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Sample records for shield mass requirements

  1. Estimating ISABELLE shielding requirements

    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

  2. Shielding requirements for particle bed propulsion systems

    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.

  3. RSMASS: A simple model for estimating reactor and shield masses

    Marshall, A.C.; Aragon, J.; Gallup, D.

    1987-01-01

    A simple mathematical model (RSMASS) has been developed to provide rapid estimates of reactor and shield masses for space-based reactor power systems. Approximations are used rather than correlations or detailed calculations to estimate the reactor fuel mass and the masses of the moderator, structure, reflector, pressure vessel, miscellaneous components, and the reactor shield. The fuel mass is determined either by neutronics limits, thermal/hydraulic limits, or fuel damage limits, whichever yields the largest mass. RSMASS requires the reactor power and energy, 24 reactor parameters, and 20 shield parameters to be specified. This parametric approach should be applicable to a very broad range of reactor types. Reactor and shield masses calculated by RSMASS were found to be in good agreement with the masses obtained from detailed calculations

  4. Nuclear data requirements for fusion reactor shielding

    Abdou, M.A.

    1979-01-01

    The nuclear data requirements for experimental, demonstration and commercial fusion reactors are reviewed. Particular emphasis is given to the shield as well as major reactor components of concern to the nuclear performance. The nuclear data requirements are defined as a result of analyzing four key areas. These are the most likely candidate materials, energy range, types of needed nuclear data, and the required accuracy in the data. Deducing the latter from the target goals for the accuracy in prediction is also discussed. A specific proposal of measurements is recommended. Priorities for acquisition of data are also assigned. (author)

  5. Technical products for radiation shielding. Shield assembled from lead blocks for radiation protection. General technical requirements

    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. Improved Nuclear Reactor and Shield Mass Model for Space Applications

    Robb, Kevin

    2004-01-01

    New technologies are being developed to explore the distant reaches of the solar system. Beyond Mars, solar energy is inadequate to power advanced scientific instruments. One technology that can meet the energy requirements is the space nuclear reactor. The nuclear reactor is used as a heat source for which a heat-to-electricity conversion system is needed. Examples of such conversion systems are the Brayton, Rankine, and Stirling cycles. Since launch cost is proportional to the amount of mass to lift, mass is always a concern in designing spacecraft. Estimations of system masses are an important part in determining the feasibility of a design. I worked under Michael Barrett in the Thermal Energy Conversion Branch of the Power & Electric Propulsion Division. An in-house Closed Cycle Engine Program (CCEP) is used for the design and performance analysis of closed-Brayton-cycle energy conversion systems for space applications. This program also calculates the system mass including the heat source. CCEP uses the subroutine RSMASS, which has been updated to RSMASS-D, to estimate the mass of the reactor. RSMASS was developed in 1986 at Sandia National Laboratories to quickly estimate the mass of multi-megawatt nuclear reactors for space applications. In response to an emphasis for lower power reactors, RSMASS-D was developed in 1997 and is based off of the SP-100 liquid metal cooled reactor. The subroutine calculates the mass of reactor components such as the safety systems, instrumentation and control, radiation shield, structure, reflector, and core. The major improvements in RSMASS-D are that it uses higher fidelity calculations, is easier to use, and automatically optimizes the systems mass. RSMASS-D is accurate within 15% of actual data while RSMASS is only accurate within 50%. My goal this summer was to learn FORTRAN 77 programming language and update the CCEP program with the RSMASS-D model.

  7. Shielded Cells D ampersand D and Dismantlement System Requirements

    Witherspoon, R.L.

    1995-01-01

    This document describes the basis for the development of the System for Highly Radioactive Equipment Dismantlement or SHRED. It is the result of a thorough investigation into current and past dismantlement practices at shielded cell facilities around the DOE complex. This information has been used to formulate the development requirements for the SHRED

  8. Technical Requirements for Fabrication and Installation of Removable Shield for CNRF in HANARO

    Ryu, Jeong Soo; Cho, Yeong Garp; Lee, Jung Hee; Shin, Jin Won

    2008-04-15

    This report details the technical requirements for the fabrication and installation of the removable shield for the Cold Neutron Research Facility (CNRF) in HANARO reactor hall. The removable shield is classified as non-nuclear safety (NNS), seismic category II, and quality class T. The main function of the removable shield is to do the biological shielding of neutrons and gamma from the CN port and the guides. The removable shield consists of block type walls and roofs that can be necessarily assembled, disassembled and moveable. These will be installed between the reactor pool wall and the CNS guide bunker in. This report describes technical requirements for the removable shield such as quality assurance, seismic analysis requirements, configuration, concrete compositions, fabrication and installation requirements, test and inspection, shipping, delivery, etc. Appendix is the technical specification of structural design and analysis. Attachments are composed of the technical specification for the fabrication of the removable shield, shielding design drawings and procurement quality requirements. These technical requirements will be provided to a contract for the manufacturing and installation.

  9. Shielding required for radiation produced by 15 GeV stored electrons

    Jenkins, T.M.; McCaslin, J.B.; Thomas, R.H.

    1974-01-01

    The first phase of PEP will consist of a 15 GeV electron and positron storage ring. This note examines the shielding required by such a facility. Shielding of neutrons and muons produced by 200 GeV protons in the second phase of PEP has been discussed in previous notes. 9 refs., 9 figs., 5 tabs

  10. High ionization radiation field remote visualization device - shielding requirements

    Fernandez, Antonio P. Rodrigues; Omi, Nelson M.; Silveira, Carlos Gaia da; Calvo, Wilson A. Pajero

    2011-01-01

    The high activity sources manipulation hot-cells use special and very thick leaded glass windows. This window provides a single sight of what is being manipulated inside the hot-cell. The use of surveillance cameras would replace the leaded glass window, provide other sights and show more details of the manipulated pieces, using the zoom capacity. Online distant manipulation may be implemented, too. The limitation is their low ionizing radiation resistance. This low resistance also limited the useful time of robots made to explore or even fix problematic nuclear reactor core, industrial gamma irradiators and high radioactive leaks. This work is a part of the development of a high gamma field remote visualization device using commercial surveillance cameras. These cameras are cheap enough to be discarded after the use for some hours of use in an emergency application, some days or some months in routine applications. A radiation shield can be used but it cannot block the camera sight which is the shield weakness. Estimates of the camera and its electronics resistance may be made knowing each component behavior. This knowledge is also used to determine the optical sensor type and the lens material, too. A better approach will be obtained with the commercial cameras working inside a high gamma field, like the one inside of the IPEN Multipurpose Irradiator. The goal of this work is to establish the radiation shielding needed to extend the camera's useful time to hours, days or months, depending on the application needs. (author)

  11. Guideline on radiation protection requirements for ionizing radiation shielding in nuclear power plants

    1988-01-01

    The guideline which entered into force on 1 May 1988 stipulates the radiation protection requirements for shielding against ionizing radiation to be met in the design, construction, commissioning, operation, and decommissioning of nuclear power plants

  12. Shielded chamber control stations: an answer to radiological protection requirements

    Delaboudiniere, Catherine.

    1979-06-01

    This study, carried out at the radiometallurgy laboratory of the Fontenay-aux-Roses Nuclear Research Centre, is based on an ergonomic approach. It shows that the shielded chamber control station, designed for human protection against ionizing radiations, demands an effort of adaptation on the part of the operator. This station does in fact meet radiological protection obligations but would benefit from certain improvements aimed at reducing the secondary fatigue of manipulators already under the strain of adapting to the machine. Fields in which improvements and/or changes are desirable have been investigated with special reference to telemanipulators. New and better designed instruments will soon appear on the telemanipulator market [fr

  13. Method for calculating required shielding in medical x-ray rooms

    Karppinen, J.

    1997-10-01

    The new annual radiation dose limits - 20 mSv (previously 50 mSv) for radiation workers and 1 mSv (previously 5 mSv) for other persons - implies that the adequacy of existing radiation shielding must be re-evaluated. In principle, one could assume that the thicknesses of old radiation shields should be increased by about one or two half-value layers in order to comply with the new dose limits. However, the assumptions made in the earlier shielding calculations are highly conservative; the required shielding was often determined by applying the maximum high-voltage of the x-ray tube for the whole workload. A more realistic calculation shows that increased shielding is typically not necessary if more practical x-ray tube voltages are used in the evaluation. We have developed a PC-based calculation method for calculating the x-ray shielding which is more realistic than the highly conservative method formerly used. The method may be used to evaluate an existing shield for compliance with new regulations. As examples of these calculations, typical x-ray rooms are considered. The lead and concrete thickness requirements as a function of x-ray tube voltage and workload are also given in tables. (author)

  14. Construction of concrete hot cells; requirements for shielding windows for concrete walls with different densities

    1987-10-01

    The shielding windows form part of the basic equipment of hot cells for remote handling, as defined in standard DIN 25 420 part 1. The draft standard in hand is intended to specify the design and manufacture requirements, especially with regard to main dimensions, sight quality, shielding effects, and radiation resistance. The standard refers to three types of shielding window with surface area design (product of density and wall thickness) corresponding to concrete walls of the densities 2.4, 3.4, and 4.0 g/cm 3 . The windows fit to three types of concrete of common usage, and the design is made for Co-60 radiation, with attenuation factors of about 10 4 , 10 6 , or 10 7 . For concrete walls with densities between these data, a shielding window suitable to the next higher density data is to be chosen. (orig./HP) [de

  15. Shielding requirements for the Space Station habitability modules

    Avans, Sherman L.; Horn, Jennifer R.; Williamsen, Joel E.

    1990-01-01

    The design, analysis, development, and tests of the total meteoroid/debris protection system for the Space Station Freedom habitability modules, such as the habitation module, the laboratory module, and the node structures, are described. Design requirements are discussed along with development efforts, including a combination of hypervelocity testing and analyses. Computer hydrocode analysis of hypervelocity impact phenomena associated with Space Station habitability structures is covered and the use of optimization techniques, engineering models, and parametric analyses is assessed. Explosive rail gun development efforts and protective capability and damage tolerance of multilayer insulation due to meteoroid/debris impact are considered. It is concluded that anticipated changes in the debris environment definition and requirements will require rescoping the tests and analysis required to develop a protection system.

  16. Muon shield requirements for ISABELLE at 400 GeV/c

    Ludlam, T.; Thorndike, A.M.

    1977-01-01

    A goodly portion of the ISABELLE ring lies above the existing contours of the land, and so a substantial earth berm is required to shield against penetrating muons which result from proton interactions within the ring. The size and shape of this shield is determined not only by the magnitude of expected proton losses from the circulating beams, but also by the geometry and magnetic structure of the machine, and the proximity of potential muon sources to the site boundary. The cost of constructing this berm is sufficiently great as to warrant detailed attention to the required shield thickness at each point around the ring. The report given updates previous discussions of the subject by incorporating the six-fold geometry and higher energy of the 400 GeV ISABELLE design, and taking advantage of a more refined study of the effects of magnetic deflection on the trajectories of muons produced within the lattice structure of the machine

  17. Shielding practice

    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

  18. Gamma-ray mass attenuation coefficient and half value layer factor of some oxide glass shielding materials

    Waly, El-Sayed A.; Fusco, Michael A.; Bourham, Mohamed A.

    2016-01-01

    The variation in dosimetric parameters such as mass attenuation coefficient, half value layer factor, exposure buildup factor, and the photon mean free path for different oxide glasses for the incident gamma energy range 0.015–15 MeV has been studied using MicroShield code. It has been inferred that the addition of PbO and Bi 2 O 3 improves the gamma ray shielding properties. Thus, the effect of chemical composition on these parameters is investigated in the form of six different glass compositions, which are compared with specialty concrete for nuclear radiation shielding. The composition termed ‘Glass 6’ in this paper has the highest mass attenuation and the smallest half value layer and may have potential applications in radiation shielding. An example dry storage cask utilizing an additional layer of Glass 6 as an intermediate shielding layer, simulated in MicroShield, is capable of reducing the exposure rate at the cask surface by over 20 orders of magnitude compared to the case without a glass layer. Based on this study, Glass 6 shows promise as a gamma-ray shielding material, particularly for dry cask storage.

  19. Comparison of radiation shielding requirements for HDR brachytherapy using 169Yb and 192Ir sources

    Lymperopoulou, G.; Papagiannis, P.; Sakelliou, L.; Georgiou, E.; Hourdakis, C. J.; Baltas, D.

    2006-01-01

    169 Yb has received a renewed focus lately as an alternative to 192 Ir sources for high dose rate (HDR) brachytherapy. Following the results of a recent work by our group which proved 169 Yb to be a good candidate for HDR prostate brachytherapy, this work seeks to quantify the radiation shielding requirements for 169 Yb HDR brachytherapy applications in comparison to the corresponding requirements for the current 192 Ir HDR brachytherapy standard. Monte Carlo simulation (MC) is used to obtain 169 Yb and 192 Ir broad beam transmission data through lead and concrete. Results are fitted to an analytical equation which can be used to readily calculate the barrier thickness required to achieve a given dose rate reduction. Shielding requirements for a HDR brachytherapy treatment room facility are presented as a function of distance, occupancy, dose limit, and facility workload, using analytical calculations for both 169 Yb and 192 Ir HDR sources. The barrier thickness required for 169 Yb is lower than that for 192 Ir by a factor of 4-5 for lead and 1.5-2 for concrete. Regarding 169 Yb HDR brachytherapy applications, the lead shielding requirements do not exceed 15 mm, even in highly conservative case scenarios. This allows for the construction of a lead door in most cases, thus avoiding the construction of a space consuming, specially designed maze. The effects of source structure, attenuation by the patient, and scatter conditions within an actual treatment room on the above-noted findings are also discussed using corresponding MC simulation results

  20. Calculating additional shielding requirements in diagnostics X-ray departments by computer

    Rahimi, A.

    2004-01-01

    This report provides an extension of an existing method for the calculation of the barrier thickness required to reduce the three types of radiation exposure emitted from the source, the primary, secondary and leakage radiation, to a specified weekly design limit (MPD). Since each of these three types of radiation are of different beam quality, having different shielding requirements, NCRP 49 has provided means to calculate the necessary protective barrier thickness for each type of radiation individually. Additionally, barrier requirements specified using the techniques stated at NCRP 49, show enormous variations among users. Part of the variations is due to different assumptions made regarding the use of the examined room and the characteristics of adjoining space. Many of the differences result from the difficulty of accurately relating information from the calculations to graphs and tables involved in the calculation process specified by this report. Moreover, the latest technological developments such as mammography are not addressed and attenuation data for three-phase generators, that are most widely used today, is not provided. The design of shielding barriers in diagnostic X-ray departments generally follow the ALARA principle. That means that, in practice, the exposure levels are kept 'as low as reasonably achievable', taking into account economical and technical factors. Additionally, the calculation of barrier requirements includes many uncertainties (e.g. the workload, the actual kVp used etc.). (author)

  1. Shield requirement estimation for pin storage room in fuel fabrication plant

    Shanthi, M.M.; Keshavamurthy, R.S.; Sivashankaran, G.

    2012-01-01

    Fast Reactor Fuel Cycle Facility (FRFCF) is an upcoming project in Kalpakkam. It has the facility to recycle the fuel from PFBR. It is an integrated facility, consists of fuel reprocessing plant, fuel fabrication plant (FFP), core subassembly plant, Reprocessed Uranium plant (RUP) and waste management plant. The spent fuel from PFBR would be reprocessed in fuel reprocessing plant. The reprocessed fuel material would be sent to fuel fabrication plant. The main activity of fuel fabrication plant is the production of MOX fuel pins. The fuel fabrication plant has a fuel pin storage room. The shield requirement for the pin storage room has been estimated by Monte Carlo method. (author)

  2. Shielding requirements for the transport of nuclear warhead components under decommissioning

    Hansen, L.F.

    1994-09-01

    The requirements to carry out accurate shielding calculations involved with the safe off-site transportation of packages containing nuclear warhead components, special assemblies and radioactive materials are discussed. The need for (a) detailed information on the geometry and material composition of the packaging and radioactive load, (b) accurate representation of the differential energy spectra (dN/dE) for the neutron and gamma spectra emitted by the radioactive materials enclosed in the packaging, (c) well-tested neutron and photon cross section libraries, (d) and accurate three-dimensional Monte Carlo transport codes are illustrated. A brief discussion of the need for reliable dose measurements is presented

  3. Improved performance of a shielded torch using ethanol in inductively coupled plasma–sector field mass spectrometry

    Chen, Tao [State Key Laboratory of Geological Processes and Mineral Resources, Faculty of Earth Sciences, China University of Geosciences, Wuhan 430074 (China); Hu, Zhaochu, E-mail: zchu@vip.sina.com [State Key Laboratory of Geological Processes and Mineral Resources, Faculty of Earth Sciences, China University of Geosciences, Wuhan 430074 (China); Liu, Shenghua [Institute of Hydrogeology and Environmental Geology, CAGS, Shijiazhuang 050061 (China); Liu, Yongsheng; Gao, Shan; Li, Ming; Zong, Keqing; Chen, Haihong; Hu, Shenghong [State Key Laboratory of Geological Processes and Mineral Resources, Faculty of Earth Sciences, China University of Geosciences, Wuhan 430074 (China)

    2015-04-01

    To improve the accuracy and precision of trace element analysis, higher analytical sensitivity and lower interference are required. In this study, we investigated the effects of the addition of ethanol in combination with a shielded torch on the signal intensities of elements from {sup 7}Li to {sup 238}U, oxide yields, doubly charged ion yields, isobaric interferences and ion distributions in inductively coupled plasma–sector field mass spectrometry (ICP–SFMS). For the 39 investigated elements in this study, using the shielded torch increases the sensitivity by a factor of 17–58. The well-known drawback of using a shielded torch is that it will increase the oxide yield. In this study, the CeO{sup +}/Ce{sup +} ratio is increased by a factor of 3.3 in the GE-on mode compared to that in the GE-off mode at normal conditions (without ethanol). In the GE-on mode, the addition of 4% ethanol in ICP–SFMS is found not only to decrease the CeO{sup +}/Ce{sup +} ratio by a factor of 4 but also to suppress the Ce{sup 2+}/Ce{sup +} ratio by a factor of 4.2. In large contrast, the effect of 2–6% ethanol on the oxide yield and doubly charged ratio is minimal in the GE-off mode. Except for As, Se, Sb, Te and Au, for which the signal intensities are increased by a factor of 1.4–3.7 in the presence of 2–6% ethanol, an increased concentration of ethanol suppresses intensities of other elements. In the GE-off mode, the suppression of the analyte signal due to increased ethanol concentration is more significant than that in the GE-on mode. Compared to the spatial profiles of the ion distributions in the normal mode (without ethanol), the addition of 2–4% ethanol leads to significantly wider axial and radial profiles. The significantly wider axial and radial ion distribution had a dilution effect on the ion densities, which subsequently reduced the ion signal intensities. The addition of 4% ethanol was also found to suppress the interferences of Xe by a factor of 2.8 and

  4. Requirements for timber and cadmium used in shielding for fissile material transport packaging

    1982-02-01

    This Code of Practice has been prepared as a guide for designers who require packaging for fissile materials. It should be noted that this document covers design requirements only and it is not a manufacturing specification which can be quoted on a manufacturing contract without qualification. Compliance with the regulations regarding the safe transport of fissile materials may be achieved by the provision of an effective shield embodying:- (a) a moderating material -usually one rich in hydrogen, such as wood - in order to thermalise incoming neutrons, and (b) a material - such as cadmium - with a large absorption cross-section for thermal neutrons, located between the moderator and the fissile material, in order to capture the incoming neutrons. This Code describes the requirements in two sections, one for each of these materials. (author)

  5. SHIELD: The Star Formation Law in Extremely Low-mass Galaxies

    Teich, Yaron; McNichols, Andrew; Cannon, John M.; SHIELD Team

    2016-01-01

    The "Survey of HI in Extremely Low-mass Dwarfs" (SHIELD) is a multiwavelength, legacy-class observational study of 12 low-mass dwarf galaxies discovered in Arecibo Legacy Fast ALFA (ALFALFA) survey data products. Here we analyze the relationships between HI and star formation in these systems using multi-configuration, high spatial (~300 pc) and spectral (0.82 - 2.46 km s-1 ch-1) resolution HI observations from the Karl G. Jansky Very Large Array, Hα imaging from the WIYN 3.5m telescope, and archival GALEX far-ultraviolet imaging. We compare the locations and intensities of star formation with the properties of the neutral ISM. We quantify the degree of local co-spatiality between star forming regions and regions of high HI column densities using the Kennicutt-Schmidt (K-S) relation. The values of the K-S index N vary considerably from system to system; because no single galaxy is representative of the sample, we instead focus on the narratives of the individual galaxies and their complex distribution of gaseous and stellar components. At the extremely faint end of the HI mass function, these systems are dominated by stochastic fluctuations in their interstellar media, which governs whether or not they show signs of recent star formation.Support for this work was provided by NSF grant AST-1211683 to JMC at Macalester College.

  6. Requirement for radiation shields of transportation pipe for on line inhalation gases from compact cyclotron in positron emission tomography

    Hachiya, Takenori; Hagami, Eiichi; Shoji, Yasuaki; Aizawa, Yasuo; Kanno, Iwao; Uemura, Kazuo; Handa, Masahiko; Mori, Junichi; Fukagawa, Akihisa.

    1989-01-01

    In the unit housing of a compact cyclotron and positron emission CT (PET), positron emitting gas such as 15 O, 11 C, C 15 O 2 , C 15 O etc. is supplied from a cyclotron to a PET room through a transportation pipe with an appropriate shield to reduce positron annihilation radiation. This paper discribes the effect of lead and concrete shields with various thickness. Using lead or concrete shield blocks with various thicknesses, radiation leakage through the shield was measured by an ionization chamber type survey meter during continuous and constant supply of 15 O gas of 1.85 GBq/min concentration which is the maximum dose for clinical use. The leakage radiation measured was 213.7, 56.0, 15.3, 5.0 μSv/week for lead shield with 1, 2, 3, and 4 cm thickness, respectively, and 193.3, 30.5 and 5.1 μSv/week for concrete shields with thickness of 10, 20, and 30 cm, respectively. The present study shows that to keep less than 300 μSv/week, which is the permissible dose rate of the boundary zone around the radiation controlled area by Japan Science and Technology Agency, it is required to use more than 8 mm thick lead shield or 7 cm thick concrete for continuous supply of 1.85 GBq/min 15 O gas. (author)

  7. Online mass storage system detailed requirements document

    1976-01-01

    The requirements for an online high density magnetic tape data storage system that can be implemented in a multipurpose, multihost environment is set forth. The objective of the mass storage system is to provide a facility for the compact storage of large quantities of data and to make this data accessible to computer systems with minimum operator handling. The results of a market survey and analysis of candidate vendor who presently market high density tape data storage systems are included.

  8. Radiation shielding lead shield

    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.)

  9. If additional shielding required for the linear accelerator room when modern treatment techniques are intensively used

    Miller, Albert V.; Atkocius, Vydmantas; Aleknavicius, Eduardas

    2001-01-01

    are different and should be considered separately. Additionally, in theory, the workloads for all energies in use have to be calculated. In practice, however, the shielding calculation needs to be performed only for the highest energy used for the particular technique. In our case it was 25 MV X-rays. Results - The primary barrier thickness is calculated with the use of direct workload value. Direct workload in our case is the sum of conventional workload, conformal workload and TBI workload. For the secondary barrier thickness calculation the leakage and scatter workloads are used. Leakage workload is determined as total number of monitor units accumulated per week. For the conventional treatment leakage workload the factor of 1cGy/MU is used to convert from direct workload. For conformal technique, however, additional factor of approximately 2 is required. This factor is about 10 for our TBI technique. Scatter workload we found less or equal to the direct workload. Further, we assumed about 40% of patients receive conformal therapy on the accelerator and in average one patient per week receive TBI. The leakage workload in this situation will be 1.6 of the direct workload. It will increase the secondary barrier thickness by approximately 0.18 of tenth value layer. Conclusion - When intensive use of conformal therapy and TBI is planned on linear accelerator one should consider increasing secondary barrier thickness for the treatment room. The calculation procedure is quite simple and does not require any additional data or measurements. (author)

  10. Study of radiation shielding requirements for n-MOS devices on the Exosat spacecraft. Final report

    1977-01-01

    The device-degradation and radiation-shielding problems presented by the probable use of an n-channel microprocessor integrated circuit of the 8080 type on the Exosat spacecraft of the European Space Agency, was studied. The radiation exposure likely for this device was calculated, using various assumptions for the amount of surrounding absorber, some being intentional shielding others being normal structure elements and device encapsulation. The conclusion was that this type of device could be used if careful engineering design and quality control were used. Mission doses vary between 5000 and 800 rads for various configurations and some patterns of MOS device will tolerate these doses. The use of specially thickened module covers was not recommended, a better method being upgrading device quality and applying internal (local) shielding when necessary and possibly modular addition of external plates in specific directions only. The result of this shielding philosophy would be much greater efficiency in weight use. The further development of a rads (reduction) per gram philosophy was strongly recommended. Throughout, the strong link between mission success and the choice (and control) of the correct MOS manufacturing technology is emphasized and some guidelines on control of manufactured MOS parts (n-channel and complementary type) with respect to tolerance to radiation are given

  11. Solar Storm GIC Forecasting: Solar Shield Extension Development of the End-User Forecasting System Requirements

    Pulkkinen, A.; Mahmood, S.; Ngwira, C.; Balch, C.; Lordan, R.; Fugate, D.; Jacobs, W.; Honkonen, I.

    2015-01-01

    A NASA Goddard Space Flight Center Heliophysics Science Division-led team that includes NOAA Space Weather Prediction Center, the Catholic University of America, Electric Power Research Institute (EPRI), and Electric Research and Management, Inc., recently partnered with the Department of Homeland Security (DHS) Science and Technology Directorate (S&T) to better understand the impact of Geomagnetically Induced Currents (GIC) on the electric power industry. This effort builds on a previous NASA-sponsored Applied Sciences Program for predicting GIC, known as Solar Shield. The focus of the new DHS S&T funded effort is to revise and extend the existing Solar Shield system to enhance its forecasting capability and provide tailored, timely, actionable information for electric utility decision makers. To enhance the forecasting capabilities of the new Solar Shield, a key undertaking is to extend the prediction system coverage across Contiguous United States (CONUS), as the previous version was only applicable to high latitudes. The team also leverages the latest enhancements in space weather modeling capacity residing at Community Coordinated Modeling Center to increase the Technological Readiness Level, or Applications Readiness Level of the system http://www.nasa.gov/sites/default/files/files/ExpandedARLDefinitions4813.pdf.

  12. Repository Waste Package Transporter Shielding Weight Optimization

    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

  13. Mass attenuation coefficients of X-rays in different barite concrete used in radiation protection as shielding against ionizing radiation

    Almeida, A. T. Jr.; Nogueira, M.S.; Santos, M.A.P.; Campos, L.L.; Araújo, F. G. S.

    2015-01-01

    The attenuation coefficient depends on the incident photon energy and the nature of the materials. In order to minimize exposure to individuals. Barite concrete has been largely used as a shielding material in installations housing gamma radiation sources as well as X-ray generating equipment. This study was conducted to evaluate the efficacy of different mixtures of barite concrete for shielding in diagnostic X-ray rooms. The mass attenuation coefficient (μ/ρ). The mass attenuation coefficients have been measured by employing the CdTe detector model XR-100T. The distance between the source and the exposed surface of all samples was measured by SSD light indicator of machine which was 350 cm. The slope of the linear plot of the intensity transmitted versus specimen thickness would yield the attenuation coefficient. The mass attenuation coefficients (μ/ρ) were compared with the tabulations based upon the results of the XCOM program. The rectangular barite concrete blocks in different thicknesses from were used for the radiation attenuation test. The experimental values were compared with theoretical values WinXcom. The plots of the logarithm of transmitted intensity versus specimen thickness were linear for all the samples and the µ/ρ was obtained from the plots by linear regression over the 25%-2% transmission range, under good geometrical condition. There is a good agreement between theoretical and experimental values, within the 9%. In fact over the entire transmission range of 25-2% the experimental and theoretical values agree well for both the energies. (authors)

  14. Problems of the power plant shield optimization

    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

  15. CHARACTERIZING THE STAR FORMATION OF THE LOW-MASS SHIELD GALAXIES FROM HUBBLE SPACE TELESCOPE IMAGING

    McQuinn, Kristen B. W.; Skillman, Evan D.; Simones, Jacob E. [Minnesota Institute for Astrophysics, School of Physics and Astronomy, University of Minnesota, 116 Church Street, S.E., Minneapolis, MN 55455 (United States); Cannon, John M. [Department of Physics and Astronomy, Macalester College, 1600 Grand Avenue, Saint Paul, MN 55105 (United States); Dolphin, Andrew E. [Raytheon Company, 1151 E. Hermans Road, Tucson, AZ 85756 (United States); Haynes, Martha P.; Giovanelli, Riccardo [Center for Radiophysics and Space Research, Space Sciences Building, Cornell University, Ithaca, NY 14853 (United States); Salzer, John J. [Department of Astronomy, Indiana University, 727 East 3rd Street, Bloomington, IN 47405 (United States); Adams, Elizabeth A. K. [Netherlands Institute for Radio Astronomy (ASTRON), Postbus 2, 7900 AA Dwingeloo (Netherlands); Elson, Ed C. [Astrophysics, Cosmology and Gravity Centre (ACGC), Department of Astronomy, University of Cape Town, Private Bag X3, Rondebosch 7701 (South Africa); Ott, Jürgen, E-mail: kmcquinn@astro.umn.edu [National Radio Astronomy Observatory, P.O. Box O, 1003 Lopezville Road, Socorro, NM 87801 (United States)

    2015-03-20

    The Survey of Hi in Extremely Low-mass Dwarfs is an on-going multi-wavelength program to characterize the gas, star formation, and evolution in gas-rich, very low-mass galaxies that populate the faint end of the galaxy luminosity function. The galaxies were selected from the first ∼10% of the Hi Arecibo Legacy Fast ALFA survey based on their low Hi mass and low baryonic mass. Here, we measure the star formation properties from optically resolved stellar populations for 12 galaxies using a color–magnitude diagram fitting technique. We derive lifetime average star formation rates (SFRs), recent SFRs, stellar masses, and gas fractions. Overall, the recent SFRs are comparable to the lifetime SFRs with mean birthrate parameter of 1.4, with a surprisingly narrow standard deviation of 0.7. Two galaxies are classified as dwarf transition galaxies (dTrans). These dTrans systems have star formation and gas properties consistent with the rest of the sample, in agreement with previous results that some dTrans galaxies may simply be low-luminosity dwarf irregulars. We do not find a correlation between the recent star formation activity and the distance to the nearest neighboring galaxy, suggesting that the star formation process is not driven by gravitational interactions, but regulated internally. Further, we find a broadening in the star formation and gas properties (i.e., specific SFRs, stellar masses, and gas fractions) compared to the generally tight correlation found in more massive galaxies. Overall, the star formation and gas properties indicate these very low-mass galaxies host a fluctuating, non-deterministic, and inefficient star formation process.

  16. Photon mass attenuation coefficients of a silicon resin loaded with WO3, PbO, and Bi2O3 Micro and Nano-particles for radiation shielding

    Verdipoor, Khatibeh; Alemi, Abdolali; Mesbahi, Asghar

    2018-06-01

    Novel shielding materials for photons based on silicon resin and WO3, PbO, and Bi2O3 Micro and Nano-particles were designed and their mass attenuation coefficients were calculated using Monte Carlo (MC) method. Using lattice cards in MCNPX code, micro and nanoparticles with sizes of 100 nm and 1 μm was designed inside a silicon resin matrix. Narrow beam geometry was simulated to calculate the attenuation coefficients of samples against mono-energetic beams of Co60 (1.17 and 1.33 MeV), Cs137 (663.8 KeV), and Ba133 (355.9 KeV). The shielding samples made of nanoparticles had higher mass attenuation coefficients, up to 17% relative to those made of microparticles. The superiority of nano-shields relative to micro-shields was dependent on the filler concentration and the energy of photons. PbO, and Bi2O3 nanoparticles showed higher attenuation compared to WO3 nanoparticles in studied energies. Fabrication of novel shielding materials using PbO, and Bi2O3 nanoparticles is recommended for application in radiation protection against photon beams.

  17. PWR upper/lower internals shield

    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.

  18. Shielding requirements on-site loading and acceptance testing on the Leksell gamma knife

    Maitz, A.H.; Lunsford, L.D.; Wu, A.; Lindner, G.; Flickinger, J.C.

    1990-01-01

    On August 14, 1987, the first stereotactic radiosurgical procedure using the gamma knife was performed in North America. Located in a self-contained radiosurgical suite in the basement of Presbyterian-University Hospital in Pittsburgh, Pennsylvania. This device uses 201 highly focused beams 60Co for the single-treatment closed-skull irradiation of brain lesions localized by stereotactic techniques (radiosurgery). One hundred and fifty-two patients with intracranial arteriovenous malformations or brain tumors were treated in the first year of operation. The Presbyterian University Hospital of Pittsburgh gamma knife is the first such unit in which the 60Co sources were loaded on-site. This effort required us to solve some difficult and unusual problems encountered during site preparation, delivery, and loading of the unit in a busy hospital setting. The solutions developed enabled installation and use of the gamma knife with minimal disruption of hospital activities while maintaining acceptable levels of exposure to radiation. Environmental surveys performed during the loading of the 201 radioactive sources (total, 219 TBq) confirmed that on-site loading is possible and practical. Our experience in the design, construction, and implementation of the first North American gamma knife supports the practicality and safety of on-site loading and may be of value in the planning and development of future gamma knife installations

  19. Space nuclear reactor shields for manned and unmanned applications

    McKissock, B.I.; Bloomfield, H.S.

    1990-01-01

    Missions which use nuclear reactor power systems require radiation shielding of payload and/or crew areas to predetermined dose rates. Since shielding can become a significant fraction of the total mass of the system, it is of interest to show the effect of various parameters on shield thickness and mass for manned and unmanned applications. Algorithms were developed to give the thicknesses needed if reactor thermal power, separation distances and dose rates are given as input. The thickness algorithms were combined with models for four different shield geometries to allow tradeoff studies of shield volume and mass for a variety of manned and unmanned missions. The shield design tradeoffs presented in this study include the effects of: higher allowable dose rates; radiation hardened electronics; shorter crew exposure times; shield geometry; distance of the payload and/or crew from the reactor; and changes in the size of the shielded area. Specific NASA missions that were considered in this study include unmanned outer planetary exploration, manned advanced/evolutionary space station and advanced manned lunar base. (author)

  20. Space nuclear reactor shields for manned and unmanned applications

    Mckissock, B.I.; Bloomfield, H.S.

    1989-01-01

    Missions which use nuclear reactor power systems require radiation shielding of payload and/or crew areas to predetermined dose rates. Since shielding can become a significant fraction of the total mass of the system, it is of interest to show the effect of various parameters on shield thickness and mass for manned and unmanned applications. Algorithms were developed to give the thicknesses needed if reactor thermal power, separation distances, and dose rates are given as input. The thickness algorithms were combined with models for four different shield geometries to allow tradeoff studies of shield volume and mass for a variety of manned and unmanned missions. Shield design tradeoffs presented in this study include the effects of: higher allowable dose rates; radiation hardened electronics; shorter crew exposure times; shield geometry; distance of the payload and/or crew from the reactor; and changes in the size of the shielded area. Specific NASA missions that were considered in this study include unmanned outer planetary exploration, manned advanced/evolutionary space station, and advanced manned lunar base

  1. Electromagnetic shielding

    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

  2. MMW [multimegawatt] shielding design and analysis

    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

  3. Space Shielding Materials for Prometheus Application

    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

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

    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.

  5. Shielding plugs

    Makishima, Kenji.

    1986-01-01

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

  6. Using the shield for thermal energy storage in pulsar

    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.)

  7. Shielding container

    Darling, K.A.M.

