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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. 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

  11. 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

  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. 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

  17. 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)

  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. 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)

  15. 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)

  16. 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.

  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. 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.

  4. 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)

  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. 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.

  18. 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

  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.

  2. 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

  3. 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

  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. 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.

  10. 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.

  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. 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

  5. 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 ...

  6. 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)

  7. 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

  8. 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

  9. 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)

  10. 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

  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. 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)

  8. 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

  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