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Sample records for 1c shield design

  1. Magnetic shielding design analysis

    Two passive magnetic-shielding-design approaches for static external fields are reviewed. The first approach uses the shielding solutions for spheres and cylinders while the second approach requires solving Maxwell's equations. Experimental data taken at LLNL are compared with the results from these shieldings-design methods, and improvements are recommended for the second method. Design considerations are discussed here along with the importance of material gaps in the shield

  2. 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 ...

  3. Design experience: CRBRP radiation shielding

    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''

  4. Radiation protection/shield design

    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.)

  5. Aladdin upgrade design study: shielding

    The object of this shielding is to examine all aspects of Aladdin operation to ensure that adequate shielding is provided to meet the design objectives. To do this, we will look at shielding necessary for radiation produced during the injection process, during normal loss of the stored beam and during accidental loss of the stored beam. It will therefore be necessary to specify shielding not only at the ring, but also along the injection line and the optical beam lines. We will also give special attention to the occupation of the accelerator Vault during injection as this may be a desirable design option. In effect, two shielding plans will be presented, permitting estimates of cost and space requirements for both

  6. New facility shield design criteria

    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

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

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

  8. MMW [multimegawatt] shielding design and analysis

    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

  9. Japanese contributions to ITER shielding neutronics design

    Shielding design for superconducting magnets and personal exposure were performed in ITER nuclear design on the basis of reports presented to the 1990 winter and summer ITER specialist meetings. Inboard shield benchmark calculation, bulk inboard shielding analysis, inboard heterogeneity effect on shielding property analysis, gap streaming analysis were discussed on shielding properties for superconducting magnets. In addition to these, transport and Monte Carlo analyses in neutral beam injector duct for biological shielding were investigated with relation to the concept of cryostat. Further biological shielding were investigated in reactor room and site boundary during the maintenance when one activated module was extracted and hanged from the ceiling. As the results of these studies, ITER shielding characteristics were evaluated and problem areas and directions for future works were shown. (author)

  10. REPOSITORY RADIATION SHIELDING DESIGN GUIDE

    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

  11. Shielding design of fusion experimental reactor (FER)

    This report first describes the basic design philosophy of radiation shields for the fusion experimental reactor (FER) which has been proposed to be the next step machine to JT-60. Next, geometrical models and calculation parameters for shielding calculations were investigated to establish the standard design calculation methods, and accuracy of the calculation was evaluated. Further, irradiation properties of in-vessel components and bulk shielding properties were summarized in the useful form for the future design works. (author)

  12. Heat-shield design for glovebox applications

    Heat shields can often be used in place of insulation materials as an effective means of insulating glovebox furnace vessels. If used properly, shields can accomplish two important objectives: thermal insulation of the vessel to maintain a desired process temperature and protection of the glovebox, equipment, and user. A heat-shield assembly can be described as an arrangement of thin, properly-spaced, metal sheets that reduce radiation heat transfer. The main problem encountered in the design of a heat shield assembly is choosing the number of shields. In determining the heat transfer characteristics of a heat-shield assembly, a number of factors must be taken into consideration. The glovebox or outside environment, material properties, geometry, and operating temperature all have varying effects on the expected results. A simple method, for planar-horizontal and cylindrical-vertical shields, allowing the approximation of the outermost shield temperature, the practical number of shields, and the net heat-transfer rate will be presented. Methods used in the fabrication of heat-shield assemblies will also be discussed

  13. Thermal design of top shield

    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/m2. 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 m

  14. MFTF-α + T shield design

    MFTF-α+T is a DT upgrade option of the Tandem Mirror Fusion Test Facility (MFTF-B) to study better plasma performance, and test tritium breeding blankets in an actual fusion reactor environment. The central cell insert, designated DT axicell, has a 2-MW/m2 neutron wall loading at the first wall for blanket testing. This upgrade is completely shielded to protect the reactor components, the workers, and the general public from the radiation environment during operation and after shutdown. The shield design for this upgrade is the subject of this paper including the design criteria and the tradeoff studies to reduce the shield cost

  15. ITER cryostat thermal shield detailed design

    The structural design and study on fabrication and assembly of the cryostat thermal shield for International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat thermal shield is attached to cover the cryostat inner wall in order to reduce the radiation heat loads applied to the superconducting coils operation at 4 K. The thermal shield consists of low-emissivity foils which are passively cooled and shield plates which are actively cooled with low temperature helium gas. The foils are multi-layered assemblies and are attached on both surfaces of the shield plates. The material of the foils are silver coated 304 stainless steel, polyimide or polyester. The silver coated stainless steel foils should be adopted to the foils at the locations where radiation dose is over 10 MGy. The route of coolant pipes for the shield plates is designed so as to keep the surface temperature of the shield plates below 100 K. This report describes the detailed design of the cryostat thermal shield, and outlines the fabrication and assembly procedures. (J.P.N.)

  16. ITER cryostat thermal shield detailed design

    Ito, Akira; Nakahira, Masataka; Hamada, Kazuya; Takahashi, Hiroyuki; Tada, Eisuke; Kato, Takashi [Department of Fusion Engineering Research, Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan); Nishikawa, Akira

    1999-03-01

    The structural design and study on fabrication and assembly of the cryostat thermal shield for International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat thermal shield is attached to cover the cryostat inner wall in order to reduce the radiation heat loads applied to the superconducting coils operation at 4 K. The thermal shield consists of low-emissivity foils which are passively cooled and shield plates which are actively cooled with low temperature helium gas. The foils are multi-layered assemblies and are attached on both surfaces of the shield plates. The material of the foils are silver coated 304 stainless steel, polyimide or polyester. The silver coated stainless steel foils should be adopted to the foils at the locations where radiation dose is over 10 MGy. The route of coolant pipes for the shield plates is designed so as to keep the surface temperature of the shield plates below 100 K. This report describes the detailed design of the cryostat thermal shield, and outlines the fabrication and assembly procedures. (J.P.N.)

  17. Shielding design for better plant availability

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

  18. Design of ITER shielding blanket

    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)

  19. Shielding options for the ITER conceptual design

    Several shield options were analyzed for the ITER conceptual design to minimize the nuclear responses in the toroidal field (TF) coils. The total nuclear heating in the physics phase and the insulator dose in the technology phase are the most critical parameters in the design process. The first shield option has type 316 stainless steel and water shielding material. Steel and water also serve as structural material and coolant, respectively. The second option is similar to the first except that borated water is used instead of ordinary water. The other two options include a small layer of lead or boron carbide (B4C) at the back of the shield. The last three shield options were considered to reduce the nuclear heating in the toroidal field coils relative to the steel/water shield. An optimization process was performed taking into consideration the thermal-hydraulics and the engineering- requirements to define the shield configuration. A careful integration was performed to calculate the total nuclear heating in the toroidal field coils which account for the neutron wall loading distribution, the change in the shield thickness in the poloidal direction, and the space between the toroidal field coils in the divertor zone. The results show that the steel/water/Pb and the steel/borated water shield options are very close in terms of the total nuclear heating in the toroidal field coils and the dose in the insulator material. The other two options, steel/water and steel/water/B4C deposit more nuclear heating in the toroidal field coils. 5 refs., 3 figs., 5 tabs

  20. Design and Analysis of HJ-1-C Satellite SAR Antenna

    Zheng Shi-kun

    2014-06-01

    Full Text Available With truss deployable mesh parabolic reflector, the HJ-1-C SAR antenna has complex structure and multiple steps during the deployed processing. The design of the antenna is difficult in terms of deployed reliability and electrical performance. This paper makes intensive research on system, structure and electrical design, and the analysis of mechanical and thermal performance in the actual space conditions is also presented. The successful deploying in orbit and high image quality of the HJ-1-C satellite indicate that the mechanical, electronic, thermal and reliability design of the antenna satisfy the project requirement, and these research provides valuable experience for the design of the centralized mesh parabolic SAR antenna.

  1. Shielding design to obtain compact marine reactor

    Yamaji, Akio; Sako, Kiyoshi (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment)

    1994-06-01

    The marine reactors equipped in previously constructed nuclear ships are in need of the secondary shield which is installed outside the containment vessel. Most of the weight and volume of the reactor plants are occupied by this secondary shield. An advanced marine reactor called MRX (Marine Reactor X) has been designed to obtain a more compact and lightweight marine reactor with enhanced safety. The MRX is a new type of marine reactor which is an integral PWR (The steam generator is installed in the pressure vessel.) with adopting a water-filled containment vessel and a new shielding design method of no installation of the secondary shield. As a result, MRX is considerably lighter in weight and more compact in size as compared with the reactors equipped in previously constructed nuclear ships. For instance, the plant weight and volume of the containment vessel of MRX are about 50% and 70% of those of the Nuclear Ship MUTSU, in spite of the power of MRX is 2.8 times as large as the MUTSU's reactor. The shielding design calculation was made using the ANISN, DOT3.5, QAD-CGGP2 and ORIGEN codes. The computational accuracy was confirmed by experimental analyses. (author).

  2. Shielding design for positron emission tomography facility

    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

  3. Preliminary Thermal Design of Cryogenic Radiation Shielding

    Li, Xiaoyi; Mustafi, Shuvo; Boutte, Alvin

    2015-01-01

    Cryogenic Hydrogen Radiation Shielding (CHRS) is the most mass efficient material radiation shielding strategy for human spaceflight beyond low Earth orbit (LEO). Future human space flight, mission beyond LEO could exceed one year in duration. Previous radiation studies showed that in order to protect the astronauts from space radiation with an annual allowable radiation dose less than 500 mSv, 140 kgm2 of polyethylene is necessary. For a typical crew module that is 4 meter in diameter and 8 meter in length. The mass of polyethylene radiation shielding required would be more than 17,500 kg. The same radiation study found that the required hydrogen shielding for the same allowable radiation dose is 40 kgm2, and the mass of hydrogen required would be 5, 000 kg. Cryogenic hydrogen has higher densities and can be stored in relatively small containment vessels. However, the CHRS system needs a sophisticated thermal system which prevents the cryogenic hydrogen from evaporating during the mission. This study designed a cryogenic thermal system that protects the CHRS from hydrogen evaporation for one to up to three year mission. The design also includes a ground based cooling system that can subcool and freeze liquid hydrogen. The final results show that the CHRS with its required thermal protection system is nearly half of the mass of polyethylene radiation shielding.

  4. Design and analysis of ITER shield blanket

    This report includes electromagnetic analyses for ITER shielding blanket modules, fabrication methods for the blanket modules and the back plate, the design and the fabrication methods for port limiter have been investigated. Studies on the runaway electron impact for Be armor have been also performed. (J.P.N.)

  5. Summary of Prometheus Radiation Shielding Nuclear Design Analysis

    J. Stephens

    2006-01-13

    This report transmits a summary of radiation shielding nuclear design studies performed to support the Prometheus project. Together, the enclosures and references associated with this document describe NRPCT (KAPL & Bettis) shielding nuclear design analyses done for the project.

  6. Summary of Prometheus Radiation Shielding Nuclear Design Analyses , for information

    This report transmits a summary of radiation shielding nuclear design studies performed to support the Prometheus project. Together, the enclosures and references associated with this document describe NRPCT (KAPL and Bettis) shielding nuclear design analyses done for the project

  7. Shielding design for ETOILE hadron therapy centre

    The Ion Beam Applications Company is developing a compact superconducting cyclotron for hadron therapy able to deliver various ion beams with an energy of 400 MeV per nucleon and proton beams with an energy of 260 MeV. This system is being proposed to equip ETOILE hadron therapy centre in Lyon. Shielding design based on PHITS and MCNPX Monte Carlo simulation codes is presented, together with some performance figures for the energy degrader. (authors)

  8. Advanced materials and design for electromagnetic interference shielding

    Tong, Xingcun Colin

    2008-01-01

    Exploring the role of EMI shielding in EMC design, this book introduces the design guidelines, materials selection, characterization methodology, manufacturing technology, and future potential of EMI shielding. It covers an array of issues in advanced shielding materials and design solutions, including enclosures and composites.

  9. Design of ITER vacuum vessel neutron shielding structure

    Neutron shielding structure of the ITER vacuum vessel (VV) will be applied to shielding neutron and gamma-ray and reducing the toroidal field ripple. The features of ITER vacuum vessel and the material selection for shielding structures are briefly discussed. A shielding conceptual design and some correlative support structures have been developed. The layout of ferromagnetic inserts was performed. Filling ratios of shielding materials between VV shells were acquired according to ITER VV physics calculation results. In term of the ITER VV design criteria, the detailed design work for library of the shielding blocks and the emulational structure have been finished based on the 3D modeling software. (authors)

  10. Active Muon Shield - Preliminary Design Report

    Bayliss, Victoria; Rawlings, T

    2015-01-01

    This report summarises the initial design study which was carried out for the SHiP magnetic muon shield – which is proposed to consist of a 40m beamline of seven magnets generating a 1.8T By field over defined cross-section. This is intended to sweep unwanted muons off the beamline to prevent them reaching the detector. The magnetic shield is an alternative to a passive tungsten shield. This work was carried out in three sections. Initially the magnets were considered in isolation to establish whether they were theoretically feasible to build and the impact of the iron yoke shape and material was considered. Next the beamline was considered as a whole; this included issues such as the impact of neighbouring magnets and the hadrons stopper, and also building a model of the complete beamline whose magnetic fields could be exported for use in particle modelling. Finally, some consideration was given to the manufacture and operational issues, including costs.

  11. Layered shielding design for an active neutron interrogation system

    Whetstone, Zachary D.; Kearfott, Kimberlee J.

    2016-08-01

    The use of source and detector shields in active neutron interrogation can improve detector signal. In simulations, a shielded detector with a source rotated π/3 rad relative to the opening decreased neutron flux roughly three orders of magnitude. Several realistic source and detector shield configurations were simulated. A layered design reduced neutron and secondary photon flux in the detector by approximately one order of magnitude for a deuterium-tritium source. The shield arrangement can be adapted for a portable, modular design.

  12. Shield design development of nuclear propulsion merchant ship

    Shielding design both in Japan and abroad for nuclear propulsion merchant ships is explained, with emphasis on the various technological problems having occurred in the shield design for one-body type and separate type LWRs as conceptual design. The following matters are described: the peculiarities of the design as compared with the case of land-based nuclear reactors, problems in the design standards of shielding, the present status and development of the design methods, and the instances of the design; thereby, the trends of shielding design are disclosed. The following matters are pointed out: Importance of the optimum design, of shielding, significance of radiation streaming through large voids, activation of the secondary water in built-in type steam generators, and the need of the guides for shield design. (Mori, K.)

  13. Fusion reactor blanket/shield design study

    A joint study of tokamak reactor first-wall/blanket/shield technology was conducted by Argonne National Laboratory (ANL) and McDonnell Douglas Astronautics Company (MDAC). The objectives of this program were the identification of key technological limitations for various tritium-breeding-blanket design concepts, establishment of a basis for assessment and comparison of the design features of each concept, and development of optimized blanket designs. The approach used involved a review of previously proposed blanket designs, analysis of critical technological problems and design features associated with each of the blanket concepts, and a detailed evaluation of the most tractable design concepts. Tritium-breeding-blanket concepts were evaluated according to the proposed coolant. The ANL effort concentrated on evaluation of lithium- and water-cooled blanket designs while the MDAC effort focused on helium- and molten salt-cooled designs. A joint effort was undertaken to provide a consistent set of materials property data used for analysis of all blanket concepts. Generalized nuclear analysis of the tritium breeding performance, an analysis of tritium breeding requirements, and a first-wall stress analysis were conducted as part of the study. The impact of coolant selection on the mechanical design of a tokamak reactor was evaluated. Reference blanket designs utilizing the four candidate coolants are presented

  14. Design of radiation shields in nuclear reactor core

    This article consists of designing radiation shields in the core of nuclear reactors to control and restrain the harmful nuclear radiations in the nuclear reactor cores. The radiation shields protect the loss of energy. caused by nuclear radiation in a nuclear reactor core and consequently, they cause to increase the efficiency of the reactor and decrease the risk of being under harmful radiations for the staff. In order to design these shields, by making advantages of the Oppenheim Electrical Networkmethod, the structure of the shields are physically simulated and by obtaining a special algorithm, the amount of optimized energy caused by nuclear radiations, is calculated

  15. ARIES-ST nuclear analysis and shield design

    A power plant design based on the spherical torus (ST) concept has been developed by the ARIES team. This paper documents the results of the nuclear and radiation protection analyses carried out for the ARIES-ST design. The nuclear analysis addresses the key neutronics issues, such as the neutron wall loading profile, radiation damage to structural components and their lifetimes, tritium breeding ratio (TBR), and nuclear heat loads to in-vessel components. The main theme of the shielding analysis is to develop guidance and recommendations on radiation protection for the TF magnet, in particular, the center-post. The need for an inboard shield, the selection of an optimal shield, the rationale for the shielding material choices, and the consequences of the shielding material choices on the overall design are reported herein. During the course of the ARIES-ST study, the design has been analyzed rigorously with a set of 3-D nuclear analyses to guide the developing design. The results of the assessment had a major impact on the design choices. For instance, the insufficient breeding along with other design constraints have ruled out the use of several attractive solid breeder blankets, the excessive neutron damage to the center-post provided strong incentives to shield the center-post, and the high radiation damage to the plasma facing components has limited the service lifetime of the ferritic steel (FS) structure to a few full power years. Furthermore, the high heat load to the inboard shield (400 MW, 10% of thermal power) forced the design to recover the inboard heating as high-grade heat to enhance the power balance. Also, the sensitivity of the outboard-only LiPb breeding blanket to the inboard shielding materials has limited the shielding options and excluded several high performance inboard-shielding materials. The performed analyses and results are reported and discussed in relation to the integrated design and the established top-level requirements for the

  16. Radiation shielding analysis for conceptual design of HIC transport package

    KHNP(Korea Hydro and Nuclear Power Ltd., Co.) is developing a HIC transport package which is satisfying domestic and IAEA regulations and NETEC(Nuclear Environment Technology Institute) is conducting a conceptual design. In this study, the shielding thickness was calculated using the data from radionuclide assay program which is currently using in nuclear sites and Micro Shield code. Considering the structural safety, carbon steel was chosen as shielding material and the shielding thickness was calculated for 500 R/hr and 100 R/hr at HIC surface, respectively. Through the shielding analysis, it was evaluated that the regulation limit is satisfied when the shielding thickness is 22 cm for 500 R/hr and 17 cm for 100/hr

  17. Optimized Design of the Shielded-Loop Resonator

    Stensgaard, Anders

    1996-01-01

    The shielded-loop resonator is known to have low capacitive sample loss due to perfect balancing. We present a new analysis of the unbalanced driven shielded-loop resonator that calculates the resonance frequencies and also determines some design considerations. The analysis enables us to optimize...

  18. Planar quadrature coil design using shielded-loop resonators

    Stensgaard, A

    1997-01-01

    The shielded-loop resonator is known to have a low capacitive sample loss due to a perfect balancing. In this paper, it is demonstrated that shielded-loop technology also can be used to improve design of planar quadrature coils. Both a dual-loop circuit and especially a dual-mode circuit may...

  19. Inhibited Shaped Charge Launcher Testing of Spacecraft Shield Designs

    Grosch, Donald J.

    1996-01-01

    This report describes a test program in which several orbital debris shield designs were impact tested using the inhibited shaped charge launcher facility at Southwest Research Institute. This facility enables researchers to study the impact of one-gram aluminum projectiles on various shielding designs at velocities above 11 km/s. A total of twenty tests were conducted on targets provided by NASA-MSFC. This report discusses in detail the shield design, the projectile parameters and the test configuration used for each test. A brief discussion of the target damage is provided, as the detailed analysis of the target response will be done by NASA-MSFC.

  20. Design of ITER vacuum vessel in-wall shielding

    Wang, X., E-mail: xiaoyu.wang@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ioki, K. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Morimoto, M. [Mitsubishi Heavy Industries, 1-1, Wadasaki-cho 1-chome, Hyogo-ku, Kobe (Japan); Choi, C.H.; Utin, Y.; Sborchia, C. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); TaiLhardat, O. [Assystem EOS, ZAC SAINT MARTIN, 23 rue Benjamin Franklin, 84120 Pertuis (France); Mille, B.; Terasawa, A.; Gribov, Y.; Barabash, V.; Polunovskiy, E.; Dani, S. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Pathak, H.; Raval, J. [ITER-India, Institute for Plasma Research, Gandhinagar 382025 (India); Liu, S.; Lu, M.; Du, S. [Institute of Plasma Physics, China Academy of Sciences, Shushanhu Road 350, Hefei (China)

    2014-10-15

    The 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 (IWS) and cooling water. The main purpose of the in-wall shielding is to provide neutron shielding together with the blanket and VV shells and water during ITER plasma operation and to reduce the ripple of the Toroidal magnetic field. Based on ITER vacuum vessel structure and related requirements, in-wall shielding are designed as about 8900 individual blocks with different sizes and several different materials distributed over nine vessel sectors and nine field joints of vessel sectors. This paper presents the design of the IWS, considering loads, structural stresses and assembly method, and also shows neutron shielding effect and TF ripple reduced by the IWS.

  1. Design of ITER vacuum vessel in-wall shielding

    The 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 (IWS) and cooling water. The main purpose of the in-wall shielding is to provide neutron shielding together with the blanket and VV shells and water during ITER plasma operation and to reduce the ripple of the Toroidal magnetic field. Based on ITER vacuum vessel structure and related requirements, in-wall shielding are designed as about 8900 individual blocks with different sizes and several different materials distributed over nine vessel sectors and nine field joints of vessel sectors. This paper presents the design of the IWS, considering loads, structural stresses and assembly method, and also shows neutron shielding effect and TF ripple reduced by the IWS

  2. Fusion reactor design towards radwaste minimum with advanced shield material

    A new concept of fusion reactor design is proposed to minimize the radioactive waste of the reactor. The main point of the concept is to clear massive structural components located outside the neutron shield from regulatory control. The concept requires some reinforcement of shielding with an advanced shield material such as a metal hydride, detriation, and tailoring of a detrimental element from the superconductor. Our assessment confirmed a large impact of the concept on radwaste reduction, in that it reduces the radwaste fraction of a fusion reactor A-SSTR2 from 92 wt.% to 17 wt.%. (author)

  3. Methodology for shielding design and evaluation in radiotherapy facilities

    The Government of the Republic of Cuba has decided to carry out a wide programme concerning the purchase of more than a dozen dual linear accelerators and, also; more than a dozen cobalt-60 units. Due to the lack of a national methodology for the design and calculation of shielding enclosures for radiotherapy units, the medical physicists from different hospitals began to use different methodologies, e.g. those in: a) Medical Physics Publishing. Shielding Techniques for Radiation Oncology Facilities. Patton H. McGinley. 1998.; b) National Council on Radiation Protection and Measurements, Structural shielding design and evaluation for medical use of X-rays and gamma-rays of energies up to 10 MeV, Report No. 49, NCRP, Washington, DC (1976).; c) National Council on Radiation Protection and Measurements, Radiation Protection Guidelines for 0.1 - 100 MeV Particle Accelerator Facilities, Report No. 51, NCRP, Washington, DC (1977). In some cases this caused the overestimation of the shielding thickness, when applying the values of dose constraints required by the Cuban regulations. The objective of the present work is to provide the medical physicists, the Radiation Safety Officers and other related professionals with a consistent methodology for the design and remodelation of bunkers hosting radiotherapy units but not using shielding doors. This work shows the validity of the above mentioned methodology, and the feasibility of designing door less bunkers for radiotherapy purposes. This methodology is considered to be self consistent and therefore no other complementary materials for its application are required. The experience so far confirms that; entry of realistic input data, and adequate application of sound engineering concepts when using this methodology leads to the achievement of enclosure shielding designs for radiotherapy units that comply with the dose constraints established by the Cuban regulations. Radiation shielding is attained having no over expenses on

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

    The Waste Safety Testing Facility (WASTEF) was planned in 1978 to test the safety performance of HLW vitrified forms under the simulated conditions of long term storage and disposal, and completed in August 1981. The designed feature of the facility is to treat the vitrified forms contain actual high-level wastes of 5 x 104 Ci in maximum with 5 units of concrete shilded hot cells (3 units : Bate-Gamma cells, 2 units : Alpha-Gamma cells) and one units of Alpha-Gamma lead shielded cell, and to store radioactivity of 106 Ci in maximum. The safety performance of this facility is fundamentally maintained with confinement of radioactivity and shielding of the radiation. This report describes the method of gamma-ray shielding design, evaluation of the shielding test performed by using sealded gamma-ray sources(Co-60). (author)

  5. Methods for the design of shielding concrete mix ratio

    Guided by general concrete mix principles, we made a comprehensive study on methods for the design of shielding concrete mix ratio as well as its related factors by means of orthogonal design experiments and regression analysis method. Then we put forward the calculating formulae and steps for the design of shielding concrete mix ratio which combined the weight-holding method with the volume method. A series of tests and practical application show that this method of mix design is accurate, efficient and reliable. (authors)

  6. Shielding design calculation of a 50 MW research reactor

    The computer code ANISN/PC has been applied to calculate the group flux distribution across different shield layers of a 50 MW light water research reactor. The code has been run in P3 approximation and S8 discrete ordinates. The calculated group fluxes multiplied by appropriate flux-to-dose rate conversion factors have been used to give the dose distribution across the shield layers. The thickness of the concrete shield has been determined to give the dose rate at the outer surface of the shield as 0.5 nSv/sec. The same calculation have been also performed in axial direction to determine the thickness of water needed above the core to reduce the dose level to 25 nSv/sec. The result of calculation shows that the contribution of capture gamma rays to the total dose at the outer surface of the shield is more than 50 percent. This simplifies the calculations to determine the shield layer thickness, especially in preliminary stages of the shield design. (author)

  7. Concrete mix design for X-and gamma shielding

    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)

  8. Fusion Engineering Device (FED) first wall/shield design

    The torus of the Fusion Engineering Device (FED) is comprised of the bulk shield and its associated spool lstructure and support system, the first wall water-cooled panel and armor systems, and the pumped limiter. The bulk shielding is provided by ten shield sectors that are installed in the spool structure in such a way as to permit extraction of the sectors through the openings between adjacent toroidal field coils with a direct radial movement. The first wall armor is installed on the inboard and top interior walls of these sectors, and the water-cooled panels are installed on the outboard interior walls and the pumped limiter in the bottom of the sectors. The overall design of the first wall and shield system is described in this paper

  9. Designing of central shield in Ca Cx cases

    This paper deals with the designing aspects of central shielding block for external radiotherapy used in treatment of Ca Cx patients who are already treated or are going to be treated with intracavitary radiation (ICR) treatments. The designing aspects are discussed in detail particularly for the region between point A and point B. (author)

  10. Shielding design for research and education reactor

    For the purpose of education and research at the University, 20-KW powered SLOWPOKE-2 research reactor has been chosen as a prototype reactor. In order to study the safety characteristics of the reactor, exposure rate has been estimated at the pool boundary. Reactor core as a radiation source is assumed to be cylindrical volume source. Thus point kernel integration method can be applied to determine the exposure rate. For the sake of simplicity, calculation was done only for the prompt fission gamma rays and fission product gamma rays. As a result, the maximum exposure rate at the pool boundary was estimated to be 18R/min at the same height of the center of the core. In order to examine the accuracy for the point kernel integration method, two shielding experiments were carried out: one for the water tank only and the other for with concrete blocks outside the water tank. Water tank was made of wood pieces which is 13.4cm wide, 1.5cm thick and 2.15m long. Thus the water tank has the total dimension of 1 m radius and 2.1 m height. The experiment was carried out for the radiation source of 0.968 mCi Co-60 at the center of the water tank and the penetrated gamma rays were measured at 5 different detector positions. For the measurement and analysis of the responses, NaI(T1) 3''x3'' detector and 256 channel multichannel analyzer was utilized. To convert pulse height distribution to the exposure rate, Moriuchi conversion factor was adopted. Data from the calculations by point kernel method were well agreed within 10% band with the data from the the experiments. (Author)

  11. D0 Silicon Upgrade: Muon Shield Conceptual Design Report

    The nominal overall dimensions are 71-inch high x 71-inch wide x 144-inch long and has a 25-inch square hole throughout. The shield consists of three different materials, steel (inner most section), polycarbonate (central section) and lead (outer most section). The material thicknesses are, steel=15-inch, poly=6-inch and lead=2-inch. The estimated weight is ∼69 tons. The shield is centered about the Tev beam line and the 25-inch square hole provides clearance to the low Beta quad, which is nominally 20-inch square. During beamline operation, the shield is in contact with Samus magnet core at the detector end and with the Main Ring shield wall on the MR side (with some small clearance ∼2-inch-3-inch). The need for the clearance will be discussed later. The shield support structure consists steel structural members appropriately sized for loads encountered in the design. The structure must not only support the shield but, must be designed for rolling the entire assembly into position in the collision hall. It must provide for cylinders to lift the assembly, Hilman rollers and also connections for moving the entire assembly. The movement is considered to be similar to that with which the calorimeters were moved from the clean room to the sidewalk staging area, i.e. hydraulic cylinder and chain (see dwg. 3740.000-ME294017,3 sheets). This method will be used for the East to West motion and a hydraulic scheme will be used for any North-South motion. Since the shield is 144-inch long and the sidewalk structural support is ∼96-inch, there is a section of the shield that is cantilevered (48-inch). Further, the EF toroid must open ∼40+ inch for access to the detector during operations and this requires that the shield or some part of it must also move. This conceptual design suggests that the shield be designed in two pieces axially. These two pieces are identical in cross section but, the lengths are divided into 48-inch nearest EF and 96-inch nearest the MR tunnel

  12. Shielding design of the beam tube in the KMRR

    The Korea Multipurpose Research Reactor (KMRR) has a core of honeycomb form surrounded by a cylindrical reflector of D2O and will be operated at maximum thermal power of 30 MW. Seven beam tubes tangentially placed to the reactor core penetrate the reflector and extend to the biological shield end. In neutronics and shielding analysis, this complex geometry usually requires a three-dimensional treatment to obtain data with reasonable accuracy. However, computer implementation of transport or Monte Carlo calculations with a realistic three-dimensional beam tube model is not easy in regard to its problem size. This paper describes a KMRR beam tube shielding design method using coupling of Monte Carlo code MORSE-CG and two-dimensional discrete ordinates code DOT4.2 with VITAMIN-C nuclear data base. The evaluated dose rate at the outside of shield turns out to be 0.135 mrem/hr, which is well within the shielding design criteria 1.25 mrem/hr

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

    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

  14. Design and Analysis of the Thermal Shield of EAST Tokamak

    Xie, Han; Liao, Ziying

    2008-04-01

    EAST (Experimental Advanced Superconducting Tokamak) is a tokamak with superconducting toroidal and poloidal magnets operated at 4.5 K. In order to reduce the thermal load applied on the surfaces of all cryogenically cooled components and keep the heat load of the cryogenic system at a minimum, a continuous radiation shield system located between the magnet system and warm components is adopted. The main loads to which the thermal shield system is subjected are gravity, seismic, electromagnetic and thermal gradients. This study employed NASTRAN and ANSYS finite element codes to analyze the stress under a spectrum of loading conditions and combinations, providing a theoretical basis for an optimization design of the structure.

  15. Design for shielding kit for local irradiation in mice and test of its shielding effect

    In order to fulfill the immediate requirement for experimental studies on biological effect of low dose radiation. A shielding kit for local irradiation in mice is designed. Its advantages are: (1) several mice can be irradiated at the same time; (2) the radiation condition is identical; (3) it is easy to control and perform; (4) during irradiation, the animals don't need any special treatments such as anaesthesia. It was proved by TLD test, that absorbed dose in areas of spleen and head in mice after shielding has decreased to 0.85% and 0.5% of the original dose in the center of radiation field respectively. The results suggest that the kit was able to satisfy the needs of the experimental studies on radiation biology

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

    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)

  17. Approximate design calculation methods for radiation streaming in shield irregularities

    Investigation and assessment are made for approximate design calculation methods of radiation streaming in shield irregularities. Investigation is made for (1) source, (2) definition of streaming radiation components, (3) calculation methods of streaming radiation, (4) streaming formulas for each irregularity, (5) difficulties in application of streaming formulas, etc. Furthermore, investigation is made for simple calculation codes and albedo data. As a result, it is clarified that streaming calculation formulas are not enough to cover various irregularities and their accuracy or application limit is not sufficiently clear. Accurate treatment is not made in the formulas with respect to the radiation behavior for slant incidence, bend part, offset etc., that results in too much safety factors in the design calculation and distrust of the streaming calculation. To overcome the state and improve the accuracy of the design calculation for shield irregularities, it is emphasized to assess existing formulas and develop better formulas based on systematic experimental studies. (author)

  18. Design Of Material Access Shielding Door Of ISFSF Building

    Base on the planning to maintain of the air pressure in the reactor building, the design of material access shielding door in the ISFSF building has been done. By the installation designed, the air pressure condition in the reactor building well meet the design criteria. The system requires 12 pieces of steel beam L 4 x 3 x 1/2 inches ASTM A36 and 6 pieces steel plate by 2400 x 1200 x 3 mm dimension ASTM A514. This paper concluded that this design is feasible to be realized

  19. Preliminary studies for the PDX tokamak radiation shield design

    Shielding calculations have been done to arrive at a preliminary radiation shield design for the Poloidal Diverter Experiment (PDX) tokamak at PPPL. The PDX, which will be used to study plasma impurities in tokamaks, may generate up to 1018 2.5-MeV neutrons per year from D--D reactions and 5 x 1015 14-MeV neutrons per year from D--T reactions. It was determined that shielding doors and a shielding roof must be added to the existing .813-m (32-in.) heavy (i.e., high density) concrete walls to reduce radiation doses to the control room adjacent to PDX and to the site boundary 150 m away to acceptable levels. A .381-m (15-in.) thick roof composed of borated water (5000 ppM) contained in steel cans was recommended. Two doors leading to the control room should be .305 m (1 ft) of polyethylene or water-filled cans backed by 5.08 cm (2 in.) of steel or lead

  20. Radiation shielding design calculation of gamma knife for therapy

    The author reports the method and results of radiation shielding calculation of the gamma knife for therapy which is composed of thirty 60Co sources each with 7.4 EBq, semi-spherical shield, lateral shielding cupboard and the shielding door. The shielding thicknesses of the back shield, the lateral shielding cupboard and the shielding door were calculated. The leakage radiation by test indicates that the shielding is sufficient safety for this Gamma knife and the Kerma rate of control calculated agrees with that by test

  1. First wall/blanket/shield design and optimization system

    First wall/blanket/shield design and optimization system (BSDOS) has been developed to provide a state-of-the-art design tool for fast accurate analysis. In addition, it has been designed to perform several other functions: (1) allowing comparison and evaluation studies for different concepts using the same data bases and ground rules, (2) permitting the use of any figure of merit in the evaluation studies, (3) optimizing the first wall/blanket/shield design parameters for any figure of merit under several design constraints, (4) permitting the use of different reactor parameters in the evaluation and optimization analyses, (5) allowing the use of improved eingineering data bases to study the impact on the design performance for planning future research and development, and (6) evaluating the effect of the data base uncertainties on the design performance. BSDOS is the first design and optimization system to couple the highly interacting neutronics, heat transfer, thermal hydraulics, stress analysis, radioactivity and decay-heat analyses, tritium balance, and capital cost. A brief description of the main features of BSDOS is given in this paper. Also, results from using BSDOS to perform design analysis for several reactor components are presented. 17 refs., 1 fig., 2 tabs

  2. Self-shielded electron linear accelerators designed for radiation technologies

    Belugin, V. M.; Rozanov, N. E.; Pirozhenko, V. M.

    2009-09-01

    This paper describes self-shielded high-intensity electron linear accelerators designed for radiation technologies. The specific property of the accelerators is that they do not apply an external magnetic field; acceleration and focusing of electron beams are performed by radio-frequency fields in the accelerating structures. The main characteristics of the accelerators are high current and beam power, but also reliable operation and a long service life. To obtain these characteristics, a number of problems have been solved, including a particular optimization of the accelerator components and the application of a variety of specific means. The paper describes features of the electron beam dynamics, accelerating structure, and radio-frequency power supply. Several compact self-shielded accelerators for radiation sterilization and x-ray cargo inspection have been created. The introduced methods made it possible to obtain a high intensity of the electron beam and good performance of the accelerators.

  3. ITER blanket module 17 shield block design and analysis

    The shield block reference design of the typical ITER blanket module has a number of grave disadvantages, precarious with relation to nuclear safety of the reactor. The main problems may arise when innage of the parallel cooling passages both in the first wall and in the shield block. Vapor locking in a radial channel with flow insert driver is very probable. Another problem, as a result of the same reason, is draining and dehydration of the coolant system. Then the highly dense packing of the radial channels in the collector array brings an essential flow irregularity. Customary as a rule, the lack of coolant is observed in the last channels, nearest to the outside, most heated surface of the shield block. A local boiling is possible in these dead spaces of coolant system. In consequence of the radial flow irregularity the cooling in the upper box header, directly under the first wall, may be extremely poor. Among the other imperfections one should note the large frontal figured lids, which overburden at welding and give to rise of stresses and shrinkages, and as a result, the large share of irreparable spoilage. The paper represents an alternative design of the shield block coolant system with predominantly sequential flow circuit. The cooling channels are drilled from the frontal side as inclined transverse holes. The open drilling ends are combined in pairs with milled grooves and welded with small lids. This gain the following advantages: the lids may have smaller thickness (7 mm instead 20 mm), the cooling passengers are placed closer to the lateral and upper sides and make cooling better, the welding stress and shrinkages are reduced, there are no any dead spaces of coolant, and the water fillup and draining are substantially improved. The listed hydraulic and thermo mechanical problems have been analysed with help of 3D models in ANSYS CFX program. The models include both the cooling space filled by water and the solid part of shield block. Thus the

  4. Nuclear data relevant to shield design of FMIT facility

    Nuclear data requirements are reviewed for the design of the Fusion Materials Irradiation Test (FMIT) facility. This accelerator-based facility, now in the early stages of construction at Hanford, will provide high fluences in a fusion-like radiation environment for the testing of materials. The nuclear data base required encompasses the entire range of neutron energies from thermal to 50 MeV. In this review, we consider neutron source terms, cross sections for thermal and bulk shield design, and neutron activation for the facility

  5. The Load Design and Implementation of HJ-1-C Space-borne SAR

    Yu Wei-dong

    2014-06-01

    Full Text Available HJ-1-C is a Synthetic Aperture Radar (SAR satellite in the Constellation of “2+1” for China environment and disaster monitoring. It works at S-band with a resolution of 5 m. SAR payload uses a reflector antenna and a high-power concentrated transmitter. Its light weight and high efficiency is very suitable for a small satellite platform. Now HJ-1-C satellite has been launched into orbit and has acquired Chinese first S-band SAR images from space, which demonstrate excellent quality and rich information about scenes imaged. This success verifies our design, testing and experiment work on the payload. With its following operation, HJ-1-C satellite is expected to make a great contribution to the applications of environment protection and disaster monitoring in China. This paper introduces the design and development of HJ-1-C SAR payload, present its main parameters and performance, describes its device details and its manufacture, testing and experiment process. Some images acquired in the orbit are showed.

  6. Design and analysis of electromagnetic interference filters and shields

    McDowell, Andrew Joel

    Electromagnetic interference (EMI) is a problem of rising prevalence as electronic devices become increasingly ubiquitous. EMI filters are low pass filters intended to prevent the conducted electric currents and radiated electromagnetic fields of a device from interfering with the proper operation of other devices. Shielding is a method, often complementary to filtering, that typically involves enclosing a device in a conducting box in order to prevent radiated EMI. This dissertation includes three chapters related to the use of filtering and shielding for preventing electromagnetic interference. The first chapter deals with improving the high frequency EMI filtering performance of surface mount capacitors on printed circuit boards (PCBs). At high frequencies, the impedance of a capacitor is dominated by a parasitic inductance, thus leading to poor high frequency filtering performance. Other researchers have introduced the concept of parasitic inductance cancellation and have applied this concept to improving the filtering performance of volumetrically large capacitors at frequencies up to 100 MHz. The work in this chapter applies the concept of parasitic inductance cancellation to much smaller surface mount capacitors at frequencies up to several gigahertz. The second chapter introduces a much more compact design for applying parasitic inductance cancellation to surface mount capacitors that uses inductive coupling between via pairs as well as coplanar traces. This new design is suited for PCBs having three or more layers including solid ground and/or power plane(s). This design is demonstrated to be considerably more effective in filtering high frequency noise due to crosstalk than a comparable conventional shunt capacitor filter configuration. Finally, chapter 3 presents a detailed analysis of the methods that are used to decompose the measure of plane wave shielding effectiveness into measures of absorption and reflection. Textbooks on electromagnetic

  7. Shielding design for a proton medical accelerator facility

    Source terms and attenuation lengths for neutrons produced by 250 MeV protons on iron, copper and soft tissue, calculated with the FLUKA Monte Carlo code, were used for the shielding calculations (walls, ceilings, and floors) for the National Centre for Oncological Hadrontherapy to be built in Italy. Appropriate hypotheses on the proton current, beam loss factors, duty factors, occupancy factors and use factors of the shields were adopted. A dose equivalent limit of 1 mSv per year in the areas where the public has access and of 2 mSv per year for facility personnel were assumed. Shielding requirements vary from 1.5 m to about 4 m of ordinary concrete. The results agree with Monte Carlo simulations of the complete geometry of the facility obtained in a previous work. The access mazes to the treatment rooms were designed by the LCS Monte Carlo code by optimizing the length and section of their legs and their wall thicknesses with the dose equivalent limit of 2 mSv per year, fixed in the areas accessed by personnel. The resulting annual neutron dose equivalent at the maze mouth is 0.6 mSv

  8. License Application Design Selection Feature Report: Waste Package Self Shielding Design Feature 13

    In the Viability Assessment (VA) reference design, handling of waste packages (WPs) in the emplacement drifts is performed remotely, and human access to the drifts is precluded when WPs are present. This report will investigate the feasibility of using a self-shielded WP design to reduce the radiation levels in the emplacement drifts to a point that, when coupled with ventilation, will create an acceptable environment for human access. This provides the benefit of allowing human entry to emplacement drifts to perform maintenance on ground support and instrumentation, and carry out performance confirmation activities. More direct human control of WP handling and emplacement operations would also be possible. However, these potential benefits must be weighed against the cost of implementation, and potential impacts on pre- and post-closure performance of the repository and WPs. The first section of this report will provide background information on previous investigations of the self-shielded WP design feature, summarize the objective and scope of this document, and provide quality assurance and software information. A shielding performance and cost study that includes several candidate shield materials will then be performed in the subsequent section to allow selection of two self-shielded WP design options for further evaluation. Finally, the remaining sections will evaluate the impacts of the two WP self-shielding options on the repository design, operations, safety, cost, and long-term performance of the WPs with respect to the VA reference design

  9. Design and Testing of Improved Spacesuit Shielding Components

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

  10. Design and Testing of Improved Spacesuit Shielding Components

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

    2002-05-08

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

  11. Design to neutron shielding for medical LINAC therapy room

    Objective: To study the source of neutron in the therapy room and the essence of neutron scattering, to analyze the change patten of neutron dose in the therapy room as well as in the maze so as to design the shielding. Methods: Based on the measurement of the neutron flux for concerning points at the patient plane caused by a running 15 MeV accelerator and referred to NCRP report 79, this paper carried out the radiation protection design and calculation. Results: When X-ray produced by medical accelerator reached certain energy, photonuclear reaction is the main source of neutron contamination in medical accelerator room. The main target of protection is the scattered neutron that comes to inside entrance of the maze and the γ-ray caused by them. As a result neutron contamination has nothing to do with the therapeutic effect, but only to increase the dose commitment for relevant person. Conclusions: Neutron contamination has nothing to do with therapeutic effect, but only to increase the dose commitment for relevant person. Under certain conditions, it may also cause radiation insult to them. Therefore, certain attention should be paid the hazard caused by neutron external exposure, at the same time shielding design and evaluation should be implemented against the neutron contamination in medical accelerator room. (authors)

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

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

  13. Shielding Design Analyses for Smart Core with 49-CEDM

    In Korea, an advanced reactor system of 330MWt power called SMART (System integrated Modular Advanced ReacTor) is being developed by KAERI to supply energy for seawater desalination as well as electricity generation. A shielding design of the SMART core with 49 CEDM is established by a two-dimensional discrete ordinates radiation transport analyses. The DORT two-dimensional discrete ordinates transport code is used to evaluate the SMART shielding designs. Three axial regions represent the SMART reactor assembly, each of which is modeled in the R-Z geometry. The BUGLE-96 library is used in the analyses, which consists of 47 neutron and 20 gamma energy groups. The results indicate that the maximum neutron fluence at the bottom of reactor vessel is 5.89 x 1017 n/cm2 and that on the radial surface of reactor vessel is 4.49 x 1016 n/cm2. These results meet the requirement, 1.0 x 1020 n/cm2, in 10 CFR 50.61 and the integrity of SMART reactor vessel during the lifetime of the reactor is confirmed

  14. Structural Integrity Evaluation of Cold Neutron Laboratory Building by Design Change of Guide Shielding Room

    Wu, Sangik; Kim, Youngki; Kim, Harkrho

    2007-06-15

    This report summarizes the results of the structural integrity evaluation for the cold neutron laboratory building by design change of guide shielding room. The design of the guide shielding room was changed by making its structure members in normal concrete (2.3 g/cc) instead of heavy concrete (3.5 g/cc) because the heavy concrete could be not supplied to meet its design specification. Therefore, it was decided that the guide shielding room is made of the normal concrete. And, the shielding performance of the normal concrete was recalculated to confirm satisfying its design specification, which is of a 9000 zone according to HANARO radiation region classification. The change makes the shielding wall thicker than existing design, and then it is caused to qualify the structural integrity evaluation of the CNLB. Finally, the structural integrity of the CNLB was re-evaluated by considering the design change of the guide shielding room.

  15. A User's Manual for the NRN Shield Design Method

    This report describes a code system for bulk shield design written for a Ferranti Mercury computer and is intended as a manual for those using the programme. The idea of an 'almost direct' flux, as in the removal theory serves as a basis for further development of the theory. An important aspiration has been to minimize the manual work of administering the codes. The codes concerned are: NECO, computing necessary group constants from primary data, REFUSE and REBOX (infinite plane or cylindrical, and box geometry, respectively), computing removal flux, NEDI a one-dimensional (plane, spherical, cylindrical) diffusion multigroup code, and SALOME a Monte Carlo code computing the gamma flux. Output tapes are constructed for direct use as input tapes, when required, for a following code

  16. Neutron beam-line shield design for the protein crystallography instrument at the Lujan Center

    We have developed a very useful methodology for calculating absolute total (neutron plus gamma-ray) dose equivalent rates for use in the design of neutron beam line shields at a spallation neutron source. We have applied this technique to the design of beam line shields for several new materials science instruments being built at the Manuel Lujan Jr. Neutron Scattering Center. These instruments have a variety of collimation systems and different beam line shielding issues. We show here some specific beam line shield designs for the Protein Crystallography Instrument. (author)

  17. Shielding design calculations for ESS activated target system

    Full text of publication follows: The European Spallation Source (ESS) is a European common effort in designing and building a next generation large-scale user facility for studies of the structure and dynamics of materials. The ESS target, moderators and reflectors system through interactions with 5 MW proton beam (2.5 GeV, 20 Hz) will produce long pulse (2.8 ms width) neutrons in sub-thermal and thermal energy range. These neutrons are further transported to a variety of neutron scattering instruments. The aim of this work is to assess the strategy to be used for the safe handling and shipping of the ESS target and associated shaft. For safe maintenance, during operation as well as handling, transport and storage of the components of the ESS target station after their lifetime, detailed knowledge is required about the activation induced by the impinging protons and secondary radiation fields. The Monte Carlo transport code MCNPX2.6.0 was coupled with CINDER90 version 07.4 to calculate the residual nuclide production in the target wheel and associated shaft. Dose equivalent rates due to the residual radiation were further calculated with the MICROSHIELD and MCNPX codes using photon sources resulting from CINDER. Various decay times after ceasing operation of the target components were considered. The activation and decay heat density distributions of the target system together with the derived dose rates were analysed to assess the best strategy to be used for their safe removal and transport to a hot cell, eventual dismantling, storage on-site and shipping off-site as intermediate level waste packages. The derived photon sources were used afterwards to design the shielded exchange flasks that are needed to remove and transport the target after its lifespan to a hot cell. Design of a multipurpose cask able to accommodate the different highly activated components of the ESS target station and ship them to external conditioning facility is intended to be developed

  18. Shielding analyses for design of the upgraded JRR-3 research reactor, 1

    Shielding analyses for design of the upgraded JRR-3 research reactor have been performed. In the report described are the design principles and the overall analytical procedures. In addition, described are the results of shielding analyses of reactor, canal, spent fuel storage pond and so on. (author)

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

    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.

