WorldWideScience

Sample records for biological shielding

  1. Mechanical properties of JPDR biological shield concrete

    International Nuclear Information System (INIS)

    Idei, Yoshio; Kamata, Hiroshi; Akutsu, Youichi; Onizawa, Kunio; Nakajima, Nobuya; Sukegawa, Takenori; Kakizaki, Masayoshi.

    1990-11-01

    Plant life of nuclear power plant will be determined by the aging degradation of main components and structures because of the difficulty and the cost of the replacement. These components are the reactor pressure vessel, concrete structures and cables. Authors have performed the investigation of JPDR biological shield which was the succeeded in first generating electricity in Japan and is now being decommissioned in JAERI. The test core samples were bored from the shield concrete and tested to obtain the mechanical properties. Test results are summarized as below, (1) Peak value of fast neutron dose was estimated as 1 x 10 18 n/cm 2 which is equivalent to the dose at the end of life for commercial power reactor. (2) Averaged compressive strength of all specimens had been increased about 20 % compared with initial design strength. (3) It was identified that the compressive strength had a little trend to increase with the increase of neutron dose within the dose range obtained in this study. (4) Tensile strength, Elastic modulus and Poisson's ratio showed little effect of neutron dose. (5) It was suggested that the inside and the mid-section liners were effective to keep the water in concrete and to avoid the reduction in strength. (author)

  2. The manufacturing of depleted uranium biological shield components

    International Nuclear Information System (INIS)

    Metelkin, J.A.

    1998-01-01

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

  3. Biological shield design for a 10 MeV Rhodotron

    International Nuclear Information System (INIS)

    Khalafi, H.; Ghane, A.; Safaei Arshi, S.; Tabakh, F.

    2012-01-01

    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.

  4. Abrasive water jet cutting technique for biological shield concrete dismantlement

    International Nuclear Information System (INIS)

    Konno, T.; Narazaki, T.; Yokota, M.; Yoshida, H.; Miura, M.; Miyazaki, Y.

    1987-01-01

    The Japan Atomic Energy Research Institute (JAERI) is developing the abrasive-water jet cutting system to be applied to dismantling the biological shield walls of the JPDR as a part of the reactor dismantling technology development project. This is a total system for dismantling highly activated concrete. The concrete biological shield wall is cut into blocks by driving the abrasive-water jet nozzle, which is operated with a remote, automated control system. In this system, the concrete blocks are removed to a container, while the slurry and dust/mist which are generated during cutting are collected and treated, both automatically. It is a very practical method and will quite probably by used for actual dismantling of commercial power reactors in the future because it can minimize workers' exposure to radioactivity during dismantling, contributes to preventing diffusion of radiation, and reduces the volume of contaminated secondary waste

  5. Development of the cutting machine for the biological shield wall

    International Nuclear Information System (INIS)

    Yokota, Mitsuo; Hasegawa, Tetsuo; Kohyama, Kazunori.

    1987-01-01

    22 years have passed since the first commercial nuclear power plant operation in Japan. At present, there were 33 commercial nuclear power plants in operation, supplying about 25 percent of total electricity. Some of them are going to be terminated in the near future and enter into the decommissioning stage. Therefore, it is now necessary to developed decommissioning technologies, including dismantling techniques of these power plants. The development of a concrete cutting machine is one of the most important items applicable to dismantling biological shield walls of the plants. This paper describes the outline of the cutting machine developed for the biological shield wall demolition of the Japan Power Demonstration Reactor (JPDR) including actual decommissioning works tested. (author)

  6. Soil biological shield exposed to high energy neutrons; Zemlja kao bioloski stit od neutrona visokih energija

    Energy Technology Data Exchange (ETDEWEB)

    Simovic, R; Marinkovic, N [Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1993-04-15

    Shielding efficiency of soil biological shield exposed to high energy neutrons was investigated. Dose rate equivalents for neutrons, secondary gamma and gamma radiation were computed on the surface of soil slabs having different thicknesses. Yields of primary and secondary nuclear radiation in the total dose were evaluated. Influence of the incident neutron spectrum, water content and chemical composition of the material on its shielding efficiency was examined. It was found that the soil density and the water content determine the quality of biological shield, the influence of other factors being less important. Comparison of shielding efficiencies for soil with sand, brick and ordinary concrete shields was done.

  7. Radiation-resistant composite for biological shield of personnel

    Science.gov (United States)

    Barabash, D. E.; Barabash, A. D.; Potapov, Yu B.; Panfilov, D. V.; Perekalskiy, O. E.

    2017-10-01

    This article presents the results of theoretical and practical justification for the use of polymer concrete based on nonisocyanate polyurethanes in biological shield structures. We have identified the impact of ratio: polymer - radiation-resistant filling compound on the durability and protection properties of polymer concrete. The article expounds regression dependence of the change of basic properties of the aforementioned polymer concrete on the absorbed radiation dose rate. Synergy effect in attenuation of radioactivity release in case of conjoint use of hydrogenous polymer base and radiation-resistant powder is also addressed herein.

  8. Removal of concrete layers from biological shields by microwaves

    International Nuclear Information System (INIS)

    Wace, P.F.; Harker, A.H.; Hills, D.L.

    1990-01-01

    A comprehensive literature review has been carried out, to provide information for an experimental programme and equipment design. Mathematical modelling of the microwave and power fields in a concrete block, both steel reinforced and unreinforced, subjected to a microwave attack at two frequencies, has been carried out and estimates of the likely temperature rise with time obtained. A method of launching microwaves into concrete has been established from theoretical considerations and from the findings of the literature review. Equipment for laboratory trials has been designed and assembled using an 896 MHz, 25 kW microwave generator. Reinforced concrete blocks, 0.6 m in dimension and representing the concrete in a Magnox reactor biological shield, have been attacked at different power levels and the surface removed to the depth of the reinforcing steel (100 mm). Outline proposals for the design of a remotely operated prototype microwave machine for stripping the surface of large concrete test panels have been prepared. (author)

  9. Bragg Curve, Biological Bragg Curve and Biological Issues in Space Radiation Protection with Shielding

    Science.gov (United States)

    Honglu, Wu; Cucinotta, F.A.; Durante, M.; Lin, Z.; Rusek, A.

    2006-01-01

    The space environment consists of a varying field of radiation particles including high-energy ions, with spacecraft shielding material providing the major protection to astronauts from harmful exposure. Unlike low-LET gamma or X-rays, the presence of shielding does not always reduce the radiation risks for energetic charged particle exposure. Since the dose delivered by the charged particle increases sharply as the particle approaches the end of its range, a position known as the Bragg peak, the Bragg curve does not necessarily represent the biological damage along the particle traversal since biological effects are influenced by the track structure of both primary and secondary particles. Therefore, the biological Bragg curve is dependent on the energy and the type of the primary particle, and may vary for different biological endpoints. To achieve a Bragg curve distribution, we exposed cells to energetic heavy ions with the beam geometry parallel to a monolayer of fibroblasts. Qualitative analyses of gamma-H2AX fluorescence, a known marker of DSBs, indicated increased clustering of DNA damage before the Bragg peak, enhanced homogenous distribution at the peak, and provided visual evidence of high linear energy transfer (LET) particle traversal of cells beyond the Bragg peak. A quantitative biological response curve generated for micronuclei (MN) induction across the Bragg curve did not reveal an increased yield of MN at the location of the Bragg peak. However, the ratio of mono-to bi-nucleated cells, which indicates inhibition in cell progression, increased at the Bragg peak location. These results, along with other biological concerns, show that space radiation protection with shielding can be a complicated issue.

  10. Attenuation of fast neutron in concretes for biological shielding

    International Nuclear Information System (INIS)

    Labrada, A.; Chavez, A.; Gonzalez Mateu, D.; Desdin, F.; Tenjeiro, J.I.; Tellez, E.

    1993-01-01

    The attenuation of neutrons emitted by an 10 6 n/s. Am-Be source, in concretes elaborated with different aggregates is discussed in this paper. Two measurement methods were used an dosimetric system with Bonner spheres and 6 LiI(Eu) detector, and LAVSAN dielectric nuclear track detectors - with 238 U converts. The concretes elaborated with magnetite is reported as the best for neutron shielding while the Bauxite is not advisable for this purpose

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

    International Nuclear Information System (INIS)

    Kansouh, W.A.

    2012-01-01

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

  12. The biological shield of a high-intensity spallation source: a monte Carlo design study

    International Nuclear Information System (INIS)

    Koprivnikar, I.; Schachinger, E.

    2004-01-01

    The design of high-intensity spallation sources requires the best possible estimates for the biological shield. The applicability of three-dimensional Monte Carlo simulation in the design of the biological shield of a spallation source will be discussed. In order to achieve reasonable computing times along with acceptable accuracy, biasing techniques are to be employed and it was the main purpose of this work to develop a strategy for an effective Monte Carlo simulation in shielding design. The most prominent MC computer codes, namely MCNPX and FLUKA99, have been applied to the same model spallation source (the European Spallation Source) and on the basis of the derived strategies, the design and characteristics of the target station shield are discussed. It is also the purpose of the paper to demonstrate the application of the developed strategy for the design of beam lines with their shielding using as an example the target-moderator-reflector complex of the ESS as the primary particle source. (author)

  13. Economic analysis of sectional concrete blocks uses in biological shieldings

    International Nuclear Information System (INIS)

    Ivanov, V.N.

    1977-01-01

    The relative economy of different structural embodiments of the biological protection of a research reactor has been evaluated. The alternatives include cast in-situ concrete and prefabricated blocks with different linear dimension tolerances (+-2, +-5 and +-7 mm). The cost-benefit estimates have been done according to the reduced cost calculated for the final products - the erected structures. It has been found that the optimum tolerances for 6 meter-long blocks are not less than +-5 mm for the other linear dimensions. The optimum concrete block volume for dismountable structures is 1 to 1.5 m 3 and for prefabricated protection structures -more than 4 m 3

  14. Characterization of Radiation Fields for Assessing Concrete Degradation in Biological Shields of NPPs

    Science.gov (United States)

    Remec, Igor; Rosseel, Thomas M.; Field, Kevin G.; Pape, Yann Le

    2017-09-01

    Life extensions of nuclear power plants (NPPs) to 60 years of operation and the possibility of subsequent license renewal to 80 years have renewed interest in long-term material degradation in NPPs. Large irreplaceable sections of most nuclear generating stations are constructed from concrete, including safety-related structures such as biological shields and containment buildings; therefore, concrete degradation is being considered with particular focus on radiation-induced effects. Based on the projected neutron fluence values (E > 0.1 MeV) in the concrete biological shields of the US pressurized water reactor fleet and the currently available data on radiation effects on concrete, some decrease in mechanical properties of concrete cannot be ruled out during extended operation beyond 60 years. An expansion of the irradiated concrete database is desirable to ensure reliable risk assessment for extended operation of nuclear power plants.

  15. Calculation of neutron fluxes in biological shield of the TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Bozic, M.; Zagar, T.; Ravnik, M.

    2001-01-01

    The complete calculation of neutron fluxes in biological shield and verification with experimental results is presented. Calculated results are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Experimental results used for comparison are available from irradiation experiment with selected type of concrete and other materials in irradiation channel 4 in TRIGA Mark II reactor. These experimental results were used as a benchmark. Homogeneous type of problem (without inserted irradiation channel) and problem with asymmetry (inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. Deviation from material data set up as original parameters is also considered (first of all presence of water in concrete and density of concrete) for type of concrete in biological shield and for selected type of concrete in irradiation channel. BUGLE-96 (47 neutron energy groups) library is used. Excellent agreement between calculated and experimental results for reaction rate is received.(author)

  16. Modelling the Influence of Shielding on Physical and Biological Organ Doses

    CERN Document Server

    Ballarini, Francesca; Ferrari, Alfredo; Ottolenghi, Andrea; Pelliccioni, Maurizio; Scannicchio, Domenico

    2002-01-01

    Distributions of "physical" and "biological" dose in different organs were calculated by coupling the FLUKA MC transport code with a geometrical human phantom inserted into a shielding box of variable shape, thickness and material. While the expression "physical dose" refers to the amount of deposited energy per unit mass (in Gy), "biological dose" was modelled with "Complex Lesions" (CL), clustered DNA strand breaks calculated in a previous work based on "event-by-event" track-structure simulations. The yields of complex lesions per cell and per unit dose were calculated for different radiation types and energies, and integrated into a version of FLUKA modified for this purpose, allowing us to estimate the effects of mixed fields. As an initial test simulation, the phantom was inserted into an aluminium parallelepiped and was isotropically irradiated with 500 MeV protons. Dose distributions were calculated for different values of the shielding thickness. The results were found to be organ-dependent. In most ...

  17. Radiological characterization of the concrete biological shield of the APSARA reactor

    OpenAIRE

    Srinivasan Priya; Srinivasan Panchapakesan; Thomas Shibu; Gopalakrishnan R.K.; Goswami A.

    2013-01-01

    The first Indian research reactor, APSARA, was utilized for various R&D programmes from 1956 until its shutdown in 2009. The biological shield of the reactor developed residual activity due to neutron irradiation during the operation of the reactor. Dose rate mapping and in-situ gamma spectrometry of the concrete structures of the reactor pool were carried out. Representative concrete samples collected from various locations were subjected to high-resolution gamma spectrometry analysis....

  18. Application of a calculational model for thermal neutrons through biological shields

    Energy Technology Data Exchange (ETDEWEB)

    Hathout, A M [Nuclear engineering safety department, national center for nuclear safety and radiation, Nasr City Cairo, (Egypt)

    1995-10-01

    In this work a computational program, based on the Boltzmann transport integrodifferential equation, is applied. The scattering kernel is represented by the synthetic scattering model. The behaviour of thermal neutron in hydrogenous materials, which can be used as biological shields, are studied. These materials are water, polyethylene, Oak-Ridge concrete, ordinary concrete and manganese concrete. The data obtained are presented in tables. The results are analysed and compared with similar experimental values. Safety evaluation and environmental impact are discussed. 2 tabs.

  19. Improvements at the biological shielding of BNCT research facility in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Souza, Gregorio Soares de

    2011-01-01

    The technique of neutron capture in boron is a promising technique in cancer treatment, it uses the high LET particles from the reaction 10 B (n, α) 7 Li to destroy cancer cells.The development of this technique began in the mid-'50s and even today it is the object of study and research in various centers around the world, Brazil has built a facility that aims to conduct research in BNCT, this facility is located next to irradiation channel number three at the research nuclear reactor IEA-R1 and has a biological shielding designed to meet the radiation protection standards. This biological shielding was developed to allow them to conduct experiments with the reactor at maximum power, so it is not necessary to turn on and off the reactor to irradiate samples. However, when the channel is opened for experiments the background radiation in the experiments salon increases and this background variation makes it impossible to perform measurements in a neutron diffraction research that utilizes the irradiation channel number six. This study aims to further improve the shielding in order to minimize the variation of background making it possible to perform the research facility in BNCT without interfering with the action of the research group of the irradiation channel number six. To reach this purpose, the code MCNP5, dosimeters and activation detectors were used to plan improvements in the biological shielding. It was calculated with the help of the code an improvement that can reduce the average heat flow in 71.2% ± 13 and verified experimentally a mean reduce of 70 ± 9% in dose due to thermal neutrons. (author)

  20. The application of semianalytic method for calculating the thickness of biological shields of nuclear reactors. Part 2. Attenuation of gamma rays. An example of shield's thickness calculation

    International Nuclear Information System (INIS)

    Lukaszek, W.; Kucypera, S.

    1982-01-01

    The semianalytic method was used for calculating the attenuation of gamma rays and the thickness of biological shield of graphite moderated reactor. A short description of computer code as well as the exemplary results of calculations are given. (A.S.)

  1. Neutron flux measurements at the TRIGA reactor in Vienna for the prediction of the activation of the biological shield

    International Nuclear Information System (INIS)

    Merz, Stefan; Djuricic, Mile; Villa, Mario; Boeck, Helmuth; Steinhauser, Georg

    2011-01-01

    The activation of the biological shield is an important process for waste management considerations of nuclear facilities. The final activity can be estimated by modeling using the neutron flux density rather than the radiometric approach of activity measurements. Measurement series at the TRIGA reactor Vienna reveal that the flux density next to the biological shield is in the order of 10 9 cm -2 s -1 at maximum power; but it is strongly influenced by reactor installations. The data allow the estimation of the final waste categorization of the concrete according to the Austrian legislation. - Highlights: → Neutron activation is an important process for the waste management of nuclear facilities. → Biological shield of the TRIGA reactor Vienna has been topic of investigation. → Flux values allow a categorization of the concrete concerning radiation protection legislation. → Reactor installations are of great importance as neutron sources into the biological shield. → Every installation shows distinguishable flux profiles.

  2. Photon spectrum behind biological shielding of the LVR-15 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Klupak, V.; Viererbl, L.; Lahodova, Z.; Marek, M.; Vins, M. [Research Centre Rez Ltd., Husinec-Rez 130 (Czech Republic)

    2011-07-01

    The LVR-15 reactor is a light water research reactor situated at the Research Centre Rez, near Prague. It operates as a multipurpose facility with a maximum thermal power of 10 MW. The reactor core usually contains from 28 to 32 fuel assemblies with a total mass of {sup 235}U of about 5 kg. Emitted radiation from the fuel caused by fission is shielded by moderating water, a steel reactor vessel, and heavy concrete. This paper deals with measurement and analysis of the gamma spectrum near the outer surface of the concrete wall, behind biological shielding, mainly in the 3- to 10-MeV energy range. A portable HPGe detector with a portable multichannel analyzer was used to measure gamma spectra. The origin of energy lines in gamma detector spectra was identified. (authors)

  3. Inhomogeneity of neutron and gamma-ray attenuation in biological shields

    Energy Technology Data Exchange (ETDEWEB)

    El-bakkoush, F A; El-Ghobary, A M; Megahid, R M [Reactor and Neutron physics Department, Nuclear Research Center, A.E.A., Cairo (Egypt)

    1997-12-31

    Measurements have been carried-out to investigate the attenuation properties of some materials which are used as biological shields around nuclear radiation sources. Investigation was performed by measuring the transmitted fast neutron and gamma-spectra through cylindrical samples of magnetite- limonite, steel and cellulose shields. The neutron and gamma spectra were measured by a neutron-gamma spectrometer with stilbene scintillator. Discrimination between neutron and gamma pulses was achieved by a discrimination method. The obtained results are displayed in the form of neutron and gamma spectra and attenuation relations which are used to derive the total macroscopic cross-sections for neutrons and total linear attenuation coefficients for gamma-rays. The values of neutron and gamma relaxation lengths are also derived for the investigated materials. 10 figs., 1 tabs.

  4. Characterisation of the inventory of radioisotopes induced in the biological shield a WWER-440 reactor

    International Nuclear Information System (INIS)

    Feher, S.; Czifrus, Sz.; Zsolnay, E.M.; Szondi, E.

    2001-01-01

    A significant part of the radwaste originating from the decommissioning of NPPs is made up of the activated concrete and steel components of the biological shield. The paper presents the results of studies aimed at the determination of the amount of radionuclides accumulating in the serpentinous and ordinary concrete shield around the WWER-440 reactors of the Paks NPP. For the calculations, the reactor, vessel and shield were modelled in detail both in terms of geometry and material composition. The spatial and energy distribution of the activating neutron spectrum was determined by certain modules of SCALE 4.3 and the code TORT in two and three dimensions, while the activation was calculated using ORIGEN-S for 22 geometrical regions. The results showed that the activity of the concrete structures at final shutdown after 30 years of operation is approximately 50 TBq, which decreases to 20, 12, 1.1 TBq and 27 GBq after 1 month, 1 year, 10 and 100 years, respectively (Authors)

  5. Biological shield around the neutral beam injector ducts in the ITER conceptual design

    International Nuclear Information System (INIS)

    Maki, Koichi; Takatsu, Hideyuki; Satoh, Satoshi; Seki, Yasushi

    1994-01-01

    There are gaps between the toroidal field coils and neutral beam injector (NBI) duct wall for the thermal insulator in tokamak reactors such as ITER (International Thermonuclear Experimental Reactor). Neutrons stream through the duct, and some of them penetrate the wall and stream through the gaps. These neutrons activate the materials composing the duct wall, toroidal field coil (TFC) case and cryostat wall surfaces. The dose rate is enhanced just outside the cryostat around the ducts in the reactor room after reactor operation by activation. We investigated the gamma-ray dose rate just outside the cryostat after shutdown due to gamma-rays from activity induced by the neutrons streaming through the gaps. By evaluating the difference between the dose rate in models with and without gaps, we decided whether the thickness of the cryostat as biological shielding is sufficient or not. From these investigations, we recommend a cryostat design suitable for radiation shielding. Dose rates after shutdown at a point just outside the cryostat around the NBI ducts in the model with gaps are two orders larger than those without gaps. The value at this point is approximately 400 mrem h -1 (4 mSv h -1 ), which is two orders larger than the design value for workers to enter the reactor room. In order to reduce the dose rate after shutdown, a method of providing the shielding function of the cryostat is suggested. ((orig.))

  6. Activation of concrete samples from the biological shield of the ASTRA reactor

    International Nuclear Information System (INIS)

    Smecka, F.

    2006-09-01

    Drill cores from the biological shield of the ASTRA reactor in Seibersdorf were taken and milled because of the different size of the Baryt crystals in the concrete in order to get homogenous samples. The powder samples were put into bore holes of a graphite block which was placed into the thermal column of the TRIGA Mark II reactor. The block was irradiated for 10 minutes at a reactor power of 25 kW. After one hour the dose rate was examined and the samples were ready for further save handling. The gamma spectrum was measured with a Ge detector and the results were compared with simulation data. (nevyjel)

  7. Biological shielding design and qualification of concreting process for construction of electron beam irradiation facility

    International Nuclear Information System (INIS)

    Petwal, V.C.; Kumar, P.; Suresh, N.; Parchani, G.; Dwivedi, J.; Thakurta, A.C.

    2011-01-01

    A technology demonstration facility for irradiation of food and agricultural products is being set-up by RRCAT at Indore. The facility design is based on linear electron accelerator with maximum beam power of 10 kW and can be operated either in electron mode at 10 MeV or photon modes at 5/7.5 MeV. Biological shielding has been designed in accordance with NCRP 51 to achieve dose rate at all accessible points outside the irradiation vault less than the permissible limit of 0.1 mR/hr. In addition to radiation attenuation property, concrete must have satisfactory mechanical properties to meet the structural requirements. There are number of site specific variables which affect the structural, thermal and radiological properties of concrete, leading to considerable difference in actual values and design values. Hence it is essential to establish a suitable site and environmental specific process to cast the concrete and qualify the process by experimental measurement. For process qualification we have cast concrete test blocks of different thicknesses up to 3.25 m and evaluated the radiological and mechanical properties by radiometry, ultrasonic and mechanical tests. In this paper we describe the biological shielding design of the facility and analyse the results of tests carried out for qualification of the process. (author)

  8. INTOR radiation shielding for personnel access

    International Nuclear Information System (INIS)

    Gohar, Y.; Abdou, M.

    1981-01-01

    The INTOR reactor shield system consists of the blanket, bulk shield, penetration shield, component shield, and biological shield. The bulk shield consists of two parts: (a) the inboard shield; and (b) the outboard shield. The distinction between the different components of the shield system is essential to satisfy the different design constraints and achieve various objectives

  9. Evaluation and Verification of a Biological Shield in a SHARS Unit

    Energy Technology Data Exchange (ETDEWEB)

    Dhlomo, S.V.; Swart, H.S. [Compliance Management Department, Nuclear Liabilities Management, South African Nuclear Energy Corporation, P.O. Box 582, Pretoria 0001 (South Africa)

    2008-07-01

    The International Atomic Energy Agency (IAEA) Waste Technology Section with additional support from the U.S. National Nuclear Security Agency (NNSA) through the IAEA Nuclear Security Fund has funded the design, fabrication, evaluation, and testing of a portable hot cell intended to address the problem of disused SHARS in obsolete irradiation devices such as teletherapy heads and dry irradiators. This unit, designed and manufactured by the South African Nuclear Energy Corporation (Necsa), can be assembled, disassembled and packed inside two ISO containers and transported to the desired destination with relative ease. The unit was also licensed by the South African Regulator, the Department of Health (DoH), Directorate Radiation Control. This facility is used for the recovery and conditioning of orphaned/ abandoned or spent high activity radioactive sources from teletherapy units, gamma irradiators, and brachytherapy units. The hot cell was designed for a 3,7 E+13 Bq (1000 Ci) activity although it was demonstrated that it can handle activities of more than 7,4 E+13 Bq (2000 Ci) with ease. The biological shield of the SHARS facility consists of river sand sandwiched between metal plates, and a viewing window filled with a 50% zinc bromide solution. The shielding effectiveness of the river sand is evaluated experimentally by determining its density. The experimentally measured dose rates are compared to the dose rates estimated by computational codes. (authors)

  10. Evaluation and Verification of a Biological Shield in a SHARS Unit

    International Nuclear Information System (INIS)

    Dhlomo, S.V.; Swart, H.S.

    2008-01-01

    The International Atomic Energy Agency (IAEA) Waste Technology Section with additional support from the U.S. National Nuclear Security Agency (NNSA) through the IAEA Nuclear Security Fund has funded the design, fabrication, evaluation, and testing of a portable hot cell intended to address the problem of disused SHARS in obsolete irradiation devices such as teletherapy heads and dry irradiators. This unit, designed and manufactured by the South African Nuclear Energy Corporation (Necsa), can be assembled, disassembled and packed inside two ISO containers and transported to the desired destination with relative ease. The unit was also licensed by the South African Regulator, the Department of Health (DoH), Directorate Radiation Control. This facility is used for the recovery and conditioning of orphaned/ abandoned or spent high activity radioactive sources from teletherapy units, gamma irradiators, and brachytherapy units. The hot cell was designed for a 3,7 E+13 Bq (1000 Ci) activity although it was demonstrated that it can handle activities of more than 7,4 E+13 Bq (2000 Ci) with ease. The biological shield of the SHARS facility consists of river sand sandwiched between metal plates, and a viewing window filled with a 50% zinc bromide solution. The shielding effectiveness of the river sand is evaluated experimentally by determining its density. The experimentally measured dose rates are compared to the dose rates estimated by computational codes. (authors)

  11. Multi-objective optimization of a compact pressurized water nuclear reactor computational model for biological shielding design using innovative materials

    Energy Technology Data Exchange (ETDEWEB)

    Tunes, M.A., E-mail: matheus.tunes@usp.br [Department of Metallurgical and Materials Engineering, Escola Politécnica da Universidade de São Paulo, Av. Prof. Mello Moraes, 2463 – CEP 05508 – 030 São Paulo (Brazil); Oliveira, C.R.E. de, E-mail: cassiano@unm.edu [Department of Nuclear Engineering, The University of New Mexico, Farris Engineering Center, 221, Albuquerque, NM 87131-1070 (United States); Schön, C.G., E-mail: schoen@usp.br [Department of Metallurgical and Materials Engineering, Escola Politécnica da Universidade de São Paulo, Av. Prof. Mello Moraes, 2463 – CEP 05508 – 030 São Paulo (Brazil)

    2017-03-15

    Highlights: • Use of two n-γ transport codes leads to optimized model of compact nuclear reactor. • It was possible to safely reduce both weight and volume of the biological shielding. • Best configuration obtained by using new composites for both γ and n attenuation. - Abstract: The aim of the present work is to develop a computational model of a compact pressurized water nuclear reactor (PWR) to investigate the use of innovative materials to enhance the biological shielding effectiveness. Two radiation transport codes were used: the first one – MCNP – for the PWR design and the GEM/EVENT to simulate (in a 1D slab) the behavior of several materials and shielding thickness on gamma and neutron radiation. Additionally MATLAB Optimization Toolbox was used to provide new geometric configurations of the slab aiming at reducing the volume and weight of the walls by means of a cost/objective function. It is demonstrated in the reactor model that the dose rate outside biological shielding has been reduced by one order of magnitude for the optimized model compared with the initial configuration. Volume and weight of the shielding walls were also reduced. The results indicated that one-dimensional deterministic code to reach an optimized geometry and test materials, combined with a three-dimensional model of a compact nuclear reactor in a stochastic code, is a fast and efficient procedure to test shielding performance and optimization before the experimental assessment. A major outcome of this research is that composite materials (ECOMASS 2150TU96) may replace (with advantages) traditional shielding materials without jeopardizing the nuclear power plant safety assurance.