    1981-01-01

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

  8. Prediction of Marginal Mass Required for Successful Islet Transplantation

    Papas, Klearchos K.; Colton, Clark K.; Qipo, Andi; Wu, Haiyan; Nelson, Rebecca A.; Hering, Bernhard J.; Weir, Gordon C.; Koulmanda, Maria

    2013-01-01

    Islet quality assessment methods for predicting diabetes reversal (DR) following transplantation are needed. We investigated two islet parameters, oxygen consumption rate (OCR) and OCR per DNA content, to predict transplantation outcome and explored the impact of islet quality on marginal islet mass for DR. Outcomes in immunosuppressed diabetic mice were evaluated by transplanting mixtures of healthy and purposely damaged rat islets for systematic variation of OCR/DNA over a wide range. The probability of DR increased with increasing transplanted OCR and OCR/DNA. On coordinates of OCR versus OCR/DNA, data fell into regions in which DR occurred in all, some, or none of the animals with a sharp threshold of around 150-nmol/min mg DNA. A model incorporating both parameters predicted transplantation outcome with sensitivity and specificity of 93% and 94%, respectively. Marginal mass was not constant, depended on OCR/DNA, and increased from 2,800 to over 100,000 islet equivalents/kg body weight as OCR/DNA decreased. We conclude that measurements of OCR and OCR/DNA are useful for predicting transplantation outcome in this model system, and OCR/DNA can be used to estimate the marginal mass required for reversing diabetes. Because human clinical islet preparations in a previous study had OCR/DNA values in the range of 100–150-nmol/min mg DNA, our findings suggest that substantial improvement in transplantation outcome may accompany increasedOCR/DNAin clinical islet preparations. PMID:20233002

  9. Meeting Radiation Protection Requirements and Reducing Spacecraft Mass - A Multifunctional Materials Approach

    Atwell, William; Koontz, Steve; Reddell, Brandon; Rojdev, Kristina; Franklin, Jennifer

    2010-01-01

    Both crew and radio-sensitive systems, especially electronics must be protected from the effects of the space radiation environment. One method of mitigating this radiation exposure is to use passive-shielding materials. In previous vehicle designs such as the International Space Station (ISS), materials such as aluminum and polyethylene have been used as parasitic shielding to protect crew and electronics from exposure, but these designs add mass and decrease the amount of usable volume inside the vehicle. Thus, it is of interest to understand whether structural materials can also be designed to provide the radiation shielding capability needed for crew and electronics, while still providing weight savings and increased useable volume when compared against previous vehicle shielding designs. In this paper, we present calculations and analysis using the HZETRN (deterministic) and FLUKA (Monte Carlo) codes to investigate the radiation mitigation properties of these structural shielding materials, which includes graded-Z and composite materials. This work is also a follow-on to an earlier paper, that compared computational results for three radiation transport codes, HZETRN, HETC, and FLUKA, using the Feb. 1956 solar particle event (SPE) spectrum. In the following analysis, we consider the October 1989 Ground Level Enhanced (GLE) SPE as the input source term based on the Band function fitting method. Using HZETRN and FLUKA, parametric absorbed doses at the center of a hemispherical structure on the lunar surface are calculated for various thicknesses of graded-Z layups and an all-aluminum structure. HZETRN and FLUKA calculations are compared and are in reasonable (18% to 27%) agreement. Both codes are in agreement with respect to the predicted shielding material performance trends. The results from both HZETRN and FLUKA are analyzed and the radiation protection properties and potential weight savings of various materials and materials lay-ups are compared.

  10. Electromagnetic shield

    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

  11. Neutron shieldings

    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.)

  12. Nuclear shields

    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

  13. REACTOR SHIELD

    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.

  14. Selective shielding device for scintiphotography

    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

  15. Shielding requirements for constant-potential diagnostic x-ray beams determined by a Monte Carlo calculation

    Simpkin, D.J.

    1989-01-01

    A Monte Carlo calculation has been performed to determine the transmission of broad constant-potential x-ray beams through Pb, concrete, gypsum wallboard, steel and plate glass. The EGS4 code system was used with a simple broad-beam geometric model to generate exposure transmission curves for published 70, 100, 120 and 140-kVcp x-ray spectra. These curves are compared to measured three-phase generated x-ray transmission data in the literature and found to be reasonable. For calculation ease the data are fit to an equation previously shown to describe such curves quite well. These calculated transmission data are then used to create three-phase shielding tables for Pb and concrete, as well as other materials not available in Report No. 49 of the NCRP

  16. Shielding requirements for constant-potential diagnostic x-ray beams determined by a Monte Carlo calculation.

    Simpkin, D J

    1989-02-01

    A Monte Carlo calculation has been performed to determine the transmission of broad constant-potential x-ray beams through Pb, concrete, gypsum wallboard, steel and plate glass. The EGS4 code system was used with a simple broad-beam geometric model to generate exposure transmission curves for published 70, 100, 120 and 140-kVcp x-ray spectra. These curves are compared to measured three-phase generated x-ray transmission data in the literature and found to be reasonable. For calculation ease the data are fit to an equation previously shown to describe such curves quite well. These calculated transmission data are then used to create three-phase shielding tables for Pb and concrete, as well as other materials not available in Report No. 49 of the NCRP.

  17. Nuclear data for radiation shielding

    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.)

  18. Radiation shielding

    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.)

  19. Concrete shielding exterior to iron

    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

  20. Gonadal shield.

    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.

  1. Hybrid Magnetic Shielding

    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.

  2. Radiation shielding

    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)

  3. Shielded container

    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

  4. Glasses impregnated with lead for radiation shielding

    Abd El Monem, A.M.; Kansouh, W.A.; Megahid, R.M.; Ismail, A.L.; Awad, E.M.

    2005-01-01

    The attenuation properties of glasses with different concentration of lead have been investigated for the attenuation of gamma-rays from cesium-137 and for total gamma rays using a beam of neutrons and gamma rays emitted from californium-252 source. Measurements have been performed using a gamma-ray spectrometer with Nal(T1) detector for gamma-rays emitted from 137 Cs and a neutron/gamma spectrometer with stilbene scintillator for measurement of total gamma-rays from 252 Cf neutron source. The latter applied the pulse shape discrimination technique to distinguish between recoil proton and recoil electron pulses. The obtained results given the form displayed pulse height spectra and attenuation relations which were used to derive the linear attenuation coefficient (μ), and the mass attenuation coefficient (mu/p) of the investigated glasses. In addition, calculations were performed to determine the attenuation properties of glass shields under investigation using XCOM code given by the others. A comparison of the shielding properties of these glasses with some standard shielding materials indicated that, the investigated glasses process the shielding advantages required for different nuclear technology applications

  5. Concrete shielding for nuclear ship 'Mutsu'

    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.)

  6. Design experience: CRBRP radiation shielding

    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''

  7. Functional Requirements for the Next Generation of Mass Notification

    Trumbo, Berkly

    2012-01-01

    While the latest update to National Fire Protection Association (NFPA) redefines mass notification as "emergency communications systems" (ECS), the end user community is formulating expectations related to the future functionality of today's alerting solutions. Numerous best practices have surfaced since alerting technology began its rapid,…

  8. Shielding research in France

    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

  9. Radiation protection/shield design

    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

  10. Modular reactor head shielding system

    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

  11. Radiation shielding material

    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.)

  12. Radiation shielding for fusion reactors

    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)

  13. Radiation shielding for neutron guides

    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)

  14. Radiation shielding for neutron guides

    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

  15. About the Scythian Shields

    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 Computational Analysis

    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)

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

    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.

  18. Shielding benchmark problems

    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)

  19. Radiation shielding device

    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.)

  20. Study of neutron and gamma shielding by lead borate and bismuth lead borate glasses: transparent radiation shielding

    Singh, Vishwanath P.; Badiger, N.M.

    2013-01-01

    Radiation shielding for gamma and neutron is the prominent area in nuclear reactor technology, medical application, dosimetry and other industries. Shielding of these types of radiation requires an appropriate concrete with mixture of low-to-high Z elements which is an opaque medium. The transparent radiation shielding in visible light for gamma and neutron is also extremely essential in the nuclear facilities as lead window. Presently various types of lead equivalent glass oxides have been invented which are transparent as well as provide protection from radiation. In our study we have assessment of effectiveness of neutron and gamma radiation shielding of xPbO.(1-x) B 2 O 3 (x=0.15 to 0.60) and xBi 2 O 3 .(0.80-x) PbO.0.20 B 2 O 3 (x=0.10 to 0.70) transparent borate and bismuth glasses by NXCOM program. The neutron effective mass removal cross section, Σ R /ρ (cm 2 /g) of the lead, bismuth and boron oxides are given. We found invariable Σ R /ρ of various combinations of the lead borate glass for x=0.15 to 0.60 and bismuth lead borate glass for x=0.10 to 0.70. It is observed that the effective removal cross-section for fast neutron (cm -1 ) of lead borate reduces significantly whereas roughly constant for bismuth borate. The gamma mass attenuation coefficients (μ/ρ) of the glasses were also compared with possible experimental values and found comparable. High (μ/ρ) for gamma radiation of the bismuth glasses shows that it is better gamma shielding compared with lead containing glass. However lead borate glasses are better neutron shielding as the neutron removal coefficient are higher. Our investigation is very useful for nuclear reactor technology where prompt neutron of energy 17 MeV and gamma photon up to 10 MeV produced. (author)

  1. A practical neutron shielding design based on data-base interpolation

    Jiang, S.H.; Sheu, R.J.

    1993-01-01

    Neutron shielding design is an important part of the construction of nuclear reactors and high-energy accelerators. Neutron shielding design is also indispensable in the packaging and storage of isotopic neutron sources. Most efforts in the development of neutron shielding design have been concentrated on nuclear reactor shielding because of its huge mass and strict requirement of accuracy. Sophisticated computational tools, such as transport and Monte Carlo codes and detailed data libraries have been developed. In principle, now, neutron shielding, in spite of its complexity, can be designed in any detail and with fine accuracy. However, in most practical cases, neutron shielding design is accomplished with simplified methods. Unlike practical gamma-ray shielding design, where exponential attenuation coupled with buildup factors has been applied effectively and accurately, simplified neutron shielding design, either by using removal cross sections or by applying charts or tables of transmission factors such as the National Council on Radiation Protection and Measurements (NCRP) 38 (Ref. 1) for general neutron protection or to NCRP 51 (Ref. 2) for accelerator neutron shielding, is still very primitive and not well established. The available data are limited in energy range, materials, and thicknesses, and the estimated results are only roughly accurate. It is the purpose of this work to establish a simple, convenient, and user-friendly general-purpose computational tool for practical preliminary neutron shielding design that is reasonably accurate. A wide-range (energy, material, and thickness) data base of dose transmission factors has been generated by applying one-dimensional transport calculations in slab geometry

  2. Concrete shielding for nuclear ship 'Mutsu'

    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.)

  3. Software Tools for Measuring and Calculating Electromagnetic Shielding Effectiveness

    Tesny, Neal

    2005-01-01

    The evaluation and the analysis of high-altitude electromagnetic pulse response of shielded enclosures require the availability of software tools able to acquire data and calculate shielding effectiveness...

  4. Radiation shielding estimates for manned Mars space flight

    Dudkin, V.E.; Kovalev, E.E.; Kolomensky, A.V.; Sakovich, V.A.; Semenov, V.F.; Demin, V.P.; Benton, E.V.

    1992-01-01

    In the analysis of the required radiation shielding for spacecraft during a Mars flight, the specific effects of solar activity (SA) on the intensity of galactic and solar cosmic rays were taken into consideration. Three spaceflight periods were considered: (1) maximum SA; (2) minimum SA; and (3) intermediate SA, when intensities of both galactic and solar cosmic rays are moderately high. Scenarios of spaceflights utilizing liquid-propellant rocket engines, low-and intermediate-thrust nuclear electrojet engines, and nuclear rocket engines, all of which have been designed in the Soviet Union, are reviewed. Calculations were performed on the basis of a set of standards for radiation protection approved by the U.S.S.R. State Committee for Standards. It was found that the lowest estimated mass of a Mars spacecraft, including the radiation shielding mass, obtained using a combination of a liquid propellant engine with low and intermediate thrust nuclear electrojet engines, would be 500-550 metric tons. (author)

  5. Shielded scanning electron microscope for radioactive samples

    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

  6. Core test reactor shield cooling system analysis

    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.)

  7. Thermal design of top shield for PFBR

    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

  8. PC based temporary shielding administrative procedure (TSAP)

    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

  9. PC based temporary shielding administrative procedure (TSAP)

    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. Handout on shielding calculation

    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)

  11. Design of emergency shield

    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

  12. A Sensitivity Study on the Radiation Shield of KSPR Space Reactor

    Cerba, S.; Lee, Hyun Chul; Lim, Hong Sik; Noh, Jae Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The idea of a space reactor was realised some decades ago and since that time several research activities have been performed into this field. The US National Aeronautics and Space Administration (NASA) has been developing a small fast reactor called as fission power system (FPS) for deep space mission, where highly enriched uranium (HEU) is used as fuel. On the other hand, other researchers have also surveyed a thermal reactor concept with low enriched uranium (LEU) for space applications. One of the main concerns in terms of a space reactor is the total size and the mass of the system including the reactor itself as well as the radiation shield. Since the reactor core is a source of neutrons and gamma photons of various energies, which may cause severe damage on the electronics of the space stations, the questions related to the development of a radiation shield should be address appropriately. The proposal of a radiation shield for a small space reactor is discussed in this paper. The requirements for the radiation shield have been addressed in terms of maximal absorbed doses and neutron flounces during 10 years of operation. In this study a radiation shield design for a small space reactor was investigated. All the presented calculations were performed using the multi-purpose stochastic MCNP code with temperature dependent continuous energy ENDF/B VII.0 neutron and photon cross section libraries. The aim of this study was to design a neutron and gamma shield that can meet the requirements of 250 Gy absorbed during 10 years of reactor operation. The comparison with a fast reactor design showed that high content of {sup 238}U strongly influences the shielding mass. This phenomenon is due to the higher photon production in case of the KSPR design and therefore the use of high {sup 235}U enrichments and the operation in fast neutron spectrum may be more desirable. In case if the KSPR space reactor the best shielding performance was achieved while utilizing a multi

  13. Electromagnetically shielded building

    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

  14. Electromagnetically shielded building

    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.

  15. Mars Exploration Rover Heat Shield Recontact Analysis

    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.

  16. Neutron shield analysis and design for the PDX fusion facility

    Grimesey, R.A.; Nigg, D.W.; Scott, A.J.; Wheeler, F.J.; Jassby, D.L.; Perry, E.D.

    1979-01-01

    The basic component of the biological shield for PDX is an existing 81 cm thick high-density concrete shielding wall surrounding the machine. The principal additional shielding requirement is a roof shield over the machine to reduce air-scattered skyshine dose into the PDX control room and to the site boundary. The roof shield is designed in removable sections on a steel support structure permitting overhead crane access to major PDX components. After analysis of a number of alternate concepts, a roof shield consisting of 50 cm of water in polyethylene tanks was selected to meet design objectives of effectiveness, weight, removability, and cost

  17. Radiation shield for nuclear reactors

    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)

  18. Hot Cell Window Shielding Analysis Using MCNP

    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.

  19. ITER shielding blanket

    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.).

  20. Radiation Exposure Analyses Supporting the Development of Solar Particle Event Shielding Technologies

    Walker, Steven A.; Clowdsley, Martha S.; Abston, H. Lee; Simon, Hatthew A.; Gallegos, Adam M.

    2013-01-01

    NASA has plans for long duration missions beyond low Earth orbit (LEO). Outside of LEO, large solar particle events (SPEs), which occur sporadically, can deliver a very large dose in a short amount of time. The relatively low proton energies make SPE shielding practical, and the possibility of the occurrence of a large event drives the need for SPE shielding for all deep space missions. The Advanced Exploration Systems (AES) RadWorks Storm Shelter Team was charged with developing minimal mass SPE storm shelter concepts for missions beyond LEO. The concepts developed included "wearable" shields, shelters that could be deployed at the onset of an event, and augmentations to the crew quarters. The radiation transport codes, human body models, and vehicle geometry tools contained in the On-Line Tool for the Assessment of Radiation In Space (OLTARIS) were used to evaluate the protection provided by each concept within a realistic space habitat and provide the concept designers with shield thickness requirements. Several different SPE models were utilized to examine the dependence of the shield requirements on the event spectrum. This paper describes the radiation analysis methods and the results of these analyses for several of the shielding concepts.

  1. Practical radiation shielding for biomedical research

    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

  2. Electromagnetic shielding formulae

    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)

  3. Shielding benchmark problems, (2)

    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)

  4. The Active Muon Shield

    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.

  5. Radiation Shielding Properties Comparison of Pb-Based Silicate, Borate, and Phosphate Glass Matrices

    Suwimon Ruengsri

    2014-01-01

    Full Text Available Theoretical calculations of mass attenuation coefficients, partial interactions, atomic cross-section, and effective atomic numbers of PbO-based silicate, borate, and phosphate glass systems have been investigated at 662 keV. PbO-based silicate glass has been found with the highest total mass attenuation coefficient and then phosphate and borate glasses, respectively. Compton scattering has been the dominate interaction contributed to the different total attenuation coefficients in each of the glass matrices. The silicate and phosphate glass systems are more appropriate choices as lead-based radiation shielding glass than the borate glass system. Moreover, comparison of results has shown that the glasses possess better shielding properties than standard shielding concretes, suggesting a smaller size requirement in addition to transparency in the visible region.

  6. How many stakes are required to measure the mass balance of a glacier?

    Fountain, A.G.; Vecchia, A.

    1999-01-01

    Glacier mass balance is estimated for South Cascade Glacier and Maclure Glacier using a one-dimensional regression of mass balance with altitude as an alternative to the traditional approach of contouring mass balance values. One attractive feature of regression is that it can be applied to sparse data sets where contouring is not possible and can provide an objective error of the resulting estimate. Regression methods yielded mass balance values equivalent to contouring methods. The effect of the number of mass balance measurements on the final value for the glacier showed that sample sizes as small as five stakes provided reasonable estimates, although the error estimates were greater than for larger sample sizes. Different spatial patterns of measurement locations showed no appreciable influence on the final value as long as different surface altitudes were intermittently sampled over the altitude range of the glacier. Two different regression equations were examined, a quadratic, and a piecewise linear spline, and comparison of results showed little sensitivity to the type of equation. These results point to the dominant effect of the gradient of mass balance with altitude of alpine glaciers compared to transverse variations. The number of mass balance measurements required to determine the glacier balance appears to be scale invariant for small glaciers and five to ten stakes are sufficient.

  7. Technical specifications for the bulk shielding reactor

    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

  8. Wake Shield Target Protection

    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

  9. Technology development for radiation shielding analysis

    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)

  10. Shielding effectiveness of superconductive particles in plastics

    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

  11. Radiation shielding concrete

    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.)

  12. Accelerator shielding benchmark problems

    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)

  13. Ultra high molecular weight polyethylene (UHMWPE) fiber epoxy composite hybridized with Gadolinium and Boron nanoparticles for radiation shielding

    Mani, Venkat; Prasad, Narasimha S.; Kelkar, Ajit

    2016-09-01

    Deep space radiations pose a major threat to the astronauts and their spacecraft during long duration space exploration missions. The two sources of radiation that are of concern are the galactic cosmic radiation (GCR) and the short lived secondary neutron radiations that are generated as a result of fragmentation that occurs when GCR strikes target nuclei in a spacecraft. Energy loss, during the interaction of GCR and the shielding material, increases with the charge to mass ratio of the shielding material. Hydrogen with no neutron in its nucleus has the highest charge to mass ratio and is the element which is the most effective shield against GCR. Some of the polymers because of their higher hydrogen content also serve as radiation shield materials. Ultra High Molecular Weight Polyethylene (UHMWPE) fibers, apart from possessing radiation shielding properties by the virtue of the high hydrogen content, are known for extraordinary properties. An effective radiation shielding material is the one that will offer protection from GCR and impede the secondary neutron radiations resulting from the fragmentation process. Neutrons, which result from fragmentation, do not respond to the Coulombic interaction that shield against GCR. To prevent the deleterious effects of secondary neutrons, targets such as Gadolinium are required. In this paper, the radiation shielding studies that were carried out on the fabricated sandwich panels by vacuum-assisted resin transfer molding (VARTM) process are presented. VARTM is a manufacturing process used for making large composite structures by infusing resin into base materials formed with woven fabric or fiber using vacuum pressure. Using the VARTM process, the hybridization of Epoxy/UHMWPE composites with Gadolinium nanoparticles, Boron, and Boron carbide nanoparticles in the form of sandwich panels were successfully carried out. The preliminary results from neutron radiation tests show that greater than 99% shielding performance was

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

    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

  15. Progress on establishing guidelines for National Ignition Facility (NIF) experiments to extend debris shield lifetime

    Tobin, M.; Eder, D.; Braun, D.; MacGowan, B.

    2002-01-01

    The survivability of the debris shields on the National Ignition Facility (NIF) are a key factor for the affordable operation of the facility. The improvements required over Nova debris shields are described. Estimates of debris shield lifetimes in the presence of target emissions with 4-8 J/cm 2 laser fluences indicate lifetimes that may contribute unacceptably to operations costs for NIF. We are developing detailed suggested guidance for target and experiment designers for NIF to assist in minimizing the damage to, and therefore the cost of, maintaining NIF debris shields. The guidance suggests a target mass quantity that as particulate on the debris shields (300 mg) may be within current operating budgets. It also suggests the amount of material that should become shrapnel on a shot (10 mg). Finally, it suggests the level of non-volatile residue (NVR) that would threaten the sol-gel coatings on the debris shields (1 μg/cm 2 ). We review the experimentation on the Nova chamber that included measuring quantities of particulate on debris shields by element and capturing shrapnel pieces in aerogel samples mounted in the chamber. We also describe computations of X-ray emissions from a likely NIF target and the associated ablation expected from this X-ray exposure on supporting target hardware. We describe progress in assessing the benefits of a pre-shield and the possible impact on the guidance for target experiments on NIF. Plans for possible experimentation on Omega and other facilities to improve our understanding of target emissions and their impacts are discussed. Our discussion of planned future work provides a forum to invite possible collaboration with the IFE community

  16. INTOR radiation shielding for personnel access

    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

  17. Jet energy measurements at ILC. Calorimeter DAQ requirements and application in Higgs boson mass measurements

    Ebrahimi, Aliakbar

    2017-11-01

    required for the Higgs boson mass measurement can only be achieved using the particle flow approach to reconstruction. The particle flow approach requires highly-granular calorimeters and a highly efficient tracking system. The CALICE collaboration is developing highly-granular calorimeters for such applications. One of the challenges in the development of such calorimeters with millions of read-out channels is their Data Acquisition System (DAQ) system. The second part of this thesis involves contributions to development of a new DAQ system for the CALICE scintillator calorimeters. The new DAQ system fulfills the requirements for the prototypes tests while being scalable to larger systems. The requirements and general architecture of the DAQ system is outlined in this thesis. The new DAQ system has been commissioned and tested with particle beams at the CERN Proton Synchrotron test beam facility in 2014,results of which are presented here.

  18. Jet energy measurements at ILC. Calorimeter DAQ requirements and application in Higgs boson mass measurements

    Ebrahimi, Aliakbar

    2017-11-15

    jet energy resolution required for the Higgs boson mass measurement can only be achieved using the particle flow approach to reconstruction. The particle flow approach requires highly-granular calorimeters and a highly efficient tracking system. The CALICE collaboration is developing highly-granular calorimeters for such applications. One of the challenges in the development of such calorimeters with millions of read-out channels is their Data Acquisition System (DAQ) system. The second part of this thesis involves contributions to development of a new DAQ system for the CALICE scintillator calorimeters. The new DAQ system fulfills the requirements for the prototypes tests while being scalable to larger systems. The requirements and general architecture of the DAQ system is outlined in this thesis. The new DAQ system has been commissioned and tested with particle beams at the CERN Proton Synchrotron test beam facility in 2014,results of which are presented here.

  19. Shielded regenerative neutron detector

    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

  20. Radiation shielding plate

    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.)

  1. Development of point Kernel radiation shielding analysis computer program implementing recent nuclear data and graphic user interfaces

    Kang, S.; Lee, S.; Chung, C.

    2002-01-01

    There is an increasing demand for safe and efficient use of radiation and radioactive work activity along with shielding analysis as a result the number of nuclear and conventional facilities using radiation or radioisotope rises. Most Korean industries and research institutes including Korea Power Engineering Company (KOPEC) have been using foreign computer programs for radiation shielding analysis. Korean nuclear regulations have introduced new laws regarding the dose limits and radiological guides as prescribed in the ICRP 60. Thus, the radiation facilities should be designed and operated to comply with these new regulations. In addition, the previous point kernel shielding computer code utilizes antiquated nuclear data (mass attenuation coefficient, buildup factor, etc) which were developed in 1950∼1960. Subsequently, the various nuclear data such mass attenuation coefficient, buildup factor, etc. have been updated during the past few decades. KOPEC's strategic directive is to become a self-sufficient and independent nuclear design technology company, thus KOPEC decided to develop a new radiation shielding computer program that included the latest regulatory requirements and updated nuclear data. This new code was designed by KOPEC with developmental cooperation with Hanyang University, Department of Nuclear Engineering. VisualShield is designed with a graphical user interface to allow even users unfamiliar to radiation shielding theory to proficiently prepare input data sets and analyzing output results

  2. Minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems

    Jordan, W.C.; Turner, J.C.

    1992-12-01

    A parametric calculational analysis has been performed in order to estimate the minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems. The analysis was performed using a version of the SCALE-4.0 code system and the 27-group ENDF/B-IV cross-section library. Water-moderated uranyl fluoride (UO{sub 2}F{sub 2} and H{sub 2}O) and hydrofluoric-acid-moderated uranium hexaflouride (UF{sub 6} and HF) systems were considered in the analysis over enrichments of 1.4 to 5 wt % {sup 235}U. Estimates of the minimum critical volume, minimum critical mass of uranium, and the minimum mass of moderator required for criticality are presented. There was significant disagreement between the values generated in this study when compared with a similar undocumented study performed in 1983 using ANISN and the Knight-modified Hansen-Roach cross sections. An investigation into the cause of the disagreement was made, and the results are presented.

  3. Minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems

    Jordan, W.C.; Turner, J.C.

    1992-12-01

    A parametric calculational analysis has been performed in order to estimate the minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems. The analysis was performed using a version of the SCALE-4.0 code system and the 27-group ENDF/B-IV cross-section library. Water-moderated uranyl fluoride (UO[sub 2]F[sub 2] and H[sub 2]O) and hydrofluoric-acid-moderated uranium hexaflouride (UF[sub 6] and HF) systems were considered in the analysis over enrichments of 1.4 to 5 wt % [sup 235]U. Estimates of the minimum critical volume, minimum critical mass of uranium, and the minimum mass of moderator required for criticality are presented. There was significant disagreement between the values generated in this study when compared with a similar undocumented study performed in 1983 using ANISN and the Knight-modified Hansen-Roach cross sections. An investigation into the cause of the disagreement was made, and the results are presented.

  4. Radiation shielding of the main injector

    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

  5. Magnet Architectures and Active Radiation Shielding Study - SR2S Workshop

    Westover, Shane; Meinke, Rainer; Burger, William; Ilin, Andrew; Nerolich, Shaun; Washburn, Scott

    2014-01-01

    Analyze new coil configurations with maturing superconductor technology -Develop vehicle-level concept solutions and identify engineering challenges and risks -Shielding performance analysis Recent advances in superconducting magnet technology and manufacturing have opened the door for re-evaluating active shielding solutions as an alternative to mass prohibitive passive shielding.Publications on static magnetic field environments and its bio-effects were reviewed. Short-term exposure information is available suggesting long term exposure may be okay. Further research likely needed. center dotMagnetic field safety requirements exist for controlled work environments. The following effects have been noted with little noted adverse effects -Magnetohydrodynamic (MHD) effects on ionized fluids (e.g. blood) creating an aortic voltage change -MHD interaction elevates blood pressure (BP) center dot5 Tesla equates to 5% BP elevation -Prosthetic devises and pacemakers are an issue (access limit of 5 gauss).

  6. HPGe detector shielding adjustment

    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

    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. New facility shield design criteria

    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

  9. Facility target insert shielding assessment

    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.

  10. Superconducting magnetic shields for neutral beam injectors. Final report

    1985-04-01

    Large high energy deuterium neutral beams which must be made from negative ions require extensive magnetic shielding against the intense fringe fields surrounding a magnetic fusion power plant. The feasibility of shielding by multilayer sheets of copper-superconducting laminated material was investigated. It was found that, if necessary fabrication techniques are developed, intrinsically stable type II superconductors will be able to shield against the magnetic fields of the fusion reactors. Among the immediate benefits of this research is better magnetic shields for neutral beam injectors in support of DOE's fusion program. Another application may be in the space vehicles, where difficulties in transporting heavy μ-metal sections may make a comparatively light superconducting shield attractive. Also, as high-field superconducting magnets find widespread applications, the need for high-intensity magnetic shielding will increase. As a result, the commercial market for the magnetic shields should expand along with the market for superconducting magnets

  11. New applications and developments in the neutron shielding

    Uğur Fatma Aysun

    2017-01-01

    Full Text Available Shielding neutrons involve three steps that are slowing neutrons, absorption of neutrons, and impregnation of gamma rays. Neutrons slow down with thermal energy by hydrogen, water, paraffin, plastic. Hydrogenated materials are also very effective for the absorption of neutrons. Gamma rays are produced by neutron (radiation retention on the neutron shield, inelastic scattering, and degradation of activation products. If a source emits gamma rays at various energies, high-energy gamma rays sometimes specify shielding requirements. Multipurpose Materials for Neutron Shields; Concrete, especially with barium mixed in, can slow and absorb the neutrons, and shield the gamma rays. Plastic with boron is also a good multipurpose shielding material. In this study; new applications and developments in the area of neutron shielding will be discussed in terms of different materials.

  12. New applications and developments in the neutron shielding

    Uğur, Fatma Aysun

    2017-09-01

    Shielding neutrons involve three steps that are slowing neutrons, absorption of neutrons, and impregnation of gamma rays. Neutrons slow down with thermal energy by hydrogen, water, paraffin, plastic. Hydrogenated materials are also very effective for the absorption of neutrons. Gamma rays are produced by neutron (radiation) retention on the neutron shield, inelastic scattering, and degradation of activation products. If a source emits gamma rays at various energies, high-energy gamma rays sometimes specify shielding requirements. Multipurpose Materials for Neutron Shields; Concrete, especially with barium mixed in, can slow and absorb the neutrons, and shield the gamma rays. Plastic with boron is also a good multipurpose shielding material. In this study; new applications and developments in the area of neutron shielding will be discussed in terms of different materials.

  13. Mechanical design of the TIBER breeding shield

    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.).

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

    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)

  15. Shielding member for thermonuclear device

    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.)

  16. Mass

    Quigg, Chris

    2007-01-01

    In the classical physics we inherited from Isaac Newton, mass does not arise, it simply is. The mass of a classical object is the sum of the masses of its parts. Albert Einstein showed that the mass of a body is a measure of its energy content, inviting us to consider the origins of mass. The protons we accelerate at Fermilab are prime examples of Einsteinian matter: nearly all of their mass arises from stored energy. Missing mass led to the discovery of the noble gases, and a new form of missing mass leads us to the notion of dark matter. Starting with a brief guided tour of the meanings of mass, the colloquium will explore the multiple origins of mass. We will see how far we have come toward understanding mass, and survey the issues that guide our research today.

  17. Shielding design for positron emission tomography facility

    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

  18. Radiation shielding member

    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.)

  19. Scintillation counter, segmented shield

    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.)

  20. Shields for nuclear reactors

    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)

  1. Radiation shielding curtain

    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

  2. Shielded cells transfer automation

    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

  3. Analytic Ballistic Performance Model of Whipple Shields

    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.

  4. REPOSITORY LAYOUT SUPPORTING DESIGN FEATURE NO.13 - WASTE PACKAGE SELF SHIELDING

    Owen, J.

    1999-01-01

    The objective of this analysis is to develop a repository layout, for Feature No. 13, that will accommodate self-shielding waste packages (WP) with an areal mass loading of 25 metric tons of uranium per acre (MTU/acre). The scope of this analysis includes determination of the number of emplacement drifts, amount of emplacement drift excavation required, and a preliminary layout for illustrative purposes

  5. CSER 95-002: ALARA shielding for IAEA SNM container movement

    Miller, E.M.

    1995-01-01

    This CSER qualifies use of a 5% borated, lead foil lined polyethylene 1 inch annulus as a bucket and in a small carrier to move sealed containers of plutonium. The containers are Oversize Cans or smaller containing plutonium limited in mass and H/Pu ratio by PFP storage and transportation CPS's. These ALARA shielding units reduce personnel exposure to the radiation from the containers as they are moved for assay and other required activities

  6. Lean mass and fat mass predict bone mineral density in middle-aged individuals with noninsulin-requiring type 2 diabetes mellitus.

    Moseley, Kendall F; Dobrosielski, Devon A; Stewart, Kerry J; De Beur, Suzanne M Jan; Sellmeyer, Deborah E

    2011-05-01

    Despite high bone mineral density (BMD), persons with type 2 diabetes are at greater risk of fracture. The relationship between body composition and BMD in noninsulin-requiring diabetes is unclear. The aim was to examine how fat and lean mass independently affect the skeleton in this population. Subjects for this cross-sectional analysis were men (n = 78) and women (n = 56) aged 40-65 years (56 ± 6 years) with uncomplicated, noninsulin-requiring type 2 diabetes. Total body fat and lean mass, total body, hip and lumbar spine BMD were measured with dual energy X-ray absorptiometry. Magnetic resonance imaging measured total abdominal, visceral and subcutaneous (SQ) fat. Subjects had normal all-site BMD and were obese to overweight (body mass index 29-41 kg/m(2)) with controlled diabetes (HbA1c women 6·6 ± 1·2%, men 6·7 ± 1·6%). Lean mass was positively associated with total body, hip, femoral neck and hip BMD in both sexes. Fat mass, abdominal total and SQ fat were associated with total body and hip BMD in women. In multivariate analyses adjusted for sex, lean mass significantly predicted total, hip and femoral neck BMD in men and women. In unadjusted models, lean mass continued to predict BMD at these sites in men; fat mass also predicted total body, femoral and hip BMD in women. In men and women with uncomplicated, noninsulin-requiring diabetes, lean mass significantly predicted BMD at the total body, hip and femoral neck. Further research is needed to determine whether acquisition or maintenance of lean mass in T2DM can prevent hip fracture in this at-risk population. © 2011 Blackwell Publishing Ltd.

  7. The Spallation Neutron Source (SNS) conceptual design shielding analysis

    Johnson, J.O.; Odano, N.; Lillie, R.A.

    1998-03-01

    The shielding design is important for the construction of an intense high-energy accelerator facility like the proposed Spallation Neutron Source (SNS) due to its impact on conventional facility design, maintenance operations, and since the cost for the radiation shielding shares a considerable part of the total facility costs. A calculational strategy utilizing coupled high energy Monte Carlo calculations and multi-dimensional discrete ordinates calculations, along with semi-empirical calculations, was implemented to perform the conceptual design shielding assessment of the proposed SNS. Biological shields have been designed and assessed for the proton beam transport system and associated beam dumps, the target station, and the target service cell and general remote maintenance cell. Shielding requirements have been assessed with respect to weight, space, and dose-rate constraints for operating, shutdown, and accident conditions. A discussion of the proposed facility design, conceptual design shielding requirements calculational strategy, source terms, preliminary results and conclusions, and recommendations for additional analyses are presented

  8. Magnetic shield effect simulation of superconducting film shield covering directly coupled HTS dc-SQUID magnetometer

    Terauchi, N.; Noguchi, S.; Igarashi, H.