  20. Engineering design and development for prototype fast breeder reactor (PFBR) shielding experiments at Apsara

    Prototype fast breeder reactor (PFBR) houses radial shields inside the reactor vessel which consists of many layers of steel and borated graphite within sodium coolant so as to reduce the neutron flux impingement on Intermediate Heat Exchanger (IHX) (also located inside the reactor vessel) to an acceptable limit. In order to cross check the uncertainties involved in theoretical shielding calculations and neutron cross-section data used, IGCAR proposed to carry out various shielding experiments at Apsara reactor to simulate the theoretical shielding configuration. The experiments would also provide bias factors for detailed shielding design calculations. The shielding experiments were planned to be carried out at Apsara shielding corner with reactor core brought to C-dash (C) position. The neutron flux intensity in the shielding corner was inadequate for the purpose of carrying out experiments. Hence the neutron flux level was enhanced to the order of 1010 n/cm2/s by replacing the water column between the core edge and SS liner of Apsara pool on the shielding corner side with an air filled leak tight aluminium box. The fuel loading in the reactor core was also modified to increase neutron flux intensity towards aluminium box. The neutron flux emerging out of the pool into the shielding corner is essentially a thermal neutron spectrum, which was converted into a typical fast reactor leakage neutron spectrum with the help of converter assemblies (CAs ). The converter assemblies were made of depleted uranium and the assemblies were installed on a CA trolley. The CA trolley was positioned outside Apsara pool in the shielding corner. The models of proposed shields manufactured from various shielding materials viz. sodium, steel, borated graphite and boron carbide were installed on a shield model (SM) trolley. The SM trolley was positioned behind CA trolley. Shield models had provisions for irradiating in any foils which were used for measuring the neutron attenuation

  1. The use of linked shielding codes to substantiate the design of the top corner shielding of a CAGR

    This paper presents a summary of the detailed design substantiation performed for the current commercial AGR internal shielding around the top corner region of the reactor. The design is required to reduce shutdown activation dose rates at accessible positions inside the reactor pressure vessel to the order of 1 mSv/h, whilst at the same time providing adequate space outside the shielding for the remote operation of in-service inspection equipment. This design substantiation work serves as an example of the use of the latest UK shielding codes and demonstrates their flexibility, user-oriented linking capabilities and their practicality as design tools. The codes used involve a variety of the standard techniques for the solution of radiation transport problems: neutron diffusion (SCORMA), kernel/albedo neutron streaming (MULTISORD), Monte Carlo neutron transport (McBEND) and point kernel neutron and gamma ray line-of-sight integration (RANKERN). The linking facilities utilised in this particular application include the automatic transfer of Monte Carlo neutron collision data as secondary gamma-ray source terms into a gamma-ray point kernel integration calculation. This technique means that the inherently accurate, but normally uneconomic, Monte Carlo method can be employed as a design tool in complex limited attentuation situations, as part of an integrated calculational route for large attenuation design problems. (author)

  2. Application of the Moyer Model to shielding design of high-energy heavy-ion accelerators

    Application of Moyer Model for evaluation of shielding design of high-energy heavy-ion accelerators is presented. Selection of Moyer parameters and calculations of shielding thickness in conditions of point and extended beam losses were described. Methods of determination of roof shielding thickness on the basis of sky shine dose are given. The calculations are compared with some results of analogue Monte Carlo calculations

  3. Design and optimization of HPLWR high pressure Turbine gamma ray shield

    Highlights: • Shield around HPLWR high pressure turbine optimized by Monte Carlo method. • The occupancy period in the turbine building was considered in the optimization. • Shield thickness is significantly reduced when heavy concretes are used. • Shield thickness for BWRs is sufficient for HPWR if heavy concrete is used. - Abstract: This work proposes the optimum gamma ray shield thickness around the HPLWR high pressure turbine for different occupancy periods in the turbine building. Monte Carlo method was employed in the design process and only radioactive nitrogen-16 was considered as the source of radiation. Five grades of concrete (ordinary, magnetite, heavy magnetite, steel magnetite and barite) were used as shielding materials. The isotope source term in the high pressure turbine was estimated by modeling the HPLWR three pass core in MCNP and tracking the inventory using a simple algorithm. The high pressure turbine was thereafter modeled in MCNP with a concrete shield arrayed in layers around it. The surface flux tally and ICRP74 dose conversion coefficients were employed to estimate the dose profile across the shield. For some shielding materials, exponential functions were fitted on the calculated data to extrapolate dose values beyond the model thickness. The optimum shield thickness was determined by comparing the calculated dose profiles with dose limit proposals in the IAEA standard (NS-G-1.13) on radiation protection considerations during nuclear power plant design. It was observed that with a 120 cm thick heavy concrete shield, the turbine building would be safe for most occupancy periods. However for ordinary concrete the shield would require some extension to guarantee safety. For very long occupancy (more than 10 person hours per week), magnetite shield may also require slight extension. It can therefore be concluded that the shield thickness recommended for BWR turbines (which operate on a direct cycle like HPLWR) could be sufficient

  4. Biological shield design for a 10 MeV Rhodotron

    Highlights: ► We evaluate the produced radiations of the Rhodotron-TT200 and their attenuation to the permitted level. ► We apply analytical calculations to determine the shield material and thickness. ► We simulate the Rhodotron accelerator and its shielding using MCNPX code to make sure of results accuracy. -- Abstract: Radiation field of the Rhodotron-TT200 electron accelerator is determined in this study. Regarding the interactions of electron with matter, the produced radiations and their attenuation to the permitted level (i.e. 0.01 mrem/h) are evaluated and calculated. For this purpose analytical calculations are applied to determine the biological shield material and thickness. In order to make sure of results accuracy, Rhodotron accelerator and its shielding are simulated using MCNPX code and the results of analytical calculations and MCNPX code are compared with the experimental ones.

  5. A Cylindrical Shielding Design Concept for the Prototype Gen-IV Sodium-cooled Fast Reactor

    In the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), a metal fueled, blanket-free, pool type SFR concept is adopted to acquire the inherent safety characteristics and high proliferation-resistance. In the pool type fast reactor, the intermediate heat exchangers (IHXs), which transfer heat from the primary sodium pool to a secondary sodium loop, are placed inside of the reactor vessel. Hence, secondary sodium passing the IHXs can be radioactivated by a 23Na(n,g)24Na reaction, and radioactivated secondary sodium causes a significant dose in the Steam Generator Building (SGB). Therefore, a typical core of a pool type fast reactor is usually surrounded by a massive quantity of shields. In addition, the blanket composed of depleted uranium plays a role as superior shielding material; a significant increase in shields is required in the blanket-free pool type SFR. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR. In a conventional shielding design, massive axial shields are required to prevent irradiation of secondary sodium passing IHXs and they should be replaced according to the subassembly replacement in spite of negligible depletion of the shielding material. The proposed shielding design concept minimizes the quantity of shields without their replacement. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR such as a PGSFR. The proposed design concept satisfied the dose limit in the steam generator building successfully without introducing a large quantity of B4C shielding inside the subassembly

  6. Comparison of deterministic and Monte Carlo methods in shielding design

    In shielding calculation, deterministic methods have some advantages and also some disadvantages relative to other kind of codes, such as Monte Carlo. The main advantage is the short computer time needed to find solutions while the disadvantages are related to the often-used build-up factor that is extrapolated from high to low energies or with unknown geometrical conditions, which can lead to significant errors in shielding results. The aim of this work is to investigate how good are some deterministic methods to calculating low-energy shielding, using attenuation coefficients and build-up factor corrections. Commercial software MicroShield 5.05 has been used as the deterministic code while MCNP has been used as the Monte Carlo code. Point and cylindrical sources with slab shield have been defined allowing comparison between the capability of both Monte Carlo and deterministic methods in a day-by-day shielding calculation using sensitivity analysis of significant parameters, such as energy and geometrical conditions. (authors)

  7. The design study of the JT-60SU device. No.8. Nuclear shielding and safety design

    Results of nuclear shielding design study and safety analysis for the steady-state tokamak device JT-60SU are described. D-T operation (option) for two years is adopted in addition to ten years operation using deuterium. Design work has been done in accordance with general laws for radioisotopes handling in Japan as a guideline of safety evaluation, which is applied to the operation of present JT-60U device. Optimization of the shielding design for the device structure including vacuum vessel has been presented to meet with allowable limits of biological shielding determined in advance. It is shown that JT-60SU can be operated safely in the present JT-60 experimental building. It is planed to use 100g/year of tritium in D-T operation phase. A concept of multiple -barrier system is applied to the facility design to prevent propagation of tritium, in which the torus hall and the tritium removal room provide the tertiary confinement. From the design of atmosphere detritiation system for accidental tritium release, it is shown that tritium concentration level can be reduced to the allowable level after two weeks with reasonable compact size components. Safety assessment related to activation of coolant/air, and atmospheric tritium effluents are discussed. (author)

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

    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.

  9. Application of Particle Swarm Optimization Algorithm in Design of Multilayered Planar Shielding Body

    FUJiwei; HOUChaozhen; DOULihua

    2005-01-01

    Based on the basic electromagnetic wave propagation theory in this article, the Particle swarm optimization algorithm (PSO) is used in the design of the multilayered composite materials and the thickness of shielding body by the existent multilayered planar composite elec-tromagnetic shielding materials model, the different shielding materials of each layer can be designed under some kinds of circumstances: the prespecified Shielding effectiveness (SE), different incident angle and the prespecified band of frequencies. Finally the algorithm is simulated. At the same time the similar procedure can be implemented by Genetic algorithm (GA). The results acquired by particle swarm optimization algorithm are compared with there sults acquired by the genetic algorithm. The results indicate that: the particle swarm optimization algorithm is much better than the genetic algorithm not only in convergence speed but also in simplicity. So a more effective method (Particle Swarm Optimization algorithm) is offered for the design of the multilayered composite shielding materials.

  10. Shielding analyses for design of the upgraded JRR-3 research reactor, 2

    Shielding analyses of neutron beam holes have been presented for the shield design of the upgraded JRR-3 research reactor. Description is given about the calculational procedures and results for the standard beam hole, the beam hole for neutron radiography and the guide tunnels. The streaming analyses are made by using the MORSE-CG and DOT 3.5 codes. (author)

  11. Myo1c is designed for the adaptation response in the inner ear

    Batters, Christopher; Arthur, Christopher P.; Lin, Abel; Porter, Jessica; Geeves, Michael A.; Milligan, Ronald A.; Molloy, Justin E.; Coluccio, Lynne M.

    2004-01-01

    The molecular motor, Myo1c, a member of the myosin family, is widely expressed in vertebrate tissues. Its presence at strategic places in the stereocilia of the hair cells in the inner ear and studies using transgenic mice expressing a mutant Myo1c that can be selectively inhibited implicate it as the mediator of slow adaptation of mechanoelectrical transduction, which is required for balance. Here, we have studied the structural, mechanical and biochemical properties of Myo1c to gain an insi...

  12. Magnetic Field Design by Using Image Effect from Iron Shield

    Quanling PENG; S.M. McMurry; J.M.D.Coey

    2004-01-01

    Permanent magnet rings are presented, which exploit the image effect in the surrounding circular iron shields. The theory is given for a general permanent ring when the magnetization orientation Ψ at each coordinate angle Ψ changes by Ψ=(n+1)Ψ,where n is a positive or negative integer. For the uniformly magnetized case n=-1, the permanent ring produces no field in its bore, and the field is that of a dipole outside. When the ring is surrounded by a soft iron shield, its field becomes uniform in the bore, and zero outside the ring. The field can be varied continuously by moving the iron shield along the magnet axis.A small variable field device was constructed by using NdFeB permanent rings, which produced a field flux density of 0~0.5 T in the central region.

  13. Design, construction and erection of the biological shield wall for the Caorso nuclear power station

    This article describes the major aspects of the design, construction and erection of the biological shield wall encircling the reactor pressure vessel of the Caorso nuclear power station (Italy) (BWR-Mark 2, 840MWe)

  14. Shielding design of a mobile electron accelerator using Monte Carlo technique

    Shielding of a mobile electron accelerator of 0.6 MeV, 33 mA has been designed and examined by Monte Carlo technique. Based on a 3-D model of electron accelerator shielding which is designed with steel and lead shield, radiation leakage was examined using the MCNP code. Calculations using two different versions (version 4C2 and version 5) of MCNP showed agreements within statistical uncertainties, and the highest leakage expected is 5.5061 x 10-1 (1 ± 0.0454) μSvh-1, which is far below the tolerable radiation dose limit of 1 mSv (week)-1

  15. Final focus shielding designs for modern heavy-ion fusion power plant designs

    Latkowski, J. F.; Meier, W. R.

    2001-05-01

    Recent work in heavy-ion fusion accelerators and final focusing systems shows a trend towards less current per beam, and thus, a greater number of beams. Final focusing magnets are susceptible to nuclear heating, radiation damage, and neutron activation. The trend towards more beams, however, means that there can be less shielding for each magnet. Excessive levels of nuclear heating may lead to magnet quench or to an intolerable recirculating power for magnet cooling. High levels of radiation damage may result in short magnet lifetimes and low reliability. Finally, neutron activation of the magnet components may lead to difficulties in maintenance, recycling, and waste disposal. The present work expands upon previous, three-dimensional magnet shielding calculations for a modified version of the HYLIFE-II IFE power plant design. We present key magnet results as a function of the number of beams.

  16. Design of the shield door and transporter for the Culham Conceptual Tokamak Reactor Mark II

    In the Culham Conceptual Tokamak Reactor MK II access to the interior for blanket maintenance is through large openings in the fixed shield structure closed by removable shield doors when the reactor is operational. This report describes the design of the 200 tonne doors and the associated special-purpose remote operating transporter manipulator. The design, which has not been optimised, generally uses available commercial equipment and state-of-the-art techniques. (U.K.)

  17. A model for the rapid evaluation of active magnetic shielding designs

    Washburn, Scott Allen

    The use of active magnetic radiation shielding designs has the potential to reduce the radiation exposure received by astronauts on deep-space missions at a significantly lower mass penalty than designs that utilize only passive shielding. One of the common techniques for assessing the effectiveness of active or passive shielding designs is the use of Monte Carlo analysis to determine crew radiation exposure. Unfortunately, Monte Carlo analysis is a lengthy and computationally intensive process, and the associated time requirements to generate results make a broad analysis of the active magnetic shield design trade space impractical using this method. The ability to conduct a broad analysis of system design variables would allow the selection of configurations suited to specific mission goals, including mission radiation exposure limits, duration, and destination. Therefore, a rapid analysis method is required in order to effectively assess active shielding design parameters, and this body of work was developed in order to address this need. Any shielding analysis should also use complete representations of the radiation environment and detailed transport analyses to account for secondary particle production mechanisms. This body of work addresses both of these issues by utilizing the full Galactic Cosmic Radiation GCR flux spectrum and a detailed transport analysis to account for secondary particle effects due to mass interactions. Additionally, there is a complex relationship between the size and strength of an active shielding design and the amount and type of mass required to create it. This mass can significantly impact the resulting flux and radiation exposures inside the active shield, and any shielding analysis should not only include passive mass, but should attempt to provide a reasonable estimate of the actual mass associated with a given design. Therefore, a survey of active shielding systems is presented so that reasonable mass quantity and composition

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

    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/m2 neutron wall loading. The primary requirement for the shield is to limit the fast neutron fluence to 1019 n/cm2. 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 H2O/LiNO3 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 0C. 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

  19. Efficient time-independent method for conceptual design optimization of the national ignition facility primary shield

    Minimum-cost design concepts of the primary shield for the National (laser fusion) Ignition Facility are sought with the help of the SWAN optimization code. The computational method developed for this search involves incorporating the time dependence of the delayed photon field within effective delayed photon production cross sections. This method enables the time-dependent problem to be addressed using time-independent transport calculations, thus significantly simplifying and accelerating the design process. The search for constituents that will minimize the shield cost is guided by the newly defined equal cost replacement effectiveness functions. The minimum-cost shield design concept consists of a mixture of polyethylene and low-cost, low-activation materials, such as CaCO3 or silicon carbide, with boron added near the shield boundaries. An alternative approach to the target chamber design is analyzed. It involves placing the shield interior, rather than exterior to the main aluminum structural wall of the target chamber. The resulting inner shield design approach was found to be more expensive but inherently safer; the overall inventory of radioactive activation product it contains is one to two orders of magnitude lower than in the conventional design approach. 21 refs., 16 figs., 15 tabs

  20. Design improvement for reduction of decommissioning waste of PWR primary biological shield

    Most operating nuclear power plants were constructed with no attention to the amount of decommissioning waste. Consequently, a large portion of the primary concrete shield wall is irradiated by neutrons escaping from the reactor core to produce high concentration of activation products. These radioactive waste comprises of around 30% of the decommissioning waste. Under the circumstance that the rad waste disposal cost is continuously increasing, reduction of decommissioning waste becomes an important issue. In this study, a design improvement was attempted to reduce activation of the primary shield wall by placing a water-filled neutron shield tank between the reactor pressure vessel and the primary shield wall. Procedure for calculating the amount of activated radionuclides remaining at different cooling times were developed by MCNP-ORIGEN coupled calculations. Particular attention was paid to correction of activation cross sections since the ORIGEN code was designed for use in calculation of isotope generation and depletion in the operating reactor core where temperature is high, while temperature of the shield wall is low. The established procedure was applied to the 1500MWe APR+ model to evaluated the effectiveness of the neutron shield tank. Distributions of activation products for many different thickness of the neutron shield tank were calculated and an effective thickness was selected. Finally, by comparing the resulting activity distributions with the exemption criteria for radioactive waste, the expected cost reduction was assessed. The model applied in this study, however, is limited to preliminary design in terms of neutronics and does not take account any engineering problems which may be caused by installation of the shield tank. Practical engineering needs further detailed design analysis including cooling and cleaning of the shield water and other related engineering issues

  1. Shielding design of electron beam accelerators using supercomputer

    The MCNP5 neutron, electron, photon Monte Carlo transport program was installed on the KISTI's SUN Tachyon computer using the parallel programming. Electron beam accelerators were modeled and shielding calculations were performed in order to investigate the reduction of computation time in the supercomputer environment. It was observed that a speedup of 40 to 80 of computation time can be obtained using 64 CPUs compared to an IBM PC

  2. Optimised design of local shielding for the IFMIF/EVEDA beam dump

    This paper describes the local shielding design process of the IFMIF/EVEDA beam dump and the most relevant results obtained from the simulations. Different geometries and materials have been considered, and the design has been optimised taking into account the origin of the doses, the effect of the walls of the accelerator vault and the space restrictions. The initial idea was to shield the beam stopper with a large water tank of easy transport and dismantling, but this was shown to be insufficient to satisfy the dose limit requirements, basically due to photon dose, and hence a denser shield combining hydrogenous and heavy materials was preferred. It will be shown that, with this new shielding, dose rate outside the accelerator vault during operation complies with the legal limits and unrestricted maintenance operations inside most of the vault are possible after a reasonable cooling time after shutdown. (authors)

  3. A database-informed approach to new plant shielding design

    Document available in abstract form only, full text of document follows: To facilitate the definition and description of radiation dose rates in the numerous rooms and areas of a new nuclear power plant, a database approach was developed. This approach offers a number of benefits over more manual methods. A key benefit is that the selection of an appropriate shielding method to use in each area of the plant is greatly facilitated by virtue of the team's improved ability to grasp the significance of each of the individual sources that are candidates for making a significant contribution to the dose rate in each area. By understanding the level of relevant contribution - if any - of each of these candidate sources, an analyst is able to select a method that will define the contribution without becoming enmired in a model representing inappropriately high degrees of accuracy and modeling time. This database method, by allowing for an evolving understanding of dose rates and sources in the neighboring rooms for each portion of the plant, leads to substantial reductions in the effort of characterizing a plant's radiation environment. As an additional benefit, the database serves as a tool for documenting the shielding calculations themselves, automatically generating formatted sections including drawing and source term references, shielding calculation types, key dimensions, and results; these sections can form the starting point of a full calculation package. The approach offers a final project management benefit: estimating, tracking, and predicting the effort associated with the many calculations involved in such a project are greatly systematized, leading to more reliable manpower estimates. (authors)

  4. Design and Test of a Blast Shield for Boeing 737 Overhead Compartment

    Xinglai Dang

    2006-01-01

    Full Text Available This work demonstrates the feasibility of using a composite blast shield for hardening an overhead bin compartment of a commercial aircraft. If a small amount of explosive escapes detection and is brought onboard and stowed in an overhead bin compartment of a passenger aircraft, the current bins provide no protection against a blast inside the compartment. A blast from the overhead bin will certainly damage the fuselage and likely lead to catastrophic inflight structural failure. The feasibility of using an inner blast shield to harden the overhead bin compartment of a Boeing 737 aircraft to protect the fuselage skin in such a threat scenario has been demonstrated using field tests. The blast shield was constructed with composite material based on the unibody concept. The design was carried out using LS-DYNA finite element model simulations. Material panels were first designed to pass the FAA shock holing and fire tests. The finite element model included the full coupling of the overhead bin with the fuselage structure accounting for all the different structural connections. A large number of iterative simulations were carried out to optimize the fiber stacking sequence and shield thickness to minimize weight and achieve the design criterion. Three designs, the basic, thick, and thin shields, were field-tested using a frontal fuselage section of the Boeing 737–100 aircraft. The basic and thick shields protected the integrity of the fuselage skin with no skin crack. This work provides very encouraging results and useful data for optimization implementation of the blast shield design for hardening overhead compartments against the threat of small explosives.

  5. Design of portable radiation-shielding device for gamma radiography of pipe welds

    Industrial radiography is a nondestructive test method that makes use of ionizing radiation, requiring the adoption of adequate radiation safety measures. Radiation protection systems adopted for industrial radiography operations must take into account radiological workers as well as members of the public. The importance of this concern increases in construction sites, oil refineries, process plants and offshore installations. In such cases, industrial radiography of several thousand welded joints may be required while members of the public are working in other construction activities nearby. An analysis was performed on radiation safety standards adopted in industrial radiography operations in construction sites. Following a critical review, performance specifications were developed for a portable radiation-shielding device to be used in gamma radiography of pipe welds. Prototypes were designed, built and tested under actual construction site conditions, performing successfully. The radiation-shielding devices were developed to fit specific pipe sizes. A support system is designed to clamp to the pipe joint under examination. A collimator and a shield are permanently attached to this support system, according to optimum geometric arrangements of source, weld and film, intended to improve radiographic quality and to reduce setup and exposure times. The special-purpose collimator is penetrated by the isotope source, shielding radiation in all directions, except for the solid angle corresponding to the film. The shield, placed behind the film, covers that solid angle, ensuring that radiation is properly shielded in all directions. (author)

  6. Validity assessment of shielding design tools for ITER through analysis of benchmark experiment on SS316/water shield conducted at FNS/JAERI

    Maekawa, Fujio; Ikeda, Yujiro; Verzilov, Y.M.; Konno, Chikara; Wada, Masayuki; Maekawa, Hiroshi; Oyama, Yukio; Uno, Yoshitomo [Japan Atomic Energy Research Inst., Ibaraki (Japan)

    1996-12-31

    To assess validity of the shielding design tools for ITER, the benchmark experiment on SS316/water shield conducted at FNS/JAERI is analyzed. As far as a simple bulk shield of SS316/water is concerned, the followings are found assuming that no uncertainty is involved in the response functions of the design parameters. Nuclear data bases of JENDL Fusion File and FENDL/E-1.0 are valid to predict all the design parameters with uncertainties less than a factor of 1.25. At the connection legs between shield blanket modules and back plates, both MCNP and DOT calculations can predict helium production rate with uncertainties less than 10%. For the toroidal field coils on the midplane, all the nuclear parameters can be predicted with uncertainties less than a factor of 1.25 by MCNP and DOT with consideration of self-shielding correction of cross sections and energy group structure of 125-n and 40-{gamma}. The uncertainties for toroidal field coils are considerably smaller than the design margins secured to the shielding designs under ITER/EDA. 22 refs., 8 figs.

  7. Japanese contribution to the design of primary module of shielding blanket in ITER-FEAT

    Kuroda, Toshimasa; Hatano, Toshihisa; Miki, Nobuharu; Hiroki, Seiji; Enoeda, Mikio; Ohmori, Junji; Akiba, Masato [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Sato, Shinichi [Kawasaki Heavy Industries, Ltd., Tokyo (Japan)

    2003-02-01

    Japanese contributions to the design activity on the shielding blanket module consisting of the separable first wall and the shield block for ITER-FEAT are compiled. Temperature and stress distributions in the first wall and the shield block are analyzed and evaluated with 2-D and 3-D models for steady state and also for transient condition according to plasma ramp-up and ramp-down. While temperatures and stresses in the first wall satisfy their allowable values, those in a front part of the shield block exceed the allowable guideline. Based on this result, design improvements are suggested. Coolant flow and pressure distributions along the complicated coolant channel in the shield block are preliminary analyzed. Though heat removal is satisfactory in all coolant channels, back flows due to choking in coolant collectors are found. Design improvements to avoid the choking are suggested. Electromagnetic forces acting on blanket modules are analyzed with detailed 3-D models of solid elements for different disruption scenarios. The maximum moment around radial axis is 1.36 MNm on module no.5 under fast upward VDE, and the maximum moment around vertical axis is 1.47 MNm on module no.1 under fast downward VDE. The supporting beam of the first wall with welded attachment to the shield block is designed. Required welding thickness and support conditions to withstand electromagnetic forces are estimated. Strength of the shield block at the region mating the flexible cartridge is also estimated. Though the shield block surface attached by the flexible cartridge shows sufficient strength, the internal thread mating the Inconel bolt would need more length. In addition, water-to-water leak detection system in case main supply/return manifolds are located within the vacuum vessel is designed. By using Kr as the tracer material, the possibility of water-to-water leak detection and the concept of the detection system are shown. The design of the shielding blanket of ITER-FEAT has

  8. Japanese contribution to the design of primary module of shielding blanket in ITER-FEAT

    Japanese contributions to the design activity on the shielding blanket module consisting of the separable first wall and the shield block for ITER-FEAT are compiled. Temperature and stress distributions in the first wall and the shield block are analyzed and evaluated with 2-D and 3-D models for steady state and also for transient condition according to plasma ramp-up and ramp-down. While temperatures and stresses in the first wall satisfy their allowable values, those in a front part of the shield block exceed the allowable guideline. Based on this result, design improvements are suggested. Coolant flow and pressure distributions along the complicated coolant channel in the shield block are preliminary analyzed. Though heat removal is satisfactory in all coolant channels, back flows due to choking in coolant collectors are found. Design improvements to avoid the choking are suggested. Electromagnetic forces acting on blanket modules are analyzed with detailed 3-D models of solid elements for different disruption scenarios. The maximum moment around radial axis is 1.36 MNm on module no.5 under fast upward VDE, and the maximum moment around vertical axis is 1.47 MNm on module no.1 under fast downward VDE. The supporting beam of the first wall with welded attachment to the shield block is designed. Required welding thickness and support conditions to withstand electromagnetic forces are estimated. Strength of the shield block at the region mating the flexible cartridge is also estimated. Though the shield block surface attached by the flexible cartridge shows sufficient strength, the internal thread mating the Inconel bolt would need more length. In addition, water-to-water leak detection system in case main supply/return manifolds are located within the vacuum vessel is designed. By using Kr as the tracer material, the possibility of water-to-water leak detection and the concept of the detection system are shown. The design of the shielding blanket of ITER-FEAT has

  9. Present status and future needs of nuclear data for high energy accelerator shielding design calculation

    With increasing application of high energy accelerators, their shielding design study has become very important. The present study discusses the current status of nuclear data, especially inclusive neutron production cross sections (thin target yields), thick target yields, neutron transmission through shield and cross sections, and activation and spallation cross sections of neutrons and protons. Their experimental and theoretical data are compared and further needs of basic nuclear data for shielding design calculation are described. Nuclear data in high energy regions above 20 MeV, such as double differential neutron production cross section, double differential neutron yield and neutron reaction cross section, are very poor both in number and in accuracy at the present stage. The shielding design calculation requires the comprehensive data set of these cross sections and/or the more accurate computer code to calculate these quantities. Further effort is expected to carry out shielding and cross section experiments using the monoenergetic neutron beam line which is under construction in Takasaki Branch of JAERI (Japan Atomic Energy Research Institute). (N.K.)

  10. Design concepts to minimize the activation of the biological shield of light-water reactors

    An investigation, concentrating on the nuclear aspects, has been made into the concept of minimizing the activation of the biological shield by substituting the material concrete with other neutron-shielding materials. This work was for nuclear plant designs which have a non-supporting inner shield wall such as that in the General Electric BWR/6 and the Kraftwerk Union PWR. The attenuation performance and activation levels have been analysed. Based on this analysis the performance of the materials in relation to that of concrete was assessed. Other non-nuclear properties were considered but the engineering problems were not addressed. The conclusion reached was that the concept was credible but would require a more rigorous examination in terms of structural design, economics and licensability

  11. Methodology for performing the design review of plant shielding and environmental qualification required after TMI-2

    After TMI-2, the NRC issued several documents requiring a design review of plant shielding and environmental qualification for spaces/systems which may be used in post-accident operations. The objective of this requirement was to confirm the shielding provided around systems that may contain highly radioactive materials as a result of an accident, which may unduly limit personnel occupancy or degrade safety equipment by radiation. The objective of this paper is to describe a methodology developed to obtain the inputs needed for shielding and dose calculations. It has to be pointed out that criteria for determination of vital areas, system flow paths and fluid data were not well defined by the NRC and the events considered were beyond design basis. The paper described (1) the approach adopted, exposing the conservative hypothesis applied to meet the intention of requirements, (2) more significant differences in this job between PWR and BWR, and (3) some complementary applications of the work developed

  12. Shielding design for target trolley for a spallation neutron source in the J-PARC project

    To pull out a mercury target horizontally and to transfer it to hot cell for replacement, a target trolley will be installed in a spallation neutron source facility in the High-intensity Proton Accelerator Project (J-PARC). According to the progress of the target trolley design and the modification of building design, the shielding performance of the mercury target trolley was evaluated. Target doses are 25 μSv/h at a manipulator operation room behind a concrete wall of 1.5 m, and 0.5 μSv/h at a non-controlled area behind another concrete wall of 1.5 m, respectively. Bending mercury piping and gaps between the target trolley and surrounding liners, etc, were modeled 3-dimensionally in order to evaluate streaming effects. Radiation doses around the target trolley were evaluated using a 3-dimensional Monte Carlo calculation code NMTC/JAM, applying the above three-dimensional model. Since concrete walls could be considered to be simple bulk shields, doses for the manipulator room and the non-controlled area were calculated using 1-dimensional spherical model with a Monte Carlo code MCNPX by using neutron fluxes at the back of the target trolley as a source. By replacing concrete shield with iron shield and reduction of gap streaming effects, the target trolley radiation shield structures were determined, which could suppress radiation doses in the manipulator room and the non-controlled area below the target doses. (author)

  13. Design of shielded encircling send-receive type pulsed eddy current probe using numerical analysis method

    An encircling send-receive type pulsed eddy current (PEC) probe is designed for use in aluminum tube inspection. When bare receive coils located away from the exciter were used, the peak time of the signal did not change although the distance from the exciter increased. This is because the magnetic flux from the exciter coil directly affects the receive coil signal. Therefore, in this work, both the exciter and the sensor coils were shielded in order to reduce the influence of direct flux from the exciter coil. Numerical simulation with the designed shielded encircling PEC probe showed the corresponding increase of the peak time as the sensor distance increased. Ferrite and carbon steel shields were compared and results of the ferrite shielding showed a slightly stronger peak value and a quicker peak time than those of the carbon steel shielding. Simulation results showed that the peak value increased as the defect size (such as depth and length) increased regardless of the sensor location. To decide a proper sensor location, the sensitivity of the peak value to defect size variation was investigated and found that the normalized peak value was more sensitive to defect size variation when the sensor was located closer to the exciter.

  14. Design of special lead shielding facilities for medium and low energy gamma radiation

    The cardinal principles of radiation protection from external sources are based on four factors: time, distance, shielding and activity. These factors are more or less rigorously observed inside the hot room of the nuclear medicine laboratory. Unfortunately the importance of radiation protection during data acquisition, storage of radioisotopes and waste products are often overlooked. The patients are a source of significant radiation and the nuclear medicine personnel must consciously take measures for protection against this source during patient handling. There are many commercial shielding materials for the partition from radioisotopes. Commercially available lead cupboard and lead glass used as partition from radioisotopes are very expensive. A project was therefore undertaken to develop practical low cost shielding against radiation sources. The present article outlines the various lead shielding facilities designed and built for medium and low energy gamma radiation. The designs were at first sketched out on paper giving specific shapes and measurement of each structure. Model structures were then made accordingly and the protective capacities of these structures were checked by mathematical calculation from the equation of gamma ray attenuation. In the design and structures lead plating thickness was in between 0.30 to 5.0 cm. Correction and restructuring of the models were undertaken to achieve the designs satisfactory. Different structures served different aspects of radiation protection. These structures sculptured as per designs are now in use for radiation protection at the Institute of Nuclear Medicine. (author) 2 tabs., 6 figs., 4 refs

  15. Guide to beamline radiation shielding design at the Advanced Photon Source

    This document is concerned with the general requirements for radiation shielding common to most Advanced Photon Source (APS) users. These include shielding specifications for hutches, transport, stops, and shutters for both white and monochromatic beams. For brevity, only the results of calculations are given in most cases. So-called open-quotes special situationsclose quotes are not covered. These include beamlines with white beam mirrors for low-pass energy filters (open-quotes pink beamsclose quotes), extremely wide band-pass monochromators (multilayers), or novel insertion devices. These topics are dependent on beamline layout and, as such, are not easily generalized. Also, many examples are given for open-quotes typicalclose quotes hutches or other beamline components. If a user has components that differ greatly from those described, particular care should be taken in following these guidelines. Users with questions on specific special situations should address them to the APS User Technical Interface. Also, this document does not cover specifics on hutch, transport, shutter, and stop designs. Issues such as how to join hutch panels, floor-wall interfaces, cable feed-throughs, and how to integrate shielding into transport are covered in the APS Beamline Standard Components Handbook. It is a open-quotes living documentclose quotes and as such reflects the improvements in component design that are ongoing. This document has the following content. First, the design criteria will be given. This includes descriptions of some of the pertinent DOE regulations and policies, as well as brief discussions of abnormal situations, interlocks, local shielding, and storage ring parameters. Then, the various sources of radiation on the experimental floor are discussed, and the methods used to calculate the shielding are explained (along with some sample calculations). Finally, the shielding recommendations for different situations are given and discussed

  16. Design and construction of a movement mechanical system for a shield detector of a neutron diffractometer

    We present the design parameters of the mechanical system for a shield movement detector of a neutron diffractometer and the calculations to determine the power required to produce the rotation. The movement of the detection system is an essential part in order to get neutron diffraction spectra of a crystal. (author)

  17. Analysis of the ORNL/TSF GCFR Grid-Plate Shield Design Confirmation Experiment

    The results of the analysis of the GCFR Grid-Plate Shield Design Confirmation Experiment are presented. The experiment, performed at the ORNL Tower Shielding Facility, was designed to test the adequacy of methods and data used in the analysis of the GCFR design. In particular, the experiment tested the adequacy of methods to calculate: (1) axial neutron streaming in the GCFR core and axial blanket, (2) the amount and location of the maximum fast-neutron exposure to the grid plate, and (3) the neutron source leaving the top of the grid plate and entering the upper plenum. Other objectives of the experiment were to verify the grid-plate shielding effectiveness and to assess the effects of fuel-pin and subassembly spacing on radiation levels in the GCFR. The experimental mockups contained regions representing the GCFR core/blanket region, the grid-plate shield section, and the grid plate. Most core design options were covered by allowing: (1) three different spacings between fuel subassemblies, (2) two different void fractions within a subassembly by variation of the number of fuel pins, and (3) a mockup of a control-rod channel

  18. MCNP benchmark calculation: GCFR grid-plate shield design, configuration II.A

    This report describes the Monte Carlo MCNP analysis of one of the GCFR Shield Design experimental configurations which has been constructed and analyzed at the Test Shielding Facility in ORNL. It is a part of the benchmarking program for MCNP, which has been agreed upon with HRB, Mannheim. The calculated response results for the selected detectors agree within 10 % with the measured ones, what can be considered as a very good agreement. The code appears to be a reliable tool for the analysis of similar systems. (author)

  19. Design, fabrication and testing results of vacuum vessel, thermal shield and cryostat of EAST

    The EAST (Experimental Advanced Superconducting Tokamak) is an advanced steady-state plasma physics experimental device, which has been approved by the Chinese government and being constructed as the Chinese national nuclear fusion Research project. The vacuum vessel as one of the key components for this device can provide ultra-high vacuum and cleanly location of plasma operation. It is a torus with 'D' shaped cross-section, double wall, upper vertical ports, lower vertical ports, horizontal ports and flexible supports. The cryostat is a large single walled vessel surrounding the entire basic machine with central cylindrical section and two end enclosures, a flat base structure with external reinforcements and dome-shaped lid structure. It provides the thermal barrier with the base pressure of 5x10-4 Pa between the ambient temperature testing hall and the liquid helium cooled superconducting magnet. The thermal shields comprise the vacuum vessel thermal shield (VVTS), between the vacuum vessel and the cold Toroidal Field (TF) coil structures, the cryostat thermal shield (CTS), covering the walls of the cryostat, thereby preventing direct line of sight of the room temperature walls to the cold structures, the vacuum port thermal shields (VPTS) that enclose the port connection ducts. This paper is a report of the structure design and stress analyses for the vacuum vessel, thermal shield and cryostat. And also some key R and D and testing results for these components have been presented. (author)

  20. Dose conversion coefficients in the shielding design calculation for high energy proton accelerator facilities

    Dose quantity in the shielding design calculation was changed from the 1 cm depth dose equivalent to effective dose on the occasion of the introduction of the International Commission on Radiological Protection (ICRP) 1990 Recommendations (ICRP Publication 60) into domestic laws. As dose conversion coefficients in the shielding design calculation for accelerator facilities, the values for front irradiation (AP irradiation geometry) of neutrons below 20 MeV based on the ICRP Publication 74 are listed in the accompanying table of the domestic laws, but the values for neutrons above 20 MeV are not shown in the accompanying table. The status of dose conversion coefficients for neutrons above 20 MeV was surveyed and the effective dose rates behind the concrete shield of proton accelerator facilities were obtained by using typical neutron spectra and various dose conversion coefficients. As a result of consideration, the effective dose conversion coefficients for front irradiation of neutrons above 20 MeV evaluated by using HERMES code system was recommended for high energy neutrons in the shielding design calculation of proton accelerator facilities and 77 energy group averaged dose conversion coefficients was produced from thermal energy to 2 GeV. (author)

  1. Application of ISOCS sourceless calibration software on designment for the shield thickness of collimator

    The designment for collimator need consider not only the factors about spatial resolution, sensitivity etc, but also the effect on shadow areas because of the shield thickness. One optimized method for the designment of the shield thickness of collimator, and the curve of the angle response and the percent of count contribution at different angle range for the infinite plane source were calculated with ISOCS when the types of collimator using in this process were 5 cm-30, 10 cm-30 and 15 cm-30. The results indicate that when the shield thickness of collimator is 5 cm, and the azimuthal angle is 90 degrees, the count countribution still exist, and the shadow area is large. And when the thickness is 15 cm,the visual field of the detector is controlled well. The method not only supply basis to the designment for the shield thickness of collimator, but also can be seen as the references of efficiency calibration and uncertainty evaluation for the detector with collimator. (authors)

  2. Shielding analysis and design of the KIPT experimental neutron source facility of Ukraine.

    Zhong, Z.; Gohar, M. Y. A.; Naberezhnev, D.; Duo, J.; Nuclear Engineering Division

    2008-10-31

    Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an experimental neutron source facility based on the use of an electron accelerator driven subcritical (ADS) facility [1]. The facility uses the existing electron accelerators of KIPT in Ukraine. The neutron source of the sub-critical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and the electron energy in the range of 100 to 200 MeV, [2]. The main functions of the facility are the production of medical isotopes and the support of the Ukraine nuclear power industry. Reactor physics experiments and material performance characterization will also be carried out. The subcritical assembly is driven by neutrons generated by the electron beam interactions with the target material. A fraction of these neutrons has an energy above 50 MeV generated through the photo nuclear interactions. This neutron fraction is very small and it has an insignificant contribution to the subcritical assembly performance. However, these high energy neutrons are difficult to shield and they can be slowed down only through the inelastic scattering with heavy isotopes. Therefore the shielding design of this facility is more challenging relative to fission reactors. To attenuate these high energy neutrons, heavy metals (tungsten, iron, etc.) should be used. To reduce the construction cost, heavy concrete with 4.8 g/cm{sup 3} density is selected as a shielding material. The iron weight fraction in this concrete is about 0.6. The shape and thickness of the heavy concrete shield are defined to reduce the biological dose equivalent outside the shield to an acceptable level during operation. At the same time, special attention was give to reduce the total shield mass to reduce the construction cost. The shield design is configured

  3. Design and fabrication of shielding for gamma spectrometry

    To have a system of gamma spectrometry in the Radiological Mobile Unit No. 1 (UMOR-1) was designed and manufactured an armor-plating appropriate to this, to make analysis of radioactive samples in place in the event of a radiological emergency, besides being able to give support to the Management of Radiological Safety, and even to give service of sample analysis of other Institutions. (Author)

  4. Design and shielding calculation for a PET/CT facility

    Following the AAPM Task Group Report No. 108, the NCRP Report No. 147 recommendations and the Cuban's local regulations for nuclear medicine practice were carried out the safety planning and design of a new PET/CT facility for the Nuclear Medicine Department of 'Hermanos Ameijeiras' Hospital. It should be installed in the top floor of the NM building (3th floor), occupied by offices, classrooms and ancillaries areas, meanwhile in the second floor is working the conventional nuclear medicine department. The radiation doses were evaluated in areas of the second, third and quarter floor taking into account the pet isotope, the workload, the occupancy factors of each place, the use factors of different sources and the dose reduction factors, warranty the accomplish of the Cuban dose restrictions associated to the nuclear medicine practice. In each point of calculation was considered the contribution from each source to the total dose, as well as the contribution of the CT in the adjacent room to the imaging room. For the proper facility design was considered the transmission factors of the existing barriers, and calculated the new ones to be added between each source and the estimation point, keeping in mind the space limitations. The PET/CT design plan meet all the needs, the development of the project is consistent with the mission of the facility and the radiation protection regulations of nuclear medicine. (Author)

  5. Design, fabrication and testing results of vacuum vessel, thermal shield and cryostat of EAST

    The EAST (Experimental Advanced Superconducting Tokamak) is an advanced steady-state plasma physics experimental device, which has been approved by the Chinese government and being constructed as the Chinese national nuclear fusion Research project. Vacuum vessel is the location for the operation of plasma as one of the key component for EAST device. During it operation the vacuum vessel will not only endure the electromagnetic force due to the plasma disruption and Halo current but also the pressure of boride water and the thermal stress owing to the 250 deg C baking out by the hot pressure nitrogen gas or the 100 deg C hot wall during plasma operation. The cryostat is a large single walled vessel surrounding the entire Basic Machine with central cylindrical section and two end enclosures, a flat base structure with external reinforcements and dome-shaped lid structure. It provides the thermal barrier with the base pressure of 5x10-4 Pa between the ambient temperature testing hall and the liquid helium cooled superconducting magnet. The thermal shields comprise the vacuum vessel thermal shield (VVTS), between the vacuum vessel and the cold TF coil structures, the cryostat thermal shield (CTS), covering the walls of the cryostat, thereby preventing direct line of sight of the room temperature walls to the cold structures, the vacuum port thermal shields (VPTS) that enclose the port connection ducts. The thermal shields are made of double-wall panels, sandwich structure consist of two stainless steel panels and weld quadrate cooling pipe in between the total surface of the thermal shields is about 351 m2. This paper is a report of the structure design and mechanical analyses on the vacuum vessel, thermal shield and cryostat. According to the allowable stress criteria of ASME, the maximum integrated stress intensity on these key components is less than the allowable design stress intensity 3 Sm. The fabrication for these components was completed in 2004 and has been

  6. Shielding Design for Adjacent, Underground Buildings of a Megavoltage Radiotherapy Facility.

    Sanz, Darío Esteban

    2016-07-01

    In a radiotherapy facility, safety in areas next to the treatment room can be of concern when irradiating downward due to oblique x-ray transmission through the floor and/or walls, especially in areas immediately adjacent or underground. Even when there is no basement underneath, a usual conservative solution is to build a thick concrete slab as the base for the treatment room. Of course, this implies deeper soil excavation and higher associated costs. As a convenient alternative, the limiting walls can be buried a certain depth below floor level to shield oblique, downward irradiation. Besides, for space considerations, laminated barriers are usually employed, and some additional shielding to the floor may be required (L-shaped barriers). In this work, the author introduces an analytical method for calculating the required wall penetration below floor level or, alternatively, the additional floor shielding for L-shaped barriers, taking into account in either case the attenuation properties of the earth underneath the vault. Interestingly, the required penetration depth for a given wall barrier (primary or secondary), relative to a reference thickness, is only a function of basic attenuation data. Likewise, for a laminated, lead-concrete barrier, the required dimensions depend on the relative amount of lead used for the wall and on the corresponding attenuation data. The shielding design criteria developed in this work to protect underground nearby sites is conservative in nature, yet it yields optimal shield dimensions for wall footing and for wall-floor shielding, avoiding the need to construct oversized concrete slab floors. PMID:27218288

  7. Analysis of the conceptual shielding design for the upflow Gas-Cooled Fast Breeder Reactor

    Slater, C.O.; Reed, D.A.; Cramer, S.N.; Emmett, M.B.; Tomlinson, E.T.

    1981-01-01

    Conceptual Shielding Configuration III for the Gas-Cooled Fast Breeder Reactor (GCFR) was analyzed by performing global calculations of neutron and gamma-ray fluences and correcting the results as appropriate with bias factors from localized calculations. Included among the localized calculations were the radial and axial cell streaming calculations, plus extensive preliminary calculations and three final confirmation calculations of the plenum flow-through shields. The global calculations were performed on the GCFR mid-level and the lower and upper plenum regions. Calculated activities were examined with respect to the design constraint, if any, imposed on the particular activity. The spatial distributions of several activities of interest were examined with the aid of isoplots (i.e., symbols are used to describe a surface on which the activity level is everywhere the same). In general the results showed that most activities were below the respective design constraints. Only the total neutron fluence in the core barrel appeared to be marginal with the present reactor design. Since similar results were obtained for an earlier design, it has been proposed that the core barrel be cooled with inlet plenum gas to maintain it at a temperature low enough that it can withstand a higher fluence limit. Radiation levels in the prestressed concrete reactor vessel (PCRV) and liner appeared to be sufficiently below the design constraint that expected results from the Radial Shield Heterogeneity Experiment should not force any levels above the design constraint. A list was also made of a number of issues which should be examined before completion of the final shielding design.