  12. Analysis of MIR-18 results for physical and biological dosimetry: radiation shielding effectiveness in LEO

    International Nuclear Information System (INIS)

    Cucinotta, F.A.; Wilson, J.W.; Williams, J.R.; Dicello, J.F.

    2000-01-01

    We compare models of radiation transport and biological response to physical and biological dosimetry results from astronauts on the Mir space station. Transport models are shown to be in good agreement with physical measurements and indicate that the ratio of equivalent dose from the Galactic Cosmic Rays (GCR) to protons is about 3/2:1 and that this ratio will increase for exposures to internal organs. Two biological response models are used to compare to the Mir biodosimetry for chromosome aberration in lymphocyte cells; a track-structure model and the linear-quadratic model with linear energy transfer (LET) dependent weighting coefficients. These models are fit to in vitro data for aberration formation in human lymphocytes by photons and charged particles. Both models are found to be in reasonable agreement with data for aberrations in lymphocytes of Mir crew members: however there are differences between the use of LET dependent weighting factors and track structure models for assigning radiation quality factors. The major difference in the models is the increased effectiveness predicted by the track model for low charge and energy ions with LET near 10 keV/μm. The results of our calculations indicate that aluminum shielding, although providing important mitigation of the effects of trapped radiation, provides no protective effect from the galactic cosmic rays (GCR) in low-earth orbit (LEO) using either equivalent dose or the number of chromosome aberrations as a measure until about 100 g/cm 2 of material is used

  13. Geochemical pathways and biological uptake of radium in small Canadian Shield lakes

    International Nuclear Information System (INIS)

    Hesslein, R.H.; Slavicek, E.

    1984-01-01

    The sediment-water interactions and biological uptake of 226 Ra are described for four small Canadian Shield lakes at the Experimental Lakes Area, Kenora, Ontario. A single addition of 226 Ra was made to each lake between 1970 and 1976. Approximately 90 percent of the added 226 Ra initially sorbed to the sediments. Outflow from the lakes showed losses of only 5-11 percent 226 Ra per year. Models are proposed for adsorption and outflow of 226 Ra from lakes. Biological uptake and long-term 226 Ra concentrations were measured in three species of macrophytes, crayfish, and five species of fish. Bioaccumulation ranged from 1100 to 5000 in macrophytes, 705 in crayfish, from 30 to 80 in large trout (Salvelinus namaycush), white sucker (Catostomus commersoni), and lake whitefish (Coregonus clupeaformis), and from 230 to 1200 in fathead minnows (Pimephales promelas), pearl dace (Semotilus margarita), and northern redbelly dace (Chrosomus eos). The concept of Ra/Ca ratio in organisms versus water and food is used to explain the differences in bioaccumulation. 226 Ra is discriminated against versus calcium by fish but favoured by macrophytes and crayfish

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

    International Nuclear Information System (INIS)

    Freund, H.U.

    1995-01-01

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

  15. Comparison of neutron fluxes obtained by 2-D and 3-D geometry with different shielding libraries in biological shield of the TRIGA MARK II reactor

    International Nuclear Information System (INIS)

    Bozic, M.; Zagar, T.; Ravnik, M.

    2003-01-01

    Neutron fluxes in different spatial locations in biological shield are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Libraries used with TORT code were BUGLE-96 library (coupled library with 47 neutron groups and 20 gamma groups) and VITAMIN-B6 library (coupled library with 199 neutron groups and 42 gamma groups). BUGLE-96 library is derived from VITAMIN-B6 library. 2-D and 3-D models for homogeneous type of problem (without inserted beam port 4) and problem with asymmetry (non-homogeneous problem; inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. The main purpose is to verify the possibility for using 2-D approximation model instead of large 3-D model in some calculations. Another purpose of this paper was to compare neutron spectral constants obtained from neutron fluxes (3-D model) determined with smaller BUGLE-96 library with new constants obtained from fluxes calculated with bigger VITAMIN-B6 library. These neutron spectral constants are used in isotopic calculation with SCALE code package (ORIGEN-S). In past only neutron spectral constants determined by neutron fluxes from BUGLE-96 library were used. Experimental results used for isotopic composition comparison are available from irradiation experiment with selected type of concrete and other materials in beam port 4 (irradiation channel 4) in TRIGA Mark II reactor. These experimental results were used as a benchmark in this paper. (author)

  16. Radiation shielding lead shield

    International Nuclear Information System (INIS)

    Dei, Shoichi.

    1991-01-01

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

  17. Effective protection of biological membranes against photo-oxidative damage: Polymeric antioxidant forming a protecting shield over the membrane.

    Science.gov (United States)

    Mertins, Omar; Mathews, Patrick D; Gomide, Andreza B; Baptista, Mauricio S; Itri, Rosangela

    2015-10-01

    We have prepared a chitosan polymer modified with gallic acid in order to develop an efficient protection strategy biological membranes against photodamage. Lipid bilayers were challenged with photoinduced damage by photosensitization with methylene blue, which usually causes formation of hydroperoxides, increasing area per lipid, and afterwards allowing leakage of internal materials. The damage was delayed by a solution of gallic acid in a concentration dependent manner, but further suppressed by the polymer at very low concentrations. The membrane of giant unilamellar vesicles was covered with this modified macromolecule leading to a powerful shield against singlet oxygen and thus effectively protecting the lipid membrane from oxidative stress. The results have proven the discovery of a promising strategy for photo protection of biological membranes. Copyright © 2015 Elsevier B.V. All rights reserved.

  18. Investigations on construction material and construction concepts in order to obtain dose-reducing effects in the dismantling of the biological shield of a 1300 MWe-PWR

    International Nuclear Information System (INIS)

    Bittner, A.; Jungwirth, D.; Knell, M.; Schnitzler, L.

    1984-04-01

    Numerical values of neutron fluxes, activations, dose rates etc. as a function of characteristic values of materials required for optimization purposes to reduce the radiation effect of the biological shield of a PWR are not available. Design concepts are presented for biological shields of PWRs made of concrete with respect to both the most suitable application of materials and the design principles aiming at reduced radiation exposure as compared to present designs during entering, waste disposal and ultimate storage. To evaluate the present-state design the above values have been calculated. Suggested alternative designs are biological shields with selective material application, built from precast elements with or without boron carbide layer arranged in front of it. (orig./HP) [de

  19. Regulatory inhibition of biological tissue mineralization by calcium phosphate through post-nucleation shielding by fetuin-A

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Joshua C., E-mail: joshchang@ucla.edu [Clinical Center, National Institutes of Health, Bethesda, Maryland 20892, USA and Mathematical Biosciences Institute, The Ohio State University, Columbus, Ohio 43210 (United States); Miura, Robert M., E-mail: miura@njit.edu [Department of Mathematical Sciences, New Jersey Institute of Technology, Newark, New Jersey 07102 (United States)

    2016-04-21

    In vertebrates, insufficient availability of calcium and inorganic phosphate ions in extracellular fluids leads to loss of bone density and neuronal hyper-excitability. To counteract this problem, calcium ions are usually present at high concentrations throughout bodily fluids—at concentrations exceeding the saturation point. This condition leads to the opposite situation where unwanted mineral sedimentation may occur. Remarkably, ectopic or out-of-place sedimentation into soft tissues is rare, in spite of the thermodynamic driving factors. This fortunate fact is due to the presence of auto-regulatory proteins that are found in abundance in bodily fluids. Yet, many important inflammatory disorders such as atherosclerosis and osteoarthritis are associated with this undesired calcification. Hence, it is important to gain an understanding of the regulatory process and the conditions under which it can go awry. In this manuscript, we extend mean-field continuum classical nucleation theory of the growth of clusters to encompass surface shielding. We use this formulation to study the regulation of sedimentation of calcium phosphate salts in biological tissues through the mechanism of post-nuclear shielding of nascent mineral particles by binding proteins. We develop a mathematical description of this phenomenon using a countable system of hyperbolic partial differential equations. A critical concentration of regulatory protein is identified as a function of the physical parameters that describe the system.

  20. The quality control for biological-shield heavy concrete construction of nuclear power project

    International Nuclear Information System (INIS)

    Sun Hongjun; Ma Xinchao

    2012-01-01

    The paper introduces the function and characteristics of biological protective heavy-concrete, and its main application scope and role in Fangjiashan nuclear power project. From the aspects of raw material selection, mixing ratio test, heavy concrete production, the paper discusses the main control points of heavy concrete construction process, points out the basic characteristics of heavy concrete construction, and put forward measures to prevent density non-uniformity during heavy concrete construction and to control slump during transportation. Results prove that reasonable construction process control can assure the engineering quality. (authors)

  1. Theoretical evaluation of the biological shielding sufficiency for the Pelletron NEC Particle Accelerator at the Ghana Atomic Energy Commission

    International Nuclear Information System (INIS)

    Amoah, P. A.

    2012-01-01

    Theoretical evaluation of the biological shielding sufficiency provided for 1.7MV Pelletron NEC Particle Accelerator yet to be installed at the Accelerator Research Centre of the Ghana Atomic Energy Commission (GAEC) has been done. Using the Beer Lambert law attenuation of radiation dose outside the walls of the facility was made for protons of energy 1.7MeV. Simulation of charged particle-matter interactions leading to bremsstrahlung radiation using Monte Carlo code (MCNP5) have been carried out. Neutron Activation Analysis (NAA) technique has also been used to identify the composition of the concrete material used during the construction of the Accelerator Research Centre (ARC) building. The NAA analysis revealed that the elemental constituents of the ordinary concrete of density 2.3g/cm 3 used for the construction of the walls included Na, Al, and Ca. Background radiation levels within and outside the facility was measured with the aid of a Sodium Iodide (NaI) identifinder and a Rados detector so as to have a practical reference datum. The weekly background radiation measurements yielded an average dose rate value of 0.05μSv/hr from recorded value range of 0.01μSv/hr to 0.07μSv/hr for an eight month period. Modeling and simulation of charged particle-matter interactions at different beam energies using Monte Carlo code (MCNP5) have yielded the dose rate of 1.58E-07μSv/hr, 1.98E-07μSv/h and 2.20E-05μSv/h outside the 22.86cm (9.0 inch) thick wall of the accelerator facility, for the beam energy range of 0.5-3.0MeV for Titanium, iron and Zirconium target samples respectively. From the Beer-Lambert law, the operational energy of 1.7MeV was used to evaluate theoretically the radiation dose rate of 1.178E-05μSv/hr, 2.656E-05μSv/hr and 4.787E-05μSv/hr outside the 22.86cm thick wall of the accelerator facility for Titanium, Iron and Zirconium targets respectively. At the operational energy energy of 3.0 MeV, the dose rate values obtained were 4.382E-05μSv/h, 9

  2. Monte carlo calculation of the neutron effective dose rate at the outer surface of the biological shield of HTR-10 reactor

    International Nuclear Information System (INIS)

    Remetti, Romolo; Andreoli, Giulio; Keshishian, Silvina

    2012-01-01

    Highlights: ► We deal with HTR-10, that is a helium-cooled graphite-moderated pebble bed reactor. ► We carried out Monte Carlo simulation of the core by MCNP5. ► Extensive use of MCNP5 variance reduction methods has been done. ► We calculated the trend of neutron flux within the biological shield. ► We calculated neutron effective dose at the outer surface of biological shield. - Abstract: Research on experimental reactors, such as HTR-10, provide useful data about potentialities of very high temperature gas-cooled reactors (VHTR). The latter is today rated as one of the six nuclear reactor types involved in the Generation-IV International Forum (GIF) Initiative. In this study, the MCNP5 code has been employed to evaluate the neutron radiation trend vs. the biological shield's thickness and to calculate the neutron effective dose rate at the outer surface. The reactor's geometry has been completely modeled by means of lattices and universes provided by MCNP, even though some approximations were required. Monte Carlo calculations have been performed by means of a simple PC and, as a consequence, in order to obtain acceptable run times, it was made an extensive recourse to variance reduction methods.

  3. Shielding practice

    International Nuclear Information System (INIS)

    Sauermann, P.F.

    1985-08-01

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

  4. Electromagnetic shielding

    International Nuclear Information System (INIS)

    Tzeng, Wen-Shian V.

    1991-01-01

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

  5. Shielding plugs

    International Nuclear Information System (INIS)

    Makishima, Kenji.

    1986-01-01

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

  6. Shielding container

    International Nuclear Information System (INIS)

    Darling, K.A.M.

    1981-01-01

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

  7. Mechanical shielded hot cell

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  8. Electromagnetic shield

    International Nuclear Information System (INIS)

    Miller, J.S.

    1987-01-01

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

  9. Neutron shieldings

    International Nuclear Information System (INIS)

    Tarutani, Kohei

    1979-01-01

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

  10. Nuclear shields

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  11. REACTOR SHIELD

    Science.gov (United States)

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

    1959-02-17

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

  12. Radiation shielding

    International Nuclear Information System (INIS)

    Aitken, D.

    1979-01-01

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

  13. Gonadal shield.

    Science.gov (United States)

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

    1975-10-01

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

  14. Radiation shielding

    International Nuclear Information System (INIS)

    Yue, D.D.

    1979-01-01

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

  15. Shielded container

    International Nuclear Information System (INIS)

    Fries, B.A.

    1978-01-01

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

  16. Radiation shielding for fusion reactors

    International Nuclear Information System (INIS)

    Santoro, R.T.

    2000-01-01

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

  17. The application of semianalytic method for calculating the thickness of biological shields of nuclear reactors. Part 1. Theoretical basis of a semianalytic method. Attenuation of neutrons' radiation

    International Nuclear Information System (INIS)

    Lukaszek, W.; Kucypera, S.

    1982-01-01

    The basis of a semianalytic method for calculating attenuation of rays (neutron, gamma) in material medium is described. The method was applied in determining the neutrons' flux density in one dimensional Cartesian geometry of the reflector and the shield. (author)

  18. Shielding calculations for NET

    International Nuclear Information System (INIS)

    Verschuur, K.A.; Hogenbirk, A.

    1991-05-01

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

  19. About the Scythian Shields

    Directory of Open Access Journals (Sweden)

    About the Scythian Shields

    2017-10-01

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

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  1. Shielding benchmark problems

    International Nuclear Information System (INIS)

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

    1978-09-01

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

  2. Radiation shielding device

    International Nuclear Information System (INIS)

    Nakagawa, Takahiro; Yamagami, Makoto.

    1996-01-01

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

  3. Thermal design of top shield for PFBR

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  4. Handout on shielding calculation

    International Nuclear Information System (INIS)

    Heilbron Filho, P.F.L.

    1991-01-01

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

  5. Design of emergency shield

    International Nuclear Information System (INIS)

    Soliman, S.E.

    1993-01-01

    Manufacturing of an emergency movable shield in the hot laboratories center is urgently needed for the safety of personnel in case of accidents or spilling of radioactive materials. In this report, a full design for an emergency shield is presented and the corresponding dose rates behind the shield for different activities (from 1 mCi to 5 Ci) was calculated by using micro shield computer code. 4 figs., 1 tab

  6. Electromagnetically shielded building

    International Nuclear Information System (INIS)

    Takahashi, T.; Nakamura, M.; Yabana, Y.; Ishikawa, T.; Nagata, K.

    1992-01-01

    This invention relates to a building having an electromagnetic shield structure well-suited for application to an information network system utilizing electromagnetic waves, and more particularly to an electromagnetically shielded building for enhancing the electromagnetic shielding performance of an external wall. 6 figs

  7. Electromagnetically shielded building

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, T; Nakamura, M; Yabana, Y; Ishikawa, T; Nagata, K

    1992-04-21

    This invention relates to a building having an electromagnetic shield structure well-suited for application to an information network system utilizing electromagnetic waves, and more particularly to an electromagnetically shielded building for enhancing the electromagnetic shielding performance of an external wall. 6 figs.

  8. Electromagnetic shielding formulae

    International Nuclear Information System (INIS)

    Dahlberg, E.

    1979-02-01

    This addendum to an earlier collection of electromagnetic shielding formulae (TRITA-EPP-75-27) contains simple transfer matrices suitable for calculating the quasistatic shielding efficiency for multiple transverse-field and axial-field cylindrical and spherical shields, as well as for estimating leakage fields from long coaxial cables and the normal-incidence transmission of a plane wave through a multiple plane shield. The differences and similarities between these cases are illustrated by means of equivalent circuits and transmission line analogies. The addendum also includes a discussion of a possible heuristic improvement of some shielding formulae. (author)

  9. Shielding benchmark problems, (2)

    International Nuclear Information System (INIS)

    Tanaka, Shun-ichi; Sasamoto, Nobuo; Oka, Yoshiaki; Shin, Kazuo; Tada, Keiko.

    1980-02-01

    Shielding benchmark problems prepared by Working Group of Assessment of Shielding Experiments in the Research Committee on Shielding Design in the Atomic Energy Society of Japan were compiled by Shielding Laboratory in Japan Atomic Energy Research Institute. Fourteen shielding benchmark problems are presented newly in addition to twenty-one problems proposed already, for evaluating the calculational algorithm and accuracy of computer codes based on discrete ordinates method and Monte Carlo method and for evaluating the nuclear data used in codes. The present benchmark problems are principally for investigating the backscattering and the streaming of neutrons and gamma rays in two- and three-dimensional configurations. (author)

  10. The Active Muon Shield

    CERN Document Server

    Bezshyiko, Iaroslava

    2016-01-01

    In the SHiP beam-dump of the order of 1011 muons will be produced per second. An active muon-shield is used to magnetically deflect these muons out of the acceptance of the spectrom- eter. This note describes how this shield is modelled and optimized. The SHiP spectrometer is being re-optimized using a conical decay-vessel, and utilizing the possibility to magnetize part of the beam-dump shielding iron. A shield adapted to these new conditions is presented which is significantly shorter and lighter than the shield used in the Technical Proposal (TP), while showing a similar performance.

  11. Fast Neutron Transport in the Biological Shielding Model and Other Regions of the VVER-1000 Mock-Up on the LR-0 Research Reactor

    Directory of Open Access Journals (Sweden)

    Košťál Michal

    2016-01-01

    Full Text Available A set of benchmark experiments was carried out in the full scale VVER-1000 mock-up on the reactor LR-0 in order to validate neutron transport calculation methodologies and to perform the optimization of the shape and locations of neutron flux operation monitors channels inside the shielding of the new VVER-1000 type reactors. Compared with previous experiments on the VVER-1000 mock-up on the reactor LR-0, the fast neutron spectra were measured in the extended neutron energy interval (0.1–10 MeV and new calculations were carried out with the MCNPX code using various nuclear data libraries (ENDF/B VII.0, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND 2009, and CENDL 3.1. Measurements and calculations were carried out at different points in the mock-up. The calculation and experimental data are compared.

  12. Wake Shield Target Protection

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  13. Radiation shielding concrete

    International Nuclear Information System (INIS)

    Kunishima, Shigeru.

    1990-01-01

    The radiation shielding concretes comprise water, cement, fine aggregates consisting of serpentines and blown mist slags, coarse aggregates consisting of serpentines and kneading materials. Since serpentines containing a relatively great amount of water of crystallization in rocks as coarse aggregates and fine aggregates, the hydrogen content in the radiation shielding concretes is increased and the neutron shielding effect is improved. In addition, since serpentines are added as the fine aggregates and blown mists slags of a great specific gravity are used, the specific gravity of the shielding concretes is increased to improve the γ-ray shielding effect. Further, by the use of the kneading material having a water reducing effect and fluidizing effect, and by the bearing effect of the spherical blown mist slags used as the fine aggregates, concrete fluidity can be increased. Accordingly, workability of the radiation shielding concretes can be improved. (T.M.)

  14. Accelerator shielding benchmark problems

    International Nuclear Information System (INIS)

    Hirayama, H.; Ban, S.; Nakamura, T.

    1993-01-01

    Accelerator shielding benchmark problems prepared by Working Group of Accelerator Shielding in the Research Committee on Radiation Behavior in the Atomic Energy Society of Japan were compiled by Radiation Safety Control Center of National Laboratory for High Energy Physics. Twenty-five accelerator shielding benchmark problems are presented for evaluating the calculational algorithm, the accuracy of computer codes and the nuclear data used in codes. (author)

  15. Shielded regenerative neutron detector

    International Nuclear Information System (INIS)

    Terhune, J.H.; Neissel, J.P.

    1978-01-01

    An ion chamber type neutron detector is disclosed which has a greatly extended lifespan. The detector includes a fission chamber containing a mixture of active and breeding material and a neutron shielding material. The breeding and shielding materials are selected to have similar or substantially matching neutron capture cross-sections so that their individual effects on increased detector life are mutually enhanced

  16. Radiation shielding plate

    International Nuclear Information System (INIS)

    Kobayashi, Torakichi; Sugawara, Takeo.

    1983-01-01

    Purpose: To reduce the weight and stabilize the configuration of a radiation shielding plate which is used in close contact with an object to be irradiated with radiation rays. Constitution: The radiation shielding plate comprises a substrate made of lead glass and a metallic lead coating on the surface of the substrate by means of plating, vapor deposition or the like. Apertures for permeating radiation rays are formed to the radiation shielding plate. Since the shielding plate is based on a lead glass plate, a sufficient mechanical strength can be obtained with a thinner structure as compared with the conventional plate made of metallic lead. Accordingly, if the shielding plate is disposed on a soft object to be irradiated with radiation rays, the object and the plate itself less deform to obtain a radiation irradiation pattern with distinct edges. (Moriyama, K.)

  17. HPGe detector shielding adjustment

    International Nuclear Information System (INIS)

    Trnkova, L.; Rulik, P.

    2008-01-01

    Low-level background shielding of HPGe detectors is used mainly for environmental samples with very low content of radionuclides. National Radiation Protection Institute (SURO) in Prague is equipped with 14 HPGe detectors with relative efficiency up to 150%. The detectors are placed in a room built from materials with low content of natural radionuclides and equipped with a double isolation of the floor against radon. Detectors themselves are placed in lead or steel shielding. Steel shielding with one of these detectors with relative efficiency of 100% was chosen to be rebuilt to achieve lower minimum detectable activity (MDA). Additional lead and copper shielding was built up inside the original steel shielding to reduce the volume of the inner space and filled with nitrogen by means of evaporating liquid nitrogen. The additional lead and copper shielding, consequent reduction of the inner volume and supply of evaporated nitrogen, caused a decrease of the background count and accordingly MDA values as well. The effect of nitrogen evaporation on the net areas of peaks belonging to radon daughters is significant. The enhanced shielding adjustment has the biggest influence in low energy range, what can be seen in collected data. MDA values in energy range from 30 keV to 400 keV decreased to 0.65-0.85 of original value, in energy range from 400 keV to 2 MeV they fell to 0.70-0.97 of original value. (authors)

  18. Shielding in experimental areas

    International Nuclear Information System (INIS)

    Stevens, A.; Tarnopolsky, G.; Thorndike, A.; White, S.

    1979-01-01

    The amount of shielding necessary to protect experimental detectors from various sources of background radiation is discussed. As illustrated an experiment has line of sight to sources extending approx. 90 m upstream from the intersection point. Packing a significant fraction of this space with shielding blocks would in general be unacceptable because primary access to the ring tunnel is from the experimental halls. (1) From basic machine design considerations and the inherent necessity to protect superconducting magnets it is expected that experimental areas in general will be cleaner than at any existing accelerator. (2) Even so, it will likely be necessary to have some shielding blocks available to protect experimental apparatus, and it may well be necessary to have a large amount of shielding available in the WAH. (3) Scraping will likely have some influence on all halls, and retractable apparatus may sometimes be necessary. (4) If access to any tunnel is needed to replace a magnet, one has 96 h (4 days) available to move shielding away to permit access without additional downtime. This (the amount of shielding one can shuffle about in 96 h) is a reasonable upper limit to shielding necessary in a hall

  19. Shielding member for thermonuclear device

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, Masanori

    1997-06-30

    In a thermonuclear device for shielding fast neutrons by shielding members disposed in a shielding vessel (vacuum vessel and structures such as a blanket disposed in the vacuum vessel), the shielding member comprises a large number of shielding wires formed fine and short so as to have elasticity. The shielding wires are sealed in a shielding vessel together with water, and when the width of the shielding vessel is changed, the shielding wires follow after the change of the width while elastically deforming in the shielding vessel, so that great stress and deformation are not formed thereby enabling to improve reliability. In addition, the length, the diameter and the shape of each of the shielding wires can be selected in accordance with the shielding space of the shielding vessel. Even if the shape of the shielding vessel is complicated, the shielding wires can be inserted easily. Accordingly, the filling rate of the shielding members can be changed easily. It can be produced more easily compared with a conventional spherical pebbles. It can be produced more easily than existent spherical shielding pebbles thereby enabling to reduce the production cost. (N.H.)

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

    International Nuclear Information System (INIS)

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

    1998-03-01

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

  1. Estimating ISABELLE shielding requirements

    International Nuclear Information System (INIS)

    Stevens, A.J.; Thorndike, A.M.

    1976-01-01

    Estimates were made of the shielding thicknesses required at various points around the ISABELLE ring. Both hadron and muon requirements are considered. Radiation levels at the outside of the shield and at the BNL site boundary are kept at or below 1000 mrem per year and 5 mrem/year respectively. Muon requirements are based on the Wang formula for pion spectra, and the hadron requirements on the hadron cascade program CYLKAZ of Ranft. A muon shield thickness of 77 meters of sand is indicated outside the ring in one area, and hadron shields equivalent to from 2.7 to 5.6 meters in thickness of sand above the ring. The suggested safety allowance would increase these values to 86 meters and 4.0 to 7.2 meters respectively. There are many uncertainties in such estimates, but these last figures are considered to be rather conservative

  2. Scintillation counter, segmented shield

    International Nuclear Information System (INIS)

    Olson, R.E.; Thumim, A.D.

    1975-01-01

    A scintillation counter, particularly for counting gamma ray photons, includes a massive lead radiation shield surrounding a sample-receiving zone. The shield is disassembleable into a plurality of segments to allow facile installation and removal of a photomultiplier tube assembly, the segments being so constructed as to prevent straight-line access of external radiation through the shield into radiation-responsive areas. Provisions are made for accurately aligning the photomultiplier tube with respect to one or more sample-transmitting bores extending through the shield to the sample receiving zone. A sample elevator, used in transporting samples into the zone, is designed to provide a maximum gamma-receiving aspect to maximize the gamma detecting efficiency. (U.S.)

  3. Shields for nuclear reactors

    International Nuclear Information System (INIS)

    Aspden, G.J.

    1984-01-01

    The patent concerns shields for nuclear reactors. The roof shield comprises a normally fixed radial outer portion, a radial inner portion rotatable about a vertical axis, and a connection between the inner and outer portions. In the event of hypothecal core disruption conditions, a cantilever system on the inner wall allows the upward movement of the inner wall, in order to prevent loss of containment. (UK)

  4. Radiation shielding curtain

    International Nuclear Information System (INIS)

    Winkler, N.T.

    1976-01-01

    A radiation shield is described in the form of a stranded curtain made up of bead-chains whose material and geometry are selected to produce a cross-sectional density that is the equivalent of 0.25 mm or more of lead and which curtain may be mounted on various radiological devices to shield against scattered radiation while offering a minimum of obstruction to the radiologist

  5. Shielded cells transfer automation

    International Nuclear Information System (INIS)

    Fisher, J.J.

    1984-01-01

    Nuclear waste from shielded cells is removed, packaged, and transferred manually in many nuclear facilities. Radiation exposure is absorbed by operators during these operations and limited only through procedural controls. Technological advances in automation using robotics have allowed a production waste removal operation to be automated to reduce radiation exposure. The robotic system bags waste containers out of glove box and transfers them to a shielded container. Operators control the system outside the system work area via television cameras. 9 figures

  6. Shielding calculations for the Intense Neutron Source Facility. Final report

    International Nuclear Information System (INIS)

    Battat, M.E.; Henninger, R.J.; Macdonald, J.L.; Dudziak, D.J.

