WorldWideScience

Sample records for gamma radiation shieldings

  1. Gamma radiation shielding analysis of lead-flyash concretes

    International Nuclear Information System (INIS)

    Singh, Kanwaldeep; Singh, Sukhpal; Dhaliwal, A.S.; Singh, Gurmel

    2015-01-01

    Six samples of lead-flyash concrete were prepared with lead as an admixture and by varying flyash content – 0%, 20%, 30%, 40%, 50% and 60% (by weight) by replacing cement and keeping constant w/c ratio. Different gamma radiation interaction parameters used for radiation shielding design were computed theoretically and measured experimentally at 662 keV, 1173 keV and 1332 keV gamma radiation energy using narrow transmission geometry. The obtained results were compared with ordinary-flyash concretes. The radiation exposure rate of gamma radiation sources used was determined with and without lead-flyash concretes. - Highlights: • Concrete samples with lead as admixture were casted with flyash replacing 0%, 20%, 30%, 40%, 50% and 60% of cement content (by weight). • Gamma radiation shielding parameters of concretes for different gamma ray sources were measured. • The attenuation results of lead-flyash concretes were compared with the results of ordinary flyash concretes

  2. Study of local Agregate for Gamma radiation concrete shield

    International Nuclear Information System (INIS)

    Tochrul-Binowo; Endro-Kismolo; Darsono

    1996-01-01

    Investigation on the composition of gamma radiation concrete shield made of local barite, manganese fine and coarse aggregates from Kulon Progo, Yogyakarta has been done. The purpose of the research was to find out the quality of these local material for an aggregate of gamma radiation concrete shield. The research was done where each mineral was used as coarse aggregate and the fine aggregate from Kulon Progo was used as fine basic aggregate. Firstly a normal concrete was made by mixing cement, fine aggregate, coarse aggregate and water at a weight ratio of cement: fine aggregate: coarse: water 1: 2.304: 3.456: 0.58. The gamma radiation absorption capacity of the concrete tested by using Cs-137 as source standard. The same method was done on barite concrete at the weight ratio of cement: fine aggregate: barite aggregate: water 1: 2.303: 3.456: 0.58 and manganese concrete at the weight ratio of cement: fine aggregate: manganese aggregate: and water 1: 1.896: 2.844: 0.58. The result of the study showed that the gamma radiation absorption capacity of barite aggregate was greater than that of normal concrete and manganese concrete. The coefficient linear attenuation (for 6.0 cm thickness) of each concrete were μ barite concrete = 0.23071 cm -1 , μ manganese concrete = 0.08401 cm -1 and μ normal concrete = 0.1669 cm -1

  3. Attenuation of gamma radiation in concrete shields

    International Nuclear Information System (INIS)

    Azevedo e Souza, A.C. de.

    1978-12-01

    The attenuation characteristics of γ radiation in concrete layers considering their mechanical resistence and densities were determined. A 137 Cs source was used in a 'good geometry' arrangement to eliminate the effects of the buildup factor. The ordinary and the heavy concrete were irradiated and for the latter it was used as additives iron ore and Fe 2 O 3 pellets in various grain sizes. The detection system consisted of a 2' x 2' NaI (Tl) crystal coupled to a photomultiplier tube and the associated electronic equipment. FORTRAN programs were used for determining the absorption coefficients and the attenuation factors. These programs calculate photopeak areas eliminating all contributions due to Compton effect and background. (Author) [pt

  4. Determination of gamma radiation shielding characteristics of some tropical woods

    International Nuclear Information System (INIS)

    Aigbosuria, E. F.

    2011-01-01

    This study compares the shielding characteristics of twenty-two tropical woods by using gamma scintillation detection method. Woods sourced are Anogeisus Leiocarpus(Ayin), Nesogordonia Papverifera(Oro), Entandrophragma Microphyllum(Anunje), Brachystagia Eurycoma(Ako), Cassia Alata(Asunrun), Afzelia Africana(Apa-Igbo), Khaya Grandifoliala(Gedu), Piptadenistrum Africana(Agbonyin), Nanclea Diderrehii(Opepe), Khaya Ivorensis(Oganwo), Chlorophora Excelsa(Iroko), Masonia altissima(Odogi), Entandrophragma Angolense(Ijebo), Altium Sativum(Ayo), Albizia Zygia(Ayunre), Terminalia Superba(Afara), Cordia Millenii(Omo), Melania(Melania), Pycnanthus Angolensis(Akomu), Triplochitons Scleroxylon(Arere), Pine(Pine), Ceiba Pentradra(Araba). The intensities of the emergent radiation were measured, when each of these woods were placed between a scintillation detector and a standard radioactive source. Analysis of result obtained shows an appreciable evidence of radiation attenuation due to the changes in the chemical composition of the woods and the dependence of the attenuation coefficient on energy and densities of these woods. The descending order of attenuation coefficient determined are; Ayin, Oro, Anuje, Ako, Asunrun, Apa-igbo Gedu, Agbonyin, Opepe Oganwo, Iroko Odogi , Ayo, Ayunre, Afara, Omo, Melania, Akomu, Arere, Pine, Araba. For a constant energy of 0.101MeV, the attenuation coefficient are 0.190cm -1 , 0.165cm -1 , 0.163cm -1 , 0.156cm -1 , 0.149cm -1 , 0.143cm -1 , 0.133cm -1 , 0.132cm -1 , 0.127cm -1 , 0.124cm -1 , 0.085cm -1 , 0.123cm -1 , 0.122cm -1 , 0.113cm -1 , 0.101cm -1 , 0.088cm -1 , 0.087cm -1 , 0.086cm -1 , 0.082cm -1 respectively. The wood in descending order of dependence of attenuation coefficient on density are: Ayin,Oro, Anunje,Ako,Asunrun,Apa-Igbo, Gedu, Agbonyin, Opepe, Oganwo, Iroko, Odogi, Ijebo, Ayo, Ayunre, Afara, Omo, Melania, Akomu, Arere, Pine and Araba. The half value layer shows the thickness at various energy regions.

  5. Radiation shield

    International Nuclear Information System (INIS)

    Hosoya, Yasuaki

    1993-01-01

    In the present invention, the thickness of the radiation shields is minimized to save the quantity of shields thereby utilizing spaces in a facility effectively. That is, the radiation shields of the present invention comprise first and second shields forming stepwise gaps. They are disposed between a high dose region and a low dose region. The first and second shields have a feature in that the thickness thereof can be set to a size capable of shielding the gaps in accordance with the strength of the radiation source to be shielded. With such a constitution, the thickness of the shields of the radiation processing facility can be minimized. Accordingly, the quantity of the shields can be greatly saved. Spaces in the facility can be utilized effectively. (I.S.)

  6. Gamma radiation shielding and health physics characteristics of diaspore-flyash concretes.

    Science.gov (United States)

    Singh, Kanwaldeep; Singh, Sukhpal; Singh, S P; Mudahar, Gurmel S; Dhaliwal, A S

    2015-06-01

    Different gamma radiation interaction parameters has been measured experimentally for the prepared diaspore-flyash concretes at 59.54, 662, 1173 and 1332 keV using narrow-beam transmission geometry and results are found to be in good agreement with theoretical values computed with a computer programme, WinXCom. The radiation exposure rate and absorbed dose rate for the gamma radiation with and without shielding of diaspore-flyash concretes have been determined using linear attenuation results. The results show that on average, there is reduction of 95%, 53% and 40% in dose rate for gamma sources (241)Am, (137)Cs and (60)Co, respectively with diaspore-flyash concretes as shielding material. Other health physics parameters namely equivalent dose, effective dose, gamma flux and energy fluence rate have also been determined.

  7. Study of gamma radiation shielding properties of ZnO−TeO2 glasses

    Indian Academy of Sciences (India)

    2017-07-25

    Jul 25, 2017 ... The addition of ZnO to tellurite increases the glass formation and thermal stability [10,11]. Gamma radiation shielding properties of different com- pounds were evaluated using parameters such as mass atten- uation coefficient (μm), half value layer (HVL), effective atomic number (Zeff), electron density (Nel) ...

  8. Analysis of coupled neutron-gamma radiations, applied to shieldings in multigroup albedo method

    International Nuclear Information System (INIS)

    Dunley, Leonardo Souza

    2002-01-01

    The principal mathematical tools frequently available for calculations in Nuclear Engineering, including coupled neutron-gamma radiations shielding problems, involve the full Transport Theory or the Monte Carlo techniques. The Multigroup Albedo Method applied to shieldings is characterized by following the radiations through distinct layers of materials, allowing the determination of the neutron and gamma fractions reflected from, transmitted through and absorbed in the irradiated media when a neutronic stream hits the first layer of material, independently of flux calculations. Then, the method is a complementary tool of great didactic value due to its clarity and simplicity in solving neutron and/or gamma shielding problems. The outstanding results achieved in previous works motivated the elaboration and the development of this study that is presented in this dissertation. The radiation balance resulting from the incidence of a neutronic stream into a shielding composed by 'm' non-multiplying slab layers for neutrons was determined by the Albedo method, considering 'n' energy groups for neutrons and 'g' energy groups for gammas. It was taken into account there is no upscattering of neutrons and gammas. However, it was considered that neutrons from any energy groups are able to produce gammas of all energy groups. The ANISN code, for an angular quadrature order S 2 , was used as a standard for comparison of the results obtained by the Albedo method. So, it was necessary to choose an identical system configuration, both for ANISN and Albedo methods. This configuration was six neutron energy groups and eight gamma energy groups, using three slab layers (iron aluminum - manganese). The excellent results expressed in comparative tables show great agreement between the values determined by the deterministic code adopted as standard and, the values determined by the computational program created using the Albedo method and the algorithm developed for coupled neutron-gamma

  9. Radiation safety aspects during nondestructive testing of reactor shielding components by gamma radiometry

    International Nuclear Information System (INIS)

    Viswanathan, S.; Jose, M.T.; Venkatraman, B.

    2016-01-01

    In nuclear facilities, effective shielding of radioactive components and structures are essential to ensure radiation protection to operating personnel. The shield structures are made of lead, steel and concrete with varying thickness of up to 1200 mm. It needs to be verified for shielding integrity, presence of voids, blowholes and defects to avoid exposure to workers and to public at large. Radiometry using gamma source serves as excellent tool for non-destructive examination of such structures and components. Gamma sources of high activity up to 50 Curies (gamma camera type) depending on the thickness of component have to be used. During the testing exposure to the operating personnel needs to be minimized, this requires certain safety procedures to be followed. This paper focuses the methodology to be adapted by means of selection of source, effective training of personnel, compliance with safety requirements and maintenance of source devices

  10. Gamma radiation shielding properties of poly (ethylene glycol) - barium oxide composite

    International Nuclear Information System (INIS)

    Hussain, R.; Haq, Z.U.; Mohammad, D.

    1995-01-01

    Shielding properties of the composites of poly (ethylene glycol) and barium oxide were studied using a cobalt-60 gamma radiation source. The constituents of the composites were physically mixed in a number and discs prepared on hydraulic press at room temperature. The results reveal that PEG-BaO composites containing 20% or more of barium oxide have better radiation attenuation properties than poly (ethylene glycol). author

  11. Toward advanced gamma rays radiation resistance and shielding efficiency with phthalonitrile resins and composites

    Science.gov (United States)

    Derradji, Mehdi; Zegaoui, Abdeldjalil; Xu, Yi-Le; Wang, An-ran; Dayo, Abdul Qadeer; Wang, Jun; Liu, Wen-bin; Liu, Yu-Guang; Khiari, Karim

    2018-04-01

    The phthalonitrile resins have claimed the leading place in the field of high performance polymers thanks to their combination of outstanding properties. The present work explores for the first time the gamma rays radiation resistance and shielding efficiency of the phthalonitrile resins and its related tungsten-reinforced nanocomposites. The primary goal of this research is to define the basic behavior of the phthalonitrile resins under highly ionizing gamma rays. The obtained results confirmed that the neat phthalonitrile resins can resist absorbed doses as high as 200 kGy. Meanwhile, the remarkable shielding efficiency of the phthalonitrile polymers was confirmed to be easily improved by preparing lead-free nanocomposites. In fact, the gamma rays screening ratio reached the exceptional value of 42% for the nanocomposites of 50 wt% of nano-tungsten loading. Thus, this study confirms that the remarkable performances of the phthalonitrile resins are not limited to the thermal and mechanical properties and can be extended to the gamma rays radiation and shielding resistances.

  12. Shielding Factors for Gamma Radiation from Activity Deposited on Structures and Ground Surfaces

    DEFF Research Database (Denmark)

    Jensen, Per Hedemann

    1985-01-01

    A computer model DEPSHIELD for the calculation of shielding factors for gamma radiation at indoor residences in multistorey and single-family houses has been developed. The model is based on the exponential point kernel that links the radiation flux density at a given detector point to a point......-source strength. The radiation sources considered in the model are fallout radioactivity deposited on roofs, outer walls, and ground surfaces. For any combination of source strength on roof, outer wall, and ground surface, the model calculates shielding factors for specified photon energies. The input data...... it possible to determine the dose reduction effect from a decontamination of the different surfaces. The model has been used in a study of the consequences of land contamination of Danish territory after hypothetical core-melt accidents at the Barseback nuclear power plant in Sweden. The model has also been...

  13. Shielding factors for gamma radiation from activity deposited on structures and ground surfaces

    International Nuclear Information System (INIS)

    Hedemann Jensen, P.

    1982-11-01

    This report describes a computer model that calculates shielding factors for indoor residence in multistorey and single-family houses for gamma radiation from activity despoited on roofs, outer walls, and ground surfaces. The dimensions of the buildings including window areas and the nearby surroundings has to be speficied in the calculations. Shielding factors can be calculated for different photon energies and for a uniform surface activity distribution as well as for separate activity on roof, outer wall, and ground surface achieved from decontamination or different deposition velocities. For a given area with a known distribution of different houses a weighted shielding factor can be calculated as well as a time-averaged one based on a given residence time distribution for work/school, home, outdoors, and transportation. Calculated shielding factors are shown for typical Danish houses. To give an impression of the sensitivity of the shielding factor on the parameters used in the model, variations were made in some of the most important parameters: wall thickness, road and ground width, percentage of outer wall covered by windows, photon energy, and decontamination percentage for outer walls, ground and roofs. The uncertainity of the calculations is discussed. (author)

  14. Characterization of barite and crystal glass as attenuators in X-ray and gamma radiation shieldings

    International Nuclear Information System (INIS)

    Almeida Junior, Airton Tavares de

    2005-03-01

    Aiming to determine the barium sulphate (BaSO 4 ) ore and crystal glass attenuation features, both utilized as shieldings against ionizing X and gamma radiations in radiographic installations, a study of attenuation using barite plaster and barite concrete was carried out, which are used, respectively, on wall coverings and in block buildings. The crystal glass is utilized in screens and in windows. To do so, ten plates of barite plaster and three of barite concrete with 900 cm 2 and with an average thickness ranging from 1 to 5 cm, and three plates of crystal glass with 323 cm 2 and with thicknesses of 1, 2 and 4 cm were analyzed. The samples were irradiated with X-rays with potentials of 60, 80, 110 and 150 kilovolts, and also with 60 Co gamma rays. Curves of attenuation were obtained for barite plaster and barite concrete (mGy/mA.min) and (mGy/h), both at 1 meter, as a function of thickness and curve of transmission through barite plaster and barite concrete as a function of the thickness. The equivalent thicknesses of half and tenth value layers for barite plaster, barite concrete and crystal glass for all X-Ray energies were also determined. (author)

  15. Shielding factors for vehicles to gamma radiation from activity deposited on structures and ground surfaces

    International Nuclear Information System (INIS)

    Lauridsen, B.; Hedemann Jensen, P.

    1982-04-01

    This report describes a measuring procedure for the determination of shielding factors for vehicles passing through areas that have been contaminated by activity released to the atmosphere from a reactor accident. A simulated radiation field from fallout has been approximated by a point source that has been placed in a matrix around and above the vehicle. Modifying factors are discussed such as mutual shielding by nearby buildings and passengers. From measurements on different vehicles with and without passengers shielding factors are recommended for ordinary cars and busses in both urban and open areas, and areas with single family houses. (author)

  16. 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.)

  17. Nano lead oxide and epdm composite for development of polymer based radiation shielding material: Gamma irradiation and attenuation tests

    Science.gov (United States)

    Özdemir, T.; Güngör, A.; Akbay, I. K.; Uzun, H.; Babucçuoglu, Y.

    2018-03-01

    It is important to have a shielding material that is not easily breaking in order to have a robust product that guarantee the radiation protection of the patients and radiation workers especially during the medical exposure. In this study, nano sized lead oxide (PbO) particles were used, for the first time, to obtain an elastomeric composite material in which lead oxide nanoparticles, after the surface modification with silane binding agent, was used as functional material for radiation shielding. In addition, the composite material including 1%, 5%, 10%, 15% and 20% weight percent nano sized lead oxide was irradiated with doses of 81, 100 and 120 kGy up to an irradiation period of 248 days in a gamma ray source with an initial dose rate of 21.1 Gy/h. Mechanical, thermal properties of the irradiated materials were investigated using DSC, DMA, TGA and tensile testing and modifications in thermal and mechanical properties of the nano lead oxide containing composite material via gamma irradiation were reported. Moreover, effect of bismuth-III oxide addition on radiation attenuation of the composite material was investigated. Nano lead oxide and bismuth-III oxide particles were mixed with different weight ratios. Attenuation tests have been conducted to determine lead equivalent values for the developed composite material. Lead equivalent thickness values from 0.07 to 0.65 (2-6 mm sample thickness) were obtained.

  18. Comparative study of lead borate and bismuth lead borate glass systems as gamma-radiation shielding materials

    International Nuclear Information System (INIS)

    Singh, Narveer; Singh, Kanwar Jit; Singh, Kulwant; Singh, Harvinder

    2004-01-01

    Gamma-ray mass attenuation coefficients have been measured experimentally and calculated theoretically for PbO-B 2 O 3 and Bi 2 O 3 -PbO-B 2 O 3 glass systems using narrow beam transmission method. These values have been used to calculate half value layer (HVL) parameter. These parameters have also been calculated theoretically for some standard radiation shielding concretes at same energies. Effect of replacing lead by bismuth has been analyzed in terms of density, molar volume and mass attenuation coefficient

  19. Shielding Factors for Gamma Radiation from Activity Deposited on Structures and Ground Surfaces

    DEFF Research Database (Denmark)

    Jensen, Per Hedemann

    1985-01-01

    -source strength. The radiation sources considered in the model are fallout radioactivity deposited on roofs, outer walls, and ground surfaces. For any combination of source strength on roof, outer wall, and ground surface, the model calculates shielding factors for specified photon energies. The input data...... are the dimensions of the house, the thickness of the walls and floors, the window dimensions, and the size of the surrounding ground surface. The fallout source strength on the surfaces is allowed to have different values due to different deposition velocities to these surfaces. This feature of the model also makes...

  20. An Analysis of Radiation Penetration through the U-Shaped Cast Concrete Joints of Concrete Shielding in the Multipurpose Gamma Irradiator of BATAN

    Science.gov (United States)

    Ardiyati, Tanti; Rozali, Bang; Kasmudin

    2018-02-01

    An analysis of radiation penetration through the U-shaped joints of cast concrete shielding in BATAN’s multipurpose gamma irradiator has been carried out. The analysis has been performed by calculating the radiation penetration through the U-shaped joints of the concrete shielding using MCNP computer code. The U-shaped joints were a new design in massive concrete construction in Indonesia and, in its actual application, it is joined by a bonding agent. In the MCNP simulation model, eight detectors were located close to the observed irradiation room walls of the concrete shielding. The simulation results indicated that the radiation levels outside the concrete shielding was less than the permissible limit of 2.5 μSv/h so that the workers could safely access electrical room, control room, water treatment facility and outside irradiation room. The radiation penetration decreased as the density of material increased.

  1. Radiation shielding member

    International Nuclear Information System (INIS)

    Mizuochi, Akira; Narita, Takuya; Omori, Tetsu; Nemezawa, Isao; Kimura, Kunihiro.

    1997-01-01

    A radiation shielding member comprising a lead plate or a lead fabrication product is covered integrally with a rubber, a synthetic resin or a flame retardant fabric having a thickness greater than that of an oxidation preventive membrane made of a copper material. Radiation rays are shielded by the lead material, and not only oxidation but also failure of the lead material and generation of lead pieces or powder can be prevented by the coating of the rubber, the synthetic resin or the flame retardant fabric. The shape of the radiation shielding member can be conformed to constitutional products, a reinforcing frame or plate is formed integrally with the radiation shielding plate, alternatively, lowering of strength of the structure by fabrication of the shape is reinforced by the reinforcing frame or plate. The radiation shielding member is suspended by hanging a rope on a grommet, or disposed on constitutional products, or adjacent radiation shielding members are combined with each other by fixing metals. The thickness of the coating made of rubber, synthetic rubber or flame retardant fabric is determined to 0.1mm or greater to prevent failures of the lead material or formation of lead powder. (N.H.)

  2. Development of a new RF coil and {gamma}-ray radiation shielding assembly for improved MR image quality in SPECT/MRI

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Seunghoon; Hamamura, Mark J; Roeck, Werner W; Muftuler, L Tugan; Nalcioglu, Orhan [Tu and Yuen Center for Functional Onco-Imaging, University of California, Irvine, CA (United States)], E-mail: seunghoh@uci.edu

    2010-05-07

    Magnetic resonance (MR)-based multimodality imaging systems, such as single-photon emission tomography (SPECT)/magnetic resonance imaging (MRI) or positron emission tomography (PET)/MRI, face many difficulties because of problems with the compatibility of the nuclear detector system with the MR system. However, several studies have reported on the design considerations of MR-compatible nuclear detectors for combined SPECT/MRI. In this study, we developed a new radiofrequency (RF) coil and {gamma}-ray radiation shielding assembly to advance the practical implementation of SPECT/MRI in providing high sensitivity while minimizing the interference between the MRI and SPECT systems. The proposed assembly consists of a three-channel receive-only RF coil and {gamma}-ray radiation shields made of a specialized lead composite powder designed to reduce conductivity and thus minimizing any effect on the magnetic field arising from the induced eddy currents. A conventional birdcage RF coil was also tested for comparison with the proposed RF coil. Quality (Q)-factors were measured using both RF coils without any shielding, with solid lead shielding, and with our composite lead shielding. Signal-to-noise ratios (SNRs) were calculated using 4 T MR images of phantoms both with and without the new {gamma}-ray radiation shields. The Q-factor and SNR measurements demonstrate the improved MRI performance due to the new RF coil/{gamma}-ray radiation shield assembly designed for SPECT/MRI, making it a useful addition to multimodality imaging technology not only for animal studies but also for in vivo study of humans.

  3. An experimental study of the shielding characteristics of the dwelling house building materials against gamma radiations in the Central Region of Syria

    International Nuclear Information System (INIS)

    Albarhoum, M.; Soufan, A.H.; Mustafa, H.

    2011-01-01

    Highlights: → We measure shielding properties of dwelling houses in the central region of Syria. → The concrete used for ceiling construction is good for shielding from gamma radiations. → Fairly high linear attenuation coefficients are obtained (from 0.173 to 0.198 cm -1 ). → Blocks used for house walls are not effective against gamma radiations. → Blocks efficiency can be improved by filling their holes with a cement paste. - Abstract: The shielding properties of the concrete and blocks used for the construction of dwelling houses in the Central Region of Syria (CRS) were measured and studied. The concrete used for the ceiling construction was found to have optimum shielding properties with 0.182 cm -1 (or equivalently 0.0859 cm 2 g -1 ) for the linear (mass) attenuation coefficient [L(M)AC]. In addition gamma radiation is attenuated by 73.221% on average, while the blocks used for the walls have smaller LACs (0.082 cm -1 for the bare blocks, and 0.118 cm -1 for the coated ones). Although the LACs for the blocks are smaller than those for the concrete their shielding properties are good to protect from the gamma radiations coming from radioactive or nuclear accidents (78.630% attenuation), even Chernobyl - like disasters, because of their big width (10-12 cm). The LACs were measured by an ionization chamber and simple theoretical calculations have been made to predict the concrete LACs. The calculations showed an average LAC for the six samples equal to 0.1664 cm -1 with 8.47% error with respect to the experimental values. The average LAC for the concrete used for ceiling construction in the CRS was found to be comparable or even better than the average of some international values for the reactor shielding concretes, which are about 0.163 cm -1 .

  4. Determination of the shielding factors for gamma-ray spectrometers

    International Nuclear Information System (INIS)

    Korun, M.; Vodenik, B.; Zorko, B.

    2014-01-01

    A method for determining the shielding factors for gamma-ray spectrometers is described. The shielding factors are expressed by decomposing the peaked background of the spectrometer into contributions of the detector, spectrometer shield and ambient radiation to the spectrometer background. The dimensions of the sample and its mass-attenuation coefficient are taken into account using a simple model. For six spectrometers, with contributions to the background quantified, the shielding factors were determined for the background based on the thorium decay series and the radon daughters. For a water sample with a diameter of 9 cm and a thickness of 4 cm and the nuclides of the thorium decay series that are in the spectrometer shields, the values of the shielding factors lie in the interval 0.95–1.00. For a spectrometer exhibiting the diffusion of radon into the shielding material, the values of the shielding factors for the same sample for gamma-rays from the radon daughters lie in the interval 0.88–1.00. - Highlights: • A model is described to assess shielding factors for gamma-ray spectrometers. • The background due to the detector, shield and ambient radiation must be known. • The sample attenuation, its dimensions and distance from the crystal are considered. • Shielding factors for gamma-rays from the 232 Th and 226 Ra decay chains are assessed. • For a water sample with a mass of 0.25 kg, shielding factors above 0.88 are obtained

  5. 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)

  6. Spacecraft Electrostatic Radiation Shielding

    Science.gov (United States)

    2008-01-01

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

  7. 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.)

  8. Study of gamma radiation shielding properties of ZnO−TeO2 glasses

    Indian Academy of Sciences (India)

    2017-07-25

    Jul 25, 2017 ... molar volume and thickness are listed in table 1. 2.2 Measurements. Mass attenuation coefficient measurements were performed using a gamma-ray spectrometer, employing a scintillation detector (3 × 3 inch) (figure 1). Its hermetically sealed assembly includes a high-resolution NaI(Tl) crystal, photo-.

  9. Radiation shielding member

    International Nuclear Information System (INIS)

    Tada, Nobuo; Ito, Masato; Nihei, Ken-ichi; Takeshi, Tetsu

    1998-01-01

    A radiation shielding member comprises a metal vessel and a liquid therein, and is disposed to the upper surface of a lower flange of a reactor core shroud. Waterproof hot wires are contained in the liquid and are connected to a power source disposed at the outside. Electric current is supplied to the hot wires to elevate the temperature of the liquid, and the temperature of the vessel is kept higher than an atmospheric temperature thereby suppressing generation of dew condensation or water droplets. In addition, a water repellent coating is applied to the shielding member itself to prevent deposition of water droplets. Further, the bottom of the shielding member is inclined, and a water droplet-recovering vessel is disposed at the lower portion of the shielding member, so that the water droplets collected by the inclination of the bottom are recovered to the water droplet recovering vessel. With such a constitution, access of an operator to the inside of a reactor pressure vessel is facilitated, and at the same time, the working circumstance at the reactor bottom can be improved. (I.N.)

  10. Effect of Flyash Addition on Mechanical and Gamma Radiation Shielding Properties of Concrete

    Directory of Open Access Journals (Sweden)

    Kanwaldeep Singh

    2014-01-01

    Full Text Available Six concrete mixtures were prepared with 0%, 20%, 30%, 40%, 50%, and 60% of flyash replacing the cement content and having constant water to cement ratio. The testing specimens were casted and their mechanical parameters were tested experimentally in accordance with the Indian standards. Results of mechanical parameters show their improvement with age of the specimens and results of radiation parameters show no significant effect of flyash substitution on mass attenuation coefficient.

  11. 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...

  12. 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)

  13. 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)

  14. 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 ...

  15. 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

  16. 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.)

  17. 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)

  18. 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...

  19. 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

  20. 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.

  1. 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.

  2. Radiation shielding aspects of compact medical cyclotrons

    International Nuclear Information System (INIS)

    Mukherjee, B.; Ruth, T.J.

    1995-01-01

    Hospital-based compact medical cyclotrons are commonly used to produce large activities of short-lived PET radioisotopes such as fluorine-18 (HL110 min) and oxygen-18 (HL= 20 min), by bombarding suitably enriched gas or liquid targets with 11-15 MeV protons. High energy prompt neutron/gamma radiation fields are generated as the nuclear reaction product. The compact medical cyclotrons are installed either inside or in the close proximity of the nuclear medicine clinic. Therefore the adequacy of the radiation shielding is vitally important for the radiological safety of the patients and members of the public. The present paper highlights the important radiation shielding aspects of some compact medical cyclotrons presently available in the international market. 2 tabs., 4 figs

  3. Radiation shielding in dental radiography.

    Science.gov (United States)

    Stenström, B; Rehnmark-Larsson, S; Julin, P; Richter, S

    1983-01-01

    The protective effect in the thyroid region from different types of radiation shielding at intraoral radiography has been studied as well as the reduction of the absorbed dose to the sternal and the gonadal regions. The shielding tested were the 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 times 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 shielding 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 microGy compared with 18 microGy (parallelling) and 31 microGy (bisecting-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 was approximately one per cent of the dose without any shielding, i.e. below 0.01 microGy per single intraoral exposure.

  4. Radiation shielding in dental radiography

    Energy Technology Data Exchange (ETDEWEB)

    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 ..mu..Gy compared with 18 ..mu..Gy (parallelling) and 31 ..mu..Gy (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 ..mu..Gy per single intraoral exposure.

  5. Radiation shielding in dental radiography

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-01-01

    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. In the sternal region the combinations of apron and collar reduced the absorbed dose from a full survey to below 2 ..mu..Gy compared with 18 ..mu..Gy (parallelling) and 31 ..mu..Gy (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 ..mu..Gy per single intraoral exposure.

  6. 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

  7. A Comprehensive Study on Gamma Rays and Fast Neutron Sensing Properties of GAGOC and CMO Scintillators for Shielding Radiation Applications

    Directory of Open Access Journals (Sweden)

    Shams A. M. Issa

    2017-01-01

    Full Text Available The WinXCom program has been used to calculate the mass attenuation coefficients (μm, effective atomic numbers (Zeff, effective electron densities (Nel, half-value layer (HVL, and mean free path (MFP in the energy range 1 keV–100 GeV for Gd3Al2Ga3O12Ce (GAGOC and CaMoO4 (CMO scintillator materials. The geometrical progression (G-P method has been used to compute the exposure buildup factors (EBF and gamma ray energy absorption (EABF in the photon energy range 0.015–15 MeV and up to a 40 penetration depth (mfp. In addition, the values of the removal cross section for a fast neutron ∑R have been calculated. The computed data observes that GAGOC showed excellent γ-rays and neutrons sensing a response in the broad energy range. This work could be useful for nuclear radiation sensors, detectors, nuclear medicine applications (medical imaging and mammography, nuclear engineering, and space technology.

  8. Attenuation of a non-parallel beam of gamma radiation by thick shielding-application to the determination of the 235U enrichment with NaI detectors

    International Nuclear Information System (INIS)

    Mortreau, Patricia; Berndt, Reinhard

    2005-01-01

    The traditional method used to determine the Uranium enrichment by nondestructive analysis is based on the 'enrichment meter principle' [1]. It involves measuring the intensity of the 186 keV net peak area of 235 U in 'quasi-infinite' samples. A prominent factor, which affects the peak intensity, is the presence of gamma absorbing material (e.g., container wall, detector cover) between the sample and the detector. Its effect is taken into consideration in a commonly called 'wall thickness' correction factor. Often calculated on the basis of approximations, its performance is adequate for small attenuation factors applicable to the case of narrow beams. However these approximations do not lead to precise results when wide non-parallel beams are attenuated through thick container walls. This paper is dedicated to the calculation by numerical integration of the geometrical correction factor (K wtc ) which describes the effective mean path length of the radiation through the absorbing layer. This factor was calculated as a function of various measurement parameters (types and dimensions of the detector, of the collimator and of the shielding) for the most commonly used collimator shapes and detectors. Both coherent scattering (Rayleigh) and incoherent scattering (Compton) are taken into account for the calculation of the radiation interaction within the detector

  9. 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

  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. 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.

  12. Designing Shelter in New Buildings. A Manual for Architects on the Preliminary Designing of Shielding from Fallout Gamma Radiation in Normally Functioning Spaces in New Buildings.

    Science.gov (United States)

    Knott, Albert

    Analysis of radiation fallout prevention factors in new construction is presented with emphasis on architectural shielding principles. Numerous diagrams and charts illustrate--(1) radiation and fallout properties, (2) building protection principles, (3) details and planning suggestions, and (4) tabular data interpretation. A series of charts is…

  13. A Novel Radiation Shielding Material Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Radiation shielding simulations showed that epoxy loaded with 10-70% polyethylene would be an excellent shielding material against GCRs and SEPs. Milling produced an...

  14. Boron filled siloxane polymers for radiation shielding

    Science.gov (United States)

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

    2018-03-01

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

  15. 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.

  16. 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)

  17. 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)

  18. MicroShield/ISOCS gamma modeling comparison.

    Energy Technology Data Exchange (ETDEWEB)

    Sansone, Kenneth R

    2013-08-01

    Quantitative radiological analysis attempts to determine the quantity of activity or concentration of specific radionuclide(s) in a sample. Based upon the certified standards that are used to calibrate gamma spectral detectors, geometric similarities between sample shape and the calibration standards determine if the analysis results developed are qualitative or quantitative. A sample analyzed that does not mimic a calibrated sample geometry must be reported as a non-standard geometry and thus the results are considered qualitative and not quantitative. MicroShieldR or ISOCSR calibration software can be used to model non-standard geometric sample shapes in an effort to obtain a quantitative analytical result. MicroShieldR and Canberras ISOCSR software contain several geometry templates that can provide accurate quantitative modeling for a variety of sample configurations. Included in the software are computational algorithms that are used to develop and calculate energy efficiency values for the modeled sample geometry which can then be used with conventional analysis methodology to calculate the result. The response of the analytical method and the sensitivity of the mechanical and electronic equipment to the radionuclide of interest must be calibrated, or standardized, using a calibrated radiological source that contains a known and certified amount of activity.

  19. Optically-transparent radiation-shielding composition

    International Nuclear Information System (INIS)

    Bolles, T.F.; Fleming, P.B.

    1976-01-01

    An optically transparent, essentially colorless radiation shielding material for high energy radiation contains a combination of lead or thallium salts of C 1 to C 5 organic acids and may contain lead or thallium salts of mineral acids. Shields of complex shapes are easily constructed

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

    Energy Technology Data Exchange (ETDEWEB)

    Joenemalm, C.; Malen, K

    1966-10-15

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

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

    International Nuclear Information System (INIS)

    Joenemalm, C.; Malen, K

    1966-10-01

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

  3. Integrated Solar Concentrator and Shielded Radiator

    Science.gov (United States)

    Clark, David Larry

    2010-01-01

    A shielded radiator is integrated within a solar concentrator for applications that require protection from high ambient temperatures with little convective heat transfer. This innovation uses a reflective surface to deflect ambient thermal radiation, shielding the radiator. The interior of the shield is also reflective to provide a view factor to deep space. A key feature of the shield is the parabolic shape that focuses incoming solar radiation to a line above the radiator along the length of the trough. This keeps the solar energy from adding to the radiator load. By placing solar cells along this focal line, the concentration of solar energy reduces the number and mass of required cells. By shielding the radiator, the effective reject temperature is much lower, allowing lower radiator temperatures. This is particularly important for lower-temperature processes, like habitat heat rejection and fuel cell operations where a high radiator temperature is not feasible. Adding the solar cells in the focal line uses the concentrating effect of the shield to advantage to accomplish two processes with a single device. This shield can be a deployable, lightweight Mylar structure for compact transport.

  4. Shielding of gamma field in residential houses

    International Nuclear Information System (INIS)

    Smejkal, Z.; Pavlata, M.; Pokorna, I.; Urban, M.

    1995-01-01

    In the past some flats were built from defective materials contained uranium-238, which radiate dangerous gamma radiation. The object of this work consisted in searching mechanical barriers, which would decrease penetrating of this radiation into a flat. The measurement was realized in system made of connecting of Ge/Li detector with multichannel analyser MCA JAK 202 and IBM PC. Plenty of building parts such as bricks, plaster slabs with/without lead dust, wasted plaster from Pocerady Electric Power Station (EPS), etc., were measured to get and compare shading abilities. Maximal intensity of gamma radiation (47.1%) is visible for energy E=609 keV radium-226, therefore the measurement was only carried out for this energy. The measurement performed in defective houses start during years 1988-1991 demonstrated that excepting higher activity radon-222 and its daughter products forms uneligible gamma field, as well. This is limited by values of rate dose equivalent. The problem was successfully solved by lead slabs fixed to wood construction that is covered by applications. The manipulation with materials and construction was difficult, therefore another materials and segments were tested, for more easy fix to defective walls. In 1995 the experiment was realised in the cooperation with the chemical department of Pocerady EPS, the plaster is outlet product from the removing sulphur process. There were made an experimental slabs, sizes 18 x 18 x 2 cm. The barrier effect of slabs were compared with other building material and parts. So that the elimination of radiation would be effective is necessary reduce the level of radiation penetrating to the smallest level. However, the the thickness of shading material is limited by economical reasons, prices of material, square weighting and reducing of living room. The results of measuring is this one: the plaster slabs with lead dust made in EPS Pocerady are suitable to reduce gamma ray, the values of reducing coefficient are

  5. 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

  6. 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.

  7. 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...

  8. Foam-Reinforced Polymer Matrix Composite Radiation Shields Project

    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...

  9. 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...

  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. 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)

  12. A low background gamma ray spectrometer with anticosmic shielding

    International Nuclear Information System (INIS)

    Nguyen Quoc Hung; Vo Hong Hai; Tran Kim Tuyet; Ho Lai Tuan

    2016-01-01

    The article describes a gamma ray spectrometer protected by a lead shield (Model 747E Canberra lead shield) and an active shield made of an 80 cm x 80 cm x 3 cm plastic scintillator plate in anticoincidence on top of the lead shield. The detector used as low background gamma-spectrometer is a high purity Germanium crystal of model GC2018 Canberra. The background count rate currently achieved (30-2400 keV) is 1.27 cps without anticoincidence. The level of background suppression obtained from the active protection is 0.80 overall and about 0.43 for the 511 keV gamma line. The gamma ray spectrometer is installed and operated in the Nuclear Laboratory, Department of Nuclear Physics, University of Science, Ho Chi Minh City Vietnam National University. (author)

  13. 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.)

  14. 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...

  15. 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

  16. 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

  17. Effectiveness of construction materials and some minerals used as radiation shielding

    International Nuclear Information System (INIS)

    Khunarak, P.; Bunnak, S.

    1988-01-01

    There are many kinds of ores in Thailand, some large amount of them are cheap and easy to obtain possess shielding properties for gamma radiation. These ores are baryte, illmenite, galena, scheelite, wolframite pyrite, cerrusite. Besides, building structure materials are also introduced for shielding properties study by using Co-60, Cs-137 and Ra-226 as gamma radiation sources in the experiments. The results turn out that those high density ores will possess a better shielding property than the low density ores. Radiation measurement equipment is G.M. tube connected to rate meter

  18. Gamma shielding by aluminum (Al-shielder manual)

    International Nuclear Information System (INIS)

    Zeb, J.; Arshad, W.; Rashid, A.; Akhter, P.

    2010-12-01

    This report is an operational manual of shielding software 'AI-Shielder', developed at Health Physics Division (HPD), PINSTECH. The AI-Shielder software estimates shielding thickness of Aluminum for photons having energy in the range 0.5 to 10 MeV. The software is helpful in safe handling or storage of gamma emitting radionuclide(s) by using Aluminum as shielding material. To compute the shield thickness, self absorption in the source has been neglected and the source has been assumed as a point source. This report is intended not only to serve as a manual of the AI-Shielder, but also to provide the user(s) with the necessary information about gamma shielding. (author)

  19. Correlated Uncertainties in Radiation Shielding Effectiveness

    Science.gov (United States)

    Werneth, Charles M.; Maung, Khin Maung; Blattnig, Steve R.; Clowdsley, Martha S.; Townsend, Lawrence W.

    2013-01-01

    The space radiation environment is composed of energetic particles which can deliver harmful doses of radiation that may lead to acute radiation sickness, cancer, and even death for insufficiently shielded crew members. Spacecraft shielding must provide structural integrity and minimize the risk associated with radiation exposure. The risk of radiation exposure induced death (REID) is a measure of the risk of dying from cancer induced by radiation exposure. Uncertainties in the risk projection model, quality factor, and spectral fluence are folded into the calculation of the REID by sampling from probability distribution functions. Consequently, determining optimal shielding materials that reduce the REID in a statistically significant manner has been found to be difficult. In this work, the difference of the REID distributions for different materials is used to study the effect of composition on shielding effectiveness. It is shown that the use of correlated uncertainties allows for the determination of statistically significant differences between materials despite the large uncertainties in the quality factor. This is in contrast to previous methods where uncertainties have been generally treated as uncorrelated. It is concluded that the use of correlated quality factor uncertainties greatly reduces the uncertainty in the assessment of shielding effectiveness for the mitigation of radiation exposure.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  1. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kebwaro, Jeremiah Monari; Zhao, Yaolin, E-mail: zhaoyaolin@mail.xjtu.edu.cn; He, Chaohui

    2015-04-01

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

  3. 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

  4. DIII-D radiation shielding procedures and experiences

    International Nuclear Information System (INIS)

    Taylor, P.L.

    1991-11-01

    The D3-D tokamak operates with a neutron radiation shield to allow enhanced plasma operations with increased neutron production while minimizing the site boundary dose level. Neutron rates as high as 4 x 10 15 neutrons/s and total neutron production as high as 4 x 10 15 neutrons per shot are obtained while maintaining the site dose below the DOE administrative level of 20 mrem per year; a much more restrictive level than the State of California radiation limits. The radiation shielding has increased by a factor of 300 over the preshield value and is in agreement with the design calculation. The maximum site neutron dose since installation of the shield has been less than 0.03 mrem for a shot and less than 0.4 mrem for a day. The site neutron and gamma dose are monitored continuously during operations by a PC-based computer system that provides the means of measuring the low dose levels that occur during a shot by including postshot background subtraction. The neutron and gamma dose are measured and archived by shot, hour, and day in a database. Activation of the machine after a run day and during vessel entries is monitored and the activated nuclides have been determined. A radiation monitoring program and procedures are used to control the exposures to facility personnel and the exposure at the site boundary

  5. Shielding application of perturbation theory to determine changes in neutron and gamma doses due to changes in shield layers

    Science.gov (United States)

    Fieno, D.

    1972-01-01

    Perturbation theory formulas were derived and applied to determine changes in neutron and gamma-ray doses due to changes in various radiation shield layers for fixed sources. For a given source and detector position, the perturbation method enables dose derivatives with respect to density, or equivalently thickness, for every layer to be determined from one forward and one inhomogeneous adjoint calculation. A direct determination without the perturbation approach would require two forward calculations to evaluate the dose derivative due to a change in a single layer. Hence, the perturbation method for obtaining dose derivatives requires fewer computations for design studies of multilayer shields. For an illustrative problem, a comparison was made of the fractional change in the dose per unit change in the thickness of each shield layer in a two-layer spherical configuration as calculated by perturbation theory and by successive direct calculations; excellent agreement was obtained between the two methods.

  6. Radiation shielding fiber and its manufacturing method

    International Nuclear Information System (INIS)

    Tanaka, Koji; Ono, Hiroshi.

    1988-01-01

    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.)

  7. Investigation of gamma ray shielding efficiency and mechanical performances of concrete shields containing bismuth oxide as an environmentally friendly additive

    Science.gov (United States)

    Yao, Ya; Zhang, Xiaowen; Li, Mi; Yang, Rong; Jiang, Tianjiao; Lv, Junwen

    2016-10-01

    Concrete has a proven ability to attenuate gamma rays and neutrons without compromising structural property; therefore, it is widely used as the primary shielding material in many nuclear facilities. Recently, there is a tendency toward using various additives to enhance the shielding properties of these concrete mixtures. However, most of these additives being used either pose hygiene hazards or require special handling processes. It would be ideal if environmentally friendly additives were available for use. The bismuth oxide (Bi2O3) additive shows promise in various shielding applications due to its proven radiation attenuation ability and environmentally friendly nature. To the best of our knowledge, however, Bi2O3 has never been used in concrete mixtures. Therefore, for this research, we fabricated the Bi2O3-based concrete mixtures by adding Bi2O3 powder in the ordinary concrete mixture. Concrete mixtures with lead oxide (PbO) additives were used for comparison. Radiation shielding parameters like the linear attenuation coefficients (LAC) of all these concrete mixtures showing the effects of the Bi2O3 additions are presented. The mechanical performances of concrete mixtures incorporated with Bi2O3 additive were also investigated. It suggested that the concrete mixture containing 25% Bi2O3 powder (B5 in this study) provided the best shielding capacity and mechanical performance among other mixes. It has a significant potential for application as a structural concrete where radiological protection capability is required.

  8. 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.)

  9. 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

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

    Science.gov (United States)

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

    2018-03-01

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

  11. Radiation shielding and health physics instrumentation for PET medical cyclotrons

    International Nuclear Information System (INIS)

    Mukherjee, B.

    2002-01-01

    Full text: Modern Medical Cyclotrons produce a variety of short-lived positron emitting PET radioisotopes, and as a result are the source of intense neutron and gamma radiations. Since such cyclotrons are housed within hospitals or medical clinics, there is significant potential for un-intentional exposure to staff or patients in proximity to cyclotron facilities. Consequently, the radiological hazards associated with Cyclotrons provide the impetus for an effective radiological shielding and continuous monitoring of various radiation levels in the cyclotron environment. Management of radiological hazards is of paramount importance for the safe operation of a Medical Cyclotron facility. This work summarised the methods of shielding calculations for a compact hospital based Medical Cyclotron currently operating in Canada, USA and Australia. The design principle and operational history of a real-time health physics monitoring system (Watchdog) operating at a large multi-energy Medical Cyclotron is also highlighted

  12. Nanocomposite for Radiation Shielding Project

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA's Advanced Extravehicular Activity (EVA) program requires the need for materials that can protect astronauts and spacecrafts from ionizing radiations such as...

  13. 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.

  14. 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)

  15. 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.)

  16. Radiation shielding conceptual design of RRI-50 reactor

    International Nuclear Information System (INIS)

    Amir Hamzah; Iman Kuntoro

    2015-01-01

    One of the parameters that must be met in the design of nuclear reactors is radiation shielding design to ensure the security and safety of workers and the surrounding community. This study has been conducted to design radiation shielding of RRI-50 with high density U 9 Mo-Al fuel elements that consist of 21 pieces of plate type fuel elements with dimension as same as RSG-GAS fuel elements but the active length is 70 cm. Core configurations consist of 16 fuel elements and 4 control elements and 5 irradiation positions to form a matrix of 5 x 5. The objective of this research is to design radiation shielding and determine the distribution of dose rates in the working area and the environment of RRI-50 reactor. The early stages of this research is to calculate source strength and inventory of radioactive materials within the reactor core with one operation pattern cycle of 50 MW for 20 days using ORGEN2.1 program. Based on core source strength and models that are created using the VisEd software, the analysis parameter of the shielding was determined iteratively using MCNPX program. In the final stage, an analysis of the dose rate distributions in the whole space inside and outside the reactor building was conducted also using MCNPX program. The results show that the height of the water surface is 1000 cm and the combination of heavy concrete thickness of 90 cm and ordinary concrete thickness of 60 cm can be used as an biological shield. This design can reduce the dose rate to 0.05 µSv/h in the Operations Room while in the Experiments Room and outside the reactor building to 4.2 µSv/h and 0.03 µSv/h during reactor operation. The results also suggest that the installation of additional radiation shield of 280 cm thickness within 300 cm in front of the open radial neutron beam tube can reduce gamma and neutron dose rate to 3.3 µSv/h and 3.1 x 10 -11 µSv/h. The results of this study indicate that the radiation shield design is made to make reactor RRI-50 to be safe

  17. Radiation safety shield for a syringe

    International Nuclear Information System (INIS)

    Tipton, H.W.