  8. Analysis of the conceptual shielding design for the upflow Gas-Cooled Fast Breeder Reactor

    Conceptual Shielding Configuration III for the Gas-Cooled Fast Breeder Reactor (GCFR) was analyzed by performing global calculations of neutron and gamma-ray fluences and correcting the results as appropriate with bias factors from localized calculations. Included among the localized calculations were the radial and axial cell streaming calculations, plus extensive preliminary calculations and three final confirmation calculations of the plenum flow-through shields. The global calculations were performed on the GCFR mid-level and the lower and upper plenum regions. Calculated activities were examined with respect to the design constraint, if any, imposed on the particular activity. The spatial distributions of several activities of interest were examined with the aid of isoplots (i.e., symbols are used to describe a surface on which the activity level is everywhere the same). In general the results showed that most activities were below the respective design constraints. Only the total neutron fluence in the core barrel appeared to be marginal with the present reactor design. Since similar results were obtained for an earlier design, it has been proposed that the core barrel be cooled with inlet plenum gas to maintain it at a temperature low enough that it can withstand a higher fluence limit. Radiation levels in the prestressed concrete reactor vessel (PCRV) and liner appeared to be sufficiently below the design constraint that expected results from the Radial Shield Heterogeneity Experiment should not force any levels above the design constraint. A list was also made of a number of issues which should be examined before completion of the final shielding design

  9. Study on shielding design methods for fusion reactors using benchmark experiments

    In this study, a series of engineering benchmark experiments have been performed on the critical issues of shielding designs for DT fusion reactors. Based on the experiments, calculational accuracy of shielding design methods used in the ITER conceptual design, discrete ordinates code DOT3.5 and Monte Carlo code MCNP-3, have been estimated, and difficulties on calculational methods have been revealed. Furthermore, the feasibility for shielding designs have been examined with respect to a discrete ordinates code system BERMUDA which is developed to attain high accuracy of calculation. As for neutron streaming in an off-set narrow gap experimental assembly made of stainless steel, DOT3.5 and MCNP-3 codes reproduced the experiments within the accuracy presumed in the ITER conceptual design. DOT3.5 and MCNP-3 codes are available for secondary γ ray nuclear heating in a type 316L stainless steel assembly and neutron streaming in a multi-layered slit experimental assembly, respectively. Moreover, BERMUDA-2DN code is an effective tool as to neutron deep penetration in a type 316L stainless steel assembly and the neutron behavior in a large cavity experimental assembly. (author)

  10. Study on shielding design methods for fusion reactors using benchmark experiments

    Nakashima, Hiroshi (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment)

    1992-02-01

    In this study, a series of engineering benchmark experiments have been performed on the critical issues of shielding designs for DT fusion reactors. Based on the experiments, calculational accuracy of shielding design methods used in the ITER conceptual design, discrete ordinates code DOT3.5 and Monte Carlo code MCNP-3, have been estimated, and difficulties on calculational methods have been revealed. Furthermore, the feasibility for shielding designs have been examined with respect to a discrete ordinates code system BERMUDA which is developed to attain high accuracy of calculation. As for neutron streaming in an off-set narrow gap experimental assembly made of stainless steel, DOT3.5 and MCNP-3 codes reproduced the experiments within the accuracy presumed in the ITER conceptual design. DOT3.5 and MCNP-3 codes are available for secondary {gamma} ray nuclear heating in a type 316L stainless steel assembly and neutron streaming in a multi-layered slit experimental assembly, respectively. Moreover, BERMUDA-2DN code is an effective tool as to neutron deep penetration in a type 316L stainless steel assembly and the neutron behavior in a large cavity experimental assembly. (author).

  11. Conceptual design of neutron shield for ECH launcher on D-T fusion reactors

    Conceptual design of Electron Cyclotron Heating and Current Drive (ECH/ECCD) launcher for fusion reactors is described. The ECH injection power of 20∼25 MW per a port and the shielding capability to protect superconducting magnets and ECH torus windows from radiation damages are required for the ECH launcher in deuterium - tritium (D-T) fusion reactors. The conceptual design study and the nuclear analysis (2D) for the ECH launcher to qualify the design specification were carried out. The guideline of the detail design for the ECH launcher was obtained. (author)

  12. Shielding experiments

    Shielding mock-up experiments for Prototype Fast Breeder Reactor (PFBR) and Advanced Heavy Water Reactor (AHWR) are carried out in shielding corner facility of APSARA reactor, to assess the overall accuracy of the codes and nuclear data used in reactor shield design. As APSARA is a swimming pool-type thermal reactor, for fast reactor experiments, typical fast reactor shielding facility was created by using uranium assemblies as spectrum converter. The flux was also enhanced by replacing water by air. Experiments have been carried out to study neutron attenuation through typical fast reactor radial and axial bulk shielding materials such as steel, sodium, graphite, borated graphite and boron carbide. A large number of reaction rates, sensitive to different regions of the neutron energy spectrum, were measured using foil activation and Solid State Nuclear Track Detector (SSNTD) techniques. These experimental results were analysed using computational tools normally used in design calculations, viz., discrete ordinate transport codes with multigroup cross section sets. Comparison of measured reaction rates with calculations provided suitable bias factors for parameters relevant to shield design, such as sodium activation, fast neutron fluence, fission equivalent fluxes etc. The measured neutron spectrum on the incident face of shield model compares well with the calculated fast reactor blanket leakage neutron spectrum. The comparison of calculated reaction rates within shield model indicate that the calculations suffer from considerable uncertainties, in shield models with boron carbide/borated graphite. For AHWR shielding experiments, no spectrum converter was used as it is also a thermal reactor. Radiation streaming studies through penetrations/ducts of various shapes and sizes relevant to AHWR shielding were carried out. (author)

  13. The study on mix radio design and construction technology of radiation-shielding and high-density concrete

    Newly-constructed nuclear facilities requires the shielding concrete with density of 4600 kg/m3 or even higher for shielding of γ rays or neutron rays. Systemic tests and studies on radiation shielding concrete (neutrons and γ-ray absorbing) were conducted in such aspects as mix ratio design, preparation, construction technology, shielding effect, uniform shielding etc. The results show concrete for γ ray could be prepared with an average density of 4670 kg/m3, compressive strength of 37 MPa and permeability-resistant grade of P10. For neutron ray shield, the prepared concrete could be at an average density of 4680 kg/m3, with crystal water of 2.65% (wt) and boron of 0.11% (wt), and compressive strength of 45.6 MPa. (authors)

  14. Aspects of the core shielding assessment for the FASTEF-MYRRHA design

    In the frame of the FP7 European project Central Design Team (CDT), an extensive simulation study has been done to assess the main shielding problems in view of the construction of the MYRRHA accelerator-driven system at SCK·CEN in Mol (Belgium). A method based on the combined use of the two Monte Carlo codes MCNPX and FLUKA has been developed, with the goal to characterize realistic neutron fields around the core barrel and build complex source terms, to be used in detailed analyses of the radiation fields due to the system in operation, and of the coupled residual radiation. The results evidenced a powerful way to analyze the shielding and activation problems, with direct and clear implications on the design solutions. (author)

  15. Design and fabrication of radiation shielded laser ablation ICP-MS system

    Ha, Yeong Keong; Han, Sun Ho; Park, Soon Dal; Park, Yang Soon; Jee, Kwang Yong; Kim, Won Ho

    2006-09-15

    In relation to high burn up and extended fuel cycle for the fuel cycle efficiency, we need to take chemical analysis of spent nuclear fuel for the integrity of nuclear fuel at high burn up. to measure the isotopic distribution of fission product in a high burn up nuclear fuel, radiation shielded laser ablation system was designed and fabricated. By probing the sample with a laser beam, micro sampling system for the mass analyzer was successfully developed. This report describes the structural design and the function of developed radiation shielded LA system. This system will be used for the analysis of isotopic distribution from core to rim of a spent nuclear fuel prepared from the hot-cell in PIE facility and/or an irradiated fuel from research reactor.

  16. Optimized design of shields for diagnostic X rays with NCRP 147 technique

    A comparison among the design techniques of shielding for X-ray diagnostic rooms with the NCRP 49 (1976) report technique, AAPM 39 (1993) Y the one of the NCRP 147 (2005) technique. The designs correspond to a room of conventional X-rays, one of fluoroscopy, one of tomography Y one of mammography. In all the cases it demonstrates that the NCRP 49 technique overestimate the shieldings. The causes of the overestimation of the NCRP 49 can be attributed to: a) high values of the work charge that don't consider the spectral fluence of the photons that are present in each room, b) to the differences in the values of the kerma in air without attenuation for the dispersed primary radiation Y of leakage among both reports. (Author)

  17. Shielding analysis of depleted uranium silicate filler concept for spent fuel canister designs

    A Depleted Uranium Silicate Container Backfill System (DUSCOBS) has been proposed at Oak Ridge National Laboratory. This concept suggests the use of small, depleted-uranium silicate glass beads as a backfill material inside storage, transportation, and repository waste packages containing spent nuclear fuel. Use of this backfdl material would substantially reduce external dose rates from a waste canister, allowing a reduction of the amount of external shielding required. This paper summarizes the results of scoping studies to estimate the dose reduction from the use of DUSCOBS in a conceptual canister design, and to determine what design modifications are required to offset the increased mass of the system, while simultaneously maintaining sufficient shielding to meet external dose rate limits

  18. Physical analysis of the shielding capacity for a lightweight apron designed for shielding low intensity scattering X-rays

    Kim, Seon Chil; Choi, Jeong Ryeol; Jeon, Byeong Kyou

    2016-07-01

    The purpose of this paper is to develop a lightweight apron that will be used for shielding low intensity radiation in medical imaging radiography room and to apply it to a custom-made effective shielding. The quality of existing aprons made for protecting our bodies from direct radiation are improved so that they are suitable for scattered X-rays. Textiles that prevent bodies from radiation are made by combining barium sulfate and liquid silicon. These materials have the function of shielding radiation in a manner like lead. Three kinds of textiles are produced. The thicknesses of each textile are 0.15 mm, 0.21 mm, and 0.29 mm and the corresponding lead equivalents are 0.039 mmPb, 0.095 mmPb, 0.22 mmPb for each. The rate of shielding space scattering rays are 80% from the distance of 0.5 m, 86% from 1.0 m, and 97% from 1.5 m. If we intend to approach with the purpose of shielding scattering X-rays and low intensity radiations, it is possible to reduce the weight of the apron to be 1/5 compared to that of the existing lead aprons whose weight is typically more than 4 kg. We confirm, therefore, that it is possible to produce lightweight aprons that are used for the purpose of shielding low dose radiations.

  19. Driving Parts Optimization Design for Radiation Shielding Doors of Proton Accelerator Research Center

    PEFP(Proton Engineering Frontier Project) was Launched in 2002 as one of the 21st Century Frontier R and D Programs of MOST(Ministry of Science and Technology). Gyeongju city was selected as the project host site in March, 2006, where 'Proton Accelerator Research Center' was going to be constructed. After starting the design in 2005, the Architectural and Civil design work has been performed by 2010. Since the Earthwork was started in 2009, the Construction works of Accelerator Facilities has been going smoothly to complete by 2012. In this paper, we describe driving Parts optimization design for radiation shielding doors of Proton Accelerator Research Center

  20. Design approach of the vacuum vessel and thermal shields towards assembly at the ITER-site

    Utin, Yu. [ITER Organization, 13108 St. Paul lez Durance (France)], E-mail: yuri.utin@iter.org; Ioki, K.; Bachmann, Ch. [ITER Organization, 13108 St. Paul lez Durance (France); Chung, W. [NFRI, 52 Yeoeundong Yuseonggu, Daejeon 305-333 (Korea, Republic of); Her, N.I.; Johnson, G. [ITER Organization, 13108 St. Paul lez Durance (France); Jones, L. [Fusion for Energy, C Josep Pla 2, Edificio B3, 08019 Barcelona (Spain); Jun, C.H. [ITER Organization, 13108 St. Paul lez Durance (France); Kim, B.C. [NFRI, 52 Yeoeundong Yuseonggu, Daejeon 305-333 (Korea, Republic of); Kuzmin, E. [NTC Sintez, Efremov Inst., Metallostroy, St. Petersburg 189631 (Russian Federation); Macklin, B.; Sannazzaro, G.; Shaw, R.; Wang, X.; Yu, J. [ITER Organization, 13108 St. Paul lez Durance (France)

    2009-06-15

    Recent progress of the ITER vacuum vessel (VV) and thermal shields (TS) design is presented. As the ITER construction phase approaches, the design of the VV and TS (in particular, the vacuum vessel TS-VVTS) has been improved and developed in more detail with the focus on better performance, improved manufacturing ability and successful assembly at the ITER-site. In addition to the design progress, the main principles and operations for assembly of the VV, VVTS and other TS components at the ITER-site are described.

  1. On an optimized neutron shielding for an advanced molten salt fast reactor design

    The molten salt reactor technology has gained renewed interest. In contrast to the historic molten salt reactors, the current projects are based on designing a molten salt fast reactor. Thus the shielding becomes significantly more challenging than in historic concepts. One very interesting and innovative result of the most recent EURATOM project on molten salt reactors – EVOL – is the fluid flow optimized design of the inner core vessel using curved blanket walls. The developed structure leads to a very uniform flow distribution. The design avoids all core internal structures. On the basis of this new geometry a model for neutron physics calculation is presented and applied for a shielding optimization. Based on these results an optimized shielding strategy is developed for the molten salt fast reactor to keep the fluence in the safety related outer vessel below expected limit values. A lifetime of 80 years can be assured, but the size of the core/blanket system has to be significantly increased and will finally be comparable to a sodium cooled fast reactor. The HELIOS results are verified against Monte-Carlo calculations with very satisfactory agreement for a deep penetration problem. (author)

  2. A robust approach to the design of an electromagnetic shield based on pyrolitic carbon

    Lamberti, Patrizia; Kuzhir, Polina; Tucci, Vincenzo

    2016-07-01

    A robust approach to the design of an electromagnetic shield based on ultra-thin pyrolytic carbon (PyC, 5 ÷ 110 nm) films is proposed. Finite Element Method (FEM) simulations and Monte Carlo based tolerance analysis are used to show that even a deviation of 15 ÷ 20% from the nominal values of the most important design parameters of the PyC film, i.e. its thickness and sheet resistance, does not significantly affect the wanted level of electromagnetic interference shielding efficiency (EMI SE). The ranges of the SE show that EMI shield based on PyC film is characterized by a robust behavior with respect to the variation of such parameters due to the production processes. Therefore, since the PyC can be produced on a scalable basis, is chemically inert, significantly transparent in the visible range and can be deposited onto both metal and dielectric substrates, including flexible polymers, it may be appropriate for the highly demanding technological needs associated to the graphene revolution and can be developed from laboratory to mass production applications.

  3. Design of Neutron Shield Used for Calibrating Sensitivity of Scintillating-Film Neutron Detector

    In order to calibrate the sensitivity of the scintillating-film neutron detector with the scintillator thickness great than 0.1 mm in 5SDH-2 accelerator at the Radiation Metrology Center of China Institute of Atomic Energy, a neutron shield was designed by modeling and optimizing with MCNP program. Experiment results show that this shield can attenuate the neutron fluence out of collimator to one-tenth of that in the collimator, and constrain the background signal to the level of photomultiplier's (PMT) dark current, therefore, it can make signal-to-noise of the detector great than 1:1 for the detector with thickness of scintllator above 0.1 mm. Calculation results indicate that sensitivity change resulted from collimator scattering is less than 5%. (authors)

  4. Designing shielded radio-frequency phased-array coils for magnetic resonance imaging

    Xu Wen-Long; Zhang Ju-Cheng; Li Xia; Xu Bing-Qiao; Tao Gui-Sheng

    2013-01-01

    In this paper,an approach to the design of shielded radio-frequency (RF) phased-array coils for magnetic resonance imaging (MRI) is proposed.The target field method is used to find current densities distributed on primary and shield coils.The stream function technique is used to discretize current densities and to obtain the winding patterns of the coils.The corresponding highly ill-conditioned integral equation is solved by the Tikhonov regularization with a penalty function related to the minimum curvature.To balance the simplicity and smoothness with the homogeneity of the magnetic field of the coil's winding pattern,the selection of a penalty factor is discussed in detail.

  5. Design study on water- and gas-cooled outboard shield blankets for NET

    The Karlsruhe Nuclear Research Center entered into an agreement with the Commission of the European Communities on execution of development work geared to shielding blankets for NET. The concept to be investigated concerned water-cooled shielding blankets and, as a backup solution, a variation with helium cooling. The NET standard version as of late 1985 was considered as the basis of the investigations. The results of the study prepared in cooperation with the Sulzer company, Winterthur, and relating to the outboard blanket are contained in this report, which shows that it is relatively easy to fabricate water-cooled blankets. The stresses acting on the material during operation as a result of temperature gradients and coolant pressure are low. By addition of lithium salts to the coolant a great potential of tritium generation is offered. On the other hand, helium cooling is associated with some difficulties in design and with higher expenditure in fabrication. However, these difficulties can probably be overcome. (orig.)

  6. A Combined Shielding Design for a Neutron Generator and a Linear Accelerator at Soreq NRC

    A new radiography facility is designed at Soreq NRC. The facility will hold a neutron generator that produces 1.73·109 n/s with an energy of 14 MeV and a linear accelerator that accelerates electrons to an energy of 9 MeV. The two radiation sources will be installed in 2 separate laboratories that will be built in an existing building. Each laboratory will have its own machine and control room. The dose rates around the sources were calculated using the FLUKA Monte Carlo code(1,2). The annual doses were calculated in several regions around the generator and the accelerator laboratories in accordance with the occupancy in each area. The calculated annual doses were compared with the dose limits specified in the Safety at Work Regulations(3) and the IAEC Standard for Protection against Ionizing Radiation. The shielding was designed to comply with the following dose constraints: 0.3 mSv/y for members of the public and 2 mSv/y for radiation workers. Each radiation source is planned to produce radiation for a maximum of 500 hours per year. The dose rate in the direct beam of the accelerator is 30 Gy/min at 1 m from the source and it will be surrounded by a collimator with an opening of 30N-tilde horizontally and 2 mm vertically, 3 m from the radiation source. The leakage radiation dose will not be greater than 1.5 mGy/min (0.005% of the direct beam, according to the manufacturer). The leakage radiation will be produced isotropically. The neutron generator will be surrounded by a shielding made of a 10 cm iron cylinder (density 7.87 g/cm3), surrounded by 50 cm of borated polyethylene (atomic percent: H (13.8%), C (82.2%), B (4%), density: 0.92 g/cm3) and 5 cm of lead (density 11.35 g/cm3). The neutron generator shielding was not designed or required in the present shielding design but was considered in the shielding calculations

  7. Mechanical design and analysis for a EPR first wall/blanket/shield system

    Continuing studies are in progress at ANL to expand upon the design of a first wall/blanket/shield FW/B/S system and power conversion for a tokamak type Experimental Power Reactor (EPR). The FW/B/S system has evolved from an earlier design for a low beta, circular cross section plasma (major radius = 6 m) to one for a higher beta elongated plasma with a 4.7 m major radius. Basic mechanical design and layout features of the old and new EPR designs showing some of the more important design developments are given. These developments are aimed at simplifying the design, reducing the costs and in addition, improving the plant thermal efficiency and overall maintainability. In the area of the reactor blanket, significant thermal hydraulic and stress analysis have been performed to substantiate the integrity of the chosen concept. This paper deals with the discussion of these improved features

  8. Mechanical design and analysis for a EPR first wall/blanket/shield system

    Stevens, H.C.; Misra, B.; Youngdahl, C.K.

    1977-01-01

    Continuing studies are in progress at ANL to expand upon the design of a first wall/blanket/shield FW/B/S system and power conversion for a tokamak type Experimental Power Reactor (EPR). The FW/B/S system has evolved from an earlier design for a low beta, circular cross section plasma (major radius = 6 m) to one for a higher beta elongated plasma with a 4.7 m major radius. Basic mechanical design and layout features of the old and new EPR designs showing some of the more important design developments are depicted. These developments are aimed at simplifying the design, reducing the costs and, in addition, improving the plant thermal efficiency and overall maintainability. In the area of the reactor blanket, significant thermal hydraulic and stress analysis have been performed to substantiate the integrity of the chosen concept. This paper deals with the discussion of these improved features.

  9. Principle of shielding design for the target cave of a high current medical cyclotron

    Full text: During routine isotope production regimen at the H-ion Medical Cyclotron of the Radiopharmaceutical Division of ANSTO a thick copper plate electroplated with enriched target material is bombarded with a 30 MeV proton beam at 250μA which results in the production of fast evaporation neutrons with an average energy of 5.1 MeV (Nakamura, T. et al, Nucl Sci Engg, 83: 444-458, 1992). This paper highlights the principle of shielding design using the empirical method (Mukherjee, B. Proc. 14th Int. Conf. on Cyclotrons and their Applications, Cape Town, South Africa, October 1995) for a new target cave proposed to house a solid target irradiation station to be bombarded with a proton beam of 500μA and thereby producing an intense flux of fast neutrons. Important nuclear data such as the neutron and gamma source terms defined as the corresponding dose equivalent rates for lμA proton beam current at 1m from the target as well as the neutron energy distribution required for the shielding calculations were previously estimated experimentally (Mukherjee, B. Proc. 14th Int. Conf. on Cyclotrons and their Applications, Cape Town, South Africa, October 1995). The neutron attenuation coefficients of the 4 legged maze were explicitly evaluated from experiments conducted in the existing cyclotron vault and beam room (Mukherjee, B. et al, Appl Radiat Isot, in print, June 1996). Low sodium content high density (2350 kg.m-3) concrete was used as the shielding material. The optimum lateral thicknesses of the shielding walls (t1 and t2) were calculated to be 2.1 m and 2.5 m and the length of the maze legs (ab, bc, cd, de) were evaluated as 1.2 m, 3.1 m, 1.5 m and 6 m respectively. The dose equivalent (gamma plus neutron) rates at the locations of interest 'P1', 'P2' and 'e' were set at 0.5 μSvh-1, 10 pSvh-1 and 10 μSvh-1 respectively. The neutron shielding calculation method presented in this paper was found to be more suitable than the Monte Carlo code generally used for

  10. Design, construction and application of a neutron shield for the treatment of diffuse lung metastases in rats using BNCT

    A model of multiple lung metastases in BDIX rats is under study at CNEA (Argentina) to evaluate the feasibility of BNCT for multiple, non-surgically resectable lung metastases. A practical shielding device that comfortably houses a rat, allowing delivery of a therapeutic, uniform dose in lungs while protecting the body from the neutron beam is presented. Based on the final design obtained by numerical simulations, the shield was constructed, experimentally characterized and recently used in the first in vivo experiment at RA-3. - Highlights: • A practical shielding device that comfortably houses a rat is presented. • The shield allows a uniform and useful dose in the rat thoracic area. • A novel computational dosimetry in animals based on Multicell is presented. • Experimental characterization evidences the good performance of the shield. • An irradiation based on a diffuse lung metastases model in rats was performed

  11. Empirical shielding design data for facilities administering I131 for thyroid carcinoma

    Retrospective review of the records for 434 post thyroidectomy patients receiving I131 therapy for thyroid carcinoma revealed approximately 75% of the patients were discharged within 48 hours and 90% within 72 hours. Criterion for discharge was an external radiation dose below 25 μSv/hr, measured at one metre anterior to the patient's neck. The time-averaged average dose rate at one metre anterior to the neck of a typical patient during the isolation period was 72 μSv/hr, with 90% of the patients below 82 μSv/hr. After correcting for the effects of patient size and scatter, the effective design dose rate from a patient in an isolation room treating two or three patients/week is 105 μSv.m2.hr-1, or 75 μSv.m2.hr-1 where only one patient is treated each week. Concrete is the most economical shielding material, with 190 mm filled concrete block walls and 150+mm concrete floors as the minimum recommended shielding for a radioiodine therapy suite. Additional shielding will be required if the suite adjoins (including areas immediately above and below) areas with a high occupancy factor. Copyright (1998) Australasian Physical and Engineering Sciences in Medicine

  12. Optimal filter design for shielded and unshielded ambient noise reduction in fetal magnetocardiography

    Comani, S [Department of Clinical Sciences and Bio-imaging, Chieti University (Italy); Mantini, D [Department of Informatics and Automation Engineering, Marche Polytechnic University, Ancona (Italy); Alleva, G [ITAB-Institute of Advanced Biomedical Technologies, University Foundation ' G. D' Annunzio' , Chieti University (Italy); Luzio, S Di [Department of Clinical Sciences and Bio-imaging, Chieti University (Italy); Romani, G L [Department of Clinical Sciences and Bio-imaging, Chieti University (Italy)

    2005-12-07

    The greatest impediment to extracting high-quality fetal signals from fetal magnetocardiography (fMCG) is environmental magnetic noise, which may have peak-to-peak intensity comparable to fetal QRS amplitude. Being an unstructured Gaussian signal with large disturbances at specific frequencies, ambient field noise can be reduced with hardware-based approaches and/or with software algorithms that digitally filter magnetocardiographic recordings. At present, no systematic evaluation of filters' performances on shielded and unshielded fMCG is available. We designed high-pass and low-pass Chebychev II-type filters with zero-phase and stable impulse response; the most commonly used band-pass filters were implemented combining high-pass and low-pass filters. The achieved ambient noise reduction in shielded and unshielded recordings was quantified, and the corresponding signal-to-noise ratio (SNR) and signal-to-distortion ratio (SDR) of the retrieved fetal signals was evaluated. The study regarded 66 fMCG datasets at different gestational ages (22-37 weeks). Since the spectral structures of shielded and unshielded magnetic noise were very similar, we concluded that the same filter setting might be applied to both conditions. Band-pass filters (1.0-100 Hz) and (2.0-100 Hz) provided the best combinations of fetal signal detection rates, SNR and SDR; however, the former should be preferred in the case of arrhythmic fetuses, which might present spectral components below 2 Hz.

  13. Methods for U.S. shielding calculations: applications to FFTF and CRBR designs

    The primary components of the U.S. reactor shielding methodology consist of: (1) computer code systems based on discrete ordinates or Monte Carlo radiation transport calculational methods; (2) a data base of neutron and gamma-ray interaction and gamma-ray-production cross sections used as input in the codes; (3) a capability for processing the cross sections into multigroup or point energy formats as required by the codes; (4) large-scale integral shielding experiments designed to test cross-section data or techniques utilized in the calculations; and (5) a ''sensitivity'' analysis capability that can identify the most important interactions in a transport calculation and assign uncertainties to the calculated result that are based on uncertainties in all of the input data. The required accuracy for the methodology is to within 5 to 10% for responses at locations near the core to within a factor of 2 for responses at distant locations. Under these criteria, the methodology has proved to be adequate for in-vessel LMFBR calculations of neutron transport through deep sodium and thick iron and stainless steel shields, of neutron streaming through lower axial coolant channels and primary pipe chaseways, and of the effects of fuel stored within the reactor vessel. For ex-vessel LMFBR problems, the methodology requires considerable improvement, the areas of concern including neutron streaming through heating and ventilation ducts, through the cavity surrounding the reactor vessel, and through gaps around rotating plugs in the reactor heat, as well as gamma-ray streaming through plant shield penetrations

  14. Design of piezoelectric transducer layer with electromagnetic shielding and high connection reliability

    Qiu, Lei; Yuan, Shenfang; Shi, Xiaoling; Huang, Tianxiang

    2012-07-01

    Piezoelectric transducer (PZT) and Lamb wave based structural health monitoring (SHM) method have been widely studied for on-line SHM of high-performance structures. To monitor large-scale structures, a dense PZTs array is required. In order to improve the placement efficiency and reduce the wire burden of the PZTs array, the concept of the piezoelectric transducers layer (PSL) was proposed. The PSL consists of PZTs, a flexible interlayer with printed wires and signal input/output interface. For on-line SHM on real aircraft structures, there are two main issues on electromagnetic interference and connection reliability of the PSL. To address the issues, an electromagnetic shielding design method of the PSL to reduce spatial electromagnetic noise and crosstalk is proposed and a combined welding-cementation process based connection reliability design method is proposed to enhance the connection reliability between the PZTs and the flexible interlayer. Two experiments on electromagnetic interference suppression are performed to validate the shielding design of the PSL. The experimental results show that the amplitudes of the spatial electromagnetic noise and crosstalk output from the shielded PSL developed by this paper are - 15 dB and - 25 dB lower than those of the ordinary PSL, respectively. Other two experiments on temperature durability ( - 55 °C-80 °C ) and strength durability (160-1600μɛ, one million load cycles) are applied to the PSL to validate the connection reliability. The low repeatability errors (less than 3% and less than 5%, respectively) indicate that the developed PSL is of high connection reliability and long fatigue life.

  15. Design of a Laboratory Hall Thruster with Magnetically Shielded Channel Walls, Phase I: Numerical Simulations

    Mikellides, Ioannis G.; Katz, Ira; Hofer, Richard R.

    2011-01-01

    In a proof-of-principle effort to demonstrate the feasibility of magnetically shielded (MS) Hall thrusters, an existing laboratory thruster has been modified with the guidance of physics-based numerical simulation. When operated at a discharge power of 6-kilowatts the modified thruster has been designed to reduce the total energy and flux of ions to the channel insulators by greater than 1 and greater than 3 orders of magnitude, respectively. The erosion rates in this MS thruster configuration are predicted to be at least 2-4 orders of magnitude lower than those in the baseline (BL) configuration. At such rates no detectable erosion is expected to occur.

  16. Design of the precast, post-tensioned concrete shielding structure for the TFTR neutral beam test cell

    At the TFTR facility, the Neutral Beam Test Cell is a room separated from the TFTR Cell by a 4-foot-thick concrete wall and devoted to testing the neutral beam injector. The function of the shielding structure is to protect personnel from radiation casued by pulsing the injector. The distance from the TFTR device to the injector is large enough to permit use of magnetic materials in the shielding structure, and the neutron flux levels are small enough so that ordinary concrete of moderate thickness may be employed. Radiation considerations are not discussed in this paper, which is devoted to a description of the structural design of the shield

  17. Design verification of large time constant thermal shields for optical reference cavities.

    Zhang, J; Wu, W; Shi, X H; Zeng, X Y; Deng, K; Lu, Z H

    2016-02-01

    In order to achieve high frequency stability in ultra-stable lasers, the Fabry-Pérot reference cavities shall be put inside vacuum chambers with large thermal time constants to reduce the sensitivity to external temperature fluctuations. Currently, the determination of thermal time constants of vacuum chambers is based either on theoretical calculation or time-consuming experiments. The first method can only apply to simple system, while the second method will take a lot of time to try out different designs. To overcome these limitations, we present thermal time constant simulation using finite element analysis (FEA) based on complete vacuum chamber models and verify the results with measured time constants. We measure the thermal time constants using ultrastable laser systems and a frequency comb. The thermal expansion coefficients of optical reference cavities are precisely measured to reduce the measurement error of time constants. The simulation results and the experimental results agree very well. With this knowledge, we simulate several simplified design models using FEA to obtain larger vacuum thermal time constants at room temperature, taking into account vacuum pressure, shielding layers, and support structure. We adopt the Taguchi method for shielding layer optimization and demonstrate that layer material and layer number dominate the contributions to the thermal time constant, compared with layer thickness and layer spacing. PMID:26931831

  18. Advancing Control for Shield Tunneling Machine by Backstepping Design with LuGre Friction Model

    Haibo Xie

    2014-01-01

    Full Text Available Shield tunneling machine is widely applied for underground tunnel construction. The shield machine is a complex machine with large momentum and ultralow advancing speed. The working condition underground is rather complicated and unpredictable, and brings big trouble in controlling the advancing speed. This paper focused on the advancing motion control on desired tunnel axis. A three-state dynamic model was established with considering unknown front face earth pressure force and unknown friction force. LuGre friction model was introduced to describe the friction force. Backstepping design was then proposed to make tracking error converge to zero. To have a comparison study, controller without LuGre model was designed. Tracking simulations of speed regulations and simulations when front face earth pressure changed were carried out to show the transient performances of the proposed controller. The results indicated that the controller had good tracking performance even under changing geological conditions. Experiments of speed regulations were carried out to have validations of the controllers.

  19. Design and performance of new type carbon fiber reinforced polyimide-based composites for X/γ photon shielding

    Background: With the rapid development of radiation technology, demands of functional and structural integration have been put forward for the photon shielding material. Purpose: To meet this need, a new type of carbon fiber reinforced polyimide composite has been designed and tested. Methods: Shielding properties of composite materials of different PbO contents are modeled based on MCNP. According to the simulation results, shielding material is designed and prepared. And its shielding properties, mechanical properties as well as radiation-resistant properties are tested. Results: Through photon shield experiment and mechanical performance experiment, the composite material has good shielding performance for photons. Its photon transmission rate at thickness of 4.80-mm is 54.13% for 137Cs (662 keV) gamma-ray, bend strength and stretch strength at l.2-mm thickness can reach 263 MPa and 369 MPa, respectively. After 90-kGy irradiation, the stretch strength can retain 83.47% of its performance. Conclusion: Therefore, the material possesses great application potential in medicine and industry such as gamma ray flaw detection. (authors)

  20. Multiphysics Engineering Analysis for an Integrated Design of ITER Diagnostic First Wall and Diagnostic Shield Module Design

    Zhai, Y. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Loesser, G. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Smith, M. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Udintsev, V. [ITER Org, F-13115 St Paul Les Durance, France.; Giacomin, T., T. [ITER Org, F-13115 St Paul Les Durance, France.; Khodak, A. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Johnson, D, [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Feder, R, [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

    2015-07-01

    ITER diagnostic first walls (DFWs) and diagnostic shield modules (DSMs) inside the port plugs (PPs) are designed to protect diagnostic instrument and components from a harsh plasma environment and provide structural support while allowing for diagnostic access to the plasma. The design of DFWs and DSMs are driven by 1) plasma radiation and nuclear heating during normal operation 2) electromagnetic loads during plasma events and associate component structural responses. A multi-physics engineering analysis protocol for the design has been established at Princeton Plasma Physics Laboratory and it was used for the design of ITER DFWs and DSMs. The analyses were performed to address challenging design issues based on resultant stresses and deflections of the DFW-DSM-PP assembly for the main load cases. ITER Structural Design Criteria for In-Vessel Components (SDC-IC) required for design by analysis and three major issues driving the mechanical design of ITER DFWs are discussed. The general guidelines for the DSM design have been established as a result of design parametric studies.

  1. Design of a magnetic shielding system for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak.

    Cui, Z Q; Chen, Z J; Xie, X F; Peng, X Y; Hu, Z M; Du, T F; Ge, L J; Zhang, X; Yuan, X; Xia, Z W; Hu, L Q; Zhong, G Q; Lin, S Y; Wan, B N; Fan, T S; Chen, J X; Li, X Q; Zhang, G H

    2014-11-01

    The novel neutron spectrometer TOFED (Time of Flight Enhanced Diagnostics), comprising 90 individual photomultiplier tubes coupled with 85 plastic scintillation detectors through light guides, has been constructed and installed at Experimental Advanced Superconducting Tokamak. A dedicated magnetic shielding system has been constructed for TOFED, and is designed to guarantee the normal operation of photomultiplier tubes in the stray magnetic field leaking from the tokamak device. Experimental measurements and numerical simulations carried out employing the finite element method are combined to optimize the design of the magnetic shielding system. The system allows detectors to work properly in an external magnetic field of 200 G. PMID:25430242

  2. Design of a magnetic shielding system for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak

    The novel neutron spectrometer TOFED (Time of Flight Enhanced Diagnostics), comprising 90 individual photomultiplier tubes coupled with 85 plastic scintillation detectors through light guides, has been constructed and installed at Experimental Advanced Superconducting Tokamak. A dedicated magnetic shielding system has been constructed for TOFED, and is designed to guarantee the normal operation of photomultiplier tubes in the stray magnetic field leaking from the tokamak device. Experimental measurements and numerical simulations carried out employing the finite element method are combined to optimize the design of the magnetic shielding system. The system allows detectors to work properly in an external magnetic field of 200 G

  3. Design of a magnetic shielding system for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak

    Cui, Z. Q.; Chen, Z. J.; Xie, X. F.; Peng, X. Y.; Hu, Z. M.; Du, T. F.; Ge, L. J.; Zhang, X.; Yuan, X.; Xia, Z. W.; Hu, L. Q.; Zhong, G. Q.; Lin, S. Y.; Wan, B. N.; Fan, T. S.; Chen, J. X.; Li, X. Q.; Zhang, G. H.

    2014-11-01

    The novel neutron spectrometer TOFED (Time of Flight Enhanced Diagnostics), comprising 90 individual photomultiplier tubes coupled with 85 plastic scintillation detectors through light guides, has been constructed and installed at Experimental Advanced Superconducting Tokamak. A dedicated magnetic shielding system has been constructed for TOFED, and is designed to guarantee the normal operation of photomultiplier tubes in the stray magnetic field leaking from the tokamak device. Experimental measurements and numerical simulations carried out employing the finite element method are combined to optimize the design of the magnetic shielding system. The system allows detectors to work properly in an external magnetic field of 200 G.

  4. Design and Simulation of Concrete Reinforced with Fiber as a Shield to Gamma and Neutron Radiations

    Marzieh Salimi

    2013-10-01

    Full Text Available In this paper, the goal is to design and simulate some concrete samples reinforced with Fiber as well as containing Bohr in order to be utilized as a shield to nuclear radiations (Gamma & Neutron.by presenting experimental composing projects for heavy and ordinary concrete containing Bohr ,heavy and ordinary concrete reinforced with Fiber , proper ratios for stuff were computed. Then using nuclear computations with computational code of MCNPX for Neutron and Gamma`s radiant spectra, the required Bohr, polypropylene Fiber, and steel percentages were obtained in order to be used in plain, Fiber, and heavy concrete material. In the next stage, ten concrete samples were analyzed in final experiments. Finally, reinforced concrete material weakening coefficients using MCNPX code were computed and the level of Neutron and Gamma radiating for every composing project was separately attained.

  5. Design of a control system for self-shielded irradiators with remote access capability

    With self-shielded irradiators like Gamma chambers, and Blood irradiators are being sold by BRIT to customers both within and outside the country, it has become necessary to improve the quality of service without increasing the overheads. The recent advances in the field of communications and information technology can be exploited for improving the quality of service to the customers. A state of the art control system with remote accessibility has been designed for these irradiators enhancing their performance. This will provide an easy access to these units wherever they might be located, through the Internet. With this technology it will now be possible to attend to the needs of the customers, as regards fault rectification, error debugging, system software update, performance testing, data acquisition etc. This will not only reduce the downtime of these irradiators but also reduce the overheads. (author)

  6. Design and verification of the shielding around the new Neutron Standards Laboratory (LPN) at CIEMAT

    The construction of the new Neutron Standards Laboratory at CIEMAT (Laboratorio de Patrones Neutronicos) has been finalised and is ready to provide service. The facility is an ∼8 m x 8 m x 8 m irradiation vault, following the International Organization for Standardization 8529 recommendations. It relies on several neutron sources: a 5-GBq (5.8 x 108 s-1) 252Cf source and two 241Am-Be neutron sources (185 and 11.1 GBq). The irradiation point is located 4 m over the ground level and in the geometrical centre of the room. Each neutron source can be moved remotely from its storage position inside a water pool to the irradiation point. Prior to this, an important task to design the neutron shielding and to choose the most appropriate materials has been developed by the Radiological Security Unit and the Ionizing Radiations Metrology Laboratory. MCNPX was chosen to simulate the irradiation facility. With this information the walls were built with a thickness of 125 cm. Special attention was put on the weak points (main door, air conditioning system, etc.) so that the ambient dose outside the facility was below the regulatory limits. Finally, the Radiation Protection Unit carried out a set of measurements in specific points around the installation with an LB6411 neutron monitor and a Reuter-Stokes high-pressure ion chamber to verify experimentally the results of the simulation. Several things have to be taken into consideration in order to analyse the obtained results: First of all, the neutron and gamma background is not taken into account in the simulation. The gamma background at CIEMAT is conservatively established to be 0.2 μSv h-1 and all the measured points around the installation are below this number and so it can be deduced that the gamma background is being measured in these places. A similar argument may be applied to neutron doses around external shielding and inside the irradiation room with the slab closed. Other problem comes from a few measured points

  7. Mars Radiation Risk Assessment and Shielding Design for Long-term Exposure to Ionizing Space Radiation

    Tripathi, Ram K.; Nealy, John E.

    2007-01-01

    NASA is now focused on the agency's vision for space exploration encompassing a broad range of human and robotic missions including missions to Moon, Mars and beyond. As a result, there is a focus on long duration space missions. NASA is committed to the safety of the missions and the crew, and there is an overwhelming emphasis on the reliability issues for space missions and the habitat. The cost-effective design of the spacecraft demands a very stringent requirement on the optimization process. Exposure from the hazards of severe space radiation in deep space and/or long duration missions is a critical design constraint and a potential 'show stopper'. Thus, protection from the hazards of severe space radiation is of paramount importance to the agency's vision. It is envisioned to have long duration human presence on the Moon for deep space exploration. The exposures from ionizing radiation - galactic cosmic radiation and solar particle events - and optimized shield design for a swing-by and a long duration Mars mission have been investigated. It is found that the technology of today is inadequate for safe human missions to Mars, and revolutionary technologies need to be developed for long duration and/or deep space missions. The study will provide a guideline for radiation exposure and protection for long duration missions and career astronauts and their safety.

  8. Implementation and display of Computer Aided Design (CAD) models in Monte Carlo radiation transport and shielding applications

    An Xwindow application capable of importing geometric information directly from two Computer Aided Design (CAD) based formats for use in radiation transport and shielding analyses is being developed at ORNL. The application permits the user to graphically view the geometric models imported from the two formats for verification and debugging. Previous models, specifically formatted for the radiation transport and shielding codes can also be imported. Required extensions to the existing combinatorial geometry analysis routines are discussed. Examples illustrating the various options and features which will be implemented in the application are presented. The use of the application as a visualization tool for the output of the radiation transport codes is also discussed

  9. Shielding Design of Cold Neutron Triple-axis Spectrometer using MCNP6

    Ryu, J. M.; Hong, K. P.; Park, J. M. Sungil; Choi, Y. H.; Lee, K. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The segmented shield was suggested while adding more radiation shield to the top cover of the monochromator chamber. To investigate the radiation, we performed MCNP6 simulations of a few different configurations of the Cold-TAS shield and obtained neutron and photon flux at 5 points that depend on reducing the height of the segmented shield and locating lead from the bottom of the top cover made of polyethylene. Cold neutron three-axis spectrometer was collapsed by weight of segmented shield. For normal condition of instrument, the 35% height of segmented shield was reduced and lead at over the monochromator was located 10cm bottom over the shield by calculating MCNP6. Photon flux increased 18% and 25% at P4 and P5 than original geometry, but absolute height was lower. Moreover, increased flux didn't reach to detector which is located 300 cm from center of C-TAS. In this paper, we got the best case of shield geometry using MCNP6.

  10. Upgrading the Neutron Radiography Facility in South Africa (SANRAD): Concrete Shielding Design Characteristics

    de Beer, F. C.; Radebe, M. J.; Schillinger, B.; Nshimirimana, R.; Ramushu, M. A.; Modise, T.

    A common denominator of all neutron radiography (NRAD) facilities worldwide is that the perimeter of the experimental chamber of the facility is a radiation shielding structure which,in some cases, also includes flight tube and filter chamber structures. These chambers are normally both located on the beam port floor outside the biological shielding of the neutron source. The main function of the NRAD-shielding structure isto maintain a radiological safe working environment in the entire beam hall according to standards set by individual national radiological safety regulations. In addition, the shielding's integrity and capability should not allow, during NRAD operations, an increase in radiation levels in the beam port hall and thus negatively affectadjacent scientific facilities (e.g. neutron diffraction facilities).As a bonus, the shielding for the NRAD facility should also prevent radiation scattering towards the detector plane and doing so, thus increase thecapability of obtaining better quantitative results. This paper addresses Monte Carlo neutron-particletransport simulations to theoretically optimize the shielding capabilities of the biological barrierfor the SANRAD facility at the SAFARI-1 nuclear research reactor in South Africa. The experimental process to develop the shielding, based on the principles of the ANTARES facility, is described. After casting, the homogeneity distribution of these concrete mix materials is found to be near perfect and first order experimental radiation shielding characteristicsthrough film badge (TLD) exposure show acceptable values and trends in neutron- and gamma-ray attenuation.

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

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

  12. Re-evaluation of structural shielding designs of X-ray and CO-60 gamma-ray scanners at the Port of Tema, Ghana

    This research work was conducted to re-evaluate the shielding designs of the 6 MeV x-ray and the 1.253 MeV Co-60 gamma ray scanners used for cargo-containerized scanning at the Port of Tema. These scanners utilize ionizing radiation, therefore adequate shielding must be provided to reduce the radiation exposure of persons in and around the facilities to acceptable levels. The purpose of radiation shielding is to protect workers and the general public from the harmful effects of ionizing radiation. Investigations on the facilities indicated that after commissioning, no work had been carried out to re-evaluate the shielding designs. However, workloads have increased over time neccessitating review of the installed shielding. There has been introduction of scanner units with higher radiation energy (as in the case of the x-ray scanner) posibily increasing dose rates at various location requiring review of the shielding. New structures have been dotted around the facilities without particular attention to their distances and locations with respect to the radiation source. Measurements of distances from the source axes to the points of concern for primary and leakage barrier shielding; source to container and container to the points of concern for scattered radiation shielding were taken. The primary and secondary thicknesses required for both scanners were determined based on current operational parameters and compared with the thickness constituted during the construction of the facilities. Calculated and measured dose rate beyond the shielding barriers were used to established the adequacy or otherwise of the shielding employed by the shielding designers. Values obtained fell below the 20 µSv/hr specified by NCRP 151 (2005) which showed that the primary and secondary shields of both facilities were adequate requiring no additional shielding. (author)

  13. Criticality safety and shielding design issues related to transport cask design

    This paper reports that the high enrichments and burnups of fuel assemblies currently being irradiated pose new problems for transport cask design. Criticality control may be assured with fixed absorbers, burnup credit, or moderator control. Fixed neutron absorbers can extend the fresh fuel enrichment limit of high density PWR fuel baskets to 2.0-2.4 w/o, but lower capacity flux trap designs are required for higher enrichments. Burnup credit can extend the irradiated enrichment limit to 5.0 w/o, but burnup credit has not yet been applied to the transport of LWR fuel. Moderator control may be implemented through moderator exclusion or by the displacement of moderator. Moderator exclusion permits LWR enrichments over 5.0 w/o but may not be licensable. Moderator displacement by inserting solid rods into PWR fuel assemblies is currently licensed. The use of borated stainless steel rods in PWR guide tubes is a practical means of increasing enrichment limits. A combination of these methods may be employed to insure subcriticality for a variety of cask designs

  14. Design of a management information system for the Shielding Experimental Reactor ageing management

    The problem of nuclear reactor ageing is a topic of increasing importance in nuclear safety recent years. Ageing management is usually implemented for reactors maintenance. In the practice, a large number of data and records need to be processed. However, there are few professional software applications that aid reactor ageing management, especially for research reactors. This paper introduces the design of a new web-based management information system (MIS), named the Shielding Experimental Reactor Ageing Management Information System (SERAMIS). It is an auxiliary means that helps to collect data, keep records, and retrieve information for a research reactor ageing management. The Java2 Enterprise Edition (J2EE) and network database techniques, such as three-tiered model, Model-View-Controller architecture, transaction-oriented operations, and JavaScript techniques, are used in the development of this system. The functionalities of the application cover periodic safety review (PSR), regulatory references, data inspection, and SSCs classification according to ageing management methodology. Data and examples are presented to demonstrate the functionalities. For future work, techniques of data mining will be employed to support decision-making.