    1978-06-01

    Results of shielding calculations for the Intnse Neutron Source (INS) facility are presented. The INS facility is designed to house two sources, each of which will produce D--T neutrons with intensities in the range from 1 to 3 x 10 15 n/s on a continuous basis. Topics covered include the design of the biological shield, use of two-dimensional discrete-ordinates results to specify the source terms for a Monte Carlo skyshine calculation, air activation, and dose rates in the source cell (after shutdown) due to activation of the biological shield

  7. Development of shielding design analysis system

    International Nuclear Information System (INIS)

    Tada, Keiko; Shiraki, Takako

    2001-03-01

    The aim of this work is to develop insufficient auxiliary routines which manage input and output data and interface the main codes and to establish a shielding design analysis system on work stations (SUN, DEC). In shielding design analyses, one- and two- dimensional (1-D and 2-D) transport Sn codes are used mainly with some auxiliary codes which generate input data of Sn calculation and edit Sn calculation outputs. The main transport calculation codes can be obtained from the Code Center of RIST (Research Organization for Information Science and Technology). In this work, peripheral codes are developed to generate cross sections, produce Sn quadrature sets, edit calculation outputs or draw contour figures. In shielding calculations around a reactor, the boot-strapping technique is often employed to treat a large area extending from the core to the biological shield to improve the calculation accuracy. When a three-dimensional (3-D) calculation for a complex geometry with shielding defects, 2-D and 3-D coupling calculation is employed frequently. To use this coupling method conversion cods are prepared which read flux file from DORT and prepare an external boundary source file for the 2-D or the 3-D calculation codes. For further conveniences well used data such as the Sn quadrature sets, the dose rate conversion factors, the reaction cross section sets are stored as a data base and code manuals including sample inputs of typical problems are prepared which are comprehensible to beginners. (author)

  8. Neutron shielding material

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  9. Concrete radiation shielding

    International Nuclear Information System (INIS)

    Kaplan, M.F.

    1989-01-01

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

  10. Method for dismantling shields

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  11. Shield support frame. Schildausbaugestell

    Energy Technology Data Exchange (ETDEWEB)

    Plaga, K.

    1981-09-17

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

  12. Radiation shielding material

    International Nuclear Information System (INIS)

    Matsumoto, Akio; Isobe, Eiji.

    1976-01-01

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

  13. Hybrid Magnetic Shielding

    Science.gov (United States)

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

    2017-09-01

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

  14. Reactor head shielding apparatus

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    Science.gov (United States)

    Sariyer, Demet; Küçer, Rahmi

    2018-02-01

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

  16. Shielding research in France

    Energy Technology Data Exchange (ETDEWEB)

    Lafore, P

    1964-10-01

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

  17. The shielding calculation for the CN guide shielding assembly in HANARO

    International Nuclear Information System (INIS)

    Kim, H. S.; Lee, B. C.; Lee, K. H.; Kim, H.

    2006-01-01

    The cold neutron research facility in HANARO is under construction. The area including neutron guides and rotary shutter in the reactor hall should be shielded by the guide shielding assembly which is constructed of heavy concrete blocks and structure. The guide shielding assembly is divided into 2 parts, A and B. Part A is about 6.4 meters apart from the reactor biological shield and it is constructed of heavy concrete blocks whose density is above 4.0g/cm 3 . And part B is a fixed heavy concrete structure whose density is above 3.5g/cm 3 . The rotary shutter is also made with heavy concrete whose density is above 4.0g/cm 3 and includes 5 neutron guides inside. It can block the neutron beam by rotating when CNS is not operating. The dose criterion outside the guide shielding assembly is established as 12.5 μSv/hr which is also applied to reactor shielding in HANARO

  18. Radiation shielding bricks

    International Nuclear Information System (INIS)

    Crowe, G.J.W.

    1983-01-01

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

  19. Hybrid Active-Passive Radiation Shielding System

    Data.gov (United States)

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

  20. Secondary gamma-ray data for shielding calculation

    International Nuclear Information System (INIS)

    Miyasaka, Sunichi

    1979-01-01

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

  1. Radiation shielding cloth

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  2. Electrostatic shielding of transformers

    Energy Technology Data Exchange (ETDEWEB)

    De Leon, Francisco

    2017-11-28

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

  3. Penetration portion shielding structure

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  4. Dosimetry and shielding

    International Nuclear Information System (INIS)

    Farinelli, U.

    1977-01-01

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

  5. Radiation shielding glass

    International Nuclear Information System (INIS)

    Kido, Kazuhiro; Ueda, Hajime.

    1997-01-01

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

  6. Gonad shielding in diagnostic radiology

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

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

  7. Shielding experiments for accelerator facilities

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-06-01

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

  8. Shielding experiments for accelerator facilities

    International Nuclear Information System (INIS)

    Nakashima, Hiroshi; Tanaka, Susumu; Sakamoto, Yukio

    2000-01-01

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

  9. EBT-P gamma-ray-shielding analysis

    International Nuclear Information System (INIS)

    Gohar, Y.

    1983-01-01

    First, a one-dimensional scoping study was performed for the gamma-ray shield of the ELMO Bumpy Torus proof-of-principle device to define appropriate shielding material and determine the required shielding thickness. The dose-equivalent results are analyzed as a function of the radiation-shield thickness for different shielding options. A sensitivity analysis for the pessimistic case is given. The recommended shielding option based on the performance and cost is discussed. Next, a three-dimensional scoping study for the coil shield was performed for four different shielding options to define the heat load for each component and check the compliance with the design criterion of 10 watts maximum heat load per coil from the gamma-ray sources. Also, a detailed biological-dose survey was performed which included: (a) the dose equivalent inside and outside the building, (b) the dose equivalent from the two mazes of the building, and (c) the skyshine contribution to the dose equivalent

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

    International Nuclear Information System (INIS)

    Kim, Younggook; Kim, Jeongdong; Lee, Yongdeok

    2015-01-01

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

  11. External dosimetry sources and shielding

    International Nuclear Information System (INIS)

    Calisto, Washington

    1994-01-01

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

  12. Radiation shielding calculation using MCNP

    International Nuclear Information System (INIS)

    Masukawa, Fumihiro

    2001-01-01

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

  13. Shielding benchmark test

    International Nuclear Information System (INIS)

    Kawai, Masayoshi

    1984-01-01

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

  14. Neutron shielding material

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  15. Radiation shielding wall structure

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  16. Radiation shielding material

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  17. Multilayer radiation shield

    Science.gov (United States)

    Urbahn, John Arthur; Laskaris, Evangelos Trifon

    2009-06-16

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

  18. Light shielding apparatus

    Science.gov (United States)

    Miller, Richard Dean; Thom, Robert Anthony

    2017-10-10

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

  19. Shielding container for radioactive isotopes

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  20. Primary shield displacement and bowing

    International Nuclear Information System (INIS)

    Scott, K.V.

    1978-01-01

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

  1. Shielding calculations using FLUKA

    International Nuclear Information System (INIS)

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

    1988-06-01

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

  2. Shielding Benchmark Computational Analysis

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  3. Muon shielding for PEP

    International Nuclear Information System (INIS)

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

    1974-01-01

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

  4. A contribution to shielding effectiveness analysis of shielded tents

    Directory of Open Access Journals (Sweden)

    Vranić Zoran M.

    2004-01-01

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

  5. Radiation shielding quality assurance

    Science.gov (United States)

    Um, Dallsun

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

  6. The SNS target station preliminary Title I shielding analyses

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    Kalcheva, S.

    2005-02-01

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

  8. Shielding from cosmic radiation for interplanetary missions Active and passive methods

    CERN Document Server

    Spillantini, P; Durante, M; Müller-Mellin, R; Reitz, G; Rossi, L; Shurshakov, V; Sorbi, M

    2007-01-01

    Shielding is arguably the main countermeasure for the exposure to cosmic radiation during interplanetary exploratory missions. However, shielding of cosmic rays, both of galactic or solar origin, is problematic, because of the high energy of the charged particles involved and the nuclear fragmentation occurring in shielding materials. Although computer codes can predict the shield performance in space, there is a lack of biological and physical measurements to benchmark the codes. An attractive alternative to passive, bulk material shielding is the use of electromagnetic fields to deflect the charged particles from the spacecraft target. Active shielding concepts based on electrostatic fields, plasma, or magnetic fields have been proposed in the past years, and should be revised based on recent technological improvements. To address these issues, the European Space Agency (ESA) established a Topical Team (TT) in 2002 including European experts in the field of space radiation shielding and superconducting magn...

  9. Neutronic reactor thermal shield

    International Nuclear Information System (INIS)

    Lowe, P.E.

    1976-01-01

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

  10. A shield against distraction

    OpenAIRE

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

    2014-01-01

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

  11. Measuring space radiation shielding effectiveness

    OpenAIRE

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

    2017-01-01

    Passive radiation shielding is one strategy to mitigate the problem of space radiation exposure. While space vehicles are constructed largely of aluminum, polyethylene has been demonstrated to have superior shielding characteristics for both galactic cosmic rays and solar particle events due to the high hydrogen content. A method to calculate the shielding effectiveness of a material relative to reference material from Bragg peak measurements performed using energetic heavy charged particles ...

  12. Selective shielding device for scintiphotography

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  13. Tax Shield, Insolvenz und Zinsschranke

    OpenAIRE

    Arnold, Sven; Lahmann, Alexander; Schwetzler, Bernhard

    2010-01-01

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

  14. SHIELD verification and validation report

    International Nuclear Information System (INIS)

    Boman, C.

    1992-02-01

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

  15. Multifunctional Hot Structure Heat Shield

    Data.gov (United States)

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

  16. Radiation shield for nuclear reactors

    International Nuclear Information System (INIS)

    Weissenfluh, J.A.

    1978-01-01

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

  17. New Toroid shielding design

    CERN Multimedia

    Hedberg V

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

  18. SHIELDS Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-10-03

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

  19. Radiation shielding analysis

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  20. Radiation shield for PWR reactors

    International Nuclear Information System (INIS)

    Esenov, Amra; Pustovgar, Andrey

    2013-01-01

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

  1. Nuclear data for radiation shielding

    International Nuclear Information System (INIS)

    Miyasaka, Shunichi; Takahashi, Hiroshi.

    1976-01-01

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

  2. Concrete shielding exterior to iron

    International Nuclear Information System (INIS)

    Yurista, P.; Cossairt, D.

    1983-08-01

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

  3. Gonad shielding in diagnostic radiology

    International Nuclear Information System (INIS)

    1975-06-01

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

  4. The Imperial Shield

    DEFF Research Database (Denmark)

    Mortensen, Simon Valentin

    2006-01-01

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

  5. Shielded Canister Transporter

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  6. ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-03-01

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

  7. Welding shield for coupling heaters

    Science.gov (United States)

    Menotti, James Louis

    2010-03-09

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

  8. Penetration shielding applications of CYLSEC

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  9. Shield calculations, optimization vs. paradigm

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  10. Modular reactor head shielding system

    International Nuclear Information System (INIS)

    Jacobson, E. B.

    1985-01-01

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

  11. Parameters calculation of shielding experiment

    International Nuclear Information System (INIS)

    Gavazza, S.

    1986-02-01

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

  12. Design experience: CRBRP radiation shielding

    International Nuclear Information System (INIS)

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

    1978-11-01

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

  13. Nanoscale microwave microscopy using shielded cantilever probes

    KAUST Repository

    Lai, Keji; Kundhikanjana, Worasom; Kelly, Michael A.; Shen, Zhi-Xun

    2011-01-01

    Quantitative dielectric and conductivity mapping in the nanoscale is highly desirable for many research disciplines, but difficult to achieve through conventional transport or established microscopy techniques. Taking advantage of the micro-fabrication technology, we have developed cantilever-based near-field microwave probes with shielded structures. Sensitive microwave electronics and finite-element analysis modeling are also utilized for quantitative electrical imaging. The system is fully compatible with atomic force microscope platforms for convenient operation and easy integration of other modes and functions. The microscope is ideal for interdisciplinary research, with demonstrated examples in nano electronics, physics, material science, and biology.

  14. Nanoscale microwave microscopy using shielded cantilever probes

    KAUST Repository

    Lai, Keji

    2011-04-21

    Quantitative dielectric and conductivity mapping in the nanoscale is highly desirable for many research disciplines, but difficult to achieve through conventional transport or established microscopy techniques. Taking advantage of the micro-fabrication technology, we have developed cantilever-based near-field microwave probes with shielded structures. Sensitive microwave electronics and finite-element analysis modeling are also utilized for quantitative electrical imaging. The system is fully compatible with atomic force microscope platforms for convenient operation and easy integration of other modes and functions. The microscope is ideal for interdisciplinary research, with demonstrated examples in nano electronics, physics, material science, and biology.

  15. Transparent fast neutron shielding material and shielding method

    International Nuclear Information System (INIS)

    Nashimoto, Tetsuji; Katase, Haruhisa.

    1993-01-01

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

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

    Science.gov (United States)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  18. Device for sealing and shielding a nuclear fuel storage tank

    International Nuclear Information System (INIS)

    Masaki, Gengo.

    1975-01-01

    Object: To provide a shield device for opening and closing a great opening in a relay-storage-tank within a hot cell for temporarily storing a nuclear fuel, in which the device is simplified in construction and which can perform the opening and closing operation in simple, positive and quick manner. Structure: A biological shield is positioned upwardly of an opening of a nuclear fuel storage tank to render an actuator inoperative. A sealing plate, which is pivotally supported by a plurality of support rod devices from the biological shield for parallel movement with respect to the biological shield, comes in contact with a resilient seal disposed along the entire peripheral edge of the opening to form an air-tight seal therebetween. In order to release the opening, the actuator is first actuated and the end of the sealing plate is horizontally pressed by a piston rod thereof. Then, the sealing plate is moved along the line depicted by the end of the support rod in the support rod devices and as a consequence, the plate is moved away from the resilient seal in the peripheral edge of the opening. When a driving device is actuated to travel the plate along the aforesaid line while maintaining the condition as described, the biological device moves along the guide. (Kamimura, M.)

  19. Morphometry of terrestrial shield volcanoes

    Science.gov (United States)

    Grosse, Pablo; Kervyn, Matthieu

    2018-03-01

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

  20. A study of gamma shielding

    International Nuclear Information System (INIS)

    Roogtanakait, N.

    1981-01-01

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

  1. Radiation protection/shield design

    International Nuclear Information System (INIS)

    Disney, R.K.

    1977-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-11

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  4. Radiation shielding member

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  5. Self-shielding factors

    International Nuclear Information System (INIS)

    Kaul, D.C.

    1982-01-01

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

  6. Shielding plug device

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  7. A novel comprehensive utilization of vanadium slag: As gamma ray shielding material

    Energy Technology Data Exchange (ETDEWEB)

    Dong, Mengge [School of Metallurgy, Northeastern University, Shenyang 110004 (China); Liaoning Key Laboratory of Metallurgical Resources Recycling Science, Shenyang 110004 (China); Xue, Xiangxin, E-mail: xuexx@mail.neu.edu.cn [School of Metallurgy, Northeastern University, Shenyang 110004 (China); Liaoning Key Laboratory of Metallurgical Resources Recycling Science, Shenyang 110004 (China); Yang, He; Liu, Dong [School of Metallurgy, Northeastern University, Shenyang 110004 (China); Liaoning Key Laboratory of Metallurgical Resources Recycling Science, Shenyang 110004 (China); Wang, Chao [Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016 (China); Li, Zhefu [Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China)

    2016-11-15

    Highlights: • A novel comprehensive utilization method for vanadium slag is proposed. • Shielding properties of vanadium slag are better than ordinary concrete. • HVL of vanadium slag is between Lead and concrete to shield {sup 60}Co gamma ray. • HVL of composite is higher than concrete when adding amount of vanadium slag is 900. • Composite can be used as injecting mortar for cracks developed in concrete shields. - Abstract: New exploration of vanadium slag as gamma ray shielding material was proposed, the shielding properties of vanadium slag was higher than concrete when the energy of photons was in 0.0001 MeV–100000 MeV. Vanadium slag/epoxy resin composites were prepared, shielding and material properties of materials were tested by {sup 60}Co gamma ray, simultaneous DSC-TGA, electronic universal testing machine and scanning electron microscopy, respectively. The results showed that the shielding properties of composite would be better with the increase of vanadium slag addition amount. The HVL (half value layer thickness) of vanadium slag was between Lead and concrete while composite was higher than concrete when the addition amount of vanadium slag was 900 used as material to shield {sup 60}Co gamma ray, also the resistance temperature of composite was about 215 °C and the bending strength was over 10 MPa. The composites could be used as injecting mortar for cracks developed in biological concrete shields, coating for the floor of the nuclear facilities, and shielding materials by itself.

  8. SNF shipping cask shielding analysis

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  9. Active Radiation Shield, Phase I

    Data.gov (United States)

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

  10. Gonad shielding in computerized tomography

    International Nuclear Information System (INIS)

    Rockstroh, G.

    1984-01-01

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

  11. Radiation shielding for neutron guides

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  12. Shielding calculations. Optimization vs. Paradigms

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  13. Radiation shielding for neutron guides

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  14. Measuring space radiation shielding effectiveness

    Directory of Open Access Journals (Sweden)

    Bahadori Amir

    2017-01-01

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

  15. Measuring space radiation shielding effectiveness

    Science.gov (United States)

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

    2017-09-01

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

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

    International Nuclear Information System (INIS)

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

    2009-07-01

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

  17. Electrodynamic Dust Shield Demonstrator

    Science.gov (United States)

    Stankie, Charles G.

    2013-01-01

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

  18. Radiation shielding in dental radiography

    International Nuclear Information System (INIS)

    Stenstroem, B.; Rehnmark-Larsson, S.; Julin, P.; Richter, S.

    1983-01-01

    The protective effect in the thyroid region from different types of radiation shieldings at intraoral radiography has been studied as well as the reduction of the absorbed dose to the sternal and the gonadal regions. The shieldings tested were five different types of leaded aprons, of which three had an attached leaded collar and the other two were used in combination with separate soft leaded collars. Furthermore one of the soft leaded collars and an unflexible horizontal leaded shield were tested separately. Two dental x-ray machines of 60 and 65 kVp with rectangular and circular tube collimators were used. The exposure time corresponded to speed group E film. The absorbed doses were measured with two ionization chambers. No significant difference in the protective effect in the thyroid gland could be found between the different types of radiation shieldings. There was a dose reduction by approximately a factor of 2 to the thyroid region down to 0.08 mGy per full survey using parallelling technique, and below 0.001 mGy per single bitewing exposure. The shieldings reduced the thyroid dose using bisecting-angle technique by a factor of 5 down to 0.15 mGy per full survey (20 exposures). In the sternal region the combinations of apron and collar reduced the absorbed dose from a full survey to below 2 μGy compared with 18 μGy (parallelling) and 31 μGy (biscting-angle) without any shielding. With the horizontal leaded shield a reduction by a factor of 6 was obtained but no significant sternal dose reduction could be detected from the soft collar alone. The gonadal dose could be reduced by a factor of 10 with the horizontal leaded shield, parallelling technique and circular collimator. Using leaded aprons the gonadal dose was approximately one per cent of the dose without any shielding, i.e. below 0.01 μGy per single intraoral exposure. (Authors)

  19. Novel shielding materials for space and air travel

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  20. Basic design of shield blocks for a spallation neutron source under the high-intensity proton accelerator project

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Katsuhiko; Maekawa, Fujio; Takada, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Under the JAERI-KEK High-Intensity Proton Accelerator Project (J-PARC), a spallation neutron source driven by a 3 GeV-1 MW proton beam is planed to be constructed as a main part of the Materials and Life Science Facility. Overall dimensions of a biological shield of the neutron source had been determined by evaluation of shielding performance by Monte Carlo calculations. This report describes results of design studies on an optimum dividing scheme in terms of cost and treatment and mechanical strength of shield blocks for the biological shield. As for mechanical strength, it was studied whether the shield blocks would be stable, fall down or move to a horizontal direction in case of an earthquake of seismic intensity of 5.5 (250 Gal) as an abnormal load. For ceiling shielding blocks being supported by both ends of the long blocks, maximum bending moment and an amount of maximum deflection of their center were evaluated. (author)

  1. Basic design of shield blocks for a spallation neutron source under the high-intensity proton accelerator project

    CERN Document Server

    Yoshida, K; Takada, H

    2003-01-01

    Under the JAERI-KEK High-Intensity Proton Accelerator Project (J-PARC), a spallation neutron source driven by a 3 GeV-1 MW proton beam is planed to be constructed as a main part of the Materials and Life Science Facility. Overall dimensions of a biological shield of the neutron source had been determined by evaluation of shielding performance by Monte Carlo calculations. This report describes results of design studies on an optimum dividing scheme in terms of cost and treatment and mechanical strength of shield blocks for the biological shield. As for mechanical strength, it was studied whether the shield blocks would be stable, fall down or move to a horizontal direction in case of an earthquake of seismic intensity of 5.5 (250 Gal) as an abnormal load. For ceiling shielding blocks being supported by both ends of the long blocks, maximum bending moment and an amount of maximum deflection of their center were evaluated.

  2. Shielding features of quarry stone

    International Nuclear Information System (INIS)

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

    2010-10-01

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

  3. Radiation Shielding Materials and Containers Incorporating Same

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-11-01

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

  4. MMW [multimegawatt] shielding design and analysis

    International Nuclear Information System (INIS)

    Olson, A.P.

    1988-01-01

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

  5. Magnetic shielding for superconducting RF cavities

    Science.gov (United States)

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

    2017-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-02-01

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

  7. Radiation shielding for fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Laboratory, Tokyo (Japan)

    2000-03-01

    Radiation shielding aspects relating fission reactors have been reviewed. Domestic activities in the past five years have been mainly described concerning nuclear data, calculation methods, shielding and skyshine experiments, Advanced Boiling Water Reactor (ABWR), Advanced Pressurized Water Reactor (APWR), High Temperature Engineering Test Reactor (HTTR), Experimental and Prototype Fast Reactors (JOYO, MONJU), Demonstration FBR, core shroud replacement of BWR, and spent fuel transportation cask and vessel. These studies have valuable information in safety and cost reduction issues of fission reactor design for not only existing reactors but also new reactor concepts in the next century. It has been concluded that we should maintain existing shielding technologies and improve these data and methods for coming generations in the next millennium. (author)

  8. Shield cost minimization using SWAN

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  9. Radiation shield for nuclear reactors

    International Nuclear Information System (INIS)

    Weissenfluh, J.A.

    1980-01-01

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

  10. Shielding walls against ionizing radiation

    International Nuclear Information System (INIS)

    1993-05-01

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

  11. Nuclear steam generator tubesheet shield

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  12. Shielding structure analysis for LSDS facility

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  13. Concrete shielding for nuclear ship 'Mutsu'

    International Nuclear Information System (INIS)

    Nagase, Tetsuo; Saito, Tetsuo

    1983-01-01

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

  14. Shielding structure analysis for LSDS facility

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  15. Accelerator shield design of KIPT neutron source facility

    International Nuclear Information System (INIS)

    Zhong, Z.; Gohar, Y.

    2013-01-01

    Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the design development of a neutron source facility at KIPT utilizing an electron-accelerator-driven subcritical assembly. Electron beam power is 100 kW, using 100 MeV electrons. The facility is designed to perform basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The biological shield of the accelerator building is designed to reduce the biological dose to less than 0.5-mrem/hr during operation. The main source of the biological dose is the photons and the neutrons generated by interactions of leaked electrons from the electron gun and accelerator sections with the surrounding concrete and accelerator materials. The Monte Carlo code MCNPX serves as the calculation tool for the shield design, due to its capability to transport electrons, photons, and neutrons coupled problems. The direct photon dose can be tallied by MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is less than 0.01 neutron per electron. This causes difficulties for Monte Carlo analyses and consumes tremendous computation time for tallying with acceptable statistics the neutron dose outside the shield boundary. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were developed for the study. The generated neutrons are banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron and secondary photon doses. The weight windows variance reduction technique is utilized for both neutron and photon dose calculations. Two shielding materials, i.e., heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total

  16. Electron accelerator shielding design of KIPT neutron source facility

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, Zhao Peng; Gohar, Yousry [Argonne National Laboratory, Argonne (United States)

    2016-06-15

    The Argonne National Laboratory of the United States and the Kharkov Institute of Physics and Technology of the Ukraine have been collaborating on the design, development and construction of a neutron source facility at Kharkov Institute of Physics and Technology utilizing an electron-accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100-MeV electrons. The facility was designed to perform basic and applied nuclear research, produce medical isotopes, and train nuclear specialists. The biological shield of the accelerator building was designed to reduce the biological dose to less than 5.0e-03 mSv/h during operation. The main source of the biological dose for the accelerator building is the photons and neutrons generated from different interactions of leaked electrons from the electron gun and the accelerator sections with the surrounding components and materials. The Monte Carlo N-particle extended code (MCNPX) was used for the shielding calculations because of its capability to perform electron-, photon-, and neutron-coupled transport simulations. The photon dose was tallied using the MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is very small, ∼0.01 neutron for 100-MeV electron and even smaller for lower-energy electrons. This causes difficulties for the Monte Carlo analyses and consumes tremendous computation resources for tallying the neutron dose outside the shield boundary with an acceptable accuracy. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were utilized for this study. The generated neutrons were banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron dose. The weight windows variance reduction technique was also utilized for both neutron and photon dose

  17. Method of constructing shielding wall

    International Nuclear Information System (INIS)

    Nagao, Tetsuya.

    1990-01-01

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

  18. Shielding Effectiveness of a Thin Film Window

    National Research Council Canada - National Science Library

    Johnson, Eric

    1998-01-01

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

  19. Gonadal Shielding in Radiography: A Best Practice?

    Science.gov (United States)

    Fauber, Terri L

    2016-11-01

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

  20. Infinite slab-shield dose calculations

    International Nuclear Information System (INIS)

    Russell, G.J.

    1989-01-01

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

  1. Using glass as a shielding material

    International Nuclear Information System (INIS)

    Yousef, S.

    2002-04-01

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

  2. Using glass as a shielding material

    International Nuclear Information System (INIS)

    Yousef, S.

    2003-01-01

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

  3. Shielding and grounding in large detectors

    International Nuclear Information System (INIS)

    Radeka, V.

    1998-09-01

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

  4. Neutron shielding properties of boron-containing ore and epoxy composites

    International Nuclear Information System (INIS)

    Li Zhifu; Xue Xiangxin

    2011-01-01

    Using the boron-containing iron ore concentrate and boron-rich slag as studying object, the starting materials were got after the specific green ore containing boron dressing in China and blast furnace separation respectively. Monte-Carlo method was used to study the effect of the boron-containing iron ore concentrate and boron-rich slag and their composites with epoxy on the neutron shielding abilities. The reasons that affecting the shielding materials properties was discussed and the suitable proportioning of boron-containing ore to epoxy composites was confirmed; the 14.1 MeV fast neutron removal cross section and the total thermal neutron attenuation coefficient were obtained and compared with that of the common used concrete. The results show that the shielding property of 14.1 MeV fast neutron is mainly concerned with the low-Z elements in the shielding materials, the thermal neutron shielding ability is mainly concerned with boron concentrate in the composite, the attenuation of the accompany γ-ray photon is mainly concerned with the high atom number elements content in the ore and the density of the shielding material. The optimum Janume fractions of composites are in the range of 0.4-0.6 and the fast neutron shielding properties are similar to concrete while the thermal neutron shielding properties are higher than concrete. The composites are expected to be used as biological concrete shields crack injection and filling of the anomalous holes through the concrete shields around the radiation fields or directly to be prepared as shielding materials.(authors)

  5. Pretinning Nickel-Plated Wire Shields

    Science.gov (United States)

    Igawa, J. A.

    1985-01-01

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

  6. Computed tomography shielding methods: a literature review.

    Science.gov (United States)

    Curtis, Jessica Ryann

    2010-01-01

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

  7. BRH Gonad Shielding Program: where it has led

    International Nuclear Information System (INIS)

    Arcarese, J.S.

    1975-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-04-15

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  10. New facility shield design criteria

    International Nuclear Information System (INIS)

    Howell, W.P.