    2011-01-01

    A superconducting film shield over a SQUID ring improves the robustness of the SQUID with respect to magnetic noise. Supercurrent in the SQUID magnetometer and the superconducting film shield were simulated. The superconducting film shield reduces the influence of the external magnetic field on the SQUID ring. An HTS SQUID is a high sensitive magnetic sensor. In recent years, the HTS SQUID is widely used in various applications. In some applications, high robustness with respect to magnetic noise is required to realize stable operation at outside of a magnetic shielding room. The target of this paper is a directly coupled HTS dc-SQUID magnetometer. To enhance the robustness of the SQUID magnetometer, use of a superconducting thin film shield has been proposed. The magnetic field directly penetrating the SQUID ring causes the change of the critical current of Josephson junction, and then the SQUID magnetometer transitions into inoperative state. In order to confirm the magnetic shield effect of the superconducting film shield, electromagnetic field simulation with 3D edge finite element method was performed. To simulate the high temperature superconductor, E-J characteristics and c-axis anisotropy are considered. To evaluate the effect of the superconducting film shield, an external magnetic field which is supposed to be a magnetic noise is applied. From the simulation results, the time transition of the magnetic flux penetrating the SQUID ring is investigated and the effect of the superconducting film shield is confirmed. The amplitude of the magnetic flux penetrating the SQUID ring can be reduced to about one-sixth since the superconducting film shield prevents the magnetic noise from directly penetrating the SQUID ring.

  9. An attenuation Layer for Electromagnetic Shielding in X- Band Frequency

    Vida Zaroushani

    2015-06-01

    Full Text Available Uncontrolled exposure to X-band frequency leads to health damage. One of the principles of radiation protection is shielding. But, conventional shielding materials have disadvantages. Therefore, studies of novel materials, as an alternative to conventional shielding materials, are required to obtain new electromagnetic shielding material. Therefore, this study investigated the electromagnetic shielding of two component epoxy thermosetting resin for the X - band frequency with workplace approach. Two components of epoxy resin mixed according to manufacturing instruction with the weight ratio that was 100:10 .Epoxy plates fabricated in three different thicknesses (2, 4 and 6mm and shielding effectiveness measured by Vector Network Analyzer. Then, shielding effectiveness measured by the scattering parameters.The results showed that 6mm thickness of epoxy had the highest and 2mm had the lowest average of shielding effectiveness in X-band frequency that is 4.48 and 1.9 dB, respectively. Also, shielding effectiveness increased by increasing the thickness. But this increasing is useful up to 4mm. Percentage shielding effectiveness of attenuation for 6, 4 and 2mm thicknesses is 64.35%, 63.31% and 35.40%. Also, attenuation values for 4mm and 6mm thicknesses at 8.53 GHz and 8.52 GHz frequency are 77.15% and 82.95%, respectively, and can be used as favourite shields for the above frequency. 4mm-Epoxy is a suitable candidate for shielding application in X-band frequency range but, in the lower section, 6mm thickness is recommended. Finely, the shielding matrix can be used for selecting the proper thickness for electromagnetic shielding in X- Band frequency.

  10. Innovative technologies for Faraday shield cooling

    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

  11. Current trends and future requirements for the mass spectrometric investigation of microbial, mammalian and plant metabolomes

    Dunn, Warwick B

    2008-01-01

    The functional levels of biological cells or organisms can be separated into the genome, transcriptome, proteome and metabolome. Of these the metabolome offers specific advantages to the investigation of the phenotype of biological systems. The investigation of the metabolome (metabolomics) has only recently appeared as a mainstream scientific discipline and is currently developing rapidly for the study of microbial, plant and mammalian metabolomes. The metabolome pipeline or workflow encompasses the processes of sample collection and preparation, collection of analytical data, raw data pre-processing, data analysis and data storage. Of these processes the collection of analytical data will be discussed in this review with specific interest shown in the application of mass spectrometry in the metabolomics pipeline. The current developments in mass spectrometry platforms (GC–MS, LC–MS, DIMS and imaging MS) and applications of specific interest will be highlighted. The current limitations of these platforms and applications will be discussed with areas requiring further development also highlighted. These include the detectable coverage of the metabolome, the identification of metabolites and the process of converting raw data to biological knowledge. (review article)

  12. Analyzing the requirements for mass production of small wind turbine generators

    Anuskiewicz, T.; Asmussen, J.; Frankenfield, O.

    Mass producibility of small wind turbine generators to give manufacturers design and cost data for profitable production operations is discussed. A 15 kW wind turbine generator for production in annual volumes from 1,000 to 50,000 units is discussed. Methodology to cost the systems effectively is explained. The process estimate sequence followed is outlined with emphasis on the process estimate sheets compiled for each component and subsystem. These data enabled analysts to develop cost breakdown profiles crucial in manufacturing decision-making. The appraisal also led to various design recommendations including replacement of aluminum towers with cost effective carbon steel towers. Extensive cost information is supplied in tables covering subassemblies, capital requirements, and levelized energy costs. The physical layout of the plant is depicted to guide manufacturers in taking advantage of the growing business opportunity now offered in conjunction with the national need for energy development.

  13. Latest experiences in inspecting the inside of BWR vessel shields

    Alberdi, R.; Gonzalez, E.

    2001-07-01

    In the last few years, the owners of BWR nuclear power plants have been forced to address new fuel shield inspection requirements, TECNATOM has responded to this situation by launching the TEIDE projects, which include development of an inspection machine and the corresponding Non-Destructive Tests to examine the inside of this shield. With these projects, TECNATOM has performed more than 12 fuel shield inspections in different countries. This article describes the experience gained in the last three years. (Author)

  14. Shear-wave elastography and greyscale assessment of palpable probably benign masses: is biopsy always required?

    Giannotti, Elisabetta; Vinnicombe, Sarah; Thomson, Kim; McLean, Dennis; Purdie, Colin; Jordan, Lee; Evans, Andy

    2016-06-01

    To establish if palpable breast masses with benign greyscale ultrasound features that are soft on shear-wave elastography (SWE) (mean stiffness masses at ultrasound. All underwent ultrasound, SWE and needle core biopsy. Static greyscale images were retrospectively assigned Breast Imaging Reporting and Data System (BI-RADS) scores by two readers blinded to the SWE and pathology findings, but aware of the patient's age. A mean stiffness of 50 kPa was used as the SWE cut-off for calling a lesion soft or stiff. Histological findings were used to establish ground truth. No cancer had benign characteristics on both modalities. 466 (99.8%) of the 467 cancers were classified BI-RADS 4a or above. The one malignant lesion classified as BI-RADS 3 was stiff on SWE. 446 (96%) of the 467 malignancies were stiff on SWE. No cancer in females under 40 years had benign SWE features. 74 (32.6%) of the 227 benign lesions were BI-RADS 3 and soft on SWE; so, biopsy could potentially have been avoided in this group. Lesions which appear benign on greyscale ultrasound and SWE do not require percutaneous biopsy or short-term follow-up, particularly in females under 40 years. None of the cancers had benign characteristics on both greyscale ultrasound and SWE, and 32% of benign lesions were BI-RADS 3 and soft on SWE; lesions that are benign on both ultrasound and SWE may not require percutaneous biopsy or short-term follow-up.

  15. Neutron shielding material

    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)

  16. Concrete radiation shielding

    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)

  17. Method for dismantling shields

    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.)

  18. Mechanical shielded hot cell

    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)

  19. Radiation shielding issues on the FMIT

    Burke, R.J.; Davis, A.A.; Huang, S.; Morford, R.J.

    1981-05-01

    The Fusion Materials Irradiation Test Facility (FMIT) is being built to study neutron radiation effects in candidate fusion reactor materials. The FMIT will yield high fluence data in a fusion-like neutron radiation environment produced by the interaction of a 0.1A, 35 MeV deuteron beam with a flowing lithium target. The design of the facility as a whole is driven by a high availability requirement. The variety of radiation environments in the facility requires the use of diverse and extensive shielding. Shielding design throughout the FMIT must accommodate the need for maintenance and operations access while providing adequate personnel and equipment protection

  20. WASTE HANDLING BUILDING SHIELD WALL ANALYSIS

    Padula, D.

    2000-01-01

    The scope of this analysis is to estimate the shielding wall, ceiling or equivalent door thicknesses that will be required in the Waste Handling Building to maintain the radiation doses to personnel within acceptable limits. The shielding thickness calculated is the minimum required to meet administrative limits, and not necessarily what will be recommended for the final design. The preliminary evaluations will identify the areas which have the greatest impact on mechanical and facility design concepts. The objective is to provide the design teams with the necessary information to assure an efficient and effective design

  1. The manufacturing of depleted uranium biological shield components

    Metelkin, J.A.

    1998-01-01

    The unique combination of the physical and mechanical properties of uranium made it possible to manufacture biological shield components of transport package container (TPC) for transportation nuclear power plant irradiated fuel and radionuclides of radiation diagnostic instruments. Protective properties are substantially dependent on the nature radionuclide composition of uranium, that why I recommended depleted uranium after radiation chemical processing. Depleted uranium biological shield (DUBS) has improved specific mass-size characteristics compared to a shield made of lead, steel or tungsten. Technological achievements in uranium casting and machining made it possible to manufacture DUBS components of TPC up to 3 tons of mass and up to 2 metres of the maximum size. (authors)

  2. LOFT shield tank steady state temperatures with addition of gamma and neutron shielding

    Kyllingstad, G.

    1977-01-01

    The effect of introducing a neutron and gamma shield into the annulus between the reactor vessel and the shield tank is analyzed. This addition has been proposed in order to intercept neutron streaming up the annulus during nuclear operations. Its installation will require removal of approximately 20- 1 / 2 inches of stainless steel foil insulation at the top of the annulus. The resulting conduction path is believed to result in increased water temperatures within the shield tank, possibly beyond the 150 0 F limit, and/or cooling of the reactor vessel nozzles such that adverse thermal stresses would be generated. A two dimensional thermal analysis using the finite element code COUPLE/MOD2 was done for the shield tank system illustrated in the figure (1). The reactor was assumed to be at full power, 55 MW (th), with a loop flow rate of 2.15 x 10 6 lbm/hr (268.4 kg/s) at 2250 psi (15.51 MPa). Calculations indicate a steady state shield tank water temperature of 140 0 F (60 0 C). This is below the 150 0 F (65.56 0 C) limit. Also, no significant changes in thermal gradients within the nozzle or reactor vessel wall are generated. A spacer between the gamma shield and the shield tank is recommended, however, in order to ensure free air circulation through the annulus

  3. Shield support frame. Schildausbaugestell

    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).

  4. Radiation shielding material

    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.)

  5. Reactor head shielding apparatus

    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

  6. MEANS FOR SHIELDING AND COOLING REACTORS

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

    1959-02-10

    Reactors of the water-cooled type and a means for shielding such a rcactor to protect operating personnel from harmful radiation are discussed. In this reactor coolant tubes which contain the fissionable material extend vertically through a mass of moderator. Liquid coolant enters through the bottom of the coolant tubes and passes upwardly over the fissionable material. A shield tank is disposed over the top of the reactor and communicates through its bottom with the upper end of the coolant tubes. A hydrocarbon shielding fluid floats on the coolant within the shield tank. With this arrangements the upper face of the reactor can be opened to the atmosphere through the two superimposed liquid layers. A principal feature of the invention is that in the event radioactive fission products enter thc coolant stream. imposed layer of hydrocarbon reduces the intense radioactivity introduced into the layer over the reactors and permits removal of the offending fuel material by personnel shielded by the uncontaminated hydrocarbon layer.

  7. A comparative study of two digestion methods employed for the determination boron in ferroboron used as an advanced shielding material

    Kamble, Granthali S.; Manisha, V.; Venkatesh, K.

    2015-01-01

    Shielding of nuclear reactor core is an important requirement of fast reactors. An important objective of future Fast Breeder Reactors (FBRs) is to reduce the volume of shields. A large number of materials have been considered for use to reduce the neutron flux to acceptable levels. A shield material which brings down the energy of neutrons by elastic and inelastic scattering along with absorption will be more effective. Ferro boron is identified as one of the advanced shielding materials considered for use in future FBRs, planned to be constructed in India. Ferroboron is an economical and indigenously available material which qualifies as a promising shield material through literature survey and scoping calculations. Experiments have been conducted in KAMINI reactor to understand the effectiveness of prospective shield material Ferro-boron as an in-core shield material for future FBRs. The Ferro boron used in these experiments contained 11.8% and 15% of boron. Precise determination of boron content in these ferro boron samples is very important to determine its effectiveness as a shield material. In this work a comparative study was carried out to determine the boron content in ferro boron samples. In the first method the sample was treated with incremental amounts of nitric acid under reflux (to prevent rigorous reaction and volatalisation of boron). The solution was gradually heated and the solution was filtered through a Whatman Filter paper no. 41. The undissolved ferro boron residue collected in the filter paper after filtration, is transferred to a platinum crucible; mixed with sodium carbonate and is ashed. The crucible is placed over a burner for 1 h to fuse the contents. The fused mass is leached in dilute hydrochloric acid, added to the nitric acid filtrate and made up to pre-determined volume

  8. Slipforming of reinforced concrete shield building

    Hsieh, M.C.; King, J.R.

    1982-01-01

    The unique design and construction features of slipforming the heavily reinforced concrete cylindrical shield walls at the Satsop nuclear plant in Washington, D.C. site are presented. The shield walls were designed in compliance with seismic requirements which resulted in the need for reinforcing steel averaging 326 kg/m/sup 3/. A 7.6 m high, three-deck moving platform was designed to permit easy installation of the reinforcing steel, embedments, and blockouts, and to facilitate concrete placement and finishing. Two circular box trusses, one on each side of the shield wall, were used in combination with a spider truss to meet both the tolerance and strength requirements for the slipform assembly

  9. Radiation shielding bricks

    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)

  10. Hybrid Active-Passive Radiation Shielding System

    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...

  11. Radiation shielding cloth

    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.)

  12. Electrostatic shielding of transformers

    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.

  13. Penetration portion shielding structure

    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.)

  14. Dosimetry and shielding

    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

  15. Radiation shielding glass

    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.)

  16. SHIELD 1.0: development of a shielding calculator program in diagnostic radiology

    Santos, Romulo R.; Real, Jessica V.; Luz, Renata M. da; Friedrich, Barbara Q.; Silva, Ana Maria Marques da

    2013-01-01

    In shielding calculation of radiological facilities, several parameters are required, such as occupancy, use factor, number of patients, source-barrier distance, area type (controlled and uncontrolled), radiation (primary or secondary) and material used in the barrier. The shielding design optimization requires a review of several options about the physical facility design and, mainly, the achievement of the best cost-benefit relationship for the shielding material. To facilitate the development of this kind of design, a program to calculate the shielding in diagnostic radiology was implemented, based on data and limits established by National Council on Radiation Protection and Measurements (NCRP) 147 and SVS-MS 453/98. The program was developed in C⌗ language, and presents a graphical interface for user data input and reporting capabilities. The module initially implemented, called SHIELD 1.0, refers to calculating barriers for conventional X-ray rooms. The program validation was performed by the comparison with the results of examples of shielding calculations presented in NCRP 147.

  17. Radiation shielding aspects for long manned mission to space - Criteria, survey study and preliminary model

    Sztejnberg, M.; Xiao, S.; Satvat, N.; Limon, F.; Hopkins, J.; Jevremovic, T.; T. Jevremovic)

    2006-01-01

    The prospect of manned space missions out side Earth's or bit is limited by the travel time and shielding against cosmic radiation. The chemical rockets currently used in the space program have no hope of propelling a manned vehicle to a far away location such as Mars due to the enormous mass of fuel that would be required. The specific energy available from nuclear fuel is a factor of 106 higher than chemical fuel; it is there fore obvious that nuclear power production in space is a must. On the other hand, recent considerations to send a man to the Moon for a long stay would require a stable, secured, and safe source of energy (there is hardly anything beyond nuclear power that would provide a useful and reliably safe sustainable supply of energy). National Aeronautics and Space Administration (NASA) anticipates that the mass of a shielding material required for long travel to Mars is the next major design driver. In 2006 NASA identified a need to assess and evaluate potential gaps in existing knowledge and understanding of the level and types of radiation critical to astronauts' health during the long travel to Mars and to start a comprehensive study related to the shielding design of a spacecraft finding the conditions for the mitigation of radiation components contributing to the doses beyond accepted limits. In order to reduce the overall space craft mass, NASA is looking for the novel, multi-purpose and multi-functional materials that will provide effective shielding of the crew and electronics on board. The Laboratory for Neutronics and Geometry Computation in the School of Nuclear Engineering at Purdue University led by Prof. Tatjana Jevremovic began in 2004 the analytical evaluations of different lightweight materials. The preliminary results of the design survey study are presented in this paper. (author)

  18. Radiation shielding aspects for long manned mission to space: Criteria, survey study, and preliminary model

    Sztejnberg Manuel

    2006-01-01

    Full Text Available The prospect of manned space missions outside Earth's orbit is limited by the travel time and shielding against cosmic radiation. The chemical rockets currently used in the space program have no hope of propelling a manned vehicle to a far away location such as Mars due to the enormous mass of fuel that would be required. The specific energy available from nuclear fuel is a factor of 106 higher than chemical fuel; it is therefore obvious that nuclear power production in space is a must. On the other hand, recent considerations to send a man to the Moon for a long stay would require a stable, secured and safe source of energy (there is hardly anything beyond nuclear power that would provide a useful and reliably safe sustainable supply of energy. National Aeronautics and Space Administration (NASA anticipates that the mass of a shielding material required for long travel to Mars is the next major design driver. In 2006 NASA identified a need to assess and evaluate potential gaps in existing knowledge and understanding of the level and types of radiation critical to astronauts' health during the long travel to Mars and to start a comprehensive study related to the shielding design of a spacecraft finding the conditions for the mitigation of radiation components contributing to the doses beyond accepted limits. In order to reduce the overall space craft mass, NASA is looking for the novel, multi-purpose and multi-functional materials that will provide effective shielding of the crew and electronics on board. The Laboratory for Neutronics and Geometry Computation in the School of Nuclear Engineering at Purdue University led by Prof. Tatjana Jevremović began in 2004 the analytical evaluations of different lightweight materials. The preliminary results of the design survey study are presented in this paper.

  19. Gonad shielding in diagnostic radiology

    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

  20. Radiation shielding activities at IDOM

    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)

  1. Magnetic shielding of a limiter

    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

  2. Radiation shielding activities at IDOM

    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)

  3. Technical specifications: Tower Shielding Reactor II

    1979-02-01

    The technical specifications define the key limitations that must be observed for safe operation of the Tower Shielding Reactor II (TSR-II) and an envelope of operation within which there is reasonable assurance that these limits cannot be exceeded. The specifications were written to satisfy the requirements of the Department of Energy (DOE) Manual Chapter 0540, September 1, 1972

  4. Shielding experiments for accelerator facilities

    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)

  5. Shielding experiments for accelerator facilities

    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)

  6. Shielding of manned space stations against Van Allen Belt protons: a preliminary scoping study

    Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M.; Corbin, J.M.

    1986-09-01

    Calculated results are presented to aid in the design of the shielding required to protect astronauts in a space station that is orbiting through the Van Allen proton belt. The geometry considered - a spherical shell shield with a spherical tissue phantom at its center - is only a very approximate representation of an actual space station, but this simple geometry makes it possible to consider a wide range of possible shield materials. Both homogeneous and laminated shields are considered. Also, an approximation procedure - the equivalent thickness approximation - that allows dose rates to be estimated for any shield material or materials from the dose rates for an aluminum shield is presented and discussed

  7. Development of HANARO ST3 shield

    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

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

    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

  9. Shielding walls against ionizing radiation. Lead bricks

    1993-04-01

    The standard contains specifications for the shape and requirements set for lead bricks such that they can be used to construct radiation-shielding walls according to the building kit system. The dimensions of the bricks are selected in such a way as to permit any modification of the length, height and thickness of said shielding walls in units of 50 mm. The narrow side of the lead bricks juxtaposed to one another in a wall construction to shield against radiation have to form prismatic grooves and tongues: in this way, direct penetration by radiation is prevented. Only cuboid bricks (serial nos. 55-60 according to Table 10) do not have prismatic tongues and grooves. (orig.) [de

  10. External dosimetry sources and shielding

    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

  11. Radiation shielding calculation using MCNP

    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)

  12. Radiation shielding calculations for the vista spacecraft

    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

  13. Shielding benchmark test

    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)

  14. Neutron shielding material

    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.)

  15. Radiation shielding wall structure

    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. Multilayer radiation shield

    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.

  17. Light shielding apparatus

    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.

  18. Evaluation of the room shielding thickness of Hi-Art tomotherapy system

    Liu Haikuan; Wu Jinhai; Gu Naigu; Gao Yiming; Wang Li; Huang Weiqin; Wang Fengxian

    2010-01-01

    In this paper, we calculate and evaluate the room shielding thickness of a Hi-Art tomotherapy system, which is a new type of radiotherapy facility. Due to the self-shielding of the accelerator,only scattered beam and beam leakage were considered in calculating the room shielding thickness. The radiation field of the tomotherapy system was used as the basic data to calculate the shielding thickness of every 15 degree solid angle. The maximum shielding thickness required of each shielding wall was at the position with the angle of 15 degree, and the calculated shielding thickness were 1023, 975, 917, 1460, 1147 and 1189 mm for the east wall,south wall,west wall, north wall, the roof and the floor,respectively. According to the calculation results, all shielding walls, ceiling and floor could meet the requirement of the radiation protection, but the north wall thickness of 1200 mm was a little thinner. (authors)

  19. Shielding container for radioactive isotopes

    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

    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

    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. Task force's viewpoint of the current BRH Program. Gonad shield, Part F

    Dolezal, A.

    1975-01-01

    A previous report on radiation protection during diagnostic exposure is reviewed. Some topics discussed are: Task Force regulations for the model by which routine use of patient shielding or gonad shielding would be required; shielding in relation to collimation, scatter radiation, and enforcement of regulations; and proposed regulations for reduction of unnecessary radiation

  3. Muon shielding for PEP

    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

  4. Shielding calculations for NET

    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

  5. Laboratory-scale shielded cell for 252Cf

    Anderl, R.A.; Cargo, C.H.

    1979-01-01

    A shielded-cell facility for storing and handling remotely up to 2 milligram quantities of unencapsulated 252 Cf has been built in a radiochemistry laboratory at the Test Reactor Area of the Idaho National Engineering Laboratory. Unique features of this facility are its compact bulk radiation shield of borated gypsum and transfer lines which permit the transport of fission product activity from 252 Cf fission sources within the cell to a mass separator and to a fast radiochemistry system in nearby rooms

  6. The SNS target station preliminary Title I shielding analyses

    Johnson, J.O.; Santoro, R.T.; Lillie, R.A.; Barnes, J.M.; McNeilly, G.S.

    2000-01-01

    The Department of Energy (DOE) has given the Spallation Neutron Source (SNS) project approval to begin Title I design of the proposed facility to be built at Oak Ridge National Laboratory (ORNL). During the conceptual design phase of the SNS project, the target station bulk-biological shield was characterized and the activation of the major targets station components was calculated. Shielding requirements were assessed with respect to weight, space, and dose-rate constraints for operating, shut-down, and accident conditions utilizing the SNS shield design criteria, DOE Order 5480.25, and requirements specified in 10 CFR 835. Since completion of the conceptual design phase, there have been major design changes to the target station as a result of the initial shielding and activation analyses, modifications brought about due to engineering concerns, and feedback from numerous external review committees. These design changes have impacted the results of the conceptual design analyses, and consequently, have required a re-investigation of the new design. Furthermore, the conceptual design shielding analysis did not address many of the details associated with the engineering design of the target station. In this paper, some of the proposed SNS target station preliminary Title I shielding design analyses will be presented. The SNS facility (with emphasis on the target station), shielding design requirements, calculational strategy, and source terms used in the analyses will be described. Preliminary results and conclusions, along with recommendations for additional analyses, will also be presented. (author)

  7. Comprehensive analysis of shielding effectiveness for HDPE, BPE and concrete as candidate materials for neutron shielding

    Dhang, Prosenjit; Verma, Rishi; Shyam, Anurag

    2015-01-01

    In the compact accelerator based DD neutron generator, the deuterium ions generated by the ion source are accelerated after the extraction and bombarded to a deuterated titanium target. The emitted neutrons have typical energy of ∼2.45MeV. Utilization of these compact accelerator based neutron generators of yield up to 10 9 neutron/second (DD) is under active consideration in many research laboratories for conducting active neutron interrogation experiments. Requirement of an adequately shielded laboratory is mandatory for the effective and safe utilization of these generators for intended applications. In this reference, we report the comprehensive analysis of shielding effectiveness for High Density Polyethylene (HDPE), Borated Polyethylene (BPE) and Concrete as candidate materials for neutron shielding. In shielding calculations, neutron induced scattering and absorption gamma dose has also been considered along with neutron dose. Contemporarily any material with higher hydrogenous concentration is best suited for neutron shielding. Choice of shielding material is also dominated by practical issues like economic viability and availability of space. Our computational analysis results reveal that utilization of BPE sheets results in minimum wall thickness requirement for attaining similar range of attenuation in neutron and gamma dose. The added advantage of using borated polyethylene is that it reduces the effect of both neutron and gamma dose by absorbing neutron and producing lithium and alpha particle. It has also been realized that for deciding upon optimum thickness determination of any shielding material, three important factors to be necessarily considered are: use factor, occupancy factor and work load factor. (author)

  8. Shielding studies and LMFBR development achievements and future trends

    Salvatores, M.

    1990-01-01

    Shielding studies in the last decade have been performed in cooperation with several European countries. Shielding has become a mature discipline that takes advantage of improvements in data and methods and supplies the designer with a better set of tools to tackle much stricter requirements. The paper describes achievements to date and the Super Phenix start-up experiments. The present trends to design (a) reduced axial/radial shields, (b) cores that allow internal storage of irradiated sub-assemblies, and (c) cores with specific axial/radial peripheral core zone architectures to improve sodium void reactivity effects require further studies and experimental validation

  9. Shielding consideration for a deuteron activated liquid lithium system

    Huang, S.T.; Shapiro, A.M.; Lee, J.B.; Miller, W.C.

    1979-09-01

    A parametric study was conducted to evaluate the potential shielding implication due to the 7 Be plateout on the lithium piping in the FMIT facility. Various parameters such as plateout percentage, hot flush efficiency and 7 Be trapping efficiency were varied to assess the overall shielding requirement relationship. The 7 Be plateout was found to place severe limitations on the hands-on maintenance access. Hot flush and 7 Be traps are effective ways of minimizing the 7 Be plateout. To be effective in reducing local shielding requirements, the combined 7 Be trapping and hot flush efficiency shall be greater than 95%

  10. A contribution to shielding effectiveness analysis of shielded tents

    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.

  11. Characteristics simulation of wireless power transfer system considering shielding distance

    Lee, Yu Kyeong; Choi, Hyo Sang; Jung, Byung Ik; Jeong, In Sung [Chosun University, Gwangju (Korea, Republic of)

    2015-03-15

    Wireless power transfer technology is using the magnetic resonance recently drawing increased attention. It uses the resonance between transmitter and receiver coils to transfer power. Thus, it can improve the transfer distance and efficiency compared with the existing magnetic induction technique. The authors found from the previous study that the application of the superconductor coil to the magnetic resonance wireless power transfer system improved its efficiency. Its application to real life, however, requires the additional study on the effects of adjacent materials. In this study, the two resonance coils made by superconductor coils were used to aluminum and plastic shielding materials was placed between the coils. S-parameters were analyzed according to the position of the shielding material between the transmitter and receiver coils. As a result, the plastic of shielding material had no effect, but the aluminum of shielding material affected the wireless power transfer due to the shielding effectiveness.

  12. Characteristics simulation of wireless power transfer system considering shielding distance

    Lee, Yu Kyeong; Choi, Hyo Sang; Jung, Byung Ik; Jeong, In Sung

    2015-01-01

    Wireless power transfer technology is using the magnetic resonance recently drawing increased attention. It uses the resonance between transmitter and receiver coils to transfer power. Thus, it can improve the transfer distance and efficiency compared with the existing magnetic induction technique. The authors found from the previous study that the application of the superconductor coil to the magnetic resonance wireless power transfer system improved its efficiency. Its application to real life, however, requires the additional study on the effects of adjacent materials. In this study, the two resonance coils made by superconductor coils were used to aluminum and plastic shielding materials was placed between the coils. S-parameters were analyzed according to the position of the shielding material between the transmitter and receiver coils. As a result, the plastic of shielding material had no effect, but the aluminum of shielding material affected the wireless power transfer due to the shielding effectiveness

  13. Optimal selection for shielding materials by fuzzy linear programming

    Kanai, Y.; Miura, N.; Sugasawa, S.

    1996-01-01

    An application of fuzzy linear programming methods to optimization of a radiation shield is presented. The main purpose of the present study is the choice of materials and the search of the ratio of mixture-component as the first stage of the methodology on optimum shielding design according to individual requirements of nuclear reactor, reprocessing facility, shipping cask installing spent fuel, ect. The characteristic values for the shield optimization may be considered their cost, spatial space, weight and some shielding qualities such as activation rate and total dose rate for neutron and gamma ray (includes secondary gamma ray). This new approach can reduce huge combination calculations for conventional two-valued logic approaches to representative single shielding calculation by group-wised optimization parameters determined in advance. Using the fuzzy linear programming method, possibilities for reducing radiation effects attainable in optimal compositions hydrated, lead- and boron-contained materials are investigated

  14. Analytic flux formulas and tables of shielding functions

    Wallace, O.J.

    1981-06-01

    Hand calculations of radiation flux and dose rates are often useful in evaluating radiation shielding and in determining the scope of a problem. The flux formulas appropriate to such calculations are almost always based on the point kernel and allow for at most the consideration of laminar slab shields. These formulas often require access to tables of values of integral functions for effective use. Flux formulas and function tables appropriate to calculations involving homogeneous source regions with the shapes of lines, disks, slabs, truncated cones, cylinders, and spheres are presented. Slab shields may be included in most of these calculations, and the effect of a cylindrical shield surrounding a cylindrical source may be estimated. Detector points may be located axially, laterally, or interior to a cylindrical source. Line sources may be tilted with respect to a slab shield. All function tables are given for a wide range of arguments

  15. γ-ray shielding behaviors of some nuclear engineering materials

    Mann, Kulwinder Singh

    2017-01-01

    The essential requirement of a material to be used for engineering purposes at nuclear establishments is its ability to attenuate the most penetrating ionizing radiations, gamma (γ)-rays. Mostly, high-Z materials such as heavy concrete, lead, mercury, and their mixtures or alloys have been used in the construction of nuclear establishments and thus termed as nuclear engineering materials (NEM). The NEM are classified into two categories, namely opaque and transparent, depending on their behavior towards the visible spectrum of EM waves. The majority of NEM are opaque. By contrast, various types of glass, which are transparent to visible light, are necessary at certain places in the nuclear establishments. In the present study, γ-ray shielding behaviors (GSB) of six glass samples (transparent NEM) were evaluated and compared with some opaque NEM in a wide range of energy (15 keV–15 MeV) and optical thickness (OT). The study was performed by computing various γ-ray shielding parameters (GSP) such as the mass attenuation coefficient, equivalent atomic number, and buildup factor. A self-designed and validated computer-program, the buildup factor-tool, was used for various computations. It has been established that some glass samples show good GSB, thus can safely be used in the construction of nuclear establishments in conjunction with the opaque NEM as well

  16. γ-ray shielding behaviors of some nuclear engineering materials

    Mann, Kulwinder Singh [Dept. of Physics, D.A.V. College, Punjab (India)

    2017-06-15

    The essential requirement of a material to be used for engineering purposes at nuclear establishments is its ability to attenuate the most penetrating ionizing radiations, gamma (γ)-rays. Mostly, high-Z materials such as heavy concrete, lead, mercury, and their mixtures or alloys have been used in the construction of nuclear establishments and thus termed as nuclear engineering materials (NEM). The NEM are classified into two categories, namely opaque and transparent, depending on their behavior towards the visible spectrum of EM waves. The majority of NEM are opaque. By contrast, various types of glass, which are transparent to visible light, are necessary at certain places in the nuclear establishments. In the present study, γ-ray shielding behaviors (GSB) of six glass samples (transparent NEM) were evaluated and compared with some opaque NEM in a wide range of energy (15 keV–15 MeV) and optical thickness (OT). The study was performed by computing various γ-ray shielding parameters (GSP) such as the mass attenuation coefficient, equivalent atomic number, and buildup factor. A self-designed and validated computer-program, the buildup factor-tool, was used for various computations. It has been established that some glass samples show good GSB, thus can safely be used in the construction of nuclear establishments in conjunction with the opaque NEM as well.

  17. γ-Ray Shielding Behaviors of Some Nuclear Engineering Materials

    Kulwinder Singh Mann

    2017-06-01

    Full Text Available The essential requirement of a material to be used for engineering purposes at nuclear establishments is its ability to attenuate the most penetrating ionizing radiations, gamma (γ-rays. Mostly, high-Z materials such as heavy concrete, lead, mercury, and their mixtures or alloys have been used in the construction of nuclear establishments and thus termed as nuclear engineering materials (NEM. The NEM are classified into two categories, namely opaque and transparent, depending on their behavior towards the visible spectrum of EM waves. The majority of NEM are opaque. By contrast, various types of glass, which are transparent to visible light, are necessary at certain places in the nuclear establishments. In the present study, γ-ray shielding behaviors (GSB of six glass samples (transparent NEM were evaluated and compared with some opaque NEM in a wide range of energy (15 keV–15 MeV and optical thickness (OT. The study was performed by computing various γ-ray shielding parameters (GSP such as the mass attenuation coefficient, equivalent atomic number, and buildup factor. A self-designed and validated computer-program, the buildup factor-tool, was used for various computations. It has been established that some glass samples show good GSB, thus can safely be used in the construction of nuclear establishments in conjunction with the opaque NEM as well.

  18. Required momentum, heat, and mass transport experiments for liquid-metal blankets

    Tillack, M.S.; Sze, D.K.; Abdou, M.A.

    1986-01-01

    Through the effects on fluid flow, many aspects of blanket behavior are affected by magnetohydrodynamic (MHD) effects, including pressure drop, heat transfer, mass transfer, and structural behavior. In this paper, a set of experiments is examined that could be performed in order to reduce the uncertainties in the highly related set of issues dealing with momentum, heat, and mass transport under the influence of a strong magnetic field (i.e., magnetic transport phenomena). By improving our basic understanding and by providing direct experimental data on blanket behavior, these experiments will lead to improved designs and an accurate assessment of the attractiveness of liquid-metal blankets

  19. Efficiency of cement - based low - weight shielding materials for Cs-137 gamma rays

    Satty, H. E. M.

    2014-10-01

    Due to the development of nuclear technology and use of technologies in various field of industry, medicine and research against ionizing radiation is one of the most important topics in this field. The purpose of this work is to reduce the dose rate from radioactive sources. The exposure to gamma radiation is leading to several health effects as the result of absorption by the human body. The frequently used shielding material for gamma rays is lead. In spite of its effectiveness and high mass attenuation coefficient, lower weight gamma shielding materials are required. In this effectiveness of three materials: carbon and mixture (50% carbon + 50% cement) was studied and compared to that of lead. The results were obtained in terms of the variations of the transmitted intensity. This is done using a gamma spectroscopy system.(Author)

  20. Radiation shielding quality assurance

    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. Stability of a radioactive waste repository in the Canadian shield

    Mahtab, M.A.; Ratigan, J.L.; McCreath, D.R.

    1977-01-01

    A nonlinear finite element analysis is presented for a radioactive waste repository room assumed to be located at a depth of 1,000 meters in the Canadian Shield. The loading of the finite element model is both due to in situ stresses which exist prior to excavation and thermomechanical stresses arising from the radiogenic heat dissipation of the waste assumed to have a half life of 30 years and a gross thermal loading of 32 watts/m 2 . The influence of in situ stress, joint cohesion and joint friction angle on the isolation room stability and support requirements is examined for a simulated period of 30 years. For the range of in situ stress conditions, properties of the jointed rock mass, and the thermal loading considered, the extent of the rock failure is within the capability of conventional rock support measures

  2. EBT-P gamma-ray-shielding analysis

    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

  3. Discussions for the shielding materials of synchrotron radiation beamline hutches

    Asano, Y.