    1976-01-01

    Safety apparatus for use in administering radioactive serums by a syringe, without endangering the health and safety of the medical operators is described. The apparatus consists of a sheath and a shield which can be retracted into the sheath to assay the radioactive serum in an assay well. The shield can be moved from the retracted position into an extended position when the serum is to be injected into the patient. To protect the operator, the shield can be constructed of tantalum or any like high density substance to attenuate the radiation, emanating from the radioactive serums contained in the syringe, from passing to the atmosphere. A lead glass window is provided so that the operator can determine the exact quantity of the radioactive serum which is contained in the syringe

  18. Effects of Shielding on Gamma Rays

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-03-13

    The interaction of gamma rays with matter results in an effect we call attenuation (i.e. ‘shielding’). Attenuation can dramatically alter the appearance of a spectrum. Attenuating materials may actually create features in a spectrum via x-ray fluorescence

  19. Enhanced radiation shielding with galena concrete

    Directory of Open Access Journals (Sweden)

    Hadad Kamal

    2015-01-01

    Full Text Available A new concrete, containing galena mineral, with enhanced shielding properties for gamma sources is developed. To achieve optimized shielding properties, ten types of galena concrete containing different mixing ratios and a reference normal concrete of 2300 kg/m3 density are studied experimentally and numerically using Monte Carlo and XCOM codes. For building galena concrete, in addition to the main composition, micro-silica and water, galena mineral (containing lead were used. The built samples have high density of 4470 kg/m3 to 5623 kg/m3 and compressive strength of 628 kg/m2 to 685 kg/m2. The half and tenth value layers (half value layer and tenth value layers for the galena concrete, when irradiated with 137Cs gamma source, were found to be 1.45 cm and 4.94 cm, respectively. When irradiated with 60Co gamma source, half value layer was measured to be 2.42 cm. The computation modeling by FLUKA and XCOM shows a good agreement between experimental and computational results.

  20. Multifunctional Structural Composites for Radiation Shielding, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Radiation shielding materials are necessary for protecting astronaut crews from the hazards of space radiation during future NASA missions. Although polyethylene...

  1. Characterization of barite and crystal glass as attenuators in X-ray and gamma radiation shieldings; Caracterizacao da barita e do vidro cristal como atenuadores na blindagem das radiacoes X e gama

    Energy Technology Data Exchange (ETDEWEB)

    Almeida Junior, Airton Tavares de

    2005-03-15

    Aiming to determine the barium sulphate (BaSO{sub 4}) ore and crystal glass attenuation features, both utilized as shieldings against ionizing X and gamma radiations in radiographic installations, a study of attenuation using barite plaster and barite concrete was carried out, which are used, respectively, on wall coverings and in block buildings. The crystal glass is utilized in screens and in windows. To do so, ten plates of barite plaster and three of barite concrete with 900 cm{sup 2} and with an average thickness ranging from 1 to 5 cm, and three plates of crystal glass with 323 cm{sup 2} and with thicknesses of 1, 2 and 4 cm were analyzed. The samples were irradiated with X-rays with potentials of 60, 80, 110 and 150 kilovolts, and also with {sup 60}Co gamma rays. Curves of attenuation were obtained for barite plaster and barite concrete (mGy/mA.min) and (mGy/h), both at 1 meter, as a function of thickness and curve of transmission through barite plaster and barite concrete as a function of the thickness. The equivalent thicknesses of half and tenth value layers for barite plaster, barite concrete and crystal glass for all X-Ray energies were also determined. (author)

  2. 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)

  3. Radiation shielded movable work station apparatus

    International Nuclear Information System (INIS)

    Meuschke, R.E.; Andrews, H.N.; Massaro, A.A. Jr.

    1982-01-01

    A movable work station includes travelling hoist-supported tools and a radiation-shielded enclosure or gondola that may be moved vertically or rotated. The enclosure is divided and accommodates four upright workers in facing pairs at opposite sides of a clearanceway observable and accessible from the gondola interior via lead glass windows and hand holes. The work station is particularly suitable for personnel involved in tube bundle replacement tasks performed within the shell of a nuclear power plant steam generator

  4. 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

  5. Determining optical and radiation characteristics of cathode ray tubes' glass to be reused as radiation shielding glass

    Science.gov (United States)

    Zughbi, A.; Kharita, M. H.; Shehada, A. M.

    2017-07-01

    A new method of recycling glass of Cathode Ray Tubes (CRTs) has been presented in this paper. The glass from CRTs suggested being used as raw materials for the production of radiation shielding glass. Cathode ray tubes glass contains considerable amounts of environmentally hazardous toxic wastes, namely heavy metal oxides such as lead oxide (PbO). This method makes CRTs glass a favorable choice to be used as raw material for Radiation Shielding Glass and concrete. The heavy metal oxides increase its density, which make this type of glass nearly equivalent to commercially available shielding glass. CRTs glass have been characterized to determine heavy oxides content, density, refractive index, and radiation shielding properties for different Gamma-Ray energies. Empirical methods have been used by using the Gamma-Ray source cobalt-60 and computational method by using the code XCOM. Measured and calculated values were in a good compatibility. The effects of irradiation by gamma rays of cobalt-60 on the optical transparency for each part of the CRTs glass have been studied. The Results had shown that some parts of CRTs glass have more resistant to Gamma radiation than others. The study had shown that the glass of cathode ray tubes could be recycled to be used as radiation shielding glass. This proposed use of CRT glass is only limited to the available quantity of CRT world-wide.

  6. Attenuation of a non-parallel beam of gamma radiation by thick shielding-application to the determination of the {sup 235}U enrichment with NaI detectors

    Energy Technology Data Exchange (ETDEWEB)

    Mortreau, Patricia [European Commission, Joint Research Centre, Institute for the Protection and Security of the Citizen, TP 800 Via Fermi, Ispra (Vatican City State, Holy See,) (Italy)]. E-mail: patricia.mortreau@jrc.it; Berndt, Reinhard [European Commission, Joint Research Centre, Institute for the Protection and Security of the Citizen, TP 800 Via Fermi, Ispra (VA) (Italy)

    2005-09-21

    The traditional method used to determine the Uranium enrichment by nondestructive analysis is based on the 'enrichment meter principle' [1]. It involves measuring the intensity of the 186 keV net peak area of {sup 235}U in 'quasi-infinite' samples. A prominent factor, which affects the peak intensity, is the presence of gamma absorbing material (e.g., container wall, detector cover) between the sample and the detector. Its effect is taken into consideration in a commonly called 'wall thickness' correction factor. Often calculated on the basis of approximations, its performance is adequate for small attenuation factors applicable to the case of narrow beams. However these approximations do not lead to precise results when wide non-parallel beams are attenuated through thick container walls. This paper is dedicated to the calculation by numerical integration of the geometrical correction factor (K {sub wtc}) which describes the effective mean path length of the radiation through the absorbing layer. This factor was calculated as a function of various measurement parameters (types and dimensions of the detector, of the collimator and of the shielding) for the most commonly used collimator shapes and detectors. Both coherent scattering (Rayleigh) and incoherent scattering (Compton) are taken into account for the calculation of the radiation interaction within the detector.

  7. Mobile robot prototype detector of gamma radiation

    International Nuclear Information System (INIS)

    Vazquez C, R.M.; Duran V, M. D.; Jardon M, C. I.

    2014-10-01

    In this paper the technological development of a mobile robot prototype detector of gamma radiation is shown. This prototype has been developed for the purpose of algorithms implementation for the applications of terrestrial radiation monitoring of exposed sources, search for missing radioactive sources, identification and delineation of radioactive contamination areas and distribution maps generating of radioactive exposure. Mobile robot detector of radiation is an experimental technology development platform to operate in laboratory environment or flat floor facilities. The prototype integrates a driving section of differential configuration robot on wheels, a support mechanism and rotation of shielded detector, actuator controller cards, acquisition and processing of sensor data, detection algorithms programming and control actuators, data recording (Data Logger) and data transmission in wireless way. The robot in this first phase is remotely operated in wireless way with a range of approximately 150 m line of sight and can extend that range to 300 m or more with the use of signal repeaters. The gamma radiation detection is performed using a Geiger detector shielded. Scan detection is performed at various time sampling periods and diverse positions of discrete or continuous angular orientation on the horizon. The captured data are geographical coordinates of robot GPS (latitude and longitude), orientation angle of shield, counting by sampling time, date, hours, minutes and seconds. The data is saved in a file in the Micro Sd memory on the robot. They are also sent in wireless way by an X Bee card to a remote station that receives for their online monitoring on a laptop through an acquisition program by serial port on Mat Lab. Additionally a voice synthesizing card with a horn, both in the robot, periodically pronounced in Spanish, data length, latitude, orientation angle of shield and detected accounts. (Author)

  8. Configuration Design of Detector Shielding for Gamma Prompt Analysis

    International Nuclear Information System (INIS)

    Elin-Nuraini; Darsono; Elisabeth

    2000-01-01

    Configuration on design of detector shielding for gamma prompt analysishas been performed. The aim of this design is to obtain effective shieldingmaterial and configuration that able to protect the detector for fastneutron. The result shown that detector shielding configuration that obtainedby configuration of water and concrete, would be able to absorb fast neutronup to 99.5 %. The neutron flux that passed through shielding configuration is2.4 x 10 3 n/cm 2 dt, in the detector position of 60 cm (forward neutron beamdirection) on the X axis and 30 cm (side ward neutron beam direction) on theZ axis of target. On this position (60,30) counting result was 104358 for Pbcollimator and 246652 for PVC collimator. From examination result shown thatthe weight of silicon is in order 175 gram. (author)

  9. 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

  10. 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.

  11. Summary of Prometheus Radiation Shielding Nuclear Design Analysis

    Energy Technology Data Exchange (ETDEWEB)

    J. Stephens

    2006-01-13

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

  12. 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

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

    Directory of Open Access Journals (Sweden)

    Watcharin Rachniyom

    2015-12-01

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

  14. Attenuation characteristics of materials used in radiation protection as radiation shielding

    Energy Technology Data Exchange (ETDEWEB)

    Almeida Junior, Airton T., E-mail: airton.almeida@fundacentro.gov.br [Fundacao Jorge Duprat Figueiredo de Seguranca e Medicina do Trabalho (FUNDACENTRO), Belo Horizonte, MG (Brazil); Araujo, F.G.S. [Universidade Federal de Ouro Preto (UFOP/REDEMAT), MG (Brazil); Nogueira, M.S. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Santos, M.A.P. [Centro Regional de Ciencias Nucleares (CRCN-NE/CNEN-PE), Recife, PE (Brazil)

    2013-07-01

    Crystal glass has been widely used as shielding material in gamma radiation sources as well as x-ray generating equipment to replace the plumbiferous glass, in order to minimize exposure to individuals. In this work, ten plates of crystal glass, with dimensions of 20cm x 20cm and range of thicknesses from 0.5 to 2.0 cm, and barite concrete were irradiated with potential constants of 60kV, 80kV, 110kV, 150kV and gamma radiation of {sup 60}Co. The curves of attenuation and of transmission were obtained for crystal glass, barite plaster and barite concrete (mGy/mA.min) at 1 meter as a function of thickness.Crystal glass has been widely used as shielding material in gamma radiation sources as well as x-ray generating equipment to replace the plumbiferous glass, in order to minimize exposure to individuals. In this work, ten plates of crystal glass, with dimensions of 20cm x 20cm and range of thicknesses from 0.5 to 2.0 cm, and barite concrete were irradiated with potential constants of 60kV, 80kV, 110kV, 150kV and gamma radiation of {sup 60}Co. The curves of attenuation and of transmission were obtained for crystal glass, barite plaster and barite concrete (mGy/mA.min) at 1 meter as a function of thickness. (author)

  15. Radiation shielding for medical compact cyclotron

    International Nuclear Information System (INIS)

    Futatsukawa, Syoji; Hatakeyama, Satoru; Saito, Yoshihiro; Sera, Kouichiro; Hatano, Kentaro; Sasaki, Toshiaki.

    1993-01-01

    We are using a medical compact cyclotron for PET (positron emission tomography) and PIXE (particle induced X-ray emission) analysis in Nishina Memorial Cyclotron Center. The cyclotron vault is covered by concrete wall of 1.5 m thickness. However, a big penetrating window of 1.8 m square was needed at the concrete wall between the cyclotron vault and the PIXE room for beam transport equipment. This window was closed by packing materials of iron shots, small grained polyethylenes mixed with boron and lead grains for shielding of neutrons and γ-rays. Several measurement data have showed that this method is useful for shielding of radiations from the compact cyclotron. (author)

  16. Gamma and X-ray shielding compositions utilizing bauxite - Red Mud regional research laboratory (CSIR), Bhopal, India

    International Nuclear Information System (INIS)

    Anshul, Avneesh; Amritphale, Sudhir Sitaram; Chandra, Navin; Ramakrishnan, N.

    2007-01-01

    Available in abstract form only. Full text of publication follows: The application spectrum of X-ray and Gamma radiation is increasing exponentially in the area of diagnostic, nuclear medicine, food preservation, nuclear power plants and strategic utilities. To prevent the harmful effects of these radiations, shielding materials based on lead metal and its compounds are being used historically, which are toxic in nature. To protect environment it has become necessary to develop non-toxic lead free shielding materials. The use of titanium metal and its compounds as synthetic rock i.e. SYNROC are reported to be very effective non-toxic shielding materials for various applications. Red mud waste generated in aluminum producing industries possesses a unique mineralogical compositions containing fairly high quantity of titanium oxide and iron oxide useful for making non toxic shielding compositions and therefore red mud has been utilized for the first time in the world for making radiation shielding materials. The red mud based compositions developed have been characterized for their various physico-mechanical properties namely compressive strength, impact strength, density and X-ray and gamma radiation shielding capacity in terms of shielding thickness i.e. HVT. Based on the characterization results it is found that the red mud based materials can be used for the construction of X-ray diagnostic and CT-Scanner room and as a substitute shielding material for concrete in the nuclear reactors and other radiation based applications. Studies on the identification of shielding phases and their morphology present, in the red mud based shielding compositions has been carried out using X-ray diffraction and SEM technique. The results of these studies are presented in this paper. (authors)

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

    International Nuclear Information System (INIS)

    Wilson, M.T.

    1975-01-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    CERN Document Server

    CERN Library

    2013-01-01

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

  20. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ingersoll, D T [comp.; Oak Ridge National Lab., TN (United States); Ingersoll, J K [comp.; Tec-Com, Knoxville, TN (United States)

    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.

  2. 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)

  3. Radiation shielding properties of high performance concrete reinforced with basalt fibers infused with natural and enriched boron

    Energy Technology Data Exchange (ETDEWEB)

    Zorla, Eyüp; Ipbüker, Cagatay [University of Tartu, Institute of Physics (Estonia); Biland, Alex [US Basalt Corp., Houston (United States); Kiisk, Madis [University of Tartu, Institute of Physics (Estonia); Kovaljov, Sergei [OÜ Basaltest, Tartu (Estonia); Tkaczyk, Alan H. [University of Tartu, Institute of Physics (Estonia); Gulik, Volodymyr, E-mail: volodymyr.gulik@gmail.com [Institute for Safety Problems of Nuclear Power Plants, Lysogirska 12, of. 201, 03028 Kyiv (Ukraine)

    2017-03-15

    Highlights: • Basalt fiber infused with natural and enriched boron in varying proportions. • Gamma-ray attenuation remains stable with addition of basalt-boron fiber. • Improvement in neutron shielding for nuclear facilities producing fast fission spectrum. • Basalt-boron fiber could decrease the shielding thickness in thermal spectrum reactors. - Abstract: The importance of radiation shielding is increasing in parallel with the expansion of the application areas of nuclear technologies. This study investigates the radiation shielding properties of two types of high strength concrete reinforced with basalt fibers infused with 12–20% boron oxide, containing varying fractions of natural and enriched boron. The gamma-ray shielding characteristics are analyzed with the help of the WinXCom, whereas the neutron shielding characteristics are modeled and computed by Monte Carlo Serpent code. For gamma-ray shielding, the attenuation coefficients of the studied samples do not display any significant variation due to the addition of basalt-boron fibers at any mixing proportion. For neutron shielding, the addition of basalt-boron fiber has negligible effects in the case of very fast neutrons (14 MeV), but it could considerably improve the neutron shielding of concrete for nuclear facilities producing a fast fission spectrum (e.g. with reactors as BN-800, FBTR) and thermal neutron spectrum (Light Water Reactors (LWR)). It was also found that basalt-boron fiber could decrease the thickness of radiation shielding material in thermal spectrum reactors.

  4. 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...

  5. TFTR radiation contour and shielding efficiency measurements during D-D operations

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, H.W.; Ascione, G.; Elwood, S.; Gilbert, J.; Hwang, D.; Lewis, M.; Levine, J.; Ku, L.P.; Rule, K. [Princeton Plasma Physics Lab., NJ (United States); Hajnal, F. [Department of Energy, New York, NY (United States)] [and others

    1994-11-01

    Extensive neutron and gamma radiation contour, shielding efficiency, and spectral measurements were performed during high power TFTR D-D operations at the tokamak Test Cell inner walls, ceiling, roof, and outer walls, in nearby control rooms, work areas, and personnel pathways, outdoors along the site fence at 125 m, and out to the nearest property lines at 180 m. The results confirmed that the efficiency of the basic radiation shielding was sufficient to allow the TFTR D-T experimental plan, and provide empirical guidance for simulating the radiation fields of future fusion reactors.

  6. Gamma ray absorption of cylindrical fissile material with dual shields

    International Nuclear Information System (INIS)

    Wu Chenyan; Cheng Yiying; Huang Yongyi; Lu Fuquan; Yang Fujia

    2005-01-01

    This work analyzed the gamma ray attenuation effect from the self-absorption and shield attenuation perspectively. An exact mathematical equation was given for the geometric factor of the cylindrical fissile material with dual shields. In addition, several approximation approaches suitable for real situation were discussed, especially in the radial and axial directions of the cylinders, since the G-factors have simple forms. Then the space distribution patterns of the G-factor were analyzed based on numerical result and effective ways to solved the geometric information of the cylindrical fissile material, the radii and the heights, were deduced. This method was checked and verified by numerical calculation. Because of the efficiency of the method, it is ideal for application in real situations, such as nuclear safeguards, which demands speed of detection and accuracy of geometric analysis. (authors)

  7. Shielding technology for high energy radiation production facility

    International Nuclear Information System (INIS)

    Lee, Byung Chul; Kim, Heon Il

    2004-06-01

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

  8. Mercuric Iodide Anticoincidence Shield for Gamma-Ray Spectrometer, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — We propose to utilize a new detector material, polycrystalline mercuric iodide, for background suppression by active anticoincidence shielding in gamma-ray...

  9. Mercuric Iodide Anticoincidence Shield for Gamma-Ray Spectrometer, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — We utilize a new detector material, polycrystalline mercuric iodide, for background suppression by active anticoincidence shielding in gamma-ray spectrometers. Two...

  10. Investigation on gamma and neutron radiation shielding parameters for BaO/SrO‒Bi2O3‒B2O3 glasses

    Science.gov (United States)

    Sayyed, M. I.; Lakshminarayana, G.; Dong, M. G.; Ersundu, M. Çelikbilek; Ersundu, A. E.; Kityk, I. V.

    2018-04-01

    In this work, mass attenuation coefficients (μ/ρ), effective atomic number (Zeff), electron density (Ne), mean free path (MFP), and half-value layer (HVL) of 20 BaO/SrO‒(x) Bi2O3‒(80‒x) B2O3 glasses (where x=10, 20, 30, 40, 50 and 60 mol%) were calculated using WinXCom program and MCNP5 code. The obtained (μ/ρ) results using both MCNP5 code and WinXCom program were in good agreement. It is found that the addition of Bi2O3 leads to increase the Zeff values in both BaO/SrO‒Bi2O3‒B2O3 glass systems. However, the Zeff values of the BaO‒Bi2O3‒B2O3 glass system are higher than those of the SrO‒Bi2O3‒B2O3 glasses. The fast neutrons effective removal cross sections (ΣR) for 20 SrO‒40 Bi2O3‒40 B2O3 glass is the highest among all studied glasses. The calculated half-value layer values were compared with different glass systems and it was found that the shielding properties of the selected glasses are comparable or even better than other glass systems such as phosphate glasses.

  11. GammaCam trademark radiation imaging system

    International Nuclear Information System (INIS)

    1998-02-01

    GammaCam trademark, a gamma-ray imaging system manufactured by AIL System, Inc., would benefit a site that needs to locate radiation sources. It is capable of producing a two-dimensional image of a radiation field superimposed on a black and white visual image. Because the system can be positioned outside the radiologically controlled area, the radiation exposure to personnel is significantly reduced and extensive shielding is not required. This report covers the following topics: technology description; performance; technology applicability and alternatives; cost; regulatory and policy issues; and lessons learned. The demonstration of GammaCam trademark in December 1996 was part of the Large-Scale Demonstration Project (LSDP) whose objective is to select and demonstrate potentially beneficial technologies at the Argonne National Laboratory-East (ANL) Chicago Pile-5 Research Reactor (CP-5). The purpose of the LSDP is to demonstrate that by using innovative and improved decontamination and decommissioning (D and D) technologies from various sources, significant benefits can be achieved when compared to baseline D and D technologies

  12. Recent trends in radiation shielding: a RSIC perspective

    International Nuclear Information System (INIS)

    Trubey, D.K.; Roussin, R.W.; Maskewitz, B.F.

    1979-01-01

    The subject of radiation transport and shielding in the nuclear power industry is reviewed, and advances in the state of the art are described. These fall into the areas of computational methods, nuclear cross sections, industry practices, and standards. Computer codes and data available from the Radiation Shielding Information Center (RSIC) representing recent advances are also described

  13. Radiation shielding phenolic fibers and method of producing same

    International Nuclear Information System (INIS)

    Ohtomo, K.

    1976-01-01

    A radiation shielding phenolic fiber is described comprising a filamentary phenolic polymer consisting predominantly of a sulfonic acid group-containing cured novolak resin and a metallic atom having a great radiation shielding capacity, the metallic atom being incorporated in the polymer by being chemically bound in the ionic state in the novolak resin. A method for the production of the fiber is discussed

  14. Gamma irradiators for radiation processing

    International Nuclear Information System (INIS)

    2006-01-01

    Radiation technology is one of the most important fields which the IAEA supports and promotes, and has several programmes that facilitate its use in the developing Member States. In view of this mandate, this Booklet on 'Gamma Irradiators for Radiation Processing' is prepared which describes variety of gamma irradiators that can be used for radiation processing applications. It is intended to present description of general principles of design and operation of the gamma irradiators available currently for industrial use. It aims at providing information to industrial end users to familiarise them with the technology, with the hope that the information contained here would assist them in selecting the most optimum irradiator for their needs. Correct selection affects not only the ease of operation but also yields higher efficiency, and thus improved economy. The Booklet is also intended for promoting radiation processing in general to governments and general public

  15. CHESS upgrade 1995: Improved radiation shielding

    International Nuclear Information System (INIS)

    Finkelstein, K.

    1996-01-01

    The Cornell Electron Storage Ring (CESR) stores electrons and positrons at 5.3 GeV for the production and study of B mesons, and, in addition, it supplies synchrotron radiation for CHESS. The machine has been upgraded for 300 mA operation. It is planned that each beam will be injected in about 5 minutes and that particle beam lifetimes will be several hours. In a cooperative effort, staff members at CHESS and LNS have studied sources in CESR that produce radiation in the user areas. The group has been responsible for the development and realization of new tunnel shielding walls that provide a level of radiation protection from 20 to approx-gt 100 times what was previously available. Our experience has indicated that a major contribution to the environmental radiation is not from photons, but results from neutrons that are generated by particle beam loss in the ring. Neutrons are stopped by inelastic scattering and absorption in thick materials such as heavy concrete. The design for the upgraded walls, the development of a mix for our heavy concrete, and all the concrete casting was done by CHESS and LNS personnel. The concrete incorporates a new material for this application, one that has yielded a significant cost saving in the production of over 200 tons of new wall sections. The material is an artificially enriched iron oxide pellet manufactured in vast quantities from hematite ore for the steel-making industry. Its material and chemical properties (iron and impurity content, strength, size and uniformity) make it an excellent substitute for high grade Brazilian ore, which is commonly used as heavy aggregate in radiation shielding. Its cost is about a third that of the natural ore. The concrete has excellent workability, a 28 day compressive strength exceeding 6000 psi and a density of 220 lbs/cu.ft (3.5 gr/cc). The density is limited by an interesting property of the pellets that is motivated by efficiency in the steel-making application. (Abstract Truncated)

  16. 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

  17. High Density Radiation Shielding Concretes for Hot Cells of 99mTc Project

    International Nuclear Information System (INIS)

    Sakr, K.

    2006-01-01

    High density concrete [more than 3.6 ton/m 3 (3.6x10 3 kg/m 3 )] was prepared to be used as a radiation shielding concrete (RSC) for hot-cells in gel technetium project at inshas to attenuate gamma radiation emitted from radioactive sources. different types of concrete were prepared by mixing local mineral aggregates mainly gravel and ilmenite . iron shots were added to the concrete mixture proportion as partial replacement of heavy aggregates to increase its density. the physical properties of prepared concrete in both plastic and hardened phases were investigated. compressive strength and radiation attenuation of gamma rays were determined. Results showed that ilmenite concrete mixed with iron shots had the highest density suitable to be use as RSC according to the chinese hot cell design requirements. Recommendations to avoid some technical problems of manufacturing radiation shielding concrete were maintained

  18. Neutron and gamma ray transport calculations in shielding system

    Energy Technology Data Exchange (ETDEWEB)

    Masukawa, Fumihiro; Sakamoto, Hiroki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    In the shields for radiation in nuclear facilities, the penetrating holes of various kinds and irregular shapes are made for the reasons of operation, control and others. These penetrating holes and gaps are filled with air or the substances with relatively small shielding performance, and radiation flows out through them, which is called streaming. As the calculation techniques for the shielding design or analysis related to the streaming problem, there are the calculations by simplified evaluation, transport calculation and Monte Carlo method. In this report, the example of calculation by Monte Carlo method which is represented by MCNP code is discussed. A number of variance reduction techniques which seem effective for the analysis of streaming problem were tried. As to the investigation of the applicability of MCNP code to streaming analysis, the object of analysis which are the concrete walls without hole and with horizontal hole, oblique hole and bent oblique hole, the analysis procedure, the composition of concrete, and the conversion coefficient of dose equivalent, and the results of analysis are reported. As for variance reduction technique, cell importance was adopted. (K.I.)

  19. Slow neutrons and secondary gamma ray distributions in concrete shields followed by reflecting layers

    International Nuclear Information System (INIS)

    Makarious, A.S.; Swilem, Y.I.; Awwad, Z.; Bayomy, T.

    1993-01-01

    Slow neutrons and secondary gamma ray distributions in concrete shields with and without a reflecting layer behind layer behind the concrete shield have been investigated first in case of using a bare reactor beam and then on using a B-4 C filtered beam. The total and capture secondary gamma ray coefficient (B gamma and B gamma C ), the ratio of the reflected thermal neutron (gamma) the ratio of the secondary gamma rays caused by reflected neutrons to those caused transmitted neutrons (Th I gamma/F I gamma) and the effect of inserting a blocking layer (a B-4 C layer) between the concrete shield and the reflector on the suppression of the produced secondary gamma rays have been investigated. It was found that the presence of the reflector layer behind the concrete shield reflects some thermal neutrons back to the concrete shields and so it increases the number of thermal neutrons at the interface between the concrete shield and the reflector. Also the capture secondary gamma rays was increased at the interface between the two medii due to the capture of the reflected thermal neutrons in the concrete shields. It was shown that B-gamma is higher than and that B g amma B gamma C and I gamma T h/ I gamma i f for the different concrete types is higher in case of using the graphite reflector than that in using either water or paraffin reflectors. Putting a blocking layer (B 4 C layer) between the concrete shield and the reflector decreases the produced secondary gamma rays due to the absorption of the reflected thermal neutrons. 17 figs

  20. Radiation shielding design for irradiation facilities

    International Nuclear Information System (INIS)

    Ito, Kazuo

    1974-01-01

    The effective atomic numbers of various kinds of concrete were experimentally determined. The effective radiation absorption coefficient μsub(M)(cm -1 ) was determined at first. μsub(M) and the theoretical absorption coefficient μsub(T) are given by μsub(M)=(1/x)2.31og(Isub(o)/I) and μsub(T)=[tausub(a)+sigmasub(A)+kappasub(a)]. rhoNsub(o)/A, respectively, where Isub(o) is the intensity of incoming gamma photon, I is the intensity of the gamma photon after transmission, x is the thickness of a mixture, rho is the density of the mixture, A is the equivalent atomic weight of the mixture, Nsub(o) is the Avogadro number, and tausub(a), sigmasub(a), and kappasub(a) are the cross sections in cm 2 /atom of photoelectric effect, Compton scattering, and electron pair formation. The effective atomic number Z is the one that makes the difference between μsub(M) and μsub(T) minimum. The effective atomic numbers thus determined were 11.3, 34.8, 18.9, 18.3, and 22.8 for ordinary concrete, baryte concrete, magnetite concrete, boron magnetite concrete, and magnetite ironball concrete, respectively. The values of specific gravity of these concretes were 2.35, 3.60, 3.89, 3.71, and 5.12, respectively. Monte Carlo simulation was made on various kinds of concrete by means of the use of the effective atomic numbers. The energy of incoming gamma ray was 1.25 MeV. The thickness of the concretes was 10 cm. The intensity of both scattered and transmitted photons of gamma ray is discussed. (Fuktomi, T.)

  1. Investigation of ionizing radiation shielding effectiveness of decorative building materials used in Bangladeshi dwellings

    Science.gov (United States)

    Yesmin, Sabina; Sonker Barua, Bijoy; Uddin Khandaker, Mayeen; Tareque Chowdhury, Mohammed; Kamal, Masud; Rashid, M. A.; Miah, M. M. H.; Bradley, D. A.

    2017-11-01

    Following the rapid growing per capita income, a major portion of Bangladeshi dwellers is upgrading their non-brick houses by rod-cement-concrete materials and simultaneously curious to decorate the houses using luxurious marble stones. Present study was undertaken to investigate the gamma-ray attenuation co-efficient of decorative marble materials leading to their suitability as shielding of ionizing radiation. A number of commercial grades decorative marble stones were collected from home and abroad following their large-scale uses. A well-shielded HPGe γ-ray spectrometer combined with associated electronics was used to evaluate the mass attenuation coefficients of the studied materials for high energy photons. Some allied parameters such as half-value layer and radiation protection efficacy of the investigated marbles were calculated. The results showed that among the studied samples, the marble 'Carrara' imported from Italy is suitable to be used as radiation shielding material.

  2. Electrically nonconductive shield for electric equipment generating ionizing radiation

    International Nuclear Information System (INIS)

    Aitken, D.

    1979-01-01

    As a radiation protection shield there is proposed a nonconductive shield fabricated from epoxides or other plastics material and containing finely dispersed radiation absorbing metal. It is to be designed in such a way that it lies in the range of a high electric gradient in the equipment, close to the radiation-producing component. As suitable metals there are mentioned tin, tungsten, and lead resp. their oxides. As an example there is used an X-ray shielding. (RW) 891 RW/RW 892 MKO [de

  3. Gamma irradiators for radiation sterilization

    International Nuclear Information System (INIS)

    Mehta, K.

    2008-01-01

    The radiation processing industry gained significant impetus with the advent of nuclear reactors, which have the capability to produce radioisotopes such as 60 Co. These gamma ray emitters became popular radiation sources for medical and industrial applications. Many gamma ray irradiators have been built, 200 of which are estimated to be currently in operation in Member States of the IAEA. In recent times, the use of electron accelerators as radiation source (and sometimes equipped with an X ray converter) is increasing. However, gamma irradiators are difficult to replace, especially for non-uniform and high density products. Currently, 60 Co is used almost solely as a gamma radiation source for industrial use now, mainly because of its easy production method and its non-solubility in water. Based on the total cumulative sale of 60 Co by all suppliers, it can be estimated that the installed capacity of cobalt is increasing at the rate of about 6% per year. It is interesting to note that the worldwide use of disposable medical devices is growing at approximately the same rate (5-6%), which seems to be driving the growth in cobalt sale

  4. Radiation shielding for 250 MeV protons

    Energy Technology Data Exchange (ETDEWEB)

    Awschalom, M.

    1987-04-01

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

  5. Radiation shielding for 250 MeV protons

    International Nuclear Information System (INIS)

    Awschalom, M.

    1987-01-01

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

  6. Investigating Radiation Shielding Properties of Different Mineral Origin Heavyweight Concretes

    Science.gov (United States)

    Basyigit, Celalettin; Uysal, Volkan; Kilinçarslan, Şemsettin; Mavi, Betül; Günoǧlu, Kadir; Akkurt, Iskender; Akkaş, Ayşe

    2011-12-01

    The radiation although has hazardous effects for human health, developing technologies bring lots of usage fields to radiation like in medicine and nuclear power station buildings. In this case protecting from undesirable radiation is a necessity for human health. Heavyweight concrete is one of the most important materials used in where radiation should be shielded, like those areas. In this study, used heavyweight aggregates of different mineral origin (Limonite, Siderite), in order to prepare different series in concrete mixtures and investigated radiation shielding properties. The experimental results on measuring the radiation shielding, the heavyweight concrete prepared with heavyweight aggregates of different mineral origin show that, are useful radiation absorbents when they used in concrete mixtures.

  7. Investigating Radiation Shielding Properties of Different Mineral Origin Heavyweight Concretes

    International Nuclear Information System (INIS)

    Basyigit, Celalettin; Uysal, Volkan; Kilincarslan, Semsettin; Akkas, Ayse; Mavi, Betuel; Guenoglu, Kadir; Akkurt, Iskender

    2011-01-01

    The radiation although has hazardous effects for human health, developing technologies bring lots of usage fields to radiation like in medicine and nuclear power station buildings. In this case protecting from undesirable radiation is a necessity for human health. Heavyweight concrete is one of the most important materials used in where radiation should be shielded, like those areas. In this study, used heavyweight aggregates of different mineral origin (Limonite, Siderite), in order to prepare different series in concrete mixtures and investigated radiation shielding properties. The experimental results on measuring the radiation shielding, the heavyweight concrete prepared with heavyweight aggregates of different mineral origin show that, are useful radiation absorbents when they used in concrete mixtures.

  8. 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.

  9. Experimental investigation on the effect of high Z front layer on saturation thickness for radiation shielding materials

    Science.gov (United States)

    Purkayastha, Biswajit

    Measurement of number and energy albedo of backscattered gamma rays has wide applications in the design of gamma ray shield involving the use of nuclear radiations. In these applications the characteristic parameter which gives an integral measurement of gamma ray scattering is the albedo of the material involved. Albedo measurements were initiated at the end of the fifth decade of the present century, just after the 2nd World War. Large number of workers worked in different fields such as (i) measurement of angular and energy distributions of backscattered in photons by applying Monte Carlo technique, (ii) experimental investigations of backscattering gamma rays by scintillation spectrometer, Uniform Sensitivity Photon Counter, Proportional Response Photon Counter etc. However, the literature on the effect of high atomic number (Z) front layer on the saturation thickness of a gamma radiation shield is almost nil. In view of the above discussions attempts have been made in the present investigation to extend the scope for optimisation of saturation thickness of a radiation shield by placing a high Z front layer material of varying thickness. Measurements of energy albedo for a homogeneous shield with and without a front layer of different high Z material of varying thickness for two energies namely 662 keV and 1250 keV have been carried out and a careful analysis is presented in the present investigation. From a detailed analysis of the present investigation the following conclusions can be drawn : (i) Anisotroplc distributions of photons as function of emerging angle. (ii) Total energy albedo values of different shielding materials decrease when high Z material is used in front of the shielding material. (iii) Saturation thickness of scattering combinations depends on Z values of the front layer. (iv) The total energy albedo values decrease exponentially as the increase in thickness of the high Z material of the front layer. (v) This experiment not only throws

  10. Multifunctional, Boron-Foam Based Radiation Shielding Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The NASA vision of Space Exploration requires new approaches to radiation shielding. Both Spiral 2 and Spiral 3 concepts are extremely sensitive to weight reduction....

  11. Characterizing and Manufacturing Multifunctional Radiation Shielding Materials, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This project addresses two vital problems for long-term space travel activities: radiation shielding and hydrogen storage for power and propulsion. While both...

  12. Magnetic Active Radiation Shielding System Using Helmholtz Coil Lattices

    Data.gov (United States)

    National Aeronautics and Space Administration — The objective of this proposal is to investigate whether a magnetic active radiation shielding systems can be designed from an optics perspective, where Helmholtz...

  13. Efficient Radiation Shielding Through Direct Metal Laser Sintering

    Data.gov (United States)

    National Aeronautics and Space Administration — We have developed a method for efficient component-level radiation shielding that can be printed by direct metal laser sintering (DMLS) from files generated by the...

  14. Radiation Shielding and Hydrogen Storage with Multifunctional Carbon, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This project addresses two vital problems for long-term space travel activities: radiation shielding and hydrogen storage for power and propulsion. While both...

  15. Improved Metal-Polymeric Laminate Radiation Shielding, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — In this proposed Phase II program, builds on the phase I feaibility where a multifunctional lightweight radiation shield composite was developed and fabricated. This...

  16. Mechanical and radiation shielding properties of mortars with additive fine aggregate mine waste

    International Nuclear Information System (INIS)

    Gallala, Wissem; Hayouni, Yousra; Gaied, Mohamed Essghaier; Fusco, Michael; Alsaied, Jasmin; Bailey, Kathryn; Bourham, Mohamed

    2017-01-01

    Highlights: • Effectiveness of mine waste as additive fine aggregate has been investigated. • Experimental results are verified by computationally from composition of synthesized samples. • Work focuses on shielding materials for nuclear systems including spent fuel storage and drycasks. - Abstract: Incorporation of barite-fluorspar mine waste (BFMW) as a fine aggregate additive has been investigated for its effect on the mechanical and shielding properties of cement mortar. Several mortar mixtures were prepared with different proportions of BFMW ranging from 0% to 30% as fine aggregate replacement. Cement mortar mixtures were evaluated for density, compressive and tensile strengths, and gamma ray radiation shielding. The results revealed that the mortar mixes containing 25% BFMW reaches the highest compressive strength values, which exceeded 50 MPa. Evaluation of gamma-ray attenuation was both measured by experimental tests and computationally calculated using MicroShield software package, and results have shown that using BFMW aggregates increases attenuation coefficient by about 20%. These findings have demonstrated that the mine waste can be suitably used as partial replacement aggregate to improve radiation shielding as well as to reduce the mortar and concrete costs.

  17. 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.)

  18. Passive radiation shielding considerations for the proposed space elevator

    Science.gov (United States)

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

    2007-02-01

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

  19. Review of diagnostic methods for TFTR D{endash}T radiation shielding and neutronics studies

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, H.W.; Ascione, G.; Gilbert, J. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States); Azziz, N.; Goldhagen, P.; Reginatto, M.; Shebell, P. [U.S. Department of Energy Environmental Measurements Laboratory, New York, New York 10014-4811 (United States); Kumar, A. [School of Engineering and Applied Science, University of California at Los Angeles, Los Angeles, California 90095 (United States)

    1997-01-01

    The methods and instrument systems used for TFTR D{endash}T radiation shielding and neutronics studies involving signal strengths ranging over 10 orders of magnitude are reviewed. Neutron and gamma dose-equivalent, fluence, spectral, and materials activation measurements have been performed at various locations from the TFTR vessel to the nearest property lines. The detection systems include {sup 3}He, BF{sub 3}, and {sup 235}U proportional counters in moderated spheres, Bonner sphere arrays, advanced thermoluminescent detectors, argon ionization chambers, intrinsic Ge gamma detectors, and activation foil spectrometry methods. {copyright} {ital 1997 American Institute of Physics.}

  20. Review of diagnostic methods for TFTR D-T radiation shielding and neutronics studies

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, H.W.; Ascione, G.; Gilbert, J. [Princeton Univ., NJ (US). Princeton Plasma Physics Lab.; Azziz, N.; Goldhagen, P.; Reginatto, M.; Shebell, P. [Dept. of Energy, New York, NY (US). Environmental Measurements Lab.; Kumar, A. [Univ. of California, Los Angeles, CA (US). School of Engineering and Applied Science

    1996-10-01

    The methods and instrument systems used for TFTR D-T radiation shielding and neutronics studies involving signal strengths ranging over 10 orders of magnitude are reviewed. Neutron and gamma dose-equivalent, fluence, spectral, and materials activation measurements have been performed at various locations from the TFTR vessel to the nearest property lines. The detection systems include {sup 3}He, BF{sub 3}, and {sup 235}U proportional counters in moderated spheres, Bonner sphere arrays, advanced thermoluminescent detectors, argon ionization chambers, intrinsic Ge gamma detectors, and activation foil spectrometry methods.

  1. A comprehensive study of the energy absorption and exposure buildup factors of different bricks for gamma-rays shielding

    Directory of Open Access Journals (Sweden)

    M.I. Sayyed

    Full Text Available The present investigation has been performed on different bricks for the purpose of gamma-ray shielding. The values of the mass attenuation coefficient (µ/ρ, energy absorption buildup factor (EABF and exposure buildup factor (EBF were determined and utilized to assess the shielding effectiveness of the bricks under investigation. The mass attenuation coefficients of the selected bricks were calculated theoretically using WinXcom program and compared with MCNPX code. Good agreement between WinXcom and MCNPX results was observed. Furthermore, the EABF and EBF have been discussed as functions of the incident photon energy and penetration depth. It has been found that the EABF and EBF values are very large in the intermediate energy region. The steel slag showed good shielding properties, consequently, this brick is eco-friendly and feasible compared with other types of bricks used for construction. The results in this work should be useful in the construction of effectual shielding against hazardous gamma-rays. Keywords: Brick, Mass attenuation coefficient, Buildup factor, G-P fitting, Radiation shielding

  2. Up-dating of the RA-0 reactor shielding. Gamma and neutron isodoses

    International Nuclear Information System (INIS)

    Murua, Carlos A.; Chautemps, Norma A.; Ackerley, Alejandro F.; Alexeiew, Vladimiro

    1999-01-01

    A comparative analysis of the historical shielding configurations of the RA-0 reactor is performed and the comparison methodology is described. The gamma and neutron dose mapping of the last two stages of the reactor shielding has been carried out and the results are analysed

  3. A Reinforcement for Multifunctional Composites for Non-Parasitic Radiation Shielding, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Innovative lightweight radiation shielding materials are enabling to shield humans in aerospace transportation vehicles and other human habited spaces....

  4. Radiation shielding rubber blend and radiation ray shielding packing material molded therefrom

    International Nuclear Information System (INIS)

    Takechi, Sanjuro; Urakami, Masamichi.

    1996-01-01

    The present invention concerns a flexible radiation shielding packing material having a good molding fabricability in a not cross-linked state, and having excellent tensile strength, stretchability and tearing resistance. The blend is formed by blending ethylene/propylene rubber (EPR (copolymer of ethylene and propylene), EPDM (terpolymer of ethylene, propylene and a slight amount of dienic ingredient)) and isopropylene rubber (IR) with lead oxide. The blending ratio is preferably from 30 to 100 parts by weight of a lead oxide powder based on 100 parts by weight of a rubber comprising ethylene/propylene rubber and isopropylene rubber at a weight ratio of from 3:7 to 7:3. If ethylene/propylene rubber is EPDM, it is more preferable. (T.M.)

  5. Investigation of gamma-ray shielding effectiveness of natural marble used for external wall cladding of buildings in Riyadh, Saudi Arabia

    Directory of Open Access Journals (Sweden)

    Ibrahim F. Al-Hamarneh

    Full Text Available Gamma-ray shielding effectiveness of different types of natural marble tiles commonly used for cladding the exterior walls of residential and non-residential buildings in Riyadh, Saudi Arabia has been investigated in the energy range 59.5–1332.5 keV. To this end, linear attenuation coefficients (μℓ have been obtained by applying narrow-beam technique. Mass attenuation coefficient (μm, transmission factor (TF and half-value layer (HVL have been employed to study the shielding performance of marbles and also compared with those of calcite. The experimental findings revealed that HVL parameter correctly determined the shielding effectiveness of the dissimilar types of marble. Compared with other marbles, Carrara marble (Italy showed preferable shielding effectiveness. Moreover, a formula, μm = aE−b, was proposed to evaluate and compare the shielding properties of the marble tiles over a broad energy range. In this formula, a and b were determined empirically. In conclusion, natural marble proves to be more reliable in reducing gamma radiation when used for external building cladding than ordinary concrete, and it could be a good alternative to lead shield against high gamma energy because it is 24% heavier than lead. Keywords: Gamma-ray shielding effectiveness, Attenuation coefficients, HVL, Transmission factor, Natural marble

  6. Gamma radiation effect to prostaglandin

    International Nuclear Information System (INIS)

    Coelho, Fernando Rodrigues; Lima, Wothan Tavares de; Rogero, Sizue Ota; Lugao, Ademar Benevolo

    2005-01-01

    Prostaglandins and their analogs are of great physiological importance used to prepare drugs by pharmaceutical industry. But the resistance to radiation sterilization process is not too much studied. This work had the objective of study the relaxation activity of irradiated prostaglandin type E1 on the muscle of respiratory tract. 1% HPMC prostaglandin dried dispersion was submitted to radiation from Co-60 gamma source with 10 kGy/h dose rate at 0, 50, 75 e 100 kGy doses. After irradiation degradation measurement was performed by HPLC analysis and the biological activity by in vitro assay of relaxation activity of muscle, in trachea isolated from rats. The results showed in the maximum radiation dose (]100 kGy) about 5% loss of prostaglandin relaxation activity and degradation of about 30% in relation to non irradiated sample. Prostaglandin dispersion in HPMC can be considered steady after irradiation in the dose used for medical products sterilization. (author)

  7. Calculation analysis of the thickness of radiation shield for the RIA equipment IP10

    International Nuclear Information System (INIS)

    Benar Bukit; Kristiyanti; Hari Nurcahyadi

    2011-01-01

    Calculation Analysis has been performed on the thickness of radiation shield for the design of the Radioimmunoassay (RIA) IP10 counters using five detectors arranged in parallel. The calculation is intended to ensure that the radiation on each detector does not influence each other. The radiation shield is made of lead. The calculation of lead thickness was based on the principle of the lead plates absorptive power toward the gamma ray of a certain energy. which is the function of linear absorption coefficient and the material thickness. Assuming the use of Iodium-125(I-125) source with an activity 10 µCi, and expecting an absorptive power of 95%, calculations showed that the required lead thickness is equal to 0,013 cm. Since lead is soft and its availability in the market is limited, lead plate of 2 mm thickness are used instead, so that counting result for the detectors do not influence each other. (author)

  8. Elevated gamma-rays shielding property in lead-free bismuth tungstate by nanofabricating structures

    Science.gov (United States)

    Liu, Jun-Hua; Zhang, Quan-Ping; Sun, Nan; Zhao, Yang; Shi, Rui; Zhou, Yuan-Lin; Zheng, Jian

    2018-01-01

    Radiation shielding materials have attracted much attention across academia and industry because of the increasing of nuclear activities. To achieve the materials with low toxicity but good protective capability is one of the most significant goals for personal protective articles. Here, bismuth tungstate nanostructures are controllably fabricated by a versatile hydrothermal treatment under various temperatures. The crystals structure and morphology of products are detailedly characterized with X-ray diffraction, electron microscope and specific surface area. It is noteworthy that desired Bi2WO6 nanosheets treated with 190 °C show the higher specific surface area (19.5 m2g-1) than that of the other two products. Importantly, it has a close attenuating property to lead based counterpart for low energy gamma-rays. Due to the less toxicity, Bi2WO6 nanosheets are more suitable than lead based materials to fabricate personal protective articles for shielding low energy radiations and have great application prospect as well as market potential.

  9. Correlation of gamma ray shielding and structural properties of PbO–BaO–P{sub 2}O{sub 5} glass system

    Energy Technology Data Exchange (ETDEWEB)

    Kaur, Kulwinder; Singh, K.J., E-mail: kanwarjitsingh@yahoo.com; Anand, Vikas

    2015-04-15

    Highlights: • Transparent glass samples of the system 55PbO{sub x}BaO(45 − x)P{sub 2}O{sub 5} (x = 1 up to 5) have been prepared in the laboratory. • Gamma ray shielding properties improve with the addition of BaO. • Number of non-bridging oxygens decrease with the increase in the content of BaO. • Investigated glass system can be potential candidate as an alternate to conventional radiation shielding ‘concrete’. - Abstract: The presented work has been undertaken to evaluate the applicability of BaO doped PbO-P{sub 2}O{sub 5} glass system as gamma ray shielding material in terms of mass attenuation coefficient and half value layer at photon energies 662, 1173 and1332 keV. A meaningful comparison of their radiation shielding properties has been made in terms of their mass attenuation coefficient and HVL parameters with standard radiation shielding concrete ‘barite’. The density, molar volume, XRD, FTIR, Raman and UV–visible techniques and mechanical properties (by Yamane and Mackenzie's procedure) have been used to study the structural properties of the prepared glass system in order to check the possibility of their commercial utility as alternate to conventional concrete for gamma ray shielding applications.