    1981-07-01

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

  11. INERT GAS SHIELD FOR WELDING

    Science.gov (United States)

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

    1958-10-14

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

  12. Facility target insert shielding assessment

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-06

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

  13. Transient shielded liquid hydrogen containers

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  14. Shielding calculational system for plutonium

    International Nuclear Information System (INIS)

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

    1975-08-01

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

  15. Radiation-shielding transparent material

    International Nuclear Information System (INIS)

    Kusumeki, Asao.

    1983-01-01

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

  16. Radiation shield vest and skirt

    International Nuclear Information System (INIS)

    Maine, G.J.

    1982-01-01

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

  17. Handbook of radiation shielding data

    International Nuclear Information System (INIS)

    Courtney, J.C.

    1976-07-01

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

  18. Shielding wall for thermonuclear device

    International Nuclear Information System (INIS)

    Uchida, Takaho.

    1989-01-01

    This invention concerns shielding walls opposing to plasmas of a thermonuclear device and it is an object thereof to conduct reactor operation with no troubles even if a portion of shielding wall tiles should be damaged. That is, the shielding wall tiles are constituted as a dual layer structure in which the lower base tiles are connected by means of bolts to first walls. Further, the upper surface tiles are bolt-connected to the layer base tiles. In this structure, the plasma thermal loads are directly received by the surface layer tiles and heat is conducted by means of conduction and radiation to the underlying base tiles and the first walls. Even upon occurrence of destruction accidents to the surface layer tiles caused by incident heat or electromagnetic force upon elimination of plasmas, since the underlying base tiles remain as they are, the first walls constituted with stainless steels, etc. are not directly exposed to the plasmas. Accordingly, the integrity of the first walls having cooling channels can be maintained and sputtering intrusion of atoms of high atom number into the plasmas can be prevented. (I.S.)

  19. Design of ITER shielding blanket

    International Nuclear Information System (INIS)

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

    1997-05-01

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

  20. Photonic Bandgap (PBG) Shielding Technology

    Science.gov (United States)

    Bastin, Gary L.

    2007-01-01

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

  1. Reactor vessel head permanent shield

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  2. ATLAS Award for Shield Supplier

    CERN Multimedia

    2004-01-01

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

  3. Concrete shielding for nuclear ship 'Mutsu'

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  4. Special concrete shield selection using the analytic hierarchy process

    International Nuclear Information System (INIS)

    Abulfaraj, W.H.

    1994-01-01

    Special types of concrete radiation shields that depend on locally available materials and have improved properties for both neutron and gamma-ray attenuation were developed by using plastic materials and heavy ores. The analytic hierarchy process (AHP) is implemented to evaluate these types for selecting the best biological radiation shield for nuclear reactors. Factors affecting the selection decision are degree of protection against neutrons, degree of protection against gamma rays, suitability of the concrete as building material, and economic considerations. The seven concrete alternatives are barite-polyethylene concrete, barite-polyvinyl chloride (PVC) concrete, barite-portland cement concrete, pyrite-polyethylene concrete, pyrite-PVC concrete, pyrite-portland cement concrete, and ordinary concrete. The AHP analysis shows the superiority of pyrite-polyethylene concrete over the others

  5. Shielding calculations for the TFTR neutral beam injectors

    International Nuclear Information System (INIS)

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

    1979-07-01

    Two-dimensional discrete ordinates calculations have been performed to determine the location and thickness of concrete shielding around the Tokamak Fusion Test Reactor (TFTR) neutral beam injectors. Two sets of calculations were performed: one to determine the dose equivalent rate on the roof and walls of the test cell building when no injectors are present, and one to determine the contribution to the dose equivalent rate at these locations from radiation streaming through the injection duct. Shielding the side and rear of the neutral beam injector with 0.305 and 0.61 m of concrete, respectively, and lining the inside of the test cell wall with an additional layer of concrete having a thickness of 0.305 m and a height above the axis of deuteron injection of 3.10 m are sufficient to maintain the biological dose equivalent rate outside the test cell to approx. 1 mrem/DT pulse

  6. Shielded room measurements, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Stanton, J.S.

    1949-02-22

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

  7. Radiation shielding activities at IDOM

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  8. Magnetic shielding of a limiter

    International Nuclear Information System (INIS)

    Brevnov, N.N.; Stepanov, S.B.; Khimchenko, L.N.; Matthews, G.F.; Goodal, D.H.J.

    1991-01-01

    Localization of plasma interaction with material surfaces in a separate chamber, from where the escape of impurities is hardly realized, i.e. application of magnetic divertors or pump limiters, is the main technique for reduction of the impurity content in a plasma. In this case, the production of a divertor configuration requires a considerable power consumption and results in a less effective utilization of the magnetic field volume. Utilization of a pump limiter, for example the ICL-type, under tokamak-reactor conditions would result in the extremely high and forbidden local heat loadings onto the limiter surface. Moreover, the magnetically-shielded pump limiter (MSL) was proposed to combine positive properties of the divertor and the pump limiter. The idea of magnetic shielding is to locate the winding with current inside the limiter head so that the field lines of the resultant magnetic field do not intercept the limiter surface. In this case the plasma flows around the limiter leading edges and penetrates into the space under the limiter. The shielding magnetic field can be directed either counter the toroidal field or counter the poloidal one of a tokamak, dependent on the concrete diagram of the device. Such a limiter has a number of advantages: -opportunity to control over the particle and impurity recycling without practical influence upon the plasma column geometry, - perturbation of a plasma column magnetic configuration from the side of such a limiter is less than that from the side of the divertor coils. The main deficiency is the necessity to locate active windings inside the discharge chamber. (author) 5 refs., 3 figs

  9. Radiation shielding activities at IDOM

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  10. Core test reactor shield cooling system analysis

    International Nuclear Information System (INIS)

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

    1971-01-01

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

  11. Highly heat removing radiation shielding material

    International Nuclear Information System (INIS)

    Asano, Norio; Hozumi, Masahiro.

    1990-01-01

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

  12. Shielding design of ITER pressure suppression system

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  13. Irrigoscopy - irrigography method, dosimetry and radiation shielding

    International Nuclear Information System (INIS)

    Zubanov, Z.; Kolarevic, G.

    1999-01-01

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

  14. Cage for shield-type support. Schildausbaugestell

    Energy Technology Data Exchange (ETDEWEB)

    Harryers, W; Blumenthal, G; Irresberger, H

    1981-08-13

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

  15. Shielded scanning electron microscope for radioactive samples

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  16. Problems of the power plant shield optimization

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  17. Neutron shielding for a 252 Cf source

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  18. Hot Cell Window Shielding Analysis Using MCNP

    International Nuclear Information System (INIS)

    Pope, Chad L.; Scates, Wade W.; Taylor, J. Todd

    2009-01-01

    The Idaho National Laboratory Materials and Fuels Complex nuclear facilities are undergoing a documented safety analysis upgrade. In conjunction with the upgrade effort, shielding analysis of the Fuel Conditioning Facility (FCF) hot cell windows has been conducted. This paper describes the shielding analysis methodology. Each 4-ft thick window uses nine glass slabs, an oil film between the slabs, numerous steel plates, and packed lead wool. Operations in the hot cell center on used nuclear fuel (UNF) processing. Prior to the shielding analysis, shield testing with a gamma ray source was conducted, and the windows were found to be very effective gamma shields. Despite these results, because the glass contained significant amounts of lead and little neutron absorbing material, some doubt lingered regarding the effectiveness of the windows in neutron shielding situations, such as during an accidental criticality. MCNP was selected as an analysis tool because it could model complicated geometry, and it could track gamma and neutron radiation. A bounding criticality source was developed based on the composition of the UNF. Additionally, a bounding gamma source was developed based on the fission product content of the UNF. Modeling the windows required field inspections and detailed examination of drawings and material specifications. Consistent with the shield testing results, MCNP results demonstrated that the shielding was very effective with respect to gamma radiation, and in addition, the analysis demonstrated that the shielding was also very effective during an accidental criticality.

  19. Radiation shielding application of lead glass

    International Nuclear Information System (INIS)

    Nathuram, R.

    2017-01-01

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

  20. Radiation dose reduction by water shield

    International Nuclear Information System (INIS)

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

    2007-06-01

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

  1. Enhancement of thermal neutron shielding of cement mortar by using borosilicate glass powder.

    Science.gov (United States)

    Jang, Bo-Kil; Lee, Jun-Cheol; Kim, Ji-Hyun; Chung, Chul-Woo

    2017-05-01

    Concrete has been used as a traditional biological shielding material. High hydrogen content in concrete also effectively attenuates high-energy fast neutrons. However, concrete does not have strong protection against thermal neutrons because of the lack of boron compound. In this research, boron was added in the form of borosilicate glass powder to increase the neutron shielding property of cement mortar. Borosilicate glass powder was chosen in order to have beneficial pozzolanic activity and to avoid deleterious expansion caused by an alkali-silica reaction. According to the experimental results, borosilicate glass powder with an average particle size of 13µm showed pozzolanic activity. The replacement of borosilicate glass powder with cement caused a slight increase in the 28-day compressive strength. However, the incorporation of borosilicate glass powder resulted in higher thermal neutron shielding capability. Thus, borosilicate glass powder can be used as a good mineral additive for various radiation shielding purposes. Copyright © 2017 Elsevier Ltd. All rights reserved.

  2. Decontaminating lead bricks and shielding

    International Nuclear Information System (INIS)

    Lussiez, G.

    1994-01-01

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

  3. Decontaminating lead bricks and shielding

    International Nuclear Information System (INIS)

    Lussiez, G.W.

    1993-01-01

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

  4. Decontaminating lead bricks and shielding

    International Nuclear Information System (INIS)

    Lussiez, G.W.

    1993-01-01

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

  5. TPX remote maintenance and shielding

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  6. Isotope effects on nuclear shielding

    International Nuclear Information System (INIS)

    Hansen, P.E.

    1983-01-01

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

  7. Continuous electrodeionization through electrostatic shielding

    International Nuclear Information System (INIS)

    Dermentzis, Konstantinos

    2008-01-01

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

  8. Improvements in or relating to nuclear shields

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  9. Shielding effectiveness of superconductive particles in plastics

    International Nuclear Information System (INIS)

    Pienkowski, T.; Kincaid, J.; Lanagan, M.T.; Poeppel, R.B.; Dusek, J.T.; Shi, D.; Goretta, K.C.

    1988-09-01

    The ability to cool superconductors with liquid nitrogen instead of liquid helium has opened the door to a wide range of research. The well known Meissner effect, which states superconductors are perfectly diamagnetic, suggests shielding applications. One of the drawbacks to the new ceramic superconductors is the brittleness of the finished material. Because of this drawback, any application which required flexibility (e.g., wire and cable) would be impractical. Therefore, this paper presents the results of a preliminary investigation into the shielding effectiveness of YBa 2 Cu 3 O/sub 7-x/ both as a composite and as a monolithic material. Shielding effectiveness was measured using two separate test methods. One tested the magnetic (near field) shielding, and the other tested the electromagnetic (far field) shielding. No shielding was seen in the near field measurements on the composite samples, and only one heavily loaded sample showed some shielding in the far field. The monolithic samples showed a large amount of magnetic shielding. 5 refs., 5 figs

  10. Preliminary radiation shielding design for BOOMERANG

    International Nuclear Information System (INIS)

    Donahue, Richard J.

    2002-01-01

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

  11. Actively shielded low level gamma - spectrometric system

    International Nuclear Information System (INIS)

    Mrdja, D.; Bikit, I.; Forkapic, S.; Slivka, J.; Veskovic, M.

    2005-01-01

    The results of the adjusting and testing of the actively shielded low level gamma-spectrometry system are presented. The veto action of the shield reduces the background in the energy region of 50 keV to the 2800 keV for about 3 times. (author) [sr

  12. Optimization of multi-layered metallic shield

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  13. Alignment modification for pencil eye shields

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  14. Several problems in accelerator shielding study

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  15. Collection shield for ion separation apparatus

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  16. Current status of methods for shielding analysis

    International Nuclear Information System (INIS)

    Engle, W.W.

    1980-01-01

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

  17. Thermal neutron shield and method of manufacture

    Science.gov (United States)

    Brindza, Paul Daniel; Metzger, Bert Clayton

    2013-05-28

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  19. Radiation Shielding Systems Using Nanotechnology

    Science.gov (United States)

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

    2011-01-01

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

  20. Radiation shielding and safety design

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-15

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

  1. Shield verification and validation action matrix summary

    International Nuclear Information System (INIS)

    Boman, C.

    1992-02-01

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

  2. Practical radiation shielding for biomedical research

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  3. PC based temporary shielding administrative procedure (TSAP)

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  4. A perturbation technique for shield weight minimization

    International Nuclear Information System (INIS)

    Watkins, E.F.; Greenspan, E.

    1993-01-01

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

  5. Mars Exploration Rover Heat Shield Recontact Analysis

    Science.gov (United States)

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

    2011-01-01

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

  6. Technology development for radiation shielding analysis

    International Nuclear Information System (INIS)

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

    1986-12-01

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

  7. PC based temporary shielding administrative procedure (TSAP)

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-03-01

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

  8. Repository Waste Package Transporter Shielding Weight Optimization

    International Nuclear Information System (INIS)

    C.E. Sanders; Shiaw-Der Su

    2005-01-01

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

  9. The removal of concrete layers from biological shields by microwaves

    International Nuclear Information System (INIS)

    Hills, D.L.

    1989-01-01

    Concrete blocks reinforced with steel bars have been subjected to microwave attack at a frequency of 896 MHz at power levels up to 25 kW. The surface concrete has been explosively removed to the depth of the reinforcement, 10 cm, at a rate of about 2 litres per kWh. Heating was localized around the point of attack, with temperatures up to 300 0 C at the fractured face being attained. A simple mathematical model of the propagation and absorption of micro-waves was used to estimate the temperature rise of concrete at microwave frequencies of 896 wand 2450 MHz, at different power levels with and without the presence of reinforcing bars. This demonstrated that reinforcement is expected to significantly increase the temperature rise in the concrete between the irradiated surface and the reinforcement, and that near-surface heating should be more rapid at the higher frequency. There was reasonable agreement between predicted and observed temperature at the higher power levels. Further desk and laboratory studies are proposed before proceeding to a fullscale practical demolition machine and the requirements for a prototype remotely-operated demonstration system have been identified. This consists of a static generator of high power (at least 50 kW) transmitting microwaves via a steerable waveguide to a remote applicator mounted on a simple three-axis manipulator capable of traversing realistically large concrete test panels

  10. The use of nipple shields: A review

    Directory of Open Access Journals (Sweden)

    Selina Chow

    2016-11-01

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

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

    Data.gov (United States)

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

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

    Data.gov (United States)

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

  13. Study on bulk shielding for a spallation neutron source facility in the high-intensity proton accelerator project

    CERN Document Server

    Maekawa, F; Takada, H; Teshigawara, M; Watanabe, N

    2002-01-01

    Under the JAERI-KEK High-Intensity Proton Accelerator Project, a spallation neutron source driven by a 3 GeV-1 MW proton beam is planed to be constructed in a main part of the Materials and Life Science Facility. This report describes results of a study on bulk shielding performance of a biological shield for the spallation neutron source by means of a Monte Carlo calculation method, that is important in terms of radiation safety and cost reduction. A shielding configuration was determined as a reference case by considering preliminary studies and interaction with other components, then shielding thickness that was required to achieve a target dose rate of 1 mu Sv/h was derived. Effects of calculation conditions such as shielding materials and dimensions on the shielding performance was investigated by changing those parameters. By taking all the results and design margins into account, a shielding configuration that was identified as the most appropriate was finally determined as follows. An iron shield regi...

  14. Shielding concerns at a spallation source

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  15. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.H.

    1992-09-01

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

  16. Development of neutron shielding material for cask

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  17. Heating profiles on ICRF antenna Faraday shields

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  18. PWR upper/lower internals shield

    Energy Technology Data Exchange (ETDEWEB)

    Homyk, W.A. [Indian Point Station, Buchanan, NY (United States)

    1995-03-01

    During refueling of a nuclear power plant, the reactor upper internals must be removed from the reactor vessel to permit transfer of the fuel. The upper internals are stored in the flooded reactor cavity. Refueling personnel working in containment at a number of nuclear stations typically receive radiation exposure from a portion of the highly contaminated upper intervals package which extends above the normal water level of the refueling pool. This same issue exists with reactor lower internals withdrawn for inservice inspection activities. One solution to this problem is to provide adequate shielding of the unimmersed portion. The use of lead sheets or blankets for shielding of the protruding components would be time consuming and require more effort for installation since the shielding mass would need to be transported to a support structure over the refueling pool. A preferable approach is to use the existing shielding mass of the refueling pool water. A method of shielding was devised which would use a vacuum pump to draw refueling pool water into an inverted canister suspended over the upper internals to provide shielding from the normally exposed components. During the Spring 1993 refueling of Indian Point 2 (IP2), a prototype shield device was demonstrated. This shield consists of a cylindrical tank open at the bottom that is suspended over the refueling pool with I-beams. The lower lip of the tank is two feet below normal pool level. After installation, the air width of the natural shielding provided by the existing pool water. This paper describes the design, development, testing and demonstration of the prototype device.

  19. Mechanical design of the TIBER breeding shield

    Energy Technology Data Exchange (ETDEWEB)

    Rathke, J.; Deutsch, L. (Grumman Corp., Bethpage, NY (USA). Space Systems Div.)

    1989-04-01

    TIBER features a segmented shield assembly that provides the nuclear shielding for the superconducting toroidal field coils. In addition to its primary function, the shield also provides tritium breeding through the use of water coolant that contains 16 wt% dissolved lithium nitrate. Because the TIBER reactor need not provide electrical power, the coolant is maintained at low pressure (0.2 MPa) and low temperature (75/sup 0/C). The shield is made in several segments to facilitate assembly and allow for replacement of high heat flux components (divertor blades). The segments are designated as inboard, outboard, upper, lower, and divertor modules. In total, there are 96 separate modules in the machine, consisting of six different types. The design features of the different modules vary primarily depending on the thickness of the shield in a given location. The very thick outboard shield has a breeding zone in the inboard portion of the module, with a shielding zone behind it. The breeding zone consists of a stainless steel casing filled with beryllium spheres. The shielding zone consists of the same casing filled with steel spheres. Both of these zones have lithiated water circulated throughout to provide cooling and breeding. In zones with minimal thickness, tungsten alloys are used to achieve the required shielding. These alloys are incoprorated in subassemblies utilizing stainless steel casings surrounding blocks of tungsten heavy metal alloy. These are infiltrated with lead on final assembly to form a thermally continuous panel. Several of these panels are then assembled into an outer stainless steel case to form an inboard module. These modules also use the lithiated coolant. The details of the design are presented and discussed. (orig.).

  20. Scale-4 shipping cask shielding applications

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  1. Radiation shielding for TFTR DT diagnostics

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  2. MFTF-α + T shield design

    International Nuclear Information System (INIS)

    Gohar, Y.

    1985-01-01

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

  3. Seismic proof test of shielding block walls

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  4. REPOSITORY RADIATION SHIELDING DESIGN GUIDE

    International Nuclear Information System (INIS)

    M. Haas; E.M. Fortsch

    1997-01-01

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

  5. REPOSITORY RADIATION SHIELDING DESIGN GUIDE

    Energy Technology Data Exchange (ETDEWEB)

    M. Haas; E.M. Fortsch

    1997-09-12

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

  6. Particle Tracing Modeling with SHIELDS

    Science.gov (United States)

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

    2017-12-01

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

  7. Thermal design of top shield

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Nurhayati Ramli; Ahmad Nabil Abdul Rahim; Ariff Shah Ismail

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  10. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.H.

    1990-03-01

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

  11. Multifunctional BHL Radiation Shield, Phase I

    Data.gov (United States)

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

  12. Radiation shielding method for pipes, etc

    International Nuclear Information System (INIS)

    Nagao, Tetsuya; Takahashi, Shuichi.

    1988-01-01

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

  13. Shielding design for better plant availability

    International Nuclear Information System (INIS)

    Biro, G.G.

    1975-01-01

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

  14. Long Duration Space Shelter Shielding, Phase I

    Data.gov (United States)

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

  15. Technical specifications for the bulk shielding reactor

    International Nuclear Information System (INIS)

    1986-05-01

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

  16. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.N.

    1990-01-01

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

  17. Radiation shielding structure for concrete structure

    International Nuclear Information System (INIS)

    Oya, Hiroshi

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    1981-01-01

    The object of this standard description is the general technological requirements of 50 and 100 mm thick radiation protection shields assembled from lead blocks. The standard contains the definitions, types, parameters and dimensions of shields, their technical and acceptance criteria with testing methods, tagging, packaging, transportation and storage requirements, producer's liability. Some illustrated assembling examples, preferred parameters and dosimetry methods for shield inspection are given. (R.P.)

  19. Status report of shielding investigation in Japan

    International Nuclear Information System (INIS)

    Shindo, M.

    1964-01-01

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

  20. Nuclear reactor shield including magnesium oxide

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  1. Shield structure for a nuclear reactor

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  2. Measurement accuracy in shielded magnetic fields

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  3. Influence of Shielding Arrangement on ECT Sensors

    Directory of Open Access Journals (Sweden)

    J. L. Fernandez Marron

    2006-09-01

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

  4. Acoustic Metacages for Omnidirectional Sound Shielding

    OpenAIRE

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

    2017-01-01

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

  5. Method to produce a neutron shielding

    International Nuclear Information System (INIS)

    Merkle, H.J.

    1978-01-01

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

  6. Shielded transport containers for reactor waste

    International Nuclear Information System (INIS)

    Grundfelt, B.; Eriksson, E.

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

  7. Radiation shielding of the main injector

    International Nuclear Information System (INIS)

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

    1995-05-01

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

  8. ANALISIS KESELAMATAN TERMOHIDROLIK BULK SHIELDING REAKTOR KARTINI

    Directory of Open Access Journals (Sweden)

    Azizul Khakim

    2015-10-01

    Full Text Available ABSTRAK ANALISIS KESELAMATAN TERMOHIDROLIK BULK SHIELDING REAKTOR KARTINI. Bulk shielding merupakan fasilitas yang terintegrasi dengan reaktor Kartini yang berfungsi sebagai penyimpanan sementara bahan bakar bekas. Fasilitas ini merupakan fasilitas yang termasuk dalam struktur, sistem dan komponen (SSK yang penting bagi keselamatan. Salah satu fungsi keselamatan dari sistem penanganan dan penyimpanan bahan bakar adalah mencegah kecelakaan kekritisan yang tak terkendali dan membatasi naiknya temperatur bahan bakar. Analisis keselamatan paling kurang harus mencakup analisis keselamatan dari sisi neutronik dan termo hidrolik Bulk shielding. Analisis termo hidrolik ditujukan untuk memastikan perpindahan panas dan proses pendinginan bahan bakar bekas berjalan baik dan tidak terjadi akumulasi panas yang mengancam integritas bahan bakar. Code tervalidasi PARET/ANL digunakan untuk analisis pendinginan dengan mode konveksi alam. Hasil perhitungan menunjukkan bahwa mode pendinginan konvekasi alam cukup memadai dalam mendinginkan panas sisa tanpa mengakibatkan kenaikan temperatur bahan bakar yang signifikan. Kata kunci: Bulk shielding, bahan bakar bekas, konveksi alam, PARET.   ABSTRACT THERMAL HYDRAULIC SAFETY ANALYSIS OF BULK SHIELDING KARTINI REACTOR. Bulk shielding is an integrated facility to Kartini reactor which is used for temporary spent fuels storage. The facility is one of the structures, systems and components (SSCs important to safety. Among the safety functions of fuel handling and storage are to prevent any uncontrolable criticality accidents and to limit the fuel temperature increase. Safety analyses should, at least, cover neutronic and thermal hydraulic calculations of the bulk shielding. Thermal hydraulic analyses were intended to ensure that heat removal and the process of the spent fuels cooling takes place adequately and no heat accumulation that challenges the fuel integrity. Validated code, PARET/ANL was used for analysing the

  9. Shielding modefication and safety review on Mutsu

    International Nuclear Information System (INIS)

    Osanai, Masao

    1978-01-01

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

  10. Shielding requirements for particle bed propulsion systems

    Science.gov (United States)

    Gruneisen, S. J.

    1991-06-01

    Nuclear Thermal Propulsion systems present unique challenges in reliability and safety. Due to the radiation incident upon all components of the propulsion system, shielding must be used to keep nuclear heating in the materials within limits; in addition, electronic control systems must be protected. This report analyzes the nuclear heating due to the radiation and the shielding required to meet the established criteria while also minimizing the shield mass. Heating rates were determined in a 2000 MWt Particle Bed Reactor (PBR) system for all materials in the interstage region, between the reactor vessel and the propellant tank, with special emphasis on meeting the silicon dose criteria. Using a Lithium Hydride/Tungsten shield, the optimum shield design was found to be: 50 cm LiH/2 cm W on the axial reflector in the reactor vessel and 50 cm LiH/2 cm W in a collar extension of the inside shield outside of the pressure vessel. Within these parameters, the radiation doses in all of the components in the interstage and lower tank regions would be within acceptable limits for mission requirements.

  11. Potential of Nanocellulose Composite for Electromagnetic Shielding

    Directory of Open Access Journals (Sweden)

    Nabila Yah Nurul Fatihah

    2017-01-01

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

  12. Innovative technologies for Faraday shield cooling

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  13. Space Shielding Materials for Prometheus Application

    Energy Technology Data Exchange (ETDEWEB)

    R. Lewis

    2006-01-20

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

  14. Shielding technology for high energy radiation production facility

    International Nuclear Information System (INIS)

    Lee, Byung Chul; Kim, Heon Il

    2004-06-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  16. Study and installation of concrete shielding in the civil engineering of nuclear construction (1960)

    International Nuclear Information System (INIS)

    Dubois, F.

    1960-01-01

    The object of this report is to give technical information about high density concretes which have become very important for radiation biological shielding. The most generally used heavy aggregates (barytes, ilmenite, ferrophosphorus, limonite, magnetite and iron punching) to make these concretes are investigated from the point of view prospecting and physical and chemical characteristics. At first, a general survey of shielding concretes is made involving the study of components, mixing and placing methods, then, a detailed investigation of some high density concretes: barytes concrete, with incorporation of iron punching or iron shot, ferrophosphorus concrete, ilmenite concrete and magnetite concrete, more particularly with regard to grading and mix proportions and testing process. To put this survey in concrete form, two practical designs are described such as they have been carried out at the Saclay Nuclear Station. Specifications are given for diverse concretes and for making the proton-synchrotron 'Saturne' shielding blocks. (author) [fr

  17. Shielding design for positron emission tomography facility

    International Nuclear Information System (INIS)

    Abdallah, I.I.

    2007-01-01

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

  18. Evaluation of the shielding integrity of end-shields in PHWR type NPPs

    International Nuclear Information System (INIS)

    Sah, B.M.L.; Ramamirtham, B.; Kutty, B.S.

    1996-01-01

    In the new plants (Narora Atomic Power Plants (NAPP) onwards) relatively higher radiation fields exist on the north and south fuelling machine (FM) vault walls of the E1 100m accessible area passages. These fields were first noticed at NAPS-1 and subsequently at NAPS-2 and KAPS-1. Such surveys done at RAPS have indicated that the fields on these walls would come out to be quite low (only 1-2 mR/h) from sources other than that arising from 41 Ar contamination. RAPS/MAPS experience pointed to adequacy of shielding of the FM vault walls and sufficient overall shielding thickness of the end-shields. Further, radiometry tests of end-shields carried out at Kaiga and RAPP 3 and 4 indicated fairly satisfactory and uniform filling of balls. Hence, incomplete filling of water column of the end-shields due to any venting problem was suspected to be one possible reason for the observed high fields in NAPS and Kakrapar Atomic Power Station (KAPS). Since the presence of high radiation fields, both neutron and gamma, is of long-term concern, a special study/measurement of radiation levels on reactor face during high power operation was undertaken. In order to compare the shielding integrity of the older (RAPS/MAPS solid plate type shielding) and newer (NAPS/KAPS steel ball-filled type) end shields, these experiments were done at MAPS-2 and NAPS-2. (author). 2 refs., 2 tabs

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

    International Nuclear Information System (INIS)

    1970-07-01

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

  20. Comparison of eye shields in radiotherapeutic beams

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  1. Hydrogen Induced Cracking of Drip Shield

    Energy Technology Data Exchange (ETDEWEB)

    G. De

    2003-02-24

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

  2. Analytic Ballistic Performance Model of Whipple Shields

    Science.gov (United States)

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

    2015-01-01

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

  3. Shielding methods development in the United States

    International Nuclear Information System (INIS)

    Mynatt, F.R.