    2006-01-01

    Many synchrotron radiation facilities are now under operation such as E.S.R.F., APS, and S.P.ring-8. New facilities with intermediated stored electron energy are also under construction and designing such as D.I.A.M.O.N.D., S.O.L.E.I.L., and S.S.R.F.. At these third generation synchrotron radiation facilities, the beamline shielding as well as the bulk shield is very important for designing radiation safety because of intense and high energy synchrotron radiation beam. Some reasons employ lead shield wall for the synchrotron radiation beamlines. One is narrow space for the construction of many beamlines at the experimental hall, and the other is the necessary of many movable mechanisms at the beamlines, for examples. Some cases are required to shield high energy neutrons due to stored electron beam loss and photoneutrons due to gas Bremsstrahlung. Ordinary concrete and heavy concrete are coming up to shield material of synchrotron radiation beamline hutches. However, few discussions have been performed so far for the shielding materials of the hutches. In this presentation, therefore, we will discuss the characteristics of the shielding conditions including build up effect for the beamline hutches by using the ordinary concrete, heavy concrete, and lead for shielding materials with 3 GeV and 8 GeV class synchrotron radiation source. (author)

  4. Neutronic reactor thermal shield

    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

  5. A shield against distraction

    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...

  6. Measuring space radiation shielding effectiveness

    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 ...

  7. Tax Shield, Insolvenz und Zinsschranke

    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...

  8. Real-time viscosity and mass density sensors requiring microliter sample volume based on nanomechanical resonators.

    Bircher, Benjamin A; Duempelmann, Luc; Renggli, Kasper; Lang, Hans Peter; Gerber, Christoph; Bruns, Nico; Braun, Thomas

    2013-09-17

    A microcantilever based method for fluid viscosity and mass density measurements with high temporal resolution and microliter sample consumption is presented. Nanomechanical cantilever vibration is driven by photothermal excitation and detected by an optical beam deflection system using two laser beams of different wavelengths. The theoretical framework relating cantilever response to the viscosity and mass density of the surrounding fluid was extended to consider higher flexural modes vibrating at high Reynolds numbers. The performance of the developed sensor and extended theory was validated over a viscosity range of 1-20 mPa·s and a corresponding mass density range of 998-1176 kg/m(3) using reference fluids. Separating sample plugs from the carrier fluid by a two-phase configuration in combination with a microfluidic flow cell, allowed samples of 5 μL to be sequentially measured under continuous flow, opening the method to fast and reliable screening applications. To demonstrate the study of dynamic processes, the viscosity and mass density changes occurring during the free radical polymerization of acrylamide were monitored and compared to published data. Shear-thinning was observed in the viscosity data at higher flexural modes, which vibrate at elevated frequencies. Rheokinetic models allowed the monomer-to-polymer conversion to be tracked in spite of the shear-thinning behavior, and could be applied to study the kinetics of unknown processes.

  9. No Assembly Required: Mergers are Mostly Irrelevant for the Growth of Low-mass Dwarf Galaxies

    Fitts, Alex; Boylan-Kolchin, Michael; Bullock, James S.; Weisz, Daniel R.; El-Badry, Kareem; Wheeler, Coral; Faucher-Giguère, Claude-André; Quataert, Eliot; Hopkins, Philip F.; Kereš, Dušan; Wetzel, Andrew; Hayward, Christopher C.

    2018-06-01

    We investigate the merger histories of isolated dwarf galaxies based on a suite of 15 high-resolution cosmological zoom-in simulations, all with masses of Mhalo ≈ 1010 M⊙ (and M⋆ ˜ 105 - 107 M⊙) at z = 0, from the Feedback in Realistic Environments (FIRE) project. The stellar populations of these dwarf galaxies at z = 0 are formed essentially entirely "in situ": over 90% of the stellar mass is formed in the main progenitor in all but two cases, and all 15 of the galaxies have >70% of their stellar mass formed in situ. Virtually all galaxy mergers occur prior to z ˜ 3, meaning that accreted stellar populations are ancient. On average, our simulated dwarfs undergo 5 galaxy mergers in their lifetimes, with typical pre-merger galaxy mass ratios that are less than 1:10. This merger frequency is generally comparable to what has been found in dissipationless simulations when coupled with abundance matching. Two of the simulated dwarfs have a luminous satellite companion at z = 0. These ultra-faint dwarfs lie at or below current detectability thresholds but are intriguing targets for next-generation facilities. The small contribution of accreted stars make it extremely difficult to discern the effects of mergers in the vast majority of dwarfs either photometrically or using resolved-star color-magnitude diagrams (CMDs). The important implication for near-field cosmology is that star formation histories of comparably massive galaxies derived from resolved CMDs should trace the build-up of stellar mass in one main system across cosmic time as opposed to reflecting the contributions of many individual star formation histories of merged dwarfs.

  10. SHIELD verification and validation report

    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

  11. Multifunctional Hot Structure Heat Shield

    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...

  12. Radiation shield for nuclear reactors

    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

  13. Shielding consideration for the SSCL experimental halls

    Bull, J.; Coyne, J.; Mokhov, N.; Stapleton, G.

    1994-03-01

    The Superconducting Super Collider which is being designed and built in Waxahachie, Texas consists Of series of proton accelerators, culminating in a 20 Te proton on proton collider. The collider will be in a tunnel which will be 87 km in circumference and. on average about 30 meters underground. The present design calls for two large interaction halls on the east side of the ring. The shielding for these halls is being designed for an interaction rate of 10 9 Hz or 10 16 interactions per year, based on 10 7 seconds per operational year. SSC guidelines require that the shielding be designed to meet the criterion of 1mSv per year for open areas off site 2mSv per year for open areas on site, and 2mSv per year for controlled areas. Only radiation workers will be routinely allowed to work in controlled areas. It should be pointed that there is a potential for an accidental full beam loss in either of the experimental halls, and this event would consist of the loss of the full circulating beam up to 4 x 10 14 protons. With the present design. the calculated dose equivalent for this event is about 10% of the annual dose equivalent for the normal p-p interactions, so that die accident condition does not control the shielding. If, for instance, local shielding within the experimental hall is introduced into the calculations, this could change. The shielding requirements presented here are controlled by the normal p-p interactions. Three important questions were addressed in the present calculations. They are (1) the thickness of the roof over the experimental halls, (2) the configuration of the shafts and adits which give access to the halls, and (3) the problem of ground water and air activation

  14. New Toroid shielding design

    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 ...

  15. SHIELDS Final Technical Report

    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.

  16. Radiation shielding analysis

    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)

  17. MPACT Subgroup Self-Shielding Efficiency Improvements

    Stimpson, Shane; Liu, Yuxuan; Collins, Benjamin S.; Clarno, Kevin T.

    2016-01-01

    Recent developments to improve the efficiency of the MOC solvers in MPACT have yielded effective kernels that loop over several energy groups at once, rather that looping over one group at a time. These kernels have produced roughly a 2x speedup on the MOC sweeping time during eigenvalue calculation. However, the self-shielding subgroup calculation had not been reevaluated to take advantage of these new kernels, which typically requires substantial solve time. The improvements covered in this report start by integrating the multigroup kernel concepts into the subgroup calculation, which are then used as the basis for further extensions. The next improvement that is covered is what is currently being termed as ''Lumped Parameter MOC''. Because the subgroup calculation is a purely fixed source problem and multiple sweeps are performed only to update the boundary angular fluxes, the sweep procedure can be condensed to allow for the instantaneous propagation of the flux across a spatial domain, without the need to sweep along all segments in a ray. Once the boundary angular fluxes are considered to be converged, an additional sweep that will tally the scalar flux is completed. The last improvement that is investigated is the possible reduction of the number of azimuthal angles per octant in the shielding sweep. Typically 16 azimuthal angles per octant are used for self-shielding and eigenvalue calculations, but it is possible that the self-shielding sweeps are less sensitive to the number of angles than the full eigenvalue calculation.

  18. AP1000 shield building: a constructability challenge

    Di Giuseppe, Giovanni; Bonanno, Domenico

    2010-01-01

    The AP1000 Shield Building, an enhanced structure which surrounds the containment vessel, consists of standard Reinforced Concrete (RC) and composite Steel and Concrete (SC) construction. In the SC module the surface steel plates, (with attached shear studs and angles) filled with concrete, act as the steel reinforcement in concrete. This is a relatively new design technology that required the appropriate use of structural codes, supplemented with information from applicable tests on similar composite steel and concrete construction. Being a newer design concept, existing codes do not provide explicit guidance on SC construction so a review of literature and test data on composite structures similar to AP1000 shield building was done in order to confirm the technical basis for the design. The SC walls, air inlet structure and roof of the Shield Building will be constructed using modular construction practices and then transported to site and lifted into place. These modules, working also as permanent form-work, will be filled with high strength Self- Consolidating Concrete. (SCC) This paper provides a focused and integrated presentation of the enhanced shield building design methodology, testing, constructability and inspection. (authors)

  19. Methods for calculating radiation attenuation in shields

    Butler, J; Bueneman, D; Etemad, A; Lafore, P; Moncassoli, A M; Penkuhn, H; Shindo, M; Stoces, B

    1964-10-01

    In recent years the development of high-speed digital computers of large capacity has revolutionized the field of reactor shield design. For compact special-purpose reactor shields, Monte-Carlo codes in two- and three dimensional geometries are now available for the proper treatment of both the neutron and gamma- ray problems. Furthermore, techniques are being developed for the theoretical optimization of minimum-weight shield configurations for this type of reactor system. In the design of land-based power reactors, on the other hand, there is a strong incentive to reduce the capital cost of the plant, and economic considerations are also relevant to reactors designed for merchant ship propulsion. In this context simple methods are needed which are economic in their data input and computing time requirements and which, at the same time, are sufficiently accurate for design work. In general the computing time required for Monte-Carlo calculations in complex geometry is excessive for routine design calculations and the capacity of the present codes is inadequate for the proper treatment of large reactor shield systems in three dimensions. In these circumstances a wide range of simpler techniques are currently being employed for design calculations. The methods of calculation for neutrons in reactor shields fall naturally into four categories: Multigroup diffusion theory; Multigroup diffusion with removal sources; Transport codes; and Monte Carlo methods. In spite of the numerous Monte- Carlo techniques which are available for penetration and back scattering, serious problems are still encountered in practice with the scattering of gamma rays from walls of buildings which contain critical facilities and also concrete-lined discharge shafts containing irradiated fuel elements. The considerable volume of data in the unclassified literature on the solution of problems of this type in civil defence work appears not to have been evaluated for reactor shield design. In

  20. Thick Galactic Cosmic Radiation Shielding Using Atmospheric Data

    Youngquist, Robert C.; Nurge, Mark A.; Starr, Stanley O.; Koontz, Steven L.

    2013-01-01

    NASA is concerned with protecting astronauts from the effects of galactic cosmic radiation and has expended substantial effort in the development of computer models to predict the shielding obtained from various materials. However, these models were only developed for shields up to about 120 g!cm2 in thickness and have predicted that shields of this thickness are insufficient to provide adequate protection for extended deep space flights. Consequently, effort is underway to extend the range of these models to thicker shields and experimental data is required to help confirm the resulting code. In this paper empirically obtained effective dose measurements from aircraft flights in the atmosphere are used to obtain the radiation shielding function of the earth's atmosphere, a very thick shield. Obtaining this result required solving an inverse problem and the method for solving it is presented. The results are shown to be in agreement with current code in the ranges where they overlap. These results are then checked and used to predict the radiation dosage under thick shields such as planetary regolith and the atmosphere of Venus.

  1. FORMATION OF CLOSE IN SUPER-EARTHS AND MINI-NEPTUNES: REQUIRED DISK MASSES AND THEIR IMPLICATIONS

    Schlichting, Hilke E., E-mail: hilke@mit.edu [Massachusetts Institute of Technology, 77 Massachusetts Avenue, Cambridge, MA 02139-4307 (United States)

    2014-11-01

    Recent observations by the Kepler space telescope have led to the discovery of more than 4000 exoplanet candidates consisting of many systems with Earth- to Neptune-sized objects that reside well inside the orbit of Mercury around their respective host stars. How and where these close-in planets formed is one of the major unanswered questions in planet formation. Here, we calculate the required disk masses for in situ formation of the Kepler planets. We find that if close-in planets formed as isolation masses, then standard gas-to-dust ratios yield corresponding gas disks that are gravitationally unstable for a significant fraction of systems, ruling out such a scenario. We show that the maximum width of a planet's accretion region in the absence of any migration is 2v {sub esc}/Ω, where v {sub esc} is the escape velocity of the planet and Ω is the Keplerian frequency, and we use it to calculate the required disk masses for in situ formation with giant impacts. Even with giant impacts, formation without migration requires disk surface densities in solids at semi-major axes of less than 0.1 AU of 10{sup 3}-10{sup 5} g cm{sup –2}, implying typical enhancements above the minimum-mass solar nebular (MMSN) by at least a factor of 20. Corresponding gas disks are below but not far from the gravitational stability limit. In contrast, formation beyond a few AU is consistent with MMSN disk masses. This suggests that the migration of either solids or fully assembled planets is likely to have played a major role in the formation of close-in super-Earths and mini-Neptunes.

  2. Thermoforming plastic in lead shield construction

    Abrahams, M.E.; Chow, C.H.; Loyd, M.D.

    1989-01-01

    Radiation treatments using low energy X-rays or electrons frequently require a final field defining shield to be placed on the patient's skin. A custom made lead cut-out is used to provide a close fit to a particular patient's surface contours. We have developed a procedure which utilizes POLYFORM thermoplastic to obtain a negative mold of the patient instead of the traditional plaster bandage or dental impression gel. The Polyform is softened in warm water, molded carefully over the patient's surface, and is removed when set or hardened, usually within five minutes. Then lead sheet cut-outs can be formed within this negative. For shielding cut-outs requiring thicker lead sheet, a positive is made from dental stone using this Polyform negative. We have found this procedure to be neat, fast and comfortable for both patient and the dosimetrist

  3. Operating manual for the Tower Shielding Facility

    1985-12-01

    This manual provides information necessary to operate and perform maintenance on the reactor systems and all equipment or systems which can affect their operation or the safety of personnel at the Tower Shielding Facility. The first four chapters consist of introductory and descriptive material of benefit to personnel in training, the qualifications required for training, the responsibilities of the personnel in the organization, and the procedures for reviewing proposed experiments. Chapter 8, Emergency Procedures, is also a necessary part of the indoctrination of personnel. The procedures for operation of the Tower Shielding Reactor (TSR-II), its water cooling system, and the main tower hoists are outlined in Chapters 5, 6, and 7. The Technical Specification surveillance requirements for the TSR-II are summarized in Chapter 9. The maintenance and calibration schedule is spelled out in Chapter 10. The procedures for assembly and disassembly of the TSR-II are outlined in Chapter 11

  4. Grounding and shielding circuits and interference

    Morrison, Ralph

    2016-01-01

    Applies basic field behavior in circuit design and demonstrates how it relates to grounding and shielding requirements and techniques in circuit design This book connects the fundamentals of electromagnetic theory to the problems of interference in all types of electronic design. The text covers power distribution in facilities, mixing of analog and digital circuitry, circuit board layout at high clock rates, and meeting radiation and susceptibility standards. The author examines the grounding and shielding requirements and techniques in circuit design and applies basic physics to circuit behavior. The sixth edition of this book has been updated with new material added throughout the chapters where appropriate. The presentation of the book has also been rearranged in order to reflect the current trends in the field.

  5. E-Alerts: Nuclear science and technology (radiation shielding, protection, and safety). E-mail newsletter

    1999-01-01

    Topics include: Shielding design, nuclear radiation transport properties of materials, decontamination; Container design and transportation requirements for radioactive materials; and Fallout shelters

  6. Detector Background Reduction by Passive and Active Shielding

    Bikit, I.; Bikit, K.; Forkapic, S.; Mrda, D.; Nikolov, J.; Slivka, J.; Todorovic, N.

    2013-01-01

    The operational problems of the gamma ray spectrometer shielded passively with 12 cm of lead and actively by five 0.5 m × 0.5 m × 0.05 m plastic veto shields are described. The active shielding effect from both environmental gamma ray, cosmic muons and neutrons was investigated. For anticoincidence gating wide range of scintillator pulses, corresponding to the energy range of 150 keV-75 MeV, were used. With the optimal set up the integral background, for the energy region of 50 - 3000 keV, of 0.31 c/s was achieved. The detector mass related background was 0.345 c/(kg s). The 511 keV annihilation line was reduced by the factor of 7 by the anticoincidence gate. It is shown that the plastic shields increase the neutron capture gamma line intensities due to neutron termalization.(author)

  7. Shielding and maintainability in an experimental tokamak

    Abdou, M.A.; Fuller, G.; Hager, E.R.; Vogelsang, W.F.

    1979-01-01

    This paper presents the results of an attempt to develop an understanding of the various factors involved. This work was performed as a part of the task assigned to one of the expert groups on the International Tokamak Reactor (INTOR). The shielding penalties for requiring personnel access are quantified. This is followed by a quantitative estimate of the benefits associated with personnel access. The penalties to the benefits and conclusions and recommendations on resolving the issue are discussed

  8. SINBAD: Shielding integral benchmark archive and database

    Hunter, H.T.; Ingersoll, D.T.; Roussin, R.W.

    1996-01-01

    SINBAD is a new electronic database developed to store a variety of radiation shielding benchmark data so that users can easily retrieve and incorporate the data into their calculations. SINBAD is an excellent data source for users who require the quality assurance necessary in developing cross-section libraries or radiation transport codes. The future needs of the scientific community are best served by the electronic database format of SINBAD and its user-friendly interface, combined with its data accuracy and integrity

  9. Radiation shield for PWR reactors

    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.)

  10. Gonad shielding in diagnostic radiology

    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

  11. SHIELD 1.0: development of a shielding calculator program in diagnostic radiology; SHIELD 1.0: desenvolvimento de um programa de calculo de blindagem em radiodiagnostico

    Santos, Romulo R.; Real, Jessica V.; Luz, Renata M. da [Hospital Sao Lucas (PUCRS), Porto Alegre, RS (Brazil); Friedrich, Barbara Q.; Silva, Ana Maria Marques da, E-mail: ana.marques@pucrs.br [Pontificia Universidade Catolica do Rio Grande do Sul (PUCRS), Porto Alegre, RS (Brazil)

    2013-08-15

    In shielding calculation of radiological facilities, several parameters are required, such as occupancy, use factor, number of patients, source-barrier distance, area type (controlled and uncontrolled), radiation (primary or secondary) and material used in the barrier. The shielding design optimization requires a review of several options about the physical facility design and, mainly, the achievement of the best cost-benefit relationship for the shielding material. To facilitate the development of this kind of design, a program to calculate the shielding in diagnostic radiology was implemented, based on data and limits established by National Council on Radiation Protection and Measurements (NCRP) 147 and SVS-MS 453/98. The program was developed in C⌗ language, and presents a graphical interface for user data input and reporting capabilities. The module initially implemented, called SHIELD 1.0, refers to calculating barriers for conventional X-ray rooms. The program validation was performed by the comparison with the results of examples of shielding calculations presented in NCRP 147.

  12. Electromagnetic analysis of the Faraday shield of the EAST ICRF antenna

    Yang Qingxi; Song Yuntao; Wu Songtao; Zhao Yanping

    2011-01-01

    Faraday shield is one of the important components of ICRF antenna for EAST. In view of the structural safety of the Faraday shield, the electromagnetic and structural analyses for the Faraday shield have been carried out by applying the finite element method and the formulas under the cases of plasma disruption and vertical displacement event (VDE). Results of the electromagnetic forces, the stresses distribution as well as the deformation in the Faraday shield have been obtained under the two cases. They meet the design requirements and provide the theoretical basis for the structural safety evaluation of the Faraday shield. (authors)

  13. The Imperial Shield

    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...

  14. Shielded Canister Transporter

    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

  15. Welding shield for coupling heaters

    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.

  16. Penetration shielding applications of CYLSEC

    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

  17. Shield calculations, optimization vs. paradigm

    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)

  18. Parameters calculation of shielding experiment

    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

  19. Histone deacetylase 3 is required for maintenance of bone mass during aging

    McGee-Lawrence, Meghan E.; Bradley, Elizabeth W.; Dudakovic, Amel; Carlson, Samuel W.; Ryan, Zachary C.; Kumar, Rajiv; Dadsetan, Mahrokh; Yaszemski, Michael J.; Chen, Qingshan; An, Kai-Nan; Westendorf, Jennifer J.

    2012-01-01

    Histone deacetylase 3 (Hdac3) is a nuclear enzyme that removes acetyl groups from lysine residues in histones and other proteins to epigenetically regulate gene expression. Hdac3 interacts with bone-related transcription factors and co-factors such as Runx2 and Zfp521, and thus is poised to play a key role in the skeletal system. To understand the role of Hdac3 in osteoblasts and osteocytes, Hdac3 conditional knockout (CKO) mice were created with the Osteocalcin (OCN) promoter driving Cre expression. Hdac3 CKOOCN mice were of normal size and weight, but progressively lost trabecular and cortical bone mass with age. The Hdac3 CKOOCN mice exhibited reduced cortical bone mineralization and material properties and suffered frequent fractures. Bone resorption was lower, not higher, in the Hdac3 CKOOCN mice, suggesting that primary defects in osteoblasts caused the reduced bone mass. Indeed, reductions in bone formation were observed. Osteoblasts and osteocytes from Hdac3 CKOOCN mice showed increased DNA damage and reduced functional activity in vivo and in vitro. Thus, Hdac3 expression in osteoblasts and osteocytes is essential for bone maintenance during aging. PMID:23085085

  20. Calcium and vitamin D requirements for optimal bone mass during adolescence

    There remains very strong interest in the calcium and vitamin D requirements of adolescents related to bone health. The Institute of Medicine (IOM) released new dietary guidelines in late 2010 for these nutrients. These guidelines were primarily based on literature published in 2009 and earlier and ...

  1. Radiation shielding aspects for long manned mission to space: Criteria, survey study, and preliminary model

    Sztejnberg Manuel; Xiao Shanjie; Satvat Nader; Limón Felisa; Hopkins John; Jevremović Tatjana

    2006-01-01

    The prospect of manned space missions outside Earth's orbit is limited by the travel time and shielding against cosmic radiation. The chemical rockets currently used in the space program have no hope of propelling a manned vehicle to a far away location such as Mars due to the enormous mass of fuel that would be required. The specific energy available from nuclear fuel is a factor of 106 higher than chemical fuel; it is therefore obvious that nuclear power production in space is a must. On th...

  2. Analysis methods for Kevlar shield response to rotor fragments

    Gerstle, J. H.

    1977-01-01

    Several empirical and analytical approaches to rotor burst shield sizing are compared and principal differences in metal and fabric dynamic behavior are discussed. The application of transient structural response computer programs to predict Kevlar containment limits is described. For preliminary shield sizing, present analytical methods are useful if insufficient test data for empirical modeling are available. To provide other information useful for engineering design, analytical methods require further developments in material characterization, failure criteria, loads definition, and post-impact fragment trajectory prediction.

  3. Multihelix rotating shield brachytherapy for cervical cancer

    Dadkhah, Hossein [Department of Biomedical Engineering, University of Iowa, 1402 Seamans Center for the Engineering Arts and Sciences, Iowa City, Iowa 52242 (United States); Kim, Yusung; Flynn, Ryan T., E-mail: ryan-flynn@uiowa.edu [Department of Radiation Oncology, University of Iowa, 200 Hawkins Drive, Iowa City, Iowa 52242 (United States); Wu, Xiaodong [Department of Radiation Oncology, University of Iowa, 200 Hawkins Drive, Iowa City, Iowa 52242 and Department of Electrical and Computer Engineering, University of Iowa, 4016 Seamans Center for the Engineering Arts and Sciences, Iowa City, Iowa 52242 (United States)

    2015-11-15

    Purpose: To present a novel brachytherapy technique, called multihelix rotating shield brachytherapy (H-RSBT), for the precise angular and linear positioning of a partial shield in a curved applicator. H-RSBT mechanically enables the dose delivery using only linear translational motion of the radiation source/shield combination. The previously proposed approach of serial rotating shield brachytherapy (S-RSBT), in which the partial shield is rotated to several angular positions at each source dwell position [W. Yang et al., “Rotating-shield brachytherapy for cervical cancer,” Phys. Med. Biol. 58, 3931–3941 (2013)], is mechanically challenging to implement in a curved applicator, and H-RSBT is proposed as a feasible solution. Methods: A Henschke-type applicator, designed for an electronic brachytherapy source (Xoft Axxent™) and a 0.5 mm thick tungsten partial shield with 180° or 45° azimuthal emission angles and 116° asymmetric zenith angle, is proposed. The interior wall of the applicator contains six evenly spaced helical keyways that rigidly define the emission direction of the partial radiation shield as a function of depth in the applicator. The shield contains three uniformly distributed protruding keys on its exterior wall and is attached to the source such that it rotates freely, thus longitudinal translational motion of the source is transferred to rotational motion of the shield. S-RSBT and H-RSBT treatment plans with 180° and 45° azimuthal emission angles were generated for five cervical cancer patients with a diverse range of high-risk target volume (HR-CTV) shapes and applicator positions. For each patient, the total number of emission angles was held nearly constant for S-RSBT and H-RSBT by using dwell positions separated by 5 and 1.7 mm, respectively, and emission directions separated by 22.5° and 60°, respectively. Treatment delivery time and tumor coverage (D{sub 90} of HR-CTV) were the two metrics used as the basis for evaluation and

  4. Transparent fast neutron shielding material and shielding method

    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.)

  5. Nuclear shielding of openings in ITER Tokamak building

    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.

  6. GRAVITATIONAL FIELD SHIELDING AND SUPERNOVA EXPLOSIONS

    Zhang, T. X.

    2010-01-01

    A new mechanism for supernova explosions called gravitational field shielding is proposed, in accord with a five-dimensional fully covariant Kaluza-Klein theory with a scalar field that unifies the four-dimensional Einsteinian general relativity and Maxwellian electromagnetic theory. It is shown that a dense compact collapsing core of a star will suddenly turn off or completely shield its gravitational field when the core collapses to a critical density, which is inversely proportional to the square of mass of the core. As the core suddenly turns off its gravity, the extremely large pressure immediately stops the core collapse and pushes the mantle material of supernova moving outward. The work done by the pressure in the expansion can be the order of energy released in a supernova explosion. The gravity will resume and stop the core from a further expansion when the core density becomes less than the critical density. Therefore, the gravitational field shielding leads a supernova to impulsively explode and form a compact object such as a neutron star as a remnant. It works such that a compressed spring will shoot the oscillator out when the compressed force is suddenly removed.

  7. DID FOMALHAUT, HR 8799, AND HL TAURI FORM PLANETS VIA THE GRAVITATIONAL INSTABILITY? PLACING LIMITS ON THE REQUIRED DISK MASSES

    Nero, D.; Bjorkman, J. E.

    2009-01-01

    Disk fragmentation resulting from the gravitational instability has been proposed as an efficient mechanism for forming giant planets. We use the planet Fomalhaut b, the triple-planetary system HR 8799, and the potential protoplanet associated with HL Tau to test the viability of this mechanism. We choose the above systems since they harbor planets with masses and orbital characteristics favored by the fragmentation mechanism. We do not claim that these planets must have formed as the result of fragmentation, rather the reverse: if planets can form from disk fragmentation, then these systems are consistent with what we should expect to see. We use the orbital characteristics of these recently discovered planets, along with a new technique to more accurately determine the disk cooling times, to place both lower and upper limits on the disk surface density-and thus mass-required to form these objects by disk fragmentation. Our cooling times are over an order of magnitude shorter than those of Rafikov, which makes disk fragmentation more feasible for these objects. We find that the required mass interior to the planet's orbital radius is ∼0.1 M sun for Fomalhaut b, the protoplanet orbiting HL Tau, and the outermost planet of HR 8799. The two inner planets of HR 8799 probably could not have formed in situ by disk fragmentation.

  8. Involution Requirement on a Boundary Makes Massless Fermions Compactified on a Finite Flat Disk Mass Protected

    Mankoč Borštnik, N. S.; Nielsen, H. B.

    2006-12-01

    The genuine Kaluza-Klein-like theories--with no fields in addition to gravity--have difficulties with the existence of massless spinors after the compactification of some space dimensions \\cite{witten}. We proposed (Phys. Lett. B 633 (2006)771) such a boundary condition for spinors in 1+5 compactified on a flat disk that ensures masslessness of spinors in d=1+3 as well as their chiral coupling to the corresponding background gauge field (which solves equations of motion for a free field linear in the Riemann curvature). In this paper we study the same toy model: M^{(1+3)} x M^{(2)}, looking this time for an involution which transforms a space of solutions of Weyl equations in d=1+5 from the outside of the flat disk in x^5 and x^6 into its inside, allowing massless spinor of only one handedness--and accordingly assures mass protection--and of one charge--1/2--and infinitely many massive spinors of the same charge, chirally coupled to the corresponding background gauge field. We reformulate the operator of momentum so that it is Hermitean on the vector space of spinor states obeying the involution boundary condition.

  9. Morphometry of terrestrial shield volcanoes

    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.

  10. Cerrobend shielding stents for buccal carcinoma patients

    Karma Yangchen

    2016-01-01

    Full Text Available Buccal carcinoma is one of the most common oral malignant neoplasms, especially in the South Asian region. Radiotherapy, which plays a significant role in the treatment of this carcinoma, has severe adverse effects. Different types of prosthesis may be constructed to protect healthy tissues from the adverse effects of treatment and concentrate radiation in the region of the tumor mass. However, the technique for fabrication of shielding stent with Lipowitz's alloy (cerrobend/Wood's alloy has not been well documented. This article describes detailed technique for fabrication of such a stent for unilateral buccal carcinoma patients to spare the unaffected oral cavity from potential harmful effects associated with radiotherapy.

  11. A study of gamma shielding

    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

  12. Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

    Ersez Tunay

    2017-01-01

    Full Text Available The shielding for the neutron high-resolution backscattering spectrometer (EMU located at the OPAL reactor (ANSTO was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

  13. Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

    Ersez, Tunay; Esposto, Fernando; Souza, Nicolas R. de

    2017-09-01

    The shielding for the neutron high-resolution backscattering spectrometer (EMU) located at the OPAL reactor (ANSTO) was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

  14. APPLICATION OF SOFT COMPUTING TECHNIQUES FOR PREDICTING COOLING TIME REQUIRED DROPPING INITIAL TEMPERATURE OF MASS CONCRETE

    Santosh Bhattarai

    2017-07-01

    Full Text Available Minimizing the thermal cracks in mass concrete at an early age can be achieved by removing the hydration heat as quickly as possible within initial cooling period before the next lift is placed. Recognizing the time needed to remove hydration heat within initial cooling period helps to take an effective and efficient decision on temperature control plan in advance. Thermal properties of concrete, water cooling parameters and construction parameter are the most influencing factors involved in the process and the relationship between these parameters are non-linear in a pattern, complicated and not understood well. Some attempts had been made to understand and formulate the relationship taking account of thermal properties of concrete and cooling water parameters. Thus, in this study, an effort have been made to formulate the relationship for the same taking account of thermal properties of concrete, water cooling parameters and construction parameter, with the help of two soft computing techniques namely: Genetic programming (GP software “Eureqa” and Artificial Neural Network (ANN. Relationships were developed from the data available from recently constructed high concrete double curvature arch dam. The value of R for the relationship between the predicted and real cooling time from GP and ANN model is 0.8822 and 0.9146 respectively. Relative impact on target parameter due to input parameters was evaluated through sensitivity analysis and the results reveal that, construction parameter influence the target parameter significantly. Furthermore, during the testing phase of proposed models with an independent set of data, the absolute and relative errors were significantly low, which indicates the prediction power of the employed soft computing techniques deemed satisfactory as compared to the measured data.

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

    Bronsch, T.

    1977-01-01

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

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

    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)

  17. Self-shielding factors

    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

  18. Shielding plug device

    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.)

  19. Is mass drug administration against lymphatic filariasis required in urban settings? The experience in Kano, Nigeria.

    Pam, Dung D; de Souza, Dziedzom K; D'Souza, Susan; Opoku, Millicent; Sanda, Safiya; Nazaradden, Ibrahim; Anagbogu, Ifeoma N; Okoronkwo, Chukwu; Davies, Emmanuel; Elhassan, Elisabeth; Molyneux, David H; Bockarie, Moses J; Koudou, Benjamin G

    2017-10-01

    The Global Programme to Eliminate Lymphatic Filariasis (GPELF), launched in 2000, has the target of eliminating the disease as a public health problem by the year 2020. The strategy adopted is mass drug administration (MDA) to all eligible individuals in endemic communities and the implementation of measures to reduce the morbidity of those suffering from chronic disease. Success has been recorded in many rural endemic communities in which elimination efforts have centered. However, implementation has been challenging in several urban African cities. The large cities of West Africa, exemplified in Nigeria in Kano are challenging for LF elimination program because reaching 65% therapeutic coverage during MDA is difficult. There is therefore a need to define a strategy which could complement MDA. Thus, in Kano State, Nigeria, while LF MDA had reached 33 of the 44 Local Government Areas (LGAs) there remained eleven 'urban' LGAs which had not been covered by MDA. Given the challenges of achieving at least 65% coverage during MDA implementation over several years in order to achieve elimination, it may be challenging to eliminate LF in such settings. In order to plan the LF control activities, this study was undertaken to confirm the LF infection prevalence in the human and mosquito populations in three urban LGAs. The prevalence of circulating filarial antigen (CFA) of Wuchereria bancrofti was assessed by an immuno-chromatography test (ICT) in 981 people in three urban LGAs of Kano state, Nigeria. Mosquitoes were collected over a period of 4 months from May to August 2015 using exit traps, gravid traps and pyrethrum knock-down spray sheet collections (PSC) in different households. A proportion of mosquitoes were analyzed for W. bancrofti, using dissection, loop-mediated isothermal amplification (LAMP) assay and conventional polymerase chain reaction (PCR). The results showed that none of the 981 subjects (constituted of West Africa may not be of significant importance

  20. Design of auxiliary shield for remote controlled metallographic microscope

    Matsui, Hiroki; Okamoto, Hisato

    2014-06-01

    The remote controlled optical microscope installed in the lead cell at the Reactor Fuel Examination Facility (RFEF) in Japan Atomic Energy Agency (JAEA) has been upgraded to a higher performance unit to study the effect of the microstructural evolution in clad material on the high burn-up fuel behavior under the accident condition. The optical pass of the new microscope requires a new through hole in the shielding lead wall of the cell. To meet safety regulations, auxiliary lead shieldings were designed to cover the lost shielding function of the cell wall. Particle and Heavy Ion Transport Code System (PHITS) was used to calculate and determine the shape and setting positions of the shielding unit. Seismic assessments of the unit were also performed. (author)

  1. Shielding assessment of the ETRR-1 Reactor Under power upgrading

    Ahmad, E E [Reactor Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    The assessment of existing shielding of the ETRR-1 reactor in case of power upgrading is presented and discussed. It was carried out using both the present EK-10 type fuel elements and some other types of fuel elements with different enrichments. The shielding requirements for the ETRR-1 when power is upgraded are also discussed. The optimization curves between the upgraded reactor power and the shield thickness are presented. The calculation have been made using the ANISN code with the DLC-75 data library. The results showed that the present shield necessitates an additional layer of steel with thickness of 10.20 and 25 cm. When its power is upgraded to 3, 6 and 10 MWt in order to cutoff all neutron energy groups to be adequately safe under normal operating conditions. 4 figs.

  2. Anode and cathode geometry and shielding gas interdependence in GTAW

    Key, J.F.