  15. Design of a management information system for the Shielding Experimental Reactor ageing management

    He Jie, E-mail: hejiejoe@163.co [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Xu Xianhong [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)

    2010-01-15

    The problem of nuclear reactor ageing is a topic of increasing importance in nuclear safety recent years. Ageing management is usually implemented for reactors maintenance. In the practice, a large number of data and records need to be processed. However, there are few professional software applications that aid reactor ageing management, especially for research reactors. This paper introduces the design of a new web-based management information system (MIS), named the Shielding Experimental Reactor Ageing Management Information System (SERAMIS). It is an auxiliary means that helps to collect data, keep records, and retrieve information for a research reactor ageing management. The Java2 Enterprise Edition (J2EE) and network database techniques, such as three-tiered model, Model-View-Controller architecture, transaction-oriented operations, and JavaScript techniques, are used in the development of this system. The functionalities of the application cover periodic safety review (PSR), regulatory references, data inspection, and SSCs classification according to ageing management methodology. Data and examples are presented to demonstrate the functionalities. For future work, techniques of data mining will be employed to support decision-making.

  16. Design and characterization of a high-Tc superconducting shielded-core transformer

    In this paper the principle of a high-Tc superconducting shielded-core transformer is offered capable of protecting from overvoltages and the results of experimental investigations with a laboratory model are described. (orig.)

  17. Nuclear analysis and shielding optimisation in support of the ITER In-Vessel Viewing System design

    Turner, A; Loughlin, M J; Ghani, Z; Hurst, G; Bue, A Lo; Mangham, S; Puiu, A; Zheng, S

    2014-01-01

    The In-Vessel Viewing System (IVVS) units proposed for ITER are deployed to perform in-vessel examination. During plasma operations, the IVVS is located beyond the vacuum vessel, with shielding blocks envisaged to protect components from neutron damage and reduce shutdown dose rate (SDR) levels. Analyses were conducted to determine the effectiveness of several shielding configurations. The neutron response of the system was assessed using global variance reduction techniques and a surface source, and shutdown dose rate calculations were undertaken using MCR2S. Unshielded, the absorbed dose to piezoelectric motors (PZT) was found to be below stable limits, however activation of the primary closure plate (PCP) was prohibitively high. A scenario with shielding blocks at probe level showed significantly reduced PCP contact dose rate, however still marginally exceeded port cell requirements. The addition of shielding blocks at the bioshield plug demonstrated PCP contact dose rates below project requirements. SDR l...

  18. The design of shielding material for ultra low-background gamma-ray spectrometry

    A study of gamma and neutron backgrounds has been performed, based on Monte Carlo simulations combined with radio purity data. With reference to other papers and reports, materials of radiation shielding structure was determined. Then, the thickness of each material is determined by the GEANT4 simulation. In the future the gamma and neutron background radiation will be measured with shielded detector and non-shielded detector to calculate the actual TR. It will be compared with the results of the GEANT4 simulation. Also it is planned an active shielding to reduce cosmic muons with an anti-coincidence of a pair of plastic scintillators and a passive shielding with nitrogen gas. Development and performance of an ultra low-background γ-ray spectrometer will be performed at Dongnam Inst. of Radiological and Medical Sciences (DIRAMS) as basic tool for various radioactivity measurements. Gamma-ray spectrometry with a HPGe detector is widely used for the identification and activity measurements of radionuclides in a sample, impurity checks of a standard source, determination of emission probabilities in radioactive decay, and low level counting's. In low-level counting's, a variety of techniques to reduce the background have been employed and makes it possible to radio assay an environmental sample containing a trace of γ-emitting radio nuclides. The application of γ-spectrometry for environmental monitoring of radioactivity requires as low detection limits as practically achievable due to the limited amount of sample provided for measurement and the relatively low concentrations. We present in this paper a study of shielding materials for ultra low-background shielding structure and a calculation of transmission rate (TR) of the shielded structure using the GEANT4 simulation code

  19. Space reactor shield technology

    The reactor shield mass contributes a large portion (10% to 25%) to the total mass of an unmanned reactor system. Different shield materials are required to attenuate neutrons and gamma rays and still obtain a minimum mass. The shield material selection should also consider structural characteristics, physical and chemical properties, fabricability and availability. Minimum mass is achieved by using a shadow shield. Neutron capture gamma ray and heat generation are extremely important considerations. Lithium hydride was selected for the neutron shield material due to its excellent properties. It has to be canned and may be compartmentalized to reduce the probability of complete shielding effectiveness loss due to meteoroid puncture of the can. The initial shield design was based on previous SNAP shield design experience. The Monte Carlo Neutron Photon code, which includes the radiation scattering with the radiator and power conversion system, was then used to ensure that the design requirements were met. Fabrication of the shield by casting techniques is recommended to maintain shield integrity during vibration and to accommodate complex penetrations. A method for casting full-scale shields is described

  20. Shield for a medical actinometer

    The shield is designed for an actinometer enabling a kidney clearance determination. It shields the radioactive radiation coming from the kidney-bladder region opposite the measuring head. The shield consists of two plates which can be pushed together so that the dimensions of the shield are variable. (DG)

  1. Shielding benchmark problems, (2)

    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)

  2. Design of a PET/CT facility considering the shielding calculation in accordance with AAPM TG-108

    A Positron Emission Tomography / Computed Tomography facility may require protection barriers on floor, ceiling and walls, because the patient becomes a radioactive source that emits photons of 0.511 MeV, after having received a radiopharmaceutical, usually F-18 fluorodeoxyglucose (F-18 FDG). This work has as objective to propose the design of a PET/CT facility, taking into account technical and radiation protection considerations applied internationally, and also develop the necessary shielding for such installation by applying as published by the American Association of Physicists in Medicine Task Group Report 108. A shielding spreadsheet in Excel program was developed with reference to the recommendations of the AAPM TG - 08, to determine the shielding required for the walls, floor and ceiling. For fixing the radiation levels in the shielding calculation has been considered the actual restrictions for the occupationally exposed personnel (100 μSv/week) as well as the people in general (20 μSv/ week). The radiopharmaceutical used as a reference for the shielding calculation was the F-18 FDG. With the assistance of an architectural plan were determined distances from potential sources of radiation in facility (uptake and image acquisition living rooms) to points of interest around them. Finally the thickness of the protective barriers in lead and concrete necessary to achieve the established radiation levels were calculated and these results were stored in a table. This paper shows that technical aspects considered in the design of the installation and environments distribution can improve work processes within the PET/CT facility, consequently resulting in a reduction of the dose levels for people in general. (author)

  3. Measurement of dose rate profile and spectra through a cylindrical duct vis-a-vis Monte Carlo simulation studies for optimisation of reactor shield design

    In the design of a nuclear reactor, penetrations are provided in the top shield to carry out some essential operations. Radiation streaming is envisaged through such penetrations. To avoid radiation streaming, complementary shielding is provided. Optimisation of complementary shielding is carried out by performing calculations using MCNP code. Uncertainties in the calculations are taken care of by incorporating a safety factor. The assumption of the safety factor, while designing the reactor shielding, has been validated by undertaking experimental measurements on a similar geometry vis-a-vis the computed values obtained using MCNP code. The results of the present work agree with the safety factor of two assumed during the shield design. The details of gamma spectral measurements carried out with high purity germanium detector to understand the pattern of the scattered spectrum are also presented

  4. A1C test

    HbA1C test; Glycated hemoglobin test; Glycosylated hemoglobin test; Hemoglobin glycosylated test; Glycohemoglobin test ... have recently eaten does not affect the A1C test, so you do not need to fast to ...

  5. Nuclear analysis and shielding optimisation in support of the ITER In-Vessel Viewing System design

    The In-Vessel Viewing System (IVVS) units proposed for ITER are deployed to perform in-vessel examination. During plasma operations, the IVVS is located beyond the vacuum vessel, with shielding blocks envisaged to protect components from neutron damage and reduce shutdown dose rate (SDR) levels. Analyses were conducted to determine the effectiveness of several shielding configurations. The neutron response of the system was assessed using global variance reduction techniques and a surface source, and shutdown dose rate calculations were undertaken using MCR2S. Unshielded, the absorbed dose to piezoelectric motors (PZT) was found to be below stable limits, however activation of the primary closure plate (PCP) was prohibitively high. A scenario with shielding blocks at probe level showed significantly reduced PCP contact dose rate, however still marginally exceeded port cell requirements. The addition of shielding blocks at the bioshield plug demonstrated PCP contact dose rates below project requirements. SDR levels in contact with the isolated IVVS cartridge were found to marginally exceed the hands-on maintenance limit. For engineering feasibility, shielding blocks at bioshield level are to be avoided, however the port cell SDR field requires further consideration. In addition, alternative low-activation steels are being considered for the IVVS cartridge

  6. Characterization of Magnetic Field Immersed Photomultipliers from Double Chooz Experiment. Design and Construction of their Magnetic Shields

    Flavour oscillations of neutrinos are a quantum-mechanical effect widely demonstrated. It is explained through interferences of their mass eigenstates, therefore, belonging to the physical area beyond the Standard Model. This work deals with the CIEMAT collaboration in the neutrino experiment Double Chooz. Such an experiment aims to measure the mixture angle θ13, one of the PMNS leptonic mixture matrix, with a un reached-before sensibility by decrease of systematic errors. For this, two identical scintillator detectors, equipped with PMT's, will be sited at different distances from two reactors located in the nuclear power plant CHOOZ B (France). The electronic neutrino flux from these reactors will be compared, explaining its deficit by flavour oscillations of these particles. The identity of both detectors will be diminished by the magnetic field effects on the PMT's response. Therefore, this study serves as for quantifying such an effects as for fitting the magnetic shields design that minimize them. Shielding measurements and final design of magnetic shields as much as the effect these ones cause in the PMT's response immersed in a monitored magnetic field are presented. (Author) 85 refs

  7. The effect of self-shielding of resonance cross sections on the performance of some promising fusion blanket designs

    The effect of self-shielding of resonance cross sections on the tritium breeding ratio was investigated for three promising fusion blanket designs with liquid lithium, lithium oxide and lithium-lead breeders. Calculations were performed using ANISN and MCNP transport codes with the ENDF/B-V based nuclear data libraries. It is found that the self-shielding effect cannot be neglected in the blanket design if the blanket is neutron leaky in the case when the blanket is thin or with lower Li-6 enrichment in Li. This may result in an underestimate of the tritium breeding ratio if the cross sections are infinitely diluted. This is due to the resonances in the structure materials in which the absorption cross sections are enhanced in the infinitely diluted case. Thus the effect of self-shielding of resonance cross sections should be considered in neutronics calculations of fusion reactors. It is shown that the MCNP results are better reproduced by those from the transport code with the infinitely diluted library. This is probably due to the weight function used to generate the library and to the number of groups considered. Thus for fusion applications it is recommanded to collapse broad group cross sections with the spectrum obtained from an accurate calculation based on many fine groups. (author)

  8. Neutron Shielding Design for 4π BaF2 Detector Facility

    HUANG; Xing; ZHANG; Qi-wei; HE; Guo-zhu; CHENG; Pin-jing; TANG; Hong-qing; ZHOU; Zu-ying

    2013-01-01

    Neutrons within energy range of 5 to 300 keV can be produced by pulsed proton beam striking thick lithium target,based on the HI-13 tandem accelerator.Neutron shielding is necessary when the Gamma-ray Total Absorption Facility(GTAF)is applied to measured(n,γ)reaction cross sections

  9. TRIPOLI-2: neutron gamma coupling - applications to shielding benchmarks and designs

    Recent additions in the on-going development of the TRIPOLI Monte Carlo code system include conversion to the ENDF/B data format and an automated coupling scheme for neutron secondary gamma-ray calculations. Two shielding calculations are presented here which feature these two new developments

  10. Design of software for calculation of shielding based on various standards radiodiagnostic calculation

    The aim of this study was to develop a software application that performs calculation shields in radiology room depending on the type of equipment. The calculation will be done by selecting the user, the method proposed in the Guide 5.11, the Report 144 and 147 and also for the methodology given by the Portuguese Health Ministry. (Author)

  11. 环境一号C星SAR天线设计与分析%Design and Analysis of HJ-1-C Satellite SAR Antenna

    郑士昆; 冀有志; 崔兆云; 方永刚; 周丽萍

    2014-01-01

    With truss deployable mesh parabolic reflector, the HJ-1-C SAR antenna has complex structure and multiple steps during the deployed processing. The design of the antenna is difficult in terms of deployed reliability and electrical performance. This paper makes intensive research on system, structure and electrical design, and the analysis of mechanical and thermal performance in the actual space conditions is also presented. The successful deploying in orbit and high image quality of the HJ-1-C satellite indicate that the mechanical, electronic, thermal and reliability design of the antenna satisfy the project requirement, and these research provides valuable experience for the design of the centralized mesh parabolic SAR antenna.%环境一号C星SAR天线采用了构架式可展开网状抛物面反射器,天线构型复杂,在轨展开步骤多,天线展开可靠性及在轨电气性能等都是设计难点。该文阐述了天线总体、结构和电气方面的设计研究,并从力学和热角度进行了实际工况的分析。环境一号C星在轨成功展开及良好的SAR成像质量表明了天线在机电热及可靠性设计方面满足型号工程的使用要求,为集中式网状抛物面SAR天线的设计研究提供了宝贵经验。

  12. Simulation Studies on the New Small Wheel Shielding of the ATLAS Experiment and Design and Construction of a Test Facility for Gaseous Detectors

    Weber, Stefan

    2016-01-01

    In this thesis two main projects are presented, both aiming at the overall goal of particle detector development. In the first part of the thesis detailed shielding studies are discussed, focused on the shielding section of the planned New Small Wheel as part of the ATLAS detector upgrade. Those studies supported the discussions within the upgrade community and decisions made on the final design of the New Small Wheel. The second part of the thesis covers the design, constructi...

  13. Biological Shielding Design Effectiveness of the Brachytherapy Unit at the Korle Bu Teaching Hospital in Ghana Using Mcnp5 Monte Carlo Code

    C.C. Arwui; E.O. Darko; P. Deatanyah; S. Wotorchi-Gordon; H. Lawluvi; Kpeglo, D. O.; G. Emi-Reynolds

    2011-01-01

    Design objectives for brachytherapy treatment facilities require sufficient shielding to reduce primary and scatter radiation to design limit in order to limit exposure to patients, staff and the general public. The primary aim of this study is to verify whether shielding of the brachytherapy unit at the Korle Bu teaching Hospital in Ghana provides adequate protection in order to assess any radiological health and safety impact and also test the suitability of other available sources. The stu...

  14. Engineering design and development of shielding door and safety shutter for transfer line-3 tunnel of Indus accelerator complex

    Indus accelerator complex houses two synchrotron radiation sources, Indus-1 and Indus-2. Electron beam from booster synchrotron is injected into lndus-2 through transfer line-3 (TL-3). In order to reduce the radiation coming from TL-3 through the door opening to Indus-1, to safe limit, a sliding shielding door has been developed. It is a close welded frame of mild steel with required thickness of shielding material in the form of interlocking lead bricks stacked inside. The door is suspended from an overhead I-beam attached to welded wall brackets, which are fixed to RCC wall using anchor fasteners. Pneumatic drive is used for moving the door. Radiation measurements carried out after the installation show substantial reduction in the radiation field at Indus-1 experimental hall. Another safety concern is the inadvertent transmission of electron beam from booster synchrotron to TL-3. For this purpose a safety shutter has been made. It comprises of a beam absorber made of high density alloy DENSIMET which can be moved in and out of electron beam path. Design, fabrication, installation and testing of sliding shielding door and the beam shutter have been described. (author)

  15. Shield evaluation and validation for design and operation of facility for treatment of legacy Intermediate Level Radioactive Liquid Waste (ILW)

    An ion exchange treatment facility has been commissioned at PRIX facility, for the treatment of legacy ILW generated at reprocessing plant, Trombay. The treatment system is based on the deployment of selective sorbents for removal of cesium and strontium from ILW. Activity concentration due to beta emitters likely to be processed is of the order of 111-1850 MBq/l. Dose rates in different areas of the facility were evaluated using shielding code and design input. Present work give details of the comparison of dose rates estimated and dose rates measured at various stages of the processing of ILW. At PRIX, the ILW treatment system comprises of shielded IX columns (two cesium and one strontium) housed in a MS cubicle the process lines inlet and outlet of IX treatment system and effluent storage tanks. The MS cubicle, prefilter and piping are housed in a process cell of 500 mm concrete shielding. Effluent storage tanks are outside processing building. Theoretical assessment of expected dose rates were carried out prior to installation of various systems in different areas of PRIX. Dose rate on IX column and MS cubicle for a maximum inventory of 3.7x107 MBq of 137Cs and its contribution in operating gallery was estimated

  16. Evaluating secondary neutron doses of a refined shielded design for a medical cyclotron using the TLD approach

    An increasing number of cyclotrons at medical centers in Taiwan have been installed to generate radiopharmaceutical products. An operating cyclotron generates immense amounts of secondary neutrons from reactions such the 18O(p, n)18F, used in the production of FDG. This intense radiation can be hazardous to public health, particularly to medical personnel. To increase the yield of 18F-FDG from 4200 GBq in 2005 to 48,600 GBq in 2011, Chung Shan Medical University Hospital (CSMUH) has prolonged irradiation time without changing the target or target current to meet requirements regarding the production 18F. The CSMUH has redesigned the CTI Radioisotope Delivery System shield. The lack of data for a possible secondary neutron doses has increased due to newly designed cyclotron rooms. This work aims to evaluate secondary neutron doses at a CTI cyclotron center using a thermoluminescent dosimeter (TLD-600). Two-dimensional neutron doses were mapped and indicated that neutron doses were high as neutrons leaked through self-shielded blocks and through the L-shaped concrete shield in vault rooms. These neutron doses varied markedly among locations close to the H218O target. The Monte Carlo simulation and minimum detectable dose are also discussed and demonstrated the reliability of using the TLD-600 approach. Findings can be adopted by medical centers to identify radioactive hot spots and develop radiation protection. - Highlights: • Neutron doses were verified using TLD approach. • Neutron doses were increased at cyclotron centers. • Revised L-shaped shield suppresses effectively the neutrons. • Neutron dose can be attenuated to 1.13×106%

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

    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

  18. Structural shielding design for a gamma ray stereotactic body radiotherapy system

    An OUR-QGD gamma ray stereotactic body radiotherapy system (body knife), made in China, is described. According to its structure and the principle of gamma radiation revolved on a focus, the energy distribution of scattered radiation in its treatment room is calculated. The structural shielding of the wall, roof, and door for a certain treatment room is calculated according to the local radiation protection law

  19. Structural shielding design for a gamma ray stereotactic body radiotherapy system.

    Xie, Xiangdong; Yang, Guoshan; Zhou, Hongmei; Qu, Decheng

    2006-09-01

    An OUR-QGD gamma ray stereotactic body radiotherapy system (body knife), made in China, is described. According to its structure and the principle of gamma radiation revolved on a focus, the energy distribution of scattered radiation in its treatment room is calculated. The structural shielding of the wall, roof, and door for a certain treatment room is calculated according to the local radiation protection law. PMID:16926472

  20. Structural shielding design for a gamma ray stereotactic body radiotherapy system

    Xie Xiangdong; Yang Guoshan; Zhou Hongmei; Qu Decheng [Institute of Radiation Medicine, National Center of Biomedical Analysis, Beijing 100850 (China)

    2006-09-15

    An OUR-QGD gamma ray stereotactic body radiotherapy system (body knife), made in China, is described. According to its structure and the principle of gamma radiation revolved on a focus, the energy distribution of scattered radiation in its treatment room is calculated. The structural shielding of the wall, roof, and door for a certain treatment room is calculated according to the local radiation protection law.

  1. Design and Test of a Blast Shield for Boeing 737 Overhead Compartment

    Xinglai Dang; Philemon C. Chan

    2006-01-01

    This work demonstrates the feasibility of using a composite blast shield for hardening an overhead bin compartment of a commercial aircraft. If a small amount of explosive escapes detection and is brought onboard and stowed in an overhead bin compartment of a passenger aircraft, the current bins provide no protection against a blast inside the compartment. A blast from the overhead bin will certainly damage the fuselage and likely lead to catastrophic inflight structural failure. The feasibil...

  2. Shielding practice

    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)

  3. UCSD High Energy X-ray Timing Experiment magnetic shield design and test results

    Rothschild, Richard E.; Pelling, Michael R.; Hink, Paul L.

    1991-01-01

    Results are reported from an effort to define a passive magnetic field concept for the High Energy X-ray Timing Experiment (HEXTE), in the interest of reducing the detector-gain variations due to 0.5-1.0-sec timescale magnetic field variations. This will allow a sensitivity of the order of 1 percent of the HEXTE background. While aperture modulation and automatic gain control will minimize effects on timescales of tens of seconds and longer, passive magnetic shielding of the photomultiplier tubes will address 1-sec timescale variations due to aperture motions.

  4. Collimator and shielding design for boron neutron capture therapy (BNCT) facility at TRIGA MARK II reactor

    The geometry of reactor core, thermal column, collimator and shielding system for BNCT application of TRIGA MARK II Reactor were simulated with MCNP5 code. Neutron particle lethargy and dose were calculated with MCNPX code. Neutron flux in a sample located at the end of collimator after normalized to measured value (Eid Mahmoud Eid Abdel Munem, 2007) at 1 MW power was 1.06 x 108 n/ cm2/ s. According to IAEA (2001) flux of 1.00 x 109 n/ cm2/ s requires three hours of treatment. Few modifications were needed to get higher flux. (Author)

  5. Electromagnetic shielding

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

  6. A1C Test

    ... to minimize the complications caused by chronically elevated glucose levels, such as progressive damage to body organs like the kidneys, eyes, cardiovascular system, and nerves. The A1c test result ...

  7. Radiation physics and shielding codes and analyses applied to design-assist and safety analyses of CANDUR and ACRTM reactors

    This paper discusses the radiation physics and shielding codes and analyses applied in the design of CANDU and ACR reactors. The focus is on the types of analyses undertaken rather than the inputs supplied to the engineering disciplines. Nevertheless, the discussion does show how these analyses contribute to the engineering design. Analyses in radiation physics and shielding can be categorized as either design-assist or safety and licensing (accident) analyses. Many of the analyses undertaken are designated 'design-assist' where the analyses are used to generate recommendations that directly influence plant design. These recommendations are directed at mitigating or reducing the radiation hazard of the nuclear power plant with engineered systems and components. Thus the analyses serve a primary safety function by ensuring the plant can be operated with acceptable radiation hazards to the workers and public. In addition to this role of design assist, radiation physics and shielding codes are also deployed in safety and licensing assessments of the consequences of radioactive releases of gaseous and liquid effluents during normal operation and gaseous effluents following accidents. In the latter category, the final consequences of accident sequences, expressed in terms of radiation dose to members of the public, and inputs to accident analysis, e.g., decay heat in fuel following a loss-of-coolant accident, are also calculated. Another role of the analyses is to demonstrate that the design of the plant satisfies the principle of ALARA (as low as reasonably achievable) radiation doses. This principle is applied throughout the design process to minimize worker and public doses. The principle of ALARA is an inherent part of all design-assist recommendations and safety and licensing assessments. The main focus of an ALARA exercise at the design stage is to minimize the radiation hazards at the source. This exploits material selection and impurity specifications and relies

  8. Shielding Effectiveness of Laminated Shields

    P. V. Y. Jayasree, V. S. S. N. S. Baba, B. P. Rao

    2008-01-01

    Shielding prevents coupling of undesired radiated electromagnetic energy into equipment otherwise susceptible to it. In view of this, some studies on shielding effectiveness of laminated shields with conductors and conductive polymers using plane-wave theory are carried out in this paper. The plane wave shielding effectiveness of new combination of these materials is evaluated as a function of frequency and thickness of material. Conductivity of the polymers, measured in previous investigatio...

  9. Research Reactor Spent Fuel Transfer/Storage Cask with Application to TRIGA Fuel - Designed Cask Shielding Independent Evaluation

    Margeanu, C.A.; Iorgulis, C. [Reactor Physics, Nuclear Fuel Performances and Nuclear Safety Department, Institute for Nuclear Research Pitesti, Campului Street, no.1, 115400 Mioveni (Romania); Ciocanescu, M. [Institute for Nuclear Research Pitesti, Campului Street, no.1, 115400 Mioveni (Romania); Prava, M. [Design Department, Institute for Nuclear Research Pitesti, Campului Str, no.1, 115400 Mioveni (Romania); Margeanu, S. [Radiation Protection Department, Institute for Nuclear Research Pitesti, Campului Street, no.1, 115400 Mioveni (Romania)

    2011-07-01

    Institute for Nuclear Research (INR) Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies (dual-core concept involves independent operation of TRIGA 14 MW Steady-State Reactor and TRIGA Annular-Core Pulsing Reactor at each end of a large pool). In May 2006, TRIGA 14 MW SSR core was fully converted to Low Enriched Uranium (LEU 20 wt% {sup 235}U) fuel, according to Reduced Enrichment for Research and Test Reactors agreements and current worldwide non-proliferation efforts. Paper presents a shielding independent evaluation applied to designed transfer/ storage cask for TRIGA INR spent fuel, a mandatory step in preparation of the documentation required for spent fuel transfer/storage cask authorisation process. Fuel elements irradiation was modelled by assuming constant power for entire residence time inside reactor core, for 14 MW reactor operation power and two different scenarios characteristic for accident calculations according to TRIGA 14 MW SSR safety report and reactor operation experience. The discharged spent LEU fuel was cooled down for 2 and 5 years, respectively. Source term assessment and spent fuel characteristic parameters estimation were done by means of ORIGEN-S burn-up code (included in Oak Ridge National Laboratory's SCALE6 package) with specific cross-sections libraries, updating data for each burn-up step. For the transfer/storage cask shielding analysis, two different cases have been considered, the main difference residing in TRIGA fuel elements loading. The radiation dose rates to the transfer/storage cask wall and in air at different distances from the cask have been estimated by means of MAVRIC/Monaco shielding 3D Monte Carlo code included in ORNL's SCALE6 package. (author)

  10. Research Reactor Spent Fuel Transfer/Storage Cask with Application to TRIGA Fuel - Designed Cask Shielding Independent Evaluation

    Institute for Nuclear Research (INR) Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies (dual-core concept involves independent operation of TRIGA 14 MW Steady-State Reactor and TRIGA Annular-Core Pulsing Reactor at each end of a large pool). In May 2006, TRIGA 14 MW SSR core was fully converted to Low Enriched Uranium (LEU 20 wt% 235U) fuel, according to Reduced Enrichment for Research and Test Reactors agreements and current worldwide non-proliferation efforts. Paper presents a shielding independent evaluation applied to designed transfer/ storage cask for TRIGA INR spent fuel, a mandatory step in preparation of the documentation required for spent fuel transfer/storage cask authorisation process. Fuel elements irradiation was modelled by assuming constant power for entire residence time inside reactor core, for 14 MW reactor operation power and two different scenarios characteristic for accident calculations according to TRIGA 14 MW SSR safety report and reactor operation experience. The discharged spent LEU fuel was cooled down for 2 and 5 years, respectively. Source term assessment and spent fuel characteristic parameters estimation were done by means of ORIGEN-S burn-up code (included in Oak Ridge National Laboratory's SCALE6 package) with specific cross-sections libraries, updating data for each burn-up step. For the transfer/storage cask shielding analysis, two different cases have been considered, the main difference residing in TRIGA fuel elements loading. The radiation dose rates to the transfer/storage cask wall and in air at different distances from the cask have been estimated by means of MAVRIC/Monaco shielding 3D Monte Carlo code included in ORNL's SCALE6 package. (author)

  11. The shielding design calculation of HWZPR using one-dimension transport method and ZPR-22 group cross section library

    The one-dimension SN method code ANISN and specific cross section library ZPR-22 have been used to perform the design calculation of dose rate distribution along the radial and axial direction of HWZPR shielding. Through multi-case calculations and optimization analysis works, a double slab cover structure is adopted. It is combined with the feasibility of structure and the possibility of boron concentration to be merged in paraffin for design case. The calculation results of axial direction: the core lattice distance is 18 cm; core radius R = 113 cm; reflector saving of radial direction is 25 cm; transfer leakage Dy = Dz = 244.6 cm. The calculation results of radial direction; the core lattice distance is 18 cm; critical water level 138.5 cm; reflector saving of axial direction is 20 cm; transfer leakage correction parameter Dy = 160 cm

  12. Calculation of neutron and gamma ray energy spectra for fusion reactor shield design: comparison with experiment

    Integral experiments that measure the transport of approx. 14 MeV D-T neutrons through laminated slabs of proposed fusion reactor shield materials have been carried out. Measured and calculated neutron and gamma ray energy spectra are compared as a function of the thickness and composition of stainless steel type 304, borated polyethylene, and Hevimet (a tungsten alloy), and as a function of detector position behind these materials. The measured data were obtained using a NE-213 liquid scintillator using pulse-shape discrimination methods to resolve neutron and gamma ray pulse height data and spectral unfolding methods to convert these data to energy spectra. The calculated data were obtained using two-dimensional discrete ordinates radiation transport methods in a complex calculational network that takes into account the energy-angle dependence of the D-T neutrons and the nonphysical anomalies of the S/sub n/ method

  13. Shields-1, A SmallSat Radiation Shielding Technology Demonstration

    Thomsen, D. Laurence, III; Kim, Wousik; Cutler, James W.

    2015-01-01

    The NASA Langley Research Center Shields CubeSat initiative is to develop a configurable platform that would allow lower cost access to Space for materials durability experiments, and to foster a pathway for both emerging and commercial-off-the-shelf (COTS) radiation shielding technologies to gain spaceflight heritage in a relevant environment. The Shields-1 will be Langleys' first CubeSat platform to carry out this mission. Radiation shielding tests on Shields-1 are planned for the expected severe radiation environment in a geotransfer orbit (GTO), where advertised commercial rideshare opportunities and CubeSat missions exist, such as Exploration Mission 1 (EM-1). To meet this objective, atomic number (Z) graded radiation shields (Zshields) have been developed. The Z-shield properties have been estimated, using the Space Environment Information System (SPENVIS) radiation shielding computational modeling, to have 30% increased shielding effectiveness of electrons, at half the thickness of a corresponding single layer of aluminum. The Shields-1 research payload will be made with the Z-graded radiation shields of varying thicknesses to create dose-depth curves to be compared with baseline materials. Additionally, Shields-1 demonstrates an engineered Z-grade radiation shielding vault protecting the systems' electronic boards. The radiation shielding materials' performances will be characterized using total ionizing dose sensors. Completion of these experiments is expected to raise the technology readiness levels (TRLs) of the tested atomic number (Z) graded materials. The most significant contribution of the Z-shields for the SmallSat community will be that it enables cost effective shielding for small satellite systems, with significant volume constraints, while increasing the operational lifetime of ionizing radiation sensitive components. These results are anticipated to increase the development of CubeSat hardware design for increased mission lifetimes, and enable

  14. THE MECHANICAL AND SHIELDING DESIGN OF A PORTABLE SPECTROMETER AND BEAM DUMP ASSEMBLY AT BNLS ACCELERATOR TEST FACILITY

    A portable assembly containing a vertical-bend dipole magnet has been designed and installed immediately down-beam of the Compton electron-laser interaction chamber on beamline 1 of the Accelerator Test Facility (ATF) at Brookhaven National Laboratory (BNL). The water-cooled magnet designed with field strength of up to 0.7 Tesla will be used as a spectrometer in the Thompson scattering and vacuum acceleration experiments, where field-dependent electron scattering, beam focusing and energy spread will be analyzed. This magnet will deflect the ATF's 60 MeV electron-beam 90o downward, as a vertical beam dump for the Compton scattering experiment. The dipole magnet assembly is portable, and can be relocated to other beamlines at the ATF or other accelerator facilities to be used as a spectrometer or a beam dump. The mechanical and shielding calculations are presented in this paper. The structural rigidity and stability of the assembly were studied. A square lead shield surrounding the assembly's Faraday Cup was designed to attenuate the radiation emerging from the 1 inch-copper beam stop. All photons produced were assumed to be sufficiently energetic to generate photoneutrons. A safety evaluation of groundwater tritium contamination due to the thermal neutron capturing by the deuterium in water was performed, using updated Monte Carlo neutron-photon coupled transport code (MCNP). High-energy neutron spallation, which is a potential source to directly generate radioactive tritium and sodium-22 in soil, was conservatively assessed in verifying personal and environmental safety

  15. Efficacy of Cosmic Ray Shields

    Rhodes, Nicholas

    2015-10-01

    This research involved testing various types of shielding with a self-constructed Berkeley style cosmic ray detector, in order to evaluate the materials of each type of shielding's effectiveness at blocking cosmic rays and the cost- and size-efficiency of the shields as well. The detector was constructed, then tested for functionality and reliability. Following confirmation, the detector was then used at three different locations to observe it altitude or atmospheric conditions had any effect on the effectiveness of certain shields. Multiple types of shielding were tested with the detector, including combinations of several shields, primarily aluminum, high-iron steel, polyethylene plastic, water, lead, and a lead-alternative radiation shield utilized in radiology. These tests regarding both the base effectiveness and the overall efficiency of shields is designed to support future space exploratory missions where the risk of exposure to possibly lethal amounts of cosmic rays for crew and the damage caused to unshielded electronics are of serious concern.

  16. The influence of the lay-out of beam diagnostic elements on the shielding design at the Tesla accelerator installation

    A number of devices are routinely used for collimation and beam diagnostics at accelerator installations. Some of them permanently cuts-off parts of the accelerated particles forming a distinct beam shape (slits and diaphragms). Others are used from time to time to give information on beam shape or current by intercepting the beam (Faraday cups, scintillators, wire grids). Due to a number of nuclear reactions resulted by the interaction of energetic particles with matter, these devices are strong sources of ionizing radiation and they must be considered in the shielding design procedure. The fast neutron equivalent dose rate is calculated for a certain lay-out of beam diagnostic elements at the 'Tesla' Accelerator Installation and their contribution to the equivalent dose is discussed. (author)

  17. Gamma dose rate calculations for conceptual design of a shield system for spent fuel interim dry storage in CNA 1

    After completing the rearrangement of the Spent Fuel Elements (SFE) into a compact arrangement in the two storage water pools, Atucha Nuclear Reactor 1 (ANR 1) will leave free position for the wet storage of the SFE discharged until December 2014. Even, in two possible scenarios, such as extending operation from 2015 or the cessation of operation after that date, it will be necessary to empty the interim storage water pools transferring the SFE to a temporary dry storage system. Because the law 25.018 'Management of Radioactive Wastes' implies for the first scenario - operation beyond 2015 - that Nucleoelectrica Argentina S.A. will still be in charge of the dry storage system and for the second - the cessation of operation after 2015 - the National Commission of Atomic Energy (CNEA) will be in charge by the National Management Program of Radioactive Wastes, the interim dry storage system of SNF is an issue of common interest which justifies go forward together. For that purpose and in accordance with the criticality and shielding calculations relevant to the project, in this paper we present the dose rate calculations for shielding conceptual design of a system for dry interim storage of the SFE of ANR 1. The specifications includes that the designed system must be suitable without modification for the SFE of the ANR 2. The results for the calculation of the photon dose rate, in touch and at one meter far, for the Transport Module and the Container of the SFE, are presented, which are required and controlled by the National Regulatory Authority (NRA) and the International Atomic Energy Agency (IAEA). It was used the SAS4 module of SCALE5.1 system and MCNP5. As a design tool for the photon shielding in order to meet current standards for allowable dose rates, a radial and axial parametric analysis were developed based on the thickness of lead of the Transport Module. The results were compared and verified between the two computing systems. Before this

  18. Transient heat flux shielding using thermal metamaterials

    Narayana, Supradeep; Savo, Salvatore; Sato, Yuki

    2013-05-01

    We have developed a heat shield based on a metamaterial engineering approach to shield a region from transient diffusive heat flow. The shield is designed with a multilayered structure to prescribe the appropriate spatial profile for heat capacity, density, and thermal conductivity of the effective medium. The heat shield was experimentally compared to other isotropic materials.

  19. Transient heat flux shielding using thermal metamaterials

    Narayana, Supradeep; Sato, Yuki

    2013-01-01

    We have developed a heat shield based on a metamaterial engineering approach to shield a region from transient diffusive heat flow. The shield is designed with a multilayered structure to prescribe the appropriate spatial profile for heat capacity, density, and thermal conductivity of the effective medium. The heat shield was experimentally compared to other isotropic materials.

  20. Alternate shield material feasibility

    The feasibility and cost/benefit of using materials other than stainless steel for in-vessel neutron shielding in large LMFBRs were investigated. Canned vibratorally compacted B4C powder shields were found to be much more economical than stainless steel (a savings of $1.1M in loop plant designs and $9.4M in pool plant designs). The helium gas pressure buildup in B4C shields placed around LMFBR in-vessel components (direct reactor heat exchangers in a loop reactor and intermediate heat exchangers in a pool reactor) would only be 0.04 atm after 40 y of reactor operation (with 80% dense powder). The irradiation-induced swelling of the B4C would only be 0.002%. No adverse reactor impact would occur if the B4C escaped from the B4C shields

  1. Alternate shield material feasibility

    Specht, E.R.; Levitt, L.B.

    1984-04-01

    The feasibility and cost/benefit of using materials other than stainless steel for in-vessel neutron shielding in large LMFBRs were investigated. Canned vibratorally compacted B/sub 4/C powder shields were found to be much more economical than stainless steel (a savings of $1.1M in loop plant designs and $9.4M in pool plant designs). The helium gas pressure buildup in B/sub 4/C shields placed around LMFBR in-vessel components (direct reactor heat exchangers in a loop reactor and intermediate heat exchangers in a pool reactor) would only be 0.04 atm after 40 y of reactor operation (with 80% dense powder). The irradiation-induced swelling of the B/sub 4/C would only be 0.002%. No adverse reactor impact would occur if the B/sub 4/C escaped from the B/sub 4/C shields.

  2. Design, Progressive Modeling, Manufacture, and Testing of Composite Shield for Turbine Engine Blade Containment

    Binienda, Wieslaw K.; Sancaktar, Erol; Roberts, Gary D. (Technical Monitor)

    2002-01-01

    An effective design methodology was established for composite jet engine containment structures. The methodology included the development of the full and reduced size prototypes, and FEA models of the containment structure, experimental and numerical examination of the modes of failure clue to turbine blade out event, identification of materials and design candidates for future industrial applications, and design and building of prototypes for testing and evaluation purposes.

  3. Shielding material

    The present invention effectively utilizes iron reinforced concrete wastes generated upon dismantling of concretes of nuclear facilities, to provide shielding material. That is, at least one of members selected from the group consisting of iron rods in iron-reinforced concretes and, regenerated aggregates regenerated from concrete wastes upon dismantling is charged in a predetermined mold. Cement pastes or cement mortars are charged therein, and solidified, cured and released from the mold. With such procedures, a block-formed shielding materials made of precast concretes can be obtained. In this case, the cements including much water of crystallization are used. Since iron reinforcing dusts and iron reinforcing dust chips are contained in the shielding materials, a great γ-ray shielding effect can be obtained. Further, since cements containing a great amount of water of crystallization are used, a great neutron shielding effect can be obtained. (I.S.)

  4. Designer's guidebook for first wall/blanket/shield assembly, maintenance, and repair

    This is the initial issue of the guidebook. Since a guidebook of this type must incorporate information concerning a wide range of subjects, much additional data will eventually be included. The guidebook will document, in summary and easily referenceable form, data, designs, design concepts, design guidelines and background information useful to the FWBS and to the Maintenance System designer. In providing guidelines for the AMR of the FWBS, the guidebook must, of necessity, include guidelines for all aspects of maintenance associated with the FWBS. These include most maintenance operations within the reactor room necessary to gain access, identify faults, and handle equipment related to FWBS maintenance. In addition, the guidelines include those required to define facility requirements for handling and repair of FWBS and related reactor components external to the reactor room. Particular emphasis is given to remote maintenance design and operations

  5. Designer's guidebook for first wall/blanket/shield assembly, maintenance, and repair

    1983-12-30

    This is the initial issue of the guidebook. Since a guidebook of this type must incorporate information concerning a wide range of subjects, much additional data will eventually be included. The guidebook will document, in summary and easily referenceable form, data, designs, design concepts, design guidelines and background information useful to the FWBS and to the Maintenance System designer. In providing guidelines for the AMR of the FWBS, the guidebook must, of necessity, include guidelines for all aspects of maintenance associated with the FWBS. These include most maintenance operations within the reactor room necessary to gain access, identify faults, and handle equipment related to FWBS maintenance. In addition, the guidelines include those required to define facility requirements for handling and repair of FWBS and related reactor components external to the reactor room. Particular emphasis is given to remote maintenance design and operations.

  6. A novel Compton camera design featuring a rear-panel shield for substantial noise reduction in gamma-ray images

    After the Japanese nuclear disaster in 2011, large amounts of radioactive isotopes were released and still remain a serious problem in Japan. Consequently, various gamma cameras are being developed to help identify radiation hotspots and ensure effective decontamination operation. The Compton camera utilizes the kinematics of Compton scattering to contract images without using a mechanical collimator, and features a wide field of view. For instance, we have developed a novel Compton camera that features a small size (13 × 14 × 15 cm3) and light weight (1.9 kg), but which also achieves high sensitivity thanks to Ce:GAGG scintillators optically coupled wiith MPPC arrays. By definition, in such a Compton camera, gamma rays are expected to scatter in the ''scatterer'' and then be fully absorbed in the ''absorber'' (in what is called a forward-scattered event). However, high energy gamma rays often interact with the detector in the opposite direction - initially scattered in the absorber and then absorbed in the scatterer - in what is called a ''back-scattered'' event. Any contamination of such back-scattered events is known to substantially degrade the quality of gamma-ray images, but determining the order of gamma-ray interaction based solely on energy deposits in the scatterer and absorber is quite difficult. For this reason, we propose a novel yet simple Compton camera design that includes a rear-panel shield (a few mm thick) consisting of W or Pb located just behind the scatterer. Since the energy of scattered gamma rays in back-scattered events is much lower than that in forward-scattered events, we can effectively discriminate and reduce back-scattered events to improve the signal-to-noise ratio in the images. This paper presents our detailed optimization of the rear-panel shield using Geant4 simulation, and describes a demonstration test using our Compton camera

  7. A novel Compton camera design featuring a rear-panel shield for substantial noise reduction in gamma-ray images

    Nishiyama, T.; Kataoka, J.; Kishimoto, A.; Fujita, T.; Iwamoto, Y.; Taya, T.; Ohsuka, S.; Nakamura, S.; Hirayanagi, M.; Sakurai, N.; Adachi, S.; Uchiyama, T.

    2014-12-01

    After the Japanese nuclear disaster in 2011, large amounts of radioactive isotopes were released and still remain a serious problem in Japan. Consequently, various gamma cameras are being developed to help identify radiation hotspots and ensure effective decontamination operation. The Compton camera utilizes the kinematics of Compton scattering to contract images without using a mechanical collimator, and features a wide field of view. For instance, we have developed a novel Compton camera that features a small size (13 × 14 × 15 cm3) and light weight (1.9 kg), but which also achieves high sensitivity thanks to Ce:GAGG scintillators optically coupled wiith MPPC arrays. By definition, in such a Compton camera, gamma rays are expected to scatter in the ``scatterer'' and then be fully absorbed in the ``absorber'' (in what is called a forward-scattered event). However, high energy gamma rays often interact with the detector in the opposite direction - initially scattered in the absorber and then absorbed in the scatterer - in what is called a ``back-scattered'' event. Any contamination of such back-scattered events is known to substantially degrade the quality of gamma-ray images, but determining the order of gamma-ray interaction based solely on energy deposits in the scatterer and absorber is quite difficult. For this reason, we propose a novel yet simple Compton camera design that includes a rear-panel shield (a few mm thick) consisting of W or Pb located just behind the scatterer. Since the energy of scattered gamma rays in back-scattered events is much lower than that in forward-scattered events, we can effectively discriminate and reduce back-scattered events to improve the signal-to-noise ratio in the images. This paper presents our detailed optimization of the rear-panel shield using Geant4 simulation, and describes a demonstration test using our Compton camera.

  8. Planar Shielded-Loop Resonators

    Tierney, Brian B.; Grbic, Anthony

    2014-01-01

    The design and analysis of planar shielded-loop resonators for use in wireless non-radiative power transfer systems is presented. The difficulties associated with coaxial shielded-loop resonators for wireless power transfer are discussed and planar alternatives are proposed. The currents along these planar structures are analyzed and first-order design equations are presented in the form of a circuit model. In addition, the planar structures are simulated and fabricated. Planar shielded-loop ...

  9. Design and testing of a magnetic shield for the Thomson scattering photomultiplier tubes in the stray fields of the ERASMUS tokamak

    Desoppere, E.; Van Oost, G.

    1983-01-01

    A multiple coaxial shield system has been designed for the photomultiplier tubes of the ERASMUS tokamak Thomson scattering diagnostic. A stray field of 75 x 10-4 T was reduced to 0.01 x 10-4 T for a field parallel to the tube axis, and to 0.03 × 10-4 T for a perpendicular field.

  10. Design and applications of an anticoincidence shielded low background gamma-ray spectrometer

    Petri, H. [Forschungszentrum Juelich GmbH (Germany). Zentralabteilung fuer Chemische Analysen

    1997-03-01

    A low background gamma-ray spectrometer has been constructed for measuring artificial and natural radioative isotopes. The design of the spectrometer, its properties and the application to the determination of natural radioactivity of dental ceramics are described. (orig.)

  11. Size Design of CdZnTe Detector Shield for Measuring Burnup of Spent Fuel

    2008-01-01

    <正>It is important to measure the burnup of spent fuel for nuclear safeguards, burnup credit and critical safety in spent-fuel reprocessing process. The purpose of this work is designing a portable device to

  12. Design and applications of an anticoincidence shielded low background gamma-ray spectrometer

    A low background gamma-ray spectrometer has been constructed for measuring artificial and natural radioative isotopes. The design of the spectrometer, its properties and the application to the determination of natural radioactivity of dental ceramics are described. (orig.)

  13. Shielding Effectiveness of Laminated Shields

    B. P. Rao

    2008-12-01

    Full Text Available Shielding prevents coupling of undesired radiated electromagnetic energy into equipment otherwise susceptible to it. In view of this, some studies on shielding effectiveness of laminated shields with conductors and conductive polymers using plane-wave theory are carried out in this paper. The plane wave shielding effectiveness of new combination of these materials is evaluated as a function of frequency and thickness of material. Conductivity of the polymers, measured in previous investigations by the cavity perturbation technique, is used to compute the overall reflection and transmission coefficients of single and multiple layers of the polymers. With recent advances in synthesizing stable highly conductive polymers these lightweight mechanically strong materials appear to be viable alternatives to metals for EM1 shielding.