    1977-01-01

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

  4. Development of HANARO ST3 shield

    International Nuclear Information System (INIS)

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

    2004-12-01

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

  5. Cosmic Ray Interactions in Shielding Materials

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  6. Bremsstrahlung converter debris shields: test and analysis

    International Nuclear Information System (INIS)

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

    1983-10-01

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

  7. Preliminary shielding analysis of VHTR reactors

    International Nuclear Information System (INIS)

    Flaspoehler, Timothy M.; Petrovic, Bojan

    2011-01-01

    Over the last 20 years a number of methods have been established for automated variance reduction in Monte Carlo shielding simulations. Hybrid methods rely on deterministic adjoint and/or forward calculations to generate these parameters. In the present study, we use the FWCADIS method implemented in MAVRIC sequence of the SCALE6 package to perform preliminary shielding analyses of a VHTR reactor. MAVRIC has been successfully used by a number of researchers for a range of shielding applications, including modeling of LWRs, spent fuel storage, radiation field throughout a nuclear power plant, study of irradiation facilities, and others. However, experience in using MAVRIC for shielding studies of VHTRs is more limited. Thus, the objective of this work is to contribute toward validating MAVRIC for such applications, and identify areas for potential improvement. A simplified model of a prismatic VHTR has been devised, based on general features of the 600 MWt reactor considered as one of the NGNP options. Fuel elements have been homogenized, and the core region is represented as an annulus. However, the overall mix of materials and the relatively large dimensions of the spatial domain challenging the shielding simulations have been preserved. Simulations are performed to evaluate fast neutron fluence, dpa, and other parameters of interest at relevant positions. The paper will investigate and discuss both the effectiveness of the automated variance reduction, as well as applicability of physics model from the standpoint of specific VHTR features. (author)

  8. Radiation shielding fiber and its manufacturing method

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, Koji; Ono, Hiroshi.

    1988-08-17

    Purpose: To manufacture radiation shielding fibers of excellent shielding effects. Method: Fibers containing more than 1 mmol/g of carboxyl groups are bonded with heavy metals, or they are impregnated with an aqueous solution containing water-soluble heavy metal salts dissolved therein. Fibers as the substrate may be any of forms such as short fibers, long fibers, fiber tows, webs, threads, knitting or woven products, non-woven fabrics, etc. It is however necessary that fibers contain more than 1 mmol/g, preferably, from 2 to 7 mmol/g of carboxylic groups. Since heavy metals having radiation shielding performance are bonded to the outer layer of the fibers and the inherent performance of the fibers per se is possessed, excellent radiation shielding performance can be obtained, as well as they can be applied with spinning, knitting or weaving, stitching, etc. thus can be used for secondary fiber products such as clothings, caps, masks, curtains, carpets, cloths, etc. for use in radiation shieldings. (Kamimura, M.).

  9. MEANS FOR SHIELDING AND COOLING REACTORS

    Science.gov (United States)

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

    1959-02-10

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

  10. Shielding design to obtain compact marine reactor

    International Nuclear Information System (INIS)

    Yamaji, Akio; Sako, Kiyoshi

    1994-01-01

    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)

  11. Design report for shielded glove box

    International Nuclear Information System (INIS)

    Ku, J. H.; Lee, J. C.; Seo, K. S.; Bang, K. S.; Lee, D. W.; Kim, J. H.; Min, D. K.; Park, S. W.

    1999-05-01

    For the examination of spent fuels and high radioactive specimens using a specially equipped scanning electron microscope, a shielded glove box was designed and constructed at PIE facility of KAERI. This glove box consisted of shielding walls, containment box, lead glasses, manipulators, gloves, ventilation systems, doors, hot-cell specimen cask adapter, etc. It was emphasized that both the easy operation and radiation safety are important factors in the shielded glove box were installed also considered as a important factor to build the basic concept of the assembling. Two sliding doors and one hinge-type door were installed for the easy installation, operation and maintenance of scanning electron microscope. Containment box which confines the radioactive material into the box consisted of reinforced transparent glasses, aluminum frames and stainless steel plate liner. Therefore everything beyond the containment box can be seen through the lead glass which installed at the front shielding wall. All shielding walls and doors were introduced separately into the room and assembled by bolting. (author). 3 refs., 5 tabs., 18 figs

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  13. Development of neutron shielding concrete containing iron content materials

    Science.gov (United States)

    Sariyer, Demet; Küçer, Rahmi

    2018-02-01

    Concrete is one of the most important construction materials which widely used as a neutron shielding. Neutron shield is obtained of interaction with matter depends on neutron energy and the density of the shielding material. Shielding properties of concrete could be improved by changing its composition and density. High density materials such as iron or high atomic number elements are added to concrete to increase the radiation resistance property. In this study, shielding properties of concrete were investigated by adding iron, FeB, Fe2B, stainless - steel at different ratios into concrete. Neutron dose distributions and shield design was obtained by using FLUKA Monte Carlo code. The determined shield thicknesses vary depending on the densities of the mixture formed by the additional material and ratio. It is seen that a combination of iron rich materials is enhanced the neutron shielding of capabilities of concrete. Also, the thicknesses of shield are reduced.

  14. Analysis of Shield Construction in Spherical Weathered Granite Development Area

    Science.gov (United States)

    Cao, Quan; Li, Peigang; Gong, Shuhua

    2018-01-01

    The distribution of spherical weathered bodies (commonly known as "boulder") in the granite development area directly affects the shield construction of urban rail transit engineering. This paper is based on the case of shield construction of granite globular development area in Southern China area, the parameter control in shield machine selection and shield advancing during the shield tunneling in this special geological environment is analyzed. And it is suggested that shield machine should be selected for shield construction of granite spherical weathered zone. Driving speed, cutter torque, shield machine thrust, the amount of penetration and the speed of the cutter head of shield machine should be controlled when driving the boulder formation, in order to achieve smooth excavation and reduce the disturbance to the formation.

  15. Shielding walls against ionizing radiation. Lead bricks

    International Nuclear Information System (INIS)

    1993-04-01

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

  16. Accelerator shielding experts meet at CERN

    CERN Multimedia

    CERN Bulletin

    2010-01-01

    Fifteen years after its first CERN edition, the Shielding Aspects of Accelerator, Targets and Irradiation Facility (SATIF) conference was held again here from 2-4 June. Now at its 10th edition, SATIF10 brought together experts from all over the world to discuss issues related to the shielding techniques. They set out the scene for an improved collaboration and discussed novel shielding solutions.   This was the most attended meeting of the series with more than 65 participants from 34 institutions and 14 countries. “We welcomed experts from many different laboratories around the world. We come from different contexts but we face similar problems. In this year’s session, among other things, we discussed ways for improving the effectiveness of calculations versus real data, as well as experimental solutions to investigate the damage that radiation produces on various materials and the electronics”, says Marco Silari, Chair of the conference and member of the DGS/RP gro...

  17. Slipforming of reinforced concrete shield building

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  18. First Wall, Blanket, Shield Engineering Technology Program

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1982-01-01

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

  19. Calculated shielding factors for selected European houses

    International Nuclear Information System (INIS)

    Hedemann Jensen, P.

    1984-12-01

    Shielding factors for gamma radiation from activity deposited on structures and ground surfaces have been calculated with the computer model DEPSHIELD for single-family and multi-storey buildings in France, United Kingdom and Denmark. For all three countries it was found that the shielding factors for single-family houses are approximately a factor of 2 - 10 higher that those for buildings with five or more storeys. Away from doors and windows the shielding factors for French, British, and Danish single-family houses are in the range 0.03 - 0.1, 0.06 - 0.4, and 0.07 - 0.3, respectively. The uncertainties of the calculations are discussed and DEPSHIELD-results are compared with other methods as well as with experimental results. (author)

  20. Radiation shielding performance of some concrete

    International Nuclear Information System (INIS)

    Akkurt, I.; Akyildirim, H.; Mavi, B.; Kilincarslan, S.; Basyigit, C.

    2007-01-01

    The energy consumption is increasing with the increased population of the world and thus new energy sources were discovered such as nuclear energy. Besides using nuclear energy, nuclear techniques are being used in a variety of fields such as medical hospital, industry, agriculture or military issue, the radiation protection becomes one of the important research fields. In radiation protection, the main rules are time, distance and shielding. The most effective radiation shields are materials which have a high density and high atomic number such as lead, tungsten which are expensive. Alternatively the concrete which produced using different aggregate can be used. The effectiveness of radiation shielding is frequently described in terms of the half value layer (HVL) or the tenth value layer (TVL). These are the thicknesses of an absorber that will reduce the radiation to half, and one tenth of its intensity respectively. In this study the radiation protection properties of different types of concrete will be discussed

  1. Magnetic shielding for MRI superconducting magnets

    International Nuclear Information System (INIS)

    Ishiyama, A.; Hirooka, H.

    1991-01-01

    This paper describes an optimal design of a highly homogeneous superconducting coil system with magnetic shielding for Magnetic Resonance Imaging (MRI). The presented optimal design method; which is originally proposed in our earlier papers, is a combination of the hybrid finite element and boundary element method for analysis of an axially symmetric nonlinear open boundary magnetic field problem, and the mathematical programming method for solving the corresponding optimization problem. In this paper, the multi-objective goal programming method and the nonlinear least squares method have been adopted. The optimal design results of 1.5- and 4.7-Tesla-magnet systems with different types of magnetic shielding for whole-body imaging are compared and the advantages of a combination of active and yoke shields are shown

  2. Search of the Griffiths shield region

    International Nuclear Information System (INIS)

    Wei, I.C.; Scott, R.L.

    1988-01-01

    The van der Walls equation of state for binary mixtures has been used to determine the location and shape of the Griffiths shield region (where three tricritical lines intersect). If one takes the geometric means for a 12 , the arithmetic mean for b 12 , and the configurational free energy as ideal, the center of the Griffiths shield region is found only when the ratio of molecular sizes is infinite. When the Flory equation for the configurational free energy for mixtures of chain molecules is substituted for the ideal form, the results appear to be somewhat different. However, for all the cases studied, with systems which approach geometric mean behavior one finds the shield region only when the ratio of molecular size is very large and when the internal pressure of the small molecule is very much greater than that of the long-chain molecule

  3. Heating profiles on ICRF antenna Faraday shields

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  4. Passive magnetic shielding in MRI-Linac systems

    Science.gov (United States)

    Whelan, Brendan; Kolling, Stefan; Oborn, Brad M.; Keall, Paul

    2018-04-01

    Passive magnetic shielding refers to the use of ferromagnetic materials to redirect magnetic field lines away from vulnerable regions. An application of particular interest to the medical physics community is shielding in MRI systems, especially integrated MRI-linear accelerator (MRI-Linac) systems. In these systems, the goal is not only to minimize the magnetic field in some volume, but also to minimize the impact of the shield on the magnetic fields within the imaging volume of the MRI scanner. In this work, finite element modelling was used to assess the shielding of a side coupled 6 MV linac and resultant heterogeneity induced within the 30 cm diameter of spherical volume (DSV) of a novel 1 Tesla split bore MRI magnet. A number of different shield parameters were investigated; distance between shield and magnet, shield shape, shield thickness, shield length, openings in the shield, number of concentric layers, spacing between each layer, and shield material. Both the in-line and perpendicular MRI-Linac configurations were studied. By modifying the shield shape around the linac from the starting design of an open ended cylinder, the shielding effect was boosted by approximately 70% whilst the impact on the magnet was simultaneously reduced by approximately 10%. Openings in the shield for the RF port and beam exit were substantial sources of field leakage; however it was demonstrated that shielding could be added around these openings to compensate for this leakage. Layering multiple concentric shield shells was highly effective in the perpendicular configuration, but less so for the in-line configuration. Cautious use of high permeability materials such as Mu-metal can greatly increase the shielding performance in some scenarios. In the perpendicular configuration, magnetic shielding was more effective and the impact on the magnet lower compared with the in-line configuration.

  5. Use of multiple radiographic techniques for nuclear shielding development

    Energy Technology Data Exchange (ETDEWEB)

    Day, S., E-mail: dayse@mcmaster.ca [McMaster Univ., McMaster Nuclear Reactor, Hamilton, Ontario (Canada)

    2016-01-15

    The McMaster Nuclear Reactor (MNR) is a medium-sized research reactor located on the campus of McMaster University in Hamilton, Ontario. The largest nuclear reactor on a Canadian university campus, MNR is an open-pool, light water research reactor, Equipped with both in-core and ex-core irradiation facilities, MNR serves research, education, and industrial needs while producing radioisotopes for medical applications. Presently the University is preparing for the installation of a positron beam line facility at one of the MNR beam ports. One of five such facilities worldwide, the McMaster Intense Positron Beam Facility (MIPBF) will provide orders of magnitude more positrons than a standard bench-top arrangement, making possible experiments not previously feasible. Funded by the Canadian Foundation for Innovation and the Ontario Research Foundation, the MIPBF project is a collaborative effort between positron groups at McMaster, Western, and York universities and researchers and staff at MNR, The project has involved design and fabrication of both the positron production and transport system and custom biological shielding. The MIPBF shielding fabrication is scheduled to be complete by fall 2015 followed shortly by installation and commissioning testing at MNR. Positron beam availability is planned for later in 2016. Applications include defect characterization and surface analysis of advanced engineering materials and fundamental science experiments on antimatter properties, The MIPBF will be the latest addition to the extensive array of materials-related facilities at McMaster University. (author)

  6. Use of multiple radiographic techniques for nuclear shielding development

    International Nuclear Information System (INIS)

    Day, S.

    2016-01-01

    The McMaster Nuclear Reactor (MNR) is a medium-sized research reactor located on the campus of McMaster University in Hamilton, Ontario. The largest nuclear reactor on a Canadian university campus, MNR is an open-pool, light water research reactor, Equipped with both in-core and ex-core irradiation facilities, MNR serves research, education, and industrial needs while producing radioisotopes for medical applications. Presently the University is preparing for the installation of a positron beam line facility at one of the MNR beam ports. One of five such facilities worldwide, the McMaster Intense Positron Beam Facility (MIPBF) will provide orders of magnitude more positrons than a standard bench-top arrangement, making possible experiments not previously feasible. Funded by the Canadian Foundation for Innovation and the Ontario Research Foundation, the MIPBF project is a collaborative effort between positron groups at McMaster, Western, and York universities and researchers and staff at MNR, The project has involved design and fabrication of both the positron production and transport system and custom biological shielding. The MIPBF shielding fabrication is scheduled to be complete by fall 2015 followed shortly by installation and commissioning testing at MNR. Positron beam availability is planned for later in 2016. Applications include defect characterization and surface analysis of advanced engineering materials and fundamental science experiments on antimatter properties, The MIPBF will be the latest addition to the extensive array of materials-related facilities at McMaster University. (author)

  7. Shielding in ungated field emitter arrays

    Energy Technology Data Exchange (ETDEWEB)

    Harris, J. R. [U.S. Navy Reserve, Navy Operational Support Center New Orleans, New Orleans, Louisiana 70143 (United States); Jensen, K. L. [Code 6854, Naval Research Laboratory, Washington, D.C. 20375 (United States); Shiffler, D. A. [Directed Energy Directorate, Air Force Research Laboratory, Albuquerque, New Mexico 87117 (United States); Petillo, J. J. [Leidos, Billerica, Massachusetts 01821 (United States)

    2015-05-18

    Cathodes consisting of arrays of high aspect ratio field emitters are of great interest as sources of electron beams for vacuum electronic devices. The desire for high currents and current densities drives the cathode designer towards a denser array, but for ungated emitters, denser arrays also lead to increased shielding, in which the field enhancement factor β of each emitter is reduced due to the presence of the other emitters in the array. To facilitate the study of these arrays, we have developed a method for modeling high aspect ratio emitters using tapered dipole line charges. This method can be used to investigate proximity effects from similar emitters an arbitrary distance away and is much less computationally demanding than competing simulation approaches. Here, we introduce this method and use it to study shielding as a function of array geometry. Emitters with aspect ratios of 10{sup 2}–10{sup 4} are modeled, and the shielding-induced reduction in β is considered as a function of tip-to-tip spacing for emitter pairs and for large arrays with triangular and square unit cells. Shielding is found to be negligible when the emitter spacing is greater than the emitter height for the two-emitter array, or about 2.5 times the emitter height in the large arrays, in agreement with previously published results. Because the onset of shielding occurs at virtually the same emitter spacing in the square and triangular arrays, the triangular array is preferred for its higher emitter density at a given emitter spacing. The primary contribution to shielding in large arrays is found to come from emitters within a distance of three times the unit cell spacing for both square and triangular arrays.

  8. Self-Shielding Of Transmission Lines

    Energy Technology Data Exchange (ETDEWEB)

    Christodoulou, Christos [Univ. of New Mexico, Albuquerque, NM (United States)

    2017-03-01

    The use of shielding to contend with noise or harmful EMI/EMR energy is not a new concept. An inevitable trade that must be made for shielding is physical space and weight. Space was often not as much of a painful design trade in older larger systems as they are in today’s smaller systems. Today we are packing in an exponentially growing number of functionality within the same or smaller volumes. As systems become smaller and space within systems become more restricted, the implementation of shielding becomes more problematic. Often, space that was used to design a more mechanically robust component must be used for shielding. As the system gets smaller and space is at more of a premium, the trades starts to result in defects, designs with inadequate margin in other performance areas, and designs that are sensitive to manufacturing variability. With these challenges in mind, it would be ideal to maximize attenuation of harmful fields as they inevitably couple onto transmission lines without the use of traditional shielding. Dr. Tom Van Doren proposed a design concept for transmission lines to a class of engineers while visiting New Mexico. This design concept works by maximizing Electric field (E) and Magnetic Field (H) field containment between operating transmission lines to achieve what he called “Self-Shielding”. By making the geometric centroid of the outgoing current coincident with the return current, maximum field containment is achieved. The reciprocal should be true as well, resulting in greater attenuation of incident fields. Figure’s 1(a)-1(b) are examples of designs where the current centroids are coincident. Coax cables are good examples of transmission lines with co-located centroids but they demonstrate excellent field attenuation for other reasons and can’t be used to test this design concept. Figure 1(b) is a flex circuit design that demonstrate the implementation of self-shielding vs a standard conductor layout.

  9. Experimental analysis of an MIM capacitor with a concave shield

    International Nuclear Information System (INIS)

    Liu Lintao; Yu Mingyan; Wang Jinxiang

    2009-01-01

    A novel shielding scheme is developed by inserting a concave shield between a metal-insulator-metal (MIM) capacitor and the silicon substrate. Chip measurements reveal that the concave shield improves the quality factor by 11% at 11.8 GHz and 14% at 18.8 GHz compared with an unshielded MIM capacitor. It also alleviates the effect on shunt capacitance between the bottom plate of the MIM capacitor and the shield layer. Moreover, because the concave shields simplify substrate modeling, a simple circuit model of the MIM capacitor with concave shield is presented for radio frequency applications.

  10. A study on the shielding element using Monte Carlo simulation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki Jeong [Dept. of Radiology, Konkuk University Medical Center, Seoul (Korea, Republic of); Shim, Jae Goo [Dept. of Radiologic Technology, Daegu Health College, Daegu (Korea, Republic of)

    2017-06-15

    In this research, we simulated the elementary star shielding ability using Monte Carlo simulation to apply medical radiation shielding sheet which can replace existing lead. In the selection of elements, mainly elements and metal elements having a large atomic number, which are known to have high shielding performance, recently, various composite materials have improved shielding performance, so that weight reduction, processability, In consideration of activity etc., 21 elements were selected. The simulation tools were utilized Monte Carlo method. As a result of simulating the shielding performance by each element, it was estimated that the shielding ratio is the highest at 98.82% and 98.44% for tungsten and gold.

  11. Safety guide data on radiation shielding in a reprocessing facility

    International Nuclear Information System (INIS)

    Sekiguchi, Noboru; Naito, Yoshitaka

    1986-04-01

    In a reprocessing facility, various radiation sources are handled and have many geometrical conditions. To aim drawing up a safety guidebook on radiation shielding in order to evaluate shielding safety in a reprocessing facility with high reliability and reasonableness, JAERI trusted investigation on safety evaluation techniques of radiation shielding in a reprocessing facility to Nuclear Safety Research Association. This report is the collection of investigation results, and describes concept of shielding safety design principle, radiation sources in reprocessing facility and estimation of its strength, techniques of shielding calculations, and definite examples of shielding calculation in reprocessing facility. (author)

  12. FLUKA shielding calculations for the FAIR project

    International Nuclear Information System (INIS)

    Fehrenbacher, Georg; Kozlova, Ekaterina; Radon, Torsten; Sokolov, Alexey

    2015-01-01

    FAIR is an international accelerator project being in construction at GSI Helmholtz center for heavy ion research in Darmstadt. The Monte Carlo program FLUKA is used to study radiation protection problems. The contribution deals with general application possibilities of FLUKA and for FAIR with respect the radiation protection planning. The necessity to simulate the radiation transport through shielding of several meters thickness and to determine the equivalent doses outside the shielding with sufficient accuracy is demonstrated using two examples under consideration of the variance reduction. Results of simulation calculations for activation estimation in accelerator facilities are presented.

  13. Thermophysical Properties of Heat Resistant Shielding Material

    International Nuclear Information System (INIS)

    Porter, W.D.

    2004-01-01

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

  14. Radiation shielding issues on the FMIT

    International Nuclear Information System (INIS)

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

    1981-05-01

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

  15. Shielding in electron beams used in radiotherapy

    International Nuclear Information System (INIS)

    Sentenac, Irenee.

    1979-01-01

    The interactions of electron beams with initial energies between 7 and 30 MeV have been studied in various materials including polystyrene, aluminium, copper and lead. The following experimental results have been found: estimation of measurement point displacement in a cylindrical chamber and of its variations with electron beam energy, empirical relations between the energy at the surface and the practical range of the electrons in various materials, an estimation of the relative ionisation due to the 'bremsstrahlung' measured behind different materials with beam complete shielding. Improvement of electron beam collimation is suggested after analysis of the dose distribution behind partial shielding [fr

  16. ICRF antenna Faraday shield plasma sheath model

    International Nuclear Information System (INIS)

    Whealton, J.H.; Ryan, P.M.; Raridon, R.J.

    1990-01-01

    A two-dimensional nonlinear formulation that explicitly considers the plasma edge near a Faraday shield in a self-consistent manner is used in the modeling of the ion motion for a Faraday shield concept and model suggested by Perkins. Two models are considered that may provide significant insight into the generation of impurities for ion cyclotron resonance heating (ICRH) antennas. In one of these models a significant sheath periodically forms next to the Faraday screen, with ion acoustic waves heating the ions in the plasma. (orig.)

  17. WASTE HANDLING BUILDING SHIELD WALL ANALYSIS

    International Nuclear Information System (INIS)

    Padula, D.

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    Kyllingstad, G.

    1977-01-01

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

  19. RadShield: semiautomated shielding design using a floor plan driven graphical user interface.

    Science.gov (United States)

    DeLorenzo, Matthew C; Wu, Dee H; Yang, Kai; Rutel, Isaac B

    2016-09-08

    The purpose of this study was to introduce and describe the development of RadShield, a Java-based graphical user interface (GUI), which provides a base design that uniquely performs thorough, spatially distributed calculations at many points and reports the maximum air-kerma rate and barrier thickness for each barrier pursuant to NCRP Report 147 methodology. Semiautomated shielding design calculations are validated by two approaches: a geometry-based approach and a manual approach. A series of geometry-based equations were derived giv-ing the maximum air-kerma rate magnitude and location through a first derivative root finding approach. The second approach consisted of comparing RadShield results with those found by manual shielding design by an American Board of Radiology (ABR)-certified medical physicist for two clinical room situations: two adjacent catheterization labs, and a radiographic and fluoroscopic (R&F) exam room. RadShield's efficacy in finding the maximum air-kerma rate was compared against the geometry-based approach and the overall shielding recommendations by RadShield were compared against the medical physicist's shielding results. Percentage errors between the geometry-based approach and RadShield's approach in finding the magnitude and location of the maximum air-kerma rate was within 0.00124% and 14 mm. RadShield's barrier thickness calculations were found to be within 0.156 mm lead (Pb) and 0.150 mm lead (Pb) for the adjacent catheteriza-tion labs and R&F room examples, respectively. However, within the R&F room example, differences in locating the most sensitive calculation point on the floor plan for one of the barriers was not considered in the medical physicist's calculation and was revealed by the RadShield calculations. RadShield is shown to accurately find the maximum values of air-kerma rate and barrier thickness using NCRP Report 147 methodology. Visual inspection alone of the 2D X-ray exam distribution by a medical physicist may not

  20. Usability evaluation through gonad shielding production of pediatric patients by gender and age rating

    International Nuclear Information System (INIS)

    Chui, Sung Hyun; Park, Jung Eun; Chun, Woon Kwan; Ju, Yong Jin; Yang, Nam Hee; Dong, Kyung Rae

    2015-01-01

    The gonad shielding is used to minimize the impact of the exposure to gonads when Abdomen, Plevis and Hip X-ray inspections are conducted on radiation impressionable pediatric patients. By the way, the gonad is palpable difficult and impossible to check visually because it's a sensitive area, so tests are conducted with the approximate location of shielding, thereby appearing problems of not shielding gonads accurately. Accordingly, this study produced shields by age and gender of pediatric patients and studied the method of positioning shields with ASIS as a reference point without palpable sensitive areas, and tried to evaluate its usability. The study surveyed 30 pediatric patients by gender and age, who came and got inspected in Department of Radiology, our hospital from February 2012 to January 2014 and obtained the value of tolerance by measuring the average size of the pelvis using the distance measurement function of Infinitt Piview with the images stored in the PACS and producing shields by age and gender of pediatric patients and specifying the areas at random for the comparative analysis of pre- and post-using. It calculated the technology statistics (mean±SD) with the value of tolerance measured the length using SPSS 12.0 statistical program. As for boys, differences in the tolerance range of pre- and post-using shields were 2.69 mm in case of 1 year old, 2.58 mm in 2 years, 2.37 mm in 3 years, 2.815 mm in 4-5 years, 2.043 mm in 7-10 years, and as for girls,1.92 mm in 1-2 years, 1.75 mm in 3-4 years, 2.52 mm in 5-6 years and 1.93 mm in 7-10. After analyzing the pre- and post-using shields for all of boys and girls, there were statistically significant differences (P<0.050). It is considered that we can minimize the exposure to gonads and get a better video for diagnosis in testing high biological impressionable pediatric, if we use shields correctly with ASIS as a reference point considering its shape and size by age and gender in Abdomen, Plevis

  1. Usability evaluation through gonad shielding production of pediatric patients by gender and age rating

    Energy Technology Data Exchange (ETDEWEB)

    Chui, Sung Hyun; Park, Jung Eun [Dept. of Radiology, Kyung Hee University Hospital at Gangdong, Seoul (Korea, Republic of); Chun, Woon Kwan; Ju, Yong Jin; Yang, Nam Hee [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of); Dong, Kyung Rae [Dept. of Radiological Technology, Gwangju Health University, Gwangju (Korea, Republic of)

    2015-05-15

    The gonad shielding is used to minimize the impact of the exposure to gonads when Abdomen, Plevis and Hip X-ray inspections are conducted on radiation impressionable pediatric patients. By the way, the gonad is palpable difficult and impossible to check visually because it's a sensitive area, so tests are conducted with the approximate location of shielding, thereby appearing problems of not shielding gonads accurately. Accordingly, this study produced shields by age and gender of pediatric patients and studied the method of positioning shields with ASIS as a reference point without palpable sensitive areas, and tried to evaluate its usability. The study surveyed 30 pediatric patients by gender and age, who came and got inspected in Department of Radiology, our hospital from February 2012 to January 2014 and obtained the value of tolerance by measuring the average size of the pelvis using the distance measurement function of Infinitt Piview with the images stored in the PACS and producing shields by age and gender of pediatric patients and specifying the areas at random for the comparative analysis of pre- and post-using. It calculated the technology statistics (mean±SD) with the value of tolerance measured the length using SPSS 12.0 statistical program. As for boys, differences in the tolerance range of pre- and post-using shields were 2.69 mm in case of 1 year old, 2.58 mm in 2 years, 2.37 mm in 3 years, 2.815 mm in 4-5 years, 2.043 mm in 7-10 years, and as for girls,1.92 mm in 1-2 years, 1.75 mm in 3-4 years, 2.52 mm in 5-6 years and 1.93 mm in 7-10. After analyzing the pre- and post-using shields for all of boys and girls, there were statistically significant differences (P<0.050). It is considered that we can minimize the exposure to gonads and get a better video for diagnosis in testing high biological impressionable pediatric, if we use shields correctly with ASIS as a reference point considering its shape and size by age and gender in Abdomen, Plevis

  2. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.H.