    1979-01-01

    Parametric analyses and high-speed photography of the interdependence of electrode (cathode) tip geometry, shielding gas composition, and groove (anode) geometry indicate that spot-on-plate tests show that blunt cathode shapes have penetration effects similar to addition of a high ionization potential inert gas (such as helium) to the argon shielding gas. Electrode shape and shielding gas composition effects are not synergistic. The time required to develop a given penetration is a function of anode and cathode geometry and shielding gas composition, in addition to other essential welding variables. Spot-on-plate tests are a valid analysis of radical pulsed GTAW. Bead-on-plate tests are a valid analysis of mild pulsed or constant current GTAW

  3. New gadolinium based glasses for gamma-rays shielding materials

    Kaewjang, S.; Maghanemi, U.; Kothan, S.; Kim, H.J.; Limkitjaroenporn, P.; Kaewkhao, J.

    2014-01-01

    Highlights: • Gd 2 O 3 based glasses have been fabricated and investigated radiation shielding properties between 223 and 662 keV. • Density of the glass increases with increasing of Gd 2 O 3. • All the glasses of Gd 2 O 3 compositions studied had been shown lower HVL than X-rays shielding window. • Prepared glasses to be utilized as radiation shielding material with Pb-free advantage. • This work is the first to reports on radiation shielding properties of Gd 2 O 3 based glass matrices. - Abstract: In this work, Gd 2 O 3 based glasses in compositions (80−x)B 2 O 3 -10SiO 2 -10CaO-xGd 2 O 3 (where x = 15, 20, 25, 30 and 35 mol%) have been fabricated and investigated for their radiation shielding, physical and optical properties. The density of the glass was found to increase with the increasing of Gd 2 O 3 concentration. The experimental values of mass attenuation coefficients (μ m ), effective atomic number (Z eff ) and effective electron densities (N e ) of the glasses were found to increase with the increasing of Gd 2 O 3 concentration and also with the decreasing of photon energy from 223 to 662 keV. The glasses of all Gd 2 O 3 compositions studied have been shown with lower HVL values in comparison to an X-rays shielding window, ordinary concrete and commercial window; indicating their potential as radiation shielding materials with Pb-free advantage. Optical spectra of the glasses in the present study had been shown with light transparency; an advantage when used as radiation shielding materials

  4. New gadolinium based glasses for gamma-rays shielding materials

    Kaewjang, S.; Maghanemi, U.; Kothan, S. [Department of Radiologic Technology, Faculty of Associated Medical Sciences, Chang Mai University, Chang Mai 50200 (Thailand); Kim, H.J. [Department of Physics, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Limkitjaroenporn, P. [Center of Excellence in Glass Technology and Materials Science (CEGM), Nakhon Pathom Rajabhat University, Nakhon Pathom 73000 (Thailand); Kaewkhao, J., E-mail: mink110@hotmail.com [Center of Excellence in Glass Technology and Materials Science (CEGM), Nakhon Pathom Rajabhat University, Nakhon Pathom 73000 (Thailand)

    2014-12-15

    Highlights: • Gd{sub 2}O{sub 3} based glasses have been fabricated and investigated radiation shielding properties between 223 and 662 keV. • Density of the glass increases with increasing of Gd{sub 2}O{sub 3.} • All the glasses of Gd{sub 2}O{sub 3} compositions studied had been shown lower HVL than X-rays shielding window. • Prepared glasses to be utilized as radiation shielding material with Pb-free advantage. • This work is the first to reports on radiation shielding properties of Gd{sub 2}O{sub 3} based glass matrices. - Abstract: In this work, Gd{sub 2}O{sub 3} based glasses in compositions (80−x)B{sub 2}O{sub 3}-10SiO{sub 2}-10CaO-xGd{sub 2}O{sub 3} (where x = 15, 20, 25, 30 and 35 mol%) have been fabricated and investigated for their radiation shielding, physical and optical properties. The density of the glass was found to increase with the increasing of Gd{sub 2}O{sub 3} concentration. The experimental values of mass attenuation coefficients (μ{sub m}), effective atomic number (Z{sub eff}) and effective electron densities (N{sub e}) of the glasses were found to increase with the increasing of Gd{sub 2}O{sub 3} concentration and also with the decreasing of photon energy from 223 to 662 keV. The glasses of all Gd{sub 2}O{sub 3} compositions studied have been shown with lower HVL values in comparison to an X-rays shielding window, ordinary concrete and commercial window; indicating their potential as radiation shielding materials with Pb-free advantage. Optical spectra of the glasses in the present study had been shown with light transparency; an advantage when used as radiation shielding materials.

  5. Highly Efficient and Reliable Transparent Electromagnetic Interference Shielding Film.

    Jia, Li-Chuan; Yan, Ding-Xiang; Liu, Xiaofeng; Ma, Rujun; Wu, Hong-Yuan; Li, Zhong-Ming

    2018-04-11

    Electromagnetic protection in optoelectronic instruments such as optical windows and electronic displays is challenging because of the essential requirements of a high optical transmittance and an electromagnetic interference (EMI) shielding effectiveness (SE). Herein, we demonstrate the creation of an efficient transparent EMI shielding film that is composed of calcium alginate (CA), silver nanowires (AgNWs), and polyurethane (PU), via a facile and low-cost Mayer-rod coating method. The CA/AgNW/PU film with a high optical transmittance of 92% achieves an EMI SE of 20.7 dB, which meets the requirements for commercial shielding applications. A superior EMI SE of 31.3 dB could be achieved, whereas the transparent film still maintains a transmittance of 81%. The integrated efficient EMI SE and high transmittance are superior to those of most previously reported transparent EMI shielding materials. Moreover, our transparent films exhibit a highly reliable shielding ability in a complex service environment, with 98 and 96% EMI SE retentions even after 30 min of ultrasound treatment and 5000 bending cycles (1.5 mm radius), respectively. The comprehensive performance that is associated with the facile fabrication strategy imparts the CA/AgNW/PU film with great potential as an optimized EMI shielding material in emerging optoelectronic devices, such as flexible solar cells, displays, and touch panels.

  6. REPOSITORY RADIATION SHIELDING DESIGN GUIDE

    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

  7. REPOSITORY RADIATION SHIELDING DESIGN GUIDE

    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.

  8. SNF shipping cask shielding analysis

    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

  9. Active Radiation Shield, Phase I

    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...

  10. Gonad shielding in computerized tomography

    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)

  11. Shielding calculations. Optimization vs. Paradigms

    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. Isolated human islets require hyperoxia to maintain islet mass, metabolism, and function.

    Komatsu, Hirotake; Kang, Dongyang; Medrano, Leonard; Barriga, Alyssa; Mendez, Daniel; Rawson, Jeffrey; Omori, Keiko; Ferreri, Kevin; Tai, Yu-Chong; Kandeel, Fouad; Mullen, Yoko

    2016-02-12

    Pancreatic islet transplantation has been recognized as an effective treatment for Type 1 diabetes; however, there is still plenty of room to improve transplantation efficiency. Because islets are metabolically active they require high oxygen to survive; thus hypoxia after transplant is one of the major causes of graft failure. Knowing the optimal oxygen tension for isolated islets would allow a transplant team to provide the best oxygen environment during pre- and post-transplant periods. To address this issue and begin to establish empirically determined guidelines for islet maintenance, we exposed in vitro cultured islets to different partial oxygen pressures (pO2) and assessed changes in islet volume, viability, metabolism, and function. Human islets were cultured for 7 days in different pO2 media corresponding to hypoxia (90 mmHg), normoxia (160 mmHg), and hyerpoxia (270 or 350 mmHg). Compared to normoxia and hypoxia, hyperoxia alleviated the loss of islet volume, maintaining higher islet viability and metabolism as measured by oxygen consumption and glucose-stimulated insulin secretion responses. We predict that maintaining pre- and post-transplanted islets in a hyperoxic environment will alleviate islet volume loss and maintain islet quality thereby improving transplant outcomes. Copyright © 2016 Elsevier Inc. All rights reserved.

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

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

    1998-01-01

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

  14. Radiation production and absorption in human spacecraft shielding systems under high charge and energy Galactic Cosmic Rays: Material medium, shielding depth, and byproduct aspects

    Barthel, Joseph; Sarigul-Klijn, Nesrin

    2018-03-01

    Deep space missions such as the planned 2025 mission to asteroids require spacecraft shields to protect electronics and humans from adverse effects caused by the space radiation environment, primarily Galactic Cosmic Rays. This paper first reviews the theory on how these rays of charged particles interact with matter, and then presents a simulation for a 500 day Mars flyby mission using a deterministic based computer code. High density polyethylene and aluminum shielding materials at a solar minimum are considered. Plots of effective dose with varying shield depth, charged particle flux, and dose in silicon and human tissue behind shielding are presented.

  15. Is mass drug administration against lymphatic filariasis required in urban settings? The experience in Kano, Nigeria.

    Dung D Pam

    2017-10-01

    Full Text Available The Global Programme to Eliminate Lymphatic Filariasis (GPELF, launched in 2000, has the target of eliminating the disease as a public health problem by the year 2020. The strategy adopted is mass drug administration (MDA to all eligible individuals in endemic communities and the implementation of measures to reduce the morbidity of those suffering from chronic disease. Success has been recorded in many rural endemic communities in which elimination efforts have centered. However, implementation has been challenging in several urban African cities. The large cities of West Africa, exemplified in Nigeria in Kano are challenging for LF elimination program because reaching 65% therapeutic coverage during MDA is difficult. There is therefore a need to define a strategy which could complement MDA. Thus, in Kano State, Nigeria, while LF MDA had reached 33 of the 44 Local Government Areas (LGAs there remained eleven 'urban' LGAs which had not been covered by MDA. Given the challenges of achieving at least 65% coverage during MDA implementation over several years in order to achieve elimination, it may be challenging to eliminate LF in such settings. In order to plan the LF control activities, this study was undertaken to confirm the LF infection prevalence in the human and mosquito populations in three urban LGAs.The prevalence of circulating filarial antigen (CFA of Wuchereria bancrofti was assessed by an immuno-chromatography test (ICT in 981 people in three urban LGAs of Kano state, Nigeria. Mosquitoes were collected over a period of 4 months from May to August 2015 using exit traps, gravid traps and pyrethrum knock-down spray sheet collections (PSC in different households. A proportion of mosquitoes were analyzed for W. bancrofti, using dissection, loop-mediated isothermal amplification (LAMP assay and conventional polymerase chain reaction (PCR.The results showed that none of the 981 subjects (constituted of <21% of children 5-10 years old tested

  16. Adaption of the LHC cold mass cooling system to the requirements of the Future Circular Collider (FCC)

    Kotnig, C.; Tavian, L.; Brenn, G.

    2017-12-01

    The cooling of the superconducting magnet cold masses with superfluid helium (He II) is a well-established concept successfully in operation for years in the LHC. Consequently, its application for the cooling of FCC magnets is an obvious option. The 12-kW heat loads distributed over 10-km long sectors not only require an adaption of the magnet bayonet heat exchangers but also present new challenges to the cryogenic plants, the distribution system and the control strategy. This paper recalls the basic LHC cooling concept with superfluid helium and defines the main parameters for the adaption to the FCC requirements. Pressure drop and hydrostatic head are developed in the distribution and pumping systems; their impact on the magnet temperature profile and the corresponding cooling efficiency is presented and compared for different distribution and pumping schemes.

  17. Measuring space radiation shielding effectiveness

    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.

  18. Measuring space radiation shielding effectiveness

    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.

  19. Why do mothers use nipple shields and how does this influence duration of exclusive breastfeeding?

    Kronborg, Hanne; Foverskov, Else; Nilsson, Ingrid

    2017-01-01

    with introducing nipple shields were lower- gestational age and birthweight. The use of nipple shields was furthermore found to be associated with a threefold increased risk of earlier cessation of exclusive breastfeeding: among primiparae odds ratio = 3.80 (confidence interval 2.61–5.53); among multiparae odds......The present study addressed the contentious discussions about the benefits and risks of nipple shield use. The objective was to explore self-reported reasons for using a nipple shield and examine associations pertaining to the mother, the infant and duration of breastfeeding. Data were collected...... breastfeeding period. Primiparae used nipple shields more often than multiparae, and early breastfeeding problems as well as background factors like lower age, education and higher body mass index were associated with a higher likelihood of using nipple shields. Characteristics of infants associated...

  20. Psychiatric emergency "surge capacity" following acts of terrorism and mass violence with high media impact: what is required?

    Claassen, Cindy; Kashner, T Michael; Kashner, Tetyana K; Xuan, Lei; Larkin, Gregory L

    2011-01-01

    Adequate preparedness for acts of terrorism and mass violence requires a thorough understanding of the postdisaster mental health needs of all exposed groups, including those watching such events from a distance. This study examined emergency psychiatric treatment-seeking patterns following media exposure to four national terrorist or mass casualty events. An event was selected for study if (a) it precipitated local front-page headlines for >5 consecutive days and (b) emergency service psychiatrists identified it as specifically precipitating help-seeking in the study hospital. Four events qualified: the Oklahoma City bombing (1995), the Columbine High School (1999) and Wedgewood Baptist Church (1999) shootings and the terrorist attacks of September 11, 2001. Time-series analyses were used to correct for autocorrelation in visit patterns during the postdisaster week, and equivalent time periods from years before and after each event were used as control years. Overall, disaster week census did not differ significantly from predisaster weeks, although 3-day nonsignificant decreases in visit rate were observed following each disaster. Treatment-seeking for anxiety-related issues showed a nonsignificant increase following each disaster, which became significant in the "all disaster" model (t=5.17; P=.006). Intensity of media coverage did not impact rate of help-seeking in any analysis. Although these sentinel US disasters varied in scope, method, geographic proximity to the study site, perpetrator characteristics, public response, sequelae and degree of media coverage, the extent to which they impacted emergency department treatment-seeking was minimal. Geographically distant mass violence and disaster events of the type and scope studied here may require only minimal mental health "surge capacity" in the days following the event. Copyright © 2011 Elsevier Inc. All rights reserved.

  1. Study on Basic Characteristics for the Development of Radiation Shielding High-Weight Concrete

    Mun, Young Bum; Lee, Jea Hyung; Choi, Hyun Kook [Sungshin Cement CO., Sejong (Korea, Republic of); Oh, Jeong Hwan; Choi, Soo Seok [Jeju National University, Jeju (Korea, Republic of)

    2016-05-15

    It is planned to build a power plant more than 6 units. Although the demand of a nuclear power plant is going to increase, the attention for radiation shielding is relatively in a low level. Concrete is one of the excellent and widely used shielding materials. Since the radiation shielding of a given material is proportional to density and thickness, a high-weight concrete with high-weight aggregate which is higher than normal concrete is used for radiation shielding. However, there are a few studies and references about radiation shielding concrete. Therefore, it is required to find a high-weight aggregate. The purpose of this paper is the development of a highweight concrete to improve radiation shielding capability. The radiation shielding rate of high-weight concrete is higher than that of reference concrete. It is confirmed that the density of aggregate and the unit weight of concreate is proportional to the radiation shielding rate. In addition, the chemical composition of aggregate has also has an important effect on γ-ray shielding. Therefore, high weight aggregates of higher density are essentially required to improve radiation shielding capability. The compressive strength of a high weight concrete is better than that of reference concrete. Slump and air contents, however, are slightly increased with by-product aggregates.

  2. Photon Shielding Features of Quarry Tuff

    Vega-Carrillo Hector Rene

    2017-01-01

    Full Text Available Cantera is a quarry tuff widely used in the building industry; in this work the shielding features of cantera were determined. The shielding characteristics were calculated using XCOM and MCNP5 codes for 0.03, 0.07, 0.1, 0.3, 0.662, 1, 2, and 3 MeV photons. With XCOM the mass interaction coefficients, and the total mass attenuation coefficients, were calculated. With the MCNP5 code a transmission experiment was modelled using a point-like source located 42 cm apart from a point-like detector. Between the source and the detector, cantera pieces with different thickness, ranging from 0 to 40 cm were included. The collided and uncollided photon fluence, the Kerma in air and the Ambient dose equivalent were estimated. With the uncollided fluence the linear attenuation coefficients were determined and compared with those calculated with XCOM. The linear attenuation coefficient for 0.662 MeV photons was compared with the coefficient measured with a NaI(Tl-based γ-ray spectrometer and a 137Cs source.

  3. RadWorks Storm Shelter Design for Solar Particle Event Shielding

    Simon, Matthew A.; Cerro, Jeffrey; Clowdsley, Martha

    2013-01-01

    In order to enable long-duration human exploration beyond low-Earth orbit, the risks associated with exposure of astronaut crews to space radiation must be mitigated with practical and affordable solutions. The space radiation environment beyond the magnetosphere is primarily a combination of two types of radiation: galactic cosmic rays (GCR) and solar particle events (SPE). While mitigating GCR exposure remains an open issue, reducing astronaut exposure to SPEs is achievable through material shielding because they are made up primarily of medium-energy protons. In order to ensure astronaut safety for long durations beyond low-Earth orbit, SPE radiation exposure must be mitigated. However, the increasingly demanding spacecraft propulsive performance for these ambitious missions requires minimal mass and volume radiation shielding solutions which leverage available multi-functional habitat structures and logistics as much as possible. This paper describes the efforts of NASA's RadWorks Advanced Exploration Systems (AES) Project to design minimal mass SPE radiation shelter concepts leveraging available resources. Discussion items include a description of the shelter trade space, the prioritization process used to identify the four primary shelter concepts chosen for maturation, a summary of each concept's design features, a description of the radiation analysis process, and an assessment of the parasitic mass of each concept.

  4. Evaluation of the Prevalence and Utility of Gonad Shielding in Pediatrics Undergoing Pelvic X-Ray

    Vahid Karami

    2016-11-01

    Full Text Available Background: Gonad shielding has been recommended during pelvic x-rays since the 1950s. The popular method of gonad shielding is placement a lead shield in the midline of the pelvis. The aim of this study was to evaluate the prevalence and utility of gonad shielding in pediatrics undergoing pelvic x-rays.Materials and MethodsFollowing study approval, we retrospectively retrieved data from the digital image library of ten radiology depertments of Khuzestan provience-Iran to identify pediatric patients who underwent pelvic x-ray (anteriorposterior [AP] view. All the images were reviewed for the probable evidence of gonad shield. If there was evidence of shielding, the accuracy positioning of the shield was also investigated by a single assistant radiologist.ResultsIn all 1745 pelvic x-rays (942 girls and 803 boys were identified of which the shield was present in 51 (5.41% radiographs of girls and 132 (16.43% radiographs of boys. When a shield was present; the shields has adequate positioning only in 8 (15.68% radiographs in girls and 59 radiographs in boys. Inaccurate placement and absence of gonad shields were more common in girls than the boys. Due to the shield has concealed the anatomical criteria of the pelvis, retakes of the examination was required in 11 (21.56% radiographs of girls and 14 (10.6% radiographs of boys.ConclusionThe current methods of gonad shielding in girls pelvic x-ray was not effective nor is justifiable. We  no longer advocate of gonad shielding during girls pelvic x-ray. However in boys it is controversial and depends on the skill and effort of radiographers.

  5. Analytic Shielding Optimization to Reduce Crew Exposure to Ionizing Radiation Inside Space Vehicles

    Gaza, Razvan; Cooper, Tim P.; Hanzo, Arthur; Hussein, Hesham; Jarvis, Kandy S.; Kimble, Ryan; Lee, Kerry T.; Patel, Chirag; Reddell, Brandon D.; Stoffle, Nicholas; hide

    2009-01-01

    A sustainable lunar architecture provides capabilities for leveraging out-of-service components for alternate uses. Discarded architecture elements may be used to provide ionizing radiation shielding to the crew habitat in case of a Solar Particle Event. The specific location relative to the vehicle where the additional shielding mass is placed, as corroborated with particularities of the vehicle design, has a large influence on protection gain. This effect is caused by the exponential- like decrease of radiation exposure with shielding mass thickness, which in turn determines that the most benefit from a given amount of shielding mass is obtained by placing it so that it preferentially augments protection in under-shielded areas of the vehicle exposed to the radiation environment. A novel analytic technique to derive an optimal shielding configuration was developed by Lockheed Martin during Design Analysis Cycle 3 (DAC-3) of the Orion Crew Exploration Vehicle (CEV). [1] Based on a detailed Computer Aided Design (CAD) model of the vehicle including a specific crew positioning scenario, a set of under-shielded vehicle regions can be identified as candidates for placement of additional shielding. Analytic tools are available to allow capturing an idealized supplemental shielding distribution in the CAD environment, which in turn is used as a reference for deriving a realistic shielding configuration from available vehicle components. While the analysis referenced in this communication applies particularly to the Orion vehicle, the general method can be applied to a large range of space exploration vehicles, including but not limited to lunar and Mars architecture components. In addition, the method can be immediately applied for optimization of radiation shielding provided to sensitive electronic components.

  6. Radiation shielding calculation for the MOX fuel fabrication plant Melox

    Lee, Y.K.; Nimal, J.C.; Chiron, M.

    1994-01-01

    Radiation shielding calculation is an important engineering work in the design of the MOX fuel fabrication plant MELOX. Due to the recycle of plutonium and uranium from UO2 spent fuel reprocessing and the large capacity of production (120t HM/yr.), the shielding design requires more attention in this LWR fuel plant. In MELOX, besides several temporary storage facilities of massive fissile material, about one thousand radioactive sources with different geometries, forms, densities, quantities and Pu concentrations, are distributed through different workshops from the PuO 2 powder reception unit to the fuel assembly packing room. These sources, with or without close shield, stay temporarily in different locations, containers and glove boxes. In order to optimize the dimensions, the material and the cost of shield as well as to limit the calculation work in a reasonable engineer-hours, a calculation scheme for shielding design of MELOX is developed. This calculation scheme has been proved to be useful in consideration of the feedback from the evolutionary design and construction. The validated shielding calculations give a predictive but reliable radiation doses information. (authors). 2 figs., 10 refs

  7. Gamma dose from activation of internal shields in IRIS reactor.

    Agosteo, Stefano; Cammi, Antonio; Garlati, Luisella; Lombardi, Carlo; Padovani, Enrico

    2005-01-01

    The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressuriser and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield.

  8. Gamma dose from activation of internal shields in IRIS reactor

    Agosteo, S.; Cammi, A.; Garlati, L.; Lombardi, C.; Padovani, E.

    2005-01-01

    The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressurizer and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60 Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield. (authors)

  9. ASOP, Shield Calculation, 1-D, Discrete Ordinates Transport

    1993-01-01

    1 - Nature of physical problem solved: ASOP is a shield optimization calculational system based on the one-dimensional discrete ordinates transport program ANISN. It has been used to design optimum shields for space applications of SNAP zirconium-hydride-uranium- fueled reactors and uranium-oxide fueled thermionic reactors and to design beam stops for the ORELA facility. 2 - Method of solution: ASOP generates coefficients of linear equations describing the logarithm of the dose and dose-weight derivatives as functions of position from data obtained in an automated sequence of ANISN calculations. With the dose constrained to a design value and all dose-weight derivatives required to be equal, the linear equations may be solved for a new set of shield dimensions. Since changes in the shield dimensions may cause the linear functions to change, the entire procedure is repeated until convergence is obtained. The detailed calculations of the radiation transport through shield configurations for every step in the procedure distinguish ASOP from other shield optimization computer code systems which rely on multiple component sources and attenuation coefficients to describe the transport. 3 - Restrictions on the complexity of the problem: Problem size is limited only by machine size

  10. Shieldings for X-ray radiotherapy facilities calculated by computer

    Pedrosa, Paulo S.; Farias, Marcos S.; Gavazza, Sergio

    2005-01-01

    This work presents a methodology for calculation of X-ray shielding in facilities of radiotherapy with help of computer. Even today, in Brazil, the calculation of shielding for X-ray radiotherapy is done based on NCRP-49 recommendation establishing a methodology for calculating required to the elaboration of a project of shielding. With regard to high energies, where is necessary the construction of a labyrinth, the NCRP-49 is not very clear, so that in this field, studies were made resulting in an article that proposes a solution to the problem. It was developed a friendly program in Delphi programming language that, through the manual data entry of a basic design of architecture and some parameters, interprets the geometry and calculates the shields of the walls, ceiling and floor of on X-ray radiation therapy facility. As the final product, this program provides a graphical screen on the computer with all the input data and the calculation of shieldings and the calculation memory. The program can be applied in practical implementation of shielding projects for radiotherapy facilities and can be used in a didactic way compared to NCRP-49.

  11. Shielding calculation for treatment rooms of high energy linear accelerator

    Elleithy, M.A.

    2006-01-01

    A review of German Institute of Standardization (DIN) scheme of the shielding calculation and the essential data required has been done for X-rays and electron beam in the energy range from 1 MeV to 50 MeV. Shielding calculation was done for primary and secondary radiations generated during X-ray operation of Linac. In addition, shielding was done against X-rays generated (Bremsstrahlung) by useful electron beams. The calculations also covered the neutrons generated from the interactions of useful X-rays (at energies above 8 MeV) with the surrounding. The present application involved the computation of shielding against the double scattered components of X-rays and neutrons in the maze area and the thickness of the paraffin wax of the room door. A new developed computer program was designed to assist shielding thickness calculations for a new Linac installation or in replacing an existing machine. The program used a combination of published tables and figures in computing the shielding thickness at different locations for all possible radiation situations. The DIN published data of 40 MeV accelerator room was compared with the program calculations. It was found that there is good agreement between both calculations. The developed program improved the accuracy and speed of calculation

  12. Electrodynamic Dust Shield Demonstrator

    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

  13. Gonad shielding in paediatric pelvic radiography: Effectiveness and practice

    Warlow, Thomas; Walker-Birch, Peter; Cosson, Philip

    2014-01-01

    The use of Gonad Shields (GS) has been advocated during pelvic radiography since the 1950's, particularly in children where the risks from radiation are higher. Previous literature reports that GS are often omitted and rarely used correctly. Objectives: Presentation of findings concerning use of GS in the context of previous data in the literature, and recommend any appropriate actions. Method: A retrospective analysis of images from an existing DICOM Digital Teaching Library (DTL) was conducted. Images of the pelvis from paediatric patients were reviewed and scored on whether a GS was present and (if present) whether the shield was considered to adequately protect the gonads. Results: 130 images were reviewed. 70 male and 60 female. The gonads were deemed to be protected by a shield in 22 images (17%), inadequately protected when a shield was used in 44 images (34%) with the remaining 64 images (49%) having no shield at all. A lack of adequate protection for the gonads was found, with females more likely to be inadequately protected than males (χ 2  = 19.009, df = 1, p < 0.001). These findings become more clinically significant when reports of ovaries lying outside of the pelvic basin (in paediatric patients) are considered. Conclusions: The current practice of gonad shielding is neither effective nor beneficial for female paediatric patients, incorrect shield placement can often require repeat exposures. This finding is commensurate with previous literature. Therefore, gonad shielding is no longer an appropriate optimization tool for female paediatric patients during conventional radiography of the pelvis, and should be abandoned

  14. Thermal design of top shield

    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

  15. Radiation shielding in dental radiography

    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)

  16. Shielding features of quarry stone

    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)

  17. Using the Monte Carlo Coupling Technique to Evaluate the Shielding Ability of a Modular Shielding House to Accommodate Spent-Fuel Transportable Storage Casks

    Ueki, Kohtaro; Kawakami, Kazuo; Shimizu, Daisuke

    2003-01-01

    The Monte Carlo coupling technique with the coordinate transformation is used to evaluate the shielding ability of a modular shielding house that accommodates four spent-fuel transportable storage casks for two units. The effective dose rate distributions can be obtained as far as 300 m from the center of the shielding house. The coupling technique is created with the Surface Source Write (SSW) card and the Surface Source Read/Coordinate Transformation (SSR/CRT) card in the MCNP 4C continuous energy Monte Carlo code as the 'SSW-SSR/CRT calculation system'. In the present Monte Carlo coupling calculation, the total effective dose rates 100, 200, and 300 m from the center of the shielding house are estimated to be 1.69, 0.285, and 0.0826 (μSv/yr per four casks), respectively. Accordingly, if the distance between the center of the shielding house and the site boundary of the storage facility is kept at >300 m, approximately 2400 casks are able to be accommodated in the modular shielding houses, under the Japanese severe criterion of 50 μSv/yr at the site boundary. The shielding house alone satisfies not only the technical conditions but also the economic requirements.It became evident that secondary gamma rays account for >60% of the effective total dose rate at all the calculated points around the shielding house, most of which are produced from the water in the steel-water-steel shielding system of the shielding house. The remainder of the dose rate comes mostly from neutrons; the fission product and 60 Co activation gamma rays account for small percentages. Accordingly, reducing the secondary gamma rays is critical to improving not only the shielding ability but also the radiation safety of the shielding house

  18. Shielding and maintainability in an experimental tokamak

    Abdou, M.A.; Fuller, G.; Hager, E.R.; Vogelsang, W.F.

    1979-01-01

    This paper presents the results of an attempt to develop an understanding of the various factors involved. This work was performed as a part of the task assigned to one of the expert groups on the International Tokamak Reactor (INTOR). However, the results of this investigation are believed to be generally applicable to the broad class of the next generation of experimental tokamak facilities such as ETF. The shielding penalties for requiring personnel access are quantified. This is followed by a quantitative estimate of the benefits associated with personnel access. The penalties are compared to the benefits and conclusions and recommendations are developed on resolving the issue

  19. Shielding design at Fermilab: Calculations and measurements

    Cossairt, J.D.

    1986-11-01

    The development of the Fermilab accelerator complex during the past two decades from its concept as the ''200 BeV accelerator'' to that of the present tevatron, designed to operate at energies as high as 1 TeV, has required a coincidental refinement and development in methods of shielding design. In this paper I describe these methods as used by the radiation protection staff of Fermilab. This description will review experimental measurements which substantiate these techniques in realistic situations. Along the way, observations will be stated which likely are applicable to other protron accelerators in the multi-hundred GeV energy region, including larger ones yet to be constructed

  20. Space Shuttle Orbiter AFT heat shield seal

    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.

  1. Combination thermal and radiation shield for well logging apparatus

    Wilson, B.F.

    1984-01-01

    A device for providing both thermal protection and radiation shielding for components such as radiation detectors within a well logging instrument comprises a thermally insulative flask containing a weldment filled with a mass of eutectic material which undergoes a change of state e.g. melting at a temperature which will provide an acceptable thermal environment for such components for extended time periods. The eutectic material which is preferably a bismuth (58%)/tin (42%) alloy has a specific gravity (> 8.5) facilitating its use as a radiation shield and is distributed around the radiation detectors so as to selectively impede the impinging of the detectors by radiation. The device is incorporated in a skid of a well logging instrument for measuring γ backscatter. A γ source is located either above or within the protective shielding. (author)

  2. Radiation Shielding Materials and Containers Incorporating Same

    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.

  3. Magnetic shielding for superconducting RF cavities

    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.

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

    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.

  5. What Bed Size Does a Patient Need? The Relationship Between Body Mass Index and Space Required to Turn in Bed.

    Wiggermann, Neal; Smith, Kathryn; Kumpar, Dee

    A bed that is too small to allow patients to turn from supine to side lying increases the difficulty of mobilizing patients, which can increase risk of musculoskeletal injury to caregivers, increase risk of pressure injuries to patients, and reduce patient comfort. Currently, no guidance is available for what patient sizes are accommodated by the standard 91cm (36 in.)-wide hospital bed, and no studies have evaluated the relationship between anthropometric attributes and space required to turn in bed. The purpose of this research was to determine how much space individuals occupy when turning from supine to side lying as predicted by their anthropometry (i.e., body dimensions) to establish guidance on selecting the appropriate bed size. Forty-seven adult participants (24 female) with body mass index (BMI) from 20 to 76 kg/m participated in a laboratory study. Body dimensions were measured, and the envelope of space required to turn was determined using motion capture. Linear regressions estimated the relationship between anthropometric attributes and space occupied when turning. BMI was strongly correlated (R = .88) with the space required to turn. Based on the linear regressions, individuals with BMI up to 35 kg/m could turn left and right within 91 cm and individuals with BMI up to 45 kg/m could turn one direction within 91 cm. BMI is a good predictor of the space required to turn from supine to lateral. Nurses should consider placing patients that are unable to laterally reposition themselves on a wider bed when BMI is greater than 35 kg/m and should consider placing all patients greater than 45 kg/m on a wider bed regardless of mobility. Hospital administrators can use historical demographic information about the BMI of their patient populations to plan facility-level equipment procurement for equipment that accommodates their patients.

  6. Radiation shielding for fission reactors

    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)

  7. Shield cost minimization using SWAN

    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

  8. Novel shielding materials for space and air travel

    Vana, N.; Hajek, M.; Berger, T.; Fugger, M.; Hofmann, P.

    2006-01-01

    The reduction of dose onboard spacecraft and aircraft by appropriate shielding measures plays an essential role in the future development of space exploration and air travel. The design of novel shielding strategies and materials may involve hydrogenous composites, as it is well known that liquid hydrogen is most effective in attenuating charged particle radiation. As precursor for a later flight experiment, the shielding properties of newly developed hydrogen-rich polymers and rare earth-doped high-density rubber were tested in various ground-based neutron and heavy ion fields and compared with aluminium and polyethylene as reference materials. Absorbed dose, average linear energy transfer and gamma-equivalent neutron absorbed dose were determined by means of LiF:Mg,Ti thermoluminescence dosemeters and CR-39 plastic nuclear track detectors. First results for samples of equal aerial density indicate that selected hydrogen-rich plastics and rare-earth-doped rubber may be more effective in attenuating cosmic rays by up to 10% compared with conventional aluminium shielding. The appropriate adaptation of shielding thicknesses may thus allow reducing the biologically relevant dose. Owing to the lower density of the plastic composites, mass savings shall result in a significant reduction of launch costs. The experiment was flown as part of the European Space Agency's Biopan-5 mission in May 2005. (authors)

  9. Post Three Mile Island shielding review - a case history

    Isakari, H.H.; Shaw, H.C.

    1983-01-01

    The radiation shielding review of the Diablo Canyon Nuclear Power Plant was performed in accordance with the requirement of the Three Mile Island Action Plan. The review covered plant shielding and environmental qualification of equipment for spaces and systems which may be used in post-accident operations. Radiation doses during postulated loss-of-coolant accident and high-energy-line-break accident were calculated for equipment located both inside and outside the containment. Vital areas, those requiring post-accident access and occupancy, were identified and their associated dose rates and integrated doses were calculated. It was found that all four of the vital areas (Control Room, Technical Support Center, Switchgear Room, and Emergency Sampling Compartment) are shielded from external sources of radiation sufficiently to permit personnel access and occupancy that would not be unduly limited by the radiation environment caused by the postulated accidents. (author)

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

    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.

  11. Evaluation of shielding parameters for heavy metal fluoride based tellurite-rich glasses for gamma ray shielding applications

    Sayyed, M. I.; Lakshminarayana, G.; Kityk, I. V.; Mahdi, M. A.

    2017-10-01

    In this work, we have evaluated the γ-ray shielding parameters such as mass attenuation coefficient (μ/ρ), effective atomic number (Zeff), half value layer (HVL), mean free path (MFP) and exposure buildup factors (EBF) for heavy metal fluoride (PbF2) based tellurite-rich glasses. In addition, neutron total macroscopic cross sections (∑R) for these glasses were also calculated. The maximum value for μ/ρ, Zeff and ∑R was found for heavy metal (Bi2O3) oxide introduced glass. The results of the selected glasses have been compared, in terms of MFP with different glass systems. The shielding effectiveness of the selected glasses is found comparable or better than of common ones, which indicates that these glasses with suitable oxides could be developed for gamma ray shielding applications.

  12. Shielding walls against ionizing radiation

    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

  13. Nuclear steam generator tubesheet shield

    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

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

    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.

  15. The UK shielding Forum. Best Practice through consensus

    Hobson, J.; Gunston, K.; Gunston, K.

    2000-01-01

    The UK national shielding Forum has been established to represent all key industry groups in the UK (including the Nuclear Installations Inspectorate (NII), the national regulatory authority). The Forum's aim is to increase awareness and confidence in the range of professional practice within the UK shielding community, with a view to having a coherent and dynamic role within the international shielding community. In the past, no comprehensive representative body covering the whole UK nuclear industry has existed, and the different industry shielding groups have developed local ways of working to address their particular requirements. Inevitably, there are common issues arising from these requirements which benefit from a wider consensus. As a result of the formation of the Forum (initiated by the NII and subsequently chaired by BNFL as an industry key player), expertise, experience and best working practice are now being actively shared between shielding professionals, and there has been a strong and successful drive to achieving consensus on key issues, which is also reflected in the increasing quality of industry-regulator relationships. (author)

  16. Status of the ITER tokamak nuclear shielding and radiological protection design

    Leichtle, D., E-mail: dieter.leichtle@f4e.europa.eu [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Chaffard, P.Y.; Izquierdo, J. [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Juarez, R. [UNED, Juan del Rosal 12, Madrid 28040 (Spain); Pampin, R.; Portone, A. [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain)

    2016-11-01

    Highlights: • Comprehensive review of design status of the ITER tokamak regarding nuclear shielding. • Investigation of shield design options and streaming mitigation measures. • Review of state-of-the-art shutdown dose rate analyses for selected port systems. - Abstract: Nuclear shielding of the ITER tokamak encompasses several systems and interfaces in a complex radiation environment. Therefore any shielding design has to involve a series of structures, systems and components in an integrated approach. This is evident for the complex ex-vessel radiation environment with streaming and leakage of plasma neutrons and subsequent activation of ex-vessel structures which give raise to excessive shutdown dose rates in accessible areas of the cryostat. The paper reviews recent nuclear analyses related to the performance of primary shields and highlights challenges toward an integrated nuclear shielding design. The general need of propagation of shielding requirements is highlighted in the context of radiation cross talk due to penetrations. Radiation streaming through gaps and penetrations is a key problem in any efficient shield design. The impact on the evolving radiation environment due to several design options along streaming paths such as port gaps, as well as their modeling for nuclear analysis, is presented. Implications regarding design integration and compliance with integrated shielding requirements and ALARA dose are finally given.

  17. Lightweight Shield for Centrifuge

    Luper, C.

    1982-01-01

    Centrifuge bowl composed of laminated aluminum offers required combination of high strength at reduced weight. Around outside wall of bowl core of 1/16 inch thick spun aluminum are wrapped two layers of aluminum, each also one-sixteenth inch thick. Layered structure prevents cracks from propagating through wall.

  18. Shielding structure analysis for LSDS facility

    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.

  19. Shielding structure analysis for LSDS facility

    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

  20. Simbol-X Mirror Module Thermal Shields: I-Design and X-Ray Transmission

    Collura, A.; Barbera, M.; Varisco, S.; Basso, S.; Pareschi, G.; Tagliaferri, G.; Ayers, T.

    2009-05-01

    The Simbol-X mission is designed to fly in formation flight configuration. As a consequence, the telescope has both ends open to space, and thermal shielding at telescope entrance and exit is required to maintain temperature uniformity throughout the mirrors. Both mesh and meshless solutions are presently under study for the shields. We discuss the design and the X-ray transmission.

  1. Simbol-X Mirror Module Thermal Shields: I - Design and X-Ray Transmission

    Collura, A.; Varisco, S.; Barbera, M.; Basso, S.; Pareschi, G.; Tagliaferri, G.; Ayers, T.

    2009-01-01

    The Simbol-X mission is designed to fly in formation flight configuration. As a consequence, the telescope has both ends open to space, and thermal shielding at telescope entrance and exit is required to maintain temperature uniformity throughout the mirrors. Both mesh and meshless solutions are presently under study for the shields. We discuss the design and the X-ray transmission.

  2. Method of constructing shielding wall

    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.)

  3. Gamma ray and neutron shielding properties of some concrete materials

    Yilmaz, E.; Baltas, H.; Kiris, E.; Ustabas, I.; Cevik, U.; El-Khayatt, A.M.

    2011-01-01

    Highlights: → This study sheds light on the shielding properties of gamma-rays and neutrons for some concrete samples. → The experimental mass attenuation coefficients values were compared with theoretical values obtained using WinXCom. → Moreover, neutron shielding has been treated in terms of macroscopic removal cross-section (Σ R , cm -1 ) concept. → The NXcom program was employed to calculate the attenuation coefficients values of neutrons. → These values showed a change with energy and composition of the concrete samples. - Abstract: Shielding of gamma-rays and neutrons by 12 concrete samples with and without mineral additives has been studied. The total mass attenuation and linear attenuation coefficients, half-value thicknesses, effective atomic numbers, effective electron densities and atomic cross-sections at photons energies of 59.5 and 661 keV have been measured and calculated. The measured and calculated values were compared and a reasonable agreement has been observed. Also the recorded values showed a change with energy and composition of the concrete samples. In addition, neutron shielding has been treated in terms of macroscopic removal cross-section (Σ R , cm -1 ) concept. The WinXCom and NXcom programs were employed to calculate the attenuation coefficients of gamma-rays and neutrons, respectively.

  4. Designing shields for KeV photons with genetic algorithms

    Asbury, Stephen; Holloway, James P.

    2011-01-01

    Shielding of x-ray sources and low energy gamma rays is often accomplished with lead aprons, comprising a thin layer (0.5 mm to 1 mm) of lead or similar high-Z material. In previous work the authors used Genetic Algorithms to explore the design of a shadow shield for space applications. Now those techniques have been applied to the problem of shielding humans from low energy gamma radiation. This paper uses a simple geometry to explore layering various materials as a method to reduce mass and dose for thin gamma shields. The genetic algorithms discover layers of materials with various Z is in fact more effective than an equivalent mass of Pb alone for lower energy gammas, but as the incident radiation energy increases the efficacy of such layering diminishes. The utility of varying Z for lower energy gammas is in part due to their complementary K-edges, where one material compensates for the transmission that would occur just below the K-edge in another material. (author)

  5. Convertible shielding to ceramic breeding blanket

    Furuya, Kazuyuki; Kurasawa, Toshimasa; Sato, Satoshi; Nakahira, Masataka; Togami, Ikuhide; Hashimoto, Toshiyuki; Takatsu, Hideyuki; Kuroda, Toshimasa.

    1995-05-01

    Four concepts have been studied for the ITER convertible blanket: 1)Layered concept 2)BIT(Breeder-Inside-Tube)concept 3)BOT(Breeder-Out of-Tube)concept 4)BOT/mixed concept. All concepts use ceramic breeder and beryllium neutron multiplier, both in the shape of small spherical pebbles, 316SS structure, and H 2 O coolant (inlet/outlet temperatures : 100/150degC, pressure : 2 MPa). During the BPP, only beryllium pebbles (the primary pebble in case of BOT/mixed concept) are filled in the blanket for shielding purpose. Then, before the EPP operation, breeder pebbles will be additionally inserted into the blanket. Among possible conversion methods, wet method by liquid flow seems expecting for high and homogeneous pebble packing. Preliminary 1-D neutronics calculation shows that the BOT/mixed concept has the highest breeding and shielding performance. However, final selection should be done by R and D's and more detail investigation on blanket characteristics and fabricability. Required R and D's are also listed. With these efforts, the convertible blanket can be developed. However, the following should be noted. Though many of above R and D's are also necessary even for non-convertible blanket, R and D's on convertibility will be one of the most difficult parts and need significant efforts. Besides the installation of convertible blanket with required structures and lines for conversion will make the ITER basic machine more complicated. (author)

  6. Radioactivity, shielding, radiation damage, and remote handling

    Wilson, M.T.

    1975-01-01

    Proton beams of a few hundred million electron volts of energy are capable of inducing hundreds of curies of activity per microampere of beam intensity into the materials they intercept. This adds a new dimension to the parameters that must be considered when designing and operating a high-intensity accelerator facility. Large investments must be made in shielding. The shielding itself may become activated and require special considerations as to its composition, location, and method of handling. Equipment must be designed to withstand large radiation dosages. Items such as vacuum seals, water tubing, and electrical insulation must be fabricated from radiation-resistant materials. Methods of maintaining and replacing equipment are required that limit the radiation dosages to workers.The high-intensity facilities of LAMPF, SIN, and TRIUMF and the high-energy facility of FERMILAB have each evolved a philosophy of radiation handling that matches their particular machine and physical plant layouts. Special tooling, commercial manipulator systems, remote viewing, and other techniques of the hot cell and fission reactor realms are finding application within accelerator facilities. (U.S.)

  7. Shielding Effectiveness of a Thin Film Window

    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...

  8. Gonadal Shielding in Radiography: A Best Practice?

    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.

  9. Infinite slab-shield dose calculations

    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

  10. Using glass as a shielding material

    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)

  11. Using glass as a shielding material

    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)

  12. Shielding and grounding in large detectors

    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

  13. RASH D - A mercury programme for neutron shielding calculations

    Bendall, D.E.

    1962-08-01

    An improved version of an earlier neutron shielding programme (RASH B) is described. The new programme is also written in Mercury Autocode and solves a set of multigroup diffusion equations in one dimension. It differs from RASH B in that distributed source terms may be introduced into all the groups if required. Some other improvements are also included. (author)

  14. Pretinning Nickel-Plated Wire Shields

    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.

  15. Computed tomography shielding methods: a literature review.

    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.

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

    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. Radiation dose reduction to the male gonads during MDCT: the effectiveness of a lead shield.

    Hohl, Christian; Mahnken, Andreas H; Klotz, Ernst; Das, Marco; Stargardt, Achim; Mühlenbruch, Georg; Schmidt, Thorsten; Günther, Rolf W; Wildberger, Joachim E

    2005-01-01

    Our study was designed to quantify the effect of a standard gonad shield on the testicular radiation exposure due to scatter during routine abdominopelvic MDCT. Routine abdominopelvic MDCT was performed in 34 patients with gonadal lead shielding and 32 patients without this shielding; the testes were not exposed to the direct beam during the examination. We estimated the testicular dose administered with thermoluminescent dosimetry, taking into account each patient's body weight and body mass index (BMI). With a 1-mm lead shield, the mean testicular dose was reduced from 2.40 to 0.32 mSv, a reduction of 87%. The difference was found to be statistically significant (p Shielding the male gonads reduces the testicular radiation dose during abdominopelvic MDCT significantly and can be recommended for routine use.

  18. Shield materials recommended for space power nuclear reactors

    Kaszubinski, L. J.

    1973-01-01

    Lithium hydride is recommended for neutron attenuation and depleted uranium is recommended for gamma ray attenuation. For minimum shield weights these materials must be arranged in alternate layers to attenuate the secondary gamma rays efficiently. In the regions of the shield near the reactor, where excessive fissioning occurs in the uranium, a tungsten alloy is used instead. Alloys of uranium such as either the U-0.5Ti or U-8Mo are available to accommodate structural requirements. The zone-cooled casting process is recommended for lithium hydride fabrication. Internal honeycomb reinforcement to control cracks in the lithium hydride is recommended.

  19. A new formulation for resonance self-shielding factors

    Palma, Daniel A.P.; Martinez, Aquilino S.; Silva, Fernando C. da [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear]. E-mail: aquilino@lmp.ufrj.br

    2007-07-01

    The activation technique allows either absolute or relative very precise neutron intensity measurements. This technique requires the knowledge of the Doppler broadening function to determine resonance self-shielding factors. In the present work a new formulation is proposed for the self-shielding factors where the Doppler broadening function is calculated using the Frobenius's method and compared to the values obtained from the four-pole Pade method. This calculation method is shown to be effective from the point of view of accuracy. (author)

  20. A new formulation for resonance self-shielding factors

    Palma, Daniel A.P.; Martinez, Aquilino S.; Silva, Fernando C. da

    2007-01-01

    The activation technique allows either absolute or relative very precise neutron intensity measurements. This technique requires the knowledge of the Doppler broadening function to determine resonance self-shielding factors. In the present work a new formulation is proposed for the self-shielding factors where the Doppler broadening function is calculated using the Frobenius's method and compared to the values obtained from the four-pole Pade method. This calculation method is shown to be effective from the point of view of accuracy. (author)

  1. BRH Gonad Shielding Program: where it has led

    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

  2. Kinematics of center of mass and center of pressure predict friction requirement at shoe-floor interface during walking.

    Yamaguchi, Takeshi; Yano, Masaru; Onodera, Hiroshi; Hokkirigawa, Kazuo

    2013-06-01

    We aimed to determine whether inclination angles of the line connecting the whole body center of mass (COM) to the center of pressure (COP) (COM-COP angle) help predict the required coefficient of friction (RCOF) in young adult males during the weight acceptance and push-off phases in transient movements such as turning, gait termination and initiation, and steady-state movements such as straight walking. Seventeen healthy young adult males were asked to (1) walk in a straight line, (2) turn 60° with either foot (step and spin turns), and (3) initiate and terminate walking on a dry level floor. Peak absolute values of the ratio between resultant horizontal and vertical ground reaction forces during the weight acceptance and push-off phases (RCOFh and RCOFt, respectively) were calculated. COM-COP angles θh and θt at the instant of RCOFh and RCOFt, respectively, were also calculated. Bivariate regression analysis demonstrated that the |θh| and |θt| tangents were significant predictors of RCOFh (R = 0.878; R(2) = 0.770; ppush-off phases in steady-state movements such as straight walking and transient movements such as turning as well as gait termination and initiation. Copyright © 2012 Elsevier B.V. All rights reserved.

  3. Semi-analytic flux formulas for shielding calculations

    Wallace, O.J.

    1976-06-01

    A special coordinate system based on the work of H. Ono and A. Tsuro has been used to derive exact semi-analytic formulas for the flux from cylindrical, spherical, toroidal, rectangular, annular and truncated cone volume sources; from cylindrical, spherical, truncated cone, disk and rectangular surface sources; and from curved and tilted line sources. In most of the cases where the source is curved, shields of the same curvature are allowed in addition to the standard slab shields; cylindrical shields are also allowed in the rectangular volume source flux formula. An especially complete treatment of a cylindrical volume source is given, in which dose points may be arbitrarily located both within and outside the source, and a finite cylindrical shield may be considered. Detector points may also be specified as lying within spherical and annular source volumes. The integral functions encountered in these formulas require at most two-dimensional numeric integration in order to evaluate the flux values. The classic flux formulas involving only slab shields and slab, disk, line, sphere and truncated cone sources become some of the many special cases which are given in addition to the more general formulas mentioned above

  4. Status of reactor-shielding research in the US

    Maienshein, F.C.

    1980-01-01

    While reactor programs change, shielding analysis methods are improved slowly. Version-V of ENDF/B provides improved data and Version-VI will be cost effective in advanced fission reactors are to be developed in the US. Benchmarks for data and methods validation are collected and distributed in the US in two series, one primarily for FBR-related experiments and one for LWR calculational methods. For LWR design, cavity streaming is now handled adequately, if with varying degrees of elegance. Investigations of improved detector response for LWRs rely upon transport methods. The great potential importance of pressure-vessel damage is dreflected in widespread studies to aid in the prediction of neutron fluences in vessels. For LMFBRS, the FFTF should give attenuation results on an operating reactor. For larger power reactors, the advantages of alternate shield materials appear compelling. A few other shielding studies appear to require experimental confirmation if LMFBRs are to be economically competitive. A coherent shielding program for the GCFR is nearing completion. For the fusion-reactor program, methods verification is under way, practical calculations are well advanced for test devices such as the TFTR and FMIT, and consideration is now given to shielding problems of large reactors, as in the ETF study

  5. Treatment vault shielding for a flattening filter-free medical linear accelerator

    Kry, Stephen F.; Howell, Rebecca M.; Polf, Jerimy; Mohan, Radhe; Vassiliev, Oleg N.

    2009-03-01

    The requirements for shielding a treatment vault with a Varian Clinac 2100 medical linear accelerator operated both with and without the flattening filter were assessed. Basic shielding parameters, such as primary beam tenth-value layers (TVLs), patient scatter fractions, and wall scatter fractions, were calculated using Monte Carlo simulations of 6, 10 and 18 MV beams. Relative integral target current requirements were determined from treatment planning studies of several disease sites with, and without, the flattening filter. The flattened beam shielding data were compared to data published in NCRP Report No. 151, and the unflattened beam shielding data were presented relative to the NCRP data. Finally, the shielding requirements for a typical treatment vault were determined for a single-energy (6 MV) linac and a dual-energy (6 MV/18 MV) linac. With the exception of large-angle patient scatter fractions and wall scatter fractions, the vault shielding parameters were reduced when the flattening filter was removed. Much of this reduction was consistent with the reduced average energy of the FFF beams. Primary beam TVLs were reduced by 12%, on average, and small-angle scatter fractions were reduced by up to 30%. Head leakage was markedly reduced because less integral target current was required to deliver the target dose. For the treatment vault examined in the current study, removal of the flattening filter reduced the required thickness of the primary and secondary barriers by 10-20%, corresponding to 18 m3 less concrete to shield the single-energy linac and 36 m3 less concrete to shield the dual-energy linac. Thus, a shielding advantage was found when the linac was operated without the flattening filter. This translates into a reduction in occupational exposure and/or the cost and space of shielding.

  6. Treatment vault shielding for a flattening filter-free medical linear accelerator

    Kry, Stephen F; Howell, Rebecca M; Polf, Jerimy; Mohan, Radhe; Vassiliev, Oleg N [Department of Radiation Physics, University of Texas M. D. Anderson Cancer Center, Houston, TX (United States)], E-mail: sfkry@mdanderson.org

    2009-03-07

    The requirements for shielding a treatment vault with a Varian Clinac 2100 medical linear accelerator operated both with and without the flattening filter were assessed. Basic shielding parameters, such as primary beam tenth-value layers (TVLs), patient scatter fractions, and wall scatter fractions, were calculated using Monte Carlo simulations of 6, 10 and 18 MV beams. Relative integral target current requirements were determined from treatment planning studies of several disease sites with, and without, the flattening filter. The flattened beam shielding data were compared to data published in NCRP Report No. 151, and the unflattened beam shielding data were presented relative to the NCRP data. Finally, the shielding requirements for a typical treatment vault were determined for a single-energy (6 MV) linac and a dual-energy (6 MV/18 MV) linac. With the exception of large-angle patient scatter fractions and wall scatter fractions, the vault shielding parameters were reduced when the flattening filter was removed. Much of this reduction was consistent with the reduced average energy of the FFF beams. Primary beam TVLs were reduced by 12%, on average, and small-angle scatter fractions were reduced by up to 30%. Head leakage was markedly reduced because less integral target current was required to deliver the target dose. For the treatment vault examined in the current study, removal of the flattening filter reduced the required thickness of the primary and secondary barriers by 10-20%, corresponding to 18 m{sup 3} less concrete to shield the single-energy linac and 36 m{sup 3} less concrete to shield the dual-energy linac. Thus, a shielding advantage was found when the linac was operated without the flattening filter. This translates into a reduction in occupational exposure and/or the cost and space of shielding.

  7. Treatment vault shielding for a flattening filter-free medical linear accelerator

    Kry, Stephen F; Howell, Rebecca M; Polf, Jerimy; Mohan, Radhe; Vassiliev, Oleg N

    2009-01-01

    The requirements for shielding a treatment vault with a Varian Clinac 2100 medical linear accelerator operated both with and without the flattening filter were assessed. Basic shielding parameters, such as primary beam tenth-value layers (TVLs), patient scatter fractions, and wall scatter fractions, were calculated using Monte Carlo simulations of 6, 10 and 18 MV beams. Relative integral target current requirements were determined from treatment planning studies of several disease sites with, and without, the flattening filter. The flattened beam shielding data were compared to data published in NCRP Report No. 151, and the unflattened beam shielding data were presented relative to the NCRP data. Finally, the shielding requirements for a typical treatment vault were determined for a single-energy (6 MV) linac and a dual-energy (6 MV/18 MV) linac. With the exception of large-angle patient scatter fractions and wall scatter fractions, the vault shielding parameters were reduced when the flattening filter was removed. Much of this reduction was consistent with the reduced average energy of the FFF beams. Primary beam TVLs were reduced by 12%, on average, and small-angle scatter fractions were reduced by up to 30%. Head leakage was markedly reduced because less integral target current was required to deliver the target dose. For the treatment vault examined in the current study, removal of the flattening filter reduced the required thickness of the primary and secondary barriers by 10-20%, corresponding to 18 m 3 less concrete to shield the single-energy linac and 36 m 3 less concrete to shield the dual-energy linac. Thus, a shielding advantage was found when the linac was operated without the flattening filter. This translates into a reduction in occupational exposure and/or the cost and space of shielding.

  8. Radiation shielding apparatus

    McCullagh, R.J.

    1977-01-01

    The disclosure pertains to a clamping apparatus having a stud capturing portion and a stud facing portion bolted together so as to compressively support a radiation-proof sheet material, such as lead sheeting, there-in-between. The interior wall covering material, such as panelling or wall board, is secured to the external surface of the stud facing portion. No nails are required to support the radiation-proof sheeting material, thereby minimizing accidental leakage due to harmful radiation passing through openings inadvertently disposed in the radiation-proof sheeting in the conventional nail securing supporting thereof. A pair of radiation-proof tracks capture the free ends of the stud capturing portion and the stud facing portion

  9. Concrete mix design for X-and gamma shielding

    Mohamad Pauzi Ismail; Noor Azreen Masenwat; Suhairy Sani; Abdul Bakhri Muhammad; Mohd Kamal Shah Shamsuddin; Rahmad Abd Rashid

    2012-01-01

    The design of X-ray or gamma ray radiographic exposure room requires some calculations on shielding to provide safe operation of the facility and minimum exposure to radiation workers. Careful design can lead to economical installations with minimal barriers. The design depends on such factors as: maximum energy, maximum intensity, permitted full-body dosage, workload, use factor, occupancy factor, maximum dose output and shielding materials. Choice of material for a barrier depends on convenience and cost. The radiographic exposure room is usually made of normal concrete with density of about 2.3 - 2.4 g/ cc. Normal concrete is often used for construction of exposure room because of cheap and ease of construction. This paper explained and discussed the optimum mix design for normal concrete used for X-and gamma shielding. (author)

  10. The ATLAS SCT grounding and shielding concept and implementation

    Bates, RL; Bernabeu, J; Bizzell, J; Bohm, J; Brenner, R; Bruckman de Renstrom, P A; Catinaccio, A; Cindro, V; Ciocio, A; Civera, J V; Chouridou, S; Dervan, P; Dick, B; Dolezal, Z; Eklund, L; Feld, L; Ferrere, D; Gadomski, S; Gonzalez, F; Gornicki, E; Greenhall, A; Grillo, A A; Grosse-Knetter, J; Gruwe, M; Haywood, S; Hessey, N P; Ikegami, Y; Jones, T J; Kaplon, J; Kodys, P; Kohriki, T; Kondo, T; Koperny, S; Lacasta, C; Lozano Bahilo, J; Malecki, P; Martinez-McKinney, F; McMahon, S J; McPherson, A; Mikulec, B; Mikus, M; Moorhead, G F; Morrissey, M C; Nagai, K; Nichols, A; O'Shea, V; Pater, J R; Peeters, S J M; Pernegger, H; Perrin, E; Phillips, P W; Pieron, J P; Roe, S; Sanchez, J; Spencer, E; Stastny, J; Tarrant, J; Terada, S; Tyndel, M; Unno, Y; Wallny, R; Weber, M; Weidberg, A R; Wells, P S; Werneke, P; Wilmut, I

    2012-01-01

    This paper describes the design and implementation of the grounding and shielding system for the ATLAS SemiConductor Tracker (SCT). The mitigation of electromagnetic interference and noise pickup through power lines is the critical design goal as they have the potential to jeopardize the electrical performance. We accomplish this by adhering to the ATLAS grounding rules, by avoiding ground loops and isolating the different subdetectors. Noise sources are identified and design rules to protect the SCT against them are described. A rigorous implementation of the design was crucial to achieve the required performance. This paper highlights the location, connection and assembly of the different components that affect the grounding and shielding system: cables, filters, cooling pipes, shielding enclosure, power supplies and others. Special care is taken with the electrical properties of materials and joints. The monitoring of the grounding system during the installation period is also discussed. Finally, after con...

  11. Optimization of the outer support in the ITER lower cryostat thermal shield

    Noh, C.H., E-mail: chnoh@nfri.re.kr [National Fusion Research Institute, Daejeon, 305-333 (Korea, Republic of); Korea Advanced Institute of Science and Technology, Daejeon, 305-338 (Korea, Republic of); Chung, W., E-mail: whchung@nfri.re.kr [National Fusion Research Institute, Daejeon, 305-333 (Korea, Republic of); Lim, J., E-mail: jongmin.lim@kaist.ac.kr [Korea Advanced Institute of Science and Technology, Daejeon, 305-338 (Korea, Republic of); Lee, B.C., E-mail: bclee@kaist.ac.kr [Korea Advanced Institute of Science and Technology, Daejeon, 305-338 (Korea, Republic of)

    2016-02-15

    Highlights: • Design methodology for the ITER lower cryostat thermal shield support is described. • Design optimization is performed using the kriging meta-model. • Single plate support design is proposed as an alternative design. • Proposed design has 24% reduced weight compared with the initial design. - Abstract: ITER Lower Cryostat Thermal Shield (LCTS) is fixed to the cryostat floor by a thin flexible plate support. Double plate made of titanium alloy is adopted as a reference design. Double plate is effective to get structural reliability for the high inertia load and buckling load. Thin plate with titanium alloy has good flexibility to the thermal movement and reduces conduction heat load from cryostat floor to the thermal shield. Double plate support has enough structural margin. In addition, titanium alloy requires high cost for fabrication. Design optimization is required to save manufacturing expenses. In addition to the mass minimization, design modification from double plate to single plate is proposed, because welding of double plate is difficult due to narrow gap between two plates. In this paper, design process to find optimal design of LCTS support is described. The sensitivities of the design variables such as thickness, height, width and gap between two plates are investigated. Optimal design solution is obtained by Sequential Quadratic Programming (SQP) algorithm based on the meta-model developed by randomly selected experimental samples. Through the design optimization process, optimal designs of the LCTS support are obtained. The weight of the support plates can be reduced to 24% compared with the initial design.

  12. Optimization of the outer support in the ITER lower cryostat thermal shield

    Noh, C.H.; Chung, W.; Lim, J.; Lee, B.C.

    2016-01-01

    Highlights: • Design methodology for the ITER lower cryostat thermal shield support is described. • Design optimization is performed using the kriging meta-model. • Single plate support design is proposed as an alternative design. • Proposed design has 24% reduced weight compared with the initial design. - Abstract: ITER Lower Cryostat Thermal Shield (LCTS) is fixed to the cryostat floor by a thin flexible plate support. Double plate made of titanium alloy is adopted as a reference design. Double plate is effective to get structural reliability for the high inertia load and buckling load. Thin plate with titanium alloy has good flexibility to the thermal movement and reduces conduction heat load from cryostat floor to the thermal shield. Double plate support has enough structural margin. In addition, titanium alloy requires high cost for fabrication. Design optimization is required to save manufacturing expenses. In addition to the mass minimization, design modification from double plate to single plate is proposed, because welding of double plate is difficult due to narrow gap between two plates. In this paper, design process to find optimal design of LCTS support is described. The sensitivities of the design variables such as thickness, height, width and gap between two plates are investigated. Optimal design solution is obtained by Sequential Quadratic Programming (SQP) algorithm based on the meta-model developed by randomly selected experimental samples. Through the design optimization process, optimal designs of the LCTS support are obtained. The weight of the support plates can be reduced to 24% compared with the initial design.

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

    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)

  14. Shield wall evaluation of hot cell facility for advanced spent fuel conditioning process

    Cho, I. J.; Kuk, D. H.; Ko, J. H.; Jung, W. M.; Yoo, G. S.; Lee, E. P.; Park, S. W.

    2002-01-01

    The future hot cell is located in the Irradiated Material Experiment Facility (IMEF) at the Korea Atomic Energy Research Institute (KAERI). It is β-γ type hot cell that was constructed on the base floor in IMEF building for irradiated material testing. And this hot cell will be used for carrying out the Advanced spent fuel Conditioning Process (ACP). The radiation shielding capability of hot cell should be sufficient to meet the radiation dose requirements in the related regulations. Because the radioactive sources of ACP are expected to be higher than radioactive sources of IMEF design criteria, the future hot cell in current status is unsatisfactory to hot test of ACP. So the shielding analysis of the future hot cell is performed to evaluate shielding ability of concrete shield wall. The shielding analysis included (a) identification of ACP source term; (b) photon source spectrum; (c) shielding analysis by QADS and MCNP-4C; and (d) enhancement of concrete shield wall. In this research, dose rates are obtained according to ACP source, geometry and hot cell shield wall thickness. And the evaluation and reinforcement thickness of the shield wall about future hot cell are concluded

  15. Software requirements, design, and verification and validation for the FEHM application - a finite-element heat- and mass-transfer code

    Dash, Z.V.; Robinson, B.A.; Zyvoloski, G.A.

    1997-07-01

    The requirements, design, and verification and validation of the software used in the FEHM application, a finite-element heat- and mass-transfer computer code that can simulate nonisothermal multiphase multicomponent flow in porous media, are described. The test of the DOE Code Comparison Project, Problem Five, Case A, which verifies that FEHM has correctly implemented heat and mass transfer and phase partitioning, is also covered

  16. Gas Shielding Technology for Welding and Brazing

    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."

  17. Boron filled siloxane polymers for radiation shielding

    Labouriau, Andrea; Robison, Tom; Shonrock, Clinton; Simmonds, Steve; Cox, Brad; Pacheco, Adam; Cady, Carl

    2018-03-01

    The purpose of the present work was to evaluate changes to structure-property relationships of 10B filled siloxane-based polymers when exposed to nuclear reactor radiation. Highly filled polysiloxanes were synthesized with the intent of fabricating materials that could shield high neutron fluences. The newly formulated materials consisted of cross-linked poly-diphenyl-methylsiloxane filled with natural boron and carbon nanofibers. This polymer was chosen because of its good thermal and chemical stabilities, as well as resistance to ionizing radiation thanks to the presence of aromatic groups in the siloxane backbone. Highly isotopically enriched 10B filler was used to provide an efficient neutron radiation shield, and carbon nanofibers were added to improve mechanical strength. This novel polymeric material was exposed in the Annular Core Research Reactor (ACRR) at Sandia National Labs to five different neutron/gamma fluxes consisting of very high neutron fluences within very short time periods. Thermocouples placed on the specimens recorded in-situ temperature changes during radiation exposure, which agreed well with those obtained from our MCNP simulations. Changes in the microstructural, thermal, chemical, and mechanical properties were evaluated by SEM, DSC, TGA, FT-IR NMR, solvent swelling, and uniaxial compressive load measurements. Our results demonstrate that these newly formulated materials are well-suitable to be used in applications that require exposure to different types of ionizing conditions that take place simultaneously.

  18. Discussion on variance reduction technique for shielding

    Maekawa, Fujio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    As the task of the engineering design activity of the international thermonuclear fusion experimental reactor (ITER), on 316 type stainless steel (SS316) and the compound system of SS316 and water, the shielding experiment using the D-T neutron source of FNS in Japan Atomic Energy Research Institute has been carried out. However, in these analyses, enormous working time and computing time were required for determining the Weight Window parameter. Limitation or complication was felt when the variance reduction by Weight Window method of MCNP code was carried out. For the purpose of avoiding this difficulty, investigation was performed on the effectiveness of the variance reduction by cell importance method. The conditions of calculation in all cases are shown. As the results, the distribution of fractional standard deviation (FSD) related to neutrons and gamma-ray flux in the direction of shield depth is reported. There is the optimal importance change, and when importance was increased at the same rate as that of the attenuation of neutron or gamma-ray flux, the optimal variance reduction can be done. (K.I.)

  19. Measurement of TFTR D-T radiation shielding efficiency

    Kugel, H.W.; Ascione G.; Elwood, S.

    1994-01-01

    High power D-T fusion reactor designs presently exhibit complex geometric and material density configurations. Simulations of the radiation shielding required for safe operation and full compliance with all regulatory requirements must include sufficient margin to accommodate uncertainties in material properties and distributions, uncertainties in the final configurations, and uncertainties in approximations employing the homogenization of complex geometries. Measurements of radiation shielding efficiency performed in a realistic D-T tokamak environment can provide empirical guidance for simulating safe, efficient, and cost effective shielding systems for future high power fusion reactors. In this work, the authors present the results of initial measurements of the TFTR radiation shielding efficiency during high power D-T operations with record neutron yields. The TFTR design objective is to limit the total dose-equivalent at the nearest PPPL property lines from all radiation pathways to 10 mrem per calendar year. Compliance with this design objective over a calendar year requires measurements in the presence of typical site backgrounds of about 80 mrem per year

  20. INERT GAS SHIELD FOR WELDING

    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.

  1. Transient shielded liquid hydrogen containers

    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

  2. Shielding calculational system for plutonium

    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.)

  3. Radiation-shielding transparent material

    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.)

  4. Radiation shield vest and skirt

    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

  5. Handbook of radiation shielding data

    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

  6. Rho-associated kinase activity is required for proper morphogenesis of the inner cell mass in the mouse blastocyst.

    Laeno, Arlene May A; Tamashiro, Dana Ann A; Alarcon, Vernadeth B

    2013-11-01

    The blastocyst consists of the outer layer of trophectoderm and pluripotent inner cell mass (ICM), the precursor of the placenta and fetus, respectively. During blastocyst expansion, the ICM adopts a compact, ovoidal shape, whose proper morphology is crucial for normal embryogenesis. Rho-associated kinase (ROCK), an effector of small GTPase RHO signaling, mediates the diverse cellular processes of morphogenesis, but its role in ICM morphogenesis is unclear. Here, we demonstrate that ROCK is required for cohesion of ICM cells and formation of segregated tissues called primitive endoderm (PrE) and epiblast (Epi) in the ICM of the mouse blastocyst. Blastocyst treatment with ROCK inhibitors Y-27632 and Fasudil caused widening or spreading of the ICM, and intermingling of PrE and Epi. Widening of ICM was independent of trophectoderm because isolated ICMs as well as colonies of mouse embryonic stem cells (mESC) also spread upon Y-27632 treatment. PrE, Epi, and trophectoderm cell numbers were similar between control and treated blastocysts, suggesting that ROCK inhibition affected ICM morphology but not lineage differentiation. Rock1 and Rock2 knockdown via RNA interference in mESC also induced spreading, supporting the conclusion that morphological defects caused by the pharmacological inhibitors were due to ROCK inactivation. When blastocysts were transferred into surrogates, implantation efficiencies were unaffected by ROCK inhibition, but treated blastocysts yielded greater fetal loss. These results show that proper ICM morphology is dependent on ROCK activity and is crucial for fetal development. Our studies have wider implication for improving efficiencies of human assisted reproductive technologies that diminish pregnancy loss and promote successful births.

  7. Shielding wall for thermonuclear device

    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.)

  8. Design of ITER shielding blanket

    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)

  9. Photonic Bandgap (PBG) Shielding Technology

    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

  10. Reactor vessel head permanent shield

    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

  11. ATLAS Award for Shield Supplier

    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...

  12. Radiation protection/shield design: a need for a systems approach

    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. The system approach to Clinch River Breeder Reactor Plant (CRBRP) radiation protection is described, and the program developed to implement this approach is defined. In addition, the principal shielding design problems for LMFBR nuclear reactor systems are discussed in relation 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 are discussed

  13. Gamma self-shielding correction factors calculation for aqueous bulk sample analysis by PGNAA technique

    Nasrabadi, M.N.; Mohammadi, A.; Jalali, M.

    2009-01-01

    In this paper bulk sample prompt gamma neutron activation analysis (BSPGNAA) was applied to aqueous sample analysis using a relative method. For elemental analysis of an unknown bulk sample, gamma self-shielding coefficient was required. Gamma self-shielding coefficient of unknown samples was estimated by an experimental method and also by MCNP code calculation. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the gamma self-shielding within the sample volume is required.

  14. A Reinforcement for Multifunctional Composites for Non-Parasitic Radiation Shielding, Phase II

    National Aeronautics and Space Administration — Radiation shielding is a requirement to protect humans from the hazards of space radiation during NASA missions. Multifunctional materials have the potential to...