  10. Radiation shielding design considerations for Doublet III

    International Nuclear Information System (INIS)

    Engholm, B.A.

    1980-06-01

    Calculations and measurements were made of the bremsstrahlung (x-ray) doses resulting from runaway electron shots at Doublet III. The analysis considered direct, wall-scattered, and skyshine contributions. Reasonably good agreement was obtained between calculations and measurements. The x-ray dose in the control room was about 1 mR per runaway shot, while that at the north boundary was undetectable, with a calculated value of 0.05 mR per shot. These low doses attest to the adequacy of the 2 ft concrete shadow shield surrounding the Doublet III room. Exploratory shielding analyses were performed for possible neutron generation if Doublet III were operated with neutral beam injection in an aggressive D-D mode

  11. Radiation shielding evaluation of vertical test stand facility at RRCAT

    International Nuclear Information System (INIS)

    Sahani, P.K.; Haridas, G.; Patel, Hemant Kumar; Kush, P.K.; Joshi, S.C.; Puntambekar, T.A.

    2015-01-01

    A vertical test stand facility (VTSF) for testing and characterizing super conducting Radio-Frequency (SCRF) cavities is set up at RRCAT, Indore. The test stand has the capacity to test multicell SCRF cavities at frequencies of 650 MHz and 1.3 GHz at liquid Helium temperature (2K). When cavity is powered, high electric field gradient up to 35 MV/m is generated inside the cavity. Because of this high electric field gradient, field emission within the cavity may produce high energy electrons followed by Bremsstrahlung emission and photo-neutrons. Within the cryostat of the VTSF, internal radiation shielding has been provided with 200 mm lead, 100 mm Steel and 100 mm borated polyethylene. Besides the internal shield, an external radiation shield is proposed to be provided at the top of the vertical pit to reduce radiation levels in the accessible areas to acceptable limits. Radiation dose outside the external shield, comprising of 150 mm steel and 480 mm concrete, due to Bremsstrahlung x-rays and photo-neutrons are simulated using the Monte-Carlo code, FLUKA. The geometry used for the simulation of dose outside the shield of VTSF setup showing the detectors and the simulated dose rates are given. The paper describes the details of the simulation and results. (author)

  12. Attenuation of neutrons and gamma-rays in homogeneous and multilayered shields

    International Nuclear Information System (INIS)

    Abdo, A.E.; Megahid, R.M.

    1997-01-01

    Measurements were carried-out to compare the attenuation properties of homogeneous shields and shields of two layers and three layers for fast neutrons and total gamma-rays. These were performed by measuring the fast neutron and total gamma-ray spectra behind homogeneous shields of magnetite-limonite, ilmenite-ilmenite and magnetite-magnetite concretes. The two layers assembly consists of iron and one of the above mentioned concretes, while the three layers shield consists of water, iron and one of the previously mentioned concretes. All measurements were carried-out using a neutron-gamma spectrometer with stilbene scintillator coupled to a fast photo multi player tube. Separation between pulses of recoil protons and recoil electrons was achieved by a pulse shape discrimination technique. 3 tabs., 10 figs., 13 refs

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  14. PMMA/MWCNT nanocomposite for proton radiation shielding applications

    Science.gov (United States)

    Li, Zhenhao; Chen, Siyuan; Nambiar, Shruti; Sun, Yonghai; Zhang, Mingyu; Zheng, Wanping; Yeow, John T. W.

    2016-06-01

    Radiation shielding in space missions is critical in order to protect astronauts, spacecraft and payloads from radiation damage. Low atomic-number materials are efficient in shielding particle-radiation, but they have relatively weak material properties compared to alloys that are widely used in space applications as structural materials. However, the issues related to weight and the secondary radiation generation make alloys not suitable for space radiation shielding. Polymers, on the other hand, can be filled with different filler materials for reinforcement of material properties, while at the same time provide sufficient radiation shielding function with lower weight and less secondary radiation generation. In this study, poly(methyl-methacrylate)/multi-walled carbon nanotube (PMMA/MWCNT) nanocomposite was fabricated. The role of MWCNTs embedded in PMMA matrix, in terms of radiation shielding effectiveness, was experimentally evaluated by comparing the proton transmission properties and secondary neutron generation of the PMMA/MWCNT nanocomposite with pure PMMA and aluminum. The results showed that the addition of MWCNTs in PMMA matrix can further reduce the secondary neutron generation of the pure polymer, while no obvious change was found in the proton transmission property. On the other hand, both the pure PMMA and the nanocomposite were 18%-19% lighter in weight than aluminum for stopping the protons with the same energy and generated up to 5% fewer secondary neutrons. Furthermore, the use of MWCNTs showed enhanced thermal stability over the pure polymer, and thus the overall reinforcement effects make MWCNT an effective filler material for applications in the space industry.

  15. A Novel Radiation Shielding Material, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — In order to safely explore space, humans must be protected from radiation. There are 2 predominant sources of extraterrestrial ionizing radiation, namely, Galactic...

  16. CAD-based radiation protection and shielding in space

    International Nuclear Information System (INIS)

    Appleby, M.H.

    1991-01-01

    In the not-too-distant future, astronauts will begin living and working on space station Freedom (SSF), eventually establishing a permanent presence in space. Beyond Freedom, the National Aeronautics and Space Administration (NASA) has set its sights on returning to and eventually establishing outposts on the moon and Mars. Without appropriate methods of identifying protection deficiencies, spacecraft designers often overestimate or defer shielding solutions in both cases burdening the program. To avoid possible penalties such as increased mass, complexity, and cost, radiation analysis should be conducted as part of the preliminary design process. An innovative radiation assessment system combining computer-aided design (CAD) capabilities with established NASA transport codes has been developed permitting fast, accurate analysis of spacecraft. The use of this automated analytical tool the Boeing Radiation Exposure Model (Brem) is discussed in this paper, relative to spacecraft design and the optimization of radiation shielding. Results obtained from recently completed radiation analysis of space station Freedom are also discussed

  17. Influence on cell proliferation of background radiation or exposure to very low, chronic gamma radiation. [Paramecium tetraurelia; Synechococcus lividus

    Energy Technology Data Exchange (ETDEWEB)

    Planel, H.; Soleilhavoup, J.P.; Tixador, R.; Richoilley, G.; Conter, A.; Croute, F.; Caratero, C.; Gaubin, Y.

    1987-05-01

    Investigations carried out on the protozoan Paramecium tetraurelia and the cyanobacteria Synechococcus lividus, which were shielded against background radiation or exposed to very low doses of gamma radiation, demonstrated that radiation can stimulate the proliferation of these two single-cell organisms. Radiation hormesis depends on internal factors (age of starting cells) and external factors (lighting conditions). The stimulatory effect occurred only in a limited range of doses and disappeared for dose rates higher than 50 mGy/y.

  18. Evaluation of rubber composites as shielding materials against ionizing radiation

    International Nuclear Information System (INIS)

    Atia, M.K.

    2010-01-01

    Styrene-butadiene rubber/lead oxide composites were prepared as γ-radiation shields.The composites were prepared with different concentration of red lead oxide (Pb 3 O 4 ) .The assessment of the linear attenuation coefficient of the SBR/lead oxide composites for γ -rays from 137 Cs 137 γ-radiation point source was studied . The factors affecting the mechanical properties and shielding capacity of the composites were also studied. These factors include the lead oxide concentration, the type of monomers added and the irradiation dose. The styrene-butadiene rubber/lead oxide composites can attain up to about 43% of the shielding capacity of pure lead. The incorporation of high concentrations of lead oxide and the effect of accumulative irradiation doses up to 3000 kGy on the physico-mechanical properties of the composites were studied . These led to hardening of the SBR rubber/lead oxide composites.

  19. Nanocomposite for Radiation Shielding, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA's Advanced Extravehicular Activity (EVA) program requires the need for materials that can protect astronauts and spacecrafts from ionizing radiations such as...

  20. Adaptation of radiation shielding code to space environment

    International Nuclear Information System (INIS)

    Okuno, Koichi; Hara, Akihisa

    1992-01-01

    Recently, the trend to the development of space has heightened. To the development of space, many problems are related, and as one of them, there is the protection from cosmic ray. The cosmic ray is the radiation having ultrahigh energy, and there was not the radiation shielding design code that copes with cosmic ray so far. Therefore, the high energy radiation shielding design code for accelerators was improved so as to cope with the peculiarity that cosmic ray possesses. Moreover, the calculation of the radiation dose equivalent rate in the moon base to which the countermeasures against cosmic ray were taken was simulated by using the improved code. As the important countermeasures for the safety protection from radiation, the covering with regolith is carried out, and the effect of regolith was confirmed by using the improved code. Galactic cosmic ray, solar flare particles, radiation belt, the adaptation of the radiation shielding code HERMES to space environment, the improvement of the three-dimensional hadron cascade code HETCKFA-2 and the electromagnetic cascade code EGS 4-KFA, and the cosmic ray simulation are reported. (K.I.)

  1. The investigation of gamma and neutron shielding properties of concrete including basalt fibre for nuclear energy applications

    International Nuclear Information System (INIS)

    Nulk, H.; Ipbuker, C.; Gulik, V.; Tkaczyk, A.; Biland, A.

    2015-01-01

    In this study, we would like to draw attention to the prospect of basalt fibre as the main component for concrete reinforcement of NPP. This work describes the computational study of gamma attenuation parameters, the effective atomic number Z(eff) and the effective electron density N e (eff), of relatively light-weight concrete with chopped basalt fibre used as reinforcement in different mixture rates. We can draw the following conclusions. Basalt fibre is a relatively cheap material that can be used as reinforcement instead of metallic fibers. Basalt fibre has a similar specific gravity to that of concrete elements. Basalt fibre has high chemical and abrasion resistance. Basalt fibre has almost 10 times the tensile strength of steel re-bars. Gamma-ray attenuation coefficients increase with addition of basalt fibre into concrete in every case. The effective atomic number of the concrete increases with the addition of basalt fibre. The results show that basalt fibre reinforced concrete have improved shielding properties against gamma rays in comparison with regular concrete. This result is based on a regular concrete with only basalt fiber reinforcement. We estimate that with addition of standard aggregates for radiation shielding concrete, such as barite, magnetite or hematite, the shielding properties will increase exponentially

  2. Measurement of TFTR D-T radiation shielding efficiency

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, H.W.; Ascione G.; Elwood, S. [Princeton Univ., NJ (United States)] [and others

    1994-12-31

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

  3. Radiation Shielding Information Center: a source of computer codes and data for fusion neutronics studies

    International Nuclear Information System (INIS)

    McGill, B.L.; Roussin, R.W.; Trubey, D.K.; Maskewitz, B.F.

    1980-01-01

    The Radiation Shielding Information Center (RSIC), established in 1962 to collect, package, analyze, and disseminate information, computer codes, and data in the area of radiation transport related to fission, is now being utilized to support fusion neutronics technology. The major activities include: (1) answering technical inquiries on radiation transport problems, (2) collecting, packaging, testing, and disseminating computing technology and data libraries, and (3) reviewing literature and operating a computer-based information retrieval system containing material pertinent to radiation transport analysis. The computer codes emphasize methods for solving the Boltzmann equation such as the discrete ordinates and Monte Carlo techniques, both of which are widely used in fusion neutronics. The data packages include multigroup coupled neutron-gamma-ray cross sections and kerma coefficients, other nuclear data, and radiation transport benchmark problem results

  4. Linear attenuation coefficient and build up factor of MCP-96 alloy for radiation shielding and protection

    Science.gov (United States)

    Hopkins, Deidre; Maqbool, Muhammad; Islam, Mohammed

    2009-10-01

    Build-up factors and linear attenuation coefficients of MCP-96 alloy are determined for radiation shielding and protection, using ^60Co and ^137Cs gamma emitters. A narrow collimated beam of γ-rays is passed through various thicknesses of MCP-96 alloy and the attenuation in the intensity of the beam is determined. The thickness of the 4 x 4 cm^2 blocks varies from 0.5 cm to 6 cm. Plotting the thickness of the alloy and the corresponding intensity of the beam allowed us to determine its linear attenuation coefficient. The narrow beam geometry is then replaced by broad beam geometry by removing the collimator and the radiation beam is able to interact with the MCP-96 alloy at all possible positions facing the radiation source. Additional radiations obtained by the detector as a result from the scattering of radiation develops the build-up factor. The buildup factor is then calculated using the attenuated beam received by the detector in the broad beam geometry and in the narrow beam geometry. The buildup factor is found to be dependent on the thickness of the MCP-96 attenuator, the beam energy and the source to attenuator distance. These values are providing ways for dose correction in radiation oncology and radiation shielding and protection when MCP-96 is used as tissue compensator or for radiation protection purposes.

  5. Shielding against the radiation effects of tactical nuclear weapons

    International Nuclear Information System (INIS)

    Hehn, G.

    1985-01-01

    The results of the one-dimensional transport calculations to determine the shielding efficiency of shelter walls to afford protection against enhanced radiation weapons are conservative. Good protective effects are obtained by structural measures such as underground shelters or filled-up soil, due to the soil's good attenuation effect. Other structural measures will concentrate on improving the radiation protective efficiency of the upper edge of shelters formed by the top cover and lateral walls. (DG) [de

  6. Radiation Shielding Materials Containing Hydrogen, Boron, and Nitrogen: Systematic Computational and Experimental Study

    Data.gov (United States)

    National Aeronautics and Space Administration — The objectives of the proposed research are to develop a space radiation shielding material system that has high efficacy for shielding radiation and also has high...

  7. Radiation Shielding Materials Containing Hydrogen, Boron, and Nitrogen: Systematic Computational and Experimental Study Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The objectives of the proposed research are to develop a space radiation shielding material system that has high efficacy for shielding radiation and also has high...

  8. Mobile robot prototype detector of gamma radiation; Prototipo de robot movil detector de radiacion gamma

    Energy Technology Data Exchange (ETDEWEB)

    Vazquez C, R.M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Duran V, M. D.; Jardon M, C. I., E-mail: raulmario.vazquez@inin.gob.mx [Tecnologico de Estudios Superiores de Villa Guerrero, Carretera Federal Toluca-Ixtapan de la Sal Km. 64.5, La Finca Villa Guerrero, Estado de Mexico (Mexico)

    2014-10-15

    In this paper the technological development of a mobile robot prototype detector of gamma radiation is shown. This prototype has been developed for the purpose of algorithms implementation for the applications of terrestrial radiation monitoring of exposed sources, search for missing radioactive sources, identification and delineation of radioactive contamination areas and distribution maps generating of radioactive exposure. Mobile robot detector of radiation is an experimental technology development platform to operate in laboratory environment or flat floor facilities. The prototype integrates a driving section of differential configuration robot on wheels, a support mechanism and rotation of shielded detector, actuator controller cards, acquisition and processing of sensor data, detection algorithms programming and control actuators, data recording (Data Logger) and data transmission in wireless way. The robot in this first phase is remotely operated in wireless way with a range of approximately 150 m line of sight and can extend that range to 300 m or more with the use of signal repeaters. The gamma radiation detection is performed using a Geiger detector shielded. Scan detection is performed at various time sampling periods and diverse positions of discrete or continuous angular orientation on the horizon. The captured data are geographical coordinates of robot GPS (latitude and longitude), orientation angle of shield, counting by sampling time, date, hours, minutes and seconds. The data is saved in a file in the Micro Sd memory on the robot. They are also sent in wireless way by an X Bee card to a remote station that receives for their online monitoring on a laptop through an acquisition program by serial port on Mat Lab. Additionally a voice synthesizing card with a horn, both in the robot, periodically pronounced in Spanish, data length, latitude, orientation angle of shield and detected accounts. (Author)

  9. Radiation shielding techniques and applications. 3. Analysis of Photon Streaming Through and Around Shield Doors

    International Nuclear Information System (INIS)

    Barnett, Marvin; Hack, Joe; Nathan, Steve; White, Travis

    2001-01-01

    Westinghouse Safety Management Solutions (Westinghouse SMS) has been tasked with providing radiological engineering design support for the new Commercial Light Water Reactor Tritium Extraction Facility (CLWR-TEF) being constructed at the Savannah River Site (SRS). The Remote Handling Building (RHB) of the CLWR-TEF will act as the receiving facility for irradiated targets used in the production of tritium for the U.S. Department of Energy (DOE). Because of the high dose rates, approaching 50 000 rads/h (500 Gy/h) from the irradiated target bundles, significant attention has been made to shielding structures within the facility. One aspect of the design that has undergone intense review is the shield doors. The RHB has six shield doors that needed to be studied with respect to photon streaming. Several aspects had to be examined to ensure that the design meets the radiation dose levels. Both the thickness and streaming issues around the door edges were designed and examined. Photon streaming through and around a shield door is a complicated problem, creating a reliance on computer modeling to perform the analyses. The computer code typically used by the Westinghouse SMS in the evaluation of photon transport through complex geometries is the MCNP Monte Carlo computer code. The complexity of the geometry within the problem can cause problems even with the Monte Carlo codes. Striking a balance between how the code handles transport through the shield door with transport through the streaming paths, particularly with the use of typical variance reduction methods, is difficult when trying to ensure that all important regions of the model are sampled appropriately. The thickness determination used a simple variance reduction technique. In construction, the shield door will not be flush against the wall, so a solid rectangular slab leaves streaming paths around the edges. Administrative controls could be used to control dose to workers; however, 10 CFR 835.1001 states

  10. Virtual Gamma Ray Radiation Sources through Neutron Radiative Capture

    Energy Technology Data Exchange (ETDEWEB)

    Scott Wilde, Raymond Keegan

    2008-07-01

    The countrate response of a gamma spectrometry system from a neutron radiation source behind a plane of moderating material doped with a nuclide of a large radiative neutron capture cross-section exhibits a countrate response analogous to a gamma radiation source at the same position from the detector. Using a planar, surface area of the neutron moderating material exposed to the neutron radiation produces a larger area under the prompt gamma ray peak in the detector than a smaller area of dimensions relative to the active volume of the gamma detection system.

  11. Novel Concepts for Radiation Shielding Materials

    Data.gov (United States)

    National Aeronautics and Space Administration — The likelihood of safely sending astronauts to Mars is becoming bleaker because of the health risks that would result from exposure to galactic cosmic radiation...

  12. Radiation Shield Optimizer using Genetic Algorithms

    Data.gov (United States)

    National Aeronautics and Space Administration — The harmful effects of radiation are one of the most significant challenges to long duration spaceflight. Therefore, we aim to develop a new approach for optimizing...

  13. 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.)

  14. The Radiation Streaming Calculation for Air Gap of the Shielding Door

    International Nuclear Information System (INIS)

    Min, Y. S.; Jeon, G. P.; Mun, K. J.; Nam, J. M.; Cho, J. S.; Cho, J. H.; Kim, Jun Yeon

    2011-01-01

    There are many penetrations and the thin air filled clearance gaps in accelerator facility, such as a cable, a cooling water pipe or an air conditioning duct as well as an air gap of between the wall and the shielding door. The estimation of the radiation streaming through these penetrations or the air filled gaps is one of the most difficult parts in shielding design. The Shin's semiempirical formula describing energy-space distributions of neutrons and gamma-rays streaming in ducts or labyrinths is very useful for application to accelerator facility. A streaming calculation code DUCT-III is based on the Shin's formula with the albedo data up to 3GeV. In this paper, the source term was calculated by MCNPX and the radiation streaming through the air gap of between the wall and the shielding door by DUCT-III. The DUCT-III code is based on the Shin's semi-empirical formula. The formula, which describes the direct and albedo components, is derived in generic straight duct geometry. It is expressed by the product of spatial distributions which are represented by twice and eight-time reflected components, and power of an albedo matrix. This formula was then extended to bent ducts. The inflow of radiations to downstream at a corner of multi-bent ducts is formulated with the flux in the upstream leg. Using the obtained inflow current as the source term to downstream, the formula predicts the radiation flux in the downstream leg

  15. ZnO-PbO-B2O3 glasses as gamma-ray shielding materials

    DEFF Research Database (Denmark)

    Singh, H.; Singh, K.; Gerward, Leif

    2003-01-01

    Values of the gamma-ray mass-attenuation coefficient, the photon mean free path (MFP), the effective atomic number and the effective electron density have been determined experimentally for xZnO.2xPbO.(1-3x)B2O3 (x = 0.1-0.26) glasses at photon energies 511, 662, 1173 and 1332 keV and compared wi...... with theoretical data. The specific volume of the glasses has been derived from density measurements and studied as a function of composition. It is pointed out that these glasses have potential applications in radiation shielding.......Values of the gamma-ray mass-attenuation coefficient, the photon mean free path (MFP), the effective atomic number and the effective electron density have been determined experimentally for xZnO.2xPbO.(1-3x)B2O3 (x = 0.1-0.26) glasses at photon energies 511, 662, 1173 and 1332 keV and compared...

  16. [The model of radiation shielding of the service module of the International space station].

    Science.gov (United States)

    Kolomenskiĭ, A V; Kuznetsov, V G; Laĭko, Iu A; Bengin, V V; Shurshakov, V A

    2001-01-01

    Compared and contrasted were models of radiation shielding of habitable compartments of the basal Mir module that had been used to calculate crew absorbed doses from space radiation. Developed was a model of the ISS Service module radiation shielding. It was stated that there is a good agreement between experimental shielding function and the one calculated from this model.

  17. Radiation protection in category III large gamma irradiators

    International Nuclear Information System (INIS)

    Costa, Neivaldo; Furlan, Gilberto Ribeiro; Itepan, Natanael Marcio

    2011-01-01

    This article discusses the advantages of category III large gamma irradiator compared to the others, with emphasis on aspects of radiological protection, in the industrial sector. This category is a kind of irradiators almost unknown to the regulators authorities and the industrial community, despite its simple construction and greater radiation safety intrinsic to the model, able to maintain an efficiency of productivity comparable to those of category IV. Worldwide, there are installed more than 200 category IV irradiators and there is none of a category III irradiator in operation. In a category III gamma irradiator, the source remains fixed in the bottom of the tank, always shielded by water, negating the exposition risk. Taking into account the benefits in relation to radiation safety, the category III large irradiators are highly recommended for industrial, commercial purposes or scientific research. (author)

  18. Radiation source shielding and collimating device

    International Nuclear Information System (INIS)

    Garrett, R.E.

    1978-01-01

    A radiation source, such as 241 Am, sealed in a capsule at atmospheric pressure especially for use in agricultural machines such as lettuce harvesters having a maturity tester is described. The capsule is disposed in a plunger movable in an evacuated chamber between a first position in alignment with a passage having a window permeable by radiation to the outside and a second position in which the capsule is out of alignment with the passage. A pressure responsive switch monitors the pressure in the chamber and affords an alarm if the chamber pressure rises

  19. Development of simplified methods and data bases for radiation shielding calculations for concrete

    Energy Technology Data Exchange (ETDEWEB)

    Bhuiyan, S.I.; Roussin, R.W.; Lucius, J.L.; Marable, J.H.; Bartine, D.A.

    1986-06-01

    Two simplified methods have been developed which allow rapid and accurate calculations of the attenuation of neutrons and gamma rays through concrete shields. One method, called the BEST method, uses sensitivity coefficients to predict changes in the transmitted dose from a fission source that are due to changes in the composition of the shield. The other method uses transmission factors based on adjoint calculations to predict the transmitted dose from an arbitrary source incident on a given shield. The BEST method, utilizing an exponential molecule that is shown to be a significant improvement over the traditional linear model, has been successfully applied to slab shields of standard concrete and rebar concrete. It has also been tested for a special concrete that has been used in many shielding experiments at the ORNL Tower Shielding Facility, as well as for a deep-penetration sodium problem. A comprehensive data base of concrete sensitivity coefficients generated as part of this study is available for use in the BEST model. For problems in which the changes are energy independent, application of the model and data base can be accomplished with a desk calculator. Larger-scale calculations required for problems that are energy dependent are facilitated by employing a simple computer code, which is included, together with the data base and other calculational aids, in a data package that can be obtained from the ORNL Radiation Shielding Information Center (request DLC-102/CONSENT). The transmission factors used by the second method are a byproduct of the sensitivity calculations and are mathematically equivalent to the surface adjoint function phi*, which gives the dose equivalent transmitted through a slab of thickness T due to one particle incident on the surface in the gth energy group and jth direction. 18 refs., 1 fig., 50 tabs.

  20. Nuclear Technology Series. Course 19: Radiation Shielding.

    Science.gov (United States)

    Center for Occupational Research and Development, Inc., Waco, TX.

    This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutions in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

  1. Shielding ability of lead loaded radiation resistant gloves

    International Nuclear Information System (INIS)

    Kawano, Takao; Ebihara, Hiroshi

    1990-01-01

    The shielding ability of radiation resistant gloves were examined. The gloves are made of lead loaded (as PbO 2 ) polyvinyl chloride resin and are about 0.4 mm of thickness (70 mg/cm 2 ). Eleven test pieces were sampled from each of three gloves (total were thirty three) and the transmission rates for radiations (X-ray or γ-ray) through the test pieces were measured with radiation sources, 99m Tc, 57 Co, 133 Ba, 133 Xe and 241 Am. The differences of the transmission rate for radiations by the positions of the gloves were smaller than 15%, and the differences by three gloves were smaller than 5% in the case of 60 keV and 141 keV radiations. The average transmission rates for radiations in thirty three test pieces were about 40% for 30 keV radiation, about 90% for 80 keV and 140 keV radiations. The shielding characteristic of the gloves could be equivalent to about 0.026 mm thick lead plate. (author)

  2. 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.

  3. Shielding and synchrotron radiation in toroidal waveguide

    Directory of Open Access Journals (Sweden)

    G. V. Stupakov

    2003-03-01

    Full Text Available We develop a new approach to the calculation of the synchrotron radiation in a toroidal vacuum chamber. Using a small parameter ϵ=sqrt[a/R], where a is the characteristic size of the cross section of the toroid and R is the bending radius, we simplify Maxwell’s equations assuming that the characteristic frequency of the modes ω∼c/aϵ and neglect terms of higher order in ϵ. For a rectangular cross section of the waveguide, we find an analytical solution of the equations and analyze their asymptotics at very high frequency. We then obtain an equation which gives radiation into each synchronous mode. We demonstrate the flexibility of the new method by calculating the frequencies and the loss factors for the lowest modes in square and round waveguides.

  4. Estimation of the shielding ability of a tungsten functional paper for diagnostic x-rays and gamma rays.

    Science.gov (United States)

    Monzen, Hajime; Kanno, Ikuo; Fujimoto, Takahiro; Hiraoka, Masahiro

    2017-09-01

    Tungsten functional paper (TFP) is a novel paper-based radiation-shielding material. We measured the shielding ability of TFP against x-rays and gamma rays. The TFP was supplied in 0.3-mm-thick sheets that contained 80% tungsten powder and 20% cellulose (C 6 H 10 O 5 ) by mass. In dose measurements for x-rays (60, 80, 100, and 120 kVp), we measured doses after through 1, 2, 3, 5, 10, and 12 TFP sheets, as well as 0.3 and 0.5 mm of lead. In lead equivalence measurements, we measured doses after through 2 and 10 TFP sheets for x-rays (100 and 150 kVp), and 0, 7, 10, 20, and 30 TFP sheets for gamma rays from cesium-137 source (662 keV). And then, the lead equivalent thicknesses of TFP were determined by comparison with doses after through standard lead plates (purity >99.9%). Additionally, we evaluated uniformity of the transmitted dose by TFP with a computed radiography image plate for 50 kVp x-rays. A single TFP sheet was found to have a shielding ability of 65%, 53%, 48%, and 46% for x-rays (60, 80, 100, and 120 kVp), respectively. The lead equivalent thicknesses of two TFP sheets were 0.10 ± 0.02, 0.09 ± 0.02 mmPb, and of ten TFP sheets were 0.48 ± 0.02 and 0.51 ± 0.02 mmPb for 100 and 150 kVp x-rays, respectively. The lead equivalent thicknesses of 7, 10, 20, and 30 sheets of TFP for gamma rays from cesium-137 source were estimated as 0.28, 0.43, 0.91, and 1.50 mmPb with an error of ± 0.01 mm. One TFP sheet had nonuniformity, however, seven TFP sheets provided complete shielding for 50 kVp x-rays. TFP has adequate radiation shielding ability for x-rays and gamma rays within the energy range used in diagnostic imaging field. © 2017 The Authors. Journal of Applied Clinical Medical Physics published by Wiley Periodicals, Inc. on behalf of American Association of Physicists in Medicine.

  5. A new lead-free radiation shielding material for radiotherapy.

    Science.gov (United States)

    Yue, Kun; Luo, Wenyun; Dong, Xiaoqing; Wang, Chuanshan; Wu, Guohua; Jiang, Mawei; Zha, Yuanzi

    2009-02-01

    Lead has recently been recognised as a source of environmental pollution, including the lead used for radiation shielding in radiotherapy. The bremsstrahlung radiation caused by the interaction between the electron beam and lead may reduce the accuracy of radiotherapy. To avoid the use of lead, a new material composed of tungsten and hydrogenated styrene-butadiene-styrene copolymer is studied with the Monte Carlo (MC) method and experiment in this paper. The component of the material is chosen after simulation with the MC method and the practical measurement is taken to validate the shielding ability of the material. The result shows that the shielding ability of the new material is good enough to fulfill the requirement for application in radiotherapy. Compared with lead alloy, the present new material is so flexible that can be easily customized into arbitrary shapes. Moreover, the material is environmentally friendly and can be recycled conveniently. Therefore, the material can be used as an effective lead substitute for shielding against electron beams in radiotherapy.

  6. Gamma rays shielding and sensing application of some rare earth doped lead-alumino-phosphate glasses

    Science.gov (United States)

    Kaur, Preet; Singh, Devinder; Singh, Tejbir

    2018-03-01

    Seven rare earth (Sm3+, Eu3+ and Nd3+) doped lead alumino phosphate glasses were prepared. The protective and sensing measures from gamma rays were analysed in terms of parameters viz. density (ρ), refractive index, energy band gap (Eg), mean free path (mfp), effective atomic number (Zeff) and buildup factors (energy absorption EABF as well as exposure buildup factor EBF). The energy dependent parameters (mfp, Zeff, EABF and EBF) were investigated in the energy region from 15 keV to 15 MeV. EABF and EBF values were observed to be maximum in the intermediate energy region. Besides, the EABF and EBF values for the prepared samples are shown to have strong dependence on chemical composition of the glass at lower energy, whereas, it is almost independent of chemical composition in higher energy region. The prepared glass samples are found to have potential applications in radiation shielding as well as radiation sensing, which further find numerous applications in the field of medicine and industry.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-07-15

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

  8. Investigation of Gamma and Neutron Shielding Parameters for Borate Glasses Containing NiO and PbO

    Directory of Open Access Journals (Sweden)

    Vishwanath P. Singh

    2014-01-01

    Full Text Available The mass attenuation coefficients, μ/ρ, half-value layer, HVL, tenth-value layer, TVL, effective atomic numbers, ZPIeff, and effective electron densities, Ne,eff, of borate glass sample systems of (100-x-y Na2B4O7 : xPbO : yNiO (where x and y=0, 2, 4, 6, 8, and 10 weight percentage containing PbO and NiO, with potential gamma ray and neutron shielding applications, have been investigated. The gamma ray interaction parameters, μ/ρ, HVL, TVL, ZPIeff, and Ne,eff, were computed for photon energy range 1 keV–100 GeV. The macroscopic fast neutron removal cross-sections (ΣR have also been calculated. Appreciable variations were noted for all the interaction parameters by varying the photon energy and the chemical composition of the glass samples. The better shielding properties of borate glass samples containing PbO were found. These results indicated that borate glass samples are a good radiation shielding material.

  9. Project Marna Natural Gamma Radiation MAP

    International Nuclear Information System (INIS)

    Suarez, E.; Fernandez, J.A.

    1997-01-01

    The confusion created by the accident that occurred in one of the Chernobyl reactors in April of 1986 made the general public and governments aware of the need for improved monitoring of environmental radiation levels. The levels of total gamma radiation or total gamma exposure rate over large areas reached values as high as 400 micro Roentgen/hour (mu R/h) and at points exceeded 1000 mu R/h. It should be borne in mind that, depending on the type of geological formations, normal values range from 5 to 30 mu R/h. The IAEA recommended to all countries that natural gamma radiation maps be made available to evaluate the levels of natural gamma radiation and possible increases, and it also indicated its concern that information be standardized. In addition, it stressed the advisability of using data obtained from uranium prospecting. (Author)

  10. ORNL shielded facilities capable of remote handling of highly radioactive beta--gamma emitting materials

    International Nuclear Information System (INIS)

    Whitson, W.R.

    1977-09-01

    A survey of ORNL facilities having adequate shielding and containment for the remote handling of experimental quantities of highly radioactive beta-gamma emitting materials is summarized. Portions of the detailed descriptions of these facilities previously published in ORNL/TM-1268 are still valid and are repeated

  11. Monte Carlo applications to radiation shielding problems

    International Nuclear Information System (INIS)

    Subbaiah, K.V.

    2009-01-01

    Monte Carlo methods are a class of computational algorithms that rely on repeated random sampling of physical and mathematical systems to compute their results. However, basic concepts of MC are both simple and straightforward and can be learned by using a personal computer. Uses of Monte Carlo methods require large amounts of random numbers, and it was their use that spurred the development of pseudorandom number generators, which were far quicker to use than the tables of random numbers which had been previously used for statistical sampling. In Monte Carlo simulation of radiation transport, the history (track) of a particle is viewed as a random sequence of free flights that end with an interaction event where the particle changes its direction of movement, loses energy and, occasionally, produces secondary particles. The Monte Carlo simulation of a given experimental arrangement (e.g., an electron beam, coming from an accelerator and impinging on a water phantom) consists of the numerical generation of random histories. To simulate these histories we need an interaction model, i.e., a set of differential cross sections (DCS) for the relevant interaction mechanisms. The DCSs determine the probability distribution functions (pdf) of the random variables that characterize a track; 1) free path between successive interaction events, 2) type of interaction taking place and 3) energy loss and angular deflection in a particular event (and initial state of emitted secondary particles, if any). Once these pdfs are known, random histories can be generated by using appropriate sampling methods. If the number of generated histories is large enough, quantitative information on the transport process may be obtained by simply averaging over the simulated histories. The Monte Carlo method yields the same information as the solution of the Boltzmann transport equation, with the same interaction model, but is easier to implement. In particular, the simulation of radiation

  12. Mathematical modeling of the radiation dose received from photons passing over and through shielding walls in a PET/CT suite

    DEFF Research Database (Denmark)

    Fog, Lotte S; Cormack, John

    2010-01-01

    Given that the financial cost of shielding PET/CT suites can be substantial, it has become increasingly important to be able to accurately assess the thickness of shielding required for barriers and whether it is necessary to extend such shielding all the way to the ceiling. The overall shielding...... dependent on the geometry of the radiation source and the resulting energy spectrum of the emitted radiation. The transmission from a patient source was found to be around half of that from a small vial and also half of that reported previously using parallel beams of mono-energetic radiation. For PET...... emissions, the dose from scatter over the barrier at waist height is relatively small but may have to be taken into account if the design dose limit is low. Shielding from floor to ceiling is probably not warranted in most instances for PET gamma emissions; in PET/CT installations, however, a thinner layer...

  13. On the honeybee resistance to gamma radiation

    International Nuclear Information System (INIS)

    Courtois, G.; Lecomte, J.

    1960-01-01

    The honeybee, when irradiated by gamma radiations from a cobalt-60 source can stand a 18000 r dose without any apparent harm. Noticeable harm is observed for 90000 r. while immediate death of 100% of the individuals is obtained with a 200000 r dose. The physiological condition of the honeybee plays an important role in its resistance to gamma radiation. Reprint of a paper published in Annales de l'abeille, IV, 1959, p. 285-290 [fr

  14. Gamma ray shielding and structural properties of Bi{sub 2}O{sub 3}−PbO−B{sub 2}O{sub 3}−V{sub 2}O{sub 5} glass system

    Energy Technology Data Exchange (ETDEWEB)

    Kaur, Kulwinder, E-mail: kanwarjitsingh@yahoo.com; Singh, K. J., E-mail: kanwarjitsingh@yahoo.com; Anand, Vikas, E-mail: kanwarjitsingh@yahoo.com [Department of Physics, Guru Nanak Dev University, Amritsar 143005 (India)

    2014-04-24

    The present work has been undertaken to evaluate the applicability of Bi{sub 2}O{sub 3}−PbO−B{sub 2}O{sub 3}−V{sub 2}O{sub 5} glass system as gamma ray shielding material. Gamma ray mass attenuation coefficient has been determined theoretically using WinXcom computer software developed by National Institute of Standards and Technology. A meaningful comparison of their radiation shielding properties has been made in terms of their half value layer parameter with standard radiation shielding concrete 'barite'. Structural properties of the prepared glass system have been investigated in terms of XRD and FTIR techniques in order to check the possibility of their commercial utility as alternate to conventional concrete for gamma ray shielding applications.

  15. Comparison of measured and calculated neutron and gamma-ray energy spectra behind an in-line shielded duct

    International Nuclear Information System (INIS)

    Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M.; Chapman, G.T.; Tang, J.S.

    1982-05-01

    Integral experiments that measure the transport of approx. 14 MeV neutrons through a 0.30-m-diameter duct having a length-to-diameter ratio of 2.83 that is partially plugged with a 0.15 m diameter, 0.51 m long shield comprised of alternating layers of stainless steel type 304 and borated polyethylene have been carried out at the Oak Ridge National Laboratory. Measured and calculated neutron and gamma ray energy spectra are compared at several locations relative to the mouth of the duct. The measured spectra were obtained using an NE-213 liquid scintillator detector with pulse shape discrimination methods used to simultaneously resolve neutron and gamma ray events. The calculated spectra were obtained using a computer code network that incorporates two radiation transport methods: discrete ordinates (with P 3 multigroup cross sections) and Monte Carlo (with continuous point cross sections). The two radiation transport methods are required to account for neutrons that singly scatter from the duct to the detectors. The calculated and measured neutron energy spectra above 850 keV agree with 5 to 50% depending on detector location and neutron energy. The calculated and measured gamma ray energy spectra above 750 keV are also in favorable agreement, approx. 5 to 50%, depending on detector location and gamma ray energy

  16. Effects of gamma radiation in tomato seeds

    International Nuclear Information System (INIS)

    Wiendl, Toni A.; Wiendl, Fritz W.; Franco, Suely S.H.; Franco, Jose G.; Althur, Valter; Arthur, Paula B.

    2013-01-01

    Tomato dry seeds of the hybrid 'Gladiador' F1 were exposed to low doses of gamma radiation from Co-60 source at 0,509 kGy tax rate in order to study stimulation effects of radiation on germination and plant growth. Eight treatments radiation doses were applied as follows: 0 (control); 2,5; 5,0; 7,5; 10,0; 12,5; 15,0; 20,0 Gy. Seed germination as well as green fruits number, harvested fruit number, fruit weight and total production were assessed to identify occurrence of stimulation. Tomato seeds and plants were handled as for usual tomato production in Brazil. Low doses of gamma radiation treatment in the seeds stimulate germination and substantially increase fruit number and total production up to 86% at 10 Gy dose. There are evidences that the use of low doses of gamma radiation can stimulate germination and plant production thus, showing hormetic effects. (author)

  17. Effects of gamma radiation in tomato seeds

    Energy Technology Data Exchange (ETDEWEB)

    Wiendl, Toni A.; Wiendl, Fritz W.; Franco, Suely S.H.; Franco, Jose G.; Althur, Valter, E-mail: tawiendl@hotmail.com, E-mail: gilmita@uol.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Arthur, Paula B., E-mail: arthur@cena.usp.br [Centro de Energia Nuclear na Agricultura (CENA/USP), Piracicaba, SP (Brazil)

    2013-07-01

    Tomato dry seeds of the hybrid 'Gladiador' F1 were exposed to low doses of gamma radiation from Co-60 source at 0,509 kGy tax rate in order to study stimulation effects of radiation on germination and plant growth. Eight treatments radiation doses were applied as follows: 0 (control); 2,5; 5,0; 7,5; 10,0; 12,5; 15,0; 20,0 Gy. Seed germination as well as green fruits number, harvested fruit number, fruit weight and total production were assessed to identify occurrence of stimulation. Tomato seeds and plants were handled as for usual tomato production in Brazil. Low doses of gamma radiation treatment in the seeds stimulate germination and substantially increase fruit number and total production up to 86% at 10 Gy dose. There are evidences that the use of low doses of gamma radiation can stimulate germination and plant production thus, showing hormetic effects. (author)

  18. Biological shielding design calculation for agricultural radiation processing facility

    International Nuclear Information System (INIS)

    Petwal, V.C.; Sandha, R.S.; Soni, H.C.; Subbaiah, K.V.

    2005-01-01

    An electron beam radiation processing facility for agricultural products is being set-up at Centre for Advanced Technology Indore. The facility will be based on a pulsed linear accelerator and will be used in electron and photon modes to process various products e.g. onion, potato, home-pack items and medical products. When electron beam interact with structural components of accelerator or high Z-target used in photon mode, it generates intense Bremsstrahlung radiation field, which poses radiation protection problem. Biological shielding has been designed to provide protection against the generated radiation. Different conveying schemes and hence design of irradiation cell have been studied and results are presented for two promising designs. (author)

  19. 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.)

  20. Outline of radiation shielding (streaming) code system used in Kawasaki Heavy Industries, Ltd. (KHI)

    International Nuclear Information System (INIS)

    Suzuki, Ikunori

    1980-01-01

    The most troublesome problem in designing nuclear reactor shielding is streaming. The paper introduces the shield computing code system of Kawasaki Heavy Industries, Ltd. (KHI) centering around the streaming. This code system is roughly composed of 9 sections, each of which is explained. Of these, the transmission computing section and the streaming computing section have no significant difference as both handle radiation transport problems. These computing code groups are divided into the one used for neutrons only, the one for gamma emission only and the one for both neutrons and gamma radiation. In the second half of the paper, the codes RASC2D and DOT3.5 are specifically described in detail, which are frequently used in KHI as the codes for streaming analysis. The RASC2D has been developed by KHI, started in 1970 in commission of the Power Reactor and Nuclear Fuel Development Corporation. The DOT code was developed in ORNL, U.S., and the present DOT3.5 is the latest one of the DOT codes usable in Japan. The code DOT3.5 is more convenient to use than the RASC2D if both are compared, and thus KHI employs DOT3.5 more frequently than RASC2D at present. (Wakatsuki, Y.)

  1. SGR-76 gamma radiation level indicator

    International Nuclear Information System (INIS)

    Chubinskij-Nadezhdin, I.V.

    1978-01-01

    The design of a gamma-radiation level indicator is described; the instrument is part of a mobile radiometric laboratory (MRL). The design of the instrument permits gamma-radiation dose rates recording at 0.2-200 R/hr, and signals on gamma-background levels. The instrument has two separate threshold levels of signalling actuation. The light signalling at the first level is precautionary, and the sound signalling at the second level indicates the necessity of taking a decision as to whether or not the MRL can remain in the gamma-radiation field. Halogenic counters operating in a current mode are used as detectors. The basic error in recording the dose rate amounts to +-25%. Overall dimensions of the instrument 150x280x100 mm; weight less than 2.5 kg

  2. Elementary computation of radiation doses and shieldings for radiochemical laboratories; Calculo Elemental de dosis y blindajes para laboratorios radioquimicos

    Energy Technology Data Exchange (ETDEWEB)

    Jimeno de Osso, F.

    1971-07-01

    Simple procedures for the calculation of radiation exposition, half thickness, shield thickness, etc. are described and equations and graphs are included for those gamma-emitting radionuclides, that are more often used in radiochemical laboratories. Application is made of these procedures to three radionuclides, bromine-82, sodium-24 and cobalt-60 which cover a rather wl.de energy range; theoretical results are compared with those obtained from experimental measurements. (Author) 23 refs.

  3. Gamma radiation effect on vulcanized synthetic rubber

    International Nuclear Information System (INIS)

    Santos, L.G. dos.

    1987-01-01

    Samples of polybutadiene were irradiated with gamma radiation, using cobalt-60 source, with time interval up to 20 days. Tensile-deformation tests carried out in physics testing machine, shown mechanical hardening induced by radiation, followed by reduction of breaking stress and ultimate elongation. (M.C.K.)

  4. A new approximating formula for calculating gamma-ray buildup factors in multilayer shields

    International Nuclear Information System (INIS)

    Assad, A.; Chiron, M.; Nimal, J.C.; Diop, C.M.; Ridoux, P.

    1999-01-01

    This study proposes a new approximating formula for calculating gamma-ray buildup factors in multilayer shields. The formula combines the buildup factors of single-layer shields with products and quotients. The feasibility of the formula for reproducing the buildup factors was tested by using point isotropic buildup factors calculated with the SN1D discrete ordinates code as reference data. The dose buildup factors of single-, double-, and multilayer shields composed of water, aluminum, iron, and lead were calculated for a spherical geometry in the energy range between 10 MeV and 40 keV and for total thicknesses of up to 30 mean free paths. The calculation of the buildup factors takes into account the bound electron effect of Compton scattering (incoherent scattering), the coherent scattering, the pair production, and the secondary sources of bremsstrahlung and fluorescence. The tests have shown that the approximating formula reproduces the reference data of double-layer shields very well for most cases. With the same parameters and with a new physical consideration that takes into account in a global way the degradation of the gamma-ray energy spectrum, the buildup factors of three- and five-layer shields were also very well reproduced

  5. Mathematical modeling of the radiation dose received from photons passing over and through shielding walls in a PET/CT suite

    DEFF Research Database (Denmark)

    Fog, Lotte S; Cormack, John

    2010-01-01

    Given that the financial cost of shielding PET/CT suites can be substantial, it has become increasingly important to be able to accurately assess the thickness of shielding required for barriers and whether it is necessary to extend such shielding all the way to the ceiling. The overall shielding...... requirement for a PET/CT installation must take into account both 511 keV gamma ray emissions from PET scans and lower energy x-ray scatter from CT scans. This paper deals with the overall impact of emissions from both modalities. Radiation exposure from both scatter over shielding barriers as well...... as transmission through these barriers is taken into account. A series of simulations of the dose received by a person positioned behind a shielding barrier in a typical PET/CT scanning suite were carried out using both Monte Carlo and analytical models. The transmission through lead barriers was found to be very...

  6. Validity of the Aluminum Equivalent Approximation in Space Radiation Shielding

    Science.gov (United States)

    Badavi, Francis F.; Adams, Daniel O.; Wilson, John W.

    2009-01-01

    The origin of the aluminum equivalent shield approximation in space radiation analysis can be traced back to its roots in the early years of the NASA space programs (Mercury, Gemini and Apollo) wherein the primary radiobiological concern was the intense sources of ionizing radiation causing short term effects which was thought to jeopardize the safety of the crew and hence the mission. Herein, it is shown that the aluminum equivalent shield approximation, although reasonably well suited for that time period and to the application for which it was developed, is of questionable usefulness to the radiobiological concerns of routine space operations of the 21 st century which will include long stays onboard the International Space Station (ISS) and perhaps the moon. This is especially true for a risk based protection system, as appears imminent for deep space exploration where the long-term effects of Galactic Cosmic Ray (GCR) exposure is of primary concern. The present analysis demonstrates that sufficiently large errors in the interior particle environment of a spacecraft result from the use of the aluminum equivalent approximation, and such approximations should be avoided in future astronaut risk estimates. In this study, the aluminum equivalent approximation is evaluated as a means for estimating the particle environment within a spacecraft structure induced by the GCR radiation field. For comparison, the two extremes of the GCR environment, the 1977 solar minimum and the 2001 solar maximum, are considered. These environments are coupled to the Langley Research Center (LaRC) deterministic ionized particle transport code High charge (Z) and Energy TRaNsport (HZETRN), which propagates the GCR spectra for elements with charges (Z) in the range I aluminum equivalent approximation for a good polymeric shield material such as genetic polyethylene (PE). The shield thickness is represented by a 25 g/cm spherical shell. Although one could imagine the progression to greater

  7. PRELIMINARY DESIGN OF CRYOGENIC HYDROGEN RADIATION SHIELD FOR HUMAN SPACE FLIGHT

    Data.gov (United States)

    National Aeronautics and Space Administration — Hydrogen is the most mass-efficient radiation shielding material for protection against the space radiation environment. The concept of Cryogenic Hydrogen Radiation...