    1993-01-01

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

  3. The Tower Shielding Facility: Its glorious past

    International Nuclear Information System (INIS)

    Muckenthaler, F.J.

    1997-01-01

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

  4. MPACT Subgroup Self-Shielding Efficiency Improvements

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  5. Ford motor company NDE facility shielding design

    International Nuclear Information System (INIS)

    Metzger, R. L.; Van Riper, K. A.; Jones, M. H.

    2005-01-01

    Ford Motor Company proposed the construction of a large non-destructive evaluation laboratory for radiography of automotive power train components. The authors were commissioned to design the shielding and to survey the completed facility for compliance with radiation doses for occupationally and non-occupationally exposed personnel. The two X-ray sources are Varian Linatron 3000 accelerators operating at 9-11 MV. One performs computed tomography of automotive transmissions, while the other does real-time radiography of operating engines and transmissions. The shield thickness for the primary barrier and all secondary barriers were determined by point-kernel techniques. Point-kernel techniques did not work well for skyshine calculations and locations where multiple sources (e.g. tube head leakage and various scatter fields) impacted doses. Shielding for these areas was determined using transport calculations. A number of MCNP [Briesmeister, J. F. MCNPCA general Monte Carlo N-particle transport code version 4B. Los Alamos National Laboratory Manual (1997)] calculations focused on skyshine estimates and the office areas. Measurements on the operational facility confirmed the shielding calculations. (authors)

  6. The Tower Shielding Facility: Its glorious past

    Energy Technology Data Exchange (ETDEWEB)

    Muckenthaler, F.J.

    1997-05-07

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

  7. Design and analysis of ITER shield blanket

    Energy Technology Data Exchange (ETDEWEB)

    Ohmori, Junji; Hatano, Toshihisa; Ezato, Kouichiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1998-12-01

    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.)

  8. EFFECTS OF INTERFACES ON GAMMA SHIELDING

    Energy Technology Data Exchange (ETDEWEB)

    Clifford, C. E.

    1963-06-15

    A survey is presented of studies of interface effects in gamma shielding problems. These studies are grouped into three types of approaches, viz.: sources at the interface; radiation backscattered from the interface; and radiation transmitted through the interface. A bibliography of 54 references is included. Limitations on the applicability of the results are discussed. (T.F.H.)

  9. A support shield for a powered support

    Energy Technology Data Exchange (ETDEWEB)

    Korczynski, P; Rojicek, P

    1982-08-31

    The subject of this invention is a support shield for a powered support. This shield resists creep of the coal from the longwall face, including an inclined longwall. It involves an arrangement wherein the distributor hinge is separated into upper and lower sections. The lower section holds a second hinge and the end of the rod of the pressure cylinder, one end of which is connected to the roof support. An advantage of the support is that compared to existing supports, it is self-advancing and holds stopes with coal which slopes in its formation or has inrushes. To a certain degree it prevents the stope space from ejecting fine pieces of rock from the longwall roof using an inter-timber support between the powered support and the stope. The support shield is moved by a simple system of levers which makes it possible, using a pressurized cylinder, to transmit a significant force to the coal stope. In this case, when the coal inrushes exceed the designed capabilities of the shield, a wooden extension piece is placed between the support and the face, through which the support pressure is applied.

  10. Technical specifications: Tower Shielding Reactor II

    International Nuclear Information System (INIS)

    1979-02-01

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

  11. Subsurface Shielding Source Term Specification Calculation

    International Nuclear Information System (INIS)

    S.Su

    2001-01-01

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

  12. Ford Motor Company NDE facility shielding design.

    Science.gov (United States)

    Metzger, Robert L; Van Riper, Kenneth A; Jones, Martin H

    2005-01-01

    Ford Motor Company proposed the construction of a large non-destructive evaluation laboratory for radiography of automotive power train components. The authors were commissioned to design the shielding and to survey the completed facility for compliance with radiation doses for occupationally and non-occupationally exposed personnel. The two X-ray sources are Varian Linatron 3000 accelerators operating at 9-11 MV. One performs computed tomography of automotive transmissions, while the other does real-time radiography of operating engines and transmissions. The shield thickness for the primary barrier and all secondary barriers were determined by point-kernel techniques. Point-kernel techniques did not work well for skyshine calculations and locations where multiple sources (e.g. tube head leakage and various scatter fields) impacted doses. Shielding for these areas was determined using transport calculations. A number of MCNP [Briesmeister, J. F. MCNPCA general Monte Carlo N-particle transport code version 4B. Los Alamos National Laboratory Manual (1997)] calculations focused on skyshine estimates and the office areas. Measurements on the operational facility confirmed the shielding calculations.

  13. Angular assymetries in a shielded pair line

    International Nuclear Information System (INIS)

    Fanchiotti, H.; Garcia Canal, C.A.; Vucetich, H.

    1979-01-01

    The capacitance matrix and surface charge density distribution of an unbalanced pair line with both longitudinal and balanced excitations is presented. In particular the case in which the axes of the inner wires are not restricted to lie on a line through the axis of the shield is discussed

  14. Surface magnetic field measurement with magnetic shielding

    Czech Academy of Sciences Publication Activity Database

    Perevertov, Oleksiy

    2010-01-01

    Roč. 61, č. 7 (2010), 66-68 ISSN 1335-3632 Grant - others:AVČR(CZ) M100100906 Institutional research plan: CEZ:AV0Z10100520 Keywords : magnetic hysteresis * magnetic field measurement * magnetic shielding * extrapolation Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 0.270, year: 2010

  15. Glasses impregnated with lead for radiation shielding

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  16. Shielding Effectiveness Measurements using a Reverberation Chamber

    NARCIS (Netherlands)

    Leferink, Frank Bernardus Johannes; Bergsma, J.G.; Bergsma, Hans; van Etten, Wim

    2006-01-01

    Shielding effectiveness measurements have been performed using a reverberation chamber. The reverberation chamber methodology as we1l as the measurement setup is described and some results are given. Samples include glass reinforced plastic panels, aluminum panels with many holes, wire mesh, among

  17. AP1000 shield building: a constructability challenge

    International Nuclear Information System (INIS)

    Di Giuseppe, Giovanni; Bonanno, Domenico

    2010-01-01

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

  18. Methods for calculating radiation attenuation in shields

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-10-01

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

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

    International Nuclear Information System (INIS)

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

    1992-11-01

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

  20. Neutron shielding material based on colemanite and epoxy resin

    International Nuclear Information System (INIS)

    Okuno, K.

    2005-01-01

    In recent years, there has been a need for compact shielding design such as self-shielding of a PET cyclotron or up-gradation of radiation machinery in existing facilities. In these cases, high performance shielding materials are needed. Concrete or polyethylene have been used for a neutron shield. However, for compact shielding, they fall short in terms of performance or durability. Therefore, a new type of neutron shielding material based on epoxy resin and colemanite has been developed. Slab attenuation experiments up to 40 cm for the new shielding material were carried out using a 252 Cf neutron source. Measurement was carried out using a REM-counter, and compared with calculation. The results show that the shielding performance is better than concrete and polyethylene mixed with 10 wt% boron oxide. From the result, we confirmed that the performance of the new material is suitable for practical use. (authors)

  1. Superconductor shields test chamber from ambient magnetic fields

    Science.gov (United States)

    Hildebrandt, A. F.

    1965-01-01

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

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

    International Nuclear Information System (INIS)

    J. Stephens

    2006-01-01

    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

  3. Software Tools for Measuring and Calculating Electromagnetic Shielding Effectiveness

    National Research Council Canada - National Science Library

    Tesny, Neal

    2005-01-01

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

  4. Safety shield for vacuum/pressure-chamber windows

    Science.gov (United States)

    Shimansky, R. A.; Spencer, R.

    1980-01-01

    Optically-clear shatter-resistant safety shield protects workers from implosion and explosion of vacuum and pressure windows. Plastic shield is inexpensive and may be added to vacuum chambers, pressure chambers, and gas-filling systems.

  5. Multihelix rotating shield brachytherapy for cervical cancer

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

  6. Radiation shielding calculations for the vista spacecraft

    International Nuclear Information System (INIS)

    Sahin, Suemer; Sahin, Haci Mehmet; Acir, Adem

    2005-01-01

    The VISTA spacecraft design concept has been proposed for manned or heavy cargo deep space missions beyond earth orbit with inertial fusion energy propulsion. Rocket propulsion is provided by fusion power deposited in the inertial confined fuel pellet debris and with the help of a magnetic nozzle. The calculations for the radiation shielding have been revised under the fact that the highest jet efficiency of the vehicle could be attained only if the propelling plasma would have a narrow temperature distribution. The shield mass could be reduced from 600 tons in the original design to 62 tons. Natural and enriched lithium were the principle shielding materials. The allowable nuclear heating in the superconducting magnet coils (up to 5 mW/cm 3 ) is taken as the crucial criterion for dimensioning the radiation shielding structure of the spacecraft. The space craft mass is 6000 tons. Total peak nuclear power density in the coils is calculated as ∼5.0 mW/cm 3 for a fusion power output of 17 500 MW. The peak neutron heating density is ∼2.0 mW/cm 3 , and the peak γ-ray heating density is ∼3.0 mW/cm 3 (on different points) using natural lithium in the shielding. However, the volume averaged heat generation in the coils is much lower, namely 0.21, 0.71 and 0.92 mW/cm 3 for the neutron, γ-ray and total nuclear heating, respectively. The coil heating will be slightly lower if highly enriched 6 Li (90%) is used instead of natural lithium. Peak values are then calculated as 2.05, 2.15 and 4.2 mW/cm 3 for the neutron, γ-ray and total nuclear heating, respectively. The corresponding volume averaged heat generation in the coils became 0.19, 0.58 and 0.77 mW/cm 3

  7. Shielding consideration for the SSCL experimental halls

    International Nuclear Information System (INIS)

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

    1994-03-01

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

  8. Flux trapping and shielding in irreversible superconductors

    International Nuclear Information System (INIS)

    Frankel, D.J.

    1978-05-01

    Flux trappings and shielding experiments were carried out on Pb, Nb, Pb-Bi, Nb-Sn, and Nb-Ti samples of various shapes. Movable Hall probes were used to measure fields near or inside the samples as a function of position and of applied field. The trapping of transverse multipole magnetic fields in tubular samples was accomplished by cooling the samples in an applied field and then smoothly reducing the applied field to zero. Transverse quadrupole and sextupole fields with gradients of over 2000 G/cm were trapped with typical fidelity to the original impressed field of a few percent. Transverse dipole fields of up to 17 kG were also trapped with similar fidelity. Shielding experiments were carried out by cooling the samples in zero field and then gradually applying an external field. Flux trapping and shielding abilities were found to be limited by two factors, the pinning strength of the material, and the susceptibility of a sample to flux jumping. The trapping and shielding behavior of flat disk samples in axial fields and thin-walled tubular samples in transverse fields was modeled. The models, which were based on the concept of the critical state, allowed a connection to be made between the pinning strength and critical current level, and the flux trapping and shielding abilities. Adiabatic and dynamic stability theories are discussed and applied to the materials tested. Good qualitative, but limited quantitative agreement was obtained between the predictions of the theoretical stability criteria and the observed flux jumping behavior

  9. Verification of radiation exposure using lead shields

    International Nuclear Information System (INIS)

    Hayashida, Keiichi; Yamamoto, Kenyu; Azuma, Masami

    2016-01-01

    A long time use of radiation during IVR (intervention radiology) treatment leads up to an increased exposure on IVR operator. In order to prepare good environment for the operator to work without worry about exposure, the authors examined exposure reduction with the shields attached to the angiography instrument, i. e. lead curtain and lead glass. In this study, the lumber spine phantom was radiated using the instrument and the radiation leaked outside with and without shields was measured by the ionization chamber type survey meter. The meter was placed at the position which was considered to be that for IVR operator, and changed vertically 20-100 cm above X-ray focus by 10 cm interval. The radiation at the position of 80 cm above X-ray focus was maximum without shield and was hardly reduced with lead curtain. However, it was reduced with lead curtain plus lead glass. Similar reduction effects were observed at the position of 90-100 cm above X-ray focus. On the other hand, the radiation at the position of 70 cm above X-ray focus was not reduced with either shield, because that position corresponded to the gap between lead curtain and lead glass. The radiation at the position of 20-60 cm above X-ray focus was reduced with lead curtain, even if without lead glass. These results show that lead curtain and lead glass attached to the instrument can reduce the radiation exposure on IVR operator. Using these shields is considered to be one of good means for IVR operator to work safely. (author)

  10. Development of radiation shielding standards in the American Nuclear Society

    International Nuclear Information System (INIS)

    Trubey, D.K.

    1975-11-01

    The American Nuclear Society (ANS) is a standards-writing organization-member of the American National Standards Institute (ANSI). The ANS Standards Committee has a subcommittee denoted ANS-6, Shielding, whose charge is to establish standards in connection with radiation protection and shielding, to provide shielding information to other standards writing groups, and to prepare recommended sets of shielding data and test problems. This paper is a progress report of this subcommittee

  11. Important aspects of radiation shielding for fusion reactor tokamaks

    International Nuclear Information System (INIS)

    Abdou, M.A.

    1977-01-01

    Radiation shielding is a key subsystem in tokamak reactors. Design of this shield must evolve from economic and technological trade-off studies that account for the strong interrelations among the various components of the reactor system. These trade-offs are examined for the bulk shield on the inner side of the torus and for the special shields of major penetrations. Results derived are applicable for a large class of tokamak-type reactors

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

    International Nuclear Information System (INIS)

    1978-01-01

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

  13. Shielding of the child's head during x-ray studies

    International Nuclear Information System (INIS)

    Tolmach, Eh.U.

    1985-01-01

    Three devices for X-ray shielding of child's head are suggested; the first one is a protective attachment for shielding a child being in horizontal position on an X-ray table; the second one is a protective stand for shielding head and body at roentgenofraphy of upper extremities of a child sitting near the X-ray table; the third one is a prot ctive suspension for shielding the head of a child being in vertical position

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

    International Nuclear Information System (INIS)

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

    1975-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1975-11-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  17. 30 CFR 56.14213 - Ventilation and shielding for welding.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Ventilation and shielding for welding. 56.14213... Equipment Safety Practices and Operational Procedures § 56.14213 Ventilation and shielding for welding. (a) Welding operations shall be shielded when performed at locations where arc flash could be hazardous to...

  18. Calculation and design for SSRF's bulk shield

    Energy Technology Data Exchange (ETDEWEB)

    Fang, K.M. [Shanghai Institute of Applied Physics, Chinese Academy of Science (China)]. E-mail: fangkm@sinap.ac.cn; Xu, X.J. [Shanghai Institute of Applied Physics, Chinese Academy of Science (China); Cai, J.H. [Shanghai Institute of Applied Physics, Chinese Academy of Science (China)

    2006-12-15

    Shielding design objectives for the SSRF are chosen, assumptions for beam loss rates are given, the methods used on the APS by Moe are summarized and introduced to make calculation and design on bulk shield, the factor of skyshine is also considered, design thicknesses for SSRF's bulk shield are presented.

  19. System for detecting and processing abnormality in electromagnetic shielding

    International Nuclear Information System (INIS)

    Takahashi, T.; Nakamura, M.; Yabana, Y.; Ishikawa, T.; Nagata, K.

    1991-01-01

    The present invention relates to a system for detecting and processing an abnormality in electromagnetic shielding of an intelligent building which is constructed using an electromagnetic shielding material for the skeleton and openings such as windows and doorways so that the whole of the building is formed into an electromagnetic shielding structure. (author). 4 figs

  20. 21 CFR 886.4750 - Ophthalmic eye shield.

    Science.gov (United States)

    2010-04-01

    ...) MEDICAL DEVICES OPHTHALMIC DEVICES Surgical Devices § 886.4750 Ophthalmic eye shield. (a) Identification. An ophthalmic eye shield is a device that consists of a plastic or aluminum eye covering intended to... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Ophthalmic eye shield. 886.4750 Section 886.4750...

  1. 3-dimensional shielding design for a spallation neutron source facility in the high-intensity proton accelerator project

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-03-01

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

  2. Improving the shielding effectiveness of a board-level shield by bonding it with the waveguide-below-cutoff principle

    OpenAIRE

    Degraeve, Andy; Pissoort, Davy; Armstrong, Keith

    2015-01-01

    This paper discusses the shielding performance or shielding effectiveness of a board-level shield in function of its bonding method. Improved shielding performance at board-level in order to harden integrated circuits against unintentional and intentional electromagnetic interference, and this under harsh environmental conditions, is getting more and more important to achieve the desired levels of functional performance and operational reliability despite an ever more aggressive electromagnet...

  3. Multi-objective optimization design method of radiation shielding

    International Nuclear Information System (INIS)

    Yang Shouhai; Wang Weijin; Lu Daogang; Chen Yixue

    2012-01-01

    Due to the shielding design goals of diversification and uncertain process of many factors, it is necessary to develop an optimization design method of intelligent shielding by which the shielding scheme selection will be achieved automatically and the uncertainties of human impact will be reduced. For economical feasibility to achieve a radiation shielding design for automation, the multi-objective genetic algorithm optimization of screening code which combines the genetic algorithm and discrete-ordinate method was developed to minimize the costs, size, weight, and so on. This work has some practical significance for gaining the optimization design of shielding. (authors)

  4. The Fabrication of Internal Shielding using Provil and Cerrobend

    International Nuclear Information System (INIS)

    Kim Jong Hwa; Lee, Kang Hyun; Son, Jeong Hye

    1996-01-01

    The skin cancer is a highly curable disease which frequently occurs in the head and neck region exposed to the sun. When the eyelid is treated usually eye shield made of lead is used to protect the eyeball as a internal shield. For the same reason on internal shield should be used when the nose is treated when electron to protect the nasal mucosa. Our hospital made an internal shield for the treatment of the skin cancer on the nose using provil and cerrobend. The characteristics of the internal shield were examined.

  5. ANS shielding standards for light-water reactors

    International Nuclear Information System (INIS)

    Trubey, D.K.

    1982-01-01

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

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

    International Nuclear Information System (INIS)

    Dhang, Prosenjit; Verma, Rishi; Shyam, Anurag

    2015-01-01

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

  7. Determination of shielding parameters for different types of concretes by Monte Carlo methods

    International Nuclear Information System (INIS)

    Aminian, A.; Nematollahi, M. R.

    2007-01-01

    The chose of a suitable concrete composition for a biological reactor shield remain as a research target up to now. In the present study the attempts has been made to estimate the influence of the concrete aggregates on the shielding parameters for three type of ordinary, serpentine and steel magnetite concrete by Monte Carlo N-Particle (MCNP ) transport code. MCNP calculations have been performed in order to obtain the leakage of neutrons, photons and electrons from dry shield. Also the mass attenuation coefficients and the liner attenuation coefficient are estimated for neutron and photon in those energies in range of actual energy which exist out of pressure vessel of power reactor in the cavity for the investigated concretes. The concrete densities ranged from 2.3 to 5.11 g/cm 3 . These calculations were done in the condition of a typical commercial Pressurized Water Reactor (PWR). The results show that Steel-magnetite concrete, with high density (5.11 g/cm 3 ) and constituents of relatively high atomic number, is an effective shield for both photons and neutrons

  8. Radiation shielding analysis of medical cyclotron at Radiation Medicine Centre, Parel

    International Nuclear Information System (INIS)

    Gathibandhe, M.V.; Agrawal, R.A.; Utge, C.G.

    2003-01-01

    Full text: PET (Positron Emission Tomography) is a diagnostic method to obtain 3-D functional images of the distribution of radio-nuclides introduced in the human body as tracers for specific biological processes. Tracers are produced by bombardment of different target nuclides by protons and deuterons of high energy produced in the cyclotron. A Wipro-GE medical cyclotron was installed in the basement of RMC, Parel. Shielding around the cyclotron is provided in the form of borated concrete walls of required thickness to limit dose rates to design values as per AERB criteria. The roof of the cyclotron room is made of heavy concrete. Entry in to the room is through a maze. Shielding analysis for the cyclotron room has been carried out using computer code ANISN. The maze has been analyzed using code MCNP. Based on the analysis carried out additional shielding was recommended to meet the design requirements. The paper discusses the shielding analysis carried out for the cyclotron room and the maze. Dose rate estimated at various locations are highlighted

  9. Symbolic math for computation of radiation shielding

    International Nuclear Information System (INIS)

    Suman, Vitisha; Datta, D.; Sarkar, P.K.; Kushwaha, H.S.

    2010-01-01

    Radiation transport calculations for shielding studies in the field of accelerator technology often involve intensive numerical computations. Traditionally, radiation transport equation is solved using finite difference scheme or advanced finite element method with respect to specific initial and boundary conditions suitable for the geometry of the problem. All these computations need CPU intensive computer codes for accurate calculation of scalar and angular fluxes. Computation using symbols of the analytical expression representing the transport equation as objects is an enhanced numerical technique in which the computation is completely algorithm and data oriented. Algorithm on the basis of symbolic math architecture is developed using Symbolic math toolbox of MATLAB software. Present paper describes the symbolic math algorithm and its application as a case study in which shielding calculation of rectangular slab geometry is studied for a line source of specific activity. Study of application of symbolic math in this domain evolves a new paradigm compared to the existing computer code such as DORT. (author)

  10. Dismantling system of concrete thermal shielding walls

    International Nuclear Information System (INIS)

    Machida, Nobuhiro; Saiki, Yoshikuni; Ono, Yorimasa; Tokioka, Masatake; Ogino, Nobuyuki.

    1985-01-01

    Purpose: To enable safety and efficient dismantling of concrete thermal shielding walls in nuclear reactors. Method: Concrete thermal shielding walls are cut and dismantled into dismantled blocks by a plasma cutting tool while sealing the top opening of bioshielding structures. The dismantled blocks are gripped and conveyed. The cutting tool is remote-handled while monitoring on a television receiver. Slugs and dusts produced by cutting are removed to recover. Since the dismantling work is carried out while sealing the working circumstance and by the remote control of the cutting tool, the operators' safety can be secured. Further, since the thermal sealing walls are cut and dismantled into blocks, dismantling work can be done efficiently. (Moriyama, K.)

  11. Future directions in shielding methods and analysis

    International Nuclear Information System (INIS)

    Goldstein, H.

    1987-01-01

    Over the nearly half century history of shielding against reactor radiation, there has been a see-saw battle between theory and measurement. During that period the capability and accuracy of calculational methods have been enormously improved. The microscopic cross sections needed as input to the theoretical computations are now also known to adequate accuracy (with certain exceptions). Nonetheless, there remain substantial classes of shielding problems not yet accessible to satisfactory computational methods, particularly where three-dimensional geometries are involved. This paper discusses promising avenues to approach such problems, especially in the light of recent and expected advances in supercomputers. In particular, it seems that Monte Carlo methods should be much more advantageous in the new computer environment than they have been in the past

  12. In-beam background suppression shield

    DEFF Research Database (Denmark)

    Santoro, V.; Cai, Xiao Xiao; DiJulio, D. D.

    2015-01-01

    The long (3 ms) proton pulse of the European Spallation Source (ESS) gives rise to unique and potentially high backgrounds for the instrument suite. In such a source an instrument's capabilities will be limited by its Signal to Noise (S/N) ratio. The instruments with a direct view of the moderator......, which do not use a bender to help mitigate the fast neutron background, are the most challenging. For these beam lines we propose the innovative shielding of placing blocks of material directly into the guide system, which allow a minimum attenuation of the cold and thermal fluxes relative...... to the background suppression. This shielding configuration has been worked into a beam line model using Geant4. We study particularly the advantages of single crystal sapphire and silicon blocks....

  13. Thermoforming plastic in lead shield construction

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  14. Electrodynamic Dust Shield for Space Applications

    Science.gov (United States)

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

    2016-01-01

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

  15. Debye shielding in a nonextensive plasma

    International Nuclear Information System (INIS)

    Ait Gougam, Leila; Tribeche, Mouloud

    2011-01-01

    The phenomenon of Debye Shielding is revisited within the theoretical framework of the Tsallis statistical mechanics. The plasma consists of nonextensive electrons and ions. Both the effective Debye length λ D q and the fall-off of the electrostatic potential Φ are considered and a parameter study conducted. Owing to electron nonextensivity, the critical Mach number derived from the modified Bohm sheath criterion may become less than unity allowing therefore ions with speed less than ion-acoustic speed to enter the sheath from the main body of the plasma. Considering the wide relevance of collective processes, our analysis may be viewed as a first step toward a more comprehensive Debye shielding and electrostatic plasma sheath in nonequilibrium plasmas.

  16. Benchmarks for evaluation of shielding calculations

    International Nuclear Information System (INIS)

    Coelho, P.R.P.; Maiorino, J.R.

    1989-01-01

    The spatial-energy neutron distribution emerging from a laminated shielding (stainless, polyethylene and lead) were measured by a fast neutron spectrometer and some experimental results were compared with those calculated by a network of codes. The source neutrons incident in the shielding were 14 MeV neutrons from a H-3(d,n)He-4 reaction coming from a Van de Graaff accelerator. Experimentally was verified a good radial symmetry of neutron energy-spectrum, and also a moderation and attenuation effect for points located out of the central axis of symmetry. These results indicate that the experiment can be well modelated by R-Z geometry. A neutron-energy spectra calculated by DOT 3.5 was compared with the measured spectra, showing a good agreement in the shape and value of the spectra (12% for an integrated spectrum from 2 to 16 MeV). (author) [pt

  17. Electronically shielded solid state charged particle detector

    International Nuclear Information System (INIS)

    Balmer, D.K.; Haverty, T.W.; Nordin, C.W.; Tyree, W.H.