  14. Design and characterization of a novel neutron shield for BNCT in an experimental model of oral cancer in the hamster cheek pouch at RA-3

    Our research group at the Radiation Pathology Division of the Department of Radiobiology (National Atomic Energy Commission) has previously demonstrated the therapeutic efficacy of different BNCT protocols to treat oral cancer in an experimental hamster cheek pouch model. In particular, to perform studies in this experimental model at the thermal facility constructed at RA-3, we designed and constructed a shielding device for thermal neutrons, to be able to expose the cheek pouch while minimizing the dose to the rest of the body. This device allowed for the irradiation of one animal at a time. Given the usage rate of the device, the aim of the present study was to design and construct an optimized version of the existing shielding device that would allow for the simultaneous irradiation of 2 animals at the thermal facility of RA-3. Taking into account the characteristics of the neutron source and preliminary biological assays, we designed the shielding device for the body of the animal, i.e. a rectangular shaped box with double acrylic walls. The space between the walls contains a continuous filling of 6Li2CO3 (95% enriched in 6Li), approximately 6 mm thick. Two small windows interrupt the shield at one end of the box through which the right pouch of each hamster is everted out onto an external acrylic shelf for exposure to the neutron flux. The characterization of the shielding device showed that the neutron flux was equivalent at both irradiation positions confirming that we were able to design and construct a new shielding device that allows for the irradiation of 2 animals at the same time at the thermal facility of RA-3. This new version of the shielding device will reduce the number of interventions of the reactor operators, reducing occupational exposure to radiation and will make the procedure more efficient for researchers. In addition, we addressed the generation of tritium as a product of the capture reaction in lithium. It was considered as a potential

  15. Conductor backed and shielded multi-layer coplanar waveguide designs on LTCC for RF carrier boards for packaging PICs

    Marraccini, Philip J.; Jezzini, Moises A.; Peters, Frank H.

    2016-05-01

    Designing photonic integrated circuits (PICs) with packaging in mind is important since this impacts the performance of the final product. In coherent optical communication applications there are a large number of DC and RF lines that need routed to connect the PIC to the outer packaging. These RF lines should be impedance matched to the devices, isolated from each other, low loss and protected against electromagnetic interference (EMI) over the frequency range of interest to achieve the performance required for the application. Multilevel low temperature co-fired ceramic (LTCC) boards can be used as a carrier board connecting the PIC to the packaging due to its good RF performance, machinability, compatibility with hermetic sealing, and ability to integrate drivers into the board. Flexibility with layer numbers enables additional layers for shielding against electromagnetic interference or increased space for routing electrical connections. In this paper the design, simulations, and measured results for a set of 4 phase matched transmission lines in LTCC that would be used with an IQ MZM are presented. The measured 3dB bandwidth for a set of four phase matched transmission lines for an IQ MZM was measured to be 19.8 GHz.

  16. Material characterisation and preliminary mechanical design for the HL-LHC shielded beam screens operating at cryogenic temperatures

    Garion, C; Koettig, T; Machiocha, W; Morrone, M

    2015-01-01

    The High Luminosity LHC project (HL-LHC) aims at increasing the luminosity (rate of collisions) in the Large Hadron Collider (LHC) experiments by a factor of 10 beyond the original design value (from 300 to 3000 fb-1). It relies on new superconducting magnets, installed close to the interaction points, equipped with new beam screen. This component has to ensure the vacuum performance together with shielding the cold mass from physics debris and screening the cold bore cryogenic system from beam induced heating. The beam screen operates in the range 40-60 K whereas the magnet cold bore temperature is 1.9 K. A tungsten-based material is used to absorb the energy of particles. In this paper, measurements of the mechanical and physical properties of such tungsten material are shown at room and cryogenic temperature. In addition, the design and the thermal mechanical behaviour of the beam screen assembly are presented also. They include the heat transfer from the tungsten absorbers to the cooling pipes and the sup...

  17. Scintillation counter, segmented shield

    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.)

  18. WE-G-17A-09: Novel Magnetic Shielding Design for Inline and Perpendicular Integrated 6 MV Linac and 1.0 T MRI Systems

    Li, X; Ma, B; Kuang, Y [University of Nevada, Las Vegas, Las Vegas, NV (United States); Diao, X [Shenzhen University, Shenzhen, Guangdong (China)

    2014-06-15

    Purpose: The influence of fringe magnetic fields delivered by magnetic resonance imaging (MRI) on the beam generation and transportation in Linac is still a major challenge for the integration of linear accelerator and MRI (Linac-MRI). In this study, we investigated an optimal magnetic shielding design for Linac-MRI and further characterized the beam trajectory in electron gun. Methods: Both inline and perpendicular configurations were analyzed in this study. The configurations, comprising a Linac-MRI with a 100cm SAD and an open 1.0 T superconductive magnet, were simulated by the 3D finite element method (FEM). The steel shielding around the Linac was included in the 3D model, the thickness of which was varied from 1mm to 20mm, and magnetic field maps were acquired with and without additional shielding. The treatment beam trajectory in electron gun was evaluated using OPERA 3d SCALA with and without shielding cases. Results: When Linac was not shielded, the uniformity of diameter sphere volume (DSV) (30cm) was about 5 parts per million (ppm) and the fringe magnetic fields in electron gun were more than 0.3 T. With shielding, the magnetic fields in electron gun were reduced to less than 0.01 T. For the inline configuration, the radial magnetic fields in the Linac were about 0.02T. A cylinder steel shield used (5mm thick) altered the uniformity of DSV to 1000 ppm. For the perpendicular configuration, the Linac transverse magnetic fields were more than 0.3T, which altered the beam trajectory significantly. A 8mm-thick cylinder steel shield surrounding the Linac was used to compensate the output losses of Linac, which shifted the magnetic fields' uniformity of DSV to 400 ppm. Conclusion: For both configurations, the Linac shielding was used to ensure normal operation of the Linac. The effect of magnetic fields on the uniformity of DSV could be modulated by the shimming technique of the MRI magnet. NIH/NIGMS grant U54 GM104944, Lincy Endowed Assistant

  19. WE-G-17A-09: Novel Magnetic Shielding Design for Inline and Perpendicular Integrated 6 MV Linac and 1.0 T MRI Systems

    Purpose: The influence of fringe magnetic fields delivered by magnetic resonance imaging (MRI) on the beam generation and transportation in Linac is still a major challenge for the integration of linear accelerator and MRI (Linac-MRI). In this study, we investigated an optimal magnetic shielding design for Linac-MRI and further characterized the beam trajectory in electron gun. Methods: Both inline and perpendicular configurations were analyzed in this study. The configurations, comprising a Linac-MRI with a 100cm SAD and an open 1.0 T superconductive magnet, were simulated by the 3D finite element method (FEM). The steel shielding around the Linac was included in the 3D model, the thickness of which was varied from 1mm to 20mm, and magnetic field maps were acquired with and without additional shielding. The treatment beam trajectory in electron gun was evaluated using OPERA 3d SCALA with and without shielding cases. Results: When Linac was not shielded, the uniformity of diameter sphere volume (DSV) (30cm) was about 5 parts per million (ppm) and the fringe magnetic fields in electron gun were more than 0.3 T. With shielding, the magnetic fields in electron gun were reduced to less than 0.01 T. For the inline configuration, the radial magnetic fields in the Linac were about 0.02T. A cylinder steel shield used (5mm thick) altered the uniformity of DSV to 1000 ppm. For the perpendicular configuration, the Linac transverse magnetic fields were more than 0.3T, which altered the beam trajectory significantly. A 8mm-thick cylinder steel shield surrounding the Linac was used to compensate the output losses of Linac, which shifted the magnetic fields' uniformity of DSV to 400 ppm. Conclusion: For both configurations, the Linac shielding was used to ensure normal operation of the Linac. The effect of magnetic fields on the uniformity of DSV could be modulated by the shimming technique of the MRI magnet. NIH/NIGMS grant U54 GM104944, Lincy Endowed Assistant

  20. Shielded syringe

    This patent specification relates to a partially disposable shielded syringe for injecting radioactive material into a patient. It is claimed that the technique overcomes the problems of non-standardisation of syringe size. (U.K.)

  1. Design and fabrication of shielding for gamma spectrometry; Diseno y fabricacion de blindaje para espectrometria gamma

    Mariano H, E

    1991-05-15

    To have a system of gamma spectrometry in the Radiological Mobile Unit No. 1 (UMOR-1) was designed and manufactured an armor-plating appropriate to this, to make analysis of radioactive samples in place in the event of a radiological emergency, besides being able to give support to the Management of Radiological Safety, and even to give service of sample analysis of other Institutions. (Author)

  2. Shielding walls against ionizing radiation

    This standard shall be applied to closed shielding facilities which, together with the lead bricks according to DIN 25 407 part 1 and the functional elements according to this standard, are designed to make possible the setting-up of complete shieldings for hot cells in beta-gamma-technique (see DIN 25 407 part 3) according to modular principles. This standard is intended to facilitate the design and construction of hot cells with shielding walls made of lead as well as the interchangeability of individual constructional elements in existing shielding walls. (orig./HP)

  3. Shielding design of electron beam stop for Dual-Axis Radiographic Hydrotest Facility (DARHT)

    An electron beam stop was designed to allow workers to be present in the experimental area while the accelerators are producing electron beam pulses. The beam stop is composed of a graphite region to stop the electron pulses and a surrounding tungsten region to attenuate photons produced by electron transport in the graphite. Radiation-transport dose calculations were performed to set the dimensions of the graphite and tungsten regions. To reduce calculational effort, electron transport in the graphite was calculated separately from photon dose transport to worker locations. The source for photon dose transport was generated by tallying photons emerging from the graphite during electron transport

  4. Design of the Fifth-Generation Target-Moderator-Reflector-Shield Assembly

    Nowicki, Suzanne Florence [Los Alamos National Lab. (LANL), Los Alamos, NM (United Sta; Mocko, Michal [Los Alamos National Lab. (LANL), Los Alamos, NM (United Sta

    2015-11-16

    The facilities at the Los Alamos Neutron Science Center are described first. The target is being redesigned so that the Flight Paths (FP) in the upper tier provide a higher intensity in the epithermal and medium energy range. It is found that a 3-piece design looks promising: intensity in epithermal and medium energy range in upper tier is an order of magnitude higher than current Mark III, and intensity in the thermal energy range is higher in the lower tier than current Mark III. Time emission spectra show a bump due to the scattering of fast neutrons. Other investigations such as the addition of wings around the upper target will be conducted.

  5. Shield For Flexible Pipe

    Ponton, Michael K.; Williford, Clifford B.; Lagen, Nicholas T.

    1995-01-01

    Cylindrical shield designed to fit around flexible pipe to protect nearby workers from injury and equipment from damage if pipe ruptures. Designed as pressure-relief device. Absorbs impact of debris ejected radially from broken flexible pipe. Also redirects flow of pressurized fluid escaping from broken pipe onto flow path allowing for relief of pressure while minimizing potential for harm.

  6. Design and Testing of a Prototype Pixellated CZT Detector and Shield for Hard X-Ray Astronomy

    Bloser, P F; Narita, T; Jenkins, J A

    1999-01-01

    We report on the design and laboratory testing of a prototype imaging CZT detector intended for balloon flight testing in April 2000. The detector tests several key techniques needed for the construction of large-area CZT arrays, as required for proposed hard X-ray astronomy missions. Two 10 mm x 10 mm x 5 mm CZT detectors, each with a 4 x 4 array of 1.9 mm pixels on a 2.5 mm pitch, will be mounted in a ``flip-chip'' fashion on a printed circuit board carrier card; the detectors will be placed 0.3 mm apart in a tiled configuration such that the pixel pitch is preserved across both crystals. One detector is eV Products high-pressure Bridgman CZT, and the other is IMARAD horizontal Bridgman material. Both detectors are read out by a 32-channel VA-TA ASIC controlled by a PC/104 single-board computer. A passive shield/collimator surrounded by plastic scintillator surrounds the detectors on five sides and provides a ~45 deg field of view. The background spectrum recorded by this instrument will be compared to that...

  7. Detector, shielding and geometric design factors for a high-resolution PET system

    The authors have evaluated the resolution, efficiency and scatter rejection on a new high resolution PET system designed for animal studies which is based on a 2-D modular detector system. A digital positioning system was evaluated by testing different encoding methods. Tungsten inter-plane septa of different thicknesses and geometries were evaluated by Monte Carlo simulations and experiments. The detector system consists of a 6 x 8 array of BGO crystals coupled to 2 dual photomultiplier tubes (PMTs). The crystals are 3.5 mm wide with 4 mm spacing transaxially and are 6.25 mm long with 6.75 mm spacing axially. PMT outputs are digitized and Anger camera type logic is used to determine the X and Y location of the scintillation event

  8. Study and implementation of the CADIS methodology to research reactor shielding design

    Souza, Gregorio S.; Shorto, Julian M.B.; Santos, Adimir dos, E-mail: greguis@gmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    The Consistent Adjoint Driven Importance Sampling (CADIS) is a methodology that basically uses source biasing and a mesh-based importance map. Therefore, to make the best use of an importance map, the map must be consistent with the source biasing. To achieve this consistency, a Sn calculation could be made to improve the importance map and the computational performance. The MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations) code does that and this work intends to study the code options to generate the importance map. A pool type 10 MW research reactor was designed in a simple way just to study the prompt gamma rays penetration in the concrete and therefore study the CADIS methodology applied to point detectors and mesh tallies. By keeping constant the simulation time and the CPU (Central Processing Unit) power a significant improvement was achieved in the relative errors for the point detectors and for the mesh tally. (author)

  9. Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP

    The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model, which has a boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245x10-4 s was recorded for the new boron carbide designed model while a value of 1.5571x10-7 s was recorded for the original MCNP design of the GHARR-1.

  10. Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP.

    Abrefah, R G; Sogbadji, R B M; Ampomah-Amoako, E; Birikorang, S A; Odoi, H C; Nyarko, B J B

    2011-01-01

    The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model, which has a boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245 × 10(-4)s was recorded for the new boron carbide designed model while a value of 1.5571 × 10(-7)s was recorded for the original MCNP design of the GHARR-1. PMID:20637646

  11. 3-dimensional shielding design for a spallation neutron source facility in the high-intensity proton accelerator project

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

    2003-03-01

    Evaluation of shielding performance for a 1 MW spallation neutron source facility in the Materials and Life Science Facility being constructed in the High-Intensity Proton Accelerator Project (J-PARC) is important from a viewpoint of radiation safety and optimization of arrangement of components. This report describes evaluated results for the shielding performance with modeling three-dimensionally whole structural components including gaps between them in detail. A Monte Carlo calculation method with MCNPX2.2.6 code and LA-150 library was adopted. Streaming and void effects, optimization of shield for cost reduction and optimization of arrangement of structures such as shutters were investigated. The streaming effects were investigated quantitatively by changing the detailed structure of components and gap widths built into the calculation model. Horizontal required shield thicknesses were ranged from about 6.5 m to 7.5 m as a function of neutron beam line angles. A shutter mechanism for a horizontal neutron reflectometer that was directed downward was devised, and it was shown that the shielding performance of the shutter was acceptable. An optimal biological shield configuration was finally determined according to the calculated results. (author)

  12. Shielding door

    An exhaust processing device disposed at the outside of a radioactive nuclide handling chamber is connected to a shielding door as an exit/inlet for the radioactive nuclide handling chamber. An exhaust chamber is disposed in the inside of the thick shielding door having a thickness. The exhaust chamber is always evacuated by an exhaustion blower and maintained at a negative pressure. The radioactive nuclides in the radiation nuclide handling facility are shielded by an inner seal of the double seals which seal the gap between the wall body and the shielding door. Even if a trace amount of radioactive nuclides leaks from the seal at the inner side, it is shielded by an outer seal, and sucked into the exhaust chamber which is maintained at the negative pressure. Then, it is passed from a ventilation channel through a flexible tube then caught and removed by the filter of the exhaust processing device. This can reduce the capacity of the exhaustion blower to reduce the scale of the exhaust processing device. (I.N.)

  13. Drip Shield Emplacement Gantry Concept

    Silva, R.A.; Cron, J.

    2000-03-29

    This design analysis has shown that, on a conceptual level, the emplacement of drip shields is feasible with current technology and equipment. A plan for drip shield emplacement was presented using a Drip Shield Transporter, a Drip Shield Emplacement Gantry, a locomotive, and a Drip Shield Gantry Carrier. The use of a Drip Shield Emplacement Gantry as an emplacement concept results in a system that is simple, reliable, and interfaces with the numerous other exising repository systems. Using the Waste Emplacement/Retrieval System design as a basis for the drip shield emplacement concept proved to simplify the system by using existing equipment, such as the gantry carrier, locomotive, Electrical and Control systems, and many other systems, structures, and components. Restricted working envelopes for the Drip Shield Emplacement System require further consideration and must be addressed to show that the emplacement operations can be performed as the repository design evolves. Section 6.1 describes how the Drip Shield Emplacement System may use existing equipment. Depending on the length of time between the conclusion of waste emplacement and the commencement of drip shield emplacement, this equipment could include the locomotives, the gantry carrier, and the electrical, control, and rail systems. If the exisiting equipment is selected for use in the Drip Shield Emplacement System, then the length of time after the final stages of waste emplacement and start of drip shield emplacement may pose a concern for the life cycle of the system (e.g., reliability, maintainability, availability, etc.). Further investigation should be performed to consider the use of existing equipment for drip shield emplacement operations. Further investigation will also be needed regarding the interfaces and heat transfer and thermal effects aspects. The conceptual design also requires further design development. Although the findings of this analysis are accurate for the assumptions made

  14. Ulexite-galena intermediate-weight concrete as a novel design for overcoming space and weight limitations in the construction of efficient shields against neutrons and photons.

    Aghamiri, S M R; Mortazavi, S M J; Razi, Z; Mosleh-Shirazi, M A; Baradaran-Ghahfarokhi, M; Rahmani, F; Faeghi, F

    2013-01-01

    Recently, due to space and weight limitations, scientists have tried to design and produce concrete shields with increased attenuation of radiation but not increased mass density. Over the past years, the authors' had focused on the production of heavy concrete for radiation shielding, but this is the first experience of producing intermediate-weight concrete. In this study, ulexite (hydrated sodium calcium borate hydroxide) and galena (lead ore) have been used for the production of a special intermediate-weight concrete. Shielding properties of this intermediate-weight concrete against photons have been investigated by exposing the samples to narrow and broad beams of gamma rays emitted from a ⁶⁰Co radiotherapy unit. Densities of the intermediate-weight concrete samples ranged 3.64-3.90 g cm⁻³, based on the proportion of the ulexite in the mix design. The narrow-beam half-value layer (HVL) of the ulexite-galena concrete samples for 1.25 MeV ⁶⁰Co gamma rays was 2.84 cm, much less than that of ordinary concrete (6.0 cm). The Monte Carlo (MC) code MCNP4C was also used to model the attenuation of ⁶⁰Co gamma-ray photons and Am-Be neutrons of the ulexite-galena concrete with different thicknesses. The ⁶⁰Co HVL calculated by MCNP simulation was 2.87 cm, indicating a good agreement between experimental measurements and MC simulation. Furthermore, MC-calculated results showed that thick ulexite-galena concrete shields (60-cm thickness) had a 7.22 times (722 %) greater neutron attenuation compared with ordinary concrete. The intermediate-weight ulexite-galena concrete manufactured in this study may have many important applications in the construction of radiation shields with weight limitations such as the swing or sliding doors that are currently used for radiotherapy treatment rooms. PMID:23019599

  15. Ulexite-galena intermediate-weight concrete as a novel design for overcoming space and weight limitations in the construction of efficient shields against neutrons and photons

    Recently, due to space and weight limitations, scientists have tried to design and produce concrete shields with increased attenuation of radiation but not increased mass density. Over the past years, the authors' had focused on the production of heavy concrete for radiation shielding, but this is the first experience of producing intermediate-weight concrete. In this study, ulexite (hydrated sodium calcium borate hydroxide) and galena (lead ore) have been used for the production of a special intermediate- weight concrete. Shielding properties of this intermediate-weight concrete against photons have been investigated by exposing the samples to narrow and broad beams of gamma rays emitted from a 60Co radiotherapy unit. Densities of the intermediate-weight concrete samples ranged 3.64-3.90 g cm-3, based on the proportion of the ulexite in the mix design. The narrow-beam half-value layer (HVL) of the ulexite-galena concrete samples for 1.25 MeV 60Co gamma rays was 2.84 cm, much less than that of ordinary concrete (6.0 cm). The Monte Carlo (MC) code MCNP4C was also used to model the attenuation of 60Co gamma-ray photons and Am-Be neutrons of the ulexite-galena concrete with different thicknesses. The 60Co HVL calculated by MCNP simulation was 2.87 cm, indicating a good agreement between experimental measurements and MC simulation. Furthermore, MC-calculated results showed that thick ulexite-galena concrete shields (60-cm thickness) had a 7.22 times (722 %) greater neutron attenuation compared with ordinary concrete. The intermediate-weight ulexite- galena concrete manufactured in this study may have many important applications in the construction of radiation shields with weight limitations such as the swing or sliding doors that are currently used for radiotherapy treatment rooms. (authors)

  16. Using glass as a shielding material

    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)

  17. Optimized design of shields for diagnostic X rays with NCRP 147 technique; Diseno optimizado de blindajes para rayos X diagnostico con tecnica NCRP 147

    Gama T, G. [Calidad XXI SA de CV, Zacatecas 67-007 Col. Roma, 06700 Mexico D.F. (Mexico)]. e-mail: cxxi@prodigy.net.mx

    2006-07-01

    A comparison among the design techniques of shielding for X-ray diagnostic rooms with the NCRP 49 (1976) report technique, AAPM 39 (1993) Y the one of the NCRP 147 (2005) technique. The designs correspond to a room of conventional X-rays, one of fluoroscopy, one of tomography Y one of mammography. In all the cases it demonstrates that the NCRP 49 technique overestimate the shieldings. The causes of the overestimation of the NCRP 49 can be attributed to: a) high values of the work charge that don't consider the spectral fluence of the photons that are present in each room, b) to the differences in the values of the kerma in air without attenuation for the dispersed primary radiation Y of leakage among both reports. (Author)

  18. Shield calculations, optimization vs. paradigm

    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)

  19. Biological Shielding Design Effectiveness of the Brachytherapy Unit at the Korle Bu Teaching Hospital in Ghana Using Mcnp5 Monte Carlo Code

    C.C. Arwui

    2011-05-01

    Full Text Available Design objectives for brachytherapy treatment facilities require sufficient shielding to reduce primary and scatter radiation to design limit in order to limit exposure to patients, staff and the general public. The primary aim of this study is to verify whether shielding of the brachytherapy unit at the Korle Bu teaching Hospital in Ghana provides adequate protection in order to assess any radiological health and safety impact and also test the suitability of other available sources. The study evaluates the effectiveness of the biological shielding design of a Cs-137 brachytherapy unit at the Korle-Bu Teaching Hospital in Ghana using MCNP5. The facility was modeled based on the design specifications for LDR Cs-137, MDR Cs-137, HDR Co-60 and HDR Ir-192 treatment modalities. The estimated dose rate ranged from (0.01-0.15 μSv/h and (0.37-3.05 μSv/h for the existing initial and decayed activities of LDR Cs-137 for the public and controlled areas respectively, (0.03-0.57 μSv/h and (1.53-8.06 μSv/h for MDR Cs-137, (7.47-59.46 μSv/h and (144.87-178.74 μSv/h for HDR Co- 60, (0.13-6.95 μSv/h and (19.47-242.98 μSv/h for HDR Ir-192 for the public and controlled areas respectively. The results were verified by dose rates measurement for the current LDR setup at the Brachytherapy unit and agreed quiet well. It was also compared with the reference values of 0.5 μSv/h for public areas and 7.5 μSv/h for controlled areas respectively. It can be concluded that the shielding is adequate for the existing source.

  20. A new design of a lead-acrylic shield for staff dose reduction in radial and femoral access coronary catheterization

    Today's standard radiation protection during coronary angiography and percutaneous coronary interventions is the combined use of lead acrylic shields and table-mounted lower body protection. Ambient dose measurements, however, have shown that these protection devices need improvement. Using an anthropomorphic physical phantom, various scenarios were investigated with respect to personnel exposure: (a) enlarging the shield (b) adding a flexible protective curtain to the bottom side of the shield, and (c) application of radioprotective patient drapes. For visualization of the dose reduction effect, Monte Carlo simulations were performed. The flexible curtain in contact with the patient's body reduces the ambient dose rate at the operator's position by up to (87.5 % ± 7.1) compared to the situation with the bare shield. The use of both the flexible curtain and the patient drape reduces the ambient dose rate by up to (90.8 % ± 7). Similar results were achieved for the assisting personnel when they were positioned next to the operator. In addition, the enlarged shield provides better protection of the head region of tall operators. Adding a flexible protective curtain to the bottom side of the shield can protect operators from high doses, especially for body parts which are not protected by lead aprons, e.g. head, and eye lenses. This may be important with respect to lower dose limits for eye lenses in future. The protective effect in real-life working conditions is still being evaluated in an ongoing clinical study.

  1. A new design of a lead-acrylic shield for staff dose reduction in radial and femoral access coronary catheterization

    Eder, H. [Deptartment of Radiation Protection (Germany); Seidenbusch, M.C.; Treitl, M. [Muenchen Univ. Clinical Center (Germany). Inst. for Clinical Radiology; Gilligan, P. [Mater Private Hospital, Dublin (Ireland). Medical Physics

    2015-10-15

    Today's standard radiation protection during coronary angiography and percutaneous coronary interventions is the combined use of lead acrylic shields and table-mounted lower body protection. Ambient dose measurements, however, have shown that these protection devices need improvement. Using an anthropomorphic physical phantom, various scenarios were investigated with respect to personnel exposure: (a) enlarging the shield (b) adding a flexible protective curtain to the bottom side of the shield, and (c) application of radioprotective patient drapes. For visualization of the dose reduction effect, Monte Carlo simulations were performed. The flexible curtain in contact with the patient's body reduces the ambient dose rate at the operator's position by up to (87.5 % ± 7.1) compared to the situation with the bare shield. The use of both the flexible curtain and the patient drape reduces the ambient dose rate by up to (90.8 % ± 7). Similar results were achieved for the assisting personnel when they were positioned next to the operator. In addition, the enlarged shield provides better protection of the head region of tall operators. Adding a flexible protective curtain to the bottom side of the shield can protect operators from high doses, especially for body parts which are not protected by lead aprons, e.g. head, and eye lenses. This may be important with respect to lower dose limits for eye lenses in future. The protective effect in real-life working conditions is still being evaluated in an ongoing clinical study.

  2. Design and performance characteristics of an electromagnetic interference shielded enclosure for high voltage Pockels cell switching system

    A K Sharma; K K Mishra; M Raghuramaiah; P A Naik; P D Gupta

    2007-06-01

    An electro-magnetic interference noise shielding enclosure for Pockels cells for high speed synchronized switching has been set-up and tested. The shielding effectiveness of the aluminum enclosures housing the Pockels cells and the electronic circuitry has been measured using a high impedance probe and is found to be $\\sim 50 dB$. This ensures a noise-free and synchronized electro-optic switching in an Nd:glass re-generative amplifier of chirped pulse amplification based table top terawatt laser system.

  3. Mechanical shielded hot cell

    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)

  4. Design and construction of a shielded process box for the production of radiopharmaceuticals labelled with 131I

    A leakproof process box, shielded with a 5 mm lead wall, for the labelling, purification, pH adjutment and dispensing of Rose Bengal 131I, Hippuran 131I, Diprocon 131I, Hipaque 131I, Bromosuphthalein 131I, etc. is described. (author)

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

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

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

    Mason, John A.; Burke, Kevin J.; Towner, Antony C.N. [ANTECH, A. N. Technology Ltd., Unit 6, Thames Park, Wallingford, Oxfordshire, OX10 9TA (United Kingdom); Beaven, Graham; Spence, Robert [Dounreay Site Restoration Ltd., Thurso, Caithness, Scotland, KW14 7TZ (United Kingdom)

    2013-07-01

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

  7. The Load Design and Implementation of HJ-1-C Space-borne SAR%HJ-1-C卫星合成孔径雷达载荷的设计与实现

    禹卫东; 杨汝良; 邓云凯; 赵凤军; 雷宏

    2014-01-01

    HJ-1-C is a Synthetic Aperture Radar (SAR) satellite in the Constellation of “2+1” for China environment and disaster monitoring. It works at S-band with a resolution of 5 m. SAR payload uses a reflector antenna and a high-power concentrated transmitter. Its light weight and high efficiency is very suitable for a small satellite platform. Now HJ-1-C satellite has been launched into orbit and has acquired Chinese first S-band SAR images from space, which demonstrate excellent quality and rich information about scenes imaged. This success verifies our design, testing and experiment work on the payload. With its following operation, HJ-1-C satellite is expected to make a great contribution to the applications of environment protection and disaster monitoring in China. This paper introduces the design and development of HJ-1-C SAR payload, present its main parameters and performance, describes its device details and its manufacture, testing and experiment process. Some images acquired in the orbit are showed.%HJ-1-C 是我国环境与灾害监测小卫星“2+1”星座中的一颗合成孔径雷达(SAR)卫星,工作于 S 波段,具有5 m分辨率。SAR载荷采用网状反射面天线和大功率放大器方案,具有重量轻、效率高的特点,适合于小卫星平台。目前HJ-1-C卫星已在轨运行,获得我国首批S波段星载SAR图像,图像质量高,地物信息丰富,表明SAR 载荷的设计合理,试验和测试充分。HJ-1-C 卫星将为我国减灾和环境应用发展做出贡献。该文将对 HJ-1-C卫星SAR载荷的设计和研制进行全面介绍,包括其主要功能和技术指标,各部分的设计,以及研制、测试和试验工程,最后给出其在轨获得的图像。

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

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

  9. Designing of epoxy composites reinforced with carbon nanotubes grown carbon fiber fabric for improved electromagnetic interference shielding

    B. P. Singh

    2012-06-01

    Full Text Available In this letter, we report preparation of strongly anchored multiwall carbon nanotubes (MWCNTs carbon fiber (CF fabric preforms. These preforms were reinforced in epoxy resin to make multi scale composites for microwave absorption in the X-band (8.2-12.4GHz. The incorporation of MWCNTs on the carbon fabric produced a significant enhancement in the electromagnetic interference shielding effectiveness (EMI-SE from −29.4 dB for CF/epoxy-composite to −51.1 dB for CF-MWCNT/epoxy multiscale composites of 2 mm thickness. In addition to enhanced EMI-SE, interlaminar shear strength improved from 23 MPa for CF/epoxy-composites to 50 MPa for multiscale composites indicating their usefulness for making structurally strong microwave shields.

  10. Designing of epoxy composites reinforced with carbon nanotubes grown carbon fiber fabric for improved electromagnetic interference shielding

    Singh, B. P.; Choudhary, Veena; Saini, Parveen; Mathur, R. B.

    2012-06-01

    In this letter, we report preparation of strongly anchored multiwall carbon nanotubes (MWCNTs) carbon fiber (CF) fabric preforms. These preforms were reinforced in epoxy resin to make multi scale composites for microwave absorption in the X-band (8.2-12.4GHz). The incorporation of MWCNTs on the carbon fabric produced a significant enhancement in the electromagnetic interference shielding effectiveness (EMI-SE) from -29.4 dB for CF/epoxy-composite to -51.1 dB for CF-MWCNT/epoxy multiscale composites of 2 mm thickness. In addition to enhanced EMI-SE, interlaminar shear strength improved from 23 MPa for CF/epoxy-composites to 50 MPa for multiscale composites indicating their usefulness for making structurally strong microwave shields.

  11. BUMPERII - DESIGN ANALYSIS CODE FOR OPTIMIZING SPACECRAFT SHIELDING AND WALL CONFIGURATION FOR ORBITAL DEBRIS AND METEOROID IMPACTS

    Hill, S. A.

    1994-01-01

    BUMPERII is a modular program package employing a numerical solution technique to calculate a spacecraft's probability of no penetration (PNP) from man-made orbital debris or meteoroid impacts. The solution equation used to calculate the PNP is based on the Poisson distribution model for similar analysis of smaller craft, but reflects the more rigorous mathematical modeling of spacecraft geometry, orientation, and impact characteristics necessary for treatment of larger structures such as space station components. The technique considers the spacecraft surface in terms of a series of flat plate elements. It divides the threat environment into a number of finite cases, then evaluates each element of each threat. The code allows for impact shielding (shadowing) of one element by another in various configurations over the spacecraft exterior, and also allows for the effects of changing spacecraft flight orientation and attitude. Four main modules comprise the overall BUMPERII package: GEOMETRY, RESPONSE, SHIELD, and CONTOUR. The GEOMETRY module accepts user-generated finite element model (FEM) representations of the spacecraft geometry and creates geometry databases for both meteoroid and debris analysis. The GEOMETRY module expects input to be in either SUPERTAB Universal File Format or PATRAN Neutral File Format. The RESPONSE module creates wall penetration response databases, one for meteoroid analysis and one for debris analysis, for up to 100 unique wall configurations. This module also creates a file containing critical diameter as a function of impact velocity and impact angle for each wall configuration. The SHIELD module calculates the PNP for the modeled structure given exposure time, operating altitude, element ID ranges, and the data from the RESPONSE and GEOMETRY databases. The results appear in a summary file. SHIELD will also determine the effective area of the components and the overall model, and it can produce a data file containing the probability

  12. Accuracy evaluation of the current data and method applied to shielding design of the Fusion Experimental Reactor (FER)

    Shielding benchmarking study of the current data and method applied to the Fusion Experimental Reactor (FER) was performed. First, neutron and gamma ray fluxes were calculated by the one-dimensional SN code using various cross section libraries and the continuous energy Monte Carlo code. The results were compared in terms of the SN/MC ratio. The worst ratios are about 0.5 and 0.25 for neutron flux and gamma ray flux, respectively. Next, the analytical calculations of the iron sphere transmission experiment of 14 MeV neutrons were performed to examine the accuracy of cross section data of iron, which is the most important material of shield. The E/C ratio is larger than 2 even if the continuous energy Monte Carlo code was used. Thirdly, the influence of geometrical representation of the shield was investigated by comparing the homogeneous model and the heterogeneous model (alternating layers of SS316 and water). As a result, it was made clear that the homogeneous model underestimates neutron flux by a factor of 2. Finally, the necessity of benchmark experiment and improvement of cross section library was pointed out as the further R and D issues. (author)

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

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

  14. Design on Electromagnetic Shielding for a Sealed Communication Terminal%一种密封通信终端的电磁屏蔽设计

    吴迪; 高驰名; 马建章

    2015-01-01

    电磁干扰严重影响着密封终端技术性能的有效发挥,电磁屏蔽是提高电子系统和电子设备电磁兼容性能力的重要措施之一。简要介绍了典型通信终端的结构形式,分析了密封通信终端电磁屏蔽设计所面临的问题,指出了影响设备电磁屏蔽效能的泄露源。以典型密封通信终端设备为实例,在不改变或降低密设备封性能的情况下,针对连接器、开关、指示灯孔、显示屏窗口、键盘开口和盒体缝隙等不同的结构特点进行电磁屏蔽设计,给出了具体的设计思路和方法。经电磁兼容测试验证,该设计方法是科学、有效的。%The electromagnetic interference(EMI)has a strong impact on the technology performance of sealed terminal.The elec⁃tromagnetic shielding is one of the important measures of raising electromagnetic compatibility of electronic system and electronic equip⁃ment.This paper introduces briefly the structure of a typical sealed communication terminal,analyzes the problems in electromagnetic shielding design of sealed communication terminals,and also points out the leak source that impacts the electromagnetic shielding effi⁃ciency of terminals.Taking a typical sealed communication terminal as an example,the electromagnetic shielding design is performed aiming at different structural characteristics of connector,switch,indicator hole,screen,keyboard interface,power connector and switch without sealing capability changing and reducing,and the typical design idea and method are proposed.The electromagnetic compatibility test results show that this design method is scientific and available.

  15. Noise Shielding Using Acoustic Metamaterials

    We exploit theoretically a class of rectangular cylindrical devices for noise shielding by using acoustic metamaterials. The function of noise shielding is justified by both the far-field and near-field full-wave simulations based on the finite element method. The enlargement of equivalent acoustic scattering cross sections is revealed to be the physical mechanism for this function. This work makes it possible to design a window with both noise shielding and air flow. (electromagnetism, optics, acoustics, heat transfer, classical mechanics, and fluid dynamics)

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

    Kramer, S. L.; Ghosh, V. J.; Breitfeller, M.

    2016-08-01

    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.

  17. Shielded cells transfer automation

    Nuclear waste from shielded cells is removed, packaged, and transferred manually in many nuclear facilities. To reduce radiation exposure to operators, technological advances in remote handling and automation were employed. An industrial robot and a specially designed end effector, access port, and sealing machine were used to remotely bag waste containers out of a glove box. The system is operated from a control panel outside the work area via television cameras

  18. Design and evaluation for the shielding system of the 9 MeV travelling wave linear electron accelerator

    The authors use EGS4 code, a generally known Monte Carlo computer simulation package, to carry out the simulation analysis of the radiation dose distribution around the head shielding system and inside the accelerator hall of the 9 MeV travelling wave linear electron accelerator. The accelerator is used for the large container inspecting system. The comparison of experience formulae evaluation and practical data was made. The results show that, at the main reference points in the accelerator hall, the dose calculated by EEG's is well coincided with the results measured. It serves as a good example for flexible application of EGS4

  19. Designing of epoxy composites reinforced with carbon nanotubes grown carbon fiber fabric for improved electromagnetic interference shielding

    Singh, B. P.; Veena Choudhary; Parveen Saini; Mathur, R.B.

    2012-01-01

    In this letter, we report preparation of strongly anchored multiwall carbon nanotubes (MWCNTs) carbon fiber (CF) fabric preforms. These preforms were reinforced in epoxy resin to make multi scale composites for microwave absorption in the X-band (8.2-12.4GHz). The incorporation of MWCNTs on the carbon fabric produced a significant enhancement in the electromagnetic interference shielding effectiveness (EMI-SE) from −29.4 dB for CF/epoxy-composite to −51.1 dB for CF-MWCNT/epoxy multiscale comp...

  20. Designing of shielding and collimator for high-energy X-ray beam: application to 3 MV Tandem Pelletron based HEX-ray spectrometry

    High-energy X-rays (60-1000 keV) or HEX-ray bear unique advantage over conventional X-rays in the field of diffraction and scattering experiments. HEX-ray is an important experimental technique because of its uniqueness to characterize technically challenging samples (liquids, glass, amorphous materials, nanomaterial, thick samples, etc.). In the present paper detail designing of the lead collimator and shielding for the recently developed HEX-Ray spectrometry experiment using characteristic PbK-lines (75-85 keV) emitted due to bombardment of 3 MeV proton beam generated from the Pelletron accelerator has been reported. Attenuation (%), transmission (%), and energy absorption (%) are calculated using the NIST data which is based on X-ray interaction cross sections and material densities. Thickness for Pb-shielding possessing highest attenuation (99.99%) for 75 keV and 85 keV X-rays are 3 mm and 4 mm respectively. Hence, shielding thickness about 5 mm is used to cover the experimental table. Again, introducing a source collimator (diameter = 8 mm, length = 35 mm) between the X-ray target (source) and the sample, the incoherent scattered radiation due to air in the path of the beam and the multiple reflection from the chamber as well as from the detector inserting tube can be suppressed. An extra energetic and intense X-ray line was observed at about 44.5 keV using a high-purity germanium (HpGe) detector (Detector Systems GmbH (DSG), Germany). This contamination line is not an escape peak (PbKGeK) and can be suppressed using 8 mm Pb-collimator between X-ray target and sample. The details of the designing, experimental setup, and its applications are discussed in the full paper.

  1. Radiation shielding for neutron guides

    Ersez, T. [Reactor Operations, ANSTO, PMB 1, Menai, NSW 2234 (Australia)]. E-mail: tez@ansto.gov.au; Braoudakis, G. [Reactor Operations, ANSTO, PMB 1, Menai, NSW 2234 (Australia); Osborn, J.C. [Reactor Operations, ANSTO, PMB 1, Menai, NSW 2234 (Australia)

    2006-11-15

    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.

  2. Spacecraft Electrostatic Radiation Shielding

    2008-01-01

    This project analyzed the feasibility of placing an electrostatic field around a spacecraft to provide a shield against radiation. The concept was originally proposed in the 1960s and tested on a spacecraft by the Soviet Union in the 1970s. Such tests and analyses showed that this concept is not only feasible but operational. The problem though is that most of this work was aimed at protection from 10- to 100-MeV radiation. We now appreciate that the real problem is 1- to 2-GeV radiation. So, the question is one of scaling, in both energy and size. Can electrostatic shielding be made to work at these high energy levels and can it protect an entire vehicle? After significant analysis and consideration, an electrostatic shield configuration was proposed. The selected architecture was a torus, charged to a high negative voltage, surrounding the vehicle, and a set of positively charged spheres. Van de Graaff generators were proposed as the mechanism to move charge from the vehicle to the torus to generate the fields necessary to protect the spacecraft. This design minimized complexity, residual charge, and structural forces and resolved several concerns raised during the internal critical review. But, it still is not clear if such a system is costeffective or feasible, even though several studies have indicated usefulness for radiation protection at energies lower than that of the galactic cosmic rays. Constructing such a system will require power supplies that can generate voltages 10 times that of the state of the art. Of more concern is the difficulty of maintaining the proper net charge on the entire structure and ensuring that its interaction with solar wind will not cause rapid discharge. Yet, if these concerns can be resolved, such a scheme may provide significant radiation shielding to future vehicles, without the excessive weight or complexity of other active shielding techniques.

  3. Performance of the improved version of Monte Carlo Code A{sup 3}MCNP for cask shielding design

    Hasegawa, T. [Mitsubishi Heavy Industries, Yokohama (Japan); Ueki, K. [Tokai Univ., Kanagawa (Japan); Sato, O. [Mitsubishi Research Inst., Tokyo (Japan); Sjoden, G.E. [Dept. of Nuclear and Radiological Engineering, Univ. of Florida, Gainesville, FL (United States); Miyake, Y.; Ohmura, M.; Haghighat, A.

    2004-07-01

    A{sup 3}MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, that automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic ''importance'' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A{sup 3}MCNP uses the 3-D Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A3MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A{sup 3}MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A{sup 3}MCNP (referred to as A{sup 3}MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A{sup 3}MCNPV for cask neutron and gamma-ray shielding problem.

  4. Shielding design of a treatment room for an accelerator-based epithermal neutron irradiation facility for BNCT

    Protecting the facility personnel and the general public from radiation exposure is a primary safety concern of an accelerator-based epithermal neutron irradiation facility. This work makes an attempt at answering the questions open-quotes How much?close quotes and open-quotes What kind?close quotes of shielding will meet the occupational limits of such a facility. Shielding effectiveness is compared for ordinary and barytes concretes in combination with and without borated polyethylene. A calculational model was developed of a treatment room, patient open-quotes scatterer,close quotes and the epithermal neutron beam. The Monte Carlo code, MCNP, was used to compute the total effective dose equivalent rates at specific points of interest outside of the treatment room. A conservative occupational effective dose rate limit of 0.01 mSv h-1 was the guideline for this study. Conservative Monte Carlo calculations show that constructing the treatment room walls with 1.5 m of ordinary concrete, 1.2 m of barytes concrete, 1.0 m of ordinary concrete preceded by 10 cm of 5% boron-polyethylene, or 0.8 m of barytes concrete preceded by 10 cm of 5% boron-polyethylene will adequately protect facility personnel. 20 refs., 8 figs., 2 tabs

  5. Design of a Laboratory Hall Thruster with Magnetically Shielded Channel Walls, Phase III: Comparison of Theory with Experiment

    Mikellides, Ioannis G.; Katz, Ira; Hofer, Richard R.; Goebel, Dan M.

    2012-01-01

    A proof-of-principle effort to demonstrate a technique by which erosion of the acceleration channel in Hall thrusters of the magnetic-layer type can be eliminated has been completed. The first principles of the technique, now known as "magnetic shielding," were derived based on the findings of numerical simulations in 2-D axisymmetric geometry. The simulations, in turn, guided the modification of an existing 6-kW laboratory Hall thruster. This magnetically shielded (MS) thruster was then built and tested. Because neither theory nor experiment alone can validate fully the first principles of the technique, the objective of the 2-yr effort was twofold: (1) to demonstrate in the laboratory that the erosion rates can be reduced by >order of magnitude, and (2) to demonstrate that the near-wall plasma properties can be altered according to the theoretical predictions. This paper concludes the demonstration of magnetic shielding by reporting on a wide range of comparisons between results from numerical simulations and laboratory diagnostics. Collectively, we find that the comparisons validate the theory. Near the walls of the MS thruster, theory and experiment agree: (1) the plasma potential has been sustained at values near the discharge voltage, and (2) the electron temperature has been lowered by at least 2.5-3 times compared to the unshielded (US) thruster. Also, based on carbon deposition measurements, the erosion rates at the inner and outer walls of the MS thruster are found to be lower by at least 2300 and 1875 times, respectively. Erosion was so low along these walls that the rates were below the resolution of the profilometer. Using a sputtering yield model with an energy threshold of 25 V, the simulations predict a reduction of 600 at the MS inner wall. At the outer wall ion energies are computed to be below 25 V, for which case we set the erosion to zero in the simulations. When a 50-V threshold is used the computed ion energies are below the threshold at both

  6. Calculation And Design Of A New Configuration For Radiation Shielding At Neutron Beam No.3 For Fundamental And Applied Researches

    The tangential horizontal channel of No. 3 of the Dalat Research Reactor has been opened and used during the 1990s. The utilizations of the thermal neutron beam at this channel were the Neutron Radiography and the Prompt Gamma Neutron Activation Analysis method (PGNAA). At present, the neutron beam used for nuclear structure data researches based on the Summing of Amplitude Coincident Pulses system (SACP). Beside, several related research equipments have been set up and operated for the research purposes. A renovation of the neutron channel, therefore, will play an important role in safe and effective utilizations of the neutron beam in fields of nuclear physic training and researches. A new configuration for radiation shielding has been simulated by MCNP code. The calculated results of dose rates for neutron and gamma at working positions are in range of dose rate limit. (author)

  7. Space reactor shielding: an assessment of the technology

    Space power reactor systems require shielding to protect payload and reactor shielding components, and also maintenance and operating personnel. Shield composition, size, and shape are important design considerations, since the shield can dominate the overall weight of the system. Techniques for space reactor shield design analysis and optimization and experimental test facilities are available for design verification. With these tools, a shielding technology in support of current and future space power reactor systems can be developed. Efforts in this direction should begin with a generic shielding program to provide information on materials properties and geometric effects and should be followed by project-specific shielding programs to provide design optimization and prototype shield verification

  8. Performance test on shielding concrete

    The cylinder of the shielding concrete is made from common Portland cement and home-made coarse or fine aggregates. Orthogonal design experiment and regression analysis are adopted to study the effects of the water content, sand percentage and water-cement ratio on the property of shielding concrete and the difference between them. The test shows that the tensile strength is in inverse proportion with water-cement ratio, and the influence is quite significant. Another factor is the type of aggregates. The effect of the age on its density is not obvious. Similarly, the concrete shielding γ rays shares the same influencing factors with that shielding neutron rays on density, slump and tensile strength. And both have the same change rules regarding to mechanical property. (authors)

  9. Shielding design of a brachytherapy unit at the Korle Bu teaching hospital in Ghana: Comparison of theoretical calculations and experimental study

    A theoretical study was carried out to re-evaluate the integrity of the biological shielding of 137Cs brachytherapy unit at the Korle Bu Teaching Hospital (Ghana), and the results were verified by measurement of the dose rates at selected locations. The primary objective was to determine the current state of protection and safety of staff and the general public. Shielding design of the brachytherapy unit at the hospital was based on postulated workload and occupancy factors of the facility. The facility has been in existence for 12 y and has accumulated operational workload data that differs from the postulated one. The results show that despite the variation in actual and postulated workloads, the dose rates were below the reference values 0.5 mSv h-1 for public areas and 7.5 μSv h-1 for controlled areas. These values were in the range of 0.10-0.12 μSv h-1 for public areas and of 0.50-2.10 μSv h-1 for controlled areas. (authors)

  10. A Study on the Radiation Source Effect to the Radiation Shielding Analysis for a Spent-Fuel Cask Design with Burnup-Credit

    The radiation shielding analysis for a Burnup-credit (BUC) cask designed under the management of Korea Radioactive Waste Management Corporation (KRMC) was performed to examine the contribution of each radiation source affecting dose rate distribution around the cask. Various radiation sources, which contain neutron and gamma-ray sources placed in active fuel region and the activation source, and imaginary nuclear fuel were all considered in the MCNP calculation model to realistically simulate the actual situations. It was found that the maximum external and surface dose rates of the spent fuel cask were satisfied with the domestic standards both in normal and accident conditions. In normal condition, the radiation dose rate distribution around the cask was mainly influenced by activation source (60 Co radioisotope); in another case, the neutron emitted in active fuel region contributed about 90% to external dose rate at 1m distance from side surface of the cask. Besides, the contribution level of activation source was dramatically increased to the dose rates in top and bottom regions of the cask. From this study, it was recognized that the detailed investigation on the radiation sources should be performed conservatively and accurately in the process of radiation shielding analysis for a BUC cask.