  8. Geant4 calculations for space radiation shielding material Al2O3

    Science.gov (United States)

    Capali, Veli; Acar Yesil, Tolga; Kaya, Gokhan; Kaplan, Abdullah; Yavuz, Mustafa; Tilki, Tahir

    2015-07-01

    Aluminium Oxide, Al2O3 is the most widely used material in the engineering applications. It is significant aluminium metal, because of its hardness and as a refractory material owing to its high melting point. This material has several engineering applications in diverse fields such as, ballistic armour systems, wear components, electrical and electronic substrates, automotive parts, components for electric industry and aero-engine. As well, it is used as a dosimeter for radiation protection and therapy applications for its optically stimulated luminescence properties. In this study, stopping powers and penetrating distances have been calculated for the alpha, proton, electron and gamma particles in space radiation shielding material Al2O3 for incident energies 1 keV - 1 GeV using GEANT4 calculation code.

  9. Geant4 calculations for space radiation shielding material Al2O3

    Directory of Open Access Journals (Sweden)

    Capali Veli

    2015-01-01

    Full Text Available Aluminium Oxide, Al2O3 is the most widely used material in the engineering applications. It is significant aluminium metal, because of its hardness and as a refractory material owing to its high melting point. This material has several engineering applications in diverse fields such as, ballistic armour systems, wear components, electrical and electronic substrates, automotive parts, components for electric industry and aero-engine. As well, it is used as a dosimeter for radiation protection and therapy applications for its optically stimulated luminescence properties. In this study, stopping powers and penetrating distances have been calculated for the alpha, proton, electron and gamma particles in space radiation shielding material Al2O3 for incident energies 1 keV – 1 GeV using GEANT4 calculation code.

  10. Effects of scattering anisotropy approximation in multigroup radiation shielding calculations

    International Nuclear Information System (INIS)

    Altiparmakov, D.

    1983-01-01

    Expansion of the scattering cross sections into Legendre series is the usual way of solving neutron transport problems. Because of the large space gradients of the neutron flux, the effects of that approximation become especially remarkable in the radiation shielding calculations. In this paper, a method taking into account the scattering anisotropy is presented. From the point od view of the accuracy and computing rate, the optimal approximation of the scattering anisotropy is established for the basic protective materials on the basis of simple problem calculations. (author)

  11. Ore sorting using natural gamma radiation

    International Nuclear Information System (INIS)

    Clark, G.J.; Dickson, B.L.; Gray, F.E.

    1980-01-01

    A method of sorting an ore which emits natural gamma radiation is described, comprising the steps of: (a) mining the ore, (b) placing, substantially at the mining location, the sampled or mined ore on to a moving conveyor belt, (c) measuring the natural gamma emission, water content and mass of the ore while the ore is on the conveyor belt, (d) using the gamma, water content and mass measurements to determine the ore grade, and (e) directing the ore to a location characteristic of its grade when it leaves the conveyor belt

  12. Spectral properties and shielding behavior of gamma irradiated MoO{sub 3}-doped silicophosphate glasses

    Energy Technology Data Exchange (ETDEWEB)

    Hamdy, Y.M., E-mail: yousry_m_h@yahoo.com [Spectroscopy Department, Physics Division, National Research Center, Dokki, Cairo (Egypt); Marzouk, M.A.; ElBatal, H.A. [Glass Research Department, National Research Center, Dokki, Cairo (Egypt)

    2013-11-15

    Combined optical and infrared absorption spectra of prepared molybdenum ions in sodium silicophosphate host glasses were investigated before and after gamma irradiation with a dose of 8 Mrad (8×10{sup 4} Gy). The undoped base sodium silicophosphate glass reveals strong charge transfer ultraviolet absorption but with no visible bands. This strong UV absorption is related to the presence of contaminated trace iron impurities (mainly Fe{sup 3+} ions) within the raw materials used for the preparation of this host glass. The MoO{sub 3} doped glasses exhibit extra characteristic absorption bands due to the presence of molybdenum ions in three possible valence states, the trivalent, pentavalent and hexavalent forms. Gamma irradiation of the base undoped glass increases the extension of optical absorption within the UV spectrum and produces an extra broad visible band centered at 520 nm. Such radiation-induced spectra are interpreted by assuming the formation of new induced color centers through the absorption of released electrons and formed positive holes during the irradiation process. Also, the possible formation of induced centers through photochemical transformation of some Fe{sup 2+} ions to Fe{sup 3+} ions by accepting positive holes. The presence of molybdenum ions is assumed to compete with the suggested irradiation reactions by capturing electrons and positive holes during the irradiation process. Infrared absorption spectra of the undoped and MoO{sub 3}-doped glasses reveal broad IR vibrational bands which are attributed to the presence of combined characteristic vibrational IR modes due to main phosphate and partner silicate groups. The addition of MoO{sub 3} (0.5–1.5%) as dopant level causes no changes in the number and position of the main characteristic absorption bands. Gamma irradiation did not cause any marked changes in the IR spectra and the maintainance of the same main IR bands due to the stability of the network containing dual compact two glass

  13. Natural fibre high-density polyethylene and lead oxide composites for radiation shielding

    CERN Document Server

    El-Sayed, A; Ismail, M R

    2003-01-01

    Study has been made of the radiation shielding provided by recycled agricultural fibre and industrial plastic wastes produced as composite materials. Fast neutron and gamma-ray spectra behind composites of fibre-plastic (rho = 1.373 g cm sup - sup 3) and fibre-plastic-lead (rho = 2.756 g cm sup - sup 3) have been measured using a collimated reactor beam and neutron-gamma spectrometer with a stilbene scintillator. The pulse shape discriminating technique based on the zero-cross-over method was used to discriminate between neutron and gamma-ray pulses. Slow neutron fluxes have been measured using a collimated reactor beam and BF sub 3 counter, leading to determination of the macroscopic cross-section (SIGMA). The removal cross-sections (SIGMA sub R) of fast neutrons have been determined from measured results and elemental composition of the composites. For gamma-rays, total linear attenuation coefficients (mu) and total mass attenuation coefficients (mu/rho) have been determined from use of the XCOM code and me...

  14. Current trends in gamma radiation detection for radiological emergency response

    Science.gov (United States)

    Mukhopadhyay, Sanjoy; Guss, Paul; Maurer, Richard

    2011-09-01

    Passive and active detection of gamma rays from shielded radioactive materials, including special nuclear materials, is an important task for any radiological emergency response organization. This article reports on the current trends and status of gamma radiation detection objectives and measurement techniques as applied to nonproliferation and radiological emergencies. In recent years, since the establishment of the Domestic Nuclear Detection Office by the Department of Homeland Security, a tremendous amount of progress has been made in detection materials (scintillators, semiconductors), imaging techniques (Compton imaging, use of active masking and hybrid imaging), data acquisition systems with digital signal processing, field programmable gate arrays and embedded isotopic analysis software (viz. gamma detector response and analysis software [GADRAS]1), fast template matching, and data fusion (merging radiological data with geo-referenced maps, digital imagery to provide better situational awareness). In this stride to progress, a significant amount of inter-disciplinary research and development has taken place-techniques and spin-offs from medical science (such as x-ray radiography and tomography), materials engineering (systematic planned studies on scintillators to optimize several qualities of a good scintillator, nanoparticle applications, quantum dots, and photonic crystals, just to name a few). No trend analysis of radiation detection systems would be complete without mentioning the unprecedented strategic position taken by the National Nuclear Security Administration (NNSA) to deter, detect, and interdict illicit trafficking in nuclear and other radioactive materials across international borders and through the global maritime transportation-the so-called second line of defense.

  15. Internal conversion of gamma radiation

    International Nuclear Information System (INIS)

    Dragoun, O.

    1982-01-01

    The process of the gamma-ray internal conversion is reviewed. The principle of the calculations of the internal conversion coefficients is outlined and methods of conversion electron measurements are described. The extensive utilization of internal conversion in nuclear physics, as well as several applications in chemistry and solid state physics are also discussed. (author)

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

    International Nuclear Information System (INIS)

    Disney, R.K.; Kulikowski, H.D.; McGinnis, C.A.; Reese, J.C.; Thomas, K.; Wiltshire, F.

    1991-01-01

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

  17. Radiation gamma-background at Kurtovo resort

    International Nuclear Information System (INIS)

    Miloslavov, V.

    2000-01-01

    This report presents the difference between the measured values of the radiation gamma-background at Kurtovo resort (located at Rila National Park). The measurements are comparative and are carried out together with the Institute for Nuclear Research (Sofia) and various equipment from the National Center of Radiobiology and Radiation Protection (Sofia). Obtained results are compared according to the precessing method. The advantages of the method for real-time computer precessing of the flowing experimental data on the values of the natural gamma-background are underlined, including the use for early detection of dose increase, due to technological dose implements

  18. Residual water treatment for gamma radiation

    International Nuclear Information System (INIS)

    Mendez, L.

    1990-01-01

    The treatment of residual water by means of gamma radiation for its use in agricultural irrigation is evaluated. Measurements of physical, chemical, biological and microbiological contamination indicators were performed. For that, samples from the treatment center of residual water of San Juan de Miraflores were irradiated up to a 52.5 kGy dose. The study concludes that gamma radiation is effective to remove parasites and bacteria, but not for removal of the organic and inorganic matter. (author). 15 refs., 3 tabs., 4 figs

  19. Sensitiveness of jasmine cuttings to gamma radiation

    International Nuclear Information System (INIS)

    Devaiah, K.A.; Srivastava, H.C.

    1989-01-01

    Half lethal dose (LD 50 ) gamma radiation for five genotypes of jasmine and the effect of such radiation on their rooting parameters were studied. The LD 50 was close to 2.5 krad for Jasminum grandiflorum var. Pink Pin, 0.5 krad for var. Pink Thrum, 2.5 krad for J. flexile Valh., 1 krad for J. calophyllum Wall and 2 krad for J. sambac Ait var. 'Gundumalli'. Percentage of rooting, number of roots per cutting, length and thickness of roots decreased with increase in intensity of gamma irradiation. (author) 8 refs.; 4 tabs

  20. Effects of gamma radiation in annatto seeds

    International Nuclear Information System (INIS)

    Franco, Camilo F. de Oliveira; Arthur, Valter; Arthur, Paula B.; Harder, Marcia N.C.; Filho, Jose C.; Neto, Miguel B.

    2015-01-01

    The annatto bixin has emerged as a major source of natural dyes used in the world notably by the substitution of synthetics harmful to human health and ecologic tendency in obtaining industrial products free of additives with applications in industries textiles; cosmetics; pharmaceutical and food mainly. The aim of this research was to obtain increased of germination rate and dormancy breaking on annatto seeds by gamma radiation. Annatto dry seeds were exposed to low doses of gamma radiation from source of Cobalt-60, type Gammecell-220, at 0.456 kGy/hour dose rate. In order to study stimulation effects of radiation on germination rate and dormancy breaking in the seeds. Five treatments with gamma radiation doses were applied as follows: 0 (control); 100; 125; 150 and 175 Gy. After irradiation the annatto seeds were planted as for usual seed production. According to the results obtained in this experiment we can conclude that the low doses of gamma radiation utilized on the annatto seeds did not presented significantly effect on the germination of plants. But the best dose to increase the germination of seeds was 150 Gy. (author)

  1. Effects of gamma radiation in annatto seeds

    Energy Technology Data Exchange (ETDEWEB)

    Franco, Camilo F. de Oliveira, E-mail: camilo.urucum@hotmail.com [Empresa Brasileira de Pesquisa Agropecuaria (EMBRAPA/EMEPA), Joao Pessoa, PB (Brazil); Arthur, Valter; Arthur, Paula B., E-mail: arthur@cena.usp.br [Centro de Energia Nuclear na Agricultura (CENA/USP), Piracicaba, SP (Brazil); Harder, Marcia N.C., E-mail: marcia.harder@fatec.sp.gov.br [Centro Paula Souza, Curso Superior de Tecnologia em Bicombustiveis (FATEC), Piracicaba, SP (Brazil); Filho, Jose C.; Neto, Miguel B., E-mail: jorgecazefilho@yahoo.com.br [Empresa Estadual de Pesquisa Agropecuaria da Paraiba (EMEPA), Joao Pessoa, PB (Brazil)

    2015-07-01

    The annatto bixin has emerged as a major source of natural dyes used in the world notably by the substitution of synthetics harmful to human health and ecologic tendency in obtaining industrial products free of additives with applications in industries textiles; cosmetics; pharmaceutical and food mainly. The aim of this research was to obtain increased of germination rate and dormancy breaking on annatto seeds by gamma radiation. Annatto dry seeds were exposed to low doses of gamma radiation from source of Cobalt-60, type Gammecell-220, at 0.456 kGy/hour dose rate. In order to study stimulation effects of radiation on germination rate and dormancy breaking in the seeds. Five treatments with gamma radiation doses were applied as follows: 0 (control); 100; 125; 150 and 175 Gy. After irradiation the annatto seeds were planted as for usual seed production. According to the results obtained in this experiment we can conclude that the low doses of gamma radiation utilized on the annatto seeds did not presented significantly effect on the germination of plants. But the best dose to increase the germination of seeds was 150 Gy. (author)

  2. Computational Design of Epoxy/ Boron Carbide Nanocomposites for Radiation Shielding Applications

    Science.gov (United States)

    Bejagam, Karteek; Galehdari, Nasim; Espinosa, Ingrid; Deshmukh, Sanket A.; Kelkar, Ajit D.

    An individual working in industries that include nuclear power plants, healthcare industry, and aerospace are knowingly or unknowingly exposed to radiations of different energies. Exposure to high-energy radiations such as α/ β particle emissions or gamma ray electromagnetic radiations enhances the health risks that can lead to carcinogenesis, cardiac problems, cataracts, and other acute radiation syndromes. The best possible solution to protect one from the exposure to radiations is shielding. In the present study, we have developed a new algorithm to generate a range of different structures of Diglycidyl Ether of Bisphenol F (EPON 862) and curing agent Diethylene Toluene Diamine (DETDA) resins with varying degrees of crosslinking. 3, 5, and 10 weight percent boron carbide was employed as filling materials to study its influence on the thermal and mechanical properties of composite. We further conduct the reactive molecular dynamics (RMD) simulations to investigate the effect of radiation exposure on the structural, physical, and mechanical properties of these Epoxy/Boron Carbide nanocomposites. Where possible the simulation results were compared with the experimental data.

  3. Optimization of NTP System Truss to Reduce Radiation Shield Mass

    Science.gov (United States)

    Scharber, Luke L.; Kharofa, Adam; Caffrey, Jarvis A.

    2016-01-01

    The benefits of nuclear thermal propulsion are numerous and relevant to the current NASA mission goals involving but not limited to the crewed missions to mars and the moon. They do however also present new and unique challenges to the design and logistics of launching/operating spacecraft. One of these challenges, relevant to this discussion, is the significant mass of the shielding which is required to ensure an acceptable radiation environment for the spacecraft and crew. Efforts to reduce shielding mass are difficult to accomplish from material and geometric design points of the shield itself, however by increasing the distance between the nuclear engines and the main body of the spacecraft the required mass of the shielding is lessened considerably. The mass can be reduced significantly per unit length, though any additional mass added by the structure to create this distance serves to offset those savings, thus the design of a lightweight structure is ideal. The challenges of designing the truss are bounded by several limiting factors including; the loading conditions, the capabilities of the launch vehicle, and achieving the ideal truss length when factoring for the overall mass reduced. Determining the overall set of mass values for a truss of varying length is difficult since to maintain an optimally designed truss the geometry of the truss or its members must change. Thus the relation between truss mass and length for these loading scenarios is not linear, and instead has relation determined by the truss design. In order to establish a mass versus length trend for various truss designs to compare with the mass saved from the shield versus length, optimization software was used to find optimal geometric properties that still met the design requirements at established lengths. By solving for optimal designs at various lengths, mass trends could be determined. The initial design findings show a clear benefit to extending the engines as far from the main

  4. Gamma Radiation Doses In Sweden

    International Nuclear Information System (INIS)

    Almgren, Sara; Isaksson, Mats; Barregaard, Lars

    2008-01-01

    Gamma dose rate measurements were performed in one urban and one rural area using thermoluminescence dosimeters (TLD) worn by 46 participants and placed in their dwellings. The personal effective dose rates were 0.096±0.019(1 SD) and 0.092±0.016(1 SD)μSv/h in the urban and rural area, respectively. The corresponding dose rates in the dwellings were 0.11±0.042(1 SD) and 0.091±0.026(1 SD)μSv/h. However, the differences between the areas were not significant. The values were higher in buildings made of concrete than of wood and higher in apartments than in detached houses. Also, 222 Rn measurements were performed in each dwelling, which showed no correlation with the gamma dose rates in the dwellings

  5. Automated TLD system for gamma radiation monitoring

    International Nuclear Information System (INIS)

    Nyberg, P.C.; Ott, J.D.; Edmonds, C.M.; Hopper, J.L.

    1979-01-01

    A gamma radiation monitoring system utilizing a commercially available TLD reader and unique microcomputer control has been built to assess the external radiation exposure to the resident population near a nuclear weapons testing facility. Maximum use of the microcomputer was made to increase the efficiency of data acquisition, transmission, and preparation, and to reduce operational costs. The system was tested for conformance with an applicable national standard for TLD's used in environmental measurements

  6. Automated TLD system for gamma radiation monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Nyberg, P.C.; Ott, J.D.; Edmonds, C.M.; Hopper, J.L.

    1979-01-01

    A gamma radiation monitoring system utilizing a commercially available TLD reader and unique microcomputer control has been built to assess the external radiation exposure to the resident population near a nuclear weapons testing facility. Maximum use of the microcomputer was made to increase the efficiency of data acquisition, transmission, and preparation, and to reduce operational costs. The system was tested for conformance with an applicable national standard for TLD's used in environmental measurements.

  7. Gamma radiation sterilization of Bactrocera invadens (Diptera ...

    African Journals Online (AJOL)

    The African invader fly, Bactrocera invadens, an invasive pest in Africa since 2003, causes damage and poses a threat to the mango and horticultural industry. Its control is therefore needed. Sterilization of males using gamma radiation doses (25, 50 and 75 Gy) as a means of population control was investigated. Irradiation ...

  8. Early Results from the Advanced Radiation Protection Thick GCR Shielding Project

    Science.gov (United States)

    Norman, Ryan B.; Clowdsley, Martha; Slaba, Tony; Heilbronn, Lawrence; Zeitlin, Cary; Kenny, Sean; Crespo, Luis; Giesy, Daniel; Warner, James; McGirl, Natalie; hide

    2017-01-01

    The Advanced Radiation Protection Thick Galactic Cosmic Ray (GCR) Shielding Project leverages experimental and modeling approaches to validate a predicted minimum in the radiation exposure versus shielding depth curve. Preliminary results of space radiation models indicate that a minimum in the dose equivalent versus aluminum shielding thickness may exist in the 20-30 g/cm2 region. For greater shield thickness, dose equivalent increases due to secondary neutron and light particle production. This result goes against the long held belief in the space radiation shielding community that increasing shielding thickness will decrease risk to crew health. A comprehensive modeling effort was undertaken to verify the preliminary modeling results using multiple Monte Carlo and deterministic space radiation transport codes. These results verified the preliminary findings of a minimum and helped drive the design of the experimental component of the project. In first-of-their-kind experiments performed at the NASA Space Radiation Laboratory, neutrons and light ions were measured between large thicknesses of aluminum shielding. Both an upstream and a downstream shield were incorporated into the experiment to represent the radiation environment inside a spacecraft. These measurements are used to validate the Monte Carlo codes and derive uncertainty distributions for exposure estimates behind thick shielding similar to that provided by spacecraft on a Mars mission. Preliminary results for all aspects of the project will be presented.

  9. Use of Existing CAD Models for Radiation Shielding Analysis

    Science.gov (United States)

    Lee, K. T.; Barzilla, J. E.; Wilson, P.; Davis, A.; Zachman, J.

    2015-01-01

    The utility of a radiation exposure analysis depends not only on the accuracy of the underlying particle transport code, but also on the accuracy of the geometric representations of both the vehicle used as radiation shielding mass and the phantom representation of the human form. The current NASA/Space Radiation Analysis Group (SRAG) process to determine crew radiation exposure in a vehicle design incorporates both output from an analytic High Z and Energy Particle Transport (HZETRN) code and the properties (i.e., material thicknesses) of a previously processed drawing. This geometry pre-process can be time-consuming, and the results are less accurate than those determined using a Monte Carlo-based particle transport code. The current work aims to improve this process. Although several Monte Carlo programs (FLUKA, Geant4) are readily available, most use an internal geometry engine. The lack of an interface with the standard CAD formats used by the vehicle designers limits the ability of the user to communicate complex geometries. Translation of native CAD drawings into a format readable by these transport programs is time consuming and prone to error. The Direct Accelerated Geometry -United (DAGU) project is intended to provide an interface between the native vehicle or phantom CAD geometry and multiple particle transport codes to minimize problem setup, computing time and analysis error.

  10. Evaluation of radiation shielding properties of the polyvinyl alcohol/iron oxide polymer composite

    Directory of Open Access Journals (Sweden)

    K Srinivasan

    2017-01-01

    Full Text Available Context: Lead is the conventional shielding material against gamma/X-rays. It has some limitations such as toxic, high density, nonflexibility, and also bremsstrahlung production during electron interaction. It may affect the accuracy of radiotherapy outcome. Aims: To theoretically analyze the radiation shielding properties of flexible polyvinyl alcohol/iron oxide polymer composite with five different concentrations of magnetite over the energy range of 15 KeV–20 MeV. Subjects and Methods: Radiological properties were calculated based on the published literature. Attenuation coefficients of pure elements are generated with the help of WinXCOM database. Results: Effective atomic numbers and electron density are increased with the concentration of magnetite. On the other hand, the number of electrons per gram decreased. Mass attenuation coefficient (μ/ϼ and linear attenuation coefficients (μ are higher in the lower energy <100 KeV, and their values decreased when the energy increased. Computed tomography numbers (CT show the significant variation between the concentrations in <60 KeV. Half-value layer and tenth-value layers are directly proportional to the energy and indirectly proportional to the concentration of magnetite. Transmission curve, relaxation length (ƛ, kinetic energy released in the matter, and elemental weight fraction are also calculated and the results are discussed. Conclusions: 0.5% of the magnetite gives superior shielding properties compared with other concentrations. It may be due to the presence of 0.3617% of Fe. Elemental weight fraction, atomic number, photon energy, and mass densities are the important parameters to understand the shielding behavior of any material.

  11. Observation of the doubly radiative decay eta ' -> gamma gamma pi(0)

    NARCIS (Netherlands)

    Haddadi, Z.; Kalantar-Nayestanaki, N.; Kavatsyuk, M.; Löhner, H.; Messchendorp, J.; Tiemens, M.

    2017-01-01

    Based on a sample of 1.31 billion J/psi events collected with the BESIII detector, we report the study of the doubly radiative decay eta' -> gamma gamma pi(0) for the first time, where the eta' meson is produced via the J/psi -> gamma eta' decay. The branching fraction of eta' -> gamma gamma pi(0)

  12. Development of radiation shielding materials and NBC pads for infantry combat vehicle and tank

    International Nuclear Information System (INIS)

    Pal, R.S.; Gautam, O.P.; Katiyar, Mohit; Tripathi, D.N.; Singh, R.K.

    2008-01-01

    Tanks have special lining materials inside, providing a certain degree of radiation protection for operation in nuclear scenario's. At present these special lining materials in the form of sheets are imported and are fitted into armoured vehicles. Three types of polymer compositions; PE(M)SE, PEC-ISE and PEC-IISE were formulated based on polymer matrix, specific fillers and anti-ageing additives. Prototype NBC pads based on polymer composition PEC-ISE was finalized for moulding of NBC pads for use in ICVs and composition PE(M)SE was finalized for T-90 tanks. The physico-mechanical properties for NBC pads have been evaluated. Radiography of test samples was conducted to ensure homogeneity of specific fillers in the polymer matrix. Radiation shielding factors against nuclear radiation sources ( 60 Co, I37 Cs and 252 Cf) were evaluated at DL Jodhpur and found to be better than imported Russian Pads designed for ICVs and T-90 Tanks. Drawings for twelve types of NBC pads for ICVs and one hundred eighteen types of pads for T-90 tank were generated with the help of design tool, Auto Desk Inventor-II and metallic moulds for moulding of NBC pads were fabricated. Prototype NBC pads were moulded through compression moulding process. Radiation protection factors of prototype NBC pads, after fitment in ICVs, were also evaluated against neutron and gamma (primary and secondary) radiation sources. Prototype NBC pads for ICVs have shown 20% improvement in overall protection level and NBC pads for T-90 tanks have been developed as per design requirements. Manufacturing facility for NBC shielding pads have been established in association with industries. (author)

  13. Possibility of using gamma radiation from HTR reactors for the processing of food and medical products

    International Nuclear Information System (INIS)

    Pahladsingh, R.R.

    2004-01-01

    During the fission process in most of the presently operating nuclear reactors nuclear energy is converted into thermal energy and transferred to common steam cycles for power generation. As part of the fission process also α, β and neutrons particles are released from the nucleus; the release of gamma-rays is also a part of the fission process. In present nuclear reactors α, β, neutrons particles and particularly gamma-rays are not gainfully used as a result of the reactor design and of the containment. These plants are built as required by regulations and international standards for safety. The inherently safe HTR reactor, by its physics and design, does not need a special reinforced containment and it is worth looking into the possibilities of this design feature to use the by-products, such as Gamma-rays, from nuclear fission. In the HTR Pebble Bed Reactors the α, and β particles will remain in the kernels of the pebbles. This means that only the neutron particles and gamma-rays will be available outside the reactor pressure vessel. In this report a proposal is presented to use the gamma-rays of the HTR reactor for irradiation of food and agricultural produce. For neutron shielding a reflector is placed inside the reactor while outside the reactor neutron- and thermal-shielding will be accomplished with water. The high energy gamma-rays will pass through the water-shield and could be harnessed for radiation processing of food and medical products. (author)

  14. FFTF reactor-characterization program: gamma-ray measurements and shield characterization

    Energy Technology Data Exchange (ETDEWEB)

    Bunch, W.L.; Moore, F.S. Jr.

    1983-02-01

    A series of experiments is to be made during the acceptance test program of the Fast Flux Test Facility (FFTF) to measure the gamma ray characteristics of the Fast Test Reactor (FTR) and to establish the performance characteristics of the reactor shield. These measurements are a part of the FFTF Reactor Characterization Program (RCP). Detailed plans have been developed for these experiments. During the initial phase of the Characteristics Program, which will be carried out in the In-Reactor Thimble (IRT), both active and passive measurement methods will be employed to obtain as much information concerning the gamma ray environment as is practical. More limited active gamma ray measurements also will be made in the Vibration Open Test Assembly (VOTA).

  15. Performances of Kevlar and Polyethylene as radiation shielding on-board the International Space Station in high latitude radiation environment.

    Science.gov (United States)

    Narici, Livio; Casolino, Marco; Di Fino, Luca; Larosa, Marianna; Picozza, Piergiorgio; Rizzo, Alessandro; Zaconte, Veronica

    2017-05-10

    Passive radiation shielding is a mandatory element in the design of an integrated solution to mitigate the effects of radiation during long deep space voyages for human exploration. Understanding and exploiting the characteristics of materials suitable for radiation shielding in space flights is, therefore, of primary importance. We present here the results of the first space-test on Kevlar and Polyethylene radiation shielding capabilities including direct measurements of the background baseline (no shield). Measurements are performed on-board of the International Space Station (Columbus modulus) during the ALTEA-shield ESA sponsored program. For the first time the shielding capability of such materials has been tested in a radiation environment similar to the deep-space one, thanks to the feature of the ALTEA system, which allows to select only high latitude orbital tracts of the International Space Station. Polyethylene is widely used for radiation shielding in space and therefore it is an excellent benchmark material to be used in comparative investigations. In this work we show that Kevlar has radiation shielding performances comparable to the Polyethylene ones, reaching a dose rate reduction of 32 ± 2% and a dose equivalent rate reduction of 55 ± 4% (for a shield of 10 g/cm 2 ).

  16. Space Station Validation of Advanced Radiation-Shielding Polymeric Materials, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — In Subtopic X11.01, NASA has identified the need to develop advanced radiation-shielding materials and systems to protect humans from the hazards of space radiation...

  17. Space Station Validation of Advanced Radiation-Shielding Polymeric Materials Project

    Data.gov (United States)

    National Aeronautics and Space Administration — In Subtopic X11.01, NASA has identified the need to develop advanced radiation-shielding materials and systems to protect humans from the hazards of space radiation...

  18. Hydrogen-rich Interpenetrating Polymer Networks for Radiation-Shield Structures, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — In Sub-topic H11.01, NASA has identified a need for advanced radiation-shielding materials and structures to protect humans from space radiation during NASA...

  19. Space Station Validation of Advanced Radiation-Shielding Polymeric Materials, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — In Subtopic X11-01, NASA has identified the need to develop advanced radiation-shielding materials and systems to protect humans from the hazards of space radiation...

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

    Data.gov (United States)

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

  1. Low doses of gamma radiation in soybean

    Energy Technology Data Exchange (ETDEWEB)

    Franco, José G.; Franco, Suely S.H.; Villavicencio, Anna L.C., E-mail: zegilmar60@gmail.com, E-mail: gilmita@uol.com.br, E-mail: villavic@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Arthur, Valter; Arthur, Paula B., E-mail: arthur@cena.usp.br [Centro de Energia Nuclear na Agricultura (CENA/USP), Piracicaba, SP (Brazil); Franco, Caio H. [Universidade Federal de São Paulo (UNIFESP), SP (Brazil). Departamento de Microbiologia, Imunologia e Parasitologia

    2017-07-01

    The degree of radiosensitivity depends mostly on the species, the stage of the embryo at irradiation, the doses employed and the criteria used to measure the effect. One of the most common criteria to evaluate radiosensitivity in seeds is to measure the average plant production. Dry soya seeds were exposed to low doses of gamma radiation from source of Cobalt-60, type Gammecell-220, at 0.210 kGy dose rate. In order to study stimulation effects of radiation on germination, plant growth and production. A treatment with four radiation doses was applied as follows: 0 (control); 12.5; 25.0 and 50.0 Gy. Seed germination and harvested of number of seeds and total production were assessed to identify occurrence of stimulation. Soya seeds number and plants were handled as for usual seed production in Brazil. The low doses of gamma radiation in the seeds that stimulate the production were the doses of 12.5 and 50.0 Gy. The results show that the use of low doses of gamma radiation can stimulate germination and plant production. (author)

  2. Effects of gamma radiation in soybean

    International Nuclear Information System (INIS)

    Franco, Jose Gilmar; Franco, Suely Salumita Haddad; Arthur, Valter; Arthur, Paula Bergamin; Franco, Caio Haddad

    2015-01-01

    The degree of radiosensitivity depends mostly on the species, the stage of the embryo at irradiation, the doses employed and the criteria used to measure the effect. One of the most common criteria to evaluate radiosensitivity in seeds is to measure the average plant production. Soya dry seeds were exposed to low doses of gamma radiation from source of Cobalt-60, type Gammecell-220, at 0.245 kGy dose rate. In order to study stimulation effects of radiation on germination, plant growth and production. Five treatments radiation doses were applied as follows: 0 (control); 25; 50; 75 and 100 Gy. Seed germination and harvest of number of seeds and total production were assessed to identify occurrence of stimulation. Soya seeds and plants were handled as for usual seed production in Brazil. The low doses of gamma radiation in the seeds that stimulate the production were doses of 25, 50 and 75 Gy. There are evidences that the use of low doses of gamma radiation can stimulate germination and plant production. (author)

  3. Low doses of gamma radiation in soybean

    International Nuclear Information System (INIS)

    Franco, José G.; Franco, Suely S.H.; Villavicencio, Anna L.C.; Arthur, Valter; Arthur, Paula B.; Franco, Caio H.

    2017-01-01

    The degree of radiosensitivity depends mostly on the species, the stage of the embryo at irradiation, the doses employed and the criteria used to measure the effect. One of the most common criteria to evaluate radiosensitivity in seeds is to measure the average plant production. Dry soya seeds were exposed to low doses of gamma radiation from source of Cobalt-60, type Gammecell-220, at 0.210 kGy dose rate. In order to study stimulation effects of radiation on germination, plant growth and production. A treatment with four radiation doses was applied as follows: 0 (control); 12.5; 25.0 and 50.0 Gy. Seed germination and harvested of number of seeds and total production were assessed to identify occurrence of stimulation. Soya seeds number and plants were handled as for usual seed production in Brazil. The low doses of gamma radiation in the seeds that stimulate the production were the doses of 12.5 and 50.0 Gy. The results show that the use of low doses of gamma radiation can stimulate germination and plant production. (author)

  4. Effects of gamma radiation in soybean

    Energy Technology Data Exchange (ETDEWEB)

    Franco, Jose Gilmar; Franco, Suely Salumita Haddad; Arthur, Valter; Arthur, Paula Bergamin, E-mail: arthur@cena.usp.br [Centro de Energia Nuclear na Agricultura (CENA/USP), Piracicaba, SP (Brazil); Franco, Caio Haddad [Centro Nacional de Pesquisa em Energia e Materiais (LNBio/CNPEM), Campinas, SP (Brazil). Laboratorio Nacional de Biociencias; Villavicencio, Anna Lucia, E-mail: zegilmar60@gmail.com, E-mail: gilmita@uol.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The degree of radiosensitivity depends mostly on the species, the stage of the embryo at irradiation, the doses employed and the criteria used to measure the effect. One of the most common criteria to evaluate radiosensitivity in seeds is to measure the average plant production. Soya dry seeds were exposed to low doses of gamma radiation from source of Cobalt-60, type Gammecell-220, at 0.245 kGy dose rate. In order to study stimulation effects of radiation on germination, plant growth and production. Five treatments radiation doses were applied as follows: 0 (control); 25; 50; 75 and 100 Gy. Seed germination and harvest of number of seeds and total production were assessed to identify occurrence of stimulation. Soya seeds and plants were handled as for usual seed production in Brazil. The low doses of gamma radiation in the seeds that stimulate the production were doses of 25, 50 and 75 Gy. There are evidences that the use of low doses of gamma radiation can stimulate germination and plant production. (author)

  5. ParShield: A computer program for calculating attenuation parameters of the gamma rays and the fast neutrons

    International Nuclear Information System (INIS)

    Elmahroug, Y.; Tellili, B.; Souga, C.; Manai, K.

    2015-01-01

    Highlights: • Description of the theoretical method used by the ParShield program. • Description of the ParShield program. • Test and validation the ParShield program. - Abstract: This study aims to present a new computer program called ParShield which determines the neutron and gamma-ray shielding parameters. This program can calculate the total mass attenuation coefficients (μ t ), the effective atomic numbers (Z eff ) and the effective electron densities (N eff ) for gamma rays and it can also calculate the effective removal cross-sections (Σ R ) for fast neutrons for mixtures and compounds. The results obtained for the gamma rays by using ParShield were compared with the results calculated by the WinXcom program and the measured results. The obtained values of (Σ R ) were tested by comparing them with the measured results,the manually calculated results and with the results obtained by using MERCSFN program and an excellent agreement was found between them. The ParShield program can be used as a fast and effective tool to choose and compare the shielding materials, especially for the determination of (Z eff ) and (N eff ), there is no other programs in the literature which can calculate

  6. Fabrication of Radiation Shielding Glasses Based on Lead-free High Refractive Index Glasses Prepared from Local Sand

    International Nuclear Information System (INIS)

    Dararutana, Pisutti; Dutchaneepet, Jirapan; Sirikulrat, Narin

    2007-08-01

    Full text: Lead glasses that show high refractive index are the best know and most popular for radiation shielding. Due to harmful effects of lead and considering the health as well as the environmental issues, lead-free glasses were developed. In this work, content of Chumphon sand was fixed at 40 % (by weight) as a main composition but concentrations of BaCO3 were varied from 6 to 30 % (by weight). It was found that the absorption coefficient of the glass samples containing 30 % BaCO3 was 0.233 cm-1 for Ba-133. The density was also measured. It can be concluded that the prepared lead free glasses offered adequate shielding to gamma radiation in comparison with the lead ones. These glasses were one of the environmental friendly materials

  7. Production of radioisotopic gamma radiation sources in JAERI

    International Nuclear Information System (INIS)

    Katoh, Hisashi; Kogure, Hiroto; Suzuki, Kyohei

    1980-04-01

    The present state of production of gamma radiation sources in Japan Atomic Energy Research Institute (JAERI) is described. Sources of 192 Ir, 60 Co and 170 Tm for industrial and 198 Au and 192 Ir for medical applications are produced and delivered routinely by JAERI. Prefabricated assembly targets are irradiated in JRR-2, JRR-3, JRR-4 or JMTR. The irradiated targets are disassembled in a heavy density concrete cave or a lead-shielded cell, depending on the level of radioactivity. The yield of radioactivity in each target is measured with the aid of an ionization chamber. Where necessary, irradiated targets are encapsulated hermetically in capsules of aluminium, stainless steel or other material. The yield of radioactivity is estimated in relation with the burn-up of target nuclide and product nuclide. (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. 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.)

  10. Radiation shielding properties of some natural rocks in upper Egypt

    International Nuclear Information System (INIS)

    Abbady, A.; Ahmed, N.K.; Saied, M.H.; Uosif, M.A.; El-kamel, A.H.

    1999-01-01

    To support the use of some natural rocks in Upper Egypt as suitable radiation materials, the attenuation of gamma - ray through destructive and nondestructive samples of alabaster, marble and limestone have been tested in the energy range from 356 keV to 1173 keV. The attenuation coefficients of the nondestructive samples are found higher than the values of the destructive samples. The half - layer values for attenuation, and the concentration of uranium and thorium in the samples were calculated and discussed

  11. Composites with carbon nanotube for radiation shielding application

    International Nuclear Information System (INIS)

    Fontainha, Críssia C.P.; Nunes, Modesto; Rosas, Víctor A.

    2017-01-01

    Polymeric composites filled with attenuating metals and functionalized with carbon nanotubes (NTC) are being largely developed. New attenuators materials have been widely investigated for radiation shielding to apply in procedures as interventional radiology, Computed Tomography (CT) and nuclear medicine. In this work composites for radiation attenuation in radiodiagnostic imaging procedures made of inorganic material as filler, by a sol-gel method, in poly(vinylidene fluoride-tryfluorethylene) [P(VDF-TrFE] copolymers that are used as the polymeric matrix. Two different metal attenuators were used as fillers: zirconia stabilized by yttria (8% wt.) and bismuth oxide. Carbon nanotubes were added with different concentrations at the solution of attenuator metal under controlled magnetic stirring. Characterization of composites by FTIR, UV-Vis, DSC and SEM-EDS were carried out. In a previous analysis of radiation attenuation, was used an incident monochromatic X-ray beam from the RIGAKU diffractometer. In this setup, one reference measure is directly exposed to the x-rays being diffracted by single crystal of Si (111). Another measure the attenuated beam is performed with the composite sample under detector. The functionalization of the carbon nanotube of multiple walls (MWNCT) in the in the P(VDF-TrFE) was evaluated. The samples present a good dispersion of the attenuator metal into presence at methacrylic acid. The cheap tube presented better dispersion in the polymer matrix than the 3100 nanotubes. Bismuth oxidation composites showed a better attenuation factor compared to Zirconia stabilized by yttria composites. (author)

  12. Composites with carbon nanotube for radiation shielding application

    Energy Technology Data Exchange (ETDEWEB)

    Fontainha, Críssia C.P.; Nunes, Modesto; Rosas, Víctor A., E-mail: crissia@gmail.com [Universidade Federal de Minas Gerais (IMA/UFMG), Belo Horizonte, MG (Brazil). Dept. de Anatomia e Imagem; Santos, Adelina P.; Furtado, Clascídia A.; Faria, Luiz O., E-mail: farialo@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    Polymeric composites filled with attenuating metals and functionalized with carbon nanotubes (NTC) are being largely developed. New attenuators materials have been widely investigated for radiation shielding to apply in procedures as interventional radiology, Computed Tomography (CT) and nuclear medicine. In this work composites for radiation attenuation in radiodiagnostic imaging procedures made of inorganic material as filler, by a sol-gel method, in poly(vinylidene fluoride-tryfluorethylene) [P(VDF-TrFE] copolymers that are used as the polymeric matrix. Two different metal attenuators were used as fillers: zirconia stabilized by yttria (8% wt.) and bismuth oxide. Carbon nanotubes were added with different concentrations at the solution of attenuator metal under controlled magnetic stirring. Characterization of composites by FTIR, UV-Vis, DSC and SEM-EDS were carried out. In a previous analysis of radiation attenuation, was used an incident monochromatic X-ray beam from the RIGAKU diffractometer. In this setup, one reference measure is directly exposed to the x-rays being diffracted by single crystal of Si (111). Another measure the attenuated beam is performed with the composite sample under detector. The functionalization of the carbon nanotube of multiple walls (MWNCT) in the in the P(VDF-TrFE) was evaluated. The samples present a good dispersion of the attenuator metal into presence at methacrylic acid. The cheap tube presented better dispersion in the polymer matrix than the 3100 nanotubes. Bismuth oxidation composites showed a better attenuation factor compared to Zirconia stabilized by yttria composites. (author)

  13. External exposure from gamma radiation in uranium mines

    International Nuclear Information System (INIS)

    Thomson, J.E.

    1982-01-01

    Radiation doses received by workers in a high ore grade uranium mine are compared to those of other radiation workers and the need to be able to calculate the exposure rate from an ore body is indicated. The uranium-238 decay chain is presented and particular reference is made to the main gamma emitters and secular equilibrium of the members of the chain. Difficulties in dealing with a self attenuating volume source, in which scattering is important, are pointed out and traditional methods of solution are mentioned. It is shown that in the special case of an infinite ore body a simple solution may be obtained using the energy conservation principle. A straightforward method for calculating the exposure rate from an arbitrarily shaped ore body is given and corrections due to air attenuation, different soil types and possible lack of secular equilibrium are dealt with. The gamma ray spectrum from the ore is discussed with specific reference to the selection of suitable exposure monitors and the calculation of transmission through shields

  14. Environmental gamma radiation measurements in Bangladeshi houses

    International Nuclear Information System (INIS)

    Idrish Miah, M.

    2002-01-01

    The indoor gamma dose rate in air measured using TLDs in the Dhaka district is not wide ranging and follows a normal distribution with an arithmetic mean of 1.54±0.26 mGy.y -1 . The result has been compared with those found by other investigators for different locations of the world. Measurements were made on a monthly basis for a year period, and a sinusoidal variation of monthly indoor gamma radiation of the type: d = 160 + 65 cos p/6 (m -1 ), where d is the indoor dose rate (nGy.h -1 ) and m the month, was observed. This might be due to seasonally varied air exchange rates of the houses. The average annual effective dose and the collective dose equivalent for the residents were estimated to be 0.86 mSv and 172.20 man-Sv respectively based on the indoor gamma exposure. (author)

  15. Environmental gamma radiation measurements in Bangladeshi houses

    International Nuclear Information System (INIS)

    Miah, M.I.

    2004-01-01

    Indoor gamma dose rate in air measured using TLDs in the Dhaka district is not wide ranging and follows a normal distribution with an arithmetic mean of 1.54±0.26 mGy y -1 . The result has been compared with those found by other investigators for different locations of the world. Measurements were made on a monthly basis for a year period, and a sinusoidal variation of monthly indoor gamma radiation of the type: d=160+65 cos π/6 (m-1), where d is the indoor dose rate (nGy h -1 ) and m the month. This might be due to the seasonally varied air exchange rates of the houses. The average annual effective dose and the collective dose equivalent for the residents were estimated to be 0.86 mSv and 172.20 man-Sv, respectively, based on the indoor gamma exposure

  16. Environmental gamma radiation measurements in Bangladeshi houses

    Energy Technology Data Exchange (ETDEWEB)

    Miah, M.I. E-mail: idrish_physics@yahoo.com

    2004-06-01

    Indoor gamma dose rate in air measured using TLDs in the Dhaka district is not wide ranging and follows a normal distribution with an arithmetic mean of 1.54{+-}0.26 mGy y{sup -1}. The result has been compared with those found by other investigators for different locations of the world. Measurements were made on a monthly basis for a year period, and a sinusoidal variation of monthly indoor gamma radiation of the type: d=160+65 cos {pi}/6 (m-1), where d is the indoor dose rate (nGy h{sup -1}) and m the month. This might be due to the seasonally varied air exchange rates of the houses. The average annual effective dose and the collective dose equivalent for the residents were estimated to be 0.86 mSv and 172.20 man-Sv, respectively, based on the indoor gamma exposure.

  17. Gamma radiation inactivation of Enterococci

    International Nuclear Information System (INIS)

    Huhtanen, C.N.

    1990-01-01

    Radiation survival curves were determined for 7 strains of Enterococcus faecium, 10 strains of E. faecalis, and 8 strains of the proteolytic variety of E. faecalis. The D values (i.e. the doses giving 90% reduction of viable counts) ranged from 5.0-47 kGy for the E. faecium strains, 3.5-21 kGy for the E. faecalis strains, and 3.0-4.5 kGy for the proteolytic variants of E. faecalis. The survival curves were linear for most strains but some exhibited significant non-linear trends

  18. Shielded Heavy-Ion Environment Linear Detector (SHIELD): an experiment for the Radiation and Technology Demonstration (RTD) Mission

    Science.gov (United States)

    Shavers, M. R.; Cucinotta, F. A.; Miller, J.; Zeitlin, C.; Heilbronn, L.; Wilson, J. W.; Singleterry, R. C. Jr

    2001-01-01

    Radiological assessment of the many cosmic ion species of widely distributed energies requires the use of theoretical transport models to accurately describe diverse physical processes related to nuclear reactions in spacecraft structures, planetary atmospheres and surfaces, and tissues. Heavy-ion transport models that were designed to characterize shielded radiation fields have been validated through comparison with data from thick-target irradiation experiments at particle accelerators. With the RTD Mission comes a unique opportunity to validate existing radiation transport models and guide the development of tools for shield design. For the first time, transport properties will be measured in free-space to characterize the shielding effectiveness of materials that are likely to be aboard interplanetary space missions. Target materials composed of aluminum, advanced composite spacecraft structure and other shielding materials, helium (a propellant) and tissue equivalent matrices will be evaluated. Large solid state detectors will provide kinetic energy and charge identification for incident heavy-ions and for secondary ions created in the target material. Transport calculations using the HZETRN model suggest that 8 g cm -2 thick targets would be adequate to evaluate the shielding effectiveness during solar minimum activity conditions for a period of 30 days or more.

  19. A Launch Requirements Trade Study for Active Space Radiation Shielding for Long Duration Human Missions

    Science.gov (United States)

    Singleterry, Robert C., Jr.; Bollweg, Ken; Martin, Trent; Westover, Shayne; Battiston, Roberto; Burger, William J.; Meinke, Rainer

    2015-01-01

    A trade study for an active shielding concept based on magnetic fields in a solenoid configuration versus mass based shielding was developed. Monte Carlo simulations were used to estimate the radiation exposure for two values of the magnetic field strength and the mass of the magnetic shield configuration. For each field strength, results were reported for the magnetic region shielding (end caps ignored) and total region shielding (end caps included but no magnetic field protection) configurations. A value of 15 cSv was chosen to be the maximum exposure for an astronaut. The radiation dose estimate over the total shield region configuration cannot be used at this time without a better understanding of the material and mass present in the end cap regions through a detailed vehicle design. The magnetic shield region configuration, assuming the end cap regions contribute zero exposure, can be launched on a single Space Launch System rocket and up to a two year mission can be supported. The magnetic shield region configuration results in two versus nine launches for a comparable mass based shielding configuration. The active shielding approach is clearly more mass efficient because of the reduced number of launches than the mass based shielding for long duration missions.

