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Sample records for rcc heat shields

  1. 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

  2. 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...

  3. 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.

  4. 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.)

  5. 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

  6. 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)

  7. 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

  8. 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.)

  9. 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.

  10. 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.

  11. 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.

  12. 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

  13. 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.

  14. 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

  15. 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.

  16. 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.

  17. 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.

  18. 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

    This NIAC project investigated an innovative approach to provide heat shield protection to spacecraft after launch and prior to each EDL thus potentially realizing significant launch mass savings. Heat shields fabricated in situ can provide a thermal-protection system for spacecraft that routinely enter a planetary atmosphere. By fabricating the heat shield with space resources from materials available on moons and asteroids, it is possible to avoid launching the heat-shield mass from Earth. Regolith has extremely good insulating properties and the silicates it contains can be used in the fabrication and molding of thermal-protection materials. Such in situ developed heat shields have been suggested before by Lewis. Prior research efforts have shown that regolith properties can be compatible with very-high temperature resistance. Our project team is highly experienced in regolith processing and thermal protection systems (TPS). Routine access to space and return from any planetary surface requires dealing with heat loads experienced by the spacecraft during reentry. Our team addresses some of the key issues with the EDL of human-scale missions through a highly innovative investigation of heat shields that can be fabricated in space by using local resources on asteroids and moons. Most space missions are one-way trips, dedicated to placing an asset in space for economical or scientific gain. However, for human missions, a very-reliable heat-shield system is necessary to protect the crew from the intense heat experienced at very high entry velocities of approximately 11 km/s at approximately Mach 33 (Apollo). For a human mission to Mars, the return problem is even more difficult, with predicted velocities of up to 14 km/s, at approximately Mach 42 at the Earth-atmosphere entry. In addition to human return, it is very likely that future space-travel architecture will include returning cargo to the Earth, either for scientific purposes or for commercial reasons

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. 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.

  4. 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

  5. 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)

  6. 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.

  7. RCC-MX (2008 edition)

    International Nuclear Information System (INIS)

    Bravo, X.; Drubay, B.

    2008-10-01

    The RCC-MX books is a compilation of design and construction rules for the mechanical materials of experimental reactors, for their auxiliaries and irradiation devices. This second edition includes the updates of references to NF, EN and ISO standards, the compliance with the regulations for nuclear pressure equipments, and the feedback since the 2005 edition. It comprises 9 books and a CD-Rom and includes a presentation document. The RCC-MX has been developed for the Jules Horowitz reactor but can be used for the design and construction of new projects of new experimental reactors or new equipments and devices for existing facilities. Content: - Book 1: general dispositions, materials for experimental reactors and their auxiliaries, for irradiation devices and for control or handling mechanisms, complementary requirements and particular dispositions; - Book 2: materials for the reactor and for its level 1 auxiliaries; - Book 3: materials for the reactor and for its level 2 and level 3 auxiliaries, control and handling mechanisms, materials for irradiation devices; - Book 4: technical appendixes - materials characteristics (steels and alloys); - Book 5: technical appendixes (design rules); - Book 6: technical specifications of materials; - Book 7: tests and control methods; - Book 8: welding; - Book 9: fabrication. (J.S.)

  8. 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

  9. 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.)

  10. 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.

  11. Simulation analysis of temperature control on RCC arch dam of hydropower station

    Science.gov (United States)

    XIA, Shi-fa

    2017-12-01

    The temperature analysis of roller compacted concrete (RCC) dam plays an important role in their design and construction. Based on three-dimensional finite element method, in the computation of temperature field, many cases are included, such as air temperature, elevated temperature by cement hydration heat, concrete temperature during placing, the influence of water in the reservoir, and boundary temperature. According to the corresponding parameters of RCC arch dam, the analysis of temperature field and stress field during the period of construction and operation is performed. The study demonstrates that detailed thermal stress analysis should be performed for RCC dams to provide a basis to minimize and control the occurrence of thermal cracking.

  12. Comparisons of ratchetting analysis methods using RCC-M, RCC-MR and ASME codes

    International Nuclear Information System (INIS)

    Yang Yu; Cabrillat, M.T.

    2005-01-01

    The present paper compares the simplified ratcheting analysis methods used in RCC-M, RCC-MR and ASME with some examples. Firstly, comparisons of the methods in RCC-M and efficiency diagram in RCC-MR are investigated. A special method is used to describe these two methods with curves in one coordinate, and the different conservation is demonstrated. RCC-M method is also be interpreted by SR (second ratio) and v (efficiency index) which is used in RCC-MR. Hence, we can easily compare the previous two methods by defining SR as abscissa and v as ordinate and plotting two curves of them. Secondly, comparisons of the efficiency curve in RCC-MR and methods in ASME-NH APPENDIX T are investigated, with significant creep. At last, two practical evaluations are performed to show the comparisons of aforementioned methods. (authors)

  13. 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.

  14. 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

  15. 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

  16. 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.

  17. 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)

  18. 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

  19. 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.

  20. RCC Plug Repair Thermal Tools for Shuttle Mission Support

    Science.gov (United States)

    Rodriguez, Alvaro C.; Anderson, Brian P.

    2010-01-01

    A thermal math model for the Space Shuttle Reinforced Carbon-Carbon (RCC) Plug Repair was developed to increase the confidence in the repair entry performance and provide a real-time mission support tool. The thermal response of the plug cover plate, local RCC, and metallic attach hardware can be assessed with this model for any location on the wing leading edge. The geometry and spatial location of the thermal mesh also matches the structural mesh which allows for the direct mapping of temperature loads and computation of the thermoelastic stresses. The thermal model was correlated to a full scale plug repair radiant test. To utilize the thermal model for flight analyses, accurate predictions of protuberance heating were required. Wind tunnel testing was performed at CUBRC to characterize the heat flux in both the radial and angular directions. Due to the complexity of the implementation of the protuberance heating, an intermediate program was developed to output the heating per nodal location for all OML surfaces in SINDA format. Three Design Reference Cases (DRC) were evaluated with the correlated plug thermal math model to bound the environments which the plug repair would potentially be used.

  1. 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)

  2. 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.

  3. 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

  4. 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

  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. 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

  8. Plutonium recycle concept for RCC - type PWRs

    International Nuclear Information System (INIS)

    Bonet, H.; Charlier, A.; Deramaix, P.; Vanderberg, C.

    1975-01-01

    Self-generated Pu recycling schemes in RCC-type PWRs have been defined. The main results of survey studies performed to compare the relative merits of various Pu recycle strategies and the merits of alternative solutions of the assembly design such as the Pu-island assembly or the all-Pu assembly are presented [fr

  9. 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.)

  10. 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.

  11. 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)

  12. 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.

  13. 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.

  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. 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

  16. 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

  17. 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

  18. 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)

  19. 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.

  20. 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.

  1. 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

  2. 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.

  3. 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.

  4. 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.

  5. 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

  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. Accelerated pavement testing of thin RCC over soil cement pavements

    Directory of Open Access Journals (Sweden)

    Zhong Wu

    2016-05-01

    Full Text Available Three full-scale roller compacted concrete (RCC pavement sections built over a soil cement base were tested under accelerated pavement testing (APT. The RCC thicknesses varied from 102 mm (4 in. to 152 mm (6 in. and to 203 mm (8 in., respectively. A bi-directional loading device with a dual-tire load assembly was used for this experiment. Each test section was instrumented with multiple pressure cells and strain gages. The objective was to evaluate the structural performance and load carrying capacity of thin RCC-surfaced pavements under accelerated loading. The APT results generally indicated that all three RCC pavement sections tested in this study possessed very high load carrying capacity; an estimated pavement life in terms of equivalent single axle load (ESAL for the thinnest RCC section (i.e., RCC thickness of 102 mm evaluated was approximately 19.2 million. It was observed that a fatigue failure would be the primary pavement distress type for a thin RCC pavement under trafficking. Specifically, the development of fatigue cracking was found to originate from a longitudinal crack at the edge or in the center of a tire print, then extended and propagated, and eventually merged with cracks of other directions. Instrumentation results were used to characterize the fatigue damage under different load magnitudes. Finally, based on the APT performance of this experiment, two fatigue models for predicting the fatigue life of thin RCC pavements were developed. Keywords: Roller compacted concrete, APT, Pavement performance, Non-destructive testing, Fatigue analysis

  8. RCC-M - Design and Conception Rules for Mechanical Components of PWR Nuclear Islands

    International Nuclear Information System (INIS)

    2007-01-01

    The design and construction rules applicable to mechanical components of PWR Nuclear Islands (RCC-M) are a part of the collection of design and construction rules for nuclear power plants. It covers the rules applicable to the design and manufacture of pressure boundaries of mechanical equipment of pressurized water reactors (PWR). The pressure components subject to the RCC-M are specified in A 4000. They include the reactor fluid systems (primary, secondary and auxiliary systems) and other components which are not subject to pressure: vessel internals, supports for pressure components subject to the RCC-M, nuclear island storage tanks. When a pressure equipment is subject to the RCC-M, all its elements subject to pressure are also, in accordance with the provisions of A 4000, and these elements are the same class as the component. In this case all the provisions of the RCC-M are applicable: design, procurement, manufacture, inspection and pressure testing. Elements which are not subject to pressure and which are subject to the RCC-M may be covered within the Code by limited specific provisions (procurement of materials for example). The other rules applicable to this equipment must be in contractual form. The assemblies comprising pressure equipment assembled by a manufacturer to constitute an integrated and functional whole, shall be subject to the rules indicated in this Code. Main objectives of Code Requirements are to ensure the integrity and mechanical stability over the equipment design life. Function ability and operability of equipment are not directly addressed in the Code. The RCC-M contributes to ensuring compliance with regulatory requirements. These requirements depend on the applicable regulatory context. The RCC-M is representative of the state of the art as concerns the design and manufacture of PWR components, ensuring an overall safety level tested through experience. The RCC-M consists of five sections, which provide rules for the design and

  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. 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

  11. Evaluation of KALIMER IHTS piping using French RCC-MR code

    International Nuclear Information System (INIS)

    Lee, Hyeong Yeon; Kim, J. B.; Lee, J. H.

    2001-12-01

    In the present report, the evaluation of design integrity for the liquid metal reactor(LMR) of KALIMER IHTS(intermediate heat transport system) piping according to the French design guideline of RCC-MR RC3600 developed for secondary piping of LMR and the evaluation procedure was presented. The evaluation results showed that the results by the simple RC-3600 procedure of design by formula were more conservative than those of ASME section III subsection NH of the design by analysis for the class I structural components

  12. 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

  13. Design and Construction Rules for Mechanical components of FBR nuclear islands: RCC-MR. Tome II: materials

    International Nuclear Information System (INIS)

    1985-06-01

    This book is the 8th of a whole set of 12 ones which constitute the edition 1985 of the RCC-MR. One deals with alloy and non-alloy steels on one hand, and with pieces and products (bars, plates, tubes, forgings and castings) on the other hand. The requirements concerning the mechanical properties, the chemical composition, the search of defects and their repair, the heat treatments, the tests to carry out and the delivery state are specified [fr

  14. Creep design rules in french ''RCC-MR'' code

    International Nuclear Information System (INIS)

    Roche, H.

    1986-04-01

    In this paper, four points enlightening the originality of the ''RCC-MR'' analysis rules in the creep range will be discussed. The three first points will concern elastic analysis, the fourth one materials data. The rules given in this paper are applicable for class 1 components and level A conditions. They are given by way of illustration in a simplified form

  15. 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.

  16. 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.

  17. Sequential Ground Motion Effects on the Behavior of a Base-Isolated RCC Building

    Directory of Open Access Journals (Sweden)

    Zhi Zheng

    2017-01-01

    Full Text Available The sequential ground motion effects on the dynamic responses of reinforced concrete containment (RCC buildings with typical isolators are studied in this paper. Although the base isolation technique is developed to guarantee the security and integrity of RCC buildings under single earthquakes, seismic behavior of base-isolated RCC buildings under sequential ground motions is deficient. Hence, an ensemble of as-recorded sequential ground motions is employed to study the effect of including aftershocks on the seismic evaluation of base-isolated RCC buildings. The results indicate that base isolation can significantly attenuate the earthquake shaking of the RCC building under not only single earthquakes but also seismic sequences. It is also found that the adverse aftershock effect on the RCC can be reduced due to the base isolation applied to the RCC. More importantly, the study indicates that disregarding aftershocks can induce significant underestimation of the isolator displacement for base-isolated RCC buildings.

  18. 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.

  19. 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

  20. Creep-fatigue rules in the RCC-MR code

    International Nuclear Information System (INIS)

    Drubay, B.

    1988-01-01

    In 1978, CEA, Electricite de France (EDF) and NOVATOME decided to draw up a complete set of design and construction rules for LMFBR components. This RCC-MR code issued in June 1985 and completed in November 1987 was chosen as a sound basis for the next European Fast Reactor (EFR). The purpose of this paper is to describe the present RCC-MR creep-fatigue design rules to be applied with elastic analysis including the modifications adopted in the first addenda. This method is based on a separate evaluation of a fatigue usage fraction V and creep rupture usage fraction W with the common linear summation rule. The fatigue usage fraction is obtained from continuous fatigue curves (without hold times) and from total strain ranges (elastic + plastic + creep). The creep rupture usage fraction W is obtained from stress to rupture curves and a stress σk evaluating the stress generated during the cycle. (author)

  1. 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)

  2. RCC1 regulates inner centromeric composition in a Ran-independent fashion.

    Science.gov (United States)

    Zhang, Michael Shaofei; Furuta, Maiko; Arnaoutov, Alexei; Dasso, Mary

    2018-01-01

    RCC1 associates to chromatin dynamically within mitosis and catalyzes Ran-GTP production. Exogenous RCC1 disrupts kinetochore structure in Xenopus egg extracts (XEEs), but the molecular basis of this disruption remains unknown. We have investigated this question, utilizing replicated chromosomes that possess paired sister kinetochores. We find that exogenous RCC1 evicts a specific subset of inner KT proteins including Shugoshin-1 (Sgo1) and the chromosome passenger complex (CPC). We generated RCC1 mutants that separate its enzymatic activity and chromatin binding. Strikingly, Sgo1 and CPC eviction depended only on RCC1's chromatin affinity but not its capacity to produce Ran-GTP. RCC1 similarly released Sgo1 and CPC from synthetic kinetochores assembled on CENP-A nucleosome arrays. Together, our findings indicate RCC1 regulates kinetochores at the metaphase-anaphase transition through Ran-GTP-independent displacement of Sgo1 and CPC.

  3. 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)

  4. 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.

  5. 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

  6. 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

  7. 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

  8. 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.

  9. 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

  10. Analysis of Pumphouse RCC Frame Structure for Soil Structure Interaction

    OpenAIRE

    Mr A.S. Thombare; Prof. V.P. Kumbhar; Prof. A.H. Kumbhar

    2016-01-01

    When structure is built on ground some elements of structure are direct contact with soil. When loads are applied on structure internal forces are developed in both the structure as well as in soil. It results in deformation of both the components which are independent to each other. This are called soil structure interaction. The analysis is done by using (Bentley STAAD.Pro V8i Version 2007) software. The analysis carried out been pump house structure R.C.C. frame structure find ...

  11. Radiometer Calibration and Characterization (RCC) User's Manual: Windows Version 4.0

    Energy Technology Data Exchange (ETDEWEB)

    Andreas, Afshin M. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Wilcox, Stephen M. [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2016-02-29

    The Radiometer Calibration and Characterization (RCC) software is a data acquisition and data archival system for performing Broadband Outdoor Radiometer Calibrations (BORCAL). RCC provides a unique method of calibrating broadband atmospheric longwave and solar shortwave radiometers using techniques that reduce measurement uncertainty and better characterize a radiometer's response profile. The RCC software automatically monitors and controls many of the components that contribute to uncertainty in an instrument's responsivity. This is a user's manual and guide to the RCC software.

  12. 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.)

  13. 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.

  14. 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.)

  15. 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''

  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. 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.

  18. 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.

  19. 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.

  20. Retrofitting Of RCC Piles By Using Basalt Fiber Reinforced Polymer BFRP Composite Part 1 Review Papers On RCC Structures And Piles Retrofitting Works.

    OpenAIRE

    R. Ananda Kumar; Dr. C. Selvamony; A. Seeni; Dr. T. R. Sethuraman

    2015-01-01

    Abstract Retrofitting works are immensely essential for deteriorated and damaged structures in Engineering and Medical fields in order to keep or return to the originality for safe guarding the structures and consumers. In this paper different types of methods of retrofitting review notes are given based on the experimental numerical and analytical methods results on strengthening the Reinforced cement concrete RCC structures including RCC piles. Soil-pile interaction on axial load lateral lo...

  1. 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

  2. t(6;11) renal cell carcinoma (RCC): expanded immunohistochemical profile emphasizing novel RCC markers and report of 10 new genetically confirmed cases.

    Science.gov (United States)

    Smith, Nathaniel E; Illei, Peter B; Allaf, Mohamed; Gonzalez, Nilda; Morris, Kerry; Hicks, Jessica; Demarzo, Angelo; Reuter, Victor E; Amin, Mahul B; Epstein, Jonathan I; Netto, George J; Argani, Pedram

    2014-05-01

    Renal cell carcinomas (RCCs) harboring the t(6;11)(p21;q12) translocation were first described in 2001 and recently recognized by the 2013 International Society of Urological Pathology Vancouver Classification of Renal Neoplasia. Although these RCCs are known to label for melanocytic markers HMB45 and Melan A and the cysteine protease cathepsin K by immunohistochemistry (IHC), a comprehensive IHC profile has not been reported. We report 10 new t(6;11) RCCs, all confirmed by break-apart TFEB fluorescence in situ hybridization. A tissue microarray containing 6 of these cases and 7 other previously reported t(6;11) RCCs was constructed and immunolabeled for 21 different antigens. Additional whole sections of t(6;11) RCC were labeled with selected IHC markers. t(6;11) RCC labeled diffusely and consistently for cathepsin K and Melan A (13 of 13 cases) and almost always at least focally for HMB45 (12 of 13 cases). They labeled frequently for PAX8 (14 of 23 cases), CD117 (10 of 14 cases), and vimentin (9 of 13 cases). A majority of cases labeled at least focally for cytokeratin Cam5.2 (8 of 13 cases) and CD10 and RCC marker antigen (10 of 14 cases each). In contrast to a prior study's findings, only a minority of cases labeled for Ksp-cadherin (3 of 19 cases). The median H score (product of intensity score and percentage labeling) for phosphorylated S6, a marker of mTOR pathway activation, was 101, which is high relative to most other RCC subtypes. In summary, IHC labeling for PAX8, Cam5.2, CD10, and RCC marker antigen supports classification of the t(6;11) RCC as carcinomas despite frequent negativity for broad-spectrum cytokeratins and EMA. Labeling for PAX8 distinguishes the t(6;11) RCC from epithelioid angiomyolipoma, which otherwise shares a similar immunoprofile. CD117 labeling is more frequent in the t(6;11) RCC compared with the related Xp11 translocation RCC. Increased pS6 expression suggests a possible molecular target for the uncommon t(6;11) RCCs that

  3. t(6;11) Renal Cell Carcinoma (RCC) Expanded Immunohistochemical Profile Emphasizing Novel RCC Markers and Report of 10 New Genetically Confirmed Cases

    Science.gov (United States)

    Smith, Nathaniel E.; Illei, Peter B.; Allaf, Mohamed; Gonzalez, Nilda; Morris, Kerry; Hicks, Jessica; DeMarzo, Angelo; Reuter, Victor E.; Amin, Mahul B.; Epstein, Jonathan I.; Netto, George J.; Argani, Pedram

    2015-01-01

    Renal cell carcinomas (RCCs) harboring the t(6;11)(p21;q12) translocation were first described in 2001 and recently recognized by the 2013 International Society of Uro-logical Pathology Vancouver Classification of Renal Neoplasia. Although these RCCs are known to label for melanocytic markers HMB45 and Melan A and the cysteine protease cath-epsin K by immunohistochemistry (IHC), a comprehensive IHC profile has not been reported. We report 10 new t(6;11) RCCs, all confirmed by break-apart TFEB fluorescence in situ hybridization. A tissue microarray containing 6 of these cases and 7 other previously reported t(6;11) RCCs was constructed and immunolabeled for 21 different antigens. Additional whole sections of t(6;11) RCC were labeled with selected IHC markers. t(6;11) RCC labeled diffusely and consistently for cathepsin K and Melan A (13 of 13 cases) and almost always at least focally for HMB45 (12 of 13 cases). They labeled frequently for PAX8 (14 of 23 cases), CD117 (10 of 14 cases), and vimentin (9 of 13 cases). A majority of cases labeled at least focally for cytokeratin Cam5.2 (8 of 13 cases) and CD10 and RCC marker antigen (10 of 14 cases each). In contrast to a prior study's findings, only a minority of cases labeled for Ksp-cadherin (3 of 19 cases). The median H score (product of intensity score and percentage labeling) for phosphorylated S6, a marker of mTOR pathway activation, was 101, which is high relative to most other RCC subtypes. In summary, IHC labeling for PAX8, Cam5.2, CD10, and RCC marker antigen supports classification of the t(6;11) RCC as carcinomas despite frequent negativity for broad-spectrum cytokeratins and EMA. Labeling for PAX8 distinguishes the t(6;11) RCC from epithelioid angiomyolipoma, which otherwise shares a similar immunoprofile. CD117 labeling is more frequent in the t(6;11) RCC compared with the related Xp11 translocation RCC. Increased pS6 expression suggests a possible molecular target for the uncommon t(6;11) RCCs that

  4. 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

  5. Comparison of elevated temperature design codes of ASME Subsection NH and RCC-MRx

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyeong-Yeon, E-mail: hylee@kaeri.re.kr

    2016-11-15

    Highlights: • Comparison of elevated temperature design (ETD) codes was made. • Material properties and evaluation procedures were compared. • Two heat-resistant materials of Grade 91 steel and austenitic stainless steel 316 are the target materials in the present study. • Application of the ETD codes to Generation IV reactor components and a comparison of the conservatism was conducted. - Abstract: The elevated temperature design (ETD) codes are used for the design evaluation of Generation IV (Gen IV) reactor systems such as sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), and very high temperature reactor (VHTR). In the present study, ETD code comparisons were made in terms of the material properties and design evaluation procedures for the recent versions of the two major ETD codes, ASME Section III Subsection NH and RCC-MRx. Conservatism in the design evaluation procedures was quantified and compared based on the evaluation results for SFR components as per the two ETD codes. The target materials are austenitic stainless steel 316 and Mod.9Cr-1Mo steel, which are the major two materials in a Gen IV SFR. The differences in the design evaluation procedures as well as the material properties in the two ETD codes are highlighted.

  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. Retrofitting Of RCC Piles By Using Basalt Fiber Reinforced Polymer BFRP Composite Part 1 Review Papers On RCC Structures And Piles Retrofitting Works.

    Directory of Open Access Journals (Sweden)

    R. Ananda Kumar

    2015-01-01

    Full Text Available Abstract Retrofitting works are immensely essential for deteriorated and damaged structures in Engineering and Medical fields in order to keep or return to the originality for safe guarding the structures and consumers. In this paper different types of methods of retrofitting review notes are given based on the experimental numerical and analytical methods results on strengthening the Reinforced cement concrete RCC structures including RCC piles. Soil-pile interaction on axial load lateral load reviews are also presented. This review paper is prepared to find out the performance of basalt fibre reinforced polymer BFRP composite retrofitted reinforced cement concrete single end bearing piles.

  8. Nano-Phase Powder Based Exothermic Braze Repair Technology For RCC Materials, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — MRi is proposing, with its partner, Exotherm Corp (Camden, NJ) to demonstrate the feasibility of using exothermic brazing to join RCC (or C:SiC) composites to itself...

  9. Nano-Phase Powder Based Exothermic Braze Repair Technology For RCC Materials, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — The Phase II project will advance innovative, cost effective and reliable nano-phase exothermic RCC joining processes (ExoBrazeTM) in order to be able to reinforce...

  10. Test and Analysis Correlation of Form Impact onto Space Shuttle Wing Leading Edge RCC Panel 8

    Science.gov (United States)

    Fasanella, Edwin L.; Lyle, Karen H.; Gabrys, Jonathan; Melis, Matthew; Carney, Kelly

    2004-01-01

    Soon after the Columbia Accident Investigation Board (CAIB) began their study of the space shuttle Columbia accident, "physics-based" analyses using LS-DYNA were applied to characterize the expected damage to the Reinforced Carbon-Carbon (RCC) leading edge from high-speed foam impacts. Forensic evidence quickly led CAIB investigators to concentrate on the left wing leading edge RCC panels. This paper will concentrate on the test of the left-wing RCC panel 8 conducted at Southwest Research Institute (SwRI) and the correlation with an LS-DYNA analysis. The successful correlation of the LS-DYNA model has resulted in the use of LS-DYNA as a predictive tool for characterizing the threshold of damage for impacts of various debris such as foam, ice, and ablators onto the RCC leading edge for shuttle return-to-flight.

  11. Application of a polycarboxylate ether admixture in RCC dam construction[ACI SP-239

    Energy Technology Data Exchange (ETDEWEB)

    Asmus, S.M.F.; Christensen, B.J.; Varley, N.J. [BASF Construction Chemicals Asia Pacific, Shanghai (China)

    2006-07-01

    Chemical admixtures are used in dam construction to improve plasticity of the dry materials mixture over time. Roller compacted concrete (RCC) has been used on many dam projects in China. However, the use of RCC has frequently resulted in water reduction problems. This paper provided details of an admixture based on polycarboxylate ether (PCE) which was developed to improve the quality of RCC constructions at the JinHong dam in China. Use of the polymer at the JinHong dam resulted in a vibration sensitive concrete that was sustained over time. Under identical mix-design and compaction conditions in the laboratory, specific gravity of the RCC was increased from 2417 kg/m{sup 3} to 2463 kg/m{sup 3}. The high specific gravity of the material resulted in satisfactory strength data from the dam project. The key-ratio of the splitting tensile strength versus compressive strength was higher than 8 per cent in all cases. A key advantage of the tailored PCE-RCC was the short Vebe times sustained over elapsed time in the RCC. Without additional compaction or vibration efforts, the specific density of RCC was better than conventional admixture technologies. The reduced viscosity provided cement paste films which formed on the surface of each layer of the RCC, which resulted in better bonding between the layers. It was concluded that the new PCE polymer is compatible with alternative retarder systems, which contributes to more extensive setting times under strict hydration regimes. 7 refs., 4 tabs., 4 figs.

  12. High volume fly ash RCC for dams - I : mixture optimization and mechanical properties

    Energy Technology Data Exchange (ETDEWEB)

    Jacobsen, S. [PEAB Construction Co., Oslo (Norway); Lahus, O. [Norwegian Building Research Inst., Oslo (Norway)

    2001-07-01

    Roller compacted concretes (RCC) were developed for the Norwegian Skjerka hydropower project. RCCs were developed to have a high-volume fly ash content to address environmental issues, including the reduction of carbon dioxide emissions associated with dam construction. They also makes good use of waste product and conserve natural resources. This study examined a series of mixtures to determine the appropriateness of using RCC as a competing alternative to the traditional rock fill dam proposed for the Skjerka hydropower project. The main advantage of RCC is speed, allowing a relatively large dam to be constructed in just one summer season, saving financial costs and providing early return on the investment. In addition, fly ash can be used in the structure, using clean and renewable energy. Several procedures to proportion RCC mixtures were proposed, including the optimal paste volume method which is based on the assumption that an optimal RCC should have just enough paste to fill the space between particles when the granular skeleton has reached its maximum density under compaction. With this assumption, RCC tests began in 1998 in the laboratories of the Norwegian Building Research Institute. An ordinary portland cement was used and combined with ordinary low lime fly ash. Both coarse and fine aggregate were used. The tests determined the optimum paste-mortar ratio, the content of coarse aggregates and the production of specimens for test on hardened and fresh concrete. The study showed that the compressive strength of RCC increased with increasing cement/(cement + fly ash) ratio. The permeability coefficient decreased with increasing cement-content and increasing cement/(cement + fly ash) ratio due to the slow pozzolanic reaction of fly ash making a more open pore structure. It was concluded that an optimized mixture can result in a high performance RCC in terms of fresh and hardened concrete properties. 15 refs., 5 tabs., 11 figs.

  13. 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

  14. 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.)

  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. 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)

  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. 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.)

  19. RCC2 over-expression in tumor cells alters apoptosis and drug sensitivity by regulating Rac1 activation.

    Science.gov (United States)

    Wu, Nan; Ren, Dong; Li, Su; Ma, Wenli; Hu, Shaoyan; Jin, Yan; Xiao, Sheng

    2018-01-10

    Small GTP binding protein Rac1 is a component of NADPH oxidases and is essential for superoxide-induced cell death. Rac1 is activated by guanine nucleotide exchange factors (GEFs), and this activation can be blocked by regulator of chromosome condensation 2 (RCC2), which binds the switch regions of Rac1 to prevent access from GEFs. Three cancer cell lines with up- or down-regulation of RCC2 were used to evaluate cell proliferation, apoptosis, Rac1 signaling and sensitivity to a group of nine chemotherapeutic drugs. RCC2 expression in lung cancer and ovarian cancer were studied using immunochemistry stain of tumor tissue arrays. Forced RCC2 expression in tumor cells blocked spontaneous- or Staurosporine (STS)-induced apoptosis. In contrast, RCC2 knock down in these cells resulted in increased apoptosis to STS treatment. The protective activity of RCC2 on apoptosis was revoked by a constitutively activated Rac1, confirming a role of RCC2 in apoptosis by regulating Rac1. In an immunohistochemistry evaluation of tissue microarray, RCC2 was over-expressed in 88.3% of primary lung cancer and 65.2% of ovarian cancer as compared to non-neoplastic lung and ovarian tissues, respectively. Because chemotherapeutic drugs can kill tumor cells by activating Rac1/JNK pathway, we suspect that tumors with RCC2 overexpression would be more resistant to these drugs. Tumor cells with forced RCC2 expression indeed had significant difference in drug sensitivity compared to parental cells using a panel of common chemotherapeutic drugs. RCC2 regulates apoptosis by blocking Rac1 signaling. RCC2 expression in tumor can be a useful marker for predicting chemotherapeutic response.

  20. Design and Construction Rules for Mechanical components of FBR nuclear islands: RCC-MR. Tome 1, Volume A: generalities

    International Nuclear Information System (INIS)

    1985-06-01

    The French Rules of Mechanical equipments of Fast Neutron nuclear Reactors (RCC-MR) aims at equipments included in a safety classification. The equipments concerned are those of the nuclear boiler and its auxiliaries: tanks, vessels, internal equipments of the reactor, exchangers, pumps, fittings, pipes, and supports. The present edition of the RCC-MR comprises 12 books presented in the present one in the volume A. The chapter RA 3000 defines the documents to be established in application of the RCC-MR rules. The chapter RA 5000 defines the requirements to take into account to establish and carry out quality Assurance programs according to the RCC-MR rules [fr

  1. Harmonization of welding qualification provisions in RCC-M and European standards

    International Nuclear Information System (INIS)

    Lemoine, M.; Lainez, B.; Anastassiades, P.

    2007-01-01

    Since a long time, numerous precautions for welding have been integrated in the nuclear codes, in particular in the RCC-M applicable to pressurized water reactors, in order to guarantee a high quality level of permanent assemblies. In parallel, European and ISO standardization works have led to a harmonisation of practices on qualification of welding processes, welders and operators. In the context of the regulatory evolutions presented during this conference, it was judged appropriate to bring closer the RCC-M practices and those of EN and ISO standards, while taking the precaution of specifying, if necessary, the complementary provisions allowing maintaining guarantees of quality consistent with the prior experience feedback. This paper presents the amendments brought to the RCC-M Code by the 2005 and 2007 addenda, in order to respond to this objective, and develops their motivations. (authors) [fr

  2. 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.).

  3. 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

  4. Material report in support to RCC-MRX code 2010 stainless steel parts and products

    International Nuclear Information System (INIS)

    Ancelet, Olivier; Lebarbe, Thierry; Dubiez-Le Goff, Sophie; Bonne, Dominique; Gelineau, Odile

    2012-01-01

    This paper presents the Material Report dedicated to stainless steels parts and products issued by AFCEN (Association Francaise pour les regles de Conception et de Construction des Materiels des Chaudieres Electro-Nucleaires) in support to RCC-MRx 2010 Code. The RCC-MRx Code is the result of the merger of the RCC-MX 2008, developed in the context of the research reactor Jules Horowitz Reactor project, in the RCC-MR 2007, which set up rules applicable to the design of components operating at high temperature and to the Vacuum Vessel of ITER (a presentation of RCC-MRx 2010 Code is the subject of another paper proposed in this Congress; it explains in particular the status of this Code). This Material Report is part of a set of Criteria of RCC-MRx (this set of Criteria is under construction). The Criteria aim at explaining the design and construction rules of the Code. They cover analyses rules as well as part procurement, welding, methods of tests and examination and fabrication rules. The Material Report particularly provides justifications and explanations on requirements and features dealing with parts and products proposed in the Code. The Material Report contains the following information: Introduction of the grade(s): codes and standards and Reference Procurement Specifications covering parts and products, applications and experience gained, - Physical properties, - Mechanical properties used for design calculations (base metal and welds): basic mechanical properties, creep mechanical properties, irradiated mechanical properties, - Fabrication: experience gained, metallurgy, - Welding: weldability, experience gained during welding and repair procedure qualifications, - Non-destructive examination, - In-service behaviour. In the article, examples of data supplied in the Material Report dedicated to stainless steels will be exposed. (authors)

  5. ASME and RCC-MR comparison for the prevention of fatigue analysis

    International Nuclear Information System (INIS)

    Autrusson, B.; Acker, D.

    1989-01-01

    The purpose of this survey is to compare the simplified methods, without reference to the safety factor allowed for the mechanical properties. An application of both codes, RCC-MR and ASME, on the design of the wall mock-up of the NET project is made and also an estimation with an elastoplastic analysis. In the case of fatigue analysis according to ASME in the plastic field, the elastic stress is magnified by a K e factor derived from stress variation, S n , disregarding geometrical discontinuities. According to RCC-MR, the elastic maximum strain will magnified by two coefficients accounting for plasticity and variation of Poisson ratio

  6. RCC-C: Design and construction rules for fuel assemblies of PWR nuclear power plants

    International Nuclear Information System (INIS)

    2015-01-01

    The RCC-C code contains all the requirements for the design, fabrication and inspection of nuclear fuel assemblies and the different types of core components (rod cluster control assemblies, burnable poison rod assemblies, primary and secondary source assemblies and thimble plug assemblies). The design, fabrication and inspection rules defined in RCC-C leverage the results of the research and development work pioneered in France, Europe and worldwide, and which have been successfully used by industry to design and build nuclear fuel assemblies and incorporate the resulting feedback. The code's scope covers: fuel system design, especially for assemblies, the fuel rod and associated core components, the characteristics to be checked for products and parts, fabrication methods and associated inspection methods. The RCC-C code is used by the operator of the PWR nuclear power plants in France as a reference when sourcing fuel from the world's top two suppliers in the PWR market, given that the French operator is the world's largest buyer of PWR fuel. Fuel for EPR projects is manufactured according to the provisions of the RCC-C code. The code is available in French and English. The 2005 edition has been translated into Chinese. Contents of the 2015 edition of the RCC-C code: Chapter 1 - General provisions: 1.1 Purpose of the RCC-C, 1.2 Definitions, 1.3 Applicable standards, 1.4 Equipment subject to the RCC-C, 1.5 Management system, 1.6 Processing of non-conformances; Chapter 2 - Description of the equipment subject to the RCC-C: 2.1 Fuel assembly, 2.2 Core components; Chapter 3 - Design: Safety functions, operating functions and environment of fuel assemblies and core components, design and safety principles; Chapter 4 - Manufacturing: 4.1 Materials and part characteristics, 4.2 Assembly requirements, 4.3 Manufacturing and inspection processes, 4.4 Inspection methods, 4.5 Certification of NDT inspectors, 4.6 Characteristics to be inspected for the

  7. 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)

  8. 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

  9. 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

  10. 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.

  11. GSK-3 inhibition in vitro and in vivo enhances antitumor effect of sorafenib in renal cell carcinoma (RCC)

    Energy Technology Data Exchange (ETDEWEB)

    Kawazoe, Hisashi; Bilim, Vladimir N. [Laboratory of Molecular Oncology, Department of Urology, Yamagata University School of Medicine, Iida-nishi 2-2-2, Yamagata 990-9585 (Japan); Ugolkov, Andrey V., E-mail: ugolkov@northwestern.edu [Tumor Biology Core, Center for Developmental Therapeutics, Chemistry of Life Processes Institute, Silverman Hall B733, Northwestern University, Evanston, IL (United States); Yuuki, Kaori; Naito, Sei; Nagaoka, Akira; Kato, Tomoyuki [Laboratory of Molecular Oncology, Department of Urology, Yamagata University School of Medicine, Iida-nishi 2-2-2, Yamagata 990-9585 (Japan); Tomita, Yoshihiko, E-mail: ytomita@med.id.yamagata-u.ac.jp [Laboratory of Molecular Oncology, Department of Urology, Yamagata University School of Medicine, Iida-nishi 2-2-2, Yamagata 990-9585 (Japan)

    2012-07-06

    Highlights: Black-Right-Pointing-Pointer Sorafenib treatment upregulated GSK-3{beta} levels in RCC cells. Black-Right-Pointing-Pointer Pharmacologic inhibition of GSK-3 suppressed xenograft RCC tumor growth. Black-Right-Pointing-Pointer Inhibition of GSK-3 enhanced antitumor effect of sorafenib in vitro and in vivo. -- Abstract: Sorafenib is a multikinase inhibitor approved for the systemic treatment of renal cell carcinoma (RCC). However, sorafenib treatment has a limited effect due to acquired chemoresistance of RCC. Previously, we identified glycogen synthase kinase-3 (GSK-3) as a new therapeutic target in RCC. Here, we observed that sorafenib inhibits proliferation and survival of RCC cells. Significantly, we revealed that sorafenib enhances GSK-3 activity in RCC cells, which could be a potential mechanism of acquired chemoresistance. We found that pharmacological inhibition of GSK-3 potentiates sorafenib antitumor effect in vitro and in vivo. Our results suggest that combining GSK-3 inhibitor and sorafenib might be a potential new therapeutic approach for RCC treatment.

  12. 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

  13. 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.

  14. Tumor mutational load and immune parameters across metastatic Renal Cell Carcinoma (mRCC) risk groups

    Science.gov (United States)

    de Velasco, Guillermo; Miao, Diana; Voss, Martin H.; Hakimi, A. Ari; Hsieh, James J.; Tannir, Nizar M.; Tamboli, Pheroze; Appleman, Leonard J.; Rathmell, W. Kimryn; Van Allen, Eliezer M.; Choueiri, Toni K.

    2016-01-01

    Patients with metastatic renal cell carcinoma (mRCC) have better overall survival when treated with nivolumab, a cancer immunotherapy that targets the immune checkpoint inhibitor programmed cell death 1 (PD-1), rather than everolimus (a chemical inhibitor of mTOR and immunosuppressant). Poor-risk mRCC patients treated with nivolumab seemed to experience the greatest overall survival benefit, compared to patients with favorable or intermediate-risk, in an analysis of the CheckMate-025 trial subgroup of the Memorial Sloan Kettering Cancer Center (MSKCC) prognostic risk groups. Here we explore whether tumor mutational load and RNA expression of specific immune parameters could be segregated by prognostic MSKCC risk strata and explain the survival seen in the poor-risk group. We queried whole exome transcriptome data in RCC patients (n = 54) included in The Cancer Genome Atlas that ultimately developed metastatic disease or were diagnosed with metastatic disease at presentation and did not receive immune checkpoint inhibitors. Nonsynonymous mutational load did not differ significantly by MSKCC risk group, nor was the expression of cytolytic genes –granzyme A and perforin – or selected immune checkpoint molecules different across MSKCC risk groups. In conclusion, this analysis found that mutational load and expression of markers of an active tumor microenvironment did not correlate with MSKCC risk prognostic classification in mRCC. PMID:27538576

  15. Telemetry Standards, RCC Standard 106-17, Chapter 4, Pulse Code Modulation Standards

    Science.gov (United States)

    2017-07-01

    A-4 Appendix 4-B. Citations ...investigation can be found in a paper by J. L. Maury, Jr. and J. Styles , “Development of Optimum Frame Synchronization Codes for Goddard Space Flight Center...Standards, RCC Standard 106-17 Chapter 4, July 2017 B-1 APPENDIX 4-B Citations Aeronautical Radio, Inc. Mark 33 Digital Information Transfer

  16. 77 FR 17569 - United States-Canada Regulatory Cooperation Council (RCC)-Transportation-Dangerous Goods Working...