  11. Development of discrete ordinates S sub(N) code in three-dimensional (X,Y,Z) geometry for shielding design

    A discrete ordinates transport code ENSEMBLE in (X,Y,Z) geometry has been developed for the purpose of shielding calculations in three-dimensional geometry. The code has some superior features, compared with THREETRAN which is the only code of the same kind so far developed. That is, the code can treat higher order anisotropic scattering and employs a coarse mesh rebalancing method. Moreover it has a negative flux fix-up routine using a variable weight diamond difference equation scheme and has a ray-effect fix-up option using a fictitious source based on S sub(N)→P sub(N-1) conversion technique. Formulations for these advanced features in three-dimensional space have been derived. As the demonstration of the capabilities of the code, several numerical analyses and an analysis of an annular duct streaming experiment in JRR-4 at Japan Atomic Energy Research Institute, have been performed. As a result of these analyses, confirmation has been obtained for the prospect of applicability of ENSEMBLE to practical shielding design. (author)

  12. Shielding integrity testing of radioactive material transport packaging

    Although this Code of Practice is intended primarily to cover shielding integrity test requirements for off-site shielded radioactive material transport packaging, it may also be partly applicable to containers and specialised handling equipment (e.g. fuelling machines) used only on site, and to radiation shielding generally. The code is not concerned with proving adequacy of shielding design or with its absolute shielding value. (author)

  13. A1C Test and Diabetes

    ... Prediabetes The A1C Test and Diabetes The A1C Test and Diabetes What is the A1C test? The A1C test is a blood test that ... management and diabetes research. How does the A1C test work? The A1C test is based on the ...

  14. Guidebook on radiation shielding safety for nuclear fuel facilities Q and A volume

    The Q and A volume of 'Guidebook on Radiation Shielding Safety for Nuclear Fuel Facilities' describes questions and answers which are commonly raised by the novices of shielding design and shielding safety evaluation. In this report, there are about 40 sets of Q and A which are classified by 7 different subjects, namely, (1) outlines of shielding, (2) methodology of shielding design, (3) shielding materials, (4) bulk shielding, (5) streaming, (6) skyshine, and (7) certification of shielding performance. The draft of the report has been discussed and summarized by the members of the specialists group for demonstration of shielding safety by analysis, committee for safety research on nuclear facilities. (author)

  15. Study on shielding design method of radiation streaming in a tokamak-type DT fusion reactor based on Monte Carlo calculation

    In tokamak-type DT nuclear fusion reactor, there are various type slits and ducts in the blanket and the vacuum vessel. The helium production in the rewelding location of the blanket and the vacuum vessel, the nuclear properties in the super-conductive TF coil, e.g. the nuclear heating rate in the coil winding pack, are enhanced by the radiation streaming through the slits and ducts, and they are critical concern in the shielding design. The decay gamma ray dose rate around the duct penetrating the blanket and the vacuum vessel is also enhanced by the radiation streaming through the duct, and they are also critical concern from the view point of the human access to the cryostat during maintenance. In order to evaluate these nuclear properties with good accuracy, three dimensional Monte Carlo calculation is required but requires long calculation time. Therefore, the development of the effective simple design evaluation method for radiation streaming is substantially important. This study aims to establish the systematic evaluation method for the nuclear properties of the blanket, the vacuum vessel and the Toroidal Field (TF) coil taking into account the radiation streaming through various types of slits and ducts, based on three dimensional Monte Carlo calculation using the MNCP code, and for the decay gamma ray dose rates penetrated around the ducts. The present thesis describes three topics in five chapters as follows; 1) In Chapter 2, the results calculated by the Monte Carlo code, MCNP, are compared with those by the Sn code, DOT3.5, for the radiation streaming in the tokamak-type nuclear fusion reactor, for validating the results of the Sn calculation. From this comparison, the uncertainties of the Sn calculation results coming from the ray-effect and the effect due to approximation of the geometry are investigated whether the two dimensional Sn calculation can be applied instead of the Monte Carlo calculation. Through the study, it can be concluded that the

  16. Design of MP3 Player Based on AT89C51SND1C Microcontroller%AT89C51SND1C单片机的 MP3播放器设计

    徐阳; 徐爱钧

    2015-01-01

    A MP3 player design scheme based on 51 microcontroller is proposed after analyzing the working principle and system compo‐sition of the portable MP3 player .The design uses the AT89C51SND1C microcontroller as the MP3 decoder ,the K9F1208 flash as exter‐nal memory ,the CS4330 as the sound playing circuit .The music file can be downloaded from PC through USB interface of the player .The scheme is simple ,cost‐effective ,low power consumption and easy to expand .%首先分析了便携式M P3播放器的工作原理及其系统构成,接着介绍了一种基于51单片机的M P3播放器设计方案。采用AT89C51SND1C单片机,其片内集成了 MP3解码器,使用 K9F1208闪存作为外存储器,放音电路采用CS4330,音乐文件通过播放器上的U SB接口从PC机上直接下载。该方案设计简单,性价比高、低功耗、易扩展。

  17. Shielding integral benchmark archive and database

    SINBAD (Shielding integral benchmark archive and database) is a new electronic database developed to store a variety of radiation shielding benchmark data so that users can easily 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. It has been designed to be able to include data from nuclear reactor shielding, fusion blankets and accelerator shielding experiments. (authors)

  18. Design, fabrication, installation and shielding integrity testing of source storage container for automatic source movement system used in TLD calibration facility

    A state-of-art TLD laboratory has been commissioned in January 2000 at Radiological Safety Division of Indira Gandhi Centre for Atomic Research (IGCAR). The laboratory provides personnel monitoring service to 2000 occupational workers from Indira Gandhi Centre for Atomic Research and Bhabha Atomic Research Centre facilities. The laboratory has been accredited by the Radiation Safety Systems Division (RSSD), Bhabha Atomic Research Centre (BARC) since year 2002. The laboratory has exclusive facility for the calibration of the TLD cards. As apart of accreditation procedure and taking into account of geometry effect, the dose rate at the card position is determined by the accreditation authorities by using graphite chamber (secondary or national standard instrument) and often re estimated by a condenser R meter (M/s Victoreen, Germany) by our laboratory. As per the regulatory requirement, the exposure protocols should be automated. Towards this an automatic source movement system has been augmented in the calibration facility. By using the system, the source will be brought to the irradiation position by pneumatically and exposures will be terminated by counter, timer and triggering system. To accomplish this task a lead container has been designed, fabricated and mounted at the beneath of the calibration table for the storage of source. As per the automation process, a lead container for the source storage has been designed and installed beneath to the Calibration Table. The container was designed to hold a 3Ci 137Cs source, but present activity of the source is 1.2Ci. Hence, the shielding integrity was tested with higher active source (1.7Ci 60Co). The dose rate measured outside on the circumference of the container at the middle of the source is found to be the same as calculated using QAD CGGP calculations. The top plug is so designed to avoid inadvertent upward movement of the source. Though, the shielding was not adequate on top of the top plug, however it does

  19. Neutron shielding for a 252 Cf source

    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 252Cf isotopic neutron source. During calculations a detailed model for the 252Cf and the shield was utilized. To compare the shielding features of water extended polyester, the calculations were also made for the bare 252Cf 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 for Nuclear Thermal Propulsion

    Caffrey, Jarvis A.

    2016-01-01

    Design and analysis of radiation shielding for nuclear thermal propulsion has continued at Marshall Space Flight Center. A set of optimization tools are in development, and strategies for shielding optimization will be discussed. Considerations for the concurrent design of internal and external shielding are likely required for a mass optimal shield design. The task of reducing radiation dose to crew from a nuclear engine is considered to be less challenging than the task of thermal mitigation for cryogenic propellant, especially considering the likely implementation of additional crew shielding for protection from solar particles and cosmic rays. Further consideration is thus made for the thermal effects of radiation absorption in cryogenic propellant. Materials challenges and possible methods of manufacturing are also discussed.

  1. Shielding research at the Hanford Site

    The original three plutonium production reactors (B, D, and F) constructed at the Hanford Site in 1943--1944 had shields consisting of alternate layers of iron and a high-density pressed-wood product called Masonite *. This design was the engineering response to the scientific request for a mixture of iron and hydrogen. The design mix was based on earlier studies using iron and water or iron and paraffin; however, these materials did not have satisfactory structural characteristics. Although the shields performed satisfactorily, the fabrication cost was high. Each piece had to be machined precisely to fit within structural webs, so as not to introduce cracks through the shield. Before 1950, two additional reactors (DR and H) were built using the same shield design. At the request of R.L. Dickeman, an experimental facility was included in the top of the DR Reactor to permit evaluation of shield materials. Concurrent with the measurement of attenuation properties of materials in this facility, a program was undertaken to investigate the structural characteristics of various high-density Portland cement concretes. This research effort continued for over a decade, and led to the use of these concretes in subsequent reactor shields at the Hanford Site and elsewhere with significant savings in construction costs. Completion of the attenuation and structural measurements on the various high-density concretes provided a database that could be used in the design of shields for new reactors. At the Hanford Site, the top shield of the C Reactor was constructed of concrete, whereas the sides were constructed of iron-Masonite. As more and more data were acquired, the later rectors, KE, KW, and NPR, had shields of various tested concretes. Using concrete in these shields materially reduced the cost of the facilities. Additionally, studies on heat damage to the masonite resulted in changes that permitted increases in production, while at the same time maintaining shield integrity

  2. Handout on shielding calculation

    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)

  3. The ITER EC H and CD upper launcher: Design, analysis and testing of a bolted joint for the Blanket Shield Module

    Gessner, Robby, E-mail: robby.gessner@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Aiello, Gaetano; Grossetti, Giovanni; Meier, Andreas [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Ronden, Dennis [DIFFER – Dutch Institute for Fundamental Energy Physics, P.O. Box 1207, NL-3430 BE Nieuwegein (Netherlands); Spaeh, Peter; Scherer, Theo; Schreck, Sabine; Strauss, Dirk; Vaccaro, Alessandro [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2013-10-15

    Highlights: ► The BSM of the ECH Launcher is attached to the Launcher Main Frame by a bolted joint. ► The bolts were designed as “captive” in order to avoid their accidental removal from the joint. ► The bolted flange connection using two sets of 15 captive bolts (M22 × 2) placed along the sides. ► The captive bolt design is based on a concept that uses a dedicated spring ring, a standard spiral spring and a tensioning screw with two threads to secure the bolts in a form-locking stop. -- Abstract: The final design of the structural system for the ITER EC H and CD upper launcher is in progress. Many design features of the preliminary design are under revision with the aim to achieve the built-to-print-status. This paper deals with design and analysis of a bolted joint for the Blanket Shield Module with special perspective on Remote Handling capability. The BSM of the ECH Launcher is attached to the Launcher Main Frame by a bolted joint conceived so that in the Hot Cell Facility, RH maintenance can be performed on internal components. The joint must be capable to resist very high Electro-Magnetic loads from disruptions, while it has to sustain substantial thermal cycling during operation. Thus the need for a rigid and reliable design is essential. Beside the set of pre-stressed bolts the flanges were therefore equipped with additional shear keys to divert radial moments away from the bolts. Main focus of the work performed was the mechanical design of the joint and the assessment of the structural integrity with respect to the loads applied and its capability for maintenance by RH procedures. To fulfill a major aspect of the RH requirements, the bolts were designed as “captive” in order to avoid their accidental removal from the joint. The captive bolt design is based on a concept that uses a dedicated spring ring, a standard spiral spring and a tensioning screw with two threads to secure the bolts in a form-locking stop. The final approval phase of

  4. Development and application of high performance liquid shielding materials

    We tried to manufacture gel shielding materials of 1mm width with good shielding performance for neutron and γ-rays using five kinds of monomers with high hydrogen density such as long chain fatty acid acrylate, isodecyl methacrylate, lauryl acrylate, stearyle acrylate and stearyle methacrylate, and then lead borate, lead nitrate and lead as lead compounds, and boric acid as neutron adsorbed materials. Some kinds of shielding materials were produced by experiments. Lamination of shielding materials with practical width was obtained. One of three kinds of high performance shielding materials in the fiscal year 1996 was selected. The compensated shielding was designed for double refracted cylindrical duct using one. (S.Y.)

  5. Enhanced Whipple Shield

    Crews, Jeanne L. (Inventor); Christiansen, Eric L. (Inventor); Williamsen, Joel E. (Inventor); Robinson, Jennifer R. (Inventor); Nolen, Angela M. (Inventor)

    1997-01-01

    A hypervelocity impact (HVI) Whipple Shield and a method for shielding a wall from penetration by high velocity particle impacts where the Whipple Shield is comprised of spaced apart inner and outer metal sheets or walls with an intermediate cloth barrier arrangement comprised of ceramic cloth and high strength cloth which are interrelated by ballistic formulae.

  6. Electromagnetically shielded building

    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

  7. ANS shielding standards for light-water reactors

    The purpose of the American Nuclear Society Standards Subcommittee, ANS-6, Radiation Protection and Shielding, is to develop standards for radiation protection and shield design, to provide shielding information to other standards-writing groups, and to develop standard reference shielding data and test problems. A total of seven published ANS-6 standards are now current. Additional projects of the subcommittee, now composed of nine working groups, include: standard reference data for multigroup cross sections, gamma-ray absorption coefficients and buildup factors, additional benchwork problems for shielding problems and energy spectrum unfolding, power plant zoning design for normal and accident conditions, process radiation monitors, and design for postaccident radiological conditions

  8. Electromagnetic shielding formulae

    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)

  9. Rotating shielded crane system

    Commander, John C.

    1988-01-01

    A rotating, radiation shielded crane system for use in a high radiation test cell, comprises a radiation shielding wall, a cylindrical ceiling made of radiation shielding material and a rotatable crane disposed above the ceiling. The ceiling rests on an annular ledge intergrally attached to the inner surface of the shielding wall. Removable plugs in the ceiling provide access for the crane from the top of the ceiling into the test cell. A seal is provided at the interface between the inner surface of the shielding wall and the ceiling.

  10. Normalization of shielding structure quality and the method of its studying

    Method for evaluation of nuclear facility radiation shield quality is suggested. Indexes of shielding structure radiation efficiency and face efficiency are used as the shielding structure quality indexes. The first index is connected with radiation dose rate during personnel irradiation behind the shield, and the second one - with the stresses in shielding structure introduction of the indexes presented allows to evaluate objectively the quality of nuclear facility shielding structure quality design construction and operation and to economize labour and material resources

  11. Iron shielded MRI optimization

    Borghi, C. A.; Fabbri, M.

    1998-09-01

    The design of the main current systems of an actively shielded and of an iron shielded MRI device for nuclear resonance imaging, is considered. The model for the analysis of the magnetic induction produced by the current system, is based on the combination of a Boundary Element technique and of the integration of two Fredholm integral equations of the first and the second kind. The equivalent current magnetization model is used for the calculation of the magnetization produced by the iron shield. High field uniformity in a spherical region inside the device, and a low stray field in the neighborhood of the device are required. In order to meet the design requirements a multi-objective global minimization problem is solved. The minimization method is based on the combination of the filled function technique and the (1+1) evolution strategy algorithm. The multi-objective problem is treated by means of a penalty method. The actively shielded MRI system results to utilize larger amount of conductor and produce higher magnetic energy than the iron shield device. On veut étudier le projet du système des courants principaux d'un MRI à écran en fer et d'un MRI à écran actif. Le modèle d'analyse du champ magnétique produit par le système de courants est basé sur la combinaison d'une technique Boundary Element et de l'intégration de deux équations intégrales de Fredholm de première et de seconde sorte. On utilise pour calculer la magnétisation produite par l'écran en fer le modèle à cou rants de magné ti sa tion équivalents. On exige une élévation uniforme du champ dans une région sphérique au cœur de l'appareil et un bas champ magnétique dispersé à proximité de l'appareil. Dans le but de répondre aux impératifs du projet, on va résoudre un problème multiobjectif de minimisation globale. On utilise une technique de minimisation obtenue par la combinaison des méthodes “Filled Function” et “(1+1) Evolution Strategy”. Le probl

  12. Proceedings of a meeting on radiation shielding and related topics

    This is a proceedings of a meeting on radiation shielding and related topics held on Feb. 22 and 23 in 1978 at Nuclear Engineering Research Laboratory of University of Tokyo. The reports includes the following items (1) studies on neutronics with accelerators (2) radiation damage (3) shielding design (4) radiation streaming (5) shielding experiments from a point of view of radiation measurements (6) shielding benchmark experiments (7) prospects on the study of neutronics. All items are written in Japanese. (auth.)

  13. Scale-4 and related modular systems for the evaluation of nuclear facilities and package design featuring criticality, shielding and transfer capabilities

    Nuclear industry, licensing and regulatory authorities need to be able to rely on good performance of computer codes and nuclear data used in calculations for design and operation of nuclear energy facilities. Given the international impact of a major nuclear accident, and the current crisis in public confidence, it is equally important that the methods, programs and data issued should be internationally accepted. The SCALE modular system has been developed and its capabilities extended during the last 15 years. The driving idea behind its development is that it should contain well established computer codes and data libraries, have an user friendly input format, combine and automate analyses requiring multiple computer codes or calculations into standard analytic sequences and to be well documented and publicly available. The fifth version called SCALE-4 has now been released through the Radiation Shielding Information Center (RSIC) to the OECD/NEA Data Bank. SCALE is now used worldwide. The NEA Data Bank alone has distributed more than one hundred copies of the different versions. The OECD/NEA Data Bank has been asked by its international management committee to hold a seminar with the purpose of exchanging information on the latest developments and experiences among code authors and users, to ensure that users have a correct understanding as to how SCALE should be used to model different problems, and to issue recommendations for further development and benchmarking

  14. Technical specifications for the bulk shielding reactor

    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

  15. ITER blanket, shield and material data base

    As part of the summary of the Conceptual Design Activities (CDA) for the International Thermonuclear Experimental Reactor (ITER), this document describes the ITER blanket, shield, and material data base. Part A, ''ITER Blanket and Shield Conceptual Design'', discusses the need for ITER of a tritium breeding blanket to supply most of the tritium for the fuel cycle of the device. Blanket and shield combined must be designed to operate at a neutron wall loading of 1MW/m2, and to provide adequate shielding of the magnets to meet the neutron energy fluence goal of 3MWa/m2 at the first wall. After a summary of the conceptual design, the following topics are elaborated upon: (1) function, design requirement, and critical issues; (2) material selection; (3) blanket and shield segmentation; (4) blanket design description; (5) design analysis; (6) shield; (7) radiation streaming analysis; and (8) a summary of benchmark calculations. Part B, ''ITER Materials Evaluation and Data Base'', treats the compilation and assessment of the available materials data base used for the selection of the appropriate materials for all major components of ITER, including (i) structural materials for the first wall, (ii) Tritium breeding materials for the blanket, (iii) plasma facing materials for the divertor and first wall armor, and (4) electric insulators for use in the blanket and divertor. Refs, figs and tabs

  16. Analysis on the shielding ability of a hot cell to accommodate advanced spent fuel conditioning process

    A design work is conducting for the IMEF's future cell which located in the basement to use it as a demonstration facility for Advanced Spent Fuel Conditioning Process (ACP). Since the total radiation source which used in ACP is expected as approximately 10 times higher than the design criteria of IMEF, the existing concrete structure cannot meet the shielding requirements. Therefore, shielding design which reinforcing the shielding capability has carried out for the ACP hot cell to satisfy the shielding criteria for the expected maximum radioactivity of ACP. This study presents a shielding analysis results using QADS code for the reinforced shielding wall with heavy concrete, steel or lead, etc. As a results of the analysis, a shielding wall reinforcing method was proposed. Additional shielding analysis was performed for the ACP hot cell with proposed reinforced shielding design using MCNP-4C code, and the validity of radiation shielding design was evaluated

  17. A study on the shielding mechanisms of SOI pixel detector

    Lu, Yunpeng; Wu, Zhigang; Ouyang, Qun; Arai, Yasuo

    2015-01-01

    In order to tackle the charge injection issue that had perplexed the counting type SOI pixel for years, two successive chips CPIXTEG3 and CPIXTEG3b were developed utilizing two shielding mechanisms, Nested-well and Double-SOI, in the LAPIS process. A TCAD simulation showed the shielding effectiveness influenced by the high sheet resistance of shielding layers. Test structures specially designed to measure the crosstalk associated to charge injection were implemented in CPIXTEG3/3b. Measurement results proved that using shielding layer is indispensable for counting type pixel and Double-SOI is superior to Nested-well in terms of shielding effectiveness and design flexibility.

  18. Design of a PET/CT facility considering the shielding calculation in accordance with AAPM TG-108; Diseno de una instalacion PET/CT considerando el calculo de blindaje segun AAPM TG-108

    Guevara R, V. Y.; Romero C, N. [Empresa QC DOSE S. A. C., Av. Tomas Marsano 1915, Surquillo, Lima 34 (Peru); Berrocal T, M., E-mail: vguevara@qcdose.com [Universidad Nacional Mayor de San Marcos, C. German Amezaga 375, Edif. Jorge Basadre, Ciudad Universitaria, Lima 1 (Peru)

    2014-08-15

    A Positron Emission Tomography / Computed Tomography facility may require protection barriers on floor, ceiling and walls, because the patient becomes a radioactive source that emits photons of 0.511 MeV, after having received a radiopharmaceutical, usually F-18 fluorodeoxyglucose (F-18 FDG). This work has as objective to propose the design of a PET/CT facility, taking into account technical and radiation protection considerations applied internationally, and also develop the necessary shielding for such installation by applying as published by the American Association of Physicists in Medicine Task Group Report 108. A shielding spreadsheet in Excel program was developed with reference to the recommendations of the AAPM TG - 08, to determine the shielding required for the walls, floor and ceiling. For fixing the radiation levels in the shielding calculation has been considered the actual restrictions for the occupationally exposed personnel (100 μSv/week) as well as the people in general (20 μSv/ week). The radiopharmaceutical used as a reference for the shielding calculation was the F-18 FDG. With the assistance of an architectural plan were determined distances from potential sources of radiation in facility (uptake and image acquisition living rooms) to points of interest around them. Finally the thickness of the protective barriers in lead and concrete necessary to achieve the established radiation levels were calculated and these results were stored in a table. This paper shows that technical aspects considered in the design of the installation and environments distribution can improve work processes within the PET/CT facility, consequently resulting in a reduction of the dose levels for people in general. (author)

  19. 某医院介入诊疗机房的屏蔽设计及防护效果验证%Verification on shielding design and radiation protection effect of interventional diagnosis and therapy machine room in a hospital

    程晋鹏

    2013-01-01

    [Objective] To discuss the method for verifying the radiation protection effect of shielding design of the interventional diagnosis and therapy machine room,provide an access to put completion acceptance of radiation protective facilities into effect.[Methods] The shielding design of an interventional diagnosis and therapy machine room was verified and assessed by field investigation,theoretical estimate and sites measurement.[Results]The shielding material was reasonable for to this project.The radiation exposure levels of 21 monitoring sites around the room ranged between 0.02 and 0.20 μGy/h.However,the shielding design values of the walls were higher than theoretical count value of the same walls by 2-3 mm lead equivalent.[Conclusion] The total radiation protection effect of shielding design of the room is good.The combination of the above methods can effectively verify the radiation protection effect of shielding design.%目的 探讨介入诊疗机房屏蔽设计效果的验证办法,以利于放射防护设施竣工验收工作的实施.方法 通过现场调查、理论估算、现场检测等对某介入诊疗机房的屏蔽设计进行核实和评价.结果 屏蔽材料选用合理.机房周围21个检测位置的辐射水平均在0.02~0.20 μGy/h范围内.但机房四周墙体的屏蔽设计值均高出理论计算值约二三毫米铅当量.结论 该机房的屏蔽设计总体防护效果良好,上述几种方法的综合应用可较好地核实介入诊疗项目屏蔽设计,验证其防护效果.

  20. Under the Rape Shield

    Roman, Denise

    2011-01-01

    This article focuses on the Rape Shield Laws and their evolution in the United States, one of the pioneers in this field. The article also discusses constitutional and feminist critiques of present Rape Shield Laws, and ends with a comparative perspective throughout the Anglo-American legal space today. Finally, although the Rape Shield Laws can be approached from a variety of discourses, this article engages specifically with a discourse that intersects legal and feminist analyses.

  1. Accelerator shielding benchmark problems

    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)

  2. Accelerator shielding benchmark problems

    Hirayama, H.; Ban, S.; Nakamura, T. [and others

    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).

  3. Optimal Shielding for Minimum Materials Cost of Mass

    Woolley, Robert D. [PPPL

    2014-08-01

    Material costs dominate some shielding design problems. This is certainly the case for manned nuclear power space applications for which shielding is essential and the cost of launching by rocket from earth is high. In such situations or in those where shielding volume or mass is constrained, it is important to optimize the design. Although trial and error synthesis methods may succeed a more systematic approach is warranted. Design automation may also potentially reduce engineering costs.

  4. 7 CFR 1c.114 - Cooperative research.

    2010-01-01

    ... 7 Agriculture 1 2010-01-01 2010-01-01 false Cooperative research. 1c.114 Section 1c.114 Agriculture Office of the Secretary of Agriculture PROTECTION OF HUMAN SUBJECTS § 1c.114 Cooperative research. Cooperative research projects are those projects covered by this policy which involve more than...

  5. AP600 Shield building

    In order to minimize capital costs and save time in the global construction time schedule for the AP600 Nuclear Power Plant, planned in 36 months from excavation up to the fuel charging, ANSALDO has developed an innovative Shield Building Conical Roof design having the following basic characteristics: i) can be erected approximately in less than two months; ii) allows the functionality of the Passive Containment Cooling System (PCSS) located in the PCCS tank and in the Valve Room anchored directly to the conical roof itself; iii) satisfies the structural loads design as Safe Shutdown Earthquake, or the Aircraft Crash and both integrated with the sloshing analysis for the tank located at the top of the conical roof. The most important aspects of this new roof are: a) use of prefabricated precast panels; b) address the erection of the formworks using temporary structures having the capability of becoming final elements; c) develop a modular rebars sizing and design in order to perform the most important portion of the job in the workshop; d) second pouring construction sequence assuring full integration with the formwork function; e) modular construction of the PCSS tank at the top of the conical roof. An interesting evaluation has been also performed in calculating sloshing phenomenon in the PCSS tank by comparing detailed 3D Finite Element Model approach and simplified qualified formulas dedicated to this phenomenon. (author). 2 figs

  6. Optimization Design of a 5kW Lift Type Vertical Axis Wind Turbine With Wind Shield-Growth Patterns%5kW遮蔽-增速升力型垂直轴风力机优化设计

    姬俊峰; 邓召义; 蒋磊; 黄典贵

    2012-01-01

    本文详细介绍了5 kW遮蔽-增速垂直轴风力机的结构特点及主要参数。利用正交优化设计方法,采用计算流体力学软件,针对5 kW风力机,在叶片个数和遮蔽板安装位置半径一定的情况下,对翼型弦长、叶片转动扫掠面的半径、风轮旋转速度、遮蔽-增速板个数、遮蔽-增速板与叶片间的气动间隙以及遮蔽-增速板的安装角六个参数进行优化计算,找出一组最佳设计参数,进而设计出5 kW遮蔽-增速升力型垂直轴风力机,并对设计出的有遮蔽板与无遮蔽板两类型风力机的变工况特性进行比较分析。%This paper is devoted to a detailed study of the characteristics of the structure and most important aerodynamic design parameters of a 5 kW lift type vertical axis wind turbine with wind shield-growth patterns.As the number of turbine blades and the installation location of wind shield-growth patterns have been previously set,the computational fluid dynamics code was adopted in this study to obtain the numerical results and then six parameters,including blade airfoil chord length,radius of turbine rotor,rotational speed of the wind turbine,number of wind shield-growth patterns,clearance between the turbine blades and surrounding wind shield-growth patterns and installation angle of wind shield-growth patterns,were all optimized by using an orthogonal optimization algorithm.Based on the optimized parameters obtained,a 5 kW lift type vertical axis wind turbine can be designed.In addition,the performance characteristics of the designed wind turbine with and without wind shield-growth patterns were analyzed and compared under variable wind conditions.

  7. 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

  8. Rotating shielded crane system

    A rotating, radiation-shielded crane system is described comprising: a generally cylindrical, radiation-shielding wall, the top of the wall forming a first annular ledge; a second annular ledge integrally attached to the inner surface of the shielding wall; a generally cylindrical ceiling made of radiation shielding material, the ceiling including a flange portion on the top thereof and a body portion, the flange portion associated with the second annular ledge such that the ceiling is supported thereby, the volume inside the wall and the ceiling forming a test cell; a rotatable crane disposed above the ceiling such that the crane is outside of the test cell; removable access means in the ceiling for allowing the crane to access the inside of the test cell from the top of the ceiling; means for sealing the interface between the inner surface of the shielding wall and the ceiling

  9. Led Astray by Hemoglobin A1c

    Chen, Jean; Diesburg-Stanwood, Amy; Bodor, Geza; Rasouli, Neda

    2016-01-01

    Hemoglobin A1c (A1c) is used frequently to diagnose and treat diabetes mellitus. Therefore, it is important be aware of factors that may interfere with the accuracy of A1c measurements. This is a case of a rare hemoglobin variant that falsely elevated a nondiabetic patient’s A1c level and led to a misdiagnosis of diabetes. A 67-year-old male presented to endocrine clinic for further management after he was diagnosed with diabetes based on an elevated A1c of 10.7%, which is approximately equivalent to an average blood glucose of 260 mg/dL. Multiple repeat A1c levels remained >10%, but his home fasting and random glucose monitoring ranged from 92 to 130 mg/dL. Hemoglobin electrophoresis and subsequent genetic analysis diagnosed the patient with hemoglobin Wayne, a rare hemoglobin variant. This variant falsely elevates A1c levels when A1c is measured using cation-exchange high-performance liquid chromatography. When the boronate affinity method was applied instead, the patient’s A1c level was actually 4.7%. Though hemoglobin Wayne is clinically silent, this patient was erroneously diagnosed with diabetes and started on an antiglycemic medication. Due to this misdiagnosis, the patient was at risk of escalation in his “diabetes management” and hypoglycemia. Therefore, it is important that providers are aware of factors that may result in hemoglobin A1c inaccuracy including hemoglobin variants. PMID:26848480

  10. Characterization of Magnetic Field Immersed Photomultipliers from Double Chooz Experiment. Design and Construction of their Magnetic Shields; Caracterizacion de los fotomultiplicadores del experimento Double Chooz bajo campo magnetico y diseno y construccion de sus blindajes magneticos

    Valdivia Valero, F. J.

    2007-12-28

    Flavour oscillations of neutrinos are a quantum-mechanical effect widely demonstrated. It is explained through interferences of their mass eigenstates, therefore, belonging to the physical area beyond the Standard Model. This work deals with the CIEMAT collaboration in the neutrino experiment Double Chooz. Such an experiment aims to measure the mixture angle {theta}{sub 1}3, one of the PMNS leptonic mixture matrix, with a un reached-before sensibility by decrease of systematic errors. For this, two identical scintillator detectors, equipped with PMT's, will be sited at different distances from two reactors located in the nuclear power plant CHOOZ B (France). The electronic neutrino flux from these reactors will be compared, explaining its deficit by flavour oscillations of these particles. The identity of both detectors will be diminished by the magnetic field effects on the PMT's response. Therefore, this study serves as for quantifying such an effects as for fitting the magnetic shields design that minimize them. Shielding measurements and final design of magnetic shields as much as the effect these ones cause in the PMT's response immersed in a monitored magnetic field are presented. (Author) 85 refs.

  11. WASTE HANDLING BUILDING SHIELD WALL ANALYSIS

    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

  12. 小型中子源高能中子照相装置准直屏蔽系统设计%Conception Design of Shielding Collimator System for High Energy Neutron Radiography with Minitype Neutron Source

    吴洋; 窦海峰; 唐彬; 霍合勇

    2013-01-01

    Shielding collimator system is necessary in the neutron radiography installation; this issue gives the conception design of shielding collimator system for FNR about high energy neutron source by MCNP.Preliminarily ascertain the material component and dimension,confirm the neutron flux at imaging position,imaging distance,imaging field range of the FNP installation in theory.%准直屏蔽系统是中子照相装置的必要设备.本文采用蒙特卡洛中子输运程序(MCNP)等软件对高能中子准直屏蔽系统进行理论设计,初步确定了其材料构成和外观尺寸,从理论上确定了装置包括成像处注量率、成像距离及相应视场等关键参数.

  13. SEA participation in the design of local shielding the beam dump of IFMIF; Participacion de sea en el diseno del blindaje local del beam dumpo de IFMIF

    Ortega, P.

    2012-07-01

    The beam dump is a very important source of radiation both during the operation and after the throttle stop, so it requires the availability of a local shielding to minimize dose in the neighboring rooms during operation as well as the dose to the teams maintenance during shutdown.

  14. 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.

  15. Radiation shielding device

    Purpose: To lower the shielding cost by providing a shielding wall having cavities and charging spherical shiedling materials in the cavities only when the shielding is required. Constitution: The structure comprises two parallel steel side plates aparting from each other to form a space therebetween and reinforcements such as H-type steels vertically provided between the side plates. The upper and the lower ends of the reinforcements are aparted from the upper and the lower edges of the side plates by a predetermined distance to form lateral passage between the top plate and the bottom plate. A guide plate having a plurality of openings is mounted on the upper ends of the reinforcements. If it is required for the structure to serve as the shield, spherical radioactive shielding materials are supplied through an injection port onto the guide plate while opening the injection port is opened and closing discharge port. The spherical radioactive shielding materials are fallen through the openings and filled in the space to thereby providing the structure with shielding performance. (Yoshino, Y.)

  16. The Role of Metformin Response in Lipid Metabolism in Patients with Recent-Onset Type 2 Diabetes: HbA1c Level as a Criterion for Designating Patients as Responders or Nonresponders to Metformin.

    Zahra Kashi

    Full Text Available In this study, we investigated whether response to metformin, the most frequently drug for diabetes treatment, influences the therapeutic effects of antilipidemic medication in newly diagnosed patients with type 2 diabetes mellitus (T2DM.A total of 150 patients with T2DM were classified into two groups following 3 months of metformin therapy (1000 mg twice daily: responders (patients showing ≥1% reduction in HbA1c from baseline and nonresponders (patients showing <1% reduction in HbA1c from baseline. The patients received atorvastatin 20 mg, gemfibrozil 300 mg, or atorvastatin 20 mg and gemfibrozil 300 mg daily.HbA1c and fasting glucose levels were significantly different between baseline and 3 months among responders receiving atorvastatin; however, these differences were not statistically significant in nonresponders. Atherogenic ratios of low-density lipoprotein cholesterol to high-density lipoprotein cholesterol (LDL-C/HDL-C; p = 0.002, total cholesterol to HDL-C (TC/HDL-C; p<0.001 and AIP (the atherogenic index of plasma; p = 0.004 decreased significantly in responders receiving atorvastatin than in nonresponders. Moreover, responders receiving atorvastatin showed a significant increase in HDL-C levels but nonresponders receiving atorvastatin did not (p = 0.007. The multivariate model identified a significant association between metformin response (as the independent variable and TG, TC, HDL-C and LDL-C (dependent variables; Wilk's λ = 0.927, p = 0.036.Metformin response affects therapeutic outcomes of atorvastatin on atherogenic lipid markers in patients newly diagnosed with T2DM. Metformin has a greater impact on BMI in responders of metformin compared to nonresponders. Adoption of better therapeutic strategies for reducing atherogenic lipid markers may be necessary for metformin nonresponders.

  17. TPX remote maintenance and shielding

    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

  18. TPX remote maintenance and shielding

    The Tokamak Physics Experiment (TPX) 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 open-quotes hands-onclose quotes 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 Plasma Facing Components (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

  19. Shielding member for thermonuclear device

    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.)

  20. Passive Magnetic Shielding in Gradient Fields

    Bidinosti, C P

    2013-01-01

    The effect of passive magnetic shielding on dc magnetic field gradients imposed by both external and internal sources is studied. It is found that for concentric cylindrical or spherical shells of high permeability material, higher order multipoles in the magnetic field are shielded progressively better, by a factor related to the order of the multipole. In regard to the design of internal coil systems for the generation of uniform internal fields, we show how one can take advantage of the coupling of the coils to the innermost magnetic shield to further optimize the uniformity of the field. These results demonstrate quantitatively a phenomenon that was previously well-known qualitatively: that the resultant magnetic field within a passively magnetically shielded region can be much more uniform than the applied magnetic field itself. Furthermore we provide formulae relevant to active magnetic compensation systems which attempt to stabilize the interior fields by sensing and cancelling the exterior fields clos...

  1. 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

  2. Consolidated fuel shielding calculations

    Irradiated fuel radiation dose rate and radiation shielding requirements are calculated using a validated ISOSHLD-II model. Comparisons are made to experimental measurements. ISOSHLD-11 calculations are documented

  3. Radiation shielding curtain

    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

  4. 14-MeV neutron streaming through shield gaps

    Monte Carlo calculations have been performed to determine the neutron streaming through straight and single-bend gaps for three different shield thicknesses. A uniform plane source emitting 14-MeV neutrons with a cosine angular distribution was used in the analyses. The results obtained are discussed in terms of how they might be used in the early stages of a shield design to obtain approximate solutions to design questions. These results have direct implications regarding neutron-streaming problems that will be encountered in the shielding analyses of tokamak fusion reactors which are constructed from pie-shaped shield/vacuum chamber segments

  5. 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...

  6. Steam generator hand hole shielding.

    Cox, W E

    2000-05-01

    Seabrook Station is an 1198 MWE Pressurized Water Reactor (PWR) that began commercial operation in 1990. Expensive and dose intensive Steam Generator Replacement Projects among PWR operators have led to an increase in steam generator preventative maintenance. Most of this preventative maintenance is performed through access ports in the shell of the steam generator just above the tube sheet known as secondary side hand holes. Secondary side work activities performed through the hand holes are typically performed without the shielding benefit of water in the secondary side of the steam generator. An increase in cleaning and inspection work scope has led to an increase in dose attributed to steam generator secondary side maintenance. This increased work scope and the station goal of maintaining personnel radiation dose ALARA led to the development of the shielding concept described in this article. This shield design saved an estimated 2.5 person-rem (25 person-Smv) the first time it was deployed and is expected to save an additional 50 person-rem (500 person-mSv) over the remaining life of the plant. PMID:10770158

  7. Shielding Development for Nuclear Thermal Propulsion

    Caffrey, Jarvis A.; Gomez, Carlos F.; Scharber, Luke L.

    2015-01-01

    Radiation shielding analysis and development for the Nuclear Cryogenic Propulsion Stage (NCPS) effort is currently in progress and preliminary results have enabled consideration for critical interfaces in the reactor and propulsion stage systems. Early analyses have highlighted a number of engineering constraints, challenges, and possible mitigating solutions. Performance constraints include permissible crew dose rates (shared with expected cosmic ray dose), radiation heating flux into cryogenic propellant, and material radiation damage in critical components. Design strategies in staging can serve to reduce radiation scatter and enhance the effectiveness of inherent shielding within the spacecraft while minimizing the required mass of shielding in the reactor system. Within the reactor system, shield design is further constrained by the need for active cooling with minimal radiation streaming through flow channels. Material selection and thermal design must maximize the reliability of the shield to survive the extreme environment through a long duration mission with multiple engine restarts. A discussion of these challenges and relevant design strategies are provided for the mitigation of radiation in nuclear thermal propulsion.

  8. Structural Monitoring of Metro Infrastructure during Shield Tunneling Construction

    L. Ran

    2014-01-01

    Full Text Available Shield tunneling construction of metro infrastructure will continuously disturb the soils. The ground surface will be subjected to uplift or subsidence due to the deep excavation and the extrusion and consolidation of the soils. Implementation of the simultaneous monitoring with the shield tunnel construction will provide an effective reference in controlling the shield driving, while how to design and implement a safe, economic, and effective structural monitoring system for metro infrastructure is of great importance and necessity. This paper presents the general architecture of the shield construction of metro tunnels as well as the procedure of the artificial ground freezing construction of the metro-tunnel cross-passages. The design principles for metro infrastructure monitoring of the shield tunnel intervals in the Hangzhou Metro Line 1 are introduced. The detailed monitoring items and the specified alarming indices for construction monitoring of the shield tunneling are addressed, and the measured settlement variations at different monitoring locations are also presented.

  9. Use of MCNP in fusion blanket design ITER magnet system shielding analysis benchmark of the EFF (European Fusion File) neutron data with the FNG (Frascati Neutron Generator) 14 MeV neutron facility

    Since eight years at our laboratory, MCNP code has been used as a fundamental tool in many fusion directed activities in which we have been or we still are involved. Mainly they are: neutronics analysis of the performances of blanket components, supporting and optimizing their design; the estimation of the nuclear heat and radiation loads on the toroidal superconducting coils to assess the system shielding performances; then, a 14 MeV neutron generator is recently operating in Frascati and an experimental programme started with a benchmark neutron transport in a stainless steel block, MCNP is used to perform calculations. Present status of these experiments are reviewed. (K.A.)

  10. Inner Shielding of the COMET Cosmic Veto System

    Markin, Oleg

    2015-01-01

    A simulation of neutrons traversing a shield beneath the COMET scintillator strip cosmic-veto counter is accomplished using the Geant4 toolkit. A Geant4 application is written with an appropriate detector construction and a possible spectrum of neutron's energy. The response of scintillator strips to neutrons is studied in detail. A design of the shield is optimized to ensure the time loss concerned with fake veto signals caused by neutrons from muon captures is tolerable. Materials of shield...

  11. A study on the shielding mechanisms of SOI pixel detector

    Lu, Yunpeng; Liu, Yi; Wu, Zhigang; Ouyang, Qun; Arai, Yasuo

    2015-01-01

    In order to tackle the charge injection issue that had perplexed the counting type SOI pixel for years, two successive chips CPIXTEG3 and CPIXTEG3b were developed utilizing two shielding mechanisms, Nested-well and Double-SOI, in the LAPIS process. A TCAD simulation showed the shielding effectiveness influenced by the high sheet resistance of shielding layers. Test structures specially designed to measure the crosstalk associated to charge injection were implemented in CPIXTEG3/3b. Measuremen...

  12. Radiation shielding effectiveness of newly developed superconductors

    Gamma ray shielding effectiveness of superconductors with a high mass density has been investigated. We calculated the mass attenuation coefficients, the mean free path (mfp) and the exposure buildup factor (EBF). The gamma ray EBF was computed using the Geometric Progression (G-P) fitting method at energies 0.015–15 MeV, and for penetration depths up to 40 mfp. The fast-neutron shielding effectiveness has been characterized by the effective neutron removal cross-section of the superconductors. It is shown that CaPtSi3, CaIrSi3, and Bi2Sr2Ca1Cu2O8.2 are superior shielding materials for gamma rays and Tl0.6Rb0.4Fe1.67Se2 for fast neutrons. The present work should be useful in various applications of superconductors in fusion engineering and design. - Highlights: • Radiation shielding properties of superconductors were investigated. • µ/ρ, mean free path, and exposure buildup factor were calculated. • CaPtSi3, CaIrSi3, and Bi2Sr2Ca1Cu2O8.2 were found superior for γ-ray shielding. • Tl0.6Rb0.4Fe1.67Se2 was found superior for fast neutron shielding

  13. Shielding Effectiveness of Composites Containing Flaky Inclusions

    WANG Qingguo; QU Zhaoming; WANG Yilong

    2013-01-01

    To investigate the quantitative relationship between the electromagnetic-shielding property of composites and the distribution of inclusions,a scheme for predicting the shielding effectiveness of composites containing variously-distributed flaky inclusions is proposed.The scheme is based on equivalent parameters of homogeneous comparison materials and the plane-wave shielding theory.It leads to explicit formulas for the shielding effectiveness of multi-layered composites in terms of microstructural parameters that characterize the shape,distribution and orientation of the inclusions.For single layer composite that contains random and aligned flaky silver-coated carbonyl-iron particles with fractions of different volume,the predicted shielding effectiveness agrees well with the experimental data.As for composites containing aligned flaky particles,the shielding effectiveness obtained by the proposed scheme and experiment data is higher than that the random case,e.g.about 20 dB higher at 750 MHz.The proposed scheme is a straightforward method for optimizing future composite designs.

  14. A1C Test and Diabetes

    ... of Diabetes Educators American Diabetes Association JDRF MedlinePlus Diabetes Disease Organizations ​There are many organizations who provide ... KB). Alternate Language URL The A1C Test and Diabetes Page Content On this page: What is the ...

  15. GCFR radial blanket and shield experiment: objectives, preanalysis, and specifications

    An integral experiment has been designed for the verification of radiation transport methods and nuclear data used for the design of the radial shield for the proposed 300 MW(e) gas-cooled fast breeder reactor (GCFR). The scope of the experiment was chosen to include a thorium oxide radial blanket mockup as well as several shield configurations in order to reduce the uncertainties in the calculated source terms for the radial shield, and to reduce the uncertainties in the calculated radiation damage to the prestressed concrete reactor vessel (PCRV). Additionally, the measurements are intended to bound the uncertainties in calculated gamma-ray heating rates within the blanket and shield. Although designed specifically for the GCFR, the experiment will provide generic data regarding deep penetration in ThO2 and common shield materials, which should also benefit LMFBR designers

  16. Neutron shielding material

    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)

  17. Innovative Neutron Shielding Material Composing of Natural Rubber-Styrene Butadiene Rubber Blends Diboron Trioxide

    The optimized flexible and lightweight neutron shielding materials were designed using the Monte Carlo N-Particle (MCNP) code. The neutron shielding materials thickness for 10 mm and 100 mm were considered for the neutron shielding performance. Results indicated that 10 mm shielding material of rubber compound between natural rubber (NR) and styrene butadiene rubber (SBR) blend (1:1) compose with diboron trioxide 60 phr and 100 mm shielding material for 4 layers structure of rubber compound between natural rubber (NR) and styrene butadiene rubber (SBR) blend compose with diboron trioxide 10 phr and rubber compound with iron oxide 100 phr offered advantages on neutron shielding compared to other designs. Experimental for these shielding material results verified the correctness of the optimal design and fabrication. The designed shielding materials are highly suitable for applications in nuclear science and technology.

  18. Small foamed polystyrene shield protects low-frequency microphones from wind noise

    Tedrick, R. N.

    1964-01-01

    A foamed polystyrene noise shield for microphones has been designed in teardrop shape to minimize air turbulence. The shield slips on and off the microphone head easily and is very effective in low-frequency sound intensity measurements.