  20. Gamma radiation effects on peanut skin antioxidants

    International Nuclear Information System (INIS)

    Camargo, Adriano Costa de; Canniatti-Brazaca, Solange Guidolin; Vieira, Thais Maria Ferreira de Souza; Regitano-d'Arce, Marisa Aparecida Bismara; Calori-Domingues, Maria Antonia

    2011-01-01

    Peanut skin, which is removed in the peanut blanching process, is rich in bioactive compounds with antioxidant properties. The viability of using natural sources of antioxidants to replace synthetic antioxidants was assessed. The aims of this study were to measure bioactive compounds in peanut skins and evaluate the effect of gamma radiation on their antioxidant activity. Peanut skin samples were treated with 0.0, 5.0, 7.5, or 10.0 kGy gamma rays at a dose rate of 7.5 kGy/h using a 60 Co source. Total phenolics, condensed tannins, total flavonoids, and antioxidant activity were evaluated. Extracts obtained from the peanut skins were added to refined-bleached deodorized (RBD) soybean oil that was free from synthetic antioxidants. The oxidative stability of the oil samples was determined using the Rancimat method and compared to a control and synthetic antioxidants (100 mg/kg BHT and 200 mg/kg TBHQ). Gamma radiation changed total phenolic content, total condensed tannins, total flavonoid content, and the antioxidant activity. Ethanolic extracts, gamma irradiated or not, presented increasing induction period (h), measured by the Rancimat method, when compared with the control. Antioxidant activity of the peanut skins was higher than BHT but lower than THBQ. The present study confirmed that gamma radiation did not affect the peanut skin extracts' antioxidative level when added to soybean oil. The induction period of the control soybean oil was 5.7 h, while soybean oil with added ethanolic peanut skin extract had an induction period of 7.2 h, on average. (author)

  1. Gamma radiation effects on peanut skin antioxidants

    Energy Technology Data Exchange (ETDEWEB)

    Camargo, Adriano Costa de [Centro de Energia Nuclear na Agricultura (CENA/USP), Piracicaba, SP (Brazil); Canniatti-Brazaca, Solange Guidolin; Vieira, Thais Maria Ferreira de Souza; Regitano-d' Arce, Marisa Aparecida Bismara; Calori-Domingues, Maria Antonia, E-mail: sgcbraza@usp.b, E-mail: tvieira@esalq.usp.b, E-mail: mabra@esalq.usp.b, E-mail: macdomin@esalq.usp.b [Escola Superior de Agricultura Luiz de Queiroz (ESALQ/USP), Piracicaba, SP (Brazil). Dept. de Agroindustria, Alimentos e Nutricao

    2011-07-01

    Peanut skin, which is removed in the peanut blanching process, is rich in bioactive compounds with antioxidant properties. The viability of using natural sources of antioxidants to replace synthetic antioxidants was assessed. The aims of this study were to measure bioactive compounds in peanut skins and evaluate the effect of gamma radiation on their antioxidant activity. Peanut skin samples were treated with 0.0, 5.0, 7.5, or 10.0 kGy gamma rays at a dose rate of 7.5 kGy/h using a {sup 60}Co source. Total phenolics, condensed tannins, total flavonoids, and antioxidant activity were evaluated. Extracts obtained from the peanut skins were added to refined-bleached deodorized (RBD) soybean oil that was free from synthetic antioxidants. The oxidative stability of the oil samples was determined using the Rancimat method and compared to a control and synthetic antioxidants (100 mg/kg BHT and 200 mg/kg TBHQ). Gamma radiation changed total phenolic content, total condensed tannins, total flavonoid content, and the antioxidant activity. Ethanolic extracts, gamma irradiated or not, presented increasing induction period (h), measured by the Rancimat method, when compared with the control. Antioxidant activity of the peanut skins was higher than BHT but lower than THBQ. The present study confirmed that gamma radiation did not affect the peanut skin extracts' antioxidative level when added to soybean oil. The induction period of the control soybean oil was 5.7 h, while soybean oil with added ethanolic peanut skin extract had an induction period of 7.2 h, on average. (author)

  2. Preliminary study for development of low dose radiation shielding material using liquid silicon and metallic compound

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Seo Goo; Lee, Sung Soo [Dept. of Medical Science, Graduate School of Soonchunhyang University, Asan (Korea, Republic of); Han, Su Chul [Div. of Medical Radiation Equipment, Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of); Kang, Sung Jin [SoonChunHyang University Hospital, Seoul (Korea, Republic of); Lim, Sung Wook [Graduate school of SeJong University, Seoul (Korea, Republic of)

    2017-09-15

    This study measured and compared the protective clothing using Pb used for shielding in a diagnostic X-ray energy range, and the shielding rates of X-ray fusion shielding materials using Si and TiO{sub 2}. For the experiment, a pad type shielding with a thickness of 1 mm was prepared by mixing Si-TiO{sub 2}, and the X-ray shielding rate was compared with 0.5 mmPb plate of The shielding rate of shielding of 0.5 mmPb plate 95.92%, 85.26 % based on the case of no shielding under each 60kVp, 100kVp tube voltage condition. When the shielding of Si-TiO{sub 2} pad was applied, the shielding rate equal to or greater than 0.5 mmPb plate was obtained at a thickness of 11 mm or more, and the shielding rate of 100% or more was confirmed at a thickness of 13 nn in 60kVp condition. When the shielding of Si-TiO{sub 2} pad was applied, the shielding rate equal to or greater than 0.5 mmPb plate was obtained at a thickness of 17 mm or more, and a shielding rate of 0.5 mmPb plate was observed at a thickness of 23 mm in 100kVp condition. Through the results of this study, We could confirm the possibility of manufacturing radiation protective materials that does not contain lead hazard using various metallic compound and liquid Si. This study shows that possibility of liquid Si and other metallic compound can harmonize easily. Beside, It is flexible and strong to physical stress than Pb obtained radiation protective clothes. But additional studies are needed to increase the shielding rate and reduce the weight.

  3. Effect of gamma radiation on polyvinylpyrrolidone hydrogels

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, M.J.A.; Vásquez, P.A.S.; Alcântara, M.T.S.; Munhoz, M.M.L.; Lugão, A.B., E-mail: mariajhho@yahoo.com.br, E-mail: pavsalva@ipen.br, E-mail: ablugao@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    Polyvinylpyrrolidone (PVP) hydrogels have been investigated as drug delivery matrices for the treatment of wounds, such as cutaneous leishmaniasis, and matrices with silver nanoparticles for chronic wounds and burns. The preparation of such hydrogels can occur by various cross-linking methods, such as gamma, chemical, physical, among others. The most feasible for wound dressings is gamma irradiation from cobalt-60, because gamma irradiation simultaneously promotes crosslinking and sterilization, leaving the wound dressing ready for use. The objective of this work was to investigate the effect on physico- chemical properties of gamma radiation on PVP hydrogel according to the radiation absorbed dose variation. The PVP hydrogels were irradiated with doses of 5, 15, 25, 35, 45, 55, 65, 75 and 95kGy at dose rate of 5 kGy/h and characterized by swelling, thermogravimetric and mechanical analysis. Results shown a favorable dose range window for processing of these hydrogels related to the application. The results showed that mechanical strength was affected at doses starting at 25 kGy. (author)

  4. Well logging with natural gamma radiation

    International Nuclear Information System (INIS)

    Ellis, D.V.

    1983-01-01

    An invention is described for use in natural gamma radiation well logging in which measurements taken in a borehole are used in the search for valuable underground resources such as oil or gas. The invention comprises deriving a log of natural gamma radiation detected in selected energy windows for a selected borehole depth interval and converting it into a log of the selected subsurface materials, e.g. Th, U, K. This log is corrected for the effects of 1) either a gamma ray emitter in the borehole fluid, e.g. potassium salts and/or 2) a gamma ray attenuator in the borehole fluid, e.g. a strong attenuator such as barite and/or hematite. The Th, U, K log is particularly useful in the exploration of oil and gas resources since the Th, U, K concentrations are a good indication as to the presence, type and volume of shale and clay in the formations surrounding the borehole. (U.K.)

  5. Development of BaO-ZnO-B2O3 glasses as a radiation shielding material

    Science.gov (United States)

    Chanthima, N.; Kaewkhao, J.; Limkitjaroenporn, P.; Tuscharoen, S.; Kothan, S.; Tungjai, M.; Kaewjaeng, S.; Sarachai, S.; Limsuwan, P.

    2017-08-01

    The effects of the BaO on the optical, physical and radiation shielding properties of the xBaO: 20ZnO: (80-x)B2O3, where x=5, 10, 15, 20 and 25 mol%, were investigated. The glasses were developed by the conventional melt-quenching technique at 1400 °C with high purity chemicals of H3BO3, ZnO, and BaSO4. The optical transparency of the glasses indicated that the glasses samples were high, as observed by visual inspections. The mass attenuation coefficients (μm), the effective atomic numbers (Zeff), and the effective electron densities (Ne) were increased with the increase of BaO concentrations, and the decrease of gamma-ray energy. The developed glass samples were investigated and compared with the shielding concretes and glasses in terms of half value layer (HVL). The overall results demonstrated that the developed glasses had good shielding properties, and highly practical potentials in the environmental friendly radiation shielding materials without an additional of Pb.

  6. Sterilization of activated sludges by gamma radiations

    International Nuclear Information System (INIS)

    Lacroix, J.P.; Boland, M.

    1978-01-01

    The purification process of a wastewater is described. It has been found that the great amount of sludge produced, the average value being 50 g of dry matter per inhabitant and per day, contains three species of pathogen microorganims: enteroviruses, bacteria and parasite eggs and cystes and that this microbial pollution is extremely harmful to human health. Therefore in order to manipulate as well as to use activated sludges in an agriculturad soil, a strong action against pathogen microorganisms is to be taken. Various treatment of sterilization are then described such as: pasteurization, incineration, liming, composting and by gamma radiations. The treatment by gamma radiations has been found to have many advantages in comparison with the other ones. An example of a sterilization plant located in Western Germany is given. (G.C.)

  7. Exploring gamma radiation effect on exoelectron emission properties of bone

    International Nuclear Information System (INIS)

    Zakaria, M.; Dekhtyar, Y.; Bogucharska, T.; Noskov, V.

    2006-01-01

    Gamma radiation is used for radiation therapy to treat carcinogenic diseases including bone cancer. Ionising radiation kills carcinogenic calls. However, there are side effects of the gamma radiation on the bone surface electron structure. One of the effects is in the form of altering electron density of states of bone that, with time, influences biomedical reactions on bone life condition. (authors)

  8. Calculation of gamma-ray attenuation parameters for locally developed shielding material: Polyboron

    Directory of Open Access Journals (Sweden)

    Ripan Biswas

    2016-01-01

    Full Text Available In the present study, the mass attenuation coefficient (μm has been calculated analytically for a locally developed shielding material, polyboron, and compared with the values obtained from the WinXCom code, a Windows version of the XCOM database at the photon energy range 0.001 MeV–20 MeV. A good agreement has been observed between these two values. The linear attenuation coefficients (μ and relaxation lengths (λ have also been calculated from the obtained μm values and their variations with photon energy have been plotted. For comparison, other four shielding materials- ordinary concrete, pure polyethylene, borated polyethylene and water have also been studied. The obtained result shows that μm, μ and λ strongly depends on the photon energy, chemical composition and density of the shielding materials. The values of μm and μ of polyboron have been found greater than those of pure polyethylene and borated polyethylene but less than those of ordinary concrete and water at low photon energy range; and at the intermediate photon energy range (0.125 MeV–6 MeV, all the sample materials have approximately the same μm values. It has also been noticed that polyboron has the medial relaxation length (λ over the entire photon energy range. The total mass attenuation coefficient (μm and linear attenuation coefficient (μ, Half Value Layer (HVL and Tenth Value Layer (TVL of the five sample materials for some common gamma sources have been worked out and the transmission curves have been plotted. The curves exhibit that the transmission factor of the sample materials decreases with the increase in shielding thickness. The results of this study can be utilized to comprehend the shielding effectiveness of this locally developed material.

  9. Effects of gamma radiation on potato meristems

    International Nuclear Information System (INIS)

    Fernandez Gonzalez, J.; Garcia Collantes, M.A.

    1976-01-01

    The development of buds in potato tubers subjected to gamma radiation at doses of 3, 6, 8 and 12 Krad is studied at histological level. The irradiation was supplied at the beguining and end of the resting period, and the irradiated buds were observed at different stages of their development. Meristem's sensitivity depends on the state of activity involved at the moment of irradiation. Different parts of the meristem present different radiosensitivity, being the most radioresistant. (author) [es

  10. Gamma radiation as stressor of tadpole growth

    International Nuclear Information System (INIS)

    Ahmad, M.; Adeeba, S.; Humera, A.; Siddiqui, P.Q.R.

    1979-01-01

    The complicated non-specific neuro-endocrine reactions for the growth of Bufo melanostictus tadpoles have been studied. Significant diversification in the growth pattern of the anuran larvae is brought about by the action of a dose of 0.52 C per kg gamma radiation. The delay in the normal STH response to the radiation stress suggests dissocation between the mechanisms regulating the hypothalamo-hypophyseal activity. The detectable difference between control and test growth seems to be due to the opposing effect of the excessive secretion of ACTH on the STH. (author)

  11. Bibliography, subject index, and author index of the literature examined by the Radiation Shielding Information Center (Reactor and Weapons Radiation Shielding). [1973--1976

    Energy Technology Data Exchange (ETDEWEB)

    1978-01-01

    An indexed bibliography is presented of literature selected by the Radiation Shielding Information Center since the previous volume was published in 1974 in the area of radiation transport and shielding against radiation from nuclear reactors, x-ray machines, radioisotopes, nuclear weapons (including fallout), and low-energy accelerators (e.g., neutron generators). In addition to lists of literature titles by subject categories (accessions 3501-4950), author and keyword indexes are given. Most of the literature selected for Vol. V was published in the years 1973 to 1976.

  12. Bibliography, subject index, and author index of the literature examined by the radiation shielding information center. Volume 6. Reactor and weapons radiation shielding

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-01

    An indexed bibliography is presented of literature selected by the Radiation Shielding Information Center since the previous volume was published in 1978 in the area of radiation transport and shielding against radiation from nuclear reactors, x-ray machines, radioisotopes, nuclear weapons (including fallout), and low energy accelerators (e.g., neutron generators). The bibliography was typeset from data processed by computer from magnetic tape files. In addition to lists of literature titles by subject categories (accessions 4951-6200), an author index is given.

  13. Bibliography, subject index, and author index of the literature examined by the radiation shielding information center. Volume 6. Reactor and weapons radiation shielding

    International Nuclear Information System (INIS)

    1980-05-01

    An indexed bibliography is presented of literature selected by the Radiation Shielding Information Center since the previous volume was published in 1978 in the area of radiation transport and shielding against radiation from nuclear reactors, x-ray machines, radioisotopes, nuclear weapons (including fallout), and low energy accelerators (e.g., neutron generators). The bibliography was typeset from data processed by computer from magnetic tape files. In addition to lists of literature titles by subject categories (accessions 4951-6200), an author index is given

  14. 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

  15. Calculation, planning, and production technology of the radiation shielding means of nuclear power plants in spacecraft

    International Nuclear Information System (INIS)

    Eremin, A.G.; Korobkov, L.S.; Dubinin, A.A.; Pyshko, A.P.; Tkach, K.G.; Bochvar, A.A.

    1993-01-01

    The development of reliable nuclear power plants for spacecraft is an iterative process in which all elements are jointly optimized. In the present development stage, there is a division into quasi-independent problems one of which is the development of radiation shielding means. The solution to this problem is a complicated multistage process involving greatly dissimilar tasks, namely the calculation, the planning, the production, and the quality control of the shielding means. In the initial stage of planning, the research work is concerned with the selection of the optimal rearrangement of the spacecraft and the associated shielding composition. The arrangement selected is thereafter entered into detailed calculations which are to define in greater detail the parameters of the radiation shielding means and the parameters of the radiation fields in the circuitry and the components of the spacecraft. The calculations proceed in parallel with further developments of the design of the shield, which implies determination of optimal design and method meeting the requirements

  16. Thermal, Radiation and Impact Protective Shields (TRIPS) for Robotic and Human Space Exploration Missions

    Science.gov (United States)

    Loomis, M. P.; Arnold, J. L.

    2005-01-01

    New concepts for protective shields for NASA s Crew Exploration Vehicles (CEVs) and planetary probes offer improved mission safety and affordability. Hazards include radiation from cosmic rays and solar particle events, hypervelocity impacts from orbital debris/ micrometeorites, and the extreme heating environment experienced during entry into planetary atmospheres. The traditional approach for the design of protection systems for these hazards has been to create single-function shields, i.e. ablative and blanket-based heat shields for thermal protection systems (TPS), polymer or other low-molecular-weight materials for radiation shields, and multilayer, Whipple-type shields for protection from hypervelocity impacts. This paper introduces an approach for the development of a single, multifunctional protective shield, employing nanotechnology- based materials, to serve simultaneously as a TPS, an impact shield and as the first line of defense against radiation. The approach is first to choose low molecular weight ablative TPS materials, (existing and planned for development) and add functionalized carbon nanotubes. Together they provide both thermal and radiation (TR) shielding. Next, impact protection (IP) is furnished through a tough skin, consisting of hard, ceramic outer layers (to fracture the impactor) and sublayers of tough, nanostructured fabrics to contain the debris cloud from the impactor before it can penetrate the spacecraft s interior.

  17. Issues in space radiation shielding for lunar base

    International Nuclear Information System (INIS)

    Oishi, Koji

    1995-01-01

    Precise estimation of production of secondary neutrons from space radiation particles in the thick shield is very important to define dose rate inside the lunar base. NASA has developed one-dimensional baryon transport code BRYNTRN, which requires only a very small fraction of computer resources. However, for neutrons, backward production and scattering are not modeled in BRYNTRN. Comparisons of the calculated secondary particle spectra in lunar concrete and regolith at the depth of 10, 50, 100, and 200 g/cm 2 between BRYNTRN and Monte-Carlo calculation code system HETC-MCNP were performed. From the comparison, large underestimation of the calculated result of BRYNTRN in the lower neutron energy region, En<10 MeV, were observed. Verification of nuclear data used in MCNP calculation for low energy neutrons were performed, and good agreement between experiment and calculation was obtained. It is concluded that careful consideration for the lower energy neutrons will be required by using BRYNTRN transport code system. (author)

  18. The effect of 60Co gamma radiation on human serum gamma-globulin

    International Nuclear Information System (INIS)

    Portakal, S.

    1984-01-01

    The effect of 60 Co gamma radiation on human serum gamma-globulin was studied in vitro experiments. Solutions of 0.5 percent gamma-globulin were exposed to 0.2, 0.5, 1.0 and 1.9 Mrad doses 60 Co gamma irradiation. Experiments showed that electrophoretic mobility of serum gamma-globulin decreased after gamma irradiation. No significant change in gamma-globulin UV absorption spectrum was observed at 0.2, 0.5, 1.0 and 1.9 Mrad doses. Gamma-globulin becomes progressively less soluble in water as the radiation doses is increased. Radiation induced transformation into insoluble gamma-globulin agregates and scission products. (author)

  19. Efficiency Studies with Gamma Ray Portion of Specialized Reactor-Shield Monte Carlo Program 18-0

    Energy Technology Data Exchange (ETDEWEB)

    Capo, M. A.

    1961-08-01

    Application studies were made with Specialized Reactor-Shield Monte Carlo Program 18-0 to determine the efficiency and feasibility of calculating energy deposition due to primary core gamma rays throughout the XNJ140E-1 reactor-shield assembly. Monte Carlo results are presented in tabular form for all geometrical regions used to describe the shield. Described here is a means of obtaining adequate and valid heating rates in about 47 hours on the IBM-704 digital computer. Comparison of Monte Carlo and point kernel data are included.

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

    Science.gov (United States)

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

    2009-01-01

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

  1. Gamma ray shielding and structural properties of PbO-P2O5-Na2WO4 glass system

    Science.gov (United States)

    Dogra, Mridula; Singh, K. J.; Kaur, Kulwinder; Anand, Vikas; Kaur, Parminder

    2017-05-01

    The present work has been undertaken to study the gamma ray shielding properties of PbO-P2O5-Na2WO4 glass system. The values of mass attenuation coefficient and half value layer parameter at photon energies 511, 662 and 1173 KeV have been determined using XCOM computer software developed by National Institute of Standards and Technology. The density, molar volume, XRD, UV-VIS and Raman studies have been performed to study the structural properties of the prepared glass system to check the possibility of the use of prepared samples as an alternate to conventional concrete for gamma ray shielding applications.

  2. Fabrication, characterization and gamma rays shielding properties of nano and micro lead oxide-dispersed-high density polyethylene composites

    Science.gov (United States)

    Mahmoud, Mohamed E.; El-Khatib, Ahmed M.; Badawi, Mohamed S.; Rashad, Amal R.; El-Sharkawy, Rehab M.; Thabet, Abouzeid A.

    2018-04-01

    Polymer composites of high-density polyethylene (HD-PE) filled with powdered lead oxide nanoparticles (PbO NPs) and bulk lead oxide (PbO Blk) were prepared with filler weight fraction [10% and 50%]. These polymer composites were investigated for radiation-shielding of gamma-rays emitted from radioactive point sources [241Am, 133Ba, 137Cs, and 60Co]. The polymer was found to decrease the heaviness of the shielding material and increase the flexibility while the metal oxide fillers acted as principle radiation attenuators in the polymer composite. The prepared composites were characterized by Fourier transform infrared spectrophotometer (FT-IR), X-ray diffraction (XRD), thermogravimetric analysis (TGA), scanning electron microscope (SEM), Brunauer-Emmett-Teller surface area (BET) and field emission transmission electron microscope (FE-TEM). The morphological analysis of the assembled composites showed that, PbO NPs and PbO Blk materials exhibited homogenous dispersion in the polymer-matrix. Thermogravimetric analysis (TGA) demonstrated that the thermal-stability of HD-PE was enhanced in the presence of both PbO Blk and PbO NPs. The results declared that, the density of polymer composites was increase with the percentage of filler contents. The highest density value was identified as 1.652 g cm-3 for 50 wt% of PbO NPs. Linear attenuation coefficients (μ) have been estimated from the use of XCOM code and measured results. Reasonable agreement was attended between theoretical and experimental results. These composites were also found to display excellent percentage of heaviness with respect to other conventional materials.

  3. Gamma radiation a help to archeological woods

    International Nuclear Information System (INIS)

    Balibar, F.

    1981-01-01

    Waterlogged archeological wood falls into dust once extracted from the water. In order to prevent this destruction several processes have been thought up. In France, the Grenoble Nuclear Study Centre has developed a method of consolidation by resin impregnation and gamma irradiation. The object is first immersed in a vessel containing liquid resin which spreads throughout the wood thereby driving off the water. During the second stage of the treatment, the impregnated objects are irradiated by gamma radiation emitted by a rectangular grid of cobalt 60, so as to polymerize the resin inside the wood. The irradiated objects are Gallo-Roman statuettes discovered during digs at the sources of the river Seine. The wood consolidated right through to the core then becomes sufficiently solid for the restorer to work on the surface of these objects [fr

  4. Gamma radiation sensitivity in Nigella sativa L

    International Nuclear Information System (INIS)

    Datta, A.K.; Biswas, A.K.; Sen, S.

    1986-01-01

    Gamma irradiation induced mutagenic sensitivity of Nigella sativa L. (black cummin) was assessed from R 1 attributes like frequency of seed germination, rate of seedling growth, chromosomal anomalies and sterility types, following exposures of dry dormant seed samples (1.8 % moisture content) to 5 KR, 10 KR, 20 KR, 30 KR, 40 KR, 50 KR and 60 KR of γ-irradiations. Results indicated that gamma radiations have induced both physiological and chromosomal disturbances. LD 50 was found to lie between 20 KR and 30 KR of γ-ray. Treatments beyond 30 KR of irradiation were found to be lethal due to complete failure of emergence of seedlings in the field conditions; although 5 KR has shown stimulation in mitotic index. Sterility types have possibly appeared as an outcome of meiotic disturbances. (author)

  5. Analysis on the steady-state coherent synchrotron radiation with strong shielding

    International Nuclear Information System (INIS)

    Li, R.; Bohn, C.L.; Bisognano, J.J.

    1997-01-01

    There are several papers concerning shielding of coherent synchrotron radiation (CSR) emitted by a Gaussian line charge on a circular orbit centered between two parallel conducting plates. Previous asymptotic analyses in the frequency domain show that shielded steady-state CSR mainly arises from harmonics in the bunch frequency exceeding the threshold harmonic for satisfying the boundary conditions at the plates. In this paper the authors extend the frequency-domain analysis into the regime of strong shielding, in which the threshold harmonic exceeds the characteristic frequency of the bunch. The result is then compared to the shielded steady-state CSR power obtained using image charges

  6. Computer program optimizes design of nuclear radiation shields

    Science.gov (United States)

    Lahti, G. P.

    1971-01-01

    Computer program, OPEX 2, determines minimum weight, volume, or cost for shields. Program incorporates improved coding, simplified data input, spherical geometry, and an expanded output. Method is capable of altering dose-thickness relationship when a shield layer has been removed.

  7. Gamma radiation detectors for safeguards applications

    International Nuclear Information System (INIS)

    Carchon, R.; Moeslinger, M.; Bourva, L.; Bass, C.; Zendel, M.

    2007-01-01

    The IAEA uses extensively a variety of gamma radiation detectors to verify nuclear material. These detectors are part of standardized spectrometry systems: germanium detectors for High-Resolution Gamma Spectrometry (HRGS); Cadmium Zinc Telluride (CZT) detectors for Room Temperature Gamma Spectrometry (RTGS); and NaI(Tl) detectors for Low Resolution Gamma Spectrometry (LRGS). HRGS with high-purity Germanium (HpGe) detectors cooled by liquid nitrogen is widely used in nuclear safeguards to verify the isotopic composition of plutonium or uranium in non-irradiated material. Alternative cooling systems have been evaluated and electrically cooled HpGe detectors show a potential added value, especially for unattended measurements. The spectrometric performance of CZT detectors, their robustness and simplicity are key to the successful verification of irradiated materials. Further development, such as limiting the charge trapping effects in CZT to provide improved sensitivity and energy resolution are discussed. NaI(Tl) detectors have many applications-specifically in hand-held radioisotope identification devices (RID) which are used to detect the presence of radioactive material where a lower resolution is sufficient, as they benefit from a generally higher sensitivity. The Agency is also continuously involved in the review and evaluation of new and emerging technologies in the field of radiation detection such as: Peltier-cooled CdTe detectors; semiconductor detectors operating at room temperature such as HgI 2 and GaAs; and, scintillator detectors using glass fibres or LaBr 3 . A final conclusion, proposing recommendations for future action, is made

  8. Propagation speed of gamma radiation in brass

    Energy Technology Data Exchange (ETDEWEB)

    Cavalcante, Jose T.P.D.; Silva, Paulo R.J.; Saitovich, Henrique

    2009-07-01

    The propagation speed (PS) of visible light -represented by a short frequency range in the large frame of electromagnetic radiations (ER) frequencies- in air was measured during the last century, using a great deal of different methods, with high precision results being achieved. Presently, a well accepted value, with very small uncertainty, is c= 299,792.458 Km/s) (c reporting to the Latin word celeritas: 'speed swiftness'). When propagating in denser material media (MM), such value is always lower when compared to the air value, with the propagating MM density playing an important role. Until present, such studies focusing propagation speeds, refractive indexes, dispersions were specially related to visible light, or to ER in wavelengths ranges dose to it, and with a transparent MM. A first incursion in this subject dealing with {gamma}-rays was performed using an electronic coincidence counting system, when the value of it's PS was measured in air, C{sub {gamma}}{sub (air)}298,300.15 Km/s; a method that went on with later electronic improvements. always in air. To perform such measurements the availability of a {gamma}-radiation source in which two {gamma}-rays are emitted simultaneously in opposite directions -as already used as well as applied in the present case- turns out to be essential to the feasibility of the experiment, as far as no reflection techniques could be used. Such a suitable source was the positron emitter {sup 22}Na placed in a thin wall metal container in which the positrons are stopped and annihilated when reacting with the medium electrons, in such way originating -as it is very well established from momentum/energy conservation laws - two gamma-rays, energy 511 KeV each, both emitted simultaneously in opposite directions. In all the previous experiments were used photomultiplier detectors coupled to NaI(Tl) crystal scintillators, which have a good energy resolution but a deficient time resolution for such purposes

  9. Multilayer Polymeric Shielding to Protect Humans from Galactic Cosmic Radiation, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — In Sub-topic X4.01, NASA has identified a need for advanced radiation-shielding materials and structures to protect humans from the hazards of galactic cosmic...

  10. Innovative, Lightweight Thoraeus RubberTM for MMOD and Space Radiation Shielding, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — NanoSonic offers an innovative manufacturing process to yield ultra-lightweight radiation shielding nanocomposites by exploiting the concept of the Thoraeus filter...

  11. Modeling, Testing and Deploying a Multifunctional Radiation Shielding / Hydrogen Storage Unit, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — This project addresses two vital problems for long-term space travel activities: radiation shielding and hydrogen storage for power and propulsion. While both...

  12. Multifunctional Carbon Nanotube/Polyethylene Complex Composites for Space Radiation Shielding, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Polyethylene (PE), due to its high hydrogen content relative to its weight, has been identified by NASA as a promising radiation shielding material against galactic...

  13. Multilayer Polymeric Shielding to Protect Humans from Galactic Cosmic Radiation, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — In Sub-topic X4.01, NASA has identified a need for advanced radiation-shielding materials and structures to protect humans from the hazards of galactic cosmic...

  14. Multifunctional B/C Fiber Composites for Radiation Shielding, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Radiation shielding is an enabling technology required for extended manned missions to the Moon, Mars and the planets beyond. Multifunctional structural must protect...

  15. Hydrogenous Polymer-Regolith Composites for Radiation-Shielding Materials, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA has identified a need in Sub-topic H11.01 for advanced radiation-shielding technologies using in situ resources, such as regolith, to protect humans from the...

  16. Hydrogenous Polymer-Regolith Composites for Radiation-Shielding Materials, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA has identified a need in Sub-topic H11.01 for advanced radiation-shielding technologies using in situ resources, such as regolith, to protect humans from the...

  17. Application of Advanced Radiation Shielding Materials to Inflatable Structures, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This innovation is a weight-optimized, inflatable structure that incorporates radiation shielding materials into its construction, for use as a habitation module or...

  18. Growth retardation of paramecium and mouse cells by shielding them from background radiation

    International Nuclear Information System (INIS)

    Kawanishi, Masanobu; Okuyama, Katsuyuki; Shiraishi, Kazunori; Matsuda, Yatsuka; Taniguchi, Ryoichi; Shiomi, Nobuyuki; Yonezawa, Morio; Yagi, Takashi

    2012-01-01

    In the 1970s and 1980s, Planel et al. reported that the growth of paramecia was decreased by shielding them from background radiation. In the 1990s, Takizawa et al. found that mouse cells displayed a decreased growth rate under shielded conditions. The purpose of the present study was to confirm that growth is impaired in organisms that have been shielded from background radiation. Radioprotection was produced with a shielding chamber surrounded by a 15 cm thick iron wall and a 10 cm thick paraffin wall that reduced the γ ray and neutron levels in the chamber to 2% and 25% of the background levels, respectively. Although the growth of Paramecium tetraurelia was not impaired by short-term radioprotection (around 10 days), which disagreed with the findings of Planel et al., decreased growth was observed after long-term (40-50 days) radiation shielding. When mouse lymphoma L5178Y cells were incubated inside or outside of the shielding chamber for 7 days, the number of cells present on the 6th and 7th days under the shielding conditions was significantly lower than that present under the non-shielding conditions. These inhibitory effects on cell growth were abrogated by the addition of a 137 Cs γ-ray source disk to the chamber. Furthermore, no growth retardation was observed in XRCC4-deficient mouse M10 cells, which display impaired DNA double strand break repair. (author)

  19. Gamma radiation and HZE treatment of seedlings in Arabidopsis

    Data.gov (United States)

    National Aeronautics and Space Administration — Plants exhibit a robust transcriptional response to gamma radiation which includes the induction of transcripts required for homologous recombination and the...

  20. Optimization of a partially non-magnetic primary radiation shielding for the triple-axis spectrometer PANDA at the Munich high-flux reactor FRM-II

    CERN Document Server

    Pyka, N M; Rogov, A

    2002-01-01

    Monte Carlo simulations have been used to optimize the monochromator shielding of the polarized cold-neutron triple-axis spectrometer PANDA at the Munich high-flux reactor FRM-II. By using the Monte Carlo program MCNP-4B, the density of the total spectrum of incoming neutrons and gamma radiation from the beam tube SR-2 has been determined during the three-dimensional diffusion process in different types of heavy concrete and other absorbing material. Special attention has been paid to build a compact and highly efficient shielding, partially non-magnetic, with a total biological radiation dose of less than 10 mu Sv/h at its outsides. Especially considered was the construction of an albedo reducer, which serves to reduce the background in the experiment outside the shielding. (orig.)

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

    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. The system approach to Clinch River Breeder Reactor Plant (CRBRP) radiation protection is described, and the program developed to implement this approach is defined. In addition, the principal shielding design problems for LMFBR nuclear reactor systems are discussed in relation to LWR nuclear reactor system shielding designs. The methodology used to analyze these problems in the U.S. LMFBR program, the resultant design solutions, and the experimental verification of these designs and/or methods are discussed

  2. Integrated evaluation of the geology, aero gamma spectrometry and aero magnetometry of the Sul-Riograndense Shield, southernmost Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Leo A.; Savian, Jairo F., E-mail: leo.hartmann@ufrgs.br [Universidade Federal do Rio Grande do Sul (UFRS), Porto Alegre, RS (Brazil). Instituto de Geociencias; Lopes, William R. [Servico Geologico do Brasil (CPRM), Porto Alegre, RS (Brazil). Gerencia de Geologia e Mineracao

    2016-03-15

    An integrated evaluation of geology, aero gamma spectrometry and aero magnetometry of the Sul-Riograndense Shield is permitted by the advanced stage of understanding of the geology and geochronology of the southern Brazilian Shield and a 2010 airborne geophysical survey. Gamma rays are registered from the rocks near the surface and thus describe the distribution of major units in the shield, such as the Pelotas batholith, the juvenile São Gabriel terrane, the granulite-amphibolite facies Taquarembo terrane and the numerous granite intrusions in the foreland. Major structures are also observed, e.g., the Dorsal de Cangucu shear. Magnetic signals register near surface crustal compositions (analytic signal) and total crust composition (total magnetic signal), so their variation as measured indicates either shallow or whole crustal structures. The Cacapava shear is outstanding on the images as is the magnetic low along the N-S central portion of the shield. These integrated observations lead to the deepening of the understanding of the largest and even detailed structures of the Sul-Riograndense Shield, some to be correlated to field geology in future studies. Most significant is the presence of different provinces and their limits depending on the method used for data acquisition - geology, aero gamma spectrometry or aero magnetometry. (author)

  3. Radiation Shielding Utilizing A High Temperature Superconducting Magnet Project

    Data.gov (United States)

    National Aeronautics and Space Administration — This project aims to leverage near-term high-temperature superconducting technologies to assess applicability of magnetic shielding for protecting against exposure...

  4. Gamma-ray shielding properties of CaO–SrO–B2O3 glasses

    DEFF Research Database (Denmark)

    Singh, K; Singh, H; Sharma, G

    2005-01-01

    Mass attenuation coefficients, effective atomic numbers and effective electron densities have been determined experimentally for the glass system xCaO . (0.3-x)SrO . 0.7B(2)O(3) at photon energies 511, 662, 1173, and 1332KeV. The results are compared with theoretical calculations. In addition......, the molar volume of the glasses has been derived from density measurements, and the excess volume has been determined as a function of composition. The reported data should be useful for potential applications of these glasses in radiation shielding. (C) 2004 Elsevier Ltd. All rights reserved....

  5. X-ray and radium gamma radiation injuries

    International Nuclear Information System (INIS)

    Fokkema, R.E.

    1993-05-01

    During the period 1896-1939 a number of maxima could be distinguished in the incidence of X-ray and radium gamma ray injuries in patients. An explanation for these fluctuations is investigated in this study. The first distinguishable maximum in the number of reported cases of X-ray injuries can be found in the period 1896-1897 and mainly concerns skin lesions, caused by the lack of shielding and ignorance of the effects. In the period 1904-1905 there was once again an apparent prevalence of radiation injuries to patients. After 1905 the incidence of radiation injuries decreased due to a wider use of dosimetric methods. The third phase of increased injuries may be subdivided into three components. In diagnostic roentgenology from 1896 to 1926 a number of causes of roentgen burns persisted: multiple or long exposures, the use of a short focus-skin-distance and a lack of suitable dosimetric methods. The reduction of complications after 1923 can be attributed to several factors: systematic training of physics who wished to become roentgenologists, greater care of doctors, the use of an alternative method of radiotherapy according to Coutard's method, the introduction of dosimetry with ionization chambers (after 1924), the consensus reached over the roentgen as a unit of applied dosage (in 1928), and the introduction of absorption curves for radiation quality (in 1933). Around 1920 a high complication rate arose as a result of exposure to radiation emitted by radium. In 1922 the first reliable radium dosimetry method came available. This applied to external radium therapy by regular shaped applicators. After 1938 reliable dosimetry was achieved in the field of interstitial radium therapy (brachytherapy). Injuries from radium therapy, however, persisted till about 1940, caused not only by the delayed availability of radium dosimetry, but also to the use of radium therapy by poorly trained radium therapists. 28 figs., 5 tabs

  6. Semiconductor scintillator detector for gamma radiation

    International Nuclear Information System (INIS)

    Laan, F.T.V. der; Borges, V.; Zabadal, J.R.S.

    2015-01-01

    Nowadays the devices employed to evaluate individual radiation exposition are based on dosimetric films and thermoluminescent crystals, whose measurements must be processed in specific transductors. Hence, these devices carry out indirect measurements. Although a new generation of detectors based on semiconductors which are employed in EPD's (Electronic Personal Dosemeters) being yet available, it high producing costs and large dimensions prevents the application in personal dosimetry. Recent research works reports the development of new detection devices based on photovoltaic PIN diodes, which were successfully employed for detecting and monitoring exposition to X rays. In this work, we step forward by coupling a 2mm anthracene scintillator NE1, which converts the high energy radiation in visible light, generating a Strong signal which allows dispensing the use of photomultipliers. A low gain high performance amplifier and a digital acquisition device are employed to measure instantaneous and cumulative doses for energies ranging from X rays to Gamma radiation up to 2 MeV. One of the most important features of the PIN diode relies in the fact that it can be employed as a detector for ionization radiation, since it requires a small energy amount for releasing electrons. Since the photodiode does not amplify the corresponding photon current, it must be coupled to a low gain amplifier. Therefore, the new sensor works as a scintillator coupled with a photodiode PIN. Preliminary experiments are being performed with this sensor, showing good results for a wide range of energy spectrum. (author)

  7. Heavy concrete exerting shielding effects particularly against gamma radiation

    International Nuclear Information System (INIS)

    Valenta, D.; Oravec, J.; Racek, M.

    1990-01-01

    The heavy concrete contains synthetic iron(III) oxide in amounts of 5 to 100% with respect to the aggregate content. The oxide has smooth grains, no more than 4 mm in size. The remaining aggregate has grains up to 32 mm in size and a specific weight of 3500 to 5200 kg.m -3 . The remaining concrete components are cement, water and plasticizer. The mixture is homogeneous and is well suited to feeding by means of concrete pumps. (M.D.)

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  9. Effect of gamma radiation on graphite - PTFE dry lubrication system

    Science.gov (United States)

    Singh, Sachin; Tyagi, Mukti; Seshadri, Geetha; Tyagi, Ajay Kumar; Varshney, Lalit

    2017-12-01

    An effect of gamma radiation on lubrication behavior of graphite -PTFE dry lubrication system has been studied using (TR-TW-30L) tribometer with thrust washer attachment in plane contact. Different compositions of graphite and PTFE were prepared and irradiated by gamma rays. Gamma radiation exposure significantly improves the tribological properties indicated by decrease in coefficient of friction and wear properties of graphite -PTFE dry lubrication system. SEM and XRD analysis confirm the physico-chemical modification of graphite-PTFE on gamma radiation exposure leading to a novel dry lubrication system with good slip and anti friction properties.

  10. Optimisation of the body composition analyser using the gamma shield and improvement of accuracy using ORNL anthropomorphic phantoms.

    Science.gov (United States)

    Araghian, N; Rafat-Motavalli, L; Miri-Hakimabad, H

    2013-07-01

    The analysis of body elements using the prompt gamma rays neutron activation method is a very useful and highly accurate method Lone, M. A., Mughabghab, S. F. and Paviotti-Corcuera, R. Development of a database for prompt gamma-ray neutron activation analysis. Summary Report of the Second Research Coordination Meeting. IAEA Headquarters. IAEA Nuclear Data Section. INDC(NDS)-424. (2001) that has many applications in different fields such as the diagnosis of specific diseases, including certain types of cancers. To protect patients from gamma rays produced by the system, an efficient strategy is to apply a gamma shield. In this study, the gamma shield was placed in three separate positions. The influence of these positions on reducing the effective dose was examined in a 5-y-old Oak Ridge National Laboratory mathematical phantom. Other parameters considered were sensitivity and coefficient of variation (CV) of thermal neutron fluence rate. With the best configuration, the total effective dose per minute (ET) was decreased ∼52 % and the sensitivity was ∼2.03-fold higher than when no shield was present.

  11. Dismountable wall for radiation shielding and screen realized from this wall

    International Nuclear Information System (INIS)

    Blomart, P.

    1988-01-01

    The wall for protection against neutrons and gamma radiations is made of bricks with a shoulder on the upper and side faces and the complementary shape on the lower face to provide a barrier to radiations. Bricks are made of a heavy material for gamma absorption and of epoxy resin, boric acid and hydrated alumina [fr

  12. X-Ray and Gamma-Ray Radiation Detector

    DEFF Research Database (Denmark)

    2015-01-01

    Disclosed is a semiconductor radiation detector for detecting X-ray and / or gamma-ray radiation. The detector comprises a converter element for converting incident X-ray and gamma-ray photons into electron-hole pairs, at least one cathode, a plurality of detector electrodes arranged with a pitch...

  13. The effect of gamma-radiation on bilirubin

    International Nuclear Information System (INIS)

    Iqbal, M.S.; Shad, M.A.; Akhtar, M.I.

    2001-01-01

    The effect of gamma-radiations on bilirubin, in vitro, has been studied. It was found that gamma-radiation causes oxidation of bilirubin to biliverdine as one of the products. The likely implication of this effect in transformation of bilirubin to excretable products, in vino, in case of jaundice is discussed. (author)

  14. Effect of gamma radiation on optical and electrical properties of ...

    Indian Academy of Sciences (India)

    Wintec

    current has, however, been found to decrease with further increase in gamma radiation dose. The observed changes in both the optical and electrical properties indicate that TeO2 thin films can be used as the real time gamma radiation dosimeter up to a certain dose, a quantity that depends upon the thickness of the film.

  15. 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.

  16. 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

  17. Bone densitometry in dogs using gamma radiation

    International Nuclear Information System (INIS)

    Oliveira, A.L.B.; Costa, V.E.; Rezende, M.A.; Grossklauss, D.B.B.F.; Oliveira, T.B.

    2010-01-01

    Full text. The purpose of this work came from the possibility of joining similar methodologies for determination of density, used in different areas, and provide more precise values of bone density by analyzing the mass attenuation coefficient. For over 20 years, The Applied Physics Laboratory, Department of Physics and Biophysics, IBB- UNESP, Botucatu campus, has been working in the determination of density in different areas, using the methods of immersion and gamma radiation attenuation. The results presented have excellent precision, due to the facility in obtaining and preparing samples, coupled to the large experience in the area. This study aims to determine the bone density of samples of mongrel dogs (dogs without defined breed) by the immersion method; to determine the mass attenuation coefficient of bones samples of mongrel dogs with a gamma radiation source; to discuss and to evaluate the methodological aspects involved in the optic densitometry used nowadays, presenting its advantages and disadvantages and, finally, to examine the effect of animal weight, age and sex on bone densitometry of medium-sized dogs. For this study, we use upper limbs samples, at the joint region humerus-radio-ulnar of after death mongrel dogs, assigned by the Department of Pathology, Faculty of Animal Science and Veterinary Medicine (UNESP-Botucatu) and by the Kennel of the city of Araras, Sao Paulo. This work is performed in three stages. In the first step is determined the density by the method of immersion in water, in the second step, is obtained the mass coefficient attenuation and, finally, in the third step are discussed the implemented methods and evaluate the density bone samples to establish correlations with the age, weight and sex parameters of each group of animals. Based on this methodology , we can find the average value for the mass attenuation coefficient of gamma radiation with energy 59,6, find variations in the values of bone densitometry in the same bone

  18. Bone densitometry in dogs using gamma radiation

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, A.L.B.; Costa, V.E.; Rezende, M.A.; Grossklauss, D.B.B.F.; Oliveira, T.B. [UNESP, Botucatu, SP (Brazil)

    2010-07-01

    Full text. The purpose of this work came from the possibility of joining similar methodologies for determination of density, used in different areas, and provide more precise values of bone density by analyzing the mass attenuation coefficient. For over 20 years, The Applied Physics Laboratory, Department of Physics and Biophysics, IBB- UNESP, Botucatu campus, has been working in the determination of density in different areas, using the methods of immersion and gamma radiation attenuation. The results presented have excellent precision, due to the facility in obtaining and preparing samples, coupled to the large experience in the area. This study aims to determine the bone density of samples of mongrel dogs (dogs without defined breed) by the immersion method; to determine the mass attenuation coefficient of bones samples of mongrel dogs with a gamma radiation source; to discuss and to evaluate the methodological aspects involved in the optic densitometry used nowadays, presenting its advantages and disadvantages and, finally, to examine the effect of animal weight, age and sex on bone densitometry of medium-sized dogs. For this study, we use upper limbs samples, at the joint region humerus-radio-ulnar of after death mongrel dogs, assigned by the Department of Pathology, Faculty of Animal Science and Veterinary Medicine (UNESP-Botucatu) and by the Kennel of the city of Araras, Sao Paulo. This work is performed in three stages. In the first step is determined the density by the method of immersion in water, in the second step, is obtained the mass coefficient attenuation and, finally, in the third step are discussed the implemented methods and evaluate the density bone samples to establish correlations with the age, weight and sex parameters of each group of animals. Based on this methodology , we can find the average value for the mass attenuation coefficient of gamma radiation with energy 59,6, find variations in the values of bone densitometry in the same bone

  19. Gamma radiation effects on nestling Tree Swallows

    International Nuclear Information System (INIS)

    Zach, R.; Mayoh, K.R.

    1984-01-01

    The sensitivity of Tree Swallows (Tachycineta bicolor) to the stress of ionizing radiation was investigated with growth analysis. Freshly hatched nestlings were temporarily removed from nests, taken to the laboratory and acutely exposed to 0.9, 2.7, or 4.5 Gy gamma radiation. Some of the unirradiated control nestlings were also taken to the laboratory whereas others were left in the nests. Growth of all the nestlings was measured daily and analyzed by fitting growth models. There was no detectable radiation-induced mortality up to fledgling, approx. = 20 d after irradiation. Radiation exposure did not affect the basic growth pattern; the logistic growth model was most suitable for body mass and foot length, and the von Bertalanffy model for primary-feather length, irrespective of treatment. Parameter values from these models indicated pronounced growth depression in the 2.7-Gy and 4.5-Gy groups, particularly for body mass. Radiation also affected the timing of development. The growth depression of the 2.7-Gy group was similar to that caused by hatching asynchrony in unirradiated nestlings. The 4.5-Cy nestlings grew as well as unexposed nestlings that died from natural causes. Chronic irradiation at approx. = 1.0 Cy/d caused more severe growth effects than acute exposure to 4.5 Gy and may have caused permanent stunting. Growth analysis is a potent tool for assessing man-made environmental stresses. Observed body-mass statistics and model parameters seem to be most sensitive to environmental stresses, but coefficients of variation are not necessarily correlated with sensitivity. 34 references, 2 figures, 4 tables

  20. Radiation shielding and dose rate evaluation at the interim storage facility for spent fuel from Cernavoda NPP

    International Nuclear Information System (INIS)

    Stanciu, Marcela; Mateescu, Silvia; Pantazi, Doina; Penescu, Maria

    2000-01-01

    At present studies necessary to license the Interim Storage Facility for the Spent Fuel (CANDU type) from Cernavoda NPP are developed in our country.The spent fuel from Cernavoda NPP is discharged into Spent Fuel Bay in Service Building of the plant, where it remains several years for cooling. After this period, the bundles of spent fuel are to be transferred to the Interim Storage Facility.The dry interim storage solution seems to be the most appropriate variant for Cernavoda NPP.The design of the Spent Fuel Interim Storage Facility must meet the applicable safety requirements in order to ensure radiological protection of the personnel, public and environment during all phases of the facility achievement. In this paper we intend to present the calculation of radiation shielding at the spent fuel interim storage facility for two technical solutions: - Concrete Monolithic Module and Concrete Storage Cask. In order to quantify the fuel composition after irradiation, the isotope generation and depletion code ORIGEN 2.1 has been used, taking into account a cooling time of 7 years and 9 years, respectively, for these two variants. The shielding calculations have been performed using the computer codes QAD-5K and MICROSHIELD-4. The evaluations refer only to gamma radiation because the resulting neutron source (from (α,n) reactions and spontaneous fission) is insignificant as compared to the gamma source. The final results consist in the minimum thickness of the shielding and the corresponding external dose rates, ensuring a design average dose rate based on national and international regulations. (authors)

  1. AA, radiation shielding curtain along the target area

    CERN Multimedia

    CERN PhotoLab

    1980-01-01

    At the far left is the beam tube for the high-intensity proton beam from the 26 GeV PS. The tube ends in a thin window and the proton beam continues in air through a hole in the shielding blocks (see also 8010308), behind which the target (see 7905091, 7905094)was located. After the target followed the magnetic horn, focusing the antiprotons, and the first part of the injection line with a proton dump. The antiprotons, deflected by a magnet, left the target area through another shielding wall, to make their way to the AA ring. Laterally, this sequence of components was shielded with movable, suspended, concrete blocks: the "curtain". Balasz Szeless, who had constructed it, is standing at its side.