    Science.gov (United States)

    2012-03-26

    ... identified in the Joint Action Plan, the Transportation--Dangerous Goods Working Group led by senior...)-- Transportation--Dangerous Goods Working Group AGENCY: Pipeline and Hazardous Materials Safety Administration...--Dangerous Goods Working Group, of the United States-Canada Regulatory Cooperation Council (RCC). Comments...

  17. 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)

  18. 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)

  19. 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.

  20. RCC-E: Design and construction rules for electrical equipment of PWR nuclear islands

    International Nuclear Information System (INIS)

    2016-01-01

    RCC-E describes the rules for designing, building and installing electrical and I and C systems and equipment for pressurized water reactors. The code was drafted in partnership with industry, engineering firms, manufacturers, building control firms and operators, and represents a collection of best practices in accordance with IAEA requirements and IEC standards. The code's scope covers: architecture and the associated systems, materials engineering and the qualification procedure for normal and accidental environmental conditions, facility engineering and management of common cause failures (electrical and I and C) and electromagnetic interference, testing and inspecting electrical characteristics, quality assurance requirements supplementing ISO 9001 and activity monitoring. Use: RCC-E has been used to build the following power plants: France's last 12 nuclear units (1,300 MWe (8) and 1,450 MWe (4)), 2 M310 reactors in Korea (2), 44 M310 (4), CPR-1000 (28), CPR-600 (6), HPR-1000 (4) and EPR (2) reactors in service or undergoing construction in China, 1 EPR reactor in France. RCC-E is used for maintenance operations in French power plants (58 units) and Chinese M310 and CPR-1000 power plants. RCC-E has been chosen for the construction of the EPR plants in Hinkley Point, UK. Contents of the 2016 edition of the RCC-E code: Volume 1 - General requirements and quality assurance; Volume 2 - Specification of requirements; Volume 3 - I and C systems; Volume 4 - Electrical systems; Volume 5 - Materials engineering; Volume 6 - Installation of electrical and I and C systems; Volume 7 - Inspection and test methods

  1. 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

  2. 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.)

  3. 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

  4. 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.)

  5. 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

  6. 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.)

  7. 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

  8. 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

  9. 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.

  10. 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.)

  11. 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)

  12. The RCC-MR design code for LMFBR components. A useful basic for fusion reactor design tools development

    International Nuclear Information System (INIS)

    Acker, D.; Chevereau, G.

    1985-11-01

    LMFBR and fusion reactors exhibit common features with regard to structural materials (Stainless steels), temperature service level (550-600 0 C), loading types. So, design and construction rules used in France for LMFBR, that is to say RCC-MR Code, can constitute a good basis for fusion reactors design. Some original aspects of RCC-MR design rules are described, relating to unsignificant creep, ratchetting effect, fatigue and creep damage limits, creep damage evaluation, fatigue damage evaluation, buckling. The main originality of RCC-MR consists to propose comprehensive simplified rules based on elastic calculations and extended from classical cold temperatures to the elevated temperature domain

  13. The RCC-MR design code for LMFBR components. A useful basis for fusion reactor design tools development

    International Nuclear Information System (INIS)

    Acker, D.; Chevereau, G.

    1986-01-01

    LMFBR and fusion reactors exhibit common features with regard to structural materials, temperature service level, loading types. So, design and construction rules used in France for LMFBR, that is to say RCC-MR Code, can constitute a good basis for fusion reactors design. Some original aspects of RCC-MR design rules are described, relating to unsignificant creep, ratchetting effect, fatigue and creep damage limits, creep damage evaluation, fatigue damage evaluation, buckling. The main originality of RCC-MR consists to propose comprehensive simplified rules based on elastic calculations and extended from classical cold temperatures to the elevated temperature domain. (author)

  14. 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.).

  15. Application of the active flexible fixture with passive RCC function to peg-in-hole task

    International Nuclear Information System (INIS)

    Yamaguchi, Tomomi; Higuchi, Masahiro

    2005-01-01

    This paper describes the application of the active flexible fixture (AFLEF) with passive RCC to the peg-in-hole task on the disk in the X-band accelerator. The AFLEF can fix any work and position the fixed work at short range. In this paper, the 2-dimensional AFLEF is proposed as the simplified type and is provided with passive RCC function to be equipped with dexterity for a peg-in-hole task. As results of the experiment on the peg-in-hole task on the X-band accelerator disks with the AFLEF, we make the ability of the AFLEF for the task clear and also the boundary conditions to the complete task clear. (author)

  16. 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

  17. Telemetry Standards, RCC Standard 106-17. Chapter 8. Digital Data Bus Acquisition Formatting Standard

    Science.gov (United States)

    2017-07-01

    incorrect word count/message and illegal mode codes are not considered bus errors. 8.6.2 Source Signal The source of data is a signal conforming to...Telemetry Standards, RCC Standard 106-17 Chapter 8, July 2017 CHAPTER 8 Digital Data Bus Acquisition Formatting Standard Acronyms...check FCS frame check sequence HDDR high-density digital recording MIL-STD Military Standard msb most significant bit PCM pulse code modulation

  18. Telemetry Standards, RCC Standard 106-17, Chapter 27, RF Network Access Layer

    Science.gov (United States)

    2017-07-01

    to the physical media (i.e., the wireless RF network). On the transmission side, it is responsible for framing IP packets for physical transmission ...resolution bandwidth of 30 kHz. It was measured during the steady power condition during a burst transmission . Telemetry Standards, RCC Standard... power levels available for modulated burst transmission . Table 27-1. Transceiver Phase Noise Mask dBc/Hz Frequency Offset −30 dBc/Hz 10 Hz −60 dBc

  19. Telemetry Standards, RCC Standard 106-17. Chapter 3. Frequency Division Multiplexing Telemetry Standards

    Science.gov (United States)

    2017-07-01

    Standard 106-17 Chapter 3, July 2017 3-5 Table 3-4. Constant-Bandwidth FM Subcarrier Channels Frequency Criteria\\Channels: A B C D E F G H Deviation ...Telemetry Standards , RCC Standard 106-17 Chapter 3, July 2017 3-i CHAPTER 3 Frequency Division Multiplexing Telemetry Standards Acronyms...Frequency Division Multiplexing Telemetry Standards ................................ 3-1 3.1 General

  20. RCC-CW - Rules for design and construction of PWR nuclear civil works

    International Nuclear Information System (INIS)

    2016-01-01

    RCC-CW describes the rules for designing, building and testing civil engineering works in PWR reactors. It explains the principles and requirements for the safety, serviceability and durability of concrete and metal frame structures, based on Eurocode design principles (European standards for the structural design of construction works) combined with specific measures for safety-class buildings. The code is produced as part of the RCC-CW Subcommittee, which includes all the parties involved in civil engineering works in the nuclear sector: clients, contractors, general and specialized firms, consultancies and inspection offices. The code covers the following areas relating to the design and construction of civil engineering works that play an important safety role: geotechnical aspects, reinforced concrete structures and galleries, pre-stressed containments with metal liner, metal containment and pool liners, metal frames, anchors, concrete cylinder pipes, containment leak tests. The RCC-CW code is available as an ETC-C version specific to EPR projects (European pressurized reactor). Contents of the 2016 edition of the RCC-CW Code: Part G - General: scope, standards, notations, quality management, general principles; Part D - Design: actions and combinations of actions, geotechnical aspects, pre-stressed or reinforced concrete structures, metal containment liners, metal pool liners, metal frames, anchors; Part C - Construction: geotechnical aspects, concrete, surface finish and formwork, reinforcement for reinforced concrete, pre-stressing processes, prefabricated concrete elements, metal containment liners, metal pool liners, metal frames, anchors, embedded pipelines, joint sealing, survey networks and tolerances; Part M - Maintenance and monitoring: containment integrity and rate tests

  1. Consideration of creep in design rules of AFCEN RCC-MRx 2012 code

    International Nuclear Information System (INIS)

    Lebarbe, T.; Petesch, C.; Lejeail, Y.; Lamagnere, P.; Dubiez-Le Goff, S.

    2014-01-01

    The 2012 edition of the RCC-MRx Code has been issued in French and English versions by AFCEN (Association Francaise pour les regles de Conception et de Construction des Materiels des Chaudieres Electro-nucleaires). This Code is the result of the merger of the RCC-MX 2008 developed in the context of the research reactor Jules Horowitz Reactor project, in the RCC-MR 2007 which set up rules applicable to the design of components operating at high temperature and to the Vacuum Vessel of ITER. This new edition is the opportunity to publish also the background of the rules. This paper is one illustration of what may be such a document, on a dedicated example, the creep rules. It contains an overview of the design rules associated to the creep damage and explains the purpose and the origins of these rules. This type of exercise is going to be generalized to all the parts of the code in AFCEN technical publications, the criteria. (authors)

  2. 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

  3. 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

  4. A comparative study of physical and chemical properties of different pozzolanic materials used for roller compacted concrete RCC dams

    OpenAIRE

    Husein Malkawi Abdallah I.; Shatnawi Ehab; Husein Malkawi Dima A.

    2017-01-01

    This paper addresses the feasibility and the efficiency of using Natural Pozzolan and/or Rock flour in Roller Compacted Concrete (RCC) gravity dams. For this purpose, five identical mortar trial mixes were prepared using five different supplementary materials, i.e., fly ash produced in South Africa (proven to be effective in RCC construction), fly ash produced in Turkey, Jordanian natural pozzolan, Saudi natural pozzolan, and rock flour from Mujib Dam basalt quarry. The physical and chemical ...

  5. 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.)

  6. 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)

  7. 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

  8. 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.)

  9. Proper understanding of relevant articles in RCC-M and PANE G-7-002

    International Nuclear Information System (INIS)

    Sun Zaozhan

    2005-01-01

    The French code, 'Design and construction rules for mechanical components of PWR nuclear island' RCC-M and the Russian code 'Regulations for strength analysis of equipment and pipelines of atomic power plants' PANE G-7-002, both for nuke plant design, are sometimes used in China, and discrepancies in understanding some articles of such foreign design codes will some times give different conclusions as 'acceptable' or 'unacceptable' the concerned material or product. Section IV of RCC-M concerns with qualification and performance of welding operations. Article S 3150 in this section covers the retesting procedures and article S 3152 in it relates to destructive tests. One sentence in Article S 3152 reads like this, 'Where an unsatisfactory result is due to the poor execution of the test or the presence of a defect in the specimen, the relevant result must be discarded and the test shall be repeated.' The 'presence of a defect' arose discrepancies in understanding the sentence because there is no definition in RCC-M to clarify their scopes such as how big a defect should be considered as a 'defect' that can invalidate the test results. Section 5.5 of PANE G-7-002 concerns the determination of critical brittle temperature. It stipulates that critical brittle temperature T k is a temperature at which the impact strength values must be higher than the specified ones. The specified criterion values of impact strength are related to the yield strength but it is not quite definite that the yield strength values in the table should be the documented minimum values for different materials or that they should be the tested results of the concerned lots of materials the specimen represents. The two different understandings of the yield strength values will sometimes give contrary conclusions. This paper introduces the different opinions in understanding the code texts, makes clear what it supports, and states in detail the reasons it so believes. (authors)

  10. RCC-M: Design and construction rules for mechanical components of PWR nuclear islands

    International Nuclear Information System (INIS)

    2017-01-01

    AFCEN's RCC-M code concerns the mechanical components designed and manufactured for pressurized water reactors (PWR). It applies to pressure equipment in nuclear islands in safety classes 1, 2 and 3, and certain non-pressure components, such as vessel internals, supporting structures for safety class components, storage tanks and containment penetrations. RCC-M covers the following technical subjects: sizing and design, choice of materials and procurement. Fabrication and control, including: associated qualification requirements (procedures, welders and operators, etc.), control methods to be implemented, acceptance criteria for detected defects, documentation associated with the different activities covered, and quality assurance. The design, manufacture and inspection rules defined in RCC-M leverage the results of the research and development work pioneered in France, Europe and worldwide, and which have been successfully used by industry to design and build PWR nuclear islands. AFCEN's rules incorporate the resulting feedback. Use: France's last 16 nuclear units (P'4 and N4); 4 CP1 reactors in South Africa (2) and Korea (2); 44 M310 (4), CPR-1000 (28), CPR-600 (6), HPR-1000 (4) and EPR (2) reactors in service or undergoing construction in China; 4 EPR reactors in Europe: Finland (1), France (1) and UK (2). Content: Section I - nuclear island components, subsection 'A': general rules, subsection 'B': class 1 components, subsection 'C': class 2 components, subsection 'D': class 3 components, subsection 'E': small components, subsection 'G': core support structures, subsection 'H': supports, subsection 'J': low pressure or atmospheric storage tanks, subsection 'P': containment penetration, subsection 'Q': qualification of active mechanical components, subsection 'Z': technical appendices; section II - materials; section III - examination

  11. 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

  12. 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.)

  13. 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)

  14. 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

  15. 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

  16. 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

  17. 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

  18. Coronin-1C and RCC2 guide mesenchymal migration by trafficking Rac1 and controlling GEF exposure

    Science.gov (United States)

    Williamson, Rosalind C.; Cowell, Christopher A. M.; Hammond, Christina L.; Bergen, Dylan J. M.; Roper, James A.; Feng, Yi; Rendall, Thomas C. S.; Race, Paul R.; Bass, Mark D.

    2014-01-01

    ABSTRACT Sustained forward migration through a fibrillar extracellular matrix requires localization of protrusive signals. Contact with fibronectin at the tip of a cell protrusion activates Rac1, and for linear migration it is necessary to dampen Rac1 activity in off-axial positions and redistribute Rac1 from non-protrusive membrane to the leading edge. Here, we identify interactions between coronin-1C (Coro1C), RCC2 and Rac1 that focus active Rac1 to a single protrusion. Coro1C mediates release of inactive Rac1 from non-protrusive membrane and is necessary for Rac1 redistribution to a protrusive tip and fibronectin-dependent Rac1 activation. The second component, RCC2, attenuates Rac1 activation outside the protrusive tip by binding to the Rac1 switch regions and competitively inhibiting GEF action, thus preventing off-axial protrusion. Depletion of Coro1C or RCC2 by RNA interference causes loss of cell polarity that results in shunting migration in 1D or 3D culture systems. Furthermore, morpholinos against Coro1C or RCC2, or mutation of any of the binding sites in the Rac1–RCC2–Coro1C complex delays the arrival of neural crest derivatives at the correct location in developing zebrafish, demonstrating the crucial role in migration guidance in vivo. PMID:25074804

  19. 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

  20. RCC-F: Design and construction rules for PWR fire protection systems

    International Nuclear Information System (INIS)

    2013-01-01

    The RCC-F code defines the rules for designing, building and installing the fire protection systems used to manage the nuclear hazards inherent in the outbreak of a fire inside the facility and thereby control the fundamental nuclear functions. The code provides fire protection recommendations in terms of: the industrial risk (loss of assets and/or operation), personnel safety, the environment. The code is divided into five main sections: generalities, design safety principles, fire protection design bases, construction provisions, rules for installing the fire protection components and equipment. The RCC-F code is available as an ETC-F version specifically for EPR projects (European pressurized reactor). Contents of the 2013 edition of the ETC-F code: Volume A - Generalities: Structure of ETC-F general points, documentation (in progress), chapter (provision) quality assurance; Volume B - Design safety principles: design nuclear safety principles; Volume C - Fire protection design bases: fire protection design bases; Volume D - Construction provisions: construction provisions; Volume E - Installation rules for fire protection: rules for installing the fire protection, components and equipment

  1. 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.)

  2. 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)

  3. Concrete radiation shielding

    International Nuclear Information System (INIS)

    Kaplan, M.F.

    1989-01-01

    The increased use of nuclear energy has given rise to a growth in the amount of artificially produced radiation and radioactive materials. The design and construction of shielding to protect people, equipment and structures from the effects of radiation has never been more important. Experience has shown that concrete is an effective, versatile and economical material for the construction of radiation shielding. This book provides information on the principles governing the interaction of radiation with matter and on relevant nuclear physics to give the engineer an understanding of the design and construction of concrete shielding. It covers the physical, mechanical and nuclear properties of concrete; the effects of elevated temperatures and possible damage to concrete due to radiation; basic procedures for the design of concrete radiation shields and finally the special problems associated with their construction and cost. Although written primarily for engineers concerned with the design and construction of concrete shielding, the book also reviews the widely scattered data and information available on this subject and should therefore be of interest to students and those wishing to research further in this field. (author)

  4. Method for dismantling shields

    International Nuclear Information System (INIS)

    Fukuzawa, Rokuro; Kondo, Nobuhiro; Kamiyama, Yoshinori; Kawasato, Ken; Hiraga, Tomoaki.

    1990-01-01

    The object of the present invention is to enable operators to dismantle shieldings contaminated by radioactivity easily and in a short period of time without danger of radiation exposure. A plurality of introduction pipes are embedded previously to the shielding walls of shielding members which contain a reactor core in a state where both ends of the introduction pipes are in communication with the outside. A wire saw is inserted into the introduction pipes to cut the shieldings upon dismantling. Then, shieldings can be dismantled easily in a short period of time with no radiation exposure to operator's. Further, according to the present invention, since the wire saw can be set easily and a large area can be cut at once, operation efficiency is improved. Further, since remote control is possible, cutting can be conducted in water and complicated places of the reactor. Biting upon starting the wire saw in the introduction pipe is reduced to facilitate startup for the rotation. (I.S.)

  5. Mechanical shielded hot cell

    International Nuclear Information System (INIS)

    Higgy, H.R.; Abdel-Rassoul, A.A.

    1983-01-01

    A plan to erect a mechanical shielded hot cell in the process hall of the Radiochemical Laboratory at Inchas is described. The hot cell is designed for safe handling of spent fuel bundles, from the Inchas reactor, and for dismantling and cutting the fuel rods in preparation for subsequent treatment. The biological shielding allows for the safe handling of a total radioactivity level up to 10,000 MeV-Ci. The hot cell consists of an α-tight stainless-steel box, connected to a γ-shielded SAS, through an air-lock containing a movable carriage. The α-box is tightly connected with six dry-storage cavities for adequate storage of the spent fuel bundles. Both the α-box, with the dry-storage cavities, and the SAS are surrounded by 200-mm thick biological lead shielding. The α-box is equipped with two master-slave manipulators, a lead-glass window, a monorail crane and Padirac and Minirag systems. The SAS is equipped with a lead-glass window, tong manipulator, a shielded pit and a mechanism for the entry of the spent fuel bundle. The hot cell is served by adequate ventilation and monitoring systems. (author)

  6. 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

    Full text of publication follows: Prototype Fast Breeder Reactor (PFBR) is a 500 MWe, sodium cooled, pool type fast reactor. The top shield forms the top cover for the main vessel (MV) and includes roof slab (RS), large rotatable plug (LRP), small rotatable plug (SRP) and control Plug (CP). RS, LRP and SRP are box type structures consisting of top and bottom plates stiffened by radial stiffeners and vertical penetration shells. TS is exposed to argon cover gas provided above sodium pool on the bottom side and reactor containment building air at the top. Heat transfer takes place through the argon cover gas to the bottom plate of TS. Annular gaps are formed between the components supported on TS and the component penetrations through which cellular convection takes place. A single thermal shield provided below TS reduces the heat flux to the bottom plate to 1.15 kW/m 2 . The MV (SS 316 LN) is welded to RS (carbon steel A48 P2) through a dissimilar metal weld. A step in RS and an anti convection barrier (ACB) outside RS are provided to limit the temperature at the MV-RS junction. The MV is surrounded by safety vessel (SV) and reactor vault made of concrete. Thermal insulation is provided outside SV to limit the heat transfer to the reactor vault. The design requirements of TS are to maintain the operating temperature at 383-393 K, limit the temperature difference (ΔT) across the height of TS to 20 / 100 K under normal operation/loss of cooling, provide minimum annular gap size at the component penetrations, provide a nearly linear temperature gradient in the CP portion within the height of TS, maintain the temperature of top plate of CP > 383 K, limit the ΔT across the top plate of CP to 2 K, limit the temperature near the inflatable / backup seal to 393 K, limit the temperature at the MV-RS junction and the heat flux to the reactor vault. The total heat transferred to TS is estimated to be 210 kW. A dedicated closed loop cooling system with a total flow rate of 10

  7. Creep-fatigue damage evaluation for SS-316LN (ORNL PLATES): - RCC-MR vs. ASME SEC III - NH

    International Nuclear Information System (INIS)

    Sati, Bhuwan Chandra; Jalaldeen, S.; Velusamy, K.; Selvaraj, P.

    2016-01-01

    Investigations of high temperature tests done on ORNL plate with deformation control loading, under creep-fatigue damage have been presented. The test results with methodology of RCC-MR and ASME-NH life prediction under creep-fatigue loading have been assessed. The stress relaxation effect in calculating the life using RCC-MR under creep-fatigue damage is found to be significant in presence of secondary stress. RCC-MR: 2007 is more realistic number of cycles (predicts 51 number of cycles) as compared to ASME-NH (predicts 312 number of cycles) which is demonstrated by the experimental work (observed 86 numbers of cycles). Between RCC-MR and experimental work, design code seems to be more conservative for life prediction due to creep-fatigue damage. For fatigue damage, the approaches are same and the difference comes from material properties and the starting stress for applying Neuber's rule. ASME approach has the limitation of stress range magnitude. ASME approach predicts lower elastic plus plastic strain for the cases having S* above the linear stress limit. For creep strain and creep damage evaluation, ASME and RCC-MR have different approaches for calculating the stress at the beginning and during the hold period. The RCC-MR takes account of cyclic hardening or softening effects (hardening in the present case of 316 LN) by means of the cyclic stress-strain curve and the benefit of symmetrization effects which are significant for this material. The ASME code neglects these effects and instead relies on an approach based on the isochronous stress-strain curves. (author)

  8. Shield support frame. Schildausbaugestell

    Energy Technology Data Exchange (ETDEWEB)

    Plaga, K.

    1981-09-17

    A powered shield support frame for coal sheds is described comprising of two bottom sliding shoes, a large area gob shield and a larg area roof assembly, all joined movable together. The sliding shoes and the gob shield are joined by a lemniscate guide. Two hydraulic props are arranged at the face-side at one third of the length of the sliding shoes and at the goaf-side at one third of the length of the roof assembly. A nearly horizontal lying pushing prop unit joins the bottom wall sliding shoes to the goaf-side lemniscate guide. This assembly can be applied to seams with a thickness down to 45 cm. (OGR).

  9. Radiation shielding material

    International Nuclear Information System (INIS)

    Matsumoto, Akio; Isobe, Eiji.

    1976-01-01

    Purpose: To increase the shielding capacity of the radiation shielding material having an abundant flexibility. Constitution: A mat consisting of a lead or lead alloy fibrous material is covered with a cloth, and the two are made integral by sewing in a kilted fashion by using a yarn. Thereafter, the system is covered with a gas-tight film or sheet. The shielding material obtained in this way has, in addition to the above merits, advantages in that (1) it is free from restoration due to elasticity so that it can readily seal contaminants, (2) it can be used in a state consisting of a number of overlapped layers, (3) it fits the shoulder well and is readily portable and (4) it permits attachment of fasteners or the like. (Ikeda, J.)

  10. Hybrid Magnetic Shielding

    Science.gov (United States)

    Royal, Kevin; Crawford, Christopher; Mullins, Andrew; Porter, Greg; Blanton, Hunter; Johnstone, Connor; Kistler, Ben; Olivera, Daniela

    2017-09-01

    The search for the electric dipole moment of the neutron requires the ambient magnetic field to be on the pT scale which is accomplished with large magnetic shielding rooms. These rooms are fitted with large mu-metal sheets to allow for passive cancellation of background magnetic fields. Active shielding technology cannot uniformly cancel background magnetic fields. These issues can be remedied by combining the methods into a hybrid system. The design used is composed of panels that have an active layer of cancellation between two sheets of mu-metal. The panels form a cube and draw in magnetic fields perpendicular to the surface which can then be reduced using active shielding. This work is supported by the Department of Energy under Contract DE-SC0008107.

  11. Reactor head shielding apparatus

    International Nuclear Information System (INIS)

    Schukei, G.E.; Roebelen, G.J.

    1992-01-01

    This patent describes a nuclear reactor head shielding apparatus for mounting on spaced reactor head lifting members radially inwardly of the head bolts. It comprises a frame of sections for mounting on the lifting members and extending around the top central area of the head, mounting means for so mounting the frame sections, including downwardly projecting members on the frame sections and complementary upwardly open recessed members for fastening to the lifting members for receiving the downwardly projecting members when the frame sections are lowered thereto with lead shielding supported thereby on means for hanging lead shielding on the frame to minimize radiation exposure or personnel working with the head bolts or in the vicinity thereof

  12. Double-layer neutron shield design as neutron shielding application

    Science.gov (United States)

    Sariyer, Demet; Küçer, Rahmi

    2018-02-01

    The shield design in particle accelerators and other high energy facilities are mainly connected to the high-energy neutrons. The deep penetration of neutrons through massive shield has become a very serious problem. For shielding to be efficient, most of these neutrons should be confined to the shielding volume. If the interior space will become limited, the sufficient thickness of multilayer shield must be used. Concrete and iron are widely used as a multilayer shield material. Two layers shield material was selected to guarantee radiation safety outside of the shield against neutrons generated in the interaction of the different proton energies. One of them was one meter of concrete, the other was iron-contained material (FeB, Fe2B and stainless-steel) to be determined shield thicknesses. FLUKA Monte Carlo code was used for shield design geometry and required neutron dose distributions. The resulting two layered shields are shown better performance than single used concrete, thus the shield design could leave more space in the interior shielded areas.

  13. 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)

  14. Using the shield for thermal energy storage in pulsar

    International Nuclear Information System (INIS)

    Sager, G.T.; Sze, D.K.; Wong, C.P.C.; Bathke, C.G.; Blanchard, J.P.; Brimer, C.; Cheng, E.T.; El-Guebaly, L.A.; Hasan, M.Z.; Najmabadi, F.; Sharafat, S.; Sviatoslavski, I.N.; Waganer, L.

    1995-01-01

    The PULSAR pulsed tokamak power plant design utilizes the outboard shield for thermal energy storage to maintain full 1000MW(e) output during the dwell period of 200s. Thermal energy resulting from direct nuclear heating is accumulated in the shield during the 7200s fusion power production phase. The maximum shield temperature may be much higher than that for the blanket because radiation damage is significantly reduced. During the dwell period, thermal power discharged from the shield and coolant temperature are simultaneously regulated by controlling the coolant mass flow rate at the shield inlet. This is facilitated by throttled coolant bypass. Design concepts using helium and lithium coolant have been developed. Two-dimensional time-dependent thermal hydraulic calculations were performed to confirm performance capabilities required of the design concepts. The results indicate that the system design and performance can accommodate uncertainties in material limits or the length of the dwell period. (orig.)

  15. Shielding research in France

    Energy Technology Data Exchange (ETDEWEB)

    Lafore, P

    1964-10-01

    Shielding research as an independent subject in France dates from 1956. The importance of these studies has been reflected in the contribution which they have made to power reactor design and in the resultant savings in expenditure for civil engineering and machinery for the removal of mobile shields. The Reactor Shielding Research Division numbers approximately 60 persons and uses several experimental facilities. These include: NAIADE I, installed near the ZOE reactor and operating with a natural uranium slab 2 cm thick (an effective diameter of 60 cm is the one most commonly used); the TRITON pool-type reactor, mainly used in shielding studies, includes an active-water loop, by means of which the secondary shields required for light-water reactors can be studied; core, NEREIDE, which is situated near a 2 m x 2 m aluminium window enables a large neutron source to be placed in a compartment without water in which large-scale mock-ups can be mounted for the study, in particular, of neutron diffusion in large cavities, and of reactor shielding of greater thickness than that in NAIADE I; SAMES 600 keV accelerator is used for monoenergetic neutron studies. Instrumentation studies are an important part of the work, mainly in the measurement of fast neutrons and their spectra by activation detectors. Of late, attention has been directed towards the use of (n, n') (rhodium) reactions and of heavy detectors for low-flux measurements. The simultaneous use of a large number of detectors poses automation problems. With our installation we can count 16 detectors simultaneously. Neutron spectrum studies are conducted with nuclear emulsions and a lithium-6 semiconductor spectrometer. As to the materials used, the research carried out in France involves chiefly graphite, iron and concrete at various temperatures up to 800 deg C. Different compounds, borated and non-borated and of densities up to between 1 and 9 are under consideration. Problems connected with applications are

  16. Radiation shielding bricks

    International Nuclear Information System (INIS)

    Crowe, G.J.W.

    1983-01-01

    A radiation shielding brick for use in building dry walls to form radiation proof enclosures and other structures is described. It is square in shape and comprises a sandwich of an inner layer of lead or similar shielding material between outer layers of plastics material, for structural stability. The ability to mechanically interlock adjacent bricks is provided by shaping the edges as cooperating external and internal V-sections. Relatively leak-free joints are ensured by enlarging the width of the inner layer in the edge region. (author)

  17. Hybrid Active-Passive Radiation Shielding System

    Data.gov (United States)

    National Aeronautics and Space Administration — A radiation shielding system is proposed that integrates active magnetic fields with passive shielding materials. The objective is to increase the shielding...

  18. Simulation of divertor targets shielding during transients in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Pestchanyi, Sergey, E-mail: serguei.pestchanyi@kit.edu [KIT, Hermann-von-Helmholtz-Platz 1, Eggenstein-Leopoldshafen (Germany); Pitts, Richard; Lehnen, Michael [ITER Organization,Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2016-11-01

    Highlights: • We simulated plasma shielding effect during disruption in ITER using the TOKES code. • It has been found that vaporization is unavoidable under action of ITER transients, but plasma shielding drastically reduces the divertor target damage: the melt pool and the vaporization region widths reduced 10–15 times. • A simplified 1D model describing the melt pool depth and the shielded heat flux to the divertor targets have been developed. • The results of the TOKES simulations have been compared with the analytic model when the model is valid. - Abstract: Direct extrapolation of the disruptive heat flux on ITER conditions predicts severe melting and vaporization of the divertor targets causing their intolerable damage. However, tungsten vaporized from the target at initial stage of the disruption can create plasma shield in front of the target, which effectively protects the target surface from the rest of the heat flux. Estimation of this shielding efficiency has been performed using the TOKES code. The shielding effect under ITER conditions is found to be very strong: the maximal depth of the melt layer reduced 4 times, the melt layer width—more than 10 times and vaporization region shrinks 10–15 times due to shielding for unmitigated disruption of 350 MJ discharge. The simulation results show complex, 2D plasma dynamics of the shield under ITER conditions. However, a simplified analytic model, valid for rough estimation of the maximum value for the shielded flux to the target and for the melt depth at the target surface has been developed.

  19. Radiation shielding cloth

    International Nuclear Information System (INIS)

    Ijiri, Yasuo; Fujinuma, Tadashi; Tamura, Shoji.

    1989-01-01

    Radiation shielding cloth having radiation shielding layers comprising a composition of inorganic powder of high specific gravity and rubber are excellentin flexibility and comfortable to put on. However, since they are heavy in the weight, operators are tired upon putting them for a long time. In view of the above, the radiation ray shielding layers are prepared by calendering sheets obtained by preliminary molding of the composition to set the variation of the thickness within a range of +15% to -0% of prescribed thickness. Since the composition of inorganic powder at high specific gravity and rubber used for radiation ray shielding comprises a great amount of inorganic powder at high specific gravity blended therein, it is generally poor in fabricability. Therefor, it is difficult to attain fine control for the sheet thickness by merely molding a composition block at once. Then, the composition is at first preliminarily molded into a sheet-like shape which is somewhat thickener than the final thickness and then finished by calendering, by which the thickness can be reduced in average as compared with conventional products while keeping the prescribed thickness and reducing the weight reduce by so much. (N.H.)

  20. 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.

  1. Penetration portion shielding structure

    International Nuclear Information System (INIS)

    Hayashi, Katsumi; Narita, Hitoshi; Handa, Hiroyuki; Takeuchi, Jun; Tozuka, Fumio.

    1994-01-01

    Openings of a plurality of shieldings for penetration members are aligned to each other, and penetration members are inserted from the openings. Then, the openings of the plurality of shielding members are slightly displaced with each other to make the penetration portions into a helical configuration, so that leakage of radiation is reduced. Upon removal of the members, reverse operation is conducted. When a flowable shielding material is used, the penetration portions are constituted with two plates having previously formed openings and pipes for connecting the openings with each other and a vessel covering the entire of them. After passing the penetration members such as a cable, the relative position of the two plates is changed by twisting, to form a helical configuration which reduces radiation leakage. Since they are bent into the helical configuration, shielding performance is extremely improved compared with a case that radiation leakage is caused from an opening of a straight pipe. In addition, since they can be returned to straight pipes, attachment, detachment and maintenance can be conducted easily. (N.H.)

  2. Dosimetry and shielding

    International Nuclear Information System (INIS)

    Farinelli, U.

    1977-01-01

    Today, reactor dosimetry and shielding have wide areas of overlap as concerns both problems and methods. Increased interchange of results and know-how would benefit both. The areas of common interest include calculational methods, sensitivity studies, theoretical and experimental benchmarks, cross sections and other nuclear data, multigroup libraries and procedures for their adjustment, experimental techniques and damage functions. This paper reviews the state-of-the-art and the latest development in each of these areas as far as shielding is concerned, and suggests a number of interactions that could be profitable for reactor dosimetry. Among them, re-evaluation of the potentialities of calculational methods (in view of the recent developments) in predicting radiation environments of interest; the application of sensitivity analysis to dosimetry problems; a common effort in the field of theoretical benchmarks; the use of the shielding one-material propagation experiments as reference spectra for detector cross sections; common standardization of the detector nuclear data used in both fields; the setting up of a common (or compatible) multigroup structure and library applicable to shielding, dosimetry and core physics; the exchange of information and experience in the fields of cross section errors, correlations and adjustment; and the intercomparison of experimental techniques

  3. Radiation shielding glass

    International Nuclear Information System (INIS)

    Kido, Kazuhiro; Ueda, Hajime.

    1997-01-01

    It was found that a glass composition comprising, as essential ingredients, SiO 2 , PbO, Gd 2 O 3 and alkali metal oxides can provide a shielding performance against electromagnetic waves, charged particles and neutrons. The present invention provides radiation shielding glass containing at least from 16 to 46wt% of SiO 2 , from 47 to 75wt% of PbO, from 1 to 10wt% of Gd 2 O 3 , from 0 to 3wt% of Li 2 O, from 0 to 7wt% of Na 2 O, from 0 to 7wt% of K 2 O provided that Li 2 O + Na 2 O + K 2 O is from 1 to 10wt%, B 2 O 3 is from 0 to 10wt%, CeO 2 is from 0 to 3wt%, As 2 O 3 is from 0 to 1wt% and Sb 2 O 3 is from 0 to 1wt%. Since the glass can shield electromagnetic waves, charged particles and neutrons simultaneously, radiation shielding windows can be designed and manufactured at a reduced thickness and by less constitutional numbers in a circumstance where they are present altogether. (T.M.)

  4. Dynamic contrast-enhanced CT (DCE-CT) as a potential biomarker in patients with metastatic renal cell carcinoma (mRCC)

    DEFF Research Database (Denmark)

    Mains, Jill Rachel; Donskov, Frede; Pedersen, Erik Morre

    Purpose To explore the impact of DCE-CT as a biomarker in mRCC.  Methods and Materials 12 patients with mRCC participating in a phase II trial with immunotherapy and bevacizumab and with a follow-up time of at least 2 years were included in this preliminary analysis. DCE-CT interpretation (max s...

  5. Gonad shielding in diagnostic radiology

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    The use of gonad shielding is an important radiation protection technique, intended to reduce unnecessary x-ray exposure of the gonads of patients from diagnostic x-ray procedures. The types of gonad shields in use are discussed as are the types of diagnostic examinations that should include gonad shielding. It was found that when properly used, most shields provided substantial gonad dose reductions

  6. Shielding experiments for accelerator facilities

    Energy Technology Data Exchange (ETDEWEB)

    Nakashima, Hiroshi; Tanaka, Susumu; Sakamoto, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    2000-06-01

    A series of shielding experiments was carried out by using AVF cyclotron accelerator of TIARA at JAERI in order to validate shielding design methods for accelerator facilities in intermediate energy region. In this paper neutron transmission experiment through thick shields and radiation streaming experiment through a labyrinth are reported. (author)

  7. Shielding experiments for accelerator facilities

    International Nuclear Information System (INIS)

    Nakashima, Hiroshi; Tanaka, Susumu; Sakamoto, Yukio

    2000-01-01

    A series of shielding experiments was carried out by using AVF cyclotron accelerator of TIARA at JAERI in order to validate shielding design methods for accelerator facilities in intermediate energy region. In this paper neutron transmission experiment through thick shields and radiation streaming experiment through a labyrinth are reported. (author)

  8. SEISMIC B E H AV I OUR OF R.C.C BUILDING WITH AND WITHOUT FLOATING COLUMNS

    OpenAIRE

    E. ANUSHA; E. V. RAGHVA RAO; N. CHENNA KESAVA

    2018-01-01

    The main purpose of the project is to study the seismic behaviour of R.C.C building with and without floating column,G+5 structures has been selected for carrying out the project work. The building models are generated using software STAAD.

  9. Stratification of clear cell renal cell carcinoma (ccRCC) genomes by gene-directed copy number alteration (CNA) analysis.

    Science.gov (United States)

    Thiesen, H-J; Steinbeck, F; Maruschke, M; Koczan, D; Ziems, B; Hakenberg, O W

    2017-01-01

    Tumorigenic processes are understood to be driven by epi-/genetic and genomic alterations from single point mutations to chromosomal alterations such as insertions and deletions of nucleotides up to gains and losses of large chromosomal fragments including products of chromosomal rearrangements e.g. fusion genes and proteins. Overall comparisons of copy number alterations (CNAs) presented in 48 clear cell renal cell carcinoma (ccRCC) genomes resulted in ratios of gene losses versus gene gains between 26 ccRCC Fuhrman malignancy grades G1 (ratio 1.25) and 20 G3 (ratio 0.58). Gene losses and gains of 15762 CNA genes were mapped to 795 chromosomal cytoband loci including 280 KEGG pathways. CNAs were classified according to their contribution to Fuhrman tumour gradings G1 and G3. Gene gains and losses turned out to be highly structured processes in ccRCC genomes enabling the subclassification and stratification of ccRCC tumours in a genome-wide manner. CNAs of ccRCC seem to start with common tumour related gene losses flanked by CNAs specifying Fuhrman grade G1 losses and CNA gains favouring grade G3 tumours. The appearance of recurrent CNA signatures implies the presence of causal mechanisms most likely implicated in the pathogenesis and disease-outcome of ccRCC tumours distinguishing lower from higher malignant tumours. The diagnostic quality of initial 201 genes (108 genes supporting G1 and 93 genes G3 phenotypes) has been successfully validated on published Swiss data (GSE19949) leading to a restricted CNA gene set of 171 CNA genes of which 85 genes favour Fuhrman grade G1 and 86 genes Fuhrman grade G3. Regarding these gene sets overall survival decreased with the number of G3 related gene losses plus G3 related gene gains. CNA gene sets presented define an entry to a gene-directed and pathway-related functional understanding of ongoing copy number alterations within and between individual ccRCC tumours leading to CNA genes of prognostic and predictive value.