  15. Micrometeoroid and Orbital Debris (MMOD) Shield Ballistic Limit Analysis Program

    Ryan, Shannon

    2013-01-01

    This software implements penetration limit equations for common micrometeoroid and orbital debris (MMOD) shield configurations, windows, and thermal protection systems. Allowable MMOD risk is formulated in terms of the probability of penetration (PNP) of the spacecraft pressure hull. For calculating the risk, spacecraft geometry models, mission profiles, debris environment models, and penetration limit equations for installed shielding configurations are required. Risk assessment software such as NASA's BUMPERII is used to calculate mission PNP; however, they are unsuitable for use in shield design and preliminary analysis studies. The software defines a single equation for the design and performance evaluation of common MMOD shielding configurations, windows, and thermal protection systems, along with a description of their validity range and guidelines for their application. Recommendations are based on preliminary reviews of fundamental assumptions, and accuracy in predicting experimental impact test results. The software is programmed in Visual Basic for Applications for installation as a simple add-in for Microsoft Excel. The user is directed to a graphical user interface (GUI) that requires user inputs and provides solutions directly in Microsoft Excel workbooks.

  16. Radiation shielding for the Super Collider West Utility region

    Meinke, R.; Mokhov, N.; Orth, D.; Parker, B.; Plant, D.

    1994-02-01

    Shielding considerations in the 20 x 20-TeV Superconducting Super Collider are strongly correlated with detailed machine specifics in the various accelerator sections. The West Utility, the most complex area of the Collider, concentrates all the major accelerator subsystems in a single area. The beam loss rate and associated radiation levels in this region are anticipated to be quite high, and massive radiation shielding is therefore required to protect personnel, Collider components, and the environment. The challenging task of simultaneously optimizing accelerator design and radiation shielding, both of which are strongly influenced by subsystem design details, requires the integration of several complex simulation codes. To this end we have performed exhaustive hadronic shower simulations with the MARS12 program; detailed accelerator lattice and optics optimization via the SYNCH, MAD, and MAGIC codes; and extensive 3-D configuration modeling of the accelerator tunnel and subsystems geometries. Our technique and the non-trivial results from such a combined approach are presented here. An integrated procedure is found invaluable in developing cost-effective radiation shielding solutions

  17. CASKCODES, Program CAPSIZE Scope KWIKDOSE for Shipping Cask Shielding

    1988-01-01

    1 - Description of program or function: CAPSIZE is an interactive program to rapidly determine the likely impact that proposed design objectives might have on the size and capacity of spent fuel casks designed to meet those objectives. 2 - Method of solution: Given the burnup of the spent fuel, its cooling time, the thickness of the internal basket walls, the desired external dose rate, and the nominal weight limit of the load cask, the CAPSIZE program will determine the maximum number of PWR fuel assemblies that may be shipped in a lead-, steel-, or uranium- shielded cask meeting those objectives. Using optimal packing arrangements and shielding requirements input by the user, SCOPE will design a cask to carry a single fuel assembly and then continue incrementing the number of assemblies until one or more of the design limits can no longer be met. KWIKDOSE queries the user for the number of PWR fuel assemblies in a cask, the type of cask and thickness of the shield. Upon getting the necessary input, KWIKDOSE prints out the total dose rate, 10 feet from the centerline of the cask, as a function of the burnup and cooling time of the spent fuel. 3 - Restrictions on the complexity of the problem: The restrictions are subject to the shielding requirements of the shipping cask

  18. Distance determinations to shield galaxies from Hubble space telescope imaging

    McQuinn, Kristen B. W.; Skillman, Evan D. [Minnesota Institute for Astrophysics, School of Physics and Astronomy, University of Minnesota, 116 Church Street, S.E., Minneapolis, MN 55455 (United States); Cannon, John M.; Cave, Ian [Department of Physics and Astronomy, Macalester College, 1600 Grand Avenue, Saint Paul, MN 55105 (United States); Dolphin, Andrew E. [Raytheon Company, 1151 E. Hermans Road, Tucson, AZ 85756 (United States); Salzer, John J. [Department of Astronomy, Indiana University, 727 East 3rd Street, Bloomington, IN 47405 (United States); Haynes, Martha P.; Adams, Elizabeth; Giovanelli, Riccardo [Center for Radiophysics and Space Research, Space Sciences Building, Cornell University, Ithaca, NY 14853 (United States); Elson, Ed C. [Astrophysics, Cosmology and Gravity Centre (ACGC), Department of Astronomy, University of Cape Town, Private Bag X3, Rondebosch 7701 (South Africa); Ott, Juërgen [National Radio Astronomy Observatory, P.O. Box O, 1003 Lopezville Road, Socorro, NM 87801 (United States); Saintonge, Amélie, E-mail: kmcquinn@astro.umn.edu [Max-Planck-Institute for Astrophysics, D-85741 Garching (Germany)

    2014-04-10

    The Survey of H I in Extremely Low-mass Dwarf (SHIELD) galaxies is an ongoing multi-wavelength program to characterize the gas, star formation, and evolution in gas-rich, very low-mass galaxies. The galaxies were selected from the first ∼10% of the H I Arecibo Legacy Fast ALFA (ALFALFA) survey based on their inferred low H I mass and low baryonic mass, and all systems have recent star formation. Thus, the SHIELD sample probes the faint end of the galaxy luminosity function for star-forming galaxies. Here, we measure the distances to the 12 SHIELD galaxies to be between 5 and 12 Mpc by applying the tip of the red giant method to the resolved stellar populations imaged by the Hubble Space Telescope. Based on these distances, the H I masses in the sample range from 4 × 10{sup 6} to 6 × 10{sup 7} M {sub ☉}, with a median H I mass of 1 × 10{sup 7} M {sub ☉}. The tip of the red giant branch distances are up to 73% farther than flow-model estimates in the ALFALFA catalog. Because of the relatively large uncertainties of flow-model distances, we are biased toward selecting galaxies from the ALFALFA catalog where the flow model underestimates the true distances. The measured distances allow for an assessment of the native environments around the sample members. Five of the galaxies are part of the NGC 672 and NGC 784 groups, which together constitute a single structure. One galaxy is part of a larger linear ensemble of nine systems that stretches 1.6 Mpc from end to end. Three galaxies reside in regions with 1-9 neighbors, and four galaxies are truly isolated with no known system identified within a radius of 1 Mpc.

  19. Evaluation of shielding parameters for heavy metal fluoride based tellurite-rich glasses for gamma ray shielding applications

    Sayyed, M.I.; Lakshminarayana, G.; Kityk, I.V.; Mahdi, M.A.

    2017-01-01

    In this work, we have evaluated the γ-ray shielding parameters such as mass attenuation coefficient (μ/ρ), effective atomic number (Z eff ), half value layer (HVL), mean free path (MFP) and exposure buildup factors (EBF) for heavy metal fluoride (PbF 2 ) based tellurite-rich glasses. In addition, neutron total macroscopic cross sections (∑ R ) for these glasses were also calculated. The maximum value for µ/ρ, Z eff and ∑ R was found for heavy metal (Bi 2 O 3 ) oxide introduced glass. The results of the selected glasses have been compared, in terms of MFP with different glass systems. The shielding effectiveness of the selected glasses is found comparable or better than of common ones, which indicates that these glasses with suitable oxides could be developed for gamma ray shielding applications. - Highlights: • μ/ρ, Z eff , HVL and MFP for PbF 2 based tellurite-rich glasses have been calculated. • µ/ρ and Z eff depend on the photon energy and chemical composition of the glasses. • EBF values of these glasses have been calculated using G-P fitting method. • The maximum value for µ/ρ and Z eff was found for Bi 2 O 3 oxide introduced glass. • New types of non-traditional radiation shielding glasses are demonstrated.

  20. Study of shielding analysis methods for casks of spent fuel and radioactive waste

    Saito, Ai

    2017-01-01

    Casks are used for storage or transport spent fuels or radioactive waste. Because high shielding performances are required, it is very important to confirm the validity of shielding analysis methods in order to evaluate cask shielding abilities appropriately. For this purpose, following studies were carried out. 1) A series of parameter survey for several codes to evaluated the difference of the results. 2) Calculations using the MCNP code are effective and theoretically have better accuracy. However setting reasonable variance reduction parameters is indispensable. Therefore, effectiveness of the ADVANTG code which produces automatically reasonable variance reduction parameters is carried out by comparison with conventional method. As a result, the validity of shielding analysis methods for casks is confirmed. The results will be taken into consideration in our future shielding analysis. (author)

  1. Global deformation of the Earth, surface mass anomalies, and the geodetic infrastructure required to study these processes

    Kusche, J.; Rietbroek, R.; Gunter, B.; Mark-Willem, J.

    2008-12-01

    Global deformation of the Earth can be linked to loading caused by mass changes in the atmosphere, the ocean and the terrestrial hydrosphere. World-wide geodetic observation systems like GPS, e.g., the global IGS network, can be used to study the global deformation of the Earth directly and, when other effects are properly modeled, provide information regarding the surface loading mass (e.g., to derive geo-center motion estimates). Vice versa, other observing systems that monitor mass change, either through gravitational changes (GRACE) or through a combination of in-situ and modeled quantities (e.g., the atmosphere, ocean or hydrosphere), can provide indirect information on global deformation. In the framework of the German 'Mass transport and mass distribution' program, we estimate surface mass anomalies at spherical harmonic resolution up to degree and order 30 by linking three complementary data sets in a least squares approach. Our estimates include geo-center motion and the thickness of a spatially uniform layer on top of the ocean surface (that is otherwise estimated from surface fluxes, evaporation and precipitation, and river run-off) as a time-series. As with all current Earth observing systems, each dataset has its own limitations and do not realize homogeneous coverage over the globe. To assess the impact that these limitations might have on current and future deformation and loading mass solutions, a sensitivity study was conducted. Simulated real-case and idealized solutions were explored in which the spatial distribution and quality of GPS, GRACE and OBP data sets were varied. The results show that significant improvements, e.g., over the current GRACE monthly gravity fields, in particular at the low degrees, can be achieved when these solutions are combined with present day GPS and OBP products. Our idealized scenarios also provide quantitative implications on how much surface mass change estimates may improve in the future when improved observing

  2. Improved shielding of the gonads of infants during comparative X-ray examination of the hip joints

    Petrik, R.; Reumuth, H.J.

    1979-01-01

    Improved shielding of the gonads of infants during comparative X-ray examination of the hip joints. The presented gonad shielding device for male and female infants aged 3 to 12 months has two small windows for the incident X-ray beam and allows to keep the gonad dose very low. The regions of the skeleton concerned are adequately displayed. The shield can be applied in routine examinations and does not require personnel to hold the patient. (author)

  3. Measurements and Monte-Carlo simulations of the particle self-shielding effect of B4C grains in neutron shielding concrete

    DiJulio, D. D.; Cooper-Jensen, C. P.; Llamas-Jansa, I.; Kazi, S.; Bentley, P. M.

    2018-06-01

    A combined measurement and Monte-Carlo simulation study was carried out in order to characterize the particle self-shielding effect of B4C grains in neutron shielding concrete. Several batches of a specialized neutron shielding concrete, with varying B4C grain sizes, were exposed to a 2 Å neutron beam at the R2D2 test beamline at the Institute for Energy Technology located in Kjeller, Norway. The direct and scattered neutrons were detected with a neutron detector placed behind the concrete blocks and the results were compared to Geant4 simulations. The particle self-shielding effect was included in the Geant4 simulations by calculating effective neutron cross-sections during the Monte-Carlo simulation process. It is shown that this method well reproduces the measured results. Our results show that shielding calculations for low-energy neutrons using such materials would lead to an underestimate of the shielding required for a certain design scenario if the particle self-shielding effect is not included in the calculations.

  4. Analysis of ferromagnetic shielding of the ITER NBI

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

    2006-01-01

    thickness, the coil shapes and the NBI geometry, without completely remaking the FEM model, a complete parametric modeling approach has been used. These features of the component together with the box length, height and width and the number of elements inside the ferromagnetic layer, can be changed, by simply changing a parameter value. This possibility has been used to optimize the shielding thickness and to face geometrical minor changes of the various NBI design options. At the end, the currents in the active shield coils have been optimized to achieve an overall shielding effectiveness compatible with the NBI requirements. (author)

  5. Shielded room measurements, Final report

    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.

  6. The ITER thermal shields for the magnet system: Design evolution and analysis

    Bykov, V.; Krasikov, Yu.; Grigoriev, S.; Komarov, V.; Krylov, V.; Labusov, A.; Pyrjaev, V.; Chiocchio, S.; Smirnov, V.; Sorin, V.; Tanchuk, V.

    2005-01-01

    The thermal shield (TS) system provides the required reduction of thermal loads to the cold structures operating at 4.5 K. This paper presents the rationale for the TS design evolution, details of the recent modifications that affect the TS cooling panels, the central TS ports and support system, interface labyrinths and TS structural joints. The modern results of thermal-hydraulic, thermal, seismic, static and dynamic structural analyses, that involve sub-modeling and sub-structuring finite element analysis techniques, are also reported. The modifications result in considerable reduction of TS mass, surface area and heat loads to/from the TS, simplification of TS assembly procedure and in-cryostat maintenance

  7. A history of radiation shielding of x-ray therapy rooms

    McGinley, P.H.; Miner, M.S.

    1996-01-01

    In this report the history of shielding for radiation treatment rooms is traced from the time of the discovery of x rays to the present. During the early part of the twentieth century the hazards from ionizing radiation were recognized and the use of lead and other materials became common place for shielding against x rays. Techniques for the calculation of the shield thickness needed for x ray protection were developed in the 1920's, and shielding materials were characterized in terms of the half value layer or simple exponential factors. At the same time, better knowledge of the interaction between radiation and matter was acquired. With the development of high energy medical accelerators after 1940, new and more complex shielding problems had to be addressed. Recently, shielding requirements have become more stringent as standards for exposure of personnel and the general public have been reduced. The art of shielding of radiation treatment facilities is still being developed, and the need for a revision of the reports on shielding of medical accelerators from the National Council on Radiation Protection and Measurements is emphasized in this article. (author). 61 Refs., 3 Tabs

  8. Hot-cell shielding system for high power transmission in DUPIC fuel fabrication

    Kim, K.; Lee, J.; Park, J.; Yang, M.; Park, H.

    2000-01-01

    This paper presents a newly designed hot-cell shielding system for use in the development of DUPIC (Direct Use of spent PWR fuel In CANDU reactors) fuel at KAERI (Korea Atomic Energy Research Institute). This hot-cell shielding system that was designed to transmit high power to sintering furnace in-cell from the out-of-cell through a thick cell wall has three subsystems - a steel shield plug with embedded spiral cooling line, stepped copper bus bars, and a shielding lead block. The dose-equivalent rates of the hot-cell shielding system and of the apertures between this system and the hot-cell wall were calculated. Calculated results were compared with the allowable dose limit of the existing hot-cell. Experiments for examining the temperature changes of the shielding system developed during normal furnace operation were also carried out. Finally, gamma-ray radiation survey experiments were conducted by Co-60 source. It is demonstrated that, from both calculated and experimental results, the newly designed hot-cell shielding system meets all the shielding requirements of the existing hot-cell facility, enabling high power transmission to the in-cell sintering furnace. (author)

  9. A Shielding Analysis of Hot Cell for a 10 MW Research Reactor

    Alnajjar, Alaaddin; Park, Chang Je; Roh, Gyuhong; Lee, Byunchul

    2013-01-01

    In this paper, a shielding analysis has been performed for the hot cell in a 10 MW research reactor. Two kinds of shielding analysis code systems are used such as MCNPX2.7 and M-Shield7. The first one is Monte Carlo stochastic code and the second one is a deterministic point kernel code. The results are compared in this study. In order to obtain source term, the ORIGEN-S code is used for different kinds of source. Four kinds of sources are taken into consideration. From the simulation, it is also proposed that the proper thickness of shielding material and the maximum source capacity in the hot cell. This study shows preliminary analysis results of hot cell shielding for 10MW research reactor. Total four different source terms are considered such as spent fuel assembly, Ir-192, Mo-99, and I-131. For shielding material, general concrete, heavy concrete, and lead are used. MCNPX code is mainly used for a simplified hot cell model and the result are nearly consistent when compared with M-Shield code. Required shielding thickness and the hot cell capacity are also obtained for various criterion of surface dose rates

  10. Highly heat removing radiation shielding material

    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.)

  11. Shielding design of ITER pressure suppression system

    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.)

  12. Comparision of γ -ray shielding properties of some borate glasses

    Thind, K.S.

    2003-01-01

    Several new glasses have been prepared in recent years to suit their increasing number of applications. Some of the glass compositions have distinct properties which make them the most preferred materials for certain applications such as shielding, optical fibers, electronics displays etc. The information of composition, processing and effect of environment on the glass properties is of great importance for their design and application. The shielding ability of pure elements and some mixtures have already been studied but limited attempts have been made on glasses. A good shielding glass should have high absorption cross - section for radiation and at the same time irradiation effects on its mechanical and optical properties should be small. By keeping in view of the importance of shielding ability of borate glasses, we have studied two series of different glass type: x PbO - (1-x) B 2 O 3 and x ZnO - 2xPbO - (1-3x) B 2 O 3 (where x is the mole fraction) by using narrow beam transmission method. A 2' x 2' NaI(Tl) crystal with an energy resolution of 12.5% at 662 keV of 137 Cs was used for the determination of attenuation coefficients and hence interaction cross-sections. Glass samples were prepared by using melt-quenching technique. Thickness measurement was carried out by micrometer and density was measured by Archimede's Principle using benzene as the immersion liquid. The densities of the glasses were found to increase linearly with the increase in the chemical composition of heavy metal oxide. Variations in mass attenuation coefficients and interaction cross ' sections were observed with the change in chemical composition and photon energy. It is found that these glasses have potential applications to be used as radiation shielding materials

  13. Adiabatic Demagnetization Refrigerator Field Mapping and Shielding Models for a 70 mK Superconducting Transition Edge Sensor Array and Associated Electronics

    Ladner, D. R.; Martinez-Galarce, D. S.; McCammon, D.

    2006-04-01

    An X-ray detection instrument to be flown on a sounding rocket experiment (the Advanced Technology Solar Spectroscopic Imager - ATSSI) for solar physics observations is being developed by the Lockheed Martin Solar and Astrophysics Laboratory (LMSAL). The detector is a novel class of microcalorimeter, a superconducting Transition-Edge Sensor (TES), that coupled with associated SQUID and feedback electronics requires high temperature stability at ~70 mK to resolve the energy of absorbed X-ray photons emitted from the solar corona. The cooling system incorporates an existing Adiabatic Demagnetization Refrigerator (ADR) developed at the University of Wisconsin (UW), which was previously flown to study the diffuse cosmic X-ray background. The Si thermistor detectors for that project required 130 K shielded JFET electronic components that are much less sensitive to the external field of the ADR solenoid than are the 1st (~70 mK) and 2nd (~2 K) SQUID stages used with TESs for solar observations. Modification of the Wisconsin ADR design, including TES focal plane and electronics re-positioning, therefore requires a tradeoff between the existing ADR solenoid nulling coil geometry and a low mass passive solenoid shield, while preserving the vibration isolation features of the existing design. We have developed models to accurately compute the magnetic field with and without shielding or nulling coils at critical locations to guide the re-design of the detector subsystem. The models and their application are described.

  14. Irrigoscopy - irrigography method, dosimetry and radiation shielding

    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)

  15. Cage for shield-type support. Schildausbaugestell

    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).

  16. Gamma rays shielding parameters for white metal alloys

    Kaur, Taranjot; Sharma, Jeewan; Singh, Tejbir

    2018-05-01

    In the present study, an attempt has been made to check the feasibility of white metal alloys as gamma rays shielding materials. Different combinations of cadmium, lead, tin and zinc were used to prepare quaternary alloys Pb60Sn20ZnxCd20-x (where x = 5, 10, 15) using melt quench technique. These alloys were also known as white metal alloys because of its shining appearance. The density of prepared alloys has been measured using Archimedes Principle. Gamma rays shielding parameters viz. mass attenuation coefficient (µm), effective atomic number (Zeff), electron density (Nel), Mean free path (mfp), Half value layer (HVL) and Tenth value layer (TVL) has been evaluated for these alloys in the wide energy range from 1 keV to 100 GeV. The WinXCom software has been used for obtaining mass attenuation coefficient values for the prepared alloys in the given energy range. The effective atomic number (Zeff) has been assigned to prepared alloys using atomic to electronic cross section ratio method. Further, the variation of various shielding parameters with photon energy has been investigated for the prepared white metal alloys.

  17. The radiation shielding potential of CI and CM chondrites

    Pohl, Leos; Britt, Daniel T.

    2017-03-01

    Galactic Cosmic Rays (GCRs) and Solar Energetic Particles (SEPs) pose a serious limit on the duration of deep space human missions. A shield composed of a bulk mass of material in which the incident particles deposit their energy is the simplest way to attenuate the radiation. The cost of bringing the sufficient mass from the Earth's surface is prohibitive. The shielding properties of asteroidal material, which is readily available in space, are investigated. Solution of Bethe's equation is implemented for incident protons and the application in composite materials and the significance of various correction terms are discussed; the density correction is implemented. The solution is benchmarked and shows good agreement with the results in literature which implement more correction terms within the energy ranges considered. The shielding properties of CI and CM asteroidal taxonomy groups and major asteroidal minerals are presented in terms of stopping force. The results show that CI and CM chondrites have better stopping properties than Aluminium. Beneficiation is discussed and is shown to have a significant effect on the stopping power.

  18. SEM/TIMS analysis trials on hotswipe samples taken from a shielded cell at Harwell

    Tushingham, J.; Vatter, I.; Cooke, R.

    1998-09-01

    The IAEA require advanced techniques and procedures for the detection of traces of actinides to be applied to their environmental sampling programme for nuclear safeguards as a means to detect undeclared activities. 'Swipe' samples taken from within nuclear facilities by IAEA inspectors require analysis to determine their actinide content and composition by bulk and particle measurements. The use of analytical equipment capable of analysing individual particles, particularly of actinides, is essential to optimise the IAEA's aim to monitor Member State's nuclear activities more proficiently. A trial has been undertaken at the Harwell Laboratory of AEA Technology to establish the efficacy of scanning electron microscopy (SEM) and thermal ionisation mass spectrometry (TIMS) for the particle and bulk characterisation, respectively, of actinides on samples taken from within a shielded cell. These measurements were supported by γ-spectrometry and α-spectrometry. 'Hotswipe' samples taken from within a shielded cell with a well-known recent history have been prepared for particle and bulk analysis. SEM has been used to characterise individual particles from the swipe samples and the results have been related to known cell activities. Samples were prepared for SEM using a simple procedure to minimise the potential for sample contamination. The method proved to be capable of identifying 1 μm particles that contained U, Pu, Pa and Np. The measurement of U/Pu ratios was limited to particles that contained >2% Pu in U by weight. TIMS, together with alpha spectrometry, has been used to determine the bulk actinide composition of the samples whilst gamma spectrometry has been used to determine the fission product composition. Further work to improve the potential of SEM, and also secondary ionisation mass spectrometry (SIMS), for the measurement of hotswipe samples has been proposed. (author)

  19. Neutron shielding for a 252 Cf source

    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)

  20. Radiation shielding application of lead glass

    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

  1. Radiation dose reduction by water shield

    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)

  2. Decontaminating lead bricks and shielding

    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

  3. Decontaminating lead bricks and shielding

    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

  4. Decontaminating lead bricks and shielding

    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

  5. TPX remote maintenance and shielding

    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

  6. Isotope effects on nuclear shielding

    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)

  7. Continuous electrodeionization through electrostatic shielding

    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

  8. Determination of gamma ray shielding parameters of rocks and concrete

    Obaid, Shamsan S.; Gaikwad, Dhammajyot K.; Pawar, Pravina P.

    2018-03-01

    Gamma shielding parameters such as mass attenuation coefficient (μ/ρ), effective atomic number (Zeff) and electron density (Neff) have been measured and calculated for rocks and concrete in the energy range 122-1330 keV. The measurements have been carried out at 122, 356, 511, 662, 1170, 1275, 1330 keV gamma ray energies using a gamma spectrometer includes a NaI(Tl) scintillation detector and MCA card. The atomic and electronic cross sections have also been investigated. Experimental and calculated (WinXCom) values were compared, and good agreement has been observed within the experimental error. The obtained results showed that feldspathic basalt, compact basalt, volcanic rock, dolerite and pink granite are more efficient than the sandstone and concrete for gamma ray shielding applications.

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

    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

  10. Dynamic Open-Rotor Composite Shield Impact Test Report

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

    2015-01-01

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

  11. Comments on shielding for dual energy accelerators

    Rossi, M. C.; Lincoln, H. M.; Quarin, D. J.; Zwicker, R. D.

    2008-01-01

    Determination of shielding requirements for medical linear accelerators has been greatly facilitated by the publication of the National Council on Radiation Protection and Measurements (NCRP) latest guidelines on this subject in NCRP Report No. 151. In the present report the authors review their own recent experience with patient treatments on conventional dual energy linear accelerators to examine the various input parameters needed to follow the NCRP guidelines. Some discussion is included of workloads, occupancy, use factors, and field size, with the effects of intensity modulated radiotherapy (IMRT) treatments included. Studies of collimator settings showed average values of 13.1x16.2 cm 2 for 6 MV and 14.1x16.8 cm 2 for 18 MV conventional ports, and corresponding average unblocked areas of 228 and 254 cm 2 , respectively. With an average of 77% of the field area unblocked, this gives a mean irradiated area of 196 cm 2 for the 18 MV beam, which dominates shielding considerations for most dual energy machines. Assuming conservatively small room dimensions, a gantry bin angle of 18 deg. was found to represent a reasonable unit for tabulation of use factors. For conventional 18 MV treatments it was found that the usual treatment angles of 0, 90, 180, and 270 deg. were still favored, and use factors of 0.25 represent reasonable estimates for these beams. As expected, the IMRT fields (all at 6 MV) showed a high degree of gantry angle randomization, with no bin having a use factor in excess of 0.10. It is concluded that unless a significant number of patients are treated with high energy IMRT, the traditional use factors of 0.25 are appropriate for the dominant high energy beam

  12. Comments on shielding for dual energy accelerators.

    Rossi, M C; Lincoln, H M; Quarin, D J; Zwicker, R D

    2008-06-01

    Determination of shielding requirements for medical linear accelerators has been greatly facilitated by the publication of the National Council on Radiation Protection and Measurements (NCRP) latest guidelines on this subject in NCRP Report No. 151. In the present report the authors review their own recent experience with patient treatments on conventional dual energy linear accelerators to examine the various input parameters needed to follow the NCRP guidelines. Some discussion is included of workloads, occupancy, use factors, and field size, with the effects of intensity modulated radiotherapy (IMRT) treatments included. Studies of collimator settings showed average values of 13.1 x 16.2 cm2 for 6 MV and 14.1 x 16.8 cm2 for 18 MV conventional ports, and corresponding average unblocked areas of 228 and 254 cm2, respectively. With an average of 77% of the field area unblocked, this gives a mean irradiated area of 196 cm2 for the 18 MV beam, which dominates shielding considerations for most dual energy machines. Assuming conservatively small room dimensions, a gantry bin angle of 18 degrees was found to represent a reasonable unit for tabulation of use factors. For conventional 18 MV treatments it was found that the usual treatment angles of 0, 90, 180, and 270 degrees were still favored, and use factors of 0.25 represent reasonable estimates for these beams. As expected, the IMRT fields (all at 6 MV) showed a high degree of gantry angle randomization, with no bin having a use factor in excess of 0.10. It is concluded that unless a significant number of patients are treated with high energy IMRT, the traditional use factors of 0.25 are appropriate for the dominant high energy beam.

  13. Calculation of the BREN house shielding experiments

    Woolson, William A.; Gritzner, Michael L.

    1987-01-01

    The BREN house transmission experiments provide an excellent set of measurements to validate the calculational procedures that will be used to derive house shielding estimates for the revised dosimetry of the survivors of the Hiroshima and Nagasaki A-bombs. The BREN experiments were performed in realistic full scale models of Japanese residences. Although the radiation spectra and relative intensities of neutrons and gamma rays incident on the houses from the HPRR and the 60 Co source are not appropriate for direct application to the A-bomb survivors, they cover the full energy range of importance. The codes and calculations required to compare with BREN experiments are the same as those needed for the A-bomb dosimetry. They consist of a two-dimensional discrete-ordinates calculation of the free field coupled to an adjoint Monte Carlo calculation in detailed house geometry. The agreement obtained between calculations and the experiments is excellent for neutrons and 60 Co gamma rays. Every house transmission calculation spanning simple to complex configurations and detector locations for the 60 Co and HPRR was within an acceptable margin of error. The gamma-ray TF calculations for the reactor source did not agree well with the experiments. Analysis of this discrepancy, however, strongly indicates that the problem probably does not reside in the calculational procedure but in the measurements themselves. In conclusion, it is believed that the excellent agreement of our calculations with the BREN experiments validates the calculational procedure which is planed to be applied o estimating the house shielding for survivors of the Hiroshima and Nagasaki A-bombs. Certainly, the calculations for Hiroshima and Nagasaki will involve modifications to the code used for the computations reported here, but to the extent that these modifications involve increased calculational complexity to treat more realistic materials and configurations, the benchmark established by these

  14. Shielding assessment for the proposed HRIBF upgrade to the National ISOL Facility

    Slater, C.O.; Olsen, D.K.; Johnson, J.O.; Lillie, R.A.; Gabriel, T.A.

    1997-04-01

    An upgrade of the existing ORNL Holifield Radioactive Ion Beam Facility (HRIBF) to the National Radioactive Ion Beam Isotope Separator On Line (RIB ISOL) Facility is being proposed. Part of the upgrade involves increasing the source proton energy and current, resulting in more intense, higher energy radiation. Shielding requirements for the proposed upgrade to the HRIBF have been assessed with respect to weight, space, and dose-rate constraints. Shielding assessments were made for operating, shutdown, and accident conditions. The results indicate reasonable shielding solutions for the target room except for the marginal dose rate on the roof. Shielding requirements in the target room were greatly reduced by decisions to move the target to a more interior room and to direct the proton beam downward into the target. A slightly more difficult shielding problem arises for proton beam extraction losses from the cyclotron. Here, the assumed isotropic beam losses (hence, neutron emissions) mean higher roof dose rates than those over the target room unless substantial localized shielding is placed over the cyclotron. Shutdown dose rates were found to present no problems. While dose rates through the sides of the facility during accident conditions will probably satisfy the accident dose-rate constraints, dose rates above the roof will be well above the constraints unless a solution is devised to shield the locations where beam losses are likely to occur. Ground activation analysis was postponed for this study

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

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

    2016-01-01

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

  16. Gamma radiation shielding and optical properties measurements of zinc bismuth borate glasses

    Yasaka, P.; Pattanaboonmee, N.; Kim, H.J.; Limkitjaroenporn, P.; Kaewkhao, J.

    2014-01-01

    Highlights: • 10ZnO:xBi 2 O 3 :(90−x)B 2 O 3 , (ZBB) glasses were prepared. • Radiation shielding and optical properties were investigated. • Higher 25 mol% of Bi 2 O 3 show better shielding property compared with concretes. • ZBB glasses can develop as a Pb-free radiation shielding material. - Abstract: In this work, the zinc bismuth borate (ZBB) glasses of the composition 10ZnO:xBi 2 O 3 :(90−x)B 2 O 3 (where x = 15, 20, 25 and 30 mol%) were prepared by the melt quenching technique. Their radiation shielding and optical properties were investigated and compared with theoretical calculations. The mass attenuation coefficients of ZBB glasses have been measured at different energies obtained from a Compton scattering technique. The results show a decrease of the mass attenuation coefficient, effective atomic number and effective electron density values with increasing of gamma-ray energies; and good agreements between experimental and theoretical values. The glass samples with Bi 2 O 3 concentrations higher than 25 mol% (25 and 30 mol%) were observed with lower mean free path (MFP) values than all the standard shielding concretes studied. These results are indications that the ZBB glasses in the present study may be developed as a lead-free radiation shielding material in the investigated energy range

  17. Development of a new measurement method for fast breeder reactor fuel burnup using a shielded ion microprobe analyzer

    Mizuno, M.; Enokido, Y.; Itaki, T.; Kono, K.; Unno, I.; Yamanouchi, S.

    1985-01-01

    A new method of burnup measurement using a shielded ion microprobe analyzer (SIMA) has been developed. The method is based on the isotope analysis of uranium, plutonium, and fission products in irradiated mixed oxide fuel by means of secondary ion mass spectrometry (SIMS). Fourteen samples irradiated in the Japanese experimental fast reactor JOYO were examined. The maximum local burnup of JOYO MK-I core fuels was about5.1 at. %. The axial burnup distribution of the fuel pin was in good agreement with that of the sibling pin in the same subassembly, measured by surface ionization mass spectrometry, which requires the chemical separation of fission products and heavy metals. The new method facilitates the rapid and accurate measurement of fast breeder reactor fuel burnup without human radiation exposure during sample preparation and analysis

  18. Comparison of different methods for shielding design in computed tomography

    Ciraj-Bjelac, O.; Arandjic, D.; Kosutic, D.

    2011-01-01

    The purpose of this work is to compare different methods for shielding calculation in computed tomography (CT). The BIR-IPEM (British Inst. of Radiology and Inst. of Physics in Engineering in Medicine) and NCRP (National Council on Radiation Protection) method were used for shielding thickness calculation. Scattered dose levels and calculated barrier thickness were also compared with those obtained by scatter dose measurements in the vicinity of a dedicated CT unit. Minimal requirement for protective barriers based on BIR-IPEM method ranged between 1.1 and 1.4 mm of lead demonstrating underestimation of up to 20 % and overestimation of up to 30 % when compared with thicknesses based on measured dose levels. For NCRP method, calculated thicknesses were 33 % higher (27-42 %). BIR-IPEM methodology-based results were comparable with values based on scattered dose measurements, while results obtained using NCRP methodology demonstrated an overestimation of the minimal required barrier thickness. (authors)

  19. INTOR first wall/blanket/shield activity

    Gohar, Y.; Billone, M.C.; Cha, Y.S.; Finn, P.A.; Hassanein, A.M.; Liu, Y.Y.; Majumdar, S.; Picologlou, B.F.; Smith, D.L.

    1986-01-01

    The main emphasis of the INTOR first wall/blanket/shield (FWBS) during this period has been upon the tritium breeding issues. The objective is to develop a FWBS concept which produces the tritium requirement for INTOR operation and uses a small fraction of the first wall surface area. The FWBS is constrained by the dimensions of the reference design and the protection criteria required for different reactor components. The blanket extrapolation to commercial power reactor conditions and the proper temperature for power extraction have been sacrificed to achieve the highest possible local tritium breeding ratio (TBR). In addition, several other factors that have been considered in the blanket survey study include safety, reliability, lifetime fluence, number of burn cycles, simplicity, cost, and development issues. The implications of different tritium supply scenarios were discussed from the cost and availability for INTOR conditions. A wide variety of blanket options was explored in a preliminary way to determine feasibility and to see if they can satisfy the INTOR conditions. This survey and related issues are summarized in this report. Also discussed are material design requirements, thermal hydraulic considerations, structure analyses, tritium permeation through the first wall into the coolant, and tritium inventory

  20. Fabrication of prototype mockups of ITER shielding blanket with separable first wall

    Kosaku, Yasuo; Kuroda, Toshimasa; Enoeda, Mikio; Hatano, Toshihisa; Sato, Satoshi; Akiba, Masato

    2002-07-01

    Design of shielding blanket for ITER-FEAT applies the first wall which has the separable structure from the shield block for the purpose of radio-active waste reduction in the maintenance work and cost reduction in fabrication process. Also, it is required to have various types of slots in both of the first wall and the shield block, to reduce the eddy current for reduction of electro-magnetic force in disruption events. This report summarizes the demonstrative fabrication of the ITER shielding blanket with separable first wall performed for the shielding blanket fabrication technology development, under the task agreement of G 16 TT 108 FJ (T420-2) in ITER Engineering Design Activity Extension Period. The objectives of the demonstrative fabrication are: to demonstrate the comprehensive fabrication technique in a large scale component (e.g the joining techniques for the beryllium armor/copper alloy and copper alloy/SS, and the slotting method of the FW and shield block); to develop an improved fabrication method for the shielding blanket based on the ITER-FEAT updated design. In this work, the fabrication technique of full scale separable first wall shield blanket was confirmed by fabricating full width Be armored first wall panel, full scale of 1/2 shield block with poloidal cooling channels. As the R and D for updated cooling channel configuration, the fabrication technique of the radial channel shield block was also demonstrated. Concluding to the all R and D results, it was demonstrated successfully that the fabrication technique and optimized conditions in the results obtained under the task agreement of G 16 TT 95 FJ (T420-1) was applicable to the prototype of the separable first wall blanket module. Additionally, basic echo data of ultra-sonic test method (UT) was obtained to show the applicability of UT method for in tube access detection of defect on the Cu alloy/SS tube interface. (author)

  1. Work for radiation shielding concrete in large-scaled radiation facilities

    Konomi, Shinzo; Sato, Shoni; Otake, Takao.