  2. Effects of gamma radiation on rabbit lens

    International Nuclear Information System (INIS)

    Ordahl, J.N.; Gorthy, W.C.

    1982-01-01

    Eyes of young New Zealand white rabbits were irradiated with 2000 rads of gamma radiation ( 60 Co) and the eyes removed at 2, 8, 16, and 30 days post-irradiation for electron microscopic analysis. Lenses were treated histochemically for acid phosphatase localization to examine the role of lysosomal enzymes in the early development of radiation cataracts. Intercellularly located acid phosphatase reaction product, noted in the epithelium and subjacent cortex, was more prevalent after irradiation, especially in central and transitional epithelial zones. In the central zone small vesicles typically were most numerous in areas of extensive intercellular reaction product. The occurrence of discrete packets of reaction product within the vesicles and the adjoining intercellular space, plus the resemblance of the peripheral vesicles to small Golgi vesicles also containing reaction product, suggested an exocytotic release of the enzyme. These appearances suggest that lysosomal hydrolases are released extracellularly by a secretory mechanism accelerated by radiation and that these hydrolases may play a role in both physiological and pathological functions of the lens. (author)

  3. Investigation of zones with increased ground surface gamma radiation

    International Nuclear Information System (INIS)

    Butkus, D.V.; Morkunas, G.S.; Styro, B.I.

    1989-01-01

    Measurements of the increased gamma radiation zones of soils were conducted in the South-Western part of the Litvinian. The shores of lakes in the north-eastern part of the Suduva high land were investigated. the maximum values of the gamma radiation dose rates were distributed along the lake shores at a distance of 1 m from the water surface, while farther than 1.5 m from it the dose rate was close to the natural value. The increased gamma radiation intensity zones on the ground surface were found only at the northern (Lake Reketija) or the western shore (other lakes under investigation). The highest values of the gamma radiation dose 200-600 μR/h (0.5-1.5 nGy/s) were observed in the comparatively small areas (up to several square metres). The gamma radiation intensity of soil surface increased strongly moving towards the point where the maximum intensity was obsered. 10 figs

  4. Differential sensitivity of Chironomus and human hemoglobin to gamma radiation

    International Nuclear Information System (INIS)

    Gaikwad, Pallavi S.; Panicker, Lata; Mohole, Madhura; Sawant, Sangeeta; Mukhopadhyaya, Rita; Nath, Bimalendu B.

    2016-01-01

    Chironomus ramosus is known to tolerate high doses of gamma radiation exposure. Larvae of this insect possess more than 95% of hemoglobin (Hb) in its circulatory hemolymph. This is a comparative study to see effect of gamma radiation on Hb of Chironomus and humans, two evolutionarily diverse organisms one having extracellular and the other intracellular Hb respectively. Stability and integrity of Chironomus and human Hb to gamma radiation was compared using biophysical techniques like Dynamic Light Scattering (DLS), UV-visible spectroscopy, fluorescence spectrometry and CD spectroscopy after exposure of whole larvae, larval hemolymph, human peripheral blood, purified Chironomus and human Hb. Sequence- and structure-based bioinformatics methods were used to analyze the sequence and structural similarities or differences in the heme pockets of respective Hbs. Resistivity of Chironomus Hb to gamma radiation is remarkably higher than human Hb. Human Hb exhibited loss of heme iron at a relatively low dose of gamma radiation exposure as compared to Chironomus Hb. Unlike human Hb, the heme pocket of Chironomus Hb is rich in aromatic amino acids. Higher hydophobicity around heme pocket confers stability of Chironomus Hb compared to human Hb. Previously reported gamma radiation tolerance of Chironomus can be largely attributed to its evolutionarily ancient form of extracellular Hb as evident from the present study. -- Highlights: •Comparison of radiation tolerant Chironomus Hb and radiation sensitive Human Hb. •Amino acid composition of midge and human heme confer differential hydrophobicity. •Heme pocket of evolutionarily ancient midge Hb provide gamma radiation resistivity.

  5. Differential sensitivity of Chironomus and human hemoglobin to gamma radiation

    Energy Technology Data Exchange (ETDEWEB)

    Gaikwad, Pallavi S. [Stress Biology Research Laboratory, Department of Zoology, Savitribai Phule University, Pune, 411007 (India); Molecular Biology Division, Bhabha Atomic Research Centre, Trombay, Mumbai, 400085 (India); Panicker, Lata [Solid State Physics Division, Bhabha Atomic Research Centre, Trombay, Mumbai, 400085 (India); Mohole, Madhura; Sawant, Sangeeta [Bioinformatics Center, Savitribai Phule Pune University, Pune, 411007 (India); Mukhopadhyaya, Rita [Molecular Biology Division, Bhabha Atomic Research Centre, Trombay, Mumbai, 400085 (India); Nath, Bimalendu B., E-mail: bbnath@gmail.com [Stress Biology Research Laboratory, Department of Zoology, Savitribai Phule University, Pune, 411007 (India)

    2016-08-05

    Chironomus ramosus is known to tolerate high doses of gamma radiation exposure. Larvae of this insect possess more than 95% of hemoglobin (Hb) in its circulatory hemolymph. This is a comparative study to see effect of gamma radiation on Hb of Chironomus and humans, two evolutionarily diverse organisms one having extracellular and the other intracellular Hb respectively. Stability and integrity of Chironomus and human Hb to gamma radiation was compared using biophysical techniques like Dynamic Light Scattering (DLS), UV-visible spectroscopy, fluorescence spectrometry and CD spectroscopy after exposure of whole larvae, larval hemolymph, human peripheral blood, purified Chironomus and human Hb. Sequence- and structure-based bioinformatics methods were used to analyze the sequence and structural similarities or differences in the heme pockets of respective Hbs. Resistivity of Chironomus Hb to gamma radiation is remarkably higher than human Hb. Human Hb exhibited loss of heme iron at a relatively low dose of gamma radiation exposure as compared to Chironomus Hb. Unlike human Hb, the heme pocket of Chironomus Hb is rich in aromatic amino acids. Higher hydophobicity around heme pocket confers stability of Chironomus Hb compared to human Hb. Previously reported gamma radiation tolerance of Chironomus can be largely attributed to its evolutionarily ancient form of extracellular Hb as evident from the present study. -- Highlights: •Comparison of radiation tolerant Chironomus Hb and radiation sensitive Human Hb. •Amino acid composition of midge and human heme confer differential hydrophobicity. •Heme pocket of evolutionarily ancient midge Hb provide gamma radiation resistivity.

  6. Adaptive planning using megavoltage fan-beam CT for radiation therapy with testicular shielding

    Energy Technology Data Exchange (ETDEWEB)

    Yadav, Poonam [Department of Human Oncology, University of Wisconsin, Madison, Madison, WI (United States); Department of Medical Physics, University of Wisconsin, Madison, Madison, WI (United States); School of Advance Sciences, Vellore Institue of Technology University, Vellore, Tamil Nadu (India); Kozak, Kevin [Department of Human Oncology, University of Wisconsin, Madison, Madison, WI (United States); Tolakanahalli, Ranjini [Department of Human Oncology, University of Wisconsin, Madison, Madison, WI (United States); Department of Medical Physics, University of Wisconsin, Madison, Madison, WI (United States); Ramasubramanian, V. [School of Advance Sciences, Vellore Institue of Technology University, Vellore, Tamil Nadu (India); Paliwal, Bhudatt R. [Department of Human Oncology, University of Wisconsin, Madison, Madison, WI (United States); Department of Medical Physics, University of Wisconsin, Madison, Madison, WI (United States); University of Wisconsin, Riverview Cancer Centre, Wisconsin Rapids, WI (United States); Welsh, James S. [Department of Human Oncology, University of Wisconsin, Madison, Madison, WI (United States); Department of Medical Physics, University of Wisconsin, Madison, Madison, WI (United States); Rong, Yi, E-mail: rong@humonc.wisc.edu [Department of Human Oncology, University of Wisconsin, Madison, Madison, WI (United States); University of Wisconsin, Riverview Cancer Centre, Wisconsin Rapids, WI (United States)

    2012-07-01

    This study highlights the use of adaptive planning to accommodate testicular shielding in helical tomotherapy for malignancies of the proximal thigh. Two cases of young men with large soft tissue sarcomas of the proximal thigh are presented. After multidisciplinary evaluation, preoperative radiation therapy was recommended. Both patients were referred for sperm banking and lead shields were used to minimize testicular dose during radiation therapy. To minimize imaging artifacts, kilovoltage CT (kVCT) treatment planning was conducted without shielding. Generous hypothetical contours were generated on each 'planning scan' to estimate the location of the lead shield and generate a directionally blocked helical tomotherapy plan. To ensure the accuracy of each plan, megavoltage fan-beam CT (MVCT) scans were obtained at the first treatment and adaptive planning was performed to account for lead shield placement. Two important regions of interest in these cases were femurs and femoral heads. During adaptive planning for the first patient, it was observed that the virtual lead shield contour on kVCT planning images was significantly larger than the actual lead shield used for treatment. However, for the second patient, it was noted that the size of the virtual lead shield contoured on the kVCT image was significantly smaller than the actual shield size. Thus, new adaptive plans based on MVCT images were generated and used for treatment. The planning target volume was underdosed up to 2% and had higher maximum doses without adaptive planning. In conclusion, the treatment of the upper thigh, particularly in young men, presents several clinical challenges, including preservation of gonadal function. In such circumstances, adaptive planning using MVCT can ensure accurate dose delivery even in the presence of high-density testicular shields.

  7. Gamma-ray Cherenkov-transition radiation

    Science.gov (United States)

    Aginian, M. A.; Ispirian, K. A.; Ispiryan, M.

    2013-10-01

    The spectral and angular distributions as well as the total number of photons of gamma-ray Cherenkov-transition radiation (GCTR) produced by charged particles in the photon energy region {}\\sim(0.8\\text{-}2)\\ \\text{MeV} are calculated. For this purpose we used the experimental results of the recent discovery according to which in the above-mentioned region the measured refractive index of silicon as well as the theoretically calculated refractive index of gold are greater than 1. Using the results of the carried out numerical calculations an experimental arrangement is discussed for the observation and experimental study of the GCTR. As our results show the GCTR photon yield is about one order of magnitude higher than the background bremsstrahlung yield. Some applications of GCTR, in particular, for comparatively easy search of new materials with refractive index n(\\omega )>1 , are proposed.

  8. Gamma radiation effects on liofilized human serum

    International Nuclear Information System (INIS)

    Padron Soler, E.; Romay Penabad, Z.; Chavez Ardanza, A.; Hernandez Gonzalez, C.; Martin Gonzalez, O.; Garcia Gonzalez, I.; Prieto Miranda, E.

    1995-01-01

    Human freeze dried serum was artificially contaminated with Flavobacterium sp. for studying the effects of gamma radiation of it. The radiobiological parameters of the contaminator were determined and the sterilization dose was set. The quality of the product irradiated at both, calculated sterilization dose (8.5 kGy) an another one about 25 kGy was determined. It was made according to: sterility testing, total proteins, pH enzymes (alanina-aminotransferase, aspartato-aminotransferase, alkaline phosphatase), protein electrophoresis, fast performance liquid chromatographic and effect on the cellular growth. From the latter was concluded that the calculated sterilization dose was adequate form keeping the biological properties and viability of the irradiated serum. Nevertheless, the dose of 25 k Gy was not adequate because of its dangerous effects on the cell culture

  9. Orchid flowers tolerance to gamma-radiation

    Energy Technology Data Exchange (ETDEWEB)

    Kikuchi, Olivia Kimiko E-mail: okikuchi@net.ipen.br

    2000-03-01

    Cut flowers are fresh goods that may be treated with fumigants such as methyl bromide to meet the needs of the quarantine requirements of importing countries. Irradiation is a non-chemical alternative to substitute the methyl bromide treatment of fresh products. In this research, different cut orchids were irradiated to examine their tolerance to gamma-rays. A 200 Gy dose did inhibit the Dendrobium palenopsis buds from opening, but did not cause visible damage to opened flowers. Doses of 800 and 1000 Gy were damaging because they provoked the flowers to drop from the stem. Cattleya irradiated with 750 Gy did not show any damage, and were therefore eligible for the radiation treatment. Cymbidium tolerated up to 300 Gy and above this dose dropped prematurely. On the other hand, Oncydium did not tolerate doses above 150 Gy.(author)

  10. Orchid flowers tolerance to gamma-radiation

    Science.gov (United States)

    Kikuchi, Olivia Kimiko

    2000-03-01

    Cut flowers are fresh goods that may be treated with fumigants such as methyl bromide to meet the needs of the quarantine requirements of importing countries. Irradiation is a non-chemical alternative to substitute the methyl bromide treatment of fresh products. In this research, different cut orchids were irradiated to examine their tolerance to gamma-rays. A 200 Gy dose did inhibit the Dendrobium palenopsis buds from opening, but did not cause visible damage to opened flowers. Doses of 800 and 1000 Gy were damaging because they provoked the flowers to drop from the stem. Cattleya irradiated with 750 Gy did not show any damage, and were therefore eligible for the radiation treatment. Cymbidium tolerated up to 300 Gy and above this dose dropped prematurely. On the other hand, Oncydium did not tolerate doses above 150 Gy.

  11. Gamma Radiation-Induced Mutations in Soybeans

    International Nuclear Information System (INIS)

    Smutkupt, S.

    1998-01-01

    The main objective of soybean radiation experiments was to create genetic variability in soybeans of various cultivars, mutants and mutation-derived lines with the aim of producing superior breeding lines with resistance to soybean rust (Phakopsora pachyhrizi Syd.) It took altogether 12 generations over six years after gamma irradiation if soybean seeds to produce the best resistant line (81-1-038) which a variety could be developed from it. This Line 81-1-038 showed a very good specific resistance to soybean rust, Thai race 2 and moderately resistance to Thai race 1. In the rainy season of 1985, Line 81-1-038 out yielded S.J.4 (a mother line) by 868 kg/ha in a yield trail at Suwan Farm, Pak Chong, Nakorn Rajchasima. This soybean rust mutant was later named D oi Kham

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

    Science.gov (United States)

    Barthel, Joseph; Sarigul-Klijn, Nesrin

    2018-03-01

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

  13. Experimental investigation of the radiation shielding of a MCP detector in the radiation environment near Europa

    Science.gov (United States)

    Tulej, Marek; Wurz, Peter; Meyer, Stefan; Lasi, Davide; Lüthi, Matthias; Galli, André; Piazza, Daniele; Desorgher, Laurent; Hajdas, Wojciech; Reggiani, Davide; Karlsson, Stefan; Kalla, Leif

    2016-04-01

    The Neutral Ion Mass spectrometer (NIM) is one of the six instruments in the Particle Environmental Package (PEP) designed for the JUICE mission of ESA to the Jupiter system. NIM will conduct detailed measurements of chemical composition of Jovian moon exospheres and is equipped with a sensitive MCP ion detector. To maintain high sensitivity of the NIM instrument, background signals arising from the presence of a large background of penetrating radiation (mostly high-energy electrons and protons) in Jupiter's magnetosphere have to be minimised. We investigate the performance of a layered-Z radiation shield, an Al-Ta-Al sandwich, as a potential shielding against high-energy electrons. The experimental investigations were performed at the PiM1 beam line of the High Intensity Proton Accelerator Facilities located at the Paul Scherrer Institute (PSI), Villigen, Switzerland. The facility delivers a particle beam containing e,  and  with an adjustable momentum ranging from 17.5 to 345 MeV/c. The measurements of the induced radiation background generated during the interaction of primary particles with Al-Ta-Al sandwich were conducted by beam diagnostic methods and a MCP detector. Diagnostic methods provided for the characterisation of the beam parameters (beam geometry, flux and intensity) and identification of individual particles in the primary beam and in the flux of secondary particles. The MCP detector measurements provided information on the effects of radiation and the results of these measurements define the performance of the shielding material in reducing the background arising from penetrating radiation. In parallel, we performed modelling studies using GEANT 4 and GRASS methods to identify products of the interaction and predict their fluxes and particle rates at the MCP detector. Combination of the experiment and modelling studies yields detailed characterisation of the radiation effects produced by the interaction of the incident e- in the

  14. Modeling gamma radiation dose in dwellings due to building materials.

    Science.gov (United States)

    de Jong, Peter; van Dijk, Willem

    2008-01-01

    A model is presented that calculates the absorbed dose rate in air of gamma radiation emitted by building materials in a rectangular body construction. The basis for these calculations is formed by a fixed set of specific absorbed dose rates (the dose rate per Bq kg(-1) 238U, 232Th, and 40K), as determined for a standard geometry with the dimensions 4 x 5 x 2.8 m3. Using the computer codes Marmer and MicroShield, correction factors are assessed that quantify the influence of several room and material related parameters on the specific absorbed dose rates. The investigated parameters are the position in the construction; the thickness, density, and dimensions of the construction parts; the contribution from the outer leave; the presence of doors and windows; the attenuation by internal partition walls; the contribution from building materials present in adjacent rooms; and the effect of non-equilibrium due to 222Rn exhalation. To verify the precision, the proposed method is applied to three Dutch reference dwellings, i.e., a row house, a coupled house, and a gallery apartment. The averaged difference with MCNP calculations is found to be 4%.

  15. Study and application of high-density concrete in radiation-shielding experiment

    International Nuclear Information System (INIS)

    Wu Chongming; Ding Dexin; Xiao Xuefu; Wang Shaolin; Lin Xingjun; Shen Yuanyuan

    2008-01-01

    According to the demand for research and construction project, a series of systematic experiments and studies on shielding γ-ray radiation concrete with the density of 4.60 t/m 3 were made in such aspects as mix ratio design, construction technology, uniformly shielding etc. Such issues as uniformity in the construction and compactness were solved. The ray test method for uniformly shielding concrete was presented and some technical steps for this high-density concrete used in the process of test design or construction were summed up. A series of tests and practical applications show that this technology of mix ratio design and construction is feasible. (authors)

  16. Radiation Shielding of Lunar Regolith/Polyethylene Composites and Lunar Regolith/Water Mixtures

    Science.gov (United States)

    Johnson, Quincy F.; Gersey, Brad; Wilkins, Richard; Zhou, Jianren

    2011-01-01

    Space radiation is a complex mixed field of ionizing radiation that can pose hazardous risks to sophisticated electronics and humans. Mission planning for lunar exploration and long duration habitat construction will face tremendous challenges of shielding against various types of space radiation in an attempt to minimize the detrimental effects it may have on materials, electronics, and humans. In late 2009, the Lunar Crater Observation and Sensing Satellite (LCROSS) discovered that water content in lunar regolith found in certain areas on the moon can be up to 5.6 +/-2.8 weight percent (wt%) [A. Colaprete, et. al., Science, Vol. 330, 463 (2010). ]. In this work, shielding studies were performed utilizing ultra high molecular weight polyethylene (UHMWPE) and aluminum, both being standard space shielding materials, simulated lunar regolith/ polyethylene composites, and simulated lunar regolith mixed with UHMWPE particles and water. Based on the LCROSS findings, radiation shielding experiments were conducted to test for shielding efficiency of regolith/UHMWPE/water mixtures with various percentages of water to compare relative shielding characteristics of these materials. One set of radiation studies were performed using the proton synchrotron at the Loma Linda Medical University where high energy protons similar to those found on the surface of the moon can be generated. A similar experimental protocol was also used at a high energy spalation neutron source at Los Alamos Neutron Science Center (LANSCE). These experiments studied the shielding efficiency against secondary neutrons, another major component of space radiation field. In both the proton and neutron studies, shielding efficiency was determined by utilizing a tissue equivalent proportional counter (TEPC) behind various thicknesses of shielding composite panels or mixture materials. Preliminary results from these studies indicated that adding 2 wt% water to regolith particles could increase shielding of

  17. Radiation streaming: the continuing problem of shield design

    International Nuclear Information System (INIS)

    Avery, A.F.

    1977-01-01

    The practical problems of shield design are reviewed and the major difficulties are shown to be those associated with streaming problems. The situations in which streaming occurs in various types of reactor are described including LMFBR's and fusion devices, and examples are given of ways in which the problems have been solved

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

    Science.gov (United States)

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

    2013-01-01

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

  19. Evaluation of sand as a shielding material for radiation therapy facilities

    Energy Technology Data Exchange (ETDEWEB)

    Walker, W.J.; Darwish, S.M. [Oncology Services Corp., Lovettsville, VA (United States); Fitzgerald, L.T. [Robert Boissoneault Oncology Inst., Ocala, FL (United States)

    1996-06-01

    Radiation protective barriers are designed to ensure that dose equivalent received by any individual does not exceed the applicable maximum permissible value. Materials used conventionally for shielding high energy radiotherapy facilities included concrete, lead, and steel. The choice of a shielding material is dictated by economic factors, the availability of space, and the energy range of the radiation to be attenuated. The use of silica sand to shield Megavoltage teletherapy rooms has only been recently considered for the purpose of lowering construction costs, compared to that of concrete, as well as reducing construction time. This work discusses the design and shielding evaluation of two radiation therapy facilities, which have recently been commissioned, in which the secondary barriers were designed using sand as the shielding material. For these facilities the exterior and interior walls as well as the roof were first constructed using slabs of concrete to form a shell in which sand was poured to fill the space between the slabs. Primary walls for the vault were constructed only of concrete. The sand used had a moisture content of approximately 5% and a density of about 100 lb/ft{sup 3}. In order to minimize settling over time, the sand was poured from a height of about 25 feet so that maximum compacting effect was obtained under gravity. Because of the lack of attenuation data for sand, barrier evaluation has first considered the concrete thickness required to provide adequate shielding, and the equivalent sand thickness was then determined based on the ratio of sand density to that of concrete. Post construction radiation surveys of both facilities have shown that radiation exposure levels are within the permissible limits and they proved that using sand as a shielding material is adequate and prudent.

  20. Environmental gamma radiation measurements on providence of Camaguey, Cuba

    International Nuclear Information System (INIS)

    Brigido, F.O.; Barrerras Caballero, A.; Montalvan Estrada, A.; Queipo Garcia, M.; Perez Sanchez, D.

    1999-01-01

    The population exposure to those living on the Camaguey Province of Cuba, was estimated by measuring the natural gamma background. Gamma spectra of soils and measurements of absorbed dose rate in air were taken. Radiological measurements carried out with a portable ionization chamber RSS-112 at the sampled sites revealed an average outdoor absorbed dose rate of 63.6 n Gy.h 1 - due to cosmic rays and terrestrial gamma radiation. Computed dose rates obtained through the UNSCEAR(1993) dose coefficients range from 5-136 n Gy.h 1 - , with a mean value of 39.2 n Gy.h 1 - , due to natural terrestrial gamma radiation

  1. Field maintenance of radiation-shielding windows at HFEF

    International Nuclear Information System (INIS)

    Tobias, D.A.

    1983-01-01

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

  2. Radiation monitoring in a self-shielded cyclotron installation

    International Nuclear Information System (INIS)

    Capaccioli, L.; Gori, C.; Mazzocchi, S.; Spano, G.

    2002-01-01

    As nuclear medicine is approaching a new era with the spectacular growth of PET diagnosis, the number of medical cyclotrons installed within the major hospitals is increasing accordingly. Therefore modern medical cyclotron are highly engineered and highly reliable apparatus, characterised with reduced accelerating energies (as the major goal is the production of fluorine 18) and often self-shielded. However specific dedicated monitors are still necessary in order to assure the proper radioprotection. At the Careggi University Hospital in Florence a Mini trace 10 MeV self-shielded cyclotron produced by General Electric has been installed in 2000. In a contiguous radiochemistry laboratory, the preparation and quality control of 1 8F DG and other radiopharmaceuticals takes place. Aim of this work is the characterisation and the proper calibration of the above mentioned monitors and control devices

  3. Radiation Protection of New Lightweight Electromagnetic Interference Shielding Materials Determined

    Science.gov (United States)

    1996-01-01

    Weight savings as high as 80 percent could be achieved by simply switching from aluminum electromagnetic interference (EMI) shielding covers for spacecraft power systems to EMI covers made from intercalated graphite fiber composites. Because EMI covers typically make up about one-fifth of the power system mass, this change would decrease the mass of a spacecraft power system by more than 15 percent. Intercalated graphite fibers are made by diffusing guest atoms or molecules, such as bromine, between the carbon planes of the graphite fibers. The resulting bromine-intercalated fibers have mechanical and thermal properties nearly identical to pristine graphite fibers, but their resistivity is lower by a factor of 5, giving them better electrical conductivity than stainless steel and making these composites suitable for EMI shielding.

  4. The calculation of some gamma shielding parameters for semiconductor CsPbBr3

    Science.gov (United States)

    Oto, Berna; Gulebaglan, Sinem Erden; Kanberoglu, Gulsah Saydan

    2017-02-01

    Recently, researchers produced perovskites structures used in optoelectronic devices as substrates, sensors. CsPbBr3 crystal is found in the cubic perovskite structure and its space group is Pm-3m. CsPbBr3 is a developing material for detection of X- and γ-ray radiations and the knowledge of the attenuation parameters of CsPbBr3 crystal is important. In this study, some photon shielding parameters such as mass attenuation coefficient (μρ), effective atomic number (Zeff) and electron density (Nel) have been investigated for CsPbBr3 compound. The theoretical values of μρ have been calculated in the energy range from 1 keV to 100 GeV using WinXCom computer code and these values have been used in order to calculate the values of Zeff and Nel in the same energy range.

  5. Gamma radiation in ceramic capacitors: a study for space missions

    Science.gov (United States)

    dos Santos Ferreira, Eduardo; Sarango Souza, Juliana

    2017-10-01

    We studied the real time effects of the gamma radiation in ceramic capacitors, in order to evaluate the effects of cosmic radiation on these devices. Space missions have electronic circuits with various types of devices, many studies have been done on semiconductor devices exposed to gamma radiation, but almost no studies for passive components, in particular ceramic capacitors. Commercially sold ceramic capacitors were exposed to gamma radiation, and the capacitance was measured before and after exposure. The results clearly show that the capacitance decreases with exposure to gamma radiation. We confirmed this observation in a real time capacitance measurement, obtained using a data logging system developed by us using the open source Arduino platform.

  6. Two-dimensional radiation shielding optimization analysis of spent fuel transport container

    International Nuclear Information System (INIS)

    Tian Yingnan; Chen Yixue; Yang Shouhai

    2013-01-01

    The intelligent radiation shielding optimization design software platform is a one-dimensional multi-target radiation shielding optimization program which is developed on the basis of the genetic algorithm program and one-dimensional discrete ordinate program-ANISN. This program was applied in the optimization design analysis of the spent fuel transport container radiation shielding. The multi-objective optimization calculation model of the spent fuel transport container radiation shielding was established, and the optimization calculation of the spent fuel transport container weight and radiation dose rate was carried by this program. The calculation results were checked by Monte-Carlo program-MCNP/4C. The results show that the weight of the optimized spent fuel transport container decreases to 81.1% of the origin and the radiation dose rate decreases to below 65.4% of the origin. The maximum deviation between the calculated values from the program and the MCNP is below 5%. The results show that the optimization design scheme is feasible and the calculation result is correct. (authors)

  7. Analytical theory of coherent synchrotron radiation wakefield of short bunches shielded by conducting parallel plates

    Energy Technology Data Exchange (ETDEWEB)

    Stupakov, Gennady; Zhou, Demin

    2016-04-21

    We develop a general model of coherent synchrotron radiation (CSR) impedance with shielding provided by two parallel conducting plates. This model allows us to easily reproduce all previously known analytical CSR wakes and to expand the analysis to situations not explored before. It reduces calculations of the impedance to taking integrals along the trajectory of the beam. New analytical results are derived for the radiation impedance with shielding for the following orbits: a kink, a bending magnet, a wiggler of finite length, and an infinitely long wiggler. All our formulas are benchmarked against numerical simulations with the CSRZ computer code.

  8. Observations of nesting avifauna under gamma-radiation exposure

    International Nuclear Information System (INIS)

    Buech, R.R.

    1977-01-01

    An opportunity arose to observe the nesting success of birds (up to the time of fledging) when the Enterprise Forest Radiation Facility was established for a study of the effects of gamma radiation on the flora and fauna of northern forest communities. The results of these observations on the fate of the nest occupants in relation to radiation exposure are presented

  9. A point-kernel shielding code for calculations of neutron and secondary gamma-ray 1cm dose equivalents: PKN

    International Nuclear Information System (INIS)

    Kotegawa, Hiroshi; Tanaka, Shun-ichi

    1991-09-01

    A point-kernel integral technique code, PKN, and the related data library have been developed to calculate neutron and secondary gamma-ray dose equivalents in water, concrete and iron shields for neutron sources in 3-dimensional geometry. The comparison between calculational results of the present code and those of the 1-dimensional transport code ANISN = JR, and the 2-dimensional transport code DOT4.2 showed a sufficient accuracy, and the availability of the PKN code has been confirmed. (author)

  10. Radiation shielding at interim storage facility for CANDU-type nuclear spent fuel

    International Nuclear Information System (INIS)

    Mateescu, S.; Radu, M. Pantazi D.; Stanciu, M.

    1997-01-01

    Technical measures in radiological protection are taken in the interim storage facility design to ensure that, during normal operation, exposures of workers and members of public to ionizing radiation are limited to levels lower than regulatory limits. The spent fuel storage design provides for radiation exposure to be as low as reasonable achievable (ALARA principles). The evaluation of radiation shields includes the most conservative provisions: - all locations which may contain spent fuel are full; - the spent fuel has reached the maximum burnup; - the post irradiation cooling period should be the minimum reasonable; - equipment for handling contains the maximum amount of spent fuel. Radiation shields should ensure that external radiation fields do not exceed limits accepted by the Regulatory Body Module. The evaluation has been performed with two computer codes, QAD-5K and MICROSHIELD-4. (authors)

  11. A study on the scattered-rays from the radiation shielding materials

    International Nuclear Information System (INIS)

    Kim, C.K.; Huh, J.

    1980-01-01

    To shield the radiation, we can make use of various materials, but the scattered rays can be caused by the shielding materials. The degree of the scattered rays production is influenced by the nature of the shielding materials and the energy of the radiation, therefore to choose the proper shielding material is the most important matter in radiation protection. Authors made an experimental study on the scattered rays generated from the shielding materials, and obtained the results as follows: In the ranking of the scattered rays production: Cement bricks, black colored fire bricks, and red colored fire bricks were marked the first, the second, and the third ranking respectly, and the last order was lead plates. In the relative ranking of the scattered rays production by energy increase: Lead plates were marked the first order, the next and third order were red colored fire bricks and black colored fire bricks respectively, and cement bricks were marked the last order. The scattered ray ratio of lateral-back point per lateral point were generally decreased by energy increment. The diminishing orders were that lead plates were the first order, and the next and the third order were red colored fire bricks and black colored fire bricks respectively, cement bricks were marked the last order (Author)

  12. Guide to verification and validation of the SCALE-4 radiation shielding software

    Energy Technology Data Exchange (ETDEWEB)

    Broadhead, B.L.; Emmett, M.B.; Tang, J.S.

    1996-12-01

    Whenever a decision is made to newly install the SCALE radiation shielding software on a computer system, the user should run a set of verification and validation (V&V) test cases to demonstrate that the software is properly installed and functioning correctly. This report is intended to serve as a guide for this V&V in that it specifies test cases to run and gives expected results. The report describes the V&V that has been performed for the radiation shielding software in a version of SCALE-4. This report provides documentation of sample problems which are recommended for use in the V&V of the SCALE-4 system for all releases. The results reported in this document are from the SCALE-4.2P version which was run on an IBM RS/6000 work-station. These results verify that the SCALE-4 radiation shielding software has been correctly installed and is functioning properly. A set of problems for use by other shielding codes (e.g., MCNP, TWOTRAN, MORSE) performing similar V&V are discussed. A validation has been performed for XSDRNPM and MORSE-SGC6 utilizing SASI and SAS4 shielding sequences and the SCALE 27-18 group (27N-18COUPLE) cross-section library for typical nuclear reactor spent fuel sources and a variety of transport package geometries. The experimental models used for the validation were taken from two previous applications of the SASI and SAS4 methods.

  13. Effect of Bismuth Breast Shielding on Radiation Dose and Image Quality in Coronary CT Angiography

    Science.gov (United States)

    Einstein, Andrew J.; Elliston, Carl D.; Groves, Daniel W.; Cheng, Bin; Wolff, Steven D.; Pearson, Gregory D. N.; Peters, M. Robert; Johnson, Lynne L.; Bokhari, Sabahat; Johnson, Gary W.; Bhatia, Ketan; Pozniakoff, Theodore; Brenner, David J.

    2011-01-01

    Background Coronary computed tomographic angiography (CCTA) is associated with high radiation dose to the female breasts. Bismuth breast shielding offers the potential to significantly reduce dose to the breasts and nearby organs, but the magnitude of this reduction and its impact on image quality and radiation dose have not been evaluated. Methods Radiation doses from CCTA to critical organs were determined using metal-oxide-semiconductor field-effect transistors positioned in a customized anthropomorphic whole-body dosimetry verification phantom. Image noise and signal were measured in regions of interest (ROIs) including the coronary arteries. Results With bismuth shielding, breast radiation dose was reduced 46–57% depending on breast size and scanning technique, with more moderate dose reduction to the heart, lungs, and esophagus. However, shielding significantly decreased image signal (by 14.6 HU) and contrast (by 28.4 HU), modestly but significantly increased image noise in ROIs in locations of coronary arteries, and decreased contrast-to-noise ratio by 20.9%.. Conclusions While bismuth breast shielding can significantly decrease radiation dose to critical organs, it is associated with an increase in image noise, decrease in contrast-to-noise, and changes tissue attenuation characteristics in the location of the coronary arteries. PMID:22068687

  14. Gamma radiation sterilized amnion: use in ophthalmology

    International Nuclear Information System (INIS)

    Martinez P, M. E.; Leon T, Y.; Vazquez M, L.

    2010-10-01

    Amnion processed at the Radio sterilized Tissue Bank at the National Institute of Nuclear Research, sterilized with 60 Co gamma radiation, have been used in Mexico since 2005 either as a graft to replace the damaged ocular surface, or as a patch to prevent unwanted inflammatory reactions. Patients from the Hospital General de Mexico (HGM) and Instituto Mexicano del Seguro Social (IMSS), suffering diverse pathologies such as keratoconjunctivitis; recurrent pterygium associated with symblepharon; corneal neuro trophic ulcers, chemical and thermal burns, and corneal thinning s, had been successfully treated with irradiated amnion. In the HGM, a clinical prospective study on lesions of the ocular surface of 17 eyes from 15 patients, affected with the above mentioned pathologies, was successful in 88.2%. The results have proven to be excellent as much for cosmetic purposes as for functional ones. Without the treatment, the patients could have suffered a healing after-effect or loss of sight. At IMSS, a controlled clinical randomized trial with 108 eyes from 100 patients, affected with primary nasal pterygium, was performed in 2009. These eyes were treated with radio sterilized amnion and intraoperative mitomycin C to prevent recurrence after excision of the primary pterygium. The preliminary results do not shown adverse reaction, inflammation and pain were significantly reduced radio sterilized amnion also offer security because they do no express antigens HLA-A, B or Dr and the sterile irradiated tissue do not provoke rejection or transmit an infective disease. (Author)

  15. Gamma radiation sterilized amnion: use in ophthalmology

    Energy Technology Data Exchange (ETDEWEB)

    Martinez P, M. E. [ININ, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico); Leon T, Y. [Hospital General Regional 220, IMSS, Paseo Tollocan No. 620, Col. Vertice, Toluca 50150, Estado de Mexico (Mexico); Vazquez M, L., E-mail: esther.martinez@inin.gob.m [Hospital General de Mexico, Dr. Balmis 148, Col. Doctores, 06720 Mexico D. F. (Mexico)

    2010-10-15

    Amnion processed at the Radio sterilized Tissue Bank at the National Institute of Nuclear Research, sterilized with {sup 60}Co gamma radiation, have been used in Mexico since 2005 either as a graft to replace the damaged ocular surface, or as a patch to prevent unwanted inflammatory reactions. Patients from the Hospital General de Mexico (HGM) and Instituto Mexicano del Seguro Social (IMSS), suffering diverse pathologies such as keratoconjunctivitis; recurrent pterygium associated with symblepharon; corneal neuro trophic ulcers, chemical and thermal burns, and corneal thinning s, had been successfully treated with irradiated amnion. In the HGM, a clinical prospective study on lesions of the ocular surface of 17 eyes from 15 patients, affected with the above mentioned pathologies, was successful in 88.2%. The results have proven to be excellent as much for cosmetic purposes as for functional ones. Without the treatment, the patients could have suffered a healing after-effect or loss of sight. At IMSS, a controlled clinical randomized trial with 108 eyes from 100 patients, affected with primary nasal pterygium, was performed in 2009. These eyes were treated with radio sterilized amnion and intraoperative mitomycin C to prevent recurrence after excision of the primary pterygium. The preliminary results do not shown adverse reaction, inflammation and pain were significantly reduced radio sterilized amnion also offer security because they do no express antigens HLA-A, B or Dr and the sterile irradiated tissue do not provoke rejection or transmit an infective disease. (Author)

  16. Gamma Radiation-Induced Template Polymerization Technique

    International Nuclear Information System (INIS)

    Siyam, T.

    2005-01-01

    Gamma radiation induced copolymerization of acrylamide sodiumacrylate (AM-AANa) in the presence and absence of the polymer additive was studied at low monomer concentration(1.4M/l). The results showed that the exponents of the dose rate for the polymerization rate was found to be 1.3 and 1.4 in the absence and in the presence of the polymer additive respectively. The molecular weight of the formed polymer increased by addition of the polymer to the system. In the presence of the polymer the comonomers polymerize on the added polymer. In the absence of the added polymer the comonomers polymerize according to the copolymerization process at the initial stage of the copolymerization. While at high conversion the residual comonomers polymerize on the formed macromolecular chains of the produced polymer. These studies showed that the copolymerization in the presence of added polymer is completely template copolymerization while in the absence of the polymer the copolymerization process is only template process with a high conversion

  17. Shielding NSLS-II light source: Importance of geometry for calculating radiation levels from beam losses

    Science.gov (United States)

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

    2016-11-01

    Third generation high brightness light sources are designed to have low emittance and high current beams, which contribute to higher beam loss rates that will be compensated by Top-Off injection. Shielding for these higher loss rates will be critical to protect the projected higher occupancy factors for the users. Top-Off injection requires a full energy injector, which will demand greater consideration of the potential abnormal beam miss-steering and localized losses that could occur. The high energy electron injection beam produces significantly higher neutron component dose to the experimental floor than a lower energy beam injection and ramped operations. Minimizing this dose will require adequate knowledge of where the miss-steered beam can occur and sufficient EM shielding close to the loss point, in order to attenuate the energy of the particles in the EM shower below the neutron production threshold (Designing supplemental shielding near the loss point using the analytic shielding model is shown to be inadequate because of its lack of geometry specification for the EM shower process. To predict the dose rates outside the tunnel requires detailed description of the geometry and materials that the beam losses will encounter inside the tunnel. Modern radiation shielding Monte-Carlo codes, like FLUKA, can handle this geometric description of the radiation transport process in sufficient detail, allowing accurate predictions of the dose rates expected and the ability to show weaknesses in the design before a high radiation incident occurs. The effort required to adequately define the accelerator geometry for these codes has been greatly reduced with the implementation of the graphical interface of FLAIR to FLUKA. This made the effective shielding process for NSLS-II quite accurate and reliable. The principles used to provide supplemental shielding to the NSLS-II accelerators and the lessons learned from this process are presented.

  18. Gamma radiation from PSR B1055-52

    DEFF Research Database (Denmark)

    Thompson, D.J.; Bailes, M.; Bertsch, D.L.

    1999-01-01

    The telescopes on the Compton Gamma Ray Observatory (CGRO) have observed PSR B1055-52 a number of times between 1991 and 1998. From these data a more detailed picture of the gamma radiation from this source has been developed, showing several characteristics that distinguish this pulsar: the light...

  19. Effect of gamma radiation and ethylene oxide on neomycin sulphate

    International Nuclear Information System (INIS)

    Gopal, N.G.S.; Rajagopalan, S.

    1981-01-01

    Neomycin is affected by ethylene oxide but not by gamma radiation (2.75 Mrad). Differential refractometry is more advantageous in quantitating neomycin A, B and C than is the ninhydrin method. (Auth.)

  20. Inactivation of rabies diagnostic reagents by gamma radiation

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, W.C.; Chappell, W.A.; George, E.H.

    1980-11-01

    Treatment of CVS-11 rabies adsorbing suspensions and street rabies infected mouse brains with gamma radiation resulted in inactivated reagents that are safer to distribute and use. These irradiated reagents were as sensitive and reactive as the nonirradiated control reagents.

  1. Inactivation of rabies diagnostic reagents by gamma radiation

    International Nuclear Information System (INIS)

    Gamble, W.C.; Chappell, W.A.; George, E.H.

    1980-01-01

    Treatment of CVS-11 rabies adsorbing suspensions and street rabies infected mouse brains with gamma radiation resulted in inactivated reagents that are safer to distribute and use. These irradiated reagents were as sensitive and reactive as the nonirradiated control reagents

  2. Mango conservation, Mangifera indica L., haden variety by gamma radiation

    International Nuclear Information System (INIS)

    Domarco, R.E.

    1989-02-01

    This paper evaluates the chemical characteristics and the mangoes sensorial quality after treatments with different doses of gamma radiation and during a period of storage, with constant conditions of temperature and relative humidity. (author)

  3. INFORMATION-MEASURING SYSTEM FOR EVALUATION OF ELECTROMAGNETIC RADIATION POWER LEVELS INFLUENCE TO ITS WEAKENED BY PROTECTIVE SHIELDS

    Directory of Open Access Journals (Sweden)

    O. V. Boiprav

    2013-01-01

    Full Text Available An information-measuring system and realized on the basis of its methodology used for evaluation of electromagnetic radiation power levels passing via the protective shielding construction are described. It’s proposed to use the developed methodology for the testing of electromagnetic radiation shields for anechoic chambers

  4. 3D Printed Composite-Z and Graded-Z Radiation Shields (CoGZ-Rad), Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Composite-Z and Graded-Z Radiation Shields (CoGZ-Rad) uses novel multi-material 3D printing techniques to fabricate a cost-effective and lightweight radiation...

  5. Neutron radiation shielding properties of polymer incorporated self compacting concrete mixes.

    Science.gov (United States)

    Malkapur, Santhosh M; Divakar, L; Narasimhan, Mattur C; Karkera, Narayana B; Goverdhan, P; Sathian, V; Prasad, N K

    2017-07-01

    In this work, the neutron radiation shielding characteristics of a class of novel polymer-incorporated self-compacting concrete (PISCC) mixes are evaluated. Pulverized high density polyethylene (HDPE) material was used, at three different reference volumes, as a partial replacement to river sand in conventional concrete mixes. By such partial replacement of sand with polymer, additional hydrogen contents are incorporated in these concrete mixes and their effect on the neutron radiation shielding properties are studied. It has been observed from the initial set of experiments that there is a definite trend of reductions in the neutron flux and dose transmission factor values in these PISCC mixes vis-à-vis ordinary concrete mix. Also, the fact that quite similar enhanced shielding results are recorded even when reprocessed HDPE material is used in lieu of the virgin HDPE attracts further attention. Copyright © 2017 Elsevier Ltd. All rights reserved.

  6. Radiation Shielding Utilizing A High Temperature Superconducting Magnet

    Data.gov (United States)

    National Aeronautics and Space Administration — Project objective is to evaluate human radiation protection and architecture utilizing existing superconducting magnet technology while attempting to significantly...

  7. A study on radiation shielding design in MACSTOR-400(CANDU spent fuel storage facility)

    International Nuclear Information System (INIS)

    Lee, Yoon Hee

    2006-02-01

    Since the spent fuel pool will be saturated in the near future, spent fuel storage facilities are urgently needed. Because of high radiation and decay heat, spent fuel management is difficult and important. In this study, the shielding thickness of MACSTOR-400 that satisfies the general surface dose rate limit has been investigated. And the radiation shielding safety at site boundary has also been evaluated. IAEA recommends the safety series as a guideline and the U.S. follows the NUREG guide for spent fuel storage facility design. In Japan, the regulation for internal transfer is applied to the spent fuel storage. In Korea, the ACT notification for radiation protection is considered. As a shielding design requirement, it is stated that the occupational exposure dose rate must not exceed 1 mSv/week. From this value, it is assumed that the surface dose rate limit is 25 μSv/hr. And for multi unit operation in same site, the dose rate limit at the controlled area boundary is 0.25 mSv/yr. MCNP code and Microshield program were used for calculating the surface dose rate and the dose rate at site boundary respectively. The shielding should be at least 90 cm thick except the air inlet to follow the surface dose rate limit. Additional shielding is needed on air inlet because the dose rate on air inlet is higher than the dose rate on concrete surface. Without the shielding structure, the shielding thickness should be at least 127 cm. In order to satisfy the surface dose rate limit with maintaining the same concrete thickness on air inlet, shielding structure is required on air inlet. The optimum shielding structure has been proposed in this study. The allowable number of MACSTORs with considering other nuclear facilities in Wolsung site is calculated at 60. It is expected that the required number of MACSTORs are 28 in order to store the total amount of spent fuel generated during NPP operation in Wolsung. Therefore, it seems to be safe in radiation point at site boundary

  8. A study on radiation shielding design in MACSTOR-400(CANDU spent fuel storage facility)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Hee

    2006-02-15

    Since the spent fuel pool will be saturated in the near future, spent fuel storage facilities are urgently needed. Because of high radiation and decay heat, spent fuel management is difficult and important. In this study, the shielding thickness of MACSTOR-400 that satisfies the general surface dose rate limit has been investigated. And the radiation shielding safety at site boundary has also been evaluated. IAEA recommends the safety series as a guideline and the U.S. follows the NUREG guide for spent fuel storage facility design. In Japan, the regulation for internal transfer is applied to the spent fuel storage. In Korea, the ACT notification for radiation protection is considered. As a shielding design requirement, it is stated that the occupational exposure dose rate must not exceed 1 mSv/week. From this value, it is assumed that the surface dose rate limit is 25 μSv/hr. And for multi unit operation in same site, the dose rate limit at the controlled area boundary is 0.25 mSv/yr. MCNP code and Microshield program were used for calculating the surface dose rate and the dose rate at site boundary respectively. The shielding should be at least 90 cm thick except the air inlet to follow the surface dose rate limit. Additional shielding is needed on air inlet because the dose rate on air inlet is higher than the dose rate on concrete surface. Without the shielding structure, the shielding thickness should be at least 127 cm. In order to satisfy the surface dose rate limit with maintaining the same concrete thickness on air inlet, shielding structure is required on air inlet. The optimum shielding structure has been proposed in this study. The allowable number of MACSTORs with considering other nuclear facilities in Wolsung site is calculated at 60. It is expected that the required number of MACSTORs are 28 in order to store the total amount of spent fuel generated during NPP operation in Wolsung. Therefore, it seems to be safe in radiation point at site boundary

  9. Gamma radiation combined with cinnamon oil to maintain fish quality

    Science.gov (United States)

    Lyu, Fei; Zhang, Jing; Wei, Qianqian; Gao, Fei; Ding, Yuting; Liu, Shulai

    2017-12-01

    Effects of gamma radiation combined with cinnamon oil on quality of Northern Snakehead fish fillets were observed during storage at 4 °C. Fish fillets were treated with 1-5 kGy gamma radiation, 0.05-0.5% cinnamon oil or the combination of radiation and cinnamon oil. The antimicrobial activity increased with radiation dose and cinnamon oil concentration. During storage, the combination of 1 kGy radiation and 0.5% cinnamon oil displayed better inhibiting activities on aerobic plate counts, total volatile basic nitrogen, thiobarbituric acid reaction substances than 1 kGy radiation or 0.5% cinnamon oil used alone. Moreover, the combination could arrive at the similar inhibiting activities of cinnamon oil with higher concentration of 0.5% or radiation with higher dose of 5 kGy. Thus, the combination could decrease the radiation dose and cinnamon oil concentration without decreasing the effect of them on maintaining fish quality.

  10. Radiation chemistry of aqueous solutions of cyanamide. [Gamma radiation

    Energy Technology Data Exchange (ETDEWEB)

    Draganic, Z.D.; Draganic, I.G.; Jovanovic, S.V.

    1978-09-01

    Oxygen-free aqueous solutions of 0.1 M NH/sub 2/CN (pH 2.4 and 5) were irradiated with ..gamma.. rays (0.01--25 Mrad). The rate constants determined in competition experiments are: k(H + NH/sub 2/CN) = 6.7 x 10/sup 6/ M/sup -1/ sec/sup -1/, k(e/sub aq//sup -/ + NH/sub 2/CN) = 1.5 x 10/sup 9/ M/sup -1/ sec/sup -1/, and k(OH + NH/sub 2/CN) = 8.5 x 10/sup 6/ M/sup -1/ sec/sup -1/. Radiation-chemical yields were determined for the decomposition of cyanamide molecules and the formation of several radiolytic products. The following compounds were identified in irradiated solutions: H/sub 2/, CO/sub 2/, NH/sub 3/, urea, biuret, arginine, and a--N=N-- molecule assigned to methylaminoazoformamide. Possible reactions of secondary free radicals leading to the formation of radiolytic products were considered by taking into account the model of water radiolysis and the measured radiation yields.

  11. Evaluation of Multi-Functional Materials for Deep Space Radiation Shielding

    Science.gov (United States)

    Rojdev, Kristina; Atwell, William; Wilkins, Richard; Gersey, Brad; Badavi, Francis F.

    2009-01-01

    Small scale trade study of materials for radiation shielding: a) High-hydrogen polymers; b) Z-graded materials; c) Fiber-reinforced polymer composites. Discussed multi-functionality of fiber-reinforced polymer composites. Preliminary results of ground testing data.

  12. Design and evaluation of an inexpensive radiation shield for monitoring surface air temperatures

    Science.gov (United States)

    Zachary A. Holden; Anna E. Klene; Robert F. Keefe; Gretchen G. Moisen

    2013-01-01

    Inexpensive temperature sensors are widely used in agricultural and forestry research. This paper describes a low-cost (~3 USD) radiation shield (radshield) designed for monitoring surface air temperatures in harsh outdoor environments. We compared the performance of the radshield paired with low-cost temperature sensors at three sites in western Montana to several...

  13. Effects of gamma radiation on some properties of Parfaite strawberries

    International Nuclear Information System (INIS)

    Visser, C.J.; Deist, J.; De Villiers, J.F.; Truter, A.B.

    1975-01-01

    The effects of a 200 krad dose of gamma radiation on firmness, pH, total soluble solids (TSS), acid concentration and colour of Parfaite strawberries were investigated. Experiments were performed at three stages of the picking season, considering three stages of ripeness for each experiment. Radiation caused severe tissue softening. Acidity decreased with radiation treatment while TSS content tended to increase with irradiation. With certain reservations, 60 Co-radiation can be regarded as stimulatory to ripening [af

  14. Shielding and Radiation Characteristics of Cylindrical Layered Bianisotropic Structures

    Directory of Open Access Journals (Sweden)

    A. Toscano

    2005-12-01

    Full Text Available In this paper we propose an analytical study in the spectral domainof cylindrical layered structures filled with general bianisotropicmedia and fed by a 3D electric source. The integrated structure ischaracterized in terms of transmission matrices leading to anequivalent circuit representation of the whole multilayered structure.Within the framework of this two-port formalism, we present a newcontribution to the computation of the Green's function arising in theanalysis of multilayered conformal integrated antennas loaded withgeneral bianisotropic materials. We also propose an analytical study ofthe shielding effectiveness of general bianisotropic materials locatedin multilayered, cylindrical configuration. The expression of theshielded fields sustained both by plane wave and arbitrary sources isobtained in a closed analytical form. Numerical results are alsopresented showing effects of electromagnetic parameters on radiationpattern, matching properties and radar cross section of the integratedstructure.

  15. Shielded coherent synchrotron radiation and its possible effect in the next linear collider

    International Nuclear Information System (INIS)

    Warnock, R.L.

    1991-05-01

    Shielded coherent synchrotron radiation is discussed in two cases: (1) a beam following a curved path in a plane midway between two parallel, perfectly conducting plates, and (2) a beam circulating in a toroidal chamber with resistive walls. Wake fields and the radiated energy are computed with parameters for the high-energy bunch compressor of the Next Linear Collider. 5 refs., 4 figs., 1 tab

  16. Shielded coherent synchrotron radiation and its possible effect in the next linear collider

    Energy Technology Data Exchange (ETDEWEB)

    Warnock, R.L.

    1991-05-01

    Shielded coherent synchrotron radiation is discussed in two cases: (1) a beam following a curved path in a plane midway between two parallel, perfectly conducting plates, and (2) a beam circulating in a toroidal chamber with resistive walls. Wake fields and the radiated energy are computed with parameters for the high-energy bunch compressor of the Next Linear Collider. 5 refs., 4 figs., 1 tab.

  17. Design optimization of radiation shielding structure for lead slowing-down spectrometer system

    International Nuclear Information System (INIS)

    Kim, Jeong Dong; Ahn, Sang Joon; Lee, Yong Deok; Park, Chang Je

    2015-01-01

    A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as 235U, 239Pu, 241Pu, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux (>101 2n /cm 2 ·s) neutron source comprised of a high-energy (30 MeV)/high-current (∼2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (<0.06 μSv/h), a few shielding materials [high-density polyethylene (HDPE)–Borax, B 4 C, and Li 2 CO 3 ] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near future.

  18. Dibasic calcium phosphate dihydrate, USP material compatibility with gamma radiation

    Science.gov (United States)

    Betancourt Quiles, Maritza

    Gamma radiation is a commonly used method to reduce the microbial bioburden in compatible materials when it is applied at appropriate dose levels. Gamma irradiation kills bacteria and mold by breaking down the organism’s DNA and inhibiting cell division. The purpose of this study is to determine the radiation dosage to be used to treat Dibasic Calcium Phosphate Dihydrate, USP (DCPD) and to evaluate its physicochemical effects if any, on this material. This material will be submitted to various doses of gamma radiation that were selected based on literature review and existing regulations that demonstrate that this method is effective to reduce or eliminate microbial bioburden in natural source and synthetic materials. Analytical testing was conducted to the DCPD exposed material in order to demonstrate that gamma radiation does not alter the physicochemical properties and material still acceptable for use in the manufacture of pharmaceutical products. The results obtained through this study were satisfactory and demonstrated that the gamma irradiation dosages from 5 to 30 kGy can be applied to DCPD without altering its physicochemical properties. These are supported by the Assay test data evaluation of lots tested before and after gamma irradiation implementation that show no significant statistical difference between irradiated and non irradiated assay results. The results of this study represent an achievement for the industry since they provide as an alternative the use of Gamma irradiation technology to control the microbial growth in DCPD.

  19. Advanced methodologies of evaluating the radiation sources and ionising radiation shieldings for reducing the irradiation in nuclear field personnel

    International Nuclear Information System (INIS)

    Pantazi, D.; Mateescu, S.; Stanciu, M.

    2003-01-01

    One of the technical measures of protection against ionizing radiations is the radiation shielding. The process of implementing modern and efficient methods of evaluating the radiation shielding implies advanced calculation methods. That means using from simpler 1-D or 2-D computing codes such as MicroShield or QAD up to systems of codes such as SCALE (containing several independent modules) or the Monte Carlo multipurpose and many particles, MCNP, transport code. The main objective of this work is to present the Monte Carlo based evaluation of the dose rates from the CANDU type spent fuel all along the path of its handling up to intermediate storage. These values will be then compared with the values obtained from calculations with different computing programs. To obtain this objective two problems were approached: - establishing geometrical models according to the definition used by MCNP code so that the characteristics of CANDU type nuclear fuel are taking into account; - checking the validity of the proposed models by comparing the MCNP results with those obtained with other computing codes specific for shielding evaluation and radiation dose calculation

  20. Propagation speed of gamma radiation in brass

    International Nuclear Information System (INIS)

    Cavalcante, Jose T.P.D.; Silva, Paulo R.J.; Saitovich, Henrique

    2009-01-01

    The propagation speed (PS) of visible light -represented by a short frequency range in the large frame of electromagnetic radiations (ER) frequencies- in air was measured during the last century, using a great deal of different methods, with high precision results being achieved. Presently, a well accepted value, with very small uncertainty, is c= 299,792.458 Km/s) (c reporting to the Latin word celeritas: 'speed swiftness'). When propagating in denser material media (MM), such value is always lower when compared to the air value, with the propagating MM density playing an important role. Until present, such studies focusing propagation speeds, refractive indexes, dispersions were specially related to visible light, or to ER in wavelengths ranges dose to it, and with a transparent MM. A first incursion in this subject dealing with γ-rays was performed using an electronic coincidence counting system, when the value of it's PS was measured in air, C γ(air) 298,300.15 Km/s; a method that went on with later electronic improvements. always in air. To perform such measurements the availability of a γ-radiation source in which two γ-rays are emitted simultaneously in opposite directions -as already used as well as applied in the present case- turns out to be essential to the feasibility of the experiment, as far as no reflection techniques could be used. Such a suitable source was the positron emitter 22 Na placed in a thin wall metal container in which the positrons are stopped and annihilated when reacting with the medium electrons, in such way originating -as it is very well established from momentum/energy conservation laws - two gamma-rays, energy 511 KeV each, both emitted simultaneously in opposite directions. In all the previous experiments were used photomultiplier detectors coupled to NaI(Tl) crystal scintillators, which have a good energy resolution but a deficient time resolution for such purposes. Presently, as an innovative improvement, were used BaF 2

  1. Automation of scanning technique by gamma radiation

    International Nuclear Information System (INIS)

    Aamira, Yahya

    2011-01-01

    The gamma scan technique is a nuclear test allowing the analysis of the internal mechanical properties of distillation columns used in petrochemical industries. Such technique is performed manually. So we propose in this work to automate the gamma scan procedure test by using a PLC. In addition, supervision and data acquisition interfaces are proposed.

  2. Advanced Radiation Protection (ARP): Thick GCR Shield Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The Advanced Radiation Project to date has focused on SEP events.  For long duration missions outside Earth’s geomagnetic field, the galactic cosmic ray...

  3. Gamma radiation effect study in polycarbonate optical and mechanics properties

    International Nuclear Information System (INIS)

    Araujo, E.S. de.

    1991-02-01

    Polycarbonates (PC) are used in different industrial applications due to their excellent dielectric characteristics, impact resistance, and high temperature resistance. In some of these applications, the polycarbonates are exposed to gamma radiation which produces molecular scissions, causing changes in the polycarbonate properties. To estimate the radiation effects in the Durolon polycarbonate, samples were irradiated with 60 Co gamma rays with doses between 0,2 kGy and 300 kGy. The results obtained showed that the PC mechanical properties are not changed due to the gamma radiation. However the results showed an expressive variation in the yellowness index for doses above 1 kGy. The results showed that it is possible to use the gamma sterilization of PC in applications where the coloration of PC is not critical. (author). 21 refs, 25 figs, 3 tabs

  4. Research progress in airborne surveys of terrestrial gamma radiation

    International Nuclear Information System (INIS)

    Burson, Z.G.

    1974-01-01

    Progress during the last few years in airborne surveys of terrestrial gamma radiation, i.e. in the measuring, recording, and interpreting of gamma ray signals in NaI(Tl) crystals, is discussed. Non-terrestrial background contributions have been accurately characterized. The feasibility of determining the water equivalent of snow cover by aerial survey techniques has been demonstrated. Repeat surveys over areas surrounding reactor sites can now be used to detect average differences of less than 1.0 μR/hr in terrestrial gamma radiation levels. New data acquisition and recording systems allow isotope concentrations and total inventories to be measured in spatial resolutions of a few hundred feet. Aerial survey data have been combined with population distribution data to obtain population exposure values from natural terrestrial gamma radiation around reactor sites

  5. Coronary calcium scoring with MDCT: The radiation dose to the breast and the effectiveness of bismuth breast shield

    Energy Technology Data Exchange (ETDEWEB)

    Yilmaz, Mehmet Halit [Istanbul University, Cerrahpasa Medical Faculty, Department of Radiology, Istanbul (Turkey)]. E-mail: drmhyilmaz@yahoo.com; Yasar, Dogan [Cekmece Nuclear Research and Training Centre, Secondary Standard Dosimetry Laboratory, Istanbul (Turkey); Albayram, Sait [Istanbul University, Cerrahpasa Medical Faculty, Department of Radiology, Istanbul (Turkey); Adaletli, Ibrahim [Istanbul University, Cerrahpasa Medical Faculty, Department of Radiology, Istanbul (Turkey); Ozer, Harun [Istanbul University, Cerrahpasa Medical Faculty, Department of Radiology, Istanbul (Turkey); Ozbayrak, Mustafa [Istanbul University, Cerrahpasa Medical Faculty, Department of Radiology, Istanbul (Turkey); Mihmanli, Ismail [Istanbul University, Cerrahpasa Medical Faculty, Department of Radiology, Istanbul (Turkey); Akman, Canan [Istanbul University, Cerrahpasa Medical Faculty, Department of Radiology, Istanbul (Turkey)

    2007-01-15

    Objective: The purpose of our study was to determine the breast radiation dose during coronary calcium scoring with multidetector computerized tomography (MDCT). We also evaluated the degree of dose reduction by using a bismuth breast shield when performing coronary calcium scoring with MDCT. Materials and methods: The dose reduction achievable by shielding the adult (35 years or older) female breasts was studied in 25 women who underwent coronary calcium scoring with MDCT. All examinations were performed with a 16-MDCT scanner. To compare the shielded versus unshielded breast dose, the examinations were performed with (right breast) and without (left breast) breast shielding in all patients. With this technique the superficial breast doses were calculated. To determine the average glandular breast radiation dose, we imaged an anthropomorphic dosimetric phantom into which calibrated dosimeters were placed to measure the dose to the breast. The phantom was imaged using the same protocol. Radiation doses to the breasts with and without the breast shielding were measured and compared using the Student's t-test. Results: The mean radiation doses with and without the breast shield were 5.71 {+-} 1.1 mGy versus 9.08 {+-} 1.5 mGy, respectively. The breast shield provided a 37.12% decrease in radiation dose to the breast with shielding. The difference between the dose received by the breasts with and without bismuth shielding was significant, with a p-value of less than 0.001. Conclusion: The high radiation during MDCT greatly exceeds the recommended doses and should not be underestimated. Bismuth in plane shielding for coronary calcium scoring with MDCT decreased the radiation dose to the breast. We recommend routine use of breast shields in female patients undergoing calcium scoring with MDCT.

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

    Science.gov (United States)

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

    2016-09-01

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

  7. Determination of half-value layers and tenth-value layer to barite as shielding against X radiation in radiological protection

    International Nuclear Information System (INIS)

    Lopes, G.A.; Aragao Filho, G.L.; Almeida Junior, A.T.; Santos, M.A.P.; Araujo, F.G.S.; Nogueira, M.S.

    2013-01-01

    The barium mortar has been widely used as radiation shielding material for X and gamma radiations in Brazil, by presenting some advantages as the high rate of efficiency in radiation shielding, the easy handling and application, the facility to be found in the national market and low cost. The determination of the half-value layers (HVL) and tenth-value layer (TVL) of different types of barite becomes the major factor to characterize the attenuation of these materials, in order to ensure the efficiency and quality of projects shielding, by ensuring the safety of workers occupationally exposed to radiation and of individuals to the public. Thus, plates of different thickness of mortar of barite were made for determination of their HVL and TVL. The plates were irradiated with X-ray qualities for radiological protection according to standard ISO 4037. A system of CdTe spectrometry was used to acquire spectra transmitted, in the presence of each plate, and their combinations. The areas of the spectra obtained, depending on the total thickness of the plates used in the arrangement were used to determine the attenuation curves. From these curves obtained in this work was to establish the HVL and TVL

  8. Preservation of yams by gamma radiation | Bansa | Journal of the ...

    African Journals Online (AJOL)

    Loss of yam in storage due to sprouting is very high. There is the need to investigate the possibility of inhibiting the sprouting of local varieties of yams using gamma radiation. The effect of radiation on the storage yams and the functionality of the irradiated yams in the Ghanaian food system was studied. Yams were ...

  9. Effect of gamma radiation on electrical and optical properties of ...

    Indian Academy of Sciences (India)

    The increase of the current with the gamma radiation dose may be attributed partly to the healing effect and partly to the lowering of the optical bandgap. Attempts are on to understand the ... A simple hand-held real-time radiation dosimeter is usually not available, though it is highly needed. Gene- rally, policemen take ...

  10. Monolithic active pixel radiation detector with shielding techniques

    Energy Technology Data Exchange (ETDEWEB)

    Deptuch, Grzegorz W.

    2018-03-20

    A monolithic active pixel radiation detector including a method of fabricating thereof. The disclosed radiation detector can include a substrate comprising a silicon layer upon which electronics are configured. A plurality of channels can be formed on the silicon layer, wherein the plurality of channels are connected to sources of signals located in a bulk part of the substrate, and wherein the signals flow through electrically conducting vias established in an isolation oxide on the substrate. One or more nested wells can be configured from the substrate, wherein the nested wells assist in collecting charge carriers released in interaction with radiation and wherein the nested wells further separate the electronics from the sensing portion of the detector substrate. The detector can also be configured according to a thick SOA method of fabrication.

  11. MDS G(N) fast differentiation between natural and artificial gamma radiation with a new class of mobile instruments

    International Nuclear Information System (INIS)

    Katzung, W.; Bottcher, J.

    2009-01-01

    A State-of-the-Art tool used for detecting and tracking artificial gamma radiation out of a helicopter or a vehicle is the MDS G(N) - Mobile Detection System. A highly sensitive scintillation detector detects a significant artificial gamma radiation on the ground even if the helicopter is travelling at high speed. The GPS-aided system visualizes the measured values on a moveable map displayed on the screen of a notebook every second. The colours of the continuously entered points do represent adjustable alarm thresholds. This way, location and intensity of an unknown radioactive source or a radioactive contamination can be determined very quickly. The NBR-technology (Natural Background Rejection) which is used here leads to expressive measurement results differentiating between artificial and natural gamma radiation. Additional He-3 detectors allow simultaneously the detection of neutrons. The NBR principle - developed by Thermo Scientific - stands out for its very short response times. Thus, artificial radiation can be detected reliably within seconds - even when the unit is operated by untrained staff. Unlike traditional analytic measuring techniques, the NBR method is able to detect artificial radiation sources hidden or strongly shielded gamma sources clearly from the natural background radiation. The measuring range from 1 nSv/h to 20 ?Sv/h and is extended to 1 Sv/h with a Geiger Mueller counting tube. The sensitivity amounts to max. 20000 cps (referred to 1 ?Sv/h for Cs-137). The NBR- technique is well-proven and tested for: tracking hidden radiation sources, even such ones with low activity or which are shielded, detection of artificial radiation portions in the range of the natural background, reliably measuring the ambient equivalent dose rate in the range of the natural background, fast detection of artificial radioactivity out of helicopters and vehicles.(author)

  12. Radiation area monitor device and method

    Science.gov (United States)

    Vencelj, Matjaz; Stowe, Ashley C.; Petrovic, Toni; Morrell, Jonathan S.; Kosicek, Andrej

    2018-01-30

    A radiation area monitor device/method, utilizing: a radiation sensor; a rotating radiation shield disposed about the radiation sensor, wherein the rotating radiation shield defines one or more ports that are transparent to radiation; and a processor operable for analyzing and storing a radiation fingerprint acquired by the radiation sensor as the rotating radiation shield is rotated about the radiation sensor. Optionally, the radiation sensor includes a gamma and/or neutron radiation sensor. The device/method selectively operates in: a first supervised mode during which a baseline radiation fingerprint is acquired by the radiation sensor as the rotating radiation shield is rotated about the radiation sensor; and a second unsupervised mode during which a subsequent radiation fingerprint is acquired by the radiation sensor as the rotating radiation shield is rotated about the radiation sensor, wherein the subsequent radiation fingerprint is compared to the baseline radiation fingerprint and, if a predetermined difference threshold is exceeded, an alert is issued.

  13. Metal Hydrides, MOFs, and Carbon Composites as Space Radiation Shielding Mitigators

    Science.gov (United States)

    Atwell, William; Rojdev, Kristina; Liang, Daniel; Hill, Matthew

    2014-01-01

    Recently, metal hydrides and MOFs (Metal-Organic Framework/microporous organic polymer composites - for their hydrogen and methane storage capabilities) have been studied with applications in fuel cell technology. We have investigated a dual-use of these materials and carbon composites (CNT-HDPE) to include space radiation shielding mitigation. In this paper we present the results of a detailed study where we have analyzed 64 materials. We used the Band fit spectra for the combined 19-24 October 1989 solar proton events as the input source term radiation environment. These computational analyses were performed with the NASA high energy particle transport/dose code HZETRN. Through this analysis we have identified several of the materials that have excellent radiation shielding properties and the details of this analysis will be discussed further in the paper.

  14. Lunar Surface Reactor Shielding Study

    International Nuclear Information System (INIS)

    Kang, Shawn; McAlpine, William; Lipinski, Ronald

    2006-01-01

    A nuclear reactor system could provide power to support long term human exploration of the moon. Such a system would require shielding to protect astronauts from its emitted radiations. Shielding studies have been performed for a Gas Cooled Reactor system because it is considered to be the most suitable nuclear reactor system available for lunar exploration, based on its tolerance of oxidizing lunar regolith and its good conversion efficiency. The goals of the shielding studies were to determine a material shielding configuration that reduces the dose (rem) to the required level in order to protect astronauts, and to estimate the mass of regolith that would provide an equivalent protective effect if it were used as the shielding material. All calculations were performed using MCNPX, a Monte Carlo transport code. Lithium hydride must be kept between 600 K and 700 K to prevent excessive swelling from large amounts of gamma or neutron irradiation. The issue is that radiation damage causes separation of the lithium and the hydrogen, resulting in lithium metal and hydrogen gas. The proposed design uses a layer of B4C to reduce the combined neutron and gamma dose to below 0.5Grads before the LiH is introduced. Below 0.5Grads the swelling in LiH is small (less than about 1%) for all temperatures. This approach causes the shield to be heavier than if the B4C were replaced by LiH, but it makes the shield much more robust and reliable

  15. Response of the Asiatic clam, Corbicula manilensis, to gamma radiation

    International Nuclear Information System (INIS)

    Tilly, L.J.; Corey, J.C.; Bibler, N.E.

    1977-01-01

    When heat exchangers for reactors were plugged by the Asiatic clam, acute gamma radiation was considered as a possible control. Clams were collected and sorted by size; during irradiation the clams were submerged in natural water. Clams of both sizes survived large doses with no radiation damage evident in 30 days. Mortality rose steeply at doses of 2.4 x 10 4 Rad and above; smaller clams showed a greater resistance than large ones. The feasibility of using periodic exposure to gamma radiation as a means for controlling corbicula infestations is discussed

  16. Combined effect of gamma radiation and stress cracking in polystyrene

    International Nuclear Information System (INIS)

    Amorim, Fernando A.; Rabello, Marcelo S.; Silva, Leonardo G.A.

    2011-01-01

    This study aimed to evaluate the combined effect of gamma radiation and stress cracking in polystyrene. Three different grades of polystyrene were analysed. The material was submitted to tensile tests and relaxation, analysis of molecular weight and determination of crosslinking. The results showed an increase in tensile strength in the specimens that had been exposed to radiation. The higher the molecular weight polystyrene showed better mechanical properties and after suffering the effects of gamma radiation there was an increase of 5.67% in the resistance to stress cracking effects. (author)

  17. 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)

  18. Effect of gamma radiation on ''in vitro''' efficiency of fungicides

    International Nuclear Information System (INIS)

    Menten, J.O.M.; Oliveira, G.C.X.

    1984-01-01

    The activity of 60 Co gamma radiation on eight fungicides used in post-harvesting treatment of agricultural products, was studied. Rhizoctonia solani was used in biological test as indicator-fungus. The fungicides were submitted to gamma radiation doses of O (control), 1, 10, 100, 1000 and 10.000 kR, samples of the fungicides were added to the PSA culture media to obtain 0, 1, 10 and 100 ppm concentrations of the active component of each product and of each radiation dose. The ED 50 (concentration of fungicide necessary to cause 50% radial reduction of the fungic mycelium) of each fungicide in the different gamma radiation doses was determined. (M.A.C.) [pt

  19. Elastic scattering of gamma radiation in solids

    International Nuclear Information System (INIS)

    Goncalves, O.D.

    1987-01-01

    The elastic scattering of gamma rays in solids is studied: Rayleigh scattering as well as Bragg scattering in Laue geometries. We measured Rayleigh cross sections for U, Pb, Pt, W, Sn, Ag, Mo, Cd, Zn, and Cu with gamma energies ranging from 60 to 660 KeV and angles between 5 0 and 140 0 . The experimental data are compared with form factor theories and second order perturbation theories and the limits of validity of both are established. In the 60 KeV experiment, a competition between Rayleigh and Bragg effects is found in the region of low momentum transfer. The Bragg experiments were performed using the gamma ray diffractometer from the Hahn-Meitner Institut (Berlin) with gammas of 317 KeV and angles up to 2 0 . In particular, we studied the effect of annealing in nearly perfect Czochralski Silicon crystals with high perfection in the crystallographic structure. The results are compared with Kinematical and Dynamical theories. (author)

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

    International Nuclear Information System (INIS)

    Singha Roy, P.K.

    1987-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Suwimon Ruengsri

    2014-01-01

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

  2. Including shielding effects in application of the TPCA method for detection of embedded radiation sources.

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, William C.; Shokair, Isaac R.

    2011-12-01

    Conventional full spectrum gamma spectroscopic analysis has the objective of quantitative identification of all the radionuclides present in a measurement. For low-energy resolution detectors such as NaI, when photopeaks alone are not sufficient for complete isotopic identification, such analysis requires template spectra for all the radionuclides present in the measurement. When many radionuclides are present it is difficult to make the correct identification and this process often requires many attempts to obtain a statistically valid solution by highly skilled spectroscopists. A previous report investigated using the targeted principal component analysis method (TPCA) for detection of embedded sources for RPM applications. This method uses spatial/temporal information from multiple spectral measurements to test the hypothesis of the presence of a target spectrum of interest in these measurements without the need to identify all the other radionuclides present. The previous analysis showed that the TPCA method has significant potential for automated detection of target radionuclides of interest, but did not include the effects of shielding. This report complements the previous analysis by including the effects of spectral distortion due to shielding effects for the same problem of detection of embedded sources. Two examples, one with one target radionuclide and the other with two, show that the TPCA method can successfully detect shielded targets in the presence of many other radionuclides. The shielding parameters are determined as part of the optimization process using interpolation of library spectra that are defined on a 2D grid of atomic numbers and areal densities.

  3. Measurement of background gamma radiation in the northern Marshall Islands.

    Science.gov (United States)

    Bordner, Autumn S; Crosswell, Danielle A; Katz, Ainsley O; Shah, Jill T; Zhang, Catherine R; Nikolic-Hughes, Ivana; Hughes, Emlyn W; Ruderman, Malvin A

    2016-06-21

    We report measurements of background gamma radiation levels on six islands in the northern Marshall Islands (Enewetak, Medren, and Runit onEnewetak Atoll; Bikini and Nam on Bikini Atoll; and Rongelap on Rongelap Atoll). Measurable excess radiation could be expected from the decay of (137)Cs produced by the US nuclear testing program there from 1946 to 1958. These recordings are of relevance to safety of human habitation and resettlement. We find low levels of gamma radiation for the settled island of Enewetak [mean = 7.6 millirem/year (mrem/y) = 0.076 millisievert/year (mSv/y)], larger levels of gamma radiation for the island of Rongelap (mean = 19.8 mrem/y = 0.198 mSv/y), and relatively high gamma radiation on the island of Bikini (mean = 184 mrem/y = 1.84 mSv/y). Distributions of gamma radiation levels are provided, and hot spots are discussed. We provide interpolated maps for four islands (Enewetak, Medren, Bikini, and Rongelap), and make comparisons to control measurements performed on the island of Majuro in the southern Marshall Islands, measurements made in Central Park in New York City, and the standard agreed upon by the United States and the Republic of the Marshall Islands (RMI) governments (100 mrem/y = 1 mSv/y). External gamma radiation levels on Bikini Island significantly exceed this standard (P = islands are below the standard. To determine conclusively whether these islands are safe for habitation, radiation exposure through additional pathways such as food ingestion must be considered.

  4. Designing equipment for use in gamma radiation environments

    International Nuclear Information System (INIS)

    Vandergriff, K.U.

    1990-05-01

    High levels of gamma radiation are known to cause degradation in a variety of materials and components. When designing systems to operate in a high radiation environment, special precautions and procedures should be followed. This report (1) outlines steps that should be followed in designing equipment and (2) explains the general effects of radiation on various engineering materials and components. Much information exists in the literature on radiation effects upon materials. However, very little information is available to give the designer a step-by-step process for designing systems that will be subject to high levels of gamma radiation, such as those found in a nuclear fuel reprocessing facility. In this report, many radiation effect references are relied upon to aid in the design of components and systems. 11 refs., 4 tabs

  5. Designing Equipment for Use in Gamma Radiation Environments

    Energy Technology Data Exchange (ETDEWEB)

    Vandergriff, K.U.

    1990-01-01

    High levels of gamma radiation are known to cause degradation in a variety of materials and components. When designing systems to operate in a high radiation environment, special precautions and procedures should be followed. This report (1) outlines steps that should be followed in designing equipment and (2) explains the general effects of radiation on various engineering materials and components. Much information exists in the literature on radiation effects upon materials. However, very little information is available to give the designer a step-by-step process for designing systems that will be subject to high levels of gamma radiation, such as those found in a nuclear fuel reprocessing facility. In this report, many radiation effect references are relied upon to aid in the design of components and systems.

  6. Radiation shielding calculation for digital breast tomosynthesis rooms with an updated workload survey.

    Science.gov (United States)

    Yang, Kai; Schultz, Thomas J; Li, Xinhua; Liu, Bob

    2017-03-20

    To present shielding calculations for clinical digital breast tomosynthesis (DBT) rooms with updated workload data from a comprehensive survey and to provide reference shielding data for DBT rooms. The workload survey was performed from eight clinical DBT (Hologic Selenia Dimensions) rooms at Massachusetts General Hospital (MGH) for the time period between 10/1/2014 and 10/1/2015. Radiation output related information tags from the DICOM header, including mAs, kVp, beam filter material and gantry angle, were extracted from a total of 310 421 clinical DBT acquisitions from the PACS database. DBT workload distributions were determined from the survey data. In combination with previously measured scatter fraction data, unshielded scatter air kerma for each room was calculated. Experiment measurements with a linear-array detector were also performed on representative locations for verification. Necessary shielding material and thickness were determined for all barriers. For the general purpose of DBT room shielding, a set of workload-distribution-specific transmission data and unshielded scatter air kerma values were calculated using the updated workload distribution. The workload distribution for Hologic DBT systems could be simplified by five different kVp/filter combinations for shielding purpose. The survey data showed the predominance of 45° gantry location for medial-lateral-oblique views at MGH. When taking into consideration the non-isotropic scatter fraction distribution together with the gantry angle distribution, accurate and conservative estimate of the unshielded scatter air kerma levels were determined for all eight DBT rooms. Additional shielding was shown to be necessary for two 4.5 cm wood doors. This study provided a detailed workload survey and updated transmission data and unshielded scatter air kerma values for Hologic DBT rooms. Example shielding calculations were presented for clinical DBT rooms.

  7. Evaluation of radiation shielding rate of lead aprons in nuclear medicine

    Energy Technology Data Exchange (ETDEWEB)

    Han, Sang Hyun; Han, Beom Heui; Lee, Sang Ho [Dept. of Radiological Science, Seonam University, Asan (Korea, Republic of); Hong, Dong Heui [Dept. of Radiological Science, Far East University, Eumseong (Korea, Republic of); Kim, Gi Jin [Dept. of Nuclear Medicine, Konyang University Hospital, Daejeon (Korea, Republic of)

    2017-03-15

    Considering that the X-ray apron used in the department of radiology is also used in the department of nuclear medicine, the study aimed to analyze the shielding rate of the apron according to types of radioisotopes, thus γ ray energy, to investigate the protective effects. The radioisotopes used in the experiment were the top 5 nuclides in usage statistics {sup 99m}Tc, {sup 18}F, {sup 131}I, {sup 123}I, and {sup 201}Tl, and the aprons were lead equivalent 0.35 mmPb aprons currently under use in the department of nuclear medicine. As a result of experiments, average shielding rates of aprons were {sup 99m}Tc 31.59%, {sup 201}Tl 68.42%, and {sup 123}I 76.63%. When using an apron, the shielding rate of {sup 13}'1I actually resulted in average dose rate increase of 33.72%, and {sup 18}F showed an average shielding rate of –0.315%, showing there was almost no shielding effect. As a result, the radioisotopes with higher shielding rate of apron was in the descending order of {sup 123}I, {sup 201}Tl, {sup 99m}Tc, {sup 18}F, {sup 131}I. Currently, aprons used in the nuclear medicine laboratory are general X-ray aprons, and it is thought that it is not appropriate for nuclear medicine environment that utilizes γ rays. Therefore, development of nuclear medicine exclusive aprons suitable for the characteristics of radioisotopes is required in consideration of effective radiation protection and work efficiency of radiation workers.

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

    International Nuclear Information System (INIS)

    Ipe, N.; Haeffner, D.R.; Alp, E.E.; Davey, S.C.; Dejus, R.J.; Hahn, U.; Lai, B.; Randall, K.J.; Shu, D.

    1993-11-01

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

  9. Development of radiation safety monitoring system at gamma greenhouse gamma facility

    International Nuclear Information System (INIS)

    Hairul Nizam Idris; Azimawati Ahmad, Ahmad Zaki Hussain; Ahmad Fairuz Mohd Nasir

    2009-01-01

    This paper is discussing about installation of radiation safety monitoring system at Gamma Greenhouse Gamma facility, Agrotechnology and Bioscience Division (BAB). This facility actually is an outdoor type irradiation facility, which first in Nuclear Malaysia and the only one in Malaysia. Source Cs-137 (801 Curie) was use as radiation source and it located at the centre of 30 metres diameter size of open irradiation area. The radiation measurement and monitoring system to be equipped in this facility were required the proper equipment and devices, specially purpose for application at outside of building. Research review, literature study and discussion with the equipment manufacturers was being carried out, in effort to identify the best system should be developed. Factors such as tropical climate, environment surrounding and security were considered during selecting the proper system. Since this facility involving with panoramic radiation type, several critical and strategic locations have been fixed with radiation detectors, up to the distance at 200 meter from the radiation source. Apart from that, this developed system also was built for capable to provide the online real-time reading (using internet). In general, it can be summarized that the radiation safety monitoring system for outdoor type irradiation facility was found much different and complex compared to the system for indoor type facility. Keyword: radiation monitoring, radiation safety, Gamma Greenhouse, outdoor irradiation facility, panoramic radiation. (Author)

  10. RSIC [Radiation Shielding Information Center] after 25 years: Challenges and opportunities

    International Nuclear Information System (INIS)

    Roussin, R.W.; Maskewitz, B.F.; Trubey, D.K.

    1988-01-01

    The Radiation Shielding Information Center (RSIC) observed its 25th year in operation in 1987. During that time numerous changes have occurred in the government programs that sponsor RSIC and in the radiation transport community which it serves. The continued need for RSIC is evident from the steady volume of requests and interactions with the user community. It is a continual challenge to adjust and adapt our operation to respond to the demands placed on RSIC by sponsors and users. Cooperation between sponsors, users, and the RSIC staff is the key to keeping RSIC as the focus of activities in the international radiation transport community. 7 refs

  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. Shielding of radiation fields generated by 252Cf in a concrete maze. Part 2 -- Simulation

    International Nuclear Information System (INIS)

    Fasso, A.; Ipe, N.E.; Reyna, A.

    1998-03-01

    A streaming experiment performed in a concrete maze of shape and size typical of a radiotherapy room was simulated with the Monte Carlo program FLUKA. The purpose of the calculation was to test the performance of the code in the low energy neutron range, and at the same time to provide additional information which could help in optimizing shielding of medical facilities. Instrument responses were calculated at different maze locations for several experimental configurations and were compared with measurements. In addition, neutron and gamma fluence, ambient dose equivalent and effective dose were calculated at the same positions. Both sources used in the experiment, namely a bare 252 Cf source and one shielded by a tungsten shell 5 cm thick, were considered in the simulation

  13. Shielding of radiation fields generated by {sup 252}Cf in a concrete maze. Part 2 -- Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Fasso, A.; Ipe, N.E.; Reyna, A. [Stanford Univ., CA (US). Stanford Linear Accelerator Center; McCall, R.C. [McCall Associates, Woodside, CA (US)

    1998-03-01

    A streaming experiment performed in a concrete maze of shape and size typical of a radiotherapy room was simulated with the Monte Carlo program FLUKA. The purpose of the calculation was to test the performance of the code in the low energy neutron range, and at the same time to provide additional information which could help in optimizing shielding of medical facilities. Instrument responses were calculated at different maze locations for several experimental configurations and were compared with measurements. In addition, neutron and gamma fluence, ambient dose equivalent and effective dose were calculated at the same positions. Both sources used in the experiment, namely a bare {sup 252}Cf source and one shielded by a tungsten shell 5 cm thick, were considered in the simulation.

  14. Investigation of epigenetic gene regulation in Arabidopsis modulated by gamma radiation

    International Nuclear Information System (INIS)

    Woo, Hye Ryun; Kim, Jae Sung; Lee, Myung Jin; Lee, Dong Joon; Kim, Young Min; Jung, Joon Yong; Han, Wan Keun; Kang, Soo Jin

    2011-12-01

    To investigate epigenetic gene regulation in Arabidopsis modulated by gamma radiation, we examined the changes in DNA methylation and histone modification after gamma radiation and investigated the effects of gamma radiation on epigenetic information and gene expression. We have selected 14 genes with changes in DNA methylation by gamma radiation, analyzed the changes of histone modification in the selected genes to reveal the relationship between DNA methylation and histone modification by gamma radiation. We have also analyzed the effects of gamma radiation on gene expression to investigate the relationship between epigenetic information and gene expression by gamma radiation. The results will be useful to reveal the effects of gamma radiation on DNA methylation, histone modification and gene expression. We anticipate that the information generated in this proposal will help to find out the mechanism underlying the changes in epigenetic information by gamma radiation

  15. The shielding against radiation produced by powder metallurgy with tungsten copper alloy applied on transport equipment for radio-pharmaceutical products

    International Nuclear Information System (INIS)

    Cione, Francisco C.; Sene, Frank F.; Souza, Armando C. de; Betini, Evandro G.; Rossi, Jesualdo L.; Rizzuto, Marcia A.

    2015-01-01

    Safety is mandatory on medicine radiopharmaceutical transportation and dependent on radiation shielding material. The focus of the present work is to minimize the use of harmful materials as lead and depleted uranium usually used in packages transportation. The tungsten-copper composite obtained by powder metallurgy (PM) is non-toxic. In powder metallurgy the density and the porosity of the compacted parts depends basically upon particle size distribution of each component, mixture, compacting pressure and sintering temperature cycle. The tungsten-copper composite, when used for shielding charged particles, X-rays, gamma photons or other photons of lower energy require proper interpretation of the radiation transport phenomena. The radioactive energy reduction varies according to the porosity and density of the materials used as shielding. The main factor for radiation attenuation is the cross section value for tungsten. The motivation research factor is an optimization of the tungsten and cooper composition in order to achieve the best linear absorption coefficient given by equation I (x) = I 0 e (-ux) . Experiments were conducted to quantify the effective radiation shielding properties of tungsten-copper composite produced by PM, varying the cooper amount in the composite. The studied compositions were 15%, 20% and 25% copper in mass. The Compaction pressure was 270 MPa and the sintering atmosphere was in 1.1 atm in N 2 +H 2 . The sintering temperature was 980 deg C for 2 h. The linear absorption coefficient factor was similar either for the green and the sintered compacts, due the amount of porosity did not affect the radiation attenuation. Thus the sintered was meant for size reduction and mechanical properties enhancement. (author)

  16. Variation of Natural Gamma Radiation in Isparta

    International Nuclear Information System (INIS)

    Akkurt, I.

    2004-01-01

    There is always a radiation in the earth, and its level is generated primarily by galactic cosmic rays (GCR), consisting of energetic nuclei of all naturally occurring elements, interacting with atmospheric constituents, through atomic and nuclear collisions. The other sources of natural radiations are global average background radiation from terrestrial sources such as soils, rocks ete. Background radiation levels in the atmosphere vary in intensity with latitude, altitude and phase of the solar cycle. Variation of natural radiation as a function of altitude, geological structure etc has been investigated. The measurements were performed using portable radiation counter which connected to NaI(Tl) probe

  17. Study of radiation detectors response in standard X, gamma and beta radiation standard beams

    International Nuclear Information System (INIS)

    Nonato, Fernanda Beatrice Conceicao

    2010-01-01

    The response of 76 Geiger-Mueller detectors, 4 semiconductor detectors and 34 ionization chambers were studied. Many of them were calibrated with gamma radiation beams ( 37 Cs and 60 Co), and some of them were tested in beta radiation ( 90 Sr+ 9' 0Y e 204 Tl) and X radiation (N-60, N-80, N-100, N-150) beams. For all three types of radiation, the calibration factors of the instruments were obtained, and the energy and angular dependences were studied. For beta and gamma radiation, the angular dependence was studied for incident radiation angles of 0 deg and +- 45 deg. The curves of the response of the instruments were obtained over an angle interval of 0 deg to +- 90 deg, for gamma, beta and X radiations. The calibration factors obtained for beta radiation were compared to those obtained for gamma radiation. For gamma radiation, 24 of the 66 tested Geiger-Mueller detectors presented results for the energy dependence according to international recommendation of ISO 4037-2 and 56 were in accordance with the Brazilian ABNT 10011 recommendation. The ionization chambers and semiconductors were in accordance to national and international recommendations. All instruments showed angular dependence less than 40%. For beta radiation, the instruments showed unsatisfactory results for the energy dependence and angular dependence. For X radiation, the ionization chambers presented results for energy dependence according to the national recommendation, and the angular dependence was less than 40%. (author)

  18. New Geant4 based simulation tools for space radiation shielding and effects analysis

    International Nuclear Information System (INIS)

    Santina, G.; Nieminen, P.; Evansa, H.; Daly, E.; Lei, F.; Truscott, P.R.; Dyer, C.S.; Quaghebeur, B.; Heynderickx, D.

    2003-01-01

    We present here a set of tools for space applications based on the Geant4 simulation toolkit, developed for radiation shielding analysis as part of the European Space Agency (ESA) activities in the Geant4 collaboration. The Sector Shielding Analysis Tool (SSAT) and the Materials and Geometry Association (MGA) utility will first be described. An overview of the main features of the MUlti-LAyered Shielding SImulation Software tool (MULASSIS) will follow. The tool is specifically addressed to shielding optimization and effects analysis. A Java interface allows the use of MULASSIS by the space community over the World Wide Web, integrated in the widely used SPENVIS package. The analysis of the particle transport output provides automatically radiation fluence, ionising and NIEL dose and effects analysis. ESA is currently funding the porting of this tools to a lowcost parallel processor facility using the GRID technology under the ESA SpaceGRID initiative. Other Geant4 present and future projects will be presented related to the study of space environment effects on spacecrafts

  19. Prompt and delayed radiation shielding calculations for the zephyr deuterium-tritium ignition experiment

    International Nuclear Information System (INIS)

    Prillinger, G.; Fischer, A.; Fischer, E.; Krause

    1982-01-01

    Results of discrete ordinates radiation transport calculations are presented for the proposed tokamak ignition and burn control experiment ZEPHYR. As a first step, baryte concrete with 0.15 wt% B 4 C was identified as an optimum concrete for the shielding fitting tightly around the torus and some attached devices. This shielding material with a maximum thickness of 70 cm allows personnel to enter the experiment hall just a few hours after termination of a worst-case burn discharge sequence. Inside the vacuum vessel, delayed dose rates amount to several tens of rem/h after only 50 s of plasma burn for waiting times that are typical for maintenance and repair, thus, remote handling equipment is required. Bootstrapped radiation transport calculations for neutral beam injectors show them to be strongly activated after the worst-case discharge sequence with typical dose rates of some rem/h. Thus shielding is required around the injector boxes and most repair tasks have to be performed remotely. Delayed dose rates outside the torus shielding in front of typical straight diagnostic ducts with diameters of 15 to 25 cm are shown to be significant but ''hands-on'' maintenance of the diagnostic equipment will be possible with some restrictions on working time

  20. A preliminary study to metaheuristic approach in multilayer radiation shielding optimization

    Science.gov (United States)

    Arif Sazali, Muhammad; Rashid, Nahrul Khair Alang Md; Hamzah, Khaidzir

    2018-01-01

    Metaheuristics are high-level algorithmic concepts that can be used to develop heuristic optimization algorithms. One of their applications is to find optimal or near optimal solutions to combinatorial optimization problems (COPs) such as scheduling, vehicle routing, and timetabling. Combinatorial optimization deals with finding optimal combinations or permutations in a given set of problem components when exhaustive search is not feasible. A radiation shield made of several layers of different materials can be regarded as a COP. The time taken to optimize the shield may be too high when several parameters are involved such as the number of materials, the thickness of layers, and the arrangement of materials. Metaheuristics can be applied to reduce the optimization time, trading guaranteed optimal solutions for near-optimal solutions in comparably short amount of time. The application of metaheuristics for radiation shield optimization is lacking. In this paper, we present a review on the suitability of using metaheuristics in multilayer shielding design, specifically the genetic algorithm and ant colony optimization algorithm (ACO). We would also like to propose an optimization model based on the ACO method.

  1. Activities of the Radiation Shielding Information Center and a report on codes/data for high energy radiation transport

    International Nuclear Information System (INIS)

    Roussin, R.W.

    1993-01-01

    From the very early days in its history Radiation Shielding Information Center (RSIC) has been involved with high energy radiation transport. The National Aeronautics and Space Administration was an early sponsor of RSIC until the completion of the Apollo Moon Exploration Program. In addition, the intranuclear cascade work of Bertini at Oak Ridge National Laboratory provided valuable resources which were made available through RSIC. Over the years, RSIC has had interactions with many of the developers of high energy radiation transport computing technology and data libraries and has been able to collect and disseminate this technology. The current status of this technology will be reviewed and prospects for new advancements will be examined

  2. Effects of chemical sensitizers on gamma radiation processing of ...

    African Journals Online (AJOL)

    The potential of gamma radiation processing in cross-linking natural rubber latex (NRL) for production of dipped goods has been studied. NRL produced in Ghana was irradiated to 10, 20, 30, 40 and 50 kGy, respectively, in a Gamma Chamber of dose rate 0.65 kGy/h. Irradiation of the NRL was also carried out in the ...

  3. Investigation of gold as a material for thermal radiation shielding

    Science.gov (United States)

    Munshi, Amit Harenkumar

    CdS/CdTe thin film solar cells technology is one of the fastest growing carbon neutral energy sources in the world today. Manufacturing of CdS/CdTe solar modules is carried out at temperature in the range of 620350°C under a vacuum of 40 millitorr using a Heated Pocket Deposition (HPD) system in the materials engineering laboratory. Since this system operates in vacuum, majority of the heat loss is due to thermal radiation. The concept here is to conserve the heat by reflecting the infrared radiation back into the deposition system thus increasing the thermal efficiency. Various metals may be used but calculations show that using a Gold thin film mirror can effectively reflect almost 97% of the incident radiation, thus conserving energy required for the manufacturing process. However, a phenomenon called thermal grooving or island formation inhibits its use. Thermal grooving occurs when the stress concentration at the grain boundaries causes grain separation. This phenomenon is observed in thin gold films that are exposed to a temperature in excess of 350°C for over 3 to 5 hours. In this study, these films are exposed to temperature upto 620350°C for cycles as long as 200 hours. The goal of this research is to explore the solutions for elimination of the phenomenon of thermal grooving and thus extract maximum life out of these thin gold films for conservation of heat. After carefully exploring literature on past research and conducting experiments it was found that within the range of the films that were tested, a 2000 A350° film with a 150 A350° of Indium underlay showed the best performance after thermal annealing and testing.

  4. Boiling water reactor radiation shielded Control Rod Drive Housing Supports

    Energy Technology Data Exchange (ETDEWEB)

    Baversten, B.; Linden, M.J. [ABB Combustion Engineering Nuclear Operations, Windsor, CT (United States)

    1995-03-01

    The Control Rod Drive (CRD) mechanisms are located in the area below the reactor vessel in a Boiling Water Reactor (BWR). Specifically, these CRDs are located between the bottom of the reactor vessel and above an interlocking structure of steel bars and rods, herein identified as CRD Housing Supports. The CRD Housing Supports are designed to limit the travel of a Control Rod and Control Rod Drive in the event that the CRD vessel attachement went to fail, allowing the CRD to be ejected from the vessel. By limiting the travel of the ejected CRD, the supports prevent a nuclear overpower excursion that could occur as a result of the ejected CRD. The Housing Support structure must be disassembled in order to remove CRDs for replacement or maintenance. The disassembly task can require a significant amount of outage time and personnel radiation exposure dependent on the number and location of the CRDs to be changed out. This paper presents a way to minimize personal radiation exposure through the re-design of the Housing Support structure. The following paragraphs also delineate a method of avoiding the awkward, manual, handling of the structure under the reactor vessel during a CRD change out.

  5. In vitro cell culture lethal dose submitted to gamma radiation

    International Nuclear Information System (INIS)

    Moreno, Carolina S.; Rogero, Sizue O.; Rogero, Jose Roberto; Ikeda, Tamiko I.; Cruz, Aurea S.

    2009-01-01

    The present study was designed to evaluate the in vitro effect of gamma radiation in cell culture of mouse connective tissue exposed to different doses of gamma radiation and under several conditions. The cell viability was analyzed by neutral red uptake methodology. This assay was developed for establish a methodology to be used in the future in the study of resveratrol radioprotection. Resveratrol (3,4',5- trihydroxystilbene), a phenolic phytoalexin that occurs naturally in some spermatophytes, such as grapevines, in response to injury as fungal infections and exposure to ultraviolet light. In the wines this compound is found at high levels and is considered one of the highest antioxidant constituents. The intense antioxidant potential of resveratrol provides many pharmacological activities including cardioprotection, chemoprevention and anti-tumor effects. Our results demonstrated that 60 Co gamma radiation lethal dose (LD50) on NCTC clone 929 cells was about 340Gy. (author)

  6. Effect of gamma radiation on honey quality control

    Energy Technology Data Exchange (ETDEWEB)

    Bera, A. [Radiation Technology Center, IPEN-CNEN/SP, A. Lineu Prestes, 2242, 05508-000 Sao Paulo (Brazil)], E-mail: berale@usp.br; Almeida-Muradian, L.B. [Av. Prof. Lineu Prestes, 580-Cidade Universitaria, Sao Paulo (Brazil); Sabato, S.F. [Radiation Technology Center, IPEN-CNEN/SP, A. Lineu Prestes, 2242, 05508-000 Sao Paulo (Brazil)], E-mail: sfsabato@ipen.br

    2009-07-15

    Honey is one of the most complex substances produced by bees, mainly from the nectar of flowers. Gamma radiation is a technique that can be used to decrease the number of microbiological problems associated with food and increase the shelf life of certain products. The objective of this study was to verify the effect of gamma radiation with source of cobalto-60 (10 kGy) on some parameters used in honey quality control. Seven samples of pure honey were obtained from local markets in Sao Paulo, Brazil, in 2007. The methods used are in accordance with Brazilian Regulations. The physicochemical parameters analyzed were: moisture, HMF, free acidity, pH, sugars and ash. The results showed that gamma radiation, in the dose mentioned above, did not cause significant physicochemical alterations.

  7. Measurement of gamma radiation doses in nuclear power plant environment

    International Nuclear Information System (INIS)

    Bochvar, I.A.; Keirim-Markus, I.B.; Sergeeva, N.A.

    1976-01-01

    Considered are the problems of measuring gamma radiation dose values and the dose distribution in the nuclear power plant area with the aim of estimating the extent of their effect on the population. Presented are the dosimeters applied, their distribution throughout the controlled area, time of measurement. The distribution of gamma radiation doses over the controlled area and the dose alteration with the increase of the distance from the release source are shown. The results of measurements are investigated. The conclusion is made that operating nuclear power plants do not cause any increase in the gamma radiation dose over the area. Recommendations for clarifying the techniques for using dose-meters and decreasing measurement errors are given [ru

  8. Effects of gamma radiation in cauliflower (Brassica spp) minimally processed

    International Nuclear Information System (INIS)

    Nunes, Thaise C.F.; Rogovschi, Vladimir D.; Thomaz, Fernanda S.; Trindade, Reginaldo A.; Villavicencio, Anna L.C.H.; Alencar, Severino M.

    2007-01-01

    Consumers demand for health interests and the latest diet trends. The consumption of vegetables worldwide has increased every year over the past decade, consequently, less extreme treatments or additives are being required. Minimally processed foods have fresh-like characteristics and satisfy the new consumer demand. Food irradiation is an exposure process of the product to controlled sources of gamma radiation with the intention to destroy pathogens and to extend the shelf life. Minimally processed cauliflower (Brassica oleraceae) exposed to low dose of gamma radiation does not show any change in sensory attributes. The aim of this study was to analyze the effects of the low doses of gamma radiation on sensorial aspects like appearance, texture and flavor of minimally processed cauliflower. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-09-15

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

  10. Preparation and characterization of tungsten/epoxy composites for γ-rays radiation shielding

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Le [Key Laboratory of Specially Functional Polymeric Materials and Related Technology, Ministry of Education, School of Materials Science and Engineering, East China University of Science and Technology, Shanghai 200237 (China); Zhang, Yan, E-mail: yzhang@ecust.edu.cn [Key Laboratory of Specially Functional Polymeric Materials and Related Technology, Ministry of Education, School of Materials Science and Engineering, East China University of Science and Technology, Shanghai 200237 (China); Liu, Yujian; Fang, Jun [Key Laboratory of Specially Functional Polymeric Materials and Related Technology, Ministry of Education, School of Materials Science and Engineering, East China University of Science and Technology, Shanghai 200237 (China); Luan, Weilin [Key Laboratory of Pressure and Safety (MOE), School of Mechanical and Power Engineering, East China University of Science and Technology, Shanghai 200237 (China); Yang, Xiangmin; Zhang, Weidong [Institute of Nuclear Technology Application, School of Science, East China University of Science and Technology, Shanghai 200237 (China)

    2015-08-01

    Tungsten/epoxy composites were prepared by blending epoxy resin with different weight percent of tungsten powder. The effect of filler loading on shielding and mechanical properties of composites was investigated by using two different activities of Co-60 source. And radiation degradation mechanism was studied according to the changes of free radicals concentrations, mechanical properties and thermal properties of composites. The results show that with the increment of tungsten loading, shielding property of composites increases. However, with the increase of irradiation dose, thermal stability and mechanical properties of composites decrease firstly, then increase slightly, and decline sharply at the end due to competition reaction between chain scission and cross-linking originated from γ-rays radiation.

  11. Preliminary Analysis of Radiation Shielding for B-type HIC Transport Package

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dohyung; Lee, Unjang; Ko, Jaehoon; Choi, Kyu-Sup [Korea Nuclear Engineering and Service Corporation, Seoul (Korea, Republic of)

    2007-10-15

    A radiation shielding analysis has been conducted using a computer program MCNP5 for a B-type HIC (High Integrated Container) Transport Package, which contains HIC with radioactive waste or Spent Resin, for transportation from nuclear power plant sites to disposal repository. Radiation source term is first carefully determined from the safety analysis reports related to HIC for appropriate calculation. And then MCNP (v.5) is performed to obtain the minimum thickness of HIC transport package, which meets the dose rate limit for HIC transport package prescribed in Korea Nuclear Law and IAEA Safety Standards for Radioactive Material Transport. In addition, some other analyses are done about the trend of dose rates depending on the thickness of shielding material and distance from the package.

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

    Directory of Open Access Journals (Sweden)

    Remec Igor

    2017-01-01

    Full Text Available 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.

  13. Radiation Shielding Materials Containing Hydrogen, Boron, and Nitrogen: Systematic Computational and Experimental Study. Phase I

    Science.gov (United States)

    Thibeault, Sheila A.; Fay, Catharine C.; Lowther, Sharon E.; Earle, Kevin D.; Sauti, Godfrey; Kang, Jin Ho; Park, Cheol; McMullen, Amelia M.

    2012-01-01

    The key objectives of this study are to investigate, both computationally and experimentally, which forms, compositions, and layerings of hydrogen, boron, and nitrogen containing materials will offer the greatest shielding in the most structurally robust combination against galactic cosmic radiation (GCR), secondary neutrons, and solar energetic particles (SEP). The objectives and expected significance of this research are to develop a space radiation shielding materials system that has high efficacy for shielding radiation and that also has high strength for load bearing primary structures. Such a materials system does not yet exist. The boron nitride nanotube (BNNT) can theoretically be processed into structural BNNT and used for load bearing structures. Furthermore, the BNNT can be incorporated into high hydrogen polymers and the combination used as matrix reinforcement for structural composites. BNNT's molecular structure is attractive for hydrogen storage and hydrogenation. There are two methods or techniques for introducing hydrogen into BNNT: (1) hydrogen storage in BNNT, and (2) hydrogenation of BNNT (hydrogenated BNNT). In the hydrogen storage method, nanotubes are favored to store hydrogen over particles and sheets because they have much larger surface areas and higher hydrogen binding energy. The carbon nanotube (CNT) and BNNT have been studied as potentially outstanding hydrogen storage materials since 1997. Our study of hydrogen storage in BNNT - as a function of temperature, pressure, and hydrogen gas concentration - will be performed with a hydrogen storage chamber equipped with a hydrogen generator. The second method of introducing hydrogen into BNNT is hydrogenation of BNNT, where hydrogen is covalently bonded onto boron, nitrogen, or both. Hydrogenation of BN and BNNT has been studied theoretically. Hyper-hydrogenated BNNT has been theoretically predicted with hydrogen coverage up to 100% of the individual atoms. This is a higher hydrogen content

  14. Radiation-shielded double crystal X-ray monochromator for JET

    International Nuclear Information System (INIS)

    Barnsley, R.; Morsi, H.W.; Rupprecht, G.; Kaellne, E.

    1989-01-01

    A double crystal X-ray monochromator for absolute wavelength and intensity measurements with very effective shielding of its detector against neutrons and hard X-rays was brought into operation at JET. Fast wavelength scans were taken of impurity line radiation in the wavelength region from about 0.1 nm to 2.3 nm, and monochromatic as well as spectral line scans, for different operational modes of JET. (author) 5 refs., 4 figs

  15. Radiation Protection Study for the Shielding Design of the LINAC4 Beam Dump at CERN

    CERN Document Server

    Blaha, Jan

    2013-01-01

    The aim of this study is to determine an optimal shielding of the LINAC4 beam dump fulfilling the radiation protection requirements. Therefore a detailed Monte-Carlo calculation using FLUKA particle transport and interaction code has been performed and the relevant physics quantities, such as particle fluences, neutron energy spectra, residual and prompt dose rates, air and water activation have been evaluated for different LINAC4 operation phases.

  16. Monte Carlo radiation shielding and activation analyses for the Diagnostic Equatorial Port Plug in ITER

    International Nuclear Information System (INIS)

    Serikov, A.; Fischer, U.; Leichtle, D.; Pitcher, C.S.

    2012-01-01

    Highlights: ► Systematic neutronics analyses were conducted to assess the ITER Equatorial Port Plug radiation shielding performance. ► Shielding optimization was achieved by parametric analyses of several design variants using the MCNP5, FISPACT-2007, and R2Smesh codes. ► Dominant effect of radiation streaming along the port plug gaps was recognized. ► Combination of the gap labyrinths and streaming stoppers or rails reduces shutdown doses by 2 orders of magnitude. ► Using the proposed shielding, the shutdown dose in the ITER port interspace is less than the personnel access limit of 100 μSv/h. - Abstract: This paper addresses neutronics aspects of the design development of the Diagnostic Generic Equatorial Port Plug (EPP) in ITER. To secure the personnel access at the EPP back-end interspace, parametric neutronics analyses of the EPP radiation environment have been performed and practical shielding solutions have been found. Radiation transport was performed with the Monte Carlo MCNP5 code. Activation calculations were conducted with the FISPACT-2007 inventory code. The R2Smesh approach was applied to couple transport and activation calculations. Newly created EPP local MCNP5 model was devised by extracting the EPP and adjacent blanket modules from the ITER Alite-4.1 model with proper modification of the EPP geometry in accordance with recent 3D CAD CATIA model. The EPP local model reproduces the EPP neutronically important features and allows investigation of the EPP neutronics effects in isolation from all other ITER components. Thorough EPP parametric analyses revealed dominant effect of gaps around EPP and several EPP design improvements were implemented as the outcomes of the analyses. Gap labyrinths and streaming stoppers inserted into the gaps were shown are capable to reduce the shutdown dose rate which is below the 100 μSv/h limit of personnel access and by 2 orders of magnitude less than the value in the model with straight gaps.

  17. Synthesis of graphene using gamma radiations

    Indian Academy of Sciences (India)

    Abstract. Considering the advantages of radiolytic synthesis such as the absence of toxic chemical as a reducing agent, uniform distribution of reducing agent and high purity of product, the synthesis of graphene (rGO) from graphene oxide (GO) by the gamma irradiation technique using a relatively low dose rate of 0.24 kGy ...

  18. Synthesis of graphene using gamma radiations

    Indian Academy of Sciences (India)

    Considering the advantages of radiolytic synthesis such as the absence of toxic chemical as a reducing agent, uniform distribution of reducing agent and high purity of product, the synthesis of graphene (rGO) from graphene oxide (GO) by the gamma irradiation technique using a relatively low dose rate of 0.24 kGy h−1 has ...

  19. EVALUATION OF BRACHYTHERAPY FACILITY SHIELDING STATUS IN KOREA OBTAINED FROM RADIATION SAFETY REPORTS

    Directory of Open Access Journals (Sweden)

    MI HYUN KEUM

    2013-10-01

    Full Text Available Thirty-eight radiation safety reports for brachytherapy equipment were evaluated to determine the current status of brachytherapy units in Korea and to assess how radiation oncology departments in Korea complete radiation safety reports. The following data was collected: radiation safety report publication year, brachytherapy unit manufacturer, type and activity of the source that was used, affiliation of the drafter, exposure rate constant, the treatment time used to calculate workload and the HVL values used to calculate shielding design goal values. A significant number of the reports (47.4% included the personal information of the drafter. The treatment time estimates varied widely from 12 to 2,400 min/week. There was acceptable variation in the exposure rate constant values (ranging between 0.469 and 0.592 (R-m2/Ci·hr, as well as in the HVLs of concrete, steel and lead for Iridium-192 sources that were used to calculate shielding design goal values. There is a need for standard guidelines for completing radiation safety reports that realistically reflect the current clinical situation of radiation oncology departments in Korea. The present study may be useful for formulating these guidelines.

  20. Gamma radiation effects on vitamin C standard solutions

    International Nuclear Information System (INIS)

    Amaro, Jose Daniel V.; Mansur Netto, Elias

    1995-01-01

    This word shows the physical - chemical effects of gamma radiation on standard solutions of vitamin C. Samples with concentration of 50 mg/ml were exposed to different doses of gamma radiations: 1,0 2,5 and 5,0 kGy, using a cobalt-60 source, with storing periods of 0,15 and 30 days. The results showed a vitamin C concentration loss, with a minimum of 17% for the dose of 1,0 kGy immediately after irradiation and a maximum of 81% for the dose of 5 kGy and 30 days after irradiation. (author). 3 refs., 2 tabs

  1. On-site installation and shielding of a mobile electron accelerator for radiation processing

    International Nuclear Information System (INIS)

    Catana, D.; Panaitescu, J.; Axinescu, S.; Manolache, D.; Matei, C.; Corcodel, C.; Ulmeanu, M..; Bestea, V.

    1995-01-01

    The development of radiation processing of some bulk products, e.g. grains or potatoes, would be sustained if the irradiation had been carried out at the place of storage, i.e. silo. A promising solution is proposed consisting of a mobile electron accelerator, installed on a couple of trucks and traveling from one customer to another. The energy of the accelerated electrons was chosen at 5 MeV, with 10 to 50 kW beam power. The irradiation is possible either with electrons or with bremsstrahlung. A major problem of the above solution is the provision of adequate shielding at the customer, with a minimum investment cost. Plans for a bunker are presented, which houses the truck carrying the radiation head. The beam is vertical downwards, through the truck floor, through a transport pipe and a scanning horn. The irradiation takes place in a pit, where the products are transported through a belt. The belt path is so chosen as to minimize openings in the shielding. Shielding calculations are presented supposing a working regime with 5 MeV bremsstrahlung. Leakage and scattered radiation are taken into account. (orig.)

  2. Characterization of EJ-200 plastic scintillators as active background shield for cosmogenic radiation

    Science.gov (United States)

    Tkaczyk, A. H.; Saare, H.; Ipbüker, C.; Schulte, F.; Mastinu, P.; Paepen, J.; Pedersen, B.; Schillebeeckx, P.; Varasano, G.

    2018-02-01

    This paper describes the characterization of commercially available plastic scintillation detectors to be used as an active shield or veto system to reduce the neutron background resulting from atmospheric muon interactions in low-level nuclear waste assay systems. The shield consists of an array of scintillation detectors surrounding a neutron detection system. Scintillation detectors with different thicknesses are characterized for their response to gamma rays, neutrons, and muons. Response functions to gamma rays were determined and measured in the energy range from 0.6 MeV to 6.0 MeV using radionuclide sources. Neutron response functions were derived from results of time-of-flight measurements at the Van de Graaff accelerator of the INFN Legnaro and from measurements with quasi mono-energetic neutron beams produced at the Van de Graaff accelerator of the JRC Geel. From these data, the light output and resolution functions for protons and electrons were derived. The response to muons was verified by background measurements, i.e. without the presence of any neutron or gamma source. It was found that the muon peak is more pronounced when the detectors are placed horizontally. The results indicate that a scintillator with a minimum thickness of 20 mm is needed to separate events due to atmospheric muons from natural gamma ray background, and contributions due to neutron production in nuclear waste based on only the total energy deposition in the detector. In addition, it was shown that muons can be identified with a coincidence pattern when the detectors are stacked. The effectiveness of the proposed system was demonstrated based on muon induced spallation reactions in a lead sample.

  3. Calculation of shielding and radiation doses for PET/CT nuclear medicine facility

    International Nuclear Information System (INIS)

    Mollah, A.S.; Muraduzzaman, S.M.

    2011-01-01

    Positron emission tomography (PET) is a new modality that is gaining use in nuclear medicine. The use of PET and computed tomography (CT) has grown dramatically. Because of the high energy of the annihilation radiation (511 keV), shielding requirements are an important consideration in the design of a PET or PET/CT imaging facility. The goal of nuclear medicine and PET facility shielding design is to keep doses to workers and the public as low as reasonably achievable (ALARA). Design involves: 1. Calculation of doses to occupants of the facility and adjacent regions based on projected layouts, protocols and workflows, and 2. Reduction of doses to ALARA through adjustment of the aforementioned parameters. The radiological evaluation of a PET/CT facility consists of the assessment of the annual effective dose both to workers occupationally exposed, and to members of the public. This assessment takes into account the radionuclides involved, the facility features, the working procedures, the expected number of patients per year, and so on. The objective of the study was to evaluate shielding requirements for a PET/CT to be installed in the department of nuclear medicine of Bangladesh Atomic Energy Commission (BAEC). Minimizing shielding would result in a possible reduction of structural as well as financial burden. Formulas and attenuation coefficients following the basic AAPM guidelines were used to calculate un-attenuated radiation through shielding materials. Doses to all points on the floor plan are calculated based primarily on the AAPM guidelines and include consideration of broad beam attenuation and radionuclide energy and decay. The analysis presented is useful for both, facility designers and regulators. (author)

  4. Meeting the Grand Challenge of Protecting Astronaut's Health: Electrostatic Active Space Radiation Shielding for Deep Space Missions

    Data.gov (United States)

    National Aeronautics and Space Administration — This study will seek to test and validate an electrostatic gossamer structure to provide radiation shielding. It will provide guidelines for energy requirements,...

  5. Disinfection of sewage sludge with gamma radiation

    Energy Technology Data Exchange (ETDEWEB)

    Melmed, L.N.; Comninos, D.K.

    1979-10-01

    Disinfection of sewage sludge by ionizing radiation, thermoradiation, and radiation combined with oxygenation was investigated in experimentation in Johannesburg, South Africa. Inactivation of Ascaris lumbricoides ova was used as the criterion of disinfection. Experimentation and methodology are explained. Complete inactivation could be obtained when 0.5 kGy radiation was applied at 50..cap alpha..C to a sludge containing 3% solids and when 0.4 kGy radiation was applied at 55..cap alpha..C to a sludge with 20% solids. (1 drawing, 5 graphs, 4 photos, 4 tables)

  6. Phenomenological study of the double radiative decay $B ->K\\gamma\\gamma$

    CERN Document Server

    Hiller, G; Safir, Salim A

    2006-01-01

    Using the operator product expansion (OPE) technique, we study the rare double radiative decay $B\\to K \\gamma\\gamma$ in the Standard Model (SM) and beyond. We estimate the short-distance (SD) contribution to the decay amplitude in a region of the phase space which is around the point where all decay products have energy $\\sim m_b/3$ in the rest frame of the $B$-meson. At lowest order in $1/m_b$, the $B\\to K \\gamma\\gamma$ matrix element is then expressed in terms of the usual $B\\to K$ form factors known from semileptonic rare decays. The integrated SD branching ratio in the SM in the OPE region turns out to be $\\Delta {\\cal{B}}(B \\to K \\gamma \\gamma)_{SM}^{OPE} \\simeq 1 \\times 10^{-9}$. We work out the di-photon invariant mass distribution with and without the resonant background through $B\\to K \\{\\eta_c,\\chi_{c0}\\}\\to K\\gamma \\gamma$. In the SM, the resonance contribution is dominant in the region of phase space where the OPE is valid. On the other hand, the present experimental upper limit on $B_s \\to \\tau^+...

  7. Graphical-based construction of combinatorial geometries for radiation transport and shielding applications

    International Nuclear Information System (INIS)

    Burns, T.J.

    1992-01-01

    A graphical-based code system is being developed at ORNL to manipulate combinatorial geometries for radiation transport and shielding applications. The current version (basically a combinatorial geometry debugger) consists of two parts: a FORTRAN-based ''view'' generator and a Microsoft Windows application for displaying the geometry. Options and features of both modules are discussed. Examples illustrating the various options available are presented. The potential for utilizing the images produced using the debugger as a visualization tool for the output of the radiation transport codes is discussed as is the future direction of the development

  8. Preparation of polymers suitable for radiation shielding and studying its properties (polyester composites with heavy metals salts)

    International Nuclear Information System (INIS)

    Kharita, M. H.; Al-Ajji, Z.; Yousef, S.

    2010-12-01

    Four composites were prepared in this work, based on polyester and heavy metals oxides and salts. The attenuation properties, as well as mechanical properties were studied, and the chemical stability was evaluated. It has been shown, that these composites can be used in radiation shielding for X-rays successfully, and the exact composition of these composites can be optimized according to the radiation energy to prepare the lightest possible shield. (author)

  9. Design optimization of radiation shielding structure for lead slowing-down spectrometer system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jeong Dong; Ahn, Sang Joon; Lee, Yong Deok [Nonproliferation System Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Chang Je [Dept. of Nuclear Engineering, Sejong University, Seoul (Korea, Republic of)

    2015-04-15

    A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as 235U, 239Pu, 241Pu, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux (>101{sup 2n}/cm{sup 2}·s) neutron source comprised of a high-energy (30 MeV)/high-current (∼2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (<0.06 μSv/h), a few shielding materials [high-density polyethylene (HDPE)–Borax, B{sub 4}C, and Li{sub 2}CO{sub 3}] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in

  10. Design optimization of radiation shielding structure for lead slowing-down spectrometer system

    Directory of Open Access Journals (Sweden)

    Jeong Dong Kim

    2015-04-01

    Full Text Available A lead slowing-down spectrometer (LSDS system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as 235U, 239Pu, 241Pu, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea is planned to utilize a high-flux (>1012 n/cm2·s neutron source comprised of a high-energy (30 MeV/high-current (∼2 A electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (<0.06 μSv/h, a few shielding materials [high-density polyethylene (HDPE–Borax, B4C, and Li2CO3] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near

  11. Exploring innovative radiation shielding approaches in space: A material and design study for a wearable radiation protection spacesuit.

    Science.gov (United States)

    Vuolo, M; Baiocco, G; Barbieri, S; Bocchini, L; Giraudo, M; Gheysens, T; Lobascio, C; Ottolenghi, A

    2017-11-01

    We present a design study for a wearable radiation-shielding spacesuit, designed to protect astronauts' most radiosensitive organs. The suit could be used in an emergency, to perform necessary interventions outside a radiation shelter in the space habitat in case of a Solar Proton Event (SPE). A wearable shielding system of the kind we propose has the potential to prevent the onset of acute radiation effects in this scenario. In this work, selection of materials for the spacesuit elements is performed based on the results of dedicated GRAS/Geant4 1-dimensional Monte Carlo simulations, and after a trade-off analysis between shielding performance and availability of resources in the space habitat. Water is the first choice material, but also organic compounds compatible with a human space habitat are considered (such as fatty acids, gels and liquid organic wastes). Different designs and material combinations are proposed for the spacesuits. To quantify shielding performance we use GRAS/Geant4 simulations of an anthropomorphic phantom in an average SPE environment, with and without the spacesuit, and we compare results for the dose to Blood Forming Organs (BFO) in Gy-Eq, i.e. physical absorbed dose multiplied by the proton Relative Biological Effectiveness (RBE) for non-cancer effects. In case of SPE occurrence for Intra-Vehicular Activities (IVA) outside a radiation shelter, dose reductions to BFO in the range of 44-57% are demonstrated to be achievable with the spacesuit designs made only of water elements, or of multi-layer protection elements (with a thin layer of a high density material covering the water filled volume). Suit elements have a thickness in the range 2-6 cm and the total mass for the garment sums up to 35-43 kg depending on model and material combination. Dose reduction is converted into time gain, i.e. the increase of time interval between the occurrence of a SPE and the moment the dose limit to the BFO for acute effects is reached. Wearing a

  12. Exploring innovative radiation shielding approaches in space: A material and design study for a wearable radiation protection spacesuit

    Science.gov (United States)

    Vuolo, M.; Baiocco, G.; Barbieri, S.; Bocchini, L.; Giraudo, M.; Gheysens, T.; Lobascio, C.; Ottolenghi, A.

    2017-11-01

    We present a design study for a wearable radiation-shielding spacesuit, designed to protect astronauts' most radiosensitive organs. The suit could be used in an emergency, to perform necessary interventions outside a radiation shelter in the space habitat in case of a Solar Proton Event (SPE). A wearable shielding system of the kind we propose has the potential to prevent the onset of acute radiation effects in this scenario. In this work, selection of materials for the spacesuit elements is performed based on the results of dedicated GRAS/Geant4 1-dimensional Monte Carlo simulations, and after a trade-off analysis between shielding performance and availability of resources in the space habitat. Water is the first choice material, but also organic compounds compatible with a human space habitat are considered (such as fatty acids, gels and liquid organic wastes). Different designs and material combinations are proposed for the spacesuits. To quantify shielding performance we use GRAS/Geant4 simulations of an anthropomorphic phantom in an average SPE environment, with and without the spacesuit, and we compare results for the dose to Blood Forming Organs (BFO) in Gy-Eq, i.e. physical absorbed dose multiplied by the proton Relative Biological Effectiveness (RBE) for non-cancer effects. In case of SPE occurrence for Intra-Vehicular Activities (IVA) outside a radiation shelter, dose reductions to BFO in the range of 44-57% are demonstrated to be achievable with the spacesuit designs made only of water elements, or of multi-layer protection elements (with a thin layer of a high density material covering the water filled volume). Suit elements have a thickness in the range 2-6 cm and the total mass for the garment sums up to 35-43 kg depending on model and material combination. Dose reduction is converted into time gain, i.e. the increase of time interval between the occurrence of a SPE and the moment the dose limit to the BFO for acute effects is reached. Wearing a

  13. Gamma radiation stability studies of mercury fulminate

    International Nuclear Information System (INIS)

    Fondeur, F.F.

    2000-01-01

    Mercury fulminate completely decomposed in a gamma source (0.86 Mrad/h) after a dose of 208 Mrad. This exposure equates to approximately 2.4 years in Tank 15H and 4 years in Tank 12H, one of the vessels of concern. Since the tanks lost the supernatant cover layer more than a decade ago, this study suggests that any mercury fulminate or closely related energetic species decomposed long ago if ever formed

  14. Gamma radiation stability studies of mercury fulminate

    Energy Technology Data Exchange (ETDEWEB)

    Fondeur, F.F.

    2000-02-17

    Mercury fulminate completely decomposed in a gamma source (0.86 Mrad/h) after a dose of 208 Mrad. This exposure equates to approximately 2.4 years in Tank 15H and 4 years in Tank 12H, one of the vessels of concern. Since the tanks lost the supernatant cover layer more than a decade ago, this study suggests that any mercury fulminate or closely related energetic species decomposed long ago if ever formed.

  15. Present and future problems of radiation shielding for maritime transport of nuclear spent fuels

    International Nuclear Information System (INIS)

    Ueki, K.; Nariyama, N.; Ohashi, A.

    2000-01-01

    The transport of spent fuels with casks began in September 1999 by the exclusive spent fuel transport vessel the 'Rokuei Maru'. The casks have been transported to the reprocessing plant at Rokkasho-village in Aomori Prefecture. The 'Rokuei Maru' is approximately 100 m-length, 16.5 m-width and 3,000 gross-tons. The 20 NFT casks can be loaded into 5 holds. At the present time, the NFT casks can carry spent fuels of up to 44,000 MWD/MTU. Serpentine concrete is employed as a neutron shields in the hatch covers, the bulkheads, and the house front of the accommodations except the wheelhouse. Polyethylene covers the side walls in each hold. The neutron shielding ability of serpentine concrete and polyethylene was investigated by a shielding experiment using a 252 Cf-neutron source. The shielding experiment was analyzed with the Monte Carlo code MCNP 4B. In the near future, on-board experiment will be carried out to measure the dose-equivalent rate distributions in the 'Rokuei Maru' and the measured data and the Monte Carlo analysis of it will establish the radiation safety of the ship. (author)

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Sztejnberg Manuel

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  19. Modeling the effectiveness of shielding in the earth-moon-mars radiation environment using PREDICCS: five solar events in 2012

    Science.gov (United States)

    Quinn, Philip R.; Schwadron, Nathan A.; Townsend, Larry W.; Wimmer-Schweingruber, Robert F.; Case, Anthony W.; Spence, Harlan E.; Wilson, Jody K.; Joyce, Colin J.

    2017-08-01

    Radiation in the form of solar energetic particles (SEPs) presents a severe risk to the short-term health of astronauts and the success of human exploration missions beyond Earth's protective shielding. Modeling how shielding mitigates the dose accumulated by astronauts is an essential step toward reducing these risks. PREDICCS (Predictions of radiation from REleASE, EMMREM, and Data Incorporating the CRaTER, COSTEP, and other SEP measurements) is an online tool for the near real-time prediction of radiation exposure at Earth, the Moon, and Mars behind various levels of shielding. We compare shielded dose rates from PREDICCS with dose rates from the Cosmic Ray Telescope for the Effects of Radiation (CRaTER) onboard the Lunar Reconnaissance Orbiter (LRO) at the Moon and from the Radiation Assessment Detector (RAD) on the Mars Science Laboratory (MSL) during its cruise phase to Mars for five solar events in 2012 when Earth, MSL, and Mars were magnetically well connected. Calculations of the accumulated dose demonstrate a reasonable agreement between PREDICCS and RAD ranging from as little as 2% difference to 54%. We determine mathematical relationships between shielding levels and accumulated dose. Lastly, the gradient of accumulated dose between Earth and Mars shows that for the largest of the five solar events, lunar missions require aluminum shielding between 1.0 g cm-2 and 5.0 g cm-2 to prevent radiation exposure from exceeding the 30-day limits for lens and skin. The limits were not exceeded near Mars.

  20. Gamma radiation hazard to miners in bituminous coal mines

    International Nuclear Information System (INIS)

    Skubacz, K.

    1986-01-01

    Radiation hazard to miners working in 16 bituminous coal mines was determined by a thermoluminescence method. While the miners exposure to gamma radiation is higher than that of the general population, the yearly dose was never found to exceed 5 mSv in any investigated person. Factors contributing to the estimate of the exposure hazard and the need for individual dose monitoring in mines are discussed in more detail. 3 refs., 4 figs. (author)

  1. Use of gamma radiation for preparation of nutrient culture media

    Energy Technology Data Exchange (ETDEWEB)

    Speranskaya, I.D.; Tumanyan, M.A.; Mironova, L.L.

    1977-01-01

    A technique was developed for sterilization of nutrient culture media using ..gamma..-radiation. For this purpose, dry preparations were exposed to 3 to 6 Mrad radiation, then dissolved in sterile distilled water. The quality of media and solutions thus obtained is as good as that of preparations sterilized by filtration. The advantage of the proposed sterilization method is that liquid media can be rapidly prepared and dry sterile media can be stored at room temperature for long periods of time.

  2. Impact of spectral smoothing on gamma radiation portal alarm probabilities

    International Nuclear Information System (INIS)

    Burr, T.; Hamada, M.; Hengartner, N.

    2011-01-01

    Gamma detector counts are included in radiation portal monitors (RPM) to screen for illicit nuclear material. Gamma counts are sometimes smoothed to reduce variance in the estimated underlying true mean count rate, which is the 'signal' in our context. Smoothing reduces total error variance in the estimated signal if the bias that smoothing introduces is more than offset by the variance reduction. An empirical RPM study for vehicle screening applications is presented for unsmoothed and smoothed gamma counts in low-resolution plastic scintillator detectors and in medium-resolution NaI detectors. - Highlights: → We evaluate options for smoothing counts from gamma detectors deployed for portal monitoring. → A new multiplicative bias correction (MBC) is shown to reduce bias in peak and valley regions. → Performance is measured using mean squared error and detection probabilities for sources. → Smoothing with the MBC improves detection probabilities and the mean squared error.

  3. Gamma radiation effects on molecular characteristic of vegetable tannins

    International Nuclear Information System (INIS)

    Garcia Velasco, F.; Luzardo, F.H.M.; Guzman, F.; Coto Hernandez, I.; Barroso, S.; Rodriguez, O.; Diaz Rizo, O.

    2014-01-01

    The influence of gamma radiation on tannins extracted from Pinus caribaea bark and on tannin acid has been investigated in this study with the aim of searching for evidences of structural and/or conformational changes. To fulfill this purpose, the samples of tannins, such as tannic acid and P. caribaea tannin bark, were irradiated at different doses (from 5 to 35 kGy) using a cobalt-60 gamma irradiator. The changes were analyzed by a Fourier transform infrared spectrometry and by high resolution liquid chromatography. The results pointed out some structural and conformational changes under the effects of gamma radiation for doses higher than 5 kGy for P. caribaea tannin bark. However, no changes were detected on the irradiated tannic acid. The observed behavior suggests the loss of carbonyl groups. This could be associated to a decarboxylation process with the corresponding release of CO 2 from the molecule. Evidences of some conformational changes were also noted. (author)

  4. Sewage sludge pasteurization by gamma radiation: Financial viability case studies

    Science.gov (United States)

    Swinwood, Jean F.; Kotler, Jiri

    This paper examines the financial viability of sewage sludge pasteurization by gamma radiation, by examining the following three North American scenarios: 1) Small volume sewage treatment plant experiencing high sludge disposal costs. 2) Large volume sewage treatment plant experiencing low sludge disposal costs. 3) Large volume sewage treatment plant experiencing high sludge disposal costs.

  5. Gamma Radiation Processing of Clam (Galatea Paradoxa Born ...

    African Journals Online (AJOL)

    Gamma Radiation Processing of Clam (Galatea Paradoxa Born 1778) from the Volta River Estuary for Microbiological Decontamination. ... The Clam (Galatea paradoxa Born 1778) is a dermesal dweller of riverine water and filter feed by passing water through gut concentrating particulate matter including bacteria in the gut ...

  6. Effect of gamma radiation on electrical and optical properties of ...

    Indian Academy of Sciences (India)

    We have studied in detail the gamma radiation induced changes in the electrical properties of the (TeO2)0.9 (In2O3)0.1 thin films of different thicknesses, prepared by thermal evaporation in vacuum. The current–voltage characteristics for the as-deposited and exposed thin films were analysed to obtain current versus dose ...

  7. Effect of gamma radiation on optical and electrical properties of ...

    Indian Academy of Sciences (India)

    Wintec

    Gamma radiation induced changes in the optical and electrical properties of tellurium dioxide. (TeO2) thin films ... markable properties related to polarization and polariza- ... aluminium. On the top of these aluminium contacts, thin films of TeO2 of thicknesses 300, 450 and 600 nm were deposited from a molybdenum boat.

  8. The secondary biogenic radiation of gamma-irradiated human blood

    International Nuclear Information System (INIS)

    Kuzin, A.M.; Surkenova, G.N.; Budagovskij, A.V.; Gudi, G.A.

    1997-01-01

    The sample of blood freshly taken from healthy men were gamma-irradiated with a dose of 10 Gy. It was shown that after the treatment the blood gained the capacity to emit secondary biogenic radiation. Emission lasted for some hours, passed through quartz-glass curette and was revealed by stimulating influence on biological detector (sprouting seeds)

  9. Effects of gamma radiation on enzymatic production of lignolytic ...

    African Journals Online (AJOL)

    This work aimed to study the effect of gamma radiation on the production of enzymes by filamentous fungi present in the seawater used for thermoelectric Termope S / A, in the vicinity of Port of Suape, Pernambuco. The isolated microorganisms were screened for their ability to produce enzymes. Subsequently, the fungi ...

  10. Sewage sludge pasteurization by gamma radiation: financial viability case studies

    International Nuclear Information System (INIS)

    Swinwood, J.F.; Kotler, J.

    1990-01-01

    This paper examines the financial viability of sewage sludge pasteurization by gamma radiation, by examining the following three North American scenarios: 1. Small volume sewage treatment plant experiencing high sludge disposal costs; 2. Large volume sewage treatment plant experiencing low sludge disposal costs; 3. Large volume sewage treatment plant experiencing high sludge disposal costs. (author)

  11. Effect of gamma radiation on Campylobacter jejuni

    International Nuclear Information System (INIS)

    Lambert, J.D.; Maxcy, R.B.

    1984-01-01

    Radiation resistance of Campylobacter jejuni in broth, ground beef, and ground turkey meat was determined using dose levels from 0-200 Krad at -30 +/- 10 0 C, at 0-5 0 C, and at 30 +/- 10 0 C. Irradiation at -30 0 C increased radiation resistance of cultures in ground meats; broth cultures were not greatly influenced by temperature. The effect of culture age on radiation resistance was also evaluated using cells in various physiological phases. Age did not have a pronounced effect on radiation resistance. The largest D 10 value for C. jejuni was 32 Krad, which was less than D 10 values commonly reported for salmonellae. 20 references, 4 figures

  12. Shielding an MCP Detector for a Space-Borne Mass Spectrometer Against the Harsh Radiation Environment in Jupiter’s Magnetosphere

    OpenAIRE

    Lasi D.; Tulej M.; Meyer S.; Luthi M.; Galli A.; Piazza D.; Wurz P.; Reggiani D.; Xiao H.; Marcinkowski R.; Hajdas W.; Cervelli A.; Karlsson S.; Knight T.; Grande M.

    2017-01-01

    Detectors of scientific instruments on spacecraft flying through Jupiter radiation belts need to be protected from high fluxes of penetrating radiation by means of radiation shields. Electrons constitute the most difficult component of Jupiter’s magnetosphere to shield from because of their abundance penetration depth in matter and intensity of bremsstrahlung radiation generated upon interaction with the shielding material. For the Neutral and Ion Mass spectrometer (NIM) of the Particle Envir...

  13. A New Microwave Shield Preparation for Super High Frequency Range: Occupational Approach to Radiation Protection.

    Science.gov (United States)

    Zaroushani, Vida; Khavanin, Ali; Jonidi Jafari, Ahmad; Mortazavi, Seyed Bagher

    2016-01-01

    Widespread use of X-band frequency (a part of the super high frequency microwave) in the various workplaces would contribute to occupational exposure with potential of adverse health effects.  According to limited study on microwave shielding for the workplace, this study tried to prepare a new microwave shielding for this purpose. We used EI-403 epoxy thermosetting resin as a matrix and nickel oxide nanoparticle with the diameter of 15-35 nm as filler. The Epoxy/ Nickel oxide composites with 5, 7, 9 and 11 wt% were made in three different thicknesses (2, 4 and 6 mm). According to transmission / reflection method, shielding effectiveness (SE) in the X-band frequency range (8-12.5 GHz) was measured by scattering parameters directly given by the 2-port Vector Network Analyzer. The fabricated composites characterized by X-ray Diffraction and Field Emission Scanning Electron Microscope. The best average of shielding effectiveness in each thickness of fabricated composites obtained by 11%-2 mm, 7%-4 mm and 7%-6 mm composites with SE values of 46.80%, 66.72% and 64.52%, respectively. In addition, the 11%-6 mm, 5%-6 mm and 11%-4 mm-fabricated composites were able to attenuate extremely the incident microwave energy at 8.01, 8.51 and 8.53 GHz by SE of 84.14%, 83.57 and 81.30%, respectively. The 7%-4mm composite could be introduced as a suitable alternative microwave shield in radiation protection topics in order to its proper SE and other preferable properties such as low cost and weight, resistance to corrosion etc. It is necessary to develop and investigate the efficacy of the fabricated composites in the fields by future studies.

  14. Guidelines for beamline and front-end radiation shielding design at the Advanced Photon Source

    International Nuclear Information System (INIS)

    Fernandez, P.; X-Ray Science Division

    2008-01-01

    Shielding for the APS will be such that the individual radiation worker dose will be as low as reasonably achievable (ALARA). The ALARA goals for the APS are to keep the total of the work-related radiation exposure (exposure coming from other than natural or medical sources) as far below 500 person-mrem per year, collective total effective dose equivalent, as reasonably achievable. For an individual APS radiation worker, the goal is to keep the maximum occupational total effective dose equivalent of any one employee as far below 200 mrem/yr as reasonably achievable. The ALARA goal for APS beamline scientists is to keep the total of the work-related radiation exposure (exposure coming from other than natural or medical sources) as far below 100 person-mrem per year, collective total effective dose equivalent, as reasonably achievable. For an individual APS beamline scientist, the goal is to keep the maximum occupational total effective dose equivalent of any one scientist as far below 50 mrem/yr as reasonably achievable. The dose is actively monitored by the radiation monitors on the storage ring wall in each sector and by the frequent area surveys performed by the health physics personnel. For cases in which surveys indicate elevated hourly dose rates that may impact worker exposure, additional local shielding is provided to reduce the radiation field to an acceptable level. Passive area monitors are used throughout the facility to integrate doses in various areas. The results are analyzed for trends of increased doses, and shielding in these areas is evaluated and improved, as appropriate. The APS policy for on-site nonradiation workers in the vicinity of the APS facilities requires that the average nonradiation worker dose be below 0.2 mSv/yr (20 mrem/yr). In addition, the dose at the site boundary from all pathways is required to be below 0.1 mSv/yr (10 mrem/yr). For future modifications of the facility, the doses shall be evaluated and additional shielding

  15. Radiation protection in category III large gamma irradiators; Radioprotecao em irradiadores de grande porte de categoria III

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Neivaldo; Furlan, Gilberto Ribeiro, E-mail: neivaldo@cena.usp.b, E-mail: gilfurlan@cena.usp.b [Centro de Energia Nuclear na Agricultura (CENA/USP), Piracicaba, SP (Brazil); Itepan, Natanael Marcio, E-mail: natanael.itepan@unianhanguera.edu.b [Universidade Anhanguera, Goiania, GO (Brazil)

    2011-07-01

    This article discusses the advantages of category III large gamma irradiator compared to the others, with emphasis on aspects of radiological protection, in the industrial sector. This category is a kind of irradiators almost unknown to the regulators authorities and the industrial community, despite its simple construction and greater radiation safety intrinsic to the model, able to maintain an efficiency of productivity comparable to those of category IV. Worldwide, there are installed more than 200 category IV irradiators and there is none of a category III irradiator in operation. In a category III gamma irradiator, the source remains fixed in the bottom of the tank, always shielded by water, negating the exposition risk. Taking into account the benefits in relation to radiation safety, the category III large irradiators are highly recommended for industrial, commercial purposes or scientific research. (author)

  16. Disinfection of sewage water and sludge using gamma radiation

    International Nuclear Information System (INIS)

    Musaad, R.M.A.

    2008-04-01

    This study has been carried out to assess the efficiency of gamma radiation in disinfecting sewage water and sludge from harmful pathogenic bacteria (e.g. Streptococcus, Salmonella, Shigella, total E-coli and total coliform), parasites (Ascaris ova) as well as its ability to degrade organic matter (BOD). Samples were exposed to gamma-radiation doses ranging from 0.5 to 8 KGy using Co''6 0 cell. Amongst pathogenic bacteria which are subjected to different doses of gamma-radiation Streptococcus faecalis revealed to be the most resistance bacterial indicator since complete elimination of these bacteria could be attained at 3.5 KGy. While total e-coli shown to be the most sensitive with lethal dose at 2 KGy. The radiation doses that required for reducing the bacterial population by 90% (D 10 ) and 50% (D 50 ) were determined for each species. The D 10 values found ranged from 0.75 KGy for Streptococcus and 2.75 KGy for total count bacteria. On the other hand, D 50 fall within the range of 0.5 KGy for total count bacteria, total coliform and Streptococcus, and 1.0 KGy for total e-coli. With regard to the efficiency of radiation treatment to destroy Ascaris ova viability it was found that no larvae were viable after exposure to 1.0 KGy following incubation of exposed ova for four weeks period.(Author)

  17. Terrestrial gamma radiation dose rate in Cienfuegos (Cuba))

    International Nuclear Information System (INIS)

    Alonso-Hernandez, C.M.; Sanchez-Llull, M.; Cartas-Aguila, H.; Diaz-Asencio, M.; Munoz-Caravaca, A.; Morera-Gomez, Y.; Acosta-Melian, R.

    2016-01-01

    This study assesses the level of background radiation for Cienfuegos Province, Cuba. Measurements of outdoor gamma radiation (of terrestrial and cosmic origin) in air were performed at 198 locations using a GPS navigator and a dose meter (SRP-68-01, 30 x 25 mm NaI detector). The average absorbed dose was found to be 73.9 nGy h -1 (17.2-293.9 nGy h -1 ), corresponding to an annual effective dose of 74.7 μSv (21-324 μSv). When compared with the data available for other places, the absorbed gamma doses obtained in this study indicate a background radiation level that falls within natural limits for the Damuji, Salado and Caonao watersheds; however, the Arimao and Gavilanes watersheds present levels of the absorbed dose and annual effective dose comparable with high background radiation areas. An isodose map of the terrestrial gamma dose rate in Cienfuegos was drawn using the GIS application 'Arc View'. This study provides important baseline data of radiation exposure in the area. (authors)

  18. Alteration of yeast activity by gamma radiation

    International Nuclear Information System (INIS)

    Chacharkar, M.P.; Tak, B.B.; Bhati, J.

    1996-01-01

    Yeast is an important component in microbe based industrial technologies. Due to the techno-economic reasons, the fermentation technique has acquired renewed interest. The effect of γ-radiation on the fermentation reaction has been investigated. The studies show that exposure of the fermentation mixture to γ-radiation at 5 kGy enhance alcohol production, whereas irradiation at higher doses, viz., 10 kGy and 25 kGy caused a considerable reduction in the alcohol yield. Therefore, low dose irradiation of fermentation mixtures can be applied for increasing the alcohol production by about 25%. (author). 13 refs., 1 fig

  19. Gamma radiation inside closed volumes with thin irradiating walls

    International Nuclear Information System (INIS)

    Karpov, V.I.

    1978-01-01

    The dose rate of gamma radiation inside a parallelepiped with thin radiating walls was calculated. The calculation was based on determining the dose rate from a rectangular plate and subsequently summing the dose rates from all the parallelepiped walls. The dose rate from the rectangular plate was calculated by reducing it to an equivalent plate of infinite length and certain fixed width. When the radiators had constant surface density, the dose rate in the geometric centre of volumes having the form of a parallelepiped was shown to have the least value in the case when the parallelepiped degenerates to a cube

  20. Scintillation counter, maximum gamma aspect

    International Nuclear Information System (INIS)

    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.)