  10. Shielding Design and Radiation Shielding Evaluation for LSDS System Facility

    International Nuclear Information System (INIS)

    Kim, Younggook; Kim, Jeongdong; Lee, Yongdeok

    2015-01-01

    As the system characteristics, the target in the spectrometer emits approximately 1012 neutrons/s. To efficiently shield the neutron, the shielding door designs are proposed for the LSDS system through a comparison of the direct shield and maze designs. Hence, to guarantee the radiation safety for the facility, the door design is a compulsory course of the development of the LSDS system. To improve the shielding rates, 250x250 covering structure was added as a subsidiary around the spectrometer. In this study, the evaluations of the suggested shielding designs were conducted using MCNP code. The suggested door design and covering structures can shield the neutron efficiently, thus all evaluations of all conditions are satisfied within the public dose limits. From the Monte Carlo code simulation, Resin(Indoor type) and Tungsten(Outdoor type) were selected as the shielding door materials. From a comparative evaluation of the door thickness, In and Out door thickness was selected 50 cm

  11. RCC-MRx: Design and construction rules for mechanical components in high-temperature structures, experimental reactors and fusion reactors

    International Nuclear Information System (INIS)

    2015-01-01

    The RCC-MRx code was developed for sodium-cooled fast reactors (SFR), research reactors (RR) and fusion reactors (FR-ITER). It provides the rules for designing and building mechanical components involved in areas subject to significant creep and/or significant irradiation. In particular, it incorporates an extensive range of materials (aluminum and zirconium alloys in response to the need for transparency to neutrons), sizing rules for thin shells and box structures, and new modern welding processes: electron beam, laser beam, diffusion and brazing. The RCC-MR code was used to design and build the prototype Fast Breeder Reactor (PFBR) developed by IGCAR in India and the ITER Vacuum Vessel. The RCC-Mx code is being used in the current construction of the RJH experimental reactor (Jules Horowitz reactor). The RCC-MRx code is serving as a reference for the design of the ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration), for the design of the primary circuit in MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) and the design of the target station of the ESS project (European Spallation Source). Contents of the 2015 edition of the RCC-MRx code: Section I General provisions; Section II Additional requirements and special provisions; Section III Rules for nuclear installation mechanical components: Volume I: Design and construction rules: Volume A (RA): General provisions and entrance keys, Volume B (RB): Class 1 components and supports, Volume C (RC): Class 2 components and supports, Volume D (RD): Class 3 components and supports, Volume K (RK): Examination, handling or drive mechanisms, Volume L (RL): Irradiation devices, Volume Z (Ai): Technical appendices; Volume II: Materials; Volume III: Examinations methods; Volume IV: Welding; Volume V: Manufacturing operations; Volume VI: Probationary phase rules

  12. External dosimetry sources and shielding

    International Nuclear Information System (INIS)

    Calisto, Washington

    1994-01-01

    A definition of external dosimetry r external sources dosimetry,physical and mathematical treatment of the interaction of gamma radiation with a minimal area in that direction. Concept of attenuation coefficient, cumulated effect by polyenergetic sources, exposition rate, units, cumulated dose,shielding, foton shielding, depth calculation, materials used for shielding.Beta shielding, consideration of range and maximum β energy , low stopping radiation by use of low Z shielding. Tables for β energy of β emitters, I (tau) factor, energy-range curves for β emitters in aqueous media, gamma attenuation factors for U, W and Pb. Y factor for bone tissue,muscle and air, build-up factors

  13. Radiation shielding calculation using MCNP

    International Nuclear Information System (INIS)

    Masukawa, Fumihiro

    2001-01-01

    To verify the Monte Carlo code MCNP4A as a tool to generate the reference data in the shielding designs and the safety evaluations, various shielding benchmark experiments were analyzed using this code. These experiments were categorized in three types of the shielding subjects; bulk shielding, streaming, and skyshine. For the variance reduction technique, which is indispensable to get meaningful results with the Monte Carlo shielding calculation, we mainly used the weight window, the energy dependent Russian roulette and spitting. As a whole, our analyses performed enough small statistical errors and showed good agreements with these experiments. (author)

  14. 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)

  15. Radiation shielding wall structure

    International Nuclear Information System (INIS)

    Nishimura, Yoshitaka; Oka, Shinji; Kan, Toshihiko; Misato, Takeshi.

    1990-01-01

    A space between a pair of vertical steel plates laterally disposed in parallel at an optional distance has a structure of a plurality of vertically extending tranks partitioned laterally by vertically placed steel plates. Then, cements are grouted to the tranks. Strip-like steel plates each having a thickness greater than the gap between the each of the vertically placed steel plates and the cement are bonded each at the surface for each of the vertically placed steel plates opposing to the cements. A protrusion of a strip width having radiation shielding performance substantially identical with that by the thickness of the cement is disposed in the strip-like steel plates. With such a constitution, a safety radiation shielding wall structure with no worry of radiation intrusion to gaps, if formed, between the steel plates and the grouted cements due to shrinkage of the cements. (I.N.)

  16. Radiation shielding material

    International Nuclear Information System (INIS)

    Kawakubo, Takamasa; Yamada, Fumiyuki; Nakazato, Kenjiro.

    1976-01-01

    Purpose: To provide a material, which is used for printing a samples name and date on an X-ray photographic film at the same time an X-ray radiography. Constitution: A radiation shielding material of a large mass absorption coefficient such as lead oxide, barium oxide, barium sulfate, etc. is added to a solution of a radiation permeable substance capable of imparting cold plastic fluidity (such as microcrystalline wax, paraffin, low molecular polyethylene, polyvinyl chloride, etc.). The resultant system is agitated and then cooled, and thereafter it is press fitted to or bonded to a base in the form of a film of a predetermined thickness. This radiation shielding layer is scraped off by using a writing tool to enter information to be printed in a photographic film, and then it is laid over the film and exposed to X-radiation to thereby print the information on the film. (Seki, T.)

  17. Multilayer radiation shield

    Science.gov (United States)

    Urbahn, John Arthur; Laskaris, Evangelos Trifon

    2009-06-16

    A power generation system including: a generator including a rotor including a superconductive rotor coil coupled to a rotatable shaft; a first prime mover drivingly coupled to the rotatable shaft; and a thermal radiation shield, partially surrounding the rotor coil, including at least a first sheet and a second sheet spaced apart from the first sheet by centripetal force produced by the rotatable shaft. A thermal radiation shield for a generator including a rotor including a super-conductive rotor coil including: a first sheet having at least one surface formed from a low emissivity material; and at least one additional sheet having at least one surface formed from a low emissivity material spaced apart from the first sheet by centripetal force produced by the rotatable shaft, wherein each successive sheet is an incrementally greater circumferential arc length and wherein the centripetal force shapes the sheets into a substantially catenary shape.

  18. Light shielding apparatus

    Science.gov (United States)

    Miller, Richard Dean; Thom, Robert Anthony

    2017-10-10

    A light shielding apparatus for blocking light from reaching an electronic device, the light shielding apparatus including left and right support assemblies, a cross member, and an opaque shroud. The support assemblies each include primary support structure, a mounting element for removably connecting the apparatus to the electronic device, and a support member depending from the primary support structure for retaining the apparatus in an upright orientation. The cross member couples the left and right support assemblies together and spaces them apart according to the size and shape of the electronic device. The shroud may be removably and adjustably connectable to the left and right support assemblies and configured to take a cylindrical dome shape so as to form a central space covered from above. The opaque shroud prevents light from entering the central space and contacting sensitive elements of the electronic device.

  19. Shielding container for radioactive isotopes

    International Nuclear Information System (INIS)

    Sumi, Tetsuo; Tosa, Masayoshi; Hatogai, Tatsuaki.

    1975-01-01

    Object: To effect opening and closing bidirectional radiation used particularly for a gamma densimeter or the like by one operation. Structure: This device comprises a rotatable shielding body for receiving radioactive isotope in the central portion thereof and having at least two radiation openings through which radiation is taken out of the isotope, and a shielding container having openings corresponding to the first mentioned radiation openings, respectively. The radioactive isotope is secured to a rotational shaft of the shielding body, and the shielding body is rotated to register the openings of the shielding container with the openings of the shielding body or to shield the openings, thereby effecting radiation and cut off of gamma ray in the bidirection by one operation. (Kamimura, M.)

  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. Shielding calculations using FLUKA

    International Nuclear Information System (INIS)

    Yamaguchi, Chiri; Tesch, K.; Dinter, H.

    1988-06-01

    The dose equivalent on the surface of concrete shielding has been calculated using the Monte Carlo code FLUKA86 for incident proton energies from 10 to 800 GeV. The results have been compared with some simple equations. The value of the angular dependent parameter in Moyer's equation has been calculated from the locations where the values of the maximum dose equivalent occur. (author)

  2. Revised neutral gas shielding model for pellet ablation - combined neutral and plasma shielding

    International Nuclear Information System (INIS)

    Houlberg, W.A.; Schuresko, D.D.; Attenberger, S.E.

    1986-01-01

    The ablation and penetration of pellets in early ORMAK and ISX-A experiments were reliably predicted by the neutral gas shielding model of Milora and Foster. These experiments demonstrated that the principle components of the model - a self-generated shield which reduces the heat flux at the plasma surface - were correct. In more recent experiments with higher temperature plasmas, this model consistently predicts greater penetration than observed in the experiments. Upgarding known limitations of the original model brings the predicted and observed penetration values into agreement. These improvements include: (1) treating the incident electrons as having distribution in energy rather than being monoenergetic; (2) including the shielding effects of cold, dense plasma extending along the magnetic field outside the neutral shield; and (3) modifying the finite plasma, self-limiting incident heat flux so that it represents a collisionless plasma limit rather than a collisional limit. Comparisons are made between the models for a selection of ISX-B Alcator-C, and TFTR shots. The net effect of the changes in the model is an increase in pellet ablation rates and decrease in penetration for current and future experiments

  3. Thermal shielding device in LMFBR type reactors

    International Nuclear Information System (INIS)

    Nakamura, Hiroshi.

    1985-01-01

    Purpose: To improve the soundness and earthquake proofness of mounting structures to a reactor vessel in a thermal shielding device comprising a plurality of tightly closed casings evacuated or shield with heat insulation gases, by reducing the wall thickness and weight of the casing. Constitution: the thermal shielding body comprises tightly closed casings and compressing core materials for preventing the deformation of the casings. The tightly closed casing is in the shape of a hollow vessel, completely sealed in gastight manner, and evacuated or sealed with heat insulation gases at a low pressure of about less than 0.5 kg/cm 2 G, such that the inner pressure is lower than the outer pressure. Compressing core materials made of porous metals or porous ceramics are contained to the inside of the casing. In this way, the wall thickness of the tightly closed casing can be reduced significantly as compared with the conventional case, whereby the mounting work on the site to the reactor container on the field can remarkably be improved and high reliability can be maintained at the mounting portion. (Kamimura, M.)

  4. 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)

  5. Muon shielding for PEP

    International Nuclear Information System (INIS)

    Jenkins, T.M.; Thomas, R.H.

    1974-01-01

    The first stage of construction of PEP will consist of electron and positron storage rings. At a later date a 200 GeV proton storage ring may be added. It is judicious therefore, to ensure that the first and second phases of construction are compatible with each other. One of several factors determining the elevation at which the storage rings will be constructed is the necessity to provide adequate radiation shielding. The overhead shielding of PEP is determined by the reproduction of neutrons in the hadron cascade generated by primary protons lost from the storage ring. The minimum overburden planned for PEP is 5.5 meters of earth (1100 gm cm/sup /minus/2/). To obtain a rough estimate of the magnitude of the muon radiation problem this note presents some preliminary calculations. Their purpose is intended merely to show that the presently proposed design for PEP will present no major shielding problems should the protons storage ring be installed. More detailed calculations will be made using muon yield computer codes developed at CERN and NAL and muon transport codes developed at SLAC, when details of the proton storage ring become settled. 9 refs., 4 figs

  6. Shielding calculations for NET

    International Nuclear Information System (INIS)

    Verschuur, K.A.; Hogenbirk, A.

    1991-05-01

    In the European Fusion Technology Programme there is only a small activity on research and development for fusion neutronics. Never-the-less, looking further than blanket design now, as ECN is getting involved in design of radiation shields for the coils and biological shields, it becomes apparent that fusion neutronics as a whole still needs substantial development. Existing exact codes for calculation of complex geometries like MCNP and DORT/TORT are put over the limits of their numerical capabilities, whilst approximate codes for complex geometries like FURNACE and MERCURE4 are put over the limits of their modelling capabilities. The main objective of this study is just to find out how far we can get with existing codes in obtaining reliable values for the radiation levels inside and outside the cryostat/shield during operation and after shut-down. Starting with a 1D torus model for preliminary parametric studies, more dimensional approximation of the torus or parts of it including the main heterogeneities should follow. Regular contacts with the NET-Team are kept, to be aware of main changes in NET design that might affect our calculation models. Work on the contract started 1 July 1990. The technical description of the contract is given. (author). 14 refs.; 4 figs.; 1 tab

  7. Upper shielding body in LMFBR type reactors

    International Nuclear Information System (INIS)

    Shoji, Koichi.

    1986-01-01

    Purpose: Preference is given to the strength and thermal insulation of a roof slab thereby ensuring axial size and improving the operationability upon inserting the control rod in the upper shielding body of LMFBR type reactors. Constitution: In an upper shielding body in which a large rotational plug is rotatably mounted to a circular hole formed at an eccentric position of a roof slab, while a small rotational plug is rotatably mounted to a circular hole disposed at an eccentric position of the large rotational plug and the reactor core upper mechanisms are supported on the small rotational plug, heat insulation layers are attached to the inside of the inner circumferential wall of the roof slab and the outer circumferential wall of the large rotational plug. By attaching the heat insulation layers, the heat conduction between the roof slab and the large rotational plug can be suppressed remarkably, by which occurrence of specific heat pass or local generation of large thermal stresses can be avoided even if difference is resulted to the temperature distribution between them. In this way, functions taking advantage of respective features of the roof slab and the small rotational plug can be obtained to achieve the purpose. (Kamimura, M.)

  8. Surgical treatment of Renal Cell Carcinoma (RCC with level III–IV tumor venous thrombosis

    Directory of Open Access Journals (Sweden)

    M. I. Davydov

    2016-01-01

    Full Text Available Objective: to assess the results of nephrectomy, thrombectomy in RCC patients with level III–IV tumor venous thrombosis with and without cardiopulmonary bypass.Materials and methods. Medical data of 167 consecutive RCC patients with level III–IV tumor venous thrombosis underwent nephrectomy thrombectomy in N.N. Blokhin Russian Cancer Research Center between 1998 and 2012 were collected. Right side tumor was in 122 (73.1 %, left side – in 42 (25.1 %, bilateral – in 3 (1.8 % cases. The extent of thrombus was defined as intrahepatic in 82 (49.1 %, supradiaphragmatic – in 85 (50.9 % (intrapericardial – in 44 (26.3 %, intraatrial – in 39 (23.4 %, intraventricular – in 2 (1.2 % cases. Nephrectomy, thrombectomy with cardiopulmonary bypass was used in 9 (5.4 %, 158 (94.6 % patients underwent radical nephrectomy with thrombectomy without CPBP and sternotomy. Intrapericardial IVC and right atrium were exposed through transdiaphragmatic approach and providing vascular control over infradiaphragmatic IVC and renal veins.Results. Median blood loss was 6000 (600–27 000 ml. Complications rate was 62.8 %, 90-day mortality – 13.2 %. Intraoperative complications were registered in 80 (47.9 %, postoperative – in 66 (40.5 % (grade II – 16 (9.8 %, grade IIIb – 1 (0.6 %, grade IVа – 28 (17.2 %, grade IVb – 3 (1.8 %, grade V – 18 (11.1 % patients. Modified thrombectomy technique insignificantly decreased blood loss compared to thrombectomy with CPB, did nоt increase complications rate including pulmonary vein thromboembolism, or mortality. Five-year overall, cancer-specific and recurrence-free survival was 46.2, 58.3 and 47.1 %, respectively. Thrombectomy technique did nоt affect survival.Conclusion. In selected patients with mobile thrombi transdiaphragmatic approach allows to avoid the use of CPBP and decrease surgical morbidity without survival compromising.

  9. 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

  10. 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.

  11. Nuclear shielding of openings in ITER Tokamak building

    Energy Technology Data Exchange (ETDEWEB)

    Dammann, A., E-mail: alexis.dammann@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Arumugam, A.P.; Beaudoin, V.; Beltran, D.; Benchikhoune, M.; Berruyer, F.; Cortes, P.; Gandini, F. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ghirelli, N. [ASSYSTEM E.O.S, ZAC Saint Martin, 23, rue Benjamin Franklin, 84120 Pertuis (France); Gray, A.; Hurzlmeier, H.; Le Page, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Lemée, A. [SOGETI High Tech, 180 Rue René Descartes, 13851 Aix en Provence (France); Lentini, G.; Loughlin, M.; Mita, Y.; Patisson, L.; Rigoni, G.; Rathi, D.; Song, I. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► Establishment of a methodology to design shielded opening in external wall of the Tokamak building. ► Analysis of the shielding requirement, case by case, depending on the localization and the context. ► Implementation of an integrated solution for shielded opening. -- Abstract: The external walls of the Tokamak building, made of thick concrete, provide the nuclear shielding for operators working in adjacent buildings and for the environment. There are a series of openings to these external walls, devoted to ducts or pipes for ventilation, waveguides and transmission lines for heating systems and diagnostics, cooling pipes, cable trays or busbars. The shielding properties of the wall shall be preserved by adequate design of the openings in order not to affect the radiological zoning in adjacent areas. For some of them, shielding properties of the wall are not affected because the size of the network is quite small or the source is far from the opening. But for most of the openings, specific features shall be considered. Even if the approach is the same and the ways to shield can be standardized, specific analysis is requested in any case because the constraints are different.

  12. 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.

  13. A contribution to shielding effectiveness analysis of shielded tents

    Directory of Open Access Journals (Sweden)

    Vranić Zoran M.

    2004-01-01

    Full Text Available An analysis of shielding effectiveness (SE of the shielded tents made of the metallised fabrics is given. First, two electromagnetic characteristic fundamental for coupling through electrically thin shield, the skin depth break frequency and the surface resistance or transfer impedance, is defined and analyzed. Then, the transfer function and the SE are analyzed regarding to the frequency range of interest to the Electromagnetic Compatibility (EMC Community.

  14. Light-refractory radiation shielding materials using diatomites and zeolites

    International Nuclear Information System (INIS)

    Murakami, Hideki

    2005-01-01

    It has been recently shown that diatomites and zeolites have some useful characteristics for radiation shielding materials. In this study, the availability of these materials for unexpected accidents in the nuclear sites is examined. The diatomites and zeolites, compared to existing shielding materials, have superior characteristics; low density and light weight, low in radiation-induced problem, high-heat resistance, remain unaltered by the addition of an acid except hydrofluoric acid, porous and large specific surface area, and also excellent water-absorbing property. These porous materials could also expand the shielding energy range applied and be used for fast- and thermal-neutrons, and γ ray. In addition, these materials are easy to store for long periods of time against emergency because of their natural rocks. From the examinations, it is cleared that diatomites and zeolites have excellent properties as radiation shielding materials for emergency use. (author)

  15. Feasibility of a superconducting FED with 50 cm of magnet shielding

    International Nuclear Information System (INIS)

    Schultz, J.H.; Montgomery, D.B.

    1981-01-01

    The feasibility of the suggestion that the cost of a Fusion Energy Device (FED) could be substantially reduced by operating with a reduced duty factor and only 50 cm of magnet shielding is evaluated here. This report examines the effect of light shielding on insulation life, matrix- and superconductor properties, refrigerator cost and steady-state heat removal. With very careful design, it appears feasible to build a device with only 50 cm of shielding

  16. Fatigue assessment by the RCC-MR design rules: remarks on the elastic analysis

    International Nuclear Information System (INIS)

    Taleb, L.; Sidoroff, F.

    1999-01-01

    According to RCC--MR (French rules for mechanical engineering design of FBR), fatigue life assessment is based on the evaluation of the equivalent elastoplastic strain range resulting from a given cyclic loading. Two methods can be used according to whether an elastoplastic or an elastic structure analysis is performed. The elastic analysis is of course more attractive for it avoids a heavy iterative elastoplastic analysis and an expensive identification of the material behavior from mechanical tests. On the other hand it relies on some empirical extrapolation rules from the elastic to the real case. The purpose of the present paper is to draw attention to some limitations of this procedure. In particular attention will be focused on two points: 1, the classification of the applied stress into primary and secondary parts is essential and it is shown that the thermal stresses which are often considered as secondary may in some cases play a primary role; 2. the Neuber's rule which is used to evaluate the plastic strain from the elastic stress will be shown to be significantly wrong for some special configurations. This is in fact essentially related to situations where the elastic follow up effect is important. (authors)

  17. Wing Leading Edge RCC Rapid Response Damage Prediction Tool (IMPACT2)

    Science.gov (United States)

    Clark, Robert; Cottter, Paul; Michalopoulos, Constantine

    2013-01-01

    This rapid response computer program predicts Orbiter Wing Leading Edge (WLE) damage caused by ice or foam impact during a Space Shuttle launch (Program "IMPACT2"). The program was developed after the Columbia accident in order to assess quickly WLE damage due to ice, foam, or metal impact (if any) during a Shuttle launch. IMPACT2 simulates an impact event in a few minutes for foam impactors, and in seconds for ice and metal impactors. The damage criterion is derived from results obtained from one sophisticated commercial program, which requires hours to carry out simulations of the same impact events. The program was designed to run much faster than the commercial program with prediction of projectile threshold velocities within 10 to 15% of commercial-program values. The mathematical model involves coupling of Orbiter wing normal modes of vibration to nonlinear or linear springmass models. IMPACT2 solves nonlinear or linear impact problems using classical normal modes of vibration of a target, and nonlinear/ linear time-domain equations for the projectile. Impact loads and stresses developed in the target are computed as functions of time. This model is novel because of its speed of execution. A typical model of foam, or other projectile characterized by material nonlinearities, impacting an RCC panel is executed in minutes instead of hours needed by the commercial programs. Target damage due to impact can be assessed quickly, provided that target vibration modes and allowable stress are known.

  18. Study on Flexural Behaviour of Ternary Blended Reinforced Self Compacting Concrete Beam with Conventional RCC Beam

    Science.gov (United States)

    Marshaline Seles, M.; Suryanarayanan, R.; Vivek, S. S.; Dhinakaran, G.

    2017-07-01

    The conventional concrete when used for structures having dense congested reinforcement, the problems such as external compaction and vibration needs special attention. In such case, the self compacting concrete (SCC) which has the properties like flow ability, passing and filling ability would be an obvious answer. All those SCC flow behavior was governed by EFNARC specifications. In present study, the combination type of SCC was prepared by replacing cement with silica fume (SF) and metakaolin (MK) along with optimum dosages of chemical admixtures. From the fresh property test, cube compressive strength and cylinder split tensile strength, optimum ternary mix was obtained. In order to study the flexural behavior, the optimum ternary mix was taken in which beam specimens of size 1200 mm x 100 mm x 200 mm was designed as singly reinforced section according to IS: 456-2000, Limit state method. Finally the comparative experimental analysis was made between conventional RCC and SCC beams of same grade in terms of flexural strength namely yield load & ultimate load, load- deflection curve, crack size and pattern respectively.

  19. Longterm results in the therapy of 100 cases of metastatic renal cell carcinoma (RCC)

    International Nuclear Information System (INIS)

    Falk, W.; Halama, J.M.; Halama, J.

    1990-01-01

    The success of renal cell carcinoma (RCC)-nephrectomy with radical lymph node dissection in stage I and II disease is undisputed. Through these measures 23% of metastases are controlled. The five-year survival time in stage III disease, however, stagnates at 35%±14% despite radical surgery. Also, the additional tumor-vaccine-therapy of the Mainz-Joint-Study-Group was successful only in stage I and II disease, whereas stage III disease did not benefit from this therapy. As 50% of all radically operated patients developed metastases within three years after surgery, the call by radiooncologists for supplementary radiotherapy beginning with stage III disease must be put foreward. The problems of therapy and chances of survival in generalized disease are demonstrated in 100 of our cases treated by surgery, radiotherapy and with MPA (medroxyprogesteroneacetate). Whereas Schmiedt et al. show a total survival time of 10.3 months after diagnosis of metastatic disease, the Offenbach patients achieved 16.5 months with a median survival time of 11.75 months. The necessity of therapeutic intervention is confirmed by the fact that the most favorable median survival time, 15.75 months, was achieved in metastatic disease involving three organs. We present here the special features of the individual organ manifestations and point out that not only the mean and median survival time, but also the very widely varying survival times in individual cases, make conscientious oncological post-treatment follow up and management a requirement. (orig.) [de

  20. Radiation shielding quality assurance

    Science.gov (United States)

    Um, Dallsun

    For the radiation shielding quality assurance, the validity and reliability of the neutron transport code MCNP, which is now one of the most widely used radiation shielding analysis codes, were checked with lot of benchmark experiments. And also as a practical example, follows were performed in this thesis. One integral neutron transport experiment to measure the effect of neutron streaming in iron and void was performed with Dog-Legged Void Assembly in Knolls Atomic Power Laboratory in 1991. Neutron flux was measured six different places with the methane detectors and a BF-3 detector. The main purpose of the measurements was to provide benchmark against which various neutron transport calculation tools could be compared. Those data were used in verification of Monte Carlo Neutron & Photon Transport Code, MCNP, with the modeling for that. Experimental results and calculation results were compared in both ways, as the total integrated value of neutron fluxes along neutron energy range from 10 KeV to 2 MeV and as the neutron spectrum along with neutron energy range. Both results are well matched with the statistical error +/-20%. MCNP results were also compared with those of TORT, a three dimensional discrete ordinates code which was developed by Oak Ridge National Laboratory. MCNP results are superior to the TORT results at all detector places except one. This means that MCNP is proved as a very powerful tool for the analysis of neutron transport through iron & air and further it could be used as a powerful tool for the radiation shielding analysis. For one application of the analysis of variance (ANOVA) to neutron and gamma transport problems, uncertainties for the calculated values of critical K were evaluated as in the ANOVA on statistical data.

  1. Neutronic reactor thermal shield

    International Nuclear Information System (INIS)

    Lowe, P.E.

    1976-01-01

    A shield for a nuclear reactor includes at least two layers of alternating wide and narrow rectangular blocks so arranged that the spaces between blocks in adjacent layers are out of registry, each block having an opening therein equally spaced from the sides of the blocks and nearer the top of the block than the bottom, the distance from the top of the block to the opening in one layer being different from this distance in adjacent layers, openings in blocks in adjacent layers being in registry. 1 claim, 7 drawing figures

  2. 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...

  3. Measuring space radiation shielding effectiveness

    OpenAIRE

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

    2017-01-01

    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 ...

  4. Selective shielding device for scintiphotography

    International Nuclear Information System (INIS)

    Harper, J.W.; Kay, T.D.

    1976-01-01

    A selective shielding device to be used in combination with a scintillation camera is described. The shielding device is a substantially oval-shaped configuration removably secured to the scintillation camera. As a result of this combination scanning of preselected areas of a patient can be rapidly and accurately performed without the requirement of mounting any type of shielding paraphernalia on the patient. 1 claim, 2 drawing figures

  5. Tax Shield, Insolvenz und Zinsschranke

    OpenAIRE

    Arnold, Sven; Lahmann, Alexander; Schwetzler, Bernhard

    2010-01-01

    Dieser Beitrag analysiert den Wertbeitrag fremdfinanzierungsbedingter Steuervorteile (Tax Shield) unter realistischen Bedingungen (keine Negativsteuer; mögliche Insolvenz) für unterschiedliche Finanzierungspolitiken. Zusätzlich wird der Effekt der sogenannten Zinsschranke auf den Wert des Tax Shield ermittelt. Die Bewertung des Tax Shield mit und ohne Zinsschranke findet im einperiodigen Fall auf der Basis von Optionspreismodellen und im mehrperiodigen Fall auf der Basis von Monte Carlo Simul...

  6. 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

  7. Radiation shield for nuclear reactors

    International Nuclear Information System (INIS)

    Weissenfluh, J.A.

    1978-01-01

    A shield for use with nuclear reactor systems to attenuate radiation resulting from reactor operation is described. The shield comprises a container preferably of a thin, flexible or elastic material, which may be in the form of a bag, a mattress, a toroidal segment or toroid or the like filled with radiation attenuating liuid. Means are provided in the container for filling and draining the container in place. Due to its flexibility, the shield readily conforms to irregularities in surfaces with which it may be in contact in a shielding position

  8. Mouse RC/BTB2, a Member of the RCC1 Superfamily, Localizes to Spermatid Acrosomal Vesicles

    Science.gov (United States)

    Shen, Xuening; Nagarkatti-Gude, David R.; Hess, Rex A.; Henderson, Scott C.; Strauss, Jerome F.; Zhang, Zhibing

    2012-01-01

    Mouse RC/BTB2 is an unstudied protein of the RCC1 (Regulator of Chromosome Condensation) superfamily. Because of the significant remodeling of chromatin that occurs during spermiogenesis, we characterized the expression and localization of mouse RC/BTB2 in the testis and male germ cells. The Rc/btb2 gene yields two major transcripts: 2.3 kb Rc/btb2-s, present in most somatic tissues examined; and 2.5 kb Rc/btb2-t, which contains a unique non-translated exon in its 5′-UTR that is only detected in the testis. During the first wave of spermatogenesis, Rc/btb2-t mRNA is expressed from day 8 after birth, reaching highest levels of expression at day 30 after birth. The full-length protein contains three RCC1 domains in the N-terminus, and a BTB domain in the C-terminus. In the testis, the protein is detectable from day 12, but is progressively up-regulated to day 30 and day 42 after birth. In spermatids, some of the protein co-localizes with acrosomal markers sp56 and peanut lectin, indicating that it is an acrosomal protein. A GFP-tagged RCC1 domain is present throughout the cytoplasm of transfected CHO cells. However, both GFP-tagged, full-length RC/BTB2 and a GFP-tagged BTB domain localize to vesicles in close proximity to the nuclear membrane, suggesting that the BTB domain might play a role in mediating full-length RC/BTB2 localization. Since RCC1 domains associate with Ran, a small GTPase that regulates molecular trafficking, it is possible that RC/BTB2 plays a role in transporting proteins during acrosome formation. PMID:22768142

  9. New Toroid shielding design

    CERN Multimedia

    Hedberg V

    On the 15th of June 2001 the EB approved a new conceptual design for the toroid shield. In the old design, shown in the left part of the figure above, the moderator part of the shielding (JTV) was situated both in the warm and cold areas of the forward toroid. It consisted both of rings of polyethylene and hundreds of blocks of polyethylene (or an epoxy resin) inside the toroid vacuum vessel. In the new design, shown to the right in the figure above, only the rings remain inside the toroid. To compensate for the loss of moderator in the toroid, the copper plug (JTT) has been reduced in radius so that a layer of borated polyethylene can be placed around it (see figure below). The new design gives significant cost-savings and is easier to produce in the tight time schedule of the forward toroid. Since the amount of copper is reduced the weight that has to be carried by the toroid is also reduced. Outgassing into the toroid vacuum was a potential problem in the old design and this is now avoided. The main ...

  10. SHIELDS Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    Jordanova, Vania Koleva [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-10-03

    Predicting variations in the near-Earth space environment that can lead to spacecraft damage and failure, i.e. “space weather”, remains a big space physics challenge. A new capability was developed at Los Alamos National Laboratory (LANL) to understand, model, and predict Space Hazards Induced near Earth by Large Dynamic Storms, the SHIELDS framework. This framework simulates the dynamics of the Surface Charging Environment (SCE), the hot (keV) electrons representing the source and seed populations for the radiation belts, on both macro- and micro-scale. In addition to using physics-based models (like RAM-SCB, BATS-R-US, and iPIC3D), new data assimilation techniques employing data from LANL instruments on the Van Allen Probes and geosynchronous satellites were developed. An order of magnitude improvement in the accuracy in the simulation of the spacecraft surface charging environment was thus obtained. SHIELDS also includes a post-processing tool designed to calculate the surface charging for specific spacecraft geometry using the Curvilinear Particle-In-Cell (CPIC) code and to evaluate anomalies' relation to SCE dynamics. Such diagnostics is critically important when performing forensic analyses of space-system failures.

  11. Radiation shielding analysis

    International Nuclear Information System (INIS)

    Moon, S.H.; Ha, C.W.; Kwon, S.K.; Lee, J.K.; Choi, H.S.

    1982-01-01

    The theoretical bases of radiation streaming analysis in power reactors, such as ducts or reactor cavity, have been investigated. Discrete ordinates-Monte Carlo or Monte Carlo-Monte Carlo coupling techniques are suggested for the streaming analysis of ducts or reactor cavity. Single albedo scattering approximation code (SINALB) has been developed for simple and quick estimation of gamma-ray ceiling scattering, where the ceiling is assumed to be semi-infinite medium. This code has been employed to calculate the gamma-ray ceiling scattering effects in the laboratory containing a Co-60 source. The SINALB is applicable to gamma-ray scattering, only where the ceiling is thicker than Σsup(-1) and the height is at least twice higher than the shield wall. This code can be used for the purpose of preliminary radiation shield design. The MORSE code has been improved to analyze the gamma-ray scattering problem with on approximation method in respect to the random walk and estimation processes. This improved MORSE code has been employed to the gamma-ray ceiling scattering problem. The results of the improved MORSE calculation are in good agreement with the SINALB and standard MORSE. (Author)

  12. Optimization of thermal design for nitrogen shield of JET cryopump

    International Nuclear Information System (INIS)

    Baxi, C.B.; Obert, W.

    1991-11-01

    The reference design of JET cryopump nitrogen shield consists of an outer section made of copper chevrons fastened to two cooling tubes and an inner stainless steel section and backing plate with two cooling tubes. These tubes are fed in a parallel flow arrangement. The inlet flow is divided into two parallel paths so that both tubes on either section are always at the same temperature. This arrangement was selected due to concern about conduction between warm and cold parts of the shield during cooldown transients. If the heat loads are unequal, such a parallel flow arrangement can result in flow starvation in the path with higher heat load. This will cause large temperature differences and, ultimately, structural failure. Hence, an analysis was undertaken to investigate the conduction effects in the shield for other flow arrangements. 4 refs., 8 figs

  13. A comparative study of physical and chemical properties of different pozzolanic materials used for roller compacted concrete RCC dams

    Directory of Open Access Journals (Sweden)

    Husein Malkawi Abdallah I.

    2017-01-01

    Full Text Available This paper addresses the feasibility and the efficiency of using Natural Pozzolan and/or Rock flour in Roller Compacted Concrete (RCC gravity dams. For this purpose, five identical mortar trial mixes were prepared using five different supplementary materials, i.e., fly ash produced in South Africa (proven to be effective in RCC construction, fly ash produced in Turkey, Jordanian natural pozzolan, Saudi natural pozzolan, and rock flour from Mujib Dam basalt quarry. The physical and chemical properties of these pozzolanic materials were determined. The effectiveness of each one of these mineral admixtures used as a cement replacement material in controlling alkali silica reaction are studied and analyzed. Correlations were made between the mechanical properties for the five proposed mixes and a control mix using the Jordanian Portland Cement. The results demonstrate that the performance of Natural Pozzolana and/or rock flour as compared with that of fly ash and other pozzolanic material is very satisfactory and can be effectively used in RCC construction.

  14. Shielding performance of the NET vacuum vessel

    International Nuclear Information System (INIS)

    Arkuszewski, J.J.; Jaeger, J.F.

    1988-01-01

    To corroborate 1-D deterministic shielding calculations on the Next European Torus (NET) vacuum vessel/shield and shielding blanket, 3-D Monte Carlo calculations have been done with the MCNP code. This should provide information on the poloidal and the toroidal variations. Plasma source simulation and the geometrical model are described, as are other assumptions. The calculations are based on the extended plasma power of 714 MW. The results reported here are the heat deposition in various parts of the device, on the one hand, and the neutron and photon currents at the outer boundary of the vacuum vessel, on the other hand. The latter are needed for the detailed design of the super-conducting magnetic coils. A reasonable statistics has been obtained on the outboard side of the torus, though this cannot be said for the inboard side. The inboard is, however, much more toroidally symmetric than the outboard, so that other methods could be applied such as 2-D deterministic calculations, for instance. (author) 4 refs., 44 figs., 42 tabs

  15. Topographic evidence for shield volcanism on Io

    International Nuclear Information System (INIS)

    Moore, J.M.; Mcewen, A.S.; Albin, E.F.; Greeley, R.

    1986-01-01

    Similarities between terrestrial shield volcanoes and a volcano on Io observed in Voyager I imagery of the satellite at 30 0 S, 246 0 W are delineated. A photoclinometry model was used to numerically estimate the slope based on the Minnaert photometric function. The slope values are accurate to within 10 deg on the sun-facing slope and 1 deg on the shadow side. As found with shield volcanoes, the feature has a central edifice, 40-50 km in diameter, and a broad, elliptical base, 77 x 90 km across. The summit of the Io volcano is 2.2-2.8 km above the surrounding plane and contains a caldera about 5 km in diameter. The similarity in shape between basaltic terrestrial shield volcanoes and the Io volcano indicates that the Io feature may also be composed of basalt. The composition could be sulfur if the heat flow was under 0.05 W/sq m, as it might have been in later stages of formation. 9 references

  16. RCC-E a Design Code for I and C and Electrical Systems

    International Nuclear Information System (INIS)

    Haure, J.M.

    2015-01-01

    The paper deals with the stakes and strength of the RCC-E code applicable to Electrical and Instrumentation and control systems and components as regards dealing with safety class functions. The document is interlacing specifications between Owners, safety authorities, designers, and suppliers IAEA safety guides and IEC standards. The code is periodically updated and published by French Society for Design and Construction rules for Nuclear Island Components (AFCEN). The code is compliant with third generation PWR nuclear islands and aims to suit with national regulations as needed in a companion document. The Feedback experience of Fukushima and the licensing of UKEPR in the framework of Generic Design Assessment are lessons learnt that should be considered in the upgrading of the code. The code gathers a set of requirements and relevant good practices of several PWR design and construction practices related to the electrical and I and C systems and components, and electrical engineering documents dealing with systems, equipment and layout designs. Comprehensive statement including some recent developments will be provided about: - Offsite and onsite sources requirements including sources dealing the total loss of off sites and main onsite sources. - Highlights of a relevant protection level against high frequencies disturbances emitted by lightning strokes, Interfaces data used by any supplier or designer such as site data, rooms temperature, equipment maximum design temperature, alternative current and direct current electrical network voltages and frequency variation ranges, environmental conditions decoupling data, - Environmental Qualification process including normal, mild (earthquake resistant), harsh and severe accident ambient conditions. A suit made approach based on families, which are defined as a combination of mission time, duration and abnormal conditions (pressure, temperature, radiation), enables to better cope with Environmental Qualifications

  17. Noncoding RNA Expression and Targeted Next-Generation Sequencing Distinguish Tubulocystic Renal Cell Carcinoma (TC-RCC) from Other Renal Neoplasms.

    Science.gov (United States)

    Lawrie, Charles H; Armesto, María; Fernandez-Mercado, Marta; Arestín, María; Manterola, Lorea; Goicoechea, Ibai; Larrea, Erika; Caffarel, María M; Araujo, Angela M; Sole, Carla; Sperga, Maris; Alvarado-Cabrero, Isabel; Michal, Michal; Hes, Ondrej; López, José I

    2018-01-01

    Tubulocystic renal cell carcinoma (TC-RCC) is a rare recently described renal neoplasm characterized by gross, microscopic, and immunohistochemical differences from other renal tumor types and was recently classified as a distinct entity. However, this distinction remains controversial particularly because some genetic studies suggest a close relationship with papillary RCC (PRCC). The molecular basis of this disease remains largely unexplored. We therefore performed noncoding (nc) RNA/miRNA expression analysis and targeted next-generation sequencing mutational profiling on 13 TC-RCC cases (11 pure, two mixed TC-RCC/PRCC) and compared with other renal neoplasms. The expression profile of miRNAs and other ncRNAs in TC-RCC was distinct and validated 10 differentially expressed miRNAs by quantitative RT-PCR, including miR-155 and miR-34a, that were significantly down-regulated compared with PRCC cases (n = 22). With the use of targeted next-generation sequencing we identified mutations in 14 different genes, most frequently (>60% of TC-RCC cases) in ABL1 and PDFGRA genes. These mutations were present in  600) of The Cancer Genome Atlas database. In summary, this study is by far the largest molecular study of TC-RCC cases and the first to investigate either ncRNA expression or their genomic profile. These results add molecular evidence that TC-RCC is indeed a distinct entity from PRCC and other renal neoplasms. Copyright © 2018 American Society for Investigative Pathology and the Association for Molecular Pathology. Published by Elsevier Inc. All rights reserved.

  18. Radiation shield for PWR reactors

    International Nuclear Information System (INIS)

    Esenov, Amra; Pustovgar, Andrey

    2013-01-01

    One of the chief structures of a reactor pit is a 'dry' shield. Setting up a 'dry' shield includes the technologically complex process of thermal processing of serpentinite concrete. Modern advances in the area of materials technology permit avoiding this complex and demanding procedure, and this significantly decreases the duration, labor intensity, and cost of setting it up. (orig.)

  19. Nuclear data for radiation shielding

    International Nuclear Information System (INIS)

    Miyasaka, Shunichi; Takahashi, Hiroshi.

    1976-01-01

    The third shielding expert conference was convened in Paris in Oct. 1975 for exchanging informations about the sensitivity evaluation of nuclear data in shielding calculation and integral bench mark experiment. The requirements about nuclear data presented at present from the field of nuclear design do not reflect sufficiently the requirements of shielding design, therefore it was the object to gather the requirements about nuclear data from the field of shielding. The nuclides used for shielding are numerous, and the nuclear data on these isotopes are required. Some of them cannot be ignored as the source of secondary γ-ray or in view of the radioactivation of materials. The requirements for the nuclear data of neutrons in the field of shielding are those concerning the reaction cross sections producing secondary γ-ray, the reaction cross sections including the production of secondary neutrons, elastic scattering cross sections, and total cross sections. The topics in the Paris conference about neutron shielding data are described, such as the methodology of sensitivity evaluation, the standardization of group constant libraries, the bench mark experiment on iron and sodium, and the cross section of γ-ray production. In the shielding of nuclear fission reactors, the γ-ray production owing to nuclear fission reaction is also important. In (d, t) fusion reactors, high energy neutrons are generated, and high energy γ-ray is emitted through giant E1 resonance. (Kako, I.)

  20. Concrete shielding exterior to iron

    International Nuclear Information System (INIS)

    Yurista, P.; Cossairt, D.

    1983-08-01

    A rule of thumb at Fermilab has been to use 3 feet of concrete exterior to iron shielding. A recent design of a shield with a severe dimensional constraint has prompted a re-evaluation of this rule of thumb and has led to the following calculations of the concrete thickness required to nullify this problem. 4 references, 4 figures

  1. Gonad shielding in diagnostic radiology

    International Nuclear Information System (INIS)

    1975-06-01

    The use of gonad shielding is an important radiation protection technique, intended to reduce unnecessary x-ray exposure of the gonads of patients from diagnostic x-ray procedures. This pamphlet will provide physicians and radiologic technologists with information which will aid their appropriate use of gonad shielding

  2. The Imperial Shield

    DEFF Research Database (Denmark)

    Mortensen, Simon Valentin

    2006-01-01

      The title of this Ph.d. dissertation is "The Imperial Shield: Imperial Overstretch, Assured Destruction, and the ban on nationwide ABM-defense with particular emphasis on the Johnson and the Nixon Administration". The dissertation set out to explain the origins of the ABM Treaty's central meaning....... Domestic spending continued to increase by more in real terms than the GDP, and the Democratically controlled Congress also made some very expensive modifications in Nixon tax bill in the fall of 1969, once again plunging the budget into the red.The economic crisis was partly caused by, and partly...... the Administration debated the deployment of new ABM-sites in early 1970, Kissinger could not prevail against these forces, but had to settle for a compromise, which he regarded as less than a definite commitment to nationwide ABM-defense.The political developments were of even greater importance. A strong link has...

  3. Shielded Canister Transporter

    International Nuclear Information System (INIS)

    Eidem, G.G. Jr.; Fages, R.

    1993-01-01

    The Hanford Waste Vitrification Plant (HWVP) will produce canisters filled with high-level radioactive waste immobilized in borosilicate glass. This report discusses a Shielded Canister Transporter (SCT) which will provide the means for safe transportation and handling of the canisters from the Vitrification Building to the Canister Storage Building (CSB). The stainless steel canisters are 0.61 meters in diameter, 3.0 meters tall, and weigh approximately 2,135 kilograms, with a maximum exterior surface dose rate of 90,000 R/hr. The canisters are placed into storage tubes to a maximum of three tall (two for overpack canisters) with an impact limiter placed at the tube bottom and between each canister. A floor plug seals the top of the storage tube at the operating floor level of the CSB

  4. Welding shield for coupling heaters

    Science.gov (United States)

    Menotti, James Louis

    2010-03-09

    Systems for coupling end portions of two elongated heater portions and methods of using such systems to treat a subsurface formation are described herein. A system may include a holding system configured to hold end portions of the two elongated heater portions so that the end portions are abutted together or located near each other; a shield for enclosing the end portions, and one or more inert gas inlets configured to provide at least one inert gas to flush the system with inert gas during welding of the end portions. The shield may be configured to inhibit oxidation during welding that joins the end portions together. The shield may include a hinged door that, when closed, is configured to at least partially isolate the interior of the shield from the atmosphere. The hinged door, when open, is configured to allow access to the interior of the shield.

  5. Penetration shielding applications of CYLSEC

    International Nuclear Information System (INIS)

    Dexheimer, D.T.; Hathaway, J.M.

    1985-01-01

    Evaluation of penetration and discontinuity shielding is necessary to meet 10CFR20 regulations for ensuring personnel exposures are as low as reasonably achievable (ALARA). Historically, those shielding evaluations have been done to some degree on all projects. However, many early plants used conservative methods due to lack of an economical computer code, resulting in costly penetration shielding programs. With the increased industry interest in cost effectively reducing personnel exposures to meet ALARA regulations and with the development of the CYLSEC gamma transport computer code at Bechtel, a comprehensive effort was initiated to reduce penetration and discontinuity shielding but still provide a prudent degree of protection for plant personnel from radiation streaming. This effort was more comprehensive than previous programs due to advances in shielding analysis technology and increased interest in controlling project costs while maintaining personnel exposures ALARA. Methodology and resulting cost savings are discussed

  6. Shield calculations, optimization vs. paradigm

    International Nuclear Information System (INIS)

    Cornejo D, N.; Hernandez S, A.; Martinez G, A.

    2006-01-01

    Many shieldings have been designed under the criteria of 'Maximum dose rates of project'. It has created the paradigm of those 'low dose rates', for the one which not few specialists would consider unacceptable levels of dose rate superior to the units of μSv.h -1 , independently of the exposure times. At the present time numerous shieldings are being designed considering dose restrictions in real times of exposure. After these new shieldings, the dose rates could be notably superior to those after traditional shieldings, without it implies inadequate designs or constructive errors. In the work significant differences in levels of dose rates and thickness of shieldings estimated by both methods for some typical facilities. It was concluded that the use of real times of exposure is more adequate for the optimization of the Radiological Protection, although this method demands bigger care in its application. (Author)

  7. Modular reactor head shielding system

    International Nuclear Information System (INIS)

    Jacobson, E. B.

    1985-01-01

    An improved modular reactor head shielding system is provided that includes a frame which is removably assembled on a reactor head such that no structural or mechanical alteration of the head is required. The shielding system also includes hanging assemblies to mount flexible shielding pads on trolleys which can be moved along the frame. The assemblies allow individual pivoting movement of the pads. The pivoting movement along with the movement allowed by the trolleys provides ease of access to any point on the reactor head. The assemblies also facilitate safe and efficient mounting of the pads directly to and from storage containers such that workers have additional shielding throughout virtually the entire installation and removal process. The flexible shielding pads are designed to interleave with one another when assembled around the reactor head for substantially improved containment of radiation leakage

  8. Parameters calculation of shielding experiment

    International Nuclear Information System (INIS)

    Gavazza, S.

    1986-02-01

    The radiation transport methodology comparing the calculated reactions and dose rates for neutrons and gama-rays, with experimental measurements obtained on iron shield, irradiated in the YAYOI reactor is evaluated. The ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system, for cross sections generation collapsed by the ANISN code were used. The transport calculations were made using the DOT 3.5 code, adjusting the boundary iron shield source spectrum to the reactions and dose rates, measured at the beginning of shield. The neutron and gamma ray distributions calculated on the iron shield presented reasonable agreement with experimental measurements. An experimental arrangement using the IEA-R1 reactor to determine a shielding benchmark is proposed. (Author) [pt

  9. 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.

  10. 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.

  11. 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

  12. Transparent fast neutron shielding material and shielding method

    International Nuclear Information System (INIS)

    Nashimoto, Tetsuji; Katase, Haruhisa.

    1993-01-01

    Polyisobutylene having a viscosity average molecular weight of 20,000 to 80,000 and a hydrogen atom density of greater than 7.0 x 10 22 /cm 3 is used as a fast neutron shielding material. The shielding material is excellent in the shielding performance against fast neutrons, and there is no worry of leakage even when holes should be formed to a vessel. Further, it is excellent in fabricability, relatively safe even upon occurrence of fire and, in addition, it is transparent to enable to observe contents easily. (T.M.)

  13. Upgrade of the LHC magnet interconnections thermal shielding

    Science.gov (United States)

    Musso, Andrea; Barlow, Graeme; Bastard, Alain; Charrondiere, Maryline; Chrul, Anna; Damianoglou, Dimitrios; Deferne, Guy; Dib, Gaëlle; Duret, Max; Guinchard, Michael; Prin, Hervé; Strychalski, Michał; Craen, Arnaud Vande; Villiger, Gilles; Wright, Loren

    2014-01-01

    The about 1700 interconnections (ICs) between the Large Hadron Collider (LHC) superconducting magnets include thermal shielding at 50-75 K, providing continuity to the thermal shielding of the magnet cryostats to reduce the overall radiation heat loads to the 1.9 K helium bath of the magnets. The IC shield, made of aluminum, is conduction-cooled via a welded bridge to the thermal shield of the adjacent magnets which is actively cooled. TIG welding of these bridges made in the LHC tunnel at installation of the magnets induced a considerable risk of fire hazard due to the proximity of the multi-layer insulation of the magnet shields. A fire incident occurred in one of the machine sectors during machine installation, but fortunately with limited consequences thanks to prompt intervention of the operators. LHC is now undergoing a 2 years technical stop during which all magnet's ICs will have to be opened to consolidate the magnet electrical connections. The IC thermal shields will therefore have to be removed and re-installed after the work is completed. In order to eliminate the risk of fire hazard when re-welding, it has been decided to review the design of the IC shields, by replacing the welded bridges with a mechanical clamping which also preserves its thermal function. An additional advantage of this new solution is the ease in dismantling for maintenance, and eliminating weld-grinding operations at removal needing radioprotection measures because of material activation after long-term operation of the LHC. This paper describes the new design of the IC shields and in particular the theoretical and experimental validation of its thermal performance. Furthermore a status report of the on-going upgrade work in the LHC is given.

  14. Upgrade of the LHC magnet interconnections thermal shielding

    Energy Technology Data Exchange (ETDEWEB)

    Musso, Andrea; Barlow, Graeme; Bastard, Alain; Charrondiere, Maryline; Deferne, Guy; Dib, Gaëlle; Duret, Max; Guinchard, Michael; Prin, Hervé; Craen, Arnaud Vande; Villiger, Gilles [CERN European Organization for Nuclear Research, Meyrin 1211, Geneva 23, CH (Switzerland); Chrul, Anna [The Henryk Niewodniczanski Institute of Nuclear Physics, Polish Academy of Sciences, ul.Radzikowskiego 152, 31-324 Krakow (Poland); Damianoglou, Dimitrios [NTUA National Technical University of Athens, Heeron Polytechniou 9, 15780 Zografou (Greece); Strychalski, Michał [Wroclaw University of Technology, Faculty of Mechanical and Power Engineering, Wyb. Wyspianskiego 27, Wroclaw, 50-370 (Poland); Wright, Loren [Lancaster University, Bailrigg, Lancaster, LA1 4YW (United Kingdom)

    2014-01-29

    The about 1700 interconnections (ICs) between the Large Hadron Collider (LHC) superconducting magnets include thermal shielding at 50-75 K, providing continuity to the thermal shielding of the magnet cryostats to reduce the overall radiation heat loads to the 1.9 K helium bath of the magnets. The IC shield, made of aluminum, is conduction-cooled via a welded bridge to the thermal shield of the adjacent magnets which is actively cooled. TIG welding of these bridges made in the LHC tunnel at installation of the magnets induced a considerable risk of fire hazard due to the proximity of the multi-layer insulation of the magnet shields. A fire incident occurred in one of the machine sectors during machine installation, but fortunately with limited consequences thanks to prompt intervention of the operators. LHC is now undergoing a 2 years technical stop during which all magnet's ICs will have to be opened to consolidate the magnet electrical connections. The IC thermal shields will therefore have to be removed and re-installed after the work is completed. In order to eliminate the risk of fire hazard when re-welding, it has been decided to review the design of the IC shields, by replacing the welded bridges with a mechanical clamping which also preserves its thermal function. An additional advantage of this new solution is the ease in dismantling for maintenance, and eliminating weld-grinding operations at removal needing radioprotection measures because of material activation after long-term operation of the LHC. This paper describes the new design of the IC shields and in particular the theoretical and experimental validation of its thermal performance. Furthermore a status report of the on-going upgrade work in the LHC is given.

  15. News from the Library: Facilitating access to a program for radiation shielding - the Library can help

    CERN Multimedia

    CERN Library

    2013-01-01

    MicroShield® is a comprehensive photon/gamma ray shielding and dose assessment programme. It is widely used for designing shields, estimating source strength from radiation measurements, minimising exposure to people, and teaching shielding principles.   Integrated tools allow the graphing of results, material and source file creation, source inference with decay (dose-to-Bq calculations accounting for decay and daughter buildup), the projection of exposure rate versus time as a result of decay, access to material and nuclide data, and decay heat calculations. The latest version is able to export results using Microsoft Office (formatted and colour-coded for readability). Sixteen geometries accommodate offset dose points and as many as ten standard shields plus source self-shielding and cylinder cladding are available. The library data (radionuclides, attenuation, build-up and dose conversion) reflect standard data from ICRP 38 and 107* as well as ANSI/ANS standards and RSICC publicat...

  16. Radiation shielding for fusion reactors

    International Nuclear Information System (INIS)

    Santoro, R.T.

    2000-01-01

    Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel. (author)

  17. Morphometry of terrestrial shield volcanoes

    Science.gov (United States)

    Grosse, Pablo; Kervyn, Matthieu

    2018-03-01

    Shield volcanoes are described as low-angle edifices built primarily by the accumulation of successive lava flows. This generic view of shield volcano morphology is based on a limited number of monogenetic shields from Iceland and Mexico, and a small set of large oceanic islands (Hawaii, Galápagos). Here, the morphometry of 158 monogenetic and polygenetic shield volcanoes is analyzed quantitatively from 90-meter resolution SRTM DEMs using the MORVOLC algorithm. An additional set of 24 lava-dominated 'shield-like' volcanoes, considered so far as stratovolcanoes, are documented for comparison. Results show that there is a large variation in shield size (volumes from 0.1 to > 1000 km3), profile shape (height/basal width (H/WB) ratios mostly from 0.01 to 0.1), flank slope gradients (average slopes mostly from 1° to 15°), elongation and summit truncation. Although there is no clear-cut morphometric difference between shield volcanoes and stratovolcanoes, an approximate threshold can be drawn at 12° average slope and 0.10 H/WB ratio. Principal component analysis of the obtained database enables to identify four key morphometric descriptors: size, steepness, plan shape and truncation. Hierarchical cluster analysis of these descriptors results in 12 end-member shield types, with intermediate cases defining a continuum of morphologies. The shield types can be linked in terms of growth stages and shape evolution, related to (1) magma composition and rheology, effusion rate and lava/pyroclast ratio, which will condition edifice steepness; (2) spatial distribution of vents, in turn related to the magmatic feeding system and the tectonic framework, which will control edifice plan shape; and (3) caldera formation, which will condition edifice truncation.

  18. Influence of the Radiation Shield on the Temperature of Rails Rolled in the Reversing Mill

    Directory of Open Access Journals (Sweden)

    Gołdasz A.

    2015-04-01

    Full Text Available The paper presents a mathematical model of heat transfer during cooling of hot-rolled rails in the reversing mill. The influence of the radiation shield on the temperature of rolled rails has been analyzed. The heat transfer model for cooling a strip covered by the thermal shield has been presented. The two types of shields build of steel and aluminum sheets separated with insulating layer have been studded. Calculations have been performed with self developed software which utilizes the finite element method.

  19. 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

  20. Radiation protection/shield design

    International Nuclear Information System (INIS)

    Disney, R.K.

    1977-01-01

    Radiation protection/shielding design of a nuclear facility requires a coordinated effort of many engineering disciplines to meet the requirements imposed by regulations. In the following discussion, the system approach to Clinch River Breeder Reactor Plant (CRBRP) radiation protection will be described, and the program developed to implement this approach will be defined. In addition, the principal shielding design problems of LMFBR nuclear reactor systems will be discussed in realtion to LWR nuclear reactor system shielding designs. The methodology used to analyze these problems in the U.S. LMFBR program, the resultant design solutions, and the experimental verification of these designs and/or methods will be discussed. (orig.) [de

  1. Shielding synchrotron light sources: Advantages of circular shield walls tunnels

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, S.L. [Design and Accelerator Operations Consulting, 568 Wintergreen Ct Ridge, NY 11961 (United States); Ghosh, V.J.; Breitfeller, M. [NSLS-II, Brookhaven National Laboratory, Upton, NY 11973 (United States)

    2016-08-11

    Third generation high brightness light sources are designed to have low emittance and high current beams, which contribute to higher beam loss rates that will be compensated by Top-Off injection. Shielding for these higher loss rates will be critical to protect the projected higher occupancy factors for the users. Top-Off injection requires a full energy injector, which will demand greater consideration of the potential abnormal beam miss-steering and localized losses that could occur. The high energy electron injection beam produce significantly higher neutron component dose to the experimental floor than lower energy injection and ramped operations. High energy neutrons produced in the forward direction from thin target beam losses are a major component of the dose rate outside the shield walls of the tunnel. The convention has been to provide thicker 90° ratchet walls to reduce this dose to the beam line users. We present an alternate circular shield wall design, which naturally and cost effectively increases the path length for this forward radiation in the shield wall and thereby substantially decreasing the dose rate for these beam losses. This shield wall design will greatly reduce the dose rate to the users working near the front end optical components but will challenge the beam line designers to effectively utilize the longer length of beam line penetration in the shield wall. Additional advantages of the circular shield wall tunnel are that it's simpler to construct, allows greater access to the insertion devices and the upstream in tunnel beam line components, as well as reducing the volume of concrete and therefore the cost of the shield wall.

  2. Survivor shielding. Part C. Improvements in terrain shielding

    International Nuclear Information System (INIS)

    Egbert, Stephen D.; Kaul, Dean C.; Roberts, James A.; Kerr, George D.

    2005-01-01

    A number of atomic-bomb survivors were affected by shielding provided by terrain features. These terrain features can be a small hill, affecting one or two houses, or a high mountain that shields large neighborhoods. In the survivor dosimetry system, terrain shielding can be described by a transmission factor (TF), which is the ratio between the dose with and without the terrain present. The terrain TF typically ranges between 0.1 and 1.0. After DS86 was implemented at RERF, the terrain shielding categories were examined and found to either have a bias or an excessive uncertainty that could readily be removed. In 1989, an improvement in the terrain model was implemented at RERF in the revised DS86 code, but the documentation was not published. It is now presented in this section. The solution to the terrain shielding in front of a house is described in this section. The problem of terrain shielding of survivors behind Hijiyama mountain at Hiroshima and Konpirasan mountain at Nagasaki has also been recognized, and a solution to this problem has been included in DS02. (author)

  3. Radiation shielding member

    International Nuclear Information System (INIS)

    Nemezawa, Isao; Kimura, Tadahiro; Mizuochi, Akira; Omori, Tetsu

    1998-01-01

    A single body of a radiation shield comprises a bag prepared by welding or bonding a polyurethane sheet which is made flat while interposing metal plates at the upper and the lower portion of the bag. Eyelet fittings are disposed to the upper and the lower portions of the bag passing through the metal plates and the flat portion of the bag. Water supplying/draining ports are disposed to two upper and lower places of the bag at a height where the metal plates are disposed. Reinforcing walls welded or bonded to the inner wall surface of the bag are elongated in vertical direction to divide the inside of the bag to a plurality of cells. The bag is suspended and supported from a frame with S-shaped hooks inserted into the eyelet fittings as connecting means. A plurality of bags are suspended and supported from the frame at a required height by way of the eyelets at the lower portion of the suspended and supported bag and the eyelet fittings at the upper portion of the bag below the intermediate connection means. (I.N.)

  4. Self-shielding factors

    International Nuclear Information System (INIS)

    Kaul, D.C.

    1982-01-01

    Throughout the last two decades many efforts have been made to estimate the effect of body self-shielding on organ doses from externally incident neutrons and gamma rays. These began with the use of simple geometry phantoms and have culminated in the use of detailed anthropomorphic phantoms. In a recent effort, adjoint Monte Carlo analysis techniques have been used to determine dose and dose equivalent to the active marrow as a function of energy and angle of neutron fluence externally incident on an anthropomorphic phantom. When combined with fluences from actual nuclear devices, these dose-to-fluence factors result in marrow dose values that demonstrate great sensitivity to variations in device type, range, and body orientation. Under a state-of-the-art radiation transport analysis demonstration program for the Japanese cities, sponsored by the Defense Nuclear Agency at the request of the National Council on Radiation Protection and Measurements, the marrow dose study referred to above is being repeated to obtain spectral distributions within the marrow for externally incident neutrons and gamma rays of arbitrary energy and angle. This is intended to allow radiobiologists and epidemiologists to select and to modify numbers of merit for correlation with health effects and to permit a greater understanding of the relationship between human and laboratory subject dosimetry

  5. Shielding plug device

    International Nuclear Information System (INIS)

    Orii, Shoichi; Hasegawa, Satoshi; Makishima, Kenji.

    1976-01-01

    Object: To reduce the size of and extend the life of a revolving bearing and facilitate the laying of driving cables and duct lines, this being accomplished by providing plug raising means of a fast breeder on a stationary plug mounting base so as to prevent the shearing force of sodium from acting upon the revolving bearing. Structure: The shield plug means comprises a stationary plug secured to the open end of the reactor container, a rotary plug rotatable with respect to the stationary plug, an annular base formed on top of the stationary plug so as to cover the rotary plug, a bearing secured to the rotary plug edge lower face and upper and lower locking plates. At the time of the rotation of the rotary plug, the upper locking plate is withdrawn, the stationary plug is raised to release the seal structure, and the lower locking plate is inserted between the bearing and stationary plug. In this way, smooth rotation of the rotary plug can be obtained. (Horiuchi, T.)

  6. Intracellular lipid in papillary renal cell carcinoma (pRCC): T2 weighted (T2W) MRI and pathologic correlation

    Energy Technology Data Exchange (ETDEWEB)

    Schieda, Nicola; Van der Pol, Christian B.; Moosavi, Bardia; McInnes, Matthew D.F. [The Ottawa Hospital, The University of Ottawa, Department of Medical Imaging, Ottawa, Ontario (Canada); Mai, Kien T.; Flood, Trevor A. [The Ottawa Hospital, The University of Ottawa, Department of Anatomical Pathology, Ottawa, Ontario (Canada)

    2015-07-15

    To evaluate if pRCCs demonstrate intracellular lipid (i-lipid) at chemical-shift (CS) MRI, and assess T2W-MRI and pathologic characteristics. Sixty-two patients with a pRCC diagnosis underwent MRI over 11 years (IRB-approved). Two radiologists independently assessed for presence of i-lipid on CS-MRI and homogeneity on T2W-MRI. Inter-observer agreement was assessed via an intraclass correlation and results were compared using the Chi-square test. Discordant cases were reviewed to establish consensus. T2W SI-ratios (SI.tumor/SI.kidney) and CS-SI index were compared using independent t-tests and Spearman correlation. Two pathologists re-evaluated the histopathology. Nine of the 62 pRCCs (14.5 %) demonstrated i-lipid; agreement was moderate (ICC = 0.63). Pathology review depicted clear cells in four tumours and foamy histiocytes in five tumours. 25.8-35.4 % (ICC = 0.65) of tumours were homogeneous on T2W-MRI. No pRCC with i-lipid was considered homogeneous (p = 0.01-0.04). Overall, T2W SI-ratio and CS-SI index were 0.89 (±0.29) and -3.63 % (-7.27 to 11.42). pRCC with i-lipid had significantly higher T2W SI-ratio (p = 0.003). There was a correlation between the CS-SI index and T2W SI-ratio, (r = 0.44, p < 0.001). Intracellular lipid is uncommonly detected in pRCCs due to clear cell changes and foamy histiocytes. These tumours are associated with heterogeneously-increased SI in T2W-MRI. (orig.)

  7. SNF shipping cask shielding analysis

    International Nuclear Information System (INIS)

    Johnson, J.O.; Pace, J.V. III.

    1996-01-01

    The Waste Management and Remedial Action Division has planned a modification sequence for storage facility 7827 in the Solid Waste Storage Area (SWSA). The modification cycle is: (1) modify an empty caisson, (2) transfer the spent nuclear fuel (SNF) of an occupied caisson to a hot cell in building 3525 for inspection and possible repackaging, and (3) return the package to the modified caisson in the SWSA. Although the SNF to be moved is in the solid form, it has different levels of activity. Thus, the following 5 shipping casks will be available for the task: the Loop Transport Carrier, the In- Pile Loop LITR HB-2 Carrier, the 6.5-inch HRLEL Carrier, the HFIR Hot Scrap Carrier, and the 10-inch ORR Experiment Removal Shield Cask. This report describes the shielding tasks for the 5 casks: determination of shielding characteristics, any streaming avenues, estimation of thermal limits, and shielding calculational uncertainty for use in the transportation plan

  8. Active Radiation Shield, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — DEC-Shield technology offers the means to generate electric power from cosmic radiation sources and fuse dissimilar systems and functionality into a structural...

  9. Gonad shielding in computerized tomography

    International Nuclear Information System (INIS)

    Rockstroh, G.

    1984-01-01

    The reduction of gonadal dose by shielding of the gonads was investigated for a Somatom 2 using an anthropomorphic phantom. For small distances from the slice examined the gonadal dose results from intracorporal secondary radiation and is only insignificantly reduced by shielding. For greater distances shielding is relatively more effective, the gonadal dose however is small because of the approximately exponential decay. Shielding of the gonads therefore does not seem adequate for the reduction of gonadal dose. From dose measurements in cylinder phantoms of several diameters it appears that no different results would be obtained for children and young adults. An effective reduction of gonadal dose is only possible with lead capsules for males. (author)

  10. Radiation shielding for neutron guides

    International Nuclear Information System (INIS)

    Ersez, T.; Braoudakis, G.; Osborn, J.C.

    2005-01-01

    Full text: Models of the neutron guide shielding for the out of bunker guides on the thermal and cold neutron beam lines of the OPAL Reactor (ANSTO) were constructed using the Monte Carlo code MCNP 4B. The neutrons that were not reflected inside the guides but were absorbed by the supermirror (SM) layers were noted to be a significant source of gammas. Gammas also arise from neutrons absorbed by the B, Si, Na and K contained in the glass. The proposed shielding design has produced compact shielding assemblies. These arrangements are consistent with safety requirements, floor load limits, and cost constraints. To verify the design a prototype was assembled consisting of 120mm thick Pb(96%)Sb(4%) walls resting on a concrete block. There was good agreement between experimental measurements and calculated dose rates for bulk shield regions. (authors)

  11. Shielding calculations. Optimization vs. Paradigms

    International Nuclear Information System (INIS)

    Cornejo Diaz, Nestor; Hernandez Saiz, Alejandro; Martinez Gonzalez, Alina

    2005-01-01

    Many radiation shielding barriers in Cuba have been designed according to the criterion of Maxi-mum Projected Dose Rates. This fact has created the paradigm of low dose rates. Because of this, dose rate levels greater than units of Sv.h-1 would be considered unacceptable by many specialists, regardless of the real exposure times. Nowadays many shielding barriers are being designed using dose constraints in real exposure times. Behind the new barriers, dose rates could be notably greater than those behind the traditional ones, and it does not imply inadequate designs or constructive errors. In this work were obtained significant differences in dose rate levels and shield-ing thicknesses calculated by both methods for some typical installations. The work concludes that real exposure time approach is more adequate in order to optimise Radiation Protection, although this method should be carefully applied

  12. Radiation shielding for neutron guides

    International Nuclear Information System (INIS)

    Ersez, T.; Braoudakis, G.; Osborn, J.C.

    2006-01-01

    Models of the neutron guide shielding for the out of bunker guides on the thermal and cold neutron beam lines of the OPAL Reactor (ANSTO) were constructed using the Monte Carlo code MCNP 4B. The neutrons that were not reflected inside the guides but were absorbed by the supermirror (SM) layers were noted to be a significant source of gammas. Gammas also arise from neutrons absorbed by the B, Si, Na and K contained in the glass. The proposed shielding design has produced compact shielding assemblies. These arrangements are consistent with safety requirements, floor load limits, and cost constraints. To verify the design a prototype was assembled consisting of 120 mm thick Pb(96%)Sb(4%) walls resting on a concrete block. There was good agreement between experimental measurements and calculated dose rates for bulk shield regions

  13. Thulium-170 heat source

    Science.gov (United States)

    Walter, Carl E.; Van Konynenburg, Richard; VanSant, James H.

    1992-01-01

    An isotopic heat source is formed using stacks of thin individual layers of a refractory isotopic fuel, preferably thulium oxide, alternating with layers of a low atomic weight diluent, preferably graphite. The graphite serves several functions: to act as a moderator during neutron irradiation, to minimize bremsstrahlung radiation, and to facilitate heat transfer. The fuel stacks are inserted into a heat block, which is encased in a sealed, insulated and shielded structural container. Heat pipes are inserted in the heat block and contain a working fluid. The heat pipe working fluid transfers heat from the heat block to a heat exchanger for power conversion. Single phase gas pressure controls the flow of the working fluid for maximum heat exchange and to provide passive cooling.

  14. The optimum shielding for a power reactor using local components

    International Nuclear Information System (INIS)

    AlHajali, S.; Kharita, M. H.; Yousef, S.; Naoom, B.; Al-Nassar, M.

    2009-07-01

    Some local concrete mixtures have been picked out (selected) to be studied as shielding concrete for prospective nuclear power reactor in Syria. This research has interested in the attenuation of gamma radiation and neutron fluxes by these local concretes in the ordinary conditions. In addition to the heat effect on the shielding and physical properties of local concrete. Furthermore the neutron activation of the elements of the local concrete mixtures have been studied that for selection the low-activation materials (low dose rate and short half life radioisotopes). In this way biological shielding for nuclear reactor can be safe during operation of nuclear power reactor, in addition to be low radioactive waste after decommissioning the reactor. (author)

  15. Conceptual design for the thermal shield bridges and multilayer insulation in the interconnect region for the SSC

    International Nuclear Information System (INIS)

    Baritchi, D.; Nicol, T.; Boroski, W.

    1991-01-01

    The interconnect region serves as the connection area between magnets. In order to minimize radiant heat transfer in the interconnect area, the authors used shield bridges which span the 80K and 20K shield gap between adjacent magnets. A sliding joint between bridge sections on adjacent magnets accommodates contraction during cool-down. An investigation was done to determine which attachment schemes (riveted or bolted versus welded) are better for heat transfer. Each shield bridge is covered with the same multilayer insulation scheme used throughout the body of the magnet. These shield bridges also contain pressure reliefs for each shield in the event of an internal piping failure. The reliefs are located in the upper half of the shield section in order to prevent liquid spills from impinging directly onto the vacuum vessel wall

  16. Single classifier, OvO, OvA and RCC multiclass classification method in handheld based smartphone gait identification

    Science.gov (United States)

    Raziff, Abdul Rafiez Abdul; Sulaiman, Md Nasir; Mustapha, Norwati; Perumal, Thinagaran

    2017-10-01

    Gait recognition is widely used in many applications. In the application of the gait identification especially in people, the number of classes (people) is many which may comprise to more than 20. Due to the large amount of classes, the usage of single classification mapping (direct classification) may not be suitable as most of the existing algorithms are mostly designed for the binary classification. Furthermore, having many classes in a dataset may result in the possibility of having a high degree of overlapped class boundary. This paper discusses the application of multiclass classifier mappings such as one-vs-all (OvA), one-vs-one (OvO) and random correction code (RCC) on handheld based smartphone gait signal for person identification. The results is then compared with a single J48 decision tree for benchmark. From the result, it can be said that using multiclass classification mapping method thus partially improved the overall accuracy especially on OvO and RCC with width factor more than 4. For OvA, the accuracy result is worse than a single J48 due to a high number of classes.

  17. Study on Viscoelastic Deformation Monitoring Index of an RCC Gravity Dam in an Alpine Region Using Orthogonal Test Design

    Directory of Open Access Journals (Sweden)

    Yaoying Huang

    2018-01-01

    Full Text Available The main objective of this study is to present a method of determining viscoelastic deformation monitoring index of a Roller-compacted concrete (RCC gravity dam in an alpine region. By focusing on a modified deformation monitoring model considering frost heave and back analyzed mechanical parameters of the dam, the working state of viscoelasticity for the dam is illustrated followed by an investigation and designation of adverse load cases using orthogonal test method. Water pressure component is then calculated by finite element method, while temperature, time effect, and frost heave components are obtained through deformation statistical model considering frost heave. The viscoelastic deformation monitoring index is eventually determined by small probability and maximum entropy methods. The results show that (a with the abnormal probability 1% the dam deformation monitoring index for small probability and maximum entropy methods is 23.703 mm and 22.981 mm, respectively; thus the maximum measured displacement of the dam is less than deformation monitoring index, which indicates that the dam is currently in a state of safety operation and (b the obtained deformation monitoring index using orthogonal test method is more accurate due to the full consideration of more random factors; the method gained from this study will likely be of use to diagnose the working state for those RCC dams in alpine regions.

  18. Design and Construction Rules for Mechanical components of FBR nuclear islands: RCC-MR. Tome 1, Volume H: Supports

    International Nuclear Information System (INIS)

    1985-06-01

    The present book is the fifth one of a whole set of 12 which constitute the present edition of the RCC-MR. The volume H applies to supports of equipments when these equipments are concerned by the RCC-MR rules in application of the chapter RA 4000 of the Tome 1, Vol A. One defines the field of application of this volume and the classification rules of supports in two levels. One defines the different types of supports and fasteners, as also the classification of the support elements in primary or resistance (or structure) elements and secondary (stability) elements. One defines the documents to be established for the components and their constitutive elements, the modalities of identification of pieces and welded joints, the rules to choose the materials of the support elements, the rules to follow for the support conception, the rules to follow for fabrication and tests, as also other rules than those given in the chapters RH 2000 to RH 4000, to homologate a standard support or a component of a standard support [fr

  19. 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.

  20. 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

  1. 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.

  2. 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

  3. Structural Design and Thermal Analysis for Thermal Shields of the MICE Coupling Magnets

    International Nuclear Information System (INIS)

    Green, Michael A.; Pan, Heng; Liu, X.K.; Wang, Li; Wu, Hong; Chen, A.B.; Guo, X.L.

    2009-01-01

    A superconducting coupling magnet made from copper matrix NbTi conductors operating at 4 K will be used in the Muon Ionization Cooling Experiment (MICE) to produce up to 2.6 T on the magnet centerline to keep the muon beam within the thin RF cavity indows. The coupling magnet is to be cooled by two cryocoolers with a total cooling capacity of 3 W at 4.2 K. In order to keep a certain operating temperature margin, the most important is to reduce the heat leakage imposed on cold surfaces of coil cold mass assembly. An ntermediate temperature shield system placed between the coupling coil and warm vacuum chamber is adopted. The shield system consists of upper neck shield, main shields, flexible connections and eight supports, which is to be cooled by the first stage cold heads of two ryocoolers with cooling capacity of 55 W at 60 K each. The maximum temperature difference on the shields should be less than 20 K, so the thermal analyses for the shields with different thicknesses, materials, flexible connections for shields' cooling and structure design for heir supports were carried out. 1100 Al is finally adopted and the maximum temperature difference is around 15 K with 4 mm shield thickness. The paper is to present detailed analyses on the shield system design.

  4. Modelling and Optimization of Four-Segment Shielding Coils of Current Transformers.

    Science.gov (United States)

    Gao, Yucheng; Zhao, Wei; Wang, Qing; Qu, Kaifeng; Li, He; Shao, Haiming; Huang, Songling

    2017-05-26

    Applying shielding coils is a practical way to protect current transformers (CTs) for large-capacity generators from the intensive magnetic interference produced by adjacent bus-bars. The aim of this study is to build a simple analytical model for the shielding coils, from which the optimization of the shielding coils can be calculated effectively. Based on an existing stray flux model, a new analytical model for the leakage flux of partial coils is presented, and finite element method-based simulations are carried out to develop empirical equations for the core-pickup factors of the models. Using the flux models, a model of the common four-segment shielding coils is derived. Furthermore, a theoretical analysis is carried out on the optimal performance of the four-segment shielding coils in a typical six-bus-bars scenario. It turns out that the "all parallel" shielding coils with a 45° starting position have the best shielding performance, whereas the "separated loop" shielding coils with a 0° starting position feature the lowest heating value. Physical experiments were performed, which verified all the models and the conclusions proposed in the paper. In addition, for shielding coils with other than the four-segment configuration, the analysis process will generally be the same.

  5. Electrodynamic Dust Shield Demonstrator

    Science.gov (United States)

    Stankie, Charles G.

    2013-01-01

    The objective of the project was to design and manufacture a device to demonstrate a new technology developed by NASA's Electrostatics and Surface Physics Laboratory. The technology itself is a system which uses magnetic principles to remove regolith dust from its surface. This project was to create an enclosure that will be used to demonstrate the effectiveness of the invention to The Office of the Chief Technologist. ONE of the most important challenges of space exploration is actually caused by something very small and seemingly insignificant. Dust in space, most notably on the moon and Mars, has caused many unforeseen issues. Dirt and dust on Earth, while a nuisance, can be easily cleaned and kept at bay. However, there is considerably less weathering and erosion in space. As a result, the microscopic particles are extremely rough and abrasive. They are also electrostatically charged, so they cling to everything they make contact with. This was first noted to be a major problem during the Apollo missions. Dust would stick to the spacesuits, and could not be wiped off as predicted. Dust was brought back into the spacecraft, and was even inhaled by astronauts. This is a major health hazard. Atmospheric storms and other events can also cause dust to coat surfaces of spacecraft. This can cause abrasive damage to the craft. The coating can also reduce the effectiveness of thermal insulation and solar panels.' A group of engineers at Kennedy Space Center's Electrostatics and Surface Physics Laboratory have developed a new technology, called the Electrodynamic Dust Shield, to help alleviate these problems. It is based off of the electric curtain concept developed at NASA in 1967. "The EDS is an active dust mitigation technology that uses traveling electric fields to transport electrostatically charged dust particles along surfaces. To generate the traveling electric fields, the EDS consists of a multilayer dielectric coating with an embedded thin electrode grid

  6. Gas Shielding Technology for Welding and Brazing

    Science.gov (United States)

    Nunes, Arthur J.; Gradl, Paul R.

    2012-01-01

    Welding is a common method that allows two metallic materials to be joined together with high structural integrity. When joints need to be leak-tight, light-weight, or free of contaminant-trapping seams or surface asperities, welding tends to be specified. There are many welding techniques, each with its own advantages and disadvantages. Some of these techniques include Forge Welding, Gas Tungsten Arc Welding, Friction Stir Welding, and Laser Beam Welding to name a few. Whichever technique is used, the objective is a structural joint that meets the requirements of a particular component or assembly. A key practice in producing quality welds is the use of shielding gas. This article discusses various weld techniques, quality of the welds, and importance of shielding gas in each of those techniques. Metallic bonds, or joints, are produced when metals are put into intimate contact. In the solid-state "blacksmith welding" process, now called Forge Welding (FOW), the site to be joined is pounded into intimate contact. The surfaces to be joined usually need to be heated to make it easier to deform the metal. The surfaces are sprinkled with a flux to melt surface oxides and given a concave shape so that surface contamination can be squeezed out of the joint as the surfaces are pounded together; otherwise the surface contamination would be trapped in the joint and would weaken the weld. In solid-state welding processes surface oxides or other contamination are typically squeezed out of the joint in "flash."

  7. 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)

  8. Shielding features of quarry stone

    International Nuclear Information System (INIS)

    Hernandez V, C.; Contreras S, H.; Hernandez A, L.; Baltazar R, A.; Escareno J, E.; Mares E, C. A.; Vega C, H. R.

    2010-10-01

    Quarry stone lineal attenuation coefficient for gamma-rays has been obtained. In Zacatecas, quarry stone is widely utilized as a decorative item in buildings, however its shielding features against gamma-rays unknown. The aim of this work is to determine the shielding properties of quarry stone against γ-rays using Monte Carlo calculations where a detailed model of a good geometry experimental setup was carried out. In the calculations 10 pieces 10 X 10 cm 2 of different thickness were utilized to evaluate the photons transmission as the quarry stone thickness is increased. It was noticed that transmitted photons decay away as the shield thickness is increased, these results were fitted to an exponential function were the linear attenuation coefficient was estimated. Also, using XCOM code the linear attenuation coefficient from several keV up to 100 MeV was estimated. From the comparison between Monte Carlo results and XCOM calculations a good agreement was found. For 0.662 MeV γ-rays the attenuation coefficient of quarry stone, whose density is 2.413 g-cm -3 , is 0.1798 cm -1 , this mean a X 1/2 = 3.9 cm, X 1/4 = 7.7 cm, X 1/10 = 12.8 cm, and X 1/100 = 25.6 cm. Having the information of quarry stone performance as shielding give the chance to use this material to shield X and γ-ray facilities. (Author)

  9. Magnetic shielding for superconducting RF cavities

    Science.gov (United States)

    Masuzawa, M.; Terashima, A.; Tsuchiya, K.; Ueki, R.

    2017-03-01

    Magnetic shielding is a key technology for superconducting radio frequency (RF) cavities. There are basically two approaches for shielding: (1) surround the cavity of interest with high permeability material and divert magnetic flux around it (passive shielding); and (2) create a magnetic field using coils that cancels the ambient magnetic field in the area of interest (active shielding). The choice of approach depends on the magnitude of the ambient magnetic field, residual magnetic field tolerance, shape of the magnetic shield, usage, cost, etc. However, passive shielding is more commonly used for superconducting RF cavities. The issue with passive shielding is that as the volume to be shielded increases, the size of the shielding material increases, thereby leading to cost increase. A recent trend is to place a magnetic shield in a cryogenic environment inside a cryostat, very close to the cavities, reducing the size and volume of the magnetic shield. In this case, the shielding effectiveness at cryogenic temperatures becomes important. We measured the permeabilities of various shielding materials at both room temperature and cryogenic temperature (4 K) and studied shielding degradation at that cryogenic temperature.

  10. Superconducting magnetic shields production. Realisation d'ecrans magnetiques supraconducteurs

    Energy Technology Data Exchange (ETDEWEB)

    Lainee, F; Kormann, R [Thomson-CSF, Domaine de Corbeville, 91 - Orsay (FR); Lainee, F [Ecole des Mines de Paris, 91 - Evry (FR)

    1992-02-01

    Low fields and low frequency shielding properties of YBCO magnetic shields are measured at 77 K. They compare favourably with shielding properties of mumetal shields. Therefore high-T{sub c} superconducting magnetic shields can already be used to shield small volumes. The case of magnetic shields for large volumes is also discussed. 3 refs; 6 figs; 4 tabs.

  11. 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

  12. Conceptual design of ITER shielding blanket

    International Nuclear Information System (INIS)

    Sato, Satoshi; Takatsu, Hideyuki; Kurasawa, Toshimasa

    1995-03-01

    The present report summarizes the design activities of the ITER first wall and shielding blanket conducted by the JA Home Team during this year (1994) in close contact with the JCT, and reported during the four Technical Meetings held at Garching ITER Co-center. These activities are based on the Task Agreement between the JCT and the JA Home Team. In the present report, a layered configuration composed of separate first walls, modular-type blanket modules and separate back plates has been proposed to realize reliable assembly and maintenance schemes as well as to realize reliable component designs under high surface heat loads, high neutron wall loading and electromagnetic loads during disruptions. Outline of the structural design, consideration on fabricability and maintainability, and the results of thermal, mechanical and electromagnetic analyses are described. (author)

  13. 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)

  14. Shield cost minimization using SWAN

    International Nuclear Information System (INIS)

    Watkins, E.F.; Annese, C.E.; Greenspan, E.

    1993-01-01

    The common approach to the search for minimum cost shield designs is open-quotes trial-and-errorclose quotes; it proceeds as follows: 1. Based on prior experience and intuition, divide the shield into zones and assume their composition. 2. Solve the transport equation and calculate the relevant performance characteristics. 3. Change the composition or the geometry of one or a few of the zones and repeat step 2. 4. Repeat step 3 many times until the shield design appears to be optimal. 5. Select a different set of constituents and repeat steps 2,3, and 4. 6. Repeate step 5 a few or many times until the designer can point to the most cost-effective design

  15. Radiation shield for nuclear reactors

    International Nuclear Information System (INIS)

    Weissenfluh, J.A.

    1980-01-01

    A reusable radiation shield for use in a reactor installation comprises a thin-walled, flexible and resilient container, made of plastic or elastomeric material, containing a hydrogenous fluid with boron compounds in solution. The container can be filled and drained in position and the fluid can be recirculated if required. When not in use the container can be folded and stored in a small space. The invention relates to a shield to span the top of the annular space between a reactor vessel and the primary shield. For this purpose a continuous toroidal container or a series of discrete segments is used. Other forms can be employed for different purposes, e.g. mattress- or blanket-like forms can be draped over potential sources of radiation or suspended from a mobile carrier and placed between a worker and a radiation source. (author)

  16. Thermal-hydraulic analysis and design improvement for coolant channel of ITER shield block

    International Nuclear Information System (INIS)

    Zhao Ling; Li Huaqi; Zheng Jiantao; Yi Jingwei; Kang Weishan; Chen Jiming

    2013-01-01

    As an important part for ITER, shield block is used to shield the neutron heat. The structure design of shield block, especially the inner coolant channel design will influence its cooling effect and safety significantly. In this study, the thermal-hydraulic analysis for shield block has been performed by the computational fluid dynamics software, some optimization suggestions have been proposed and thermal-hydraulic characteristics of the improved model has been analyzed again. The analysis results for improved model show that pressure drop through flow path near the inlet and outlet region of the shield block has been reduced, and the total pressure drop in cooling path has been reduced too; the uniformity of the mass flowrate distribution and the velocity distribution have been improved in main cooling branches; the local highest temperature of solid domain reduced considerably, which could avoid thermal stress becoming too large because of coolant effect unevenly. (authors)

  17. Shielding considerations for advanced space nuclear reactor systems

    International Nuclear Information System (INIS)

    Angelo, J.P. Jr.; Buden, D.

    1982-01-01

    To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO 2 ) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications

  18. Shielding walls against ionizing radiation

    International Nuclear Information System (INIS)

    1993-05-01

    Hot-cell shielding walls consist of building blocks made of lead according to DIN 25407 part 1, and of special elements according to DIN 25407 part 2. Alpha-gamma cells can be built using elements for protective contamination boxes according to DIN 25480 part 1. This standards document intends to provide planning engineers, manufacturers, future users and the competent authorities and experts with a basis for the design of hot cells with lead shielding walls and the design of hot-cell equipment. (orig./HP) [de

  19. Nuclear steam generator tubesheet shield

    International Nuclear Information System (INIS)

    Nickerson, J.H.D.; Ruhe, A.

    1982-01-01

    The invention involves improvements to a nuclear steam generator of the type in which a plurality of U-shaped tubes are connected at opposite ends to a tubesheet and extend between inlet and outlet chambers, with the steam generator including an integral preheater zone adjacent to the downflow legs of the U-shaped tubes. The improvement is a thermal shield disposed adjacent to an upper face of the tubesheet within the preheater zone, the shield including ductile cladding material applied directly to the upper face of the tubesheet, with the downflow legs of the U-shaped tubes extending through the cladding into the tubesheet

  20. Shielding structure analysis for LSDS facility

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hong Yeop; Kim, Jeong Dong; Lee, Yong Deok; Kim, Ho Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The nuclear material (Pyro, Spent nuclear fuel) itself and the target material to generate neutrons is the LSDS system for isotopic fissile assay release of high intensity neutron and gamma rays. This research was performed to shield from various strong radiation. A shielding evaluation was carried out with a facilities model of LSDS system. The MCNPX 2.5 code was used and a shielding evaluation was performed for the shielding structure and location. The radiation dose based on the hole structure and location of the wall was evaluated. The shielding evaluation was performed to satisfy the safety standard for a normal person (1 μSv/h) and to use enough interior space. The MCNPX2.5 code was used and a dose evaluation was performed for the location of the shielding material, shielding structure, and hole structure. The evaluation result differs according to the shielding material location. The dose rate was small when the shielding material was positioned at the center. The dose evaluation result regarding the location of the shielding material was applied to the facility and the shielding thickness was determined (In 50 cm + Borax 5 cm + Out 45cm). In the existing hole structure, the radiation leak is higher than the standard. A hole structure model to prevent leakage of radiation was proposed. The general public dose limit was satisfied when using the concrete reinforcement and a zigzag structure. The shielding result will be of help to the facility shielding optimization.

  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. Shielding structure analysis for LSDS facility

    International Nuclear Information System (INIS)

    Choi, Hong Yeop; Kim, Jeong Dong; Lee, Yong Deok; Kim, Ho Dong

    2014-01-01

    The nuclear material (Pyro, Spent nuclear fuel) itself and the target material to generate neutrons is the LSDS system for isotopic fissile assay release of high intensity neutron and gamma rays. This research was performed to shield from various strong radiation. A shielding evaluation was carried out with a facilities model of LSDS system. The MCNPX 2.5 code was used and a shielding evaluation was performed for the shielding structure and location. The radiation dose based on the hole structure and location of the wall was evaluated. The shielding evaluation was performed to satisfy the safety standard for a normal person (1 μSv/h) and to use enough interior space. The MCNPX2.5 code was used and a dose evaluation was performed for the location of the shielding material, shielding structure, and hole structure. The evaluation result differs according to the shielding material location. The dose rate was small when the shielding material was positioned at the center. The dose evaluation result regarding the location of the shielding material was applied to the facility and the shielding thickness was determined (In 50 cm + Borax 5 cm + Out 45cm). In the existing hole structure, the radiation leak is higher than the standard. A hole structure model to prevent leakage of radiation was proposed. The general public dose limit was satisfied when using the concrete reinforcement and a zigzag structure. The shielding result will be of help to the facility shielding optimization

  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. Protected isotope heat source

    International Nuclear Information System (INIS)

    Burns, R.K.; Shure, L.I.; Katzen, E.D.

    1975-01-01

    A radioactive isotope capsule is disposed in a container (heat shield) which will have a single stable trim attitude when reentering the earth's atmosphere and while falling to earth. The center of gravity of the heat source is located forward of the midpoint between the front face and the rear face of the container. The capsule is insulated from the front face of the container but not from the rear surface of the container. (auth)

  6. A study on the characteristics of modified and novolac type epoxy resin based neutron shielding material

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Soo Haeng; Hong, Sun Seok; Oh, Seung Chul; Do, Jae Bum [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-10-01

    Because the exposure to radiation in the nuclear facilities can be fatal to human, it is important to reduce the radiation dose level to a tolerable level. The purpose of this study is to develop highly effective neutron shielding materials for the shipping and storage cask of radioactive materials or in the nuclear/radiation facilities. In this study, we developed modified and novolac type epoxy resin based neutron shielding materials and their various material properties, including neutron shielding ability, prolonged time heat resistance, thermal and mechanical properties were evaluated experimently. (author). 31 refs., 27 figs., 16 tabs.

  7. TORE-SUPRA: design of thermal radiation shield at 80 K

    International Nuclear Information System (INIS)

    Aymar, R.; Cordier, J.J.; Deschamps, P.; Gauthier, A.; Perin, J.P.

    1982-09-01

    The TORE-SUPRA superconducting toroidal magnet operating at liquid helium temperature, must be protected against thermal radiation from the vessels. For this purpose, stainless steel heat shields, cooled at 80 K, are positioned between coil casings at 4.5 K and the vessels, and constitute a double stiff toroid which completely surrounds the magnet. Mockups have been manufactured to study their design and operating problems. Calculations have also been made to analyse the mechanical behaviour of these shields

  8. Method of constructing shielding wall

    International Nuclear Information System (INIS)

    Nagao, Tetsuya.

    1990-01-01

    For instance, surfaces of lead particles each formed into a sphere of about 0.5 to 0.3 mm grain size are coated with a coating material of a synthetic resin comprising a polymeric material such as teflon. Subsequently, the floated lead particle are kneaded with concrete materials and then poured into a molding die by way of a hose. After coagulation, the molding die is removed to complete shielding walls in which lead particles are scattered substantially at an equal distance. In this way, since the lead particles are mixed into the shielding walls, shielding effects can be improved by so much as the lead particles are mixed, thereby enabling to reduce the thickness of the shielding walls. Further, since the lead particles are coated with the coating material, the lead particles are insulated from the concrete materials, thereby enabling to prevent the corrosion of the lead particles. Furthermore, since the lead particles and the concrete materials can be transported with ease, operation labors can be reduced. (T.M.)

  9. Shielding Effectiveness of a Thin Film Window

    National Research Council Canada - National Science Library

    Johnson, Eric

    1998-01-01

    .... The predicted shielding effectiveness was 29 dB based on theoretical calculations. The error analysis of the shielding effectiveness showed that this predicted value was within the measurement error...

  10. 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.

  11. Infinite slab-shield dose calculations

    International Nuclear Information System (INIS)

    Russell, G.J.

    1989-01-01

    I calculated neutron and gamma-ray equivalent doses leaking through a variety of infinite (laminate) slab-shields. In the shield computations, I used, as the incident neutron spectrum, the leakage spectrum (<20 MeV) calculated for the LANSCE tungsten production target at 90 degree to the target axis. The shield thickness was fixed at 60 cm. The results of the shield calculations show a minimum in the total leakage equivalent dose if the shield is 40-45 cm of iron followed by 20-15 cm of borated (5% B) polyethylene. High-performance shields can be attained by using multiple laminations. The calculated dose at the shield surface is very dependent on shield material. 4 refs., 4 figs., 1 tab

  12. Using glass as a shielding material

    International Nuclear Information System (INIS)

    Yousef, S.

    2002-04-01

    Different theoretical and technological concepts and problems in using glass as a shielding material was discussed, some primarily designs for different types of radiation shielding windows were illustrated. (author)

  13. Using glass as a shielding material

    International Nuclear Information System (INIS)

    Yousef, S.

    2003-01-01

    Different theoretical and technological concepts and problems in using glass as a shielding material was discussed, some primarily designs for different types of radiation shielding windows were illustrated. (author)

  14. 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.

  15. 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.

  16. Sensitivity and uncertainty analysis of NET/ITER shielding blankets

    International Nuclear Information System (INIS)

    Hogenbirk, A.; Gruppelaar, H.; Verschuur, K.A.

    1990-09-01

    Results are presented of sensitivity and uncertainty calculations based upon the European fusion file (EFF-1). The effect of uncertainties in Fe, Cr and Ni cross sections on the nuclear heating in the coils of a NET/ITER shielding blanket has been studied. The analysis has been performed for the total cross section as well as partial cross sections. The correct expression for the sensitivity profile was used, including the gain term. The resulting uncertainty in the nuclear heating lies between 10 and 20 per cent. (author). 18 refs.; 2 figs.; 2 tabs

  17. Shielding and grounding in large detectors

    International Nuclear Information System (INIS)

    Radeka, V.

    1998-09-01

    Prevention of electromagnetic interference (EMI), or ''noise pickup,'' is an important design aspect in large detectors in accelerator environments. Shielding effectiveness as a function of shield thickness and conductivity vs the type and frequency of the interference field is described. Noise induced in transmission lines by ground loop driven currents in the shield is evaluated and the importance of low shield resistance is emphasized. Some measures for prevention of ground loops and isolation of detector-readout systems are discussed

  18. Pretinning Nickel-Plated Wire Shields

    Science.gov (United States)

    Igawa, J. A.

    1985-01-01

    Nickel-plated copper shielding for wires pretinned for subsequent soldering with help of activated rosin flux. Shield cut at point 0.25 to 0.375 in. (6 to 10 mm) from cut end of outer jacket. Loosened end of shield straightened and pulled toward cut end. Insulation of inner wires kept intact during pretinning.

  19. Computed tomography shielding methods: a literature review.

    Science.gov (United States)

    Curtis, Jessica Ryann

    2010-01-01

    To investigate available shielding methods in an effort to further awareness and understanding of existing preventive measures related to patient exposure in computed tomography (CT) scanning. Searches were conducted to locate literature discussing the effectiveness of commercially available shields. Literature containing information regarding breast, gonad, eye and thyroid shielding was identified. Because of rapidly advancing technology, the selection of articles was limited to those published within the past 5 years. The selected studies were examined using the following topics as guidelines: the effectiveness of the shield (percentage of dose reduction), the shield's effect on image quality, arguments for or against its use (including practicality) and overall recommendation for its use in clinical practice. Only a limited number of studies have been performed on the use of shields for the eyes, thyroid and gonads, but the evidence shows an overall benefit to their use. Breast shielding has been the most studied shielding method, with consistent agreement throughout the literature on its effectiveness at reducing radiation dose. The effect of shielding on image quality was not remarkable in a majority of studies. Although it is noted that more studies need to be conducted regarding the impact on image quality, the currently published literature stresses the importance of shielding in reducing dose. Commercially available shields for the breast, thyroid, eyes and gonads should be implemented in clinical practice. Further research is needed to ascertain the prevalence of shielding in the clinical setting.

  20. BRH Gonad Shielding Program: where it has led

    International Nuclear Information System (INIS)

    Arcarese, J.S.

    1975-01-01

    Some topics discussed are: Bureau of Radiological Health guidelines; types of gonad shields; specific area shielding; gonad shielding guidelines; and publication of pamphlet on types of shields and circumstances under which they should be used

  1. Radiological shielding of low power compact reactor: calculation and design

    International Nuclear Information System (INIS)

    Marino, Raul

    2004-01-01

    The development of compact reactors becoming a technology that offers great projection and innumerable use possibilities, both in electricity generation and in propulsion.One of the requirements for the operation of this type of reactor is that it must include a radiological shield that will allow for different types of configurations and that, may be moved with the reactor if it needs to be transported.The nucleus of a reactor emits radiation, mainly neutrons and gamma rays in the heat of power, and gamma radiation during the radioactive decay of fission products.This radiation must be restrained in both conditions of operation to avoid it affecting workers or the public.The combination of different materials and properties in layers results in better performance in the form of a decrease in radiation, hence causing the dosage outside the reactor, whether in operation or shut down, to fall within the allowed limits.The calculations and design of radiological shields is therefore of paramount importance in reactor design.The choice of material and the design of the shield have a strong impact on the cost and the load capacity, the latter being one of the characteristics to optimize.The imposed condition of design is that the reactor can be transported together with the decay shield in a standard container of 40 foot [es

  2. The status of shielding research at Tajoura research center

    International Nuclear Information System (INIS)

    El-Bakkoush, F.A.

    2005-01-01

    This paper gives a description to the shielding research activities which have been carried-out at the radiation shielding group ,Tajoura Research Center. This includes the design of different types of concrete shields made from local aggregates which have suitable radiation attenuation properties. These include, Ordinary Concrete(with density p = 2.3 ton/m3) heavy weight concrete (with density p =3.6 ton/m3) and heat resistant concrete with aggregates having bound- in water. Investigation have been carried -out by measuring the neutron and gamma-rays spectra which have been transmitted through barriers having different thickness. These were performed using a collimated beam of reactor neutrons and gamma-ray transmitted from the horizontal channel no 1 of Tajoura-Research reactor with 10 MW Max ape rating power. The transmitted fast neutron and gamma spectra were measured by neutron-gamma spectrometer employing NE-213 liquid organic scintillater. Discrimination of against undesired pulses of neutrons or gamma-ray was achieved by a pulse shape discrimination method based on differences in the shape of the decay part of the emitted pulses. The obtained results are presented in the form of displayed neutron and gamma spectra measured behind different thickness of the investigated concrete shield. These spectra were used to derive the macroscopic cross section for at different energy for material under investigation

  3. Radiation shield analysis for a manned Mars rover

    International Nuclear Information System (INIS)

    Morley, N.J.; ElGenk, M.S.

    1991-01-01

    Radiation shielding for unmanned space missions has been extensively studied; however, designs of man-rated shields are minimal. Engle et al.'s analysis of a man-rated, multilayered shield composed of two and three cycles (a cycle consists of a tungsten and a lithium hydride layer) is the basis for the work reported in this paper. The authors present the results of a recent study of shield designs for a manned Mars rover powered by a 500-kW(thermal) nuclear reactor. A train-type rover vehicle was developed, which consists of four cars and is powered by an SP-100-type nuclear reactor heat source. The maximum permissible dose rate (MPD) from all sources is given by the National Council on Radiation Protection and Measurements as 500 mSv/yr (50 rem/yr) A 3-yr Mars mission (2-yr round trip and 1-yr stay) will deliver a 1-Sv natural radiation dose without a solar particle event, 450 mSv/yr in flight, and an additional 100 mSv on the planet surface. An anomalously large solar particle event could increase the natural radiation dose for unshielded astronauts on the Martian surface to 200 mSv. This limits the MPD to crew members from the nuclear reactor to 300 mSv

  4. Analysis of ferromagnetic shielding of the ITER NBI

    International Nuclear Information System (INIS)

    Roccella, M.; Lucca, F.; Roccella, R.; Cocilovo, V.; Ramogida, G.; Portone, A.; Tanga, A.; Formisano, A.; Martone, R.

    2006-01-01

    In ITER two heating and one diagnostic Neutral Beam Injectors (NBIs) are foreseen [P. L. Mondino et al., ''ITER neutral beam system '', Nucl. Fus., vol. 40, p. 501 (2000)]. Inside these components there are very stringent limits on the magnetic field (the flux density must be below some Gauss (G) along the ion path and below 20 G in the neutralizing region). To achieve these performances in an environment with high stray field due to the plasma and the poloidal field coils, both passive and active shielding systems are foreseen. The present design of the Magnetic Field Reduction System (MFRS) is made of seven active coils and of a box surrounding the NBI region, consisting of ferromagnetic plates 15 cm thick. The electromagnetic analysis of the effectiveness of these shields has been performed by a full 3D FEM model using the ANSYS code. To perform the FEM modeling of the component special care has been used to face the particular geometrical features of the component (a box of about 15 x 5 x 5 m vs. a ferromagnetic layer of only 15 cm thick). To insert an adequate number of FEM elements (at least 5) in the thickness of the ferromagnetic layer, without a prohibitive increase in the total FEM elements number, a particular modeling approach (a sort of '' Chinese boxes '' technique) has been developed. Due to this technique the FEM model enclosing the ferromagnetic box results completely independent on the fine FEM structure inside the shielding layer. It has been even possible, using this technique, introducing a thin (below 1 cm thick) slot all through the shielding plates, without perturbing the rest of the model. This slot has been used to analyze the effects of possible manufacturing lacks on the residual magnetic field inside the component. This technique has allowed the use of only structured meshes made by brick elements, much more accurate than the tetra elements, needed in the usual free meshing techniques. To have the possibility of changing the shielding

  5. Somatostatin receptor scintigraphy in advanced renal cell carcinoma. Results of a phase II-trial of somatostatine analogue therapy in patients with advanced RCC

    International Nuclear Information System (INIS)

    Freudenberg, L.S.; Goerges, R.; Stergar, H.; Bockisch, A.; Gauler, T.; Bauer, S.; Antoch, G.; Schuette, J.

    2008-01-01

    Aims: objective of this prospective study was to evaluate the role of somatostatin receptor scintigraphy (SRS) in advanced renal cell carcinoma (RCC) with respect to potential therapy with somatostatin analogue (SST-A) and to assess the response rate under therapy with SST-A. Patients, methods: 16 patients with documented progression of histologically confirmed advanced RCC were included. Planar whole-body SRS was performed 4, 24 and 48h post i.v. injection of 175-200 MBq 111 In-pentetreoide. 5 and 25 h p.i. SPECT of thorax and abdomen were performed. Documentation of somatostatin receptor expression via SRS in > 50% of known tumour lesions was the criteria for treatment start with SST-A (Sandostatin LAR registered -Depot 30mg i.m. every four weeks). Results: in 9/16 of the patients SRS showed at least one metastasis with moderate (n = 5) or intense (n = 4) tracer uptake. Lesion-based SRS evaluation showed only 12.1% (20/165) of all metastases. Most false-negative lesions were located in the lungs. In too patients, the majority of the known metastases was SRS positive and these patients received SST-A therapy. The first radiographic evaluation after a two-month interval showed progressive disease in both patients. Conclusions: we conclude that SRS is of limited value in staging of advanced RCC. In our patients SST-A did not result in a growth control of RCC. Consequently, the use of SST-A in advanced RCC seems to be no relevant therapeutic option. (orig.)

  6. Survivor shielding. Part A. Nagasaki factory worker shielding

    International Nuclear Information System (INIS)

    Santoro, Robert T.; Barnes, John M.; Azmy, Yousry Y.; Kerr, George D.; Egbert, Stephen D.; Cullings, Harry M.

    2005-01-01

    Recent investigations based on conventional chromosome aberration data by the RERF suggest that the DS86 doses received by many Nagasaki factory workers may have been overestimated by as much as 40% relative to those for other survivors in Japanese-type houses and other shielding configurations (Kodama et al. 2001). Since the factory workers represent about 25% of the Nagasaki survivors with DS86 doses in excess of 0.5 Gy (50 rad), systematic errors in their dose estimates can have a major impact on the risk coefficients from RERF studies. The factory worker doses may have been overestimated for a number of reasons. The calculation techniques, including the factory building modeling, weapon source spectra and cross-section data used in the DS86 shielding calculations were not detailed enough to replicate actual conditions. The models used did not take into account local shielding provided by machinery, tools, and the internal structure in the buildings. In addition, changes in the disposition of shielding following collapse of the building by the blast wave were not considered. The location of large factory complexes may be uncertain, causing large numbers of factory survivors, correctly located relative to each other, to be uniformly too close to the hypocenter. Any or all of these reasons are sufficient to result in an overestimate of the factory worker doses. During the DS02 studies, factory worker doses have been reassessed by more carefully modeling the factory buildings, incorporating improved radiation transport methods and cross-section data and using the most recent bomb leakage spectra (Chapter 2). Two-dimensional discrete ordinates calculations were carried out initially to estimate the effects of workbenches and tools on worker doses to determine if the inclusion of these components would, in fact, reduce the dose by amounts consistent with the RERF observations (Kodama et al. 2001). (author)

  7. Comparative evaluation of structural integrity for ITER blanket shield block based on SDC-IC and ASME code

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Hee-Jin [ITER Korea, National Fusion Research Institute, 169-148 Gwahak-Ro, Yuseong-Gu, Daejeon (Korea, Republic of); Ha, Min-Su, E-mail: msha12@nfri.re.kr [ITER Korea, National Fusion Research Institute, 169-148 Gwahak-Ro, Yuseong-Gu, Daejeon (Korea, Republic of); Kim, Sa-Woong; Jung, Hun-Chea [ITER Korea, National Fusion Research Institute, 169-148 Gwahak-Ro, Yuseong-Gu, Daejeon (Korea, Republic of); Kim, Duck-Hoi [ITER Organization, Route de Vinon sur Verdon - CS 90046, 13067 Sant Paul Lez Durance (France)

    2016-11-01

    Highlights: • The procedure of structural integrity and fatigue assessment was described. • Case studies were performed according to both SDC-IC and ASME Sec. • III codes The conservatism of the ASME code was demonstrated. • The study only covers the specifically comparable case about fatigue usage factor. - Abstract: The ITER blanket Shield Block is a bulk structure to absorb radiation and to provide thermal shielding to vacuum vessel and external vessel components, therefore the most significant load for Shield Block is the thermal load. In the previous study, the thermo-mechanical analysis has been performed under the inductive operation as representative loading condition. And the fatigue evaluations were conducted to assure structural integrity for Shield Block according to Structural Design Criteria for In-vessel Components (SDC-IC) which provided by ITER Organization (IO) based on the code of RCC-MR. Generally, ASME code (especially, B&PV Sec. III) is widely applied for design of nuclear components, and is usually well known as more conservative than other specific codes. For the view point of the fatigue assessment, ASME code is very conservative compared with SDC-IC in terms of the reflected K{sub e} factor, design fatigue curve and other factors. Therefore, an accurate fatigue assessment comparison is needed to measure of conservatism. The purpose of this study is to provide the fatigue usage comparison resulting from the specified operating conditions shall be evaluated for Shield Block based on both SDC-IC and ASME code, and to discuss the conservatism of the results.

  8. Comparative evaluation of structural integrity for ITER blanket shield block based on SDC-IC and ASME code

    International Nuclear Information System (INIS)

    Shim, Hee-Jin; Ha, Min-Su; Kim, Sa-Woong; Jung, Hun-Chea; Kim, Duck-Hoi

    2016-01-01

    Highlights: • The procedure of structural integrity and fatigue assessment was described. • Case studies were performed according to both SDC-IC and ASME Sec. • III codes The conservatism of the ASME code was demonstrated. • The study only covers the specifically comparable case about fatigue usage factor. - Abstract: The ITER blanket Shield Block is a bulk structure to absorb radiation and to provide thermal shielding to vacuum vessel and external vessel components, therefore the most significant load for Shield Block is the thermal load. In the previous study, the thermo-mechanical analysis has been performed under the inductive operation as representative loading condition. And the fatigue evaluations were conducted to assure structural integrity for Shield Block according to Structural Design Criteria for In-vessel Components (SDC-IC) which provided by ITER Organization (IO) based on the code of RCC-MR. Generally, ASME code (especially, B&PV Sec. III) is widely applied for design of nuclear components, and is usually well known as more conservative than other specific codes. For the view point of the fatigue assessment, ASME code is very conservative compared with SDC-IC in terms of the reflected K_e factor, design fatigue curve and other factors. Therefore, an accurate fatigue assessment comparison is needed to measure of conservatism. The purpose of this study is to provide the fatigue usage comparison resulting from the specified operating conditions shall be evaluated for Shield Block based on both SDC-IC and ASME code, and to discuss the conservatism of the results.

  9. New facility shield design criteria

    International Nuclear Information System (INIS)

    Howell, W.P.

    1981-07-01

    The purpose of the criteria presented here is to provide standard guidance for the design of nuclear radiation shields thoughout new facilities. These criteria are required to assure a consistent and integrated design that can be operated safely and economically within the DOE standards. The scope of this report is confined to the consideration of radiation shielding for contained sources. The whole body dose limit established by the DOE applies to all doses which are generally distributed throughout the trunk of the body. Therefore, where the whole body is the critical organ for an internally deposited radionuclide, the whole body dose limit applies to the sum of doses received must assure control of the concentration of radionuclides in the building atmosphere and thereby limit the dose from internal sources

  10. INERT GAS SHIELD FOR WELDING

    Science.gov (United States)

    Jones, S.O.; Daly, F.V.

    1958-10-14

    S>An inert gas shield is presented for arc-welding materials such as zirconium that tend to oxidize rapidly in air. The device comprises a rectangular metal box into which the welding electrode is introduced through a rubber diaphragm to provide flexibility. The front of the box is provided with a wlndow having a small hole through which flller metal is introduced. The box is supplied with an inert gas to exclude the atmosphere, and with cooling water to promote the solidification of the weld while in tbe inert atmosphere. A separate water-cooled copper backing bar is provided underneath the joint to be welded to contain the melt-through at the root of the joint, shielding the root of the joint with its own supply of inert gas and cooling the deposited weld metal. This device facilitates the welding of large workpieces of zirconium frequently encountered in reactor construction.

  11. Facility target insert shielding assessment

    Energy Technology Data Exchange (ETDEWEB)

    Mocko, Michal [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-10-06

    Main objective of this report is to assess the basic shielding requirements for the vertical target insert and retrieval port. We used the baseline design for the vertical target insert in our calculations. The insert sits in the 12”-diameter cylindrical shaft extending from the service alley in the top floor of the facility all the way down to the target location. The target retrieval mechanism is a long rod with the target assembly attached and running the entire length of the vertical shaft. The insert also houses the helium cooling supply and return lines each with 2” diameter. In the present study we focused on calculating the neutron and photon dose rate fields on top of the target insert/retrieval mechanism in the service alley. Additionally, we studied a few prototypical configurations of the shielding layers in the vertical insert as well as on the top.

  12. Shielding calculational system for plutonium

    International Nuclear Information System (INIS)

    Zimmerman, M.G.; Thomsen, D.H.

    1975-08-01

    A computer calculational system has been developed and assembled specifically for calculating dose rates in AEC plutonium fabrication facilities. The system consists of two computer codes and all nuclear data necessary for calculation of neutron and gamma dose rates from plutonium. The codes include the multigroup version of the Battelle Monte Carlo code for solution of general neutron and gamma shielding problems and the PUSHLD code for solution of shielding problems where low energy gamma and x-rays are important. The nuclear data consists of built in neutron and gamma yields and spectra for various plutonium compounds, an automatic calculation of age effects and all cross-sections commonly used. Experimental correlations have been performed to verify portions of the calculational system. (23 tables, 7 figs, 16 refs) (U.S.)

  13. Radiation-shielding transparent material

    International Nuclear Information System (INIS)

    Kusumeki, Asao.

    1983-01-01

    Purpose : To obtain radiation-shielding transparent material having a high resistivity to the radioactive rays or light irradiation which is greater at least by two digits as compared with lead glass. Constitution : The shielding material is composed of a saturated aqueous solution zinc iodide. Zinc iodide (specific gravity of 4.2) is dissolved by 430 g into 100 cc of water at a temperature of 20 0 C and forms a heavy liquid with a specific gravity of 2.80. The radiation length of the heavy liquid is 3.8 cm which is 1.5 times as large as lead glass. The light transmission is greater than 95% in average. Furthermore, by adding hypophosphorous acid as a reducing agent to the aqueous solution of the lead iodide, the material is stabilized against the irradiation of light or radioactive rays and causes no discoloration for a long time. (Moriyama, K.)

  14. Radiation shield vest and skirt

    International Nuclear Information System (INIS)

    Maine, G.J.

    1982-01-01

    A two-piece garment is described which provides shielding for female workers exposed to radiation. The upper part is a vest, overlapping and secured in the front by adjustable closures. The bottom part is a wraparound skirt, also secured by adjustable closures. The two parts overlap, thus providing continuous protection from shoulder to knee and ensuring that the back part of the body is protected as well as the front

  15. Handbook of radiation shielding data

    International Nuclear Information System (INIS)

    Courtney, J.C.

    1976-07-01

    This handbook is a compilation of data on units, conversion factors, geometric considerations, sources of radiation, and the attenuation of photons, neutrons, and charged particles. It also includes related topics in health physics. Data are presented in tabular and graphical form with sufficient narrative for a least first-approximation solutions to a variety of problems in nuclear radiation protection. Members of the radiation shielding community contributed the information in this document from unclassified and uncopyrighted sources, as referenced

  16. Heat-pipe thermionic reactor concept

    DEFF Research Database (Denmark)

    Storm Pedersen, E.

    1967-01-01

    Main components are reactor core, heat pipe, thermionic converter, secondary cooling system, and waste heat radiator; thermal power generated in reactor core is transported by heat pipes to thermionic converters located outside reactor core behind radiation shield; thermionic emitters are in direct...

  17. Design of ITER shielding blanket

    International Nuclear Information System (INIS)

    Furuya, Kazuyuki; Sato, Satoshi; Hatano, Toshihisa; Tokami, Ikuhide; Kitamura, Kazunori; Miura, Hidenori; Ito, Yutaka; Kuroda, Toshimasa; Takatsu, Hideyuki

    1997-05-01

    A mechanical configuration of ITER integrated primary first wall/shield blanket module were developed focusing on the welded attachment of its support leg to the back plate. A 100 mm x 150 mm space between the legs of adjacent modules was incorporated for the working space of welding/cutting tools. A concept of coolant branch pipe connection to accommodate deformation due to the leg welding and differential displacement of the module and the manifold/back plate during operation was introduced. Two-dimensional FEM analyses showed that thermal stresses in Cu-alloy (first wall) and stainless steel (first wall coolant tube and shield block) satisfied the stress criteria following ASME code for ITER BPP operation. On the other hand, three-dimensional FEM analyses for overall in-vessel structures exhibited excessive primary stresses in the back plate and its support structure to the vacuum vessel under VDE disruption load and marginal stresses in the support leg of module No.4. Fabrication procedure of the integrated primary first wall/shield blanket module was developed based on single step solid HIP for the joining of Cu-alloy/Cu-alloy, Cu-alloy/stainless steel, and stainless steel/stainless steel. (author)

  18. Photonic Bandgap (PBG) Shielding Technology

    Science.gov (United States)

    Bastin, Gary L.

    2007-01-01

    Photonic Bandgap (PBG) shielding technology is a new approach to designing electromagnetic shielding materials for mitigating Electromagnetic Interference (EM!) with small, light-weight shielding materials. It focuses on ground planes of printed wiring boards (PWBs), rather than on components. Modem PSG materials also are emerging based on planar materials, in place of earlier, bulkier, 3-dimensional PBG structures. Planar PBG designs especially show great promise in mitigating and suppressing EMI and crosstalk for aerospace designs, such as needed for NASA's Constellation Program, for returning humans to the moon and for use by our first human visitors traveling to and from Mars. Photonic Bandgap (PBG) materials are also known as artificial dielectrics, meta-materials, and photonic crystals. General PBG materials are fundamentally periodic slow-wave structures in I, 2, or 3 dimensions. By adjusting the choice of structure periodicities in terms of size and recurring structure spacings, multiple scatterings of surface waves can be created that act as a forbidden energy gap (i.e., a range of frequencies) over which nominally-conductive metallic conductors cease to be a conductor and become dielectrics. Equivalently, PBG materials can be regarded as giving rise to forbidden energy gaps in metals without chemical doping, analogous to electron bandgap properties that previously gave rise to the modem semiconductor industry 60 years ago. Electromagnetic waves cannot propagate over bandgap regions that are created with PBG materials, that is, over frequencies for which a bandgap is artificially created through introducing periodic defects

  19. Reactor vessel head permanent shield

    International Nuclear Information System (INIS)

    Hankinson, M.F.; Leduc, R.J.; Richard, J.W.; Malandra, L.J.

    1989-01-01

    A nuclear reactor is described comprising: a nuclear reactor pressure vessel closure head; control rod drive mechanisms (CRDMs) disposed within the closure head so as to project vertically above the closure head; cooling air baffle means surrounding the control rod drive mechanisms for defining cooling air paths relative to the control rod drive mechanisms; means defined within the periphery of the closure head for accommodating fastening means for securing the closure head to its associated pressure vessel; lifting lugs fixedly secured to the closure head for facilitating lifting and lowering movements of the closure head relative to the pressure vessel; lift rods respectively operatively associated with the plurality of lifting lugs for transmitting load forces, developed during the lifting and lowering movements of the closure head, to the lifting lugs; upstanding radiation shield means interposed between the cooling air baffle means and the periphery of the enclosure head of shielding maintenance personnel operatively working upon the closure head fastening means from the effects of radiation which may emanate from the control rod drive mechanisms and the cooling air baffle means; and connecting systems respectively associated with each one of the lifting lugs and each one of the lifting rods for connecting each one of the lifting rods to a respective one of each one of the lifting lugs, and for simultaneously connecting a lower end portion of the upstanding radiation shield means to each one of the respective lifting lugs

  20. ATLAS Award for Shield Supplier

    CERN Multimedia

    2004-01-01

    ATLAS technical coordinator Dr. Marzio Nessi presents the ATLAS supplier award to Vojtech Novotny, Director General of Skoda Hute.On 3 November, the ATLAS experiment honoured one of its suppliers, Skoda Hute s.r.o., of Plzen, Czech Republic, for their work on the detector's forward shielding elements. These huge and very massive cylinders surround the beampipe at either end of the detector to block stray particles from interfering with the ATLAS's muon chambers. For the shields, Skoda Hute produced 10 cast iron pieces with a total weight of 780 tonnes at a cost of 1.4 million CHF. Although there are many iron foundries in the CERN member states, there are only a limited number that can produce castings of the necessary size: the large pieces range in weight from 59 to 89 tonnes and are up to 1.5 metres thick.The forward shielding was designed by the ATLAS Technical Coordination in close collaboration with the ATLAS groups from the Czech Technical University and Charles University in Prague. The Czech groups a...

  1. Shield or not to Shield: Effects of Solar Radiation on Water Temperature Sensor Accuracy

    Directory of Open Access Journals (Sweden)

    Robert L. Wilby

    2013-10-01

    Full Text Available Temperature sensors are potentially susceptible to errors due to heating by solar radiation. Although this is well known for air temperature (Ta, significance to continuous water temperature (Tw monitoring is relatively untested. This paper assesses radiative errors by comparing measurements of exposed and shielded Tinytag sensors under indirect and direct solar radiation, and in laboratory experiments under controlled, artificial light. In shallow, still-water and under direct solar radiation, measurement discrepancies between exposed and shielded sensors averaged 0.4 °C but can reach 1.6 °C. Around 0.3 °C of this inconsistency is explained by variance in measurement accuracy between sensors; the remainder is attributed to solar radiation. Discrepancies were found to increase with light intensity, but to attain Tw differences in excess of 0.5 °C requires direct, bright solar radiation (>400 W m−2 in the total spectrum. Under laboratory conditions, radiative errors are an order of magnitude lower when thermistors are placed in flowing water (even at velocities as low as 0.1 m s−1. Radiative errors were also modest relative to the discrepancy between different thermistor manufacturers. Based on these controlled experiments, a set of guidelines are recommended for deploying thermistor arrays in water bodies.

  2. 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.)

  3. Preparation of a basic data base for shielding design. 4

    International Nuclear Information System (INIS)

    Nakao, Makoto; Takemura, Morio

    1999-03-01

    With use of a standard groupwise shielding design library JSSTDL produced from the latest evaluated nuclear data library JENDL-3.2, experimental analyses for the JASPER experiments were performed. In order to verify the new version of JSSTDL, whose cross sections of thermal energy region was updated, the polyethylene transmission experiments of the Special Materials Experiment was analysed again, and also zirconium transmission experiment of the Experiment was newly analysed. JSSTDL was applied to the analysis of neutron multiplicative region of the in-vessel fuel storage mockup configurations in the IVS Experiment. Also it was applied to the analyses of neutron streaming effect through the mockup of sodium window in B 4 C shield in the Flux Monitor Experiment and also the mockup of narrow gaps in thick concrete shield in the Gap streaming Experiment. The results were compared with those obtained by the same analysis method and input data using the JSDJ2 library that had been applied consistently to the JASPER experiment analyses. Although the analysis with the new version of JSSTDL resulted in a little reduction of overestimation in the polyethylene transmission configuration, the results obtained with JSSTDL are, in general, higher than those with JSDJ2 as had been found in analyses in preceding years for the Radial Shield Attenuation Experiment, the Axial Shield Experiment, the Intermediate Heat Exchanger Experiment and so on. Compilation of the input data necessary for future reanalyses of important configurations in JASPER experiments, that were selected at the first stage of this study, were continued and new data were added into the computer disk holding previously accumulated data. (author)

  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.

    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

  5. Loss of coolant analysis for the tower shielding reactor 2

    International Nuclear Information System (INIS)

    Radcliff, T.D.; Williams, P.T.

    1990-06-01

    The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs

  6. Electromagnetic Shielding Characteristics of Eco-Friendly Foamed Concrete Wall

    Directory of Open Access Journals (Sweden)

    Sung-Sil Cho

    2017-01-01

    Full Text Available The electromagnetic shielding characteristics according to the material composition of foamed concrete, which was manufactured to reduce environmental pollution and to economically apply it in actual building walls, were researched herein. Industrial by-products such as ladle furnace slag (LFS, gypsum, and blast furnace slag (BFS were added to manufacture foamed concrete with enhanced functionalities such as lightweight, heat insulation, and sound insulation. The electrical characteristics such as permittivity and loss tangent according to the foam and BFS content were calculated and measured. Free space measurement was used to measure the electromagnetic shielding characteristics of the actually manufactured foamed concrete. It was confirmed that electromagnetic signals were better blocked when the foam content was low and the BFS content was high in the measured frequency bands (1–8 GHz and that approximately 90% of the electromagnetic signals were blocked over 4 GHz.

  7. Development of ITER shielding blanket prototype mockup by HIP bonding

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Satoshi; Furuya, Kazuyuki; Hatano, Toshihisa; Kuroda, Toshimasa; Enoeda, Mikio; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Takatsu, Hideyuki [Japan Atomic Energy Research Inst., Office of ITER Project Promotion, Tokyo (Japan)

    2000-07-01

    A prototype ({approx}900{sup H} x 1700{sup W} x 350{sup T} mm) of the ITER shielding blanket module has been fabricated following the previous successful fabrication of a small-scale ({approx}500{sup H} x 400{sup W} x 150{sup T} mm) and mid-scale ({approx}800{sup H} x 500{sup W} x 350{sup T} mm) mock-ups. This prototype incorporates most of key design features essential to the fabrication of the ITER shielding blanket module such as 1) the first wall heat sink made of Al{sub 2}O{sub 3} dispersion strengthened Cu (DSCu) with built-in SS316L coolant tubes bonded to a massive SS316LN shield block, 2) toroidally curved first wall with a radius of 5106 mm while straight in poloidal direction, 3) coolant channels oriented in poloidal direction in the first wall and in toroidal direction in the shield block, 4) the first wall coolant channel routing to avoid the interference with the front access holes, 5) coolant channels drilled through the forged SS316LN-IG shield block, and 6) four front access holes of 30 mm in diameter penetrated through the first wall and the shield block. For the joining method, especially for the first wall/side wall parts and the shield block, the solid HIP (Hot Isostatic Pressing) process was applied. It is difficult to apply conventional joining methods such as field welding, brazing, explosion bonding and mechanical one-axial diffusion bonding to a wide area bonding because sufficient mechanical strengths can not be obtained and excessive deformations occurs. In order to solve these fabrication issues, HIP bonding was applied. The first wall stainless steel (SS) coolant tubes of 10 mm in inner diameter and l mm in thickness were sandwiched by semi-circular grooved DSCu plates at the first wall and the front region of the side wall, and by semi-circular grooved SS plates at the back region of the side wall. After assembling of these first wall/side wall parts with the shield block, they were simultaneously bonded by single step HIP in order to

  8. 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.

  9. 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

  10. 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.)

  11. NSS for an RCC in a patient with renal insufficiency after heart transplant because of right ventricular tumor.

    Science.gov (United States)

    Prokopowicz, Grzegorz; Zyczkowski, Marcin; Nowakowski, Krzysztof; Bryniarski, Piotr; Paradysz, Andrzej

    2013-01-01

    The effect of the immunosuppressive therapy on the development of neoplasms has become the object of an ever increasing interest for clinicians all over the world. The literature on neoplasms development in the course of therapy following transplants has confirmed a considerable increase in the incidence of neoplasms of the skin and lymph nodes. Organ neoplasms developing in patients after transplants are characterized by increased progression, poor cellular diversification and a more unfavorable prognosis than in the general population The aim of the study is to present the case of a nephron-sparing surgery of a renal tumor (NSS) without any intraoperative ischaemia in a 55-year-old female patient with an orthotopic heart transplant and renal insufficiency following a prolonged immune suppression. It is estimated that the patients at the highest risk of neoplasm development are those in the first months after transplant, especially heart transplant. They require maximum doses of immunosuppressive drugs. In the case of patients with initial renal insufficiency the duration of ischaemia of the organ operated on should be minimized, and if possible, surgery should be conducted without clamping the renal pedicle. The surgical treatment of RCC (renal cell carcinoma) in transplant patients does not require any reduction in the amount of the immunosuppressive drugs.

  12. Magnetic field shielding system in a tokamak experimental power reactor (EPR): concept and calculations

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Marcus, F.B.; Dory, R.A.; Moore, J.R.

    1975-01-01

    A poloidal magnetic field shielding system is proposed for a tokamak EPR. This coil system minimizes the pulsed poloidal field that intersects the TF (toroidal field) coils and hence reduces the risk of superconductor quenching and structural failure of the coils. Based on an idealized shielding model, we have determined the configurations for the OH (ohmic heating), the S-VF (shield-vertical field), and the T-VF (trimming-vertical field) coils in a typical tokamak EPR. It is found that the pulsed poloidal field strength is greatly reduced in the TF coil region. The overall requirement in stored plasma and vertical field energy is also substantially reduced when compared with conventional EPR designs. Use of this field shielding system is expected to enhance reliability of the superconducting TF coils in a tokamak EPR

  13. Evaluation of ilmenite serpentine concrete and ordinary concrete as nuclear reactor shielding

    International Nuclear Information System (INIS)

    Abulfaraj, W.H.; Kamal, S.M.

    1994-01-01

    The present study involves adapting a formal decision methodology to the selection of alternative nuclear reactor concrete shielding. Multiattribute utility theory is selected to accommodate decision maker's preferences. Multiattribute utility theory (MAU) is here employed to evaluate two appropriate nuclear reactor shielding concretes in terms of effectiveness to determine the optimal choice in order to meet the radiation protection regulations. These concretes are Ordinary concrete (O.C.) and Illmenite Serpentile concrete (I.S.C.). These are normal weight concrete and heavy weight heat resistive concrete, respectively. The effectiveness objective of the nuclear reactor shielding is defined and structured into definite attributes and subattributes to evaluate the best alternative. Factors affecting the decision are dose received by reactor's workers, the material properties as well as cost of concrete shield. A computer program is employed to assist in performing utility analysis. Based upon data, the result shows the superiority of Ordinary concrete over Illmenite Serpentine concrete. (Author)

  14. Shielded room measurements, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Stanton, J.S.

    1949-02-22

    The attenuation of electro-statically and electro-magnetically shielded rooms in the ``E,`` ``R,`` ``I,`` and ``T`` Buildings was measured so that corrective measure could be taken if the attenuation was found to be low. If remedial measures could not be taken, the shortcomings of the rooms would be known. Also, the men making the measurements should oversee construction and correct errors at the time. The work was performed by measuring the attenuation at spot frequencies over the range of from 150 kilocycles to 1280 megacycles with suitable equipment mounted in small rubber-tried trucks. The attenuation was determined by ``before and after`` shielding and/or ``door open and door closed`` measurements after installation of copper shielding. In general, attenuation in the frequency range of approximately 10 to 150 mc. was good and was of the order expected. At frequencies in the range of 150 mc. to 1280 mc., the attenuation curve was more erratic; that is, at certain frequencies a severe loss of attenuation was noted, while at others, the attenuation was very good. This was mainly due to poor or faulty seals around doors and pass windows. These poor seals existed in the ``T,`` ``E,`` and ``I`` Buildings because the doors were fitted improperly and somewhat inferior material was used. By experience from these difficulties, both causes were corrected in the ``R`` Building, which resulted in the improvement of the very high frequency (v.h.f.) range in this building. In some specific cases, however, the results were about the same. For the range of frequencies below approximately 10 mc., the attenuation, in almost all cases, gradually decreased as the frequency decreased and reached a minimum at .3 to 1.0 mc. This loss of attenuation was attributed to multiple grounding caused by moisture in the insulating timbers and will gradually decrease as the wood dries out.

  15. Radiation shielding activities at IDOM

    Energy Technology Data Exchange (ETDEWEB)

    Ordóñez, César Hueso; Gurpegui, Unai Cano; Valiente, Yelko Chento; Poveda, Imanol Zamora, E-mail: cesar.hueso@idom.com [IDOM, Consulting, Engineering and Architecture, S.A.U, Vizcaya (Spain)

    2017-07-01

    When human activities have to be performed under ionising radiation environments the safety of the workers must be guaranteed. Usually three principles are used to accomplish with ALARA (As Low As Reasonably Achievable) requirements: the more distance between the source term and the worker, the better; the less time spent to arrange any task, the better; and, once the previous principles are optimized should the exposure of the workers continues being above the regulatory limits, shielding has to be implemented. Through this paper some different examples of IDOM's shielding design activities are presented. Beginning with the gamma collimators for the Jules Horowitz Reactor, nuclear fuel's behaviour researching facility, where the beam path crosses the reactor's containment walls and is steered up to a gamma detector where the fuel spectrum is analysed and where the beam has to be attenuated several orders of magnitude in a short distance. Later it is shown IDOM’s approach for the shielding of the Emergency Control Management Center of Asociación Nuclear Ascó-Vandellòs-II NPPs, a bunker designed to withstand severe accident conditions and to support the involved staff during 30 days, considering the outside radioactive cloud and the inside source term that filtering units become as they filter the incoming air. And finally, a general approach to this kind of problems is presented, since the study of the source term considering all the possible contributions, passing through the material selection and the thicknesses calculation until the optimization of the materials. (author)

  16. Radiation shielding activities at IDOM

    International Nuclear Information System (INIS)

    Ordóñez, César Hueso; Gurpegui, Unai Cano; Valiente, Yelko Chento; Poveda, Imanol Zamora

    2017-01-01

    When human activities have to be performed under ionising radiation environments the safety of the workers must be guaranteed. Usually three principles are used to accomplish with ALARA (As Low As Reasonably Achievable) requirements: the more distance between the source term and the worker, the better; the less time spent to arrange any task, the better; and, once the previous principles are optimized should the exposure of the workers continues being above the regulatory limits, shielding has to be implemented. Through this paper some different examples of IDOM's shielding design activities are presented. Beginning with the gamma collimators for the Jules Horowitz Reactor, nuclear fuel's behaviour researching facility, where the beam path crosses the reactor's containment walls and is steered up to a gamma detector where the fuel spectrum is analysed and where the beam has to be attenuated several orders of magnitude in a short distance. Later it is shown IDOM’s approach for the shielding of the Emergency Control Management Center of Asociación Nuclear Ascó-Vandellòs-II NPPs, a bunker designed to withstand severe accident conditions and to support the involved staff during 30 days, considering the outside radioactive cloud and the inside source term that filtering units become as they filter the incoming air. And finally, a general approach to this kind of problems is presented, since the study of the source term considering all the possible contributions, passing through the material selection and the thicknesses calculation until the optimization of the materials. (author)

  17. Shielding design of ITER pressure suppression system

    International Nuclear Information System (INIS)

    Yamauchi, Michinori; Sato, Satoshi; Nishitani, Takeo; Kawasaki, Hiromitsu

    2006-01-01

    The duct shield from streaming D-T neutrons has been designed for the ITER pressure suppression system. Streaming calculations are performed with the DUCT-III code for the region from the inlet of the pressure relief line to the rupture disk. Next, the neutron permeation through the shield is studied by Monte Carlo calculations with the MCNP code. It is found that 0.15 m thick iron shield is enough to suppress the permeating component from the outside. In addition, it is suggested that the volume of the shield can be reduced by about 30% if the optimized iron shield structure having localized thickness across intense permeation paths is employed to shield the pressure suppression line. (T.I.)

  18. Irrigoscopy - irrigography method, dosimetry and radiation shielding

    International Nuclear Information System (INIS)

    Zubanov, Z.; Kolarevic, G.

    1999-01-01

    Use of patient's radiation shielding during radiology diagnostic procedures in our country is insufficiently represent, so patients needlessly receive very high entrance skin doses in body areas which are not in direct x-ray beam. During irrigoscopy, patient's radiation shielding is very complex problem, because of the organs position. In the future that problem must be solved. We hope that some of our suggestions about patient's radiation shielding during irrigoscopy, can be a small step in that way. (author)

  19. Cage for shield-type support. Schildausbaugestell

    Energy Technology Data Exchange (ETDEWEB)

    Harryers, W; Blumenthal, G; Irresberger, H

    1981-08-13

    A cage for shield-type support containing a fracture shield supported by a hydraulic stamp and a projecting roof bar was constructed in such a way that no cellular shirt is needed to timber the caved room. The roof bar which is linked at a joint axis at the face-side end of the fracture shield is formed at the face side as a multiply foldable bar. (HGOE).

  20. First Delayed Resection Findings After Irreversible Electroporation (IRE) of Human Localised Renal Cell Carcinoma (RCC) in the IRENE Pilot Phase 2a Trial

    Energy Technology Data Exchange (ETDEWEB)

    Wendler, Johann Jakob, E-mail: johann.wendler@med.ovgu.de [Otto von Guericke University of Magdeburg, Department of Urology, University Hospital (Germany); Ricke, Jens, E-mail: jens.Ricke@med.ovgu.de; Pech, Maciej, E-mail: macej.pech@med.ovgu.de; Fischbach, Frank, E-mail: frank.fischbach@med.ovgu.de; Jürgens, Julian, E-mail: julian.juergens@med.ovgu.de [University of Magdeburg, Department of Radiology (Germany); Siedentopf, Sandra, E-mail: sandra.siedentopf@med.ovgu.de; Roessner, Albert, E-mail: albert.roessner@med.ovgu.de [University of Magdeburg, Institute of Pathology (Germany); Porsch, Markus, E-mail: markus.porsch@med.ovgu.de; Baumunk, Daniel, E-mail: daniel.baumunk@med.ovgu.de; Schostak, Martin, E-mail: martin.schostak@med.ovgu.de [Otto von Guericke University of Magdeburg, Department of Urology, University Hospital (Germany); Köllermann, Jens, E-mail: jens.koellermann@sana.de [Sana Klinikum Offenbach Am Main, Institute of Pathology (Germany); Liehr, Uwe-Bernd, E-mail: uwe-bernd.liehr@med.ovgu.de [Otto von Guericke University of Magdeburg, Department of Urology, University Hospital (Germany)

    2016-02-15

    IntroductionIt is postulated that focal IRE affords complete ablation of soft-tissue tumours while protecting the healthy peritumoral tissue. Therefore, IRE may be an interesting option for minimally invasive, kidney-tissue-sparing, non-thermal ablation of renal tumours.AimWith this current pilot study (“IRENE trial”), we present the first detailed histopathological data of IRE of human RCC followed by delayed tumour resection. The aim of this interim analysis of the first three patients was to investigate the ablation efficiency of percutaneous image-guided focal IRE in RCC, to assess whether a complete ablation of T1a RCC and tissue preservation with the NanoKnife system is possible and to decide whether the ablation parameters need to be altered.MethodsFollowing resection 4 weeks after percutaneous IRE, the success of ablation and detailed histopathological description were used to check the ablation parameters.ResultsThe IRE led to a high degree of damage to the renal tumours (1 central, 2 peripheral; size range 15–17 mm). The postulated homogeneous, isomorphic damage was only partly confirmed. We found a zonal structuring of the ablation zone, negative margins and, enclosed within the ablation zone, very small tumour residues of unclear malignancy.ConclusionAccording to these initial, preliminary study results of the first three renal cases, a new zonal distribution of IRE damage was described and the curative intended, renal saving focal ablation of localised RCC below <3 cm by percutaneous IRE by the NanoKnife system appears to be possible, but needs further, systematic evaluation for this treatment method and treatment protocol.

  1. Local Control Rates of Metastatic Renal Cell Carcinoma (RCC) to Thoracic, Abdominal, and Soft Tissue Lesions Using Stereotactic Body Radiotherapy (SBRT)

    International Nuclear Information System (INIS)

    Altoos, Basel; Amini, Arya; Yacoub, Muthanna; Bourlon, Maria T.; Kessler, Elizabeth E.; Flaig, Thomas W.; Fisher, Christine M.; Kavanagh, Brian D.; Lam, Elaine T.; Karam, Sana D.

    2015-01-01

    We report the radiographic response rate of SBRT compared to conventional fractionated radiotherapy (CF-EBRT) for thoracic, abdominal, skin and soft tissue RCC lesions treated at our institution. Fifty three lesions where included in the study (36 SBRT, 17 CF-EBRT), treated from 2004 to 2014 at our institution. We included patients that had thoracic, skin & soft tissue (SST), and abdominal metastases of histologically confirmed RCC. The most common SBRT fractionation was 50 Gy in 5 fractions. The median time of follow-up was 16 months (range 3–97 months). Median BED was 216.67 (range 66.67–460.0) for SBRT, and 60 (range 46.67–100.83) for CF-EBRT. Median radiographic local control rates at 12, 24, and 36 months were 100, 93.41, and 93.41 % for lesions treated with SBRT versus 62.02, 35.27 and 35.27 % for those treated with CF-EBRT (p < 0.001). Predictive factors for radiographic local control under univariate analysis included BED ≥ 100 Gy (HR, 0.048; 95 % CI, 0.006–0.382; p = 0.005), dose per fraction ≥ 9 Gy (HR, 0.631; 95 % CI, 0.429–0.931; p = 0.021), and gender (HR, 0.254; 95 % CI, 0.066–0.978; p = 0.048). Under multivariate analysis, there were no significant predictors for local control. Toxicity rates were low and equivalent in both groups, with no grade 4 or 5 side effects reported. SBRT is safe and effective for the treatment of RCC metastases to thoracic, abdominal and integumentary soft tissues. Radiographic response rates were greater and more durable using SBRT compared to CF-EBRT. Further prospective trials are needed to evaluate efficacy and safety of SBRT for RCC metastases

  2. French RSE-M and RCC-MR code appendices for flaw analysis: Presentation of the fracture parameters calculation-Part I: General overview

    International Nuclear Information System (INIS)

    Marie, S.; Chapuliot, S.; Kayser, Y.; Lacire, M.H.; Drubay, B.; Barthelet, B.; Le Delliou, P.; Rougier, V.; Naudin, C.; Gilles, P.; Triay, M.

    2007-01-01

    Two French nuclear codes include flaw assessment procedures: the RSE-M code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction rules for mechanical components of FBR nuclear islands and high temperature applications'. An important effort of development of these analytical methods has been made for the last 10 years in the frame of a collaboration between CEA, EDF and AREVA-NP, and in the frame of R and D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress intensity factor K I and the J integral, has been widely developed for industrial configurations. All developments have been integrated in the 2005 edition of RSE-M and in the 2007 edition of RCC-MR. This series of articles is composed of five parts: this first one presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide compendia for specific components: plates (part II), pipes (part III) and elbows (part IV). Finally, part V presents the validation elements of the methods, with details on the process followed for their development and on evaluation of the accuracy of the proposed analytical methods. This first article of the series presents an overview of the calculation of K I and J in these two codes and describes briefly the defect assessment analyses. Specific details in the Appendix A16 of RCC-MR (LBB procedure and creep analyses) are also introduced in this article

  3. Shielded scanning electron microscope for radioactive samples

    International Nuclear Information System (INIS)

    Crouse, R.S.; Parsley, W.B.

    1977-01-01

    A small commercial SEM had been successfully shielded for examining radioactive materials transferred directly from a remote handling facility. Relatively minor mechanical modifications were required to achieve excellent operation. Two inches of steel provide adequate shielding for most samples encountered. However, samples reading 75 rad/hr γ have been examined by adding extra shielding in the form of tungsten sample holders and external lead shadow shields. Some degradation of secondary electron imaging was seen but was adequately compensated for by changing operating conditions

  4. Neutron shielding for a 252 Cf source

    International Nuclear Information System (INIS)

    Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Eduardo Gallego, Alfredo Lorente

    2006-01-01

    To determine the neutron shielding features of water-extended polyester a Monte Carlo study was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through inelastic collisions and absorption reactions. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide production induced by neutron activation must be considered. In this investigation the Monte Carlo method was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a 252 Cf isotopic neutron source. During calculations a detailed model for the 252 Cf and the shield was utilized. To compare the shielding features of water extended polyester, the calculations were also made for the bare 252 Cf in vacuum, air and the shield filled with water. For all cases the calculated neutron spectra was utilized to determine the ambient equivalent neutron dose at four sites around the shielding. In the case of water extended polyester and water shielding the calculations were extended to include the prompt gamma rays produced during neutron interactions, with this information the Kerma in air was calculated at the same locations where the ambient equivalent neutron dose was determined. (Author)

  5. 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.

  6. Radiation shielding application of lead glass

    International Nuclear Information System (INIS)

    Nathuram, R.

    2017-01-01

    Nuclear medicine and radiotherapy centers equipped with high intensity X-ray or teletherapy sources use lead glasses as viewing windows to protect personal from radiation exposure. Lead is the main component of glass which is responsible for shielding against photons. It is therefore essential to check the shielding efficiency before they are put in use. This can be done by studying photon transmission through the lead glasses. The study of photon transmission in shielding materials has been an important subject in medical physics and is potential useful in the development of radiation shielding materials

  7. Radiation dose reduction by water shield

    International Nuclear Information System (INIS)

    Zeb, J.; Arshed, W.; Ahmad, S.S.

    2007-06-01

    This report is an operational manual of shielding software W-Shielder, developed at Health Physics Division (HPD), Pakistan Institute of Nuclear Science and Technology (PINSTECH), Pakistan Atomic Energy Commission. The software estimates shielding thickness for photons having their energy in the range 0.5 to 10 MeV. To compute the shield thickness, self absorption in the source has been neglected and the source has been assumed as a point source. Water is used as a shielding material in this software. The software is helpful in estimating the water thickness for safe handling, storage of gamma emitting radionuclide. (author)

  8. A dose-response relationship for time to bone pain resolution after stereotactic body radiotherapy (SBRT) for renal cell carcinoma (RCC) bony metastases

    Energy Technology Data Exchange (ETDEWEB)

    Jhaveri, Pavan M. [Dept. of Radiology, Section of Radiation Oncology, Baylor College of Medicine, Houston (United States); Teh, Bin S.; Paulino, Arnold C.; Blanco, Angel I.; Butler, E. Brian [Dept. of Radiation Oncology, The Methodist Hospital/The Methodist Hospital Research Inst., Houston (United States)], email: bteh@tmhs.org; Lo, Simon S. [Dept. of Radiation Oncology, Univ. Hospitals Seidman Cancer Center, Case Western Reserve Univ., Cleveland (United States); Amato, Robert J. [Dept. of Internal Medicine, Div. of Oncology, Univ. of Texas Health Sciences Center, Houston (United States)

    2012-05-15

    Background. To investigate the utility of stereotactic body radiotherapy (SBRT) in the treatment of painful renal cell carcinoma (RCC) bone metastases, and for a possible dose effect on time to symptom relief. Material and methods. Eighteen patients with 24 painful osseous lesions from metastatic RCC were treated with SBRT. The most common treatment regimens were 24 Gy in 3 fractions and 40 Gy in 5 fractions. The times from treatment to first reported pain relief and time to symptom recurrence were evaluated. Median follow-up was 38 weeks (1-156 weeks). Results. Seventy-eight percent of all patients had pain relief. Patients treated with a BED > 85 Gy achieved faster and more durable pain relief compared to those treated with a BED < 85 Gy. There was decrease in time to pain relief after a change in treatment regimen to 8 Gy x 5 fractions (BED = 86). There was only one patient with grade 1 skin toxicity. No neurological or other toxicity was observed. Conclusions. SBRT can safely and effectively treat painful RCC bony metastases. There appears to be a relationship between radiation dose and time to stable pain relief.

  9. French RSE-M and RCC-MR code appendices for flaw analysis: Presentation of the fracture parameters calculation-Part V: Elements of validation

    International Nuclear Information System (INIS)

    Marie, S.; Chapuliot, S.; Kayser, Y.; Lacire, M.H.; Drubay, B.; Barthelet, B.; Le Delliou, P.; Rougier, V.; Naudin, C.; Gilles, P.; Triay, M.

    2007-01-01

    French nuclear codes include flaw assessment procedures: the RSE-M Code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction Rules for Mechanical Components of FBR Nuclear Islands and High Temperature Applications'. Development of analytical methods has been made for the last 10 years in the framework of a collaboration between CEA, EDF and AREVA-NP, and by R and D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, in particular the stress intensity factor K I and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of RSE-M and in 2007 edition of RCC-MR. This series of articles consists of 5 parts: the first part presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide the compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). This part presents validation of the methods, with details on the process followed for their development and of the evaluation accuracy of the proposed analytical methods

  10. French RSE-M and RCC-MR code appendices for flaw analysis: Presentation of the fracture parameters calculation-Part IV: Cracked elbows

    International Nuclear Information System (INIS)

    Marie, S.; Chapuliot, S.; Kayser, Y.; Lacire, M.H.; Drubay, B.; Barthelet, B.; Le Delliou, P.; Rougier, V.; Naudin, C.; Gilles, P.; Triay, M.

    2007-01-01

    Two French nuclear codes include flaw assessment procedures: the RSE-M Code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction rules for mechanical components of FBR nuclear islands and high temperature applications'. Development of analytical methods has been made for the last 10 years through a collaboration between CEA, EDF and AREVA-NP, and through R and D actions involving CEA and IRSN. These activities have led to unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress intensity factor K I and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of RSE-M and in 2007 edition of RCC-MR. This series of papers is composed of five parts: the first presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). Part V presents validation of the methods, with details on their accuracy. This paper presents the stress intensity factor and J calculation for cracked elbows. General data applicable for all defect geometries are first presented, and then, compendia for K I and σ ref calculations are provided for the available defect geometries

  11. Decontaminating lead bricks and shielding

    International Nuclear Information System (INIS)

    Lussiez, G.

    1994-01-01

    Lead used for shielding is often surface contaminated with radioisotopes and is therefore a RCRA D008 mixed waste. The technology-based standard for treatment is macroencapsulation. However, decontaminating and recycling the clean lead is a more attractive solution. Los Alamos National Laboratory decontaminates material and equipment contaminated with radioisotopes. Decontaminating lead poses special problems because of the RCRA hazard classification and the size of the inventory, now about 50 tons and likely to grow substantially of planned decommissioning operations. Thus lead, in the form of bricks and other shield shapes, is surface contaminated with fission products. One of the best methods for contaminated lead is removing the superficial layer of contamination with an abrasive medium under pressure. For lead, a mixture of alumina with water and air at about 40 psig rapidly and effectively decontaminates the lead. The abrasive medium is sprayed onto the lead in a scaled-off area. The slurry of abrasive and particles of lead falls through a floor and is collected in a sump. A pump sends the slurry mixture back to the spray gun, creating a continuous process. The process generates small volumes of lead slurry that can be solidified and, because it passes the TCLP, is not a mixed waste. The decontaminated lead can be released for recycling

  12. Decontaminating lead bricks and shielding

    International Nuclear Information System (INIS)

    Lussiez, G.W.

    1993-01-01

    Lead used for shielding is often surface contaminated with radionuclides and is therefore a Resource Conservation and Recovery Act (RCRA) D008 mixed waste. The technology-based standard for treatment is macroencapsulation. However, decontaminating and recycling the clean lead is a more attractive solution. Los Alamos National Lab. decontaminates material and equipment contaminated with radioisotopes. Decontaminating lead poses special problems because of the RCRA hazard classification and the size of the inventory, now about 100 metric tons and likely to grow substantially because of planned decommissioning operations. This lead, in the form of bricks and other shield shapes, is surface contaminated with fission products. One of the best methods for decontaminating lead is removing the thin superficial layer of contamination with an abrasive medium under pressure. For lead, a mixture of alumina with water and air at about 280 kPa (40 psig) rapidly and effectively decontaminates the lead. The abrasive medium is sprayed onto the lead in a sealed-off area. The slurry of abrasive and particles of lead falls through a floor grating and is collected in a pump. A pump sends the slurry mixture back to the spray gun, creating a continuous process

  13. Decontaminating lead bricks and shielding

    International Nuclear Information System (INIS)

    Lussiez, G.W.

    1993-01-01

    Lead used for shielding is often surface contaminated with radioisotopes and is therefore a RCRA D008 mixed waste. The technology-based standard for treatment is macroencapsulation. However, decontaminating and recycling the clean lead is a more attractive solution. Los Alamos National Laboratory decontaminates material and equipment contaminated with radioisotopes. Decontaminating lead poses special problems because of the RCRA hazard classification and the size of the inventory, now about 50 tons and likely to grow substantially because of planned decommissioning operations. This lead, in the form of bricks and other shield shapes, is surface contaminated with fission products. One of the best methods for decontaminating lead is removing the thin superficial layer of contamination with an abrasive medium trader pressure. For lead, a mixture of alumina with water and air at about 40 psig rapidly and effectively decontaminates the lead. The abrasive medium is sprayed onto the lead in a sealed-off area. The slurry of abrasive and particles of lead falls through a floor grating and is collected in a sump. A pump sends the slurry mixture back to the spray gun, creating a continuous process. The process generates small volumes of contaminated lead slurry that can be solidified and, because it passes the TCLP, is not a mixed waste. The decontaminated lead can be released for recycling

  14. TPX remote maintenance and shielding

    International Nuclear Information System (INIS)

    Rennich, M.J.; Nelson, B.E.

    1994-01-01

    The Tokamak Physics Experiment machine design incorporates comprehensive planning for efficient and safe component maintenance. Three programmatic decisions have been made to insure the successful implementation of this objective. First, the tokamak incorporates radiation shielding to reduce activation of components and limit the dose rate to personnel working on the outside of the machine. This allows most of the ex-vessel equipment to be maintained through conventional ''hands-on'' procedures. Second, to the maximum extent possible, low activation materials will be used inside the shielding volume. This resulted in the selection of Titanium (Ti-6Al-4V) for the vacuum vessel and PFC structures. The third decision stipulated that the primary in-vessel components will be replaced or repaired via remote maintenance tools specifically provided for the task. The component designers have been given the responsibility of incorporating maintenance design and for proving the maintainability of the design concepts in full-scale mockup tests prior to the initiation of final fabrication. Remote maintenance of the TPX machine is facilitated by general purpose tools provided by a special purpose design team. Major tools will include an in-vessel transporter, a vessel transfer system and a large component transfer container. In addition, tools such as manipulators and remotely operable impact wrenches will be made available to the component designers by this group. Maintenance systems will also provide the necessary controls for this equipment

  15. Isotope effects on nuclear shielding

    International Nuclear Information System (INIS)

    Hansen, P.E.

    1983-01-01

    This review concentrates upon empirical trends and practical uses of mostly secondary isotope effects, both of the intrinsic and equilibrium types. The text and the tables are arranged in the following fashion. The most 'popular' isotope effect is treated first, deuterium isotope effects on 13 C nuclear shielding, followed by deuterium on 1 H nuclear shieldings, etc. Focus is thus on the isotopes producing the effect rather than on the nuclei suffering the effect. After a brief treatment of each type of isotope effect, general trends are dealt with. Basic trends of intrinsic isotope effects such as additivity, solvent effects, temperature effects, steric effects, substituent effects and hyperconjugation are discussed. Uses of isotope effects for assignment purposes, in stereochemical studies, in hydrogen bonding and in isotopic tracer studies are dealt with. Kinetic studies, especially of phosphates, are frequently performed by utilizing isotope effects. In addition, equilibrium isotope effects are treated in great detail as these are felt to be new and very important and may lead to new uses of isotope effects. Techniques used to obtain isotope effects are briefly surveyed at the end of the chapter. (author)

  16. Continuous electrodeionization through electrostatic shielding

    International Nuclear Information System (INIS)

    Dermentzis, Konstantinos

    2008-01-01

    We report a new continuous electrodeionization cell with electrostatically shielded concentrate compartments or electrochemical Faraday cages formed by porous electronically and ionically conductive media, instead of permselective ion exchange membranes. Due to local elimination of the applied electric field within the compartments, they electrostatically retain the incoming ions and act as 'electrostatic ion pumps' or 'ion traps' and therefore concentrate compartments. The porous media are chemically and thermally stable. Electrodeionization or electrodialysis cells containing such concentrate compartments in place of ion exchange membranes can be used to regenerate ion exchange resins and produce deionized water, to purify industrial effluents and desalinate brackish or seawater. The cells can work by polarity reversal without any negative impact to the deionization process. Because the electronically and ionically active media constituting the electrostatically shielded concentrate compartments are not permselective and coions are not repelled but can be swept by the migrating counterions, the cells are not affected by the known membrane associated limitations, such as concentration polarization or scaling and show an increased current efficiency

  17. Heavy density concrete for nuclear radiation shielding and power stations: [Part]2

    International Nuclear Information System (INIS)

    Singha Roy, P.K.

    1987-01-01

    This article is the second part of the paper entitled 'Heavy density concrete for nuclear radiation shielding and power stations'. In this part, some of the important properties of heavy density concrete are discussed. They include density, water retentivity, air content, permeability with special reference to concrete mixes used in India's nuclear power reactors. All these properties are affected to various extents by heating. Indian shield concrete is rarely subjected to temperatures above 60degC during its life, because of thermal shield protection. During placement, the maximum anticipated rise in temperature due to heat of hydration is restricted to around 45degC by chilling, if necessary to reduce shrinkage stresses and cracks. (M.G.B.)

  18. 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)

  19. Shielding behavior of multi-transformation phase change materials (MTPCM) against nuclear radiations

    International Nuclear Information System (INIS)

    Kumar, Ravindra; Goplani, Deepak; Kumar, Rohitash; Das, Mrinal Kumar; Kumar, Pramod; Jodha, Ajay Singh; Misra, Manoj; Khatri, P.K.

    2008-01-01

    In nuclear hardened structures and AFV's, special shielding materials are being used to provide protection from radiations generated in nuclear blast. However, in blast an intense heat pulse is also generated along with radiation. Currently used shield does not take care of this heat pulse. Defence Laboratory, Jodhpur has developed multi transformation phase change materials (MTPCM) based cool panels for passive moderation of temperature in severe desert heat. The MTPCM contains light nuclei of hydrogen, carbon and oxygen, and thus can absorb good amount of neutrons. MTPCM can also absorb intense heat pulse along with heat generated by secondary fires during blast as its latent heat (160-170 J/g) without significant rise in temperature (melting point 36-38 deg. C). Thus MTPCM can provide protection against both radiation as well as heat pulse generated in a nuclear blast along with its designed regular function of passively moderating temperature below 40 deg C during severe desert summer. A study has been undertaken to explore multiple applications of MTPCM panel. Protection factor provided by standard MTPCM panels against neutron and gamma radiations (both initial and fall out) were measured and results compared with PF provided by special lining pad currently being used in AFV's and field structures for nuclear protection. It is observed that MTPCM provides good PF (2.17) against neutron which is better than currently used shield pads (PFP%1.8). Present paper discusses results of this study. (author)

  20. An overview of Pazopanib in metastatic renal cell carcinoma (mRCC in Ministry of Health (MOH, Malaysia: A multicentre experience

    Directory of Open Access Journals (Sweden)

    May Feng Chen

    2017-12-01

    Full Text Available Background: Pazopanib was listed in MOH, Malaysia’s drug formulary on July 2013 for the treatment of mRCC. Two landmark trials were used to support the listing i.e. VEG105192 and COMPARZ which showed a median progression free survival (PFS of 9.2 and 8.4 months respectively. The efficacy and tolerability of Pazopanib in patients with mRCC have been found to differ in Western and Asian populations. Due to the difference in tolerability, dose adjustment in Asian patients is required. In most MOH hospitals in Malaysia, we tend to start mRCC patients with a lower dose as most patients can’t tolerate 800 mg of Pazopanib. We aim to prognosticate and evaluate PFS in patients with mRCC treated in MOH, Malaysia. We also aim to correlate PFS of patients on different doses of Pazopanib. Methods: The largest referral centre in central and southern Malaysia was selected namely Hospital Kuala Lumpur and Hospital Sultan Ismail. A retrospective analysis was done for patients using hospital supply of Pazopanib from 2010 to 2016. PFS is defined as the time of Pazopanib initiation till 1st documented disease progression. Prognostic variables such as age, sex, race, histology, common sites of metastasis, number of organs involved, ECOG, modified MSKCC risk category, prior history of nephrectomy or systemic treatment and initial dose of Pazopanib was analysed. Results: There was a total of 87 patients with mRCC registered. Median time since diagnosis to initiation of Pazopanib was 5 months. PFS for the overall study population was 8.0 months (95%CI, 5.6–10.4 months. PFS based on dose showed 400 mg vs. 800 mg of Pazopanib was 9.0 months (95%CI 4.5–13.4 months vs. 6.0 months (95%CI 1.9–10.1 months respectively (p = 0.209. Modified MSKCC risk category was a significant prognostic variable (p = 0.045. Conclusion: This retrospective review shows no significant difference in PFS between 400 mg vs. 800 mg of Pazopanib. MOH hospitals in Malaysia have similar PFS

  1. Structural design of DEMO Divertor Cassette Body: provisional FEM analysis and introductive application of RCC-MRx design rules

    Energy Technology Data Exchange (ETDEWEB)

    Frosi, Paolo, E-mail: paolo.frosi@enea.it [Unità Tecnica Fusione-ENEA C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy); Mazzone, Giuseppe [Unità Tecnica Fusione-ENEA C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy); You, Jeong-Ha [Max Planck Institute of Plasma Physics, Boltzmann Str. 2, 85748 Garching (Germany)

    2016-11-01

    This paper deals with the early steps in developing a structural fem model of DEMO Divertor. The study is focused on the thermal and structural analysis of the Cassette Body: a new geometry has been developed for this component: it is foreseen that the plasma facing component (PFC) will be directly placed on the cassette but for the Dome no choice has been adopted yet. For now the model contains only a suitable schematization of the Cassette Body and its objective is to analyze the effect produced by the main loads (electromagnetic loads, coolant pressure, thermal neutron and convective loads) on itself: an available estimate of loads is that one derived from ITER: for a proper translation some assumptions have been made and they are described in the paper. Now it is not a primary purpose to obtain some definitive statements about stresses, displacements, temperatures and so on; the authors want to construct a set of FEM models that will help all the decisions of DEMO Divertor design in its future development. This set is conceived as a tool that shall be improved to account for all the main enhancements that will be found in geometry, in material properties data and in load evaluations. Moreover, the main design variables (loads, material properties, some geometric items, mesh element size) are defined as parameters. This work considers also an introductive approach for future structural verification of the Divertor Cassette Body: so a concern of the Design and Construction Rules for Mechanical Components of Nuclear Installation (RCC-MRx) has been implemented. The FEM code used is Ansys rel. 15.

  2. Preparation of a basic data base for shielding design. 3

    International Nuclear Information System (INIS)

    Takemura, Morio

    1998-03-01

    With use of a standard groupwise shielding design library JSSTDL produced from the latest evaluated nuclear data library JENDL-3.2, experimental analyses for the In-Vessel Fuel Storage (IVS) Experiment and the Intermediate Heat Exchanger (IHX) Experiment were performed. The results were compared with those obtained by the same analysis method and input data using the JSDJ2 library that had been applied consistently to the JASPER experiment analyses. In general, the results obtained with JSSTDL are higher than those with JSDJ2 as were found in analyses in last two years for the Radial Shield Attenuation Experiment and the Special Materials Experiment and also the Axial Shield Experiment. The calculation-to-experiment ratios of the fast neutron flux just behind deep penetration in sodium were obtained first by these IVS and IHX experimental analyses with the JSSTDL library. However, it was confirmed not to be easy to evaluate the accuracy of sodium cross section because of its dependency on how to model the swelled sodium slabs and tanks. The analyses with JSSTDL library were verified by comparison with other analyses with another library based on JENDL-3.2. Compilation of the input data necessary for future reanalyses of important configurations in JASPER experiments, that were selected at the first stage of this study, were continued and new data were added into the computer disk holding previously accumulated data. (author)

  3. Piping structural design for the ITER thermal shield manifold

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Chang Hyun, E-mail: chnoh@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Chung, Wooho, E-mail: whchung@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Nam, Kwanwoo; Kang, Kyoung-O. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Bae, Jing Do; Cha, Jong Kook [Korea Marine Equipment Research Institute, Busan 606-806 (Korea, Republic of); Kim, Kyoung-Kyu [Mecha T& S, Jinju-si 660-843 (Korea, Republic of); Hamlyn-Harris, Craig; Hicks, Robby; Her, Namil; Jun, Chang-Hoon [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2015-10-15

    Highlights: • We finalized piping design of ITER thermal shield manifold for procurement. • Support span is determined by stress and deflection limitation. • SQP, which is design optimization method, is used for the pipe design. • Benchmark analysis is performed to verify the analysis software. • Pipe design is verified by structural analyses. - Abstract: The thermal shield (TS) provides the thermal barrier in the ITER tokamak to minimize heat load transferred by thermal radiation from the hot components to the superconducting magnets operating at 4.2 K. The TS is actively cooled by 80 K pressurized helium gas which flows from the cold valve box to the cooling tubes on the TS panels via manifold piping. This paper describes the manifold piping design and analysis for the ITER thermal shield. First, maximum allowable span for the manifold support is calculated based on the simple beam theory. In order to accommodate the thermal contraction in the manifold feeder, a contraction loop is designed and applied. Sequential Quadratic Programming (SQP) method is used to determine the optimized dimensions of the contraction loop to ensure adequate flexibility of manifold pipe. Global structural behavior of the manifold is investigated when the thermal movement of the redundant (un-cooled) pipe is large.

  4. Thermal and flow considerations for the 80 K shield of the SSC magnet cryostats

    International Nuclear Information System (INIS)

    Abramovich, S.; Yuecel, A.; Demko, J.; Thirumaleshwar, M.

    1994-01-01

    The nominal temperatures in the SSC magnets range between 4.2 K in the superconducting coils and 300 K on the cryostat outer wall. To minimize the 4 K heat load, one thermal shield cooled by liquid and vapor nitrogen flows at 84 K, and another cooled by helium flow at 20 K are incorporated in the cryostat. Tubes attached to the shields serve as conduits for the cryogens. The liquid nitrogen tube in the cryostat is used for shield refrigeration and also for liquid distribution around the SSC rings. The second nitrogen line is used to return the vapor to the helium refrigerators for helium precooling. The nominal LN2 flow from a 4.3 km long cryogenic string (4 sections) to the surface is 64 g/s. The total liquid nitrogen consumption of approximately 5000 g/s will be supplied at one, two or more locations on the surface. The total heat load of the 80 K shield is estimated as 3.2 W/m: about 50% is composed of infrared radiation; the remaining 50% is by heat conduction through supports, vacuum barriers and other thermal connections between the shield and the 300 K outer wall. The required LN2 flow rate depends on the distribution and circulation schemes. The LN2 temperature will in turn vary depending on the flow rate and on the recooling method used. For example, with a massflow of 400 g/s of LN2 the temperature rises from 82 K to 86 K between two compact recoolers 1 km apart. This temperature is higher than desired. The temperature can be reduced by increasing the flow rate of the liquid or by using the continuous recooling scheme. This paper discusses some thermal problems caused by certain mechanical designs of the 80 K shield and the possible improvement by using continuous recooling. The authors present results of the 80 K shield temperature distribution analysis, the 20 K shield heat load augmentation resulting from the increased 80 K shield temperatures, the continuous nitrogen recooling scheme and some flow timing related analysis

  5. Risk of Peripheral Nerve Disease in Military Working Dogs Deployed in Operations Desert Shield/Storm

    Science.gov (United States)

    2003-01-01

    two cohorts where not discussed except for deaths caused by hostile action, gastric dilation volvulus , heat stroke, and death due to other reasons......4. TITLE AND SUBTITLE Risk of Peripheral Nerve Disease in Military Working Dogs Deployed in Operations Desert Shield/Storm 5a. CONTRACT NUMBER 5b

  6. Radiation shielding design for the VISTA space craft

    Energy Technology Data Exchange (ETDEWEB)

    Pahyn, S.; Pahyn, H.M. [Gazi Univ., Teknik Eoitim Fakultesi, Ankara (Turkey)

    2001-07-01

    An innovative concept for the direct utilisation of fusion energy with laser ignited (D,T) capsules for propulsion is presented with the so called VISTA (Vehicle for Interplanetary Space Transport Applications) concept. VISTA's overall geometry is that of a 50 degrees-half-angle cone to avoid massive radioactive shielding. The 50 degrees-half-angle maximizes the jet efficiency, and is determined by selecting the optimum pellet firing position along the axis of the cone with respect to the plane of the magnet coil. The pellet firing position is in the vacuum. By a total fusion power production of 17 500 MW with a repetition rate of 5 Hz and 3 500 MJ per shot, the propulsion power in form of charged particles has been calculated as {approx} 7 000 MW, making {approx} 40 % of the total fusion power. About 60 % of the fusion energy is carried by the leaking neutrons out of the pellet. Most of them (96 %) escape into vacuum without striking the space ship. Only 4 % enter the frozen hydrogen exhaust cone (about 50 gr.). Total peak nuclear heat generation in the coils is calculated as 4.7 mW/cm{sup 3}. The peak neutron heating is 1.9 mW/cm{sup 3} and the peak {gamma}-ray heating density is 2.8 mW/cm{sup 3}. However, volume averaged nuclear heat generation in the coils is much lower. It is calculated as 0.18, 0.48 and 0.66 mW/cm{sup 3} for neutron, {gamma}-ray and total nuclear heating, respectively. Net shielding mass is found as 170 ton, making < 3 % of the vehicle mass. (authors)

  7. Radiation shielding design for the VISTA space craft

    International Nuclear Information System (INIS)

    Pahyn, S.; Pahyn, H.M.

    2001-01-01

    An innovative concept for the direct utilisation of fusion energy with laser ignited (D,T) capsules for propulsion is presented with the so called VISTA (Vehicle for Interplanetary Space Transport Applications) concept. VISTA's overall geometry is that of a 50 degrees-half-angle cone to avoid massive radioactive shielding. The 50 degrees-half-angle maximizes the jet efficiency, and is determined by selecting the optimum pellet firing position along the axis of the cone with respect to the plane of the magnet coil. The pellet firing position is in the vacuum. By a total fusion power production of 17 500 MW with a repetition rate of 5 Hz and 3 500 MJ per shot, the propulsion power in form of charged particles has been calculated as ∼ 7 000 MW, making ∼ 40 % of the total fusion power. About 60 % of the fusion energy is carried by the leaking neutrons out of the pellet. Most of them (96 %) escape into vacuum without striking the space ship. Only 4 % enter the frozen hydrogen exhaust cone (about 50 gr.). Total peak nuclear heat generation in the coils is calculated as 4.7 mW/cm 3 . The peak neutron heating is 1.9 mW/cm 3 and the peak γ-ray heating density is 2.8 mW/cm 3 . However, volume averaged nuclear heat generation in the coils is much lower. It is calculated as 0.18, 0.48 and 0.66 mW/cm 3 for neutron, γ-ray and total nuclear heating, respectively. Net shielding mass is found as 170 ton, making < 3 % of the vehicle mass. (authors)

  8. Improvements in or relating to nuclear shields

    International Nuclear Information System (INIS)

    Hawkins, R.J.; Riley, K.; Powell, C.

    1981-01-01

    A nuclear radiation shield comprises two pieces of steel held together edge to edge by a weld, the depth of which is less than the thickness of either of the edges. As the radiaion shielding effect of the weld will be less than the steel, an insert is bolted or welded over the weld. (U.K.)

  9. 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

  10. Preliminary radiation shielding design for BOOMERANG

    International Nuclear Information System (INIS)

    Donahue, Richard J.

    2002-01-01

    Preliminary radiation shielding specifications are presented here for the 3 GeV BOOMERANG Australian synchrotron light source project. At this time the bulk shield walls for the storage ring and injection system (100 MeV Linac and 3 GeV Booster) are considered for siting purposes

  11. 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

  12. Optimization of multi-layered metallic shield

    International Nuclear Information System (INIS)

    Ben-Dor, G.; Dubinsky, A.; Elperin, T.

    2011-01-01

    Research highlights: → We investigated the problem of optimization of a multi-layered metallic shield. → The maximum ballistic limit velocity is a criterion of optimization. → The sequence of materials and the thicknesses of layers in the shield are varied. → The general problem is reduced to the problem of Geometric Programming. → Analytical solutions are obtained for two- and three-layered shields. - Abstract: We investigate the problem of optimization of multi-layered metallic shield whereby the goal is to determine the sequence of materials and the thicknesses of the layers that provide the maximum ballistic limit velocity of the shield. Optimization is performed under the following constraints: fixed areal density of the shield, the upper bound on the total thickness of the shield and the bounds on the thicknesses of the plates manufactured from every material. The problem is reduced to the problem of Geometric Programming which can be solved numerically using known methods. For the most interesting in practice cases of two-layered and three-layered shields the solution is obtained in the explicit analytical form.

  13. Alignment modification for pencil eye shields

    International Nuclear Information System (INIS)

    Evans, M.D.; Pla, M.; Podgorsak, E.B.

    1989-01-01

    Accurate alignment of pencil beam eye shields to protect the lens of the eye may be made easier by means of a simple modification of existing apparatus. This involves drilling a small hole through the center of the shield to isolate the rayline directed to the lens and fabricating a suitable plug for this hole

  14. Several problems in accelerator shielding study

    International Nuclear Information System (INIS)

    Nakamura, Takashi; Hirayama, Hideo; Ban, Shuichi.

    1980-01-01

    Recently, the utilization of accelerators has increased rapidly, and the increase of accelerating energy and beam intensity is also remarkable. The studies on accelerator shielding have become important, because the amount of radiation emitted from accelerators increased, the regulation of the dose of environmental radiation was tightened, and the cost of constructing shielding rose. As the plans of constructing large accelerators have been made successively, the survey on the present state and the problems of the studies on accelerator shielding was carried out. Accelerators are classified into electron accelerators and proton accelerators in view of the studies on shielding. In order to start the studies on accelerator shielding, first, the preparation of the cross section data is indispensable. The cross sections for generating Bremsstrahlung, photonuclear reactions generating neutrons, generation of neutrons by hadrons, nuclear reaction of neutrons and generation of gamma-ray by hadrons are described. The generation of neutrons and gamma-ray as the problems of thick targets is explained. The shielding problems are complex and diversified, but in this paper, the studies on the shielding, by which basic data are obtainable, are taken up, such as beam damping and side wall shielding. As for residual radioactivity, main nuclides and the difference of residual radioactivity according to substances have been studied. (J.P.N.)

  15. Collection shield for ion separation apparatus

    International Nuclear Information System (INIS)

    Ford, K.L.; Pugh, R.A.

    1981-01-01

    The ion separation electrodes in isotope separation apparatus are provided with removable collection shields to collect neutral particles which would normally pass through the ionization region. A preferred collection shield comprises a u-shaped section for clipping onto the leading edge of an electrode and a pair of flanges projecting substantially perpendicular to the clipping section for collecting neutral particles

  16. Current status of methods for shielding analysis

    International Nuclear Information System (INIS)

    Engle, W.W.

    1980-01-01

    Current methods used in shielding analysis and recent improvements in those methods are discussed. The status of methods development is discussed based on needs cited at the 1977 International Conference on Reactor Shielding. Additional areas where methods development is needed are discussed

  17. Thermal neutron shield and method of manufacture

    Science.gov (United States)

    Brindza, Paul Daniel; Metzger, Bert Clayton

    2013-05-28

    A thermal neutron shield comprising concrete with a high percentage of the element Boron. The concrete is least 54% Boron by weight which maximizes the effectiveness of the shielding against thermal neutrons. The accompanying method discloses the manufacture of Boron loaded concrete which includes enriching the concrete mixture with varying grit sizes of Boron Carbide.

  18. Shielding augmentation of roll-on shield from NAPS to Kaiga-2

    International Nuclear Information System (INIS)

    Pradhan, A.S.; Kumar, A.N.

    2000-01-01

    Extensive radiation field surveys were conducted in NAPS and KAPS reactor buildings as a part of commissioning checks on radiation shielding. During such surveys, dose rate higher than the expected values were noticed in fuelling machine service areas. A movable shield, separating high field area fuelling machine vault and low field area fuelling machine service area, known as roll-on shield was identified as one of the causes of high field in fuelling machine service area along with weaker end-shield. This paper discusses systematic approach adopted in bringing down the dose rates in fuelling machine service area by augmentation of roll-on shield. (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. Radiation Shielding Systems Using Nanotechnology

    Science.gov (United States)

    Chen, Bin (Inventor); McKay, Christoper P. (Inventor)

    2011-01-01

    A system for shielding personnel and/or equipment from radiation particles. In one embodiment, a first substrate is connected to a first array or perpendicularly oriented metal-like fingers, and a second, electrically conducting substrate has an array of carbon nanostructure (CNS) fingers, coated with an electro-active polymer extending toward, but spaced apart from, the first substrate fingers. An electric current and electric charge discharge and dissipation system, connected to the second substrate, receives a current and/or voltage pulse initially generated when the first substrate receives incident radiation. In another embodiment, an array of CNSs is immersed in a first layer of hydrogen-rich polymers and in a second layer of metal-like material. In another embodiment, a one- or two-dimensional assembly of fibers containing CNSs embedded in a metal-like matrix serves as a radiation-protective fabric or body covering.

  2. Radiation shielding and safety design

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Ouk; Gil, C. S.; Cho, Y. S.; Kim, D. H.; Kim, H. I.; Kim, J. W.; Lee, C. W.; Kim, K. Y.; Kim, B. H. [KAERI, Daejeon (Korea, Republic of)

    2011-07-15

    A benchmarking for the test facility, evaluations of the prompt radiation fields, evaluation of the induced activities in the facility, and estimation of the radiological impact on the environment were performed in this study. and the radiation safety analysis report for nuclear licensing was written based on this study. In the benchmark calculation, the neutron spectra was measured in the 20 Mev test facility and the measurements were compared with the computational results to verify the calculation system. In the evaluation of the prompt radiation fields, the shielding design for 100 MeV target rooms, evaluations of the leakage doses from the accidents and skyshine analysis were performed. The evaluation of the induced activities were performed for the coolant, inside air, structural materials, soil and ground-water. At last, the radiation safety analysis report was written based on results from these studies

  3. Hollow-Wall Heat Shield for Fuel Injector Component

    Science.gov (United States)

    Hanson, Russell B. (Inventor)

    2018-01-01

    A fuel injector component includes a body, an elongate void and a plurality of bores. The body has a first surface and a second surface. The elongate void is enclosed by the body and is integrally formed between portions of the body defining the first surface and the second surface. The plurality of bores extends into the second surface to intersect the elongate void. A process for making a fuel injector component includes building an injector component body having a void and a plurality of ports connected to the void using an additive manufacturing process that utilizes a powdered building material, and removing residual powdered building material from void through the plurality of ports.

  4. Shield verification and validation action matrix summary

    International Nuclear Information System (INIS)

    Boman, C.

    1992-02-01

    WSRC-RP-90-26, Certification Plan for Reactor Analysis Computer Codes, describes a series of action items to be completed for certification of reactor analysis computer codes used in Technical Specifications development and for other safety and production support calculations. Validation and verification are integral part of the certification process. This document identifies the work performed and documentation generated to satisfy these action items for the SHIELD, SHLDED, GEDIT, GENPRT, FIPROD, FPCALC, and PROCES modules of the SHIELD system, it is not certification of the complete SHIELD system. Complete certification will follow at a later date. Each action item is discussed with the justification for its completion. Specific details of the work performed are not included in this document but can be found in the references. The validation and verification effort for the SHIELD, SHLDED, GEDIT, GENPRT, FIPROD, FPCALC, and PROCES modules of the SHIELD system computer code is completed

  5. Practical radiation shielding for biomedical research

    International Nuclear Information System (INIS)

    Klein, R.C.; Reginatto, M.; Party, E.; Gershey, E.L.

    1990-01-01

    This paper reports on calculations which exist for estimating shielding required for radioactivity; however, they are often not applicable for the radionuclides and activities common in biomedical research. A variety of commercially available Lucite shields are being marketed to the biomedical community. Their advertisements may lead laboratory workers to expect better radiation protection than these shields can provide or to assume erroneously that very weak beta emitters require extensive shielding. The authors have conducted a series of shielding experiments designed to simulate exposures from the amounts of 32 P, 51 Cr and 125 I typically used in biomedical laboratories. For most routine work, ≥0.64 cm of Lucite covered with various thicknesses of lead will reduce whole-body occupational exposure rates of < 1mR/hr at the point of contact

  6. PC based temporary shielding administrative procedure (TSAP)

    International Nuclear Information System (INIS)

    Olsen, D.E.; Pederson, G.E.; Hamby, P.N.

    1995-01-01

    A completely new Administrative Procedure for temporary shielding was developed for use at Commonwealth Edison's six nuclear stations. This procedure promotes the use of shielding, and addresses industry requirements for the use and control of temporary shielding. The importance of an effective procedure has increased since more temporary shielding is being used as ALARA goals become more ambitious. To help implement the administrative procedure, a personal computer software program was written to incorporate the procedural requirements. This software incorporates the useability of a Windows graphical user interface with extensive help and database features. This combination of a comprehensive administrative procedure and user friendly software promotes the effective use and management of temporary shielding while ensuring that industry requirements are met

  7. A perturbation technique for shield weight minimization

    International Nuclear Information System (INIS)

    Watkins, E.F.; Greenspan, E.

    1993-01-01

    The radiation shield optimization code SWAN (Ref. 1) was originally developed for minimizing the thickness of a shield that will meet a given dose (or another) constraint or for extremizing a performance parameter of interest (e.g., maximizing energy multiplication or minimizing dose) while maintaining the shield volume constraint. The SWAN optimization process proved to be highly effective (e.g., see Refs. 2, 3, and 4). The purpose of this work is to investigate the applicability of the SWAN methodology to problems in which the weight rather than the volume is the relevant shield characteristic. Such problems are encountered in shield design for space nuclear power systems. The investigation is carried out using SWAN with the coupled neutron-photon cross-section library FLUNG (Ref. 5)

  8. Technology development for radiation shielding analysis

    International Nuclear Information System (INIS)

    Ha, Jung Woo; Lee, Jae Kee; Kim, Jong Kyung

    1986-12-01

    Radiation shielding analysis in nuclear engineering fields is an important technology which is needed for the calculation of reactor shielding as well as radiation related safety problems in nuclear facilities. Moreover, the design technology required in high level radioactive waste management and disposal facilities is faced on serious problems with rapidly glowing nuclear industry development, and more advanced technology has to be developed for tomorrow. The main purpose of this study is therefore to build up the self supporting ability of technology development for the radiation shielding analysis in order to achieve successive development of nuclear industry. It is concluded that basic shielding calculations are possible to handle and analyze by using our current technology, but more advanced technology is still needed and has to be learned for the degree of accuracy in two-dimensional shielding calculation. (Author)

  9. PC based temporary shielding administrative procedure (TSAP)

    Energy Technology Data Exchange (ETDEWEB)

    Olsen, D.E.; Pederson, G.E. [Sargent & Lundy, Chicago, IL (United States); Hamby, P.N. [Commonwealth Edison Co., Downers Grove, IL (United States)

    1995-03-01

    A completely new Administrative Procedure for temporary shielding was developed for use at Commonwealth Edison`s six nuclear stations. This procedure promotes the use of shielding, and addresses industry requirements for the use and control of temporary shielding. The importance of an effective procedure has increased since more temporary shielding is being used as ALARA goals become more ambitious. To help implement the administrative procedure, a personal computer software program was written to incorporate the procedural requirements. This software incorporates the useability of a Windows graphical user interface with extensive help and database features. This combination of a comprehensive administrative procedure and user friendly software promotes the effective use and management of temporary shielding while ensuring that industry requirements are met.

  10. Repository Waste Package Transporter Shielding Weight Optimization

    International Nuclear Information System (INIS)

    C.E. Sanders; Shiaw-Der Su

    2005-01-01

    The Yucca Mountain repository requires the use of a waste package (WP) transporter to transport a WP from a process facility on the surface to the subsurface for underground emplacement. The transporter is a part of the waste emplacement transport systems, which includes a primary locomotive at the front end and a secondary locomotive at the rear end. The overall system with a WP on board weights over 350 metric tons (MT). With the shielding mass constituting approximately one-third of the total system weight, shielding optimization for minimal weight will benefit the overall transport system with reduced axle requirements and improved maneuverability. With a high contact dose rate on the WP external surface and minimal personnel shielding afforded by the WP, the transporter provides radiation shielding to workers during waste emplacement and retrieval operations. This paper presents the design approach and optimization method used in achieving a shielding configuration with minimal weight

  11. Experimental simulation and numerical modeling of vapor shield formation and divertor material erosion for ITER typical plasma disruptions

    International Nuclear Information System (INIS)

    Wuerz, H.; Arkhipov, N.I.; Bakhtin, V.P.; Konkashbaev, I.; Landman, I.; Safronov, V.M.; Toporkov, D.A.; Zhitlukhin, A.M.

    1995-01-01

    The high divertor heat load during a tokamak plasma disruption results in sudden evaporation of a thin layer of divertor plate material, which acts as vapor shield and protects the target from further excessive evaporation. Formation and effectiveness of the vapor shield are theoretically modeled and are experimentally analyzed at the 2MK-200 facility under conditions simulating the thermal quench phase of ITER tokamak plasma disruptions. ((orig.))

  12. 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.

  13. The use of nipple shields: A review

    Directory of Open Access Journals (Sweden)

    Selina Chow

    2016-11-01

    Full Text Available A nipple shield is a breastfeeding aid with a nipple-shaped shield that is positioned over the nipple and areola prior to nursing. Nipple shields are usually recommended to mothers with flat nipples or in cases in which there is a failure of the baby to effectively latch onto the breast within the first two days postpartum. The use of nipple shields is a controversial topic in the field of lactation. Its use has been an issue in the clinical literature since some older studies discovered reduced breast milk transfer when using nipple shields, while more recent studies reported successful breastfeeding outcomes. The purpose of this review was to examine the evidence and outcomes with nipple shield use. Methods: A literature search was conducted in Ovid MEDLINE, OLDMEDLINE, EMBASE Classic, EMBASE, Cochrane Central Register of Controlled Trials and CINAHL. The primary endpoint was any breastfeeding outcome following nipple shield use. Secondary endpoints included the reasons for nipple shield use and the average/median length of use. For the analysis, we examined the effect of nipple shield use on physiological responses, premature infants, mothers’ experiences, and health professionals’ experiences. Results: The literature search yielded 261 articles, 14 of which were included in this review. Of these 14 articles, three reported on physiological responses, two reported on premature infants, eight reported on mothers’ experiences, and one reported on health professionals’ experiences. Conclusion: Through examining the use of nipple shields, further insight is provided on the advantages and disadvantages of this practice, thus allowing clinicians and researchers to address improvements on areas that will benefit mothers and infants the most.

  14. Shielding of Sensitive Electronic Devices in Magnetic Nanoparticle Hyperthermia Using Arrays of Coils

    International Nuclear Information System (INIS)

    Spirou, S V; Tsialios, P; Loudos, G

    2015-01-01

    In Magnetic Nanoparticle Hyperthermia (MNH) an externally applied electromagnetic field transfers energy to the magnetic nanoparticles in the body, which in turn convert this energy into heat, thus locally heating the tissue they are located in. This external electromagnetic field is sufficiently strong so as to cause interference and affect sensitive electronic equipment. Standard shielding of magnetic fields involves Faraday cages or coating with high-permeability shielding alloys; however, these techniques cannot be used with optically sensitive devices, such as those employed in Optical Coherence Tomography or radionuclide imaging. In this work we present a method to achieve magnetic shielding using an array of coils. The magnetic field generated by a single coil was calculated using the COMSOL physics simulation toolkit. Software was written in C/C++ to import the single-coil data, and then calculate the positions, number of turns and currents in the shielding coils in order to minimize the magnetic field strength at the desired location. Simulations and calculations have shown that just two shielding coils can reduce the magnetic field by 2-3 orders of magnitude. (paper)

  15. Shielding of Sensitive Electronic Devices in Magnetic Nanoparticle Hyperthermia Using Arrays of Coils

    Science.gov (United States)

    Spirou, S. V.; Tsialios, P.; Loudos, G.

    2015-09-01

    In Magnetic Nanoparticle Hyperthermia (MNH) an externally applied electromagnetic field transfers energy to the magnetic nanoparticles in the body, which in turn convert this energy into heat, thus locally heating the tissue they are located in. This external electromagnetic field is sufficiently strong so as to cause interference and affect sensitive electronic equipment. Standard shielding of magnetic fields involves Faraday cages or coating with high-permeability shielding alloys; however, these techniques cannot be used with optically sensitive devices, such as those employed in Optical Coherence Tomography or radionuclide imaging. In this work we present a method to achieve magnetic shielding using an array of coils. The magnetic field generated by a single coil was calculated using the COMSOL physics simulation toolkit. Software was written in C/C++ to import the single-coil data, and then calculate the positions, number of turns and currents in the shielding coils in order to minimize the magnetic field strength at the desired location. Simulations and calculations have shown that just two shielding coils can reduce the magnetic field by 2-3 orders of magnitude.

  16. Dual polarized, heat spreading rectenna

    Science.gov (United States)

    Epp, Larry W. (Inventor); Khan, Abdur R. (Inventor); Smith, R. Peter (Inventor); Smith, Hugh K. (Inventor)

    1999-01-01

    An aperture coupled patch splits energy from two different polarization components to different locations to spread heat. In addition, there is no physical electrical connection between the slot, patch and circuitry. The circuitry is located under a ground plane which shields against harmonic radiation back to the RF source.

  17. Heat flow and heat generation in greenstone belts

    Science.gov (United States)

    Drury, M. J.

    1986-01-01

    Heat flow has been measured in Precambrian shields in both greenstone belts and crystalline terrains. Values are generally low, reflecting the great age and tectonic stability of the shields; they range typically between 30 and 50 mW/sq m, although extreme values of 18 and 79 mW/sq m have been reported. For large areas of the Earth's surface that are assumed to have been subjected to a common thermotectonic event, plots of heat flow against heat generation appear to be linear, although there may be considerable scatter in the data. The relationship is expressed as: Q = Q sub o + D A sub o in which Q is the observed heat flow, A sub o is the measured heat generation at the surface, Q sub o is the reduced heat flow from the lower crust and mantle, and D, which has the dimension of length, represents a scale depth for the distribution of radiogenic elements. Most authors have not used data from greenstone belts in attempting to define the relationship within shields, considering them unrepresentative and preferring to use data from relatively homogeneous crystalline rocks. A discussion follows.

  18. Foam-Reinforced Polymer Matrix Composite Radiation Shields, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — New and innovative lightweight radiation shielding materials are needed to protect humans in future manned exploration vehicles. Radiation shielding materials are...

  19. Improved Metal-Polymeric Laminate Radiation Shielding, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — In this proposed Phase I program, a multifunctional lightweight radiation shield composite will be developed and fabricated. This structural radiation shielding will...

  20. Shielding concerns at a spallation source

    International Nuclear Information System (INIS)

    Russell, G.J.; Robinson, H.; Legate, G.L.; Woods, R.

    1989-01-01

    Neutrons produced by 800-MeV proton reactions at the Los Alamos Neutron Scattering Center spallation neutron source cause a variety of challenging shielding problems. We identify several characteristics distinctly different from reactor shielding and compute the dose attenuation through an infinite slab/shield composed of iron (100 cm) and borated polyethylene (15 cm). Our calculations show that (for an incident spallation spectrum characteristic of neutrons leaking from a tungsten target at 90/degree/) the dose through the shield is a complex mixture of neutrons and gamma rays. High-energy (> 20 MeV) neutron production from the target is ≅5% of the total, yet causes ≅68% of the dose at the shield surface. Primary low-energy (< 20 MeV) neutrons from the target contribute negligibly (≅0.5%) to the dose at the shield surface yet cause gamma rays, which contribute ≅31% to the total dose at the shield surface. Low-energy neutrons from spallation reactions behave similarly to neutrons with a fission spectrum distribution. 6 refs., 8 figs., 1 tab

  1. 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

  2. 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.

  3. Radiation distribution through serpentine concrete using local materials and its application as a reactor biological shield

    International Nuclear Information System (INIS)

    Kansouh, W.A.

    2012-01-01

    Highlights: ► New serpentine concrete was made and examined as a reactor biological shield. ► Ilmenite–limonite concrete is a better reactor biological shield. ► New serpentine concrete is a better reactor fast neutrons shield than ordinary and hematite–serpentine concretes. ► Serpentine concrete has lower properties as a reactor total gamma rays shields. - Abstract: In the present work attempt has been made to estimate the shielding parameters of the new serpentine concrete (density = 2.4 g/cm 3 ) using local materials on the shielding parameters for two types of heat resistant concretes, namely hematite–serpentine (density = 2.5 g/cm 3 ) and ilmenite–limonite (density = 2.9 g/cm 3 ). Shielding parameters for ordinary concrete (density = 2.3 g/cm 3 ) were also discussed. These parameters were determined experimentally for serpentine concrete and compared with previously published values for other concretes, which had also been obtained using local materials. The leakage spectra of reactor fast neutrons and total gamma photon beams from cylindrical samples of these concrete shields were also investigated using a collimated beam from ET-RR-1 reactor. A neutron–gamma spectrometer was used in order to obtain pulse height spectra of reactor fast neutrons and the total gamma rays leakage through the investigated concrete samples. These spectra were utilized to obtain the energy spectra required in these investigations. Removal cross section Σ R (E n ) and linear attenuation coefficient μ(E g ) for reactor fast neutrons and total gamma rays and their relative coefficients were evaluated and presented. Measured results were compared with those previously measured for other concretes. The results show that ilmenite–limonite concrete is a better reactor biological shield than the other three concretes. Serpentine concrete under investigation is a better reactor fast neutrons shield than ordinary and hematite–serpentine concretes. Serpentine concrete

  4. Scale-4 shipping cask shielding applications

    International Nuclear Information System (INIS)

    Broadhead, B.L.; Parks, C.V.

    1991-01-01

    This paper reports the application of the SCALE-4 shielding sequences SAS1 and SAS4 to the problem set distributed by the Organization for Economic Cooperation and Development (OECD) Working Group on Shielding Assessment of Transportation Packages. In many cases, additional comparison are made with MCNP and QADS solutions to provide a complete cross-check of methods, cross sections, geometry, etc. The results from this effort permit the evaluation of a number of approximations and effects that must be considered in a typical shielding analysis of a transportation cask

  5. Radiation shielding for TFTR DT diagnostics

    International Nuclear Information System (INIS)

    Ku, L.P.; Johnson, D.W.; Liew, S.L.

    1994-01-01

    The authors illustrate the designs of radiation shielding for the TFTR DT diagnostics using the ACX and TVTS systems as specific examples. The main emphasis here is on the radiation transport analyses carried out in support of the designs. Initial results from the DT operation indicate that the diagnostics have been functioning as anticipated and the shielding designs are satisfactory. The experience accumulated in the shielding design for the TFTR DT diagnostics should be useful and applicable to future devices, such as TPX and ITER, where many similar diagnostic systems are expected to be used

  6. 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

  7. 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)

  8. A novel mutation causing nephronophthisis in the Lewis polycystic kidney rat localises to a conserved RCC1 domain in Nek8

    Directory of Open Access Journals (Sweden)

    McCooke John K

    2012-08-01

    Full Text Available Abstract Background Nephronophthisis (NPHP as a cause of cystic kidney disease is the most common genetic cause of progressive renal failure in children and young adults. NPHP is characterized by abnormal and/or loss of function of proteins associated with primary cilia. Previously, we characterized an autosomal recessive phenotype of cystic kidney disease in the Lewis Polycystic Kidney (LPK rat. Results In this study, quantitative trait locus analysis was used to define a ~1.6Mbp region on rat chromosome 10q25 harbouring the lpk mutation. Targeted genome capture and next-generation sequencing of this region identified a non-synonymous mutation R650C in the NIMA (never in mitosis gene a- related kinase 8 ( Nek8 gene. This is a novel Nek8 mutation that occurs within the regulator of chromosome condensation 1 (RCC1-like region of the protein. Specifically, the R650C substitution is located within a G[QRC]LG repeat motif of the predicted seven bladed beta-propeller structure of the RCC1 domain. The rat Nek8 gene is located in a region syntenic to portions of human chromosome 17 and mouse 11. Scanning electron microscopy confirmed abnormally long cilia on LPK kidney epithelial cells, and fluorescence immunohistochemistry for Nek8 protein revealed altered cilia localisation. Conclusions When assessed relative to other Nek8 NPHP mutations, our results indicate the whole propeller structure of the RCC1 domain is important, as the different mutations cause comparable phenotypes. This study establishes the LPK rat as a novel model system for NPHP and further consolidates the link between cystic kidney disease and cilia proteins.

  9. Structural damages prevention of the ITER vacuum vessel and ports by elasto-plastic analysis with regards to RCC-MR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, Jean-Marc, E-mail: jean-marc.martinez@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Jun, Chang Hoon; Portafaix, Christophe; Alekseev, Alexander; Sborchia, Carlo; Choi, Chang-Ho [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Albin, Vincent [SOM Calcul – Groupe ORTEC, 121 ancien Chemin de Cassis – Immeuble Grand Pré, 13009 Marseille (France); Borrelly, Stephane [Sogeti High Tech, RE2, 180 rue René Descartes, Le Millenium – Bat C, 13857 Aix en Provence (France); Cambazar, Magali [Assystem EOS, 117 rue Jacquard, 84120 Pertuis (France); Gaucher, Thomas [SOM Calcul – Groupe ORTEC, 121 ancien Chemin de Cassis – Immeuble Grand Pré, 13009 Marseille (France); Sfarni, Samir; Tailhardat, Olivier [Assystem EOS, 117 rue Jacquard, 84120 Pertuis (France)

    2015-10-15

    Highlights: • ITER vacuum vessel (VV) is a part of the first barrier to confine the plasma. • ITER VV as NPE necessitates a third party organization authorized by the French nuclear regulator to assure design, fabrication, and conformance testing and quality assurance, i.e. ANB. • Several types of damages have to be prevented in order to guarantee the structural integrity with regards to RCC-MR. • It is usual to employ non-linear analysis when the “classical” elastic analysis reaches its limit of linear application. • Several structural analyses were performed with many different global and local models of the whole ITER VV. - Abstract: Several types of damages have to be prevented in order to guarantee the structural integrity of a structure with regards to RCC-MR; the P-type damages which can result from the application to a structure of a steadily and regularly increasing loading or a constant loading and the S-type damages during operational loading conditions which can only result from repeated application of loadings associated to the progressive deformations and fatigue. Following RCC-MR, the S-type damages prevention has to be started only when the structural integrity is guaranteed against P-type damages. The verification of the last one on the ITER vacuum vessel and ports has been performed by limit analysis with elasto-(perfectly)plastic material behavior. It is usual to employ non-linear analysis when the “classical” elastic analysis reaches its limit of linear application. Some elasto-plastic analyses have been performed considering several cyclic loadings to evaluate also more realistic structural margins of the against S-type damages.

  10. French RSE-M and RCC-MR code appendices for flaw analysis: Presentation of the fracture parameters calculation-Part II: Cracked plates

    International Nuclear Information System (INIS)

    Marie, S.; Chapuliot, S.; Kayser, Y.; Lacire, M.H.; Drubay, B.; Barthelet, B.; Le Delliou, P.; Rougier, V.; Naudin, C.; Gilles, P.; Triay, M.

    2007-01-01

    French nuclear codes include flaw assessment procedures: the RSE-M Code 'Rules for In-service Inspection of Nuclear Power Plant Components' and the RCC-MR code 'Design and Construction rules for mechanical components of FBR nuclear islands and high temperature applications'. An important effort of development of these analytical methods has been made for the last 10 years in the frame of a collaboration between CEA, EDF and AREVA-NP, and in the frame of R and D actions involving CEA and IRSN. These activities have led to a unification of the common methods of the two codes. The calculation of fracture mechanics parameters, and in particular the stress intensity factor K I and the J integral, has been widely developed for industrial configurations. All the developments have been integrated in the 2005 edition of RSE-M and in the 2007 edition of RCC-MR. This series of articles is composed of 5 parts: the first part presents an overview of the methods proposed in the RCC-MR and RSE-M codes. Parts II-IV provide compendia for specific components. The geometries are plates (part II), pipes (part III) and elbows (part IV). Finally, part V presents the validation elements of the methods, with details on the process followed for the development and evaluation of the accuracy of the proposed analytical methods. This second article in the series presents all details for the stress intensity factor and J calculations for cracked plates. General data applicable for all defect geometries are first presented, and then, available defect geometries where compendia for K I and σ ref calculation are provided are given

  11. A PHASE II STUDY OF GEMCITABINE AND CAPECITABINE IN PATIENTS WITH ADVANCED RENAL CELL CANCER (RCC): SOUTHWEST ONCOLOGY GROUP STUDY S0312

    Science.gov (United States)

    Van Veldhuizen, Peter J.; Hussey, Michael; Lara, Primo N.; Mack, Philip C.; Gandour-Edwards, Regina; Clark, Joseph I.; Lange, Marianne K.; Crawford, E. David

    2010-01-01

    Background Gemcitabine plus capecitabine has modest efficacy in patients with advanced RCC but has considerable toxicity. We evaluated the efficacy and toxicity of a modified dose-schedule of this doublet in patients with advanced unresectable or metastatic RCC. Methods Chemotherapy-naïve patients were treated with gemcitabine at 900mg/m2 on days 1,8,15 and capecitabine at 625mg/m2 twice daily on days 1 through 21, every 28 days. Eligible patients must have adequate performance status and end-organ function. The primary endpoint was tumor response rate (RR). No further evaluation of this regimen would be pursued if the RR was ≤ 5%. In an exploratory manner using archival specimens, we also evaluated potential markers of prognosis and treatment response including thymidylate synthase (TS) gene polymorphisms and tumor expression of p53, PTEN, pAKT, pmTOR, and ERCC1. Results Of 43 patients registered, 1 was ineligible and 2 were not analyzable. There was 1 confirmed complete response (CR) and three unconfirmed partial responses (PR), for an overall response rate of 10% (95% CI: 3, 24). Nineteen patients (48%) had stable disease (SD). The six-month freedom-from-treatment-failure and overall survival rates were 20% (95% CI: 8, 32) and 75% (95% CI: 62, 88), respectively. Median survival time was 23 months (95% CI: 10, 37). One patient each experienced Grade 4 neutropenia, fatigue, thrombocytopenia and hemolysis with renal failure. The most common Grade 3 toxicities were neutropenia (12 patients), fatigue (5), and leucopenia (4). Patients with a best response of stable disease or better were more likely to have a decrease in expression of PTEN and an increased expression of pmTOR. Conclusions Gemcitabine plus capecitabine at this reduced dose-schedule benefits a small percentage of patients with RCC with an acceptable toxicity profile. The combination of gemcitabine and capecitabine may serve as a base regimen for combination therapy with targeted agents in select RCC

  12. Design and Construction Rules for Mechanical equipments of FBR nuclear islands: RCC-MR. Tome 1, Volume Z: Other technical appendixes

    International Nuclear Information System (INIS)

    1985-06-01

    This book is the 7th of a whole set of 12 which constitute the present edition of the RCC-MR. The technical appendixes of this volume deal with the following characteristics and rules: calculation of screwed assemblies, analysis taking into account creep, welded joint characteristics, elastoplastic analysis of a structure under cyclic loads, elasto-visco-plastic analysis under cyclic loads, calculation rules of revolution shells under external presure and of cylinders under axial compression, design rules of linear supports, calculation rules of convex bottoms under internal pressure [fr

  13. REPOSITORY RADIATION SHIELDING DESIGN GUIDE

    International Nuclear Information System (INIS)

    M. Haas; E.M. Fortsch

    1997-01-01

    The scope of this document includes radiation safety considerations used in the design of facilities for the Yucca Mountain Site Characterization Project (YMP). The purpose of the Repository Radiation Shielding Design Guide is to document the approach used in the radiological design of the Mined Geologic Disposal System (MGDS) surface and subsurface facilities for the protection of workers, the public, and the environment. This document is intended to ensure that a common methodology is used by all groups that may be involved with Radiological Design. This document will also assist in ensuring the long term survivability of the information basis used for radiological safety design and will assist in satisfying the documentation requirements of the licensing body, the Nuclear Regulatory Commission (NRC). This design guide provides referenceable information that is current and maintained under the YMP Quality Assurance (QA) Program. Furthermore, this approach is consistent with maintaining continuity in spite of a changing design environment. This approach also serves to ensure common inter-disciplinary interpretation and application of data

  14. REPOSITORY RADIATION SHIELDING DESIGN GUIDE

    Energy Technology Data Exchange (ETDEWEB)

    M. Haas; E.M. Fortsch

    1997-09-12

    The scope of this document includes radiation safety considerations used in the design of facilities for the Yucca Mountain Site Characterization Project (YMP). The purpose of the Repository Radiation Shielding Design Guide is to document the approach used in the radiological design of the Mined Geologic Disposal System (MGDS) surface and subsurface facilities for the protection of workers, the public, and the environment. This document is intended to ensure that a common methodology is used by all groups that may be involved with Radiological Design. This document will also assist in ensuring the long term survivability of the information basis used for radiological safety design and will assist in satisfying the documentation requirements of the licensing body, the Nuclear Regulatory Commission (NRC). This design guide provides referenceable information that is current and maintained under the YMP Quality Assurance (QA) Program. Furthermore, this approach is consistent with maintaining continuity in spite of a changing design environment. This approach also serves to ensure common inter-disciplinary interpretation and application of data.

  15. Particle Tracing Modeling with SHIELDS

    Science.gov (United States)

    Woodroffe, J. R.; Brito, T. V.; Jordanova, V. K.

    2017-12-01

    The near-Earth inner magnetosphere, where most of the nation's civilian and military space assets operate, is an extremely hazardous region of the space environment which poses major risks to our space infrastructure. Failure of satellite subsystems or even total failure of a spacecraft can arise for a variety of reasons, some of which are related to the space environment: space weather events like single-event-upsets and deep dielectric charging caused by high energy particles, or surface charging caused by low to medium energy particles; other space hazards are collisions with natural or man-made space debris, or intentional hostile acts. A recently funded project through the Los Alamos National Laboratory (LANL) Directed Research and Development (LDRD) program aims at developing a new capability to understand, model, and predict Space Hazards Induced near Earth by Large Dynamic Storms, the SHIELDS framework. The project goals are to understand the dynamics of the surface charging environment (SCE), the hot (keV) electrons on both macro- and microscale. These challenging problems are addressed using a team of world-class experts and state-of-the-art physics-based models and computational facilities. We present first results of a coupled BATS-R-US/RAM-SCB/Particle Tracing Model to evaluate particle fluxes in the inner magnetosphere. We demonstrate that this setup is capable of capturing the earthward particle acceleration process resulting from dipolarization events in the tail region of the magnetosphere.

  16. Shielding Calculations for PUSPATI TRIGA Reactor (RTP) Fuel Transfer Cask with Micro shield

    International Nuclear Information System (INIS)

    Nurhayati Ramli; Ahmad Nabil Abdul Rahim; Ariff Shah Ismail

    2011-01-01

    The shielding calculations for RTP fuel transfer cask was performed by using computer code Micro shield 7.02. Micro shield is a computer code designed to provide a model to be used for shielding calculations. The results of the calculations can be obtained fast but the code is not suitable for complex geometries with a shielding composed of more than one material. Nevertheless, the program is sufficient for As Low As Reasonable Achievable (ALARA) optimization calculations. In this calculation, a geometry based on the conceptual design of RTP fuel transfer cask was modeled. Shielding material used in the calculations were lead (Pb) and stainless steel 304 (SS304). The results obtained from these calculations are discussed in this paper. (author)

  17. Optimal beta-ray shielding thicknesses for different therapeutic radionuclides and shielding materials

    International Nuclear Information System (INIS)

    Cho, Yong In; Kim, Ja Mee; Kim, Jung Hoon

    2017-01-01

    To better understand the distribution of deposited energy of beta and gamma rays according to changes in shielding materials and thicknesses when radionuclides are used for therapeutic nuclear medicine, a simulation was conducted. The results showed that due to the physical characteristics of each therapeutic radionuclide, the thicknesses of shielding materials at which beta-ray shielding takes place varied. Additional analysis of the shielding of gamma ray was conducted for radionuclides that emit both beta and gamma rays simultaneously with results showing shielding effects proportional to the atomic number and density of the shielding materials. Also, analysis of bremsstrahlung emission after beta-ray interactions in the simulation revealed that the occurrence of bremsstrahlung was relatively lower than theoretically calculated and varied depending on different radionuclides. (authors)

  18. Characterisation of superconducting capillaries for magnetic shielding of twisted-wire pairs in a neutron electric dipole moment experiment

    Energy Technology Data Exchange (ETDEWEB)

    Henry, S., E-mail: s.henry@physics.ox.ac.uk; Pipe, M.; Cottle, A.; Clarke, C.; Divakar, U.; Lynch, A.

    2014-11-01

    The cryoEDM neutron electric dipole moment experiment requires a SQUID magnetometry system with pick-up loops inside a magnetically shielded volume connected to SQUID sensors by long (up to 2 m) twisted-wire pairs (TWPs). These wires run outside the main shield, and therefore must run through superconducting capillaries to screen unwanted magnetic pick-up. We show that the average measured transverse magnetic pick-up of a set of lengths of TWPs is equivalent to a loop area of 5.0×10{sup −6} m{sup 2}/m, or 14 twists per metre. From this we set the requirement that the magnetic shielding factor of the superconducting capillaries used in the cryoEDM system must be greater than 8.0×10{sup 4}. The shielding factor—the ratio of the signal picked-up by an unshielded TWP to that induced in a shielded TWP—was measured for a selection of superconducting capillaries made from solder wire. We conclude the transverse shielding factor of a uniform capillary is greater than 10{sup 7}. The measured pick-up was equal to, or less than that due to direct coupling to the SQUID sensor (measured without any TWP attached). We show that discontinuities in the capillaries substantially impair the magnetic shielding, yet if suitably repaired, this can be restored to the shielding factor of an unbroken capillary. We have constructed shielding assemblies for cryoEDM made from lengths of single core and triple core solder capillaries, joined by a shielded Pb cylinder, incorporating a heater to heat the wires above the superconducting transition as required.

  19. Multifunctional BHL Radiation Shield, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Advances in radiation shielding technology remain an important challenge for NASA in order to protect their astronauts, particularly as NASA grows closer to manned...

  20. Radiation shielding method for pipes, etc

    International Nuclear Information System (INIS)

    Nagao, Tetsuya; Takahashi, Shuichi.

    1988-01-01

    Purpose: To constitute shielding walls of a dense structure around pipes and enable to reduce the wall thickness thereof upon periodical inspection, etc. for nuclear power plants. Constitution: For those portions of pipes requring shieldings, cylindrical vessels surrounding the portions are disposed and connected to a mercury supply system, a mercury discharge system and a freezing system for solidifying mercury. After charging mercury in a tank by way of a supply hose to the cylindrical vessels, the temperature of the mercury is lowered below the freezing point thereof to solidify the mercury while circulating cooling medium, to thereby form dense cylindrical radioactive-ray shielding walls. The specific gravity of mercury is greater than that of lead and, accordingly, the thickness of the shielding walls can be reduced as compared with the conventional wall thickness of the entire laminates. (Takahashi, M.)

  1. Shielding design for better plant availability

    International Nuclear Information System (INIS)

    Biro, G.G.

    1975-01-01

    Design methods are described for providing a shield system for nuclear power plants that will facilitate maintenance and inspection, increase overall plant availability, and ensure that man-rem exposures are as low as practicable

  2. Long Duration Space Shelter Shielding, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Physical Sciences Inc. (PSI) has developed fiber reinforced ceramic composites for radiation shielding that can be used for external walls in long duration manned...

  3. Technical specifications for the bulk shielding reactor

    International Nuclear Information System (INIS)

    1986-05-01

    This report provides information concerning the technical specifications for the Bulk Shielding Reactor. Areas covered include: safety limits and limiting safety settings; limiting conditions for operation; surveillance requirements; design features; administrative controls; and monitoring of airborne effluents. 10 refs

  4. Radiation shielding structure for concrete structure

    International Nuclear Information System (INIS)

    Oya, Hiroshi

    1998-01-01

    Crack inducing members for inducing cracks in a predetermined manner are buried in a concrete structure. Namely, a crack-inducing member comprises integrally a shielding plate and extended plates situated at the center of a wall and inducing plates vertically disposed to the boundary portion between them with the inducing plates being disposed each in a direction perforating the wall. There are disposed integrally a pair of the inducing plate spaced at a predetermined horizontal distance on both sides of the shielding plate so as to form a substantially crank-shaped cross section and extended plates formed in the extending direction of the shielding plate, and the inducing plates are disposed each in a direction perforating the wall. Then, cracks generated when stresses are exerted can be controlled, and generation of cracks passing through the concrete structure can be prevented reliably. The reliability of a radiation shielding effect can be enhanced remarkably. (N.H.)

  5. 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.)

  6. Status report of shielding investigation in Japan

    International Nuclear Information System (INIS)

    Shindo, M.

    1964-01-01

    The Japan Atomic Energy Research Institute (JAERI) was established in 1954, and immediately proceeded with the construction of a research reactor. The first symposium in Japan on nuclear energy was held in 1957. Most of the papers presented in the field of reactor shielding were limited to shielding materials and their fabrication. In the first stage of our investigations, our efforts were devoted to practical design studies of reactor shielding. As a result of these studies, it was found that the formulae at hand for calculations were inadequate, but at that time no electronic computer was available in Japan nor were theoretical calculations very actively undertaken. Problems on nuclear ship shielding had been investigated at the Ship Research Institute, since 1956 and many fruitful results had been obtained. About that time the Japan Atomic Industry Forum started activities and took the initiative in organizing shielding research. Research workers in the shipbuilding industry in particular have been seriously studying shielding problems. Few years after the first symposium, problems concerning more fundamental studies were treated by many research workers. Shielding experiments using radioisotopes were carried out and many fruitful results were obtained. They are described in the this paper. Medium size electronic computers became available in Japan, permitting a theoretical study group to make an active contribution. They produced some codes, and their results are also described in the following sections. This constituted the second stage of our investigations. A swimming-pool reactor, JRR-4 (Japan Research Reactor-4), has been under construction at JAERI since 1962 and will become critical in autumn 1964. After characteristic tests it will be a very powerful tool for the shielding investigations. This id the beginning of the third stage of investigations

  7. Nuclear reactor shield including magnesium oxide

    International Nuclear Information System (INIS)

    Rouse, C.A.; Simnad, M.T.

    1981-01-01

    An improvement is described for nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux. The reactor shielding includes means providing structural support, neutron moderator material, neutron absorber material and other components, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron

  8. Shield structure for a nuclear reactor

    International Nuclear Information System (INIS)

    Rouse, C.A.; Simnad, M.T.

    1979-01-01

    An improved nuclear reactor shield structure is described for use where there are significant amounts of fast neutron flux above an energy level of approximately 70 keV. The shield includes structural supports and neutron moderator and absorber systems. A portion at least of the neutron moderator material is magnesium oxide either alone or in combination with other moderator materials such as graphite and iron. (U.K.)

  9. Measurement accuracy in shielded magnetic fields

    International Nuclear Information System (INIS)

    Bottauscio, Oriano; Chiampi, Mario; Crotti, Gabriella; Zucca, Mauro

    2005-01-01

    The measurement error due to both the probe size averaging effect and the coil arrangement is investigated when magnetic field measurements are performed in close proximity to different planar shields. The analysis is carried on through a hybrid FEM/BEM model which employs the 'thin shield' technique. Ferromagnetic, pure conductive and multilayer screens are taken into consideration and an estimation of the errors for concentric and non-concentric coil probes is given. The numerical results are validated by experiments

  10. Influence of Shielding Arrangement on ECT Sensors

    Directory of Open Access Journals (Sweden)

    J. L. Fernandez Marron

    2006-09-01

    Full Text Available This paper presents a full 3D study of a shielded ECT sensor. The spatialresolution and effective sensing field are obtained by means of Finite Element Methodbased simulations and are the compared to a conventional sensor's characteristics. Aneffective improvement was found in the sensitivity in the pipe cross-section, resulting inenhanced quality of the reconstructed image. The sensing field along the axis of the sensoralso presents better behaviour for a shielded sensor.

  11. Acoustic Metacages for Omnidirectional Sound Shielding

    OpenAIRE

    Shen, Chen; Xie, Yangbo; Li, Junfei; Cummer, Steven A.; Jing, Yun

    2017-01-01

    Conventional sound shielding structures typically prevent fluid transport between the exterior and interior. A design of a two-dimensional acoustic metacage with subwavelength thickness which can shield acoustic waves from all directions while allowing steady fluid flow is presented in this paper. The structure is designed based on acoustic gradient-index metasurfaces composed of open channels and shunted Helmholtz resonators. The strong parallel momentum on the metacage surface rejects in-pl...

  12. Method to produce a neutron shielding

    International Nuclear Information System (INIS)

    Merkle, H.J.

    1978-01-01

    The neutron shielding for armoured vehicles consists of preshaped plastic plates which are coated on the armoured vehicle walls by conversion of the thermoplast. Suitable plastics or thermoplasts are PVC, PVC acetate, or mixtures of these, into which more than 50% B, B 4 C, or BN is embedded. The colour of the shielding may be determined by the choice of the neutron absorber, e.g. a white colour for BN. The plates are produced using an extruder or calender. (DG) [de

  13. Shielded transport containers for reactor waste

    International Nuclear Information System (INIS)

    Grundfelt, B.; Eriksson, E.

    The report presents that part of risk analysis which deals with the frequency of breakdowns and the damage on containers. The report focusses on shielded containers made of reinforced concrete. Also a container made of steel is referred to the cases of breakdown are closely allied to collisions with ships. The frequency of breakdowns which might damage the containers is low in all respects, namely 1.10 -5 per year or lower for the shielded container. (G.B.)

  14. Radiation shielding of the main injector

    International Nuclear Information System (INIS)

    Bhat, C.M.; Martin, P.S.

    1995-05-01

    The radiation shielding in the Fermilab Main Injector (FMI) complex has been carried out by adopting a number of prescribed stringent guidelines established by a previous safety analysis. Determination of the required amount of radiation shielding at various locations of the FMI has been done using Monte Carlo computations. A three dimensional ray tracing code as well as a code based upon empirical observations have been employed in certain cases

  15. Polymers Advance Heat Management Materials for Vehicles

    Science.gov (United States)

    2013-01-01

    For 6 years prior to the retirement of the Space Shuttle Program, the shuttles carried an onboard repair kit with a tool for emergency use: two tubes of NOAX, or "good goo," as some people called it. NOAX flew on all 22 flights following the Columbia accident, and was designed to repair damage that occurred on the exterior of the shuttle. Bill McMahon, a structural materials engineer at Marshall Space Flight Center says NASA needed a solution for the widest range of possible damage to the shuttle s exterior thermal protection system. "NASA looked at several options in early 2004 and decided on a sealant. Ultimately, NOAX performed the best and was selected," he says. To prove NOAX would work effectively required hundreds of samples manufactured at Marshall and Johnson, and a concerted effort from various NASA field centers. Johnson Space Center provided programmatic leadership, testing, tools, and crew training; Glenn Research Center provided materials analysis; Langley Research Center provided test support and led an effort to perform large patch repairs; Ames Research Center provided additional testing; and Marshall provided further testing and the site of NOAX manufacturing. Although the sealant never had to be used in an emergency situation, it was tested by astronauts on samples of reinforced carbon-carbon (RCC) during two shuttle missions. (RCC is the thermal material on areas of the shuttle that experience the most heat, such as the nose cone and wing leading edges.) The material handled well on orbit, and tests showed the NOAX patch held up well on RCC.

  16. Potential of Nanocellulose Composite for Electromagnetic Shielding

    Directory of Open Access Journals (Sweden)

    Nabila Yah Nurul Fatihah

    2017-01-01

    Full Text Available Nowadays, most people rely on the electronic devices for work, communicating with friends and family, school and personal enjoyment. As a result, more new equipment or devices operates in higher frequency were rapidly developed to accommodate the consumers need. However, the demand of using wireless technology and higher frequency in new devices also brings the need to shield the unwanted electromagnetic signals from those devices for both proper operation and human health concerns. This paper highlights the potential of nanocellulose for electromagnetic shielding using the organic environmental nanocellulose composite materials. In addition, the theory of electromagnetic shielding and recent development of green and organic material in electromagnetic shielding application has also been reviewed in this paper. The use of the natural fibers which is nanocelllose instead of traditional reinforcement materials provides several advantages including the natural fibers are renewable, abundant and low cost. Furthermore, added with other advantages such as lightweight and high electromagnetic shielding ability, nanocellulose has a great potential as an alternative material for electromagnetic shielding application.

  17. Shielding repair of N.S. Mutsu and related safety features

    International Nuclear Information System (INIS)

    Kishimoto, K.; Miyakoshi, J.

    1978-01-01

    The abnormal radiation level observed on the upper deck of N.S. Mutsu was caused by neutrons streaming through an annular air gap between the reactor pressure vessel and the primary shield. In order to lower this level, a modification of shielding has been planned, for which a shielding mock-up experiment was carried. The foregoing modifications brought some change to the expected behavior of the reactor plant under ship accident situations, and studies were performed to verify plant safety, such as calculations to determine containment vessel integrity and decay heat removal after sinking, and calculations supported by experiment to ascertain the structural strength of the double bottom upon stranding of the ship

  18. Radiation shielding considerations for the repair and maintenance of a swimming pool-type tokamak reactor

    International Nuclear Information System (INIS)

    Seki, Y.; Mori, S.

    1984-01-01

    The radiation shielding relevant to the repair and maintenance of a swimming pool-type tokamak reactor is considered. The dose rate during the reactor operation can be made low enough for personnel access into the reactor room if a 2m thick water layer is installed above the magnet cryostat. The dose rate 24 h after shutdown is such that the human access is allowed above the magnet cryostat. Sufficient water layer thickness is provided in the inboard space for the operation of automatic welder/cutter while retaining the magnet shielding capability. Some forced cooling is required for the decay heat removal in the first wall. The penetration shield thickness around the neutral beam injector port is estimated to be barely sufficient in terms of the magnet radiation damage. (orig.)

  19. 46 CFR 190.20-50 - Heating and cooling.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Heating and cooling. 190.20-50 Section 190.20-50... CONSTRUCTION AND ARRANGEMENT Accomodations for Officers, Crew, and Scientific Personnel § 190.20-50 Heating and... the space. (b) Radiators and other heating apparatus must be so placed and shielded, where necessary...

  20. Double-layer rotor magnetic shield performance analysis in high temperature superconducting synchronous generators under short circuit fault conditions

    Science.gov (United States)

    Hekmati, Arsalan; Aliahmadi, Mehdi

    2016-12-01

    High temperature superconducting, HTS, synchronous machines benefit from a rotor magnetic shield in order to protect superconducting coils against asynchronous magnetic fields. This magnetic shield, however, suffers from exerted Lorentz forces generated in light of induced eddy currents during transient conditions, e.g. stator windings short-circuit fault. In addition, to the exerted electromagnetic forces, eddy current losses and the associated effects on the cryogenic system are the other consequences of shielding HTS coils. This study aims at investigating the Rotor Magnetic Shield, RMS, performance in HTS synchronous generators under stator winding short-circuit fault conditions. The induced eddy currents in different circumferential positions of the rotor magnetic shield along with associated Joule heating losses would be studied using 2-D time-stepping Finite Element Analysis, FEA. The investigation of Lorentz forces exerted on the magnetic shield during transient conditions has also been performed in this paper. The obtained results show that double line-to-ground fault is of the most importance among different types of short-circuit faults. It was revealed that when it comes to the design of the rotor magnetic shields, in addition to the eddy current distribution and the associated ohmic losses, two phase-to-ground fault should be taken into account since the produced electromagnetic forces in the time of fault conditions are more severe during double line-to-ground fault.

  1. Neutronics and shielding issues of ADS

    International Nuclear Information System (INIS)

    Abderrahim, H. A.; Aoust, T.; Haeck, W.; Malambu, E.; Van den Eynde, G.; Gonzalez, E.; Vicente, C.; Martinez-Val, J. M.; Romanets, Y.; Vaz, P.

    2007-01-01

    implementation and deployment have in common the fact that they raise cutting edge scientific and technological problems, associated to the operation of the high-intensity proton accelerator, the high-power (in the multi-MegaWatt range) delivered to the target and the material damage in the target and surrounding structures. The thermal power in the core, the thermal-hydraulic aspects associated to the heat removal in steady state and also in transient mode, the subcriticality level of the system and the efficiency of the transmutation process, is particularly sensitive to the core design (geometry, number of subassemblies, fuel composition, among many other aspects). Neutronic and shielding issues and the computation and mapping of neutron fluxes and doses are important throughout all stages of design of these systems. In this paper, i) the main characteristics and parameters of the ADS systems previously alluded to will be reviewed ii) the neutronics and shielding calculations of relevance for the design of the ADS systems, for radiation damage and for radiation protection purposes will be extensively described

  2. Space Shielding Materials for Prometheus Application

    Energy Technology Data Exchange (ETDEWEB)

    R. Lewis

    2006-01-20

    At the time of Prometheus program restructuring, shield material and design screening efforts had progressed to the point where a down-selection from approximately eighty-eight materials to a set of five ''primary'' materials was in process. The primary materials were beryllium (Be), boron carbide (B{sub 4}C), tungsten (W), lithium hydride (LiH), and water (H{sub 2}O). The primary materials were judged to be sufficient to design a Prometheus shield--excluding structural and insulating materials, that had not been studied in detail. The foremost preconceptual shield concepts included: (1) a Be/B{sub 4}C/W/LiH shield; (2) a Be/B{sub 4}C/W shield; (3) and a Be/B{sub 4}C/H{sub 2}O shield. Since the shield design and materials studies were still preliminary, alternative materials (e.g., {sup nal}B or {sup 10}B metal) were still being screened, but at a low level of effort. Two competing low mass neutron shielding materials are included in the primary materials due to significant materials uncertainties in both. For LiH, irradiation-induced swelling was the key issue, whereas for H{sub 2}O, containment corrosion without active chemistry control was key, Although detailed design studies are required to accurately estimate the mass of shields based on either hydrogenous material, both are expected to be similar in mass, and lower mass than virtually any alternative. Unlike Be, W, and B{sub 4}C, which are not expected to have restrictive temperature limits, shield temperature limits and design accommodations are likely to be needed for either LiH or H{sub 2}O. The NRPCT focused efforts on understanding swelting of LiH, and observed, from approximately fifty prior irradiation tests, that either casting ar thorough out-gassing should reduce swelling. A potential contributor to LiH swelling appears to be LiOH contamination due to exposure to humid air, that can be eliminated by careful processing. To better understand LiH irradiation performance and

  3. Edge modelling of ICFR heated plasmas on PLT

    International Nuclear Information System (INIS)

    Lehrman, I.S.

    1990-01-01

    Theoretical models are presented to explain the edge plasma-antenna interaction that occurs during ICRF heating. The periodic structure of the Faraday shield is found to result in strong ponderomotive force in the vicinity of the antenna. A fluid model, which incorporates the ponderomotive force, predicts an increase in particle transport to the Faraday shield. Kinetic modelling shows that the strong antenna near-fields act to increase the energy of deuterons that strike the shield, thereby increasing the sputtering of shield material. In addition, kinetic modelling shows that E parallel induced between adjacent shield elements acts to heat edge electron that transit close to the antenna. The predictions of the models are shown to be consistent with measurements of enhanced transport on PLT. (author). 27 refs, 17 figs

  4. Determination of ICRF antenna fields in the vicinity of a 3-D Faraday shield structure

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, P M; Rothe, K E; Whealton, J H; Shepard, T D [Oak Ridge National Lab., TN (USA)

    1990-04-01

    A three-dimensional (3-D) magnetostatic analysis developed at Oak Ridge National Laboratory has been used to calculate the electromagnetic transmission properties of representative Faraday shield designs. The analysis uses the long-wavelength approximation to obtain a 3-D Laplace solution for the magnetic scalar potential over one poloidal period of the Faraday shield, from which the complete magnetic field distribution may be obtained. Once the magnetic field distributions in the presence and absence of a Faraday shield are known, the flux transmission coefficient can be found, as well as any change in the distributed inductance of the current strap. The distrbuted capacitance of the strap can be found from an analogous 3-D electrostatic calculation, enabling the phase velocity of the slow-wave structure to be determined. Power dissipation in the shield may be estimated by equating the surface current on a perfect conductor with the surface magnetic field and using this surface current in conjunction with the finite conductivities of the shield materials to obtain the power distribution due to eddy current heating. (orig.).

  5. The lithospheric shear-wave velocity structure of Saudi Arabia: Young volcanism in an old shield

    KAUST Repository

    Tang, Zheng

    2016-05-11

    We investigate the lithospheric shear-wave velocity structure of Saudi Arabia by conducting H-κ stacking analysis and jointly inverting teleseismic P-receiver functions and fundamental-mode Rayleigh wave group velocities at 56 broadband stations deployed by the Saudi Geological Survey (SGS). The study region, the Arabian plate, is traditionally divided into the western Arabian shield and the eastern Arabian platform: The Arabian shield itself is a complicated mélange of crustal material, composed of several Proterozoic terrains separated by ophiolite-bearing suture zones and dotted by outcropping Cenozoic volcanic rocks (locally known as harrats). The Arabian platform is primarily covered by 8 to 10 km of Paleozoic, Mesozoic and Cenozoic sedimentary rocks. Our results reveal high Vp/Vs ratios in the region of Harrat Lunayyir, which are interpreted as solidified magma intrusions from old magmatic episodes in the shield. Our results also indicate slow velocities and large upper mantle lid temperatures below the southern and northern tips of the Arabian shield, when compared with the values obtained for the central shield. We argue that our inferred patterns of lid velocity and temperature are due to heating by thermal conduction from the Afar plume (and, possibly, the Jordan plume), and that volcanism in western Arabia may result from small-scale adiabatic ascent of magma diapirs.

  6. ICRS1, Proceedings of the First Radiation Shielding Symposium, Cambridge, UK 1958

    International Nuclear Information System (INIS)

    Goebelbecker, Hans-Juergen

    2008-01-01

    Description: The papers of the European Atomic Energy Society Symposium VI-58 on radiation shielding (ICRS1) held at Caius College, Cambridge England from 26 to 29 August 1958 are collected here for the first time in electronic form. This symposium was organised in connection with the Second Atoms for Peace Conference held in Geneva Held in Geneva from 1 to 13 September 1958. The Topics discussed covered gamma rays and neutron radiation; the Methods discussed were analytical approaches, semi-empirical Methods, simple computer codes, Monte Carlo method. Little quality nuclear data for shielding calculations was available and the presentations would concentrate on removal cross-sections and build-up factors. Experimental techniques in support to estimate the effective shielding properties of materials were discussed such as general experimental shielding techniques and experiments on neutron attenuation in different materials and on concrete as shield. Foil detectors for spectra measurements and determination of dose rates were mainly used. The typical issues addressed were gamma-heating, gamma spectra, neutron induced gammas, fission products gamma spectra, skyshine radiation and neutron ducts - streaming. Most participants were researchers from the naval and aeronautics sector

  7. Shielding technology for high energy radiation production facility

    International Nuclear Information System (INIS)

    Lee, Byung Chul; Kim, Heon Il

    2004-06-01

    In order to develop shielding technology for high energy radiation production facility, references and data for high energy neutron shielding are searched and collected, and calculations to obtain the characteristics of neutron shield materials are performed. For the evaluation of characteristics of neutron shield material, it is chosen not only general shield materials such as concrete, polyethylene, etc., but also KAERI developed neutron shields of High Density PolyEthylene (HDPE) mixed with boron compound (B 2 O 3 , H 2 BO 3 , Borax). Neutron attenuation coefficients for these materials are obtained for later use in shielding design. The effect of source shape and source angular distribution on the shielding characteristics for several shield materials is examined. This effect can contribute to create shielding concept in case of no detail source information. It is also evaluated the effect of the arrangement of shield materials using current shield materials. With these results, conceptual shielding design for PET cyclotron is performed. The shielding composite using HDPE and concrete is selected to meet the target dose rate outside the composite, and the dose evaluation is performed by configuring the facility room conceptually. From the result, the proper shield configuration for this PET cyclotron is proposed

  8. 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)

  9. Shielding design for positron emission tomography facility

    International Nuclear Information System (INIS)

    Abdallah, I.I.

    2007-01-01

    With the recent advent of readily available tracer isotopes, there has been marked increase in the number of hospital-based and free-standing positron emission tomography (PET) clinics. PET facilities employ relatively large activities of high-energy photon emitting isotopes, which can be dangerous to the health of humans and animals. This coupled with the current dose limits for radiation worker and members of the public can result in shielding requirements. This research contributes to the calculation of the appropriate shielding to keep the level of radiation within an acceptable recommended limit. Two different methods were used including measurements made at selected points of an operating PET facility and computer simulations by using Monte Carlo Transport Code. The measurements mainly concerned the radiation exposure at different points around facility using the survey meter detectors and Thermoluminescent Dosimeters (TLD). Then the set of manual calculation procedures were used to estimate the shielding requirements for a newly built PEF facility. The results from the measurement and the computer simulation were compared to the results obtained from the set manual calculation procedure. In general, the estimated weekly dose at the points of interest is lower than the regulatory limits for the little company of Mary Hospital. Furthermore, the density and the HVL for normal strength concrete and clay bricks are almost similar. In conclusion, PET facilities present somewhat different design requirements and are more likely to require additional radiation shielding. Therefore, existing shields at the little Company of Mary Hospital are in general found to be adequate and satisfactory and additional shielding was found necessary at the new PET facility in the department of Nuclear Medicine of the Dr. George Mukhari Hospital. By use of appropriate design, by implying specific shielding requirements and by maintaining good operating practices, radiation doses to

  10. 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

  11. Production of neutron shielding material

    International Nuclear Information System (INIS)

    Roszler, J.J.

    1979-01-01

    A neutron-absorbing material consisting of a layer of boron carbide sandwiched between layers of aluminum is produced by constructing a rectangular box from aluminum plate leaving one end open. The box is filled with a uniform mixture of finely-divided boron carbide and anodized aluminum powders and the open end is sealed by welding an aluminum plate in place. The box is then heated to 800-850 deg F and rolled to reduce its thickness to the desired amount. The hot rolling bonds or sinters the particles of metal powder or boron carbide. (LL)

  12. Performance of the JT-60 ICRF antenna with an open type Faraday shield

    International Nuclear Information System (INIS)

    Fujii, T.; Saigusa, M.; Kimura, H.; Moriyama, S.; Annoh, K.; Kawano, Y.; Kobayashi, N.; Kubo, H.; Nishitani, T.; Ogawa, Y.; Shinozaki, S.; Terakado, M.

    1992-01-01

    Performance of the JT-60 ICRF antenna in second and third harmonic heating schemes (f=120, 131 MHz) over past four years of operation is presented. The antenna is mainly composed of phased 2x2 loops, an open type Faraday shield and a metallic casing, forming a plug-in type. The antenna is operated for wide plasma parameters: anti n e =1-7x10 19 m -3 , I P =1-2.8 MA and B T =2.2-4.8 T. The open type Faraday shield shows no deterioration for impurity production and heating efficiency up to the maximum injected power of 3.1 MW (the power density of 16 MW/m 2 ) except the following particular condition. Only for (0, 0) phasing and less than 30 mm of the distance between the outermost magnetic surface and the antenna guard limiter, the radiation loss increases abruptly from ΔP rad /P IC ∝0.3 to ΔP rad /P IC ∝4 in carbon limiter discharges when the injected power exceeds a threshold value of ∝0.5 MW. Strong titanium impurity release from the Faraday shield is observed in coincidence with the increase in the radiation loss. This suggests that strong ion sputtering is induced on the Faraday shield by RF sheaths. (orig.)

  13. Final report of Shield System Trade Study. Volume II. WANL support activities for shielding trade study

    International Nuclear Information System (INIS)

    1970-07-01

    Based on the trades made within this study BATH (mixture of B 4 C, aluminum and TiH 1 . 8 ) was selected as the internal shield material. Borated titanium hydride can also meet the criteria with a competitive weight but was rejected because of schedular constraints. A baseline internal shield design was accomplished. This design resulted in a single internal shield weighing about 3300 lb for both manned and unmanned missions. WANL checks on ANSC calculations are generally in agreement, but with some difference in the prediction of the effectiveness of the Boral liner. All of the alternate NSS concepts in the system weight reduction program were rejected. While some did save shield weight, they complicated the NSS design to an unacceptable degree. Studies were made of the feasibility of manual maintenance of NSS components outside of the pressure vessel. The requirements of the NSS components located forward of the internal shield were considered from a thermal and radiation damage standpoint. (auth)

  14. Comparison of eye shields in radiotherapeutic beams

    International Nuclear Information System (INIS)

    Currie, B.E.; Wellington Hospital, Wellington; Johnson, A.D.

    2004-01-01

    Full text: Both MeV electrons and kV photons are used in the treatment of superficial cancers. The advantages and disadvantages for each of these modalities have been widely reported in the literature (See for example [1-2]). Of particular note in the literature is the use of lead and tungsten eye shields to protect ocular structures during radiotherapy. An investigation addressing issues raised in the literature that are relevant to the Wellington Cancer Centre method of treatment of lesions near the eye shall be summarised. Various small sized fields were irradiated to determine depth dose and profile curves in a water phantom shielded by various commercially available eye shields. Transmission factors relevant to critical ocular structures and particle distribution theories are used to further elucidate the comparison between the use of MeV electrons and kV photons in the treatment of superficial cancers. Superficial X-rays from a Pantak Therapax unit SXT 150 model of HVL 4.90mm Al were used for the lead eye shield measurements and electrons from a Varian Clinac 2100C nominal energies 6MeV and 9MeV (R p 3.00cm and 4.34cm respectively) were used for the tungsten eye shield measurements. For the photon measurements circular applicators of 3cm, 4cm and 5cm diameter were used and for the electrons standard 6x6cm and 10x 10cm applicators were used, with no custom inserts. A Scanditronix RFA-300 water phantom and Scanditronix RFAplus version 5.3 software application were used to collect and collate all data. The eye shields were the Radiation Products Design Inc. medium lead eye shield (item 934-014) and the MED-TEC tungsten eye shields MT-T-45 M and MT-T-45 S. It is demonstrated that electron fields have appreciably greater scatter into the area directly under the eye shields than the photon fields. Similarly at the region of d max for the electron fields the relative dose is appreciably greater than the photon fields at similar depth. The relative merits for

  15. Heat production in an Archean crustal profile and implications for heat flow and mobilization of heat-producing elements

    Science.gov (United States)

    Ashwal, L. D.; Morgan, P.; Kelley, S. A.; Percival, J. A.

    1987-01-01

    Concentrations of heat producing elements (Th, U, and K) in 58 samples representative of the main lithologies in a 100-km transect of the Superior Province of the Canadian Shield have been obtained. The relatively large variation in heat production found among the silicic plutonic rocks is shown to correlate with modal abundances of accessory minerals, and these variations are interpreted as premetamorphic. The present data suggest fundamental differences in crustal radioactivity distributions between granitic and more mafic terrains, and indicate that a previously determined apparently linear heat flow-heat production relationship for the Kapuskasing area does not relate to the distribution of heat production with depth.

  16. Hydrogen Induced Cracking of Drip Shield

    Energy Technology Data Exchange (ETDEWEB)

    G. De

    2003-02-24

    One potential failure mechanism for titanium and its alloys under repository conditions is via the absorption of atomic hydrogen in the metal crystal lattice. The resulting decreased ductility and fracture toughness may lead to brittle mechanical fracture called hydrogen-induced cracking (HIC) or hydrogen embrittlement. For the current design of the engineered barrier without backfill, HIC may be a problem since the titanium drip shield can be galvanically coupled to rock bolts (or wire mesh), which may fall onto the drip shield, thereby creating conditions for hydrogen production by electrochemical reaction. The purpose of this scientific analysis and modeling activity is to evaluate whether the drip shield will fail by HIC or not under repository conditions within 10,000 years of emplacement. This Analysis and Model Report (AMR) addresses features, events, and processes related to hydrogen induced cracking of the drip shield. REV 00 of this AMR served as a feed to ''Waste Package Degradation Process Model Report'' and was developed in accordance with the activity section ''Hydrogen Induced Cracking of Drip Shield'' of the development plan entitled ''Analysis and Model Reports to Support Waste Package PMR'' (CRWMS M&O 1999a). This AMR, prepared according to ''Technical Work Plan for: Waste Package Materials Data Analyses and Modeling'' (BSC 2002), is to feed the License Application.

  17. Analytic Ballistic Performance Model of Whipple Shields

    Science.gov (United States)

    Miller, J. E.; Bjorkman, M. D.; Christiansen, E. L.; Ryan, S. J.

    2015-01-01

    The dual-wall, Whipple shield is the shield of choice for lightweight, long-duration flight. The shield uses an initial sacrificial wall to initiate fragmentation and melt an impacting threat that expands over a void before hitting a subsequent shield wall of a critical component. The key parameters to this type of shield are the rear wall and its mass which stops the debris, as well as the minimum shock wave strength generated by the threat particle impact of the sacrificial wall and the amount of room that is available for expansion. Ensuring the shock wave strength is sufficiently high to achieve large scale fragmentation/melt of the threat particle enables the expansion of the threat and reduces the momentum flux of the debris on the rear wall. Three key factors in the shock wave strength achieved are the thickness of the sacrificial wall relative to the characteristic dimension of the impacting particle, the density and material cohesion contrast of the sacrificial wall relative to the threat particle and the impact speed. The mass of the rear wall and the sacrificial wall are desirable to minimize for launch costs making it important to have an understanding of the effects of density contrast and impact speed. An analytic model is developed here, to describe the influence of these three key factors. In addition this paper develops a description of a fourth key parameter related to fragmentation and its role in establishing the onset of projectile expansion.

  18. Shielding methods development in the United States

    International Nuclear Information System (INIS)

    Mynatt, F.R.

    1977-01-01

    A generalized shielding methodology has been developed in the U.S.A. that is adaptable to the shielding analyses of all reactor types. Thus far used primarily for liquid-metal fast breeder reactors, the methodology includes several component activities: (1) developing methods for calculating radiation transport through reactor-shield systems; (2) processing cross-section libraries; (3) performing design calculations for specific systems; (4) performing and analyzing pertinent integral experiments; (5) performing sensitivity studies on both the design calculations and the experimental analyses; and, finally, (6) calculating shield design parameters and their uncertainties. The criteria for the methodology are a 5 to 10 percent accuracy for responses at locations near the core and a factor of 2 accuracy for responses at distant locations. The methodology has been successfully adapted to most in-vessel and ex-vessel problems encountered in the shield analyses of the Fast Flux Test Facility and the Fast Flux Test Facility and the Clinch River Breeder Reactor; however, improved techniques are needed for calculating regions in which radiation streaming is dominant. Areas of the methodology in which significant progress has recently been made are those involving the development of cross-section libraries, sensitivity analysis methods, and transport codes

  19. Development of HANARO ST3 shield

    International Nuclear Information System (INIS)

    Park, K. N.; Lee, J. S.; Shim, H. S.

    2004-12-01

    This report contains the design, fabrication and accurate installation of ST3 shield, which would be installed at ST3 beam port of HANARO. At first, we designed and fabricated ST3 shield casemate composed of 14 blocks. We filled it with heavy concrete, lead ingot and polyethylene that mixed B 4 C powder and epoxy. The average filling density of total shield casemate was 4.7g/cm 3 . The developed ST3 shield was installed at the ST3 beam port and the accuracy of installation for each beam path and channel was evaluated. We found that the extraction of neutron beam to meet the requirement of neutron spectrometer is possible. Also, we developed ancillary equipment such as BGU, quick shutter and exterior shield door for the effective opening and closing of neutron beam. As a result of this study, it was found that neutron spectrometer such as neutron reflectometer and high intensity powder diffractomater can be installed at the ST3 beam port

  20. Cosmic Ray Interactions in Shielding Materials

    International Nuclear Information System (INIS)

    Aguayo Navarrete, Estanislao; Kouzes, Richard T.; Ankney, Austin S.; Orrell, John L.; Berguson, Timothy J.; Troy, Meredith D.

    2011-01-01

    This document provides a detailed study of materials used to shield against the hadronic particles from cosmic ray showers at Earth's surface. This work was motivated by the need for a shield that minimizes activation of the enriched germanium during transport for the MAJORANA collaboration. The materials suitable for cosmic-ray shield design are materials such as lead and iron that will stop the primary protons, and materials like polyethylene, borated polyethylene, concrete and water that will stop the induced neutrons. The interaction of the different cosmic-ray components at ground level (protons, neutrons, muons) with their wide energy range (from kilo-electron volts to giga-electron volts) is a complex calculation. Monte Carlo calculations have proven to be a suitable tool for the simulation of nucleon transport, including hadron interactions and radioactive isotope production. The industry standard Monte Carlo simulation tool, Geant4, was used for this study. The result of this study is the assertion that activation at Earth's surface is a result of the neutronic and protonic components of the cosmic-ray shower. The best material to shield against these cosmic-ray components is iron, which has the best combination of primary shielding and minimal secondary neutron production.