    1980-01-01

    This paper reports the radiation shielding concrete work in the construction of radiation laboratory facilities of Electrotechnical Laboratory, a Japanese Government agency for the research and development of electronic technology. The radiation shielding walls of the facilities are made of ordinary concrete, heavy weight concrete and raw iron ore. This paper particularly relates the use of ordinary concrete which constitutes the majority of such concretes. The concrete mix was determined so as to increase its specific gravity for better shielding effect, to improve mass concrete effect and to advance good workability. The tendency of the concrete to decrease its specific gravity and the temperature variations were also made on how to place concrete to secure good shielding effect and uniform quality. (author)

  2. Shielding of Medical Facilities. Shielding Design Considerations for PET-CT Facilities

    Cruzate, J.A.; Discacciatti, A.P.

    2011-01-01

    The radiological evaluation of a Positron Emission Tomography (PET) facility consists of the assessment of the annual effective dose both to workers occupationally exposed, and to members of the public. This assessment takes into account the radionuclides involved, the facility features, the working procedures, the expected number of patients per year, and so on. The evaluation embraces the distributions of rooms, the thickness and physical material of walls, floors and ceilings. This work detail the methodology used for making the assessment of a PET facility design taking into account only radioprotection aspects. The assessment results must be compared to the design requirements established by national regulations in order to determine whether or not, the facility complies with those requirements, both for workers and for members of the public. The analysis presented is useful for both, facility designers and regulators. In addition, some guidelines for improving the shielding design and working procedures are presented in order to help facility designer's job. (authors)

  3. Nomogram for Determining Shield Thickness for Point and Line Sources of Gamma Rays

    Joenemalm, C.; Malen, K

    1966-10-01

    A set of nomograms is given for the determination of the required shield thickness against gamma radiation. The sources handled are point and infinite line sources with shields of Pb, Fe, magnetite concrete (p = 3.6), ordinary concrete (p = 2.3) or water. The gamma energy range covered is 0.5 - 10 MeV. The nomograms are directly applicable for source and dose points on the surfaces of the shield. They can easily be extended to source and dose points in other positions by applying a geometrical correction. Also included are data for calculation of the source strength for the most common materials and for fission product sources

  4. Nomogram for Determining Shield Thickness for Point and Line Sources of Gamma Rays

    Joenemalm, C; Malen, K

    1966-10-15

    A set of nomograms is given for the determination of the required shield thickness against gamma radiation. The sources handled are point and infinite line sources with shields of Pb, Fe, magnetite concrete (p = 3.6), ordinary concrete (p = 2.3) or water. The gamma energy range covered is 0.5 - 10 MeV. The nomograms are directly applicable for source and dose points on the surfaces of the shield. They can easily be extended to source and dose points in other positions by applying a geometrical correction. Also included are data for calculation of the source strength for the most common materials and for fission product sources.

  5. Simbol-X Mirror Module Thermal Shields: II-Small Angle X-Ray Scattering Measurements

    Barbera, M.; Ayers, T.; Collura, A.; Nasillo, G.; Pareschi, G.; Tagliaferri, G.

    2009-05-01

    The formation flight configuration of the Simbol-X mission implies that the X-ray mirror module will be open to Space on both ends. In order to reduce the power required to maintain the thermal stability and, therefore, the high angular resolution of the shell optics, a thin foil thermal shield will cover the mirror module. Different options are presently being studied for the foil material of these shields. We report results of an experimental investigation conducted to verify that the scattering of X-rays, by interaction with the thin foil material of the thermal shield, will not significantly affect the performances of the telescope.

  6. Simbol-X Mirror Module Thermal Shields: II-Small Angle X-Ray Scattering Measurements

    Barbera, M.; Ayers, T.; Collura, A.; Nasillo, G.; Pareschi, G.; Tagliaferri, G.

    2009-01-01

    The formation flight configuration of the Simbol-X mission implies that the X-ray mirror module will be open to Space on both ends. In order to reduce the power required to maintain the thermal stability and, therefore, the high angular resolution of the shell optics, a thin foil thermal shield will cover the mirror module. Different options are presently being studied for the foil material of these shields. We report results of an experimental investigation conducted to verify that the scattering of X-rays, by interaction with the thin foil material of the thermal shield, will not significantly affect the performances of the telescope.

  7. Shielding calculations using computer techniques; Calculo de blindajes mediante tecnicas de computacion

    Gonzalez Portilla, M. I.; Marquez, J.

    2011-07-01

    Radiological protection aims to limit the ionizing radiation received by people and equipment, which in numerous occasions requires of protection shields. Although, for certain configurations, there are analytical formulas, to characterize these shields, the design setup may be very intensive in numerical calculations, therefore the most efficient from to design the shields is by means of computer programs to calculate dose and dose rates. In the present article we review the codes most frequently used to perform these calculations, and the techniques used by such codes. (Author) 13 refs.

  8. Effect of low-Z absorber's thickness on gamma-ray shielding parameters

    Mann, Kulwinder Singh, E-mail: ksmann6268@gmail.com [Department of Applied Sciences, Punjab Technical University, Kapurthala 144601 (India); Department of Physics, D.A.V. College, Bathinda 151001, Punjab (India); Heer, Manmohan Singh [Department of Physics, Kanya Maha Vidyalaya, Jalandhar 144001 (India); Rani, Asha [Department of Applied Sciences, Ferozpur College of Engineering and Technology, Ferozshah, Ferozpur 142052 (India)

    2015-10-11

    Gamma ray shielding behaviour of any material can be studied by various interaction parameters such as total mass attenuation coefficient (μ{sub m}); half value layer (HVL); tenth value layer (TVL); effective atomic number (Z{sub eff}), electron density (N{sub el}), effective atomic weight (A{sub eff}) and buildup factor. For gamma rays, the accurate measurements of μ{sub m} (cm{sup 2} g{sup −1}) theoretically require perfect narrow beam irradiation geometry. However, the practical geometries used for the experimental investigations deviate from perfect-narrowness thereby the multiple scattered photons cause systematic errors in the measured values of μ{sub m}. Present investigation is an attempt to find the optimum value of absorber thickness (low-Z) for which these errors are insignificant and acceptable. Both experimental and theoretical calculations have been performed to investigate the effect of absorber's thickness on μ{sub m} of six low-Z (10shielding parameters of any material. Good agreement of theoretical and measured values of μ{sub m} was observed for all absorbers with thickness ≤0.5 mean free paths, thus considered it as optimum thickness for low-Z materials in the selected energy range. White cement was found to possess maximum shielding effectiveness for the selected gamma rays. - Highlights: • Optimum thickness value is 0.5 mfp for low-Z absorbers in energy range 662–1332 keV. • For accurate measurement of μ{sub m} absorber's thickness should be ≤optimum thickness. • GRIC2-toolkit is useful for γ-ray shielding analysis of composite materials.

  9. Analysis of Slewing and Attitude Determination Requirements for CTEx

    2010-03-01

    shield for radiation protection. • Mass : 375 g • Dimensions: 108 x 106 x 49 mm3 • Field of View: 138o x 138o unobstructed • Power Consumption: ɘ.25 W... Mass : 24 g • Dimensions: 30 x 30 x 14.5 mm3 • Field of View: 160o, full cone angle • Power Consumption: Passive • Operating Temperature: -50oC to +85oC...12 0.43 – 1.05 * 1.50 – 1.80 2.00 – 2.50 8.00 – 12.00 * Customised according user requirements FLI/PMI Fluorescence Line Imager

  10. Improved Assembly for Gas Shielding During Welding or Brazing

    Gradl, Paul; Baker, Kevin; Weeks, Jack

    2009-01-01

    An improved assembly for inert-gas shielding of a metallic joint is designed to be useable during any of a variety of both laser-based and traditional welding and brazing processes. The basic purpose of this assembly or of a typical prior related assembly is to channel the flow of a chemically inert gas to a joint to prevent environmental contamination of the joint during the welding or brazing process and, if required, to accelerate cooling upon completion of the process.

  11. Safety aspects and shield design of a Poton irradiator

    Mehta, S.K.; Nayak, A.R.; Bongirwar, D.R.; Modi, R.K.; Ramkumar, M.S.

    1998-01-01

    An irradiation plant, POTON, for irradiation of potatoes and onions is being set up at Nashik. Shield design and safety features of this plant incorporate some novel and innovative features like a compact cell, curved cell boundaries for smooth conveyor movement though the cell labyrinth and conform to ICRP and AERB design safety requirements. The safety features include multiple safety interlocks, audio-visual alarms, scram switches and trip wire for avoiding accidental exposures. (author)

  12. Rational Design of Tunnel Supports: An Interactive Graphics Based Analysis of the Support Requirements of Excavations in Jointed Rock Masses.

    1979-09-01

    joint orientetion and joint slippage than to failure of the intact rock mass. Dixon (1971) noted the importance of including the confining influence of...dedicated computer. The area of research not covered by this investigation which holds promise for a future study is a detailed comparison of the results of...block data, type key "W". The program writes this data on Linc tapes for future retripval. This feature can be used to store the consolidated block

  13. Improvements in or relating to nuclear shields

    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.)

  14. Preliminary radiation shielding design for BOOMERANG

    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

  15. Actively shielded low level gamma - spectrometric system

    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

  16. Optimization of multi-layered metallic shield

    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.

  17. Alignment modification for pencil eye shields

    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

  18. Several problems in accelerator shielding study

    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.)

  19. Collection shield for ion separation apparatus

    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

  20. Current status of methods for shielding analysis

    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

  1. Thermal neutron shield and method of manufacture

    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.

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

    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)

  3. Passive radiation shielding considerations for the proposed space elevator

    Jorgensen, A. M.; Patamia, S. E.; Gassend, B.

    2007-02-01

    The Earth's natural van Allen radiation belts present a serious hazard to space travel in general, and to travel on the space elevator in particular. The average radiation level is sufficiently high that it can cause radiation sickness, and perhaps death, for humans spending more than a brief period of time in the belts without shielding. The exact dose and the level of the related hazard depends on the type or radiation, the intensity of the radiation, the length of exposure, and on any shielding introduced. For the space elevator the radiation concern is particularly critical since it passes through the most intense regions of the radiation belts. The only humans who have ever traveled through the radiation belts have been the Apollo astronauts. They received radiation doses up to approximately 1 rem over a time interval less than an hour. A vehicle climbing the space elevator travels approximately 200 times slower than the moon rockets did, which would result in an extremely high dose up to approximately 200 rem under similar conditions, in a timespan of a few days. Technological systems on the space elevator, which spend prolonged periods of time in the radiation belts, may also be affected by the high radiation levels. In this paper we will give an overview of the radiation belts in terms relevant to space elevator studies. We will then compute the expected radiation doses, and evaluate the required level of shielding. We concentrate on passive shielding using aluminum, but also look briefly at active shielding using magnetic fields. We also look at the effect of moving the space elevator anchor point and increasing the speed of the climber. Each of these mitigation mechanisms will result in a performance decrease, cost increase, and technical complications for the space elevator.

  4. Improvement of top shield analysis technology for CANDU 6 reactor

    Kim, Kyo Yoon; Jin, Young Kwon; Lee, Sung Hee; Moon, Bok Ja; Kim, Yong Il

    1996-07-01

    As for Wolsung NPP unit 1, radiation shielding analysis was performed by using neutron diffusion codes, one-dimensional discrete ordinates code ANISN, and analytical methods. But for Wolsung NPP unit 2, 3, and 4, two-dimensional discrete ordinates code DOT substituted for neutron diffusion codes. In other words, the method of analysis and computer codes used for radiation shielding of CANDU 6 type reactor have been improved. Recently Monte Carlo MCNP code has been widely utilized in the field of radiation physics and other radiation related areas because it can describe an object sophisticately by use of three-dimensional modelling and can adopt continuous energy cross-section library. Nowadays Monte Carlo method has been reported to be competitive to discrete ordinate method in the field of radiation shielding and the former has been known to be superior to the latter for complex geometry problem. However, Monte Carlo method had not been used for radiation streaming calculation in the shielding design of CANDU type reactor. Neutron and gamma radiations are expected to be streamed from calandria through the penetrations to reactivity mechanism deck (R/M deck) because many reactivity control units which are established on R/M deck extend from R/M deck to calandria within penetrations, which are provided by guide tube extensions. More precise estimation of radiation streaming is required because R/M deck is classified as an accessible area where atomic worker can access when necessary. Therefore neutron and gamma dose rates were estimated using MCNP code on the R/M deck in the top shield system of CANDU 6 reactor. 9 tabs., 17 figs., 21 refs. (Author)

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

    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.

  6. Radiation Shielding Systems Using Nanotechnology

    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.

  7. Radiation shielding and safety design

    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

  8. Evaluation of the radiation field and shielding assessment of the experimental area of HIE-ISOLDE

    Romanets, Y.; Goncalves, I.F.; Maria, S. di; Vaz, P.; Vollaire, J.; Bernardes, A.P.; Dorsival, A.; Kadi, Y.; Vlachoudis, V.

    2014-01-01

    The ISOLDE facility at CERN is one of the first facilities in the world dedicated to the production of the radioactive ion beams (RIB) and during all its working time underwent several upgrades. The goal of the latest proposed upgrade, 'The High Intensity and Energy ISOLDE' (HIE-ISOLDE), is to provide a higher performance facility in order to approximate it to the level of the next generation ISOL facilities, like EURISOL. The HIE-ISOLDE aims to improve significantly the quality of the produced RIB and for this reason the increasing of the primary beam power is one of the main objectives of the project. An increase in the nominal beam current (from 2 to 6 μA proton beam intensity) and energy (from 1.4 GeV to 2 GeV) of the primary proton beam will be possible due to the upgrade of CERN's accelerator infrastructure. The current upgrade means reassessment of the radiation protection and the radiation safety of the facility. However, an evaluation of the existing shielding configuration and access restrictions to the experimental and supply areas must be carried out. Monte Carlo calculations were performed in order to evaluate the radiation protection of the facility as well as radiation shielding assessment and design. The FLUKA-Monte Carlo code was used in this study to calculate the ambient dose rate distribution and particle fluxes in the most important areas, such as the experimental hall of the facility. The results indicate a significant increase in the ambient dose equivalent rate in some areas of the experimental hall when an upgrade configuration of the primary proton beam is considered. Special attention is required for the shielding of the target area once it is the main and very intensive radiation source, especially under the upgrade conditions. In this study, the access points to the beam extraction and beam maintenance areas, such as the mass separator rooms and the high voltage room, are identified as the most sensitive for the experimental hall from

  9. Shield verification and validation action matrix summary

    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

  10. A perturbation technique for shield weight minimization

    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)

  11. Isotopic dependence of GCR fluence behind shielding

    Cucinotta, Francis A.; Wilson, John W.; Saganti, Premkumar; Hu, Xiaodong; Kim, Myung-Hee Y.; Cleghorn, Timothy; Zeitlin, Cary; Tripathi, Ram K.

    2006-01-01

    In this paper we consider the effects of the isotopic composition of the primary galactic cosmic rays (GCR), nuclear fragmentation cross sections, and isotopic-grid on the solution to transport models used for shielding studies. Satellite measurements are used to describe the isotopic composition of the GCR. For the nuclear interaction data-base and transport solution, we use the quantum multiple scattering theory of nuclear fragmentation (QMSFRG) and high-charge and energy (HZETRN) transport code, respectively. The QMSFRG model is shown to accurately describe existing fragmentation data including proper description of the odd-even effects as function of the iso-spin dependence on the projectile nucleus. The principle finding of this study is that large errors (±100%) will occur in the mass-fluence spectra when comparing transport models that use a complete isotopic-grid (∼170 ions) to ones that use a reduced isotopic-grid, for example the 59 ion-grid used in the HZETRN code in the past; however, less significant errors (<+/-20%) occur in the elemental-fluence spectra. Because a complete isotopic-grid is readily handled on small computer workstations and is needed for several applications studying GCR propagation and scattering, it is recommended that they be used for future GCR studies

  12. The use of nipple shields: A review

    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.

  13. A shielding chamber for the Rossendorf whole-body counter; Eine Abschirmkammer fuer den Rossendorfer Ganzkoerperzaehler

    Ebert, Stephan; Loehnert, Daniela [VKTA - Strahlenschutz, Analytik und Entsorgung Rossendorf e.V., Dresden (Germany).; Schreiber, Wolfgang [Ingenieurbedarf G. Schoene und W. Schreiber GmbH, Radeberg (Germany)

    2017-08-01

    A large part of radionuclides to be evaluated in the frame of incorporation monitoring are gamma emitter that can be directly measured (in vivo). In order to reach the required detection limits for the relevant radionuclides in the short measuring time high effort is necessary for the shielding of natural radioactivity. The contribution describes planning, construction and installation of the shielding chamber for the new in vivo whole body counter in Rossendorf. First experiences and preliminary results are discussed.

  14. Lessons from shielding retrofits at the LAMPF/LANSCE/PSR accelerator, beam lines and target facilities

    Macek, R.J.

    1994-01-01

    The experience in the past 7 years to improve the shielding and radiation control systems at the Los Alamos Meson Physics Facility (LAMPF) and the Manuel Lujan Jr. Neutron Scattering Center (LANSCE) provides important lessons for the design of radiation control systems at future, high beam power proton accelerator facilities. Major issues confronted and insight gained in developing shielding criteria and in the use of radiation interlocks are discussed. For accelerators and beam lines requiring hands-on-maintenance, our experience suggests that shielding criteria based on accident scenarios will be more demanding than criteria based on routinely encountered beam losses. Specification and analysis of the appropriate design basis accident become all important. Mitigation by active protection systems of the consequences of potential, but severe, prompt radiation accidents has been advocated as an alternate choice to shielding retrofits for risk management at both facilities. Acceptance of active protection systems has proven elusive primarily because of the difficulty in providing convincing proof that failure of active systems (to mitigate the accident) is incredible. Results from extensive shielding assessment studies are presented including data from experimental beam spill tests, comparisons with model estimates, and evidence bearing on the limitations of line-of-sight attenuation models in complex geometries. The scope and significant characteristics of major shielding retrofit projects at the LAMPF site are illustrated by the project to improve the shielding beneath a road over a multiuse, high-intensity beam line (Line D)

  15. Securing the data economy: translating privacy and enacting security in the development of DataSHIELD.

    Murtagh, M J; Demir, I; Jenkings, K N; Wallace, S E; Murtagh, B; Boniol, M; Bota, M; Laflamme, P; Boffetta, P; Ferretti, V; Burton, P R

    2012-01-01

    Contemporary bioscience is seeing the emergence of a new data economy: with data as its fundamental unit of exchange. While sharing data within this new 'economy' provides many potential advantages, the sharing of individual data raises important social and ethical concerns. We examine ongoing development of one technology, DataSHIELD, which appears to elide privacy concerns about sharing data by enabling shared analysis while not actually sharing any individual-level data. We combine presentation of the development of DataSHIELD with presentation of an ethnographic study of a workshop to test the technology. DataSHIELD produced an application of the norm of privacy that was practical, flexible and operationalizable in researchers' everyday activities, and one which fulfilled the requirements of ethics committees. We demonstrated that an analysis run via DataSHIELD could precisely replicate results produced by a standard analysis where all data are physically pooled and analyzed together. In developing DataSHIELD, the ethical concept of privacy was transformed into an issue of security. Development of DataSHIELD was based on social practices as well as scientific and ethical motivations. Therefore, the 'success' of DataSHIELD would, likewise, be dependent on more than just the mathematics and the security of the technology. Copyright © 2012 S. Karger AG, Basel.

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

    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

  17. Conservative method for determination of material thickness used in shielding of veterinary facilities

    Lava, Deise D.; Borges, Diogo da S.; Affonso, Renato R.W.; Moreira, Maria de L.; Guimaraes, Antonio C.F.

    2014-01-01

    For determination of an effective method for shielding of veterinary rooms, was provided shielding methods generally used in rooms which works with X-ray production and radiotherapy. Every calculation procedure is based in traditional variables used to transmission calculation. The thickness of the materials used for primary and secondary shieldings are obtained to respect the limits set by the Brazilian National Nuclear Energy Commission (CNEN). This work presents the development of a computer code in order to serve as a practical tool for determining rapid and effective materials and their thicknesses to shield veterinary facilities. The code determines transmission values of the shieldings and compares them with data from transmission 'maps' provided by NCRP-148 report. These 'maps' were added to the algorithm through interpolation techniques of curves of materials used for shielding. Each interpolation generates about 1,000,000 points that are used to generate a new curve. The new curve is subjected to regression techniques, which makes possible to obtain nine degree polynomial, and exponential equations. These equations whose variables consist of transmission of values, enable trace all the points of this curve with high precision. The data obtained from the algorithm were satisfactory with official data presented by the National Council of Radiation Protection and Measurements (NCRP) and can contribute as a practical tool for verification of shielding of veterinary facilities that require using Radiotherapy techniques and X-ray production

  18. Pb-free Radiation Shielding Glass Using Coal Fly Ash

    Watcharin Rachniyom

    2015-12-01

    Full Text Available In this work, Pb-free shielding glass samples were prepared by the melt quenching technique using subbituminous fly ash (SFA composed of xBi2O3 : (60-xB2O3 : 10Na2O : 30SFA (where x = 10, 15, 20, 25, 30 and 35 by wt%. The samples were investigated for their physical and radiation shielding properties. The density and hardness were measured. The results showed that the density increased with the increase of Bi2O3 content. The highest value of hardness was observed for glass sample with 30 wt% of Bi2O3 concentration. The samples were investigated under 662 keV gamma ray and the results were compared with theoretical calculations. The values of the mass attenuation coefficient (μm, the atomic cross section (σe and the effective atomic number (Zeff were found to increase with an increase of the Bi2O3 concentration and were in good agreement with the theoretical calculations. The best results for the half-value layer (HVL were observed in the sample with 35 wt% of Bi2O3 concentration, better than the values of barite concrete. These results demonstrate the viability of using coal fly ash waste for radiation shielding glass without PbO in the glass matrices.

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

    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...

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

    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...

  1. Gamma ray shielding study of barium-bismuth-borosilicate glasses as transparent shielding materials using MCNP-4C code, XCOM program, and available experimental data

    Bagheri, Reza; Yousefinia, Hassan [Nuclear Fuel Cycle Research School (NFCRS), Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of); Moghaddam, Alireza Khorrami [Radiology Department, Paramedical Faculty, Mazandaran University of Medical Sciences, Sari (Iran, Islamic Republic of)

    2017-02-15

    In this work, linear and mass attenuation coefficients, effective atomic number and electron density, mean free paths, and half value layer and 10th value layer values of barium-bismuth-borosilicate glasses were obtained for 662 keV, 1,173 keV, and 1,332 keV gamma ray energies using MCNP-4C code and XCOM program. Then obtained data were compared with available experimental data. The MCNP-4C code and XCOM program results were in good agreement with the experimental data. Barium-bismuth-borosilicate glasses have good gamma ray shielding properties from the shielding point of view.

  2. Gamma Ray Shielding Study of Barium–Bismuth–Borosilicate Glasses as Transparent Shielding Materials using MCNP-4C Code, XCOM Program, and Available Experimental Data

    Reza Bagheri

    2017-02-01

    Full Text Available In this work, linear and mass attenuation coefficients, effective atomic number and electron density, mean free paths, and half value layer and 10th value layer values of barium–bismuth–borosilicate glasses were obtained for 662 keV, 1,173 keV, and 1,332 keV gamma ray energies using MCNP-4C code and XCOM program. Then obtained data were compared with available experimental data. The MCNP-4C code and XCOM program results were in good agreement with the experimental data. Barium–bismuth–borosilicate glasses have good gamma ray shielding properties from the shielding point of view.

  3. Role of Outgassing of ITER Vacuum Vessel In-Wall Shielding Materials in Leak Detection of ITER Vacuum Vessel

    Maheshwari, A.; Pathak, H. A.; Mehta, B. K.; Phull, G. S.; Laad, R.; Shaikh, M. S.; George, S.; Joshi, K.; Khan, Z.

    2017-04-01

    ITER Vacuum Vessel is a torus-shaped, double wall structure. The space between the double walls of the VV is filled with In-Wall Shielding Blocks (IWS) and Water. The main purpose of IWS is to provide neutron shielding during ITER plasma operation and to reduce ripple of Toroidal Magnetic Field (TF). Although In-Wall Shield Blocks (IWS) will be submerged in water in between the walls of the ITER Vacuum Vessel (VV), Outgassing Rate (OGR) of IWS materials plays a significant role in leak detection of Vacuum Vessel of ITER. Thermal Outgassing Rate of a material critically depends on the Surface Roughness of material. During leak detection process using RGA equipped Leak detector and tracer gas Helium, there will be a spill over of mass 3 and mass 2 to mass 4 which creates a background reading. Helium background will have contribution of Hydrogen too. So it is necessary to ensure the low OGR of Hydrogen. To achieve an effective leak test it is required to obtain a background below 1 × 10-8 mbar 1 s-1 and hence the maximum Outgassing rate of IWS Materials should comply with the maximum Outgassing rate required for hydrogen i.e. 1 x 10-10 mbar 1 s-1 cm-2 at room temperature. As IWS Materials are special materials developed for ITER project, it is necessary to ensure the compliance of Outgassing rate with the requirement. There is a possibility of diffusing the gasses in material at the time of production. So, to validate the production process of materials as well as manufacturing of final product from this material, three coupons of each IWS material have been manufactured with the same technique which is being used in manufacturing of IWS blocks. Manufacturing records of these coupons have been approved by ITER-IO (International Organization). Outgassing rates of these coupons have been measured at room temperature and found in acceptable limit to obtain the required Helium Background. On the basis of these measurements, test reports have been generated and got

  4. Shielding concerns at a spallation source

    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

  5. Thermal testing of solid neutron shielding materials

    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

  6. Development of neutron shielding material for cask

    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

  7. Heating profiles on ICRF antenna Faraday shields

    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

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

    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

  9. Shielding design for the front end of the CERN SPL.

    Magistris, Matteo; Silari, Marco; Vincke, Helmut

    2005-01-01

    CERN is designing a 2.2-GeV Superconducting Proton Linac (SPL) with a beam power of 4 MW, to be used for the production of a neutrino superbeam. The SPL front end will initially accelerate 2 x 10(14) negative hydrogen ions per second up to an energy of 120 MeV. The FLUKA Monte Carlo code was employed for shielding design. The proposed shielding is a combined iron-concrete structure, which also takes into consideration the required RF wave-guide ducts and access labyrinths to the machine. Two beam-loss scenarios were investigated: (1) constant beam loss of 1 Wm(-1) over the whole accelerator length and (2) full beam loss occurring at various locations. A comparison with results based on simplified approaches is also presented.

  10. Quantum Devices Bonded Beneath a Superconducting Shield: Part 2

    McRae, Corey Rae; Abdallah, Adel; Bejanin, Jeremy; Earnest, Carolyn; McConkey, Thomas; Pagel, Zachary; Mariantoni, Matteo

    The next-generation quantum computer will rely on physical quantum bits (qubits) organized into arrays to form error-robust logical qubits. In the superconducting quantum circuit implementation, this architecture will require the use of larger and larger chip sizes. In order for on-chip superconducting quantum computers to be scalable, various issues found in large chips must be addressed, including the suppression of box modes (due to the sample holder) and the suppression of slot modes (due to fractured ground planes). By bonding a metallized shield layer over a superconducting circuit using thin-film indium as a bonding agent, we have demonstrated proof of concept of an extensible circuit architecture that holds the key to the suppression of spurious modes. Microwave characterization of shielded transmission lines and measurement of superconducting resonators were compared to identical unshielded devices. The elimination of box modes was investigated, as well as bond characteristics including bond homogeneity and the presence of a superconducting connection.

  11. Fe-based bulk metallic glasses used for magnetic shielding

    Serban, Va; Codrean, C; UTu, D [Politehnica University of Timisoara, Depart for Materials Science and Welding, 1, M. Viteazu Bvd., 300222, Timisoara (Romania); ErcuTa, A, E-mail: serban@mec.upt.r [West University of Timisoara, Faculty of Physics, 4, Vasile Parvan Bdv., Timisoara 300223 (Romania)

    2009-01-01

    The casting in complex shapes (tubular) and the main magnetic properties of bulk metallic glasses (BMG) alloys from the ferromagnetic Fe-Cr-Ni-Ga-P-Si-C system, with a small addition of Ni (3%) were studied. Samples as rods and sockets having the thickness up to 1 mm were obtained from master alloys by melt injection by low cooling rates into a Cu mold and annealed in order to ensure adequate magnetic requirements. The structure was examined by X-ray diffraction (XRD) and the basic magnetic properties (coercivity, magnetic remanence, initial susceptibility, etc.) were determined by conventional low frequency induction method. The experimental investigations on producing of BMG ferromagnetic alloys with 3% Ni show the possibility to obtain magnetic shields of complex shape with satisfactory magnetic properties. The presence of Ni does not affect the glass forming ability, but reduce the shielding capacity.

  12. Homogeneous versus heterogeneous shielding modeling of spent-fuel casks

    Carbajo, J.J.; Lindner, C.N.

    1992-01-01

    The design of spent-fuel casks for storage and transport requires modeling the cask for criticality, shielding, thermal, and structural analyses. While some parts of the cask are homogeneous, other regions are heterogeneous with different materials intermixed. For simplicity, some of the heterogeneous regions may be modeled as homogeneous. This paper evaluates the effect of homogenizing some regions of a cask on calculating radiation dose rates outside the cask. The dose rate calculations were performed with the one-dimensional discrete ordinates shielding XSDRNPM code coupled with the XSDOSE code and with the three-dimensional QAD-CGGP code. Dose rates were calculated radially at the midplane of the cask at two locations, cask surface and 2.3 m from the radial surface. The last location corresponds to a point 2 m from the lateral sides of a transport railroad car

  13. Shielding design of radioactive contaminated metal waste packaging

    Zou Wenhua; Dong Zhiqiang; Yao Zhenyu; Xu Shuhe; Wang Wen

    2015-01-01

    Focusing on the cylindrical source model to calculate γ dose field of waste packages with the relative formulae then derived. By comparing the calculated data of waste packages of type Ⅷ steel box with the monitoring data, it is found that the cylinder source model could accurately reflect the distributions of γ dose of the waste package. Based on the results of the cylindrical source model, a reasonable shielding technology applicable to waste package containers was designed to meet relevant requirements prescribed in standards about the transport and disposal of radioactive materials. The cylinder source model calculated dose distributions for single package in this paper is simple and easy to implement but slightly larger than the monitoring data providing a certain safety margin for the shielding design. It is suitable for radiological engineering practices. (authors)

  14. SATIF-2 shielding aspects of accelerators, targets and irradiation facilities

    1995-01-01

    Particle accelerators have evolved over the last 50 years from simple devices to powerful machines, and will continue to have an important impact on research, technology and lifestyle. Today they cover a wide range of applications, from television and computer displays in households to the investigation of the origin and structure of matter. It has become common practice to use them for material science and medical applications. In recent years, requirements from new technological and research applications have emerged, such as increased particle beams intensities, higher flexibility, etc., giving rise to new radiation shielding aspects and problems. These proceedings review recent progress in radiation shielding of accelerator facilities, and evaluate advancements with respect to international co-operation in this field

  15. Challenges in commercial manufacture of radiation shielding glasses

    Gupta, R.K.

    2011-01-01

    Radioactive hot-cells employ Radiation Shielding Windows (RSWs), assembled from specialty glasses, developed exclusively for nuclear industry. RSWs serve the twin purpose of direct viewing and shielding protection to the operator and use various types of radiation resistant and optically compatible glasses, such as low-density borosilicate glass; medium-density glass with up to 45% Lead and high-density glass with over 70% lead. Some glasses are Ceria-doped for enhancing their resistance threshold to radiation browning. A clear view of future requirement, capital and environmental costs could be the driving force towards bringing about changes in melting practices, encourage melting development, and enhancing collaboration. With DAE and CGCRI working in tandem, production of the entire range of RSW glasses by an Indian glass industry participant may no longer be a distant dream

  16. Field maintenance of radiation-shielding windows at HFEF

    Tobias, D.A.

    1983-01-01

    The achievement of excellent viewing through hot-cell shielding windows does not occur by chance. Instead, it requires a well planned and executed program of field maintenance. The lack of such a program is a major factor when a hot-cell facility has poor window viewing. At HFEF, all preventive maintenance is performed by one group of trained technical-support personnel under the immediate direction of a Systems Engineer, who has responsibility for the shielding windows. Window maintenance is prescheduled and recorded by being incorporated into the computerized Maintenance Data System (MDS). Measurements of window light transmission are scheduled annually to determine glass browning or oil cloudiness conditions within the window tank. The tank oil is sampled and chemically analyzed annually to determine the moisture content, the acidity, and the probable deterioration rate caused by irradiation

  17. Comparative study of radiation shielding parameters for bismuth borate glasses

    Kaundal, Rajinder Singh

    2016-01-01

    Melt and quench technique was used for the preparation of glassy samples of the composition x Bi 2 O 3- (1-x) B 2 O 3 where x= .05 to .040. XCOM computer program is used for the evaluation of gamma-ray shielding parameters of the prepared glass samples. Further the values of mass attenuation coefficients, effective atomic number and half value layer for the glassy samples have been calculated in the energy range from 1KeV to 100GeV. Rigidity of the glass samples have been analyzed by molar volume of the prepared glass samples. (author)

  18. Comparative study of radiation shielding parameters for bismuth borate glasses

    Kaundal, Rajinder Singh, E-mail: rajinder_apd@yahoo.com [Department of Physics, School of Physical Sciences, Lovely Professional University, Phagwara, Punjab (India)

    2016-07-15

    Melt and quench technique was used for the preparation of glassy samples of the composition x Bi{sub 2}O{sub 3-}(1-x) B{sub 2}O{sub 3} where x= .05 to .040. XCOM computer program is used for the evaluation of gamma-ray shielding parameters of the prepared glass samples. Further the values of mass attenuation coefficients, effective atomic number and half value layer for the glassy samples have been calculated in the energy range from 1KeV to 100GeV. Rigidity of the glass samples have been analyzed by molar volume of the prepared glass samples. (author)

  19. Comparative Study of Radiation Shielding Parameters for Bismuth Borate Glasses

    Kaundal, Rajinder Singh

    2016-01-01

    Melt and quench technique was used for the preparation of glassy samples of the composition x Bi2O3-(1-x) B2O3 where x= .05 to .040. XCOM computer program is used for the evaluation of gamma-ray shielding parameters of the prepared glass samples. Further the values of mass attenuation coefficients, effective atomic number and half value layer for the glassy samples have been calculated in the energy range from 1KeV to 100GeV. Rigidity of the glass samples have been analyzed by molar volume of...

  20. Scale-4 shipping cask shielding applications

    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