    1996-01-01

    An electronically shielded solid state charged particle detector system having enhanced radio frequency interference immunity includes a detector housing with a detector entrance opening for receiving the charged particles. A charged particle detector having an active surface is disposed within the housing. The active surface faces toward the detector entrance opening for providing electrical signals representative of the received charged particles when the received charged particles are applied to the active surface. A conductive layer is disposed upon the active surface. In a preferred embodiment, a nonconductive layer is disposed between the conductive layer and the active surface. The conductive layer is electrically coupled to the detector housing to provide a substantially continuous conductive electrical shield surrounding the active surface. The inner surface of the detector housing is supplemented with a radio frequency absorbing material such as ferrite. 1 fig

  18. Operating manual for the Tower Shielding Facility

    International Nuclear Information System (INIS)

    1985-12-01

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

  19. Methodology of shielding calculation for nuclear reactors

    International Nuclear Information System (INIS)

    Maiorino, J.R.; Mendonca, A.G.; Otto, A.C.; Yamaguchi, Mitsuo

    1982-01-01

    A methodology of calculation that coupling a serie of computer codes in a net that make the possibility to calculate the radiation, neutron and gamma transport, is described, for deep penetration problems, typical of nuclear reactor shielding. This net of calculation begining with the generation of constant multigroups, for neutrons and gamma, by the AMPX system, coupled to ENDF/B-IV data library, the transport calculation of these radiations by ANISN, DOT 3.5 and Morse computer codes, up to the calculation of absorbed doses and/or equivalents buy SPACETRAN code. As examples of the calculation method, results from benchmark n 0 6 of Shielding Benchmark Problems - ORNL - RSIC - 25, namely Neutron and Secondary Gamma Ray fluence transmitted through a Slab of Borated Polyethylene, are presented. (Author) [pt

  20. Grounding and shielding circuits and interference

    CERN Document Server

    Morrison, Ralph

    2016-01-01

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

  1. Radiation shielding activities at the OECD/Nuclear Energy Agency

    International Nuclear Information System (INIS)

    Sartori, Enrico; Vaz, Pedro

    2000-01-01

    The OECD Nuclear Energy Agency (NEA) has devoted considerable effort over the years to radiation shielding issues. The issues are addressed through international working groups. These activities are carried out in close co-ordination and co-operation with the Radiation Safety Information Computational Center (RSICC). The areas of work include: basic nuclear data activities in support of radiation shielding, computer codes, shipping cask shielding applications, reactor pressure vessel dosimetry, shielding experiments database. The method of work includes organising international code comparison exercises and benchmark studies. Training courses on radiation shielding computer codes are organised regularly including hands-on experience in modelling skills. The scope of the activity covers mainly reactor shields and spent fuel transportation packages, but also fusion neutronics and in particular shielding of accelerators and irradiation facilities. (author)

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

    International Nuclear Information System (INIS)

    1985-04-01

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

  3. New applications and developments in the neutron shielding

    Directory of Open Access Journals (Sweden)

    Uğur Fatma Aysun

    2017-01-01

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

  4. New applications and developments in the neutron shielding

    Science.gov (United States)

    Uğur, Fatma Aysun

    2017-09-01

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

  5. Shielded serpentine traveling wave tube deflection structure

    Science.gov (United States)

    Hudson, C.L.; Spector, J.

    1994-12-27

    A shielded serpentine slow wave deflection structure is disclosed having a serpentine signal conductor within a channel groove. The channel groove is formed by a serpentine channel in a trough plate and a ground plane. The serpentine signal conductor is supported at its ends by coaxial feed through connectors. A beam interaction trough intersects the channel groove to form a plurality of beam interaction regions wherein an electron beam may be deflected relative to the serpentine signal conductor. 4 figures.

  6. Grounding and shielding in the accelerator environment

    International Nuclear Information System (INIS)

    Kerns, Q.

    1991-01-01

    Everyday features of the accelerator environment include long cable runs, high power and low level equipment sharing building space, stray electromagnetic fields and ground voltage differences between the sending and receiving ends of an installation. This paper pictures some Fermilab installations chosen to highlight significant features and presents practices, test methods and equipment that have been helpful in achieving successful shielding. Throughout the report are numbered statements aimed at summarizing good practices and avoiding pitfalls

  7. Shielding analyses: the rabbit vs the turtle?

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1996-01-01

    This paper compares solutions using Monte Carlo and discrete- ordinates methods applied to two actual shielding situations in order to make some general observations concerning the efficiency and advantages/disadvantages of the two approaches. The discrete- ordinates solutions are performed using two-dimensional geometries, while the Monte Carlo approaches utilize three-dimensional geometries with both multigroup and point cross-section data

  8. Shielding and maintainability in an experimental tokamak

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  9. TORE SUPRA vacuum vessel and shield manufacturing

    International Nuclear Information System (INIS)

    Blateyron, J.; Lepez, R.

    1984-01-01

    TORE SUPRA vacuum vessel and vacuum chamber shield manufacturing in progress at Jeumont-Schneider consists of three main phases: - Detail engineering and manufacturing fixture construction; - Prototype section manufacturing and process preparation; - Construction of the 6 production modules. The welding techniques adopted, call for three special automatic processes: TIG, MIG and PLASMA welding which guarantee mechanical strength, vacuum tightness and absence of distortion. Production of the modules began July 1984. (author)

  10. Monte Carlo methods for shield design calculations

    International Nuclear Information System (INIS)

    Grimstone, M.J.

    1974-01-01

    A suite of Monte Carlo codes is being developed for use on a routine basis in commercial reactor shield design. The methods adopted for this purpose include the modular construction of codes, simplified geometries, automatic variance reduction techniques, continuous energy treatment of cross section data, and albedo methods for streaming. Descriptions are given of the implementation of these methods and of their use in practical calculations. 26 references. (U.S.)

  11. SINBAD: Shielding integral benchmark archive and database

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  12. Beta Bremsstrahlung dose in concrete shielding

    Energy Technology Data Exchange (ETDEWEB)

    Manjunatha, H.C., E-mail: manjunatha@rediffmail.com [Department of Physics, Government college for women, Kolar 563101, Karnataka (India); Chandrika, B.M. [Shravana, 592, Ist Cross, Behind St.Anne s School, PC Extension, Kolar 563101, Karnataka (India); Rudraswamy, B. [Department of Physics, Bangalore University, Bangalore 560056, Karnataka (India); Sankarshan, B.M. [Shravana, 592, Ist Cross, Behind St.Anne s School, PC Extension, Kolar 563101, Karnataka (India)

    2012-05-11

    In a nuclear reactor, beta nuclides are released during nuclear reactions. These betas interact with shielding concrete and produces external Bremsstrahlung (EB) radiation. To estimate Bremsstrahlung dose and shield efficiency in concrete, it is essential to know Bremsstrahlung distribution or spectra. The present work formulated a new method to evaluate the EB spectrum and hence Bremsstrahlung dose of beta nuclides ({sup 32}P, {sup 89}Sr, {sup 90}Sr-{sup 90}Y, {sup 90}Y, {sup 91}Y, {sup 208}Tl, {sup 210}Bi, {sup 234}Pa and {sup 40}K) in concrete. The Bremsstrahlung yield of these beta nuclides in concrete is also estimated. The Bremsstrahlung yield in concrete due to {sup 90}Sr-{sup 90}Y is higher than those of other given nuclides. This estimated spectrum is accurate because it is based on more accurate modified atomic number (Z{sub mod}) and Seltzer's data, where an electron-electron interaction is also included. Presented data in concrete provide a quick and convenient reference for radiation protection. The present methodology can be used to calculate the Bremsstrahlung dose in nuclear shielding materials. It can be quickly employed to give a first pass dose estimate prior to a more detailed experimental study. - Highlights: Black-Right-Pointing-Pointer Betas released in a nuclear reactor interact with shielding concrete and produces Bremsstrahlung. Black-Right-Pointing-Pointer The present work formulated a new method to evaluate the Bremsstrahlung spectrum and dose in concrete. Black-Right-Pointing-Pointer Presented data in concrete provide a quick and convenient reference for radiation protection.

  13. Shielding design for PWR in France

    International Nuclear Information System (INIS)

    Champion, G.; Charransol; Le Dieu de Ville, A.; Nimal, J.C.; Vergnaud, T.

    1983-05-01

    Shielding calculation scheme used in France for PWR is presented here for 900 MWe and 1300 MWe plants built by EDF the French utility giving electricity. Neutron dose rate at areas accessible by personnel during the reactor operation is calculated and compared with the measurements which were carried out in 900 MWe units up to now. Measurements on the first French 1300 MWe reactor are foreseen at the end of 1983

  14. Computer control of shielded cell operations

    International Nuclear Information System (INIS)

    Jeffords, W.R. III.

    1987-01-01

    This paper describes in detail a computer system to remotely control shielded cell operations. System hardware, software, and design criteria are discussed. We have designed a computer-controlled buret that provides a tenfold improvement over the buret currently in service. A computer also automatically controls cell analyses, calibrations, and maintenance. This system improves conditions for the operators by providing a safer, more efficient working environment and is expandable for future growth and development

  15. Radiation and shielding around beam absorbers

    International Nuclear Information System (INIS)

    Hurkmans, A.; Maas, R.

    1978-12-01

    During operational conditions it is anticipated that a fair amount of the total available beam power is dumped in either the slit system on one of the beam dumps. Thses beam absorbers therefore become strong radioactive sources. The radiation level due to the absorption of a 100 kW electron beam is estimated and the problem of residual activity is treated. Proposed shielding materials are discussed. (C.F.)

  16. The TPC shielding of the CAST experiment

    International Nuclear Information System (INIS)

    Ruz, J; Luzon, G; Beltran, B; Carmona, J M; Cebrian, S; Gomez, H; Irastorza, I G; Morales, J; Ortiz de Solorzano, A; RodrIguez, A; Villar, J A

    2006-01-01

    Sunset solar axions traversing the intense magnetic field of the CERN Axion Solar Telescope (CAST) experiment may be detected in a TPC detector, placed at one side of the magnet, as point-like X-rays signals. This signal could be masked, however, by the inhomogeneous radioactive background of materials and experimental site. Here we present the shielding built to reduce and homogenize the radioactive background levels of the TPC detector

  17. Final flush of the shielded cells melter

    International Nuclear Information System (INIS)

    Marshall, K.M.; Fellinger, T.L.; Harbour, J.R.

    1997-01-01

    A flush of the Savannah River Technology Center (SRTC) Shielded Cells melter was performed after the completion of a campaign to vitrify loaded crystalline silicotitanate (CST) ion exchange medium. The purpose of the flush was to lower levels of radioisotopes accumulated during the campaign and to lower the level of titanium dioxide present in the glass. This in turn would ready the melter for future campaigns involving the Defense Waste Processing Facility (DWPF)

  18. Thermal shielding device in LMFBR type reactors

    International Nuclear Information System (INIS)

    Nakamura, Hiroshi.

    1985-01-01

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

  19. GRAVITATIONAL FIELD SHIELDING AND SUPERNOVA EXPLOSIONS

    International Nuclear Information System (INIS)

    Zhang, T. X.

    2010-01-01

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

  20. Shielding design of RIKEN RI Beam Factory

    Energy Technology Data Exchange (ETDEWEB)

    Uwamino, Yoshitomo; Fukunishi, Nobuhisa [Institute of Physical and Chemical Research, Wako, Saitama (Japan); Oishi, Koji [Shimizu Corp., Tokyo (Japan)

    2000-03-01

    Construction of the RIKEN RI Beam Factory is started, and the phase 1 will be finished by the end of March 2003. Two ring cyclotrons including one superconducting machine and two Big RIPSs will be constructed in the phase 1. Heavy ions of proton to uranium will be accelerated up to 400 MeV/u (A<40) and 150 MeV/u for uranium at an intensity of 10{sup 13} pps. Neutron production by the 400 MeV/u {sup 20}Ne beam was measured at HIMAC of NIRS and it was used for the source term of the shielding calculations. The deep penetration of high-energy neutrons was calculated by using the ANISN code with the DLC-119/HILO86R group constants and also by using the HETC code. The ANISN results were modified by using the HETC results and the shielding experiment at ISIS, and they were fitted by a simple formula for practical use. High-energy neutron penetrations of slantwise injection and the reflection probabilities of iron slab were calculated with the HETC code, and these results were used for the estimation of the thickness for the iron local shielding of Big RIPSs. Induced radioactivity in the air, accelerator components and the building, and the skyshine effect were also estimated. (author)

  1. Shielding performance of the NET vacuum vessel

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  2. Topographic evidence for shield volcanism on Io

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  3. The effect of some organic and non-organic additions on the shielding and mechanical properties of radiation shielding concrete

    International Nuclear Information System (INIS)

    Kharita, M. H.; Yousef, S.; Al-Nassar, M.

    2011-04-01

    Few studies on the effect of some additives on the shielding properties of concrete have been carried out in this research. These studies included the effect of carbon powder, boron compounds, and waste polyethylene. The effect of water to cement ratio has been studied too. The research results showed that carbon powder and some boron compounds could be used to improve shielding concrete properties, and the possibility to add waste polyethylene in shielding concrete without effects on shielding properties. No significant effect for water to cement ratio on shielding properties of concrete. (author)

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

    Directory of Open Access Journals (Sweden)

    Vida Zaroushani

    2015-06-01

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

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

    International Nuclear Information System (INIS)

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

    1981-08-01

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

  6. SU-F-I-71: Fetal Protection During Fluoroscopy: To Shield Or Not to Shield?

    International Nuclear Information System (INIS)

    Joshi, S; Vanderhoek, M

    2016-01-01

    Purpose: Lead aprons are routinely used to shield the fetus from radiation during fluoroscopically guided interventions (FGI) involving pregnant patients. When placed in the primary beam, lead aprons often reduce image quality and increase fluoroscopic radiation output, which can adversely affect fetal dose. The purpose of this work is to identify an effective and practical method to reduce fetal dose without affecting image quality. Methods: A pregnant patient equivalent abdominal phantom is set on the table along with an image quality test object (CIRS model 903) representing patient anatomy of interest. An ion chamber is positioned at the x-ray beam entrance to the phantom, which is used to estimate the relative fetal dose. For three protective methods, image quality and fetal dose measurements are compared to baseline (no protection):1. Lead apron shielding the entire abdomen; 2. Lead apron shielding part of the abdomen, including the fetus; 3. Narrow collimation such that fetus is excluded from the primary beam. Results: With lead shielding the entire abdomen, the dose is reduced by 80% relative to baseline along with a drastic deterioration of image quality. With lead shielding only the fetus, the dose is reduced by 65% along with complete preservation of image quality, since the image quality test object is not shielded. However, narrow collimation results in 90% dose reduction and a slight improvement of image quality relative to baseline. Conclusion: The use of narrow collimation to protect the fetus during FGI is a simple and highly effective method that simultaneously reduces fetal dose and maintains sufficient image quality. Lead aprons are not as effective at fetal dose reduction, and if placed improperly, they can severely degrade image quality. Future work aims to investigate a wider variety of fluoroscopy systems to confirm these results across many different system geometries.

  7. SU-F-I-71: Fetal Protection During Fluoroscopy: To Shield Or Not to Shield?

    Energy Technology Data Exchange (ETDEWEB)

    Joshi, S; Vanderhoek, M [Henry Ford Health System, Detroit, MI (United States)

    2016-06-15

    Purpose: Lead aprons are routinely used to shield the fetus from radiation during fluoroscopically guided interventions (FGI) involving pregnant patients. When placed in the primary beam, lead aprons often reduce image quality and increase fluoroscopic radiation output, which can adversely affect fetal dose. The purpose of this work is to identify an effective and practical method to reduce fetal dose without affecting image quality. Methods: A pregnant patient equivalent abdominal phantom is set on the table along with an image quality test object (CIRS model 903) representing patient anatomy of interest. An ion chamber is positioned at the x-ray beam entrance to the phantom, which is used to estimate the relative fetal dose. For three protective methods, image quality and fetal dose measurements are compared to baseline (no protection):1. Lead apron shielding the entire abdomen; 2. Lead apron shielding part of the abdomen, including the fetus; 3. Narrow collimation such that fetus is excluded from the primary beam. Results: With lead shielding the entire abdomen, the dose is reduced by 80% relative to baseline along with a drastic deterioration of image quality. With lead shielding only the fetus, the dose is reduced by 65% along with complete preservation of image quality, since the image quality test object is not shielded. However, narrow collimation results in 90% dose reduction and a slight improvement of image quality relative to baseline. Conclusion: The use of narrow collimation to protect the fetus during FGI is a simple and highly effective method that simultaneously reduces fetal dose and maintains sufficient image quality. Lead aprons are not as effective at fetal dose reduction, and if placed improperly, they can severely degrade image quality. Future work aims to investigate a wider variety of fluoroscopy systems to confirm these results across many different system geometries.

  8. Female gonadal shielding with automatic exposure control increases radiation risks

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, Summer L.; Zhu, Xiaowei [Children' s Hospital of Philadelphia, Department of Radiology, Philadelphia, PA (United States); University of Pennsylvania, Perelman School of Medicine, Philadelphia, PA (United States); Magill, Dennise; Felice, Marc A. [University of Pennsylvania, Environmental Health and Radiation Safety, Philadelphia, PA (United States); Xiao, Rui [University of Pennsylvania, Department of Biostatistics and Epidemiology, Philadelphia, PA (United States); Ali, Sayed [Temple University Hospital, Department of Radiology, Philadelphia, PA (United States)

    2018-02-15

    Gonadal shielding remains common, but current estimates of gonadal radiation risk are lower than estimated risks to colon and stomach. A female gonadal shield may attenuate active automatic exposure control (AEC) sensors, resulting in increased dose to colon and stomach as well as to ovaries outside the shielded area. We assess changes in dose-area product (DAP) and absorbed organ dose when female gonadal shielding is used with AEC for pelvis radiography. We imaged adult and 5-year-old equivalent dosimetry phantoms using pelvis radiograph technique with AEC in the presence and absence of a female gonadal shield. We recorded DAP and mAs and measured organ absorbed dose at six internal sites using film dosimetry. Female gonadal shielding with AEC increased DAP 63% for the 5-year-old phantom and 147% for the adult phantom. Absorbed organ dose at unshielded locations of colon, stomach and ovaries increased 21-51% in the 5-year-old phantom and 17-100% in the adult phantom. Absorbed organ dose sampled under the shield decreased 67% in the 5-year-old phantom and 16% in the adult phantom. Female gonadal shielding combined with AEC during pelvic radiography increases absorbed dose to organs with greater radiation sensitivity and to unshielded ovaries. Difficulty in proper use of gonadal shields has been well described, and use of female gonadal shielding may be inadvisable given the risks of increasing radiation. (orig.)

  9. Evaluation of Neutron shielding efficiency of Metal hydrides

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Sang Hwan; Chae, San; Kim, Yong Soo [Hanyang University, Seoul (Korea, Republic of)

    2012-05-15

    Neutron shielding is achieved of interaction with material by moderation and absorption. Material that contains large amounts hydrogen atoms which are almost same neutron atomic weight is suited for fast neutron shielding material. Therefore, polymers containing high density hydrogen atom are being used for fast neutron shielding. On the other hand, composite materials containing high thermal neutron absorption cross section atom (Li, B, etc) are being used for thermal neutron shielding. However, these materials have low fast neutron absorption cross section. Therefore, these materials are not suited for fast neutron shielding. Hydrogen which has outstanding neutron energy reduction ability has very low thermal neutron absorption cross section, almost cannot be used for thermal neutron shielding. In this case, a large atomic number material (Pb, U, etc.) has been used. Thus, metal hydrides are considered as complement to concrete shielding material. Because metal hydrides contain high hydrogen density and elements with high atomic number. In this research neutron shielding performance and characteristic of nuclear about metal hydrides ((TiH{sub 2}, ZrH{sub 2}, HfH{sub 2}) is evaluated by experiment and MCNPX using {sup 252}Cf neutron source as purpose development shielding material to developed shielding material

  10. Female gonadal shielding with automatic exposure control increases radiation risks

    International Nuclear Information System (INIS)

    Kaplan, Summer L.; Zhu, Xiaowei; Magill, Dennise; Felice, Marc A.; Xiao, Rui; Ali, Sayed

    2018-01-01

    Gonadal shielding remains common, but current estimates of gonadal radiation risk are lower than estimated risks to colon and stomach. A female gonadal shield may attenuate active automatic exposure control (AEC) sensors, resulting in increased dose to colon and stomach as well as to ovaries outside the shielded area. We assess changes in dose-area product (DAP) and absorbed organ dose when female gonadal shielding is used with AEC for pelvis radiography. We imaged adult and 5-year-old equivalent dosimetry phantoms using pelvis radiograph technique with AEC in the presence and absence of a female gonadal shield. We recorded DAP and mAs and measured organ absorbed dose at six internal sites using film dosimetry. Female gonadal shielding with AEC increased DAP 63% for the 5-year-old phantom and 147% for the adult phantom. Absorbed organ dose at unshielded locations of colon, stomach and ovaries increased 21-51% in the 5-year-old phantom and 17-100% in the adult phantom. Absorbed organ dose sampled under the shield decreased 67% in the 5-year-old phantom and 16% in the adult phantom. Female gonadal shielding combined with AEC during pelvic radiography increases absorbed dose to organs with greater radiation sensitivity and to unshielded ovaries. Difficulty in proper use of gonadal shields has been well described, and use of female gonadal shielding may be inadvisable given the risks of increasing radiation. (orig.)

  11. Female gonadal shielding with automatic exposure control increases radiation risks.

    Science.gov (United States)

    Kaplan, Summer L; Magill, Dennise; Felice, Marc A; Xiao, Rui; Ali, Sayed; Zhu, Xiaowei

    2018-02-01

    Gonadal shielding remains common, but current estimates of gonadal radiation risk are lower than estimated risks to colon and stomach. A female gonadal shield may attenuate active automatic exposure control (AEC) sensors, resulting in increased dose to colon and stomach as well as to ovaries outside the shielded area. We assess changes in dose-area product (DAP) and absorbed organ dose when female gonadal shielding is used with AEC for pelvis radiography. We imaged adult and 5-year-old equivalent dosimetry phantoms using pelvis radiograph technique with AEC in the presence and absence of a female gonadal shield. We recorded DAP and mAs and measured organ absorbed dose at six internal sites using film dosimetry. Female gonadal shielding with AEC increased DAP 63% for the 5-year-old phantom and 147% for the adult phantom. Absorbed organ dose at unshielded locations of colon, stomach and ovaries increased 21-51% in the 5-year-old phantom and 17-100% in the adult phantom. Absorbed organ dose sampled under the shield decreased 67% in the 5-year-old phantom and 16% in the adult phantom. Female gonadal shielding combined with AEC during pelvic radiography increases absorbed dose to organs with greater radiation sensitivity and to unshielded ovaries. Difficulty in proper use of gonadal shields has been well described, and use of female gonadal shielding may be inadvisable given the risks of increasing radiation.

  12. Radiation attenuation by lead and nonlead materials used in radiation shielding garments

    International Nuclear Information System (INIS)

    McCaffrey, J. P.; Shen, H.; Downton, B.; Mainegra-Hing, E.

    2007-01-01

    The attenuating properties of several types of lead (Pb)-based and non-Pb radiation shielding materials were studied and a correlation was made of radiation attenuation, materials properties, calculated spectra and ambient dose equivalent. Utilizing the well-characterized x-ray and gamma ray beams at the National Research Council of Canada, air kerma measurements were used to compare a variety of commercial and pre-commercial radiation shielding materials over mean energy ranges from 39 to 205 keV. The EGSnrc Monte Carlo user code cavity.cpp was extended to provide computed spectra for a variety of elements that have been used as a replacement for Pb in radiation shielding garments. Computed air kerma values were compared with experimental values and with the SRS-30 catalogue of diagnostic spectra available through the Institute of Physics and Engineering in Medicine Report 78. In addition to garment materials, measurements also included pure Pb sheets, allowing direct comparisons to the common industry standards of 0.25 and 0.5 mm 'lead equivalent'. The parameter 'lead equivalent' is misleading, since photon attenuation properties for all materials (including Pb) vary significantly over the energy spectrum, with the largest variations occurring in the diagnostic imaging range. Furthermore, air kerma measurements are typically made to determine attenuation properties without reference to the measures of biological damage such as ambient dose equivalent, which also vary significantly with air kerma over the diagnostic imaging energy range. A single material or combination cannot provide optimum shielding for all energy ranges. However, appropriate choice of materials for a particular energy range can offer significantly improved shielding per unit mass over traditional Pb-based materials

  13. Method for limiting movement of a thermal shield for a nuclear reactor, and thermal shield displacement limiter therefor

    International Nuclear Information System (INIS)

    Meuschke, R.E.; Boyd, C.H.

    1989-01-01

    This patent describes a method of limiting the movement of a thermal shield of a nuclear reactor. It comprises: machining at least four (4) pockets in upper portions of a thermal shield circumferentially about a core barrel of a nuclear reactor to receive key-wave inserts; tapping bolt holes in the pockets of the thermal shield to receive bolts; positioning key-wave inserts into the pockets of the thermal shield to be bolted in place with the bolt holes; machining dowel holes at least partially through the positioned key-way inserts and the thermal shield to receive dowel pins; positioning dowel pins in the dowel holes in the key-way insert and thermal shield to tangentially restrain movement of the thermal shield relative to the core barrel; sliding limiter keys into the key-way inserts and bolting the limiter keys to the core barrel to tangentially restrain movement of the thermal shield relative and the core barrel while allowing radial and axial movement of the thermal shield relative to the core barrel; machining dowel holes through the limiter key and at least partially through the core barrel to receive dowel pins; positioning dowel pins in the dowel holes in the limiter key and core barrel to restrain tangential movement of the thermal shield relative to the core barrel of the nuclear reactor

  14. Measurements and calculations of neutron fluxes through a simulation of the CRBR upper axial shielding

    International Nuclear Information System (INIS)

    Maerker, R.E.; Muckenthaler, F.J.

    1976-01-01

    Measurements, using a 4-in. Bonner Ball, have been made of the neutron fluxes penetrating a simulation of CRBR upper axial biological shielding at the Tower Shielding Facility. The simulation consisted of a 45.7 cm thick slab of SS-304 followed by a series of sodium tanks having a total thickness of 457 cm followed by slabs of carbon steel up to 61.0 cm thick. Measurements were made behind the stainless steel, behind intermediate thicknesses of 152 cm, 305 cm, and 457 cm of sodium (with the stainless steel in place), and behind various thicknesses of the carbon steel following both 305 cm and 457 cm of sodium (also with the stainless steel in place). Calculated and measured data are presented and compared

  15. Developing light nano-composites with improved mechanical properties for neutron shielding

    Energy Technology Data Exchange (ETDEWEB)

    Jamali, F. [Shiraz Univ. of Medical Sciences (Iran, Islamic Republic of). School of Medicine; Mortazavi, S.M.J. [Shiraz Univ. of Medical Sciences (Iran, Islamic Republic of). Dept. of Medical Physics and Medical Engineering; Shiraz Univ. of Medical Sciences (Iran, Islamic Republic of). The Center for Radiological Research; Kardan, M. [Nuclear Science and Technology Institute, Tehran (Iran, Islamic Republic of). Radiation Application School; Mosleh-Shirazi, M.A. [Shiraz Univ. of Medical Sciences (Iran, Islamic Republic of). Radiotherapy Dept.; Sina, S. [Shiraz Univ. of Medical Sciences (Iran, Islamic Republic of). Radiation Research Center; Rahpeyma, J.

    2017-12-15

    Although radiation exposures in manned space missions are normally below the limits recommended to NASA by NCRP, in long-duration deep space exploratory missions astronauts may receive relatively high doses of ionizing radiation. Novel light polyethylene-based composites can be considered as effective radiation shields in space explorations. However, normally these composites cannot provide desired mechanical properties. Over the past several years our laboratories have focused on developing efficient methods for both physical and biological protection of the crew in long term space missions. In this study carbon nanotubes and either nano-sized or micro-sized boron carbide (B{sub 4}C) fillers were incorporated into the continuous phase of low density polyethylene (LDPE). In the next phase, the mechanical characteristics of the composites as well as their neutron attenuation properties were studied. Findings of this study indicated enhanced mechanical properties accompanied by an enhanced shielding efficiency for neutrons at some specific weight fraction of the fillers.

  16. Shielding benchmark tests of JENDL-3

    International Nuclear Information System (INIS)

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

    1994-03-01

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

  17. Evaluation of root-knot nematode disease control and plant growth promotion potential of biofertilizer Ning shield on Trichosanthes kirilowii in the field.

    Science.gov (United States)

    Jiang, Chun-Hao; Xie, Ping; Li, Ke; Xie, Yue-Sheng; Chen, Liu-Jun; Wang, Jin-Suo; Xu, Quan; Guo, Jian-Hua

    Biofertilizer Ning shield was composed of different strains of plant growth promotion bacteria. In this study, the plant growth promotion and root-knot nematode disease control potential on Trichosanthes kirilowii in the field were evaluated. The application of Ning shield significantly reduced the diseases severity caused by Meloidogyne incognita, the biocontrol efficacy could reached up to 51.08%. Ning shield could also promote the growth of T. kirilowii in the field by increasing seedling emergence, height and the root weight. The results showed that the Ning shield could enhance the production yield up to 36.26%. Ning shield could also promote the plant growth by increasing the contents of available nitrogen, phosphorus, potassium and organic matter, and increasing the contents of leaf chlorophyll and carotenoid pigment. Moreover, Ning shield could efficiently enhance the medicinal compositions of Trichosanthes, referring to the polysaccharides and trichosanthin. Therefore, Ning shield is a promising biofertilizer, which can offer beneficial effects to T. kirilowii growers, including the plant growth promotion, the biological control of root-knot disease and enhancement of the yield and the medicinal quality. Copyright © 2017 Sociedade Brasileira de Microbiologia. Published by Elsevier Editora Ltda. All rights reserved.

