unit was fabricated and tested. Additionally, methods for controlling void location in order to reduce the risk of damage due to ice expansion are investigated. This paper presents the results to date of this investigation. Nomenclature

  18. Heat removal tests for pressurized water reactor containment spray by largescale facility

    International Nuclear Information System (INIS)

    Motoki, Y.; Hashimoto, K.; Kitani, S.; Naritomi, M.; Nishio, G.; Tanaka, M.

    1983-01-01

    Heat removal tests for pressurized water reactor (PWR) containment spray were carried out to investigate effectiveness of the depressurization by Japan Atomic Energy Research Institute model containment (7-m diameter, 20 m high, and 708-m 3 volume) with PWR spray nozzles. The depressurization rate is influenced by the spray heat transfer efficiency and the containment wall surface heat transfer coefficient. The overall spray heat transfer efficiency was investigated with respect to spray flow rate, weight ratio of steam/air, and spray height. The spray droplet heat transfer efficiency was investigated whether the overlapping of spray patterns gives effect or not. The effect was not detectable in the range of large value of steam/air, however, it was better in the range of small value of it. The experimental results were compared with the calculated results by computer code CONTEMPT-LT/022. The overall spray heat transfer efficiency was almost 100% in the containment pressure, ranging from 2.5 to 0.9 kg/cm 2 X G, so that the code was useful on the prediction of the thermal hydraulic behavior of containment atmosphere in a PWR accident condition

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  20. Magnetic shielding of a limiter

    International Nuclear Information System (INIS)

    Brevnov, N.N.; Stepanov, S.B.; Khimchenko, L.N.; Matthews, G.F.; Goodal, D.H.J.

    1991-01-01

    Localization of plasma interaction with material surfaces in a separate chamber, from where the escape of impurities is hardly realized, i.e. application of magnetic divertors or pump limiters, is the main technique for reduction of the impurity content in a plasma. In this case, the production of a divertor configuration requires a considerable power consumption and results in a less effective utilization of the magnetic field volume. Utilization of a pump limiter, for example the ICL-type, under tokamak-reactor conditions would result in the extremely high and forbidden local heat loadings onto the limiter surface. Moreover, the magnetically-shielded pump limiter (MSL) was proposed to combine positive properties of the divertor and the pump limiter. The idea of magnetic shielding is to locate the winding with current inside the limiter head so that the field lines of the resultant magnetic field do not intercept the limiter surface. In this case the plasma flows around the limiter leading edges and penetrates into the space under the limiter. The shielding magnetic field can be directed either counter the toroidal field or counter the poloidal one of a tokamak, dependent on the concrete diagram of the device. Such a limiter has a number of advantages: -opportunity to control over the particle and impurity recycling without practical influence upon the plasma column geometry, - perturbation of a plasma column magnetic configuration from the side of such a limiter is less than that from the side of the divertor coils. The main deficiency is the necessity to locate active windings inside the discharge chamber. (author) 5 refs., 3 figs