  19. Manufacture of blanket shield modules for ITER

    Lorenzetto, P. [EFDA CSU Garching, Boltzmannstr. 2, D-85748 Garching (Germany)]. E-mail: Patrick.Lorenzetto@tech.efda.org; Boireau, B. [AREVA Centre Technique de Framatome, BP181, F-71200 Le Creusot (France); Boudot, C. [AREVA Centre Technique de Framatome, BP181, F-71200 Le Creusot (France); Bucci, P. [CEA, DTEN/S3ME/LMIC, 17 rue des Martyrs, F-38054 Grenoble (France); Furmanek, A. [EFDA CSU Garching, Boltzmannstr. 2, D-85748 Garching (Germany); Ioki, K. [ITER IT, Boltzmannstr. 2, D-85748 Garching (Germany); Liimatainen, J. [Metso Powdermet, P.O. Box 306, FIN-33101 Tampere (Finland); Peacock, A. [EFDA CSU Garching, Boltzmannstr. 2, D-85748 Garching (Germany); Sherlock, P. [NNC Ltd., Booths Hall, Knutsford, Cheshire WA16 8QZ (United Kingdom); Taehtinen, S. [VTT Industrial Systems, P.O. Box 1704, Espoo, FIN-02044 VTT (Finland)

    2005-11-15

    A research and development programme for the ITER blanket shield modules has been implemented in Europe to provide input for the design and the manufacture of the full-scale production components. It involves in particular the fabrication and testing of mock-ups (small scale and medium scale) and full-scale prototypes of shield blocks (SB) and first wall (FW) panels. The manufacturing feasibility of FW panels has been demonstrated for two copper alloy candidates. Two designs have been developed for the manufacture of the SB, one for a conventional fabrication route and one for a fabrication route based on the hot isostatic press technology. This paper presents the fabrication routes developed in Europe for the manufacture of the ITER Shield modules.

  20. Radiation shielding of 131I therapeutic ward in department

    Objective: To rebuild an 131I therapeutic ward of the department of nuclear medicine in the hospital, and design the radiation shielding to make the radiation safety achieve the national standards. Methods: According to the protection demands of national relating standards, the design of the ward was based on the basic principles and methods of radiation shielding of 131I, and combined the distribution of radioactive sources, rooms and people. Results: The design parameters and radioprotection data of the rebuilt ward were obtained and the radiation shielding was safe by monitoring. Conclusion: The design of the 131I therapeutic ward achieved the anticipated target that the radiation safety could be controlled. (authors)

  1. Investigation of shielding analysis method for fusion reactors

    An investigation has been made, at the shielding laboratory, into the status of shielding analysis method for fusion reactor based on conceptual designs of a variety of fusion power reactors and fusion experimental facilities, in cooperation with the Fusion Reactor Shielding Working Group in the Research Committee on Fast Neutron Shielding of the Atomic Energy Society of Japan. The reactors and facilities considered are CULHAM MKII(U.K), SPTR (Japan), TFTR(U.S.A.), STARFIRE(U.S.A.) and INTOR-USA(U.S.A.). (author)

  2. Cross Section Evaluation Group shielding benchmark compilation. Volume II

    Rose, P.F.; Roussin, R.W.

    1983-12-01

    At the time of the release of ENDF/B-IV in 1974, the Shielding Subcommittee had identified a series of 12 shielding data testing benchmarks (the SDT series). Most were used in the ENDF/B-IV data testing effort. A new concept and series was begun in the interim, the so-called Shielding Benchmark (SB) series. An effort was made to upgrade the SDT series as far as possible and to add new SB benchmarks. In order to be designated in the SB class, both an experiment and analysis must have been performed. The current recommended benchmark for Shielding Data Testing are listed. Until recently, the philosophy has been to include only citations to published references for shielding benchmarks. It is now our intention to provide adequate information in this volume for proper analysis of any new benchmarks added to the collection. These compilations appear in Section II, with the SB5 Fusion Reactor Shielding Benchmark as the first entry.

  3. Radiation shielding bricks

    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)

  4. Subsurface Shielding Source Term Specification Calculation

    The purpose of this calculation is to establish appropriate and defensible waste-package radiation source terms for use in repository subsurface shielding design. This calculation supports the shielding design for the waste emplacement and retrieval system, and subsurface facility system. The objective is to identify the limiting waste package and specify its associated source terms including source strengths and energy spectra. Consistent with the Technical Work Plan for Subsurface Design Section FY 01 Work Activities (CRWMS M and O 2001, p. 15), the scope of work includes the following: (1) Review source terms generated by the Waste Package Department (WPD) for various waste forms and waste package types, and compile them for shielding-specific applications. (2) Determine acceptable waste package specific source terms for use in subsurface shielding design, using a reasonable and defensible methodology that is not unduly conservative. This calculation is associated with the engineering and design activity for the waste emplacement and retrieval system, and subsurface facility system. The technical work plan for this calculation is provided in CRWMS M and O 2001. Development and performance of this calculation conforms to the procedure, AP-3.12Q, Calculations

  5. Onboard radiation shielding estimates for interplanetary manned missions

    The main focus of space related shielding design is to protect operating systems, personnel and key structural components from outer space and onboard radiation. This paper summarizes the feasibility of a lightweight neutron radiation shield design for a nuclear powered, manned space vehicle. The Monte Carlo code MCNP5 is used to determine radiation transport characteristics of the different materials and find the optimized shield configuration. A phantom torso encased in air is used to determine a dose rate for a crew member on the ship. Calculation results indicate that onboard shield against neutron radiation coming from nuclear engine can be achieved with very little addition of weight to the space vehicle. The selection of materials and neutron transport analysis as presented in this paper are useful starting data to design shield against neutrons generated when high-energy particles from outer space interact with matter on the space vehicle. (authors)

  6. Homogeneity test on heavy concrete shield wall for ACP facility

    The hot cell facility for research activities related to the electrolytic reduction of spent fuel, which is designed to permit a safe handling of radioactive materials up to 1,385 TBq, is scheduled to be constructed in 2005. The design features of the radiation safety are reviewed for the shield wall, rear door, shielding window, penetrations, toboggan, and the storage vault. The calculations by QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts and the gamma scanning test is described to examine the integrity of the shielding structure for the hot cell. The gamma scanning test is especially good at detecting any void and cracks in a heavy concrete wall and finding crevices between the wall and the devices frames. The shielding effectiveness and homogeneity of the hot cell wall, shield window, rear door etc., shall be measured by reading the activity level of the radiation

  7. Radiation shielding for 250 MeV protons

    This paper is targetted at personnel who have the responsibility of designing the radiation shielding against neutron fluences created when protons interact with matter. Shielding of walls and roofs are discussed, as well as neutron dose leakage through labyrinths. Experimental data on neutron flux attenuation are considered, as well as some calculations using the intranuclear cascade calculations and parameterizations

  8. Where have the neutrons gone: A history of the Tower Shielding Facility

    In the early 1950's, the concept of the unit shield for the nuclear powered aircraft reactor changed to one of the divided shield concept where the reactor and crew compartment shared the shielding load. Design calculations for the divided shield were being made based on data obtained in studies for the, unit shield. It was believed that these divided shield designs were subject to error, the magnitude of which could not be estimated. This belief led to the design of the Tower Shielding Facility where divided-shield-type measurements could be made without interference from ground or structural scattering. This paper discusses that facility, its reactors, and some chosen experiments from the list of many that were performed at that facility during the past 38 years

  9. Radiation shielding materials

    Purpose: To obtain putty-like shielding materials excellent in the radiation shielding and packing workability for use in penetrations of electrical wires or pipeways in a nuclear installation. Constitution: A putty-like material is prepared from 100 parts by weight of a binder comprising a grease or the like having viscosity of greater than 5000 cst or an immiscible consistency of greater than 100 (JIS K 2220 (1980) para. 5.3.4) at 25 0C and from 1200 to 4000 parts by weight of high density inorganic powder such as lead powder or lead oxide powder having a density of greater than 5 g/cm3 and such a particle size that more than 95 % thereof passes through a 145 mesh sieve. The putty-like material is adjusted such that it has 1 - 35 mm of softness (JIS A 5752) at normal temperature, more than 1 g/5 sec of injection amount and a density of greater than 4 g/cm3. In this way, non-curable radiation shielding agent with excellent X-ray or γ-ray shielding property and being capable of packed densely to void portions can be obtained. (Ikeda, J.)

  10. Methods for calculating radiation attenuation in shields

    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

  11. Lightweight Shield Against Space Debris

    Redmon, John W., Jr.; Lawson, Bobby E.; Miller, Andre E.; Cobb, W. E.

    1992-01-01

    Report presents concept for lightweight, deployable shield protecting orbiting spacecraft against meteoroids and debris, and functions as barrier to conductive and radiative losses of heat. Shield made in four segments providing 360 degree coverage of cylindrical space-station module.

  12. Hinged Shields for Machine Tools

    Lallande, J. B.; Poland, W. W.; Tull, S.

    1985-01-01

    Flaps guard against flying chips, but fold away for tool setup. Clear plastic shield in position to intercept flying chips from machine tool and retracted to give operator access to workpiece. Machine shops readily make such shields for own use.

  13. Scale-PC shielding analysis sequences

    The SCALE computational system is a modular code system for analyses of nuclear fuel facility and package designs. With the release of SCALE-PC Version 4.3, the radiation shielding analysis community now has the capability to execute the SCALE shielding analysis sequences contained in the control modules SAS1, SAS2, SAS3, and SAS4 on a MS- DOS personal computer (PC). In addition, SCALE-PC includes two new sequences, QADS and ORIGEN-ARP. The capabilities of each sequence are presented, along with example applications

  14. Radioimmunological determination of hemoglobin Asub(1c)

    The antibodies fighting human haemaglobin A sub(1c) to diagnose diabetes is obtained by immunisation of cats, goats or sheep. The acquisition of antigen antibody complexes in a blood sample is done by radioimmunological determination using antigen labelled with I 125 (several examples). (DG)

  15. A1C Test and Diabetes

    ... Top ] Will the A1C test show changes in blood glucose levels? Large changes in a person’s blood glucose levels ... test that provides information about a person’s average levels of blood glucose, also called blood sugar, over the past 3 ...

  16. Shielding during x-ray examination of pediatric female patients with developmental dysplasia of the hip

    Patients with developmental dysplasia of the hip (DDH) generally undergo multiple x-ray examinations of both hip joints. During these examinations, the gonads are completely exposed to radiation, unless shielded. Although many types and sizes of gonad shields exist, they often do not provide adequate protection because of size and placement issues; additionally, these shields are frequently omitted for female patients. Our aim was to assess gonad protection during x-ray examination that is provided by gonad shields designed for individual female patients with DDH. We retrospectively retrieved data from the Picture Archiving and Communication System database; pelvic plain x-ray films from 766 females, 18 years old or younger, were included in our analysis. Based on x-ray measurements of the anterior superior iliac spine, we developed a system of gonad shield design that depended on the distance between anterior superior iliac spine markers. We custom-made shields and then examined shielding rates and shielding accuracy before and after these new shields became available. Standard (general-purpose) shields were used before our custom design project was implemented. The shielding rate and shielding accuracy were, respectively, 14.5% and 8.4% before the project was implemented and 72.7% and 32.2% after it was implemented. A shield that is more anatomically correct and available in several different sizes may increase the likelihood of gonad protection during pelvic x-ray examinations. (paper)

  17. Research And Design Of A Shielding System For Improving Dose Uniformity Ratio (DUR) And Setting Up A Set Of Operational Instruction For Running The Electron Beam Accelerator UELR-10-15S2

    The project “Research and design of a shielding system for improving dose uniformity ratio and setting up a set of operational instruction for the electron beam accelerator UELR-10-15S2 at Research and Development Center for Radiation Technology - VINAGAMMA” has been set up and implemented to improve dose uniformity ratio (DUR) for products with the areal densities in the range of 4.5 - 7.0 g/cm2 in the irradiation by electron beam of 10 MeV energy and to set up the first set of operational instruction of the accelerator UELR-10-15S2. By using the shielding system with the areal densities of 1.5 g/cm2 and 2.5 g/cm2, the DUR values for products with the above range are reduced lower than 2 for food irradiation and the utilization of the accelerator has been significantly enhanced. The set of operational instruction of the accelerator UELR-10-15S2 has been set up and approved by the center director in order to put it in force. (author)

  18. Calculation system analysis for radiation shielding

    This work consists of the computational system implementation for nuclear reactor shielding analysis. The system has as objectives to facilitate the installation of the calculation framework, problem set-up, and results analysis. Several computational programmes commonly used for cross-section preparation and radiation transport were chosen for the system. This work represents the capacity necessary for nuclear reactor and particle accelerator shielding design, to aid in nuclear experiments and in the utilization of nuclear techniques that require the radiation field calculation. The system was implemented in PC-DOS environment and consists of the necessary and sufficient programs and data for generation of the cross sections, groups constants, self-shielding factors, activation sources, for the calculation of neutron and gamma-ray fluence, dose rates, and other types of response functions. (author). 11 refs., 8 figs

  19. Radiation shielding effectiveness of newly developed superconductors

    Singh, Vishwanath P.; Medhat, M. E.; Badiger, N. M.; Saliqur Rahman, Abu Zayed Mohammad

    2015-01-01

    Gamma ray shielding effectiveness of superconductors with a high mass density has been investigated. We calculated the mass attenuation coefficients, the mean free path (mfp) and the exposure buildup factor (EBF). The gamma ray EBF was computed using the Geometric Progression (G-P) fitting method at energies 0.015-15 MeV, and for penetration depths up to 40 mfp. The fast-neutron shielding effectiveness has been characterized by the effective neutron removal cross-section of the superconductors. It is shown that CaPtSi3, CaIrSi3, and Bi2Sr2Ca1Cu2O8.2 are superior shielding materials for gamma rays and Tl0.6Rb0.4Fe1.67Se2 for fast neutrons. The present work should be useful in various applications of superconductors in fusion engineering and design.

  20. Secondary gamma-ray data for shielding calculation

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

  1. The heterogeneous anti-radiation shield for spacecraft*

    Telegin, S. V.; Draganyuk, O. N.

    2016-04-01

    The paper deals with modeling of elemental composition and properties of heterogeneous layers in multilayered shields to protect spacecraft onboard equipment from radiation emitted by the natural Earth’s radiation belt. This radiation causes malfunctioning of semiconductor elements in electronic equipment and may result in a failure of the spacecraft as a whole. We consider four different shield designs and compare them to the most conventional radiation-protective material for spacecraft - aluminum. Out of light and heavy chemical elements we chose the materials with high reaction cross sections and low density. The mass attenuation coefficient of boron- containing compounds is 20% higher than that of aluminum. Heterogeneous shields consist of three layers: a glass cloth, borated material, and nickel. With a protective shield containing heavy metal the output bremsstrahlung can be reduced. The amount of gamma rays that succeed to penetrate the shield is 4 times less compared to aluminum. The shields under study have the thicknesses of 5.95 and 6.2 mm. A comparative analysis of homogeneous and multilayered protective coatings of the same chemical composition has been performed. A heterogeneous protective shield has been found to be advantageous in weight and shielding properties over its homogeneous counterparts and aluminum. The dose characteristics and transmittance were calculated by the Monte Carlo method. The results of our study lead us to conclude that a three-layer boron carbide shield provides the most effective protection from radiation. This shield ensures twice as low absorbed dose and 4 times less the number of penetrated gamma-ray photons compared to its aluminum analogue. Moreover, a heterogeneous shield will have a weight 10% lighter than aluminum, with the same attenuation coefficient of the electron flux. Such heterogeneous shields can be used to protect spacecraft launched to geostationary orbit. Furthermore, a protective boron-containing and

  2. Shielding consideration for the SSCL experimental halls

    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 109 Hz or 1016 interactions per year, based on 107 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 1014 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

  3. Guidebook on radiation shielding safety for nuclear fuel facilities practical volume

    The practical volume of 'Guidebook on Radiation Shielding Safety for Nuclear Fuel Facilities' is prepared as a report which describes practical designing procedures of shielding calculation for fuel cycle facilities. In this report, the facilities of uranium fuel fabrication, MOX fuel fabrication and fuel reprocessing, and a fuel transport cask are taken up as typical facilities in the fuel cycle. The practical procedures for these facilities have been divided by four subjects, namely, (1)practical methodology of shielding design, (2)procedures of shielding calculation, (3)examples of shielding calculations and (4)check sheet of shielding calculation. The draft of the report has been discussed and summarized by the members of 'Working group on methodology of shielding safety' chaired by Dr.T.Kosako of the University of Tokyo. The working group belongs to the specialists group for demonstration of shielding safety by analysis, committee for safety research on nuclear facilities. (author)

  4. A high-performance magnetic shield with large length-to-diameter ratio

    Dickerson, Susannah; Hogan, Jason M.; Johnson, David M. S.; Kovachy, Tim; Sugarbaker, Alex; Chiow, Sheng-wey; Kasevich, Mark A.

    2012-06-01

    We have demonstrated a 100-fold improvement in the magnetic field uniformity on the axis of a large aspect ratio, cylindrical, mumetal magnetic shield by reducing discontinuities in the material of the shield through the welding and re-annealing of a segmented shield. The three-layer shield reduces Earth's magnetic field along an 8 m region to 420 μG (rms) in the axial direction, and 460 and 730 μG (rms) in the two transverse directions. Each cylindrical shield is a continuous welded tube which has been annealed after manufacture and degaussed in the apparatus. We present both experiments and finite element analysis that show the importance of uniform shield material for large aspect ratio shields, favoring a welded design over a segmented design. In addition, we present finite element results demonstrating the smoothing of spatial variations in the applied magnetic field by cylindrical magnetic shields. Such homogenization is a potentially useful feature for precision atom interferometric measurements.

  5. Shielding Structures for Interplanetary Human Mission

    Tracino, Emanuele; Lobascio, Cesare

    2012-07-01

    Since the end of Apollo missions, human spaceflight has been limited to the Low Earth Orbit (LEO), inside the protective magnetic field of the Earth, because astronauts are, to the largest degree, protected from the harsh radiation environment of the interplanetary space. However, this situation will change when space exploration missions beyond LEO will become the real challenge of the human exploration program. The feasibility of these missions in the solar system is thus strongly connected to the capability to mitigate the radiation-induced biological effects on the crew during the journey and the permanence on the intended planet surface. Inside the International Space Station (ISS), the volumes in which the crew spends most of the time, namely the crew quarters are the only parts that implement dedicated additional radiation shielding made of polyethylene tiles designed for mitigating SPE effects. Furthermore, specific radiation shielding materials are often added to the described configuration to shield crew quarters or the entire habitat example of these materials are polyethylene, liquid hydrogen, etc. but, increasing the size of the exploration vehicles to bring humans beyond LEO, and without the magnetosphere protection, such approach is unsustainable because the mass involved is a huge limiting factor with the actual launcher engine technology. Moreover, shielding against GCR with materials that have a low probability of nuclear interactions and in parallel a high ionizing energy loss is not always the best solution. In particular there is the risk to increase the LET of ions arriving at the spacecraft shell, increasing their Radio-Biological Effectiveness. Besides, the production of secondary nuclei by projectile and target fragmentation is an important issue when performing an engineering assessment of materials to be used for radiation shielding. The goal of this work is to analyze different shielding solutions to increase as much as possible the

  6. Comparison of eye shields in radiotherapeutic beams

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

  7. Capacitive Proximity Sensors With Additional Driven Shields

    Mcconnell, Robert L.

    1993-01-01

    Improved capacitive proximity sensors constructed by incorporating one or more additional driven shield(s). Sensitivity and range of sensor altered by adjusting driving signal(s) applied to shield(s). Includes sensing electrode and driven isolating shield that correspond to sensing electrode and driven shield.

  8. 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.

  9. Shielding benchmark test

    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)

  10. 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.

  11. Combustor bulkhead heat shield assembly

    Zeisser, M.H.

    1990-06-19

    This paper describes a gas turbine engine having an annular combustion chamber defined by an annular, inner liner, a concentric outer liner, and an upstream annular combustor head, wherein the head includes a radially extending bulkhead having circumferentially distributed openings for each receiving an individual fuel nozzle therethrough. It comprises: a segmented heat shield assembly, disposed between the combustion chamber interior and the bulkhead, including generally planar, sector shaped heat shields, each shield abutting circumferentially with two next adjacent shields and extending radially from proximate the inner liner to proximate the outer liner, the plurality of shields collectively defining an annular protective barrier, and wherein each sector shaped shield further includes an opening, corresponding to one of the bulkhead nozzle openings for likewise receiving the corresponding nozzle therethrough, the shield opening further including an annular lip extending toward the bulkhead and being received within the bulkhead opening, raised ridges on the shield backside, the ridges contacting the facing bulkhead surface and defining a flow path for a flow of cooling air issuing from a sized supply opening disposed in the bulkhead, the flow path running ultimately from adjacent the annular lip to the edges of each shield segment, wherein the raised edges extend fully along the lateral, circumferentially spaced edges of each shield segment and about the adjacent shield segments wherein the raised ridges further extend circumferentially between the annular lip and the abutting edge ridges.

  12. Shielding calculations for SSC

    Monte Carlo calculations of hadron and muon shielding for SSC are reviewed with emphasis on their application to radiation safety and environmental protection. Models and algorithms for simulation of hadronic and electromagnetic showers, and for production and transport of muons in the TeV regime are briefly discussed. Capabilities and limitations of these calculations are described and illustrated with a few examples. 12 refs., 3 figs

  13. Induction of quinone reductase activity by stilbene analogs in mouse Hepa 1c1c7 cells.

    Heo, Y H; Kim, S; Park, J E; Jeong, L S; Lee, S K

    2001-12-01

    Based on the potential cancer chemopreventive activity of resveratrol, a trihydroxystilbene with the induction of quinone reductase activity, this study was designed to determine if stilbene-related compounds were inducers of phase II detoxifying metabolic enzyme quinone reductase (QR) in the mouse hepatoma Hepa 1c1c7 cells. Among the thirteen compounds tested, several compounds including 3,4,5,3',5'-pentamethoxy-trans-stilbene were found to potentially induce QR activity in this cell line. In addition, substitution with 3-thiofurane ring instead of phenyl ring in the stilbene skeleton also exhibited potential induction of QR activity. This result will give primary information to design the potential inducers of QR activity in the stilbene analogs. PMID:11794542

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

    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)

  15. Shielding Benchmark Computational Analysis

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

    2000-09-17

    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).

  16. Shielding calculations for the antiproton target area

    Shielding calculations performed in conjunction with the design of the Fermilab antiproton target hall are summarized. The following radiological considerations were examined: soil activation, residual activity of components, and beam-on radiation. In addition, at the request of the designers, the energy deposition in the proposed graphite beam dump was examined for several targeting conditions in order to qualitatively determine its ability to survive

  17. Magnetic shielding tests for MFTF-B neutral beamlines

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

  18. Calmodulin binding to recombinant myosin-1c and myosin-1c IQ peptides

    Cyr Janet L

    2002-11-01

    Full Text Available Abstract Background Bullfrog myosin-1c contains three previously recognized calmodulin-binding IQ domains (IQ1, IQ2, and IQ3 in its neck region; we identified a fourth IQ domain (IQ4, located immediately adjacent to IQ3. How calmodulin binds to these IQ domains is the subject of this report. Results In the presence of EGTA, calmodulin bound to synthetic peptides corresponding to IQ1, IQ2, and IQ3 with Kd values of 2–4 μM at normal ionic strength; the interaction with an IQ4 peptide was much weaker. Ca2+ substantially weakened the calmodulin-peptide affinity for all of the IQ peptides except IQ3. To reveal how calmodulin bound to the linearly arranged IQ domains of the myosin-1c neck, we used hydrodynamic measurements to determine the stoichiometry of complexes of calmodulin and myosin-1c. Purified myosin-1c and T701-Myo1c (a myosin-1c fragment with all four IQ domains and the C-terminal tail each bound 2–3 calmodulin molecules. At a physiologically relevant temperature (25°C and under low-Ca2+ conditions, T701-Myo1c bound two calmodulins in the absence and three calmodulins in the presence of 5 μM free calmodulin. Ca2+ dissociated nearly all calmodulins from T701-Myo1c at 25°C; one calmodulin was retained if 5 μM free calmodulin was present. Conclusions We inferred from these data that at 25°C and normal cellular concentrations of calmodulin, calmodulin is bound to IQ1, IQ2, and IQ3 of myosin-1c when Ca2+ is low. The calmodulin bound to one of these IQ domains, probably IQ2, is only weakly associated. Upon Ca2+ elevation, all calmodulin except that bound to IQ3 should dissociate.

  19. Internally shielded beam transport and support system

    Due to environmental concerns, the Advanced Photon Source has a policy that disallows any exposed lead within the facility. This creates a real problem for the beam transport system, not so much for the pipe but for the flexible coupling (bellows) sections. A complete internally shielded x-ray transport system, consisting of long transport lines joined by flexible coupling sections, has been designed for CARS sector 14 to operate either at high vacuum or as a helium flight tube. It can effectively shield against air scattering of wiggler or undulator white beam with proper placement of apertures, collimators, and masks for direct beam control. The system makes use of male- and female-style fittings that create a labyrinth allowing for continuous shielding through the flexible coupling sections. These parts are precision machined from a ternary hypereutectic lead alloy (cast under 15 inches of head pressure to assure a pinhole-free casting) then pressed into either end (rotatable vacuum flanges) of the bellows assembly. The transport pipe itself consists of a four part construction using a stepped transition ring (Z-ring) to connect an inner tube to the vacuum flange and also to a protective and supportive outer tube. The inner tube is wrapped with 1/16 double-prime pure lead sheet to a predetermined thickness following the shape of the stepped transition ring for continuous shielding. This design has been prototyped and radiation tested. copyright 1996 American Institute of Physics

  20. Shielding in ungated field emitter arrays

    Cathodes consisting of arrays of high aspect ratio field emitters are of great interest as sources of electron beams for vacuum electronic devices. The desire for high currents and current densities drives the cathode designer towards a denser array, but for ungated emitters, denser arrays also lead to increased shielding, in which the field enhancement factor β of each emitter is reduced due to the presence of the other emitters in the array. To facilitate the study of these arrays, we have developed a method for modeling high aspect ratio emitters using tapered dipole line charges. This method can be used to investigate proximity effects from similar emitters an arbitrary distance away and is much less computationally demanding than competing simulation approaches. Here, we introduce this method and use it to study shielding as a function of array geometry. Emitters with aspect ratios of 102–104 are modeled, and the shielding-induced reduction in β is considered as a function of tip-to-tip spacing for emitter pairs and for large arrays with triangular and square unit cells. Shielding is found to be negligible when the emitter spacing is greater than the emitter height for the two-emitter array, or about 2.5 times the emitter height in the large arrays, in agreement with previously published results. Because the onset of shielding occurs at virtually the same emitter spacing in the square and triangular arrays, the triangular array is preferred for its higher emitter density at a given emitter spacing. The primary contribution to shielding in large arrays is found to come from emitters within a distance of three times the unit cell spacing for both square and triangular arrays

  1. Recent developments in fusion first wall, blanket, and shield technology

    This brief overview of first wall, blanket and shield technology reviews the changes and trends in important design issues in first wall, blanket and shield design and related technology from the 1970's to the 1980's. The emphasis is on base technology rather than either systems engineering or materials development. The review is limited to the two primary confinement systems, tokamaks and mirrors, and production of electricity as the primary goal for development

  2. Optimal selection for shielding materials by fuzzy linear programming

    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

  3. Justification for Shielded Receiver Tube Additional Lead Shielding

    In order to reduce high radiation dose rates encountered when core sampling some radioactive waste tanks the addition of 240 lbs. of lead shielding is being considered to the shielded receiver tube on core sample trucks No.1, No.3 and No.4. The lead shielding is 4 inch diameter x 1/2 inch thick half rounds that have been installed around the SR tube over its' full length. Using three unreleased but independently reviewed structural analyses HNF-6018 justifies the addition of the lead shielding

  4. Measurement of the transient shielding effectiveness of shielding cabinets

    H. Herlemann

    2008-05-01

    Full Text Available Recently, new definitions of shielding effectiveness (SE for high-frequency and transient electromagnetic fields were introduced by Klinkenbusch (2005. Analytical results were shown for closed as well as for non closed cylindrical shields. In the present work, the shielding performance of different shielding cabinets is investigated by means of numerical simulations and measurements inside a fully anechoic chamber and a GTEM-cell. For the GTEM-cell-measurements, a downscaled model of the shielding cabinet is used. For the simulations, the numerical tools CONCEPT II and COMSOL MULTIPHYSICS were available. The numerical results agree well with the measurements. They can be used to interpret the behaviour of the shielding effectiveness of enclosures as function of frequency. From the measurement of the electric and magnetic fields with and without the enclosure in place, the electric and magnetic shielding effectiveness as well as the transient shielding effectiveness of the enclosure are calculated. The transient SE of four different shielding cabinets is determined and discussed.

  5. Dynamic magnetic shield for the CLAS12 central TOF detector photomultiplier tubes

    The Central Time-of-Flight detector for the Jefferson Laboratory 12-GeV upgrade is being designed with linear-focused photomultiplier tubes that require a robust magnetic shield against the CLAS12 main 5-T solenoid fringe fields of 100 mT (1 kG). Theoretical consideration of a ferromagnetic cylinder in an axial field has demonstrated that its shielding capability decreases with increasing length. This observation has been confirmed with finite element analysis using POISSON model software. Several shields composed of coaxial ferromagnetic cylinders have been studied. All difficulties caused by saturation effects were overcome with a novel dynamical shield, which utilizes a demagnetizing solenoid between the shielding cylinders. Basic dynamical shields for ordinary linear-focused 2-in. photomultiplier tubes were designed and tested both with models and experimental prototypes at different external field and demagnetizing current values. Our shield design reduces the 1 kG external axial field by a factor of 5000.

  6. (n,γ)反应实验研究中的中子屏蔽设计%Neutron Shielding Design for Experiment Research of (n,γ) Reaction

    黄兴; 贺国珠; 程品晶; 张奇纬; 周祖英

    2015-01-01

    Neutron capture cross section can be measured by Gamma-ray Total Absorption Facility (GTAF) with high precision. To reduce the background of experiments, the neutron source must be collimated and shielded, and the neutrons scattered from the sample must be absorbed to minimise interference after they go into the detector. The shield, collimator and absorber were simulated and designed with MCNP code. Boron-containing polyethylene with 3%BC4 and lead are used as the materials for the neutron collimator and shield. The diameter of the collimating aperture is 13 mm, and the length of the collimator is 500 mm. After being collimated, the diameter of neutron beam plateau at the sample position is 21 mm. The neutron absorber is made of polyethylene and BC4, and the thickness of polyethylene shell and BC4 shell are 60 and 10 mm, respectively. The simulated result shows that neutrons scattered from the sample can decay 93.7%through the neutron absorber.%用γ射线全吸收型装置(Gamma-ray Total Absorption Facility,GTAF),可以对中子俘获反应截面进行高精度测量。为了降低实验本底,实验中需要对源中子进行准直和屏蔽,还要对被样品散射的中子进行吸收以减少它们进入探测器后所形成的干扰。采用MCNP对中子的准直器、屏蔽体和中子吸收体进行了模拟设计,中子准直屏蔽体材料选用含硼聚乙烯(BC4的质量分数为3%)和铅。准直孔直径为13 mm,长度为500 mm,经准直后样品处中子束斑坪顶直径为21 mm。中子吸收体材料选用聚乙烯和碳化硼,吸收体球壳内腔半径30 mm,聚乙烯壳层厚度60 mm,碳化硼壳层厚度10 mm,被样品散射的中子经吸收体后衰减93.7%。

  7. General Corrosion and Localized Corrosion of the Drip Shield

    F. Hua

    2004-09-16

    The repository design includes a drip shield (BSC 2004 [DIRS 168489]) that provides protection for the waste package both as a barrier to seepage water contact and a physical barrier to potential rockfall. The purpose of the process-level models developed in this report is to model dry oxidation, general corrosion, and localized corrosion of the drip shield plate material, which is made of Ti Grade 7. This document is prepared according to ''Technical Work Plan For: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The models developed in this report are used by the waste package degradation analyses for TSPA-LA and serve as a basis to determine the performance of the drip shield. The drip shield may suffer from other forms of failure such as the hydrogen induced cracking (HIC) or stress corrosion cracking (SCC), or both. Stress corrosion cracking of the drip shield material is discussed in ''Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material'' (BSC 2004 [DIRS 169985]). Hydrogen induced cracking of the drip shield material is discussed in ''Hydrogen Induced Cracking of Drip Shield'' (BSC 2004 [DIRS 169847]).

  8. Performance study of galactic cosmic ray shield materials

    Kim, Myung-Hee Y.; Wilson, John W.; Thibeault, Sheila A.; Nealy, John E.; Badavi, Francis F.; Kiefer, Richard L.

    1994-01-01

    The space program is faced with two difficult radiation protection issues for future long-term operations. First, retrofit of shield material or conservatism in shield design is prohibitively expensive and often impossible. Second, shielding from the cosmic heavy ions is faced with limited knowledge on the physical properties and biological responses of these radiations. The current status of space shielding technology and its impact on radiation health is discussed herein in terms of conventional protection practice and a test biological response model. The impact of biological response on the selection of optimum materials for cosmic ray shielding is presented in terms of the transmission characteristics of the shield material. Although the systematics of nuclear cross sections are able to demonstrate the relation of exposure risk to shield-material composition, the current uncertainty in-nuclear cross sections will not allow an accurate evaluation of risk reduction. This paper presents a theoretical study of risk-related factors and a pilot experiment to study the effectiveness of choice of shield materials to reduce the risk in space operations.

  9. Analysis and improvement of cyclotron thallium target room shield

    Because of high neutron and gamma ray intensities during thallium-203 target bombardment, thallium target room shield and its improvement have been investigated. Leakage neutron and gamma-ray dose rates in various points behind the shield are calculated by simulating the transport of neutrons and photons using Monte Carlo MCNP4C computer code. By considering target room geometry, its associated shield, neutron and gamma rays source strengths and spectra, three designs for enhancing shield performance have been analyzed; A door as a shield in maze entrance, covering maze walls with layers of some effective materials and adding a shadow shield in target room in front of the radiation source, have been considered as the parallel to the maze. Dose calculations carried out for each kind of suggested shields separately for different materials and dimensions, then the shield with better than has been constructed and It has been found that the deviation between calculated and measured dose values after upgrading is less than 20%

  10. Passive Shielding in CUORE

    The nature of neutrino mass is one of the friontier problems of fundamental physics. Neutrinoless Double Beta Decay (0νDBD) is a powerful tool to investigate the mass hierarchy and possible extensions of the Standard Model. CUORE is a 1-Ton next generation experiment, made of 1000 Te bolometers, aiming at reaching a background of 0.01 (possibly 0.001) counts keV-1kg-1y-1 and therefore a mass sensitivity of few tens of meV The background contribution due to environmental neutrons, muon-induced neutrons in the shieldings and external gamma is discussed

  11. Walls shielding against ionizing radiation

    These specifications are to help the users of lead bricks as under DIN 25407, leaf 1, with the construction of walls shielding against ionizing radiation by examples for the uses of the different types of lead bricks and by recommendations for the construction of shielding walls and for the determination of the wall thickness necessary for shielding against γ-radiation as a function of energy. (orig./AK)

  12. 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.

  13. 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.

  14. The Tower Shielding Facility: Its glorious past

    Muckenthaler, F.J.

    1997-05-07

    The Tower Shielding Facility (TSF) is the only reactor facility in the US that was designed and built for radiation-shielding studies in which both the reactor source and shield samples could be raised into the air to allow measurements to be made without interference from ground scattering or other spurious effects. The TSF proved its usefulness as many different programs were successfully completed. It became active in work for the Defense Atomic Support Agency (DASA) Space Nuclear Auxiliary Power, Defense Nuclear Agency, Liquid Metal Fast Breeder Reactor Program, the Gas-Cooled and High-Temperature Gas-Cooled Reactor programs, and the Japanese-American Shielding Program of Experimental Research, just to mention a few of the more extensive ones. The history of the TSF as presented in this report describes the various experiments that were performed using the different reactors. The experiments are categorized as to the programs which they supported and placed in corresponding chapters. The experiments are described in modest detail, along with their purpose when appropriate. Discussion of the results is minimal, but references are given to more extensive topical reports.

  15. The Tower Shielding Facility: Its glorious past

    The Tower Shielding Facility (TSF) is the only reactor facility in the US that was designed and built for radiation-shielding studies in which both the reactor source and shield samples could be raised into the air to allow measurements to be made without interference from ground scattering or other spurious effects. The TSF proved its usefulness as many different programs were successfully completed. It became active in work for the Defense Atomic Support Agency (DASA) Space Nuclear Auxiliary Power, Defense Nuclear Agency, Liquid Metal Fast Breeder Reactor Program, the Gas-Cooled and High-Temperature Gas-Cooled Reactor programs, and the Japanese-American Shielding Program of Experimental Research, just to mention a few of the more extensive ones. The history of the TSF as presented in this report describes the various experiments that were performed using the different reactors. The experiments are categorized as to the programs which they supported and placed in corresponding chapters. The experiments are described in modest detail, along with their purpose when appropriate. Discussion of the results is minimal, but references are given to more extensive topical reports

  16. A Monitoring System for the Assessment of Reactor Shield Performance

    This paper describes the objectives of a shield survey, presents the results of intercomparisons and makes recommendations on the selection of a monitoring system. The performance of a reactor shield must be assessed efficiently to enable immediate repairs to be made to defective shields. Correlation of survey results with shield designer's predictions has proved difficult and the factors contributing to the disagreement have been examined. The shield designer uses flux to dose conversion factors which neglect the radiation equilibrium situation at the shield-air interface and the effect of attenuation and scattering external to the shield. The presence of the operator can significantly influence a measurement. The use of the ''maximum exposure dose'' (MED) concept is recommended for gamma photon predictions and this is compared with absorbed dose measurements in phantoms. Shield survey instruments must have a flat response over a wide range of energies. For example, many gamma dosimeters are designed for use at energies up to about 2 MeV whereas around most reactors a proportion of dose is due to 6 MeV gamma photons, with some contribution from higher energies. Tests were carried out to select suitable detectors using actual reactor operating conditions and simulated conditions at specific energies, notably 6 MeV. In practice discrepancies exceeding a factor of 3 were frequently found. A comparison was made between thermoluminescent dosimeters and film badges for the purpose of fixed position integrating dosimeters. The film dosimeter (AERE/RPS) was found to over-estimate by a factor of 2 for 6 MeV gamma radiation and in practical situations over-estimated by 20 to 80%. Thermoluminescent dosimetry is recommended for shield surveys provided that a build-up cap is used to achieve charged particle equilibrium. (author)

  17. Enzymatic syntheses of (1-(C-11))pyruvic acid and L-(1-(C-11))lactic acid via DL-(1-(C-11))alanine

    Ropchan, J.R.; Barrio, J.R.

    1984-01-01

    L-(1-(C-11)) Lactic acid was prepared in three steps using a remote, semi-automated procedure: (1) production of DL-(1-(C-11)) alanine (2) enzymatic conversion of DL (1-(C-11)) alanine to (1-(C-11)) pyruvate and (3) enzymatic transformation of (1-(C-11)) pyruvate to L-(1-(C-11)) lactic acid. DL-(1-(C-11)) Alanine was synthesized from NCA C-11 HCN using a modification of the Bucherer-Strecker reaction. The DL-isomers were converted to (1-(C-11)) pyruvate by passage through (1) immobilized D-amino acid oxidase enzyme column followed by (2) immobilized L-alanine dehydrogenase (l-ADH) enzyme column. (1-(C-11)) Pyruvate was then transformed to L-(1-(C-11)) lactic acid by elution through a L-lactic dehydrogenase enzyme column. These enzyme columns are reusable beyond three months, give high radiochemical purity (>98%), eliminate the possibility of protein contamination, assure sterile, pyrogen-free products and allow rapid separation and quantitative conversion of DL-isomers to the desired products. Typically the synthesis required 30-40 min after cyclotron production of NCA C-11 HCN with radiochemical yields of 15-25 mCi (23%) of L-(1-(C-11)) lactic acid and 20-35 mCi (33%) of (l-(C-11)) pyruvic acid starting with 250-400 mCi of C-11 HCN. Also 10-20 mCi (19%) of L-(1-(C-11)) alanine was produced by resin separation (AG50W-X8), H/sup +/ form of (1-(C-11)) pyruvate and L-(1-(C-11)) alanine following elution through D-AAO enzyme column. The radiochemical purities of (1-(C-11)) pyruvic acid, L-(1-(C-11)) lactic acid and L-(1-(C-11)) alanine were verified routinely by reversed-phase HPLC.

  18. Activation of the concrete in the bio shield of ITER

    Calculations of neutron spectra in different parts of the tokamak building of ITER are performed. A computational geometry model of the tokamak building is prepared using MCNP-4C. The model includes adequate material composition and geometry description of the main parts of the tokamak for PPCS plant model A: toroidal field coils, vacuum vessel, shield, blanket structure, first wall, divertor, 14.1 MeV neutron source. The design and the dimensions of the bio shield are taken from the current ITER design. MCNP calculations of the neutron spectra in the bio shield (concrete) of ITER are performed, using the neutron spectra in TF coils calculated at UKAEA as external neutron source. The neutron spectra in the concrete calculated by MCNP are used as input data in the code EASY99 for estimations of the activation of the concrete in the bio shield around the tokamak. The time evolutions of the maximum (in the bio shield floor) and minimum (in the bio shield side walls) specific activity (Bq/kg) and dose rate (Sv/h.) of the main dominant nuclides in the concrete are evaluated and compared for 3 different concrete types, used as biological shield in the PWR and BR3 reactors. (author)

  19. 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.

  20. Nuclear shielding of openings in ITER Tokamak building

    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

  1. HbA1c in Nondiabetic Dutch Infants Aged 8–12 Months

    Jansen, Hanneke; Huiting, Haika G.; Scholtens, Salome; Sauer, Pieter J.J.; Stolk, Ronald P

    2011-01-01

    OBJECTIVE An international committee of experts recommended using HbA1c for diagnostic testing for diabetes. Little is known about normal values of HbA1c in infants. The aim of this study is to describe the distribution of HbA1c in 8- to 12-month-old nondiabetic infants. RESEARCH DESIGN AND METHODS HbA1c was measured in 86 infants participating in the Groningen Expert Center for Kids with Obesity (GECKO)-Drenthe birth cohort study. Anthropometric measurements were performed at Well Baby Clini...

  2. Simplified shielding calculation system for high-intensity proton accelerators

    A simplified shielding calculation system is developed for applying conceptual shielding design of facilities in the joint project for high-intensity proton accelerators. The system is composed of neutron transmission calculation part for bulk shielding using simplified formulas: Moyer model and Tesch's formula, and neutron skyshine calculation part using an empirical formula: Stapleton's formula. The system is made with the Microsoft Excel software for user's convenience. This report provides a manual for the system as well as calculation conditions used in the calculation such as Moyer model's parameters. In this report preliminary results based on data at December 8, 1999, are also shown as an example. (author)

  3. EBT-P gamma-ray shielding system

    Gohar, Y.

    1981-12-01

    An elaborate study was carried out for the coil and biological shield of the ELMO Bumpy Torus proof-of-principle (EBT-P) device. 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 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 machine room, and (c) the skyshine contribution to the dose equivalent.

  4. EBT-P gamma-ray shielding system

    An elaborate study was carried out for the coil and biological shield of the ELMO Bumpy Torus proof-of-principle (EBT-P) device. 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 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 machine room, and (c) the skyshine contribution to the dose equivalent

  5. Technical Aspect of Shielded SIMS Installation in CEA Cadarache

    A shielded IMS 6f has been installed in the LECA facility, CEA Cadarache France. The nuclearization was performed by CAMECA Company, which sells the standard IMS 6f. Working on nuclear materials requires in depth modifications of the apparatus itself. Despite these modifications, the shielded SIMS has the same level of performance as the standard apparatus. The design of the modified apparatus is presented and the safety aspects are emphasised. The shielded SIMS should be allowed to handle irradiated samples at the end of 2001. (Author)

  6. High frequency electromagnetic interference shielding magnetic polymer nanocomposites

    He, Qingliang

    Electromagnetic interference is one of the most concerned pollution and problem right now since more and more electronic devices have been extensively utilized in our daily lives. Besides the interference, long time exposure to electromagnetic radiation may also result in severe damage to human body. In order to mitigate the undesirable part of the electromagnetic wave energy and maintain the long term sustainable development of our modern civilized society, new technology development based researches have been made to solve this problem. However, one of the major challenges facing to the electromagnetic interference shielding is the relatively low shielding efficiency and the high cost as well as the complicated shielding material manufacture. From the materials science point of view, the key solutions to these challenges are strongly depended on the breakthrough of the current limit of shielding material design and manufacture (such as hierarchical material design with controllable and predictable arrangement in nanoscale particle configuration via an easy in-situ manner). From the chemical engineering point of view, the upgrading of advanced material shielding performance and the enlarged production scale for shielding materials (for example, configure the effective components in the shielding material in order to lower their usage, eliminate the "rate-limiting" step to enlarge the production scale) are of great importance. In this dissertation, the design and preparation of morphology controlled magnetic nanoparticles and their reinforced polypropylene polymer nanocomposites will be covered first. Then, the functionalities of these polymer nanocomposites will be demonstrated. Based on the innovative materials design and synergistic effect on the performance advancement, the magnetic polypropylene polymer nanocomposites with desired multifunctionalities are designed and produced targeting to the electromagnetic interference shielding application. In addition

  7. SQUID holder with high magnetic shielding

    Rigby, K. W.; Marek, D.; Chui, T. C. P.

    1990-01-01

    A SQUID holder designed for high magnetic shielding is discussed. It is shown how to estimate the attenuation of the magnetic field from the normal magnetic modes for an approximate geometry. The estimate agrees satisfactorily with the attenuation measured with a commercial RF SQUID installed in the holder. The holder attenuates external magnetic fields by more than 10 to the 9th at the SQUID input. With the SQUID input shorted, the response to external fields is 0.00001 Phi(0)/G.

  8. Final Technical Report [Cosmogenic background and shielding R&D for a Ge Neutrinoless Double Beta Decay Experiment

    Guiseppe, Vince

    2013-10-01

    The USD Majorana group focused all of its effort in support of the MAJORANA DEMONSTRATOR (MJD) experiment. Final designs of the shielding subsystems are complete. Construction of the MJD shielding systems at SURF has begun and the proposed activities directly support the completion of the shield systems. The PI and the group contribute heavily to the onsite construction activities of the MJD experiment. The group led investigations into neutron and neutron-­induced backgrounds, shielding effectiveness and design, and radon backgrounds.

  9. Internal Calibration of HJ-1-C Satellite SAR System

    Yang Zhen

    2014-06-01

    Full Text Available The HJ-1-C satellite is a Synthetic Aperture Radar (SAR satellite of a small constellation for environmental and disaster monitoring. At present, it is in orbit and working well. The SAR system uses a mesh reflector antenna and centralized power amplifier, and has an internal calibration function in orbit. This study introduces the internal calibration modes and signal paths. The design and realization of the internal calibrator are discussed in detail. Finally, the internal calibration data acquired in orbit are also analyzed.

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

    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 PuO2 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

  11. Relationship of HbA1c variability, absolute changes in HbA1c, and all-cause mortality in type 2 diabetes

    Skriver, Mette Vinther; Sandbæk, Annelli; Kristensen, Jette Kolding;

    2015-01-01

    was defined as the mean absolute residual around the line connecting index value with closing value. Cox proportional hazard models with restricted cubic splines were used, with all-cause mortality as the outcome. RESULTS: Variability between 0 and 0.5 HbA1c percentage point was not associated with......OBJECTIVE: We assessed the relationship of mortality with glycated hemoglobin (HbA1c) variability and with absolute change in HbA1c. DESIGN: A population-based prospective observational study with a median follow-up time of 6 years. METHODS: Based on a validated algorithm, 11 205 Danish individuals...

  12. The shield effect

    Toft, Søren; Albo, Maria J

    2016-01-01

    Several not mutually exclusive functions have been ascribed to nuptial gifts across different taxa. Although the idea that a nuptial prey gift may protect the male from pre-copulatory sexual cannibalism is attractive, it has previously been considered of no importance based on indirect evidence and...... rejected by experimental tests. We reinvestigated whether nuptial gifts may function as a shield against female attacks during mating encounters in the spider Pisaura mirabilis and whether female hunger influences the likelihood of cannibalistic attacks. The results showed that pre-copulatory sexual...... cannibalism was enhanced when males courted without a gift and this was independent of female hunger. We propose that the nuptial gift trait has evolved partly as a counteradaptation to female aggression in this spider species....

  13. Sulphate resistant shielding material

    The shielding material of the present invention is provided with sulfuric acid resistance and contains bentonite put to ion exchange treatment with barium ions as an effective ingredient. When mortars and concretes are exposed to the circumstance of sulfate, the effective ingredient functions to take place reaction between intruding sulfate and the barium ions to form insoluble barium sulfate thereby reducing chemical corrosion of mortars and concretes caused by sulfate. Cement materials, water and aggregates can optionally be contained in addition to bentonite and bentonite put to ion exchange treatment. Chemical corrosion of concretes and mortars due to intrusion of the sulfate can be prevented, and it is useful as an artificial barrier, for example, in radioactive active waste processing facilities. (T.M.)

  14. 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.

  15. Parameters calculation of shielding experiment

    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)

  16. Analysis list: Myo1c [Chip-atlas[Archive

    Full Text Available Myo1c Embryonic fibroblast + mm9 http://dbarchive.biosciencedbc.jp/kyushu-u/mm9/target/My...o1c.1.tsv http://dbarchive.biosciencedbc.jp/kyushu-u/mm9/target/Myo1c.5.tsv http://dbarchive.bioscienc...edbc.jp/kyushu-u/mm9/target/Myo1c.10.tsv http://dbarchive.biosciencedbc.jp/kyushu-u/mm9/colo/Myo1c.Embryonic

  17. Analysis list: Jmjd1c [Chip-atlas[Archive

    Full Text Available Jmjd1c Pluripotent stem cell + mm9 http://dbarchive.biosciencedbc.jp/kyushu-u/mm9/target/Jmj...d1c.1.tsv http://dbarchive.biosciencedbc.jp/kyushu-u/mm9/target/Jmjd1c.5.tsv http://dbarchive.biosc...iencedbc.jp/kyushu-u/mm9/target/Jmjd1c.10.tsv http://dbarchive.biosciencedbc.jp/kyushu-u/mm9/colo/Jmjd1c.Plu

  18. Multihelix rotating shield brachytherapy for cervical cancer

    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 (D90 of HR-CTV) were the two metrics used as the basis for evaluation and

  19. 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

  20. 泥水盾构机泥水循环从站控制系统的设计%Design of slave station control system for slurry circulation in slurry shield machine

    安冬云; 廖潇; 寇宝庆; 张华芬; 张树凯

    2013-01-01

    Taking the Herrenknecht shield machine as an example,a slave station control system for slurry circulation,independent from the master station control system,is designed to seamlessly connect to the shield machine control system.The slave station is composed of slurry pump control cabinet,valve control box and ground monitoring station.The S7200PLC is selected for the control and the Profibus-DP is used for communication with the master station.In ground monitoring station,a combination of S7-200PLC+Profibus-DP modulus+computer terminal is designed to synchronously display the operation conditions of pumps and valves.The slurry pump is driven by AC motor,with its speed controlled by a schnielder Altivar 71 frequency convertor,a Proflbus-DP communication card and a S7-200PLC.The Profibus-DP network and Siemens S7 series PLC are utilized to construct the overall system with twin-core CAT 5 cable as the transmission media in the bus,with the transmission speed up to 12 MBP.Optoelectronic isolation relay amplifiers are added in every local control cabinet,which can solve the problems in long-distance shield tunneling communication signal transmission attenuation and electrical safety in the tunnel.(3 Figures,2 References)%以海瑞克盾构机为例,设计了泥水循环从站控制系统,使其与盾构机控制系统无缝连接,并独立于主站控制系统.各泥浆泵控制柜、阀门控制箱、地面监控站为从站结构,选用S7-200PLC控制,通过Profibus-DP现场总线与主站通讯;地面监控站采用S7-200PLC+ Profibus-DP模块+计算机终端,同步显示远程控制站的泵、阀运行状态;泥浆泵驱动电机选用交流电机,调速控制采用schnielder Altivar 71变频器+ Profibus-DP通讯卡+S7-200PLC;采用Profibus-DP网络和西门子S7系列PLC搭建整个系统,总线采用2芯5类电缆作为传输介质,通讯速度可达12 MBP;在本地控制柜增加光电隔离中继放大器,解决长距离盾构隧道掘进通讯信号

  1. Practical aspects of shielding high-energy particle accelerators

    The experimental basis of shielding design for high-energy accelerators that has been established over the past thirty years is described. Particular emphasis is given to the design of large accelerators constructed underground. The first data obtained from cosmic-ray physics were supplemented by basic nuclear physics. When these data proved insufficient, experiments were carried out and interpreted by several empirical formulae -- the most successful of which has been the Moyer Model. This empirical model has been used successfully to design the shields of most synchrotrons currently in operation, and is still being used in preliminary design and to check the results of neutron transport calculations. Accurate shield designs are needed to reduce external radiation levels during accelerator operations and to minimize environmental impacts such as open-quotes skyshineclose quotes and the production of radioactivity in groundwater. Examples of the cost of minimizing such environmental impacts are given

  2. Electrodynamic Dust Shield for Space Applications

    Mackey, Paul J.; Cox, Rachel E.; Calle, Carlos I.; Johansen, Michael R.; Olsen, Robert C.; Raines, Matthew G.; Phillips, James R., III; Hogue, Michael D.; Pollard, Jacob R. S.

    2016-01-01

    Dust mitigation technology has been highlighted by NASA and the International Space Exploration Coordination Group (ISECG) as a Global Exploration Roadmap (GER) critical technology need in order to reduce life cycle cost and risk, and increase the probability of mission success. The Electrostatics and Surface Physics Lab in Swamp Works at the Kennedy Space Center has developed an Electrodynamic Dust Shield (EDS) to remove dust from multiple surfaces, including glass shields and thermal radiators. Further development is underway to improve the operation and reliability of the EDS as well as to perform material and component testing outside of the International Space Station (ISS) on the Materials on International Space Station Experiment (MISSE). This experiment is designed to verify that the EDS can withstand the harsh environment of space and will look to closely replicate the solar environment experienced on the Moon.

  3. Radioisotope Power System Facility shielding analysis

    A series of calculations for the Radioisotope Power System Facility have been performed. These analyses have determined the shielding required for storage, testing, and transport of 238Pu heat source modules using the Monte Carlo code MCNP3B. The source terms and the assumptions used have been verified by comparison of calculated dose rates with measured ones. This paper describes the methodology used for shielding designs and the utilization of available variance reduction techniques to improve the computational efficiency. The new version of MCNP (MCNP3B) with a repeated structure capability was used. It decreased the chance for computer model errors and greatly decreased the model setup time. 2 refs., 3 figs., 2 tabs

  4. COG: A particle transport code designed to solve the Boltzmann equation for deep-penetration (shielding) problems: Benchmark problems: Volume 4

    Wilcox, T.P. Jr.; Lent, E.M.

    1988-12-02

    COG is a Monte Carlo computer code designed to solve the Boltzmann equation for transporting neutrons and photons and in future versions, charged particles. Sixty-four different problems were run using the current versions of the COG code on Cray-1 and Cray/X-MP computers. In all cases, the calculated COG results either agree with the values known analytically for some problems or are within the statistical and uncertainties determined experimentally for the others. Problems such as these are referred to benchmark problems and form an important part of the validation of any new computer code. Benchmark problems are of value in that they are used to: check that the code works correctly; check that the physical data used in the code are correct; check that the user has learned to run the code properly; and understand the inherent errors associated with the calculated results. 22 refs., 21 figs., 10 tabs.

  5. Designs of Antimagnetic Shield Box to Use Micro Injection Pump in Magnetic Resonance Image Formation%一种核磁共振成像中微量注射泵防磁屏蔽箱的设计

    徐新民; 胡军智

    2011-01-01

    Objective To design one kind of antimagnetic shield box which is able to use the micro injection pump's in the MR1 strong magnetic field environment. Methods With electromagnetic screen mechanism, ihe non -ferromagnetic material was used to process the cuboid box. Whose size was proper to contain a conventional micro injection pump. Protective measures were taken to the operation hole, infusion connecting pipe and the seam beside the door, ihus enabling it to be antimagnetie and isolated. Results This method avoided the conventional MR1 magnetic field on the magnetization of the role of micro-injection pump, and the MRI image quality were not affected. Conclusion In the MRI environment, this antimagnelic shield box can cause the conventional micro injection pump to use normally under the effective addressing MRI environment, which solve the problem of inaccurate medicine given by manpower. It also expands the use scope of conventional micro injection pump, and enhances the working efficiency as well as saves the massive manpower and themedical resources.%目的:设计一种在MRI强磁场环境中能使用微量注射采的防磁屏蔽箱.方法:利用电磁屏蔽机制,采用非铁磁性材料加工成长方体箱子,其大小适合,正好能放进一台常规微量注射泵,对操作孔、输注连接管出孔、门边缝加装防护措施,使其对磁场具有很好的防磁、隔离作用.结果:该装置避免了MRI磁场对常规微量注射泵的磁化作用,使其在MRI环境中正常使用并对MRI图像质量无任何影响.结论:该防磁屏蔽箱有效解决了MRI环境下人工给药不精准的难题,扩大了常规微量注射泵的使用范围;而且提高了MRI强磁场环境中输注给药的工作效率,节约了大量的人力和医疗资源.

  6. [An individual facial shield for a sportsman with an orofacial injury].

    de Baat, C; Peters, R; van Iperen-Keiman, C M; de Vleeschouwer, M

    2005-05-01

    Facial shields are used when practising contact sports, high speed sports, sports using hard balls, sticks or bats, sports using protective shields or covers, and sports using hard boardings around the sports ground. Examples of facial shields are commercially available, per branch of sport standardised helmets. Fabricating individual protective shields is primarily restricted to mouth guards. In individual cases a more extensive facial shield is demanded, for instance in case of a surgically stabilised facial bone fracture. In order to be able to fabricate an extensive individual facial shield, an accurate to the nearest model of the anterior part of the head is required. An accurate model can be provided by making an impression of the face, which is poured in dental stone. Another method is producing a stereolithographic model using computertomography or magnetic resonance imaging. On the accurate model the facial shield can be designed and fabricated from a strictly safe material, such as polyvinylchloride or polycarbonate. PMID:15932045

  7. The construction of radiation shielding for baby ebm

    The construction of radiation shielding for electron beam machine, Baby EBM is necessary for prevention from x-ray (Bremstrahlung) that produced when electron bombarded the target material. The strength of produced x-ray is depending on electron energy and the atomic number of target material. In the construction process of radiation shielding, a few aspects need to be considered such as shielding material and its thickness to be used, mainframe for radiation shielding and the way fabrication to be done. In this project, the thickness of radiation shielding is calculated manually following the NCRP 51 guidelines whereas for frame design, shielding walls and fabrication is considered that the accelerator devices (accelerating tube, focusing device and neck) is vertically and the whole weight of Baby EBM. From the calculations, the thickness and the material for radiation shielding is to be used are 6mm lead. This radiation shielding has been tested (using the parameters that have been considered) to know the leak of radiation (at all surfaces) and direct radiation below 5 cm from the window. The value of high voltage that applied at accelerating tube is 80 kV and the voltage, current supply at electron gun is 3.0 V, 7.1 A respectively. The result of the testing found that dose rate under the window foil is more than 2000 mSv/hr and at all shielding surfaces are less than 0.5 mSv/hr, which is background reading and this is acceptable as compared to the theoretical calculation. The measurement was done using a survey meter typed Ludlum-model 3. (Author)

  8. Radiation shields for a shelter

    A simple and cheap closure and radiation shield arrangement is described for the entrance of an underground shelter. The shelter can serve as a blast-proof, biological or nuclear shelter. The radiation shield is positioned above the habitable space of the shelter and below a blast-proof, dust-proof outer cover. The shield consists of a box containing a filling, e.g. coke with a concrete screed, is closed by bolted panels and is horizontally moveable by sliding on castors. (author)

  9. New Materials for EMI Shielding

    Gaier, James R.

    1999-01-01

    Graphite fibers intercalated with bromine or similar mixed halogen compounds have substantially lower resistivity than their pristine counterparts, and thus should exhibit higher shielding effectiveness against electromagnetic interference. The mechanical and thermal properties are nearly unaffected, and the shielding of high energy x-rays and gamma rays is substantially increased. Characterization of the resistivity of the composite materials is subtle, but it is clear that the composite resistivity is substantially lowered. Shielding effectiveness calculations utilizing a simple rule of mixtures model yields results that are consistent with available data on these materials.

  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. Radiation shielding and safety analysis for SPring-8

    The methods of shielding design and safety analysis applied to SPring-8 are summarized. SPring-8, a third generation synchrotron radiation facility, is the facility with the highest stored electron energy of 8 GeV and very low beam emittance of 5.5 nm·rad. Because of these distinguished features, a variety of radiation issues have to be taken up, requiring the latest information for analyses. In this technical report are described the calculational methods and the conditions for the following shielding matters as well as verification of the validity; a bulk shielding, synchrotron radiation beamline shielding, skyshine, streaming through ducts and mazes, induced activities in air, cooling water and targets, and incident analysis due to abnormal beam losses. (author)

  12. Radiation shielding and safety analysis for SPring-8

    Asano, Yoshihiro; Sasamoto, Nobuo [Japan Atomic Energy Research Inst., Kamigori, Hyogo (Japan). Kansai Research Establishment

    1998-03-01

    The methods of shielding design and safety analysis applied to SPring-8 are summarized. SPring-8, a third generation synchrotron radiation facility, is the facility with the highest stored electron energy of 8 GeV and very low beam emittance of 5.5 nm{center_dot}rad. Because of these distinguished features, a variety of radiation issues have to be taken up, requiring the latest information for analyses. In this technical report are described the calculational methods and the conditions for the following shielding matters as well as verification of the validity; a bulk shielding, synchrotron radiation beamline shielding, skyshine, streaming through ducts and mazes, induced activities in air, cooling water and targets, and incident analysis due to abnormal beam losses. (author)

  13. Reduction of the background in the new passive shield

    A shield was designed and tested for low level gamma spectroscopy. It consists of 10 mm of iron, 100 mm of lead, 10 mm of copper, 80 mm of NEUTROSTOP bricks (polyethylene with boric acid) and of 10 mm of plexiglas. The inner dimensions of the shield are 380x380x600 mm. The characteristics of the low-background shielding were determined by measuring the background at the inside and outside. For measurements, a coaxial large-volume Ge(Li) detector, Canberra NIM modules and multichannel analyzer ICA-70 were used. The background spectra were measured in an energy range of 200 to 3050 keV for 250,000 s. The shield was found to reduce the background values in the energy range E <= 2,600 keV by a factor of 10 to 50. (E.S.)

  14. Monte Carlo simulations for optimization of neutron shielding concrete

    Piotrowski, Tomasz; Tefelski, Dariusz; Polański, Aleksander; Skubalski, Janusz

    2012-06-01

    Concrete is one of the main materials used for gamma and neutron shielding. While in case of gamma rays an increase in density is usually efficient enough, protection against neutrons is more complex. The aim of this paper is to show the possibility of using the Monte Carlo codes for evaluation and optimization of concrete mix to reach better neutron shielding. Two codes (MCNPX and SPOT — written by authors) were used to simulate neutron transport through a wall made of different concretes. It is showed that concrete of higher compressive strength attenuates neutrons more effectively. The advantage of heavyweight concrete (with barite aggregate), usually used for gamma shielding, over the ordinary concrete was not so clear. Neutron shielding depends on many factors e.g. neutron energy, barrier thickness and atomic composition. All this makes a proper design of concrete as a very important issue for nuclear power plant safety assurance.

  15. Design, fabrication, and testing of gadolinium-shielded metal fuel samples in the hydraulic tube of the high flux isotope reactor

    The use of hydraulic rabbit capsules inserted into and ejected from the core of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) during full power operation allows for precise control of the neutron fluence in fueled experiments. Rabbit capsules with strong thermal neutron absorbers must be used to screen out thermal neutrons, thereby reducing the heat generation rate while maintaining the fast neutron flux that produces displacement damage similar to fast reactor type conditions. However, rapid insertion and ejection of rabbit capsules containing a strong neutron absorber causes a reactivity response in the reactor that has the potential to engage the HFIR safety response system which could result in an unplanned shutdown. Therefore, a set of tests were performed to provide the data needed to establish limits on the reactivity worth that can be ejected from the hydraulic facility without causing a reactor shutdown. This paper will describe the design, operation, and results of the reactivity measurements undertaken to understand the reactor response to insertion of the gadolinium-lined rabbit capsules. (author)

  16. Cloud immersion building shielding factors for US residential structures

    This paper presents validated building shielding factors designed for contemporary US housing-stock under an idealized, yet realistic, exposure scenario within a semi-infinite cloud of radioactive material. The building shielding factors are intended for use in emergency planning and level three probabilistic risk assessments for a variety of postulated radiological events in which a realistic assessment is necessary to better understand the potential risks for accident mitigation and emergency response planning. Factors are calculated from detailed computational housing-units models using the general-purpose Monte Carlo N-Particle computational code, MCNP5, and are benchmarked from a series of narrow- and broad-beam measurements analyzing the shielding effectiveness of ten common general-purpose construction materials and ten shielding models representing the primary weather barriers (walls and roofs) of likely US housing-stock. Each model was designed to scale based on common residential construction practices and include, to the extent practical, all structurally significant components important for shielding against ionizing radiation. Calculations were performed for floor-specific locations as well as for computing a weighted-average representative building shielding factor for single- and multi-story detached homes, both with and without basement, as well for single-wide manufactured housing-units. (paper)

  17. Radiation shielding for the Fermilab Vertical Cavity Test Facility

    Ginsburg, Camille; Rakhno, Igor; /Fermilab

    2010-03-01

    The results of radiation shielding studies for the vertical test cryostat VTS1 at Fermilab performed with the codes FISHPACT and MARS15 are presented and discussed. The analysis is focused on operations with two RF cavities in the cryostat. The vertical cavity test facility (VCTF) for superconducting RF cavities in Industrial Building 1 at Fermilab has been in operation since 2007. The facility currently consists of a single vertical test cryostat VTS1. Radiation shielding for VTS1 was designed for operations with single 9-cell 1.3 GHz cavities, and the shielding calculations were performed using a simplified model of field emission as the radiation source. The operations are proposed to be extended in such a way that two RF cavities will be in VTS1 at a time, one above the other, with tests for each cavity performed sequentially. In such a case the radiation emitted during the tests from the lower cavity can, in part, bypass the initially designed shielding which can lead to a higher dose in the building. Space for additional shielding, either internal or external to VTS1, is limited. Therefore, a re-evaluation of the radiation shielding was performed. An essential part of the present analysis is in using realistic models for cavity geometry and spatial, angular and energy distributions of field-emitted electrons inside the cavities. The calculations were performed with the computer codes FISHPACT and MARS15.

  18. Shielding of moving line charges

    Wang, Youmei; He, Bingyu [Department of Physics, School of Science, Hangzhou Dianzi University, Hangzhou 310018 (China); Yu, Wei [Shanghai Institute of Optics and Fine Mechanics, Chinese Academy of Sciences, Shanghai 201800 (China); Yu, M.Y., E-mail: myyu@zju.edu.cn [Institute for Fusion Theory and Simulation and Department of Physics, Zhejiang University, Hangzhou 310027 (China); Institute for Theoretical Physics I, Ruhr University, D-44780 Bochum (Germany)

    2015-07-03

    A charged object moving in plasma can excite plasma waves that inevitably modify its Debye shielding characteristics. When the excited waves propagate sufficiently fast, the shielding can even break down. Here the properties of finite amplitude plasma waves excited by a moving line charge are investigated. It is found that when the speed of the latter is close to but less than the thermal speed of the background plasma electrons, only a localized disturbance in the form of a soliton that moves together with the line charge is excited. That is, the line charge is well shielded even though it is moving at a high speed and has generated a large local electrostatic field. However, for a pair of line charges moving together, such complete shielding behavior could not be found.

  19. 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.

  20. SNF shipping cask shielding analysis

    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

  1. Upgrade of the LHC magnet interconnections thermal shielding

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

  2. Shielding vacuum fluctuations with graphene

    Ribeiro, Sofia; Scheel, Stefan

    2013-01-01

    The Casimir-Polder interaction of ground-state and excited atoms with graphene is investigated with the aim to establish whether graphene systems can be used as a shield for vacuum fluctuations of an underlying substrate. We calculate the zero-temperature Casimir-Polder potential from the reflection coefficients of graphene within the framework of the Dirac model. For both doped and undoped graphene we show limits at which graphene could be used effectively as a shield. Additional results are...

  3. Composite Aerogel Multifoil Protective Shielding

    Jones, Steven M.

    2013-01-01

    New technologies are needed to survive the temperatures, radiation, and hypervelocity particles that exploration spacecraft encounter. Multilayer insulations (MLIs) have been used on many spacecraft as thermal insulation. Other materials and composites have been used as micrometeorite shielding or radiation shielding. However, no material composite has been developed and employed as a combined thermal insulation, micrometeorite, and radiation shielding. By replacing the scrims that have been used to separate the foil layers in MLIs with various aerogels, and by using a variety of different metal foils, the overall protective performance of MLIs can be greatly expanded to act as thermal insulation, radiation shielding, and hypervelocity particle shielding. Aerogels are highly porous, low-density solids that are produced by the gelation of metal alkoxides and supercritical drying. Aerogels have been flown in NASA missions as a hypervelocity particle capture medium (Stardust) and as thermal insulation (2003 MER). Composite aerogel multifoil protective shielding would be used to provide thermal insulation, while also shielding spacecraft or components from radiation and hypervelocity particle impacts. Multiple layers of foil separated by aerogel would act as a thermal barrier by preventing the transport of heat energy through the composite. The silica aerogel would act as a convective and conductive thermal barrier, while the titania powder and metal foils would absorb and reflect the radiative heat. It would also capture small hypervelocity particles, such as micrometeorites, since it would be a stuffed, multi-shock Whipple shield. The metal foil layers would slow and break up the impacting particles, while the aerogel layers would convert the kinetic energy of the particles to thermal and mechanical energy and stop the particles.

  4. Shielding requirements in helical tomotherapy

    Helical tomotherapy is a relatively new intensity-modulated radiation therapy (IMRT) treatment for which room shielding has to be reassessed for the following reasons. The beam-on-time needed to deliver a given target dose is increased and leads to a weekly workload of typically one order of magnitude higher than that for conventional radiation therapy. The special configuration of tomotherapy units does not allow the use of standard shielding calculation methods. A conventional linear accelerator must be shielded for primary, leakage and scatter photon radiations. For tomotherapy, primary radiation is no longer the main shielding issue since a beam stop is mounted on the gantry directly opposite the source. On the other hand, due to the longer irradiation time, the accelerator head leakage becomes a major concern. An analytical model based on geometric considerations has been developed to determine leakage radiation levels throughout the room for continuous gantry rotation. Compared to leakage radiation, scatter radiation is a minor contribution. Since tomotherapy units operate at a nominal energy of 6 MV, neutron production is negligible. This work proposes a synthetic and conservative model for calculating shielding requirements for the Hi-Art II TomoTherapy unit. Finally, the required concrete shielding thickness is given for different positions of interest

  5. Shielded electron microprobe analyzer for plutonium fuel

    Design, construction and performance test of a shielded electron microprobe analyzer for plutonium fuel are described. In the analyzer, the following modifications were made to Shimadzu ASM-SX (analyzer): (1) a shield of tungsten alloy is incorporated between the sample and the X-ray detector to examine highly radioactive fuel, (2) a magnetic shield against β-rays from the fuel is fitted to the electron detector, (3) a small sample-loading glove box is installed to transfer plutonium fuel safely to the analyzer, (4) a glove box containing a sample-surface treatment apparatus and a balance is connected to the sample-loading glove box, (5) for maintenance and repair of the analyzer by means of closed method, about thirty modifications are made. The performance test with nonradioactive materials showed that despite the above modifications, abilities of the original analyzer are all retained. And furthermore, the simulation test for irradiated fuel with 226Ra of dose rate 40 mR/hr at 30 cm showed that the X-ray peaks to noise ratios are unchanged by using a pulse height selector of the X-ray detector. (author)

  6. Integral Face Shield Concept for Firefighter's Helmet

    Abeles, F.; Hansberry, E.; Himel, V.

    1982-01-01

    Stowable face shield could be made integral part of helmet worn by firefighters. Shield, made from same tough clear plastic as removable face shields presently used, would be pivoted at temples to slide up inside helmet when not needed. Stowable face shield, being stored in helmet, is always available, ready for use, and is protected when not being used.

  7. 7 CFR 1c.112 - Review by institution.

    2010-01-01

    ... 7 Agriculture 1 2010-01-01 2010-01-01 false Review by institution. 1c.112 Section 1c.112 Agriculture Office of the Secretary of Agriculture PROTECTION OF HUMAN SUBJECTS § 1c.112 Review by institution... review and approval or disapproval by officials of the institution. However, those officials may...

  8. Shielding benchmark experiments through concrete and iron with high-energy proton and heavy ion accelerators

    The deep penetration of neutrons through thick shield has become a very serious problem in the shielding design of high-energy, high-intensity accelerator facility. In the design calculation, the Monte Carlo transport calculation through thick shields has large statistical errors and the basic nuclear data and model used in the existing Monte Carlo codes are not well evaluated because of very few experimental data. It is therefore strongly needed to do the deep penetration experiment as shielding benchmark for investigating the calculation accuracy. Under this circumference, we performed the following two shielding experiments through concrete and iron, one with a 800 MeV proton accelerator of the rutherford appleton laboratory (RAL), England and the other with a high energy heavy iron accelerator of the national institute of radiological sciences (NIRS), Japan. Here these two shielding benchmark experiments are outlined. (orig.)

  9. Shielding benchmark experiments through concrete and iron with high-energy proton and heavy ion accelerators

    Nakamura, T.; Sasaki, M.; Nunomiya, T.; Iwase, H. [Tohoku Univ., Sendai (Japan). Dept. of Quantum Science and Energy Engineering; Nakao, N.; Shibata, T. [High Energy Accelerator Research Organization (KEK), Ibaraki (Japan); Kim, E. [Japan Atomic Energy Research Inst. (JAERI), Ibaraki (Japan). Tokai Establishment; Kurosawa, T. [Japan Synchrotron Radiation Research Inst. (JASRI), Hyogo (Japan); Taniguchi, S. [Electrotechnical Lab. (ETL), Tsukuba, Ibaraki (Japan); Uwamino, Y.; Ito, S. [The Inst. of Physical and Chemical Research (RIKEN), Saitama (Japan); Fukumura, A. [National Inst. of Radiological Sciences (NIRS), Chiba (Japan); Perry, D.R.; Wright, P. [Rutherford Appleton Lab. (RAL), Didcot, Oxfordshire (United Kingdom). Health and Safety Group

    2001-07-01

    The deep penetration of neutrons through thick shield has become a very serious problem in the shielding design of high-energy, high-intensity accelerator facility. In the design calculation, the Monte Carlo transport calculation through thick shields has large statistical errors and the basic nuclear data and model used in the existing Monte Carlo codes are not well evaluated because of very few experimental data. It is therefore strongly needed to do the deep penetration experiment as shielding benchmark for investigating the calculation accuracy. Under this circumference, we performed the following two shielding experiments through concrete and iron, one with a 800 MeV proton accelerator of the rutherford appleton laboratory (RAL), England and the other with a high energy heavy iron accelerator of the national institute of radiological sciences (NIRS), Japan. Here these two shielding benchmark experiments are outlined. (orig.)

  10. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor.

  11. Calmodulin binding to recombinant myosin-1c and myosin-1c IQ peptides

    Cyr Janet L; Gillespie Peter G

    2002-01-01

    Abstract Background Bullfrog myosin-1c contains three previously recognized calmodulin-binding IQ domains (IQ1, IQ2, and IQ3) in its neck region; we identified a fourth IQ domain (IQ4), located immediately adjacent to IQ3. How calmodulin binds to these IQ domains is the subject of this report. Results In the presence of EGTA, calmodulin bound to synthetic peptides corresponding to IQ1, IQ2, and IQ3 with Kd values of 2–4 μM at normal ionic strength; the interaction with an IQ4 peptide was much...

  12. FIRE-RESISTANT SHIELDING COATING BASED ON SHUNGITE-CONTAINING PAINT

    BELOUSOVA Elena Sergeevna; NASONOVA Natalia Viktorovna; LYNKOV Leonid Mihailovich; BORBOTKO Timofei Valentinovich; LISOVSKIY Dmitriy Nikolaevich

    2013-01-01

    Today when specific shielded facilities are designed the construction materials and shields should meet a range of fire safety requirements. A composite coating on the basis of a water-based fire-resistant paint filled with shungite nanopowder can be applied onto walls, floors, ceilings and other surfaces in the shielded areas to reduce electromagnetic radiation and simultaneously to ensure fire safety. Shungit is a mineral with multilayer carbon fullerene globules which diameter is 10–30 nm....

  13. Where are we and where are we going in reactor shielding

    We can note substantial recent progress in most all aspects of reactor shielding and we anticipate continued gains. Methods and data may tend to assume greater stability in the future although the expected concern with new reactor systems such as those based on fusion make this less likely. Recent problems with neutron streaming may aid in gaining consideration of shielding earlier in the reactor design process. A set of possible challenges in reactor shielding is given below. (orig.)

  14. Analysis list: JMJD1C [Chip-atlas[Archive

    Full Text Available JMJD1C Blood + hg19 http://dbarchive.biosciencedbc.jp/kyushu-u/hg19/target/JMJD1C.1....tsv http://dbarchive.biosciencedbc.jp/kyushu-u/hg19/target/JMJD1C.5.tsv http://dbarchive.biosciencedbc.jp/kyushu-u/hg19/target/JM...JD1C.10.tsv http://dbarchive.biosciencedbc.jp/kyushu-u/hg19/colo/JMJD1C.Blood.tsv http://dbarchive.biosciencedbc.jp/kyushu-u/hg19/colo/Blood.gml ...

  15. Applicability of the three-dimensional transport code Tort to the shielding analysis of the prototype FBR Monju

    Shielding design of Monju was performed in 1980's by using the two-dimensional discrete ordinates transport code, DOT3.5. In view of complexity of the three-dimensional shielding geometry of Monju, the three-dimensional discrete ordinates transport code, TORT(2), has been applied to shielding measurement analyses of Monju in an attempt to prove practical usefulness of the code and to learn how much margin is associated with the shielding design performed by DOT3.5. This study has confirmed that TORT can practically be applied to the shielding measurement analyses of Monju, and has provided significant improvement in calculation accuracy thanks to its three-dimensional geometry employed, making the code applicable to the Monju shielding design evaluation analyses together with pre- and post-analyses of the shielding measurement now being planned. (authors)

  16. Pre-evaluation of fusion shielding benchmark experiment

    Hayashi, K.; Handa, H. [Hitachi Engineering Company, Ltd., Ibaraki (Japan); Konno, C. [Japan Atomic Energy Research Inst. Ibaraki (Japan)] [and others

    1994-12-31

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

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

    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.

  18. Radiation Shielding Analysis of Electron Beam Accelerator Facility

    The objective of this technical report are to establish the radiation shielding technology of a high-energy electron accelerator to the facilities which utilize with electron beam. The technologies of electron beam irradiation(300 KeV -10 MeV) demand on the diverse areas of material processing, surface treatment, treatments on foods or food processing, improvement of metal properties, semiconductors, and ceramics, sterilization of medical goods and equipment, treatment and control of contamination and pollution, and so on. In order to acquire safety design for the protection of personnel from the radiations produced by electron beam accelerators, it is important to develop the radiation shielding analysis technology. The shielding analysis are carried out by which define source term, calculation modelling and computer calculations for 2 MeV and 10 MeV accelerators. And the shielding analysis for irradiation dump shield with 10 MeV accelerators are also performed by solving the complex 3-D geometry and long computer run time problem. The technology development of shielding analysis will be contributed to extend the further high energy accelerator development

  19. Study of Neutron and Gamma Radiation Protective Shield

    Eskandar Asadi Amirabadi

    2013-08-01

    Full Text Available Due to the development of nuclear technology and use of these technologies in various fields of industry, medicine, research and etc, protection against radioactive rays is one of the most important topics in this field .The purpose of this is to reduce the dose rate from radioactive sources. The sources in terms of components are emitted various types of nuclear radiation with different energies. These radiations are involving of alpha particles, beta, and neutron and gamma radiation. Given that alpha and beta particles can be fully absorbed by the shield, the main issue in the debate protection radioactive rays is stopping of gamma rays and neutrons. Accordingly in shield design usually two types of radiation should be considered. First, X-rays and gamma rays, which have great influence, and by the mass of any suitable material, can be more efficiently attenuate the higher the density, the better the potential attenuation effect against gamma rays and the required shielding thickness decreases. The second type of radiation is neutrons. Often a combination of three materials is desirable that include heavy metals, light metals, and neutron-absorbing material to omit the slow neutrons through adsorption to the neutron shield. There are different materials that can be used to shielding against radioactive rays. The main materials that are used in protection include: water, lead, graphite, iron, compounds that contains B, concrete, and polyethylene. Accordingly, the main objective of this paper is evaluating the kind of shield against gamma and neutrons rays.

  20. Thermal Protection System (Heat Shield) Development - Advanced Development Project

    Kowal, T. John

    2010-01-01

    The Orion Thermal Protection System (TPS) ADP was a 3 1/2 year effort to develop ablative TPS materials for the Orion crew capsule. The ADP was motivated by the lack of available ablative TPS's. The TPS ADP pursued a competitive phased development strategy with succeeding rounds of development, testing and down selections. The Project raised the technology readiness level (TRL) of 8 different TPS materials from 5 different commercial vendors, eventual down selecting to a single material system for the Orion heat shield. In addition to providing a heat shield material and design for Orion on time and on budget, the Project accomplished the following: 1) Re-invigorated TPS industry & re-established a NASA competency to respond to future TPS needs; 2) Identified a potentially catastrophic problem with the planned MSL heat shield, and provided a viable, high TRL alternate heat shield design option; and 3) Transferred mature heat shield material and design options to the commercial space industry, including TPS technology information for the SpaceX Dragon capsule.

  1. Measurements for the JASPER program Axial Shield Experiment

    Muckenthaler, F.J.; Spencer, R.R.; Hunter, H.T. [Oak Ridge National Lab., TN (United States); Shono, A.; Chatani, K. [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)

    1991-08-01

    The Axial Shield Experiment was conducted at the Oak Ridge National Laboratory (ORNL) during 1990--1991 as part of the continuing series of eight experiments planned for the Japanese-American Shielding Program for Experimental Research (JASPER) program starting in 1986. The program is intended to provide support for the development of current designs proposed for advanced liquid metal reactor (LMR) system both in Japan and the United States. As in the previous two experiments, the same spectrum modifier was used to alter the Tower Shielding Reactor source spectrum to one representing the LMR neutron spectra directly above the core in the area of the fission-gas plenum. In one of the measurements the spectrum was further modified by the fission gas plenum. In all cases the modified spectrum was followed by combinations of seven hexagon assemblies that represented different coolant flow and shielding patterns within the assemblies. The varied configuration permitted not only a study of the different designs, but also allowed a comparison to be made of the relative neutron attenuation effectiveness of boron carbide and stainless steel in such designs. This experiment was the third in a series of eight experiments to be performed as part of a cooperative effort between the United States Department of Energy (US DOE) and the Japan Power Reactor and Nuclear Fuel Development Corporation (PNC). This experiment, as was the previous Radial Shield Attenuation and Fission Gas Plenum Experiments, intended to provide support for the development of advanced sodium-cooled reactors. 5 refs.

  2. Lead casting process of shielding container for transporting nuclear assembly

    The main radiation shielding of transporting container for the reactor coolant pump hydraulic assembly is a lead casting Layer. The authors investigate in detail the lead casting process. In order to assure casting quality, the different technical design for different parts is adopted. The shielding case is completed on its first cast by bottom casting, sequential solidify etc. But the lead layer of chassis base is cast at first, then pressed, followed by machining to the right size. Casting system assure the realization of technical design. Two parts cast finally prove that the performance is in accord with requirement through NDT

  3. Deep-penetration calculations in concrete and iron for shielding of proton therapy accelerators

    Proton accelerators in the energy range of approximately 200 MeV have become increasingly popular for cancer treatment in recent years. These proton therapy facilities usually involve bulky concrete or iron in their shielding design or accelerator structure. Simple shielding data, such as source terms or attenuation lengths for various proton energies and materials are useful in designing accelerator shielding. Understanding the appropriateness or uncertainties associated with these data, which are largely generated from Monte Carlo simulations, is critical to the quality of a shielding design. This study demonstrated and investigated the problems of deep-penetration calculations on the estimation of shielding parameters through an extensive comparison between the FLUKA and MCNPX calculations for shielding against a 200-MeV proton beam hitting an iron target. Simulations of double-differential neutron production from proton bombardment were validated by comparison with experimental data. For the concrete shielding, the FLUKA calculated depth–dose distributions were consistent with the MCNPX results, except for some discrepancies in backward directions. However, for the iron shielding, if FLUKA is used inappropriately then overestimation of neutron attenuation can be expected as shown by this work because of the multigroup treatment for low-energy neutrons in FLUKA. Two neutron energy group structures, three degrees of self-shielding correction, and two iron compositions were considered in this study. Significant variation of the resulting attenuation lengths indicated the importance of problem-dependent multigroup cross sections and proper modeling of iron composition in deep-penetration calculations.

  4. Near zero magnetic fields with superconducting shields

    The author's original motivation for developing the ultra low magnetic shielding was for an experiment to precisely determine h/me using rotating superconducting rings. The author first used the technique for precise magnetic charge measurements of the niobium sphere fractional electrical charge candidates from the Fairbank--Hebard--LaRue--Phillips experiments. A brief description of the technique is presented, together with a summary of work on absolute magnetometry using SQUID sensors and its application to the design of other instruments which use the ultra low field environment. Prospects for future improvements are discussed

  5. Shielding analysis for ITER equatorial bio-shield plug

    ITER equatorial port cell outside bio-shield plug is a place for allowing free personnel access after shutdown which accommodates various sensitive equipment and pipes. To ensure the personnel safety in port cell after shutdown, the distribution of dose rate in port cell was studied. Based on VisualBUS (CAD-Based Multi-Functional 4D Neutronics Simulation System), dose rate calculations were completed in port cell after shutdown. The result showed that dose rates in port cell are still 2 orders of magnitude more than desired limit (10 μSv/h) after one day shutdown. The optimization of bio-shield was needed. (authors)

  6. EMPLACEMENT DRIFT SHIELDING CALCULATION

    A. Nielsen

    1999-10-13

    The purpose of this analysis is to determine the structural response of a TRIGA Department of Energy (DOE) spent nuclear fuel (SNF) codisposal canister placed in a 5-Defense High Level Waste (DHLW) waste package (WP) and subjected to a tipover design basis event (DBE) dynamic load; the results will be reported in terms of displacements and stress magnitudes. This activity is associated with the WP design.

  7. System for imaging plutonium through heavy shielding

    A single pinhole can be used to image strong self-luminescent gamma-ray sources such as plutonium on gamma scintillation (Anger) cameras. However, if the source is weak or heavily shielded, a poor signal to noise ratio can prevent acquisition of the image. An imaging system designed and built at Los Alamos National Laboratory uses a coded aperture to image heavily shielded sources. The paper summarizes the mathematical techniques, based on the Fast Delta Hadamard transform, used to decode raw images. Practical design considerations such as the phase of the uniformly redundant aperture and the encoded image sampling are discussed. The imaging system consists of a custom designed m-sequence coded aperture, a Picker International Corporation gamma scintillation camera, a LeCroy 3500 data acquisition system, and custom imaging software. The paper considers two sources - 1.5 mCi 57Co unshielded at a distance of 27 m and 220 g of bulk plutonium (11.8% 240Pu) with 0.3 cm lead, 2.5 cm steel, and 10 cm of dense plastic material at a distance of 77.5 cm. Results show that the location and geometry of a source hidden in a large sealed package can be determined without having to open the package. 6 references, 4 figures

  8. Photon attenuation characteristics of radiation shielding materials

    In the design and construction of installation housing high intensity radiation sources and other radiation generating equipment, a variety of shielding materials are used to minimise exposure to individual. Among the materials, lead is best known for radiation shielding characteristics due to their high density and atomic number. Commercial and barium enriched cement, apart from better compressive strength, smoother surface finish and high abrasion resistance, offers adequate shielding to gamma radiations. Although photon attenuation data are available in literature, it is necessary to test these commercially available material experimentally for their radiation shielding efficiency before putting them in to regular use. In the present work, attenuation characteristics of lead. commercial cement and barium enriched cement supplied by a manufacturing firm have been studied for photons of 662 and 1250keV from Cs-137 and Co-60. The radiographic sources of Cs-137 and Co-60 of radioactive strength of 260 and 30 mCi respectively were utilised in the present investigation. Experimental measurements were done with gamma radiography survey meter MR 4500A placed at a distance of 2 meters from the source. Attenuation coefficients for photons in commercial cement, barite and lead were determined experimentally through photon transmission measurements performed under broad beam counting geometry. The absorbers used were in form of thin sheets of lead, commercial cements and barite of uniform thicknesses. These thin sheets were weighed accurately on an analytical balance and from their measured area, thicknesses proportional to area density in gram.cm-2 were determined. The average thickness of each absorber varied from a few milligram to several gram per cm-2. Higher thicknesses were obtained by stacking the absorbers with each other. Each absorber of specified thickness was interposed between the source and detector such that the primary beam is incident normally on its

  9. 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.

  10. Analytical-HZETRN Model for Rapid Assessment of Active Magnetic Radiation Shielding

    Washburn, S. A.; Blattnig, S. R.; Singleterry, R. C.; Westover, S. C.

    2014-01-01

    The use of active radiation shielding designs has the potential to reduce the radiation exposure received by astronauts on deep-space missions at a significantly lower mass penalty than designs utilizing only passive shielding. Unfortunately, the determination of the radiation exposure inside these shielded environments often involves lengthy and computationally intensive Monte Carlo analysis. In order to evaluate the large trade space of design parameters associated with a magnetic radiation shield design, an analytical model was developed for the determination of flux inside a solenoid magnetic field due to the Galactic Cosmic Radiation (GCR) radiation environment. This analytical model was then coupled with NASA's radiation transport code, HZETRN, to account for the effects of passive/structural shielding mass. The resulting model can rapidly obtain results for a given configuration and can therefore be used to analyze an entire trade space of potential variables in less time than is required for even a single Monte Carlo run. Analyzing this trade space for a solenoid magnetic shield design indicates that active shield bending powers greater than 15 Tm and passive/structural shielding thicknesses greater than 40 g/cm2 have a limited impact on reducing dose equivalent values. Also, it is shown that higher magnetic field strengths are more effective than thicker magnetic fields at reducing dose equivalent.

  11. DECOVALEX ll PROJECT. Technical Report Task 1C

    Jing, L.; Stephansson, O. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Civil and Environmental Engineering; Knight, L.J. [United Kingdom Nirex Ltd., Harwell (United Kingdom); Kautsky, F. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Tsang, C-F. [Lawrence Berkeley National Lab., Berkeley, CA (United States). Earth Science Division

    1999-05-01

    due to shaft excavation and a reinforcement study, which made the task relevant to issues of constructability and repository design. Presented in this report are the definition, conceptual models, comparison and analysis of the numerical results and concluding remarks of Task 1C. There were six research teams who participated in this task with different conceptual models and numerical models.

  12. DECOVALEX ll PROJECT. Technical Report Task 1C

    due to shaft excavation and a reinforcement study, which made the task relevant to issues of constructability and repository design. Presented in this report are the definition, conceptual models, comparison and analysis of the numerical results and concluding remarks of Task 1C. There were six research teams who participated in this task with different conceptual models and numerical models

  13. Radiation shield for nuclear reactors

    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)

  14. Shielding walls against ionizing radiation

    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.)

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

    CERN Library

    2013-01-01

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

  16. Effectiveness of custom neutron shielding in the maze of radiotherapy accelerators

    An investigation was performed to examine the neutron dose equivalent in a radiotherapy maze lined with a customised neutron shielding material. The accelerator investigated was a Varian Clinac 2100C/D using 18 MV photons, and the neutron shielding utilised at this centre was PremadexTM commercially available neutron shielding. Based on Monte Carlo simulations, properly installed customised neutron shielding may reduce the neutron dose equivalent by up to a factor of 8 outside the maze, depending upon the installation. In addition, it was determined that the neutron dose near the entrance to the maze may be reduced by approximately 40% by using customised neutron shielding in the maze, as compared with a facility not using this shielding. This would have a positive dose-saving effect in doorless maze designs. (author)

  17. A comparative study on space radiation shielding materials for LEO satellite

    In this study, a proton shielding design was optimized for choice of shielding materials by using a Monte Carlo transport code system LCS(LAHET code system). Proton shielding calculation was done for a mono-directional, mono-energetic beam impinging on the slab shields. Effect of muti-layer shields(Al-Ta-Al, Al-borated polyethelene(BPT)-Al) for secondary particle reflecting was compared with single layer of aluminium. Al-BPT-Al is better than others. And also, production of secondary neutron was smaller. Based on this result, proton shielding calculation was applied for KISAT-1. The best choice was Al-BPT-AL and the worst was Al-Ta-Al among three. Besides, it was found that there is no significant dose effect from secondary neutron in LEO

  18. Survivor shielding. Part B. Improvements in building shielding

    Most atomic-bomb survivor doses are affected by the shielding provided by wooden structures, either in which the survivor resides or which lie between him or her and the epicenter. In the dosimetry system, this shielding of survivors can be described by a transmission factor (TF), which is the ratio of the dose with and without the structures being present. The TF typically ranges between 0.3 and 1.0. After DS86 was implemented at RERF, several of the shielding categories were examined and found to either have a bias or an excessive uncertainty that could readily be removed. In 1989, a large bimodal uncertainty in the 9-parameter category 'FS=0' was identified. Corrective action was proposed and is now implemented in DS02. In 2002, a dose bias in large wooden buildings, such as schools, was identified and a correction is implemented in DS02. A correction is also implemented in DS02 to take care of a large uncertainty in the globe-house shielding. (J.P.N.)

  19. Design of software for calculation of shielding based on various standards radiodiagnostic calculation; Diseno de un software para el calculo de blindajes en radiodiagnostico basado en varios estandares de calculo

    Falero, B.; Bueno, P.; Chaves, M. A.; Ordiales, J. M.; Villafana, O.; Gonzalez, M. J.

    2013-07-01

    The aim of this study was to develop a software application that performs calculation shields in radiology room depending on the type of equipment. The calculation will be done by selecting the user, the method proposed in the Guide 5.11, the Report 144 and 147 and also for the methodology given by the Portuguese Health Ministry. (Author)

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

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