  18. Shielding evaluation of moving bed onion irradiator by radiometry

    International Nuclear Information System (INIS)

    Venkataramani, R.; Sangurdekar, P.R.; Sarangapani, R.; Raipurkar, D.R.; Mehta, S.K.; Shastri, S.P.; Patil, K.B.; Bongirwar, D.R.

    1994-01-01

    A moving bed onion irradiator made from m.s. cladded lead slab shields designed to hold 20 kCi of 60 Co source was evaluated by radiometry with an 8 Ci 60 Co source from CRC-2 radiography camera. Some shielding losses in the irradiator noted by radiometry could be visualized by a thermocole model of the complex shielding assembly. These were rectified by appropriate lead filling. Significant shielding losses noted at cladding layer positions of slabs were attributed to lack of interlocking features in the slabs. These had to be rectified by provision of 3 TVL of additional all round shielding supplemented by local shielding at some positions. (author). 1 fig., 1 tab

  19. Radiation shielding properties of barite coated fabric by computer programme

    Energy Technology Data Exchange (ETDEWEB)

    Akarslan, F.; Molla, T. [Suleyman Demirel University, Engineering Fac. Textile Dep., Isparta (Turkey); Üncü, I. S. [Suleyman Demirel University, Technological Fac. Electrical-Electronic Eng. Dep., Isparta (Turkey); Kılıncarslan, S., E-mail: seref@tef.sdu.edu.tr [Suleyman Demirel University, Engineering Fac. Civil Eng. Dep., Isparta (Turkey); Akkurt, I. [Suleyman Demirel University, Art and Science Fac., Physics Dep., Isparta (Turkey)

    2015-03-30

    With the development of technology radiation started to be used in variety of different fields. As the radiation is hazardous for human health, it is important to keep radiation dose as low as possible. This is done mainly using shielding materials. Barite is one of the important materials in this purpose. As the barite is not used directly it can be used in some other materials such as fabric. For this purposes barite has been coated on fabric in order to improve radiation shielding properties of fabric. Determination of radiation shielding properties of coated fabric has been done by using computer program written C# language. With this program the images obtained from digital Rontgen films is used to determine radiation shielding properties in terms of image processing numerical values. Those values define radiation shielding and in this way the coated barite effect on radiation shielding properties of fabric has been obtained.

  20. Magnetic shielding structure optimization design for wireless power transmission coil

    Science.gov (United States)

    Dai, Zhongyu; Wang, Junhua; Long, Mengjiao; Huang, Hong; Sun, Mingui

    2017-09-01

    In order to improve the performance of the wireless power transmission (WPT) system, a novel design scheme with magnetic shielding structure on the WPT coil is presented in this paper. This new type of shielding structure has great advantages on magnetic flux leakage reduction and magnetic field concentration. On the basis of theoretical calculation of coil magnetic flux linkage and characteristic analysis as well as practical application feasibility consideration, a complete magnetic shielding structure was designed and the whole design procedure was represented in detail. The simulation results show that the coil with the designed shielding structure has the maximum energy transmission efficiency. Compared with the traditional shielding structure, the weight of the new design is significantly decreased by about 41%. Finally, according to the designed shielding structure, the corresponding experiment platform is built to verify the correctness and superiority of the proposed scheme.

  1. Neutron shielding performance of water-extended polyester

    International Nuclear Information System (INIS)

    Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M.; Vega Carrillo, H.R.; Hernandez-Davila, V.M.; Gallego, E.; Lorente, A.

    2006-01-01

    A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester (WEP) was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (MCNP code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a 252 Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

  2. Neutron shielding performance of water-extended polyester

    International Nuclear Information System (INIS)

    Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M.; Vega Carrillo, H.R.; Gallegoc, E.; Lorentec, A.; Hernandez-Davila, V.M.

    2006-01-01

    A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (M.C.N.P. code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a 252 Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

  3. Development of neutron shielding material using metathesis-polymer matrix

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Yoshinori E-mail: ysakurai@rri.kyoto-u.ac.jp; Sasaki, Akira; Kobayashi, Tooru

    2004-04-21

    A neutron shielding material using a metathesis-polymer matrix, which is a thermosetting resin, was developed. This shielding material has characteristics that can be controlled for different mixing ratios of neutron absorbers and for formation in the laboratory. Additionally, the elastic modulus can be changed at the hardening process, from a flexible elastoma to a mechanically tough solid. Experiments were performed at the Kyoto University Research Reactor in order to determine the important characteristics of this metathesis-polymer shielding material, such as neutron shielding performance, secondary gamma-ray generation and activation. The metathesis-polymer shielding material was shown to be practical and as effective as the other available shielding materials, which mainly consist of thermoplastic resin.

  4. Development and application of high performance liquid shielding materials

    International Nuclear Information System (INIS)

    Miura, Toshimasa; Omata, Sadao; Otano, Naoteru; Hirao, Yoshihiro; Kanai, Yasuji

    1998-01-01

    Development of liquid shielding material with good performance for neutron and γ-ray was investigated. Lead, hydrogen and boron were selected as the elements of shielding materials which were made by the ultraviolet curing method. Good performance shielding materials with about 1 mm width to neutron and gamma ray were produced by mixing lead, boron compound and ultraviolet curing monomer with many hydrogens. The shielding performance was the same as a concrete with two times width. The activation was very small such as 1/10 6 -1/10 8 of the standard concrete. The weight and the external appearance did not charged from room temperature to 100degC. Polyfunctional monomer had good thermal resistance. This shielding material was applied to double bending cylindrical duct and annulus ring duct. The results proved the shielding materials developed had good performance. (S.Y.)

  5. Neutron shielding performance of water-extended polyester

    Energy Technology Data Exchange (ETDEWEB)

    Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M. [Zacatecas Univ. Autonoma, Nuclear Studies (Mexico); Vega Carrillo, H.R.; Hernandez-Davila, V.M. [Zacatecas Univ. Autonoma, Electric Engineering Academic Units (Mexico); Gallego, E.; Lorente, A. [Madrid Univ. Politecnica, cNuclear Engineering Department (Mexico)

    2006-07-01

    A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester (WEP) was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (MCNP code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a {sup 252}Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

  6. RADSHI: shielding calculation program for different geometries sources

    International Nuclear Information System (INIS)

    Gelen, A.; Alvarez, I.; Lopez, H.; Manso, M.

    1996-01-01

    A computer code written in pascal language for IBM/Pc is described. The program calculates the optimum thickness of slab shield for different geometries sources. The Point Kernel Method is employed, which enables the obtention of the ionizing radiation flux density. The calculation takes into account the possibility of self-absorption in the source. The air kerma rate for gamma radiation is determined, and with the concept of attenuation length through the equivalent attenuation length the shield is obtained. The scattering and the exponential attenuation inside the shield material is considered in the program. The shield materials can be: concrete, water, iron or lead. It also calculates the shield for point isotropic neutron source, using as shield materials paraffin, concrete or water. (authors). 13 refs

  7. Shield design development of nuclear propulsion merchant ship

    International Nuclear Information System (INIS)

    Tanaka, Yoshihisa

    1975-01-01

    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.)

  8. Shielding reproductive organs of orthopaedic patients during pelvic radiography.

    Science.gov (United States)

    Wainwright, A. M.

    2000-01-01

    The use of gonadal shielding has been advocated for patients undergoing pelvic radiography before and during the reproductive years. The aim of this study is to look at the adequacy of gonadal shielding used in a district general hospital for young patients having pelvic radiographs. A total of 200 radiographs were reviewed of 49 patients below the age of 45 years. Full coverage was achieved in only 36% of cases. Amongst females, only 22% received adequate shielding. None of the patients in their reproductive years (16-45 years) had gonad shields. The reasons for inadequate coverage were, in order of frequency: (i) no shielding was used; (ii) malposition of the shielding device; and (iii) the use of inappropriately shaped or sized devices. Suggestions for improvement are proposed. Images Figure 3 Figure 4 PMID:11041029

  9. Estimation of temperature distribution in a reactor shield

    International Nuclear Information System (INIS)

    Agarwal, R.A.; Goverdhan, P.; Gupta, S.K.

    1989-01-01

    Shielding is provided in a nuclear reactor to absorb the radiations emanating from the core. The energy of these radiations appear in the form of heat. Concrete which is commonly used as a shielding material in nuclear power plants must be able to withstand the temperatures and temperature gradients appearing in the shield due to this heat. High temperatures lead to dehydration of the concrete and in turn reduce the shielding effectiveness of the material. Adequate cooling needs to be provided in these shields in order to limit the maximum temperature. This paper describes a method to estimate steady state and transient temperature distribution in reactor shields. The results due to loss of coolant in the coolant tubes have been studied and presented in the paper. (author). 5 figs

  10. Interstitial rotating shield brachytherapy for prostate cancer

    International Nuclear Information System (INIS)

    Adams, Quentin E.; Xu, Jinghzu; Breitbach, Elizabeth K.; Li, Xing; Rockey, William R.; Kim, Yusung; Wu, Xiaodong; Flynn, Ryan T.; Enger, Shirin A.

    2014-01-01

    Purpose: To present a novel needle, catheter, and radiation source system for interstitial rotating shield brachytherapy (I-RSBT) of the prostate. I-RSBT is a promising technique for reducing urethra, rectum, and bladder dose relative to conventional interstitial high-dose-rate brachytherapy (HDR-BT). Methods: A wire-mounted 62 GBq 153 Gd source is proposed with an encapsulated diameter of 0.59 mm, active diameter of 0.44 mm, and active length of 10 mm. A concept model I-RSBT needle/catheter pair was constructed using concentric 50 and 75 μm thick nickel-titanium alloy (nitinol) tubes. The needle is 16-gauge (1.651 mm) in outer diameter and the catheter contains a 535 μm thick platinum shield. I-RSBT and conventional HDR-BT treatment plans for a prostate cancer patient were generated based on Monte Carlo dose calculations. In order to minimize urethral dose, urethral dose gradient volumes within 0–5 mm of the urethra surface were allowed to receive doses less than the prescribed dose of 100%. Results: The platinum shield reduced the dose rate on the shielded side of the source at 1 cm off-axis to 6.4% of the dose rate on the unshielded side. For the case considered, for the same minimum dose to the hottest 98% of the clinical target volume (D 98% ), I-RSBT reduced urethral D 0.1cc below that of conventional HDR-BT by 29%, 33%, 38%, and 44% for urethral dose gradient volumes within 0, 1, 3, and 5 mm of the urethra surface, respectively. Percentages are expressed relative to the prescription dose of 100%. For the case considered, for the same urethral dose gradient volumes, rectum D 1cc was reduced by 7%, 6%, 6%, and 6%, respectively, and bladder D 1cc was reduced by 4%, 5%, 5%, and 6%, respectively. Treatment time to deliver 20 Gy with I-RSBT was 154 min with ten 62 GBq 153 Gd sources. Conclusions: For the case considered, the proposed 153 Gd-based I-RSBT system has the potential to lower the urethral dose relative to HDR-BT by 29%–44% if the clinician allows

  11. An analytical theory of transmission line shielding

    International Nuclear Information System (INIS)

    Pettersson, Per

    1993-01-01

    The classical electrogeometric model of shielding failure flashovers on transmission lines is investigated by analytical methods. Most of the basic elements that has appeared in the literature on the subject have been incorporated and put into a comprehensive model. These elements are: tower top geometry, structure height above ground, line insulation, lateral slope of ground, probability distribution of lightning currents, ratio of striking distances to ground wire and earth relative to conductor, and probability distribution of lightning leader approach angle to ground. Departing from a basic idealistic case, the sensitivity of the model to variations in these parameters is studied. Numerical examples are given. 8 refs, 8 figs, 1 tab

  12. Shielding and maintainability in an experimental tokamak

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  13. Shielding design at Fermilab: Calculations and measurements

    International Nuclear Information System (INIS)

    Cossairt, J.D.

    1986-11-01

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

  14. Intrinsically disordered proteins as molecular shields.

    Science.gov (United States)

    Chakrabortee, Sohini; Tripathi, Rashmi; Watson, Matthew; Schierle, Gabriele S Kaminski; Kurniawan, Davy P; Kaminski, Clemens F; Wise, Michael J; Tunnacliffe, Alan

    2012-01-01

    The broad family of LEA proteins are intrinsically disordered proteins (IDPs) with several potential roles in desiccation tolerance, or anhydrobiosis, one of which is to limit desiccation-induced aggregation of cellular proteins. We show here that this activity, termed molecular shield function, is distinct from that of a classical molecular chaperone, such as HSP70 - while HSP70 reduces aggregation of citrate synthase (CS) on heating, two LEA proteins, a nematode group 3 protein, AavLEA1, and a plant group 1 protein, Em, do not; conversely, the LEA proteins reduce CS aggregation on desiccation, while HSP70 lacks this ability. There are also differences in interaction with client proteins - HSP70 can be co-immunoprecipitated with a polyglutamine-containing client, consistent with tight complex formation, whereas the LEA proteins can not, although a loose interaction is observed by Förster resonance energy transfer. In a further exploration of molecular shield function, we demonstrate that synthetic polysaccharides, like LEA proteins, are able to reduce desiccation-induced aggregation of a water-soluble proteome, consistent with a steric interference model of anti-aggregation activity. If molecular shields operate by reducing intermolecular cohesion rates, they should not protect against intramolecular protein damage. This was tested using the monomeric red fluorescent protein, mCherry, which does not undergo aggregation on drying, but the absorbance and emission spectra of its intrinsic fluorophore are dramatically reduced, indicative of intramolecular conformational changes. As expected, these changes are not prevented by AavLEA1, except for a slight protection at high molar ratios, and an AavLEA1-mCherry fusion protein is damaged to the same extent as mCherry alone. A recent hypothesis proposed that proteomes from desiccation-tolerant species contain a higher degree of disorder than intolerant examples, and that this might provide greater intrinsic stability

  15. Neutron streaming studies along JET shielding penetrations

    Science.gov (United States)

    Stamatelatos, Ion E.; Vasilopoulou, Theodora; Batistoni, Paola; Obryk, Barbara; Popovichev, Sergey; Naish, Jonathan

    2017-09-01

    Neutronic benchmark experiments are carried out at JET aiming to assess the neutronic codes and data used in ITER analysis. Among other activities, experiments are performed in order to validate neutron streaming simulations along long penetrations in the JET shielding configuration. In this work, neutron streaming calculations along the JET personnel entrance maze are presented. Simulations were performed using the MCNP code for Deuterium-Deuterium and Deuterium- Tritium plasma sources. The results of the simulations were compared against experimental data obtained using thermoluminescence detectors and activation foils.

  16. Cerrobend shielding stents for buccal carcinoma patients

    Directory of Open Access Journals (Sweden)

    Karma Yangchen

    2016-01-01

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

  17. Equivalent-spherical-shield neutron dose calculations

    International Nuclear Information System (INIS)

    Russell, G.J.; Robinson, H.

    1988-01-01

    Neutron doses through 162-cm-thick spherical shields were calculated to be 1090 and 448 mrem/h for regular and magnetite concrete, respectively. These results bracket the measured data, for reinforced regular concrete, of /approximately/600 mrem/h. The calculated fraction of the high-energy (>20 MeV) dose component also bracketed the experimental data. The measured and calculated doses were for a graphite beam stop bombarded with 100 nA of 800-MeV protons. 6 refs., 2 figs., 1 tab

  18. Space Shuttle Orbiter AFT heat shield seal

    Science.gov (United States)

    Walkover, L. J.

    1979-01-01

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

  19. ORNL fusion reactor shielding integral experiments

    International Nuclear Information System (INIS)

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

    1980-01-01

    Integral experiments that measure the neutron and gamma-ray energy spectra resulting from the attenuation of approx. 14 MeV T(D,n) 4 He reaction neutrons in laminated slabs of stainless steel type 304, borated polyethylene, and a tungsten alloy (Hevimet) and from neutrons streaming through a 30-cm-diameter iron duct (L/D = 3) imbedded in a concrete shield have been performed. The facility, the NE-213 liquid scintillator detector system, and the experimental techniques used to obtain the measured data are described. The two-dimensional discrete ordinates radiation transport codes, calculational models, and nuclear data used in the analysis of the experiments are reviewed

  20. Nuclear data requirements for fusion reactor shielding

    International Nuclear Information System (INIS)

    Abdou, M.A.

    1979-01-01

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

  1. Re-evaluation of the shielding adequacy of the brachytherapy treatment room at Korle-Bu teaching hospital, Ghana

    International Nuclear Information System (INIS)

    Arwui, C. C.

    2009-06-01

    Staff and the general public's safety during the operation of the 137 Cs brachytherapy unit at the Korle Bu teaching hospital depends on the adequacy of the shielding of the facility. Shielding design of the brachytherapy unit at the hospital was based on postulated workload and postulated occupancy factors to critical locations at the facility where the public and staff may occupy. This facility has been in existence for the past twelve (12) years and has accumulated operational workload data which differs from the postulated one. A study was carried out to re-evaluate the integrity of the biological shielding of the 137 Cs brachytherapy unit. This study analyzed the accumulated workload data and used the information to perform shielding calculations to verify the adequacy of the biological shielding thicknesses to provide sufficient protection of staff and the public. Dose rate calculations were verified by measurements with calibrated dose rate meters. This provided the basis for determining the current state of protection and safety for staff and the general public. The results show that despite the variation in actual and postulated workloads, the dose rates were below the reference values of 0.5μSv/h for public areas and 7.5μSv/h for controlled areas. It was confirmed that the present shielding thickness of 535 mm can accommodate a high dose rate (HDR) 192 Ir source with activity in the range 370 - 570 GBq with an operational workload of 30 patients per week and an average treatment time of 10 minutes.

  2. PEP radiation shielding tests in SLAC A Beam

    International Nuclear Information System (INIS)

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

    1977-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Alberdi, R.; Gonzalez, E.

    2001-07-01

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

  4. Shielding measurements and augmentation for high power operations of FBTR

    International Nuclear Information System (INIS)

    Jose, M.T.; Baskar, S.; Viswanathan, S.; Balasundar, S.; Subramanian, V.; Ravi, T.; Sundaram, V.M.; Raghunath, V.M.; Varadarajan, S.; Jena, A.K.

    1996-01-01

    Fast breeder test reactor (FBTR) at Kalpakkam is a 40 MWt loop type fast reactor with sodium coolant. Since criticality in 1985, radiation surveys were carried out at all accessible locations at different power levels to find out the hot spots and evaluate the shielding adequacy. This paper gives the details of findings of these measurements, consequent changes in shielding, and the present status of dose profile after the augmentation of shielding. (author). 1 ref., 1 fig., 1 tab

  5. Shielding design method for LMFBR validation on the Phenix factor

    International Nuclear Information System (INIS)

    Cabrillat, J.C.; Crouzet, J.; Misrakis, J.; Salvatores, M.; Rado, V.; Palmiotti, G.

    1983-05-01

    Shielding design methods, developed at CEA for shielding calculations find a global validation by the means of Phenix power reactor (250 MWe) measurements. Particularly, the secondary sodium activation of pool type LMFBR such as Super Phenix (1200 MWe) which is subject to strict safety limitation is well calculated by the adapted scheme, i.e. a two dimension transport calculation of shielding coupled to a Monte-Carlo calculation of secondary sodium activation

  6. Study and application of construction technology of shielding concrete

    International Nuclear Information System (INIS)

    Wu Chongming; Ding Dexin; Chen Liangzhu; Zhao Jingfa; Li Shilong

    2008-01-01

    Process and techniques such as mixing,transportation and pouring have been studied. The construction technology for the shielding concrete with different densities has been summarized. The technology for the common concrete is quite different from that of shielding concrete, especially when its density is more than 4000 kg/m3. Application and practices have shown that different construction technologies shall be used for shielding concretes with different densities, and thus to ensure its uniformity and construction quality. (authors)

  7. Methods and procedures for shielding analyses for the SNS

    International Nuclear Information System (INIS)

    Popova, I.; Ferguson, F.; Gallmeier, F.X.; Iverson, E.; Lu, Wei

    2011-01-01

    In order to provide radiologically safe Spallation Neutron Source operation, shielding analyses are performed according to Oak Ridge National Laboratory internal regulations and to comply with the Code of Federal Regulations. An overview of on-going shielding work for the accelerator facility and neutrons beam lines, methods used for the analyses, and associated procedures and regulations are presented. Methods used to perform shielding analyses are described as well. (author)

  8. Shielding effectiveness research of window panes in microwave frequency range

    OpenAIRE

    Bilotas, Evaldas

    2016-01-01

    The purpose of this work is to investigate microwave shielding effectiveness (SE) of modern window panes. In addition, it will be made sure of what is the main mechanism behind the electromagnetic shielding by investigating three different glasses reflection coefficient. In order to achieve these goals, shielding effectiveness of window panes and their components will be measured in semi-anechoic and anechoic chambers. Furthermore, these measurements will be done in near and far field conditi...

  9. Development of special radiation shielding concretes using natural local materials and evaluation of their shielding characteristics

    International Nuclear Information System (INIS)

    Kharita, M. H.; Takeyeddin, M.; Al-Nassri, M.; Yousef, S.

    2008-01-01

    Concrete is one of the most important materials used for radiation shielding in facilities containing radioactive sources and radiation generating machines. The concrete shielding properties may vary depending on the composite of the concrete. Aggregates is the largest constituent (about 70-80% of the total weight of normal concrete). The aim of this work is to develop special concrete with good shielding properties for gamma and neutrons, using natural local materials. For this reason two types of typical concrete widely used in Syria (in Damascus and Aleppo) and four other types of concrete, using aggregates from different regions, have been prepared. The shielding properties of these six types were studied for gamma ray (from Cs-137 and Co-60 sources)and for neutrons (from am-Be source). A reduction of about 10% in the HVL was obtained for the concrete from Damascus in comparison with that from Aleppo, for both neutrons and gammas. One of the other four types of concrete (from Rajo site, mostly Hematite), was found to further reduce the HVL by about 10% for both neutrons and gamma rays.(author)

  10. Using natural local materials for developing special radiation shielding concretes, and deduction of its shielding characteristics

    International Nuclear Information System (INIS)

    Kharita, M. H.; Takeyeddin, M.; Al-Nassar, M.; Yousef, S.

    2006-06-01

    Concrete is considered as the most important material to be used for radiation shielding in facilities contain radioactive sources and radiation generating machines. The concrete shielding properties may vary depending on the construction of the concrete, which is highly relative to the composing aggregates i.e. aggregates consist about 70 - 80% of the total weight of normal concrete. In this project tow types of concrete used in Syria (in Damascus and Aleppo) had been studied and their shielding properties were defined for gamma ray from Cs-137 and Co-60 sources, and for neutrons from Am-Be source. About 10% reduction in HVL was found in the comparison between the tow concrete types for both neutrons and gammas. Some other types of concrete were studied using aggregates from different regions in Syria, to improve the shielding properties of concrete, and another 10% of reduction was achieved in comparison with Damascene concrete (20% in comparison with the concrete from Aleppo) for both neutrons and gamma rays. (author)

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  12. Graphene shield enhanced photocathodes and methods for making the same

    Science.gov (United States)

    Moody, Nathan Andrew

    2014-09-02

    Disclosed are graphene shield enhanced photocathodes, such as high QE photocathodes. In certain embodiments, a monolayer graphene shield membrane ruggedizes a high quantum efficiency photoemission electron source by protecting a photosensitive film of the photocathode, extending operational lifetime and simplifying its integration in practical electron sources. In certain embodiments of the disclosed graphene shield enhanced photocathodes, the graphene serves as a transparent shield that does not inhibit photon or electron transmission but isolates the photosensitive film of the photocathode from reactive gas species, preventing contamination and yielding longer lifetime.

  13. Polyolefin-Nanocrystal Composites for Radiation Shielding, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — EIC Laboratories Inc. is proposing a lightweight multifunctional polymer/nanoparticle composite for radiation shielding during long-duration lunar missions. Isolated...

  14. Radiation protection and shielding standards for the 1980s

    International Nuclear Information System (INIS)

    Trubey, D.K.

    1982-01-01

    The American Nuclear Society (ANS) is a standards-writing organization member of the American National Standards Institute (ANSI). The ANS Standards Committee has a subcommittee denoted ANS-6, Radiation Protection and Shielding, whose charge is to develop standards for radiation protection and shield design, to provide shielding information to other standards-writing groups, and to develop standard reference shielding data and test problems. This paper is a progress report of this subcommittee. Significant progress has been made since the last comprehensive report to the Society

  15. Design of radiation shields in nuclear reactor core

    International Nuclear Information System (INIS)

    Mousavi Shirazi, A.; Daneshvar, Sh.; Aghanajafi, C.; Jahanfarnia, Gh.; Rahgoshay, M.

    2008-01-01

    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 O ppenheim Electrical Network m ethod, 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

  16. Electromagnetic shielding effectiveness of 3D printed polymer composites

    Science.gov (United States)

    Viskadourakis, Z.; Vasilopoulos, K. C.; Economou, E. N.; Soukoulis, C. M.; Kenanakis, G.

    2017-12-01

    We report on preliminary results regarding the electromagnetic shielding effectiveness of various 3D printed polymeric composite structures. All studied samples were fabricated using 3D printing technology, following the fused deposition modeling approach, using commercially available filaments as starting materials. The electromagnetic shielding performance of the fabricated 3D samples was investigated in the so called C-band of the electromagnetic spectrum (3.5-7.0 GHz), which is typically used for long-distance radio telecommunications. We provide evidence that 3D printing technology can be effectively utilized to prepare operational shields, making them promising candidates for electromagnetic shielding applications for electronic devices.

  17. Neutron shielding for a {sup 252} Cf source

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M. [Unidades Academicas de Estudios Nucleares e Ingenieria Electrica, Universidad Autonoma de Zacatecas, C. Cipres 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Eduardo Gallego, Alfredo Lorente [Depto. de Ingenieria Nuclear, ETS Ingenieros Industriales, Universidad Politecnica de Madrid, C. Jose Gutierrez Abascal 2, 28006 Madrid (Spain)]. e-mail: fermineutron@yahoo.com

    2006-07-01

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

  18. Effect of metal shielding on a wireless power transfer system

    Directory of Open Access Journals (Sweden)

    Jiacheng Li

    2017-05-01

    Full Text Available In this paper, the effect of non-ferromagnetic metal shielding (NFMS material on the resonator of wireless power transfer (WPT is studied by modeling, simulation and experimental analysis. And, the effect of NFMS material on the power transfer efficiency (PTE of WPT systems is investigated by circuit model. Meanwhile, the effect of ferromagnetic metal shielding material on the PTE of WPT systems is analyzed through simulation. A double layer metal shield structure is designed. Experimental results demonstrate that by applying the novel double layer metal shielding method, the system PTE increases significantly while the electromagnetic field of WPT systems declines dramatically.

  19. Effect of metal shielding on a wireless power transfer system

    Science.gov (United States)

    Li, Jiacheng; Huang, Xueliang; Chen, Chen; Tan, Linlin; Wang, Wei; Guo, Jinpeng

    2017-05-01

    In this paper, the effect of non-ferromagnetic metal shielding (NFMS) material on the resonator of wireless power transfer (WPT) is studied by modeling, simulation and experimental analysis. And, the effect of NFMS material on the power transfer efficiency (PTE) of WPT systems is investigated by circuit model. Meanwhile, the effect of ferromagnetic metal shielding material on the PTE of WPT systems is analyzed through simulation. A double layer metal shield structure is designed. Experimental results demonstrate that by applying the novel double layer metal shielding method, the system PTE increases significantly while the electromagnetic field of WPT systems declines dramatically.

  20. Gravity Scaling of a Power Reactor Water Shield

    International Nuclear Information System (INIS)

    Reid, Robert S.; Pearson, J. Boise

    2008-01-01

    Water based reactor shielding is being considered as an affordable option for potential use on initial lunar surface reactor power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxillary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2006). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa n . These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined